NUREG-0683
Vol. 1

Final ProgrammaticC:)

Environmental Impact Statement
related to decontamination and disposalof radioactive wastes resulting fromMarch 28, 1979, accident
Three Mile Island Nuclear Station, Unit 2
Docket No. 50-320
Metropolitan Edison Company
Jersey Central Power and Light Company
Pennsylvania Electric Company

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U.S. Nuclear Regulatory
Commission
Office of Nuclear Reactor Regulation
March 1981

Available fromGPO Sales Program
Division of Technical Information and Document Control
U.S. Nuclear Regulatory Commission
Washington, DC 20555
Printed copy price:$10.00andNational Technical Information Service
Springfield, VA 22161

NUREG-0683
Vol. 1

Final Programmatic
Environmental Impact Statement
related to decontamination and disposalof radioactive wastes resulting fromMarch 28, 1979, accident
Three Mile Island Nuclear Station, Unit 2
Docket No. 50-320
Metropolitan Edison Company
Jersey Central Power and Light Company
Pennsylvania Electric Company

U.S. Nuclear Regulatory
Commission
Office of Nuclear Reactor Regulation
March 1981

1

I

COVER SHEET AND ABSTRACT
1.

Proposed Action and Location:
DECONTAMINATION AND DISPOSAL OF RADIOACTIVE WASTES RESULTING FROM
THE MARCH 28, 1979, ACCIDENT AT THREE MILE ISLAND NUCLEAR STATION,
UNIT 2, LOCATED IN LONDONDERRY TOWNSHIP, DAUPHIN COUNTY, PENNSYLVANIA.

2.

Messrs. Oliver Lynch and Paul Leech are the Project Managers for thisstatement. They may be contacted at the Three Mile Island Program Office,U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 or at 301-492-7258.

3.

A Final Programmatic Environmental Impact Statement (PEIS) related to thedecontamination and disposal of radioactive wastes resulting from the March 28,
1979, accident at Three Mile Island Nuclear Station, Unit 2 (Docket No. 50-320)
has been prepared by the Office of Nuclear Reactor. Regulation of the Nuclear
Regulatory Commission in response to a directive issued by the Commission on
November 21, 1979. This statement is an overall study of the activities
necessary for decontamination of the facility, defueling, and disposition of
the radioactive wastes. The available alternatives considered ranged from
implementation of full cleanup to no action other than continuing to maintainthe reactor in a safe shutdown condition. Also included are comments ofgovernmental agencies, other organizations, and the general public on the Draft
PEIS on this project, and staff responses to these comments.

I

SUMMARYThis programmatic environmental impact statement (PEIS) by the staff of the U.S. Nuclear
Regulatory Commission (NRC)
is an overall study of the activities necessary for decontaminationof the facility, defueling, and disposition of the radioactive wastes which resulted from theaccident on March 28, 1979, at Unit 2 of the Three Mile Island Nuclear Station (TMI-2).
The
following summary has been prepared by the staff for those who prefer to follow the main themesof the statement without referring to the technical descriptions, calculations, data, and otherdetails that provide a basis for assessing the cleanup alternatives and their impacts.
In response to a directive issued by the Commission on November 21, 1979, to prepare this PEIS,
the staff has reviewed the status of the contaminated facilities and their surroundings, surveyed
the methods available to carry outthe cleanup operations, and analyzed the impacts of the cleanup
activities on the environment, members of the public, and plant workers.
In summary, the staff
has reached the following major conclusions and findings (see Section 12 for a complete listing):
Full cleanup of the TMI-2 facilities should proceed as expeditiously as reasonably possible
to reduce the potential for uncontrolled releases of radioactive materials to theenvironment.
Existing methods are adequate, or can be suitably modified, to perform virtually all of thenecessary operations without incurring environmental impacts that exceed acceptable limits;
where special tools or methods are found necessary for operations such as defueling,
engineering expertise is available to cope with such requirements.
An early decision to decommission TMI-2 will have very little
effect on the choice ofalternatives for the cleanup tasks because most of the same tasks must be performed in orderto remove and dispose of the damaged fuel.
The time needed to complete the cleanup will be 5 to 9 years from the time of the accident.
The most significant environmental impact associated with the cleanup will result from theradiation doses received by the entire work force from cleanup activities.
These doses are
estimated to be in the range from 2000 to 8000 person-rem.
It is predicted that less than one additional cancer death attributable to exposure toradiation will occur among the entire work force engaged in cleaning up TMI-2.
(The death
rate from cancer among the U.S. population averages approximately 200 deaths per 1000 people.)
Not more than two additional genetic defects are expected in descendents of exposed workers.
(Among the U.S. population, approximately 60 genetic defects can be expected per 1000 people.)Throughout the cleanup, any anticipated releases to the environment must be controlled by.
the licensee in accordance with the staff's proposed effluent criteria to conform to theindividual dose design objectives listed in 10 CFR Part 50, Appendix I, as mandatory limits.
The total-body dose design objectives are 15 mrem/year from airborne particulate releases
and 3 mrem/year from liquid releases.
Implementation of the criteria in this manner is more
stringent than for normally operating plants in recognition of the condition of TMI-2.
Assuming the cleanup is conducted in accordance with the staff's proposed effluent criteria,
the staff estimates that, for the entire cleanup, the cumulative total body dose to the
maximum exposed individual offsitewill range from 0.8 to 2.3 mrem for gaseous effluents.
An individual offsite receiving the maximum estimated dose resulting from atmosphericreleases during the entire cleanup (0.8 to 2.3 mrem) would-incur an estimated increased riskof dying from cancer of between 1 in 2 million and I in 600,000, and an increased risk of agenetic effect to offspring over the next five generations of between 1 in 300,000 and 1 in
100,000.

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ii

Assuming the cleanup is conducted in accordance with the staff's proposed effluent criteria,
the total cumulative dose received by the entire population within a 50-mile radius of TMI-2
due to both gaseous and liquid releases would range from 10 to 30 person-rem for the entire
cleanup.
This is a small fraction (about .01%) of the background radiation dose received
annually by the population from causes other than releases from TMI (annual population
background radiation dose = 116 mrem/yr x 2.2 x 106 people = 255,000 person-rems).
The psychological distress caused by the accident and operations necessary to proceed withthe cleanup has declined, but there is a potential for temporary increases in distress as
various cleanup activities are undertaken.
The contaminated accident-generated water in the reactor building basement (sump) and in thereactor primary system cannot be left in its present condition and location if the cleanup
effort is to proceed.
Removal of this contaminated accident water will reduce the airborneand direct radiation levels in the building sufficiently to permit other cleanup operations
to be accomplished with greater safety.
Treatment of the contaminated accident water will transform the entrained radioactivity fromits current mobile state to a more manageable form by concentrating and immobilizing theactivity by an appropriate process.
The cleanup activity will eliminate the risks associatedwith leaving the contaminated accident water radionuclide inventory in the mobile unprocessedstate.
A decision on the ultimate disposal of the processed water can be deferred until after thewater has been processed.
Then, the concentration of radionuclides remaining in the water
will be low enough for the water to be stored safely onsite until the disposal decision is
made.
Processing the water to immobilize most of the radionuclides and storage of theprocessed water will not foreclose any reasonable options for disposition of the water orconcentrated wastes.
The staff regards the transfer of high-specific-activity waste to facilities operated by the
Department of Energy to be the most appropriate course of action for processing and final
disposal of this material.
In the interim; radioactive fuel and high-specific-activity
wastes from TMI-2 must be packaged and will have to be stored at the site temporarily until
a suitable disposal site is established elsewhere.
No significant environmental effects are
expected from these activities.
The staff has concluded that TMI should not become a permanent radioactive waste disposalsite.
If the damaged fuel and radioactive wastes are not removed, the Island would, in
effect, become a permanent waste disposal site.
The location, geology, and hydrology ofThree Mile Island are among the factors that do not meet current criteria for a safe
long-term waste disposal facility.
Removing the damaged fuel and radioactive waste tosuitable storage sites is the only reliable means for eliminating the risk of widespread
uncontrolled contamination of the environment by the accident wastes.
The staff has based its analysis on the licensee's plans,* where they are available, as well as
on alternatives the staff has independently developed and assessed.
The alternatives considered
are, in general, dependent upon radiological and technological conditions encountered.
Becausethe precise conditions of the reactor core and other parts of the system are not known, the staff
has described and assessed probable or bounding situations. -When more information becomes available, appropriate supplements to the PEIS will be issued if the affected operations are found tobe significantly beyond the scope of these assessments.
The ultimate disposition of TMI-2 is of interest to the Federal, State, and local governments, aswell as to the licensee and the public.
However, the disposition of the facility--whether todecommission or restore it to a condition acceptable for licensed operation--is not within thescope of this PEIS.
The March 28, 1979, accident and its associated environmental impacts also
are not within the scope of this PEIS.
*The term 'licensee" or "Met-Ed" in this document refers to Metropolitan Edison Company, theprincipal owner (50 percent) and operator of the plant, Jersey Central Power and Light Company, Pennsylvania Electric Company, each of which owns 25 percent.

iii

S.1

The Situation

During the accident at TMI-2, the reactor coolant water level dropped, uncovering the upper
portion of the reactor core. This produced temperatures in the core in excess of 2500'F, which
may have had the following consequences:
Reaction of possibly 50 percent of the Zircaloy fuel cladding tubes (in the uncooled uppercore region) with the water vapor and steam, thereby causing the tubes to fail and exposing
uranium oxide fuel pellets containing fission products.
Possible melting and fusing together of various stainless steel parts on adjacent fuelassemblies, such as the top end fittings and spacer grids that are located along the fuelassembly.
Cracking and crumbling and possibly melting of uranium oxide fuel pellets in the overheatedsection of the core.
'Possible damage, caused by overheating, of other reactor parts.
It is possible that theoverheating produced local distortions and warping of some of these components.
Small pieces of fuel and other radioactive material may have been carried from the core by theflow of coolant.
Larger fragments may have settled out in parts of the primary coolant system,
smaller particles may be in suspension, and some will be dissolved in the cooling water.
Radioactive material also plated out, forming a thin layer on the inside surfaces of the coolant
system components.
Although the total quantity of radioactivity in these various forms is notknown with any precision, the upper limit on total radioactivity currently in the reactor coolant
system (exclusive of fuel) is estimated at about 140,000 Ci.
Some of the radioactive gases leaked out of the reactor coolant system along with a large amountof water.
Some of these gases escaped to the environment, but substantial amounts of radioactive
gases remained in the reactor building.
Shortly after the accident, xenon and iodine gases
accounted for most of the radioactivity in the reactor building atmosphere, but these decayed
rapidly to nonradioactive forms.
The radioactivity remaining in the reactor building atmosphereup to June 27, 1980, consisted almost entirely of an estimated 57,000 Ci of krypton (Kr-85)
gas.** Following authorization by the Commission, the gas was purged to the outside atmosphere
during the period June 28 to July 11, 1980.
Subsequently, the building has been purged several
times.
The release of Kr-85 has not exceeded 100 Ci for any purge; as of December 1980, theamount purged was less than 15 Ci per month and is decreasing.
Some of the purges were made in
conjunction with entries into the reactor building.
Several hundred thousand gallons of highly contaminated water were released from the primarysystem when the reactor pressurizer relief valve stuck open early in the accident and the coolant
overflow tank ruptured.
Additionally, primary system coolant leaked from the letdown and makeup
system into the auxiliary and fuel handling building (AFHB) contaminating the floors, walls, andstorage tanks.
About 700,000 gallons of contaminated water (termed sump'water) are standing
about 8 feet deep in the reactor building basement.
This sump water contains about 500,000 Ci ofradioactivity.
There also are about 100,000 gallons of water containing an estimated 20,000 Ci
stillcirculating in the reactor coolant system.
At the present time, heat from the reactor islost to the building and ultimately to the environment.
Backup cooling systems are available ifneeded.
Tritiated water and dissolved radionuclides of cesium and strontium are the dominantradioactive materials of concern in the accident water.
The other radionuclides are in low
concentrations.
Because the tanks then available for storing contaminated water were rapidly being filled, thenecessity for decontaminating the radioactive water in the AFHB tanks and sumps was recognizedsoon after the March 28 accident.
The use of a demineralizer system, designated as EPICOR II,
was authorized for this purpose by the NRC on October 16, 1979, and cleanup of the water in the
AFHB has been completed.
This processed water is being stored on the site in accordance withdirection of the Commission and an agreement among the NRC, the licensee, and the City of
Lancaster.
The processed water still contains tritium, which is not removed by the EPICOR II
system.
"XAnalysis after completion of the purging showed that, if instrument errors and uncertainties
in the building free volume are considered, the actual amount of Kr-85 purged wasapproximately 44,000 Ci.

iv

A demineralizer system designed for decontaminating water containing higher levels of radioactivity than EPICOR II is under construction by the licensee for processing the water in thereactor building.
This system, known as the submerged demineralizer system (SDS), andalternatives to SDS are evaluated in Section 7.1 of the PEIS.
Approval by the NRC would be
required before any of these systems could be placed in operation.
Exposed interior surfaces and equipment in the AFHB and the reactor building were contaminatedduring the accident.
The AFHB was contaminated by primary coolant leakage from the makeup andletdown system, and the reactor building was contaminated by hot water and steam carrying radionuclides released to the building under pressure.
The interior exposed surfaces of both buildings
were coated with thin deposits (known as plateout) of radioactive material.
Removal of theplateout in the AFHB began in April 1979 and about two-thirds of the interior surfaces had been
decontaminated by September 1980.
Very little
has been accomplished since then because of thelicensee's limited funds.
The largest portion of the radioactive contamination in the AFHB was
deposited in the sludge in the sump and several tanks in the auxiliary building.
These radiation
sources have not yet been removed.
Five entries into the reactor building for radiation mapping and damage assessment had been madeby January 1981, but work on decontaminating the reactor building had not yet started.
No significant impacts have been identified as a result of the low-activity solid waste handlingand shipment operations to date.
Wastes shipped by truck to the commercial low-level disposal
facility near Richland, Washington, have consisted of immobilized decontamination solutions,compacted trash and noncompactible solid materials.
As of February 5, 1981, 2013 drums and273 LSA boxes of low-level waste had been transferred off the island in 36 truck shipments.
An interim radwaste storage facility has been constructed onsite to store temporarily some of thehigher activity wastes, such as the spent demineralizer beds from the water treatment systems.
This storage facility will be used until the evaluation of alternatives for offsite disposal ofthese wastes has been completed and an appropriate one is selected.
S.2

Reasons for Cleanup

The cleanup operations will remove sources of potential radiation exposure that currently poserisks to the health and safety of station workers and the public.
Radiation sources are present
in the form of airborne contamination, wastewater contaminated by radioactive materials during
the accident, plateout of radioactive material on building and equipment surfaces, contaminatedsludge, contaminated filter cartridges and demineralizer resins, and damaged fuel.
As long as
water with radioactive substances in it is allowed to occupy sumps and tanks, there exists a
small probability of leakage into the groundwater and subsequently into the Susquehanna River.
The contaminated water is also a source of direct radiation to workers requiring access to thebuilding to perform critical maintenance (e.g., repair of nuclear instrumentation) or otherrepair to maintain the reactor in safe shutdown condition.
The reactor has been in a safe shutdown state since April 1979.
The primary system temperature
is about 120 0 F, and the small and decreasing amount of decay heat still being generated is being
lost to the building.
A new forced circulation system for the primary coolant, the mini-decay
heat-removal-system (MDHRS),
has been installed but is not in use because loss of heat to thebuilding has been shown to be adequate.
As long as the damaged fuel in the reactor core iscooled and remains relatively undisturbed and surrounded by boron-rich coolant, there isessentially no chance that the nuclear chain reaction, which was abruptly stopped at the time of
the accident, could start again.
But, the staff believes that as time passes, there will be anincreasing potential for failure of essential equipment.
Even though improbable, if the core
were accidentally to begin a chain reaction once more, radioactivity could be released to thereactor building.
The amount of radioactivity released during an accidental recriticality would
be much less than that released in the initial accident.
Even so,. timely removal of the damaged
fuel to safe storage is a paramount objective of the cleanup of TMI-2.
The feasibility of partial cleanup alternatives in which the reactor building would be sealedwith some or all of the radioactive sources left in place was examined.
It was found that all
such alternatives, including taking no action other than maintaining the reactor in safe shutdown
condition, either would not eliminate the potential risks or would convert part of the TMI-2 site
into a long-term or permanent waste repository.

The staff concluded that all of the cleanup operations must be performed whether TMI-2 isdecommissioned or refurbished to generate electricity in the future.
The environmental impacts
would be essentially the same regardless of whether the cleanup alternatives were chosen on theassumption that the plant would be decommissioned or on the assumption that the plant would be
refurbished and restarted.
The differences are less than the uncertainties in the best estimatesthat can be made for these impacts oh the basis of the information presently available.
Cleanup of the facility should proceed in a timely manner, not only to mitigate any risks to thephysical health of workers and nearby residents, but also to complete those activities which cancause psychological distress for residents in the area.
The sooner the cleanup process iscompleted, the sooner the sources of concern will cease to exist.
S.3

The Cleanup Operations

The cleanup comprises four fundamental activities:building and equipment decontamination; fuelremoval and decontamination of the primary coolant-system; treatment of radioactive liquids; andpackaging, handling, storage and transportation of radioactive wastes.
The current schedule for
conducting these activities is indicated in Figure 1.
Figure 2 illustrates how the wastes
resulting from the cleanup activities, and from the accident itself, would be separated foreventual packaging and disposal.
The removal of unwanted radioactive contamination from materials and equipment is a familiar androutine operation for reducing radiation levels.
Decontaminations of various types have been
conducted since the 1940s and a considerable amount of experience and technology is available.
These experiences illustrate that available techniques can be modified to suit the conditions at
TMI-2.
Applicable experience in removing damaged fuel and core components is limited, hence
development of specific techniques will be required.
Building and Equipment Decontamination
Cleanup of the AFHB started with the general areas where contamination was relatively light andis proceeding to rooms (cubicles) containing tanks and other equipment which are more heavily
contaminated.
The methods in use are not essentially different from those used in the process ofremoving dirt from any surface except that care must be taken to protect workers from radioactive
contamination, and more stringent methods must be used to remove most of the contamination.
The
methods used for decontaminating building and fixed equipment surfaces include washing with ahigh-pressure water jet, wet and dry vacuuming, and manual wiping.
Clean surfaces are often
protected by applying strippable coatings which are easily removed if the surface becomesrecontaminated.
Small demountable equipment items can be cleaned by electrochemical or
ultrasonic techniques.
The methods considered to be most practicable for removing the sludge
that accumulated in the sump, pipes, tanks and other vessels containing water involve
resuspension of the sludge in water by agitation to form a slurry which is pumped out andfiltered.
As of September 1, 1980, the amount of labor by workers directly involved in the AFHB decontamination effort was about 500,000 person-hours.
The average exposure rate for these workers
generally allowed normal shift operation.
Shielding is used to protect workers from the ambient
radiation fields, and only a fraction of a worker's time on the job is actually spent in theradiation field.
The staff estimates that a total of about 750,000 person-hours of work effort
will be needed to decontaminate the AFHB.
The methods used for decontaminating the reactor building will be similar to those used in the
AFHB, although the strategies will be different because the reactor building consists primarilyof large open spaces, while the AFHB is divided into many small cubicles.
Surface decontamination may be easier in the reactor building than in the AFHB because most of the reactor
building surfaces are painted, whereas most of the surfaces in the AFHB are untreated concrete.
The major tasks that must be coordinated and carried out are removal of the contaminated waterfrom the basement, removal of the sludge and debris, and removal of the plateout from thebuilding and equipment surfaces.
The staff estimates that decontamination of the reactor building (excluding any additional
decontamination that might be required in connection with decommissioning or refurbishing

yI7

1179

REACTOR

DECAY HEAT REMOVAL - STEAMING "A"GENERATORDECAY HEAT REMOVAL - LONG TERMAUXILIARY AND
FUEL HANDLINGBUILDING

INITIAL GENERAL AREA DECONTAMINATIONINDIVIDUAL AREA DECONTAMINATIONFINAL GENERAL AREA AND SUMP DECONTAMINATION

1/80

-

1/81

b-

-

1/82

* 6-4-

1/83

1/84

1/85

1/86

-

DECONTAMINATIONCONTAINMENTDECONTAMINATION

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CONTAINMENT ENTRY AND DATA ACQUISITIONCONTAINMENT DECONTAMINATION SUPPORT SYSTEMSCONTAINMENT DECONTAMINATION - GROSSCONTAINMENT DECONTAMINATION - MANUAL

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PREPARATION FOR RPV HEAD REMOVALREACTOR EXAMINATION
RPV HEAD AND UPPER INTERNALS REMOVALCORE INSPECTION
FUEL REMOVALREACTOR LOWER INTERNALS REMOVALREACTORDEFIJELING
AND RCS
DECONTAMINATION-INITIALRCS DECONTAMINATION<

CLEAN UP AUXILIARY BUILDING WATERCONTAINMENT PURGE
CLEAN UP CONTAINMENT SUMP WATER
CLEAN UP RC WATERPROCESSING OF
RADIOACTIVE
WASTE

I._________________RADIOACTIVE
SOLID WASTE
MANAGEMENT

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CLEAN UP CONTAINMENT DECONTAMINATION WATER____________________________________________

INTERIM LINERSTAGING
INTERIM WASTE STAGING,

EQUIPMENT AND MATERIAL STAGINGEPICOR-II LINER GENERATIONEPICOR-l LINER SHIPPINGSOLID RADWASTE SHIPPING

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CONSTRUCTION, PROCESSING ACTIVITIES-=-fACTIVITYCONTINUED AS NEEDEDMILESTONE: 'WACCIDENT,T CONTAINMENT PURGE, 7 INITIAL CONTAINMENT ENTRY,' 2 7 PEIS ISSUED, W7 START CONTAINMENTDECONTAMINATION, 'TRPV HEAD REMOVED, \'TFUEL REMOVED, TRCS DECONTAMINATION COMPLETE.
Figure 1.

Licensee's Planning Schedule for TMI-2 Cleanup,of support facilities not included.)Phase I and II.

(Construction

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vii

PRIMARY FORMS

Figure 2.

Characterization of TMI-2 Radioactive Waste.

SECONDARY FORMS

viii

operations) would require a work effort in the range of 300,000 to 900,000 person-hours.

This

includes both in-building and out-of-building time for workers with assignments that require
entry into the building; it does not include support work by others with assignments that do not
require entry.
Fuel Removal and Primary System Decontamination
The ultimate objective of the reactor defueling and primary system decontamination is to remove
all fuel, damaged reactor parts, and radioactive plateout in the coolant system.
Because theexact condition of the reactor core and some of the other reactor parts will not be known until
thorough inspections have been performed, it is impossible to predict or plan the defueling andcleanup operations in complete detail at this time.
However, the major steps and the order inwhich they will be conducted are reasonably certain.
The coolant system must first be connected to a cleanup system so that the coolant can be
decontaminated.
It also will be necessary to remove any additional radioactive materials
released to the water during fuel removal operations.
This could be done by continuously running
the water through a cleanup system consisting of filters and demineralizers.
Obtaining access to the fuel requires removal of the reactor vessel head and components above thefuel.
These tasks can be performed in much the same manner as during normal reactor refueling
operations, namely by direct contact methods.
If minimal warpage or mechanical damage occurred
during the accident, these operations should proceed with relative ease.
However, if warpage or'mechanical damage is extensive, considerable difficultycould be encountered in fuel removal
operations, and underwater cutting and machining operations might be needed.
The fuel will be moved under water to the spent fuel pool for interim storage.
assemblies may not be all in one piece.

Some of the fuelOperations necessary to remove the fuel assemblies are:Detailed inspection of the core.
Removal of loose debris.
Removal of fuel assemblies using special equipment.
During defueling, it is very important to maintain the boron concentration in the circulatingwater at the proper level in order to prevent reactor recriticality.
Since some of the water
treatment processes remove boron along with the radionuclides, boron may have to be addedto thewater during the defueling.
After the fuel has been removed, the support structure for the fuel must be removed.
Normally,
removal of the support structure is a straightforward procedure requiring hook up with the crane,
lifting it out of the pressure vessel, and moving it to the fuel transfer canal.
Because of thepossibility that overheating has caused distortion and warping of the support structure, removing
it may not be easy.
Accordingly, planning allows for the contingency of having to remove thisstructure by cutting it into smaller pieces while still under water.
After removal of all the fuel from the reactor pressure vessel (hence removing of any further
source of radioactivity which could recontaminate the system), the final step will be to cleanout the residual radioactivity from the system.
This would be accomplished by a method quite
analogous to flushing the cooling system of an automobile.
The staff assumes that decontamination of the reactor building will be largely completed before
fuel removal activities begin.
Thus, during these activities the contribution to worker exposure
from building background radiation should be small.
The major contribution to worker exposure
will be the general background of 2 to 3 mR/hr at the surface of the transfer canal water during
underwater operations to disassemble the reactor.
In addition, it will be necessary to work inradiation fields as high as 150 to 200 mR/hr in performing some of the hands-on activities.
The
staff has estimated the time-averaged field for a typical worker during the defueling activities
over a work shift to be 10 mR/hr. While some persons will be working in higher average fields
and others in lower fields, it is the staff's judgment that this average value is appropriate for
estimating radiation exposures for workers performing the defueling and primary cooling systemcleanup activities.
A total of about 100,000 to 300,000 person-hours of effort will be required,depending upon the conditions found during inspection.

ix

Treatment of Liquid Waste
Liquids involved in the TMI-2 decontamination will require further processing to permit theirsafe disposal in accordance with the staff's proposed use of the effluent criterion inl10 CFR 50
Appendix I as discussed in Section 1.6.3.2; these liquids include those directly generated during
the March 28, 1979, accident (accident water) as well as liquids contaminated during the cleanup
operations.
The accident-generated water in the reactor building sump and the primary system cannot be leftin its present condition and location if the cleanup effort is to proceed.
Some of the alternatives considered for disposition of this water involve its cleanup through the use of filtration,
Others include transfer from its present
ion exchange, evaporation and bitumenization techniques.
location to onsite storage facilities or processing the water for transport and disposal at a
low-level radioactive waste disposal facility.
As decontamination solutions are generated, they
too must be either cleaned up, stored, or processed and shipped offsite.
The alternatives
considered for accident-generated water and decontamination solutions are discussed below.
Long-term onsite storage of the unprocessed accident water involves transfer from its present
locations in the reactor building and primary system to storage tanks.
This water could betransferred to tanks within the reactor or auxiliary fuel handling buildings, if available, or to
newly constructed exterior tanks.
In either case, the storage tanks would have to be heavily
shielded to reduce radiation levels in areas near the tanks.
Storage of the accident water
onsite would defer cleanup and complicate the cleanup operation without contributing to its end
goals.
For these reasons, long-term onsite storage of unprocessed accident water is notconsidered a reasonable alternative.
Direct immobilization involves mixing unprocessed accident water with a binder materal such as
Portland cement or vinyl ester styrene for either temporary onsite storage or offsite shipment to
a commercial shallow land burial facility.
Immobilization of accident water with cement would
take about 5 years, produce about 7400 cubic yards of concrete, and require about 1900 shielded
shipments from the TMI-2 site.
This was not considered suitable for unprocessed accident water
but may be used for the relatively small quantitives of decontamination liquids.
Several processes and systems are available for treating the liquid to remove the contaminants.
The following processes were considered:(1) filtration, (2) ion exchange, (3) evaporation, and(4) bitumenization.
Filtration is applicable to TMI-2 liquid wastes as an initial step in a process.
It is not aappropriate treatment process by itself because much of the radioactivity is in solution and thus
is not removed by filtering.
Ion exchange, the same process used in household water softeners, involves the removal of ionicspecies from an aqueous phase.
The ion-exchange media considered for use at TMI-2 include
inorganic zeolites and other minerals and organic resins.
Ion exchange is appropriate for
accident water and some decontamination solutions.
It is not appropriate for treatment ofchemical decontamination solutions because the chemical nature of these liquids would lead to
rapid breakdown and plugging of ion-exchange media.
Evaporation would separate the water from the non-volatile radionuclides and other impurities
dissolved in the liquid waste.
Most of the contaminants are retained in the concentrated
solution (or bottoms) while the relatively clean water vapor is condensed to liquid which
requires further processing in an ion-exchange system.
Additionally, the concentrated solutions
would have to be immobilized in a solidification system.
Evaporation is only appropriate fortreatment of TMI-2 liquid wastes with low to moderate concentrations of dissolved solids and low
radionuclide concentrations.
Bitumenization combines evaporation and immobilization in one step.
The radionuclides areimmobilized in an asphalt-like material (bitumen) and the vaporized water removed from the liquid
waste is condensed for further treatment in an ion-exchange system.
Bitumenization is only
appropriate for TMI-2 liquids of low to moderate radioactivity concentration with at least
5 percent solids content by weight.

x

The basic processes described above were combined for purposes of analysis into nine different
treatment systems.
The zeolite-based ion-exchange systems are best suited for cleanup of
accident water and water used for decontamination which would contain a high concentration ofradionuclides.
Evaporator/resin and bitumen/resin systems are considered suitable only forcleanup of chemical decontamination solutions due to the complex chemical nature of theseliquids, their solids content, and moderate radionuclide concentrations.
There are no practicalsystems for removal of the tritium present in the accident water.
Ten alternatives were evaluated for disposition of processed accident water containing tritium.
These alternatives consisted of:Long-term onsite storage either as bulk liquid or as cement slabs.
Since it would takeabout 200 years for the contained radioactivity to decay to innocuous levels,and storage
merely defers ultimate disposition, these alternatives were not considered suitable.
Onsite disposal of cement slabs by placement in shallow land bur.ial trenches or injection ofliquid water deep underground.
These alternatives would present novel technical andregulatory issues and would also convert the site to a permanent repository.
For thesereasons, onsite disposal was not considered suitable.
Shipment offsite for disposal in a shallow land burial facility, injection deep undergroundor disposal in the ocean.
There are no regulatory obstacles to transport of bulk liquids
containing low-level radioactivity but special permits would have to be obtained to disposeof the liquid in a deep injection well or the ocean.
Immobilization of the liquid forshallow land burial is within the scope of current regulations but states may prevent theuse of burial sites, due to space limitations, for such a large quantity of material withsuch a low level of radioactivity.
Controlled release as liquid to the Susquehanna River, or as a vapor to the atmosphere or
simultaneous releases to the river and to the atmosphere.
Processed water can be diluted
with available cooling tower blowdown water and discharged to the river at rates controlled
to satisfy 10 CFR Part 20 and 10 CFR 50 Appendix I criteria.
Placement of the water in a
lined pond on the Island where it would evaporate would reduce the tritium concentration tolevels approaching background in the river over a relatively short time (3-5 years); theremaining water could be retained in the pond for the lifetime of the plant.
The water
could also be injected into the mechanical draft cooling tower where some of the water would
be vaporized and released to the atmosphere and the rest released to the river via thecooling tower blowdown.
Packaging,

Handling,

Storage, Transportation and Disposal of Radioactive Wastes and Fuel

The wastes resulting from the accident and from decontamination activities are not all in a form
acceptable for onsite storage or offsite disposal.
It will therefore be necessary to treat someof these wastes.
The treatment alternatives considered depend on the physical form of thematerial.
Combustible trash can be incinerated to reduce volume by factors of 80 to 100, but
incineration leads to the generation of ash that must be immobilized prior to disposal.
Theaddition of an immobilizing agent results in a net effective volume reduction of 40 to 50.
Noncombustible trash can be compacted to reduce volumes by a factor of about 5.
Contaminatedequipment and hardware can be disassembled and mechanically sectioned for volume reduction.
Treatment of contaminated liquids by filtration, ion-exchange, and evaporation leads to thegeneration of spent filter cartridge assemblies, contaminated ion-exchange media, and evaporator
bottoms or sludges.
At present, filters, ion-exchange media and sludges are accepted in the
dewatered condition at some commercial low-level waste disposal facilities, but some others
require that these materials be in.a solidified form.
Evaporator bottoms also must be solidified
for disposal at all sites. An NRC order currently requires that contaminated ion-exchange
resins from EPICOR II be solidified before shipment to disposal facilities.
However, there areother alternatives to solidification that immobilize the waste, such as shipping the high-specificactivity portion of these and other wastes generated in the future to Department of Energy
facilities for processing and disposal.
The wastes will be packaged prior to onsite storage or shipment to a waste disposal facility.
The disposable containers considered for packaging include 55-gallon drums, wooden boxes forxi

low-activity material, and large steel containers with capacities as great as 200 ft 3 . Special
casks are required for fuel assemblies and could be required for certain spent filter cartridges,
expended ion-exchange media, and sludges.
The package handling techniques depend on the radiation level, size, and weight of the disposable container.
Low-activity materials in drums orwooden boxes are handled manually and are transferred within the facility using forklifts,
motorized pallets, or other package-handling equipment.
For higher-activity disposable
containers, semiremote and remote handling systems are used so as to reduce radiation exposure topersonnel.
In general, all waste will be transferred to an onsite staging/storage facility prior to shipment
offsite.
These facilities may be shielded enclosures.
The facilities are divided by container
type and surface radiation level.
Shielded packages will be used for shipping spent fuel, contaminated hardware, certain spent
filter cartridges, expended ion-exchange media, and other higher radiation level wastes.
The
number of shielded waste shipments is estimated to be between 220 and 690, and the number ofunshielded shipments is estimated to range from 130 to 310.
The range reflects the uncertainties
inherent in the estimates as a result of the fact that the actual volume of waste generated will
depend on which decontamination and waste treatment alternatives are implemented.
Disposal options considered include:
Disposal of high-level wastes, high-specific-activity
special offsite waste repository,
Disposal of the lower
burial; and,wastes and transuranic wastes in arange of the high-specific-activity

wastes by intermediate

depth

Disposal of the low-activity wastes in shallow land burial.
Storage of irradiated fuel either onsite or offsite pending final disposal or reprocessing.
S.4

Environmental Impacts of the Cleanup

The principal environmental impacts that can be expected to occur as a consequence of the cleanup
activities at TMI-2 are indicated below.
Occupational Doses and Health Effects
Decontamination workers at the plant will receive a total cumulative radiation dose estimated at
between 2000 and 8000 person-rem for the whole cleanup program.
Using the NRC staff's healtheffect risk estimators, the health effect estimates corresponding to these doses range from 0.3
to 1 additional deaths among these workers due to cancer and from 0.5 to 2 additional genetic
effects among their descendents.
A summary of the minimum and maximum estimates by the staff isgiven below in Table S-1.
These ranges are broad because of uncertainties of the plant conditions
and the amount of work that will be needed to decontaminate the reactor building and its contents.
The occupational dose to each worker will be limited to 3 rem/quarter in accordance with 10 CFRPart 20; however, the exact dose to any one individual cannot be predicted because work
assignments have not been made.
Offsite Doses and Health Effects from Projected Cleanup Activities
The total-body dose to the individual that may receive the maximum exposure offsite from gaseousand liquid releases from the cleanup operations during any year will not exceed about 15 mrem(Appendix R).
The increased risk that this dose would cause a fatal cancer in the individual who
received it is about one chance in 100,000.
The increased chance of that dose causing genetic
These
effects to offspring of the maximum exposed individual is about one chance in 20,000.
risks are small compared to the current normal incidence rates for fatal cancers and genetic
effects among the poulation.
Public health statistics indicate that in the United States, one
person in five probably will die of cancer and that the normal occurrence of hereditary disease
in offspring is about 1 in 17.
For the general population within 50 miles of the plant, thexiiTable S-I. Estimated Occupational Doses and Resulting Health
Effects as a Result of TMI-2 Cleanup Operations

Major Cleanup Operation

Cumulative
Occupational
Dose(person-rem)
Minimum Maximum

Building and equipment
decontamination

1,000

3,600

0.13

0.5

0.3

0.9

Fuel removal and primarysystem decontamination

900

4,100

0.1

0.5

0.2

1.1

Treatment of radioactive
liquids

40

120

0.006

0.016

0.01

0.03

Packaging, handling, storage
and transportation ofradioactive wastes

75

520

0.009

0.07

0.02

0.14

2,000

8,000

0.3

1

0.5

2

TotalaPotential
Fatal Cancers
Minimum Maximum

Potential
Genetic Effects
in Offspring
Minimum Maximum

aTotals are rounded to one significant digit.

total cumulative dose from expected releases would be in the range of 10 to 30 person-rem; thisis less than 0.01 percent of the 255,000 person-rem annual radiation dose to the same populationfrom background.
An individual onlooker who spent three minutes at an average distance of 3 ft from a truck loadedwith radioactive waste might receive a dose of up to 1.3 mrem. The probability that this dosewould cause an increased chance in fatal cancer in the individual who received the dose is about 1
in 1 million. The added probability of genetic effects to offspring of the exposed individual isabout 1 in 200,000. The estimated 700,000 persons who reside along the longest (2750-mile) routefrom TMI to the farthest disposal site might receive a cumulative population dose within therange of 20 to 50 person-rem for all TMI decontamination waste and fuel shipments.
The potential for effects on aquatic organisms in the Susquehanna River and in Chesapeake Bay has
been evaluated for both controlled and accidental discharges of processed and unprocessed water
to the river. Low concentrations- of Cs-137 could persist in sediments in both the river and the
Bay for some years following an accidental discharge of unprocessed water from TMI-2, but thelevels would be so low as to have no radiation effects on aquatic species or on man and would not
be detectable in sediments beyond the immediate site vicinity. For controlled releases ofprocessed water there would be no significant effects on man or aquatic species.
Postulated Accidents
The accidents considered possible by the staff range from the more likely but low consequence
failure of a HEPA filter to the extremely unlikely occurrence of a leak in the reactor vesselthat leads to uncovering and overheating of the core with the reactor vessel head removed.
Building ventilation systems are equipped with two or more HEPA filters in series to remove small
particles of dust from the air being exhausted to the atmosphere. Failure of a HEPA filter would
allow some of the radioactive dust to escape until the failure was detected and the ventilation
system shut down.
An accident of this nature in the AFHB could result in a total-body dose of
0.0084 mrem to the maximum exposed individual offsite.

xiii

Although extremely unlikely, the unprocessed reactor building sump water might leak out of thereactor building.
In the improbable event that all the water reaches the river, the concentration of radionuclides would be orders of magnitude below maximum permissible concentrations
for unrestricted areas.
Percolation through the ground to the river is calculated to take morethan one year, and monitoring wells around the reactor building will provide early indication ofgroundwater contamination, thus giving sufficient time to take action to mitigate the consequences.
Only tritium would reach the river within 1 year, whereas other isotopes, such as Sr-90 andCs-137, would be greatly delayed because of adsorption onto the soil.
While the staff could not hypothesize a credible mechanism for a leak in the reactor vessel, theconsequences of such an occurrence were evaluated.
Since from many hours to several days would
be available to take corrective action should loss of cooling be developing, the leakage from thereactor vessel into the reactor building could be limited to a small fraction of the current
inventory of radioactivity.
The worst conceivable situation would occur if all the Kr-85(45,000 Ci) and most of the cesium (470,000 Ci) still in the core were released to the reactor
building.
The resulting activity levels inside the reactor building would then be comparable to,or less than, the levels inside the reactor building following the accident on March 28, 1979.
The releases and offsite doses for this incredible bounding event would be comparable to thosewhich have been estimated for cleanup of the reactor building following the accident (i.e.,
overall releases and offsite doses for the reactor building cleanup would be doubled).
Potential Releases Caused by External Events
The potential for the release of radioactive materials as a result of flood, tornado, or aircraft
impact was evaluated.
A flood somewhat la'rger than the Hurricane Agnes flood of June 1972 would
overtop the levee surrounding the plant, but waste storage facilities are protected by thickconcrete, and it is not very likely that they would be breached.
Nor are these facilities likely
to be affected by tornado or impact of an aircraft; however, the exposed processed water storage
tanks would be vulnerable if these low-probability events occurred.
Release of this watercontaining less than 0.6 pCi/mL of tritium would not present a public health hazard if the waterreached the Susquehanna River because maximum permissible concentration limits (10 CFR Part 20)
are not likely to be exceeded at the nearest public drinking water supply.
Psychological-Socioeconomic

Effects

The level of psychological distress among some members of the communities surrounding the plantincreased immediately following the accident but had considerably diminished by midsummer 1979.
Low levels of distress will probably continue during the cleanup process, but no long-term
psychological effects on the great majority of the community are predicted.
Nevertheless, thelong-term nature of the cleanup program presents the potential for increasing psychologicaldistress for some people; consequently, completing the cleanup as expeditiously as safety will
permit is desirable.
Social impacts during the cleanup could include possible resistance to consumption of agricultural and fishery products that the public may think are radioactively contaminated.
Those who
make all or part of their living from seafood taken from Chesapeake Bay or those involved inagricultural production are likely to be affected to the largest degree; but losses, if any,
should be of short duration.
The disposition of processed accident water is of concern to many individuals.
The potential
economic impact on Chesapeake Bay activities (sea food industry and recreational uses) if theprocessed accident water were diluted and a controlled release made to the Susquehanna River is amatter of special concern.
The State of Maryland is undertaking a study of the possible economic
impacts on the marketing of seafood from the Chesapeake Bay if the public were to perceive ahealth hazard from controlled release of the processed accident water.
The NRC staff is of theopinion that until such a study is completed, no decision should be made regarding the dispositionof the processed accident water (unless an emergency arises which requires early disposition).
Adequate storage capacity exists for interim retention of the processed accident water onsite.
Although the number of truck shipments necessary to carry solid radioactive wastes to disposalsites will be large (ranging from 350 to 1000), the shipments will be made over a long period oftime and reduction of the marketability of residential property near the route through Middletown,
Pennsylvania, should be temporary.

xivRadiological Environmental Monitoring Program
Monitoring around the TMI site and in nearby communities during decontamination of TMI is being
performed by (1) the U.S. Environmental Protection Agency as the lead federal agency, (2) the
Commonwealth of Pennsylvania, (3) the U.S. Department of Energy, (4) the Nuclear Regulatory
Commission, (5) the State of Maryland, and6(6) Metropolitan Edison Company (the licensee).
In addition to their own direct monitoring, the Department of Energy and the Commonwealth ofPennsylvania have sponsored a Community Radiation Monitoring Program that involved people in 12
communities within about five miles of TMI.
The very comprehensive radiological monitoring program provided by the cooperation of the aboveorganizations consists of direct and indirect measurements of exposure rates and sampling of all
appropriate media at numerous locations in the offsite area within 21 miles of TMI-2.
Exposurerate measurements are made using recording and nonrecording rate meters and thermoluminescentdosimeters (TLDs).
Samples of air, soil, vegetation, milk, fish, aquatic plants, sediments, andwater are collected and analyzed for specific radionuclides and for gross beta and gamma emissions.
The results of all monitoring programs are reported to the EPA, which is responsible for coordinating offsite monitoring around TMI-2 and for compilation and dissemination of the resulting
data to the public.
Economic Costs for Cleanup of TMI-2
Estimates of the relative cost of alternatives considered for each phase of the cleanup of TMI-2
have been developed, excluding costs not directly associated with the actual decontamination
work.
Examples of the costs not included are the costs of the support activities essential to
the whole process, cost of replacement power, interest charges and inflation.
The estimates also
are in constant 1980 dollars and do not include the effect of inflation. These cost estimates
were made for the purpose of comparing the various alternatives that were considered feasible.
A
total cost estimate for the cleanup of TMI-2 has not been made by the staff.

PREFACE

The NRC staff appreciates the many helpful comments received from thepublic and government agencies on the draft Programmatic Environmental
Impact Statement (PEIS)
comment.

which was issued in August 1980 for public

Our efforts to be responsive to these comments have resulted

in a variety of changes in the PEIS,

particularly in Chapters 5 through8 which some persons found confusing.

While the organization of thosechapters in the draft statement enabled the reader to follow the expectedchronologicalsequence of the cleanup activities, italso had the dis-advantage of scattering information on particular subjects,

such asprocessing contaminated water from several plant locations,through various

parts of the document.

We have therefore reorganized the material in thesechapters so that discussions of similar activities in this final PEIS aregrouped together.
To further aid the reader we have simplified some of the technical descriptions,

updated information,and included additional

illustrations.

Relative

economic cost estimates of the alternative methods of performing the cleanupof TMI-2 have also been included,
issued.

as promised when the draft statement was

However, we do not regard the addition of cost information (or theother modifications mentioned above) as a substantial change in the content
of the document.

Our primary responsibility is

to ensure that the cleanup

activities are conducted in accordance with NRC's mandate to ensure thehealth and safety of the public and to protect the environment.

Bernard J. Snyder, Program Director
Three Mile Island Program Office
Office of Nuclear Reactor Regulation

XV

xvi

CONTENTSPage
SUMMARY .........................
.........................................
i
PREFACE . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xv
TABLE OF CONTENTS ....................
....................................
.. xvi
LIST OF TABLES .............................
..................................
xxii
LIST OF FIGURES ...........................
....................................
xxxii
FOREWORD ......................
........
................................
.. xxxv
INTRODUCTION .........
.......
....................
. .....
.............
1.1 Purpose and Scope ......................
................................
1.2 History of the Plant and Its Present Status ....
...................
1.3 Summary of Licensee's Objectives, Proposed Actions, and Schedule .....
........
1.4 Public Concerns..................................
1.5 Decontamination Experience at Other Nuclear Facilities.............
1.5.1 Experience with Initial Entries into Highly Contaminated Facilities ........
1.5.2 Experience with Decontamination of Building Interior Surfaces andEquipment .................
.....................
1.5.3 Experience with Removal of Damaged Cores .........
..................
...
1.5.4 Experience with Decontamination of Reactor Coolant Systems ... .........
...
1.5.5 Environmental Impact ................
............................
...
1.5.6 Comparison with TMI-2 ...............
............................
..
1.6 Regulatory Requirements, Other Constraints, and Future Criteria ............
...
1.6.1 Regulatory Requirements .................
...........................
1.6.2 Other Constraints ......
.......................
. .. ..........
..
1.6.3 Future Criteria .....................
...............................
References--Section 1 ....................
.................................
...

1-1
1-1
1-2
1-6
1-7
1-9
1-10

2.

2-1
2-2
2-3
2-4

1.

MAJOR DECONTAMINATION AND WASTE DISPOSAL ALTERNATIVES .......
...............
..
2.1 Decontamination of the Facility .............
.........................
..
2.1.1 No Action--Maintain Safe Shutdown .............
......................
2.1.2 Decontamination of Buildings and the Assorted Support Facilities ..... .....
2.1.3 Reactor Core Removal and Decontamination of the Nuclear Steam Supply
System .....................
...................................
...
2.2 Decommissioning..................................
2.2.1 Rationale of Initial Cleanup Activities .............. .............
...
2.2.2 Decommissioning Alternatives ..............................
2.2.3 Waste Volumes Generated during Decommissioning .......
...............
..
2.2.4 Effluents and Releases to the Environment ............
..................
2.2.5 Environmental Impacts ...............
............................
..
2.2.6 Decommissioning Costs ...............
............................
..
2.3 Disposal of Radioactive Waste .............
........
.................
...
2.3.1 Waste Characterization ................
...........................
...
2.3.2 Classification...............................
2.3.3 Alternatives for Disposition of Radioactive Waste .....
..............
..
References--Section 2 ....................
.................................
...
3.

THE ENVIRONMENT AND POPULATION WHICH MAY BE AFFECTED ......
................
3.1 Location of the Site and Access Facilities ...........
...................
3.2 Geology .......
....................
..............................
3.3 Meteorology .....................................
3.3.1 Regional Climatology .
.. .
......................
.-3.3.2 Local Meteorology .................
..............................
3.4 Hydrology and Water Use . ....................
;...........................
3.4.1 Surface Water Hydrology .............
...........................
3.4.2 Surface Water Uses . . .................
.
..........................
3.4.3 Surface Water Quality .............
................
...............
3.4.4 Groundwater Hydrology ...............
............................

..
...
..
.

1-12
1-13
1-14
1-15
1-16
1-17
1-17
1-23
1-24
1-28

2-8
2-10
2-12
2-13
2-15
2-15
2-15
2-18
2-18
2-19
2-23
2-25
2-32

3-1
3-1
3-1
3-5
3-5
.. 3-6376... 3-6... 3-12
3-14
..
3-16

xvii

Page

CONTENTS (continued)
3.5 Ecology .................
...........................
3.5.1 Aquatic .............
.........................
3.5.2 Terrestrial ........
.......................
3.6 Population, Land Use, and Other Socioeconomic Considerations
3.6.1 Population Groups and Geographic Areas of Interest . . .
3.6.2 Socioeconomic Status of the Region ...........
3.6.3 Psychological Setting ........
...................
3.6.4 Susquehanna River/Chesapeake Bay
References--Section 3 . . . . . . . . . . . . . . . . . . . . . . .

3-17
3-17
.3-18
3-20
3-20
3-21
3-24
3-25
3-32

4.

MAINTENANCE OF THE REACTOR IN SAFE CONDITION .... ...........
4.1 Objectives and Actions
4.2 Monitoring of Reactor and Reactor Building .........
. . . . . . . . . . . . . . . . . . . . .
4.3 Decay Heat Removal
4.3.1 Decay-Heat Removal Mechanisms .....
...............
4.3.2 Available Decay-Heat-Removal Systems ..........
4.4 Effluents and Releases to the Environment .... ...........
4.4.1 Normal Releases ...
.....................
4.4.2 Accident Scenarios and Associated Releases.......
4.5 Environmental Impacts ..........
.... ................
4.5.1 Occupational Doses . . . . . . . . . . . . . . . . . . .
4.5.2 Offsite Doses ....
......................
4.5.3 Postulated Accident Effects ..................
4.5.4 Psychological-Socioeconomic Effects .... ............
References--Section 4 . . . . . . . . . . . . . . . . . . . . . . .

4-1
4-1
4-1
4-1
4-2
4-2
4-3
4-3
4-5
4-8
4-8
4-8
4-9
4-9
4-9

5.

5-1
5-4
5-4
5-6
5-8
5-10
5-13
5-19
5-19
5-20
5-27
5-33
5-42
5-45
5-50
5-50

BUILDING AND EQUIPMENT DECONTAMINATION ....................
5.1 Auxiliary and Fuel Handling Buildings ..................
5.1.1 Status and Specific Considerations ...........
5.1.2 Methods Used and Alternatives Considered .....
..
5.1.3 Details of Feasible Methods and Facilities.......
5.1.4 Effluents and Releases to the Environment .... ........
5.1.5 Environmental Impacts .........
.......
...........
5.1.6 Economic Costs . . . . . I. . . .. . . . . . . . . ...
5.2 Reactor Building . . . . . . . . . . . . . . . . . . . . . .
5.2.1 Status and Specific Considerations...........
5.2.2 Alternative Methods Considered .............
5.2.3 Details of Methods and Facilities .... .............
5.2.4 Effluents and Releases to the Environment .... ........
5.2.5 Environmental Impacts ..........
..................
5.2.6 Economic Costs . . . . . . . . . . . . . . . . . . . . .
References--Section 5 . . . . . . . . . . . . . . . . . . . . . . .
6.

REACTOR DEFUELING AND PRIMARY SYSTEM DECONTAMINATION .........
6.1 Handling of Primary Coolant System Water ..........
6.1.1 Preparation for Reactor Head Removal..........
6.1.2 Removal of RPV Head and Internals . . .
. .. ..........
6.1.3 Reactor Defueling ..........
....................
6.1.4 Primary System Flush and Drain .............
6.1.5 Decontamination of the Primary System Components . ...
6.2 Reactor Coolant System Inspection .... .....
.....
.....
...........
6.2.1 Status and Specific Considerations
6.2.2 Alternative Methods Considered ..... ...............
6.2.3 Details of Methods and Facilities .... .............
6.2.4 Effluents and Releases to the Environment .... ........
6.2.5 Environmental Impacts ..........
..................
6.2.6 Economic Costs . . . . . . . . . . . . . . . . . . .. .
6.3 Removal of the RPV Head and Internals . . .
. ..
..........
6.3.1 Status and Specific Considerations ...........
6.3.2 Alternative Methods Considered
6.3.3 Details of Alternative Methods and Facilities .........

. . . .
. . . .

6-1
6-1

. . . .

6-1

.
.
.
.

6-1
6-2
6-2
6-2

.
.
.
.

.
.
.
.

.
.
.
.

. . . .

6-2

. . . .
. . . .
. . . .

6-2
6-3
6-3

. . . .

6-5
6-5

. . . .

6-6

. . . .
. . .. ..

6-6

6-6
6-6
6-8

xviii

CONTENTS (continued)
6.3.4 Effluents and Releases to the Environment ............
..................
6.3.5 Environmental Impacts ...................
............................
6.3.6 Economic Costs .........
............
........................
6.4 Core Examination and Defueling ................
................
........
6.4.1 Status and Specific Considerations ...........
.....................
6.4.2 Alternative Methods Considered . ............
......................
6.4.3 Details of Methods and Facilities ..............
.......................
6.4.4 Effluents and Releases to the Environment ............
..................
6.4.5 Environmental Impacts ...........
..........
....................
6.4.6 Economic Costs ......................
...............................
6.5 Decontamination of Primary System Components .............
..................
6.5.1 Status and Specific Considerations. . ..........
.
..................
6.5.2 Alternative Methods Considered ................
.......... ...........
6.5.3 Details of Methods and Facilities ..............
......................
6.5.4 Effluents and Releases to the Environment ............
..................
6.5.5 Environmental Impacts .....
..............
....................
6.5.6 Economic Costs ......................
...............................
References--Section 6 ....................
.................................

Page
6-13
6-14
.
6-15
.6-18
... 6-18.. 6-18
6-19
6-23
..
6-24
6-25
6-27
.. 6-27
6-28
6-29
6-34
... 6-38
6-41... 6-43

7.

7-1
7-1
7-2
7-8
7-12
7-28
7-31
7-42
7-43
7-43
7-45
7-46
7-51
7-55
7-60
7-81
7-82
7-84

8.

SOLID WASTE PACKAGING AND HANDLING ..............
.........................
.. 8-1
8.1 Process Solid Wastes .....
..............................
8-1
8.1.1 Status and Specific Considerations ...........
.....................
... 8-1
8.1.2 Alternative Methods Considered ...........
.......................
..
8-4
8.1.3 Details of Methods and Facilities ..........
......................
... 8-14
8.1.4 Effluents and Releases to the Environment ............
..........
.....
.. 8-218.1.5 Environmental Impacts ...............
............................
..
8-26
8.1.6 Economic Costs ......................
................................
8-40
8.2 Chemical Decontamination Solutions .............
.......................
... 8-40
8.2.1 Status and Specific Considerations ...........
.....................
... 8-40
8.2.2 Alternative Methods Considered .............
.......................
... 8-40
8.2.3 Details of Methods and Facilities ..............
......................
8-43
8.2.4 Effluents and Releases to the Environment ....... . ... ...............
8-43
8.2.5 Environmental Impacts ...
...............
........... ...........
8-45
8.2.6 Economic Costs .......................
........................
.....
8-47
8.3 Solid Materials ..........
................................
.........
8-48
8.3.1 Status and Specific Considerations ...........
......................
.. 8-48
8.3.2 Alternative Methods Considered .............
.......................
... 8-51
8.3.3 Details of Methods and Facilities ......
........................
...
8-55
8.3.4 Effluents and Releases to the Environment ............
..................
8-60
8.3.5 Environmental Impacts ..............
.............................
... 8-63
8.3.6 Economic Costs ......................
...............................
8-66

WATER TREATMENT PROCESSES AND PROCESSED WATER DISPOSAL .....
...............
..
7.1 Liquid Waste Treatment ..............
.......................
..........
7.1.1 Status and Specific Considerations ...........
.....................
...
7.1.2 Alternatives Considered....................
.....
7.1.3 Details of Liquid Waste Treatment Alternatives .......
...............
...
7.1.4 Effluents and Releases to the Environment ............
..................
7.1.5 Environmental Impacts ...............
............................
..
7.1.6 Liquid Processing Costs .................
...........................
7.1.7 Comparison of Liquid Waste Treatment Alternatives .....
..............
...
7.2 Disposal of Processed Water .................
...........................
7.2.1 Current Status and Projected Water Disposal Requirements .... ..........
..
7.2.2 Alternative Methods Considered .............
.......................
...
7.2.3 Details of Alternative Methods .............
...................
.... ..
7.2.4 Effluents and Releases to the Environment ............
..................
7.2.5 Environmental Impacts ...............
........................
....
.
7.2.6 Liquid Disposal Costs ...............
............................
..
7.2.7 Comparison of Processed Water Disposal Alternatives ..... .............
...
References--Section 7 ..............
..................................
......

xix

Page

CONTENTS (continued)
8.4 Fuel Assemblies and Core Debris .............
.........................
...
8.4.1 Status and Specific Considerations ..............
.....................
8.4.2 Alternative Methods Considered ...............
.....................
...
8.4.3 Details of Methods and Facilities ..........
......................
...
8.4.4 Effluents and Releases to the Environment ...............
.............
8.4.5 Environmental Impacts ...................
.... ........................
8.4.6 Economic Costs .......................................
References--Section 8 .......
........
.............
............
. .. ........

8-75
8-75
8-76
8-77
8-78
8-78
8-78
8-78

9.

9-1
9-1
9-1
9-1
9-5
9-14
9-14
9-16
9-17
9-18
9-18
9-26
9-26
9-27
9-27
9-27
9-32
9-36
9-36
9-37

STORAGE, TRANSPORTATION, AND DISPOSAL OF FUEL AND SOLID WASTE ....
...........
9.1 Current Status and Applicable Constraints ..............
....................
9.1.1 Waste Management Activities to Date .............
.....................
9.1.2 Waste Package and Shipment Parameter ..............
....................
9.1.3 Regulatory and Technical Constraints ..............
....................
9.2 Alternative Methods Considered ..............
.........................
9.2.1 Storage ............................ ..........................
9.2.2 Disposal .......
....................... .......................
9.2.3 Transportation ..................................
9.3 Details of Methods and Facilities ............
........................
9.3.1 Storage .....................
...................................
9.3.2 Disposal ....................
...................................
9.3.3. Transportation ..................
...............................
9.4 Effluents and Releases to the Environment ..........
........
...........
9.5 Environmental Impacts .................
..............................
9.5.1 Normal Conditions .................
..............................
9.5.2 Accident Effects .......
........
..............
..
.. ...........
9.5.3 Psychological-Socioeconomic Effects ..........
.....................
9.6 Economic Costs ....................
.................................
References--Section 9 .... ........
. .........
...........
..................

..

..
...
...
...
...
...
...
...
..
..
...
..
...

10.
SUMMARY OF ENVIRONMENTAL IMPACTS OF THE PROPOSED ACTIVITIES, .
.........
10.1 Summary of Effluents and Releases to the Environment .....
..............
...
10.2 Summary of Occupational Doses and Health Effects ......
................
..
10.2.1 Occupational Dose.
.....................
.........................
...
10.2.2 Duration of Effort .............
..............................
..
10.2.3 Health Effects ...................................................
/...
10.3 Offsite Doses and Health Effects from Normal Operation ........
.............
10.3.1 Dose Commitments from Radioactive Releases to the Atmosphere ...........
...
10.3.2 Dose Commitments from Radioactive Liquid Releases to the Hydrosphere . . .
10.3.3 Transportation of Radioactive Material ............
...................
10.3.4 Radiological Impact on Man ..............
.........................
...
10.4 Doses from Postulated Accidents ....
...................
....
10.4.1 Dose Commitments from Accidental Radioactive Releases to the Atmosphere10.4.2 Dose Commitments from Accidental Radioactive Releases to theHydrosphere .....
......................
..........................
10.5 Potential Releases Due to External Events ...........
...................
...
10.5.1 Potential Releases due to Flooding ..........
.....................
...
10.5.2 Potential Releases Due to Tornado ..........................
10.5.3 Potential for Releases Due to Aircraft Impact ..............
... .......
.10.6 Psychological-Socioeconomic Effects .. . ............... ...................
10.7 Irreversible and Irretrievable Commitments of Resources ..... ............
...
10.7.1 Commitments Considered .................
...........................
10.7.2 Material Resources ...............
.............................
...
10.7.3 Water and Air Resources ..................
...........................
10.7.4 Land Resources ...
.................................
10.8 Summary of Economic Costs of the TMI-2 Cleanup ......
.................
...
References--Section 10 ...................
.................................
...

10-1
10-1
10-5
10-5
10-5
10-9
10-10
10-10
10-14
10-15
10-16
10-20
10-20

11.
ENVIRONMENTAL RADIOLOGICAL MONITORING ...............
.......................
11.1 Introduction ...................................................
11.2 Metropolitan Edison Company Radiological Monitoring Program .... ..........

11-1
11-1
11-1

..

10-23
10-25
10-25
10-27
10-28
10-29
10-30
10-30
10-31
10-31
10-31
10-31
10-32

XX

CONTENTS (continued)

Page

11.3 U.S. Environmental Protection Agency Radiological Monitoring Program .... ......
11-1... 11-3
11.4 Commonwealth of Pennsylvania Radiological Monitoring Program ..............
11-4
11.5 U.S. Department of Energy Radiological Monitoring Program .......... .......
11.5.1 Radiological Monitoring Program ............
......................
. .11-4
11.5.2 Atmospheric Release Advisory Capacity .........
................
. . ..
11-4
11.5.3 Community Monitoring Program ............
.........................
... 11-4
11.6 U.S. Nuclear Regulatory Commission Radiological Monitoring Program .........
... 11-4
11.7 State of Maryland Radiological Monitoring Program .......
...............
.. 11-5
11.8 Contingency Surveillance Procedures ..............
.......................
.11-5
11.9 Reporting Procedures .................
..............................
.. 11-5
11.9.1 Immediate Reporting Procedures ............
.......................
... 11-6
11.9.2 Reporting Data into the Data Base ...........
.....................
... 11-6
11.9.3 Reporting Information to the News Media .....
..............
11-6
11.10 Monitoring of Aquatic Ecology and Recreational Fisheries of theSusquehanna River .... ...................
. .........
................
11-6
12.

CONCLUSIONS ...... ............................

....................

....

.12-1

13.

DISCUSSION OF COMMENTS ON THE DRAFT PROGRAMMATIC ENVIRONMENTAL
IMPACT STATEMENT ................................................
.......
13.1 General ...............
...................................
. . .... .
13.2 Major Programmatic Alternatives ..............
........................
...
13.3 TMI Vicinity and Downstream Area ............
........................
...
13.4 Reactor Safety Prior to Defueling ............
.......................
...
13.5 Decontamination of Buildings and Equipment ........
....................
...
13.6 Reactor Defueling and Primary System Cleanup . . ...........
................
13.7 Water Processing and Processed Water Disposal .........
.................
...
13.8 Solid Waste Management ...............
.............................
...
13.9 Radiological Effects .....
.....................
.........
13.10 Accidents and Severe Natural Phenomena ...........
....................
..
13.11 Socioeconomic and Psychological Effects ..........
....................
...
13.12 Financial Aspects of TMI-2 Cleanup ................
......................

13-1
13-2
13-15
13-18
13-22
13-24
13-27
13-31
13-36
13-55
13-80
13-83
13-85
GL-l

GLOSSARY
APPENDICES
Appendix A. Comments on the Draft Programmatic Environmental
Impact Statement ....................
.................................

... A-1

Appendix B. Commission's Statement of Policy and Notice of Intent to
Prepare a Programmatic Environmental Impact Statement .....
...............

..

Appendix C. "Final Environmental Assessment for Decontamination of the Three
Mile Island Unit 2 Reactor Building Atmosphere, Final NRC Staff Report,"
U.S. Nuclear Regulatory Commission, NUREG-0662, May 1980 .....
.............

... C-1

Appendix 0. "Environmental Assessment for Use of EPICOR-II at Three Mile Island,
Unit 2," U.S. Nuclear Regulatory Commission, NUREG-0591, October 3, 1979 .........
Appendix E. Fish and Fisheries of York Haven Pond and Conowingo Pond of theSusquehanna River and Upper Chesapeake Bay ...........
....................
Appendix F.

Reuse of Accident Water .................

..........................

B-1

D-1
...

E-1
F-1

Appendix G. Engineering Considerations for Treatment of TMI-2 Accident-Generated
Liquid Waste .....................
....................................

.. G-1

Appendix H. Engineering Considerations Related to Immobilization ofRadioactive Wastes ....................
................................

.. H-1xxi

Page

CONTENTS (continued)
Appendix I.

Justification for Radiation Fields Used in Section 6

Appendix K.

Economic Cost Basis .............

. . .

........

•. . K-I.........................

Appendix L. Average Individual Quarterly Dose Limits Used in Determinations...........................
of Work Force Estimates .................
Appendix M. "Long-Term Environmental
Mile Island," U.S. Environmental

I-1

L-1

Radiation Surveillance Plan for Three
Protection Agency, 1981 ..........
.

. . . .

M-1

Appendix N.

Occupational Radiation Exposure during Onsite Waste Handling

. . .

N-1

Appendix 0.

Decontamination Status of Auxiliary and Fuel Handling Buildings . . . . . .

0-1

Appendix P.

Chemical Systems for Decontamination of Primary System Components

P-1

Appendix Q.

Onsite Storage-Facility .

...

. . .

Q-1

Appendix R. Proposed Additions to Technical Specifications for
TMI-2 Cleanup Program ............
........................

R-1

Appendix S. Calculations of Di-scharge of Processed Accident Water
to the Atmosphere .........
....
..................

S-1

Appendix T. The Behavior of Sorbable Radionuclides
Susquehanna River and Chesapeake Bay ...............

T-1

Appendix U.

Decommissioning of TMI-2

in theU-1.................

Appendix V. Assessment of Groundwater Liquid Pathway from Leakage ofContainment Water at Three Mile Island, Unit 2 ..........

. . . . . . . .

V-I

Appendix W. Calculation Models and Parameters Used in Estimating
Doses, and Interpretation of Model Results .. ..........

. . . . . . . .

W-IAppendix X.

. . . . . . . .

X-1

Appendix Y. Scheduled Meetings for Discussion of TMI-2 Cleanup Draft
Programmatic Environmental Impact Statement (1980)........

. . . . . ..

..

Y-I

Appendix Z.

. . . . .

..

Z-1

Contributors to the PEIS....

Health Effects Estimators ......

.

.

. ..

..................

..

xxii

TABLES

Page

Tables1.1

Acceptable Removal Surface Contamination Levels for Unrestricted Access

....

1-27

2.1
2.2

Estimated Volumes of Radioactive Waste for Decommissioning Alternatives ....
Postulated Releases of Airborne Radioactivity to the Environment during
Decommissioning Operations .................
...........................
Estimated Cumulative Radiation Doses Resulting from Decommissioning
Operations ......................
. .....
. .............
Estimated Cumulative Radiation Doses Received by Direct Decommissioning
Workers ....
Estimated Cumulative Radiation Doses Received by Transportation Workers ....
Estimated Costs for Decommissioning TMI-2 via DECON, SAFSTOR, and ENTOMB ....
Alternatives for Management of Solid Materials ......
..................
Alternatives for Management of Liquids ........
.....
.................
Alternatives for Management of Process Solids .................

2-15

Magnitude and Frequency of Low Flow, 1892-1972 ......
..................
Percentage of the Time that the Susquehanna River Flow at Harrisburg
Equals or Is Less than Specified Values . . . . . . . . . . . . . . . . . .
Mean Monthly Flows of the Susquehanna River at Harrisburg for the
Period 1891-1979..................................
Water Quality of the Susquehanna River at Several Locations ..........
Population and Population Density Distribution in the Vicinity of
TMI, 1978 ....
Municipalities in the Vicinity of TMI . . . . . . . . . . . . . . . . . . .
Status of Local Government Insurance Claims Against Met Ed Related
to TMI Accident (as of April 30, 1980) ..........
......................
Chesapeake Fisheries.
Summary of Catch for Selected Years .... ............
Registered Full-Time Watermen in Maryland and Virginia, 1977-1979 .......
Chesapeake Bay Fisheries Processing and Wholesale Activity for Several Years
Economic Impact of Chesapeake Bay Commercial Fisheries .....
..... . ........
Chesapeake Bay Sport and Commercial Harvests, 1976 ......
................
Number of Boat Trips and Out-of-Pocket Expenditures by Sport Fishermen
in Maryland, 1979 . . . . . . . . .. . . . . . . . . . . . . . . . . . . . .

3-11

2.3
2.4
2.5
2.6
2.7
2.8
2.9
3.1
3.2
3.3
3.4
3.5
3.6
3.7
3.8
3.9
3.10
3.11
3.12
3.13

4.1
.4.2
5.1
5.2
5.3
5.4
5.5

5.6
5.7

2-16
2-16
2-17
2-17
2-19
2-27
2-29
2-31

. .

3-11
3-12
3-15

. .

.

..

Routine Operational Liquid Waste Sources ..........
....................
Releases to the Susquehanna River of Treated Routine Operational
Liquids . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Airborne Releases of Principal Radionuclides Expected to Occur from
Decontamination Operations in the Auxiliary and Fuel Handling Buildings
during a 15-Month Period after September 1, 1980 ......
.................
Airborne Releases from HEPA Filter Failure during Decontamination of
the AFHB ........................................
Whole-Body Occupational Doses Received by Workers during Decontaminationof the Auxiliary and Fuel Handling Buildings ..........
..................
Dose Estimates for the Maximum Exposed Individual for AFHB Decontamination
and Desludging Activities . . . . . . . . . . . . . . .. . . . . . . . . . . .
Estimates of Offsite Doses to the Maximum Exposed Individual Caused by
Failure of a HEPA Filter during Surface Decontamination Operations in the
AFHB.......................................
Estimates of Offsite Doses to the Maximum Exposed Individual Caused by
Failure of a HEPA Filter during Desludging Operations in the AFHB .......
Economic Costs for AFHB Decontamination ....................

3-20
3-21
3-23
3-26
3-27
3-28
3-29
3-30
3-31
4-4
4-4

5-12
5-14
5-15
5-16
5-17
5-18
5-19

xxiii

Table
5.8
5.9
5.10
5.11
5.12
5.13
5.14
5.15
6.1

6.2
6.3
6.4
6.5
6.6
6.7
6.8
6.9
6.10
6.11
6.12
6.13
6.14
6.15
6.16
6.17
6.18
7.1
7.2
7.3
7.4
7.5
7.6
7.7
7.8
7.9
7.10
7.11

Page

TABLES (continued)
Composition of the Atmosphere in the Reactor Building Shortly before
the Purge on June 28 - July 11, 1980 .....
....................
Relative Amounts of the Three Principal Radionuclides for Different
Sources . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . .
Airborne Releases of Principal Radionuclides Expected to Occur from
Decontamination of Surfaces in the Reactor Building under Normal
Conditions..................................
Estimated Airborne Releases from a HEPA Filter Failure during
Decontamination of the Reactor Building Surfaces .................
....
Dose Estimates for the Maximum Exposed Individual for Reactor Containment
Building Decontamination for Lower Estimated Tritium Release ....... . . .
Dose Estimates for the Maximum Exposed Individual for Reactor Containment
Building Decontamination for Higher Estimated Tritium Release.......
Estimates of Offsite Doses to the Maximum Exposed Individual Caused by
Failure of a HEPA Filter during Decontamination of the Reactor Building
Economic Costs for Reactor Building Decontamination ............

5-21
*

5-43
5-44
5-47
5-48
5-49
5-50

Average Dose Rates, Work Times, and Cumulative Occupational Doses
for Reactor Coolant System Inspection Operations under Best- andWorst-Case Conditions. . . . . . . . . . .. . . . . . . . . . . . . . . . . .
Economic Costs for RCS Inspection
Work Effort, Crew Size, and Radiation Levels for Removal of RPVH
and Internals.....
Wastes Generated during Removal of the RPV Head and Internals .........
Work Effort, Crew Size, and Calendar Time Required to Remove TMI-2
RPV Head and Internals..............................
Estimates of Offsite Doses to the Maximum Exposed Individual due to
Tritium and Krypton Released during Removal of RPV Head and Internals .....
Economic Costs for Removal of RPV Head and Internals .....
...............
Dose Estimates for the Maximum Exposed Individual from Core
Examination and Defueling Operations .........
...... ................
Economic Costs for Defueling ....
..........................
Work Effort, Crew Size, and Radiation Field for Activities in
Decontamination of Reactor Primary System ...................
Work Crews Required for Activities in Primary System Decontamination .........
Work Effort Estimates for Subalternatives for Out-of-Containment
Decontamination of Reactor Coolant Pump Components ...
...............
Wastes Generated from Activities in Primary System Decontamination ..........
Airborne Releases of Principal Radionuclides due to a Spill of
Decontamination Liquids from the RCS into the Reactor Building .............
Occupational Dose, Calendar Time, and Personnel Required for Activities
in Decontamination of the Reactor Primary System ......
..........
.......
Dose Estimates to the Maximum Exposed Individual due to Tritium
during Decontamination of Primary System Components ..............
Dose Estimates for the Maximum Exposed Individual Due to Spilling
RCS Water in the Reactor Building
Economic Costs of Decontamination of Primary System Components .............
Summary of Estimated TMI-2 Liquid Waste ...................
Composition of AFHB Water Prior to Start of Processing .....
........
.....
Storage Locations, Volume, and Radionuclide Concentrations of Processed AFHB
Accident Water .....
.................................
Radioactivity in TMI-2 Untreated Liquid Waste
Estimated Radionuclide Concentrations of Dissolved Constituents of Reactor
Building Sump Water as of September 30, 1980 ........
...................
Concentration of Radionuclides in Filterable Solids in Reactor Building Sump
Water as of September 30, 1980 .............
....................
Estimated Concentrations of the Principal Constituents of the Reactor Coolant
System Water as of September 30, 1980 .....................
Impacts of Immobilization and Offsite Disposal of Accident Water .... ........
Treatment Alternatives Suitable for TMI Liquids ................
Comparison of Treatment Systems.........
Gaseous Effluents and Releases Attributable to Aerosol Formation from
Processing of Various Liquids (Normal Operations)...............

5-26

.

6-4
6-6
6-10
6-13
6-16
6-17
6-17
6-26
6-26
6-35
6-36
6-36
6-37
6-38
6-39
6-40
6-42
6-42
7-2
7-4
7-4
7-5
7-5
7-6
7-7
7-14
7-27
7-29
7-30

xxiv

Table
7.12
7.13
7.14
7.15
7.16
7.17
7.18
7.19
7.20
7.21
7.22
7.23
7.24
7..25
7.26
7.27
7.28
7.29
7.30
7.31
7.32
7.33
7.34
7.35
7.36

7.37
7.38

7.39
7.40
7.41
7.42
7.43

TABLES (continued)
Gaseous Effluents and Releases from Failure of a HEPA Filter during
Processing of Various Liquids ..............
.........................
...
Summary of EPICOR II System Occupational Exposure Experience ..............
...
Dose Estimates for the Maximum Exposed Individual due to Processing of Accident
and Decontamination Water ..............................
Dose Estimates for the Maximum Exposed Individual Caused by Failure of a HEPA
Filter during Processing of Reactor Building Sump Accident Water ..... ........
Dose Estimates for the Maximum Exposed Individual Caused by Failure of a HEPA
Filter during Processing of Reactor Building Decontamination Water .........
...
Dose Estimates for the Maximum Exposed Individual Caused by Failure of a HEPA
Filter during Processing of Reactor Building Decontamination Water ...........
Dose Estimates for the Maximum Exposed Individual Caused by Failure of a HEPA
Filter during Processing of RCS Accident Water . . ...........
..............
Dose Estimates for the Maximum Exposed Individual Caused by Failure of a HEPA
Filter during Processing of RCS Flush and Drain Water .....
.............
...
Dose Estimates for the Maximum Exposed Individual Caused by Failure of a HEPA
Filter during Processing of RCS Decontamination Water or RCS Chemical
Decontamination Water ....
.............................
Parameters for Comparison of Liquid Waste Treatment Alternatives .....
........
Tritium Evaporation Losses from Potential Reuse Applications andOther Facilities .................
................................
...
Radionuclide Inventory - Processed Water Best-Case SDS/EPICOR II Treatment ...............................
Radionuclide Inventory - Processed Water Worst-Case SDS Treatment..................................
Releases to the Susquehanna River for the Best-Case--SDS /EPICOR II ....
......
Release to the Susquehanna River for the Worst-Case SDS .........
............
Normal Releases from Cooling Tower Evaporation of Process Water ...........
..
Maximum Activity Released to the Susquehanna River due to a Rupture of aProcessed Water Storage Tank.. .........
. .
. . . . . . . . . . . . . ..
.
Dose Estimates to the Maximum Exposed Individual for Releasing Processed
Water to the River - SDS/EPICOR-II Option ..................
......
. . .
Dose Estimates to the Maximum Exposed Individual for Releasing Processed
Water to the River - SDS Option ..............
........................
...
Dose Estimates to the Maximum Exposed Individual for Releasing Processed
Water to the Atmosphere via Natural Evaporation .......................
....
Dose Estimates to the Maximum Exposed Individual for Releasing Processed
Water Via the Forced Evaporation Option - SDS/EPICOR II Processing Option
Dose Estimates to the Maximum Exposed Individual for Releasing Processed
Water Via the Forced Evaporation Option - SDS Processing Option .....
........
Dose Estimates to the Maximum Exposed Individual Caused by Rupture of aProcessed Water Storage Tank during High River Flows - SDS Processing Option
Dose Estimates for Consumption of Drinking Water or Fish Flesh from the East
Channel due to an Accidental Processed Water Storage Tank Rupture during
Low River Flows .......
......................
......
.. ...........
...
Background and Added Concentrations of Radionuclides in Susquehanna River
Below York Haven Dam due to Controlled Releases from SDS/EPICOR II
and SDS Treatment Systems............................
Conservative Dilution of Susquehanna River Water in Chesapeake Bay .........
...
Average Annual Doses to Fishes in the Susquehanna River Downstreamof York Haven Dam Resulting from Normal or Accidental Releases ofProcessed Water ..........................................
Bioaccumulation Factors Estimated from Measured Concentrations in Fish
and Waters of the Susquehanna River .... .................
. . .......
...
Doses to Adult Fishes, Eggs, Embryos and Larvae in Red Hill Pond Caused by
Failure of a Processed Water Storage Tank ...........
...................
...
Comparison of Acute Lethal Radiation Doses for Various Groups of Organisms .
Disposal Costs for TMI-2 Liquids ...............................
Comparison of Alternatives for Disposal of Processed Water ...............
...

Page

7-32
7-33
7-36
7-37
7-38
7-39
7-40
7-41

7-42
7-44
7-46
7-47
7-47
7-57
7-57
7-58
7-59
7-62
7-62
7-63
7-64
7-65
7-67

7-67

7-69
7-72

7-74
7-75
7-76
7-77
7-82
7-83

XXV

Table

TABLES (continued)

8.1
8.2
8.3
8.4
8.5
8.6
8.7

Process Solid Waste Summary ..................
..........................
8-2
Summary of EPICOR II Waste Packages ................
......................
8-4
Minimum Generation of Process Solid Waste .............
....................
8-5
Maximum Generation of Process Solid Waste ..... ..............
.........
8-6
Accident Sludge Packages--Minimum and Maximum Waste Generation ............
... 8-7
Generation of Zeolite Waste--Minimum and Maximum Alternatives ... .........
... 8-8
Generation of High-Specific-Activity Organic Resins--Minimumand Maximum Alternatives ..................................
8-9
Generation of Low-Activity Organic Resins--Minimumand Maximum Alternatives ...............
............................
.. 8-11
Generation of Spent Filter Cartridges . . .. .. .........
.. ...........
... 8-12
Characteristics of Packaged Evaporator Bottoms--Minimumand Maximum Generation ...............
.............................
... 8-13
Characteristics of Waste Generated from Bitumen/Resin System .... .....
....
.
8-14
Characteristics of Packaged Accident Sludge .........
..................
... 8-15
Characteristics of Packaged Zeolite Liners ........
...................
... 8-17
Characteristics of Packaged High-Specific-Activity EPICOR II Resins .... ......
8-18
Characteristics of Packaged Low-Activity EPICOR II Organic Resins .........
... 8-19
Characteristics of Packaged Low-Activity Organic Resins ..... ............
... 8-20
Maximum Estimated Airborne Effluents Released to the Environment during
Immobilization of Process Solid Waste . .....
...................
8-22
Estimated Releases during SDS-Based System Package Handling Accidents .......
.. 8-23
Estimated Releases during EPICOR II and Modified EPICOR II System
Package Handling Accidents .................
...........................
8-24
Estimated Releases during Other Waste. Type Package Handling Accidents .....
... 8-25
Estimated Worker Exposure to Package and Handle Accident Sludge andFilters ......................
..........................
.............
8-27
DoseEstimates for the Maximum Exposed Individual Due to Processing Solid.Wastes ......
........................
........... ......
............
8-28
Dose Estimates for the Maximum Exposed Individual Due to Breaching of a Package
Containing SDS Waste--Reactor Building Sump Water - Zeolite Liner .........
... 8-30
Dose Estimates for the Maximum Exposed Individual Due to Breaching of a Package
Containing SDS Waste--Reactor Building Sump Water - Filter Assembly ........
.. 8-31
Dose Estimates for the Maximum Exposed Individual Due to Breaching of a Package
Containing SDS Waste--Primary Water - Cation Liner ........
...............
8-32
Dose Estimates for the Maximum Exposed Individual due to Breaching of a Package
Containing SDS Waste--Primary Water - Mixed-Bed Liner .....
.............
... 8-33
Dose Estimates for the Maximum Exposed Individual due to Breaching of a Package
Containing EPICOR II Waste--AFHB Water - Prefilter Liner ......... ..........
8-34
Dose Estimates for the Maximum Exposed Individual due to Breaching of a Package
Containing EPICOR II Waste--AFHB Water - Cation Liner ....... ............
... 8-35
Dose Estimates for the Maximum Exposed Individual due to Breaching of a Package
Containing EPICOR II Waste--AFHB Water - Mixed Bed Liner .... ............
... 8-36
Dose Estimates for the Maximum Exposed Individual due to Breaching of a Package
Containing Modified EPICOR II Waste--Primary Water - Zeolite Liner ..........
8-37
Dose Estimates for the Maximum Exposed Individual due to Breaching of a Package
Containing Modified EPICOR' II Waste--Primary Water - Cation Liner .........
... 8-38
Dose Estimates for the Maximum Exposed Individual due to Breaching of a Package
Containing Modified EPICOR II Waste--Primary Water - Mixed Bed Liner .... ......
8-39
Characteristics and Disposition of Chemical Decontamination Solution Wastes
Generated through September 22, 1980 .............
......................
8-41
Projected Volumes and Characteristics of Chemical Decontamination Solutions
8-41
Chemical Decontamination Solution Waste Generation--Minimum andMaximum Cases ....................
.................................
... 8-42
Packaged Decontamination Solution Waste--Minimum and Maximum Cases .........
... 8-44
Estimated Releases to Environment under Normal Conditions during Immobilizationof AFHB and Reactor Building Decontamination Solutions .................
8-44
Estimated Accident Releases to Environment from Breach of Packaged
Immobilization AFHB and Reactor Building Decontamination Solutions .........
... 8-45
Estimated Occupational Doses from Handling and Packaging Chemical
Decontamination Solution Wastes ...
........................
8-46
Dose Estimates for the Maximum Exposed Individual due to Immobilizing Chemical
Decontamination Solutions ....................
...........................
8-46

8.8
8.9
8.10
8.11
8.12
8.13
8.14
8.15
8.16
8.17
8.18
8.19
8.20
8.21
8.22
8.23
8.24
8.25
8.26
8.27
8.28
8.29
8.30
8.31
8.32
8.33
8.34
8.35
8.36
8.37
8.38
8.39
8.40

Page

XXV i

Table
8.41
8.42
8.43
8.44
8.45
8.46
8.47
8.48
8.49
8.50
8.51
8.52
8.53
8.54
8.55
8.56
8.57
8.58
8.59
8.60
8.61
8.62
8.63
8.64
8.65
8.66
8.67
8.68
8.69
8.70
9.1
9.2
9.3
9.4
9.5
9.6
9.7
9.8
9.9
9.10
9.11

TABLES (continued)

Page

Estimates of Offsite Doses to the Maximum Exposed Individual Caused by Breachof Package Containing AFHB and Reactor Building Immobilized Decontamination
Solution Waste ............................................
Solid Waste Generated through September 22, 1980 ......
................
...
Solid Waste Projections ...............................
Minimum and Maximum Alternatives for Trash Generation .....................
Partial Listing of Reactor Building Contaminated Equipment for Disposal. . ..
Primary System Contaminated Equipment ..............
.....................
..
Contaminated Equipment--Minimum and Maximum Waste Generation .... ..........
Reactor Vessel.......................
...........
Irradiated Hardware--Minimum and Maximum Waste Generation .... ...........
...
Packaged Trash Summary..
. . . ..........................
Comparison of Incineration Versus Compaction of Combustible Trash .. .......
...
..
Packaged Contaminated Equipment Summary ...........
....................
Estimated Radionuclide Releases to Building Atmosphere during Trash Compaction .
Estimated Radionuclide Releases to the Environment during Trash Incineration.
Estimated Radionculide Releases to the Environment during Low-Level Waste
Storage Area Fire
Postulated Worst-Case Package Handling Accidents Involving Solid Waste ........
Estimated Accident Releases to the Environment ..........
.................
Estimated Occupational Radiation Doses from Handling and Packaging of AFHB
Solid Wastes ......................................
Estimated Occupational Radiation Doses from Handling and Packaging ofReactor Building Solid Wastes .......
.......................
.
.......
Estimated Occupational Radiation Doses from Handling and Packaging of Solid
Wastes from Defueling and Primary System Decontamination .... ............
...
Summary of Estimated Occupational Radiation Doses from Handling and Packagingof Solid Wastes ..........................................
Dose Estimates for the Maximum Exposed Individual for Trash Compaction of
AFHB Wastes ......................................
Dose Estimates for the Maximum Exposed Individual for Trash Compaction of
Reactor Building Wastes ................
............................
...
Dose Estimates for the Maximum Exposed Individual for Releases Made due toTrash Incineration ......
...............................
Dose Estimates to the Maximum Exposed Individual Caused by a Low-Level Storage
Area Fire .......................................
Dose Estimates to the Maximum Exposed Individual Caused by Breaching a Package
Containing Compactible Trash .............
..........................
..
Dose Estimates to the Maximum Exposed Individual Caused by Breaching a Package
Containing Noncompactible Trash ..............
...................
. . . .
Dose Estimates to the Maximum Exposed Individual Caused by Breaching a Package
Containing Trash Mirror Insulation ............
.......................
...
Dose Estimates for the Maximum Exposed Individual Caused by Breaching a Package
Containing Incinerator Ash .................
...........................
Cost Estimates for Solid Materials Processing (thousands of dollars) .... ......
Estimated Number of Unshielded Waste Shipments .... ..................
Estimated Number of Shielded Ion-Exchange Material Shipments ..........
Estimated Number of Shielded Drum Shipments .........
...................
Estimated Number of Miscellaneous Shielded Shipments .....
..............
Estimated Number of Shielded Shipments in Fuel Casks ..... ............
Summary of Estimated Number of Shipments ..........
....................
Characteristics of Disposable Containers .....
.........................
Unshielded Shipping Configurations ............
.......................
Shielded Shipping Configurations ............
........................
Occupational Radiation Doses during Transfer of Waste Packages from
Storage and Truck Loading ...
...........................
Transportation Parameters for Calculation of Occupational Doses for
Waste and Spent Fuel Shipments from TMI-2 ...........
...................

8-47
8-49
8-50
8-51
8-53
8-54
8-54
8-55
8-56
8-57
8-58
8-59
8-60
8-61
8................................8-61
8-62
8-62
8-64
8-64
8-65
8-65
8-67
8-68
8-69
8-70
8-71
8-72
8-73
8-74
8-75

. ..

9-2
9-3
.. 9-4... 9-4. ... 9-4.. 9-6
9-8... 9-9... 9-9
9-29... 9-30

xxvii

Table
9.12
9.13
9.14
9.15

TABLES (continued)
Volumes of Packaged Solid Waste to Be Disposed of at a Commercial
Low-Level Disposal Site ....................
.............................
Volumes of Waste Packages to Be Stored ..........
...............
.......
Estimated Airborne Releases from Selected Type-B Waste Packages under...............
Accident Conditions for 1 x 10-5 Fractional Release .....
Estimated Waterborne Releases from Selected Type-B Waste Packages
under Accident Conditions for 1 x 10-6 Fractional Release .......
. .......

Page
9-33
9-34
... 9-35
...

10.17
10.18
10.19

Maximum Estimated Amounts of the Principal Radionuclides Released to the
Atmosphere as a Consequence of Normal Cleanup Operations at TMI-2 ....
.......
Amounts of Principal Radionuclides that Will Be Present in All of theStored Water....................................
Summary of Estimated Number of Waste Shipments ......
........
.........
...
Amounts of the Principal Radionuclides Released to the Atmosphere as a.
Consequence of Postulated Accidents during Cleanup of TMI-2...........
Summary of Cumulative Dose and Health Effects to Workers from Cleanup of
TMI-2 ......
.....................................
Dose Estimates for the Maximum Exposed Individual for Normal Radioactive
Releases to the Atmosphere during Decommissioning and Decontamination of the
Facility .....................
.....................................
.
Dose Estimates for the Projected Population in Year 2010 Residing within
50-Mile Radius of TMI Resulting from Atmospheric Radioactive Releases during
Decontamination and Decommissioning of the Facility ..
..............
Dose Estimates for the Maximum Exposed Individual for Normal Radioactive
Releases to the River Resulting from Disposal of Processed Water ..... ........
Dose Estimates to the Population Residing Downstream of TMI Resulting from
Normal Radioactive Releases of Processed Water to the River
Dose Estimates to the Population Residing Along Transport Routes Enroute to
Waste Burial Grounds - Decontamination and Decommissioning............
Dose Estimates for the Maximum Exposed Individual for Entire Decontamination
Program for Each Processed Accident Water Disposition Alternative, andComparison with 10 CFR 50 Requirements .....
...............
......
Estimates of Cancer Death Rate and Genetic Abnormality Rate for Exposure of
the Population around TMI due to Releases from Decontamination for theRange of Processing and Disposal Alternatives ...
.................
Estimate of Cancer Death Rate and Genetic Abnormality Rate for Exposure of
the Population Residing along Transport Routes ......
.................
...
Dose Estimates for the Maximum Exposed Individual Caused by a HEPA Filter
Failure during Specific Decontamination Operations ......
...............
...
Dose Estimates for the Maximum Exposed Individual Caused by Breaching a
Package Containing Radioactive Waste .......................
...........
Dose Estimates for the Maximum Exposed Individual Caused by Breaching a Processed Water Storage Tank and Releasing Contents into the East Channel ofthe-Susquehanna River ................
..............................
...
Summary of Levee and Flood Elevations at Three Mile Island .... ...........
..
Significant Design Basis Tornado Missiles ....
...................
Summary of Alternative Economic Costs for the Cleanup of TMI-2 ............
...

11.1
11.2

Monitor Well Locations at TMI-2 .............
.........................
Sampling of Susquehanna River .........
...............................

1.1

Environmental Impacts of Alternatives for Removing the Krypton-85 fromthe Reactor Building Atmosphere .....
...........................
...
Hydrogen Control Subsystem ...............
.............................
Comparison of Krypton-85 Containment Techniques .......
..........
......
..
Summary of Age-Specific Cancer Mortality Risk Estimators and AssociatedLife-Shortening .....
.................................
Summary of Lifetime Risks of Mortality Numerically Equivalent to 0.2 mrem . .
Summary of Latent Radiogenic Cancer Risks Comparable to 0.2 mrem ..... ........
Three Mile Island EPA Long-Term Surveillance Stations, Air Samplers, Gamma
Rate Recorders, TLDs ......
..............................
Description of NRC TLD Locations ............
.................
.......
..

10.1
10.2
10.3
10.4
10.5
10.6
10.7
10.8
10.9
10.10
10.11

10.12
10.13
10.14
10.15
10.16

6.2-1
6.8-1
7.1
7.2
7.3
8.1
8.2

9-36

10-2
10-4
10-5
10-6
10-8
10-11
10-13
10-14
1...........
0-15
10-16
10-17

10-19
10-19
10-21
10-23

10-24
10-26
10-27
10-32

... 11-2
11-7

C-9
C-15
C-29
C-32
C-32
C-33
C-37
C-37

xxviii

Table
1
2

3
E.1
E.2
E.3
E.4
E.5

Page

TABLES (continued)
Radioactive Water Volumes for TMI Unit 2 Which Will Be Processed By EPICOR-II
Concentrations of Principal Nuclides In TMI Unit 2 Auxiliary Building Tanks
To Be Processed By EPICOR-II Corrected for Radioactive Decay to 6/15/79,.....................................
pCi/mL ..........................
...........
. .......
Radioactive Half-Lives of Principal Nuclides ........
Estimates of Recreational Fishing in Vicinity of TMI, 1974-1979.... . .
......................
Food Habits of Conowingo Pond Fishes .............
Species Composition of Fish Kept by Anglers in Conowingo Pond.
Based
on 1973-1977 Winter Fishery Survey ............
.....................
Winter Fishing Pressure and Catch per Hour of Fishes-in Conowingo Pond,
1973-1977 ..............
. ....... .........
............................
Species Composition and Estimated Harvest by Anglers in Conowingo Pond,
August 1977-August 1978 ................
............................

.

D-16
D-16

......

E-4
E-6

. ..

E-8
E-8

... E-9F.1
F.2

Radionuclide Inventory of Processed AFHB Water in Storage .... ...........
Summary of Potential Uses of Processed Accident Water .....
.............

G.1
G.2
G.3

Treatment Alternatives Suitable for TMI Liquids .......
................
...
Estimated TMI-2 Liquid Waste Summary .............
......................
Estimated Concentrations of Dissolved Contaminants in Reactor Building Sump
Water as of September 1, 1980 ..............
.........................
...
Filterable Solids in Reactor Building Sump Water as of September 1, 1980 ....
Estimated Concentrations of Contaminants in the Reactor Coolant System............................
as of September 1, 1980 ....
Estimated Concentrations of Contaminants in Reactor Coolant System
Flush and Drain................................
Estimated Radioactivity Contained in Solid Debris in Reactor Coolant System................
Flush and Drain Water as of September 1, 1980 .....
Summary of Liquid Waste Radionuclide Inventories ..........
. .........
..
Process Decontamination Factors and Breakthrough Volumes
for Submerged Demineralizer-Type Systems ..........
....................
...
EPICOR-II Performance for Batch No. 41--AFHB Water ......... . .............
Decontamination Factors Assumed for EPICOR-II System .....
..............
...
Decontamination Factors, Volume Reduction Factors and Breakthrough Volumes
Assumed for Zeolite/Evaporation Processing ........
.....................
Decontamination Factors and Breakthrough Volume
Assumed for Zeolite/EPICOR II Process ..............
.....................
Decontamination Factors and Breakthrough Volumes
Assumed for SDS/EPICOR II Process ..................
...........
Volume Reduction Factor and Decontamination Factors
Assumed for Evaporation/Resin Process ..............
................
....
Decontamination Factors and Volume Reduction Factors Assumed
for Bitumen/Resin Process ..................
..........................
.
Summary of Decontamination Factors for TMI-2 Treatment System Components ....
Summary of Overall Decontamination Factors for TMI-2 Treatment Systems ........
Radionuclides in Liquid Effluents from Treatment of Reactor Building
Sump Water ........................................
Radionuclides in Liquid Effluents from Treatment of RCS Primary System Water
Radionuclides in Liquid Effluents from Treatment of RCS Flush andDrain Water .....................................
Radionuclides in Liquid Effluents from Treatment of
RCS Chemical Decontamination Solutions ..........................
Radionuclides in Liquid Effluents from Treatment of AFHB/Reactor Building
Decontamination Solutions ................
.............................

G.4
G.5
G.6
G.7
G.8
G.9
G.10
G.11
G.12
G.13
G.14
G.15
G.16
G.17
G.18
G.19
G.20
G.21
G.22
G.23
H.1
H.2
H.3
H.4
H.5

D-15

... F-2... F-4
G-7
G-8
G-8
G-9
G-9
G-10
G-.O
G-I1
G-14
G-15
G-15
G-18
G-21
G-23
G-24
G-27
G-28
G-29
G-30
G-31
G-32
G-33
G-33

Comparison of Available Radwaste Immobilization Technologies .... ..........
... H-2
Summary of Bitumen and Cement Immobilization System Characteristics ....
......
H-4
Typical Cement/Radwaste Formations ......
.......................
H-9
Formulations for Solidifying Radwastes with Portland Cement-Sodium
Silicate ....................
....................................
.. H-10Portland Cement Admixtures to Overcome Boric Acid Inhibiting Properties . . ..
H-10

xxix

Table
1.1

J.1
J.2
J.3

TABLES (continued)

Page

Summary of Results ol Gamma Exposure Rate Calculations for TMI-2 Primary......................................
System ......................

1-2

Analysis of Potential Dose Contribution of Radionuclides in Airborne
Effluents (due to aerosol formation) from Processing Reactor Building....................................
Sump Water .....................
Analysis of Potential Dose Contribution of Radionuclides in Airborne
Effluents (due to aerosol formation) from Processing of Primary Waterfrom the Reactor Coolant System . . . . . . . . . . . . . . . . . . . . . . . .
Analysis of Potential Dose Contribution of Radionuclides in Airborne
Effluents (due to aerosol formation) from Processing AFHB Sump Water ......

K. 1
K. 2
K. 3
K.4
K. 5
K. 6
K. 7
K. 8
K. 9
K. 10
K. 11
K. 12
K. 13
K. 14

Costs of Liquid Processing per Gallon of Influent .
..
...
..
Costs of Liquid Processing Facilities
Indirect Manpower Cost Estimates ...........
.....
Unit Operating Cost for Disposal of TMI-2 Liquids
Transportation and Disposal Costs for TMI-2 Liquids
Facility Costs for Disposal of TMI-2 Liquids .......
Waste Conditioning--Unit Costs .... .............
Waste Packaging and Handling--Unit Costs ...........
Process Solid Waste Volumes Considered in Costing .
Waste Facility Costs ......
..................
Decontamination Solution Immobilization Costs ....
Estimated Unit Costs for Management of Trash .......
Low-Level Waste Transportation and Disposal Costs .
General Support Facilities .....
...............

1.

Concentrations of Major Radionuclides in Water Within the Reactor
Containment Building, 7/14/80 . . . . . . . . . . . . . . . . . . . . . .
Concentrations of Major Radionculides in the Primary Reactor Coolant',
7/21/80 . . . . . . . . . . . . .
Distance and Azimuth of Sampling Locations For The Three Mile Island
Nuclear Station Offsite Emergency Radiological Environmental Monitoring. . . . . . . . . . . . . . . . . . .. . . . . . . . . .
Program . . . .

2.
B-1

N.I
N.2
N.3
N.4
N.5
N.6
N.7
N.8
N.9
N.1O

. . .
. . .
. .
. .
. .
. .
. .
. .
. . .
. .
. .
. .
. . .
. .

.
.
.
.
.
.
.
.
.
.
.
.
.
.

.
.
.
.
.
.
.
.
.
.
.
.
.
.

.
.
.
.
.
.
.
.
.
.
.
.
.
.

.
.
.
.
.
.
.
.
.
.
.
.
.
.

.
.
.
.
.
.
.
.
.
.
.
.
.
.

.
.
.
.
.
.
.
.
.
.
.
.
.
.

.
.
.
.
.
.
.
.
.
.
.
.
.
.

.
.
.
.
.
.
.
.
.
.
.
.
.
.

Occupational Exposure from Handling Low-Activity Trash Drums .... ...........
Occupational Exposure from Handling of LSA Boxes ......
.................
Occupational Exposure from Handling of Spent Filter Cartridges .............
Occupational Exposure from Handling of Irradiated Hardware andContaminated Equipment ......
..................................
Occupational Exposure from Handling of Immobilized Incinerator Ash ..........
.........
Occupational Exposure from Handling of EPICOR-II Resin Liners
Occupational Exposure from Handling of Zeolite Resin System Liners ..........
Occupational Exposure from Handling Drums of Evaporator Bottoms
.
.
.
.
.
.
.
. . ..
and Accident Sludge .
Occupational Exposure from Handling Drums of Low-Activity Decontamination
SolutionsOccupational Exposure from Handling Drums of Intermediate-Level
Decontamination Solutions . . . . . . . . . . . . . . . . . . . . . . . . . . .
1980 . . .
1980 . . .
1980 . . .

J-2
J-3
J-3
K-5
K-6
K-7
K-8
K-8
K-9
K-9
K-IO
K- 10
K-lI
K-lI
K- 12
K-13
K- 14
M-3
M-4

M-10
N-2
N-3
N-4
N-4
N-5
N-5
N-7
N-7
N-8
N-8

0.1
0.2
.0.3

Decontamination Status for Elevation 281 as of September 1,
Decontamination Status for Elevation 305 as of September 1,
Decontamination Status for Elevation 328 as of September 1,

S.1
S.2
S.3

Mean Monthly Meteorological Data at Harrisburg Airport ...............
. . . .S-4
HTO Releases to Air for July Ponding ........
....................
. . . S-4
Density of 100% Saturated Water Vapor, Relative Humidity, and HTO Air
as a Function of Time . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . S-8

U.1
U.2
U.3

Specialty Contractors for Decommissioning ..............
Essential Systems and Services for Decommissioning ..........
Special Tools and Equipment for Decommissioning ...........

0-4
0-9
0-11

. . . . .

. . . . .
. . . . .

U-3

U-5

U-6

XXX

Table
U.4
U.5
U.6
U.7
U.8
U.9
U.10
U.11
U.12
U.13
U.14
U.15
U.16
U.17
U.18
U.19
U.20
U.21
U.22
U.23
U.24
U.25
U.26
U.27
U.28
U.29
U.30
U.31
U.32
U.33
U.34
U.35
U.36
U.37
U.38
U.39
W.1

TABLES (continued)
Comparison of Major Concrete Surface Removal Techniques .. ....
.
.......
Estimated Specific Activities and Surface Doses from Reactor VesselInternals of TMI-2 . .........
. ....
.............................
Estimated Volumes of Radioactive Waste Arising from DECON of TMI-2 .........
...
Postulated Releases of Airborne Radioactivity to the Environment During
DECON Operations ..................... .............................
...
Reference Radionuclide Inventory 2, Carbon Steel Activation Products Lower Vessel ...........................................
Reference Radionu~lide Inventory 3, Concrete Activation Products Biological Shield .........................................
Reference Radionuclide Inventory 4, Neutron-Activated Corrosion Products
Deposited on Piping Internal Surfaces ...........
.....................
...
Reference Radionuclide Inventory 5, Radioactive Surface Contaminationin the Reference PWR ...............
..............................
...
Estimated Cumulative Radiation Doses Received by Workers During DECON
Operations .....
...................................
Estimated Cumulative Radiation Doses Received by Transport Workers During
DECON Operations .....................
.................................
Estimated Doses to the Maximum-Exposed Individual from Normal Decommissioning
Activities (DECON) ....................................
Estimated Cumulative Radiation Dose Received by the Public During Transportof Radioactive Wastes from DECON ......
.........................
....
Summary of Estimated DECON Costs ...........
.............. ..............
Estimated Labor Costs for DECON .....
.......................
..........
Estimated Costs for Disposal of Radioactive Wastes from DECON ... .........
...
Systems and Services Required During the Continuing Care Period .....
........
Estimated Volumes of Radioactive Waste from Preparing TMI-2 for Safe
Storage .....................
...........................
. .........
Estimated Cumulative Radiation Doses Received by Workers During SAFSTOR
Operations ...................
...................................
...
Estimated Cumulative Radiation Doses Received by Transport Workers During
SAFSTOR Operations .....
...............................
Estimated Doses to the Maximum-Exposed Individual from Normal Decommissioning
Activities (SAFSTOR) ....
..............................
Estimated Radiation Dose Received by the Public During Transport ofRadioactive Waste from SAFSTOR Operations ...........
...................
...
Summary of Estimated Costs for SAFSTOR ..........
.....................
...
Estimated Labor Costs for SAFSTOR .............................
Estimated Costs for Disposal of Radioactive Wastes from SAFSTOR .....
........
Comparison of DECON and Deferred Decontamination Costs from Reference
Decommissioning Studies ................
............................
...
Comparison of DECON and Deferred Decontamination Costs of TMI-2 .....
........
Estimated Volumes of Radioactive Waste Arising from ENTOMB of TMI-2 .... ......
Apportionment of Radioactive Waste Arising from ENTOMB of TMI-2 .......
......
Postulated Releases of Airborne Radioactivity to the Environment During
ENTOMB Operations .....................
...............................
Estimated Cumulative Radiation Doses Received by Workers During ENTOMBOperations ...................
...................................
...
Estimated Cumulative Radiation Doses Received by Transport Workers During
ENTOMB Operations .....................
...............................
Dose Estimates to the Maximum-Exposed Individual from Normal Decommissioning
Activities (ENTOMB) ..................................
Estimated Cumulative Radiation Dose Received by the Public During Transportof Wastes from ENTOMB .........
.........
.............................
..
Summary of Estimated Costs for ENTOMB ..............
.....................
Estimated Labor Costs for ENTOMB ....
........................
Estimated Costs for Offsite Disposal of Radioactive Wastes from ENTOMB ........
Table Numbers of Regulatory Guide Y.109 where Input Parameters Can Be
Found, and Problem-Specific Parameters Used in the PEIS .........
............

Page
U-9
U-15
U-21
U-22
U-22
U-23
U-23
U-24
U-25
U-25
U-26
U-27
U-28
U-28
U-29
U-33
U-39
U-41
U-42
U-43
U-44
U-45
U-45
U-46
U-48
U-48
U-52
U-53
U-53
U-55
U-55
U-56
U-56
U-57
U-58
U-58
W-2

xxxi

Table
W.2
W.3
W.4
W.5
W.6
W.7

Z.1

Z.2
Z.3
Z.4
Z.5

Z.6
Z.7
Z.8
Z.9
Z. 10

Page

TABLES (continued)
Locations for Maximum Exposed Individual Dose Calculation andAtmospheric Dispersion Parameters. . . . . . . . . . . . . . . . . . . . . . .
x/Q Values Used in Population Dose Estimates (no decay - no depletion) .......
x/Q Values Used in Population Dose Estimates (2.26-day decay - nodepletion) .....................
...................................
X/Q Values Used in Population Dose Estimates (8-day decay with plumedepletion) ...................
.....................................
D/Q Values Used in Population Dose Estimates ...........................
Parameters Used in Maximum Individual Dose Calculations and Dose to
Biota Calculation for Liquid Pathways . . . . . . . . . . . . . . . . . . . . .
Calculation of Expected Leukemia Deaths for External Exposure . ...
Calculation of Expected Lung Cancer Deaths for External Exposure . . .
Calculation of Expected Stomach Cancer Deaths for External Exposure
Calculation of Expected Pancreas and Alimentary Canal Cancer Deaths
for External Exposure . . . . . . . . . . . . . . . . . . . . . . . .
Calculation of Expected Breast Cancer Deaths for External Exposure . .
Calculation of Expected Bone Cancer Deaths for External Exposure . . .
Calculation of Expected Thyroid Cancer Deaths for External Exposure .
Calculation of All Other Expected Deaths for External Exposure . . .
Maximum Expected Latent Cancer Deaths per Million Person-rem .......
Comparison of Fatal Cancer Risk Estimators .....
...............

W-3
W-4
W-5
W-6
W-7
W-8
Z-2
Z-3

Z-4
. . . .
. . . .
. . . .

Z-5
7.
Z-6
7.
Z-7
7.

. .
. .
. .
• .

7.
Z-8
Z-9
7.
Z-10
7.
Z-10

.
.
.
.

.
.
.
. .

xxxii

FIGURES

Page
1.1
1.2
1.3
1.4

............
Reactor Building and Major Components of Primary System .........
...
..................
Aerial View of the Three Mile Island Station ........
Schematic Plan of the TMI-2 Site ...............................
Licensee's Planning Schedule for TMI-2 Cleanup, Phase I and Phase II .... ......

2.1
2.2
2.3
2.4

.......................
Simplified Decision-Point Diagram .............
.
..............
Characterization of TMI-2 Radioactive Waste Forms . ......
......................
TMI-2 Waste Forms and Classification .............
.......................
Radioactive Waste Management Cycle ............

3.1
3.2
3.3
3.4
3.5

Map of the Area within 100 Miles of the TMI Site ......
.........
......
.
Map of the Area within 20 Miles of the TMI Site ...........
................
..............................
Aerial Photo of the TMI Site ....
TMI Annual Average Wind Rose at 100 ft, 1972-1975 Data ........
.............
York Haven Pond of the Susquehanna River, Showing the Flow Patterns in the
Channels and over the Dam at Three Mile Island .....
..................
Major Inputs and Outputs of the Susquehanna Showing Principal Water
Users below the TMI Plant ...................................
Major Types of Terrestrial Communities within One Mile of-the Three
Mile Island Station ..................
..............................

3.6
3.7
5.1
5.2
5.3

1-3
1-4
1-5
1-8

... 2-11
2-20
2-24... 2-26
3-2
3-3
3-4
3-7
3-9
3-10
..

3-19

......

5.4
5.5
5.6

Wet Mopping of TMI Auxiliary Building....... .
.................
Auxiliary Building Equipment Cleanup .............
......................
View Outside the Inner Airlock Door of the Reactor Building Just
Prior to Entry ...................
.................................
View Inside the Reactor Building during Entry .............
.................
Minimum Size Containment Services Building ........
...................
Larger Containment Services Building .............
......................

5-2
5-3

6.1
6.2
6.3
6.4

Cutaway View of a Typical PWR Similar to the TMI-2 Reactor Pressure Vessel
Reactor Vessel Internals and Core Cross Section .......
................
TMI Unit 2 Reactor Fuel Assembly ............
........................
Schematic Drawing of Primary System .............
......................

. .
...
...
...

6-7
6-20
6-21
6-30

7.1
7.2
7.3
7.4
7.5
7.6
7.7
7.8
7.9
7.10
7.11
7.12
7.13

7-9
7-11
7-15
7-17
7-19
7-20
7-22
7-23
7-25
7-48
7-56
7-71

7.14

Typical Sequence for Liquid Waste Processing ........
..................
...
Alternative Liquid Waste Treatment Systems ........
...................
...
SDS Process Diagram ..................
..............................
..
Alternative Treatment System Configuration for RB Sump Water .....
..........
Schematic Diagram of TMI-2 SDS Radwaste Treatment System .... ............
...
EPICOR II Process Diagram ................
...........................
...
TMI-2 EPICOR II Radwaste System ..............
........................
..
Evaporator/Resin Process Diagram ............
........................
...
Bitumen/Resin Process Diagram ..............
.......................
. ..
Decision Tree for Disposition of Processed Water ......
................
...
Simplified Process Diagram Forced Evaporation through Cooling Tower ....
......
Chesapeake Bay Area ..................
..............................
..
Longitudinal Salinity Distribution along Axis of Chesapeake Bay during(A) Period of High River Flow and (B) Period of Low to Moderate River Flow . . .
Tolerance of Aquatic Organisms to Radiation .............. ..................

9.1
9.2
9.3
9.4
9.5

Construction of Interim Radwaste Storage Facility at TMI .... ............
Route from TMI through Harrisburg .............
.......................
Routes to DOE Treatment and Storage Facilities .........
................
Routing in Washington to Hanford/Richland Facilities .........
...........
Routing in South Carolina to Savannah River Plant .......
...............

9-15
9-19
9-20
9-21
9-21

... 5-23
5-24... 5-40
5-40

...
...
..
...
..

7-73
7-77

xxxiii

Page

FIGURES
9.6
9.7
9.8
9.9
9.10
9.11
9.12
9.13

Routing in Idaho to Idaho National Engineering Laboratory ..........
Routes to Potential Spent Fuel Storage Facilities ..............
Routes to LLW Disposal Facilities. . . . ... . . . . . . . . . . . . . . . .
Routing in Nevada to Beatty LLW Burial Site
Routing in South Carolina to Barnwell LLW Disposal Site ...........
Routing in New York to West Valley LLW Burial Site .....
...............
Routing in Kentucky to Maxey Flats LLW Burial Site .....
...............
Routing in New Mexico to Los Alamos Site ........
....................

9-22
9-22
9-23
9-24
9-24
9-25
9-25
9-26

1

Flow Schematic of EPICOR-II Processing System

D-15

E.1

York Haven Pond of the Susquehanna River Showing the Location ofIsland and the Nuclear Station, the Generalized Flow Patterns inand over the Dams, and the Peripheral Facilities that Contributethe Pond by Fishermen and Boaters ...............
Generalized Categories of "Eater Groups" of Fishes in York Haven
Susquehanna River . . . . . . . . . . . . . . . . . . . . . . .

E.2................
Three Mile
the Channels
to Use of

E-2
Pond of the
E-3

F.1
F.2

Processed Accident Water Availability. . . . . . . . . . . . . . . . . . . . .
Potential Processed Accident Water Needs . . . . . . ...
. . . . . . . . . . .

F-3
F-5

G. 1
G. 2
G. 3
G.4
G. 5
G. 6
G. 7
G. 8
G. 9

Process
Process
Process
Process
Process
Process
Process
Process
Process

G- 12
G-13
G-16
G-17
G- 19
G-20
G-22
G-25
G-26

H.1

In-Line and In-Drum Mixing Process for Incorporating Radwaste in Cement

1

Long Term Air Monitoring (Three Mile Island) ........

0.1
0.2
0.3

Auxiliary and Fuel Handling Buildings,
Auxiliary and Fuel Handling Buildings,
Auxiliary and Fuel Handling Buildings,

S.1
S.2
S.3
S.4

HTO Concentration Versus Time for Water Ponded in July ...............
HTO Concentration Versus Time for Water Ponded in January ........
Density of 100% Saturated Water Vapor versus Temperature .........
Monthly HTO Concentration above the Pond as Calculated Using Model Two

T.1
T.2

Fraction of Radionuclides Remaining in Dissolved State ......
Chesapeake Bay ...................
..................................

U.1
U.2
U.3

Simplified Decision-Point Diagram . . . . . . . . . . . . . . . . . . . . . . .
Duration of DECON Activities...
.
........................
Postulated Sequence and Schedule of Tasks for DECON of the Containment
Building . . .......
. ..........
.................................
Postulated Sequence and Schedule of Tasks for DECON of the AFH
Building .....
...............
..........
.....................
Duration of Activities Preparing for Safe Storage ...............
Postulated Sequence and Schedule of Tasks for SAFSTOR of the Reactor
Containment Building .................
..............................
Postulated Sequence and Schedule of Tasks for SAFSTOR of the AFH
Building ......................
....................................
Duration of ENTOMB Activities . . . . . . . . . . . . . . . . . . . . . . . . .

U.4
U.5
U.6
U.7
U.8

flow Diagram for Submerged Demineralizer System ....
Flow Diagram for Modified Submerged Demineralizer System
Flow Diagram for EPICOR II System ...........
Flow Diagram for Modified EPICOR II System ..........
Flow Diagram for Zeolite/Evaporation System ......
Flow Diagram for Zeolite/EPICOR II System .......
Flow Diagram for SDS/EPICOR II System .........
Flow Diagram for Evaporator/Resin System ............
Flow Diagram for Bitumen/Resin System .........

281-Foot Elevation
305-Foot Elevation
328-Foot Elevation

....

H-7

...................

M-5

. . .
. . .
. . .

0-2
0-3
0-3
. .
. .
. .
. .

.
.
.
.

.............

S-5
S-6
S-8
S-9
T-3
T-5
U-2
U-16
U-17
U-19
U- 34
U-36
U-37
U-50

xxxiv

FIGURES
V.1
V.2
V.3
V.4
V.5

................................
Reactor Building .......................
Elution of Radionuclides ...............................
Spread of Contaminated Water from Large Volume Releases ..................
.....................
Radius of Contaminant Circle vs. Time ...............
.....................
Flux of Tritium into River ...................

Page
V-2
V-5
... V-1O
V-12
V-13
....

XXXV

FOREWORD
This programmatic environmental impact statement was prepared by the U.S. Nuclear Regulatory
Commission, TMI Program Office, Office of Nuclear Reactor Regulation (the staff), pursuant to theCommission's November 21, 1979, Statement of Policy and Notice of Intent to Prepare a Programmatic Environmental Impact Statement and the requirements of the National Environmental Policy
Act of 1969 (NEPA).
The NEPA states, among other things, that it is the continuing responsibility of the Federal
Government to use all practicable means, consistentwith other essential considerations ofnational policy, to improve and coordinate Federal plans, functions, programs, and resources tothe end that the Nation may:
Fulfill the responsibilities of each generation as trustee of the environment for succeeding
generations.
Assure for all Americans safe, healthful,ing surroundings.

productive,and esthetically and culturally pleas-

Attain the widest range of beneficial uses of the environment without degradation,
health or safety, or other undesirable and unintended consequences.
Preserve important historic,
maintain, wherever possible,
vidual choice.

risk tocultural, and natural aspects of our national heritage, andan environment which supports diversity and variety of indi-

Achieve a balance between population and resource use which will permit high standards ofliving and a wide sharing of life's amenities.
Enhance the quality of renewable resources and approach the maximum attainable recycling ofdepletable resources.
Further, with respect to every recommendation or report on proposals for legislation andother major Federal actions significantly affecting the quality of the human environment,
Section 102(2)(C) of the NEPA calls for preparation of a detailed statement on:Mi)the environmental

impact of the proposed action;(ii)

any adverse environmental
implemented;(iii)

alternatives to the proposed action;(iv)

the relationship between local short-term uses of man's environment and the maintenanceand enhancement of long-term productivity; and,(v)

any irreversible and irretrievable commitments of resources which would be involved in
the proposed action should it be implemented.

effects which cannot be avoided should the proposal be

Information for this statement was obtained from the licensee's Environmental Report and Final
Safety Analysis Report or from the staff's Final Environmental Statement for the Operating
License, dated December 1976, and from new information provided by the licensee or independently
developed by the staff. This information is available to the public.
Any comments by interested
persons received on this information have been considered by the staff.
In conducting therequired NEPA review, the staff met with the licensee to discuss items of information provided,
to seek new information from the licensee that might be needed for an adequate assessment, andgenerally to ensure that the staff had a thorough understanding of the proposed cleanup operations.
In addition, the staff sought information from other sources that would assist in thexxxvievaluation and visited and inspected the project site and surrounding vicinity.
Members of thestaff met with State and local officials charged with protecting State and local interests andheld scoping meetings with the public in Harrisburg, Pennsylvania, on January 29, 1980, in
Middletown, Pennsylvania, on February 12, 1980, and in Baltimore, Maryland, on February 15 andMarch 20, 1980.
On the basis of the foregoing and other such activities or inquiries as were deemed useful andappropriate, the staff made an independent evaluation of the TMI-2 cleanup plans and operations.
This evaluation led to the publication of a draft programmatic environmental impact statement,
prepared by the Office of Nuclear Reactor Regulation.
It was circulated to Federal, State, andlocal governmental agencies for comment, and a summary notice of the availability of the draftenvironmental statement was published in the Federal Register.
Interested persons were invited
to comment on the draft statement.
In addition, the staff held a total of 31 meetings with the
public, local officials, and interested organizations during the comment period on the draft
statement in order to explain the staff's assessments and to solicit verbal comments from individual members of the public and to respond to these comments at that time.
Verbatim transcripts
were made at ten of the major public meetings from which the staff, after review, obtained further comments which were expressed verbally by the public.
These verbal comments were received
as if they were written submittals.
After receipt and consideration of comments on the draft statement, the staff prepared this final
environmental statement, which includes a discussion of comments generated on the draft statement
and the responses thereto.
The comment letters received on the draft statement appear in
Appendix A; responses to the comments are contained in Section 13.
Single copies of this statement may be obtained as indicated on the inside front cover.
Messrs. Paul Leech and Oliver Lynch are the Environmental Project Managers for this statement.
Should there be any questions regarding its contents, they may be contacted at the following
address:
Three Mile Island Program Office
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, DC 20555
Phone:(301)

492-7258

1.
1.1

INTRODUCTION

THE PURPOSE AND SCOPE OF THIS STATEMENTOn November 21, 1979, the Nuclear Regulatory Commission announced its decision to prepare a programmatic environmental impact statement (PEIS) on the decontamination and disposition of radioactive wastes resulting from the March 28, 1979, accident at Three Mile Island Nuclear Station
Unit 2 (see Appendix B).
The PEIS is intended to provide an overall evaluation of the environmental impacts that could result from these activities, beginning when the plant conditions werestabilized after the accident and continuing through completion of the cleanup from the accident.
For purposes of this document, the term "cleanup" is used to mean decontaminating and defuelingthe plant, and disposition of radioactive wastes.
As stated in the Commission's notice, "An overall study of the decontamination and disposal
process will assist the Commission in carrying out its regulatory responsibilities under the
Atomic Energy Act to protect the publichealth and safety as decontamination progresses.
It willalso be in keeping with the purposes of the National Environmental Policy Act to engage thepublic in the Commission's decision-making processes, and to focus on environmental issues andalternatives before commitments to specific cleanup choices are made."
The Commission directed the staff to include in the PEIS an overall description of the activities
and a schedule for their completion, along with a discussion of alternatives considered and therationale for choices made.
This information has been included to the extent it is-presently
available from the licensee.* However, there are many areas of uncertainty regarding the cleanup
operation, as the Commission recognized (Appendix B).
For example, the precise condition of thereactor core will not be known until the reactor vessel has been opened.
Where such is the case,the staff has described and assessed the probable or bounding situations.
If, when more information becomes available, proposed activities are found to be significantly beyond the scope ofthese assessments, appropriate supplements to the PEIS will be issued.
The proposed scope of the statement was discussed with representatives of the President's Council
on Environmental Quality, the licensee, and several state agencies.
Early in the process ofdeveloping the draft PEIS, scoping sessions were also held with the public in Harrisburg andMiddletown, Pennsylvania, and in Baltimore, Maryland.
The comments received were particularly
helpful to the staff in understanding public concerns (see Sec. 1.4) about certain methods ofdecontamination and waste disposal being proposed by the licensee.
Similar comments and many
more specific to the Draft PEIS (see Appendix A) were received during the public comment period
which followed the issuance of that document in August 1980.
After publication of the draft PEIS
for comment, the staff held 31 meetings with the public, local officials, and interested organizations to obtain first-hand the comments and concerns of participants at these meetings and tohave an interchange of ideas.
One major issue, the ultimate disposition of TMI-2, is of interest to the federal, state, andlocal governments as well as the licensee and the public.
However, the disposition of the
facility--whether to decommission or restore it to a condition acceptable for licensed operation-is not within the scope of this statement.
If a decision to decommission Unit 2 is made before
decontamination of the reactor building is far along, it is possible that different methods andchemicals would be used to speed the cleanup of some portions of the plant, as opposed to thosethat might be chosen to minimize damage to the equipment.
Such possibilities are considered in
Section 2.2.
However, a decision to either restore or decommission the facility would probably
not occur until a detailed inspection and engineering assessment is made of the nuclear steam
supply system and further information is known about its condition.
To make this inspection*The term "licensee" or "Met-Ed" in this document refers to Metropolitan Edison Company, theprincipal owner (50%) and operator of the plant; Jersey Central Power and Light Company, andPennsylvania Electric Company, each of which own 25%.

1-1

1-2
requires that the core be removed and the reactor cooling system be decontaminated to a levelthat will not cause excessive radiation exposure to people either onsite or offsite.
However,the staff has assessed in Section 2.2 the alternative of decommissioning even in the event a
decision to decommission is made early in the cleanup.
Decontamination of the auxiliary and fuel handling buildings began prior to the Commission's
decision to issue the PEIS.
For this reason, some of the discussion in Section 5 is historical,
and no alternative methods are presented for work already done.
With regard to decontamination
efforts not yet undertaken, the staff has based its analysis on the licensee's plans where they
are available, as well as alternatives that the staff has independently developed and assessed.
Current estimates of economic costs for the proposed activities and alternatives were not available from Met-Ed or other sources in time for inclusion in the draft statement.
Cost estimates
are provided in this final statement.
To enable preparation of this document, specific cutoff dates were selected upon which to basethe assessments.
Changes in plant conditions that may have occurred since those dates do notmaterially affect the evaluations.
1.2

HISTORY OF THE PLANT AND ITS PRESENT STATUS

On November 4, 1969, following a public hearing, a construction permit for TMI-2 was issued by
the Atomic Energy Commission.
An operating license for Unit 2 was issued on February 8, 1978.*
Between issuance of its operating license and March 28, 1979, TMI-2 had operated for about 95
effective full-power days (or the equivalent of 3165 MW-days per metric ton of low-enricheduranium oxide).
Prior to the accident, the unit had been operating without interruption since
March 7, 1979.
At full power, the TMI-2 system would supply 890 MW (2770 MW thermal) of electric
power to the utility's transmission system; together, the two units had a 1700-MW electric generating capacity.
TMI-2 was operating at 97% power when the accident occurred.
TMI-2 has a pressurized-water reactor with the nuclear steam supply system (NSSS) provided, by
Babcock & Wilcox Company.
Under normal reactor operation, the primary coolant water inside thereactor vessel is maintained at around 575 0 F and 2200 psi.
Heat generated by the fission process
within the reactor core is removed by means of the primary coolant to two steam generators wheresteam is produced to operate a turbine.
The reactor vessel and components (Fig. 1.1) that are exposed to radioactivity are located within
the reactor building (also referred to as the containment building).
The turbine generator,
feedwater system, and electrical generation equipment are housed inside the turbine building.
The remaining support systems, e.g. the high-pressure injection pumps and the makeup and let-downsystems, are located in the auxiliary building adjacent to the reactor building (Figs. 1.2 and
1.3).
At about 4:00 a.m., March 28, 1979, a series of feedwater pumps in the turbine building 'tripped'(stopped operating), resulting in a turbine-generator trip and automatic shutdown of the reactor.
An increase in pressure within the pressurizer activated the pilot operated relief valve (PORV),
which served to prevent overpressure of the primary coolant system by releasing steam.
However,
after the primary system pressure was reduced to the PORV closure set point, the PORV failed toclose.
The light on the control room panel indicating the PORV status showed that electric
current to the PORV solenoid had been terminated, which would normally indicate that closing of
the valve had been accomplished.
This inadequate and misrepresented information prompted theoperators to wrongly assume, for more than two hours, that the PORV had been closed.
During thistime, the continual loss of primary coolant and reduction of pressure in the system, coupled withthe operators' shutting off of the high pressure injection (HPI) system (which had been adding
water to the primary system), led to a drop in the water level, uncovering some of the reactor
core.
Rupture of fuel rod cladding was indicated by the release of excessive radioactivity into
the primary coolant as well as the accumulation in the reactor vessel of hydrogen gas produced
from the chemical interaction between the fuel rod zircaloy cladding and steam.
After initial
uncertainties regarding the presence of a hydrogen bubble in the reactor vessel, and its implications, forced cooling was established at 8 p.m. on March 28, 1979.
On April 3, 1979, the decay
heat level was already down to 5 MW and the primary system was secured to operate at 281°F and
1050 psi with reactor coolant pump IA. On April 7, 1979, the system pressure was further reduced*TMI-1,located at the same site, was issued an operating license on April 12,

1974.

I
(JJ

Figure 1.1.

Reactor Building and Major Components of Primary System.

I-.

Figure 1.2.

Three Mile Island Nuclear Generating Station
Aerial View of the Three Mile Island Station.
Damaged Unit-2
is located on Susquehanna River, about 10 miles south of Harrisburg, Pa.
reactor is housed in the cylindrical building (marked by arrow) on the right of central
plant complex.

R IV

S U SO0 U E H A N N A

UNIT

E R

(E ASTIC HARN N EL )SO SQ-0 E±MATERIAL

UNITi

UNITI

COOLINGTOWER

TOWER
COOLING

LOW-LE
WASTESTAGINGVELRAD
EQUIPMENTAND
FACILITY *UNIT2INTERIMSOLIDWASTE
TOWER
FACILITYW COOLINGSTAGING
INTERIMRADWASTE

F-1STORAGEELECRICAFACILITY

SWITC
HYARD
AND
MID-HIGHLEVEL EQUIPMENT

PROCESSED.. .X .WATERTANKSCHEMICALDo
IS( TNKSFACILITT*
CLEAMING
LOWLEVELWATERmPROCESSING U

BUILDING
EPICOR )BOATD
AT R~OSTORAGETANK(PRESENT
SITEOFTEMPORARY
RADWASTE
STORAGEFACILITV)UNIT2REACTORY B0 LDINGHOTCOOLINGTýOWER

[Ed
IARY

NE

T

REACTOR UNIT

PO

BUI
LDNG
D

2

BUILDINGRESINSOLIDIFICATIONFACXLTYWII

BULaOCDINGNORTHMATERIALRADWASTESTAGING

1..

PERSONNELACCESSEVAPORATOR*OFFICETOWER
COOLFNG
UNIT2 MECHANICALTOWER
DRAFTCOOLING

WATERTOWCOOLING--UIT 2 MEHN

•

I UI

--

NOT TO SCALE--

S

Figure 1.3.

U S

O U

E

H

A

-

-

N

N

A

R I V

E

NOT ALL FACILITIES SHOWN*• FUTURE

-

.

R

NOTE:

ICENTER

Schematic Plan of the TMI-2 Site.

CHANNEL)(Not all features shown.)

1-6
to 400 psi, and on April 27, 1979, cooling through natural recirculation was established without
the need of primary system pumps for operation.
A more detailed description of the accident and its consequences between March 28 and April 3,
1979, can be found in the "Report of the President's Commission on the Accident 2 at Three Mile
Island,"' and "Three Mile Island: A Report to the Commissioners and the Public."
As of December 13, 1980, TMI-2 had been shut down for 628 days and decay heat had decreased toabout 70 kW.
The primary system temperature was then about 1201F and pressure was about 94 psi.
Decay heat was removed by natural recirculation, with the "A" steam generator steaming to themain condenser and the reactor coolant system loops releasing heat to the reactor building atmosphere.
On January 5, 1981, the licensee stopped steaming the "A" steam generator by shutting aturbine bypass valve.
This put the RCS in a "loss-to-ambient" mode of cooling, which is thetransfer of reactor decay heat from both RCS loops to the reactor building ambient. 3 The reactor
is subcritical, with a substantial shutdown margin below criticality provided by the reactor
control rods and boron dissolved in primary coolant water.
Radioactive gases (primarily krypton-85) contaminated the reactor building atmosphere as a resultof the accident.
On June 12, 1980, the Nuclear Regulatory Commission approved the staff's recommendation that the licensee be allowed to decontaminate the building atmosphere by controlled
purging.
The purging operation began on June 28, 1980, and was completed on July 11, 1980,
removing approximately 44,000 Ci of Kr-85.
The first entry into the reactor building since theaccident was made by two Met-Ed personnel on July 23, 1980, and by December 13, 1980, a total of
five entries into the reactor building had been made to survey damage and acquire radiation data.
This information is discussed further in Section 5.2.
The reactor building is flooded to a depth of about 8 ft with water containing radionuclides(about 500,000 Ci, primarily the cesium (Cs) isotopes 137 and 134); and most of the approximately
300,000 ft 2 of exposed surfaces are contaminated with a thin layer of radionuclides (of the orderof 0.01 Ci/ft 2 , primarily Cs-137 and Cs-134).
The auxiliary building was contaminated to a lesser extent by water and gas from the primarysystem and the reactor building.
All of the "accident" water* in the auxiliary building has beenprocessed, and the processing of water used in washing and flushing of contaminated equipment
surfaces is continuing.
As of December 1980, the surfaces of nearly all of the general areas in
the auxiliary building had been decontaminated, and decontamination of the cubicles containingtanks and other equipment was about 80 percent complete.
The level of atmospheric contamination
was below the threshold for unrestricted worker access.
1.3

SUMMARY OF THE LICENSEE'S OBJECTIVES,PROPOSED ACTIONS,AND SCHEDULE

The major objectives of the licensee's TMI-2 decontamination and defueling plan4 are to maintainthe reactor in a safe state; decontaminate the plant; process and immobilize dispersed fission
products; and remove and dispose of the reactor core, with maximum assurance of public health andsafety.
In December 1979, Met-Ed estimated that decontamination and defueling could be accomplished within a time span of about 2 to 2½ years from working entry into the reactor building,
given no unusual technical, regulatory, political, or financial constraints, 4 at an approximate
cost of $400 million.5
In November 1980, Met-Ed extended its schedule by 28 months 6 to reflect the impact of regulatory
constraints and approval processes, and the company's diminished availability of funds forcleanup resulting from an order by the Pennsylvania Public Utility Commission to cease using
operating revenues for this purpose. 7 In consideration of these developments, the utility company has also increased its estimate of6 the cleanup costs, including funds spent in 1979 and1980, to $750 million (in 1980 dollars).
Accordingly, the licensee has adjusted the cleanup efforts "to a level appropriate to the present
and indicated situation while being careful not to adversely impact public health and safety."
The objectives of its revised program8 are to:1.

Maintain the plant in a safe condition with minimum but adequate operating personnel
and site support staff.

*"Accident" water is defined in Section 1.6.2.1.

1-7
2.

Continue limited decontamination of the Auxiliary Building areas,

3.

Continue activities directed at cleanup of the reactor building water (sump and reactor
coolant system).

4.

Continue carefully selected planning, engineering and licensing activities aimed at
Reactor Building decontamination, fuel removal, support of licensing submittals.

5.

Support finalization of the Programmatic Environmental Impact Statement (PEIS).

6.

Continue development of an appropriate Unit 2 Radiological Controls Program.

lines,

tanks.

The revised schedule shown in Figure 1.4 represents the licensee's planning schedule, as ofNovember 7, 1980, for completion of the first two major phases of the effort at TMI-2 by the
spring of 1986.
The licensee's third phase, reconstruction for operation, is not within thescope of this PEIS.
Phase I commenced shortly after the accident on March 28, 1979, with plant cooldown.
The key
events of Phase I include auxiliary building decontamination, accident water processing,
krypton-85 purging from the reactor containment building, containment entry, construction of
additional site support facilities, and reactor building decontamination.
Several of these
activities will extend in time past the start of Phase II.
Phase II will commence with preparation for reactor pressure vessel head removal.
The primary
milestones for Phase II are reactor pressure vessel head removal, fuel removal, and completion of
the reactor coolant system decontamination.
The schedule will be significantly influenced by many factors that are not precisely known at
this time.
These include, but are not limited to, the exact radiation environment and core
conditions encountered, continuing financial limitations and regulatory activities, craft laborand material availability, availability of offsite or onsite radwaste storage and processing, andavailability of offsite disposal capability on a timely basis.
1.4

PUBLIC CONCERNSIn planning this environmental statement, members of the NRC staff held scoping meetings withpublic officials and interested members of the public to provide them an opportunity to identify
subjects which should be addressed in the PEIS.
These meetings were conducted at the locations
listed below on the dates shown:
Harrisburg, Pennsylvania
Middletown, Pennsylvania
University of Maryland Baltimore Campus
Johns Hopkins University, Baltimore, Maryland

January 29, 1980
February 12, 1980
February 15, 1980
March 20, 1980

The comments expressed at these sessions clearly revealed a general concern about some of thepossible decontamination methods and their potential impacts.
Many people in the plant vicinity
indicated apprehension about the relatively small releases of krypton gas from the reactor building which occasionally occur and stated their opposition to purging the gas.
Other participants,
notably those who attended the meetings in Baltimore, were more concerned about the possibility
that releases of "accident water" to the river, even though largely decontaminated and.greatly
diluted, might eventually be authorized by the NRC.
Possible adverse effects on drinking watersupplies downstream and on the marketability of Chesapeake Bay seafood were the primary concerns
relative to such releases of water from TMI-2.
The availability of appropriate disposal sites,
particularly for high-level waste, also was questioned and a number of comments indicated opposition to making TMI into a permanent waste repository.
Additional subjects for consideration in the PEIS were identified from transcripts of the meetings and subsequent letters as follows:* The geographic area evaluated should include downwind and downstream areas to the extent
that they will be affected by any releases.
* There should be a central
toring data.

agency to receive and correlate all air, water,and biotic moni-

1/79
REACTOR

DECAY
DECAY HEATHEAT REMOVALREMOVAL -- STEAMING
LONG TERM "A" GENERATORAUXILIARY AND
FUEL HANDLINGBUILDINGDECONTAMINATIONINITIAL GENERAL AREA DECONTAMINATIONINDIVIDUAL AREA DECONTAMINATIONFINAL GENERAL AREA AND SUMP DECONTAMINATION

1/80

1/81

1/82

1/83

1/84

CONTAINMENT ENTRY AND DATA ACQUISITIONCONTAINMENT DECONTAMINATION SUPPORT SYSTEMSCONTAINMENT DECONTAMINATION - GROSSCONTAINMENTDECONTAMINATIONCONTAINMENT DECONTAMINATION - MANUALREACTOR
DEFUELING
AND RCS
DECONTAMINATION

1/85

tm

PREPARATION FOR RPV HEAD REMOVALINITIAL REACTOR EXAMINATIONHEAD AND UPPER INTERNALS REMOVAL-RPVCORE INSPECTION
FUEL REMOVALREACTOR LOWER INTERNALS REMOVALRCS DECONTAMINATIONPROCESSING OF
RADIOACTIVE
WASTE

CLEAN UP AUXILIARY BUILDING WATERCONTAINMENT PURGE
CLEAN UP CONTAINMENT SUMP WATER
CLEAN UP RC WATER
CLEAN UP CONTAINMENT DECONTAMINATION WATERRADIOACTIVE
SOLID WASTE
MANAGEMENTI INTERIM LINER STAGINGI INTERIM WASTE STAGINGIEQUIPMENT AND MATERIAL STAGINGlEPICOR-11 LINER GENERATIONEPICOR-I LINER SHIPPINGI SOLID RADWASTE SHIPPINGI-----------

E

I

7I
1

ýT'

CONSTRUCTION, PROCESSING ACTIVITIESACTIVITY CONTINUED AS NEEDEDMILESTONE:.
7ACCIDENT, 7 CONTAINMENT PURGE, W7 INITIAL CONTAINMENT ENTRY,4 7 PEIS ISSUED, T7START CONTAINMENTDECONTAMINATION, '7 RPV HEAD REMOVED, '7 FUEL REMOVED, f7 RCS DECONTAMINATION COMPLETE.
ý Figure 1.4.

Licensee's Planning Schedule for TMI-2 Cleanup, Phase I and Phase II.
(Construction of support facilities not included.)

I+4

1/86

1-9
* Geology should be addressed both as relevant to hydrology and to storage of waste material
at the site.
* Ecological studies should include food chains and possible cumulative effects on the biotaof the affected areas, both terrestrial and marine.
* Independent sources should be considered

in determining which of the various alternative

methods should be used at various stages.
• The bibliography should. include the methods,

designs, and authorities considered.

* A separate chapter or appendix should deal with what are determined to be acceptable levelsof radioactive or toxic emissions.
* The NRC should establish a citizens' advisory committee with funding for independent expert
consultants to work with Argonne National Laboratory in preparation of the statement.
* The psychologicalstress issue needs to be considered.

* The impact of the restart of Unit 1 should be considered.
The role of primary and secondary safety systems in the cleanup operation should be minimized.
* Releases that have-occurred since the accident should be addressed.
* Section on offsite exposure and health effects should be expanded.
There should be coherence in the overall decontamination plan.
Acceptance by Department of Energy of all wastes until they become ready for shipment should
be addressed.
Specific radiologicaleffluent release criteria for the cleanup process should be included.

In case Met-Ed does not have sufficient funds to.complete the cleanup,
identify other sources of funds and the authorizations needed.

the PEIS should

Costs of alternative methods for decontamination should not be the basis for selection
versus lowest environmental impact; the usual cost-benefit balance, in which the benefits ofelectrical energy generated usually outweigh the environmental costs, is not appropriate forthis PEIS.
Most of the above subjects were considered by the staff in its preparation of the Draft PEIS,
which was issued for public comment in August 1980, and in this Final PEIS.
As stated in Section 1.1, a major concern regarding Unit 2--whether to restart or decommission it--is not within
the scope of this document.
However, decommissioning of TMI-2 and the resultant environmental
impacts are briefly discussed in Section 2.2.
During the 90-day public comment period, which ended on November 20, 1980, a great variety ofcomments were received.
They are reproduced in Appendix A. The staff's consideration of thesecomments is reflected in the revisions made for this Final PEIS and in specific responses whichappear in Chapter 13.
To afford the public with increased opportunities to make their views known on the cleanup plans
for TMI-2, the comment period was extended beyond the required 45 days to 90 days, and the staff
conducted 31 meetings on the Draft PEIS during that period with various groups of people in
Pennsylvania and Maryland. These meetings and their locations are listed in Appendix Y. A TMIAdvisory Panel of citizens and public officials has also been established to provide communications directly to the NRC Commissioners throughout the cleanup of TMI-2.
1.5

DECONTAMINATION EXPERIENCE AT OTHER NUCLEAR FACILITIESThe removal of unwanted radioactive contamination from materials and equipment is a familiar androutine operation in the nuclear field for reducing radiation levels. 9
Decontaminations of

1-10
various types have been conducted since the 1940s and a vast expanse of experience and technology
is available.
Decontamination operations have been accomplished in the laboratory, in in-reactor
and out-of-reactor test loops, on tools and building surfaces, and in entire nuclear reactor
coolant systems.
Many tests have been conducted to improve decontamination processes and techniques, and these improvements have been utilized on materials and equipment wherever radioactive
contamination is found, for example, universities, hospitals, space satellites, nuclear laboratories, operating nuclear reactors, and fuel cycle facilities and laboratories.
The decontamination techniques employed can generally be classified into the broad categories ofmechanical methods and chemical methods.
The majority of cases usually require only the simpler
mechanical techniques similar to those in ordinary housekeeping, i.e., wiping, brushing, scrubbing with or without detergents or scouring compounds, and wet or dry vacuuming.
More complex
mechanical techniques include water flushes, high-pressure water jets, surface-strippable films,
sandblasting, ultrasonic removal, and even controlled removal of the actual surfaces of contaminated objects.
While not strictly definable as techniques for decontamination, immobilizing
contamination with coatings or covering with shielding materials are often employed for radiationlevel reduction.
Chemical decontamination techniques employ solvents to dissolve or suspend the radioactive contamination.
The solvents can range from simple applications of water or water-containing detergents to complex formulations containing sequestering and chelating agents, surfactants, corrosion inhibitors, oxidizing or reducing agents, and inorganic or organic acids.
Solvents havebeen formulated to remove specific types of radionuclides and for use on specific types of contaminated surfaces.
Decontamination experiences directly relevant to TMI-2 are described in this section to indicate
the extent to which techniques and technologies are available and have been employed successfully(in many cases under conditions much more severe) for the various types of decontaminations thatcan be anticipated for TMI-2.
The relevant experiences include:
Entry into highly contaminated areas.
Examples of building and equipment surface decontaminations applicable to anticipateddecontaminations inside the TMI-2 containment.
Cleanup and decontamination following fuel spills and core failures.
Decontamination of entire reactor coolant systems to remove radioactive
product films and absorbed fission product radioactivities.

corrosion

Post-decontamination operational history.
These experiences illustrate that the basic technologies for decontamination are well establishedand that available techniques can be modified to suit the conditions at TMI-2.
Applicable experience in removing damaged fuel and core components is limited, and development of techniques
specific to TMI-2 will be required.
Some of the past experience relevant to the TMI-2 situation
is summarized below.
1.5.1

Experience with Initial Entries into Highly Contaminated Facilities

The main objectives of the initial entry into a highly contaminated facility are to assess boththe sources and magnitudes of the radiation fields present and-to determine the physical condition of the facility and extent of damage so that the requirements needed for decontamination canbe established.
Thorough planning is required to effectively execute the initial entry.
1,5.1.1

Canadian NRX Reactor

A power surge in the NRX reactor occurred in 1952 causing failure of about 10% of the uranium
fuel rods and releasing approximately 10,000 Ci of fission products to the cooling system, withsubsequent spread throughout the reactor structure.
The building basement acted as a reservoir
for the highly radioactive water, and as the various wetted surfaces dried out, airborne contamination became a major problem.
About one million gallons of radioactive water was generated.
Immediately after the incident, personnel were sent into the reactor building wearing masks andprotective clothing to measure exposure rates.
Rates of 10 roentgen per hour (R/hr) were

1-11
measured at the lower header room along concrete walls and floors and on the basement area walls
and floors.
Rates at the upper area ceilings, walls, and floors were 50 mR/hr, with local hot
spots of up to 3 R/hr.
During decontamination operations, improved procedures for preparing personnel for entry into
radiation field work areas evolved.
These included background seminars describing the incident
and delineating work areas and cleanup objectives; training in the use of protective clothing andequipment; use of mockups areas; use of closed-circuit TV (CCTV) to preview the contaminatedarea; and detailed descriptions of the work to be done, including procedures and estimated times

required. 1o-12
1.5.1.2

Canadian NRU Reactor

In May 1958 during transfer of a failed fuel rod from the NRU reactor to a refueling flask, therod stuck, coolant was lost and the fuel ignited.
Portions of the rod fell into a maintenancepit where it set fire to materials in the pit.
The portion of the rod that fell into the pit
contained 200,000 curies (Ci) of fission products.
Available monitoring instruments at the pit
were offscale (> 1000 R/hr).
Exposure rates elsewhere were 1 R/hr near the fuel flask, 1000 R/hr
two ft above the reactor, 2.5 R/hr in the west reactor hall walls, 100 R/hr on the floor, and30 mR/hr in offices, corridors, and change rooms.
Airborne contamination was too high to bemeasured with accuracy.
When the uranium began burning, the main exhaust fan on the roof wasrunning but was shut off immediately to confine the contamination; nevertheless, some contamination was later found up to 5000 ft downwind.
A team of workers carrying buckets of wet sand buried the burning fuel in the pit within 15 minutes, and the highest dose received was 5.3 rem.
A team of 35 workers working in one-minuteshifts, using long-handled (24-ft) rakes, hoes and shovels, raked the fuel rod onto a tray,
covered it with sand, and then removed it to a solid-waste burial ground.
Despite the highradiation field (as much as 1000 R/hr), the average dose received was 1.4 rem and the maximum
individual dose was 6.4 rem.
The record indicates that extreme care was taken in timing eachexposure and that the workers took advantage of all available shielding. The cleanup was done
under emergency conditions without benefit of detailed preplanning.
During the week after theincident a systematic attack on the decontamination problem was organized, and -during the next
two months some 600 workers took part in the cleanup.13
1.5.1.3

SL-1 Reactor

On January 3, 1961, a nuclear accident occurred at the SL-1, a military reactor at the National
Reactor Testing Station (NRTS) in Idaho, resulting in fatal injuries to three reactor operators.
During the period from January 3 to January 10, an NRTS disaster plan was in effect for therecovery of the three casualties and for the assessment of the reactor shutdown condition.
The
reactor core and pressure vessel were damaged beyond repair. The reactor building incurred only
minor damage but was grossly contaminated.
Only insignificant contamination was released to theenvironment.
Initial operations started within minutes of the accident and were limited to
extremely short entries (seconds) for determining radiation levels and casualty recovery.
Healthphysicists wearing air packs measured exposure rates of 500 to 700 R/hr two ft inside the reactor
room, 500 R/hr at the doorway to the operating floor, 10 R/hr in the control room, 250 mR/hr at
the bottom of the outside steps, and 100 mR/hr at the guard house.
Recovery teams recovered thefirst casualties within six days.
Twenty-two persons received radiation exposures in the rangeof 3 to 27"R whole-body.
The second phase of the recovery operations was undertaken with the objectives of establishing
the nuclear status of the reactor core, assuring that a renewed nuclear excursion would notoccur, and planning for methods and equipment for the cleanup operations.
This work included
radiation surveys, the installation of neutron and gamma monitoring instruments, viewing the topof the pressure vessel head and the vessel interior to determine the condition of the core, anddetermining the water level in the pressure vessel.
Because of the concern over the possibilityof renewed nuclear excursions, all of these operations were performed by means remote to thereactor building interior.
This phase was concluded in May 1961, with the determination that thepressure vessel contained no water and that subsequent nuclear excursions could be prevented by
5
4
keeping the vessel dry.f ''

1-12
1.5.2

Experience with Decontamination of Building Interior Surfaces and Equipment

Methods and experience with routine low-level decontamination are well documented. 9 ' 1 6 '1 7 Items
such as tools, equipment, walls, floors, glove boxes, and hot cells can be readily decontaminatedin place.
Most large nuclear facilities not only practice in-place decontamination, but also
have special facilities to which smaller contaminated articles can be brought for decontamination.
Decontamination of small or completely contained but highly contaminated materials and equipment
also is routinely conducted in the nuclear industry.
However, experience is limited with highlevel decontamination of building interior surfaces and equipment where the contamination has
spread over large areas such as the entire interior surfaces of a reactor building.
1.5.2.1

Canadian NRX Reactor

Because of the large volumes of water involved (1 million gallons) in the NRX accident and thehigh activity (10,000 Ci), the water was stored in the reactor basement.
The basement contained
considerable equipment, including many instruments.
The water was pumped through a pipeline to adisposal area 1¼ miles away and allowed to seep through the soil, which very effectively removed
the activity.12
The containment surfaces were decontaminated by use of the following procedures:
Flush with water,

and pump water to disposal area.

Flush with high-velocity hot water,sometimes with detergents.

Remove all possible equipment for disposal or decontamination elsewhere.
Flush a third time.
Clean or shield local hot spots.
(A "pinhole" camera using bothlight- and gamma ray-sensitive films was useful in finding hot spots.)
Decontaminate concrete by removing surfaces by means of flame priming, chioping, sandblasting and grinding, using a vacuum system to remove the dust. Seal in any remaining
activity.
Finish surface decontamination by scrubbing with cotton wipers using detergents andacids.
(This transfer technique worked better than scrubbing with brushes andflushing.)*

Protect cleaned surfaces from recontamination by covering with paper.

During the course of the decontamination, serious airborne contamination occurred; this wascorrected by improved ventilation, increased hygiene practice by the workers, and continuous
cleaning and mopping of all working areas.
Inhalation problems were solved by closer control of
the use of respirators.
Following decontamination the reactor was repaired and returned toservice 14 months after the incident.12
1.5.2.2

SL-1 Reactor

For this operation, photographic, motion picture, pinhole-camera, and television pictures (manyobtained remotely with a crane) were used extensively, together with the radiation measurements,
to assess the reactor condition and plan the decontamination effort. Training with mockups was
conducted, a control point established, equipment designed, and a new entry made to the fan room
to permit better access.
All operations were planned with the requirement of keeping water fromentering the reactor.
Manual and remote cleanup were used.
Shielding of the reactor head was
installed remotely, tanks were drained, and all loose items were removed.
Next, vacuuming andsweeping were employed to reduce exposure rates to the level at which the reactor vessel could be
removed to a hot cell for examination without incurring excessive worker doses.
Equipment andwall sections were cut out and sent to a disposal area.
Concrete support columns were removed
using a bulldozer, and the steel reinforcing rods were cut.
In slightly contaminated adjoining
buildings, wiping, vacuuming and the use of surface-strippable films were sufficient for thedecontamination.14'15

1-13
1.5.2.3

Surry

The Surry reactor steam generators, in Virginia, were removed in 1978 and 1979 because of multiple tube failures.
Before carrying out the work, it was necessary to decontaminate the reactor
building interior surfaces and the sump.
Initial radiation levels were 500 mR/hr on the interior building surfaces and 2 to 3 R/hr in thesump.
The decontamination was carried out by hot-water scrubbing with household abrasive detergent.
All decontamination solutions were collected and processed through the Surry plant waste
treatment system.
At the end of two weeks the residual radiation levels were reduced to 2 to 3 mR/hr on the surfaces, and 500 mR/hr in the sump.
These low levels allowed the programmed work on the steam
generators and primary coolant piping to continue with a minimum exposure of personnel.
Thedecontamination work resulted in a dose of 23 person-rem and involved 25 people.
In addition,
400,000 pounds of lead was installed to provide temporary shielding in the heat exchanger cells
to reduce exposure rates from 400 mR to about 40 mR/hr.
Total cumulative dose required to instal
1 the shielding was 143 person-rem.18
1.5.3

Experience with Removal of Damaged Cores

While there is considerable experience with the removal of single failed fuel elements,
ence is limited where significant core damage has been incurred.
1.5.3.1

experi-

Canadian NRX Reactor

After a considerable decontamination effort was completed in the reactor building, particularlyin the area beneath the reactor, electrical and piping connections were removed in preparation
for core removal.
All undamaged fuel rods were withdrawn from the reactor.
Many others were fused together.
Anattempt to pull out the defective fuel rods was unsuccessful because of breakage.
The tops ofreactor concrete shields and top thermal shield were removed to provide greater access to thefuel rods.
Three fuel rods were pulled using a cutting procedure.
The third rod being removed
in this fashion broke and allowed radioactive sludge to fall into the basement.
The remainder ofthe fuel rods were grasped near the bottom and pulled out, with the work crew no longer trying tokeep the lower portion of the calandria, or vessel, intact. All the fuel rods were pulled into astandard fuel rod carrying flask or fuel can and transported to a water trench leading to an
extraction plant.
The next major operation was removal of the calandria.
The tank was isolated from the coolingsystems, rigged, lifted, and placed into a large canvas bag for containment.
It then was placedon a skid and towed to a burial ground.
There the calandria, with radiation levels measured to
be about 700 R/hr, was covered with sand for shielding.
The whole calandria removal task was rehearsed using a nonradioactive "dummy" calandria.
The
trained removal crew was able to perform the difficult remote operation efficiently in about
0'
30 minutes.'12
1.5.3.2

Enrico Fermi I Reactor

On October 5, 1966, the Enrico Fermi I nuclear reactor, in Michigan, sustained partial meltdownof two fuel assemblies as the result of coolant-flow blockage.
An estimated 10,000 Ci of fission
products was released to the primary coolant and reactor cover gas.
Special wedging tools of achisel design were employed to break the bond between the two fused elements, and the elements
were removed from the core for examination.
Decontamination operations were successful and wereclimaxed by full-power operation on October 16, 1970.19
1.5.3.3

Canadian NRU Reactor

Following removal of the fuel rod fragments described in Section 1.5.1.2, a large cleanup effort
was organized.
An emergency headquarters was set up with CCTV and public address systems to thereactor hall, and an emergency change room was established.
A vacuum system with a cyclone-typesand filter to prevent clogging was used to remove residual fuel debris.
This operation reduced
exposure rates to 1 R/hr per person and by utilizing 300 workers resulted in an average exposure

1-14
of 1.5 R per person.
Wet mopping with detergent and wiping,provided the remainder of the cleanup.13
1.5.4

with some additional vacuuming,

Experience with Decontamination of Reactor Coolant Systems

Chemical decontamination experience to remove fuel failure debris, including fuel fragments, isvery limited.
Only one reactor, the Plutonium Recycle Test Reactor (PRTR), at Hanford, Washing9 20
ton, has undergone such a decontamination. 'The PRTR reactor system was decontaminated in 1962 after the failure of an experimental
element.
Radiation levels were up to 250 R/hr at low spots.

fuel

There has been little
experience with removing fuel debris from large reactors; however, tests
have been conducted on a number of loops and pilot plants. 2 1 - 2 7 Nitric acid was applied alternately with sodium hydroxide at the decontamination of the Homogeneous Reactor Experiment at Oak
Ridge National Laboratory. 9
A variety of solvents have been applied to desorb and solubilize
fission products from metal surfaces at the Idaho Chemical Processing Plant, 28 indicating that
fuel removal technology is available for application at nuclear plants.
The removal of corrosion product oxide films and accumulated sludge has been a continuing concernof non-nuclear electrical generating plants and chemical process industries for many years.
Mechanical and chemical cleaning processes have been used routinely, and systems are returned toactive service after having been cleaned.
These processes are generally applicable to cleaning
nuclear plants.
The principles and experience have been improved and applied in the nuclear
power industry where similar problems exist plus the added complication of radioactive impurities
incorporated into films and sludges.
In many cases the removal of the radioactive impurities are
the primary reason for the cleaning (decontamination) operation.
The following case histories ofdecontamination and return to active service are pertinent to TMI-2.
1.5.4.1

PRTR

As a result of the reactor's experimental fuel element failure, the normal corrosion productoxide film absorbed a portion of the fission products.
First, mechanical removal of the debris
was accomplished by high-velocity flushing and filtration followed by draining and flushingdead-legs.
Chemical removal of the oxides was then conducted using two applications of buffered
oxalic-peroxide compounds.
Finally, alkaline permanganate followed by inhibited oxalic acid wasapplied to remove residual fission product and corrosion product activities.
The final exposure
rates were near 5 mR/hr and the entire decontamination sequence took ten days.
A second fullreactor decontamination was accomplished in three days using alkaline permanganate solution
9 29 3 1
followed by a citric and oxalic acid mixture. ' 1.5.4.2

Shippingport

The Shippingport, Pennsylvania, pressurized-water reactor (PWR) primary coolant system was chemically decontaminated in 1964 to remove corrosion product oxides and reduce radiation levels.
A
two-step, modified, low-concentration alkaline permanganate-citric acid treatment was used.
The
unique feature of this decontamination was the treatment of all chemical solutions, flushingwater, and dilution water by ion exchange so that no liquid wastes were generated.
Averagedecontamination factors of about 50 were achieved for the pipe walls and steam generators.
As aresult of the decontamination, modifications to the primary coolant piping were made with minimalradiation exposure to personnel.
The Shippingport reactor was returned to active service and has
9 32
operated to the present time. '1.5.4.3

N Reactor

The N Reactor at Hanford, Washington, is a larger and more complex system than a commercial PWR.
This system has been successfully decontaminated internally 11 times to remove the corrosion
product oxides and reduce the radiation levels at the reactor face working areas.
In eachinstance, an inhibited phosphoric acid was employed.
Substantial reductions in radiation levels
were achieved, and the decontamination procedure has become an annual routine that poses no33 3 5
serious problems to reactor operations.

1-15
1.5.4.4

SGHWR Prototype

Both primary loop's of the Steam Generating Heavy Water Reactor (SGHWR), Winfrith, England, were
decontaminated in 1964 and 1971 using a citric and oxalic acid compound and since then have been
decontaminated almost on an annual basis.
In this case, a citric and oxalic acid compound waschosen because the primary concern was the removal of radioactive corrosion product oxides.
In
each case the total decontamination took six days.
This included the chemical contacts, theinitial preparations, and the rinses.
More than 90 percent of the activity was removed in thefirst chemical contact. The decontamination factors of from3 6 1.3 to 39 were achieved. The reactor
system was returned to service with no subsequent problems.
1.5.4.5

BR-3

A two-step decontamination process was carried out in October 1975 in the BR-3 reactor in Mol,
Belgium, to reduce the radiation levels, facilitate steam generator inspection, and provide
access for eventual repairs if required.
The decontamination chemicals used were alkaline permanganate followed by a citric and oxalic acid mixture.
A total of about 230 Ci of corrosion
product activities was removed, and the steam generator radiation levels were reduced from an
initial value of 10 R/hr to an average of 70 to 80 mR/hr.
1.5.4.6

Lingen Power Plant (KWL)

Several critical components of the primary coolant loop of the KWL reactor in the Federal Republic of Germany were removed and decontaminated in 1970.
The decontamination solutions used were:(a) 0.25 percent oxalic acid + 0.25 percent citric acid, (b) 0.2 percent NaOH + 0.2 percent
KMn0 4 , followed by (c) 0.25 percent oxalic acid + 0.25 percent citric acid.
Decontamination3 7
factors of 3 to 120 were reported, and most of the radioactivity removed was cobalt (Co)-60.
1.5.4.7

SENA

The Societe d'energie Nucleaire Franco-Belge des Ardennes (SENA), Chooz, France, pressurizedwater plant was decontaminated in 1968 to facilitate repairs, after 2400 equivalent full-power
hours.
Four steam generators and associated parts of their primary loops were decontaminated forthe removal of radioactive corrosion products using ztwo-step process.
The first step consistedof circulating alkaline permanganate solution.
This-was followed by a 6 percent solution of
citric acid and oxalic acid.
Decontamination factors ranging from 16 to 107 were achieved at thesteam generator inlet and outlet. 3 8
1.5.4.8

Douglas Point Reactor

A recent development in decontamination involves use of dilute solutions and regeneration ofthese solutions while in use with a strong acid cation exchanger in the H+ form.
This process
was used on the Douglas Point heavy-water-cooled PWR at Tiverton, Ontario, Canada.
The decontamination solution used was a mixture of citric acid, oxalic acid and ethylenediaminetetraacetic
acid (EDTA). 3 9 ' 4 0 The decontamination factors realized ranged from 4 to 10, and it is estimatedthat 150 to 180 person-rem dosage was avoided during the maintenance work as a result of thedecontamination.
More recently, the process has been used to decontaminate a portion of the reactor water cleanup
system and let-down heat exchanger of the Vermont Yankee Reactor and Brunswick Reactor plants,
located in Vermont and North Carolina, respectively.
1.5.5
1.5.5.1

Environmental

ImpactOccupationalDoses

The principal environmental impact of the decontamination operations discussed above has been theoccupational doses incurred by the personnel conducting the decontaminations.
For those decontaminations where significant quantities of radioactive materials escaped the reactor coolingsystem into the reactor building (i.e., NRU, NRX, and SL-1), the total occupational doses were
700, 250, and 998 person-rem, and involved 600, 200, and 463 people, respectively.
The maximum
total individual exposure to decontamination personnel was 19 rem.

1-16
1.5.5.2

Offsite Doses

During the subsequent decontamination efforts of even the more severe accidents reported in thepreceding text, there were no measurable offsite doses.
However, in some cases the offsite dose
was not measured.
1.5.5.3

Nonradiological

Effects

The most significant nonradiological environmental impact of the discussed decontaminations was
the use of materials and energy to complete the work.
In comparison to the facilities recovered
and the occupational doses avoided, the use of the materials and energy was insignificant.
Wastes generated were removed to existing radioactive waste disposal sites and the additional
volume had insignificant environmental impact.
1.5.6

Comparison with TMI-2

Much of the experience on decontamination has been with systems smaller than TMI-2 in volume andsurface area.
However, gross contamination and radiation fields have been greater following theaccidents at NRX, NRU, and SL-1 than expected at TMI-2 (based on analyses of samples of sump
water and on the extent of surface contamination of small areas which were removed from thereactor building).
From swipe samples taken during the reactor building entries to date, noparticulate uranium or other fuel debris have been detected.
The surfaces are contaminated withsoluble fission products, primarily Cs-134, Cs-137, Sr-89, and Sr-90 which leached from thebreached fuel into the water and were subsequently distributed throughout the containment.
Pastexperience in decontaminating reactor buildings indicates that early removal of all loose
objects, water washing, and scrubbing will significantly reduce this contamination.
Exposure
rates at the floor levels of the TMI-2 reactor building, excluding "hot spots", are estimated torange from 250 to 500 mR/hr.
Decontamination of the Surry reactor containment in 1979 was performed at exposure rates of up to 500 mR/hr (see Sec. 1.5.2.3).
The first major difference at TMI-2 from past experience is that the reactor building at TMI-2
contained 44,000 Ci of Kr-85, which restricted access to the containment building for more than ayear after the accident.
At all the previous accidents, the workers began recovery operations
immediately and got the job done quickly and effectively.
With passage of time since the accident, there is concern that the surface contamination is becoming more difficult to remove.
The
second major difference is that 700,000 gallons of contaminated water remain at the bottom of theTMI-2 reactor building.
The problem is further complicated by the large quantities of sodium in
the TMI-2 water.
Waste disposal requirements are much more restrictive and complex than for pastexperience.
Experience at other nuclear facilities in removing damaged fuel and core components has beenrather limited, and much of the existing experience is not directly applicable to TMI-2.
Perhapsthe most applicable experience occurred at the NRX reactor in Canada.
The NRX experience has
shown that difficult defueling operations can be completed in a reasonable length of time withfairly low occupational radiation exposure.
The important lesson from NRX is that detailed
planning and the use of mockups for training can be very beneficial in reducing the occupational
exposure.
Other plants have had fuel removed after severe damage; generally these incidents have involved
only a single or a few fuel assemblies.
For the most part these fuel assemblies have been constructed of stainless steel-clad uranium metal fuel.
The TMI-2 fuel is quite different in thatthe fuel is uranium oxide pellets with zircaloy cladding, which is more susceptible to oxidationand embrittlement.
Decontamination of the reactor cooling system to remove residual fuel debris and fission products
can be accomplished with well-established techniques.
Based on PRTR experience, it probably will
be necessary to remove the corrosion product films because the fission products may have beenabsorbed into them.
Removing these films can be accomplished readily with existing solvents; incontrast, aged films are sometimes difficult to remove.
For this special decontamination, it
might be necessary to develop a special oxidizing solvent to condition the corrosion productfilms (replacing the conventional alkaline permanganate solvent) and at the same time dissolve

1-17
the residual uranium oxide fuel debris.
tamination are currently evident:

Two major obstacles to the reactor coolant system decon-

If significant fuel cladding oxidation has occurred, insoluble Zirconium dioxide (Zr0 2 )could be distributed throughcut the reactor coolant system.
Hydrofluoric acid, one of
the few ZrO2 solvents, is too corrosive to be used as a general decontaminant; therefore, the ZrO2 will have to be removed by mechanical means.
Commercial nuclear power plants are not designed with special considerations for largescale decontamination operations.
1.6

REGULATORY REQUIREMENTS,

OTHER CONSTRAINTS,

AND FUTURE CRITERIAThe NRC is responsible for the regulation of TMI-2 cleanup operations to ensure protection of thehealth and safety of the public.
In line with this responsibility, the staff has concluded that
expeditious decontamination of TMI-2, including removal of the fuel from the accident-damaged
reactor, and disposition of the radioactive wastes are necessary.
For this reason, for all
post-accident operations at TMI-2 the NRC has maintained the following regul'atory objectives:(a)

Maintain reactor safety and reactor building integrity,(b)

Assure that environmental impacts are minimized, and that radiation exposures toworkers, to the public, and to the environment are within regulatory limits and are as
low as reasonably achievable (ALARA),(c)

Assure the safe defueling of the reactor and storage of the fuel,(d)Assure the safe collection, packaging,from the decontamination activities.

andstorage and/or disposal of radioactive wastes

Implementing cleanup activities is the responsibility of the licensee.
However, NRC's role inthe decontamination operations may change if the licensee should go bankrupt or otherwise default
on its obligation to decontaminate the TMI-2 facility. NRC's objectives in the cleanup operations would remain the same because of its mandate to protect public health, safety, and theenvironment.
In this section, current regulatory requirements, constraints, and criteria are discussed.
Future criteria pertaining to the cleanup of TMI-2 are also proposed.
These topics include
radiological requirements (occupational exposure limits, radioactive effluent limitations, andradioactive waste disposal requirements) and nonradiological requirements (sewage and industrial
wastewater discharge, and dredging permits).
1.6.1

Regulatory Requirements

Current regulatory requirements affecting the TMI-2 cleanup activities include the applicableprovisions of 10 CFR Parts 20, 30, 40, 50, 70, and 71 of the Commission's Regulations; the terms
and conditions of TMI-2 Operating License No. DPR-73; and the terms and conditions of several
Commission orders and actions relevant to the postaccident stage of TMI-2 and reflecting thenonoperable condition of the facility.
1.6.1.1

10 CFR Part 20 and Part 50

10 CFR Part 20, Standards for Protection Against Radiation, establishes standards for permissible
doses, levels, and concentrations of radiation, precautionary procedure's, and waste disposal.
Sections 20.101 through 20.104 'relate to occupational exposure, and indicate the permissible
doses of radiation and levels and concentrations of radioactive materials in air and water towhich individual radiation workers may be exposed.
Depending on conditions, these regulations
allow maximum individual occupational doses of 3 rem per quarter to the whole body, head andtrunk and active blood-forming organs, lens of eyes, or gonads; 18-3/4 rem per quarter to thehands and forearms or feet and ankles; and 7-1/2 rem per quarter to the skin of the whole body.
Limitations also are placed on the use of respiratory protection and the concentrations of radioactive materials in air and water in restricted areas where workers may be occupationally exposed.
These concentrations for individual radionuclides are specified in Table I, Column 1 of Appendix B to 10 CFR Part 20.
Additional restrictions on the concentrations of radioactive material

1-18
in effluents to unrestricted areas are contained in Table II of Appendix B to 10 CFR Part 20.
Formulae and procedures for applying these concentrations to mixtures of radionuclides also are
specified.
Sections 20.301 through 20.305 deal with disposal of solid, liquid, and airborne wastes.
These
sections establish standards and procedures governing disposal of wastes by (1) transfer toauthorized recipients, (2) release to sanitary sewage systems, (3) burial in soil, and (4) othermethods approved on a case-by-case basis.
Specific approval is required for disposal at sea andincineration of wastes, as well as for other disposal activities that do not meet the generallyprescribed criteria.
Generally, wastes discharged to sewage systems must be soluble or dispersable in water and meet the concentration and quantity limits (e.g., 1 Ci/yr) specified in
Section 20.303.
Appendix I to 10 CFR Part 50 provides numerical guides to meet the ALARA criterion of 10 CFRPart 20 for releases of radioactive materials in liquid and gaseous effluents to unrestricted
areas.
The numerical guidelines of Appendix I to 10 CFR Part 50 have been selected as the appropriate effluent discharge limits for the cleanup operations and are reflected in Appendix R.
1.6.1.2

10 CFR Parts 30,

40 and 70

Commercial land burial facilities presently are licensed on a case-by-case basis under 10 CFR
Parts 30-, 40, and 70, or equivalent regulations in the case of Agreement State licensing actions.
Such facilities must be located on federal- or state-owned land under the provisions of 10 CFR
Section 20.302(b).
NRC intends to allow the disposal of TMI-2 wastes at existing disposal sites provided that therisk to the public (both present and future generations) is similar to that presented by disposalof wastes routinely generated and disposed of at those disposal sites. Clearly, the existing
disposal site requirements would need to be met.
At TMI-2, organic resin wastes will be generated.
The upper levels of the bulk specific activity
organic resin wastes routinely generated at other nuclear power plants have been on the order of
1 to 10 Ci/ft 3 of Cs-137 total activity.
In this range of specific activities the resins will be
exposed to approximately 108 rads after 300 years, or ten half-lives of Cs-137, assuming all of
the energy is absorbed in the waste form.
At exposures of 108 to 109 rads, organic resins could
undergo significant decomposition, which could result in increased mobility of the collectedradionuclides, gas evolution, and potential acid formation.
Therefore, to ensure adequate radiation stability of the organic resin over ten half-lives, resins having bulk specific activities
greater than 10 Ci/ft 3 will undergo specific evaluations to ensure that radionuclide mobility andother impacts within the waste container are minimized over the hazardous lifetime of the wastes.
Some of the TMI-2 wastes fail to fit established patterns for nuclear power plant wastes.
Therefore, the proposed TMI waste forms need to be separately assessed by the NRC with respect to theobjectives of the proposed 10 CFR Parts 60 and 61.
The assessments of these nonroutinely generated wastes can best be performed on a case-by-case basis since there are many uncertainties
regarding the characteristics of the TMI-2 wastes, and the processing and disposal options which
are appropriate must be individually determined.
Until the uncertainties of the waste characteristics are resolved, non-routinely generated wastes
would have to be retrievably stored.
The waste container designs would be required to provide
mechanical and structural stability over the intended storage period, during onsite and offsite
handling, and throughout the transportation phase.
For wastes found to be unacceptable for routine disposal at the existing burial sites, moresophisticated disposal methods would be required.
These methods might include additional container requirements, different waste processing modes, modifications to routine burial methods
and procedures, or disposal in a high-level-waste repository.
The NRC staff intends to address the disposal criteria for nonroutine wastes on a case-by-casebasis, giving consideration to the waste forms proposed by the licensee.
This approach ensures asuitable basis on which to continue to protect the public health and safety and the environment
and at the same time offers the licensee adequate flexibility in processing and package designs.
In addition to criteria placed on TMI-2 wastes based on the staff assessments, specific requirements based on Commission orders need to be complied with.
As an example, the Commission has

1-19
ordered that EPICOR-II resins must be solidified prior to shipment offsite.
This order is compatible with recently imposed disposal site license conditions at the Hanford, Washington, andBarnwell, South Carolina, burial sites requiring that after June 30, 1981, all resins having
specific activities greater than 1 microcurie per milliliter (1 pCi/mL) (0.028 Ci/ft 3 ) for isotopes with half-lives greater than five years must be stabilized by solidification.
The Barnwellsite, however, will also consider disposal of dewatered resins in a high-integrity container.
Ahigh-integrity container is one which will maintain its mechanical and structural characteristics
over the hazardous lifetime of the wastes.
1.6.1.3

10 CFR Part 71

NRC regulations on the packaging and transport of radioactive materials are contained in 10 CFRPart 71.
This part also points out the applicability of regulations of the Department ofTransportation (DOT)
regarding radioactive material shipments.
All shipments of radioactive
materials--sources, samples, fuel, and wastes--will be made in accordance with this part and170-189).
applicable DOT regulations (49 CFR Parts
1.6.1.4

TMI-2

Operating

License,Commission Statements and Orders

Related to TMI-2 Cleanup

Several Commission Statements, Orders and amendments to the TMI-2 Facility Operating License(DPR-73) have been issued subsequent to the March 28, 1979, accident. The following listis asummary of these actions.

Date

Description

May 25, 1979

Statement - preparation of an Environmental Assessment for the operation of
the EPICOR-2 system.
Prohibited the discharge of accident-contaminatedwater.

July 20,

Order for Modification of License - Authority to operate facility suspended,
licensee to maintain facility in a shutdown condition in accordance withapproved operating and contingency procedures.

1979

October 16, 1980

Memorandum and Order-Commission approved operation of EPICOR-II system andprovided its rationale for this decision.

October 18, 1979

Order for Modification of License -- Authorized operation of EPICOR-II system,
added technical§ specifications requiring monitoring of EPICOR-II discharge
paths, required licensee to maintain tankage at TMI-1 for possible storage of
waste water from TMI-2, and prohibited offsite shipment of spent resins until
approved by the Director of NRR.

October 26, 1979

Clarifying Amendment to Order for Modification of License - Stated that ahearing is not necessary prior to operation of EPICOR-II.

November 21, 1979

Statement of Policy and Notice of Intent to Prepare a Programmatic Environmental Impact Statement - Stated that the Commission had decided to prepare aprogrammatic environmental impact statement on the decontamination and disposal of radioactive wastes resulting from the March 28, 1979, accident atThree Mile Island Unit 2.

February 11, 1980

Order - Issued new proposed Technical Specifications which superseded all
pre-accident Appendix A Technical Specifications and Appendix B Technical
Specifications 5.1, 5.2, and 5.3 which specified the licensee's pre-accident
management organization.

March 12, 1981

License Amendment No. 10 - Formally amended license to add provisions ofOctober 18, 1979, EPICOR-** Order for Modification of License.

April 7,

Approved - Interim Criteria for Radiological Effluents from TMI-2 Data
Gathering and Maintenance Operations.

1980

1-20

Date

Description

June 12, 1980

Memorandum andiOrder - Commission approved decontamination of the reactor
building atmosphere by controlled purging to the atmosphere and provided its
rationale for this decision.

June 12, 1980

Order for Temporary Modification of License - For duration of reactor building purge, deleted certain sections of Appendix B Technical Specifications
which limited gaseous effluent releases based upon Curies/second and imposed
limits based upon exposure.

June 24, 1980

License Amendment No. 11 - Permitted bypassing of the interlocks from certain
radiation monitors during purging of the reactor building.

August 11,

Modification of Order - Deleted proposed Technical Specifications operability
requirements (which has been imposed by the Order of February 11, 1980) forthe balance of plant diesel generators and the 13.2 kv circuit from theMiddletown Junction Substation.

1980

September 26, 1980

Statement of Policy - Emphasized most strongly that all of our health, safety
and environmental requirements applicable to Three Mile Island Unit 2 must be
fully complied with by Met Ed. Also stated that in the event of any conflict,
NRC health, safety and environmental requirements must supersede State agency
requirements that result in a lesser degree of protection to the public.

November 14,

1980

Amendment of Order - Added proposed Technical Specification operability
requirements for the Mini Decay Heat Removal System and deleted proposed
technical Specification operability requirements for certain balance of plant
equipment no longer required with addition of MDHRS.

January 29, 1981

License Amendment No. 12 - Revised license condition 2.E(3) so that tankage
to store waste water would no longer be required to be reserved in TMI-1 but
would rather be required to be reserved in TMI-2.

1.6.1.5

Order Regarding Proposed Technical Specifications

By Order of the Director, NRR, dated February 11, 1980, a new set of formal license requirements
was imposed to reflect the postaccident condition of the facility and to ensure the continued
maintenance of the current safe, stable, long-term cooling condition of the facility(45 F.R. 11282).
The requirements:(1)

Define operating parameters for the current safe, stable, long-term cooling mode forthe facility (defined as the recovery mode), and delete all other permissible operating
modes so as to assure that operation of the facility in other than the stable shutdown
condition of the recovery mode is precluded;(2)

Impose functional, operability, redundancy and surveillance requirements as well as
safety limits and limiting conditions with regard to those structures, systems, equipment and components necessary to maintain the facility in the current safe, stable,
shutdown condition and to cope with foreseeable off-normal conditions;(3)

Prohibit venting or purging or other treatment of the reactor building atmosphere, thedischarge of water decontaminated by EPICOR-II system, and the treatment and disposal,of high-level radioactively contaminated water in the reactor building, until each ofthese activities has been approved by the NRC,* consistent with the Commission's Statement of Policy and Notice of Intent to Prepare a Programmatic Environmental Impact
Statement (44 F.R. 67738) (see Appendix B of this document).

*By Memorandum and Order, dated June 12, 1980, the Commission gave the approval contemplated bythis restriction insofar as necessary for the licensee to conduct a purging of the TMI-2 reactor
building in accordance with procedures approved by the NRC.

1-21
These requirements were set forth in a new set of proposed Technical Specifications contained in
an attachment to the Order.
Section 6.8, Procedures, of these Technical Specifications requires that the licensee shallestablish, implement and maintain written procedures covering the following activities:a.

The applicable proceduression 2, February 1978.

b.

Recovery Operations Plan implementation.

c.

Surveillance and test activities of safety related equipment and radioactive waste
management equipment.

d.

Security Plan implementation.

e.

Emergency Plan implementation.

f.

Radiation Protection Plan implementation.

g.

Recovery Mode implementation.
(Specifically Recovery Mode procedures which involve a
reduction in the margin of safety, including those which:)

recommended in Appendix "A" of Regulatory Guide 1.33, Revi-

1.

Directly relate to core cooling.

2.

Could cause the magnitude of radiologicalthe NRC.

3.

Could increase the likelihood of failures in systems important to nuclear safety
and radioactive waste processing or storage.

4.

Alter the distribution or processing of significant quantities of stored radioactivity or radioactivity being released through known flow paths.

releases to exceed limits established byPending approval by the NRC of written procedures required by a.
shall follow the previously approved procedures.

through g.

above,

the licensee

Each procedure above, and changes thereto, shall be reviewed by the Plant Operations Review
Committee and approved by the Manager Site Operations prior to implementation and reviewed periodically as set forth in administrative procedures. . Each procedure of b. and g. above, andchanges thereto, shall be submitted to the NRC prior to implementation; these procedures, andchanges thereto, shall also be subject to approval by the NRC prior to implementation.
The
Manager Site Operations shall have responsibility for determining which procedures are in the b.
and g. category.
Temporary changes to procedures a.,

c.,

d.,

e.

and f. above may be made provided:a.

The intent of the original procedure is not altered.

b.

The change is approved by two members of the unit management staff, at least one ofwhom holds a Senior Reactor Operator's License on the unit affected.

c.

The change is documented, reviewed by the Plant Operations Review Committee andapproved by the Manager Site Operations within 14 days of implementation.

Temporary changes to procedures of b. and g. above may be made provided the provisions above aresatisfied and the change is submitted to the NRC for review within 72 hours following approval by
the Manager Site Operations.
1.6.1.6

EPICOR II Waste Treatment System

On March 12, 1980, Amendment No. 10 to License No. DPR-73 was issued (page 1 of Ref. 41).
This
amendment revised the Technical Specifications contained in Appendix B to the license and established new requirements for radioactive waste treatment, storage, and shipment involving the
EPICOR II waste treatment system.

1-22This amendment also added the following license conditions:*
2.E.(2)The licensee shall promptly begin the process of decontaminating the intermediate-level
waste water from TMI-2 by operating EPICOR-II.
Prior to operation, the licensee shallconsult the Director of (the Office of Nuclear Reactor Regulation) NRR for approval of
the final operating procedures and design and construction details.
In order to reducethe inherent risk from the contaminated water most expeditiously and prudently, thelicensee should, to the extent possible, process all the water once through the
EPICOR-II system.

2.E.(3)The licensee shall maintain suitable tankage at TMI-1 that could be used to store waste
water from TMI-2 at an appropriate state of readiness, should additional storage become
necessary.

2.E.(4)The licensee shall not ship spent resins offsite unless they have been solidified, andonly then with the prior approval of the Director of NRR, provided however, that thelicensee may ship nonsolidified but dewatered spent resins offsite if it determines,
and the Director of NRR concurs, that such a shipment is required to assure continued
operation of EPICOR-II or otherwise required to protect public health and safety. The
licensee shall expeditiously construct a facility for solidification of the spentresins and shall use such facilities for resin solidification upon receiving the
Director of NRR's concurrence with the design and operating procedures.

1.6.1.7

Interim Criteria for Radiological Effluents from Data Gatheringand Maintenance Operations

On April 7, 1980, the NRC approved specific radiological effluent criteria for the purpose ofdata gathering and maintenance operations for TMI-2 during the period prior to issuance of the
PEIS. 4 2(Releases which are specifically not covered by these criteria are purging of thereactor building atmosphere, disposal of EPICOR-II water, and treatment and disposal of highlevel radioactively contaminated water in the reactor building.)
In essence, these interim criteria established a modification of the 10 CFR Part 50, Appendix I,
design criteria for specific TMI-2 activities as effluent limits.
The criteria indicate what
information the licensee must submit to the NRC for approval prior to performing these activities, and the type of review that the staff will perform in determining whether or not to approveeach request.
These criteria are as follows:
The licensee must request approval from the NRC to perform data gathering and maintenance operations.
In addition, separate procedures must be developed for each operation and submitted to the NRC for approval.
These procedures must contain a description of the need for the operation, estimates of radioactivity that may be released,and estimates of onsite and offsite doses that may occur as a result of the operation.
The procedures for each operation should be designed to conform to the existing NRC
technical specifications as well as to the "As Low As Reasonably Achievable" (ALARA)
concepts of 10 CFR Parts 20 and 50.
The procedures developed by the licensee should
not interfere with the applicability of other limitations, conditions, or agreementsthat the licensee may have regarding the releases of radioactive gaseous or liquid
effluents with NRC, or with other federal, state, or local authorities.
These procedures will be reviewed by the NRC to ensure that they meet the existing
technical specifications, that the ALARA concepts of 10 CFR Part 20 and 10 CFR Part 50
are met, and that the existing Appendix I to 10 CFR Part 50 design objectives areconformed to, and that they conform to agreements to which the NRC is a party.
The Deputy Program Director, TMI Program Office, onsite now has the authority to permit weekly
releases which result in offsite doses that are not greater than 5 percent of the annual
Appendix 1 10 CFR Part 50 design objectives normalized to a weekly rate (i.e., 0.05 times theannual design objective divided by 52).
These permitted releases will allow the onsite TMI
manager the flexibility to continue or authorize decontamination procedures while keeping*Page 2 of Reference 41.

1-23
releases at a small fraction of those evaluatedSupplement to the FES of 1976 for Unit 2.

in the FES of 1972 for Units 1 and 2 and the

The Director of the Office of Nuclear Reactor Regulation (NRR) now has the authority to permit
weekly releases which result in offsite doses that are. not greater than 50% of the annual Appendix I to 10 CFR Part 50 design objectives normalized to a weekly rate (i.e., 0.50 times theannual design objective divided by 52).
Releases which may result in offsite doses in excess of those described above require approval by
the Commission.
Any impacts that will occur under the interim criteria are expected to be a small fraction of
those described in the TMI Final Environmental Statement (FES) of 1972 and the Supplement to the
FES in 1976.
Nevertheless, recognizing the sensitivities of the local citizens to any activities
at TMI, the NRC staff will inform the local governmental officials of any releases before allowing them to take place.
The public will also be notified through appropriate press channels.
This notification will be done in conjunction with established working relations with the Environmental Protection Agency.
If faced with an emergency, the staff may find it necessary to makeadjustments in these procedures and would use whatever means are available to keep the public
informed.
1.6.2

Other Constraints

In addition to NRC regulatory requirements, there are other constraints that affect activities at
TMI.
These constraints include the Settlement Agreement with the City of Lancaster and various
nonradiological constraints expressed in dredging permits and discharge permits issued by various
Federal and Commonwealth of Pennsylvania agencies.
1.6.2.1

Settlement of the City of Lancaster Lawsuit

In the case of the City of Lancaster v. United States Nuclear Regulatory Commission, Civil Action
No. 79-1368 before the U.S. District Court for the District of Columbia, a settlement was enteredon February 27, 1980, between the parties to that action:
City of Lancaster, City of Lancaster
Authority, Albert B. Wohlsen, Jr. , -the United States Nuclear Regulatory Commission, and theMetropolitan Edison Company, Jersey Central Power and Light Company, and Pennsylvania ElectricCompany.*
As a condition of the settlement agreement, the NRC and the utilities agreed, among
other things, that "... no accident-generated wastewater will be discharged into the SusquehannaRiver from the date of this Settlement Agreement through December 31, 1981, or until the NRC
completes its Programmatic Environmental Impact Statement ....
or until the NRC completes suchother environmental review referred to in its November 21, 1979, Statement ....
regarding thedischarge of accident-generated wastewater 'into the Susquehanna River, whichever is earlier.""
However, nothing contained in the Settlement Agreement precludes either the NRC or the Directorof the Office of Nuclear Reactor Regulation from authorizing whatever measures it or he deems
necessary to cope with any emergency situation.j
The agreement provided specific definitions for "accident-generated water":**(a)

Water that existed in the TMI-2 auxiliary, fuel handling, and containment buildings
including the primary system as of October 16, 1979, with the exception of water which
as a result of decontamination operations becomes commingled with non-accidentgenerated water such that the commingled water has a tritium content of 0.025P Ci/mL
or less before processing;(b)

Water that has a total activity of greater than 1 pCi/mL prior to processing except
where such water is originally non-accident water and becomes contaminated by use incleanup;(c)

Water that contains greater than 0.025 pCi/mL of tritium before processing.

*Settlement Agreement,

p.l.

"*Settlement Agreement,p.

2.

tSettlement Agreement,

p.

3.

1-24
The agreement also established requirements for NRC notification of appropriate public authorities in the event a release occurred or was contemplated.
1.6.2.2

Dredging Permits

Two dredging permits presently are in effect for TMI, one issued by the U.S. Army Corps of Engineers (COE) (Permit NABOP-F/4 75-1048, with amendment) and one issued by the Pennsylvania Department of Environmental Resources (PDER) (Permit 2275724).
The COE permit was issued on January 19,
1976, for maintenance dredging around the cooling water intakes for a period of ten years, expiring on December 31, 1986.43 It was amended on March 15, 1976, to allow for relocation of spoil
disposal sites on the island within the dike system surrounding the TMI facility. 4 4 The PDER
dredging permit was issued on January 13, 1976, and expires on December 31, 1981 (Sec. 1, p. 8,
item 10 of Ref. 45).
1.6.2.3

Discharge Permits

The following discharge permits regulating sewage and industrial wastewater are currently ineffect for the TMI facility:
Pennsylvania Department of Environmental Resources Sewage Permit No.2275419.This permit wasissued on January 23, 1976, for the purpose of approving the construction of a sewage treatment
facility and for the discharge of treated sewage to the Susquehanna River. 4 6
Soil and erosion
control conditions were also made part of the permit. The permit was amended on August 10, 1977,
to extend it to January 23, 1980.17 This permit was again extended on August 23, 1979, to expire
on January 23, 1982.48 However, the TMI sewage plant currently is not in service. 4 9
Industrial Waste Permits.
Four industrial waste permits issued by the Pennsylvania Department ofEnvironmental Resources currently are in effect for TMI (Sec. 1, p. 8, item 10 of Ref. 45):
Permit 2270202 (issued June 8, 1973), Permit 2270204 (issued August 17, 1971), Permit 2270209(issued February 15, 1977), and Permit 2277206 (issued February 22, 1978).
These industrial
waste permits cover thermal effluent criteria for discharge from mechanical draft cooling towers,
treated industrial wastes, and other industrial wastewater to the Susquehanna River.
National Pollutant Discharge Elimination System (NPDES) Permits. The NPDES permits" issued by the
USEPA for the TMI facility would have expired on January 30, 1980.
However, a timely application
for a new permit was made on August 15, 1979, to the Commonwealth of Pennsylvania, which now has
NPDES permit authority. 5 1 The EPA permit will remain in effect while that application is still
pending before the Pennsylvania Department of Environmental Resources.
The NPDES permit is acomprehensive discharge permit covering a variety of discharge points from the TMI facility:
surface water, sanitary, storm water transport system, surface impoundment, waste acceptanceforms, evaporation, consumption, and other miscellaneous points.
There are 27 discharge points
indicated in the application for TMI (pp. 2-5, Ref. 45).
Specific discharge limitations on heat,chemicals, oil and grease, and other pollutants are identified in the NPDES permit.
1.6.3

Future Criteria

Criteria affecting the TMI-2 cleanup proposed herein are expected to be developed over time as aresult of issuance of this PEIS or otherwise.
Future waste management and disposal criteria forlow-level radioactive wastes and proposed radiological effluent limits for decontamination activities are presented.
Decontamination criteria for removable radioactive contamination are also
indicated. These decontamination criteria have been applied to the cleanup activities discussed
in the body of this statement.
1.6.3.1

Future Radioactive Waste Disposal Criteria

Low-Level Waste
The staff currently is developing comprehensive regulations for low-level waste (LLW) disposal
which will contain technical requirements and procedures for licensing LLW disposal facilities.
A preliminary draft of these regulations, 10 CFR Part 61, was made available for public commentin the Federal Register (FR, Vol. 45, No. 41, February 28, 1980).
An environmental impact statement (EIS) relative to 10 CFR Part 61 is being prepared.

1-25
Part 61 will establish the overall performance objectives to be achieved in the disposal of LLW,
dealing with long-term protection of the groundwater, protection from an inadvertent intruder,and assurance of protection of public health and safety during site operations.
Specific technical requirements will be established regarding (1) the form and content of waste that would be
acceptable for disposal at a near-surface disposal facility, (2) the design, operation and closure of the facility, and (3) institutional controls on monitoring, surveillance and use of thesite after closure.
Procedures for the filing, content, and commission action on applications
will also be established.
The technical requirements on waste form and content, facility designand operations will address specific waste disposal concerns currently being considered in analyzing disposal of particular TMI wastes..
It is scheduled for 10 CFR Part 61 to be issued as a proposed rule during 1981 and as a final
rule in 1982.
In the interim, the staff would evaluate on a case-by-case basis the acceptabilityof wastes for disposal and applications for new disposal sites following the objectives andconsiderations set out in the draft of Part 61.
High-Level Waste
The NRC is vested with licensing and regulatory authority over certain U.S. Department of Energy(DOE) facilities by Sections 202(3) and (4) of the Energy Reorganization Act of 1974.
These
sections refer to:(1) facilities used primarily for the receipt and storage (including disposal) of high-level radioactive wastes (HLW) and (2) retrievable surface storage and otherfacilities authorized for the express purpose of long-term storage of HLW.
Geologic repositories would not be licensed as "production" or "utilization" facilities.
Rather,they would be licensed under those provisions of the Atomic Energy Act dealing with receipt andpossession of "byproduct" and "special nuclear" materials.
The NRC has the responsibility to evaluate all aspects of repository performance which could
affect public health and safety.
As a first step in carrying out its responsibility, the NRC
staff is developing the regulations under which a geologic repository will be licensed.
This
regulation will be codified in 10 CFR Part 60 - Disposal of High-Level Radioactive Wastes in Geologic Repositories.
In 10 CFR Part 60, which is currently waiting approval before the Commission, high-level wastes
are defined as:(1)

irradiated reactor fuel,(2)

liquid wastes resulting from the operation of the first cycle solvent extraction system, or equivalent, and the concentrated wastes from subsequent extraction cycles, or
equivalent, in a facility for reprocessing irradiated reactor fuel, and(3)solids into which such liquid wastes have been converted.

The overall performance objective for the high-level waste repository (i.e., the maximum allowable release of radionuclides to the biosphere) will be established by the U.S. Environmental
Protection Agency (EPA) in their Environmental Radiation Protection Standards.
The NRC staff
will implement these requirements in 10 CFR Part 60.
The procedural part of 10 CFR Part 60 isscheduled to be noticed in the "Federal Register" as a final rule in early 1981.
The technicalpart was noticed in the "Federal Register" as an advanced notice of proposed rulemaking in May
1980 and is scheduled to be noticed as a final rule in late 1981.
The technical criteria of
10 CFR Part 60 will apply to all wastes emplaced in a geologic repository for high-level wastes
and, therefore, would apply to TMI wastes disposed of in this manner.
Until very recently the designs of deep geologic repositories have placed almost total reliancefor containment of the radionuclides on the site characteristics and surrounding geology.
As aresult, little
attention has been given to waste forms, containers, and other engineered barriers
to significantly contribute to the containment and isolation of the radionuclides for extensive
periods of time.
For example, credit was not normally given for the waste form and its associated packaging being able to contain the radionuclides for more than a few tens of years.
In order to compensate for the uncertainty in predicting the behavior of geologic systems overlong periods of time, the technical criteria of 10 CFR Part 60 are based on a conservative

1-26
multi-barrier approach.
In this approach, the repository consists of three major barriers:(1) the waste package, (2) the underground facility, and (3) the geologic setting.
The NRC will
establish minimum performance objectives for each of these major barriers.
The major performance objectives for the engineered system are (1) a waste package that isdesigned to contain wastes for at least 1000 years; (2) after package failure, the system isdesigned to limit the annual rate of release of radioactive material to less than one part in one
hundred thousand of the maximum amount of each radionuclide calculated to be present in theunderground facility at any time after 1000 years; and (3) the repository must be designed sothat all the waste could be removed starting at any time up to 50 years after the termination ofwaste emplacement.
In addition to these performance objectives, the NRC staff is developing
guidance for the technical community regarding how to meet these performance objectives.
1.6.3.2

Proposed Criteria for Radiological Effluents from Decontamination Activities

The NRC staff proposes modifications of the TMI-2 technical specifications relating to radiological effluents resulting from the decontamination program.
These modifications are discussed
below:The TMI-2 Technical Specifications will be supplemented to require that radioactive effluents from the cleanup and decontamination operation of TMI-2 be limited so that the numerical design objectives of Appendix I to 10 CFR Part 50 will not be exceeded.
Additionally,
the licensee will be required to maintain radioactive effleunts as far below the Appendix I
objectives as practicable.
The licensee will have to maintain cumulative estimates ofoffsite population doses and maximum individual doses.
Quarterly, the licensee must submit to NRC dose estimates based on all actual releases which
occurred over the previous three months, including accidental and routine releases.
Dataand methods used to make these estimates should be provided or appropriately referenced.
Actual environmental conditions should be used in making these estimates.
Also, calculations as described above should be submitted to NRC within 30 days after each accidental
release.
These proposed modifications to the technical specificationare indicated in Appendix R. These
modifications are proposed for the purpose of implementing the requirements of Appendix I to
10 CFR Part 50 and to assure that the offsite doses that may occur are as low as reasonably
achievable, while at the same time, do not exceed the numerical design objectives of Appendix I.
It is necessary to evaluate each release for conformance to the ALARA principles of 10 CFRPart 50, Appendix I, for two reasons.
First, the numerical design objectives of 10 CFR Part 50,
Appendix I, were designed for routine operation of a light water reactor and not for a cleanup
and decontamination operation.
They were based on the evaluation of years of reactor operational
data.
Since there is much less experience in large-scale decontamination activities there islittlebasis on which to establish numerical design objectives.
Thus, it is necessary to evaluate each release for consistency with the "as low as reasonably achievable" (ALARA) principle.
Second, doses that have occurred offsite during the calendar year must be taken into consideration.
If accidents such as those described in the PEIS occur, doses offsite could be significantly larger than those described for routine decontamination operation, and could even exceed
the numerical design objectives of 10 CFR Part 50, Appendix I, for normal operations of a lightwater reactor.
If such a case occurs, the NRC may require a temporary termination of the cleanup
operation or other action as deemed necessary.
The basis used in approving latter programs will
depend in part on the success of the earlier ones in achieving ALARA releases.
Requiring that the numerical design objectives of Appendix I to 10 CFR Part 50 are met willassure that the radiation dose received by the public during the cleanup operation is sufficient
to protect the public and is equivalent to or below that of a normal operating reactor'.
These
doses, even when added to the doses which occurred during the TMI-2 accident (maximum dose lessthan 100 mrem) are likely to have negligible health effects to individuals of the population.
The background radiation in the area amounts to about 116 mrem per year, 36 percent of whichcomes from cosmic radiation, 39 percent from terrestrial radiation, and 24 percent from internalradiation (mainly K-40 deposited in the body).
On the basis of comparison of the doses calculated here to those of natural background radiation, the health effects through the completion of
the cleanup operation will be non-existent, especially in consideration of the fact that natural
background radiation in the United States varies from one location to another within a range ofabout 70 to 310 mrem per year.

1-27
Adherence to the proposed modifications to the existing technical specifications will assure thatthe cleanup operation will proceed in a manner that provides a wide margin of public safety. The
total doses offsite to the maximum exposed individual or to the population throughout the courseof the cleanup operation should be a small fraction of that received due to background radiation.
These requirements will also assure that any new releases are as low as reasonably achievable.
The accident reporting requirements will provide the staff with the information necessary to beable to make an informed decision should it be necessary to terminate or modify an operation.
1.6.3.3

Surface Contamination Criteria

Guidance for acceptable surface contamination levels is provided in Regulatory Guide 1.86, "Termination of Operating Licenses for Nuclear Reactors" (June 1971); and in Regulatory Guide 8.21,"Health Physics Surveys for Byproduct Material at NRC-Licensed Processing and Manufacturing
Plants" (May 1978).
Table 1 of Regulatory Guide 1.86 provides specific guidance for acceptable
removable surface contamination levels for unrestricted areas.
These levels are shown in
Table 1.1.
Table 2 of Regulatory Guide 8.21 provides similar (but slightly less stringent) guidance forunrestricted areas. 5 2 These limits are considered compatible in level of safety with those forrelease of facilities and equipment for unrestricted use, as given in Regulatory Guide 1.86. In
accordance with this guidance, the staff has developed this programmatic environmental impact
statement using the levels indicated in Table 1.1 as those levels at which decontamination would
be considered complete for unrestricted access or unrestricted release of decontaminated equipment or facilities.
Fixed contamination (e.g., radioactivity that is not detected by smear surveys for loose surface
contamination) also must be limited.
Consequently, in addition to Table 2 criteria, the average
fixed contamination will be limited to less than 0.1 mrad/hr.

Table 1.1.

Acceptable Removable Surface Contamination Levelsa
for Unrestricted Accessdpm/100 cm2Radionuclideb
Alpha radiation from U-natural,
decay products
Transuranics,
1-125, 1-129
Th-nat, Th-232,
1-133

Ra-226,

Ra-228,U-235,

U-238,

and associated1000

Th-230,Th-228,

Pa-231, Ac-227,
20

Sr-90, Ra-223,

Ra-224,

U-232,

1-126,

1-131,
200

Beta-gamma emitters (nuclides with decay modes other thanalpha emission or spontaneous fission) except Sr-90 and others
noted above
aReg. Guide 1.86, Table 1,

p.

1000

1.86-5.

bWhere surface contamination by both alpha- and beta-gamma emitting nuclides
exists, the limits established for alpha- and beta-gamma emitting nuclides
should apply independently.

1-28
References--Section 1
1.

J.G. Kemeny et al., 'Report of the President's Commission on the Accident at Three Mile
Island. The Need for Change: The Legacy of TMI," Washington, DC, October 1979.

2.

M. Rogovin and G.T. Frampton, Jr., "Three Mile Island: A Report to the Commissioners andthe Public," Nuclear Regulatory Commission Special Inquiry Group, Vol. 1, Washington, DC,
1980.

3.

Memorandum from L.H. Barrett to H.R. Denton, U.S. Nuclear Regulatory Commission, Subject:
NRC TMI Program Office Weekly Status. Report (Week of Jan. 4-10, 1981), dated January 12,
1981, p. 3.

4.

"Summary Technical Plan for TMI-2 Decontamination and Defueling," Metropolitan Edison
Company, December 12, 1979.

5.

Handouts to presentation by GPU personnel at staff site visit, January 1980.

6.

Letter (TLL-578) from G.K. Hovey, GPU/Met-Ed, to Dr. B.J. Snyder, U.S. Nuclear Regulatory
Commission, dated November 7, 1980, Attachment C (the letter is reproduced in Appendix A ofthis PEIS as comment letter #50).

7.

Prehearing statement and Order issued by the Pennsylvania Public Utility Commission on
September 18, 1980, relative to the use of revenues by the Metropolitan Edison Company.

8.

Letter from H. Dieckamp,
September 12, 1980.

9.

J.A. Ayres, "Decontamination of Nuclear Reactors and Equipment," Ronald Press Co.,NY, 1970.

10.

F.W. Gilbert,"Decontamination of the Canadian Reactor," Chem. Eng. Progr. 50(5), May 1974.

11.

S. Glasstone,"Principles of Nuclear Engineering," D. Van Nostrand Co. , New York, NY, 1955.

12.

J.L. Gray, "Reconstruction of the NRX Reactor at Chalk River," Eng. J.,

13.

E.O. Hughs and J.W. Greenwood,

14.

"Nuclear Incident at SL-1 Reactor," U.S. Atomic Energy Commission, Idaho Operations Office,ID0-19302, 1962.

15.

"Final Report of SL-1 Recovery Operation, May 1961 through July 1962," U.S. Atomic Energy
Commission, Idaho Operations Office, IDO-19211.

16.

H.J. Blythe, "Proceedings of the First International Symposium on the Decontamination of
Nuclear Installations," Cambridge University Press, New York, NY, 1967.

17.

R.W. Staehle, "Seminar on Decontamination of Nuclear Plants," Electric Power Research Institute, EPRI-ASME, OSU, May 7, 1975.

18.

A.L. Parrish, III, "Decontamination Experience at Surry Plant," Paper presented at ElectricPower Research Institute Facility Decontaminaton Technology Workshop, Hershey, PA, November
28, 1979.

19.

R.L. Scott, Jr., "Fuel Melting Incident at the Fermi Reactor on October 5, 1966," Nucl.
12(2):123, March-April 1971.

20.

L.D. Perrigo et al. , "Plutonium Recycle Test Reactor Decontamination following a MgO-Pu02
Fuel Element Failure," Battelle Pacific Northwest Laboratories, BNWL-SA-636, March 31, 1966.

21.

J.M. Skarpelos and R.J. Lobsinger, "Decontamination of H Loop Following a Fuel Element
Failure," General Electric Co., Hanford Atomic Products Organization, Richland, WA,
HW-42081, March 23, 1956.

GPU,

to John F. Ahearne, U.S. Nuclear Regulatory Commission, dated"Contamination and Cleanup of NRU,"

New York,

October 1953.

Nucleonics, January 1960.

Saf.

1-29
22.

J.M. Lojek and W.T. Lindsay, "Attempted Decontamination of the Chalk River CR-VI
Westinghouse Atomic Power Division, WAPD, COA-AO-50, February 1959.

Loop,"

23.

G.M. Allison et al., "Decontamination of the X-3 Loop Pressure Tube Using APAC,"
Energy of Canada Limited, CRRL-940, 1960.

Atomic24.

R.D. Weed and L.D. Perrigo, Jr., "Decontamination of the KER Rupture Product Prototype,"
General Electric Co.,
Hanford Atomic Products Organization, Richland, WA, HW-59659,
March 16, 1959.

25.

G.E. Neibaur and N.D. Stice, "Dissolution of Uranium Oxides Formal During a Fuel Element
Failure," General Electric Co.,
Hanford Atomic Products Organization, Richland, WA,
HW-64594, May 15, 1961.

26.

R.D. Weed, "Decontamination of the M3G7 Loop in the Engineering Test Reactor," General
Electric Co. ,. Hanford Atomic Products Organization, Richland, WA, HW-77837, June 11, 1963.

27.

G.E. Neibaur and R.D. Weed, "Decontamination of the Irradiated Rupture Prototype with
Peroxide-Carbonate and APACE," General Electric Co. , Hanford Atomic Products Organization,
Richland, WA, HW-63346, January 13, 1960.

28.

J.E. Johnson and D.J..Bradford, "Decontamination of the Waste Calcining Facility - A Historical Review," ICP-1173, December 1978.

29.

J.A.

30.

J.A. Ayres, L.D. Perrigo, and R.D. Weed, "Contamination and Decontamination of a Pressurized
Water Reactor," Battelle Pacific Northwest Laboratories, BNWL-SA-938, October 14, 1966.

31.

R.D. Weed, "Decontamination of the Plutonium Recycle Test Reactor (PRTR)
Battelle, Pacific Northwest Laboratories, BNWL-711, March 1968.

32.

C.S. Abrams and E.A. Salterelli, "Decontaminati'on of the Shippingport Atomic Power Station,"
Westinghouse Atomic Power Division, WAPD-299, 1966.

33.

W.K. Kratzer, "Single Pass Through Reactor Decontaminaton of N Reactor Carbon Steel Piping,"
Paper presented at the National Association of Corrosion Engineers Meeting, Houston, TX, May
1975.

34.

W.D. Bainard, "Decontamination of N Reactor Stainless Steel Steam Generators," Paper presented at the National Association of Corrosion Engineers Meeting, Houston, TX, May 1965.

35.

W.K. Kratzer and W.D. Bainard, "Development of Waste Disposal Method for N Reactor's Decontamination Wastes," Paper presented at the National Association of Corrosion Engineers
Meeting, Houston, TX, May 1975.

36.

G.J.C. Nash, "Decontamination of the SGHWR Prototype," Meeting of the EURATOM Working Groupon Chemistry of Water Cooled Reactors at Ispra, Italy, May 1971.

37.

W. Ahlfanger, G. Herbsleb, and G. Resch,
Speisewassertagung, pp. 41-47, 1971.

38.

Y. Solomon, "Decontamination of a PWR Primary System, SENA Plant," Seminar on Decontaminatio'n of Nuclear Plants, Sponsored by the Electric Power Research Institute, held at Ohio
State University, May 1975.

39.

P.J. Pettit et al.,Mater. Perform., p.

40.

B. Montford, "Decontamination by Cycling Techniques at the Douglas Point Nuclear Generating
Station," Proc. Am. Power Conf. 35:902, 1973.

41.

Amendment No.

Ayres,

L.D.

Perrigo,and R.D.

Weed,"Decontamination of a PWR,"

Nucleonics,

April 1967.

Primary System,""Dekontamination Nulkearer Andagentede,"

BGP -"Decontamination of the Douglas Point Reactor by the Can-Decon Process,"
34, January 1980.

10 to License No.

DPR-73, U.S.

Nuclear Regulatory Commission,

March 12,

1980.

1-30
42.

"Environmental Assessment of Radiological Effluents from Data Gathering and MaintenanceOperation on Three Mile Island Unit 2," U.S. Nuclear Regulatory Commission, NUREG-0681,
Appendix A, page A-i, May 1980.

43.

U.S. Department of the Army, Corps of Engineers, U.S. Army Engineering District, Baltimore,
Permit NABOB-F/4 (Metropolitan Edison Co.) 75-1048, pp. 1-4, January 19, 1976.

44.

Letter from John P. O'Hagan, U.S. Army Corps of Engineers,
tan Edison Company, March 15, 1976.

45.

Application of Metropolitan Edison Company for National Pollutant Discharge Elimination
System permit from Pennsylvania Department of Environmental Resources, Section I, Attached
Page 8, Item 10, August 15, 1979.

46.

Pennsylvania Department of Environmental
2275419, p. 1, January 23, 1976.

47.

Letter from Terry R. Fabian, Pennsylvania Department of Environmental
Lawyer, Metropolitan Edison Company, August 10, 1977.

Resources,

to L.L.

48.

Letter from James V. Donato, Pennsylvania Department of Environmental
Hilbish, Metropolitan Edison Company, August 23, 1979.

Resources,

to J.F.

49.

Letter from J.G. Herbein, Metropolitan Edison Company, to R.W. Reid, United States Nuclear
Regulatory Commission, Subject:
Three Mile Island Nuclear Station, Units I and II (TMI-1
and TMI-2) Operating License Nos. DPR-50 and DPR-73, Docket Nos. 50-289 and 50-320, Permit
Modifications, p. 2, March 10, 1980.

50.

U.S.

51.

Letter from J.G. Herbein, Metropolitan Edison Company, to Frederick Marrocco, Pennsylvania
Department of Environmental Resources, Subject:
NPDES Permit No. 0009920, Three Mile Island
Nuclear Station (TMI), August 15, 1979.

52.

Regulatory Guide 8.21, "Health Physics Surveys for Byproduct Material at NRC-Licensed Processing and Manufacturing Plants," U.S. Nuclear Regulatory Commission, May 1978.

Environmental

Protection Agency,

Resources,

Baltimore District, to Metropoli-

Water Quality Management Permit No.

NPDES Permit PA0009920,p.

1,

December 30,

1974.

2.

MAJOR DECONTAMINATION AND WASTE DISPOSAL ALTERNATIVES

The alternatives addressed in this statement for dealing with the aftermath of the accident at
TMI-2 are primarily concerned with decontamination, core removal,.and waste disposal.
The
broad-scope, postaccident alternatives range from implementation of full cleanup to no action
beyond maintaining the facility in safe shutdown condition, and may be grouped as follows:
1.
2.

Full cleanup, salvage and decontaminate usable equipment,
Full cleanup, remove equipment with minimal or no decontamination,

3.
4.
5.

Partial cleanup with defueling,
Partialcleanup, fix core in place,
No action--maintain safe shutdown.

The impacts that would result from subsequent use of' the site are beyond the primary scope of
this statement; however, the staff recognizes that there may be a relation between the cleanup
decision and decisions on future use of the site.
Because an early decision could be made to
decommission Unit 2, the staff has examined the impacts associated with various decommissioning
alternatives.
The results of this effort indicate that an early decision to decommission would
not significantly affect the need for decontamination activities nor the resulting impacts discussed in this impact statement.
The possible options for future use of the site include (a) decommission the TMI-2 plant,(b) refurbish, requalify and use the existing nuclear power system, (c) use the TMI-2 buildings
and site as permanent waste repositories, or (d) use the TMI-2 buildings and site as temporary
waste repositories until a final decision is made.
The decommissioning alternatives include(a) removal and dismantling (DECON), (b) safe storage with deferred decontamination (SAFSTOR),
and (c) entombment (ENTOMB).`-3 The decommissioning alternatives all depend upon prior cleanup
and defueling and are discussed in Section 2.2.
To determine the effect that different future-use options might have on cleanup operations, it isimportant to consider separately the two operational phases identified in Section 1.3: Phase I,
the decontamination phase that extends from shortly after the accident up to removal of thereactor pressure vessel head (RPVH); and Phase II, the defueling phase, which starts with removalof the RPVH and extends through removal of the fuel and decontamination of the reactor coolant
system (RCS).
The critical factor in determining the effect of an early decision regarding future use of thesite on cleanup operations is whether or not the fuel must be removed from the reactor pressure
vessel and stored elsewhere. For reasons discussed below, the staff concludes that defueling anddecontamination of the RCS are essential operations that should not be omitted regardless of theoutcome of a decision on future use of the site.
A consequence of this position is that the distinction between the first and second alternatives
listed above is unimportant for Phase I and Phase II operations. The large, permanently installed
equipment items, such as the reactor pressure vessel, reactor coolant pumps, steam generators,
and associated piping, would not be removed or refurbished until completion of these phases.
If
the future use of the site were known prior to decontamination of the reactor building this could
affect the manner in which some smaller pieces of equipment are handled.* However, these operations constitute only a small part of the decontamination operations, and the difference in the*For example, some items that were specifically designed for the TMI-2 plant, such as instrument
panels, would have no salvage value if TMI-2 were decommissioned. These items might be removed
immediately and discarded if the plant were decommissioned, whereas they might be decontaminated in place if the plant were refurbished.
2-1

2-2
impacts that might result from different alternatives for equipment decontamination/removal
operations during Phases I and II are smaller than the uncertainty with which the impacts for all
decontamination operations can be estimated. Much of the equipment will be needed during cleanup
operations regardless of the decision on future use of the site (equipment for the building
ventilation system is an example), and some equipment that was not too severely contaminatedmight be decontaminated for reuse elsewhere even if it were not used again at TMI-2 (electricmotors are examples).
The third, fourth and fifth alternatives are very dependent on the options for future use of thesite.
None of these three options is compatible with refurbishment, and the last two are also
incompatible with the decommissioning alternatives that are allowed by current federal guidelines
and regulations. 1 ' 2 ' 3 However, for reasons given below and in Section 2.1, the staff concludes
that none of these three alternatives is acceptable for TMI-2.
The alternatives that would involve permanent waste disposal onsite appear not to be technicallyfeasible considering the location of TMI-2 (see Sec. 2.2).
Such alternatives are also not compatible with current national policies and regulatory guidelines for radioactive waste disposal.
For example, in his message to the Congress on February 12, 1980, President Carter noted that thepresent strategy for disposal of high-level waste and unreprocessed commercial spent fuel isfocused on use of deep geologic repositories.
A similar strategy applies to transuranic waste.
The location of such repositories is to be based on detailed site-specific evaluations, and thelocation of TMI-2, on an island in the middle of a river, also raises the question about relying
on long-term institutional controls beyond periods of 100 years to ensure the integrity of such a
disposal site.
Three Mile Island has not been selected by the U.S. Government as a candidate site for evaluationof its suitability for location of a high-level waste repository. Therefore, unless major policychanges are made by appropriate governmental agencies, Three Mile Island would not qualify as asuitable site for permanent disposal of spent commercial fuel or other high-level waste.
In anycase, it is unlikely that the site could be qualified as a candidate high-level waste repositorysite because of such factors as nearby population densities and hydrology.
Similar considerations apply to permanent disposal of low-level radioactive waste.
As in thecase of high-level waste disposal, Three Mile Island has not been evaluated by federal and stateauthorities as a candidate regional site for permanent disposal of low-level waste.
It isunlikely that the site could be qualified for this purpose because of the same factors mentioned
previously for the case of high-level waste disposal.
These considerations do not preclude, however, interim storage of radioactive waste at the TMI-2
site.
Since there currently are no waste repositories open for the disposal of high-level activity and transuranic wastes, it will be necessary to place these wastes in temporary storage until
a permanent waste repository becomes available.
Both onsite and offsite storage are considered
for this temporary measure.
The alternatives available for full cleanup with usable equipment salvage, environmental impactsof each in terms of offsite releases radiation exposure to decontamination workers, and socialand economic effects, are considered in detail in Sections 5 through 9.
The decontamination andcore disposal alternatives, for partial cleanup are discussed below in Section 2.1. A preliminary
analysis of the decommissioning alternatives is included in Section 2.2 in order to provide aclearer perspective of the relation between these alternatives and the alternatives for Phase I
and Phase II cleanup operations.*
A description of the classification schemes and disposal
methods for radioactive wastes as they apply to the wastes generated by different cleanup options
is given in Section 2.3.
2.1

DECONTAMINATION OF THE FACILITY

Cleanup of the facility includes decontamination of building and support facilities, removal of
the core, and decontamination of the nuclear steam supply system.
The operations involved may bedescribed as follows:*It was not possible to provide even a preliminary analysis of the refurbishment alternatives
because there is insufficient information available on the condition of the equipment.
This
information will have to be obtained during the Phase I and Phase II cleanup operations.

2-3
Processing the contaminated water generated by the accident
operations) to remove radioactive materials
Decontamination of surfaces inside
reactor building,(and by decontamination

the auxiliary and fuel handling buildings and theand decontamination or disposal of equipment.

Removal and disposal of internal

parts and damaged fuel from the reactor vessel,

andCleanup of the reactor coolant system (including components such as steam generatorsand pressurizer).
For full cleanup, all cleanup operations would be carried through to the point that the facilities
were ready to initiate decommissioning or refurbishment operations.
For partial cleanup,one ormore cleanup operations might be replaced by partial action or no action.
Partial action in thecore removal operation includes immobilization and storage of the reactor fuel within the reactor
vessel (core fixation).
Consideration of the partial cleanup options requires an examination of
the impacts on other operations or modification or omission of an operation as well as impacts on
the environment.
The primary criteria used by the NRC staff for judging the acceptability of partial cleanup
alternatives are that the radioactivity left onsite and the resultant reactor core configuration
should not pose a hazard of continuing radiation exposure to the public or to the workers after
cessation of cleanup operations, and that all requirements for public health and safety should be
met.
Except as noted in Section 2.1.1, below, alternatives that imply use of the site for longterm storage of radioactive materials (beyond the normal operating lifetime of a power reactor,
which is approximately 40 years) are not ruled out as unacceptable; however, they are given only
limited treatment on the grounds that the scope of this document is restricted to alternatives
that provide for eventual removal of all radioactive materials from the site, even if a decision
to do so should be delayed beyond the cleanup period.
It is the staff's position that Three Mile Island should not become a permanent waste repository
site.
Hence, alternatives involving temporary onsite storage of waste that would greatly
increase the effort required for subsequent offsite storage, processing, or disposal are regarded
by the staff as unacceptable.
2.1.1

No Action--Maintain Safe Shutdown

For the purposes of this environmental statement, the terminology "no action" is to be understoodas representing an alternative for which the reactor would be maintained in a safe shutdown
condition, but no attempt would be made to decontaminate the TMI-2 facility.
Under present
regulations, licensing of the facility would be required and all equipment inside the reactor
building would be operated as necessary but, because of the expected high radiation levels in thebuilding, servicing of equipment inside would be strictly limited.
The no-action alternative
would, in effect, turn the reactor building into a long-term, limited-maintenance repository for
nuclear waste--a function for which the facility was not designed.
In time, some equipment andcomponents within the reactor building could be expected to malfunction or fail as a.result oflimited maintenance.
Several implications of implementing such a no-action alternative are
discussed in Sections 2.1.1.1 and 2.1.1.2.
The NRC staff considers the no-action alternative unacceptable because the public health. andsafety cannot be adequately assured until the radioactive materials decay to innocuous levels.
The times required for the predominant radionuclides to decay range from a minimum of about
300 years for Cs-137 and Sr-90 to thousands ,of years for the transuranic radioisotopes in thedamaged fuel.
In the staff's judgment,'adequate institutional controls cannot be assured forthese time periods, and thus the no-action alternative could pose a serious threat to future
generations.
2.1.1.1

The Reactor

As of December 13, 1980, the decay heat level of the TMI-2 reactor was reported by the licensee
to be about 70 kW.
The primary coolant was being maintained at about 120°F and 94 psi, andnatural recirculation was being used to remove decay heat from the reactor core.
About half ofthe total decay heat was transmitted by convection to the reactor building, while the remaining
decay heat was transferred to the cooling towers through steam generator "A".
On January 5,
1981, the licensee stopped steaming the "A" steam generator by shutting a turbine bybass valve.

2-4This put the RCS in a "loss-to-ambient" mode of cooling in which reactor decay heat is transferred from both RCS loops to the reactor building ambient environment. 4
When the rate at which neutrons are generated from nuclear fission is just sufficient to sustain
a nuclear chain reaction at a constant rate, a reactor is said to be "critical."A certain
amount of neutron-absorbing material within movable control rod assemblies inside the core acts.
as a "brake" on the ability of the neutrons to sustain the chain reaction.
When a rod is moved
further into the core, more neutrons are absorbed, decreasing the rate of fission and resulting
in a "subcritical" reactor state.
Currently, several neutron-absorbing materials are used tocontrol criticality, the most common being silver-indium-cadmium control rods and borated water(water containing boron, a neutron absorber, in the form of boric acid).
Maintaining a sufficient amount of neutron-absorbing materials inside the TMI-2 core will keep it in its present
subcritical condition.
It has been estimated by Met-Ed and independently analyzed by NRC 5 ' 6 that the. shutdown margin of
the TMI-2 reactor is quite substantial (there are about 15 percent too few neutrons to sustain
nuclear chain reaction at a constant rate), with the primary coolant containing about 3850 ppm ofboron.
Preliminary evidence indicates that some control rod material may have been relocated toregions of the reactor coolant system out of the core.
This reduction in in-core control material can be compensated for by adjustments of the boron concentration in the primary coolant.
Excess boron available in the primary coolant must be adequate to ensure subcriticality of thecore at all times.
Of the two low-range neutron flux instruments, one is no longer operational.
(The higher-range
instruments are not sensitive enough to provide a reading at the current low neutron flux levels.)
The criticality condition of the core is closely monitored by the remaining low-range instrument
and by weekly analysis of the boron concentration in the primary coolant water.
Since, under theno-action alternative, decontamination of radioactivity inside the reactor building would nottake place, maintenance required by failure of neutron or boron monitors or other essential
equipment would be carried out in relatively high contamination conditions (compared to postdecontamination) and radiation fields.
Additionally, a limited preventive maintenance program
necessitated by these higher radiation fields would increase the probability of equipment failure
and the length of resulting equipment downtime.
2.1.1.2

Water

As a result of leakage from the primary system valves and pipe seals, the water level in thereactor building sump has been slowly, but continually, rising.
Some instruments and electric
cables have been affected by the water and are inoperative.
As of January 7, 1981, the water
depth was 8.05 ft above the basement floor (282.5-ft elevation) and increasing at the rate ofabout 0.04 ft/month due to leakage of about 3000 gallons/month from the primary system.
Unlessthe water is transferred from inside the building to a different location, the plant status may
deteriorate from the existing conditions as the water rises.
The rate of leakage could increase
with time as the system deteriorates.
2.1.2
2.1.2.1

Decontamination of Buildings and the Associated Support Facilities
Processing Contaminated Water

The contaminated water consists of water released during and after the accident in the reactor
building and water released during the accident and generated by cleanup operations in the tanks
and sump of the auxiliary and fuel handling buildings (AFHB).
On October 16, 1979, the Commission ordered the cleanup of the AFHB contaminated water to begin,using EPICOR II.
The processing of contaminated water in the AFHB was essentially complete as ofJune 27, 1980.
The processed water was being used at that time for washing and flushing contaminated components and piping in the AFHB.
It also could provide a source of water for washing
surfaces in the reactor building.
There were about 700,000 gallons of contaminated water in the bottom of the reactor building asof February 1981.
This contaminated water (referred to as the sump water)-overflowed the sump at
the time of the accident and covered the entire floor of the reactor building, including.theD-rings and reactor vessel cavity.
Reasons for processing this water include:2-5Transfer radioactivity from a mobile form in the water to a less mobile form with areduced volume to facilitate disposal.
Eliminate the potential of leakage of radioactivity from the reactor building sump to
the environment.
Reduce the inventory of sump water to levels so that equipment needed to maintain safe
shutdown will not be adversely affected by rising water levels.
Provide surge capacity
for collection and temporary storage of water generated by leakage from the primarysystem or from decontamination activities involving building and equipment surfaces.
Reduce radiation exposure to workers on cleanup of building and equipment surfaces anddefueling.
Alleviate potential psychologicalof cleanup progress.

stress to the local population associated with lack

A criterion used by the NRC staff for establishing the acceptability of decontamination alternatives is that no unprocessed accident water should remain after cleanup activities have been
terminated or suspended for an indefinite period of time.
Alternatives for managing the reactor building sump water that could be used for the full-cleanup
option all involve processing the water to remove dissolved radionuclide ions, and are discussed
in Section. 7.
Alternatives that could be used only for the partial-cleanup options allow otherapproaches that include:(1) leaving the sump water in place in liquid form, and (2) immobilizing the sump water in solid form by chemical combination with cement or absorption in solids.
These alternatives are briefly discussed below.
The alternative of leaving the sump water in place in liquid form is not acceptable to the NRC
staff because it leaves unprocessed accident generated water which may eventually leak from thereactor building.
This water also results in high ambient radiation levels that would makecleanup difficult in the vicinity of the sump.
Details of the contamination levels (concentrations of dissolved radionuclides) and exposure levels may be found in Section 7.
Immobilization processes using cement are complicated by the geometry and equipment on the sumpfloor.
It would be almost impossible to introduce a portland cement slurry into some of theinaccessible spaces and to obtain adequate mixing with the water.
Even if attempted, it would be
impossible to assure that complete solidification had been achieved.
Therefore, some water would
probably remain and would continue to pose a leakage hazard to the public.
If complete solidification could be achieved and assured, the resultant product would require extended (e.g;, greaterthan 100 years) isolation from the environment, which could not be assured by existing institutional controls.
The use of cement to solidify the sump water has the further disadvantage that
it eliminates the sump as a useful means of removing decontamination liquids from the reactor
building.
Because of the difficulty in removing the solidified material and the subsequent high occupational exposures, the implementation of this option would probably result in leaving the solidified material in place.
Leaving the solidified material in place then implies that the TMI site
will become a waste disposal facility.
On the basis of the reasons given, above, the staff considers the use Of cement to solidify thesump water in place to be an unacceptable alternative.
There are a number of solids commercially available for immobilizing the sump water in the formof a wet solid; however, for all the absorptive media known and considered by the staff, thevolume of the wet solid would be larger than the volume of the sump water by a factor of at least
three; for some the factor would be greater than 50.
A cap of concrete would be needed in orderto provide shielding protection.
Subsequent removal would not be as difficult as for cement;
however, it'would still be much more difficult than pumping out the sump water, and the volume of
the reactor building in the sump area within which the equipment was entombed would be about
twice as large as for cement.
The lower part of the steam generators and part of the steam
generator piping would be covered, and defueling and cleanup of the reactor cooling system would
be severely complicated.
The problems of achieving complete filling due to complicated geometry,
noted above for the cement alternative, occur also for the absorptive media alternative, and makelikely the occurrence of pockets of contaminated water.

2-6
In view of the potential problems associated with absorptive media immobilization of the sump
water and the lack of any significant advantages, the staff regards inplace immobilization of
unprocessed sump water to be an unacceptable alternative.
2.1.2.2

Building Decontamination

The auxiliary and fuel handling buildings (AFHB) and the reactor building were contaminatedduring the accident.
The AFHB contain tanks and equipment for preparing, storing and transferring water used as the primary system coolant and facilities and equipment for refueling thereactor.
The reactor building contains the reactor, the reactor cooling system and the nuclear
steam supply system.
The tanks and equipment in the auxiliary building are needed for maintaining the reactor in safe
shutdown and for processing the contaminated water.
The equipment in the fuel-handling building
would be needed for defueling the reactor.
The licensee, with the concurrence of the NRC,
started work on decontamination of the AFHB in April 1979 in order to remove hazards from possible releases of radioactive contamination and to maintain the reactor in safe shutdown condition.
As of January 1981, the work on decontaminating the AFHB was about two-thirds complete.
The methods that have been considered and used for the decontamination are discussed in Section 5.
The considerations that led to the decision to begin decontamination of the AFHB are still valid.
In the NRC staff's judgment, any alternative provisions for ensuring protection of the publicfrom radioactivity dispersed in the AFHB would involve at least as much risk to workers andpublic as would continuation of current decontamination of the AFHB until it is completed asplanned.
Hence, alternatives for AFHB cleanup, other than the full cleanup alternatives examinedin Section 5, will not be given further consideration.
The partial cleanup alternatives for decontamination of the exposed interior surfaces in thereactor building may be divided into three categories:(1) partial decontamination (decontamination of all surfaces, but leaving a higher level of contamination on completion than with fullcleanup), (2) selective decontamination (decontaminating selected work areas, isolating theremaining areas), and (3) destructive decontamination (using methods without regard to the damage
they may do to surfaces or equipment).
Alternatives for these categories are considered individually below.
Partial Decontamination
The advantage of stopping decontamination efforts short of full decontamination is that the
exposure to radiation of workers doing the decontamination and defueling would be decreased.
The
primary disadvantage is that the exposure of workers who subsequently would be engaged in disassembling and defueling the reactor and cleaning up the primary system would remain high (seeSec. 6).
A point of diminishing returns is eventually reached at which further reduction in the radiation
fields (during defueling and primary system decontamination) is not sufficient to justify theadditional decontamination effort. There are not sufficient data on TMI-2 to enable an accurate
determination of this crossover point.
However, on the basis of the estimates of the work
efforts and worker exposures given in Section 5, the staff estimates that this point will be very
close to the stage of full decontamination.
By this stage the radiation fields would be rela.tively low so that the net increase or decrease in overall exposure from any additional decontamination work effort would be relatively small.
On the basis of the expected overall reduction in worker exposure, and the fact that partial
decontamination would merely defer the cleanup problem at the expense of a continuing hazard of
releases to the environment, the staff considers the partial overall decontamination alternatives
to be unacceptable if the reactor is to be dissassembled and defueled and the primary system is
to be cleaned up.
If the alternative of core fixation rather than defueling were chosen, and if it'were decided not
to clean up the primary system (steam generators, pressurizer and associated piping and equipment), then partial decontamination might be a more attractive alternative.
This alternative has
not been explored because examination of the core fixation alternatives by the staff has shown
them to be unacceptable (see Sec. 2.1.3.1).

2-7
Selective Decontamination
One alternative for this category would involve the decontamination of a corridor to the reactor
vessel or the construction of a shielded corridor to the reactor vessel, without upper-level
decontamination.
For this alternative, it is presumed that access to a local area around thereactor would suffice for reactor defueling.
However, the confined space of the restricted areas
provided by localized decontamination or a shielded corridor does not allow for the use of theoverhead crane that is required for the reactor head lift
nor does it take into account the areas
needed for equipment laydown, the shield slab laydown or movement of the fuel-handling machines.
The corridor concept is not practical since a much greater volume of the containment building isneeded for defueling and therefore will not be considered further.
Another alternative in the selective decontamination alternative would be to cover contaminatedsurfaces with a high-density fixitive.
For this alternative it is presumed that covering asurface with a high-density fixitive would reduce the dose rate by an appreciable amount.
However, based on estimates of existing dose rates from gamma radiation, a one- or two-inch-thickcoating with Gunnite, a typical fixitive, would reduce dose rates very little.
The staff estimates that the worker-dose cost of applying such a coating to all of the interior surfaces would
be comparable to the worker-dose cost of reducing the surface decontamination to ALARA levels;
hence, the staff does not consider the use of high-density fixitives to be an acceptable alternative.
The preceding considerations indicate that the only acceptable variants of selective decontamination, if there be any, would be limited to isolating and shielding selected contaminated areasthat were not needed for reactor defueling and primary system decontamination operations.
The
staff considers it to be unlikely that the gain from bypassing a limited number of selected
locations would offset the impact of isolating and shielding them, arid the extra effort in keeping track of and avoiding contaminated pockets.
Hence, no further consideration will be given to
any selective decontamination alternatives.
Destructive Decontamination
Alternatives in this category are of two kinds:(1) use of corrosive chemicals for surface
decontamination without regard to damage to the surface, and (2) removal of equipment for burial
offsite without decontamination (scrapping).
These alternatives are not exclusive; both may be
used to varying extents.
An advantage of using corrosive chemicals for cleaning is that they remove part of the surface sothat radionuclides that are strongly adsorbed or chemically bonded to the surface also will be
removed.
The disadvantages are that there is an increased health and safety hazard for workers;
the spent liquid must be collected by means of special facilities that isolate it from the sump,or special measures must be taken to ensure that the processing equipment for removing dissolved
radionuclides from the sump water is not adversely affected; special measures must be taken toprotect equipment that may be needed during the defueling operations; and difficulties may be
encountered in waste treatment processes.
Experience gained in decontaminating the auxiliary andfuel handling buildings (Sec. 5.1.1) and information gained on the condition, of the reactor
building interior surfaces and environment (Sec. 5.2.1) have shown that the use of corrosive
chemicals will probably be unnecessary.
The staff considers the balance between the advantages
and disadvantages to be on the disadvantageous side and will not, therefore, give further consideration to the use of very corrosive chemicals.
Less harsh chemical methods are considered inconnection with the full cleanup alternatives in Sections 5 and 6.
The removal of equipment without decontamination (scrapping) would, under the assumption thatdefueling of the reactor would take place, be limited to those pieces of equipment that were notneeded during decontamination or defueling.
Most of the building services, such as the electrical system, ventilating and cooling systems, and polar crane, would be needed during thedefueling operations.
Equipment that was specific to the operation of the power plant, such as
steam generators, pressurizer, and associated pumps and piping are not in this category; however,these items would not be removed prior to defueling because defueling operations require continued cooling and criticality control provided by the primary system.
Thus, defueling must becompleted before the primary system can be safely'dismantled.
The first equipment for which the "scrap or refurbish" issue would be likely to arise would besalvageable equipment removed in the process of removing the reactor pressure vessel head andupper internals.
By that time, progress should be close to a point where there would probably be2-8
enough information on the state of the reactor to consider the issue of whether to decommission
There does not, therefore, appear to beor rebuild, which is beyond the scope of this document.
any need to give further consideration to the alternative of scrapping all unneeded equipment.
2.1.3

Reactor Core Removal and Decontamination of the Nuclear Steam Supply System

Cleanup of the reactor and the nuclear steam supply system (NSSS) comprises the major stages forThe
which the prior, stages of building and water decontamination would merely be preparatory.
full-cleanup operations for these major stages involve opening and inspecting the reactor pressure vessel (RPV), which contains the fuel and control rod mechanisms; removing the mechanicalparts and fuel from the RPV; draining and cleaning the piping and equipment on the primary sideof the NSSS (i.e., the piping and equipment through which water from the RPV circulates, referred
to as the reactor coolant system (RCS)); and processing the water that is now in the RPV and RCS
The full-cleanup alternatives
in order to remove dissolved and suspended solid radionuclides.
for these operations are considered in Section 6.
Taking no action on any of the preceding operations is considered unacceptable by the staff forreasons discussed in Section 2.1.1 and will not, therefore, be given any further consideration.
The partial-cleanup alternatives considered in this section for the reactor and NSSS are of two
kinds:
alternatives that might be considered if an early decision were made to discard all of
the equipment, and alternatives that might be considered if the decision to rebuild or decommission were deferred indefinitely and it was decided to place the entire plant in safe storage
for an indefinite period with a minimum immediate expenditure of effort.
The primary public health and safety criteria used by the staff for assessing the acceptabilityof partial-cleanup alternatives are:(1) there should be essentially no risk of occurrence of
criticality, (2) there shouldbe essentially no risk of release of radioactivity in excess ofthat which is allowed from an operating reactor, and (3) the core should remain in a physicallyThestable configuration with adequate means for removing heat generated by decay products.
staff dismissed from consideration at this time, for reasons noted in the introduction to Section 2, those core-fixation methods that would imply use of the site as a waste repository bygreatly increasing the effort that would be required for subsequent completion of cleanup operations.
Since there
The partial-cleanup alternatives for the different operations are not independent.
would be little
justification for undertaking full action for one operation and partial action
for another, the staff will consider only certain combinations of partial cleanup alternatives
that can be examined independently for each operation.
2.1.3.1

Core Fixation

It is unlikely that any of the fuel or internal parts in the RPV are salvageable for use in theirpresent condition.
Hence, the only alternative to defueling that would be opened up by an early
decision not to rebuild would be that of immobilizing the core and removing the entire RPV, with
In view of the size
its contents, and transporting it to some other site for ultimate disposal.
(32 feet high and 14 feet in diameter) and weight of the RPV and its contents (greater than
1500 tons), the high radiation levels, the weight of shielding required for transportation andthe danger of spillage, the staff considers this option to be unfeasible and unacceptable.
The partial-cleanup alternatives to defueling that have been considered by the staff also involve
For these alternatives, the fuel would be immobilized by means intended to precore fixation.
vent occurrence of criticality and then sealed inside the RPV by means intended to prevent leakage of radioactivity to the environment.
(1) ensuring that the neutronThe problems that must be dealt with in core fixation are:absorbing material is uniformly distributed and the fuel is immobilized in a permanently stableconfiguration so that criticality cannot occur, and (2) ensuring that the vessel remains completely sealed.
The alternatives considered by the staff,
are as follows.

together with brief assessments of their feasibility,

2-9Fill the Reactor Pressure Vessel with Concentrated Boric Acid Solution
In order to prevent loss of the boric acid,. which is es'sential for preventing criticality, it
would be necessary to seal off the openings into the RPV.
This would be difficult to do without
draining the RPV first. The radiation levels and worker exposure would then be very high and it
is unlikely that the seals could be made secure enough to ensure that no leaks would occur.
Lackof knowledge of the reactor vessel integrity increases this uncertainty.
In addition, the elimination of an external recirculating heat removal system would increase the average core temperature (i.e., heat removal would be effected by conduction only).
Therefore, since the long-term
physical stability of the core could not be ensured, the staff does not regard this alternative
as an acceptable one.
Fill the Reactor Vessel with Solid Neutron Absorber Additives in Pellet or Particulate Form
If this alternative were implemented, there would be no way to ensure that the pellets or grains
were uniformly distributed throughout the core, especially in view of the lack of knowledge ofthe damaged core configuration.
There may be regions of the core which have been fused or compacted together in such a manner that the pellets or grains could not, be properly distributed
throughout the core, and there would be no reliable way of being sure that there were no voids.
Under these conditions, it would not be possible to ensure that the core would not go critical.
To avoid the necessity of providing watertight seals and to avoid the risk of pressure buildup orleakage of radioactivity, it would be necessary to drain the reactor vessel after the pellets or
grains had been distributed.
The dry core would have poor heat conductivity, and high local
temperatures could occur that would make the core physically unstable.*
The water provides
shielding from the radiation emitted by the core; hence radiation fields outside the pressure
vessel would be much higher.
For these reasons, the staff does not consider this alternative tobe sufficiently safe and reliable to be acceptable.
Fill the Reactor Vessel with Cement Containing a Neutron-Absorbing Material
Distributing a cement slurry throughout the core,'without voids, to ensure the physical stability
needed to prevent criticality would be extremely difficult. High local temperatures that mightcause the cement to pulverize could occur.
Removing the damaged core when it was immersed only
in water would be difficult enough.
Removing the damaged core at some future time after it had
been encased in cement would be far more difficult.
Removing the fuel from a solid block of
cement, 14 ft in diameter and 20 ft high, filled with fuel elements and fuel rods would require
the use of saws and air hammers that generate a great deal of debris, and would likely generatemuch higher occupational radiation doses than removing the fuel from water.
The staff does not,therefore, consider immobilization in cement to be an acceptable alternative.
Partial Core Removal and Dry Layup
If a sufficiently large fraction of the fuel were removed from the core, enough to ensure thatthe amount remaining was below the threshold for which criticality could occur, then if thereactor vessel were drained the remaining fuel could be left in place without danger of coolant
leakage to the environment.
However, integrity of all seals to the reactor vessel could not be
assured for long periods of time, and escape of some radionuclides would be likely. Anotherdisadvantage of this alternative is that with the water removed, the radiation levels outside thepressure vessel would become very high.
However, the primary reasons that the staff does notregard this as an acceptable alternative is that after the work effort and worker exposure forremoving most of the fuel had been incurred, there would be no advantage gained by not'completingthe task and removing all of the fuel and thus removing a potential hazard for the future.
On the basis of the preceding considerations, the staff finds core fixation to be an unacceptableoption that does not merit further consideration.

*The TMI-2 core would not go critical, regardless of the configuration, if it were perfectly
dry and no other neutron moderator were introduced.
Precautions would be needed to ensurethat there is sufficient boron in the water to prevent recriticality before it was fully
drained, and to ensure that the core could not go critical from inleakage of unborated water.

2-10
2.1.3.2

Reactor Coolant System Treatment

Full decontamination of the reactor coolant system (RCS) would involve draining the contaminatedwater; removing the fuel-failure debris from the system by flushing, chemical treatment, or othermeans; and flushing with appropriate solutions in order to remove any remaining contamination
adhering to inside surfaces.
Decontamination of the RCS is treated in Section 6.
An early decision to decommission the plant would open up the alternative of dismantling andscrapping all of the RCS equipment.
Even if the decommissioning alternative were selected, itwould be necessary to drain the system and remove any fuel debris that might spill during dismantling and removal.
In addition, an analysis would be needed to determine whether additional
decontamination, beyond that needed to avoid spillage, would decrease the overall occupational
exposure.
The exposure to decontamination workers could be less than the exposure that workers
would receive while dismantling the equipment and packaging it for transport without further
decontamination.
There is not sufficient information on the condition of the RCS to carry out
such an analysis at this time, and it extends into a post-decision period that is beyond thescope of this document; hence, the analysis of this alternative will not be carried through inthis statement.
By the time the cleanup work has reached the point where decontamination of the
RCS can begin, the information needed to make a decision between rebuilding or decommissioning
should be available.
The no-action option for cleanup of the RCS would be reasonable only in the context of a complete
no-action plan.
This option is unacceptable for reasons presented in Section 2.1.1; hence, itwill not be given further consideration.
Of the partial-cleanup alternatives that might be considered, any that would leave contaminatedwater in the system can be ruled out as unacceptable for the same reasons that were given in
discussing treatment of the sump water. (Sec. 2.1.2.1) and core fixation (Sec. 2.1.3.1).
The
acceptable alternatives for the partial-cleanup options would, therefore, involve draining thecontaminated water from the steam generator and piping and removing the water from dead legs
(i.e., sections of pipe that are connected to the system in such a manner that they do not drain
when the system is drained).
This leaves a range of alternatives that differ only in the amountof contami'nation removed.
Debris in the form of sludge would dry and cake with the water removed
and become more difficult to remove later.
Insofar as the staff has been able to determine, there are no advantages to be gained by partial
cleanup of the RCS if the fuel and internal mechanisms must be removed from the RPV.
In view ofthis circumstance and the above-noted hazards associated with partial RCS cleanup, the staff
considers the partial-cleanup alternatives for the RCS to be unacceptable.
2.1.3.3

RCS Water Decontamination

The only alternatives that have been found to be acceptable for defueling the reactor and decontaminating the reactor coolant system require draining of the system.
The only alternatives open
for managing the contaminated water are to store it without processing, or to process it toremove the solid radionuclides.
The first alternative is considered unacceptable by the staff
for the reasons given in Section 2.1.2.1. The second alternative is examined in detail in Section 7.1.
2.2

DECOMMISSIONING

The purpose of this section is to:(1) review briefly the steps in the cleanup process thatprecede the alternatives of restarting or decommissioning in order to determine whether an early
decision to decommission has any impact on the nature or sequence of those steps, and (2) evaluate the impacts on the environment that would result from the decommissioning of TMI-2 sometimein the near future.
Under normal circumstances, decommissioning follows the orderly shutdown ofa facility at the end of its planned operating life.
The situation at TMI-2 is significantly
different from normal, with the reactor building and the AFHB severely contaminated, and much of
the fuel core damaged.
In addition, a major cleanup effort is currently underway.
Figure 2.1 is a simplified decision point diagram which illustrates the decision flow and activities which would be involved in arriving at either a refurbished facility, which could berestarted, or a decommissioned facility, which would eventually. result in a site without any
significant residual radioactivity from the previous existence of TMI-2.
Several decommissioning
alternatives, evaluated in this respect, are also indicated.

2-11

:IiI

Figure 2.1.

Simplified Decision-Point Diagram.

2-12
It should be emphasized at this point that neither the decision to refurbish or decommission
TMI-2, nor the evaluation of the full environmental impacts of these activities, per se, arewithin the scope of the PEIS.
However, because of the question of whether the consequences of anearly decision to decommission affect alternatives in the cleanup of TMI-2, the impacts of suchan early decision to decommission have been evaluated and sufficient material on decommissioningin general has been provided in this section and in Appendix U to make the discussion understandable.
The cleanup processes and the rationales for performing the various steps are presented in
Section 2.2.1.
The alternative approaches to decommissioning are discussed in Section 2.2.2,
with the impacts of these alternative approaches given in successive sections: waste volumes
generated (Sec. 2.2.3), effluents and releases to the environment (Sec. 2.2.4), environmental
impacts (Sec. 2.2.5), and decommissioning costs (Sec. 2.2.6).
Details are given in Appendix U.
2.2.1

Rationale of Initial Cleanup Activities

As discussed in Sections 1.1 and 2.1, the initial cleanup campaign at TMI-2 has two principal
goals:(1) to safely defuel the reactor, placing the fuel in a storage configuration that issafe from nuclear criticality and/or fuel meltdown, thus precluding the possibility of an inadvertent uncontrolled nuclear reaction occurring in the fuel, and (2) to collect and package for
disposal the large quantities of water-soluble and otherwise readily dispersible radioactivity
present in the plant and to ultimately remove the radioactivity from the site, thus precluding
the possibility of its inadvertent release to the environment.
For this reason it is postulated
that prior to initiating any decommissioning activities, the reactor building sump water has beenprocessed and the sludge has been removed, the wash down of the reactor building and the installation of temporary shielding has resulted in general-area radiation exposure rates on the operating floor (347-ft level) in the 5-10 mR/hr range, and in the basement of the reactor buildingin the 30 mR/hr range, and building surfaces have smearable contamination levels in the 30004000 dpm/100 cm2 range, exclusive of hot spots.
The irradiated fuel elements and debris have been removed and are assumed to be stored in the
spent fuel pool in the AFHB.
Shipment of the irradiated fuel to an away-from-reactor fuel storage facility (AFR), reprocessing plant, or some other disposal facility is assumed to begin when
decommissioning begins and to continue until all irradiated fuel has been removed from TMI-2.
However, lack of a suitable facility to receive the fuel may result in its retention in the spent
fuel pool in the AFHB for an extended period of time, in effect, converting that portion of TMI-2
into an AFR.
In terms of the schedule shown in Figure 1.4, active decommissioning efforts would begin at theconclusion of the fuel debris dissolution and the chemical decontamination of the reactor coolant
system and associated systems.
As a practical matter, the earlier cleanup efforts contribute to
the total decommissioning effort, but for convenience in this analysis, decommissioning is necessarily treated separately from the initial cleanup.
To obtain estimates of the total impact, theimpacts from the initial cleanup (to allow safe defueling) should be added to the impacts fromdecommissioning.
Summaries of waste volumes generated during decommissioning operations, effluents and releases,
doses, and costs are provided for each decommissioning alternative in this section.
These values
are separate from the values provided in other sections of the PEIS for the cleanup and decontamination steps.
2.2.1.1

Building Decontamination

To accomplish the first goal, the removal of the irradiated fuel from the reactor vessel, it isnecessary to reduce the amount of penetrating radiation in those work areas pertinent to thedefueling operations to levels that will permit reasonable occupancy times for the workers without undue radiation exposure.
In keeping with ALARA considerations, decontamination and shielding work in the containment building prior to the defueling operations will be limited to those
activities that produce the greatest reductions in area radiation dose rates per unit radiationdose absorbed by the workers doing the decontamination and shielding (first phase decontamination, as discussed in Section 5.2).
It is anticipated that these activities will include washdown of the containment building surfaces using water jet equipment, draining and flushing of thecontainment building sumps, and installation of temporary shielding around localized hot spots to
reduce general area radiation dose rates.
Hands-on decontamination work using mops, wipes andassorted cleansers will be done only where significant reductions in local area radiation dose

z-13
rates can be achieved.
Surface decontamination efforts beyond those outlined above will generally be performed as part of the second phase of operations discussed in Section 5.2.
2.2.1.2

Fuel Fragment Dissolution

It is presumed that many of the fuel rods in the reactor core were damaged during the March 28
accident, with resulting loss of cladding integrity and potential for dispersion of fuel particulates into the reactor coolant system (RCS).
While most of the larger fuel fragments probably
remain within the reactor pressure vessel, it is anticipated that many small fuel fragments werecarried out into those portions of the RCS that are external to the reactor pressure vessel.
Thus, there may be collections of small fuel fragments at various locations within the RCS, which
may create strong local radiation fields due to the intense radioactivity present in the irradiated fuel.
These local hot spots create two problems regardless of which choice is made for the
future disposition of TMI-2:(1) the strong local radiation fields will cause increased radiation dose to the personnel working in the vicinity of these spots unless carefully shielded, anaction which in itself will result in additional radiation dose to the workers, and (2) the
possible dispersion of these very small and very radioactive fuel fragments outside of the RCS in
the event of equipment disassembly or replacement represents an additional hazard to the workers.
Past experience has shown simple flushing to be relatively ineffective for removing very small
fuel particles from complex fluid handling systems.
Therefore, it is postulated that an effort
is made, following defueling of the reactor pressure vessel, to chemically dissolve and remove
any fuel fragments not previously removed.
The fuel dissolution step is followed by a chemical
decontamination of the RCS, to remove most of the remaining fission products and activated corrosion products plated out on the interior surfaces of the system.
Nuclear criticality considerations would be addressed in assessing fuel dissolution options.
2.2.1.3

RCS Chemical Decontamination

The chemical decontamination step is required to further reduce the radiation dose to workers
within the containment building, regardless of which alternative for facility disposition is
eventually implemented.
The rationale for doing the chemical decontamination at this time isthat doing so preserves all of the options for future disposition of the facility, in a way thatis quite effective in terms of radiation exposure reduction.
Failing to perform the chemical
decontamination before placing the facility in safe storage might preclude ever taking suchaction, since the pumps, valves, and associated equipment required to handle the decontamination
solutions would most likely be unusable after an extended storage period, and the impacts of theequipment refurbishment that would be needed to accomplish a chemical decontamination after an
extended period could be significant.
2.2.2

Decommissioning Alternatives

Decommissioning means to remove the facility from service and to safely dispose of the radioactive residues. 7 Once the initial surface decontaminations are completed, the reactor defueled,
the fuel debris removed, and the fission product and activation product depositions removed fromthe coolant systems, decommissioning of the nuclear steam-supply system can begin.
Severalalternative approaches to decommissioning are possible.
These alternatives range from minimal
further cleanup' with subsequent physical security under appropriate nuclear licensing restrictions, to complete cleanup and removal of all radioactivity from the facility and its site withrelease of the plant from all nuclear licensing restrictions.
Three alternatives can be used for decommissioning reactor facilities:
Immediate removal ofradioactive materials (DECON),safe storage with deferred removal of radioactive materials(SAFSTOR),
and entombment of radioactive materials (ENTOMB).
DECON permits termination of the
facility operating license, while SAFSTOR and ENTOMB require the continuance of an amended version of the license for extended periods of time, since licensable quantities of radioactive
material remain onsite (i*e., the facility becomes an interim waste storage site).
The amended
nuclear license allows the licensee to possess but not operate the facility.
Each of thesealternatives, as applied to TMI-2, is defined and discussed in the following subsections, withsupporting analyses and details given in Appendix U.
2.2.2.1

Immediate Dismantlement (DECON)

DECON is a pseudo-acronym defined by the NRC as the immediate removal of all radioactive material
to permit unrestricted release of the property. 7 To achieve this condition, the residual radioactivity levels in the facility must be sufficiently small such that members of the public who

2-14
may occupy the decommissioned facility would receive a negligible amount of radiation.
This
condition is much more stringent than would be required for refurbishment and restart of the
facility.
DECON meets the requirements for termination of the facility operating license and renders thefacility and site available for unrestricted release within a finite period of time.
In exchange
for prompt availability of the facility and site for other purposes, large commitments of personnel radiation exposure, disposal-site space, and money are required.
Another factor favoring
DECON is the availability of the facility operations staff to form a decommissioning work force
that is highly knowledgeable about the facility.
Elimination of continuing security, maintenance, and surveillance requirements (i.e., continuing care for SAFSTOR or ENTOMB) is of small
benefit on a site such as TMI which has another nuclear power unit on the property.
2.2.2.2

Safe Storage with Deferred Decontamination (SAFSTOR)

SAFSTOR means to fix and maintain the property so that risk to safety is acceptable for a periodof storage followed by deferred decontamination and/or decay to an unrestricted level. 7 SAFSTOR
consists of:(1) a period of facility and site preparation, which includes removal of fuel andconcentration and immobilization of dispersible radioactive materials, followed by (2) an interim
period of continuing care (i.e., safe storage) that encompasses security, surveillance, andmaintenance, and concludes with (3) the deferred removal of any remaining radioactivity.
The
nuclear license remains in force throughout the safe storage period since materials having radioactivity levels above unrestricted release levels are still onsite.
SAFSTOR could be modified toinclude storage of the irradiated fuel on the site for a finite period of time.
Deferred decontamination includes whatever actions are required at the end of the period ofcontinuing care to terminate the nuclear license and to release the property for unrestricteduse.
Some disassembly and disposal of activated components is still required, but the personnel
radiation exposure and the disposal-site space requirements are potentially greatly diminished.
Deferred decontamination cannot, however, rely on the availability of facility operations staff
for personnel familiar with the facility.
2.2.2.3

Entombment (ENTOMB)ENTOMB means to encase and maintain property in a strong and structurally long-lived material(e.g.,
concrete) to assure retention and isolation from the environment until the contained
radioactivity decays to an unrestricted level. 7 Depending on the nature of the entombed radioactivity, the necessary entombment period can range from about 100 years for short-lived contaminants such as Co-60 to many thousands of years for transuranic nuclides in spent fuel.
ENTOMB is similar in nature to SAFSTOR in that it also consists of a period of facility and site
preparation, which includes removal of fuel and concentration and immobilization of dispersible
radioactive materials, followed by a period of continuing care that includes security, surveillance, and maintenance activities.
ENTOMB also requires a nuclear license to remain in force aslong as the entombed radioactivity exceeds unrestricted release levels.
The facility and site
preparations include comprehensive cleanup and decontamination outside of the entombment structure and confinement of nonreleasable materials within the monolithic structure.
Continuing care
activities are minimal, unless the irradiated fuel is stored onsite.
In the ENTOMB strategy considered for TMI-2, the fuel is removed, but the reactor vessel andinternals are left in place.
As much as possible of the radioactive equipment from outside theentombed structure is consolidated and entombed within.
Under existing regulations, the nuclear
license must remain in force for an indefinite period of continuing care, until either the entombedradioactivity has decayed to unrestricted release levels or the entombment structure is dismantled
and the entombed radioactivity removed.
When it becomes desirable to terminate the nuclear license for ENTOMB, dismantling of the entombment structure will probably be required.
This represents a task that is much more difficult
than dismantling the unentombed facility, since the entombment structure is built to endure for along period of time.
Therefore, while dismantlement of the entombment structure is not technically impossible, ENTOMB would be an almost irreversible creation of a radioactive waste repository on the site and commitment to long-term maintenance of the nuclear license.

2-15
2.2.3

Waste Volumes Generated during Decommissioning

The quantity of radioactive waste requiring disposal varies significantly with the decommissioning alternative selected, with DECON producing the largest volume, followed by ENTOMB, and then
SAFSTOR.
Estimates of the radioactive waste volumes requiring disposal are developed in Appendix U and are summarized in Table 2.1.

Estimated Volumes of Radioactive Waste
Table 2.1.
from Decommissioning Alternatives
Burial Volume (ft3)a
Alternative

Activated

ContaminatedRadwaste

Total

DECON

40,600609,000

22,000671,000

22,00022,00022,000394,000

SAFSTORb....

ENTOMBb--

372,000

aValues are rounded to three significant figures.
bValues include only those wastes from initial decommissioning activiCumulative volumes including those from deferred decontaminaties.
tion should not exceed the DECON values.

The NRC staff considers ENTOMB to be an unacceptable alternative because its implementation would(1) create a long-term radioactive waste repository on the TMI site and (2) require maintenanceand administrative controls for a length of time considered nonfeasible (100 years or longer).
In the case of DECON, all of the radioactive wastes must be removed from the facility and itssite to satisfy the conditions for termination of the nuclear license and unrestricted use of theproperty.
In the case of SAFSTOR, all or part of the packaged wastes could be placed in storage
within the facility for the duration.of the storage period, with transfer to a low-level waste
burial ground when deferred decontamination begins.
In the case of ENTOMB, all of the residual
radioactivity must be confined within the entombment barriers. Storage of packaged wastes within
the facility but outside of the barriers would not be in compliance with the conditions definedfor entombment.
Thus, all of the packaged wastes outside the barriers would have to be shipped
to a disposal site during the ENTOMB operations.
2.2.4

Effluents and Releases to the Environment

Decommissioning operations are designed such that the containment/confinement capability of the
facility is maintained while the operations are in progress.
As a result, the airborne releasesof radioactive materials from decommissioning operations are quite small, and the associatedradiation doses to the public are also quite small.
As discussed in elsewhere in Section 2.2,
accident-related radioactive liquids are already processed and the radioactivity concentrated andsolidified.
Therefore, no significant releases of radioactivity in liquids are postulated tooccur during any of the decommissioning operations.
The estimated releases of airborne radioactivity resulting from decommissioning operations are
summarized in Table 2.2. Not all types of operations take place in all decommissioning alternatives. The releases listed represent upper-bound estimates for each type of operation.
2.2.5

Environmental

Impacts

The principal environmental impacts resulting from decommissioning are the radiation doses
received by the decommissioning workers and the volume of space required at a low-level waste
disposal site for burial of the radioactive wastes.
Some small radiation doses to the populace

2-16
in the vicinity of the TMI site could result from the releases of airborne radioactivity postulated in Section 2.2.4.
Likewise, some small radiation doses to the populace along the transportation routes between the site and the low-level waste burial ground could result from theradiation emanated by the packaged wastes while in transit. These impacts are summarized inTable 2.3, with brief discussions of the development of these dose estimates presented in thefollowing subsections.

Table 2.2. Postulated Releases of Airborne Radioactivity toto the Environment during Decommissioning Operations
ReferenceRadionuclige
Inventory

DECON

Segmenting ContaminatedEquipment

4

6 x 10-6

Activated Concrete

3

0.2 x 10-6

....

5

1.6 x 10-9--

1.4 x 10-959.8 x 10-6

1.14 x 10-5

9.8 x 10-6

Decommissioning Operation

RemovalContaminated Concrete
Removal
Water-Jet Cleaning

Airborne Release (Ci)SAFSTOR

ENTOMB5 x 10-6--

aSee Tables U.8 through U.11 in Appendix U for characterization of these inventories.

Table 2.3.

Estimated Cumulative Radiation Doses Resulting
from Decommissioning Operations

Radiation Dose Recipient

Cumulative
Radiation Dose (person-rem)a
ENTOMBb
SAFSTORb
DECON

Decommissioning Worker

1800

220

1300

Truck

160

13

85

Rail

6

6

6

PublicNormal Decommissioning
Activities

6 x 10-s

7 x 10-5

6 x 10-5

Radwaste Transport

74

6

40

Spent Fuel Transport

1

1

1

Transport Worker

aResults rounded to two significant figures.
bValues include only doses from initial decommissioning activities.
Doses from deferred decontamination will be no more than the doses
from DECON, decreasing with increased deferral time.

2-17
2.2.5.1

Occupational Radiation Doses

The radiation doses received by decommissioning workers are estimated by multiplying the numberof exposure hours experienced by the direct decommissioning crews while performing a given task
times the average local radiation dose rate in the area where the task is performed, and summingover all tasks.
A detailed estimate for each of the decommissioning alternatives is developed inAppendix U, with the summaries for each major structure presented in Table 2.4.
The estimatedexposure hours and doses do not include any contributions from supervision and support staff notdirectly engaged in the decommissioning activities in radiation zones.
Radiation doses received
by these types of personnel are postulated to be small, not exceeding a few percent of the amountreceived by the direct decommissioning crews, and well within the accuracy of the estimates madefor the work crews.
The radiation doses received by transportation workers resulting from the shipment of radioactive
materials from the site to a low-level waste burial ground and from the shipment of the irradiated reactor fuel from the site to a storage or disposal facility are estimated in detail in
Appendix U. The number and type of shipments, together with the associated radiation doses, arepresented in Table 2.5 for each of the decommissioning alternatives.

Table 2.4.

Estimated Cumulative Radiation Doses Received by
Direct Decommissioning Workers
ENTOMB a

SAFSTORa

DECON
Structure

Exp.

Containment

110,300

1,500

15,40082

110,000

1,020

AFH Bldg.

139,000

278

53,900121

107,000

213

Other Bldgs.

18,000

35

5,000

15

18,00035

Totalsb

267,000

1,810

74,300218

235,000

1,270

Hrs.

Person-Rem

Exp.

Hrs.

Person-Rem

Exp.

Hrs.

Person-Rem

Bldg.

aValues include only doses from initial decommiss'inning activities.
decontamination will be no more than the doses from DECON,
deferral time.
bTotals rounded to three significant figures.

Table 2.5.

Decommissioning
Alternative

Doses from deferred

decreasing with increased

Estimated Cumulative Radiation Doses Received byTransportation Workers

No.

Truck Transport
Shipments
Person-Rem

Rail Transport
No.

Shipments

Person-Rem.1,400

157.0

26

6.2

SAFSTORa

114

12.8

26

6.2

ENTOMBa

755

84.6

26

DECON

6.2

aValues include only doses from initial decommissioning activities.

Doses fromdeferred decontamination will be no more than the doses from DECON,
with increased deferral time.

decreasing

2-18
2.2.5.2

Offsite Radiation Doses

The dose estimates presented in Section 2.2.5 and listed in Table 2.3 for normal decommissioning
activities are developed in Appendix U. Also in Appendix U are dose estimates to the maximumThe significance of these doses and their human health and environmental
exposed individual.
consequences are discussed in Section 10.3.
Radiation doses to the populace residing along the transportation routes between TMI and disposalsites are estimated assuming each shipment emits gamma radiation at the maximum level permitted
The resulting estimated radiation doses, which are upper-limit estimates,by DOT regulations.
Actual experience indicates that these doses will be much less.
are also shown in Table 2.3.
The significance of these doses is discussed in Section 10.3.
2.2.5.3

Other Environmental Impacts

Other environmental impacts resulting from the decommissioning of TMI-2 are very similar to thosefor the cleanup effort discussed in Section 10.6, but of a lesser magnitude since fewer workers
will be involved (-.500,000 person-hours) and since the gross decontamination and cleanup efforts
will have been completed prior to the start of decommissioning.
Removing all of the radioactive materials from TMI-2, as would occur for DECON, may tend to
However, transporting
reduce the anxiety level and psychological stress of the local residents.
material from TMI to a disposal site may heighten the anxiety level and psychological stress of
those people residing along the transport routes.
Placing TMI-2 in safe storage will require even fewer workers than DECON or ENTOMB
The retention of the bulk of(-125,000 person-hours), thus resulting in a smaller local payroll.
the radioactive materials onsite during the safe storage period might tend to continue the existing levels of anxiety in the local community, even though the readily dispersible radioactive
Since transport of the radioactive materials frommaterials have been solidified and packaged.
TMI to a disposal site produces the largest estimated radiation dose to the public of all of thedecommissioning operations, not shipping that material would tend to reduce the anxiety level
among the populace along the transport routes.
Entombing the lower levels of the containment building would have an impact on the local payroll
In terms of public anxiety and psychological stress, ENTOMB
very similar to that of DECON.
combines the worst features of DECON and SAFSTOR, since a large amount of radioactive material istransported offsite but the entombed reactor containment building remains as essentially a permanent waste repository.
2,2.6

Decommissioning Costs

The cost estimates for decommissioning TMI-2 following the initial cleanup effort, removal of theirradiated fuel from the reactor, and dissolution of fuel debris and chemical decontamination of
the RCS and associated fluid-handling systems are developed in Appendix U (Sec. U.3.5 for DECON,
These
Sec. U.4.7 for SAFSTOR, and Sec. U.5.5 for ENTOMB), and are summarized in Table 2.6.
estimates are based in large part on information developed in earlier studies of decommissioninga large pressurized water reactor power station, 8 ' 9 with labor and materials escalated by 17
percent to bring their costs to mid-1980 levels. Energy costs estimated earlier 8 are escalated
Costs for transporting radioactive materials by truck are based on current rates
by 32 percent.
Handling and disposal charges at a low-level waste burial ground arefor exclusive-use vehicles.
(dated November 17, 1980) issued by the Nuclear Engineering Company forbased on a price list
their Richland, Washington, site.
2.3

DISPOSAL OF RADIOACTIVE WASTE

The accident at TMI-2 led to the generation of materials that contain or are contaminated with
These radioactive materials, in the form of solids, liquids, semisolids, andradionuclides.
These radiogases for which no further use is foreseen, are referred to as radioactive wastes.
active wastes must be managed from their initial generation through their ultimate disposition ina manner consistent with personnel and public safety, minimum spread of contamination, and compliance with regulatory requirements.
This section presents a discussion of the major alternatives for disposal of radioactive wastes
The forms of waste arethat could be generated by the decontamination and defueling of TMI-2.

2-19characterized, regulatory requirements that define the constraints for various steps within thewaste management cycle are discussed, and the alternatives considered for waste treatment,
conditioning, packaging, shipment, and ultimate disposition are discussed.

Table 2.6.
Estimated Costs for Decommissioning
TMI-2 Via DECON, SAFSTOR, and ENTOMB

Costs (millions o•
mid-1980 dollars)
SAFSTOR
ENTOMB

Cost Category

DECON

Decommissioning Labor
DirectSupportRadwaste Disposal
Spent Fuel Shipment

7.200
8.900
14.600
2.500

1.890
3.048
1.210
2.500

7.200
8.920
8.500
2.500

Energy

4.620

2.640

4.620

Other Costs
Supplies
Equipment
Contractors
Nuclear Insurance
Licensing Fees

1.820
0.960
0.640
0.940
0.060

1.040
0.090
0.123
0.344
0.045

1.820
0.960
0.510
0.940
0.050

Subtotal
25% Contingency
Totals.42.200
10.60052.8

12.9003.230
16.2

36.000
9.000
45.0

0.060

0.040

Annual Continuing--

Care Cost
Deferred Decontamination

43.8-.44.0avalues rounded to three significant figures.

2.3.1

Waste Characterization

The radioactive wastes at TMI-2 arise from the accident and from decontamination operations.
Each source generates primary waste, which is the form of the radioactive material at the time itis generated, and secondary waste, which is the form of the material that arises from treatmentof the primary waste.
The types of waste that could be generated at TMI-2 are shown by source inFigure 2.2.
A major factor governing the alternatives to management of radioactive wastes is its physicalform.
The waste types shown in Figure 2.2 can be divided into four categories which reflect
their physical characteristics:(1) solid materials (dry materials in bulk form), (2) processsolids (wet solids in the form of sludges, high-solids-content slurries, or granular materials
and ion-exchange media and spent filter cartridges), (3) liquids (liquids contaminated withradioactive materials), and (4) irradiated fuel assemblies, (5) gases (the Kr-85 in the reactor
building).
The characteristics of each waste type within these categories are described below.
2.3.1.1

Solid Materials

Irradiated Hardware
These wastes consist of the structural hardware and other components within the reactor vessel at
the time of the accident and the debris generated as a result of fuel assembly and component

2-20

PRIMARY FORMSSECONDARY FORMS

WATER-BASEDDECON SOLUTIONSLOADEDEXCHANGE MEDIARUBBISH AND
TRASH
EVAPORATOR
BOTTOMS

Figure 2.2.

Characterization of TMI-2 Radioactive Waste Forms.

2-21
damage during the accident.
The wastes generally are solid metals and represent a low- to relatively high-activity source of solid waste material.
Core internals and components have been
removed from other LWRs and disposed of as waste.
However, the possible extent of damage tothese components and the nature of the debris that could be present in the reactor vessel makethese wastes unique to TMI-2.
Trash
Radioactive trash generated during the decontamination operations will consist of compactible andnoncompactible solid material, some of which also is combustible.
The compactible and combustible solids consist of disposable clothing, rags, plastic covers, laydown pads, and miscellaneous
trash.
The amount of material generated is proportional to the number of personnel engaged indecontamination and the time required to complete decontamination operations.
The noncompactible
solids consist of tools, hoses, safety goggles, miscellaneous construction materials, and othersmall items of equipment used by decontamination personnel.
The amount of noncompactible solids
generated also is dependent on the number of personnel involved and the time required for decontamination.
The form and specific activity of the solid waste generated by the decontaminationcrew is comparable to the solid waste generated from decontamination operations at LWRs and othernuclear facilities.
Contaminated Equipment
It may not be possible to decontaminate some of the equipment contaminated during the accident.
This equipment, in the form of motors, pumps, valves, instrumentation, and other components, will
be handled as radioactive waste.
The form and specific activity of this type of solid radioactive waste is comparable to that generated from decontamination operations at other nuclear
facilities.
2.3.1.2

Liquids

Accident Water
Accident water, which will be handled as radioactive waste, is present in two locations:thereactor building (RB) and the reactor coolant system (RCS).
The accident water which collectedin the AFHB after the accident has been processed and the processed water is presently stored intanks onsite.
The water in the reactor building sump is somewhat complex in both chemical andphysical makeup.
The reactor building water includes sludge, high total suspended and dissolved
solids, and colloids.
It also contains tritium, radioactive fission products, sodium, and boron.
The specific activity of the reactor building water is higher than that of the water in the RCS.
The water in the RCS contains tritium, radioactive fission products, sodium, and boron. The
water is contaminated with radioactive fission products and for some radionuclides has specificactivity levels higher than the reactor building sump water and may contain fuel particles andtransuranics.
Processing of accident water using ion exchange, evaporation, or filtration techniques could lead to the generation of secondary waste forms consisting of tritiated water,process solids in the form of loaded ion-exchange materials and evaporator bottoms, and solid
waste in the form of spent filter cartridges.
Tritiated Water
There are no practical large-scale techniques available to decontaminate accident water to remove
the tritium originally present and all techniques will leave trace amounts of fission products.
Therefore the processing will result in wastewater containing tritium at concentrations essentially the same as before treatment.
This tritiated water will represent a fairly high volume ofvery low-specific-activity liquid waste.
Water-Based. Decontamination Solutions
Some of the decontamination procedures that will be used involve the use of water to wash downinternal building surfaces and equipment.
This water will become contaminated with the radioactive materials it washes off the surfaces.
The chemical and physical characteristics of thiscontaminated water will depend on the contamination levels of the surfaces washed, the procedureused for application of the water, and the extent to which detergents are used.
Generally, such
water-based decontamination solutions contain suspended solids, fission products, and small
amounts of chemical contaminants and detergents.
The solutions represent a relatively large
volume of low-specific-activity liquid waste comparable to the solutions generated from

2-22
Processing of these solutions using ion
decontamination of other types of nuclear facilities.
exchange or evaporation could lead to the generation of secondary waste forms consisting ofprocess solids in the form of loaded ion-exchange materials and evaporator bottoms.
Chemical Decontamination Solutions
Decontamination operations also can be performed with chemical solutions (e.g., strong detergent
The physical and chemical characteristics ofsolutions or foam-type decontamination agents).
these decontamination solutions are more complex than those arising from the use of water. Their
effectiveness could be greater than that of water-based solutions so that the resultant liquid
may have higher specific activities than would water-based solutions. The volume of chemicalsolutions generated could be less than the volumes of water-based liquids generated. Whilerelatively large volumes could be generated, the nature of these chemical solution liquid wastes
and their specific activity would be comparable to that generated from decontamination operationsat other types of nuclear facilities. Processing of these chemical solutions using evaporation
could lead to the generation of secondary waste in the form of evaporator bottoms. Bituminization could lead to secondary waste in the form of bituminized solids.
2.3.1.3

Process Solids

The process solid wastes that arose from the accident and could arise from the treatment ofaccident water and decontamination liquids are discussed in this section.
Accident Sludge
This form of process solids arose from the accident and is present in the sump of the reactor
building and in tanks and sumps in the auxiliary and fuel handling buildings. The chemical andphysical characteristics of the sludge will be relatively complex and variable. This sludge iscontaminated with radioactive fission products and will represent a relatively low-volume, highThe nature of this radioactive waste is not
specific-activity form of radioactive waste.
directly comparable in form or specific activity to sludges generated by other LWRs.
Loaded Ion-Exchange Materials
The use of ion-exchange media in the form of organic resins or zeolites to remove fission productcontaminants from the liquids could lead to the generation of process solids in the form of
The specific activities and radionuclide contents of the loaded
loaded ion-exchange materials.
ion-exchange media will vary with the liquids processed through the media, the contaminants
selectively removed from the liquids, and the capacity of the media to retain contaminants. Someof the liquids that could be treated using ion-exchange techniques will generate wastes withspecific activities well above those normally generated in LWRs, while processing of other.liquids will lead to wastes with specific activities well within the range normally generated bya PWR.
Thus, loaded ion-exchange materials could represent a relatively high volume of processsolid waste with specific activities in the low to very high range relative to the expended
ion-exchange material generated by other PWRs. The relatively high specific activity'and natureof the fission product contaminants on some wastes will make them unique to TMI-2.
Evaporator Bottoms or Sludges
The use of evaporation techniques to reduce liquid waste volumes could lead to the generation ofprocess solids in the form of evaporator bottoms or sludges. The physical characteristics ofthese process solids would depend on the solids content of the liquids evaporated and the equipThese characteristics could range from slurries containing
ment used for evaporation.
10-20 weight percent solids to sludges with solids contents in excess of 50 weight percent.
The specific activities of these process solid wastes also could vary over a wide range, andcould be above those normally generated by LWRs. Evaporation of decontamination solutions could
lead to higher-than-normal concentrations of chelating agents in these process solids.
Filter Cartridge Assemblies
This form of solid waste will arise from the treatment of liquids. Filter cartridge assemblies
are typically right-circular cylinders which are used to remove particulates from liquid waste;
the contaminated particulates are deposited on the filter. They represent a low- to very-highspecific-activity form of solid waste, with their specific activity dependent on the contaminants

2-23
in the waste stream processed.
Most of these cartridges are similar to those used for liquid
treatment in other LWRs, and their specific activity could be comparable to the higher-activity
filter assemblies generated at other pressurized water reactors (PWRs).
The nature of some of
the contaminants in certain of the TMI-2 liquid wastes may lead to the generation of spent filter
cartridges with unique contaminants and specific activities higher than normal.
Bituminized Solids
The use of bituminization could lead to the generation of process materials in the form ofbituminized solids.
These wastes will be in the form of a monolithic solid containing about
50 weight percent evaporated salts and 50 weight percent bitumen.
2.3.1.4

Irradiated Fuel Assemblies

The fuel assemblies that make up the reactor core were damaged in varying degrees during theaccident.
These irradiated fuel assemblies contain uranium, plutonium, and assorted fission
products and are high level wastes.
The possible extent of damage to these fuel assemblies makes
them unique relative to the irradiated fuel assemblies at other light water reactors (LWRs).
2.3.1.5

Gases

Since essentially all of the Kr-85 gas in the reactor building atmosphere that arose from theaccident was vented* into the outside atmosphere, no secondary waste forms will arise.
2.3.2

Classification

Regulatory classification of radioactive waste is a factor in evaluating alternatives for ultimate disposition of these waste materials.
Classification will determine the standards applicable to packaging and transportation and the types of facilities that can be used for theirultimate disposition.
Federal regulations concerning radioactive materials are discussed in
Section 1.6.
Historically, three classes of waste commonly have been referred to:High-level waste (HLW).
These wastes are spent fuel or the wastes resulting from
irradiated nuclear fuel during reprocessing.
Transuranic (TRU) waste.
These wastes are made up of materials which contain or arecontaminated by transuranic elements, with atomic numbers higher than uranium. Whenwaste material contains concentrations of transuranic elements in excess of 10 nanocuries per gram, they are considered TRU waste.
Low-level waste (LLW).
These wastes are made up of all radioactive waste materials
that are not HLW or TRU waste.
It should be noted that these three broad waste classifications have evolved over the years based
upon operations of the nuclear industry.
In this regard, no regulatory framework was developed
to specifically address the types of unique waste forms that may be generated at TMI-2 as aresult of the accident.
Accordingly, the wastes resulting from TMI-2 cleanup will have to becarefully reviewed on a case-by-case basis with regard to important characteristics such as
specific activity, radionuclide content, total radioactivity inventory, and waste forms andstability.
Ultimate disposition of such wastes will depend on the unique characteristics they possess and on
suitable facilities for their-handling and disposal.
For example, because of the high concentrations of certain long-lived nuclides (approximately 30-year half-lives), particularly Cs-137 andSr-90, certain forms of TMI-2 wastes may not be suitable for handling by routine shallow land
burial techniques.
The waste materials affected could include very high specific activity loaded
ion-exchange materials, accident sludges, and spent filter cartridges.
The primary and secondary waste forms and their potential dispositions considering the current
regulatory framework are shown in Figure 2.3.
Mary of the wastes can likely be classified under*Residual amounts of Kr-85 from offgassing in the reactor building continue to be vented from~thereactor building (i.e., prior to personnel entries).
In addition, the fuel still contains approximately 45,000 additional curies of Kr-85, some of which could be released during defueling.

2-24

PR IMARY/SECONDARYWASTEFORM
REGULATORY
CLASSIFICATION

Low-

IEVEL

CHEMICAL________________

DECON SOLUTIONSWATER-BASED .+"DECON SOLUTIONSTHESE WASTE FORMS MAY REQIJIRE MORE RESTRICTIVE HANDLINGAND DISPOSAL THAN APPLIFD IN ROUTINE SHALLOWLAND BURIAL.

KNOWNCLASSIFICATION FOR WASTEFORM
POTENTIAL
CLASSIFICATION FOR A PORTIONOF THEWASTEFORM

Figure 2.3.

TMI-2 Waste Forms and Classification.

LEVELi

LOW
L
0 ,-

2-25

However, because various charpresent interpretations as low-level waste or transuranic waste.
acteristics of some of the wastes that will be generated are unique to TMI-2, higher standards ofcare than those required for routine standard treatment of low-level waste might have to be
applied. Thus, certain batches of loaded ion-exchange materials, accident sludges, and individual spent filter cartridges are likely to require special consideration and treatment. The
standards applied to these unique wastes will be considered on a case-by-case basis and, where
warranted, these wastes will be handled, packaged, and disposed of in accordance with special
requirements.
2.3.3

Alternatives for Disposition of Radioactive Waste

The alternatives considered for ultimate disposition of TMI-2 radioactive waste depend on thephysical and chemical characteristics of the radioactive materials, their radionuclide content
and specific activity, and the regulations governing packaging, transportation and disposal.
These factors establish the boundary conditions for consideration of the alternatives or options
available within the waste management cycle. The steps within the radioactive waste management
cycle at TMI-2 are shown in Figure 2.4 and are discussed below.
In the initial step, waste materials are accumulated and segregatedCollection and Segregation.
according to chemical and physical form and radionuclide content.
Treatment.
Treatment is a step which refers to the processing operations leading to the concentration of radionuclides into a smaller volume.
For liquids, the radioactivity of the treatedliquid is substantially reduced, enabling it to be disposed of by various means or reused.
Conditioning is a step which refers to those operations that transform the concenConditioning.
trates produced during treatment or untreated materials into forms suitable for transportation or
disposal.
Conditioning includes immobilization, which converts radioactive waste materials in
the form of liquids and process solids into a stable immobile form with the radioactive materials
homogeneously dispersed within it.
Conversion to this form minimizes radionuclide release to the
A detailed description of the techenvironment during storage, transportation, and disposal.
niques used to immobilize radioactive waste and the characteristics of the resultant forms using
a variety of binder materials is presented in Appendix H.
Packaging.
Packaging refers to placement of the radioactive material into a disposable container. The container is a barrier to radionuclide release and a means of handling the waste
during interim storage, transportation, and disposal.
The container also may serve as the vesselin which immobilization operations are performed.
Package Handling.
Package handling refers to those operations that move containers within thefacility. The methods used for handling depend on the radiation level of the container and its
These methods include hands-on, semiremote, and remote handling, as well as
size and weight.
placement of the container within a transfer shield to reduce radiation levels.
Storage.
Storage is the containment of radioactive wastes in a manner that provides for theirsubsequent retrieval.
Storage onsite could be required for varying time periods depending on theavailability of suitable offsite disposal, processing, and storage facilities.
Shipment.
Shipment refers to placement of radioactive waste on transport vehicles and transfer
to another facility.
Unshielded shipments contain radioactive materials in containers withradiation levels that comply with transportation regulations without the addition of shielded
overpacks.
Shielded shipments are those where the containers are placed within shipping casksthat provide shielding and additional mechanical integrity during transport.
Disposal.
Disposal refers to the emplacement of the radioactive waste containers in facilities
Discharge, which is a form of disposal, refers to the conwithout intending to retrieve them.
trolled release of liquids or gases to the environs.
As shown in Figure 2.4, some liquids may be suitable for discharge following treatment. AddiSome of the wastes
tionally, shipment to another facility would not necessarily entail disposal.
will have characteristics of routine low-level waste and may be disposed of by routine shallow
land burial.
Some of the wastes might contain transuranic contaminants with concentrations inexcess of 10 nCi/g, which would require their shipment to a transuranic waste storage facility.
Other wastes expected to be generated could have specific characteristics which preclude theirdisposal at a commercial low-level disposal facility by routine methods and special measures may

PACKAGING-I-HANDLING

SHIPMENT

Fiur-24.Rat
C

Figure 2.4.

Radioactive Waste Management Cycle.

M e MENT._

2-27
have to be taken.
handle spent fuel.

Fuel assemblies also could be shipped to treatment or storage facilities which

Within the steps shown on Figure 2.4, the alternatives considered for each waste type during eachstep are variable but can be summarized for each of the five waste form categories described in
Section 2.3.1.
A summary of these waste management alternatives, which are described in detailin subsequent sections of this document, is presented below.
2.3.3.1

Alternatives for Management of Solid Materials

The alternatives considered for solid materials are dependent on their physical form, specificcharacteristics, and classification, where relevant. The major steps within the waste management
cycle that were considered as a function of waste type are shown in Table 2.7.
Only those waste
management steps applicable to solid materials are shown.
Accordingly, discharge to the environs
and conditioning are not included.

Table 2.7.

Alternatives for Management of Solid Materials
Waste Typesa

Waste Management Alternatives

Trash

ContaminatedEquipment

IrradiatedHardware

Fuel
Assemblies

Treatment
Incineration
Compaction
Disassembly/sectioning

X
X
X

Packaging
55-gallon drums
LSA boxes
Steel liners
Special containers

X
X

X
X

X
X
0

Package Handling
Hands-onRemoteTransfer shield

X
X

X
X
X

X

X

X

X

X
X

X
X

X

X

X

X

Onsite storage
InterimLong-term

X
0

Shipment
Unshielded
Shielded

X

DisposalCommercial low-level
Transuranic
High-level treatment/storageX

aX denotes alternative considered for waste type.
0 denotes alternative applicable to special waste only.

0

2-28
The treatment alternatives considered depend entirely on the physical form of the material.
Incineration
Combustible trash can be incinerated to reduce volume by factors of 80 to 100.
leads to the generation of ash which is handled as a process solid. Noncombustible compactible
Contaminated equipment andtrash can be compacted to reduce volumes by factors of about five.
irradiated hardware consists of metallic solids, and disassembly and mechanical sectioning are
the only techniques considered for volume reduction.
The disposable containers considered for packaging include 55-gallon drums, boxes for lowspecific-activity material, and rectangular or cylindrical steel containers with capacities aslarge as 200 ft 3 . Special containers are required for fuel assemblies and could be required forspent filter cartridges if they are contaminated with transuranic contaminants with concentrations above 10 nCi/g or if they have high specific activities.
The package handling techniques considered depend on the radiation level of the disposable conLow-specific-activity materials in drums or wooden boxes aretainer, its size and weight.
handled manually and transferred within the facility using forklifts, motorized pallets, or otherFor higher-activity disposable containers, semiremote and remote
package-handling equipment.
Individual containers
handling systems are used to minimize radiation exposure to personnel.
also can be placed within transfer shields and thus permit the use of manual handling techniques
Fuel assemblies are handled undersimilar to those used for low-specific-activity containers.
water using remote techniques.
Except for
Generally, all waste is transferred to an onsite storage facility to await shipment.
trash boxes, these storage facilities are shielded enclosures, and facilities are divided by
Fuel assemblies would be stored in the spent fuelcontainer type and surface radiation level.
Long-termpool, and interim, as well as long-term, storage of these assemblies was considered.
storage may also be required for spent filter cartridge assemblies that might have to be handled
as transuranic or high-level waste.
The waste types shown will be transported offsite in both unshielded and shielded containers.
Low-activity drums and LSA boxes can be shipped unshielded, while higher-activity drums and theother container types shown will be transported in licensed, shielded shipping casks.
The destination of each shipment depends on the characteristics of the materials being transported.
Some waste will be transported to a commercial low-level waste disposal facility.
Certain high-specific-activity wastes might require transport to special waste storagefacilities.
Special destinations were considered for fuel assemblies, including high-level waste
storage.
A detailed discussion of the above alternatives and their impacts issections of this document.
2.3.3.2

presented

in subsequent

Alternatives for Management of Liquid Waste

The alternatives considered for liquid waste are dependent primarily on the chemical characterTable 2.8 contains a sumistics, radionuclide content, and specific activity of those wastes.
mary of the alternatives for liquid waste considered for each step within the waste management
cycle shown in Figure 2.4.
The treatment alternatives considered for liquids include filtration, ion exchange, evaporation,
The use of theseand bitumenization.
All four techniques were considered for accident water.
techniques substantially reduces the specific activity of the water and leads to the generationof secondary wastes in the form of spent filter cartridge assemblies, loaded ion-exchange materiUse of these techniques will notals, and evaporator bottoms or sludges and bitumenized solids.
reduce the tritium concentration of the treated water; hence, discharge of tritiated water tosurface waters or through evaporation, packaging for disposal as low-level radioactive waste, andThese treatment techniques weredeep-well injection were considered for this tritiated water.
However, the highly complex nature
also considered for the waterbased decontamination solutions.
of some of the chemical decontamination solutions eliminates the use of ion-exchange techniques
as feasible alternatives.
A conditioning step involving immobilization was considered for liquids that were not treatedbecause with only few exceptions liquids cannot be transported or disposed of at commercial
A detailed discussion of four different immobilization
low-level waste disposal facilities.
agents and the techniques used to implement them is presented in Appendix H. Of the four agents

2-29

described in that appendix, only three were considered for TMI-2 wastes. The use of urea formaldehyde was eliminated because of the problems associated with obtaining a liquid-free product.
The remaining three--cement, vinyl ester styrene, and bitumen--were considered for most of theTMI-2 waste forms requiring immobilization.
In cases of immobilization of liquids, disposable
containers (drums or cylindrical steel liners) were considered for packaging.
The packagehandling techniques considered for these drums and steel liners are similar to those described in
Section 2.3.3.1.
Packaged, immobilized liquids will be stored onsite to await shipment withother packaged waste by container type and radiation level.

Table 2.8.

Alternatives for Management of Liquids

Waste Management Alternatives

Accident
Water

Waste Typea
Water-Based
Decon Solutions

Chemical Decon
Solutions

Treatment
Filtration
Ion exchange
Evaporation
Bituminization
Discharge

X
X
X
X
0

X
X

Conditioning
Immobilize/cement
Immobilize/vinyl ester styrene

X
X

X
X

X
X

Packaging
55-gallon drums
Steel liners

X
X

X
X

X
X

Package Handling
Hands-onRemoteTransfer shield

0
X
X

X
X
X

X
X
X

Onsite storage
InterimLong-term

X
0

X

X

Shipment
Unshielded
Shielded

0
X

X
X

X
X

Disposal
Deep-well injection
Commercial low-level

0
X

X

X

X
X

aX denotes alternative considered for waste type.
0 denotes alternative for tritiated water only.

Shipment would be made either shielded or unshielded, as dictated by container radiation level.
If tritiated water is immobilized, the radiation levels of this waste may be compatible withunshielded shipment.
Some immobilized decontamination solutions also might be shipped unshielded.
The destination of all shipments will be a commercial low-level waste disposal facility.

2-30
Detailed discussion of these alternatives and their impacts is presented in subsequent sectionsof this document.
2.3.3.3

Alternatives for Management of Process Solids

The alternatives considered for process solids are dependent primarily on their physical form,chemical characteristics, radionuclide content, and specific activity.
Table 2.9 contains asummary of the alternatives considered for each step within the waste management cycle for eachtype of process solid.
Only those steps applicable to this waste form are shown, and treatment
and discharge were not considered.
The conditioning alternatives considered include dewatering and use of the three immobilization
techniques considered for liquids.
To permit disposal at a commercial low-level waste disposal
facility, evaporator bottoms and incinerator ash must be immobilized.
All three immobilization
techniques were considered for evaporator bottoms and incinerator ash.
Ion-exchange materials
and sludges can be shipped to some commercial low-level waste disposal facilities in the dewatered condition, while other disposal facilities require these materials be in an immobilized
form.
EPICOR II ion-exchange materials used to process AFHB liquids are currently required to besolidified in accordance with Commission Order.
The use of bitumen to immobilize accident
sludges was not considered because of the expected high specific activity of the sludges and the
low threshold for radiation damage of bitumen.
The disposable containers considered for packaging include 55-gallon drums and large cylindricalsteel liners.
Where ion exchange treatment techniques incorporated disposable demineralizer
vessels, such vessels ranging in size from 10 ft 3 to 195 ft 3 also were considered as disposable
containers.
These disposable demineralizer vessels also were considered as the. vessel within
which immobilization of expended ion-exchange materials could be performed.
Special containers
also were considered for special ion-exchange materials and accident sludge.
These waste materials could contain transuranic contaminants with concentrations above 10 nCi/g, requiring theirhandling as transuranic waste, or they also could have characteristics which might require special packaging, handling, storage, and disposal methods.
Drums were the only disposable containers considered for incinerator ash because of the relatively low volume of this waste
expected to be generated.
The surface radiation levels of all these disposable containers will require their handling by
semiremote and remote techniques to minimize personnel exposure.
Some of the handling techniques
considered include underwater handling of ion-exchange material containers, as well as placementof individual containers within transfer shields for movement within the facility.
All these waste types will be transferred to an onsite storage facility.
Special facilities tostore ion-exchange material containers have been considered, including surface and subsurface
cells and storage under water within the spent fuel pool.
Where ion-exchange materials, evaporator bottoms, or accident sludges warrant special handling, long-term onsite storage also wasconsidered.
All containers will likely be shipped in a shielded configuration.
Some of these waste containers will be shipped to a commercial low-level waste disposal facility. Those containers with
ion-exchange materials, evaporator bottoms or accident sludges which require special handling
could be stored onsite for periods of 20 to 30 years or shipped to special facilities for either
storage or treatment.
A detailed discussion of the above alternatives and their impacts
sections of this document.

is presented

in subsequentTable 2.9.

Alternatives for Management of Process Solids
Waste Typea

Waste Management Alternatives

Loaded IonExchange Materials

Evaporator
Bottoms

Accident
SludgeIncinerator
Ash

X
X
X

X
X

X
X
X

X

Spent
Filter
Cartridge
Assemblies

Conditioning
Dewatering
Immobilize/cement
Immobilize/vinyl ester styrene
Packaging
55-gallon drums
Steel liners
Special containers

X
X
0

X
X

X
X
0

X

X
X
0

Package handling
Hands-onRemoteTransfer shield

I-

X
X

X
X

X
X

X

X
0

X
0

X
0

XX

X

X

X

X

X

X
0
0

X

X
0
0

X

X
0
0

X
X

Onsite storage
InterimLong-term

0

Shipment
Unshielded
Shielded
DisposalCommercial low-level
Transuranic
High-level treatment/storageaX denotes alternative considered for waste type.
0 denotes alternative applicable to special waste only.

2-32

References--Section 2
1.

Regulatory Guide 1.86, U.S. Atomic Energy Commission, June 1974.

2.

"Plan for Reevaluation of NRC Policy on Decommissioning of Nuclear Facilities," U.S. Nuclear
Regulatory Commission, NUREG-0436, Revision 1, December 1978 and Supplement 1, August 1980.

3.

"Thoughts on Regulation Changes for Decommissioning,"
Commission, NUREG 0590, Rev. 2, 1980.

4.

U.S. Nuclear Regulatory Commission, TMI Program Office Weekly Status Report,
January 4-10, 1981, dated January 12, 1981.

5.

Memorandum from C. Marotta, Office of Nuclear Materials Safety and Safeguards, to K. Kniel,
Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, "Recriticality
Potential of TMI-2," May 14, 1979.

6.

R. DiSalvo et al., "A Further Evaluation of the Risk of Recriticality at TMI-2," U.S.
Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, April 14, 1980.

7.

G. D. Calkins, "Plan for Reevaluation of NRC Policy on Decommissioning of Nuclear Facilities,"
NUREG-0436, Rev. 1, Suppl. 1, August 1980.

8.

R. I. Smith, G. J. Konzek, W. E. Kennedy, Jr., "Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Station," NUREG/CR-0130, June 1978.

9.

R. I. Smith and L. M. Polentz, "Technology, Safety and Costs of Decommissioning a
Reference Pressurized Water Reactor Power Station," NUREG/CR-O130 Addendum, August 1979.

Draft Report,

U.S.

Nuclear Regulatory
for period

3.
3.1

THE ENVIRONMENT AND POPULATION WHICH MAY BE AFFECTEDLOCATION OF THE SITE AND ACCESS FACILITIESThe TMI nuclear power station is located on the northern end of Three Mile Island, one of a groupof islands on York Haven Pond of the Susquehanna River.
Harrisburg, the state capitol of Pennsylvania, is about 12 miles northwest, and Baltimore, Maryland, is about 65 miles south of thestation (Fig. 3.1).
TMI is in Londonderry Township of Dauphin County (Fig. 3.2), at coordinates 4009'10'1 north latitude and 76043'25'' west longitude.
Middletown Borough is three miles north of theisland andHarrisburg International Airport is a short distance to the west of Middletown; both are on theeast bank of the Susquehanna.
Goldsboro Borough is about one mile west of the station, along thewest riverbank in York County.
The surrounding lands are predominantly rural, supporting dairy,
poultry, and agricultural uses, as well as forestry.
The topography undulates slightly, with
1 2
maximum relief of about 200 ft. 'About 200 acres of TMI's 472 acres are occupied by the station, with the facilities for Unit 2
immediately south of Unit 1 (Fig. 3.3).
The island is about 11,000 ft long and 1700 ft wide andis aligned north-south, paralleling the east riverbank 900 ft away.
The west bank is 6500 ft
away.
In between are Beech Island, Shelley Island, and several smaller islands.
Three Mile Island is linked near its center to the east riverbank by Red Hill Dam (Fig. 3.3), andYork Haven Dam curves southward from the end of the island to the west bank.
The island also isconnected to the east bank of the mainland dam by two bridges--a permanent bridge at the northern
end which is used primarily for Unit 1, and a temporary bridge at the southern end for Unit 2.
Access to both bridges is provided by State Highway 441 which parallels the river in this area.
A one-track railroad spur crosses the permanent bridge from the rail line, which also runs close
to the river on the eastern bank.
3.2

GEOLOGY

The TMI site is within the Gettysburg Basin section of the Piedmont Physiographic Province, which
is bounded to the north and west by the Northern Valley and Ridge Province and to the south andeast by the Atlantic Coastal Plain.
The Gettysburg Basin is one of a series of long, narrowbasins of Triassic (225 to 190 million years before present) deposits which extend in broken
patches from Connecticut to North Carolina. The portion of the Gettysburg Basin in Pennsylvania
is referred to as the Triassic Lowland.
North and west of the Triassic Lowland are the folded
and thrust-faulted Paleozoic rocks of the Appalachian Mountains.
The Piedmont southeast of theTriassic Lowland consists of igneous and metamorphic rocks of the Precambrian to Early Paleozoic
Age (more than 600 to 400 million years B.P.). 3
The topography of the area immediately surrounding the site is slightly undulating, with a maximum relief of about 200 ft and highest elevation seldom above 500 ft. From the east, drainage isprimarily by Swatara Creek, which has its mouth near Middletown and flows southwest, and by themore westerly flowing Conewago Creek, which empties into the Susquehanna River at the southernend of the island.
Fishing Creek flows into the Susquehanna west of the site, and the northwesterly flowing Conewago Creek terminates at York Haven.
Three Mile Island has very little
relief, with elevations above sea level ranging from about 280 ft (+280 ft MSL) at the water'sedge to slightly more than +300 ft MSL in the north-central portion of the island.
The island is composed of stratified sand and gravel Containing varying amounts of silt,clay,
and clean sand.
Density values range from loose to very dense.
Boulders are present at depth
and are mainly confined to the lower portions of the-soil zone on the northern end of the island.
Soil thicknesses vary from about 6 ft at the southern end of the island to a maximum of 30 ft

3-1

3-2

Figure 3.1.

Map of the Area within 100 Miles of the TMI Site.

3-3

A

Figure 3.2.

withinMap of the Area

20 Miles of the

TMI Site,3-4

Figure 3.3.

Aerial Photo of TMI Site.

f'i'~'~'~'

3-5
near the center of the island.
Depth of soil is relatively constant at about 20 ft in the vicinity of the plant site.
From one-half to one foot of topsoil, composed of sandy siltwith organic
material, covers the island.
Beneath the soil overburden the, site is underlain by the Gettysburg
shale.
Bedrock surface beneath the site is essentially flat, with an approximate average elevationof +277 ft MSL.
The rock consists of red to brown, interbedded, fine- to medium-grained sandstone,
shaly siltstone, and shaly claystone.
Bedrock ranges from medium-hard to hard, with seismic
compressional wave velocities ranging from 8,500 to 11,1,500 ft/s. The upper 1 to 3 ft of bedrockis weathered.
Based on mapping of surface outcrops in the site area, bedrock strata strike from
651N to 801E and dip 30° to 701 NW.
However,
fairly consistent northerly dips of 37.5O to 45°
4were observed in core borings at the site.
Jointing in bedrock is near vertical and strikes N1O°E.
Many of the joints are healed, and 4
The jointing becomes tighter with depth.
others which are not have been altered by oxidation.
Drilling water loss was noted in two of the core borings (DH-1 and DH-4), suggesting the possibility of open fractures, although there are other causes of water loss.
The water level of theSusquehanna River controls the groundwater level on TMI and most groundwater moves to the riverunder the influence of the hydraulic gradient.
Contamination of groundwater in the Gettysburg
Formation would be highly unlikely since it is confined and under artesian pressure.
The
Gettysburg Formation is a major source of base flow to the Susquehanna River.
Any contaminationof this aquifer would flow to the Susquehanna River and would not affect groundwater users.s
A healed normal fault was encountered in one boring at a depth of about 95 ft. This minor fault
exhibited three inches of normal displacement and a dip of 800 in a northerly direction, the same
direction as the bedding.
Its age has been interpreted as being Late Triassic (190 million years
B.P.). 4 The staff concluded during the Operating License (OL) review that under Appendix A,
10 CFR Part 100, this fault is not capable.
Fractures are present in the rock parallel to thefault. These fractures have been completely healed with calcite.
All Category I structures and tanks are founded in bedrock, except the Category I storage tanks,
which are founded in compacted backfill.
Results of subsurface investigations show that thefoundation conditions in sound bedrock underlying the site were adequate for the proposed facility.
There were no extraordinary geologic engineering problems associated with construction of TMI-2. 6
The maximum seismic intensity that the site has experienced in historic times is intensity V on
the Modified Mercalli (MM) scale based on the March 8, 1899, earthquake at York, Pennsylvania.
There are no capable faults or other geologic structures that could be expected to localize
earthquakes in the immediate vicinity of the site. The maximum credible earthquake for this siteis the largest earthquake that has occurred in the Piedmont Province, intensity VII (MM).
Using
the Trifunac-Brady 7 empirical correlation between horizontal ground acceleration and intensity,
intensity VII (MM) corresponds to an acceleration of 0.12 g.
This value should be used as thehigh-frequency anchor point for the design response spectra. Thus, the Safe Shutdown Earthquake(SSE) was selected to have a horizontal ground acceleration of 0.12 g.
The Operating Basis
Earthquake was selected to be one-half of the SSE, or 0.06 g.
The staff's seismology consultant,
the USGS, concluded that the ground accelerations considered for the site are adequate based on
the seismic history of the site and the surrounding area. 8
3.3

METEOROLOGY

Meteorological conditions presently existing in the vicinity of TMI-2 do not differ significantly
from those described in the Final Environmental Report and the Safety Evaluation Reports relative
to operation of these units. 1 ' 9 Portions of the previously published descriptions are reprinted
below.
3.3.1

Regional Climatology

The climate of southeastern Pennsylvania is primarily continental in character.
Although theproximity of Chesapeake and Delaware Bays, and to a lesser extent the Atlantic Ocean, tends toexert a moderating influence on air temperatures over much of the region, these effects are weak
as far inland as Harrisburg.
Continental polar air originating in Canada is the predominant typeof air mass over the region in winter.
However, these air masses are usually modified and warmed
somewhat as the air descends the eastern slopes of the Appalachians before reaching the southeastern section of Pennsylvania.
Maritime tropical air masses originating over the Gulf of
Mexico or the Carribbean Sea predominate over this region in summer.
Winters are relatively mild
for the latitude; summers are warm and humid.1 0

3-6
TemTemperatures of 90OF or higher may be reached on 20 to 25 days annually over the region.
peratures of OF or lower may be expected on only one or two days annually; temperatures of 32'F
Precipitation is generally well distributedor lower may be expected on about 108 days per year.
throughout the year, but the greatest monthly amounts occur in the summer in association with0
On an annual basis, relative humidity averages around 70 percent.'
thundershowers.
3.3.2

Local Meteorology

Long-term weather records from Harrisburg, Pennsylvania, record the extreme maximum and minimum
Maximum 24-hour precipitation
temperatures as 107 0 F in July 1966 and -14 0 F in January 1912.
totaled 12.55 inches in June 1972 and maximum 24-hour snowfall totaled 21.0 inches in January
1945.10 The normal annual snowfall for the Harrisburg area is 37 inches, while freezing precipiHeavy fog (visibility one-quarter oftation occurs on an average of two to three days per year.
0
a mile or less) occurs an average of 21 days annually.'
Onsite wind data at the 100-ft level (10 m above nearby obstructions) indicate that the predominant wind flow is from the northwest with a frequency of 12.7 percent (Fig. 3.4).
Monitoring of wind speed and direction onsite, as well as atmospheric stability, has been conThe data will allow real-time evaluation of atmospheric
ducted since 1967 and is continuing.
The highest offsite x/Q's
relative concentrations (x/Q) for locations surrounding the site.
should lie within a range from 2.3 x 10-6 sec/m 3 (the annual average continuous release x/Q) to
8.3 x 10-4 sec/m 3 (the 0- to 2-hour x/Q that would be exceeded 5 percent or less of the time at
the exclusion area boundary).
It is possible to place meteorological controls on the cleanup operations in order to optimize
Diffusion optimization can be accomatmospheric diffusion of any gaseous effluents released.
plished by limiting releases to periods having moderate to strong winds, 10 mph or greater, and a
Pasquill-Gifford stability classification of D through A. Although such stability conditions are
preferable, whenever winds are 10 mph or greater, the relative concentration offsite would be
Another factor that would aid insubstantially less than the 0- to 2-hour value given above.
optimizing (reducing) possible x/Q at places where public exposure is likely would be to release
the gases when winds are blowing toward the southern end of the island, where no actual nearby
receptors are located.
Evaporational methods for the disposal of water from the site were also reviewed by staff meteorologists.
Forced evaporation, through addition of heat to holding ponds, may result in fog overthe ponds and the river channels adjacent to the island, depending on air temperature, humidity,
and wind speed.
Natural evaporation of water from the onsite ponds would have a lesser environmental impact than forced evaporation.
However, the rate of natural evaporation would be controlled by air motion over the pond and by increasing the water temperature by solar radiation toincrease the evaporation rate.
In this area of Pennsylvania, the average annual evaporation iswithin the range of 33 inches (lake evaporation) to 45 inches (evaporation-pan measure)," depenSeventy percent of the evaporation occurs from May
ding on the volume of the body of water.
through October.
Because annual total precipitation in this area is expected to exceed
40 inches, significant net water loss to the atmosphere is not expected from closed ponds.
3.4
3.4.1

HYDROLOGY AND WATER USE
Surface Water Hydrology

The major stream affected by liquid discharges from TMI is the Susquehanna River, which has atotal drainage area of 27,400 square miles where it enters the Chesapeake Bay, and of about
25,000 square miles near the TMI site. The TMI plant is about 11 river miles downstream from theThis station has had a continuous period of record since
Harrisburg gaging station of the USGS.
1890.
Data for the Harrisburg gage are assumed to be applicable at the plant site and show theriver's highly variable characteristics:12
Characteristic

Value, cfs

Minimum daily flow (9-18-64)
Average annual discharge
Mean annual flood
Maximum flood of record (1972)

1,700
34,500260,000
1,020,000

3-7

N

NW12

12

9.4

NE

8
WNW

IC

S

5.I.

WSW

SSW

I•

.1 37.

T

SS'•E

INDICATES PERCENT OF TIME WIND BLOWS FROMDIRECTION SHOWN

Figure 3.4.

TMI Annual Average Wind Rose at 100 ft, 1972-1975 Data.

3-8
No large dams or reservoirs exist immediately upstream from the site. The Corps of Engineers hasconstructed a new dam for flood control on the Raystown Branch of the Juniata River, which enters
The project also is used for low-flow augmentathe Susquehanna about 25 miles upstream of TMI.
For low-flow control, the project increases dependable flow at the Raystown
tion and recreation.
Whenever possible, a flow ofThis benefit will also be felt at TMI.
dam site to 480 cfs.
480 cfs is released from the reservoir, provided that a significant drop in the upstream reserHowever, the Corps plans to release at least 200 cfs at all times tovoir level does not occur.
maintain adequate flow downstream.
At the southern end of TMI, York Haven Pond on
At TMI the Susquehanna is about 1.5 miles wide.
the Susquehanna is impounded by York Haven Dam and Red Hill Dam."'1 3 The river channel is
The western
braided and flows around several islands, forming several channels (Fig. 3.5).
channel is shallow and conducts about two-thirds of the normal flow; 1 4 the eastern channel,
approximately 5.5 ft deep, conducts a minor portion of the flow,, which overtops Red Hill Dam partof the year. 1 4 The central channel is about 10 ft deep under normal flow conditions and joins
the western channel to spill over the central portion of York Haven Dam.
Some flow is diverted
along the dam to the York Haven Generating Station, a hydroelectric facility.
Downstream from York Haven Dam, the Susquehanna is impounded by three dams at Safe Harbor,
Holtwood, and Conowingo before entering Chesapeake Bay (Fig. 3.6).
Just downstream from the
Holtwood Dam is Muddy Run, a pumped storage facility capable of altering river discharge byremoving water for storage or releasing stored water to rejoin the5 river discharge.
The Muddy
16
Run removal rate is 24,500 cfs and the release rate is 32,000 cfs.s'
At the head of Chesapeake Bay, water from the Susquehanna River mixes with the discharge of othertributaries in a wide, shallow area.
From the head to the mouth, Chesapeake Bay is about
195 miles long and 3 to 35 miles wide, with a mean depth of about 30 ft, although a deep channel(120 ft) is present.
The complex water circulation and salinity patterns are influenced by thetides, tributary river discharges, and weather." 7
Low flow studies of the Susquehanna River have been conducted using data from the Harrisburg(upstream of TMI) and Marietta (downstream) gaging stations.
Table 3.1 contains flow-duration
data for the Susquehanna River at Harrisburg, based upon daily flows.
These data indicate thefrequency of flows for given durations.
For example, a flow of 5000 cfs or less, lasting
183 days, could be expected about once in 20 years.
The minimum flow of 1600 cfs occurred on November 29, 1930, as the result of an ice blockage ofthe river.
However, the minimum daily discharge since the construction of a dam downstream fromthe Harrisburg water filtration plant was 1700 cfs on September 18, 1964.
The percentage of thetime that the flow is equal to or less than specified discharge is summarized in Table 3.2.
The mean monthly flows for the period 1891-1979 are shown in Table 3.3.
the monthly variation in average flows.

These values illustrate

The amount of river dilution received by liquid waste from the TMI plant depends primarily upon
the river flow at the time of discharge.
The maximum capacity of the turbines at York Haven Dam
is about 16,000 cfs.
During periods of low flow when the river discharge is less than
16,000 cfs, essentially all of the river flow will pass through the generating units.
Therefore,
any discharged wastewater downstream of the dam will be fully mixed with the river after itpasses through the turbines.
For higher river flows, some flow will also pass over the dam.
This case is much less critical for the purposes of computing concentrations of contaminants.
However, mixing will occur downstream aided by the turbulent flow over the dam.
The flow of theSusquehanna is split into several segments at the head of TMI and Shelley Island.
The amount ofwater flowing past the discharge point (in the middle channel between TMI and Shelley Island) is
about 25 percent of the total river flow.
Below York Haven Dam, additional mixing occurs and thefull flow of the river may be used in determining dilution factors.
For steady flow in the river, dilution of a contaminant released from the plant is approximately
inversely proportional to the flow rates.
For the purposes of estimating an average concentration in the Susquehanna River for a given release, a "plug flow" model with no longitudinal
mixing is used.
The correct flowrate for this model is not the "arithmetic mean flow," Q, but
the "reciprocal mean" flow, (-1/•-).
The arithmetic mean flow is heavily weighted by flood
events, while the reciprocal mean flow is weighted in favor of low flows, for which concentrations would be higher.
The reciprocal mean flow has been calculated for flow measurements at
Harrisburg to be about 12,000 cfs.
The reciprocal mean is conservative because deviations from

3-9

Figure 3.5.

York Haven Pond of the Susquehanna River,Showing the Flow Patterns in the Channels
and over the Dam at Three Mile Island.

3-10

HAVEN DAM--- 4
SAFE HARBOR
DAM
MUDDY
aA,,- RRU N
HOLTWOODPENNSYLVANIAMARYLANDIDAMNORMALMAXIMUM WATERELEVATION (ft)109.2

CONOWINGOMUDDY RUN
(PUMPED
STORAGE_520

HOLT WOOD
SAFE HARBORYORK HAVEN

171
228
255

INDUSTRIAL USERS
I. PENNSYLVANIA
SUPPLY CO.
2. YORK HAVEN POWER CO.
3. BRUNNER ISLANDSTATION
4. WRIGHTSVILLE WATERSUPPLY CO.
5. PEACH BOTTOM STATIONDOMESTIC WATER SUPPLIES

6.COLUMBIA BOROUGH
13. BAINBRIDGE NAVALTRAINING STATION
7. CITY OF LANCASTERINCLUDING PORT
8. SAFE HARBOR VILLAGEDEPOSIT9. HOLTWOOD VILLAGE
14. PERRY POINT
10. CITY OF CHESTERVETERANS HOSPITAL
II. CITY OF BALTIMOREHAVRE DE GRACE
15.
12. CONOWINGO VILLAGE

Figure 3.6.

Major Inputs and Outputs of the Susquehanna Showing Principal
Water Users below the TMI Plant. (Modified from "Final
Environmental Impact Statement Related to Operation of theFulton Generating Station, Units 1 and 2," U.S. NRC Docket
Nos. 50-317 and 50-318, 1973.)

3-11

Table 3.1.

1892- 1972aMagnitude and Frequency of Low Flow,

3
Discharge (ft /s) for Indicated
Recurrence Interval in Years
5
10
20
30

Period ofConsecutive
Days

2

7

3,900

2,800

2,400

2,200

14

4,200

3,000

2,600

2,300.

2,100

2,000

30

4,900

3,400

2,900

2,600

2,400

2,200

60

6,300

4,000

3,300

2,800

2,600

2,400

120

9,000

5,400

4,400

3,600

3,200

2,900

183

13,000

8,100

6,300

5,000

4,400

3,800

aFrom L.V.

Page and L.C.

Shaw,

2,100

Table 3.2.
Percentage of the Timethat the Susquehanna River
Flow at Harrisburg Equals
or Is Less thanSpecified Valuesa
Percentof Time

cfs

2,000

Negligible

3,200

2

4,000

5

5,200

10

7,600

20

11,00030

15,00040

19,00050

35,00070

50,00080

79,00090

110,000

95

aFrom L.V.

1,900

"Low Flow Characteristics ofPennsylvania Streams," Commonwealth of Pennsylvania,
of Environmental Resources, 1977.

Flow,

50

Page and L.C.

Shaw,

"Low Flow Characteristics of Pennsylvania Streams," Commonwealthof Pennsylvania, Department ofEnvironmental Resources, 1977.

Department

3-12

Table 3.3. Mean Monthly Flowsof the Susquehanna Riverat Harrisburg for the
Period 1891-1979

Month

Mean Flow(cfs)

January

37,700February

40,100March

77,700April

82,600May45,700June

26,300July

15,200August11,700

September

11,700

October

17,300November

25,900December

33,300the simple plug flow model, such as reservoir storage and longitudinal mixing, tend to desensitize the average concentration to very low flows, which essentially increases the effective mean
flow.
3.4.2

Surface Water Uses

The surface water of the Susquehanna River downstream from Harrisburg is acceptable for all
general uses, e.g., supporting aquatic life, recreation, and primary contact.
It is not an
attractive source of public water supply because of occasional high sulfate levels and highamounts of wastewater-derived coliform bacteria. 1 8
Below Harrisburg, late summer algal blooms
occur that are indicative of high nutrient levels, primarily phosphates and nitrates.1 9 This is
attributable both to wastewater treatment and runoff from agricultural areas.
The river and the streams in the vicinity of TMI presently are used for water supplies, bothpublic and industrial; power generation; boating; sport fishing; and recreation.
Sport fishingis done in all streams in the general area of the site; however, there is no commercial fishing.
Brunner Island steam-electric generating station, five miles downstream, is the nearest user ofsome of the river as potable water.
The locations of downstream surface water users are indicated in Figure 3.6.
The available information on water supplies downstream from TMI issumers include public water supplies, industries, and utilities:

summarized below; the con-- The Pennsylvania Supply Company takes 2200 gpm per nine-hour work day from the mouth ofConewago Creek (near the southern tip of TMI) for sand and gravel processing.
The water isreleased into the'Susquehanna River after clarification.
- York Haven Power Company (a wholly owned subsidiary of Met-Ed) owns a hydroelectric generating station at York Haven with a total installed capacity of 20,000 kWe.
The station isabout two miles downstream of the TMI discharge structure.
A dam across the SusquehannaRiver impounds 8000 acre-ft of water for power generation, air conditioning, cooling, andfire protection.
Potable water is obtained from two wells.

3-13- PP&L owns and operates Brunner Island, a 1415-MWe-capacity steam-electric generating stationon the west shore of the Susquehanna River five miles downstream from TMI.
The station
utilizes 1155 cfs of river water for the circulating water system and ash removal; the water
is returned to the river without reconcentration.
River water for potable and cycle makeup
demineralizer requirements is treated prior to use.
- The Wrightsville Water Supply Company has a public water supply intake on the SusquehannaRiver 16.2 miles downstream from the nuclear station.
The intake is utilized as a summer
reserve supply, with treatment facilities provided in the system.
- The borough of Columbia takes an average supply of two million gallons per day from theSusquehanna River by an intake located 16.7 miles downstream from TMI.
The water is treatedprior to distribution.
- The city of Lancaster withdraws an average of 8 million gallons per day from the Susquehanna
by-an intake located about 17 miles downstream from TMI.
The water is treated prior to
distribution.
Additional water is taken from the Conestoga River.
- The Safe Harbor Water Power Corporation owns and operates a hydroelectric generating stationon the Susquehanna River about 27.2 miles downstream from TMI.
The dam across the river
impounds 92,000 acre-ft for power generation.
Water for a public water supply system forthe village of Safe Harbor is withdrawn from the reservoir at a rate of 25,000 gallons per
day and receives complete treatment prior to distribution.
- A public water supply system serves the village of Holtwood on the eastern side of theSusquehanna River 34.7 'miles downstream from TMI at the Holtwood Hydroelectric Station.
About 22,000 gallons per day is withdrawn from the 19,300 acre-ft Holtwood Reservoir andtreated prior to distribution.
- The Muddy Run pumped storage generating station is 38 miles downstream from the TMI nuclear
station.
It is a remotely controlled 800-MWe-capacity station, operating between an upper
reservoir and Conowingo Reservoir.
River water is used for fire protection.
Potable water
is supplied by deep wells.
- Peach Bottom Nuclear Generating Station is on the west bank of the Susquehanna River about
41 miles downstream from TMI.
Unit 1, rated at 40 MWe, was part of the Atomic Energy
Commission's Power Reactor Demonstration Program and has been decommissioned.
Two additional nuclear units, each rated at 1000 MWe, are now operating and withdraw a total of3450 cfs from the river for cooling purposes.
- The city of Baltimore has an intake on the river which draws water from Conowingo Reservoir
about 49 miles downstream from TMI.
Baltimore is permitted to withdraw up to 250 million
gallons per day when the river flow exceeds 5000 cfs, but is limited to 65 million gallons
per day at lower river flow.
Two water-treatment plants provide treatment prior to distribution.
The last significant withdrawals occurred in 1968; future use of potable waterfrom the Susquehanna is expected to be limited to periods of severe drought conditions which
necessitate the use of alternative supplies such as the Susquehanna.
- Philadelphia Electric Company owns and operates a hydroelectric generating station at
Conowingo Dam, which impounds 321,500 acre-ft in Conowingo Reservoir.
The generating station
has a capacity of 513,000 kWe; it is located 50 miles downstream from TMI.
Potable water
for the station and for Conowingo Village is obtained from the reservoir.
Treatment isprovided to an average supply of 12,000 gallons per day.
- Arundel Sand and Gravel Co. has been issued a permit to withdraw 900,000 gallons per day forindustrial use from intakes located about four miles upstream from Havre de Grace in thevicinity of Susquehanna Park.
- The city of Havre de Grace is permitted to withdraw about 12.5 million gallons per day fromintakes located on the west side of the river near its mouth.
About 1.8 million.gallons per
day are currently used for potable supplies and are treated prior to distribution.
Futureplans are to expand water supplies to other areas of Harford County.
- Bainbridge Naval Training Station and the town of Port Deposit have been withdrawing about
225,000 gallons per day for potable supplies from intakes located on the east side of theriver at Port Deposit.
The water receives treatment prior to distribution.

3-14- Perry Point Veterans Hospital and the town of Perryville withdraw about 400,000 gallons per
day from intakes located on the east side of the river. The water is treated before use.
3.4.3

Surface Water Quality

3.4.3.1

Chemical and Bacteriological

Water quality data have been given in monitoring records provided by the licensee over the period
1967-1976 for CP and OL licensing actions,' 9 and in USGS Water Resource Data for water year 1977
(October 1976 through September 1977).13 Water quality data for the Susquehanna River near TMI
and at three other locations are presented in Table 3.4.
The important characteristics are amoderately high total hardness, averaging about 130 mg!L as CaC0 3 (typically due to about 35 mg/Lof calcium and 10 mg/L of magnesium), and a high and variable sulfate concentration, averaging
about 80 mg/L, but occasionally exceeding 200 mg/L. The sulfate limit set by the revised statewater quality criteria is 250 mg/L. 20 The sulfate contamination is largely attributable to aciddrainage from old coal mines within the watershed, which also contribute a high and variable
total iron concentration, ranging from 0.4 to 5 mg/L, often in excess of the state limit of
1.5 mg/L.
The high sulfate concentration results in a relatively low alkalinity, averaging about 60 mg/L of
CaCO 3 . As a result, about half of the hardness is noncarbonate or permanent (not deposited as
scale on heating), but the high total hardness would severely impair the detergent action of
conventional soaps.
Oxygen concentrations, water temperature, and sediment loads of the river have been altered by
the construction of impoundments.
The impounded pools allow suspended sediment to settle out andincrease the surface area exposed to solar radiation.
Increased retention time within the impoundments allows increased solar exposure and subsequent heat gain.
Temperatures recorded atHarrisburg during water year 1977 ranged from 37.4 0 F to 85.1 0 F.
Discharge over spillways andthrough hydroelectric facilities increases turbulence, providing greater oxygenation of thewater.
The fecal coliform count, attributed to domestic and agricultural waste, often is in excess ofthe state limit of 2000 colonies per 100 mL (mean of five consecutive days).19
Water quality throughout the Chesapeake Bay is generally adequate for supporting aquatic life andrecreation.
In some areas, poor tributary water quality causes local degradation and limiteduse.
Chesapeake Bay is an estuary, a dynamic physicochemical system in which freshwater fromtributary rivers mixes with seawater.
The water quality at any location in the Bay is a functionof the relative contributions from and mixing of the different water masses (Table 3.4).
Generally, salinity depends on the amount of freshwater discharged, precipitation, and evaporation.
In Chesapeake Bay salinity is lower in spring and summer and higher in autumn and winter.21
Tributary discharge and tides affect the intrusion of seawater toward the head of the Bay.
Because of the tides and river discharges, the Bay is constantly receiving nutrients and salts,
the concentrations of which vary over time and area.
3.4.3.2

Radioactivity

The USGS records' 3 include measurements of radioactivity, both dissolved and suspended, atHarrisburg.
The tritium concentration was measured during the 1977 water year and found to be
fairly constant, 178 pCi/L.
Gross beta activity was measured on November 8, 1976, and reported
as follows:
Dissolved gross beta: 2.4 pCi/L as Cs-137
1.9 pCi/L as Sr-90/Y-90
Suspended gross beta:

0.4 pCi/L as Cs-137<0.4 pCi/L as Sr-90/Y-90

The method of reporting suggests that these are direct beta counting results using cesium andstrontium standards, and not specific radiochemical measurements of these radionuclides.

3-15

Table 3.4.

Water Quality of the Susquehanna River at Several Locationsa
Susquehanna River
Near TMIb

Parameter
GeneralHoltw~odDamConowiogo
Pond

ChesapaakeBay

6.6 - 7.9

7.3 - 8.4
1.7 - 7.4
5.4 - 9.9

Characteristics

pH (units)
BOD
DO
TDS
Alkalinity (as CaCO 3 )Hardness (as CaC0 3 )Specific conductance, pmhos
Coliforms (cols/lO0 mL)

6.5 - 8 .2e
1.0 - 5.4
5.1 - 15.2
78 - 397
23 - 17246 - 242
77 - 356
15 - 21,000

6 - 8
0 - 4.4

3.8 - 9.2
167--

Cations
SodiumPotassium
Magnesium
CalciumIron

2.3 1.3 3.5 12 0.01 -5 2 .9e5.7 14 -

20e
204e

2.4

11
37
0.17

8 2.1 7.2 21 0.2 -

17
3.3
18
49
0.6

2.6 - 10.8
1.3 - 6.1
2.4 - 2213.5 - 44
0.0 - 0.23200 115 375 130 0.27 -

4700
150
550
190
1.9

10 - 2270 - 205
0.06 - 1.1

4.1 - 2030 - 130
0.0 - 0.87

5800 - 7800
770 - 1150
0.02 - 0.14

AnionsChlorideSulfatePhosphate
Radiologicalf
Dissolved gross u as Unat,
na'
pg/L

<1.6

Suspended gross a as Unat0
pg/L

0.7

Dissolved gross p as Cs-137,
pCi/L
Suspended gross p as Cs-137,
pCi/L
Dissolved uranium,

pg/L

2.4
0.4

-

-

0.06

-

-

aThe two values presented are the maximum and minimum in mg/L unless otherwise stated.
bFrom "Water Resources Data for Pennsylvania, Water Year 1977,"

Vol.

2,

"Susquehanna andPotomac River Basins," U.S. Geological Survey, Water Data Report PA-77-2, Harrisburg, Pa.,1978.
cFrom "Final Environmental Statement Related to Operation of the Fulton Generating Station,
Units 1 and 2," U.S. Nuclear Regulatory Commission,.NUREG-75/033, 1975.
dFrom "Final Environmental Statement Related to Operation of the Calvert Cliffs Nuclear Power
Plant, Units 1 and 2," U.S. Atomic Energy Commission, Docket Nos. 50-317 and 50-318, 1973.
eFrom "Final Supplement to the Final Environmental Statement Related to Operation of the Three
Mile Island Nuclear Station, Unit 2,"' U.S. Nuclear Regulatory Commission,
No. 50-320, December 1976.
fNovember 8, 1976, from Footnote b above.

NUREG-0112,

Docket3-16
Radium-226 was measured on the same date by the radon method as 0.08 pCi (alpha)
Gross alpha activity on the same date is reported as:
Gross dissolved alpha:<1.6 pg/L as natural U (<1.08 pCi/L)

Gross suspended alpha:

0.7 pg/L as natural U (0.5 pCi/L)

per liter.

The conversions to pCi(a)/L assume the specific alpha activity of chemically pure natural uranium(U-234, U-235, and U-238 in their natural abundances) to be 0.6735 pCi/pg (approximately 1.5
A measurement of uranium concentration, presumalpha disintegrations per minute per microgram).
ably by the chemical (fluorimetric) method, also made on November 8, 1976, gave a value ofThis is presumably included in the dissolved gross alpha, but it would contribute
0.06 pg/L.
only about 0.04 pCi/L, assuming natural composition.
The following limits for radioactivity in the Susquehanna River are proposed by the revised Stateof Pennsylvania criteria:19
Total alpha:

Not to exceed 3 pCi/L over natural backgroundTotal beta:

Not to exceed 1000 pCi/L over natural backgroundTritium:

Not specified, but may be included in total beta

It is difficult to define a strictly natural background for fission products and tritium, becauseof small but significant contributions from nuclear weapons testing, which depend on latitude.
The radioactivity
The contribution from the commercial nuclear fuel cycle is negligible.
observed in the Susquehanna River at Harrisburg during.1977 is below the level regarded as normal
for this latitude zone.*
3.4.4

Groundwater Hydrology

The site has a water table elevation of about 280 ft MSL, depending upon the Susquehanna River
Site borings and observation wells indicate that waterstage, which is normally at 277 ft MSL.
The water
table elevations vary about 5 ft from a high at the island's center to the shores.
The nearest potable water supplies are threetable gradient is about 0.006 toward the river.
All of thesewells located on the east bank of the Susquehanna River directly across from TMI.
Since
wells have groundwater elevations above the river and above the groundwater level at TMI.
they are upgradient, these wells are not affected by site activities.
As stated in Section 3.2 the site is underlain by sandy silts, sands, gravels, weathered bedrock,
The Gettysburg Formation has basic artesian characand hard siltstone (Gettysburg Formation).
Groundwater flow is highly anisotropic along the strike direction,
teristics in the site area.
The leakage ofwith specific capacities ranging from 0.33 to 15.0 gpm per foot of drawdown.
groundwater from the Gettysburg Formation would be expected to be upward but would vary considerTherefore, effluents
ably with the degree of jointing and relationship to strike direction.
released accidentally from the plant should not migrate into the Gettysburg Formation.s
At theMeasures have been taken by the licensee to monitor the quality of groundwater at TMI.
request of the NRC staff, the licensee installed a series of eight monitoring wells and, subsequently, nine observation wells at the TMI site. These wells were sited so as to detect leakageof contaminated water from the Unit 2 reactor and auxiliary buildings and outside storage tanks.
The monitoring wells have been sampled at about one-week intervals since installation in late
The observation wells, installed to clarify anomalous data from some of the moniJanuary 1980.
toring wells, also have been sampled at about one-week intervals since completion in late
The samples have been tested for radioApril 1980.
Weekly sampling of all wells is continuing.
To check for errors, the analyses were done by several
activity and chemical contamination.
different laboratories.

*In Ref. 22, the average radioactivity levels in surface water in the Chicago area are given
In Ref. 23, an average tritium level of 287 pCi/L
as: alpha, 0.1-3 pCi/L; beta, 5-10 pCi/L.
in surface water is cited for latitudes 30°-504N.

3-17
Initial analyses of water samples from some monitoring wells showed concentrations of tritium
above normally occurring background concentrations in the vicinity of major plant structures.
However, these tritium levels are all less than the Maximum Permissible Concentration (MPC) of
3 x 10-3 pCi/mL, as specified in 10 CFR Part 20, Appendix B, for unrestricted locations.
Concentrations of other radionuclides also were found to be below MPC for unrestricted areas.
The highest tritium levels occurred at wells that were near the Unit 2 borated water storage tank
(BWST).
The fittings and valving appurtenant to the BWST have leaked onto the immediately surrounding ground surface.
At the request of the NRC staff, additional observation wells were
drilled to confirm that the BWST was the source of contamination.
Soil samples were taken near
the BWST during the drilling of the observation wells.
The concentrations of various radionuclides were measured at various depths in the soil column, from ground level to a depth of over
25 ft. The tests on these soil samples show higher concentrations of tritium at locations above
the water table and close to the BWST.
Infiltration of precipitation has carried the contamination downward to the water table, through which it has been transported to other locations.
Based on the well data, soil samples, and the fact that the BWST has leaked, the analysis of thetest data supports the hypothesis of the BWST leakage through the fittings as the source ofcontamination.
If the monitoring program indicated that significant concentrations of nuclides were present in
the groundwater, there would be several methods to prevent additional contamination of the site
groundwater and the Susquehanna River.
Examples include:- The monitor wells could be pumped so as to remove the contaminated groundwater.
are located to permit a large volume of groundwater to be pumped.

The wells- A slurry wall of bentonite or other relatively impermeable material could be installed tosurround the contaminated area. The wall would inhibit the passage of groundwater both toand from the site.
A minimum of about a year's time would be available to install the wall,
because it would take the contaminated groundwater at least this long to migrate to theSusquehanna River.
(See Appendix V for concentration estimates in the event of an accidental release.)
3.5
3.5.1

ECOLOGY
Aquatic*

The biota of the Susquehanna River include organisms usually associated with flowing waters and(because of the impoundments) with standing waters.
Aquatic macrophytes are rare because of fluctuating flows and water levels and the type of riverbottom substrates, which in most of the free-flowing areas are sand or rock. Most macrophytes
are found in the fluctuating littoral zone of the impoundments. 1 6 A dominant source of primary
production are algae.
The annual cycle of algal succession follows a pattern indicative of a,mesotrophic to eutrophic aquatic system:a spring bloom of diatoms, a summer abundance of green
algae, and a late summer/early fall increase in blue-green algae and flagellates.1 6 ' 2 4 Thispattern is also representative of algal succession in a lake and indicates the importance of theimpoundments in the trophic structure of the river.
Zooplankton composition and abundance are variable.
The dominant groups are rotifers(Brachionus sp.), cladocerans (Bosmina sp.), and copepods (Cyclops sp.).1 9 The periodic highabundance of rotifers also suggests excessive domestic waste loadings to the river.
The composition and abundance of the benthic invertebrate communities are influenced strongly by
the quantity and quality of available substrate.
Fluctuating flows and impoundments that act as
settling basins control the deposition of sand, silt, and finer-sized particles in the various
sections of the river.
The finer particles tend to settle in the impoundments; the larger,
heavier particles settle in the flowing sections.
The available sediments are primarily sand,with varying amounts of silt,clay, and6 organic
matter. The most abundant benthic invertebrates
24
are tubificid worms and insect larvae.1 '*For selected details on York Haven Pond, Conowingo Pond, and upper Chesapeake Bay relative tothis environmental impact statement, see Appendix E.

3-18
The fish community can be characterized as a warm water assemblage, and is dominated by membersof the minnow, perch, and sunfish families. 1 ' 1 5 - 1 7 '1 9 The lower portion of the river (below theConowingo Dam) receives spawning migrations of some anadromous species, primarily members of theherring family and striped bass.
There is an active
sport fishery throughout the river. Major
9 24
64
species sought are crappie, bass, and sunfish.1 '1 '1 'In the shallow water of upper Chesapeake Bay, aquatic macrophytes are abundant, and plants suchas cord grass (Spartina sp.) and wild celery (Vallisneria sp.) are quite productive as well as an
attractive food source for waterfowl.
Planktonic algae, primarily diatoms, dominate throughout
the Bay, although green, blue-green, red, and brown algae also are present.17
The invertebrate fauna is diverse and includes a gradation from freshwater to marine types,
depending on the salinity and bottom substrate.
Important commercial fisheries are oysters,
clams, and bluecrabs.1 7 ' 25 In 1976, the harvest of important shellfish exceeded 11,000 tons. 25
The fish fauna of the Bay also is diverse, and dominant species change with the season and migratory patterns.
The commercial fishery of the Bay is important to the Bay states; major species
harvested from Maryland waters include menhaden, striped bass, and bluefish.' 5
3.5.2

Terrestrial

Land use and major vegetation types within a one-mile radius of the plant are shown in Figure 3.7.
The northern two-thirds of Three Mile Island is occupied by the plant and the periphery of mostof the northern portion is forested. Only the eastern half of the lower, third and the western
periphery of this lower third are forested; the remainder of this portion of the island is mostly
covered by grasses and low shrubs.
The plant community is less than 80 years old and of a typecommon to the area.
A 400-ft-wide bulldozed strip parallels the construction road that runs fromthe plant to the temporary access bridge at the southern end of the island.
The nearby islands
9 26
are similarly forested, except for Shelley Island, which is partly used for agriculture. 1'1 , 6
Downriver,1 6 the riparian, bottomland and wetland vegetation are similar to vegetation in com9 24 26
parable habitats in the vicinity of Three Mile Island.1'1 ' 'The composition of the forest is indicative of a stage in the succession of a floodplain sere
between the cottonwood-willow and oak-hickory stages.
The majority of the forest land on Three
Mile Island can be classified as bottomland hardwood forest, stream terrace hardwood forest, or
black locust forest.
The distribution of terrestrial habitats is shown in Figure 3.7.
The bottomland hardwood forests,
about 230 acres, are in low-lying areas where flooding from the Susquehanna River has been frequent.
Silver maple, ash, and river birch are the most common overstory species; they are foundin association with American elm, sycamore, catalpa, cottonwood, and tulip poplar.
The stream hardwood forests are found on the higher bottoms and stream terraces. About 90 acresof this forest type were reported.
Red oak is the most abundant species, although one stand hadchestnut oak and beech as dominant overstory species.
Species occurring less frequently include
black cherry, black locust, black walnut, mockernut hickory, silver maple, tulip poplar, and ash.
The black locust forest is found primarily in a woodlot of about 50 acres on the southeastern
portion of Three Mile Island.
This is a relatively young second-growth forest.
In some areas
black cherry and black walnut occur with black locust as the dominant species; in others, black
locust alone is dominant.
Less frequently occurring species include sassafras, cottonwood, ash,staghorn sumac, and box elder.
The miscellaneous areas include a very old 3-acre stand of beech forest on Beech Island, hedgerows andwindbreaks planted between agricultural fields, and a double row of red pine bordering a
cart road on Shelley Island.
Of the 212 species of terrestrial vertebrates found in the TMI vicinity, there were 179 birds,
19 mammals, 8 reptiles, and 6 amphibians (Ref. 11, Tables 2.7-5 through 2.7-8).
Small-gameanimals were eastern cottontail rabbit and gray squirrel.
Mammalian predators were longtail.weasel and red fox.
The largest mammal on the site was the white-tail deer.
Four species of
upland game bird were found onsite:
ring-necked pheasant, American woodcock, mourning dove, androck dove.
Whistling swan, Canada goose, nine species of dabbling duck, seven species of diving
duck, and three species of mergansers also were reported. This sampling is typical of: the fauna

3-19

4-,
4-'

~0
CL

So)

4-u
4-

3-20
found downriver.1 6
Because the Susquehanna River is a major flyway, large numbers and many
species of migratory and resident waterfowl nest and feed on the ponds and reservoirs along theriver.
No endangered species are known to occur on the plant site.11'27 The site lies within the rangesof occurrence of three endangered species--southern bald eagle (Haliaetus leucocephalus),
peregrine falcon (Falco peregrinus), and Indiana bat (Myotis sodalis)--and it is possible that
individuals could visit the site, particularly during periods of migration.
Several American
osprey, a species designated as of undetermined status, have been seen onsite, but no nesting
activity has been observed." 1
The bald eagle and osprey, which are likely to take food from theriver, can be seen downriver, as well as at the site.
3.6

POPULATION,

3.6.1

LAND USE,

AND OTHER SOCIOECONOMIC CONSIDERATIONSPopulation Groups and Geographic Areas of Interest

The staff has identified four population groups and geographic areas that have the potential forbeing affected by decontamination activities at TMI-2.
The area within 15 miles of TMI, hereafter referred to as the "vicinity" of TMI, is expected to experience such impacts as are generated by in-moving technicians, traffic to and from the site, and various cleanup activities.
The second subarea subject to potential impact is the Susquehanna River and those who use the°river and adjacent lands for recreational purposes.
The Chesapeake Bay is the third subarea
identified by the staff.
Potential impacts to the Bay would affect those who make their livingfrom the Bay's resources, those who use the Bay for recreation, and those who consume Bay shellfish and finfish in marketing areas across the nation.
Finally, the transportation of waste from
TMI could impact people and areas adjacent to the selected transportation routes.
For illustrative purposes, the analysis in Section 9 features two feasible routes having extremes in distancetraveled; these routes are to the commercial burial site at Richland, Washington, a distance ofabout 2750 miles, and to the waste management facilities at West Valley, New York, 370 miles from
TMI.
3.6.1.1

Population and Land Use in the Vicinity of TMI

An estimated 489,000 people live in the vicinity of TMI (see Table 3.5).
The population densityof 570 persons per square mile *is substantially higher than the state as a whole.
The relatively
high population density in the 10- to 15-mile zone is attributable to Harrisburg, the state
capital, which is located 12 miles to the northwest of TMI; in 1970 Harrisburg had a populationof 68,000.
Goldsboro, with 576 people, is the municipality closest to TMI (about 1 mile) and islocated directly opposite the nuclear station on the west shore of the Susquehanna.
Other communities within the vicinity of TMI are listed in Table 3.6.

Table 3.5.

Population and Population Density Distribution
in the Vicinity of TMI, 1978

RadiusDensity(miles)Population%(pop./sq.

0-537,842

7.7

353.1

5-10

116,973

23.9

487.8

10-15

334,305

68.3

654.0

Total

489,120

100.0

570.0

Source:Commonwealth of Pennsylvania, "The.SocioEconomic Impacts of the Three Mile Island
Accident."

mi.)

3-21

Table 3.6.

Municipalities

in the Vicinity of TMIDistancefrom
TMI (miles)

Municipality

County1970
Population

Goldsboro
RoyaltonYork
Dauphin

576
1,040

1
2

Middletown
Highspire
Yorkhaven

Dauphin
Dauphin
York

9,080
2,947
671

2½
4
4

Elizabethtown

Lancaster

8,072

6

Manchester
Steelton
New Cumberland

York
Dauphin
Cumberland

2,391
8,555
9,803

6½
7
9

Hummelstown
Hershey

Dauphin
Dauphin

4,723
7,407

10

Harrisburg

Dauphin

68,061

12

Source:U.S. Nuclear Regulatory Commission, "Final Environmental
Statement Related to Operation of Three Mile Island
Nuclear Station Units 1 and 2," December 1972.

While urban development is concentrated around population centers and along major transportation
corridors, much of the land in the vicinity of TMI is devoted to farming. The soils, which
exhibit good to excellent characteristics, combined with favorable physiographic and climatological features produce higher than average crop yields in the state. Field crops such as corn
and wheat, and dairy, poultry, and livestock operations are prominent. 28
3.6.1.2

Transportation

The vicinity of TMI is broadly delineated by four transportation routes that encompass an irregularly shaped area. Interstate 83, which is oriented north-south, connects Baltimore, York, andHarrisburg.
Interstate 76, the Pennsylvania Turnpike, connects Harrisburg with urban centers to
the eastand west.
Interstate 95 forms the southern border of the area and connects Baltimore
City with other East Coast cities. State Route 10, which is oriented north-south, connects 1-76
and 1-95 on the eastern boundary.
With the exception of the state route, the roadway net provides high-speed and high capacity access to the area.
U.S. Route 30, another high capacity
road, connects York with Lancaster and divides the region of impact into two subareas.
3.6.2

Socioeconomic Status of the Region

The immediate and short-term socioeconomic impacts of the accident at TMI have been reported in a
number of studies.
In this section, the status and continuing social and economic impacts of theaccident are discussed in terms of relocation, property. values, the economy, government, andrecreational fishing.
Most of the discussion that follows is in terms of county-level data.
3.6.2.1

Relocation

Of 14,000 households within five miles of TMI prior to the accident, 1504 experienced at least
one member relocating to a new permanent address between March 28, 1979, and March 1980.
Followup studies of those who relocated have not yet been undertaken to determine the extent towhich the accident was a causative factor in the relocation decision. The data indicate that the3-22
mobility rate is less than the rates for both the nation and the region.
In addition, these data
indicate that 44 percent of those who relocated within the 5-mile zone; an additional 30 percent
relocated within the greater Harrisburg area. 2 9 Although specific reasons for relocation have notbeen established, the data appear to indicate that relocation to escape believed dangers of TMI
may have been initially overstated by the staff (see DPEIS for TMI Unit 2, p. 3-19).
3.6.2.2

Housing and Property Values

In the vicinity of TMI, the latest available census data indicate that 152,600 housing units were
occupied.
Of this total, two-thirds of the units were occupied by their owners and nearly thesame proportion of the units were at least 20 years old in 1970.30 Housing studies of Dauphin andLancaster Counties indicate that construction is meeting or perhaps exceeding demand.3 1 ' 3 2 The
housing stock in York County has increased by an average of 1.8 percent annually since 1970; thisrate accommodates household growth but not vacancies and housing losses. 3 3
Two independent studies evaluated the impact of the accident on property values, sales, thenumber of days required to sell a property, and perceptions of mortgage lenders, realtors, andcontractors. 2 8 ' 3 4 In general, both studies conclude that the accident did not have a significant
and continuing negative effect within 20 miles of TMI.
With respect to sales price, historical.
and cross-sectional analyses indicate no adverse impact even when distance from the nuclear
station is used as a category with which to classify properties.
For a period of at least one
month after the accident, the number of sales within 5 miles of the station did decline dramatically and brought the real estate market in this area to a virtual standstill.
However, during
the last half of 1979 sales resumed a more normal pattern.
By the end of 1979, sales had dropped
by only 7.3 percent below the 1978 rate despite high interest rates and the limited availabilityof mortgage funds in the last two quarters of 1979.
During 1979, an average of 84.3 days was
required to sell a residential property, which compares with 79.3 days required in 1978.
This
6.3 percent increase is not considered significant in view of the depressed mortgage market inlate 1979.
Interviews with realtors indicate that the accident may have had an adverse impact insecond quarter sales figures but that economic conditions had primary importance in explaining
overall trends for the year.
Some realtors have indicated that visibility of the towers has had
a negative effect on the sale of individual homes; however, such instances are not numerous, orsevere enough to affect the housing market within 5 miles of TMI.
3.6.2.3

The Economy

The economy of the region (Cumberland, Dauphin, Lancaster, Lebanon, and York Counties) is diversified, although manufacturing is the dominant activity.
In 1978 manufacturing firms employed.
31 percent of the total labor force in the five counties.
The three leading manufacturers are inmachinery (except electrical), fabricated metals, and food and kindred products.
The retailtrade industry is the second leading source of employment.
Census data for 1970 indicate thatthe occupational pattern of the region closely resembles that of the nation in the high proportion of professional, managerial, craftsman, and operative categories.
Dauphin and Cumberland
Counties lead the other counties in the proportion of their labor forces employed in white-collar
occupations, a reflectionof the concentrated opportunities for government employment in the27
greater Harrisburg area.
In the paragraphs that follow,
reviewed.

the effects of the TMI accident on four areas of the economy are

An estimated 65 percent of all manufacturing firms within 20 miles of TMI (1141 firms employing
101,700 people) experienced no adverse economic impact for the 12-month period ending March 1980.
Of the 245 firms which experienced some adverse impact, 68 percent (167 firms) reported no continuing effect by July 1979; 17 firms reported economic losses for the full year after the accident.
In addition to economic impacts, an estimated 80 percent of the firms (918) indicated that
their products suffered no adverse "image" effects in the market place; almost 7 percent of thefirms (79) indicated that their products suffered one year after the accident.
Of the 20,197 nonmanufacturing units within a six-county area surrounding TMI,* an estimated66 percent (13,418 units) experienced no economic loss as a result of the accident; 28 percent(5727 units) experienced adverse economic impact.
By July 1979, 70 percent of those previously*The six counties are Dauphin,

Lancaster, Cumberland,

Lebanon,Perry, and York.

3-23
An additional 26 percent experienced no losses
experiencing losses (4000 units) ceased to do so.
after December 1979, and 4 percent experienced economic adversity for the full one-year period.
The services and retail sales sectors accounted for 80 percent of all lost sales, which amounted
The proportion of
to $74 million; wages lost in all sectors were estimated to be $5.5 million.
nonmanufacturing units that indicated TMI was having a negative image effect has remained stable
at 10 percent during the year.
The accident has been cited as being the cause of an estimated $5 million loss to touristThis loss occurred during the four-week periodoriented businesses in the six-county area.
following the accident.
Tourist industry losses continued through the summer of 1979 when thetrend peaked and continued into the fall at reduced levels.
Other factors, such as the outbreakof polio in the Amish Community, the gasoline shortage, adverse weather conditions, and worsening
national economic conditions, also were seen as having a depressing effect on tourism, particularly during the summer and fall seasons.
Indicators for the 1980 tourist season--hotel/motel
occupancy rates and attendance figures--are approaching preaccident levels.
Even among thosebusinesses indicating continuing adverse impact, the impact is evaluated as being slight.
Economic losses in the agricultural sector were estimated to be less than $500,000; these losses
were suffered during the five-week period after the accident.
Farmers who sell milk directly tocustomers--"milk juggers"--are a notable exception to the general experience.
For juggers, milk
sales continued to linger 10 percent below 1978 levels as late as October 1979.
However, thisexperience does not extend to farmers directly selling fresh fruit and vegetables to customers.
3.6.2.4

State and Local Government

The jurisdictional structure of the area in the vicinity of TMI is notable for its complexity.
Parts of five counties--Cumberland, Dauphin, Lancaster, Lebanon, and York--lie within 15 miles ofthe station, and these jurisdictions are further subdivided into 37 townships and 25 cities orboroughs.
One of the cities, Harrisburg, is the capital of Pennsylvania.
As part of theirgeneral administrative responsibilities,
county-level governments are charged with planning and35
managing emergency situations.
Between March 30, 1979, and mid-August 1980, the State of Pennsylvania incurred TMI-related
personnel and operating costs of $761,482.28 Increased local government expenditures for emergency services and employee overtime costs amounted to $148,355.
As of April 30, 1980,
Metropolitan Edison's insurance covered 95 percent of the claims filed and 57 percent of theexpenses reported (see Table 3.7).

Table 3.7.
Status of Local Government Insurance Claims
Against Met Ed Related to TMI Accident(as of April 30, 1980)

Government
Level

Reported
ExpensesInsurance
ClaimsFiled

Insurance
ClaimsPaid

Municipal$ 90,319$49,981$45,444

County58,03639,55839,429Total$148,355$89,539$84,873

Source:Commonwealth of Pennsylvania.

3-24
3.6.2.5

Recreational

Fishing

In the months following the accident, surveys of fishing activity on the York Haven Pond of theSusquehanna indicated below normal or low normal levels of fish caught, fish kept (harvested),
number of anglers, and total hours spent fishing (pp. 56-61 of Ref. 14).
By July 1979, fishingconditions had returned to historically normal levels.
The post-accident depression of fishingindicators has been attributed to altered behavior by local anglers--the bulk of fisherman on the
York Haven Pond--who either ceased fishing activity or fished elsewhere.
In late July 1979, about 4000 gallons of water were released to the Susquehanna River from TMI.
The amount of radiation was insignificant and posed no threat to public health and safety. The
release was widely publicized by the news media. 3 6 Harvest indices declined in August andreturned to normal in September.
3.6.2.6

Summary and Conclusions

The socioeconomic effects of the accident were mainly precipitated by the evacuation of households and individuals and by the altered behavior of those who remained in their communities.
In
retrospect, the socioeconomic effects, as indicated by lost sales and wages, decline in services,
and increased expenditures, were, with some notable exceptions, short-lived.
For a small numberof business firms, tourist-oriented businesses, state and local government agencies, and milk
juggers, the cost imposed by the accident continued for months after the accident.
However, evenamong these groups the adverse socioeconomic impact appears to have declined by the end of 1979.
3.6.3

Psychological

Setting

While some level of stress exists in all communities, the unique situation of the TMI-2 accident
has sensitized the TMI population to plant-related activities.
This results in increased levelsof stress in some TMI community members, with greater awareness of the psychological consequences of the accident to themselves and the unknown prospects of the decontamination operations. The Kemeny Commission concluded, "The major health effects of the accident appear to havebeen on the mental health of the people living in the region
Stress is a complex process that i~nvolves environmental, social and psychological events that cancause harm, disruption, or lOss.4°'68'69 These events, called stressors, are subject to people's
appraisal of them.
An event .will elicit a stress response only if it implies threat or danger.
If it does, the stressor will evoke physiological and psychological arousal that attempts to copewith or reduce believed threat.
The consequences of arousal and response may include performance
deficits, somatic distress, depression or anxiety.
Thus, the experience of stress is determinedby appraisal of the stressor.
If the event is believed benign, a stress reaction is unlikely.
If the event implies threat, a stress response is likely. Distress is a disruptive reaction to a
stressor.
The length of the stress response is.determined by a person's ability to reduce thethreat and the duration of the stressor.
Some stressors, however, have consequences that last
beyond their duration.
Stress is a complex phenomena that also presents a number of obstacles in studying it, especiallydefinitional and measurement problems.
The experience of stress varies widely across individuals
and events.
Low levels of stress may facilitate some task performances and inhibit others.
In
examining stress, it is useful to have a baseline measure with which to compare subsequent stressresponses.
Retrospective questions about past experiences or reactions are susceptible to bias,
such as memory or mental status. 5'
In spite of these problems, a wealth of information about
stress has accumulated.
Various psychological studies of the TMI communities have attempted todiscern the nature and the extent of distress associated with the accident and with continued
concern about decontamination.
Various studies agree that immediately following the TMI accident there was an increase inpsychological distress in the communities surrounding the plant and that symptoms diminished by
mid-summer 1979.37'38'42-4s
Some researchers conclude that the drop indicates the threat linked
to the accident is acute or event specific. 3 7 ' 4 5 ' 4 7 Others suggest that while several items
drop, others related to anxiety continue to remain high after the accident because of continuing
uncertainty. 4 4 ' 45 ' 4 7 ' 4 8 The continuing tension is largely due to the continued presence of acrippled reactor and anticipation of decontamination activities as well as to the accident experience.
Often-mentioned evidences of stress are the public's expressed distrust of those responsible for these activities and their fears of possible, unanticipated, or underestimated health3-25
consequences, or of delays in cleanup.
DuPont, 4 9 in his analysis of ten years of nuclear energy
TV-news coverage, suggests that public concern is amplified by the national TV media's generally
negative coverage.
Baum4 7 suggests that a history of negative images linked to radiation and theatomic bomb is an important determinant in the public's assessment of nuclear power as unsafe orvery risky.
Other research indicates that a portion of the public fears nuclear power for many of the same
reasons. 4 1 The KemenyCommission 3 7 suggests that psychological distress in the TMI community is
induced and exacerbated by a lack of confidence in those parties directly involved in thecleanup.
They conclude, however, that distress associated with the TMI accident is acute, withno significant increase of long-term mental or physical health problems anticipated.
Recent studies, 50 ' 5 1 however, suggest measurable long-term psychological impact.
Bromets° finds
distress, measured by high levels of anxiety and depression, among the sample of mothers livingin the vicinity of TMI.
Mothers who exhibited symptoms of distress were likely to have one ormore of the following characteristics:
to have prior psychiatric history before the accident, tolive within 5 miles of the plant, to not have extensive social support networks, and to be pregnant during the accident.
Symptoms reported for the mothers living in the plant vicinity, however, are only marginally greater than those reported for the control group.
Houtss5
finds long-term psychological differences compatible to Bromet's 5 0 findings when hecompares a TMI sample with a sample of people who live over 40 miles from the station.
He is,
however, very cautious in his interpretation because of the use of retrospective data.
In conclusion, the available TMI psychological studies 3 7 ' 3 8 )'4 2 - 4 8 ' 5 0 i'5
suggest to the staff thatsome portion of the population in the vicinity of TMI is stressed. There is probably more stressmeasured at TMI than found in other communities.
The accident and post-accident events, however,
have sensitized the population to psychological consequences of activities associated with the
TMI nuclear station.
This sensitivity may amplify stressors or make stressors appear to berelated to TMI.
The current stress levels in downriver communities have not been specificallystudied.
Houts, 4 2 ' 5' however, uses a sample of the population over 40 miles from TMI as his
reference for normal stress levels in his analysis.
His sample is assumed to reflect stresslevels downriver.
3.6.4

Susquehanna River/Chesapeake Bay

The most notable physical features of the impact region are the Susquehanna River and the Chesapeake Bay.
Draining an area of 27,510 square miles, the 450-mile-long Susquehanna is a majorriver in the eastern United States and supplies about 50 percent of the fresh water in the bay.
The Chesapeake Bay is one of the largest estuaries in the world, having a surface of about 4400
square miles, a length of nearly 200 miles, and more than 7000 miles of shoreline. 5 2 ' 5 3 The
Susquehanna River-Chesapeake Bay system is the focus of intense commercial and recreational
fishing, boating, and water consumption.
3.6.4.1

Commercial Fishing

In 1978, the latest year for which summary data are available, the commercial harvest of shellfish and finfish
in the Chesapeake fisheries* was 214 million pounds, with a landed value of
54 55$89 million. 'Data on harvests for selected years are summarized in Table 3.8.
rently earn their living as "watermen" (see Table 3.9).

About 11,150 individuals cur-For both the Maryland and Virginia portions of the Bay, shellfish are more important economicallythan finfish, representing more than 90 percent of the landed value in 1977 and 1978.
In*Chesapeake fisheries summaries incorporate data on Chesapeake Bay and its tributaries and theAtlantic Ocean off the Maryland-Virginia coast.
Since 1970, the contribution from the Atlantic
Ocean has grown as a percentage of the total Chesapeake fisheries harvest.
In 1970, the ocean
harvest represented 14 percent of the weight and 20 percent of the value; by 1978, these percentages had increased to 26 percent and 45 percent, respectively.

Table.3.8.

Chesapeake Fisheries:

Summary of Catch for Selected Yearsa
Virginia

Maryl andFinfishShellfish

FinfishGrand Total

Shellfish

Year

Millionsof Pounds

Millions of
Dollars

Millionsof pounds

Millions of
Dollars

Millionsof Pounds

Millions of
Dollars

Millionsof Pounds

Millions of
Dollars

Millionsof Pounds

Miilions of
Dollars

1960

21

2

48

12

307

6

59

14

436

35

1965

35

2

52

11

435

9

70

18

592

40

1970

16

2

64

17

495

12

55

10

630

41

1975

15

2

49

21

360

13

85

20

509

56

1977

16

3

46

29

544

27

64

29

670

87

1978

14

3

47

32

84

10

68

44

214

89

aData refer to Chesapeake Bay, including its tributaries,

and the Atlantic Ocean.

Numbers may not add to totals becauseof rounding.
Source:U.S. Department of the Interior, Fish and Wildlife Service, "Fishery Statistics of the United States" for 1960,
1965, 1970, and 1975; U.S. Department of Commerce, National Marine Fisheries Service, "Maryland Landings Annual
Summary 1978," Current Fisheries Statistics No. 7814 and "Virginia Landings Annual Summary 1978," Current Fisheries
Statistics No. 7815.

3-27

Table 3.9.
Registered Full-Time
Watermen in Maryland andVirginia, 1977-1979
Year

Maryland

Virginia

Total

1977

4,003

4,700

8,703

1978

5,235

5,060

10,295

1979

5,349

5,803

11,152

Source:National Marine Fisheries Service.

Maryland waters, commercial fishing relies principally on blue crabs, oysters, soft-shelledcrabs, surf clams, menhaden, and striped bass.
Oysters represent the single most important cash
crop in Maryland, exceeding $14 million and $15 million in 1977 and 1978, respectively.
The
Virginia commercial fishery relies principally on harvests of hard blue crabs, oysters, surf
clams, sea scallops, croaker, menhaden, and flounder.
Until 1978 the top Virginia fishery interms of pounds caught and dollar value was the menhaden.
Landings of menhaden peaked in 1977
with a harvest exceeding 501 million pounds that had a landed value in excess of $19 million.
Harvests of blue crabs exceeded 36 million
pounds in both 1977 and 1978, with a dollar value of
54 5',$6.7 million for each of those years.
That portion of the finfish and shellfish harvest not sold directly to restaurants, markets, andindividual consumers is sold to processors and wholesalers.
In 1978, the value of processed fish
products in Virginia and Maryland was $221 million. 5 6 's 7
(An undetermined portion of fish entering the processing/wholesaling system originates in the Atlantic Ocean off Virginia and Maryland
and in water bodies elsewhere.)
The processing and wholesaling system employs an average of6900 people throughout the year in 350 establishments.
Each year during the peak months thenumber of employees added to establishment payrolls rises.
For instance, in 1978, the average
employment for the peak months was 8900, an increase of 29 percent above the yearly average.
These data are summarized for selected years in Table 3.10.
The economic benefits generated by harvesting Chesapeake seafood include stimulation of economic
activity in other segments of the fishery industry and in other economic sectors.
These benefits
accrue to areas throughout the nation in addition to the Maryland-Virginia area. A recent study
58
concluded that the multipliers for income and for employment were 3.47 and 3.13, respectively.
That is, for every dollar of seafood harvested from the Chesapeake fishery, an additional $2.47
is generated in other sectors, and each full-time waterman supports an additional 2.13 jobs.
As
indicated in Table 3.11, the 1978 Chesapeake harvest had a landed value of $89 million and produced an additional impact of more than $219 million; the 11,150 full-time, licensed watermen of
the bay helped to create employment for an additional 23,700 people.
The Chesapeake-

Bay commercial

fishing industry can be characterized

in the following ways:*

Harvesting, processing, and wholesaling is organized around widely scattered individuals or
small groups using simple--often homemade--gear.
The waterman operates on a low-capital,
low-overhead basis that reflects the need to function in an environment of great uncertainty
and flux.
In many instances harvesting and processing operations are supported by family
members, are labor intensive, and, with respect to processing and packaging facilities, aremarginally profitable.

*Information in this section is based on References 59-64.

Table 3.10.

Chesapeake Fisheries Processing and Wholesale Activity for Selected Years

MarylandYear

Value ofProcesssd
Product

Seasonal b
Employment

1960

31

1965

37

1970

Virqinia

Yearly b
Employment

NumberofPlantsValue ofProcesssd
Product

Seasonal b
Employment

6,687

4,271

253

24

4,965

4,112

285

30

58

4,812

3,790

205

1975

98

4,264

3,395

175

1977

126

3,831

2,761

1978

118

3,632

2,775

Total

Yearly b
Employment

NumberofPlantsValue ofProcessad
Product

Seasonal b
Employmentb

Yearly b
Employment

NumberofPlants

7,123

2,726

378

55

13,810

6,997

631

4,714

2,914

336

66

9,679

7,026

621

42

5,834

4,021

260

100

10,646

7,811

465

63

6,048

4,513

186

161

10,312

7,908

361

169

99

5,226

4,041

180

225

9,057

6,802

349

169

103

5,281

4,101

185

221

8,913

6,876

354

aData are in millions of dollars.
bEmployment averages for both the season and the year.
Source:From U.S. Department of the Interior, Fish and Wildlife Service, "Fishery Statistics of the United States," for 1960, 1965, 1970,
and 1975; U.S. Department of Commerce, National Marine Fisheries Service, "Processed Fishery Products, Annual Summary 1977," Current Fisheries Statistics No. 7503 and Annual Summary 1978, Current Fisheries Statistics No. 7803.

co

3-29

Table 3.11.

Economic Impact of Chesapeake Bay
Commercial Fisheries
Dollars

Employment

89,000,000

11,150Multiplier

3.47

3.13

Gross impact

308,830,000

34,900Net impact

219,830,000

23,750Basea

aLanded value and number of employees.

-

The industry is subject to seasonal cycles and biological and environmental fluxes.
men work closely within these natural constraints.

-

Recent productivity data indicate a trend toward a general, long-term decline as well as acyclical low point in the Bay's resources.
A general decline occurring simultaneously withincreased pollution represents the single most important threat to continued fisherystability.

-

The industry is faced with increasing competition as products from other states supplant
Maryland and Virginia products.
Crabs from the entire Gulf of Mexico and the Southeast
Atlantic Coast and oyster imports are particularly notable in this regard.
In addition, themarket for fresh Maryland seafood is being subjected to pressures from frozen seafood
products.

-

Despite the rise in capital costs and operating expenses, gross incomes appear to be reasonable.
Income has not experienced greater downward pressure because of the existence of an
extensive non-cash economy and because of self-imposed limitations on the modernization ofequipment.
Among processors and wholesalers, profit margins are low and variable and are
subject to pressure from rising transportation and labor costs.

-

The markets for Chesapeake fishery products are widely scattered and are composed of many
individual outlets:
restaurants, retail stores, chain stores, and institutions.
Distributors have individual marketing systems based on personal knowledge and geographic routes.
Regular markets are found as far north as Montreal, as far south as Texas, and west to
Chicago and Los Angeles.

-

The Chesapeake fishing industry is supported by small, scattered, and self-sufficient community structures that are characterized by extended family settlement, kinship and informal
relationships, and non-cash exchanges for materials and services.

3.6.4.2

Water-

Sport FishingSport fishing on the Chesapeake Bay is an activity that rivals commercial fishing in both intensity and value.
Estimates based on the 1976 sport fishing survey conducted by the State of
Maryland indicate that almost 2 million finfishing and crabbing trips were made to the Maryland
tidewater (comparable data for Virginia waters do not exist). 6 5 Approximately 88 percent ofthese trips were on private, non-charter boats; the remainder, 242,000 trips, were on charterboats.
Almost 70 percent of Maryland residents on private fishing boats were residents of Anne
Arundel, Baltimore, and Prince Georges Counties and of Baltimore City.
Of nonresident interviewees, 93 percent originated from four jurisdictions:
Washington, D.C., Delaware, Pennsylvania, and Virginia.

3-30
The total harvest of finfish and crabs during the period May through October--the peak sport
fishing season--amounted to 18.6 million pounds.
As indicated in Table 3.12, which compares the
sport and commercial harvest, overall sport activity is equivalent to a considerable percentageof the commercial harvest.
Of the total commercial finfish catch, more than 2.7 million pounds
were menhaden and gizzard shad, species not sought by sport fishermen. 6 5
If the commercial catchof these species is ignored, the sport catch was greater than the commercial finfish harvest.

Table 3.12 Chesapeake Bay Sportand Commercial Harvests, 1976
Harvest (pounds)
CategorySport

Finfish15,434,600

17,581,700

Crabs

3,160,900

20,885,500

18,595,500

38,467,200

Total
Source:Commercial

Maryland Department of
Natural Resources, Fisheries
Administration, "1976 Maryland Chesapeake Bay Sport
Fishing Survey," June 1977.

One economic measure of the significance of sport fishing is the expenditure on variable costs.
Variable costs are out-of-pocket payments for transportation, food, bait, boat gas, lodging, andboat rentals.
These expenditures maintain the shore-support service industry for sport fishingin the Chesapeake Bay area.
Based on Maryland's 1976 survey of fishing trips, average party
size, resident/non-resident proportions, and adjusted variable cost data, the staff estimatesthat the total out-of-pocket expenditures in Maryland for 1979 were $26.6 million (seeIt should be noted that the calculation does not take into account fixed costs,*
Table 3.13).
which could easily double the staff's estimate.
Sport fishing in the Susquehanna River occurs from the vicinity of TMI to Havre De Grace (seeAppendix E).
The sport fishery resource in the river primarily serves local residents, although
fishermen from the Baltimore area and from Pennsylvania are attracted to the river in sizable
numbers.
Much of the fishing effort is concentrated in the ponds, particularly Conowingo Pond,
and near the tailwaters of the dams and hydroelectric facilities. Because of the variety of fish
species (freshwater and anadromous, lotic and lentic, bottom and pelagic), the fishery resource
offers the sport fishermen a variety of experiences.
3.6.4.3

Wildlife Resource Use**In addition to Chesapeake Bay's importance to the state's commercial fishing industry and to
sport fishing, the Bay's marshes and woodlands provide many thousands of acres of natural habitat*Fixed costs include expenditures for rods, reels, boats, trailers, and clothing.
However, suchcosts do not represent money spent on the fishing day nor do such costs necessarily representmonies retained within Maryland.
**Except as noted, information in this section is from Reference 52.

3-31

Table 3.13.
Number of Boat Trips and Out-of-Pockel
Expenditures by Sport Fishermen in Maryland, 1979
Total

FactorResident

Non-Resident

Angler trips inprivate boats

1,503,017

263,687

1,766,704

9.71

16.08

10.66

14,594,295

4,240,087

18,834,382

Half Day

Full Day

Total

Charter boat trips

12,47735,88348,360

Variable cost/boat trip (dollars)

121.32

175.24161.32

1,513,710

6,288,137

7,801,847

Variable cost/angler (dollars)
Annual expenditure (dollars)

Annual expenditure (dollars)aVariable cost data for 1976 were inflatedby 1.348 to account for change
in the Consumer Price Index up to December 1979.
Source:H.J. Speir,
Survey."et al.,"1976 Maryland Chesapeake Sport Fishing

The Bay area is in the path of the Atlantic Flyway and provides
for a diversity of wildlife.
About 75 percent of the wintering
wintering areas and feeding grounds for migrating waterfowl.
Other waterfowl species
population of Canada geese are attracted to the Delmarva Peninsula.
attracted to the peninsula in large numbers include ducks, whistling swans, birds requiring thewetlands of the bay area for food and other habitat requirements, and a variety of game birds.
The interspersion of forest and farmland and the presence of shoreline and wetlands provides avariety of food chains and habitat niches that support a diverse population of small mammals,
reptiles, and amphibians.
Consumptive
The wildlife resources of the bay region are utilized by man in two basic ways.
utilization involves hunting and trapping to provide food and clothing for personal use and forsale.
Nonconsumptive activities are undertaken purely for enjoyment and include bird watching,nature walking, and nature photography.
Although detailed information on the economic value of hunting and trapping activities is not
Based onavailable for the bay area, expenditures by sportsmen are thought to be significant.
data from the 1975 national survey of hunting and fishing, the typical migratory bird hunter inMaryland and Virginia spends about $147 annually for such items as license fees, rentals, food,lodging, gasoline, and guides.* Applying this figure to the total number of licensed hunters in
Maryland and Virginia, total out-of-pocket expenditures would be more than $38.4 million annually
in 1979 dollars.
Maryland is second only to Louisiana in terms of the value of pelts and food*Dollar values cited have been calculated by multiplying values cited in Reference 66 by 1.426
to account for changes in the Consumer Price Index up to December 1979.
As in the estimate
for sport fishing expenditures, the expenditures cited here for hunters do not include fixed
costs.

3-32
trapped.
Nutria and muskrat--mostof which are trapped around the Chesapeake Bay--support a67$3 million industry in Maryland.
With respect to nonconsumptive activities, the 1975 survey indicated that the total number of
person-days engaged in bird watching, nature walking, and photographing wildlife was 30 percent
more than the total number of person-days spent fishing and hunting.
3.6.4.4

Boating

Since World War II, there has been a large growth in use of the bay by water-oriented recreationalists, such- as power boat and sailboat owners.
Population growth, urban development around thebay, increased disposable income and leisure time, and improved road systems have been responsible for increased boating activity.
Although specific boat ownership and use data are notavailable, the bay is second to Long Island Sound in the intensity of boating use. 6 7 Persons
participating in nonconsumptive activities spend large amounts of money on food, lodging, andrentals; however, estimates of these expenditures do not exist.
3.6.4.5

Summary

The preceeding sections described the Chesapeake Bay as a socioeconomic resource.
However, dataon landed values of commercial fisheries, employment in fish harvesting and processing, sport
fishing expenditures, and the consumptive and nonconsumptive utilization of wildlife only begin
to dimension the character, even the economic character, of the Bay.
Dollar values can only
inadequately express the subjective enjoyment of the recreational boater, hunter, and fishermanor the waterman's sense of independence.
Nor can economic multipliers adequately portray thethousands, of small businesses, families, and communities that support the hunting and fishingactivities on the Bay.
It is also clear that a comprehensive and exact accounting of the economic values associated with
Chesapeake Bay fishing and hunting activities--even if such an accounting were possible--would bea small percentage of the value of agricultural products and even less significant when comparedwith the value added in the manufacturing sector. However, the socioeconomic significance of the
Chesapeake Bay lies not in economic comparisons but rather in the area's distinctiveness.
Thefishing industry, the waterman, the waterfowl hunting activity, and the small rural communities
existing on the Bay's resources represent a unique culture and tradition that enhance the diversity of Maryland and Virginia.
And the benefit of such diversity may well be impossible tomeasure.

References--Section 3
1.

"Final Environmental Statement Related to Operation of Three Mile Island Nuclear Station
Units 1 and 2," U.S. Atomic Energy Commission, Docket Nos. 50-289 and 50-320, December 1972.

2.

"Draft Supplement to the Final Environmental

Statement,

Three Mile Island Nuclear Station

Unit 2," U.S.

Nuclear Regulatory Commission,

NUREG-0066,

July 1976.

3.

A.J.

"Structural Geology of North America," Harper & Row,

4.

"Final Safety Analysis Report, Three Mile Island Nuclear Station, Unit 2," Metropolitan
Edison Co. and Jersey Central Power & Light Co., Chapter 2.5 and Appendix C, 1974.

5.

T. Nicholson, "Three Mile Island Nuclear Station Unit 2 - Groundwater Study," Memorandum to
TMI Unit 2 File, Hydrologic Engineering Section, U.S. Nuclear Regulatory Commission,
April 19, 1979.

6.

"Safety Evaluation Report for Three Mile Island, Unit 2," U.S. Atomic Energy Commission,
September 1969.

7.

M.D. Trifunac and A.G. Brady, "On the Correlation of Seismic Intensity Scales with the Peaks
of Recorded Strong Ground Motion," Bull. Seismol. Soc. Am. 65(2):307-322, 1975.

Eardley,

New York, NY,

1972.

3-33
8.

"Safety Evaluation Report for Three Mile Island, Unit 2," U.S.
sion, November 1975.

9.

"Safety Evaluation Report on Three Mile Island," U.S.
1974.

10.

"Local Climatological Data, Annual Summary with Comparative Data, Harrisburg,
Department of Commerce, Environmental Data Service, published annually.

11.

"Supplement II, Environmental Report, Three Mile Island Nuclear Station - Unit 2,"
Metropolitan Edison Company, U.S. NRC Docket No. 50-320, 1974.

12.

Water Supply Papers of the United States Geological Survey (continuously published andupdated).

13.

"Water Resources Data for Pennsylvania, Water Year 1977," Vol. 2, "Susquehanna and Potomac
River Basins," U.S. Geological Survey, Water Data Report PA-77-2, Harrisburg, PA, 1978.

14.

C.R. Hickey and R.B. Samworth, "Non-Radiological Consequences to the Aquatic Biota andFisheries of the Susquehanna River from the 1979 Accident at Three Mile Island Nuclear
Station," U.S. Nuclear Regulatory Commission, NUREG-0596, November 1979.

15.

T.W. Robbins, and 0. Mathur, "The Muddy Run Pumped Storage Project:Amer. Fish. Soc. 105:165-172, 1976.

16.

"Final Environmental Statement Related to Operation of the Fulton Generating Station,
Units 1 and 2," U.S. Nuclear Regulatory Commission, NUREG-75/033, 1975.

17.

"Final Environmental Statement Related to Operation of Calvert Cliffs Nuclear Power Plant,
Units 1 and 2," U.S. Atomic Energy Commission, Docket Nos. 50-317 and 50-318, 1973.

18.

"Water Quality Criteria, 1972, Report of the Committee on Water Quality Criteria," National
Academy of Sciences, National Academy of Engineering, Washington, DC, 1972.

19.

"Final Supplement to the Final Environmental Statement Related to Operation of Three Mile,
Island Nuclear Station, Unit 2," U.S. Nuclear Regulatory Commission, NUREG-0112,
Docket No. 50-320, December 1976.

20.

"Report to the Environmental Quality Board Concerning Proposed Revision to Water Quality
Criteria (Chapter 93), Wastewater Treatment (Chapter 95) and Industrial Wastes(Chapter 97)," Pennsylvania Department of Environmental Resources, Harrisburg, PA, 1979.

21.

E.J.

22.

"Formerly Utilized MED/AEC Sites Remedial Action Program; Radiological Survey of Site A,
Palos Park, Chicago, Illinois," U.S. Department of Energy, Final Report, April 1978.

23.

"Tritium in the Environment," National Council on Radiation Protection and Measurements,
NCRP Report No. 62, Washington, DC, March 1979.

24.

W.A. Potter et al., "An Ecological Study of the Susquehanna River in the Vicinity of theThree Mile Island Nuclear Station," Annual Report for 1975, prepared by Ichthyological
Associates, Inc., Etters, PA, for Metropolitan Edison Company, February 1976.

24.

"Chesapeake Fisheries, Statistics No. 7702, Annual Summary, 1976," National Oceanic andAtmospheric Administration, National Marine Fisheries Service, Washington, DC, 1979.

26.

"Environmental Report, Operating License Stage, Three Mile Island Nuclear Station Units 1
and 2," Metropolitan Edison Company and Jersey Central Power and Light Company, October
1971.

27.

"List of Endangered and Threatened Wildlife and Plants," U.S. Department of the Interior,Fed. Reg. 44(12):3636-3654, January 17, 1979.

Perkins,

Nuclear Regulatory Commis-

Atomic Energy Commission,"The Biology of-Estuaries and Coastal Waters,"

June 3,

PA,"

U.S.

A Case History," Trans.

Academic Press,

London,

1974.

3-34
Commonwealth of

28.

"The Socio-Economic Impacts of the Three Mile Island Accident,"
Pennsylvania, undated.

29.

Interview with Marilyn King Goldhaber,

30.

Data supplied by Urban Data Processing of Burlington, Massachusetts.

31.

"Housing Market Analysis," Lancaster County Planning Commission, 1976.

32.

"Housing Plan,"
1979.

33.

"York County Housing Plan Update,"

34.

Unpublished data supplied by Dr. Hays B. Gamble, The Institute for Research on Land andWater Resources, the Pennsylvania State University.

35.

William W. Chenault et al., "Evacuation Planning in the TMI Accident," prepared for theFederal Emergency Management Agency, January 1980.

36.

The Washington Post, August 1,

37.

J.G. Kemeny, "Report of the President's Commission on the Accident at Three Mile Island,"
U.S. Government Printing Office, Washington, DC, 1979.

38.

Report of the Governor's Commission on Three Mile Island, Commonwealth of Pennsylvania,
February 26, 1980.

39.

G. Mandler, "Anxiety," in D.L. Sells (Ed.), International
the MacMillan Co. and the Free Press, New York, 1968.

40.

H.J. Eysenck, W. Arnold, and R. Meile,
Seabury Press, New York, 1979.

41.

P. Slovic, B. Fischhoff, and S.
21(4):14-20, 36-39, 1979.

42.

ReportP.S. Houts et al., "Health-Related Behavioral Impact of the TMI Nuclear Incident:
submitted to the TMI Advisory Panel on Health Research Studies of the Pennsylvania Department of Health," Part I, April 1979.

43.

Preliminary
C.B. Flynn, Mountain West Research, Inc., "Three Mile Island Telephone Survey:
Report on Procedure of Findings," U.S. Nuclear Regulatory Commission, NUREG/CR-1093,
October 1979.

44.

M. Rogovin and G.T. Frampton, Jr. (NRC Special Inquiry Group),
Report to the Commissioners and to the Public," April 5, 1979.

45.

B.P. Dohrenwend et al., "Technical Staff Analyses Report on Behavioral Effects:
President's Commission on the Accident at Three Mile Island," October 1979.

46.

Unpublished data supplied by Pennsylvania Governor's Office of Policy and Planning.
collected in March 1980.

47.

A. Baum, "Psychological Stress for Alternatives of Decontamination of Reactor Building
Atmosphere," U.S. Nuclear Regulatory Commission, 1980.

48.

"Public Opinion in Pennsylvania Toward the Accident at TMI and its Aftermath," prepared
for General Public Utilities Corporation, by Field Research Company, October 1980.

49.

R.L. DuPont, "Television Evening News Covers Nuclear Energy:
Media Institute, Washington, DC, March 1980.

Pennsylvania Department of Health,

Tri-County Regional Planning Commission,

Phase VIII:York County Planning Commission,

November 13,

1980.

1978, and Phase IX:

1977.

1979.

"Stress",

Lichtenstein,

Encyclopedia of Social Science,"Anxiety,"in Encyclopedia of Psychology,"Rating the Risks," Environment"Three Mile Island, aTo theData

A Ten-Year Perspective",

3-35
50.

E. Bromet, "Three Mile Island: Mental Health Findings," Western Psychiatric Institute andClinic, Pittsburgh, PA, October 1980.

51.

P.S. Houts et al., "Health-Related Behavioral Impact of the TMI Nuclear Incident:
ReportSubmitted to the TMI Advisory Panel on Health Research Studies of the Pennsylvania Department of Health," Part II, December 1980.

52.

"Chesapeake Bay:
Existing Cor..itions Report-Summary,"
Baltimore District, December 1973.

53.

"Chesapeake Bay:
Future Conditions Report, Volume 1,neers, Baltimore District (n.d.).

54.

"Maryland Landings Annual Summary 1978," U.S.
Current Fisheries Statistics No. 7814.

55.

"Virginia Landings Annual Summary 1978,"
Current Fisheries Statistics No. 7815.

56.

"Fisheries Statistics of the United States," for 1960, 1965, 1970,
U.S. Department of Commerce, Fish and Wildlife Service.

57.

"Processed-Fishery Products, Annual Summary," for 1977 and 1978, U.S. Departmentof Commerce, National Marine Fisheries Service, Current Fisheries Statistics No.
and 7803.

U.S.

Army Corps of Engineers,

Summary," U.S.

Army Corps of Engi-

Department of Commerce,

U.S. Department of Commerce,and 1975,

7503

58.

"Economic Impacts of the U.S. Commercial Fishing Industry," Centaur Management Consultants,
Inc., prepared for the National Marine Fisheries Service, pp. 254-257, 269, January 1975.

59.

"A Comprehensive Review of the Commercial Oyster Industries in the United States,"
Department of Commerce, National Marine Fisheries Service, p. 18, March 1977.

60.

"Maryland's Chesapeake Bay Commercial Fisheries," prepared for Maryland Department of
Natural Resources, Tidewater Administration, by Development Sciences, Inc., November 1978.

61.

"Report to Congress, the Molluscan Shellfish Industries and Water Quality:
Problems andOpportunities,"
U.S. Department of Commerce, National Marine Fisheries Service,
September 1977.

62.

Interview with W.R.

Prier,and Community Development,

Office of Seafood Marketing,

U.S.

Maryland Department of Economics

October 30, 1980.'

63.

Interview with L. Simms,

64.

Interview with T.J. Billy and D. Dressel,
ment of Commerce, November 5, 1980.

65.

H.J. Speir et al., "1976 Maryland Chesapeake Bay Sport Fishing Survey," Maryland Department of Natural Resources, Fisheries Administration, June 1977.

66.

"1975 National Survey of Hunting, Fishing and Wildlife-Associated Recreation,"
preparedby National Analysts for U.S. Department of the Interior, U.S. Fish and Wildlife Service.

67.

Telephone conversation with W. Jabine,
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68.

R.S. Lazarus, "Stress," in: D.L. Sells (Ed.), International
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69.

E.J. McGrath (Ed.), "Social
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Office of the Secretary, Maryland Department ofEncyclopedia of Social Science,

and Psychological Factors in Stress," Holt,
New York, 1970.

4.
4.1

MAINTENANCE OF THE REACTOR IN SAFE CONDITIONOBJECTIVES AND ACTIONS

The objectives of maintaining the TMI-2 reactor in a safe condition can be summarized as: achievement of a thermally stable primary system in which the decay heat from fission products is continually being removed; maintenance of subcriticality of the reactor core; and confinement of theradioactivity within the reactor building.
Beginning in late April 1979, natural recirculation was used to remove-the decay heat from thereactor core to steam generator "A", where subsequent cooling by the secondary water transferred
the heat to the atmosphere.
Throughout most of 1980, the natural recirculation was periodic as
the decay heat level decreased.
A new forced-circulation system, the mini-decay-heat-removal
system, was proposed by Met-Ed for use in placing the reactor under a long-term cooling mode inwhich temperatures can be more effectively controlled.
This system has been installed, but is
now retained as an additional available system.
Since January 5, 1981, the reactor has beencooled by a loss-to-ambient cooling mode which has been shown to be adequate to cool the reactor
at its present decay heat rate.
This cooling mode transfers the decay heat directly to thereactor building atmosphere from the reactor system through natural heat loss.
Subcriticality of the reactor is being ensured by the maintenance of sufficient boron in solution
in the reactor primary coolant.
The one operable source range neutron detector is used to monitor subcriticality.
A small amount of control rod material is believed to have melted during theaccident.
The shutdown margin available at TMI-2 is estimated to be about 15% 6k/k.*
The potential for recriticality under various hypothetical circumstances has been examined independently by several groups.
Based on these analyses'-s it can be concluded that with 3500 ppm
of boron in the primary coolant, the reactor can be maintained in a subcritical state even in the
total absence of other control materials.
The most probable (although very unlikely) cause of
recriticality was found to be boron dilution, which would be a slow enough process that any
approach to criticality can be detected and remedied.
4.2

MONITORING OF REACTOR AND REACTOR BUILDING

The reactor and the reactor building are being monitored by instruments measuring the reactor andbuilding temperature and pressure and the water level in the sump inside the building.
As ofDecember 1980, five entries had been made into the reactor building.
Radiation level readings,
equipment inspections, and some repair work have been performed.
The chemical analysis of thereactor coolant water continues to be checked weekly, and an inventory balance on the reactor
coolant water is currently performed every four hours.
One instrument channel is still functional
for monitoring the low neutron flux level and the licensee is working on making a redundant
instrument operational.
In-core thermocouples, hot-leg and cold-leg resistance temperaturedevices, and pressure gauges are available to monitor temperature and pressure inside the reactor
coolant system.
4.3

DECAY HEAT REMOVALThe term "decay heat" refers to thermal energy generated by radioactive fission products andother in-core materials after the shutdown of a reactor.
The decay heat power of the TMI-2
reactor at the time it was first shut down a few seconds after 4:00 a.m. on March 28, 1979, was160,000 kW.
By January 31, 1980, it had decayed to 200 kW and continued to diminish to about
43 kW by February 1, 1981.

*This means that there are approximately 15% too few neutrons to sustain nuclear chain reaction
at a constant rate (see Ref. 1).

4-1

4-2
4.3.1

Decay-Heat-Removal Mechanisms Employed Since the Accident

From the onset of the accident until 8:00 p.m. March 28, 1979, when reactor coolant pump 1A wasrestarted, heat removal from the reactor core was inadequate and was effected primarily by
releases of primary water to the reactor building through the pressurizer relief valve that was
Between 8:00 p.m. on March 28, 1979, and April 27, 1979, decay heat was removed
stuck open.
through steam generator "A" by forced circulation with a reactor coolant pump (see Fig. 1.1).3
From April 27, 1979, through December 1980, decay heat was removed by natural convection circulation of the primary water through steam generator "A" in a steaming mode. Since January 5, 1981,
the reactor has been cooled by heat loss from the reactor coolant system loops and reactor vesselto the reactor building air, with the building air cooled by the building cooling system.
4.3.2

Available Decay-Heat-Removal Systems

Four of the preferred modes are
There are several methods by which decay heat can be removed.
(1) the loss to ambient cooling mode, (2) the Mini-Decay-Heat-Removal System (MDHRS), (3) theAll
long-term "B" steam generator cooling system, and (4) steaming through steam generator "A".
four methods have been approved by the NRC and are available to Met-Ed as alternative means by
In addition to the above methods, the normal decay
which decay heat removal may be effected.
heat removal system is also available if needed.
4.3.2.1

Long-Term "B" Steam Generator Cooling System

Following the accident, provisions were made for long-term decay heat removal through steam
generator "B".
Additional pumps, piping, and heat exchangers were added to the steam generator "B" system to facilitate heat removal for indefinite periods of time.
4.3.2.2

Mini-Decay-Heat-Removal

SystemThe MDHRS has been installed as an additional method of transferring decay heat from the fuel in
It also may be used during defuelingthe reactor system to the nuclear service water system.'
operations when the reactor coolant system will not be configured to maintain forced cooling.
The system includes two pumps and two heat exchangers, arranged in a manner that will permit
independent operation and thereby provide redundant decay heat removal capability. Each heatexchanger has the capability to remove the total decay heat. Two of the valves that connect thereactor coolant system to the MDHRS (see Sec. 2.1.1.2) have been opened. The remaining isolation
valves are closed and will be opened just before the system will be used. The MDHRS is operational and is available for use if needed.
4.3.2.3

Loss-to-Ambient Cooling ModeOn November 6, 1980, a test of the loss-to-ambient cooling mode for controlling temperatures inr
the reactor was initiated. The turbine bypass valve from the "A" steam generator to the conThe
denser was closed, isolating the reactor cooling system from all active cooling modes.
reactor was then cooled by heat losses from the system to the air inside the reactor building.
The test was terminated after a month on December 9, 1980; the core and reactor cooling systemtemperatures had increased by about 100 F. The technical specifications were then modified torecognize loss-to-ambient as an acceptable means for long-term cooling of the reactor core.
The loss-to-ambient cooling mode was further tested during December 1980. Operating procedures
were then prepared, and loss-to-ambient cooling was subsequently adopted as the primary means bywhich the reactor core would be cooled, with the long-term "B" steam generator and MDHRS alsoavailable as alternative cooling modes. Since January 5, 1981, loss-to-ambient cooling has beenin effect.
4.3.2.4

Steaming to Steam Generator A

Since January 5, 1981, this mode has been discontinued as the primary mode for core cooling;
however, return to this method of decay heat removal is still feasible and acceptable to the NRC.
4.3.2.5

Decay Heat Removal System

The Decay Heat Removal (DHR) System has not been used during the postaccident period, but isstill an available mode of cooling. The MDHRS was designed to take the place of the DHR System

4-3
since the high-flow-rate DHR System is not necessary at the very low decay heat levels.
Use ofthe DHR System is not expected under any likely circumstances, but is available for operation.
4.4

EFFLUENTS AND RELEASES TO THE ENVIRONMENT

4.4.1
4.4.1.1

Normal Releases
Gaseous Releases

From the time when steam generator "A" started operating in the steaming mode (April 27, 1979)
until June 28, 1980, when the reactor building atmosphere was vented, about 66 to 80 Ci of Kr-85
had been leaking out of the TMI-2 reactor building every month. 6 In the steaming mode of steam
generator operation, the turbine side of the steam generator is maintained in a partial vacuum by
the plant air ejectors.
Consequently, the pressure difference between the reactor building
atmosphere and the turbine side of the steam generator enhanced leakage of Kr-85 from the reactor
building through the packing of various steam valves to the secondary system.
The Kr-85 gas andother gases (nonradioactive) were subsequently discharged from the secondary system through theauxiliary building ventilation system to the environment.
The technical specification amendment
accompanying the NRC order to permit the controlled purging requires that the dose to maximally
exposed individuals during the purge be limited to: (a) 15 mrem skin dose, (b) 5 mrem total bodydose and (c) 20 percent of the limits in (a) and (b) above may not be exceeded over any one-hour
period.
Those technical specification requirements were met.
From mid-July 1980 through November 1980, under the authority of the Interim Release Criteria,
the reactor building was purged 13 times, sometimes in conjunction with the various entries thathave been made.
The largest single release was made August 14-15, 1980, when about 84 Ci were
purged.
The average release rate from August to December 9, 1980, was about 67 Ci per month,which is equivalent to about 15 Ci per week.
In addition to releases during periodic purging, small amounts of Kr-85 have been released fromother sources within the plant (e.g., the auxiliary building ventilation system).
The release
rates are near or at equipment detection limits and are less than one curie per month.
4.4.1.2

Routine Operational

Liquid Wastes.TMI Unit 2 is similar to other commercial nuclear power plants in that reactor support facilities, e.g., the auxiliary and fuel handling building, have special floor drain systems to collectmiscellaneous liquids generated within these facilities.
These drain systems do not collect TMI"accident water" (as defined in Section 1.6.2.1) but receive initially nonradioactive water, suchas, river water inleakage, that becomes contaminated by radioactive material that is present in
the buildings.
This type of liquid radwaste is commonly referred to as "Routine Operational
Liquid Wastes" (ROLW).
As a result of cleanup efforts, the volume and radioactivity content of
ROLW at TMI is now similar to that of a normally operating power reactor. The only difference isthat the source of the relatively low levels of radioactive contamination in the drain system isthe March 28, 1979, accident, as opposed to fuel leakage and activated corrosion products at
normally operating power plants.
Table 4.1 is a summary of the expected sources, volumes andcurie content of untreated ROLW.
To date, there has been only a limited amount of ROLW generated at TMI Unit 2.
The initial AFHB
cleanup water has been processed as accident water because of its concentration of radioactive
materials.
Between March 28, 1979, and September 1980, ROLW from the contaminated drain tanks(see Table 4.1) was transferred to the Unit 1 EPICOR I radwaste system for processing and routinedischarges.
In September 1980, the Unit 1 EPICOR I system was disconnected from Unit 2.
As aresult, all ROLW is presently being collected and stored in Unit 2.
It is expected that as cleanup operations progress, the amount of ROLW generated will approach
the amounts listed in Table 4.1.
This waste will be treated in a typical radioactive liquid
waste treatment system similar to that found in any power reactor. For planning purposes, it isassumed that an ion exchange system consisting of particulate filters followed by two demineralizer beds in series will be employed.
Based on past performance of such systems, it is expected that the system decontamination factor will be at least 1000 for isotopes other than tritium.
Tritium will not be removed by the system.
The radioactivity content of the treated water would
be less than that at a normally operating nuclear power plant and is listed in Table 4.2.
It isanticipated that this treated water will be discharged after sampling and analysis to assurecompliance with all regulatory criteria.

4-4Table 4.1.

Routine Operational Liquid Waste Sources

Source
AFHB Floor Drains

Average
Radioactivity
Concentration(pCi/mL)

Estimated
Volume(gallons per day)

10-2

100

10-4

50

10-4

100

-River Water Leakage from Pump Seals-Air Conditioning Condensation-Closed (non-radioactive)
System Leakage

Cooling Water•Demineralized Water (non-radioactive)
Flush Water
AFHB Contaminated Drain Tank
-Personnel Showers.Secondary Plant Sampling-Chemistry Laboratory Drains(low-level, e.g., glassware washing)-Instrument Repair & Calibration Shop
Service/Personnel Access Facilityb-Chemistry Laboratory Drains-Personnel

Showers*Air Conditioning Condensation•Demineralized Water (non-radioactive)
Flush WateraIncludes an estimated average tritium concentration of 1 x 10-4 PCi/mL.
The
remainder has an approximate isotopic distribution of Cs-137 (89%), Cs-134 (9%)
and Sr-90 (2%).
bProposed facility.

Table 4.2.
Release to the Susquehanna Riverof Treated Routine Operational Liquids

Amount of
Release(Ci)

Average
Concentration
Prior to Discharge(pCi/mL)

Average
Concentration
Entering River a(pCi/mL)

Licensee'sLower Limit ofDetection(pCi/mL)

H-3

3 x 10-2

9 x I0-5

3 x 10-8

3 x I0-7

Cs-1341 x 10-4

4 x I0-7

1 x 10-16

6 x 10-9

Cs-137I X 10-3

4 x 10-6

I x 10-9

6 x 10-9

Sr-90

3 x I0-s

9 x 10-8

3 x 10-11

i X 10-9

Radionuclide

aBased on dilution factor of 3400 from service water. discharge.

4-5
The operation of the ROLW treatment system will produce a relatively small amount of solid waste.
It is expected that two 100 ft. 3 spent demineralizer liners containing approximately a total of
one curie of radioactivity will be generated per year.
These should be disposed at a low-level
radioactive waste disposal site.
4.4.2

Accident Scenarios and Associated Releases

There are two broad categories of core-related accidents that could result in the release ofadditional radioactive fission products from the damaged fuel in the reactor core.
The first is
sufficient overheating or mechanical damage (fracturing) to lead to the escape of some of theradioactive fission products still held within the core.
Most of the remaining fission products
are still trapped within the fuel 'particles in the core and would require'very high temperatures(on the order of those reached during the original accident) to be released from the fuel.
However, there may be small pockets of more readily released fission products (e.g. Kr-85 in asmall gas bubble) that could be released by mechanical damage.
The second broad category of
core-related accidents is an inadvertent restart of the reactor with the associated generation of
new fission products and heat.
This latter category is referred to as a recriticality accident.
Recriticality and accidents related to overheating of the whole core are discussed in the sections that follow.
In addition,

the staff has considered the leakage of the sump water from the reactor building

through the ground and into the river.
4.4.2.1

This accident is also discussed.

Recriticality

Some neutrons, from extraneous sources and the spontaneous fission of uranium, are present in thecore even when the reactor is shut down.
These neutrons do cause some uranium atoms to fission;
however, the rate of fission when the reactor is shut down is minute.
For the reactor to restart(recriticality), the rate at which new neutrons are supplied by fission must be greater than therate at which they are removed by capture in the core or by leakage from the core and capture in
the material surrounding the core.
Thus, to keep the reactor shut down (subcritical), it isnecessary to ensure a high neutron-removal rate.
Prior to the accident the neutron-removal rate
was controlled by adjusting the concentration of boric acid in the reactor coolant and by adjusting the position of the silver-indium-cadmium control, rods (both boric acid and the control rod
material are strong neutron absorbers).
In addition, fixed alumina-boron carbide (A1 2 0 3 -B 4C)
absorber rods also were used to adjust the local neutron-removal rate.
Trace amounts of silver, indium, and cadmium have been detected in a primary coolant sample bylaboratory analysis, 5 suggesting that some of the reactor control rods have been damaged becauseof melting.
Precise information on their condition* is not available.
It has been speculated
that any control material that may have melted probably has not left the core region, but has
been redistributed in the fashion of resolidified candle wax drippings. 6 There also is a possibility that boron from the boric acid in the coolant might have precipitated in the core when
portions of it were uncovered during the accident.
Furthermore, it is likely that the fixed(A12 0 3 -B 4 C) absorber rods in the less damaged portions of the core are reasonably intact because
they have a high melting point (20401C).
Thus, while it is reasonable to expect that substantial
quantities of the original neutron absorber materials remain in the core, there may be regions of
the core with insufficient quantities of the fixed or movable absorber rods to maintain a subcritical condition without additional boric acid.
Therefore, to ensure subcriticality, theamount of boric acid in the reactor coolant has been increased to about 3850 ppm and a new lower
limit of 3500 ppm has been established.
There is agreement in numerous criticality analyses'- 5 ' 7 ' 8 that the core would remain subcritical with a boric acid concentration of 3500 ppm inany physically possible geometry even if all the fixed and movable absorber rods were removed.
To ensure that the reactor remains subcritical throughout the decontamination program, it will benecessary to maintain control of the boric acid concentration in the reactor coolant system until
the defueling is completed.
Since the solubility of boron as boric acid is 4400 ppm at 351F and
7100 ppm at 59 0 F, the possibility of boric acid precipitating out of solution because of a reduction in temperature can be ruled out altogether.
Other means (such as chemical reactions or pH
effects) by which boron concentration inside the primary system would be compromised are addressed
in Reference 3; for example, large amounts of strong acid would have to be added before significant decreases in soluble boron concentration were observed.
However, there are no foreseeable
circumstances under which theseconditions would occur.

4-6
The only concern indicated by the studies regarding recriticality in the TMI-2 situation is theintroduction of water to the reactor core with a boron level of much less than 3500 ppm. 3 ' 4 ' 8
Calculations supported by experimental data from the Westinghouse Reactor Evaluation Center haveshown that the introduction of 1000 ppm borated water in the outer region of the 2.96-percentenriched square array of the fuel assemblies could result in recriticality. 4 Original undamaged
core geometry has been assumed in those calculations.
Lack of information on the current stateof the core makes it difficult to accurately calculate the critical boron concentration.
Calculations do show, however, that the introduction of underborated water could result in the core
becoming critical.
The TMI-2 reactor was designed with the intention of using boric acid as a means to control
reactivity (neutron balance in the core).
Thus, prevention of a boron dilution accident wasincorporated in the design through limiting the sources of pure water and the rate at which it
could be added to the reactor coolant system.
Although very unlikely, if a boron dilution event
were to occur, several methods of detection are available to alert the operators so that thedilute water source could be terminated prior to core criticality.
These methods include:
excore nuclear instrumentation, periodic boron analysis of reactor coolant water, and pressure
and temperature readings of the reactor coolant system.
In addition, boric acid is normally used
as a neutron absorber in the fuel transfer canal and the spent fuel storage pool.
While thepresent boric acid levels are much higher than normal and the geometry of the core is not known,
the basic approach of using boric acid to ensure shutdown is both a sound and normal practice.
Control of the chemistry of the reactor coolant system and all sources of other water added to or
mixed with the reactor coolant during the cleanup will prevent a dilution accident.
Initiation of a boron dilution event in the current reactor system is very unlikely. The standby
pressure control system (SPCS), which is currently connected to the reactor coolant system tomaintain RCS pressure and provide makeup water, is fully borated and sampled weekly for'boron
concentration.
Boron concentration is maintained between 3000 and 4500 ppm.
SPCS tank levels
are monitored every four hours to determine the quantity of water injected to calculate a reactor
coolant system leak rate (which is currently less than 0.1 gpm at about 100 psig reactor coolant
system pressure).
Other pumping systems connected to the RCS, such as makeup or decay heatremoval systems, were previously borated or will be borated prior to startup so that system
activation would not add deborated water to the RCS.
A closed valved connection from the demineralized water system to the makeup pump suction does exist but is located in a high-radiation
area to which access is administratively controlled.
In addition, the power supply breakers to
the makeup pumps are danger-tagged open to prevent inadvertent pump operation.
Even if a pressurized demineralized water source were introduced to the reactor coolant system in its current
configuration (i.e., filled solid with water), an increase in reactor coolant system pressure
indication should alert the operator-to terminate the source.
The present core inventory of radioactive fission products is very low.
In fact, it is manyorders of magnitude less than the inventory of the reactor during normal operation, which is used
as a basis in hypothetical accident studies and for which public safety is assured. Most of thefission products produced in a recriticality accident would be extremely short-lived, i.e., theywould decay away very rapidly.
For example, the total amount of curies released to the atmosphere in the March 28, 1979, accident was greater than the total fission product inventory thatwould be in the current core ten hours after a hypothetical severe recriticality transient. 8 The
reactor building is specifically designed to contain fission product inventories that are ordersof magnitude greater than the current core inventory at containment pressures far greater thanthose which would exist during defueling.
Nevertheless, protection of working personnel dictatesthat all safeguards be taken to preclude recriticality.
4.4.2.2

Core Cooling Failures

Since the accident, electric power for equipment used to remove decay heat from the core has been
provided by circuits from offsite transmission networks and onsite power supplies.
A 13.2-kV
circuit from the Middletown junction substation and two redundant diesel generators were installed
to increase reliability of power supplied to the TMI-2 system.
In March 1980, Met-Ed proposed to
the NRC to reconfigure the power system such that the two diesel generators and the 13.2-kVtransmission line would be replaced by the existing 230-kV grid system.
This system has now been
approved by NRC and has been put into effect. Three separate, redundant, and independent sets of
combustion turbines located in three geographical locations would be used to supply the grid.
Consequently, the probability of complete loss of power needed to sustain decay heat removal
operations is extremely small.

4-7
An accident scenario involving a leak in the reactor vessel which leads to uncovering and overheating of the core is discussed in Section 10.4. At the present decay heat levels, ample time
is available to take corrective action if such a condition develops.
In addition, with the low
heat-generation rate and low system pressure, it would not be difficult to provide sufficientwater to prevent overheating.
However, if overheating is assumed to occur and all the remaining
Kr-85 and cesium were released to the reactor building, the total activity from this hypotheticalreleasewould be about the same as the activity inside the reactor building prior to purging.
Given the long time available to take corrective actions (many hours to several days), the reactor
building, if occupied at the time, could be evacuated and isolated before any significant releasefrom the core had occurred.
4.4.2.3

Leakage of Reactor Building Sump Water

The largest amount of contaminated water currently on the site is the 700,000 gallons of water inthe bottom of the reactor building.
This water contains an estimated 500,000 Ci of radionuclides.
It is postulated that if this water should leak through the steel-lined concrete base of thereactor building, it would ultimately reach the Susquehanna River.
As discussed below, such an
accidental release is considered highly unlikely.
A detailed discussion is also provided inAppendix V.
The reactor building is constructed to prevent water in the sump from leaking to the environment.
The reactor building was designed and fabricated to withstand an internal pressure of 60 psig andto withstand major earthquakes.
The foundation mat for the reactor building is 11½ ft thick andrests on bedrock.
The reactor building has a carbon steel liner, 3/8 inch thick on the sides,
1/2-inch-thick dome, and 1/4-inch-thick base.
A concrete slab 2 ft thick was poured above thebase liner plate. The exposed face of the liner is coated with a prime and finish coat of phenoline 368.
An additional leakage barrier is provided by a 40-mil-thick PVC wasteproofing membrane
which is installed over the outer surfaces of the foundation mat.
In the unlikely event that water from the reactor building sump should leak through these barriers, several methods for early detection and control of this leakage are available.
One
indication of leakage would be a decrease in the sump water level, which is monitored daily.
The
detection of radioactivity in the tendon gallery and its sump would also be an indication of aleak.
The tendon gallery is located outside the steel liner but in the perimeter of the foundation base mat.
Detection of radioactivity in the site monitoring wells (see Sec. 11) would also
indicate leakage from the reactor building sump.
A final indication of leakage would be throughthe detection of radioactivity in the Susquehanna River from the licensees' environmental monitoring program and at the water intakes from the City of Lancaster through its radiologicalmonitoring program.
Leakage from the reactor building sump could be controlled by any of several measures.
Upon
detecting a leak from the reactor building sump, the remaining water in the sump could be pumped
into various available storage tanks in accordance with the licensee's existing contingency planfor this action.
Another method for controlling such leakage would be to pump the contaminatedwater out of the monitoring wells which are located around the reactor building.
A furthermethod for controlling the release of activity from this postulated leakage would be the installation of a-grout curtain between the reactor building and the Susquehanna River.
For the purposes of analyzing the transport of Sr-90 and Cs-137 from a reactor building sumpleak, it is conservatively assumed that the water is released into the ground over an area thesize of the entire base of the reactor building.
Such a postulated release could drain only that
portion of the contaminated water that lies above the water table.
To simplify the calculation,
this volume (470,000 gallons) is assumed to enter the groundwater instantaneously.
For tritium,
whose analysis depends on a somewhat different set of phenomena, the entire 470,000 gallons wasconservatively assumed to be released instantaneously from a point at the center of the building.*
Water containing tritium would begin entering the river after a minimum of 350 days and continue
for about 130 days.

*The rapid rates of release from the containment are, of course, highly unrealistic but werechosen for computational expediency.
If no consideration is given to interdiction, theresulting concentrations would be fairly insensitive to the rate of release to the ground.

4-8
After the initial holdup in the ground, the wave front of the Sr-90 and Cs-137 would reach theriver in 23 years and 284 years and would continue to enter the river over periods of about 8.5
and 140 years, respectively.
By the time the Sr-90 reached the river, radioactive decay would
When the Cs-137 reaches the river, its
have reduced the total available activity to 2660 Ci.
total activity will have decayed to 416 Ci.
The maximum annual average concentration during agiven year would be close to the peak concentrations for these two nuclides.
The release of Sr-90 and Cs-137 to the Susquehanna River would continue over a long period if
Since doses are usually calculated over a period of at least one year, theleft unchecked.
logical choice of a stream flow would be the reciprocal mean flow of 12,600 cubic feet per second(the derivation of this value is discussed in Section 3.4.1).
All downstream drinking water
users on the Susquehanna River are located far enough downstream that total mixing of the effluTravel time to downstream users would be negligible
ent across the channel would be expected.
The peak radionuclide concentrations in thecompared to that of the groundwater pathway.
Susquehanna River, based on the annual average flow, would be 4.05 x 10-8 pCi/mL for Sr-90,
5.1 x 10-1o pCi/mL for Cs-137, and 5.2 x 10-7 pCi/mL for tritium.
Furthermore, the peaks of
Maximum permissible
these three radionuclides would occur at different times because of sorption.
concentrations (MPC) for unrestricted drinking water from 10 CFR Part 20 are 3 x 10-7 PCi/mL for
Sr-90, 2 x 10-5 pCi/mL for Cs-137, and 3 x 10-3 pCi/mL for tritium.
The calculated river concentrations are thus orders of magnitude below MPC.
4.5

ENVIRONMENTAL IMPACTS

4.5.1

Occupational Doses

During the decontamination of the reactor building and primary cooling system, equipment vital tomonitoring and control of the reactor may be serviced and repaired.
Tasks such as the repair of
the malfunctioning low neutron flux level instrument channel and servicing of the building fancoolers will result in substantial doses to workers.
The details of these tasks and their associated occupational doses are discussed in later sections.
Only those tasks that are directly
related to monitoring the condition of the reactor and its attendant systems will be addressed inthis section.
The operations associated with the maintenance of the reactor in safe condition have included
monitoring the primary system for boron concentration, water chemistry, temperature, and pressure; monitoring of the neutron level in the core; and monitoring of the temperature and pressurein the reactor building.
The temperature, pressure, and neutron-level readings can be made fromthe reactor control room.
To measure the boron level and water chemistry, weekly samples of the
primary coolant have been taken and analyzed.
The radiation levels in the reactor control room are essentially at background levels, and personnel in the control room would not be expected to incur an occupational dose.
In sampling andanalyzing the primary coolant, the occupational dose has been about 20 to 30 mrem per sample.
For purposes of calculating occupational dose, it is assumed that only one person carries out the
sampling and analyzing of the primary system water and that the sampling frequency is once per
week. Using the higher dose of 30 mrem per sample, the dose in 13 weeks (one quarter) would be
390 mrem.
The maintenance activities would be needed until the primary system is defueled.
This is estimated to take three to five years.
Assuming the longer time of five years, an accumulative
occupational dose of about 8 person-rem can be expected. 'The added probability that the individual worker would die of cancer from this dose would be 1 in 950.
The probability of an additional genetic effect in the offspring of the worker would be 1 in 480.
The operation of a liquid radwaste treatment system for routine operational liquid wastes (ROLW)
will result in some slight increase in occupational doses.
Based on past experience with similarsystems at both TMI and other locations, it is estimated that this dose will be significantly
less.than one (1) man-rem per. year.
4.5.2

Offsite Doses

The dose estimates presented here for maintaining the reactor in a safe condition are based on
Kr-85 source terms described in Section 4.4.1. On the basis of historical data described in that
section, releases of Kr-85 are assumed to be made at the level of 67 Ci/month, which is equivalent

4-9The estimate of total-body dose to the maxito an annual level of approximately 800 Ci/year.
mally exposed individual offsite due to release of this amount of Kr-85 is expected to be less
The 50-mile total-body population' dose for this realease is estimated to bethan 10-3 mrem.
These individuals and population doses are based on the external exposure
3 x 10-2 person-rems.
pathway.
Lung doses are expected to be higher than total-body doses by approximately a factor of3, and skin doses are expected to be higher by a factor of about 80 over total-body doses.
The dose estimates from the release of treated routine operational liquid wastes are based. upon
The calculational models used to make these
the source terms discussed in Section 4.4.1.2.
estimates and the interpretation of their results are described in Appendix W. The dose estimates to the maximally exposed individual are 0.0015 mrem per year total body dose and 0.002 mrem
The downstream total body population dose is estimated to be lessper year bone and liver dose.
than 0.02 man-rem per year.
4.5.3

Postulated Accident Effects

However, if a recritNo accident having immediate offsite consequences has been postulated.
This
icality accident were to occur, cleanup of the resulting contamination would be necessary.
cleanup effort would probably cause effects similar, although smaller, to those estimated in theremainder of this PEIS.
4.5.4

Psychological-Socioeconomic

Effects

Maintenance of the reactor in a safe shutdown condition will continue until the primary system is
The staff estimates the period of this maintenance to be from three to five years.
defueled.
For the majority of people in the community, maintenance of the reactor in a safe condition will
be understood to be an essential procedure in the process of decontaminating Unit 2, and will notSome members of the community may believe that the possibilexacerbate existing anxiety levels.
For others,
ity of accidents involving offsite releases will increase with any delay of cleanup.
the long duration of this required maintenance will lead to the conclusion that maintenance will
actually be long-term waste disposal on the site. In either case, the length of the maintenanceperiod is expected to increase the probability of chronic anxiety. The staff is unable to ascribe
any other resulting socioeconomic impacts to maintenance procedures.
The release of treated routine operational liquids from TMI Unit 2 may result in increased anxiety
The impact may be insignificant if theamong members of the public who utilize downstream water.
public understands that the treatment and controlled discharge of these wastes is routine at a
The impacts may be greater ifnormally operating nuclear power reactor, e.g., Peach Bottom.
large segments of the public erroneously believes that the discharge is a release of accident
water.
The psychological-socioeconomic effects of release of processed accident water are
discussed in Section 7.2.5.5.
(1) overheating
Reactor core accidents hypothesized by the staff fall into two broad categories:
or mechanical damage leading to an escape of radioactive fission products presently in the core,and (2) the inadvertent restart of the reactor with the associated generation of new fission
products and heat.
In both instances, the physical effects would be limited to the reactor
Accidents during the maintenancebuilding and would not impact offsite areas or population.
period, although not directly affecting the public, would nonetheless confirm for a segment of
the community the inability of Met-Ed and the NRC to manage the technical process of cleanup.
This perception had previously been identified in postaccident studies as being an importantcause of stress, therefore, any accident would likely lead to a measurable increase in community
stress.

References--Section 4
Metropolitan Edison Co.,

1.

Letter from G. Kulynych, Babcock & Wilcox Co., to R. Harding,"Basis for Tech Spec Boron Limits," May 1, 1979.

2.

Memorandum from C. Marotta, Nuclear Materials Safety and Safeguards, U.S. Nuclear Regulatory
Commission, to K. Kniel, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory
Commission, "Recriticality Potential of TMI-2 Core," May 14, 1979.

3.-

Memorandum from D. Cokinos to D.J. Diamond,
National Laboratory, May 18, 1979.

"Recriticality Calculations for TMI,"

Brookhaven

4-10
4.

E.W. Barr et al.,"TMI-2

Post-Accident Criticality

Analysis,"

GeneralPublic Utilities

Service Company, TDR-049, August 31, 1979.
5.

R.M. Westfall et al., "Criticality Analyses of Disrupted Core Models of Three Mile Island
Unit 2," Oak Ridge National Laboratory, ORNL/CSD/TM-106, (Prepared for the President's
December 1979.
Commission on the Accident at Three Mile Island.)

6.

N.M. Haller et al., "Evaluation of the Cleanup Activities at Three Mile Island," U.S. Nuclear
Regulatory Commission, February 28, 1980.

7.

Memorandum of D.J. Malloy to J.B. Heineman, "Some Theoretical Limiting Constraints for TMI-2
Recriticality," Argonne National Laboratory, April 4, 1980.

8.

C.J. Mueller, Argonne National Laboratory, Memorandum to Distribution, "Review of Draft - An
Evaluation of the Potential for and Consequences of Recriticality During Cleanup and Defueling at TMI-2," November 18, 1980.

5.

BUILDING AND EQUIPMENT DECONTAMINATION

The interiors of the auxiliary and fuel handling buildings (AFHB) and of the reactor building (RB),as well as the equipment in those buildings, were severely contaminated by radioactive
material during and shortly after the accident on March 28, 1979.
The auxiliary building contains tanks, pumps., piping, and'other equipment used to process and store water for the reactor
and primary cooling system and to treat radioactive wastes.
The fuel handling building, which isadjacent to the auxiliary building and separated from it by a common wall, contains fuel handlingand storage equipment.
The general layout of these buildings is shown in Figures 0.1 through 0.3
of Appendix 0. The reactor building contains the reactor pressure vessel (RPV), which holds thefuel elements, and the primary system (steam generators, reactor cooling pumps, pressurizer tank,
and associated equipment).*
The building is a cylindrical structure about 170 ft high and 140 ft
in diameter with reinforced concrete walls 4 ft thick.
The radioactive contamination in the AFHB and reactor building must be removed and disposed of ina safe manner in order to permit defueling and maintenance work that may be needed to ensure
continued safe shutdown.
In this section, the procedures for decontaminating the buildings are
outlined, and the effluents and releases to the environment, as well as the environmental impacts
that might occur as a result of these decontamination activities, are discussed.
The decontamination operations may be divided into two phases.
The objective of the initial
phase is to reduce radiation levels to an extent that the reactor fuel can be removed.
This
initial phase is necessary regardless of whether the facilities are decommissioned or refurbishedfor reuse, and the manner in which it must be carried out is largely independent of the final
disposition of the facilities.
In the final phase, decontamination operations will continue
until radiation levels are as low as reasonably achievable (ALARA).
The manner in which thefinal phase is implemented will depend on whether the facilities are decommissioned or refurbished (see Sec. 2.1).
The target radiation field and surface contamination levels are given in Sections 5.1 and 5.2 for
the AFHB and reactor building, respectively.
Decontamination of the AFHB started shortly after the accident and was about two-thirds complete
as of September 1, 1980.
Decontamination activities since that time have been limited due to thelicensee's financial situation (see Sec. 1.3).
Decontamination of the reactor building had not
yet started.
Precautions must be taken to- protect the workers from excessive exposure to radiation and to
avoid releases of contaminated material that might adversely affect the environment or endanger
public health and safety.
The special clothing and breathing apparatus used to protect workers
from radioactive contamination are shown in Figures 5.1 and 5.2, which are scenes from actual
decontamination operations in the auxiliary building.
The amount of radioactive contamination
decreases with time even if nothing is done.
The rate depends on the half-life of the radionuclide.
For two of the commonly occurring radionuclides, Cs-137 and Sr-90, this decrease issmall, less than 3 percent per year.
It is somewhat greater for Cs-134, about 29 percent per
year, and quite large for Sr-89, which decreases 0.1 percent of its original amount in a periodof a year.
The sequence of decontamination operations that must be carried out may be summarized as follows.
Debris and heavy deposits, such as sludge, must be removed; then the exposed surfaces must be
cleaned.
Loose dirt can be removed by vacuuming or by hosing down the surfaces.
Dirt thatadheres too firmly to be removed by these means can be removed by high-pressure water jets or by
scrubbing.
Wiping of surfaces with specially designed wipes (similar to household dust cloths)*Removal of the fuel from the reactor pressure vessel (defueling)inside of the primary system are treated in Section 6.

5-1

and decontamination of the5-2

Figure 5.1.

Wet Mopping:of TMI-2 Auxiliary Building. Wet mopping is
one technique used to decontaminate portions of theWorkers are shown wearing special
Auxiliary Building.
clothing and breathing apparatus for protection from
(Official TMI Photo)radioactive contamination.

5-3

Figure 5.2.

Auxiliary Building Equipment Cleanup.
In this photo, aworker follows typical procedures for decontaminating
equipment and piping systems in the TMI-2 Auxiliary
Building. Protective clothing and breathing apparatus
are worn as safety precautions.
(Official TMI Photo)

5-4also is a commonly used procedure.
Finally, for contamination that has penetrated into thesurface, like a stubborn stain, methods that remove some of the surface are necessary.
5.1

AUXILIARY AND FUEL HANDLING BUILDINGSThe overall objectives of the AFHB decontamination effort are to permit access without restriction because of surface or airborne contamination, reduce radiation exposure from gamma sources
to ALARA levels, and prevent recontamination from other activities or system leaks. The following guidelines are applied to determine whether these objectives have been satisfied:
Smearable P + y surface contamination is
Airborne contamination
andGeneralis2less than 1000 dpm/100 cm ,'2less than the 10 CFR Part 20 limits for restricted areas,

radiation levels are at plant design values--generally 0.4 mR/hr.

These objectives are largely determined by the need to use the equipment in the AFHB for defueling operations and are, therefore, applicable regardless of whether the facilities are decommissioned or refurbished.
It is expected that they will be attained before defueling operations
begin in the reactor building.
In addition to these general objectives, there also is a need to clean certain systems and items
of equipment that may be needed in the efforts to decontaminatethe reactor building and theprimary coolant system.
These AFHB systems and equipment include:
Reactor bleed holdup tanks and associated equipment
Reactor coolant pump water cooling and seal water systems
Miscellaneous waste holdup tanks and associated equipment
Coolant evaporator system
Degasification system
Fuel transfer ports and machines
Fuel storage pool and handling cranes
Canal cooling and purification systems
Other handling equipment, such as cranes
Make and purification demineralizers and filter.
Decontamination of the AFHB was started by May 1, 1979, and is expected to be far enough along by
the time that decontamination of the reactor building begins to permit use of these systems (seeFig. 1.4).
Decontamination work is expected to continue at a reduced level of effort through thesummer of 1982.
5.1.1

Status and Specific Considerations

AFHB decontamination operations considered in this section are (1) cleaning of radioactive material from the interior surfaces of the building and from the surfaces of equipment, and(2) removal of radioactive sludge deposited in tanks and pipes and in the building sump.
5.1.1.1

Building and Equipment Surfaces

Surveys of the general access areas (corridors and normally nonrestricted areas) shortly after
the accident showed radiation levels of 150 to 500 mR/hr in the fuel handling building and 50 to5000 mR/hr in the auxiliary building.
At certain locations in the general areas (referred to as"hot spots") and in individual areas (cubicles) containing contaminated filters, demineralizers,
tanks, and pumps the radiation levels were much higher--hot spots up to 125 R/hr were measured in
the access area of the auxiliary building and levels
exceeded 1000 R/hr in some cubicles, such as
3
the reactor coolant bleed holdup tank cubicles.
When cleanup operations began, about 510,000 ft 2 of surface in the AFHB required decontamination.
As of September 1, 1980, about 340,000 ft 2 had been decontaminated, with an expenditure of
500,000 person-hours of effort by personnel with assignments that required entry into contaminated areas in the AFHB.
The radiation dose received by these workers during this interval was
about 250 person-rem.
These totals include support personnel (e.g., radiation technicians andmaintenance and construction workers) as well as decontamination workers.

5-5
The amount of radioactive contamination removed from the AFHB prior to September 1980, excluding
only the radionuclides in the accident water, was estimated by the licensee to be 128 Ci.
Theradioactivity in some of the more highly contaminated cubicles is not known; on the basis of thelimited data available, the staff estimates that the radioactive contamination remaining in the
AFHB on September 1, 1980, consisted of less than 60 Ci in the form of plateout and deposits onbuilding and equipment surfaces and less than 9000 Ci in the form of sludge in the auxiliarybuilding sump and tanks.
Decontamination of building and equipment surfaces initially was started in areas with low radiation levels.
The decontamination team left many high-radiation areas, such as the reactor coolant. bleed tank cubicles, until later because it is first necessary to remove the radioactive
liquids from the tanks and piping.
After these tanks and pipes are flushed and the filterschanged, radiation levels are much lower and decontamination of surfaces can proceed with muchless radiation exposure to personnel.
The general access areas in the AFHB had been decontaminated by April 1980; however, some ofthose areas have to be recleaned periodically because of recontamination.
For example, installation of new equipment needed for various tasks has resulted in some contaminated material being
tracked into previously decontaminated areas.
In addition, leakage of barrels containing contaminated industrial detergent at one time was causing similar problems, but this has been corrected.
The movement of contamination has required the routine monitoring of previously decontaminated areas and repeated decontamination of some areas.
It is expected that some additional
decontamination will be needed periodically in the general access areas (corridors, stairwells,etc.) because of airborne dispersion and tracking by workers from areas still undergoing decontamination.
Another factor affecting decontamination efforts is the increased potential fordevelopment of leaks from systems that have not received routine maintenance because of contamination.
This problem will become more acute with time and is continuing to receive attention.
Decontamination of the cubicles in the AFHB began in February 1980 and is expected to continue
through the spring of 1982; after then a reduced level of effort will be needed to maintain thedecontaminated levels.
The cubicle areas tend to be more difficult to decontaminate because of
special shielding requirements and higher contamination levels.
As of September 1, 1980, many of
the tanks, filters, and much of the piping in the AFHB still contained radioactively contaminatedliquids.
As indicated in Appendix 0, of the 57 cubicles requiring decontamination, 6 had been
completely decontaminated (except for fluid transfer and filter changes in 2 of the 6), 32 hadbeen partially decontaminated, 15 required only light decontamination, and 6 had received no
decontamination.
The makeup and purification demineralizer cubicles are likely to be the most
severely contaminated in the AFHB.
The contamination consists primarily of radionuclides thatwere deposited in the inline filters and demineralizer resins during and immediately following
the accident.
Based upon gamma surveys by the licensee, each of the two demineralizers could
contain about 2000 Ci. Transfer of fluids, changing of filters, removing of ion exchange resins,
and/or flushing of lines still were needed in 22 of the 57 cubicles.
The completion date for decontamination of the AFHB will depend on the number of persons assigned
to the effort. This number was severely reduced during September 1980, to the point that very
littlework beyond essential maintenance was being done.
A reliable estimate for the completion
date cannot be given until steps are taken to resolve financial difficulties that are (as ofJanuary 1981) limiting expenditures for decontamination work.
An estimate of the work effort
that will be required to complete the AFHB decontamination is given in Section 5.1.3.2.
5.1.1.2

SludgeSome of the contamination in the auxiliary building is in the form of sludge, which requires
special attention because of its bulk and high radioactivity content.
The sludge is mostly
cement residue (from an unfinished part of the floor and from floor drains and drain piping) withan admixture of ion-exchange resin from small amounts of resin carried over during demineralized
water system flushing operations prior to the accident.
During the accident, which caused flooding to a depth of about 6 inches on the lower level of the building, a large portion of thesludge was washed into the 9000-gallon sump.
From there the sludge was pumped into the sump tankand the miscellaneous waste holdup tank (see Figs. 0.1 through 0.3, Appendix 0, for the locationsof these tanks).
Smaller amounts of sludge may be present in the other tanks in the auxiliarybuilding, but this has not been confirmed by direct examination.
Pumps, valves, and lines also
may contain small amounts of sludge.
The estimated total volume of wet sludge in the tanks and5-6
the sump in the auxiliary building is about 150 ft 3 .
For sludge with a density of 1.6 g/mL(100 lb/ft 3 ), the total weight of wet sludge to be removed is on the order of 15,000 pounds.*
The total radioactivity in the sludge is estimated to be about 9000 Ci.
The sump is 14 ft wide by 24 ft long. It is 3 ft deep at one end and slopes to 4 ft deep at anoutlet at the other end.
The sludge is about 2 inches deep near the outlet pipe.
At the opposite (shallow) end it is 4 to 6 inches deep.
The total volume of sludge in the sump is estimatedto be about 90 ft 3 . Debris also are present in the sump.
A 6-inch hose, a 2.5-inch firehose,tie rods and a temporary sump pump have been noted. 4 Radiation levels from 3 R/hr to 5 R/hr were
measured at different locations in the sump cubicle prior to January 1980.
Radiation levels inthe sump cubicle can be controlled to some extent by pumping out the contaminated water in thesump and replacing it with processed water from which most of the contamination has been removed.
Cleanup of the sump cubicle had not been started as of January 12, 1981.
Estimates by the licensee indicate that the miscellaneous waste holdup tank contains 10 to 15 ft 3of sludge.
This tank is being used as a feed tank for the EPICOR II water decontamination
operations (see Sec. 7).
During the period August 1979 to April 1980, the radiation level near
the tank rose from about 5 R/hr to about 30 R/hr because of the accumulation of sludge as contaminated liquids were transferred in and out of the tank. The continuing need to use the tankmay delay its desludging and decontamination.
The volume of sludge in the sump tank also is quite uncertain because the interior is not visible; it is thought by the licensee to be on the order of 5 to 10 ft 3 . The radiation level at
one end of the tank is about 20 R/hr.
The total sludge in the remaining tanks is estimated by
the licensee to be about 35 ft 3 or less.
These tanks include the three reactor coolant bleed
tanks (which constitute about three-quarters of the total liquid storage capacity in the auxiliary building and probably contain about 10 ft 3 of sludge per tank 4 ), the spent resin tank, theconcentrated waste tank, and other tanks identified in Figures 0.1 through 0.3 of Appendix 0.
Desludging of a tank cannot begin until most of the liquid within it has been processed or
removed.
Liquid from reactor coolant bleed tanks A, B, and C has been processed, and flushing
the tanks with processed water has decreased the radiation levels and removed some of the sludge.
5.1.2

Methods Used and Alternative Methods Considered

5.1.2.1

Methods Used for Building and Equipment Surface Decomtamination

Appropriate combinations of well-known methods have been used in the decontamination of building
and equipment surfaces in the AFHB.
The following methods have seen at least limited use, withthe first six used predominantly:
Removalof allnonessential

items,

such as wood, tools,

hoses,

cords, and loose equipment.

Dry vacuuming of dry floors and equipment (piping, valves, cable trays,
efficiency particulate air (HEPA) filter on the vacuum exhaust.

etc.)with a highLow-pressure hot water misting and washing.
Using high-pressure water jets on floors,
Wet vacuuming of industrial

tanks, piping,and valves.

detergent after hand scrubbing.

Manual wiping with disposable towels or oil-impregnated wipes.
Removing strippable coatings on floors, walls,
Electrochemical

portable shields, and other surfaces.

decontamination of tools and small equipment.

Freon cleaning and Freon ultrasonic cleaning of electrical equipment and tools.
Concrete removal by means of pneumatic impact devices.

*The volume of dewatered sludge is smaller than the volume of wet sludge by a factor of about
4 to 5; the weight is smaller by a factor of about 2.

5-7
All of these methods, except perhaps the water jet, require the work crew to be relatively close
to the contamination.
In areas of high concentrations of airborne radioactive particulates,
personnel must wear respiratory protectors, such as air packs or filter respirators.
Workers
also must wear special clothing for protection and to control the spread of contamination.
Various combinations of decontamination methods are selected by decontamination personnel tominimize exposure and maximize effectiveness.
The decontamination personnel rely on experienceand testing to determine the best r'ethod to use for each particular task.
5.1.2.2

Methods Used and Considered for Sludge Removal

The procedure now being used for removing sludge from tanks and piping in the AFHB starts withreplacement of the inlet and outlet filters, as required.
The system then is flushed several
times, and the loose sludge remaining in a tank is collected with a portable recirculating vacuum
filter system (RVFS) with disposable filter cartridges.
Sludge caked to the walls of tanks canbe dislodged with a high-pressure water jet and then collected in the RVFS.
Standpipes (drain
pipes that extend above the bottom of the tank) may complicate the desludging effort.
The alternatives considered by the staff for removing sludge from the AFHB sump and tanks may be
divided into the following categories:-

Dissolve the sludge by chemical means so that it can be removed as liquid waste,-

Remove the sludge by mechanicalmeans (scooping and scraping),-

Resuspend the sludge so that it

can be removed as a slurry.

or

Dissolution of the sludge, which consists largely of calcium sulfate and various silicates, would
require strong acids, such as concentrated hydrochloric (muriatic) acid, that would increase thehazards for decontamination workers and damage the underlying surfaces and the drain hardware
without providing any significant advantages.
The process is not efficient and usually leaves anintractable residue; there are no practical means for dissolving cement, the primary constituentof the sludge, on a large scale.
Dissolving the sludge would greatly complicate subsequent
problems of processing the water to remove dissolved radioactive contamination for disposal assolid waste.
This approach was not, therefore, considered further.
If mechanical removal were done by hand using shovels and scrapers, the radiation exposure to theworkers would be higher than for the other alternatives.
Mechanical devices for scooping andscraping that would permit semiremote or remote operations in restricted spaces are not, insofar
as the staff is aware, available as standard equipment and thus would have to be designed, constructed, and tested.
This design and development effort could lead to delays of unpredictable
duration.
Other methods, such as resuspension (including hose washing), are simpler and more
effective and require no development work; hence, mechanical removal methods were not consideredfurther.
The following alternative methods for removal of the sludge by resuspension have been considered:
Resuspension and removal of the sludge by means of a portable RVFS.
Resuspension by flushing and/or backflushing and entrapment of the sludge on the inline
filters using existing piping and pumps.
The second resuspension alternative is very effective for tanks containing small amounts ofsludge.
The worker radiation dose that would be incurred in changing the inline filters on
which the sludge was collected would be essentially the same as for the RVFS system.
However,
the agitation would be less than for the first alternative; hence, sludge removal would be lesseffective.
Additionally, there would be a greater likelihood of spreading the sludge to otherparts of the system.
A high-pressure water jet also could be used to loosen and remove cakedsludge from surfaces and, if necessary, to assist in resuspending the sludge when agitation fromrecirculation or from the vacuum system is insufficient for this purpose.
The water remaining after removal of the sludge from the slurry could either be returned to thetanks and sump or stored in some other tank until it could be processed to remove dissolved
radionuclides.

5-8
5.1.3
5.1.3.1

Details of Feasible Methods and Facilities
Details of Methods Used

Application of the decontamination procedures listed in Section 5.1.2.1 proceeds in a logicalsequence from removal of tools, equipment, and other loose items and debris to methods for removing loose contamination that does not adhere strongly to surfaces (for which spraying and washingis effective), to methods for removing fixed contamination that adheres to the surface sostrongly that scrubbing or chemical solvents are needed.
Some of the methods used for decontaminating surfaces in the AFHB are described below. Further discussion of decontamination methods
may be found in Section 5.2.3.1.
Removal of Nonessential Items
The decontamination crews have found tools, loose equipment, barrels, boxes, staging, cables,
hoses, wood pallets, and other miscellaneous items in many areas.
Rather than decontaminate
these items in place, the crews have moved them to staging areas for cleaning and eventual storage or disposal.
Some of the items have been disposed of as low-specific-activity waste.
Mostof the items can be handled by cleanup personnel.
Some items with fixed contamination have been
stored for future use in contaminated areas.
Large pieces of equipment may be decontaminated inplace.
Worker exposure comes from the contaminated items when they must be moved, as well asfrom the contaminated environment.
Dry Vacuuming of Dry Floors and Equipment
In areas where dry dust has accumulated, dry vacuuming can be effective.
Dry vacuuming does notwork well on crusted deposits; therefore, it is used primarily in areas where dust has not beenwetted or crusted.
The vacuuming involves the use of a specially equipped machine with a HEPA
filter in the exhaust stream.
Radioactive particles are retained in the filters. Worker exposure from airborne activity may be increased by the vacuuming activity even though most of theparticles are picked up by suction.
Dry vacuuming has been used in the AFHB only for removing
the normal accumulation of dust from equipment that was above the flood line and did not,.therefore, become contaminated during the accident.
Low Pressure Water Washing
Low pressure hot water misting and washing is used to dissolve chemicals on the walls, floors andcomponents.
This method is effective in dissolving dried boron deposits and controlling airborne.
contamination.
It is used prior to the high pressure water jet spray.
Use of High-Pressure Water Jet
A high-pressure, low-flow-rate, water-jet spray system can be used to remove surface contaminants.
The system is effective and fast in removing contaminants and reducing personnel exposure
time.
Large areas can be cleaned quickly and thoroughly.
Two water-jet units are available onsite at TMI-2 for decontaminating the AFHB.
One operates at
a moderate pressure of 1200 psi and the other at a high pressure of 10,000 psi.
The water flowis relatively low, about 7 gpm for the 1200 psi unit. As of January 2, 1981, only the moderatepressure jet had been used for decontaminating the AFHB.
Use of the water jet was limited prior to March 1980 because of restricted water usage.
It isnow being used extensively with water processed by EPICOR II.
Use of this processed water,
rather than additional water pumped into the facility, has reduced the water inventory buildup
and permits extensive use of the water jet. Tests of the atmospheric contamination caused by useof processed water have been conducted with the water jet in closed quarters.
Under the worst
conditions (i.e., with the atmosphere saturated with water vapor), the concentration of tritium
2in the air would be below the level specified in 10 CFR Part 20 as the safe limit for workers.
Wet Vacuuming
Wet vacuuming has been the primary method for decontaminating areas where contaminants adhere
tightly to surfaces.
The method involves scrubbing with water and industrial detergents and thenwet vacuuming the resulting solution.
The wash solution is stored in barrels until it can besolidified for disposal.
There have been limitations on the types of cleaning compounds used

5-9
since their effect on the solidification process has not been fully evaluated.
To date, only one
commercial detergent has been used.
The scrubbing is a slow and tedious process that brings
workers into close contact with contaminated wash solution.
Manual Wiping
Manual wiping, a worker-intensive technique, may be used to remove dust and accumulated contamination that cannot be vacuumed.
Disposable towels or oil-impregnated wipes ordinarily are used.
This technique requires workers to be close to contaminated areas; thus the exposure may behigher than for other techniques, such as vacuuming, for a given level of contamination.
Manualwiping is normally used only after radiation levels have been reduced by gross decontamination
using other methods.
This method is also used for contolling contamination in areas previously
contaminated.
Removing Strippable Coatings
This method involves the application and subsequent removal of a strippable coating.
As thecoating is removed, it takes with it the surface contamination.
Strippable coatings are usefulon portable shielding, making it easy to decontaminate.
It involves close worker proximity tocontamination, but there is less likelihood of the contamination being spread to other surfaces,such as clothing or gloves, than for manual wiping.
Strippable coatings are commonly applied todecontaminated areas to facilitate subsequent decontamination if recontamination should occur.
Electrochemical Decontamination
Electrochemical decontamination is an electropolishing procedure used on metal objects to remove
a thin layer of the exterior surface and attached contamination.
The method employs a tankcontaining an acid solution and a low-voltage, high-current source.
At TMI this method can beused only for small objects because the electrochemical decontamination tank is about 3 ft by
3 ft by 2 ft deep.
Small tools and parts can be cleaned in a very short time.
The phosphoric acid in the decontamination tanks is recirculated through a filter that accumulates much of the contaminated solids removed from the surface.
The major limitations of thismethod are that the objects must be metal, they must be small, and they must be removable.
The electrochemical decontamination apparatus is set up in a facility removed from the AFHB.
Vapors from the facility are circulated through HEPA filters to limit radioactive releases tonegligible amounts.
Freon Cleaning and Freon Ultrasonic CleaningThis decontamination method involves the use of Freon, either by spray, brush, or ultrasonicbath, to clean electrical tools and small, intricate parts.
The ultrasonic bath using Freon as a
fluid will remove most contaminants, but only small parts can be cleaned in this manner.
Largerelectrical components, such as motors and switchgear, can be decontaminated by spray cleaning
with Freon.
Precautions are taken to ensure adequate ventilation in order to minimize inhalationof Freon vapor by workers.
About five gallons per week in the form of vapors are removed by thefacility ventilation system. The Freon cleaning and ultrasonic Freon cleaning equipment islocated in the same facility as the electrochemical decontamination equipment and uses the same
ventilation system.
Concrete Removal
The floors and walls in the AFHB are dense, uncoated concrete.
At some locations the contaminated water was in contact with uncoated concrete for up to a year and the contamination diffused
into the surface.
Removal of a layer of the surface was the only way to eliminate this contamination.
This was done by means of a pneumatic impact device.
The maximum surface removal
required was 3/16 of an inch.
5.1.3.2

Work Effort Required

As of September 1, 1980, about 280,000 person-hours had been expended on AFHB decontamination bydecontamination workers, i.e., those workers who plan, prepare for, and carry out the actual
decontamination work, such as operating the water jet and scrubbing and wet vacuuming. 5 An

5-10additional work effort of about 220,000 person-hours was expended by other workers with assignments that required entry into the contaminated areas in the AFHB.
These assignments included
radiation surveys and monitoring by radiation technicians, engineering and construction work,
maintenance, and plant operations.
The work effort by personnel with assignments that did not
require entry into contaminated areas (such as clerical workers, planning staff, etc.) is notincluded.
During early stages of the work, about 30 to 50 percent of the work time of workers
with assignments in the contaminated areas was spent working inside the building; the remaining
time was spent in putting on and taking off protective clothing and gear, planning, training, andsimilar activities.
The licensee expects, as a consequence of experience and improvements inorganization, to be able to maintain a performance level of 60 percent productive in-building
work for current and future operations.
Decontamination work had been completed on about two-thirds of the total surface area (about
340,000 ft 2 of a total of 510,000 ft 2 ) by September 1, 1980.
The average work effort per unit
area was about 1.5 person-hours/ft 2 during this time.
If this ratio does not change during theremainder of the decontamination effort, the total work effort for decontaminating the AFHB byworkers with assignments in the contaminated areas will be about 750,000 person-hours.
The staff
estimates that 40 percent to 50 percent of this time will be spent in contaminated areas inside
the AFHB.
Some of the more difficult tasks, such as sludge removal and decontamination of the sump cubicle,
have been left until last.
The work effort per unit area will be greater for these tasks.
However, experience gained during the first 16 months of decontamination work should decrease thework effort per unit area required for decontamination tasks of comparable difficulty. Taking
both of these considerations into account, the staff regards the estimate of 750,000 person-hours
to be a reasonable one.
The schedule for completing decontamination of the AFHB will depend on the level of work effort.
If the work effort by decontamination workers could have been maintained at a level of two
12-hour shifts per day, five days per week, with a work force of 19 persons per shift (as originally planned by the licensee), with a comparable level by other workers with in-building assignments, decontamination could probably have been completed by the end of 1981.5 However, financial difficulties forced the licensee to reduce the level of effort by decontamination workers to
one 8-hour shift per day, five days per week, with a work force of 8 in September 1980.
The
current level of effort (as of January 1981), which has been maintained since early in September
1980, limits decontamination work in the AFHB to maintenance and decontamination of selected
critical areas.
The current schedule of the licensee shows a completion date of about
September 1982 (see Fig. 1.4).
This implies an appreciable increase from the current level ofeffort some time during 1981.
The number of workers that would, on the average, be needed to provide 250,000 person-hours ofwork effort over a period from May 1, 1979 to September 1, 1982, assuming that the workers spentan average of 40 hours/week on the job, would be about 85.
This is the average number of workers
that would have to be employed in order to carry out the work that involves assignments in contaminated areas in the AFHB.
The actual number at any time will fluctuate as the level of work
effort fluctuates.
Support workers who do not have to enter contaminated areas are not included.
5.1.4
5.1.4.1

Effluents and Releases to the Environment
Normal Operations

Effluents and releases associated with decontamination of the AFHB and equipment include solids,
liquids, gases, and airborne particulates.
The solid wastes include debris removed during initial cleanup operations, contaminated filtersfrom dry vacuuming operations, solid residue removed from comtaminated liquids by liquid processing operations, and contaminated laundry and materials (e.g., wipes) used for decontamination
operations.
The activity of the radioactive waste generated by decontamination operations up to
September 1, 1980, (not including EPICOR II liners generated by water processing) was estimatedto be 128 Ci. 5 Further information concerning releases from solid waste handling may be found in
Section 8.
Liquids from use of the water jet from use of the water jet, from wet vacuuming, and fr6m
desludging operations will be processed to remove dissolved-radioactive contaminants. -The processing operations and the alternatives for the ultimate disposition of the processed liquids are
discussed in Section 7.

5-11
The estimates of concentrations of effluents and releases are based on the licensee's scheduleprior to September 1980, which called for completion of AFHB decontamination work by December 1,
1981.
According to this schedule, the total elapsed time for decontaminating the AFHB was
30 months and September 1, 1980, was the approximate midpoint in time between the start andfinish of the AFHB decontamination work. Any extension of that schedule would lead to a decreasein the concentrations; hence, the values for the release concentrations given below are stillvalid upper bounds for the concentrations that may be expected under the revised schedule(Fig. 1.3-1), which extends the completion date by almost a year. The total releases are independent of the schedule.
During the period from May 1, 1979, when decontamination of the AFHB had started, to September 1,
1980, when about two-thirds of the 510,000 ft 2 of contaminated surface had been cleaned, thereleases of airborne contamination through the AFHB building ventilation systems from decontamination operations did not exceed technical specification limits.* The area that must still becleaned is equivalent to only half the area decontaminated during the prior 15 months.
However,the levels of contamination (see Tables 0.1 through 0.3 in Appendix 0) are higher, so the release
rate of airborne effluents might be essentially the same during a subsequent 15-month decontamination period.
Estimates of the normal airborne releases that are expected to occur during a 15-month period
following September 1, 1980, are summarized in Table 5.1. The desludging operations are expected
to take only about three months; therefore, the calculations of airborne release concentrations
for desludging operations are based on this time interval./rather than 15 months.
Tritium releases during normal operations are primarily from evaporation of the processed waterused for water jet and other washing operations.
The ventilation system fans will be on when
workers are inside the building; hence water vapor in the building atmosphere will be released tothe outside atmosphere when work within the building is in progress.
The processed water has atypical tritium concentration of 0.13 pCi/mL.
The staff estimates that up. to about 2 percent of
the water used for water jet operations will be lost by evaporation; this will generate about
0.1 Ci of tritiated water vapor that will be vented to the outside atmosphere by the building
ventilation system.
The principal normal airborne- releases from surface decontamination operations result from wet
vacuuming operations.
During wet vacuuming operations, 0.1 percent of the material which may
contain up to 60 Ci of plateout and deposits on building and equipment surfaces, is estimated tobe released to the building atmosphere.
This material passes to the air-cleaning system, which
consists of two stages of HEPA filtration.
The staff has based its calculation of airborne releases that might be generated by desludging
operations on experience acquired in complex chemical processing of nuclear fuels indicating thatabout 0.01 percent of the total material processed may become airborne. 6 This estimate is judged
by the staff to be high relative to that expected in the desludging operations.
The auxiliary building and the fuel handling building have separate air filtration systems, eachconsisting of two trains in parallel.
Each train consists of four filters in series: a prefilter, a HEPA filter, a charcoal filter, and another HEPA filter. The maximum flow rate for theauxiliary, building is 65,000 cfm; the maximum flow rate for the fuel handling building is
36,000 cfm.

*The technical specifications limit the releases of 1-131 and particulate activity with halflives greater than eight days in gaseous effluents to an instantaneous release rate of 0.3 pCi/sand to 0.024 pCi/s when averaged over any calendar quarter.
Most of the contribution to particulate contaminants from the AFHB is from Cs-137,-Cs-134, and Sr-90.
The only significant
gaseous. contaminants are H-3 and Kr-85.
The technical specifications limit the instantaneous
release rates of H-3 and Kr-85 to 3 x 104 pCi/s and 4.5 x 104 pCi/s, respectively.
The technical specifications also limit the release rates of H-3 and Kr-85 to 4.8 x 101 pCi/s and.
7.2 x 103 pCi/s, respectively, when averaged over any calendar quarter. When both isotopes
occur together, the individual release limits are less (see 10 CFR Part.20, Appendix B).

5-12
Table 5.1. Airborne Releases of Principal Radionuclides Expected to
Occur from Decontamination Operations in the Auxiliary and FuelHandling Buildings during a 15-M~nth Period
after September 1, .1980
ib
Radionuclide

H-3

Surface
Decontamination

Desludging

All

Concentrations of Releasesc (pCi/mL)
5 x 10-l1
1.9 x 10-12

5 x 10-11

Cs-1372.8 x 10-14

2.8 x 10-12,

2.8 x 10-12

Cs-1344.7 x 10-"s5.5 x 10-13

5.5 x 10-13

Sr-90

3.0 x 10-16

Sr-89d

3.0 x 10-7

3.8 x 10-13
6.3 x 10-14

3.8 x 10-13
6.3 x 10-14

Total Releases (Ci)
H-3

0.1

4.5 x 10-4

Cs-1375.1 x 10-5

6.7 x 10-4

Cs-1348.6 x 10-6

1.3 x 10-4

1.4 x 10-4

0.1
7.2 x 10-4

Sr-90

5.6 x 10-7

9.0 X 10-s

9.1x 10-5

Sr-89d

5.6 x 10-8

1.5 x 10-s

1.5 x 10-s

aReleases during prior decontamination operations (May 1,

1979,

to

September 1, 1980) did not at any time exceed technical specification
limits.
bThe duration of the release is assumed to be 15 minutes; the flow rate
for the building ventilation sysltems are 65,000 cfm for the auxiliarybuilding and 36,000 cfm for the fuel handling building.
cThe concentrations for surface decontamination operations are calculated
averages over a period of 15 months.
The concentrations for desludging
operations are calculated averages over a period of three months.
The
concentrations for all operations are those that also would be applicableduring the desludging operations.
The auxiliary building and fuelhandling building have independent ventilation systems and separate
HEPA filter systems with maximum flow rates of 65,000 cfm and 36,000 cfm,
respectively.
The ducts from both systems join into a common duct tothe plant stack at a point past the HEPA filter systems.
dThe releases for Sr-89 are based on estimates of the amounts of Sr-89
present on September 1, 1980.
About 99.98 percent of this amount will
disappear by radioactive decay by December 1, 1981.

An overall penetration factor of 10-3 has been used in calculating the releases shown in
Table 5.1.* NRC Regulatory Guide 1.140, which develops guidelines for operating nuclear powerplants, specifies a very conservative penetration factor of 10-2 (corresponding to 99 percent
efficiency) for the entire exhaust system and then only if the HEPA filters test in-place to an
efficiency of 99.95 percent or greater. The Regulatory Guide gives no additional credit for HEPA*The penetration factor is the fraction of material entering a filter that passes through thefilters.
Thus, for a given filter, the penetration factor equals 1 minus the filter efficiency
when the filter efficiency is expressed as a fraction.

5-13stages in series.
Application of these criteria to the cleanup would be unnecessarily conservative in that actual penetrations through a single stage of HEPA filters are much less than
one percent (even for particle sizes having maximum penetration) and in that further substantial
reductions of penetration are achieved with a second stage of HEPA filters. Using a conservative
penetration factor of 0.001 for each of two HEPA stages (1 x 10-6 overall), the values shown in
Table 5.1 (except for H-3) would be lowered by a factor of 1000.
5.1.4.2

Accident Scenarios

Possible accidents during surface decontamination activities would include spills, rupture of the
HEPA filter attached to a vacuum cleaner, or rupture of a HEPA filter in the building ventilation
system.
The spilling of liquids inside the AFHB would not result in immediate release to theenvironment and probably would not result in increases in ultimate releases to the river, to theatmosphere, or as solid waste.
The only pathway to the river for a liquid spill inside the AFHB would be through a leak in thebuilding structure and subsequent percolation through the rock and soil to the river.
A hypothetical accident for such a liquid spill and leakage is covered in Section 7.4.2.
The rupture of a HEPA filter in the building ventilation system would increase the airborne
release rate described under normal operations (Sec. 5.1.4.1) until the failure was detected (inabout 15 minutes 7 ) and the ventilation system secured.
An analysis of similar failures in a fuelreprocessing plant 7 indicates that 1 x I0-1 of the filter .inventory could be released to theairstream.
The -resultant concentrations of the principal radionuclides released due to ruptureof a HEPA filter are listed in Table 5.2.
The accidents that could occur during desludging activities are similar to those that could occur
during surface decontamination.
Estimates of releases that could be expected as a consequence ofdesludging operations if a HEPA filter were to rupture are given in Table 5.2.
5.1.5
5.1.5.1

Environmental

Impacts

Occupational Doses

From May 1, 1979, through September 1, 1980, about two-thirds of the AFHB was decontaminated.
The cumulative whole-body dose received by decontamination workers (the workers who operate thewater jets and do the scrubbing, wet-vacuuming, wiping, and similar decontamination operations)was 142 person-rems.
The highest dose received by an individual decontamination worker over thisperiod was 2.5 rem.
Cumulative and average individual decontamination worker doses are given in
Table 5.3.
Other personnel with assignments that required entry and work in contaminated areas in the AFHB
included radiation technicians (who survey and monitor radiation levels and accompany each work
team to monitor radiation levels and provide assistance if needed) engineers, construction
workers, maintenance workers, and operations personnel.
The total cumulative dose during theperiod from April 27, 1979 through September 1, 1980, for all personnel with assignments that
required them to enter contaminated areas within the AFHB is estimated by the staff to be about
250 person-rem.
The staff estimates, on the basis of extrapolations from experience during the period from
April 27, 1979, to September 1,1980, that the cumulative dose for decontaminating the remaining
one-third of the contaminated surfaces in the AFHB will be in the range of 125 to 300 person-rem,
for a total cumulative whole-body occupational dose of 375 to 550 person-rem.
The lower bound isbased on the assumption that the occupational dose per unit area of surface decontaminated willremain the same.
The upper bound is based on the assumption that it will increase by a factorof 3 because some of the more difficult and highly contaminated areas have been left until last.
The factor 3 is based on an extrapolation of the trend in the average dose per worker per month.
The expected number of additional cancer mortalities in a work force exposed to.a cumulative
whole-body dose in the range of 375 to 550 person-rem is 0.05 to 0.07 for a work force of 85
individuals.
The average added probability that an individual worker would eventually die fromcancer would range from about 1 in 1200 to 1 in 1700 if he/she participated in the AFHB decontamination work for the full duration of the cleanup effort. The expected number of additional
genetic effects in the offspring of the exposed workers ranges from 0.10 to 0.14.

5-14

Table 5.2.

Airborne Releases from HEPA Filter Failure
during Decontamination of the AFHB
Releases from HEPA Filter FailureRadionuclidea

Surface Decontamination
Concentrations-of Releases

Desludging(PCi/mL)

Cs-137

1.2 x 10-9

2.4 x 10-8

Cs-134

2.0 x 10-1o

4.7 x 10-9

Sr-90
Sr-89c1.3 x 10-11
1.3 x 10-12

3.3 x 10-9
5.4 x 10-1o

Cs-137

Total Releases (Ci)
5.1 x 10-s

6.7 x 10-4

Cs-1348.6 x 10-6

1.3 x 10-4

Sr-90

5.6 x 10-7

9.0 x 10-5

Sr- 8 9 c

5.6 x 10-8

1.5 x 10-5aother fission products are present; however,on thebasis of analysis of dissolved radionuclides in thereactor building sump water, for which Sr-89, Sr-90,
Cs-134, and Cs-137 constitute greater than 99 percent of the MPC contribution (see Appendix J), andwhich has a radionuclide distribution similar to thatof the AFHB sump water, the contributions of each of
the other radionuclides to airborne releases are
expected to be less than 0.1 percent and have not,therefore, been listed.
bThe duration of the release is assumed to be 15 minutes; the flow rate for the building ventilation
systems are 65,000 cfm for the auxiliary building and
36,000 cfm for the fuel handling building.
cThe releases for Sr-89 are based on estimates of theamounts of Sr-89 present on September 1, 1980.
99.98 percent of this amount will disappear byradioactive decay by December 1, 1981.

AboutTable 5.3.

Whole-Body Occupational Doses Received by Worke sbd ring Decontamination of the Auxiliaryand Fuel Handling Buildings

PeriodApril 27-

June 30,

Average
Dose perWorker (mrem)

Number ofWorkers

Cumulative
Dose(person-rem)

Number ofMonths per
Period

158

183

28.9

2

79

104

192

20.0

3

35

1979-

July 1 - September 30,

1979

Average Doseper Worker
per Month (mrem)

October 1 - December 31,

1979

121

174

21.1

3

40

January 1 - February 26,

1980

182

82

14.9

2

91

February 27 - April 28, 1980

365

52

19.0

2

183

April 29 - September 1,

958

40

38.3

4

240

1980

Highest individual dose - 2.5 rem

U,Total cumulative dose - 142 person-rem

U,aMemo from W.K.

Lehto to J.H. Opelka,

Argonne National Laboratory,"Compilation of Information and Data from

Visit to TMI-2 on May 6, 1980," May 13, 1980.
bFrom T.R. Block, "Auxiliary and Fuel Handling Building Decontamination Program,"presented at an information

briefing by Met-Ed Co. at Crawford Station, Middletown, PA, September 23, 1980.
cThe doses listed are for decontamination workers only; radiation technicians, engineers,and maintenance andoperations personnel are not included.
The total cumulative dose for all personnel with assignments that
required entry into the contaminated areas of the AFHB is estimated by the staff to be 250 person-rem.
dCumulative doses for each period; the total cumulative dose is the sum of the numbers in the column.

5-16
5.1.5.2

Offsite Doses

The estimates for offsite doses to an individual subjected to the maximum exposure from AFHB
decontamination and desludging are listed in Table 5.4. These estimates are based on the source
terms developed in Section 5.1.4.1 and listed in Table 5.1.
The calculational models used tomake these estimates and the interpretation of their results are described in Appendix W. The
significance of these doses and their human health and environmental consequences are discussed
in Section 10.3.
The total-body population dose received by the human population within a
50-mile radius from normal AFHB decontamination activities, is estimated to be 0.02 person-rem.

Table 5.4. Dose Estimates for the Maximum Exposed Individual
for AFHB Decontamination and Desludging Activities
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

4.5 x 10-s
8.8 x 10-s
3.1 x 10-3

5.3 x 10-4
8.8 x 10-5
1.3 x 10-2

4.6 x 10-5
8.8 x 10-5
2.0 x 10-3

TOTAL

3.2 x 10-3

1.4 x 10-2

2.1 x 10-3

Inhalation
Ground Shine
Goat Milk Use

5.4 x 10-5
8.4 x 10-5
1.5 x 10-3

2.0 x 10-4
8.4 x 10-5
1.1 x 10-2

3.0 x 10-s
8.4 x 10-s
1.1 x 10-2

TOTAL

1.6 x 10-3

1.1 x 10-2

1.1 x 10-2

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

4.9 x
1.3 x
4.5 x
5.7 x

5.9 x 10-4
1.3 x 10-4
1.9 x 10-2
2.9 x 10-3

5.1 x 10-s
1.3 x 10-4
2.9 x 10-3
2.2 x 10-3

TOTAL

5.2 x 10-3

2.3 x 10-2

GI-tract, bone,

liver,

Nearest milk
goat

Nearest cow
and garden

aDoses were calculated for total-body,

10-s
10-4

10-3
10-4

kidney,5.3 x 10-3
thyroid,lung, and skin.
The maximum three organ doses are listed in this table. Doses
were calculated for four age groups: adults, teenagers, children and infants.
The highest dose estimates for each age group are listed. The dose estimates
for the nearest garden and the nearest cow and garden are for children, and thedose estimates for the nearest milk goat are for adults per total-body and for
infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W. The
actual locations are: nearest garden = 1.05 miles east-northeast, nearest
milk goat = 1.02 mile north, and nearest cow and garden = 1.05 mile east.

5.1.5.3

Postulated Accident Effects

The type of accident for which dose estimates are made here is rupture of a HEPA filter.
This
accident scenario is described in Section 5.1.4.2, and the source terms are listed for it in
Table 5.2.
Doses are estimated here for a HEPA filter failure occurring during the surface
decontamination operation and during desludging operations.
The calculational models used tomake these, estimates and the interpretation of their results are described in Appendix W. The
dose estimates for individuals subjected to the maximum exposure are listed in Tables 5.5 and
5.6.
The significance of these doses is discussed in Section 10.4.

5-17

Table 5.5.
Estimates of Offsite Doses to the Maximum Exposed Individual
Caused by Failure of a HEPA Filter During Surface Decontamination
Operations in the AFHB
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearestgardenb

Inhalation
Ground Shine
Vegetable Use

2.9 x 10-5
3.5 x 10-4
2.4 x 10-3

1.0 x 10-4
3.5 x 10-4
1.2 x 10-2

4.7 x 10-5
3.5 x 10-4
7.8 x 10- 3

TOTAL

2.8 x 10-3

1.2 x 10-2

8.2 x 10-3

Inhalation
Ground Shine
Goat Milk Use

2.9 x 10-5
3.5 x 10-4
5.7 x 10-3

1.0 x 10- 4
3.5 x 10-4
4.0 x 10-2

4.7 x 10-5
3.5 x 10-4
4.7 x 10-2

TOTAL

6.1 x 10-3

4.0 x 10-2

4.7 x 10-2

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

2.0 x
3.5 x
2.4 x
1.4 x

10-3
10-3

6.9 x 10-5
3.5 x 10-4
1.2 x 10-2
6.0 x 10-3

3.2 x 10-5
3.5 x 10-4
7.8 x 10-3
6.0 x 10-3

TOTAL

4.2 x 10-3

1.8 x 10-2

1.4 x 10-2

Nearest milk
goat

Nearest cow
and garden

10-5
10-4

aDoses were calculated for total-body, GI-tract, bone,

liver, kidney, thyroid,

lung, and skin. The maximum three-organ doses are listed in this table.
Doses
were calculated for four age groups: adults, teenagers, children and infants.
The highest dose estimates for each age group are listed. The total-body dose
estimates for all locations are for adults.
For the bone and liver doses, thethe estimates for the nearest garden and nearest cow are for children, and theestimates for the nearest goat are for infants.
bThe basis for selecting the special locations is described in Appendix W.
Theactual locations are: nearest garden = 1.05 miles east-northeast, nearest
milk goat = 1.02 mile north, and nearest cow and garden = 1.05 mile east.

5-18
Table 5.6.
Estimates of Offsite Doses to the Maximum Exposed Individual
Caused by Failure of a HEPA Filter during Desludging Operationsin the AFHB
Dose (mrem)a
Location

Pathway

Total-body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

6.4 x 10-4
4.7 x 10-3
1.7 x 10-1

9.0 x 10-3
4.7 x 10-3
7.3 x 10-1

6.3 x i0-4
4.7 x 10-3
1.0 X 10-1

TOTAL

1.8 x 10-1

7.4 x 10-1

1.1 X 10-1

Inhalation
Ground Shine
Goat Milk Use

8.6 x 10-4
4.7 x 10-3
8.7 x 10-2

3.8 X 10-3
4.7 x 10-3
6.2 x 10-1

4.6 x 10-4
4.7 x 10-3
6.5 x 10-1

TOTAL

9.2 x 10-2

6.3 x 10-1

6.6 x 10-1

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

4.4 x 10-4
4.7 x 10-3
1.7 x 10-1
2.1 x 10-2

6.2 x 10-3
4.7 x 10-3
7.3 x 10-1
1.1 x 10-1

4.-3 x 10-4
4.7 x 10-3
1.0 x 10-1
8.1 x 10-2TOTAL

2.0 x 10-1

Nearest milk
goat

Nearest cow
and garden

aDoses were calculated for total-body, GI-tract, bone,

8.5 x 10-1
liver,

kidney,1.9 x 1-1
thyroid,lung, and skin.
The maximum three-organ doses are listed in this table. Doses
were calculated for four age groups:
adults, teenagers, children and infants.
The highest dose estimates for each age group are listed. The dose estimates
for the nearest garden and the nearest cow and garden are for children, and thedose estimates for the nearest milk goat are for adults for total-body and for
infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W. The
actual locations are:nearest garden = 1.05 miles east-northeast, nearest
milk goat = 1.02 mile north, and nearest cow and garden = 1.05 mile east.

5.1.5.4

Psychological-Socioeconomic

Effects

The foregoing analysis indicates that the health effects of decontaminating the AFHB are expected
to be negligible under normal operations.
If the planned actions are understood and believed
safe by the general public, the majority of people living in the vicinity of TMI will not experience additional psychological distress or adverse socioeconomic impact.
Once initiated, thisphase of decontamination would be expected to result in a decrease in anxiety levels among those
who are concerned that cleanup was unnecessarily delayed.
The accident scenarios described in Section 5.1.4.2 involve spills, rupture of the HEPA filter
attached to a vacuum cleaner, and failure of the building filter system.
These hypothesized
accidents could have psychological impact.
Although the accident scenarios would have negligible
radiological consequences, some segment of the local community would believe the hypotheticalaccident to be threatening, with a resulting increase in the symptoms of anxiety.
The level ofanxiety would be expected to increase in relation to believed threats, NRC's credibility, thelevel of controversy or uncertainty, and adverse media coverage.
Accidents involving airborne
releases, whether or not such releases reach offsite areas, could result in-short-term consumer
avoidance of agricultural products, particularly dairy products.

5-19
5.1.5.5

Other Environmental

Effects

Discharge of about five gallons per week of Freon vapor to the atmosphere could occur if Freon isused to clean equipment.
Such a release is negligible compared to the global releases of Freonand would have no significant effect in the local area nor on the global ozone layer.
5.1.6

Economic Costs

Based on the number of person-hours of work effort that was required for the portion of AFHB
decontamination completed as of September 1, 1980, (about two-thirds of the total) the staff
estimates that the total direct costs for decontamination of the AFHB will be between about
A cost range is presented to allow for uncertainties in the effort$16 million and $22 million.
required to complete the remaining third of the decontamination.
The cost of desludging operations constitutes less than 5 percent of the total.
The costs for AFHB decontamination are presented in Table 5.7.
The majority of the cost is for
direct labor associated with decontamination and with maintaining~the decontaminated areas.*
Estimated equipment costs include items such as the recirculation vacuum filtration system,
The bases offilter cartridges, filter processing equipment, special casks, shielding, etc.
these cost estimates are described in Appendix K. Costs of waste packaging and handling, andwaste transportation are covered in Sections 8 and 9, respectively.

Table 5.7.
Economic Costs for
AFHB Decontamination(thousands of dollars)
Item

5.2

Best Case

Worst Case

Labor
Equipment &Materials

15,45021,150450650

TOTALS

15,90021,800

REACTOR BUILDING

The initial objective of the reactor building decontamination is to establish and maintain radiological conditions (general area radiation, airborne gaseous and particulate activities, andsurface contamination levels) that will permit reactor defueling and primary system decontamination to proceed.
Defueling is a prerequisite for decommissioning or refurbishment.
Thefollowing judgmental values have been applied as guidelines to determine whether this initial
objective has been achieved:
Reactor building sump water processed and the sludge removed.
General area radiation exposure rates less than-

30 mR/hr below the 305-ft elevation
10 mR/hr at the 305-ft elevation
10 mR/hr average at and above the 347-ft elevation, with exposure rates to about
5 mR/hr at the defueling location.
The lower exposure rates at the defuelingstation could be achieved by suitable placement of shielding and additional local
cleanup.

*Direct labor is here defined as the labor of decontamination and other workers with assignments that require work in radiation zones (see Sec.

5.1.3.2).

5-20
Smearable contamination on building surfaces
4000 dpm/100 cm2 -exclusive of hot spots.

and equipment

approximately 3000 to

Processing of the primary system water and flushing and decontamination of the primary coolingsystem may be necessary in order to bring the radiation exposure rates down to 30 mR/hr below the305-ft elevation.
The foregoing guidelines will not be affected by the outcome of the decision on future use of thesite.
The cleanup operations that must be carried out in order to achieve the initial decontamination objectives are also, for the most part, independent of the outcome.
5.2.1

Status and Specific Considerations

Preliminary reactor building decontamination operations that have taken place during the periodprior to January 1981 include purging of the Kr-85 from the atmosphere in the reactor building,followed by preliminary entries into the building in order to map radiation fields, measuresurface-contamination, and inspect the extent of any damage to the building and equipment.
These
operations were prerequisites for the remaining building and equipment decontamination operations, which would have to be carried out regardless of subsequent decisions on the choice ofcleanup alternatives or the decision on future use of the site. The operations were also necessary to ensure continued safe shutdown of the reactor.
These operations, and the informationthat was gained from them regarding the condition of the building and equipment, are described inthis section.
5.2.1.1

Reactor Building AtmosphereSignificant quantities of radioactive fission products and particulates were released into thereactor building atmosphere as a consequence of the accident on March 28, 1979 (see Table 5.8).
The total atmospheric radioactivity in the reactor building just prior tothe purge was about
44,000 Ci.
Nearly all of this was krypton-85 (Kr-85) gas.
The contribution from tritium (in theform of HTO vapor) was less than 3 Ci, and the contribution from all other radionuclides in theatmosphere was less than 0.001 Ci.
By July 11, 1980, the atmospheric contamination had been
reduced to a level such that full protective equipment is no longer required by workers.
The
sequence of events by which this reduction occurred are described below.
In consideration of the licensee's specific request dated November 13, 1979, for permission toremove the krypton gas, the NRC staff prepared and published a draft environmental assessment inMarch 1980 on decontamination of the building atmosphere and then a final environmental assessment (see Appendix C) in May 1980.
On June 12, 1980, the Nuclear Regulatory Commission, following its own review of the staff recommendations and comments from the public, the Governor of
Pennsylvania, and many other organizations and individuals, issued a Memorandum and Order(CLI-80-25) 8 which authorized Met-Ed to conduct a controlled purge of the reactor building atmosphere, commencing no sooner than June 22, 1980.
The Commission also issued on June 12, 1980,
its Order for Temporary Modification of Operating License No. DPR-73, 9 which established offsite
dose limits for *the purge.
Purging of the reactor building atmosphere by venting to the outside atmosphere started onJune 28, 1980, and continued until the morning of July 11, 1980.
The radioactivity vented during
this period was estimated to be in the range of 38,000 to 50,000 Ci of Kr-85, with a median valueof 44,000 Ci.*
Additional purges were made on August 1, 8, 14-15, and 22, September 19-20,
October 10 and 15, November 8 and 12, and December 4 and 9, 1980, in order to vent the Kr-85
released by offgasing from the water in the reactor building sump and from other sources.
The
venting was done to facilitate manned entries for inspection, sampling, and testing of reactor
building conditions.
Releases during August ventings were less than 60 Ci each, except for thepurge of August 14-15, which resulted in releases of less than 84 Ci.
The releases during themonths of September, October, November, and December were 27, 15, 12, and 7.5 Ci, respectively.
The Kr-85 concentration immediately prior to purging was about 0.8 pCi/mL. The purging operations reduced the airborne radioactivity in the reactor building by a factor of about 107 to thepoint that the concentrations were less than the limits for worker exposure stated-in Appendix B,of 10 CFR Part 20.
Some radioactivity remains because of resuspension of radionuclides frombuilding surfaces by air currents, outgassing of sump water and equipment, and from inleakage
from the primary coolant system.
Periodic purging is needed to prevent the airborne radioactivity from rising above the limits noted above.

5-21

Table 5.8.
Composition of the Atmosphere in the Reactor Building
Shortly before the Purge on June 28 - July 11, 1980

Constituent

Half-Life(years)Activity
Concentration(PCi/mL)

Total Activity in
Reactor Building
Atmosphere (Ci)Volume inReactor Building
Atmosphere (L)
5.7 x 107

Air
H2 Ob
at 2 0 C-

0at 40 C

1000 (liquid)

10.8

0 .8c

44,0003000 (liquid)
31 (gas)d

at 20 0 C

12.3

1.7 x 10-51

7 x 10-7 (liquid)

at 40'C.12.3

5 x 10-5

3

2 x 10-6 (liquid)

Kr-85

-

-

HTOb,e30< 1 x 10-9f< 6 x 10-s

< 10-6 (solid)

Cs-1342.1< Ix

1-129h

1.7 x 107

Sr-90

28.8

Cs-137

1 0 - 9g

< 6 x 10-5<

< 6 x 10-9i< 4 x 10-4

< 10-6 (solid)< 1 x< 6 x 10-5

< 10-6 (solid)

1 0 - 9g

10-6 (solid)aThe reactor containment building internal volume is about 5.7 x 101 liters.
bThe H2 0 and HTO tritiated water were present in vapor form.

A saturated vapor (100%relative humidity) at the temperature indicated was used as a basis for the calculations. The volumes listed in the last column are the liquid volumes that would beobtained if the vapor were separated out and condensed.
cMemo:
J.T. Collins to B.J. Snyder and H.R. Denton, (NRCTMI Program Office Weekly
Status Report" September 8, 1980.
dThe volume of gas that would be obtained if all of the Kr-85 were separated out andstored in a container-at room temperature and one atmospheric pressure.
eThe tritium atoms released initially with the hydrogen gas exchanged with the hydrogen atoms in H2 0 so that almost all of the tritium was present as HTO.
The tritium
activity in the atmosphere was calculated from the tritium activity in the sump water(1 pCi/mL (see J.A. Daniels, "Containment Sump Analyses," presented at the EPRI/DOE
Seminar/Workshop on Decontamination/Dose Reduction Technology, November 17, 1979))
which gives the HTO/H 2 0 ratio, and the amount of water vapor in the reactor building
atmosphere.
fCalculated from measurements of particulates absorbed in filters through which atmospheric samples are passed (reactor building particulate sample data, from continuous
tabulation by the Metropolitan Edison Co., recorded January 30, 1980, and February 13,1980).
gcalculated from the measured Cs-137 activity in the reactor building atmosphere andthe measured Cs-134/Cs-137 and Sr-90/Cs-137 ratios in the containment sump (see "Environmental Assessment for Decontamination of the Three Mile Island Unit 2 Reactor
Building Atmosphere," U.S. NRC, NUREG-0662, March 1980; J.A. Daniels, "Containment
Sump Analyses," presented at the EPRI/DOE Seminar/Workshop on Decontamination/Dose
Reduction Technology, November 27, 1979; and reactor building particulate sampling
data, from continuous tabulation by Metropolitan Edison Co., recorded on January 30,
1980, and February 13, 1980).
hThe 1-131 radionuclide, which has a half-life of eight days, has decayed to an undetectable and completely negligible level.
iCalculated from the concentration of 1-129 in the sump (see Daniels, cited above) andthe atmosphere/sump concentration ratio for 1-129.
The latter was obtained from theatmosphere/sump ratio for 1-131, which was determined from measurements prior toAugust 31, 1979, before the 1-131 had decayed to unmeasurably low levels.

5-22
Environmentaltion 5.2.5.2.
5.2.1.2

impacts that occurred as a consequence of the purge are described in Sec-

Entries

These
Five successful entries had been made into the reactor building as of January 2, 1981.
entries were preceded by an attempted entry on May 20, 1980 (prior to purging of the reactor
building atmosphere) which was aborted because the entry team was unable to open the inner doorof personnel airlock No. 2.
The inner airlock door was successfully opened on July 23, 1980, andtwo people entered the reactor building to conduct a general radiation survey on the 305-ft level
For the successive entries on July 23,
and inspect for damage resulting from the accident.
August 15, October 16, November 13, and December 11, 1980, the size of the entry team was 2, 4,
12, 5, and 14 persons; entry times varied from 20 to 121 minutes; the total cumulative occupational exposure for all team members for each entry was 0.42, 1.03, 2.28, 4.11, and 5.9 personrem; and the maximum whole-body dose received by any individual for each entry was 220, 340, 570,
460, and 650 mrem, respectively."1'1 2 No skin exposure from beta radiation was detected until
the fourth and fifth entries, when the maximum beta skin exposures were 170 mrad and 65 mrad,respectively.
The entry team members wore protective clothing for all entries.
On the firstentry breathing air was supplied by individual tanks of compressed air; on succeeding entries,
air from the containment building was drawn through individual HEPA filters by battery-poweredfans into pressurized full-face masks.
The entry teams made radiation measurements, took samples of surface radioactivity by wiping andscraping surfaces, inspected for damage, took color photographs of the interior, and removed a
few small loose items for subsequent laboratory analysis.
During the third entry, essential
maintenance on the No. 1 personnel airlock and on equipment used for monitoring from outside thereactor building was carried out.
During the fourth entry, a high-volume air sample was taken
and tests were made to determine the effectiveness of three different hands-on surface decontamination procedures.
During the fifth entry the reactor pressure vessel head was surveyed frominside the empty refueling pool and the polar crane was inspected.
The results" from these
activities are summarized in Sections 5.2.1.3 through 5.2.1.5. Typical views of the inside of
the reactor building, taken during the entries, are shown in Figures 5.3 and 5.4.
Before the start of decontamination, additional entries will be needed on a continuing basis toobtain more information on the distribution of radioactive contamination and radiation field
intensities, and on the condition of the equipment.
This information is needed to plan for thecleanup operations and for inspection and possible maintenance of instrumentation used to monitor
the condition of the reactor.
Procedures for subsequent entries are expected to be similar to
those already used.
5.2.1.3

Building and Equipment Surfaces

Radioactive deposits (plateout) are present on most of the approximately 300,000 ft 2 of exposedbuilding and equipment surfaces inside the reactor building.
Measurements of plateout fromsamples obtained by wiping or scraping at 28 locations during the first and second entries
yielded the following results for the plateout activity (in units of pCi/100 cm2 ):** measurements at eight locations on the 305-ft elevation floor ranged from 0.3 to 47, with an average of
14; measurements at four locations on the 305-ft elevation walls ranged from 0.006 to 0.02, withan average of 0.01; measurements at two locations on the 347-ft elevation floor ranged from 6.4
to 6.9, with an average of 6.7; and measurements at four locations on the 347-ft elevation walls
ranged from 0.01 to 0.04, with an average of 0.02. The remaining measurements of miscellaneous
objects at ten locations, such as stairwell surfaces, tool chest, glass sample, steel plate,funnel, cable trays, entry door, etc., ranged from 0.04 to 29, with an average of 5.

*The initial estimate of 57,000 Ci of Kr-85 was based on a conservative interpretation of measurements of small samples of the reactor building atmosphere obtained through penetrations in
the outer wall of the reactor building. The values quoted above are based on measurements ofthe plant stack gas velocity and Kr-85 concentration in the vented gas.
The uncertainty of±6000 Ci is due to inherent difficulties in making accurate measurements of gas velocities inlarge vent stacks and the concentration of Kr-85 in the gas stream.
These difficulties
limited the accuracy to about ±14 percent.' 0**An activity of I pCi/100 cm2 is approximately equal to 9.3 pCi/ft 2 or 2.2 x 106 dpm/100 cm2 .

5-23

I

Figure 5.3.

View Outside the Inner Airlock Door of the Reactor Building Just
The photograph shows Met-Ed's personnel in proPrior to Entry.
tective clothing and carrying communications and radiation detection equipment making final preparations prior to entering theinner door of the personnel access hatch of the reactor building
at the 305-ft elevation (ground level).

(N)Figure 5.4.

View Inside the Reactor Building during Entry. A view of the entry teaminspecting the condition of equipment and making radiation measurementson the 305-ft elevation (ground floor).

5-25
In a decontamination test conducted during the fourth entry, a small area on the floor of the
305-ft elevation with a beta-gamma surface activity of about 14 pCi/100 cm2 was wiped with a
towel saturated with demineralized water, then wiped with a dry towel, then wiped with a towel
saturated with a detergent solution, then wiped dry again.
After the first dry wiping the surface activity was reduced by a factor of 87; after the second dry wiping it was reduced by anadditional factor of 2.4.
The plateout in the reactor building came from fallout as the steam escaping from the primarysystem condensed during and shortly after the accident.
It was probably distributed much more
uniformly on the walls and floor immediately after the accident.
The staff infers that the muchsmaller plateout values on the walls is primarily due to flushing that occurred after the accident during the period that moisture evaporating from sump water heated by decay heat from thereactor was condensing on the walls and draining down on to the floors and back into the basement.
The principal radionuclides in the plateout are Cs-137 (83 percent) Cs-134 (13 percent) andSr-90 (4 percent).
Trace amounts of Co-60, Nb-95, Sb-125, and Ce-144 were measured at somelocations.
The Cs-137/Sr-90 ratios varied, depending on the location at which the sample wastaken.
Ratios as high as 53 and as low as 3 were measured, with an average of about 16.
Thevariation is due to the nature and concentration of the liquids in contact with various surfaces
and subsequent washing by condensation.
The relative amounts of the three principal solid radionuclides are also different for different sources.
The proportions for major sources at different
locations are shown in Table 5.9.
The total plateout on the floors, walls, and ceilings on the 305-ft and higher elevations can be
estimated from the numbers given above and the surface areas to be about 2 Ci.
This does notinclude the radioactivity in drains and other locations not accessible for swipe measurements
where contamination could accumulate.
Many of the "hot spots" are due to such sources.
Nor doesthe 2-Ci estimate include the plateout in the basement (i.e., the adherent film of contaminationthat would remain after the sludge, which consists of a layer of nonadhering particles, isremoved.*) or inside the D rings.
There are not sufficient data available as to the amount ofradioactivity in drains or other inaccessible locations, in the basement, or inside the D rings,
but the staff considers it unlikely that the total would be greater than 100 Ci; it probably ismuch less.
Most of the surfaces at the 282.5 ft (basement) level are submerged.
The floor is covered to adepth of about 8 ft with about 700,000 gallons of sump water.
This sump water and any sludge
that has accumulated on the basement floor under the water must be removed before the surfaces inthe basement can be decontaminated.
The condition of the building and equipment surfaces in thebasement and the amount of sludge present were not known as of January 2, 1981.
5.2.1.4

Radiation Levels

Plateout, the contaminated sump water, and suspended particles are the major sources of worker
radiation exposure for both beta and gamma radiation. The general gamma radiation levels (excluding "hot spots") were measured during the first three entries in the range of 100 to 700 mR/hr
at the 305-ft elevation (grade level) and 30 to 600 mR/hr at the 347-ft elevation.
The average
gamma radiation levels are estimated to be 500 mR/hr and 250 mR/hr at the 305-ft and 347-ft
elevations, respectively.
The staff estimates that the average gamma radiation level at the
305-ft elevation would drop to about 250 mR/hr if the basement were drained and decontaminated tothe extent that radiation from the basement through the floor and drain openings was appreciably
less than the contribution from plateout on the floor and walls at the 305-ft elevation.
Measurements made of beta radiation levels by instruments in contact with the surface at various points
on the walls and floors gave readings that ranged from 2 to 4 rad/hr.
Higher radiation fields occur at localized "hot spots" where concentrations of plateout occur andabove the open stairwell, which is not protected by an intervening floor or wall from radiation*Both sludge and plateout are removed from surfaces primarily by means that utilize water; theformer by resuspension and the latter by going into solution or by detergent action. The
important distinction between the two is that the former can be removed from the water byfiltration; methods utilizing ion-exchange resins or minerals, or evaporation, are necessary
for the latter.

5-26

Table 5.9.
Relative Amounts of the Three Principal
Radionuclides for Different Sources

Radionuclide
SourceCs-137

Cs-134

Sr-90

Plateout in reactor building

83%

13%

4%

Dissolved solids in reactor building sump water

86

13

1

Sample A

8

1

91

Sample B

11

2

87

Primary system water

51

8

41

Dissolved solids in water in AFHB sump

81

16

3

Filterable solids in water in AFHB sump

79

16

5

Dissolved solids in water in AFHb sump tank

82.

16

2

Filterable solids in water in AFHB sump tank

71

14

15

Filterable solids in reactor building sump waterfrom the sump water.
Typical measurements of the gamma radiation in such areas were 1-2 R/hr at
the air coolers,* 2-5 R/hr at floor drains, 10 R/hr over the metal deck for the covered floor
hatch, and 18 R/hr at the open stairwell.
The highest measurement, obtained remotely, was 40-45
R/h at 5 to 7 ft from the sump water in the. basement (below the'305-ft elevation).
5.2.1.5

Condition of Equipment

The containment spray system was actuated on March 28, 1979, following a hydrogen burn within thereactor building; hence, the system was operable at that time and may still be operable.1 3 In
addition, some of the building lights were activated during the second entry. Otherwise,-there
is no basis for assuming that any of the other reactor building equipment, services, or protective instrumentation needed for decontamination of the building surfaces and equipment would be
operable at the beginning of decontamination operations.
Provisions for supplying all of therequired services and equipment by repair, replacement, or other means would be needed.
It isbelieved that the reactor building equipment hatch at the 305-ft elevation can be used to transfer equipment to and from the reactor building before decontamination operations are begun. The
status of certain key pieces of equipment is noted below.
The electrically-powered polar crane is assumed to be inoperable because of the adverse environment in the reactor building and because of the long period without operation or maintenance.
During the fifth entry, visual observation from a platform immediately below the crane cab
revealed considerable heat damage to electrical components.
The reactor building air cooling units, located adjacent to the equipment hatch at the 305-ft
elevation, are operable but are grossly contaminated and will require extensive decontamination,
shielding, or removal to reduce radiation levels for personnel working in the area.
In-place
repair or replacement would be required if these units were to 'be used for building cooling
during cleanup.
The extensive amount of air ducting at this elevation in the building may behighly contaminated and require special decontamination efforts, or replacement.
(Some priordecontamination to reduce worker exposure would probably be needed even if the ducti.ng were
replaced.)*The most detailed measurements around the air coolers were made during the third entry on
October 10, 1980.

5-27
There was evidence at some locations within the building of the pulse of abnormally high temperatures and pressures which occurred during or after the accident:
a dent in the enclosed stairwell door at the 305-ft elevation; a collapsed metal storage drum; a melted plastic telephonehandset on the 347-ft elevation; and a charred sign.
This damage probably occurred during thehydrogen burn.
Aside from these observations, visual inspection did not reveal any obvious
damage to the equipment at the 305-ft and 347-ft elevations.
Because the basement is flooded with 8 ft of contaminated water,probably will have to be removed and disposed of.
5.2.2

most of the equipment there

Alternative Methods Considered

Alternatives other than those listed below may be suggested during the decontamination operations
and should be considered if there is reason to believe that, within the ALARA concepts, theywould reduce worker exposure, the amount of waste generated, decontamination time, or occupational hazards.
5.2.2.1

Building Decontamination

Decontamination below the 305-ft elevation (the "basement" level) and on the 305-ft and higher
elevations (the "upper" levels) present different problems, primarily because of the 700,000 gallons of contaminated water that overflowed the sump and now covers the basement floor (the282.5-ft elevation) to a depth of about 8 ft.
Decontamination of the basement will involve
pumping out the contaminated water and removing the sludge and debris, as well as decontaminating
the building and equipment surfaces. . Decontamination of surfaces is the major building decontamination task required for the upper levels.
The alternatives for the basement and upper
levels are considered separately below, but they are not independent of each other--some of the
major alternatives concern coordination of decontamination tasks on different, levels of thebuilding.
Alternatives for Decontaminating the Basement
The contaminated water in the basement (the sump water) must be removed before other decontamination tasks in the basement can be started.
After the water has been removed, removal of thesludge and debris and then decontamination of the underlying surfaces can proceed.
Alternatives for Removing the Sump Water.
The alternatives for removing the sump water dffer inthe rate of removal, transfer methods, timing, and clean-water backfilling of the basement.
Rate Alternatives.
Three major alternatives that differ in the rate at which the water
would be removed were considered by the staff: (1) rapid removal to a storage tank at some otherlocation on the site; (2) removal at a rate that would permit the water to be processed before it
was stored at some other location; and (3) a feed-and-bleed operation in which the water was
returned to the sump after processing until the dilution reduced the contamination to a predetermined level.
The rapid removal alternative is considered by the staff to be the least desirable of the threealternatives.
In order to accommodate all of the contaminated water in suitable existing tanks
it would be necessary to utilize, the Unit 2 reactor coolant bleed holdup tanks (231,750 gallons),
the tank farm in spent fuel pool A of Unit 2 (110,000 gallons), the Unit 1 reactor coolant bleed
holdup tanks (247,000 gallons), and spent fuel pool "A" of Unit 2 (320,000 gallons). 1 4 The
precautions needed to avoid exposure to personnel engaged in other operations, the increased
radiation levels in the storage areas, and the loss of spare tank capacity for managing theoverall water inventory would hamper cleanup operations.
The greater complexity (number ofvalves,, pumps, etc.) and length of lines utilized compared to other alternatives would increase
the transfer hazards. 'The storage hazards are also increased; there would be much less likelihood of leakage or spillage of the contaminated water (which has a radioactivity concentration of
0.75 Ci/gallon) if it could remain in the reactor building basement until it was processed.
Useof Unit 1 facilities is very undesirable and would by transferring radioactive accident water
increase the extent of contamination and increase potential hazards from accidental releases bysignificant amounts.
For these reasons, rapid transfer of the contaminated water into existingstorage facilities should be considered only as an emergency measure in case some unforeseen
situation (such as major leakage from the building to the environment) made necessary the rapid
removal of the contaminated-water from the basement for temporary storage elsewhere.

5-28
Some of the preceeding objections could be eliminated by building new. storage facilities specially designed to receive the contaminated water in the reactor building basement.
Heavilyshielded and leakproof tank, piping, and valves that would withstand earthquakes, tornados, andfloods would have to be designed and constructed at some location outside any of the existingbuildings but within a controlled ventilation structure.
An extended period would be requiredfor planning and construction, and even with the utmost care, the hazards from leakage and spillage from a tank located outside of the AFHB or reactor building would be much greater than for
the alternative of leaving the water in the reactor building basement until the water was processed.
Rapid removal of a fraction of the sump water to one or more of the Unit-2 reactor coolant bleed
holdup (RCBH) tanks is a less hazardous and more practical operation.
There are three RCBH
tanks, each with a capacity of 77,250 gallons, located in shielded cubicles in the auxiliarybuilding.
Even though this option does involve some increase in the spread of radioactivity andin the hazards of leakage or spillage, such spills would be limited to a part of the AFHB.
This
alternative of partial removal could be used (1) to permit the addition of water from otherdecontamination activities to the basement inventory without increasing the water level, or(2) to accelerate the basement draining schedule above that which would be possible if the other,slower water-removal alternatives were used.
Processing of the sump water as it is removed is considered by the staff to be' a practical alternative.
For a processing rate of 2.5 gallons per minute, which is the expected average processing rate for planned processing systems (see Sec. 7.1.3), the processing time for the water now
in the basement would be about seven months.
This time would be increased by one month for every
100,000 gallons of water added from decontamination operations at the 305-ft and higher elevations in the building.
If the feed-and-bleed alternatives were used to reduce the contamination in the sump water (otherthan tritium) by a factor of 106, which is the approximate reduction in dissolved radionuclides
in the effluents from a processing system with decontamination factors comparable to EPICOR II,
seven years would be required.*
A variant of this alternative would be to use a feed-and-bleed
operation for a shorter period of time to partially reduce contaminant concentrations and thenimplement a second alternative to complete the processing.
Transfer Alternatives.
Four transfer alternatives for removing water from the reactor
building basement were'considered by the staff and judged to be practicable:(1) remove thewater through the existing 4-inch-diameter sump pump discharge pipeline from the reactor building
sump into the.AFHB; (2) install a new submersible pump and discharge system in the sump or on thebasement floor; (3) adapt the 18-inch-diameter sump discharge lines that feed the reactor building spray system and decay heat removal system to divert sump water to the liquid waste processing system, or (4) use a surface suction system inside the reactor building to transfer'sump
water to the tank farm.
The existing sump pumps in the reactor building basement cannot be used because they are underwater and are not operable.
The pump in Alternative 1 would be located in the auxiliary building
and could discharge into a liquid waste collection header in the same area. The use of an existing pump, WG-P-1, located in the auxiliary building at an elevation of 281 ft was considered.
Use of this particular pump presents two disadvantages:(1) when the sump water level dropped
below 281 ft, air could be drawn in through the spray header and valve BS-V-1 (via air leakage),which could air-bind pump WG-P-1; and (2) sump water and sludge would pass through the reactor
building spray pump, BS-P-1, and sludge would accumulate in this pump, thereby creating a significant source of radioactivity.
A more practical variant of alternative 1 would'be to use a pumpsimilar to WG-P-1 in a reactor building spray pump vault, which is located in the auxiliarybuilding at the 258½-ft elevation.
This would eliminate the potential air-binding problem(because the pump- would be below the bottom of the reactor building sump at 276½-ft) and sludgebuildup problem (because the BS-P-1 pump could be bypassed).
Timing Alternatives.
The bounding cases for timing alternatives that involve basement
decontamination operations are:(1) coordinate water removal and decontamination operations sothat the water surface does not rise above the current level (which was 290.55 ft on January 8,*If the water used for decontamination operations at higher building-elevations were allowed todrain into the basement, this time would be increased.

5-291980, and rising at a rate of about 0.06 ft/month because of leakage of about 5000 gallons permonth from the reactor cooling system); or (2) proceed.with decontamination activities without
regard to the water level in the basement.
A net increase in the amount of water in the basement would cause the water level to reach the
40,000 gallons would bring the water level to the centerline of one
following critical points:of the valves, DH-VI-171, that controls the flow of water from the primary system to the minidecay-heat-removal system (MDHRS); 115,000 gallons would bring the water level to the bottom of
the Penetration 401,* 170,00Q gallons would bring the water level to the bottom of equipment that
operates DH-VI-171; and 1,270,000 gallons would bring the water level to the 305-ft elevation.
The decay heat valves inside the reactor building that control water flow to the MDHRS are now
These valves might operate under water, although they wereopen so that the MDHRS is operable.
not designed for such operation, but they could not be closed if the water level reached the
The only valve that would then be available to control flow to the MDHRS
operating equipment.
and isolate any leaks that might develop downline from the decay heat valves would be DHV-3,
which is located outside of the reactor building.
Coordination of water removal with remote wash operations in such a manner that the amount ofwater in the basement did not increase by more than 100,000 gallons (in order to avoid floodingof penetration 401) and preferably less than 40,000 gallons, would be the major concern for the
A remote wash could add from 60,000 to 360,000 gallons of water to the sumptiming alternatives.
in a period of a few days; hence, it might be necessary to remove up to 260,000 gallons before aremote wash could be initiated. The staff considers it unlikely that a remote wash will be done,
for reasons discussed below under remote decontamination alternatives.
If water removal started before semiremote decontamination operations on the upper levels, theThe waterwater level in the basement should not rise as a consequence of these operations.
The water jet used in semiremote washing operations
removal rate is expected to average 2.5 gpm.
has a flow rate of 7 gpm, but it is operated intermittently; draining of water into the basement
as a result of water jet use presents no problem relative to increasing the water level as long
If wateras the water jet is not turned on more than about 8 hours during each 24-hour period.
removal were delayed until after semiremote decontamination operations were underway, it might benecessary to transfer some of the sump water to the reactor coolant bleed holdup tanks.
Other timing alternatives that would not directly affect decontamination operations in the basement are discussed below in connection with decontamination operations on the upper levels.
Currently the sump water is the major source of radiation in the
Backfilling Alternatives.
After the water has been pumped out, the major source of radiation in the basement
basement.
The two principal alterwill be exposed deposits of sludge and other surface contamination.
natives for reducing the radiation exposure from sludge and contaminated surfaces exposed byremoval of the water would be to (1) proceed immediately with removal of the sludge and decontamination of the exposed surfaces in the basement, or (2) backfill the basement with clean orprocessed water in order to reduce radiation levels while decontamination work on other levels is
The staff has estimated that the radiation levels at the underside of the floor atin progress.
the 305-ft elevation from the exposed deposits in the basement could be up to three times less
The radiation levels at thethan the radiation levels from the sump water before it was removed.
305-ft and higher elevations are, therefore, expected to decrease uniformly as the water is
Backfilling the basement would delay cleanup of
pumped out and the basement is decontaminated.
the basement and increase the amount of water that required processing.
The thickest deposits of sludge are expected to be in the reactor building sump, which is a pit
6 ft deep, 7 ft wide and 8 ft long with a capacity of 2500 gallons located in the basement floor*Penetration 401 is a penetration through the 4-ft-thick outer wall of the reactor building.
It is normally closed with an
It has a diameter of about 2.7 inches at the inside wall.
It can be used to insert instruments or retrieve air or water
airtight and watertight seal.
It is prosamples; it is currently being used for instruments that measure the water level.
posed to be modified by adding a welded closure assembly to eliminate the potential forflooding through the penetration without eliminating the capability for making water-levelmeasurements. s

5-30
just north of the center of the D-ring.
Backfilling only the sump with clean or processed waterto reduce exposure from this source would be a practical and desirable procedure that would notinterfere with cleanup operations in the remainder of the basement area.
The alternatives for removing the sludge from the reactor
Alternatives for Removing the Sludge.
building are essentially the same as for the AFHB, except that in the reactor building the sludge
might be deposited over a much larger area.
However, neither the actual amount nor distributionof the sludge in the reactor building is known.
If it is assumed that the amount of sludge, which consists of particulate matter washed down frombuilding and equipment surfaces, is proportional to the building and equipment surface area, andthat conditions are comparable in the AFHB and reactor building, then the amount of wet sludge inthe reactor building will be about two-thirds of the amount in the AFHB, or about 100 ft 3 . Thisis probably an overestimate because almost all of the surfaces in the reactor building had been
painted or otherwise finished in a manner that would facilitate decontamination, while a large
fraction of the surfaces in the AFHB is unpainted concrete, which sloughs off more particulatematter.
In addition, there was an unfinished part of the floor in the AFHB that contributed alarge part of the cement residue, and there were ion-exchange resin beads present.
There were no
unfinished sections of floor in the reactor building and no means by which ion-exchange resins
could contribute to the sludge.
The concentration of radioactivity in the sludge can be expected to be about the same as theconcentration of radioactivity in solids filtered from the sump water.
The radioactivity offilterable solids in the reactor building sump water was 1.6 Ci/ft 3 on July 1, 1980.16
The
isotopic distribution for the reactor building sludge is assumed by the staff to be the same as
for the filterable solids in the reactor building sump water.*
Agitation of the sump water as it was being removed in order to resuspend as much of the sludge
as possible and remove it with the water was considered by the staff; however, it appears
unlikely that this method would be useful when the water level was high enough to cover theentire basement floor. Agitation of the water in the basement would have to be by remote means,
and it would be impractical to provide sufficient agitation to remove sludge deposits from all
surfaces in the basement as the water receded.
Agitation is expected to be most useful toward the end of the water removal operations when the
only remaining water is in the sump, which may contain an appreciable accumulation of sludge.
Awater jet is expected to be the most effective means for washing the sludge from the remainder ofthe basement into the sump where it can be removed by any of the alternative methods discussed in
Section 5.1.2.2.
Alternatives for Surface Decontamination.
After the water and sludge had been removed from thebasement, the alternatives for decontaminating the surfaces would be the same as those discussed
below for decontaminating the surfaces at the upper levels, except that the remote wash alternative would not be applicable, and it might be desirable to use a high-pressure (10,000 psi) water
jet in addition to the moderate-pressure (1200 psi) water jet used in the AFHB.
Alternatives for Decontaminating the Upper Levels (305-ft and Higher)
The alternatives considered by the staff for decontamination operations on the 305-ft and higher
levels in the reactor building involve (1) coordination of tasks on the basement and upper
levels; and (2) the methods used for surface decontamination.
Coordination Alternatives.
The major coordination alternatives concern the timing of the startof decontamination operations on the upper levels relative to completion of initial cleanup
operations in the basement.
The reference time for distinguishing alternatives is the time atwhich cleanup of the basement has proceeded to the point that the contribution to the radiation
fields on the upper elevations from sources in the basement is small compared to the contribution*Two measurements of the filterable solids in the reactor building sump were available as ofJanuary 30, 1981; one reported in February 1980;17 the other reported in July 1980.16 The
July 1980 measurements were on a sample taken from near the bottom of the sump and are expected
to be more representative of the properties of the sludge.

5-31
from plateout on the walls and floors of the upper elevations.
Tenfold reduction in the overall
strength of radiation sources in the basement would be sufficient for this purpose and could beaccomplished either by removing the water and sludge from the basement (except for the sludge inthe sump itself, which could be shielded by backfilling or by other means) and washing down someof the most highly contaminated surfaces (if water and sludge removal alone were not sufficient)or by a feed-and-bleed operation at an average rate of 2.5 gpm over a period of 15 months (seediscussion above under "Rate Alternatives").
This reduction of the radiation source strength in
the basement by a factor of 10 or more, by whatever means, is referred to in this section as the"initial cleanup" of the basement.
The major coordination alternatives considered by the staff are:(1) defer the start of decontamination tasks on the upper elevations of the buildings until the initial cleanup of the basement is completed; or (2) begin decontamination of the upper elevations before initial cleanup
operations in the basement have started.
There is, of course, an intermediate range of alternatives in which decontamination at upper elevations would begin after initial basement cleanup had
been started but before it had been completed.
Another intermediate alternative would be an
immediate preliminary decontamination of the floors alone (on which the plateout activity is
100 times larger than on the wells--(see Sec. 5.2.1.3), with further decontamination of the upper
levels deferred until initial cleanup of the basement was completed.
The alternatives (1) and(2) above represent bounding cases.
The only restrictions on coordination alternatives that seem appropriate in the light of information available at the time this document was prepared (other than ALARA considerations foroverall planning) are that the water level in the reactor building basement should not be allowed
to reach an elevation of 291.85 ft (which can be accomplished by not allowing the amount of waterin the basement to increase by more than about 100,000 gallons), and that the reactor pressure
vessel head should not be removed until about 300,000 gallons of sump water from the basement(together with any water that might be added from decontamination activities) have been processed.
This amount of water, plus the 743,000 gallons of water already processed by EPICOR II,
are needed for the spent fuel pools, cask pit, and fuel transfer canal.
These facilities must be
filled with about 1,040,000 gallons of water before the reactor pressure vessel head is removed
and defueling operations commence.
If defueling operations began before 300,000 gallons ofprocessed water from the basement were available, it would be necessary to use a corresponding
amount of clean water.
This would cause an undesirable increase in the amount of tritium-contaminatedwater that would have to be stored or disposed of after cleanup operations were completed.
Surface Decontamination Alternatives.
The primary decontamination task on the 305-ft and higher
elevations of the reactor building is decontamination of surfaces.
The methods available forthis task may be classified as remote, semiremote, and hands-on categories.
Remote Decontamination Alternatives.
In the staff's judgment and based on currently available information, remote decontamination is not an essential step in the cleanup of the upper
elevations of the reactor building; however, in the event that a decision is made to carry out aremote decontamination step, the staff has examined two alternatives:(1) remote wash, and(2) use of robots.
The reactor building spray system, which can be operated remotely from outside the building,
could be used to carry out a remote wash.
This system is similar to the sprinkler systems usedin public buildings, except that the flow of water is much greater.
It can be modified for steam
injection, although this involves some worker exposure because the spray nozzles must be changed.
The staff considers the following alternatives to be practicable:(1) overhead spray deluge;(2) overhead steam injection through the building spray system; (3) a low-elevation saturated
steam injection into the building atmosphere (which would also involve worker exposure during
installation) and (4) a combination of the first and third alternatives.
The primary advantages of a remote wash is that it would provide an initial reduction in radiation levels with very little
worker exposure.
The disadvantages are that a large volume of water
would be generated and only the dome area could be decontaminated; there is no spray systeminstalled below the 347-ft elevation, which is the floor of the dome area.
The walls in the dome
area are much less contaminated than the floors, probably because they already have been partially washed by condensation from the building atmosphere (which has been at 100 percent humidity for much of the time); hence the decontamination factor for a remote wash may be lower thanis usual (i.e., 2 to 10) for such operations.

5-32
The use of detergents or chemicals rather than processed water for the remote wash was consideredby the staff, but will not be pursued further because the large amounts of detergent or chemicals
required would severely complicate the eventual processing of the wash water.
The use of robots for remote decontamination was not considered further for the following reasons.
The usefulness of industrial robots is limited because they are unable to move over obstacles; the field of view, resolution, and depth perception of their optical systems are limited;
and the manipulative capacity of mechanical arms and grippers is limited.
Robots speciallydesigned for the TMI-2 environment would require considerable development and a long lead time.
The staff does not, however, rule out the use of robots in specific circumstances when thesedisadvantages are not a severe handicap and when remote means are needed to cope with decontamination of high radiation areas where shielding is not feasible.
Semiremote Decontamination Alternatives.
Semiremote decontamination operations require thepresence of people inside the reactor building to operate the decontamination equipment.
The useof appropriate equipment would limit occupational radiation exposure by allowing the workers to
be farther from the radiation source and by decreasing the worker exposure time.
The various
methods could be used individually or in combination, or alternatively, one method might follow
another on the same surface, with the choice depending on the success of the application of thefirst option.
The options considered by the staff are:hose washdown, using low-pressure water;
low-pressure hot water misting and washing; high-pressure water jet washdown; steam jet washdown
using steam to propel other liquid cleaners; steam cleaning using steam alone; sandblasting; andvacuum blasting.
These alternatives are not regarded by the staff as being exclusive nor is it intended to exclude
other methods, such as ice blasting or the use of foam.
All of the methods listed are feasible
and any of them may be used.
Some of the procedures that normally involve hands-on techniques could also be done semiremotely.
For example, commercial equipment that can perform scrubbing and wet-vacuuming operations on
floors automatically under remote control are available and may provide a practical means for
reducing operator exposure for these operations.
Hands-on Decontamination.
The hands-on phase of decontamination would be instituted when remote(if any) and semiremote decontamination methods had been productively exhausted.
Most hands-on
approaches are close in concept to industrial janitorial services--workers use wet cloths,
brooms, scrub brushes, mops, vacuum cleaners, etc., to clean surfaces and equipment.
These
methods are the same as those used for AFHB decontamination (see Sec. 5.1.2.1).
Most of the surfaces in the reactor building are painted, and it is expected that washing andscrubbing with water or detergent, or, at most, removal of the paint film will be sufficient to
reduce the plateout to acceptable levels.
However, there are some areas (for example, part ofthe surface area of the wall in the enclosed stairwell) where removal of some of the material
beneath the surface may be required.
Some areas in the basement that have been immersed incontaminated water for almost two years also may rpquire such treatment. Methods that remove alayer of material include the use of sandblasting, jackhammer, pneumatic or hydraulic impactor,
or concrete spaller.
The discussion in Appendix U is relevant to the foregoing on these techniques.
5.2.2.2

Equipment Decontamination

The equipment in the reactor building can be decontaminated by a variety of methods.
The alternatives listed below are all feasible and have been used in previous decontamination operations(see Sec. 1.5).
The alternative methods for decontamination of surfaces are the same as those
discussed in Section 5.1.2.1.
In-place decontamination.

This alternative would involve use of hands-on techniques.

Local demounting and reinstallation.
before using-hands-on techniques.

This

technique would

incorporate

disassembly

Removal for decontamination outside of the reactor building.
The equipment would be
removed to a special decontamination area where it could be decontaminated remotely orwith hands-on techniques.

5-33
DDisposal.
This option would be appropriate where replacement was more cost effective
than decontamination.
5.2.2.3

New Facilities Needed

A large-scale equipment handling and transfer capability in a containment service building would
be required for temporary storage and movement of equipment and decontamination of the reactor
building.
The minimum requirements of this containment service building would be to serve as an
The alternatives considered byextended barrier, as an airlock, and as an interface building.
the staff for fulfilling these requirements are:A medium-size, medium-integrity, steel-frame, metal-sided building designed to provide
proper control of contamination and to provide minimal processing area.
A larger,medium-integrity building similar to that above designed to provide spacethat would permit more processing and ease congestion.
5.2.3
5.2.3.1

Details of Methods and Facilities
Descriptions of Methods and Facilities

Descriptions of those feasible methods for decontaminating the reactor building that were listed
Descripand described briefly in Section 5.2.2 and are most likely to be used are given below.
tions of the enclosures that would be needed to control the spread of contamination, details of
the equipment and of the decontamination procedures -for the equipment, and descriptions anddiagrams of the containment service building needed to control access to the reactor building andExamples of successful decontaminations using someprovide a staging area also are given below.
of these techniques are given in Section 1.5.
Semiremote Decontamination
Semiremote decontamination techniques involve the use of equipment that permits the operators toExperience has shown that it is most effective tostay some distance from the radiation source.
Hot spots
remove all loose materials and debris before these techniques are used (see Sec. 1.5).
are shielded either before or while the semiremote decontamination is being conducted and thenare decontaminated later when the general background dose rates are lower.
For the semiremote decontamination methods, the use of a holdback-carrier solution consisting of
one milligram of nonradioactive cesium per liter of reprocessed water is advantageous in largeenough nonradioactive cesium is added toscale cleanups.
The holdback carrier works as follows:
the base liquid so that the number of atoms of inactive cesium in the solution will far exceed
As the holdback-carrier solution is
the number of atoms of radioactive cesium on the surface.
washed down a contaminated wall, the radioactive cesium on the wall is replaced by nonradioactive
cesium.
By using holdback carriers, the contaminant transferred to the wash is more likely to
Hose wash and high-pressure water jet
remain in solution than to be redeposited on the surface.
Other semiremote operations
operations are suitable for use of holdback-carrier solutions.
considered are steam cleaning, sandblasting, and vacuum blasting.
Hose Wash.
As a decontamination method, hose wash offers some advantages in terms of flow rate
These factors are especially advantageous forcontrol, flow pattern, and directional properties.
However, because of low impact forces, hose wash is lessdecontamination of hard-to-reach areas.
In fact, if
effective as a general decontamination technique than the other methods considered.
Depending on
the surface being cleaned is covered with oil or grease, hose wash is ineffective.
Flow rates for hose wash areconditions, hose wash decontamination factors range from 2 to 100.
typically about 50 gpm.
SolidThe hose-wash flow pattern can be a solid stream, a spray fan, or surface irrigation.
stream washing is used to forcefully penetrate and dissolve encrusted material and to transport
A portion of reactor building surface contaminantspooled or piled contaminants toward drains.
will be combined with borate films and sodium hydroxide to form "Borax", which is soluble inwater.
Hose wash is, therefore, a major cleanup tool in the building/equipment cleanup task andwill be particularly effective on those surfaces that were not adequately cleaned by means ofremote decontamination.
The spray fan flow pattern provides wide area coverage with low splatThe fan
ter.
Surface irrigation nozzle patterns provide a mild flush action with no splatter.
Surface irrigation may
pattern is expected to provide rinse action following solid-stream wash.
be used to gently flush contamination to lower elevations.

5-34
Low Pressure Hot Water Misting and Washing.
In this method the water is applied by means of anozzle that generates a fine mist.
It is similar to a hose wash, with the difference that theuse of hot water makes it more effective for surfaces covered with oil or grease and the use of amist rather than a stream of water reduces the volume of water used.
Water Jet.
High-pressure water jets* consist of a high-pressure positive-displacement pump witha lance or gun for delivery.
The unit can produce water pressures as high as 15,000 psi, but
pressures of less than 10,000 psi are more typical.
Pressures of 1000 psi are common in lowpressure units.
The high impact force of these units makes them effective tools for removal of
hard-to-remove contamination.
The volume of water emitted per unit time is a function of the delivery pressure and the nozzle
opening.
However, the total volume of water required to complete a given task is approximately
equal using either high- or low-pressure water jet units.
The selective use of nozzles improves decontamination efficiency by matching nozzle flow patterns
to specific work tasks.
For example, the slotted fan spray nozzle with its knife edge pattern istypically used for descaling applications, while the circular nozzle (using high pressures)provides a high impact stream at distances up to five meters.
The higher pressures will descale
paints and remove oxide films.
High-pressure water jets create splatter and can redistributecontaminants to other locations on the same surface.
This characteristic can be minimized bycareful pressure selection.
A water jet unit can be equipped with a nozzle brake for more precise control at the sacrifice of some delivery pressure.
Water jets can be turret-mounted toreduce operator fatigue during long hours of use.
Unit positioning may be assisted by crane oruse of movable platforms on wheels.
While the water jet may yield high decontamination factors (typically exceeding 1000), ultimate
deposition of scattered contaminants is not completely controllable, and at high pressures thepotential for creating high concentrations of airborne radionuclides is strong.
The use of lower
pressures to improve control might extend the length of the immediate task, but as a tradeoff itwill reduce unwanted side effects.
The use of radiological confinement, such as temporary barriers, isolation, and collection sheets, is essential during high-pressure water washing.
Steam Jet.
Steam-jet decontamination is similar to water-jet decontamination.
Water or reagentsare introduced into an injector system and mixed with steam. The water-steam mixture is expelled
through a nozzle.
The delivery pressure is controllable from atmospheric to very high, but istypically less than 1000 psi.
The volume of water delivered is usually greater than that forwater jet. The delivery rate is about 16 gpm at 250 psi and 19 gpm at 400 psi.
Steam-jet application of decontamination reagents is best done at low velocity.
The paths of the waste solutions and removed contamination from steam-jet decontamination are thesame as for the previously described hose and water-jet washing.
The potential for greaterliquid-waste generation must be balanced against the potential for a higher decontamination
factor.
The heated solution may provide a higher decontamination factor and the lower pressure reduces
splatter and subsequent dispersion of the removed contaminants.
This technique provides better
control than a water jet but may not provide the necessary abrasive force to remove contaminants.
The driving unit is usually the water and steam pressure; whereas the water jet requires a positive displacement pump.
Steam-jet cleaning is best applied to cleanup of areas containing oil and grease deposits.
Tothis end the steam jet may be more effective for cleaning fixed pump and oil-lubricated equipment, as well as areas near such components.
Borate and sodium tetraborate deposits may be moreeasily cleaned or dissolved using this method as a result of the elevated temperature of thesolution.
Steam Cleaning.
The steam-cleaning method involves ejection of steam through a nozzle or lance
directed toward the surface to be cleaned.
Typical pressures are less than 500 psi.
The generation of liquid waste is considerably less for this method than for the previously described*The high pressure water jet will sometimes be referred to merely as a "water jet."5-35
methods; however, a means must be provided to flush or vacuum the loosened wet contaminants.
250 psi the rate of water delivery is less than 2 gpm.

At

Steam-cleaning techniques are frequently used for degreasing and, therefore, may lend themselvesto cleaning the same areas and components as discussed for steam jet cleaning.
For effectiveness, the nozzle must be close to the surface, and as with the other methods, contaminant dispersion is the major disadvantage.
Steam cleaning is especially effective over irregular surfaces and provides minimal runoff where other methods may cause dripping or collection of excessive amounts of water in areas where liquids are undesirable.
Sandblasting.
Sandblast decontamination is an abrasive cleaning technique involving use of airpressure to drive sand particles at high velocity against a surface.
Most often it is a dry
technique with little
or no liquids produced.
The abrasive action can be adjusted by grit size,
air pressure, and standoff distance to provide an action that will scour, peen, polish, or remove
the surface material.
Sandblasting is a rapid and effective decontamination method, but becauseof the abrasive action the method is limited to use on surfaces where roughing does not destroy
the usefulness of the item worked on.
A disadvantage of sandblasting is that a large volume ofcontaminated grits are generated and must be disposed of as radioactive waste.
Typical uses forsandblasting are cleanup of odd-shaped equipment, paint removal, and cleanup of metal and porous
surfaces such as concrete where deep contaminant penetration cannot be removed by liquid cleaning
methods.
Abrasive action during sandblasting is somewhat dependent on the skill of the operator.
Typically, sandblast air pressures for steel and concrete run up to 90 psi.
Pressures of about
60 psi provide scouring action, and still lower pressures may be used on nonferrous metals.
Asand grit size of 20 mesh is usually quite damaging to surfaces, while grits above 100 meshcreate a particulate airborne contamination problem.
Sandblasting techniques require confinementof the abrasives.
Movable tenting is generally used to control dust.
The contaminated sand may
be collected and removed by vacuuming.
Wet sandblasting is milder than dry sandblasting and lends itself better to complex shapes;
cracks can be irrigated when water is used as the driving force.
The water cycle of wet sandblasting also acts as a rinse so that the generation of airborne radionuclides is low.
Vacuum Blasting.
The techniques of vacuum blasting, with abrasives other than sand, is a provenmethod and provides for the reuse of the abrasive material.
In most cases, the abrasives aresteel shot or wire cuts.
The action is harsh and very effective.
Reduced pressures can be
employed to scour rather than remove surfaces.
The removed surfaces, contaminants, and pellets
or cuts are simultaneously vacuumed up and separated by a cyclone separator. The contaminants
are removed to a specially protected unit chamber.
On metals, the action can have a scouring or
peening effect.
Since the unit can be floor mounted or hand held, its versatility lends it
either to semiremote or hands-on methods of decontamination.
Vacuum blasting is most useful when
the unit can be held perpendicular to the surface being cleaned.
On irregular surfaces, reflection of abrasives can be a problem and vacuum pickup is less efficient.
The principle advantageof this method is control of blast dust, which is nearly all returned to the unit. Very little
dust is produced when a metal abrasive is used, and the life of the abrasive material can be as
much as 100 times that of sand.
Confinement, tents, barriers, etc., are recommended and should
be used as standard practice.
Decontamination factors with vacuum blast exceed 1000.
Hands-on Decontamination
The hands-on effort should be minimized by first using semiremote decontamination methods to the
maximum extent possible so as to minimize dose to the workers.
The effective use of barriers,
partitions, screens, curtains, and enclosures is desirable to separate work areas (clean fromdirty), permit several activities to proceed simultaneously, and minimize recontamination.
General surfaces (such as floors) are expected to require scrubbing either by industrial floor
scrubbers or by hand, followed by wet vacuuming, and possible detergent cloth wiping.
A finalreagent/rinse mopping then would complete the effort. Any liquid or solid waste generated would
be collected and disposed of by solidification or compaction in a form acceptable for disposal.
Overhead areas may require damp scouring with reagents followed by rinses and cloth wipes.
High-elevation work above floors probably will involve use of bosun chairs, scaffolding, andtelescoping platforms to reach all surfaces.
The area above the polar crane may be reached byusing the crane beams as a staging platform.

5-36
Aids to provide more efficiency in the hands-on effort include industrial janitorial equipment
and hand-held automobile scrubbers and polishers. Application of strippable coatings or foamedon cleaning agents followed by wet vacuuming and wiping reduces scrubbing requirements.
Generally, a final wiping to remove residual radioactivity (transfer-technique) is more effective than
repeated rinses.
Areas where paint is removed may require immediate sealing after decontamination.
Depending on
the status of the reactor building decontamination operations, decontaminated floors either
should be covered or kept clean with oil-cloth brooms.
To reduce the potential for resuspended radioactivity in the basement from migrating to otherelevations, this level should be isolated by means of barriers placed over penetrations in the
305-ft-elevation floor.
Portable blower systems with HEPA filters could be strategically placedat the 305-ft elevation with "elephant trunk" ducting inlets lowered through the grating areas.
These blower systems would create a negative pressure (in relation to the 305-ft elevation), andthe local air flow pattern would be from less to more contaminated areas.
Concurrent with the above task, eductors could be introduced into the water in the basement.
The
sediment/sludge would be agitated by the eductor system suction, and dishcarge from the system
would pass through screens and shielded roughing filters to remove sediments and sludge.
Themain objective would be to remotely remove the sludge and particulate contamination from thebasement with as small a contribution to the personnel dose as reasonably achievable.
Equipment
similar to that used to vacuum sludges from reactor fuel storage pools could be used, but the
exposure of workers to radiation would be considerably higher.
An alternative to the agitator system is a high-pressure water jet coupled to suction dredges.
The dredged solution would be recycled through a filter bank system and a portion of the solution
used as the liquid supply for the water jet. This method requires an agitation pool at least one
foot deep to provide shielding and to prevent resuspension.
During the dredging operation,ceilings and walls could be decontaminated by semiremote techniques.
The details of methods used for concrete removal are described in Appendix U.
Enclosures and Barriers.
Enclosures and barriers consisting of portable, semi-airtight screens
and tents would be needed in order to inhibit the spread of contamination to pre-cleaned areas.
The first radiological enclosure needed would be for isolating the encroachment area.
This
enclosure would be located at the equipment hatch.
The equipment hatch location has a monorailavailable and is contiguous with the planned containment services building.
The enclosure erected for the encroachment area would be typical of other enclosures used in thereactor building except for size.
The enclosure might be a waterproof room constructed of afire-retardant wooden framework, with polyvinyl chloride roof and walls.
The floor could be
fitted and glued in place after the walls were up.
The walls of such structures customarily haveample clear material for viewing windows inserted in opague wall panels.
All seams would be
double stitched, then coated with vinyl glue to cover stitch holes.
Each compartment could beequipped with a blower and roughing filter so that the air would flow from the direction of clean
areas to contaminated areas.
An alternative would be to provide a blower fitted with high efficiency filters (HEPA) in the last of several connected compartments which, when actuated, moved
air down the chain of compartments.
This system would establish a pressure difference between
compartments, keeping the dirty sections at negative pressure with respect to the cleaner areas.
Air flow would then always be from clean to dirty, thereby inhibiting the spread of contamination.
If enclosures were needed in an area before decontami'nation could proceed, shielding might berequired to protect the workers erecting the enclosures and those who would work inside.
A
low-density shielding material (e.g., rubber, wood) would shield out low-energy beta radiation.
A thin layer of aluminum or steel would shield out higher-energy beta radiation.
If reduction of
gamma radiation were required, lead blankets, lead sheet, lead brick, or high-density concrete
blocks would have to be inserted between the gamma source and the enclosure.
Shielding materials
should be packaged or covered to prevent their contamination.
This would reduce the amount ofsolid wastes.
Fabricated stack tanks or movable five-gallon containers filled with water mightalso serve as temporary shielding materials.
To reduce the possibility of airborne contamination from affecting other work zones, the 305-ft
and 347-ft levels should be sealed from one another at the appropriate time.
Floor gaps could be5-37
In areas where liquids
sealed with fire-retardant materials, polyethylene, or waterproof tape.
are required to flow down the reactor building wall and through the floor gap, the seal could
Blower systems with HEPA filters located on thecontain a sluicing system that could be opened.
347-ft and 305-ft elevations would be needed to maintain the desired air flow rate and air direction (from areas of lower contamination to areas of higher contamination) between elevations andDecontamination of the spaces
allow the reactor building cooling system to function effectively.
Semibelow the 305-ft level then could proceed without recontamination of the upper levels.
remote methods then could be employed to decontaminate ceiling and walls exposed above the sump.
The sump would serve as a holding reservoir for spent decontamination liquids and reduce resuspension of airborne particles.
Equipment Decontamination Alternatives
In-place decontamination includes hosing, washing, brushing, or steam
In-Place Decontamination.
This is the fastest andcleaning of equipment without removing it from the installed position.
least expensive method, and involves the least total radiation exposure to operating personnel;
however, the method would likely ruin some of the equipment and would leave contamination increvices, threaded connections, and areas to which adequate access is not possible.
Examples of equipment that probably could be decontaminated in place are sealed or enclosed
motors above the flooded level that are in protected locations, such as the elevator drive systems at the 370-ft elevation level and motorized actuators between the 305-ft and 347-ft levels
that would be protected from contaminated "rainfall" by the floor at the 347-ft level.
This procedure would include the same methods as discussed above, but in
Local Demounting.
addition would involve, as needed, demounting of the equipment or adjacent equipment to permit
Increased radiation dose would result from thecleaning of otherwise inaccessible locations.
additional time required for removing the components, especially if these procedures were complex.
The equipment would remain demounted until a decision on ultimate use of the facilities
were made.
It would then be removed if the facilities were decommissioned, or remounted if the
facilities were reused.
Equipment that could be decontaminated locally after demounting are enclosed devices like motors
or actuators subjected to dripping contaminated water, or equipment such as handrails, simple
seismic restraints, or hatch cover gratings.
Components too complex or delicate
Removal for Decontamination Outside of the Reactor Building.
for quick cleaning in place or local decontamination after demounting. could be removed from theIt would provide storage for the
A building is being established for this.
reactor building.
It could be used for refurbishing equipment
equipment until the start of Phase III operations.
either for salvage (if the facilities were decommissioned) or reuse.
Equipment that would require removal from the reactor building include such items as instrument
cabinets, cabling, hydraulic snubbers, and some electrical switchgear.
Disposal.
Any component that had been submerged or contaminated to a point that the decontamination effort and resulting occupational radiation exposure would be excessive with respect to
its replacement costs would be a candidate for disposal.
Likely candidates for the disposal category are items that have been submerged, such as ducting,cabling, and instruments at the 282-ft level, or items for which past experience indicated that
Examples are crane lifting cables and exposed mirror insudecontamination was not feasible.
lation on the primary system.
For much of the equipment, detailed advanced planning is not possible because the nature of thecontamination problems is not known and the effectiveness of in-place decontamination methods as
An example is the cooling coils in the air-coolingthey apply to specific items is not certain.
These are certain to be highly contamiequipment at the 305-ft level of the reactor building.
nated because of their long-term operation in the highly contaminated building atmosphere; however, the primary contaminant is likely to be cesium, which is water soluble and might be removed
without the need to disassemble the air-cooling units.
There are various pieces of equipment in each of these categories, but most will probably not beclassifiable in advance of actual work because of the unknown contamination and condition of eachcomponent.
Since considerable equipment would have to be removed from the reactor building,5-38
regardless of whether it were decommissioned or refurbished, an extensive parts control andinventory system would be required.
There would be extensive use of the facilities immediately
outside the equipment hatch, so that some quick method of making a disposition decision (salvage
or scrap) decision would have to be developed, or a holding area would be required at some placeout of the way of the activity at the hatch.
This would require packaging of the components fortransport through noncontaminated areas, at the cost of additional handling and increased occupational dose.
Certain equipment recovery activities can be predicted at this time and are listed below by theirlocation in the reactor building.
Details of Equipment at Different Levels
430-Ft Elevation--Polar Crane Level.
The polar crane would have to be decontaminated and at
least partially refurbished early in the reactor building decontamination program to permit its
use to move heavy equipment between levels in the building, including installation of an auxiliary crane and small boom cranes to aid in the decontamination effort.
Additional activities
likely to be required are removal of the trolleys on the fuel handling bridge, removal of theshield blocks above the reactor, removal and handling of the reactor coolant pump motors, cablereplacement, and handling of any casks or shielding required during the decontamination.
Theeffort required might vary from refurbishment of only the bridge drive components in place tocomplete repl.acement of the trolley. The polar crane is needed for cleanup and defueling activities, regardless of whether the facilities are decommissioned or reused.
367-Ft Elevation--Top of Shield Walls.
Additional lifting and handling equipment (in the form ofsmall boom cranes and possibly a bridge crane with rails) would need to be installed at thislevel before most of the work involving defueling or work on the primary system inside the shieldwalls could start.
347-Ft Elevation--Main Working Floor Level.
This elevation would have to be cleared of all
unnecessary equipment to make room for handling of components being moved out of the building andfor laydown of components during defueling operations.
Refurbishment or modification of the fuelhandling bridges at this elevation might also be a high priority task.
305-Ft Elevation--Equipment Hatch Level.
The area around the equipment hatch and a path to thevertical access hatches on the south side of the building would have to be cleared and established as relatively low-radiation-level areas early in the decontamination effort since mostequipment transfer and personnel entry and exit would be at this location.
The reactor building
air cooling units located on this level immediately adjacent to the equipment hatch probably arehighly contaminated and would require extensive decontamination, shielding, or removal.
Refurbishment in place or replacement would be required in order to provide building ventilation.
The
extensive amount of air ducting at this elevation also might require special effort if it ishighly contaminated.
Most essential building services would be brought in through the equipment
hatch or spare reactor building penetrations at the 305-ft level.
282-Ft Elevation--Sump Level.
Because about 8 ft of highly contaminated water now floods thesump level, most of the equipment at this level probably would have to be scrapped.
Associated Facilities
Large-scale equipment handling and transfer capability would be required for decontamination anddecommissioning or refurbishment of equipment and the reactor building.
The 23-ft diameterequipment hatch on the southwestern side of the reactor building at ground level is the only
feasible exit/entry point for these transfers.
Since it would be necessary to use this hatch
before the inside of the reactor building is decontaminated, an enclosure of some type would be
required on the outside of the equipment hatch to serve as an interface between the reactor
building and the outside environment.
The minimum requirements for this service building would
be as follows:
Serve as an extended contamination barrier so that contaminated items can be broughtout of the reactor building for staging, decontamination, examination, and packaging.
Serve as an airlock so that material and personnel transfer through the equipment hatch
is independent of uncontaminated transfers in and out of the containment service
building.

5-39Serve as an interface building to permit rigid control of equipment,
personnel entering and leaving the contaminated zone.

supplies,and

The containment service building also would include other facilities needed for working withcontaminated material and equipment, such as a laundry facility for decontamination of clothing,equipment repair areas, and areas for temporary storage of contaminated waste, and also related
facilities that do not involve contaminated material, such as a personnel access facility and awork-effort management and control facility.
A conceptual design for a minimum-size building to meet
Minimum-Size Medium-Integrity Building.
the facility requirements is shown in Figure 5.5.
It would include an airlock outside the equipment hatch.
The airlock would have to be large enough to accommodate the largest piece of equipment, including its transport device.
This might be a reactor coolant pump motor, the fuelhandling bridge trolley, or the auxiliary bridge crane that has been proposed for installation toaid in defueling efforts.
Beyond the equipment hatch airlock would be another isolation arealarge enough to permit loading any of the equipment onto a truck or perhaps a rail car.
Equipment would be packaged in the equipment hatch airlock, transferred to the loading area and verified as being free of surface contamination before the building's exterior doors could be opened.
Interlocking would be required to ensure that the airlock/loading area door could not be opened
when any of the exterior doors were open.
Some limited decontamination or equipment survey andevaluation work could be conducted in the air lock, but this would obstruct personnel access andmaterial transfer through the hatch.
To preclude contamination of the previously clean areas andequipment, contaminated items would not be brought into the loading area before packaging.
Most
equipment would need to be transported to other facilities for decontamination,. refurbishment or
disposal.
Contaminated water, as well as particulate solids that were suspended in the water, would bepumped out through pipelines into the AFHB.
The solids would be filtered out, immobilized, andpackaged for final disposal in the AFHB.
Sludge and other wet solids that were not removed inthis manner would have to be packaged in a form suitable for moving through the hatch and staging
area without introducing any appreciable increase in the radiation exposure.
This could beaccomplished either by immobilizing and packaging the wet solids within the reactor building sothat they would be in the final packaged form before being moved through the equipment hatch, orby using shielded, watertight containers to move the wet, unprocessed solids through the equipment hatch to a location outside the reactor building where they could be immobilized and packaged for final disposal.
A steel-frame, metal-sided building could be built at the equipment hatch to meet the minimum
facility needs identified above.
Such a building might be amenable to complete decontamination
and, hence, disposal as uncontaminated waste after all decontamination operations had been completed.
It would need to be seismically isolated from the containment building and could bedesigned for Uniform Building Code (UBC) seismic and wind loading conditions.
Building integrity
would be sufficient to permit staging and processing of reasonable quantities of contaminatedmaterials, but versatility would be low because of the minimal space provided.
Larger Medium-Integrity Building.
A larger version of'the building described above could beconstructed to provide more space and greater versatility.
A diagram of a larger building of thesame general construction is shown in Figure 5.6.
The building would cost more, but should nottake significantly longer to design and construct.
The additional space would permit more of theprocessing of contaminated material to be performed in the building; this, in turn, would
decrease the effort that would have to be expended in packaging contaminated materials for transfer to processing operations elsewhere.
Use pressure and congestion should also be more controllable because of the greater versatility and increased space of the building.
More storage
space for equipment would also be available while waiting for a decision on decommissioning
versus restart.
The conceptual design has a large work area between the equipment hatch airlock and the various
loading and shipping areas.
This would permit significant waste staging and processing to occur
within the contamination control boundary, but outside the high radiation levels of the reactor
building and away from the congested area immediately outside the equipment hatch.
Since there
would be less packaging for transport and unpackaging for processing, contaminated packaging
waste generation and occupational radiation dose should be less.

5-40

~600

Figure 5.5.

Minimum Size Containment Services Building.
EXISTIW. ROAD

Figure 5.6.

Larger Containment Services Building.

5-41

5.2.3.2

Work Effort Required

Decontamination of the reactor building will require a~coordinated effort between decontamination
The work.effort estimates presented increws and support personnel from several disciplines.
this section include decontamination workers and workers with assignments that require entry into
the reactor building (radiation protection, engineering, construction, maintenance, and plant
operations personnel with in-building assignments), but not workers with assignments that do not
require entry (personnel engaged in training, suitup assistance, supply, clerical, planning,
The staff assumes that for
management, and similar assignments that do not require entry).
personnel with assignments that require entry into the building, the proportion of time spent inproductive work on the inside will be about half of the total.* The remainder of the time willbe spent outside the building in putting on and removing protective clothing and equipment,planning and instruction sessions, and similar activities.
Work effort estimates were obtained by two methods: (1) by a detailed analysis of the times
required for individual decontamination tasks, and (2) by extrapolation from experience in the
AFHB.
The detailed analysis led to an estimate of approximately 240,000 person-hours for the inbuilding work.
The staff assumes that an approximately equal amount of time will be expended by
the same workers in activities outside the reactor building; hence, the estimated work effort fordecontaminating the reactor building is about 480,000 person-hours.
Estimates of certain
Many of the time estimates for individual tasks were quite conservative.
tasks (e.g., semiremote and hands-on decontamination operations) using alternative, less conserOn the basis of such judgmental considerations, thevative, procedures gave lower estimates.
staff infers that the work effort for decontaminating the reactor building probably will not beless than about 300,000 person-hours.
The total accountable work effort by workers with assignments inside the AFHB amounted to about
This number
1.5 person-hours for each square foot of surface decontaminated (see Sec. 5.1.3.2).
is based on the known work effort by decontamination workers and an estimate that each personhour of decontamination work (setting up and operating the water jet, scrubbing, wet vacuuming,etc.) required an additional 3/4 hour of support operations (radiation surveys, monitoring andstand-by duty by radiation control personnel, and plant operations, maintenance, and construction
activities).
The proportion of support to decontamination activities will be greater in thereactor building because of the work involved in providing access to surfaces that cannot bereached from existing structures,** constructing, barriers and rerouting ventilation to prevent
spread of decontamination in the large open areas, providing essential building services (e.g.,
additional lighting, electrical outlets and water for decontamination operations), and managing
The staff assumes that thisthe materials and large pieces of equipment that must be moved.
additional work may increase the amount of support effort to as much as 2½ hours of support workfor each hour of decontamination work, which would increase the work effort per unit of decontaminated area from 1.5 to 3 person-hours/ft 2 . This leads to an estimate of 900,000 person-hours ofwork effort for cleaning up the 300,000 ft 2 of contaminated building and equipment surfaces
inside the reactor building.
The staff regards this estimate as an approximate upper bound.
The preceeding estimates do not include the work effort for defueling operations or for cleanupof the primary system.
The staff assumes, on the basis of the schedule provided by the licensee (Fig. 1.4), that work on
initial decontamination of the reactor building (i.e., decontamination to the level at which
defueling activities can start) will extend over a period of at least 18 months and, in view of
the uncertainty in the work effort that will be required and in the choice of water removal
On the same basis, completion of decontamination
alternatives, could take three times as long.
activities to the point where ALARA criteria have been attained will require at least 36 months*The licensee's objective is to maintain a 60 percent productive in-building effort; on thebasis of experience in the AFHB, the staff considers an estimate of 50 percent to be more
realistic.
**Essentially all of the decontamination work in the AFHB has been to date on floors and walls
within easy reach; ceilings and upper walls have not yet beendecontaminated.

5-42
and will probably take much longer.
Decontamination work will not be complete when Phase II
operations begin.
A useful estimate of the time that will be required to complete decontamination to ALARA levels cannot be given at this time because of the following unknown factors:
thedelays that will be introduced by post Phase II operations, the contamination that will be generated by these operations, and the level of effort, which is largely controlled by the rate at
which funds can be made available to support the decontamination work.
The staff assumes, forcalculational purposes, that decontamination work performed in connection with Phase I and II
operations will be carried out over a period that may range from 18 to 54 months.
This assumption allows for the possibility that decontamination work that goes beyond the initial objectives
described in the introduction to Section 5 may be deferred.
5.2.4
5.2.4.1

Effluents and Releases to the Environment
Normal Operations

To date, the largest release to the environment from the reactor building due to cleanup activities was the 44,000 Ci of Kr-85 that was vented to the outside atmosphere during the period from
June 28, 1980, to July 11, 1980.
Most of the other radionuclides present in the reactor building
atmosphere before purging (approximately 1 Ci of tritium in the form of tritiated water vapor(HTO),
less than 4 x 10-4 Ci of 1-129 and less than 6 x 10-5 Ci of each of the radionuclides
Sr-90, Cs-137, and Cs-137--see Table 5.8) were removed by the HEPA filters in the venting system.
Releases of the Kr-85 gas were controlled to limit the cumulative maximum individual offsite
doses to the public resulting from the purge to less than the annual dose design objectives(15 mrem to the skin, 5'mrem whole-body) of Appendix I to 10 CFR Part 50.
The offsite dose that resulted from the purge is given in Section 5.2.5.2.
Periodic releases are made of much smaller amounts of Kr-85 generated by outgassing-, primarily
from the contaminated water in the basement.
The dates of the releases, limits for August andmonthly totals for September through November are given in Section 5.2.1.1.
Venting is now being
done on a monthly basis with one release a few days before each entry and another during theentry period.
After decontamination operations are under way, continuous Venting may be desirable to minimize worker exposure during decontamination operations.
The staff assumes, on thebasis of the data given in Section 5.2.1.1, that 15 Ci/month is a conservative upper bound for
the Kr-85 releases from the reactor building during building decontamination- operations.
The
releases decreased each month during the period from August 1980 through December 1980, and willprobably be around 1 Ci/month before July 1981.
Other effluents and releases associated with decontamination of the reactor building and equipment include solids, liquids, and airborne particles.
Solids generated by decontamination activities in the reactor building come from sludge debris
and contaminated equipment removed from the the building, contaminated filters from vacuuming
operations, solid residue removed from contaminated liquids by liquid processing operations, andcontaminated laundry and materials (e.g., wipes) used for decontamination operations.
The effluents and releases from these materials are associated with waste disposal operations which aretreated in Section 8.
Decontamination activities within the reactor building initially will not result in any direct
releases of liquid to the environment.
The volumes of liquids that will be used for decontaminating the building and equipment surfaces inside the reactor building are estimated by the staff
to be zero to 370,000 gallons of water for remote wash operations 70,000 to 230,000 gallons ofwater for semiremote wash operations and 30,000 to 50,000 gallons of detergent solution, about
11 percent liquid detergent of unknown composition* and 89 percent water.
The staff assumes thatthe most likely volume of liquids that will be generated by all building and equipment surface
decontamination operations is 150,000 gallons of water and 40,000 gallons of detergent solution.
The staff estimates that at least 90 percent of the plateout in the reactor building will be*The detergent that was used for scrubbing operations in the AFHB was Radiac Wash, a proprietarychemical solution that contains ethylenediaminetetraacetic acid (EDTA) and other chemicals.
This same detergent probably will be used for scrubbing operations in the reactor building.

5-43
removed in the water, and most of the remainder will be removed in the detergent solution.
Processing and disposal of these liquids are discussed in Section 7.
The operations are expectedto last 18 to 54 months.
If the water in the basement were not removed prior to building decontamination operations or
isolated by a vapor barrier from the rest of the building, most of the tritium released from thereactor building would be tritiated water vapor from evaporation of the water in the basement.
The maximum possible release of 2500 Ci of tritium would occur if all of the water in the basement were removed by evaporation.
If saturation conditions (100 percent humidity) could bemaintained with the building ventilation system in operation (at 50,000 cfm), the entire
700,000 gallons of water would evaporate in about a year.
(This would correspond to an evaporation rate of about 0.001 pounds of water'per square foot per minute.)
The concentration oftritium in the air under these conditions would be 2 x 10-s pCi/mL, which is larger than theguidelines for restricted areas in 10 CFR 20 by a factor of 4.
These limiting conditions are unrealistic and Unattainable; however, the staff condiders an evaporation rate of about 20 percent of this limit to be a reasonable bounding case if the building
ventilation system were in operation and no special precautions were taken to prevent evaporationof the water.
If this situation continued for a period of 18 months, approximately 750 Ci of
tritium in the form of tritiated water vapor would be released to the atmosphere.
Under these
conditions the average concentration of tritium in the reactor building atmosphere would be
4 x 10-6 pCi/mL.
The staff estimates that the amount of tritiated water that will evaporate and be released fromsemiremote operations will, be about 2 Ci.
This represents a lower bound to the release oftritium that would occur if the water in the basement were removed prior to decontamination oreffectively isolated. Under these conditions the average concentration of tritium in the reactor
building atmosphere would be negligible.
Total airborne radioactivity discharged to the environment as a consequence of decontamination
operations in the reactor building is estimated by the staff to be about 10-4 Ci (excluding H-3).
The ventilation flow rate is 50,000 cfm.
The air cleaning system contains two stages of HEPA
filtration.
A penetration factor of 10-3 has been used in calculating the releases (,seeSec. 5.1.4.1).
The release concentrations of principal radionuclides are listed in Table 5.10.

Table 5.10.
Airborne Releases of Principal Radionuclides Expected to Occur from Decontaminationof Surfaces in Reactor Building
under Normal Conditions
Total Release(Ci)
2 to 750

Concentrations(PCi/mL)1 x 10- 8 toCs-1378.1.x 10-s

7.4 x 10-14

Cs-134

1.4 x 10-5

1.3 x 10-14

Sr-90

5.0 x 10-6

4.5 x 10-1s

Radionuclidea
H-3

4 x 10-6

aother fission products are present; however,on thebasis of analysis of dissolved radionuclides in thereactor building sump water, for which Sr-89, Sr-90,
Cs-134, and Cs-137 constitute greater than 99 percent of the MPC contribution (see Appendix J), andwhich has a radionuclide distribution similar to thatof the AFHB sump water, the contributions of each of
the other radionuclides to airborne releases are
expected to be less than 0.1 percent and have not,therefore, been listed.

5-44
These estimates are regarded as upper bounds.
If a more realistic estimate, based on the use ofa penetration factor of 0.001 for each of the two filters in series (i.e., a total penetration
factor of 10-6) were used, all numbers in Table 5.10 would be reduced by a factor of 1000.
The staff estimates the amount of plateout on readily accessible surfaces on the 305-ft andhigher elevations of the reactor building to be 2 Ci.
An uncertain additional amount, assumed by
the staff to be less (probably much less) than 100 Ci, is present in less accessible locations,such as in floor drains and in the basement (counting only the contaminated film that will remain
after the sludge has been removed) and inside the D rings.
This plateout will be dissolved orsuspended in the decontamination liquids and removed with them.
For computational purposes, thestaff conservatively assumes that about 0.1 percent of the surface plateout on the 305-ft andhigher elevations may become airborne during decontamination operations (primarily during water
jet operations).
The fraction of contamination from less accessible locations on the 305-ft
elevation and above that becomes airborne will be less; the staff has made the conservative
assumption that an equal contribution to airborn contamination might occur from these sources.
Conditions in the basement are largely unknown.
In view of the relatively small volume of thebasement (less than 5 percent of the total volume of the reactor building) and area (about 20 percent of the total area), the staff assumes that the contribution to airborne contamination fromthe basement area will not exceed the contribution from the 305-ft and higher elevations.
The sludge, which has been estimated to have a wet volume of about 100 ft 3 and a radioactive
content of about 160 Ci (see subsection in Sec. 5.2.2.1 on alternatives for removing the sludge),
will be removed primarily by resuspension in the sump water before it is pumped out.
The
contribution of this operation to airborne releases will be negligible compared to other contributions.
It may be necessary to dislodge some fraction, up to about 5 percent of the total, by useof a water jet. The staff again has conservatively assumed that up to 0.1 percent of this residue
might become airborne during this operation.
5.2.4.2

Accident Scenarios

The types of accidents that could occur during reactor building decontamination include a liquid
spill, either before or after processing the liquid, dropping of a solid waste package, a fire ina barrel of contaminated trash, and failure of HEPA filters in the ventilation system.
Effluents
and releases from accidents associated with processing the liquids from the reactor building
decontamination are discussed in Section 7.
Effluents and releases from accidents in the handlingof solid wastes are discussed in Section 8.
A failure of one of the HEPA filters in the ventilation system would increase the airborne particulate release described in Table 5.10 until the ventilation system is secured (15 minutes).
Based on the analysis of Section 7, the accident releasesare shown in Table 5.11.

Table 5.11.
Estimated Airborne Releases from a HEPA
Filter Failure during Decontamination of theReactor Building Surfaces

Radionuclidea

Total Release(Ci)

Concentrations(pCi/mL)

Cs-1378.1 x 10-5

3.8 x 10-9

Cs-134

1.4 x 10-5

6.6 x 10-1o

Sr-90

5.0 x 10-6

2.4 x 10-1oaother fission products are present;

however,on the basis ofanalysis of dissolved radionuclides in the reactor building sump
water, for which Sr-89, Sr-90, Cs-134, and Cs-137 constitutegreater than 99 percent of the MPC contribution (see Appendix J),
and which has a radionuclide distribution similar.to that of the
AFHB sump water, the contributions of each of the other radionuclides to airborne releases are expected to be less than 0.1 percent and have not, therefore, been listed.

5-45
5.2.5
5.2.5.1

Environmental

Impacts

Occupational Doses

The same two methods used to estimate the work effort were used to estimate the occupational dosethat will be incurred by workers who must enter the reactor building to carry out decontamination
operations:(1) by extrapolation from experience in the AFHB, and (2) by a detailed analysis of
the doses expected from individual decontamination tasks.
The first of these methods gave thehigher work effort estimate.
For the occupational dose the reverse was true; the first method
gives the lower occupational dose estimate.
The staff infers that this result may be due to more
careful attention by the licensee to the use of strategies that would minimize worker exposureand less careful attention to efficient deployment of the work force.
Experience in the AFHB up to September 1, 1980, gave a cumulative personnel dose per unit area of
surface decontaminated of 0.8 person-rem/ft 2 .
Some of the more difficult cubicles with higher
radiation fields (e.g., the sump cubicle) have been left until last; the staff estimates thatthese cubicles may increase the dose to as much as 1.1 person-rem/ft 2 by the time the AFHB decontamination is completed (see Sec. 5.1.5.1).
This dose rate is based on the estimate that 3/4 hourof support work inside the AFHB was required for each hour of decontamination work.
The staff
estimates that about 2.5 hours of support work will be required for each hour of decontamination2
This would increase the dose per unit area from 1.1 person-rem/ft
work in the reactor building.
to 2.2 person-rem/ft 2 . The initial radiation fields in the AFHB were higher than the initialradiation fields in the reactor building on the 305-ft and higher elevations; however, in view of
the unknown (and probably much higher) radiation fields to which workers will be exposed duringdecontamination operations in the reactor building basement, the staff assumes that the AFHB andreactor building radiation levels are, on the average, comparable.
On the basis of these considerations, the estimated occupational dose for decontamination and related work inside the reactor
building is 660 person-rem.
A detailed analysis of the occupational dose for the individual tasks involved in decontaminating
the reactor building leads to an initial estimate of 22,000 person-rem for the total occupationaldose for decontaminating the reactor building.
This is an initial bounding estimate, and for two
reasons, it is much higher than the actual expected dose.
First, the work effort estimates arevery conservative.
Estimates of the work effort required for semiremote wash, scrubbing, and wet
vacuuming operations on the 305-ft and higher elevations by alternative and probably more realistic methods gave much smaller work-effort and occupational dose estimates for these tasks. On
the basis of these considerations, the staff believes that 15,000 person-rem would be a more
realistic upper bound even if no credit was allowed for the use of effective shielding and work
location strategies.
A considerable reduction in occupational dose can, however, be gained by such strategies as theuse of shielding and careful planning to choose work stations where the exposure is minimum
during the decontamination operations.
Experience in the AFHB has shown that the actual exposure
is less than what would be calculated from the product of the radiation field and work duration
alone by a factor of 8 for decontamination work, and in some cases as much as 100 for construction and maintenance work.' 8
Arguments can be given to support either the assumption that thegain will be larger for the open spaces in the reactor building or that it will be smaller; thestaff has made the conservative assumption that it will be smaller and assumes that an overallfactor of 5 is reasonable.
On the basis of the preceding considerations, the staff believes that 3000 person-rem is a probable upper bound for the cumulative occupational dose that will be incurred in decontaminating
the reactor building.*
Bounding values for the estimated cumulative occupational dose are,therefore, 660 and 3000 person-rem.

*If a factor of 3 were used to compensate for the conservatism in the dose estimates for theindividual tasks (which is consistent with the alternative decontamination task assessments), anda factor of 10 were used to take into account the gain by use of a well-planned decontamination
strategy, the estimated occupational dose would be reduced to 733 person-rem, which is close to
the estimated lower bound obtained by extrapolation from the AFHB experience.

5-46
The occupational dose includes an estimated 20 to 60 person-rem due to inhalation of tritiatedwater vapor. The tritiated water will be produced by evaporation from the water in the basement,
which has a tritium concentration of about 1 pCi/mL. The estimate for the dose from inhalation
is based on the expectation that the water will remain in the basement for about one year, andthat during this time, 50,000 to 150,000 person-hours of work effort will be spent in areas inwhich the tritium concentratin from tritiated water vapor will be about 2 x 10-6 pCi/mL.*
The estimated bounds for the work effort and duration of the decontamination operations were
The staff
300,000 to 900,000 person-hours and 18 to 54 months, respectively (see Sec. 5.2.3.2).
assumes that these two sets of numbers correspond; i.e., an average rate of expenditure of effort
of 50,000 person-hours per quarter for a period of 6 to 18 quarters is assumed.
Based on the cumulative occupational doses and work effort given above, the staff estimates thatthe average dose rate for individual workers lies in the range from 0.73 mrem/hr to 10 mrem/hr.
At the lower rate, a total cumulative dose of 660 person-rem would be incurred with a work effort
of 900,000 person-hours over a period of 18 quarters, for which the worker contingent would be
100 workers if a 40-hour work-week were maintained.
At the higher dose rate (10 mrem/hr) a total
cumulative dose of 3000 person-rem would be incurred with a work effort of 300,000 person-hours
over a period of six quarters, for which a contingent of 500 workers limited to 100 hours perquarter would be needed.
The expected number of additional cancer mortalities in this work force will range from 0.09
to 0.4. The average added probability that a worker in this group would eventually die of cancer
would range from about one in 1100 to one in 1300 if he/she participated in the reactor building
decontamination work for the full duration of the cleanup effort.
The expected number of additional genetic effects in the offspring of these exposed workers ranges from 0.2 to 0.8.
5.2.5.2

Offsite Doses

The largest offsite dose that is a consequence of decontamination operations in the reactor
building is that which occurred during the period from June 28 to July 11, 1980, when the Kr-85
was purged from the reactor building atmosphere.
The point of maximum exposure during the purge was at a location about 0.4 mile from the site inan east-southeast direction.
If a person had remained at this location throughout the purge
he/she would have received a beta skin dose of 4.5 mrem and a whole-body gamma dose of 0.05'mrem.
Dose estimates for decontamination operations in the reactor building subsequent to September 1,
1980 are based on the source terms developed in Section 5.2.4.1 and .listed in Table 5.10.
The
calculational models used to make these estimates and the interpretation of their results aredescribed in Appendix W. The significance of these doses and their human health and environmental consequences are discussed in Section 10.3.
The dose estimates are listed in Table 5.12
for the lower tritium range in Table 5.10, and Table 5.13 is for the upper tritium range inTable 5.10.
The 50-mile whole-body cumulative population dose received by the human population
during these activities is estimated to be 0.02 to 6 person-rem.
The lower value is for the
lower tritium range in Table 5.10, and the upper value is for the upper tritium range inTable 5.10.
5.2.5.3

Postulated Accident Effects

No accidental releases occurred during purging of the reactor building atmosphere.
The type of
accident for which dose estimates are made here is rupture of a HEPA filter during reactor building decontamination.
This accident scenario is described in Section 5.2.4.2 and the source terms
are listed for it in Table 5.11.
Doses are estimated here for a HEPA filter failure occuring
during the surface decontamination operation in the reactor building.
The calculational models
used to make these estimates and the interpretation of their results are described in Appendix W.
The significance of these doses is discussed in Section 10.4. The dose estimates to the maximum
exposed individual are listed in Table 5.14.
*In order to maintain the average atmospheric concentration at this level or below, it will benecessary to arrange barriers and ventilation patterns so that the fraction of water vapor in
the reactor building atmosphere that comes from the water in the basement does not exceed about
10 percent of the saturation concentration at 751F.

5-47

Table 5.12.
Dose.Estimates for the Maximum Exposed Individualfor Reactor Containment Building Decontamination
for Lower Estimated Tritium Releasea
Dose (mrem)b
Location

Pathway

Total-Body

Bone

Liver

Nearestgardenc

Inhalation
Ground Shine
Vegetable Use

1.4 x 10-4
9.5 x 10-6
6.7 x 10-4

3.1 x 10-5
9.5 x 10-6
8.0 x 10-4

1.4 x 10-4
9.5 x 10-6
6.9 x 10-4

TOTAL

8.2 x 10-4

8.4 x 10-4

8.4 x 10-4

Inhalation
Ground Shine
Goat Milk Use

7.3 x 10-5
9.1 x 10-6
3.0 x 10-4

1.2 x 10-5
9.1 x 10-6
2.8 x 10-4

7.5 x 10-s
9.1 x 10-6
5.7 x 10-4

TOTAL

3.8 x 10-4

3.0 x 10-4

6.5 x 10-4

Inhalation
Ground shine
Vegetable Use
Cow Milk Use

1.6 x 10-4
1.4 x 10- 5
8.0 x 10-4
2.6 x 10-4

3.5 x
1.4 x
1.2 x
2.8 x

10-5
10-s
I0-3
10-4

1.6 x 10-4
1.4 x 10-5
8.4 x 10-4
4.6 x 10-4

TOTAL

1.2 x 10-3

1.5 x 10-3

1.5 x 10-3

Nearestmilk goat

Nearest cow
and garden

aDose estimates in this table are based on the lower range of tritium inTable 5.10.
bDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin.
The maximum three organ doses are listed in thistable.
Doses were calcuated for four age groups: adults, teenagers,
children, and infants.
The highest dose estimates for each age group are
listed.
The dose estimates for the nearest gardenlare for children thedose estimates for the nearest milk goat for total-body are for adults andfor bone and liver are for infants, and the dose estimates for the nearest
cow and garden are for children.
cThe basis for selecting the special locations is described in Appendix W.
The actual locations are:nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

5-48

Table 5.13.
Dose Estimates for the Maximum Exposed Individualfor Reactor Containment Building Decontaminationfor Higher Estimated Tritium Releasea
Dose (mrem)b
Location

Pathway

Total-Body

Bone

Liver

Nearestgardenc

Inhalation
Ground Shine
Vegetable Use

5.2 x 10- 2
9.5 x 10-6
1.8 x 10-1

3.1 x 10-s
9.5 x 10-6
8.0 x 10-4

5.2 x 10-2
9.5 x 10-6
1.8 x 10-1

TOTAL

2.3 x 10-1

8.4 x 10-4

2.4 x 10-1

Inhalation
Ground Shine
Goat Milk Use

2.7 x 10-2
9.1 x 10-6
2.0 x 10-1

1.2 x 10-s
9.1 x 10-6
1.1 x 10-3

2.7 x 10-2
9.1 x 10-6
2.1 x 10-1

TOTAL

2.3 x 10-1

1.1 x 10-3

2.4 x 10-1

2

Nearestmilk goat

Nearest cow
and garden

Inhalation
Ground shine
Vegetable Use
Cow Milk Use

5.7 x 101.4 x 10-s
2.0 x 10-1
8.0 x 10-2

3.5 x 10-5
1.4 x 10-5
1.2 x 10-3
2.8 x 10-4

5.7 x
1.4 x
2.0 x
8.1 x

TOTAL

3.4 x 10-1

1.5 x 10'-

3.4 x 10-1

10-2

10-s
10-1
10-2

aDose estimates in this table are based on the upper range of tritium inTable 5.10.
bDoses were calculated for total-body,GI-tract, bone,

liver,

kidney,thyroid, lung, and skin.
The maximum three organ doses are listed in thistable.
Doses were calcuated for four age groups: adults, teenagers,
children, and infants.
The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden are for children thedose estimates for the nearest milk goat for total-body are for adults andfor bone and liver are for infants, and the dose estimates for the nearest
cow and garden are for children.
cThe basis for selecting the special locations is described in Appendix W.
The actual locations are:nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

5-49

Table 5.14.
Estimates of Offsite Doses to the Maximum
Exposed Individual Caused by Failure of a HEPA Filter
during Decontamination of the Reactor Building
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

4.2 x 10-5
5.5 x 10-4
1.1 x 10-2

5.4 x 10-4
5.5 x 1 0 -4
4.7 x 10-2

7.5 x 10-s
5.5 x 10-4
1.2 x 10-2

TOTAL

1.2 x 10-2

4.8 x 10-2

1.3 x 10-2

Inhalation
Ground Shine
Goat Milk Use

6.9 x 10-s
5.5 x 10-4
9.7 x 10-3

2.3 x 10-4
5.5 x 10-4
6.6 x 10-2

5.5 x 10-s
5.5 x 10-4
7.6 x 10-2

TOTAL

1.0 x 10-2

6.7 x 10-2

7.7 x 10-2

Inhalation
Ground shine
Vegetable Use
Cow Milk Use

2.9 x 10-s
5.5 x 10-4
1.1 x 10-2
1.9 x I0-3

3.7 x
5.5 x
4.7 x
1.1 x

10-4
10-4
10-2
10-2

5.1 x 10-5
5.5 x 10-4
1.2 x 10-2
9.7 x 10-3

TOTAL

1.3 x 10-2

5.9 x 10-2

2.2 x 10-2

Nearestmilk goat

Nearest cow
and garden

aDoses were calculated for total-body, GI-tract, bone,

liver, kidney,

thyroid, lung, and skin.
The maximum three organ doses are listed in thisDoses were calcuated for four age groups:
table.
adults, teenagers,
children, and infants.
The dose estimates for the nearest garden and forthe nearest cow and garden locations are for c-hildren, and the dose estimates for the nearest milk goat location are for adults for total-body andfor infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are:nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

5.2.5.4

Psychological-Socioeconomic

Effects

As indicated in Section 5.2.3.2, the staff estimates that from 300,000 to 900,000 person-hours
would be required to complete the reactor building decontamination process.
Assuming a 4.5-year
schedule--winter of 1982 to spring of 1986--the number of workers employed at any given time forwork on decontaminating the reactor building is estimated to average 100, with a peak of 200 in
1983-1984.
Although the manpower requirement for reactor building and AFHB decontamination are large relative to other cleanup operations, the community level impacts will not be significant. Most
workers, particularly those involved in the actual decontamination work, will be subject to highjob turnover rates because of occupational dose limitations.
These relatively transient workers
will not move into the community with their families, but will also seek to "double-up" in hotel,
motel, and rooming house accommodations.
Moreover, because of their transient status, such
workers will tend to spend money only for necessities and recreation, retaining most of theirincomes for their families and for spending in their hometown areas.
Finally, relocating workers
will seek to minimize housing and transportation costs by selecting locations that offer acceptable amenities.
For most workers, the greater Harrisburg area will provide adequate housing
choice and community amenity levels.
Because of the small size of worker housing demand relative
to the population of the Harrisburg area, the staff judgment is that in-moving workers will not
impose a significant impact on the housing market or on community facilities.

5-50
Considering the procedures to be undertaken and the offsite releases under normal operations andunder hypothesized accident conditions, the staff believes that the possible psychological andsocioeconomic impacts during reactor building decontamination would be similar to those describedin Section 5.1.5.4.
5.2.6

Economic Costs

In addition to the work effort described in Section 5.2.3.2, costs will be incurred for suchitems as:(1) construction and operation of the containment service building, (2) installationof essential services (electrical power, steam, air and localized ventilation), (3) refurbishmentof equipment (including the polar crane and building coolers), (4) materials (chemicals, shielding, venting, filters, etc.), (5) equipment (jib cranes hoisting and rigging gear, hydrolasers,
casks, steam cleaners etc.), (6) technical services to prepare the detailed procedures for implementation of the decontamination, and (7) training of personnel.
Numerous alternatives have been discussed, and in reality, the reactor building cleanup probably
will entail, a combination of these alternatives and can therefore be estimated as a range ofcosts.
For the purpose of estimating costs, the staff has examined two cases:(1) a case inwhich extensive shielding is required and greater than expected difficulty is encountered inremoving the contamination from surfaces and equipment (worst case) and (2) a case in which
conditions are essentially as expected and supported by recent surveys (best case).
The staff estimates that the costs for reactor building decontamination will range from
$25 million for the best case to over $60 million for the worst case.
The principle catagoriesof these costs are listed in Table 5.15.
The assumptions used to make these estimates are given
in Appendix K.

Table 5.15.
Economic Costs for Reactor
Building Decontamination(thousands of.dollars)
ItemBest Case

Worst CaseLabor

13,40040,100Equipment

3,500

5,900

Facilities

8,500

17,000TOTAL

25,40063,000

References--Section 5
1.

Regulatory Guide 1.86, U.S.

Nuclear Regulatory Commission.

2.

Code of Federal Regulations,

Title 10,

3.

Memorandum from B.J. Bouchard, Vikem Industries, Inc., to Thomas Block, Metropolitan Edison
Company, relative to status of decontamination of AFHB, May 5, 1980.

4.

Memorandum from B.A. Irving and V.F.
GPU/Met Ed, October 27, 1979.

5.

T.R. Block, "Auxiliary and Fuel Handling Building Decontamination Program," presented at an
information briefing by Metropolitan Edison Co. at Crawford Station, Middletown, PA,
September 23, 1980.

Part 20, Office of the Federal Register.

Bouchard,

Vikem Industries,

Inc.,

to J.

Renshaw,

5-51
6.

W.J. Seefeldt, W.J. Mecham, and M.J. Steindler, "Characterization of Particulate Plutonium
Released in Fuel Cycle Operations," Argonne National Laboratory, ANL-75-78, p.71, May 1976.

7.

"Technology for Commercial Radioactive Waste Management,"
DOE/ET-0028, Vol. 2, p. 4.11.11, Table 4.11.8, May 1979.

8.

Memorandum and Order (CLI-80-25) issued by the U.S. Nuclear Regulatory Commission on June
12, 1980, in the matter of Metropolitan Edison Company et al. , Three Mile Island Nuclear
Station Unit 2.

9.

Order for Temporary Modification of Operating License No. DPR-73, issued by the U.S. Nuclear
Regulatory Commission on June 12, 1980, relative to the Technical Specifications for Three
Mile Island Unit 2.

10.

"Three Mile Island Unit 2 Recovery Quarterly Progress Report for the Period Ending September
30, 1980", Section 4 Metropolitan Edison Co./General Public Utilities, 1980.

11.

James Langenbach, "Preliminary Summary, Data Acquisition, TMI-2 Reactor Building, Entry No.
l,'July'23, 1980 and Entry No. 2, August 15, 1980," presented at an information briefing
meeting conducted by Metropolitan Edison Co. at Crawford Station, Middletown, PA, September
23, 1980.

12.

Memorandum from T.L. Gilbert to W.K.
December 8, 1980.

13.

"Planning Study for Containment Entry and Decontamination,"
1979.

14.

Letter (TLL 541) from R.C. Arnold,
Commission, November 4, 1980.

Met-Ed/GPU,

15.

Letter (TLL 684) from G.K. Hovey,
Commission, December 23, 1980.

Met-Ed/GPU,

16.

D.O. Campbell, E.D. Collins, L.J. King and J.B. Knaver, "Evaluation of the Submerged
Dimineralizer System (SDS) Flowsheet for Decontamination of High-Activity-Level Water at the
TMI-2 Nuclear Power Station," Oak Ridge National Laboratory, ORNL/TM-7448, July 1980.

17.

R.E. Brooksbank and W.J. Armento, "Post-Accident Cleanup of Radioactivity at the Three Mile
Island Nuclear Power Station," Oak Ridge National Laboratory, ORNL/TM-708, February 1980.

18.

Memorandum
1981.

from T.L. Gilbert

to J.E.

Lehto,

ArgonneRobinson,National

to H.R.
to L.H.

U.S. Department of Energy,

Laboratory,"Entry Data,"

BechtelPower Corp. , July 2,

Denton U.S.

Nuclear Regulatory

Barrett,

Argonne National

U.S.

Nuclear Regulatory

Laboratory,

January 12,

6.

REACTOR DEFUELING AND PRIMARY SYSTEM DECONTAMINATIONThe operations leading to and incl-iing removal of the damaged reactor core from the pressure
vessel and subsequent decontaminatn of the reactor primary cooling system (RCS) components(reactor pressure vessel, steam generators, pressurizer, pumps and associated piping) are considered in this section. The activities discussed are:-Handling of the radioactive water in the primary cooling system during reactor defueling andprimary system decontamination (Sec. 6.1).
The water processing and disposal alternatives
are discussed in Section 7.

-

Inspection of reactor system components prior to defueling and decontamination to determine
the condition of the primary cooling system and reactor core (Sec. 6.2).

-

Removalof the reactor pressure vessel (RPV)

the reactor core (Sec.

head and internal components to gain access to

6.3).

-

Removal of the reactor core fuel elements (Sec.

-Decontamination of the primary system components to remove fuel debris and other particulatematter from the system and to reduce fission product plateout on internal surfaces (Sec. 6.5).

6.1

6.4).

HANDLING OF PRIMARY COOLANT SYSTEM WATER

Processing and treatment of the primary water must precede and continue during the operations
leading to reactor defueling and primary system decontamination (see Sec. 7 for alternatives anddetails).
Following the initial decontamination of the primary water, inventories will varydepending upon the particular phase of the defueling and decontamination activities.
Appendix F
discusses potential for water reuse to minimize increasing the total inventory of contaminatedwater.
The various phases of water handling for reactor defueling and primary system decontaminationare:Preparations for reactor head removal
Removal of RPV head and internals
Reactor defuelingPrimary system flush and drain
Decontamination of primary system components.
6.1.1

Preparation for Reactor Head Removal

Following initial primary water processing (see Sec. 7.1.5) the primary system would be drained
from an initial inventory of 96,000 gallons to about 26,000 gallons.
Primary system circulation
would be maintained by existing TMI-2 piping systems.
The primary water chloride concentration
would be reduced to acceptable levels prior to opening the reactor coolant system to air.
6.1.2

Removal of RPV Head and Internals

Either after or during reactor pressure vessel (RPV) head removal, the fuel transfer canal (FTC)
would be filled with primary-system-grade water to the normal refueling level.
Filling the FTC
would add about 360,000 gallons of water.
Circulation of this water would be maintained byexisting TMI-2 systems.
Should the radionuclide concentration increase above the target value of

6-1

6-2
0.01 pCi/mL (exclusive of tritium),

additional water processing would be conducted using either

one of the alternatives described in Section 7.1.2 or the existing purification
supplemental purification system.
6.1.3

systems or a

Reactor DefuelingAfter removal of the upper internals (see Sec. 6.2) the spent fuel pool (SFP) would be filled, if
not already filled, and connected to the FTC.
The SFP contains about 700,000 gallons of water.
This would increase the total inventory of water to about 1,100,000 gallons.
Circulation of thiswater would be maintained using existing TMI-2 systems.
Should the radionuclide concentration
increase above the targeted value of 0.01 pCi/mL (exclusive of tritium), additional water processing would be conducted using either one of the alternatives described in Section 7.1.2.
6.1.4

Primary System Flush and Drain

After completion of reactor defueling, the SFP will be isolated from the FTC and the latter will
be drained.
At this point, the primary system water inventory will have been reduced to about
26,000 gallons.
The RPV head then will be installed and the primary system will be filled withwater to bring the inventory up to 96,000 gallons.
The flushing and draining operations thenwill be initiated (see Sec. 6.5).
The staff assumes that during the flush and drain operations,an additional 150,000 gallons of water will be added as makeup water during the flushing operations, to provide about 2½ system volumes.
Processing of this water is covered in Section 7.
6.1.5

Decontamination of the Primary System Components

After completion of primary system flush and drain, the primary system will once again be filled(96,000 gallons) and chemical decontamination will be initiated.
Depending upon the process
selected, between 100,000 and 500,000 gallons of water will be required.
The processing of thiswater is discussed in Section 7.1.3.
The physical operations and alternatives are discussed in
Section 6.5.
6.2

REACTOR COOLANT SYSTEM (RCS)INSPECTION

Since current knowledge about the condition of the TMI-2 reactor core and reactor coolant system
is limited, it is necessary for workers to enter the reactor building to inspect components
before plans can be finalized for primary system decontamination and reactor defueling.
Inspections will be coordinated with the reactor building decontamination (Sec. 5.2), removal of the
RPV head (Sec. 6.3), defueling (Secs. 6.3 and 6.4), and primary coolant system decontamination(Sec. 6.5).
The objectives of these inspections would be to determine the condition of the coreand provide ongoing monitoring of important reactor coolant system parameters during the succeeding
steps in the dismantling and cleanup process.
6.2.1

Status and Specific Considerations

The data available concerning the status of the TMI-2 reactor do not provide sufficient information to determine the extent of damage in the reactor core.
The best information on the possible
internal condition of the reactor was derived from attempts to reconstruct the progress of theaccident, based on data gathered during the accident. 1 ' 2 This information leads to the conclusion that a large fraction of the fuel rods have ruptured and there has been oxidation of Zircaloyin the core (about 50% of the total core inventory of Zircaloy, i.e., fuel cladding, control-rod
guide tubes, and instrument tubes, has oxidized).
There has been structural damage in the core as deduced from fission product release, hydrogengeneration, core thermocouple readings, core pressure drop, and a significant system pressuretransient.
The reactor pressure vessel, internals, and head have been exposed to high temperatures, as indicated on incore thermocouples and deduced from the behavior of the neutron detectors.
The high temperatures may have caused fragmentation of some of the fuel pellets.
Thisfragmented fuel may~have been distributed throughout the core region and could obstruct the viewneeded for visual inspection and defueling.
The temperature of parts of the core during theaccident may have been sufficiently high to have caused some melting and fusing of steel end
fittings and spacer grids on the fuel assemblies.
There is, however, no indication that significant melting of the uranium oxide fuel pellets has occurred.
Care must be taken during in-vessel
activities to avoid unnecessary scattering of damaged fuel.
There may have been some hydridingof the Zircaloy in the core; however, large-scale hydriding of the Zircaloy in the core is not6-3
expected, and any zirconium hydrides that did form should not be present as finely divided particles.
(Finely divided zirconium hydride is known to be pyrophoric and could thus become a hazard
if present during the cleanup operation.)
Thus, a pyrophoric hazard is highly unlikely.
In anycase the core will be kept constantly covered with water throughout the defueling so that even if
significant hydriding exists, pyrophoric events are not possible.
6.2.2

Alternative Methods Considered

The number of inspections required will depend on the condition of the core.
Since the environmental impact of the examinations is expected to be small relative to .the impact of the defueling/
decontamination operations, and since any attempt to consider different examination methods fordifferent probable core conditions would generate an unmanageable number of combinations, thefollowing discussion is limited to a single examination schedule for each of two core-condition
possibilities--the best case and the worst case.
The best case is an estimate of the most favorable core conditions (for which a smaller number of examinations would be needed).
The worst
case is for a severely damaged core (for which a larger number of examinations would be needed).
The planning and damage-assessment inspections will be conducted during all stages of the decontamination/defueling operation.
The principal stages are:1.

Entry into the reactor building

2.

Preparations for reactor vessel head removal

3.

Removal of the reactor vessel head

4.

Removal of the upper internals from the reactor vessel

5.

Removal of the fuel from the reactor vessel to the spent fuel pool (this operation will
involve several stages, such as fuel encapsulation).

No inspections of the core have yet been made.
Because of the requirement to satisfy ALARA considerations, the work must be conducted in as low
a radiation field as is practical.
Decontamination of the reactor building should therefore beunderway before the first significant inspection activities.
6.2.3

Details of Methods and Facilities

Standard techniques (e.g., gamma scans, visual inspections, dimensional measurements, chemical
analysis) will be used for the examination and measurement tasks.
The only details consideredhere are those needed to estimate (for best- and worst-case conditions) occupational radiation
doses,* the amounts of contaminated waste that would be generated offsite, environmental effects,
and direct economic costs.
The estimates given are based on the judgment of persons experienced
in reactor design, operations, and maintenance.
Examinations required for requalification ofreactor coolant system components are excluded.
6.2.3.1

Best-Case Conditions

The bases used in making estimates for best-case circumstances are:(1) the reactor building
would be decontaminated and hot spots would be shielded (see Sec. 5.2); (2) only a minimum numberof examinations would be required; and (3) the workers entering the reactor building would be
required to wear protective clothing and use respirators at least part of the time.
6.2.3.2

Worst-Case Conditions

The bases used in making estimates for worst-case conditions are:(1) during the period afterreactor building entry but before opening the primary system, reactor building decontamination
would be in progress and not all hot spots would be shielded; (2) for examinations after opening*Radiation values given are average values for all workers.
These values were calculated byevaluating the work to be performed, where the workers would be, how long the work might take,conducting scoping calculations for hot spots, and then using engineering judgment to evaluate
the average radiation field.

6-4
the reactor coolant system, the reactor building decontamination would be essentially complete
and hot spots would be shielded; and (3) the workers entering the reactor building would be
required to wear protective clothing and use respirators (full or half-masks) at least part of
the time.
6.2.3.3 Logistics, Crew Size, Work Effort, Radiation Levels, Waste Generated, and AssociatedFacilities
Estimates for the relevant factors are given in Table 6.1 and discussed below for the two cases.
Crew Size.
The average work crew under best-case conditions would consist of three
persons--an instrumentation and control technician, a health physics technician, and a
specialist/supervisor.
The average work crew under worst-case conditions would consistof four people: two instrumentation and control technicians, a health physics technician, and a specialist/supervisor.
The crew size would depend on the type of activity
and on the location of the effort to be performed.
Inspection activities would be
performed on an intermittent basis beginning with reactor building entry and continuing
through defueling and RCS decontamination activities.
Work Effort. The probable work effort required for reactor primary system inspection
under this alternative is 5200 person-hours under assumed best-case conditions and41,000 person-hours under worst-case conditions (see Table 6.1).
Inspections andexaminations associated with activities supporting control-rod drive uncoupling, incoredetector removal or installation, gas sampling, reactor internal and external evaluations, core evaluation, and fuel removal would comprise about 85 percent of the effort
under best-case conditions and 80 percent under worst-case conditions.

Table 6.1. Average Dose Rates, Work Times, and Cumulative Occupational Doses for Reactor
Coolant System Inspection Operations under Best- and Worst-Case Conditions

Operation Stage

Average
Dose Rate(mrem/hr)

Time in Radiation a
Areas (person-hr)

Cumulative
Occupational
Dose (person-rem)aBest-Case Conditions
1. Reactor building entry
2. Preparation for vessel head removal

10
10

1,700
.800

17
8

3. Primary system head removal

10

300

3

4. Plenum removal
5. Fuel removal

10
10

500
1,900

5
19

5,200

52

TOTAL
Worst-Case Conditions
1. Reactor building entry
2. Preparation for vessel head removal
3. Primary system head removal

25
10
10

11,000
6,900
4,800

280
69
48

4. Plenum removal
5. Fuel removal

10
10

8,500
9,900

85
99

41,000580

TOTAL
avalues rounded to two significant figures.

6-5
Radiation Levels.
Under best-case conditions, the average dose rate associated withprimary system inspection is estimated to be 10 mrem/hr.*
It is estimated that the
average worker dose rate prior to reactor coolant system breach under worst-case conditions would be 25 mrem/hr because this activity would be accomplished before reactor
building decontamination; all examinations made after the primary system breach underworst-case conditions would be conducted after completion of essential reactor building
decontamination, so a 10 mrem/hr average dose rate is estimated.*
Wastes Generated.
Wastes that would be generated include 52 incore detector assemblies, contaminated clothing, contaminated inspection equipment, and four out-of-core
nuclear detectors under best-case conditions, and eight out-of-core nuclear detectors
and 25 contaminated probes under worst-case conditions.
Special Facility Needs.
A closed circuit television system and communications system
would be needed, plus various types of probes for examinations (e.g., boroscopes,
miniature T.V. cameras, radiation detectors).
6.2.4

Effluents and Releases to the Environment

Virtually no radioactive effluents will be released as a result of inspection activities.
6.2.5

Environmental Impacts

6.2.5.1

Occupational Doses

The expected number of additional cancer mortalities in the work force exposed to the cumulative
occupational doses given in Table 6.1 would range from 0.0068 for the best case (52 person-rem)
to 0.076 for the worst case (580 person-rem).
The expected number of additional genetic effects
in the offspring of the work force exposed to these levels would range from 0.014 to 0.151.
Since the inspections will occur intermittently throughout the decontamination and defueling
operations, there will not be a separate work force for these inspections per se. The crews of
three to four people (Sec. 6.2.3.3) will ideally perform inspection-related tasks along withother core-removal and primary coolant system efforts since inspection and follow up are integrally related, and inspection requires specific engineering skills (Sec. 6.2.2). However, if
the inspections were done at the same level of effort during each quarter over a three-year
cleanup period, up to 3400 person-hours per quarter would be devoted to inspections under the
worst-case conditions.
Conversely, if the cleanup takes six years and best-case conditions
apply, about 200 person-hours per quarter would be required.
An individual quarterly (three-month) occupational dose of 1 rem has been selected by the staff
for calculation of total work force requirements.
The reasons the staff selected this value
rather than limits as high as 3 rem per quarter given in 10 CFR Part 20 are provided inAppendix L. However, individuals could approach a dose of 3 rem per quarter, if necessary.
With an average radiation field of 10 mR/hr, each person would spend only 100 hours per quarteron inspection efforts so as not to exceed 1 rem/quarter. This means that because of dose considerations, each crew of four would expend only 400 person-hours per quarter; thus, about nine suchcrews (36 individuals) would be needed to carry out the 3400 person-hours per quarter of inspection work that the staff anticipates would be needed under worst-case conditions during a threeyear cleanup period. However, under best-case conditions for inspection, a single crew of four
could perform the inspections and perform additional defueling work as well.
6.2.5.2

Offsite Doses

Offsite impacts are indistinguishable compared to defueling (Section 6.4.5.3).

*See Appendix I for justification.

6-6
6.2.5.3

Postulated Accident Effects

The staff has not identified any potential accidents from this activity that would have offsite
effects.
6.2.5.4

Psychological-Socioeconomic

Psychological-socioeconomic
6.2.6

Effects

effects relative to RCS inspection are discussed in Section 6.5.5.4.

Economic Costs

To evaluate the economic costs associated with a particular case, all labor costs, as well as
costs for equipment and facilities, need to be included.
Estimates of the economic costs for the
best-and worst-case conditions are given in Table 6.2.
These costs are estimated on a consistent, relative basis, and do not necessarily reflect all costs on an absolute basis.
Thus, thesecosts should be considered only as relative costs for the two scenarios (best case versus worstcase). The basis for the costs shown here are given in Appendix K.

Table 6.2 Economic Costs for RCS
Inspection (thousands of dollars)

ItemBest Case

Worst Case

Labor

1230

4910

Equipment
Facilities
TOTAL

160
0
1390

980
0
5890

About 85% of the labor costs shown in Table 6.2 are estimated to be incurred in preparation forwork inside the reactor building. The majority of this effort consists of training and procedure
preparation.
Estimated equipment costs consist primarily of special instruments for the necessary
evaulations at various stages of plant cleanup.
Instrumentation needs will be better defined as
detailed planning gets underway and, in some cases, as work is performed. However, the costestimates are not particularly sensitive to these needs and should not be significantly changed
regardless of specific instrument needs. Facility needs are as discussed under Special Facility
Needs in Section 6.2.3.3.
6.3

REMOVAL OF THE RPV HEAD AND INTERNALS

Removal of the reactor pressure vessel head (RPVH) and the internal components is necessary togain access to the reactor core for defueling, and to the bottom of the reactor vessel to remove
fuel debris.
Before the RPVH is removed, other decontamination activities directly related to RPVH removal
will have been completed:(1) the primary water will have been decontaminated and the chloride
concentration reduced to a level sufficient to proceed with RPVH removal, (2) the principal
reactor building support equipment and systems needed to proceed with head removal will have been
decontaminated and put into satisfactory working condition, and (3) work area decontamination
will have been accomplished in accordance with ALARA principles. Also, inspections of the core
may be accomplished by insertion of viewing devices through a port in the head, prior to headremoval.
6.3.1

Status and Specific Considerations

The reactor pressure vessel head (RPVH) is bolted to the top of the vessel. As shown in Figure 6.1, the control rod drives and the axial power shaping rod (APSR) drives extend through the6-7

CONTROL
DRIVESHEAD
REACTOR STRUCTURESERVICERoo

AND NUTS(20 TONS)
60 STUDSSEAL PLATE
PRESSURE
REACTORHEAD
VESSEL

(81 TONS),

PLENUMUPPER
ASSEMBLY )(55 TONSGRID

CORE SUPPORTSTRUCTURE(112.5 TONS)
3 6FEET

LOWER

FLOW

GRID

DISTRIBUTOR

V'l
Figure 6.1.

a Typical
View of
Cutaway Pressure Vessel.
ReactorPWR Similar

TMI-2
to the6-8
RPVH.
Before any core damage had occurred during the accident, the control rod drives "scrammed"
the control rods to the fully inserted position so as to shut down the reactor. However, the
APSRs were maintained at a partially withdrawn position.
Because of the accident damage, thecontrol rod and APSR drives may now be inoperable.
There may be fuel and fuel assembly debris and fragments in cavities, crevices, and on surfaces
on the upper plenum assembly and the RPVH.
There may also be fuel and fuel assembly debris andfragments in the lower grid and flow distributor of the core support assembly and throughout thereactor coolant system.
High temperatures and rapidly changing pressures resulting from theaccident may have caused:distortion of the RPVH; distortion, warping, and physical dislocationof the upper plenum assembly; binding of the upper plenum assembly to the core support assembly,
fuel assemblies, and the control rod assemblies; distortion and warping of the core support
assembly and binding of core support assembly to RPV.
The staff has recognized that the actual conditions and extent of damage in the RPV are not known
and could cover a broad spectrum.
Conditions and damage cannot be accurately determined until
the RPVH has been removed.
For evaluational purposes, two sets of postulated conditions (worst
case and best case, as outlined below) have been developed.
The staff is of the opinion that thecourses of action developed below for each postulated set of conditions should encompass therange of anticipated conditions.
6.3.2

Alternative Methods Considered

Regardless of the exact conditions found inside the RPV, the RPVH and internals most likely would
be removed by the following seven steps (see Fig. 6.1 for identification of components).
The
core will be constantly kept covered with water throughout the operation.
1.

Preparation of Equipment.

2.

Installation of the seal plate assembly.
(The seal plate assembly is a large flat ringthat is fastened between the reactor vessel flange and the canal floor. When this ringis installed it forms a watertight seal allowing the fuel transfer canal (FTC) to be
flooded.
The fuel transfer canal must be flooded for shielding purposes in order toremove the fuel from the reactor vessel to the spent fuel pool or elsewhere.)

3.

Removal of the RPVH closure nuts and studs.
that secure the RPVH to the reactor vessel.)4.

Uncoupling of the APSR/CR drive assemblies.
(The APSR and CR drives are electromechanical devices for raising and lowering the reactor control elements.
Before removal of
the reactor head, these drives will be separated 'from the control element so that therods remain in the reactor core when the RPVH is lifted.)

5.

Removal of the RPVH.
(The RPVH is the cover that isreactor internals and fuel.)

6.

Removal of the upper plenum assembly.
(The upper plenum assembly is a large cylinder
in the reactor vessel for guiding the control rods and pressing down on the fuel assemblies to maintain proper alignment.)

7.

Removal of the core support structure.
(This will be done after defueling (seeSec. 6.4).
The core support structure is a large basket-like component in the reactor
vessel supporting the fuel elements and directing the entering reactor coolant to thebottom of the reactor vessel.)(The

closure nuts and studs are the bolts

removed to gain access to the

The best-case (most favorable) conditions are based upon the possibility that damage is largely
limited to the core (Ref. 1, Appendix CI, Sec. 8), with minor damage to the other RPV internal
components.
(The postaccident survival of most of the 52 core thermocouples provides this indication.)
The limited damage would not preclude the use of modified existing tooling and normal* or
near-normal methods.
Core debris attached to component surfaces could be removed manually byworkers using long-handled tools.
Debris trapped in component cavities could be removed and*The word "normal" refers to tooling or methods used in the preaccident mode.

6-9
handled manually using long-handled pliers, scoops, etc. Removal methods could be characterized
as nondestructive and would include any system, tooling, or other means now available to facilitate removal and mechanical disassembly of components when required for removal.
The removed components can be stored in the fuel transfer canal water or on the operating floor
behind a temporary shielding wall.
The worst-case (least favorable) conditions are based upon the belief that during the morning ofMarch 28, 1979 (Ref. 1, Appendix CI, Secs. 11 and 15 and Appendix Tl1, Sec. 2.1 summary).
theinternal RPV environment at that time may have been drying out, experiencing high temperatures,and being thermally shocked by water quenching.
Some of the core temperature measurements (Ref. 1,
Fig. CI-12) ranged up to the order of 2000OF to 2500 0 F. If RPV internal components reached thosetemperatures, partial local melting and sagging of the components could have resulted. Water
quenching at these temperatures also could have caused warping and deformation of the components.
Thus, it is assumed that: (1) some components to be removed could have incurred accident damage
that precludes the use of existing tooling and normal or near-normal methods; (2) component
removal may not be possible until core debris attached to the component surfaces or trapped incomponent cavities has been removed; and (3) the component removal may be difficult because of
deformation or warpage or because of self-welding to adjoining components.
Removal methods used
could be characterized as destructive and would include engineered equipment, such as: special
tooling to cut the component into fragments in order to remove trapped core debris or accomplish
removal; special tooling to machine out items, such as studs, that have become fixed; and special
tooling to force warped or deformed components into a configuration that could be removed.
Any parts of cut up components would be removed from the RPV and temporarily stored in the fueltransfer canal water until the fragments could be transported out of the canal in containers.
The handling of core debris from the RP -AP either the best or worst case requires special control
methods and separate temporary storage in the canal water until the debris can be transported outof the canal in containers.
The staff has determined that criticality control should be maintained by limiting the container size and shape.* Contamination control can be maintained bydesigning the container to allow free circulation of water through ports covered by fine screens,
if necessary.
6.3.3
6.3.3.1

Details of Alternative Methods and Facilities
Best-Case and Worst-Case Conditions for Steps in the Removal of the RPVH and Internals

The various operations required to remove the RPVH and internal
Table 6.3. A discussion of each-of these operations follows.
1.

RPV components are listed in

Preparation of Equipment

The preparations required are essentially the same for the best and worst case conditions. Thiseffort involves installing working platforms, underwater lights; and temporary radiation shielding.
Additionally, there will be some preliminary decontamination of the external surfaces of thereactor head service structure, disconnection of electrical cables and cooling water lines, andcheck out of APSR drive operation.
These operations may generate contaminated waste, such as RPVH insulation, rod drive cabling, andminor amounts of debris.
The radiation environment is expected to be variable, ranging from alow of 2 to 3 mR/hr to hot spot levels of perhaps about 2 R/hr (see Appendix I). Work near these
hot spots would consist of installation of shielding or possibly operations that would require
only a few minutes of worker exposure time. The staff has used an average dose rate of 10 mrem/hr
for these operations (see Appendix I).
2.

Installation of Seal Plate

The details for this step are common for both the best and worst conditions.
*Calculations of the minimum critical size for an optimum array of U02 fuel in nonborated water
give a value of 1.8 ft 3 for a spherical shape. A non-spherical shape with a larger volume would
also be safe. These calculations are described in Reference 3.

6-10

Table 6.3.

Work Effort, Crew Size, and Radiation Levels for Removal of RPVH and Internals

Operation

NormalWork Effort(person-hours)

Estimated
Work Efforta(person-hours)
Best Case
Worst Case

Estimated
Crew Size

Radiation
Field

1. Preparation of equipment
2. Installation of seal plate

150
40

6,000
400

6,000
400

15
5

<10 mR/hrc<20 mR/hrd

3.

Removal of RPVH studs & nuts

100

1,200

2,400

5

<100 mR/hr

4.

Uncoupling of APSR & CR
drives
RPVH lift
Removal of upper plenum

80
50
100

2,000
900
2,000

5,000
1,200
13,0004
5
6<20 mR/hr
10-100 mR/hr<10 mR/hr

Removal of core support
structure

100

2,500

17,0006<10 mR/hr

600-700

15,00045,000(averageof 6)'-10 mR/hr

5.
6.
7.

Total

aEstimated work effort in a radiation environment.
bSee Appendix I.
CRadiation environment is expected to be variable, ranging from a low of 2 to 3 mR/hr to hot
spot levels of perhaps about 2 R/hr. Work near these hot spots would consist of installationof shielding or possibly operations that would require only a few minutes of worker exposure
time.
dWith installation of shielding.

Normally the seal plate is used during short-term refueling operations.
For the proposed longterm postaccident defueling operations, the seal should be upgraded.
A seal-welded diaphragm or
improved sealant would be examples of the type of improvement anticipated.
Although initiallylittlewaste should be produced by this activity, eventually the seal plate diaphragm and sealingmaterial will require disposal.
The worker radiation environment is unknown at this time, but
the installation of shielding should reduce the field to 20 mR/hr or less.
3.

Removal of RPV Head Closure Nuts and Studs

Under normal conditions this step involves loosening the closure nuts from the studs after cleaning corrosion products from the threads and lubricating the threads by applying a tensioning
force to the stud and loosening and removing the nuts.
Under worst case conditions, however,loosening may require splitting or stripping jammed nuts and cutting off and machining out someof the difficult studs.
As shown in Table 6.3, there is considerable variation in the labor
hours for this step to allow for these potential difficulties.
4.

Uncoupling of APSR & CR Drives

In order to remove the RPV head it is necessary to uncouple the control rods from the drives,mounted on the head, so the rods will stay in the core when the head is lifted.
Under the best circumstances the control rod drives can be disconnected from the control rods
using normal or near normal procedures.
Under worst case conditions, however, the control rods
may be stuck in the reactor core or plenum grid, thus complicating their disconnection from thecontrol rod drives.
This may require cutting the drive housings and/or lead screws that attach
to the control rods.
As shown by the data in Table 6;3 there is considerable variation allowedin the labor hours for this step to allow for these contingencies.

6-11
5.

RPVH Lift

Prior to lifting the RPVH, a visual inspection of the core will be made to establish as far as
possible the extent of internal core damage, such as hybriding, and physical interference ofcomponents.
An underwater television camera inserted through a control rod port could possibly
be used for this inspection.
Under the best-case conditions, it is assumed that the APSR/CR drives have not been removed andthe RPVH with the drives attached can be lifted using the normal head lifting fixture and polar
crane. The fuel transfer canal is filled with water and the inside of the head is decontaminatedby flushing. Then the RPVH is placed on a storage stand behind a temporary shielding wall on theoperating floor.
Under the worst-case conditions the APSR/CR drives have been removed by cutting them off, sodummy control rod followers must be installed to make sure the control rods are not lifted with
the RPVH.
It is assumed that, to lift the head, a force exceeding a predetermined limit above
the weight of the RPVH would be required, so precision jacking equipment is installed to replace
the rigging and the head is jacked up until separated from the RPV. The head is then liftedusing the same procedure described above.
As shown in Table 6.3, a range of from 900 to 1200 person-hours has been allowed for this operation, which is considerably above the normal requirement of 50 person-hours. Waste produced fromthis activity for both cases could consist of minor amounts of fuel and clad debris. This material would be collected under water in specially designed filters.
Worker radiation fields would be variable and range from a few mR/hr for underwater operations toperhaps 100 mR/hr.
6.

Removal of Upper Plenum Assembly

Under the best-case conditions, the plenum assembly can be lifted using the internals-handling
fixture and polar crane.
Prior to being lifted, it may have to be jacked up until separated fromthe RPV guide keys.
Before the lift is attempted, a TV camera and a periscope may be installed
to visually inspect plenum surfaces and cavities for core debris or damage.
Core debris can beremoved by using water suction vacuum equipment, grapple, tongs, and water flushing techniques.
Dummy control rod followers can also be installed to hold down the APSR/CR assemblies during
plenum removal.
Under the worst-case conditions, the upper plenum assembly could be stuck in place, making it
necessary to cut it out in pieces, decontaminate the pieces, and transport them out of the containment building in shielded containers.
Debris would be placed in specially sized containersthat can be handled by the fuel transfer equipment.
As shown in Table 6.3, there is a range of 2000-13,000 person-hours allowed for these contingencies.
This compares with about 100 person-hours normally required for this operation.
Forworst-case conditions, the entire upper plenum assembly, about 110,000 pounds of stainless steel,is considered waste. It would be cut up, perhaps into as many as 150 pieces, and transported outof the reactor building in either shielded containers or through the fuel transfer ports. Worker
radiation fields should average less than 10 mR/hr.
7.

Removal of Core Support Structure

After the fuel has been removed from the reactor vessel, removal of the core support structure isaccomplished the same as outlined above for the removal of the upper plenum assembly, i.e. thecore support assembly is lifted using the internals lifting fixture and polar crane in the best
case, while it is cut up into pieces for removal under the worst-case conditions.
As shown in Table 6.3, removal of the core support structure could be the most labor-consumptive
step in dissassembling the reactor.
Because of uncertainties a range of 2500 to 17,000 personhours has been allowed, compared with 100 required under normal conditions.
Under worst-caseconditions, the core support structure, consisting of about 225,000 pounds of stainless steel, isconsidered waste and will be cut into perhaps about 200 pieces under water and transported out of
the reactor building in either shielded containers or through the fuel transfer ports.

6-12
6.3.3.2

Logistics, Crew Size, Work Effort,
Facilities

Crew size, work effort (person-hours),
cussed below for the two cases.

Radiation Levels, Waste Generated,

and Associatedand radiation fields are summarized in Table 6.3 and dis-

Crew Size
The crew size will vary depending upon the individual activity and the conditions encountered.
Some of the activities, such as preparation, can effectively use large crews--as many as 10 to
15 workers--since many of the preparation tasks can be accomplished simultaneously. Other activities, such as inspections, may require only two to three workers. After assessing the activity
and the work that would be accomplished,the staff judged that an average crew size of six workers
would be used. The difference between best- and worst-case conditions, therefore, would not be
the size of the crew, but the time and work effort required to accomplish a given task.
Work Effort
Person-hours of work effort have been estimated by breaking each case down into major work activities, then estimating the crew size needed to perform the work and the duration of the work (seeTable 6.3).
Extensive training will have been performed before work in the reactor building
actually begins.
The staff assumes that operations associated with the reactor coolant system, i.e., RPVH andinternals removal, will be conducted under conditions similar to a normal refueling. Workers
would typically be suited up in a double set of coveralls with boots and gloves. Some operations
might require respirators for protection against airborne particulates, such as wire-brushing the
RPVH stud threads or cutting the control rod drive lead screws. Based upon these considerations,
it is estimated by the staff that 15,000 and 45,000 person-hours may be needed for best and worst
cases, respectively.
For the best case, about 40 percent of the effort would be for preparation.
For worst-caseconditions, removal of the upper plenum and core support structure would account for about 80 percent of the total work effort, not including defueling. This compares with the normal effort of
600-700 person-hours for these steps.
Radiation Fields
It has been assumed by the staff that most of the reactor building decontamination will have been
completed before any major work is started on the RPVH.
Reactor building decontamination criteria,
as discussed in Section 5, will reduce background radiation to low levels, and the major radiation source will be from within the reactor coolant system and from the free water surface of thefuel transfer canal.
Hot spots also may be encountered in such areas as around the control rod
stators and RPVH closure studs.
Calculations made to estimate the levels of radiation that may
be associated with these hot spots (see Appendix I) indicate that shielding can be provided toreduce the gamma radiation to acceptable levels; thus, installation of such shielding is includedin the estimate of work effort.
Two factors will have a major bearing on the level of radiation that will be encountered:(1) thetotal core fission product inventory is quite low because the reactor had operated at significant
power levels for only a few months before the accident, and (2) by the time these removal operations are conducted, perhaps three to five years after the accident, the reactor fission products
will have decayed far below the radiation level encountered during a normal refueling operation.
Taking into consideration that reactor building decontamination of the 347-ft level that will
have preceded these operations and the fact that shielding can be provided around hot spots, thestaff has used an average gamma dose rate of 10 mrem/hr to workers for determining exposure
during work associated with removal of the RPVH and internal components (Appendix I).
Waste Generated
The contaminated waste materials produced from the activities are as shown in Table 6.4.

6-13

Table 6.4.

Wastes Generated during Removal of the RPV Head and Internals
Waste Generateda

Operation

RPVH insulation, rod drive cabling, minor

1.

Preparation2.

Installation of seal plate

Seal plate diaphram, sealing material.

3.

Removal of RPVH studs and nuts
Best case
Worst case

None.

4.

RPVH insulation,

rod drive cabling, minor*amounts of debris.

Stud and nut fragments.

Uncoupling of APSR and CR drives
Best Case
Worst CaseAPSR stators, drive lead screws.
All control drive stators, lead screws,
control drive housings, minor amounts ofchips and seal material.

5.

RPHV lift

Minor amounts of fuel and cladding debris.

6.

Removal of upper plenum

7.

Best case

Fuel and cladding debris.

Worst case

Fuel-and cladding debris plus the entire
stainless-steel upper plenum assembly
weighing 110,000 pounds.

Removal of core support
Best case

Small portion of total core inventoryof fuel and cladding.

Worst casePart of the fuel and cladding plus thestainless steel core support structure.

aSome of the wastes may be recoverable after decontamination.

Associated Facilities
Alternatives discussed in Section 5.2 will require support facilities that also can be used forthe removal of the RPVH and internals.
The only new facility that might be needed would be awater-filled pool for special tooling development, checkout and operator training.
6.3.4
6.3.4.1

Effluents and Releases to the Environment
Normal Releases

Before the RPVH is removed, attachments will be made at various points on the head to remove all
pockets of radioactive gas that may be trapped inside the RPV.
The vessel will be purged to
sweep out these gas pockets, and the gas will be stored or vented to the environment as deemedadvisable.
It is estimated that a maximum of 100 Ci of Kr-85 gas, essentially the only radioactive gas present, would be released from the RPV by this process.
If vented to the environment, the concentration of Kr-85 at the building vent would be 1.77 x 10-3 pCi/mL for about
40 minutes at the building ventilation rate of 50,000 cfm.
Such a release is not significantly
different from previous low-level Kr-85 releases vented from TMI-2 from time to time.

6-14
During certain portions of these operations, the fuel transfer canal and possibly the spent fuelstorage pool will be filled.
One option is to fill these structures with the decontaminatedwater from processing of the liquids in the AFHB tanks, the reactor building sump, and thereactor coolant system.
If this water is used, it would contain a tritium concentration of about
0.5 pCi/mL.
Evaporation from the surface of these storage pools thus would contain tritium which
would pass which would pass through the filters in the ventilation systems.
For the reactor
building, with a ventilation flow of 50,000 cfm, an evaporation rate of 10-3 pounds per minute2
per square foot, and a surface area of 1630 ft(transfer canal), the effluent concentration
would be 2.6 x 10-7 pCi/mL.
In the fuel handling building the ventilation flow is 36,000 cfm andthe storage pool area is 2350 ft 2 . Using the same evaporation rate (10-3 lb/min-ft 2 ), the effluent concentration of tritium would be 5.3 x 10-7 pCi/mL.
Based on these ventilation rates andeffluent concentrations of tritium, the total tritium that would be released to the atmospherefrom the reactor building and fuel handling building during one year would be 500 Ci.
Because almost all the operations discussed in this section will be conducted under water, few if
any new sources of airborne particulate activity are expected.
However, depending on the levelof decontamination achieved during cleanup of the reactor building, some airborne particulate
activity may be generated simply by people and equipment moving around in the building during
removal of the head and internals.
The resulting effluents, if any, will be much less thanduring the original building decontamination effort described in Section 5.
6.3.4.2

Accident Scenarios

The main concerns during this operation are drop accidents which could lead to a loss of water ormechanical damage to the core.
Of course, careful monitoring and control of boric acid concentration to ensure that the reactor remains shut down as discussed in Section 4.4 is absolutely
essential.
The only fission product that might be released and become airborne as a result of mechanicaldamage to the core would be Kr-85, which is a gas.
Although the Zircaloy cladding on many of thefuel elements is expected to have failed, a few of the elements could be intact and still containtrapped Kr-85.
The inventory of Kr-85 in a single fuel element is about 1.5 Ci, of which only afew percent is free gas in the gap that could be released by rupturing the cladding.
The other possibility is a bubble of fission gas trapped in a damaged region of the core that isreleased by dropping something on top of the core.
Since the pressure in such a bubble would be
about two atmospheres (i.e., under about 30 ft of water) a bubble one liter in volume would
contain about 0.1 moles of gas.
The composition of fission gas from low-burnup fuel after one
year of decay is 99 percent stable isotopes of Kr and Xe and about 1 percent Kr-85.
Thus aone-liter bubble might contain up to 10-3 moles, or 35 Ci, of Kr-85 if it were pure fissionproduct gas.
Since the volume of the reactor building is about 2 million ft 3 , the release of
35 Ci of Kr-85 would result in a concentration of about 6 x 10-4 pCi/mL inside the building.
Assuming a ventilation rate of 50,000 cfm, 90 percent of this Kr-85 would be discharged to theatmosphere in about 100 hours.
At this stage of the operation, a drop accident that caused a failure of the seal plate would
result in drainage of the transfer canal to the reactor building basement, but would have nosignificant radiological consequences since the canal water would contain very little
radioactivity (0.01 pCi/mL, exclusive of tritium), and there would be no fuel in the transfer canal.
6.3.5
6.3.5.1

Environmental

Impacts

Occupational Doses

Removal of the RPVH and internals includes operations to prepare and modify equipment, to remove
the reactor components from the reactor vessel, and to remove components that cannot be reusedand other waste products from the reactor building.
It is the staff's judgment, as indicated in Appendix I, that the average radiation exposure rate
to worker during an entire shift will be about 10 mR/hr.
Therefore, since it has been estimated(Table 6.3) that these operations will require 15,000 to 45,000 person-hours of work effort, the6-15
total cumulative occupational radiation dose will range from 150 to 450 person-rem.
The occupational dose includes an estimated 4 to 14 person-rem from inhalation of tritiated water vapor
that evaporates from the fuel transfer canal during removal of the PPVH and internals.
The fuel
canal will be filled with processed water having a tritium concentration of about 0.5 pCi/mL,
which is expected to produce an average tritium concentration in the atmosphere of about 2 x 10-6
pCi/mL in areas where workers will spend about 80 percent of their time.
In order to maintain the average atmosphere concentration at 2 x 10-6 pCi/mL or below, it will benecessary to arrange barriers and building ventilation patterns so that the fraction of water
vapor at work stations that comes from the fuel transfer canal or any other source of tritiatedwater vapor does not exceed about 20 percent of the saturation concentration at 75'F.
The expected number of additional cancer mortalities in-the work force exposed to this cumulative
dose of radiation ranges between 0.02 and 0.06.
The expected number of additional genetic
effects in the offspring of the work force exposed to this cumulative dose of radiation ranges
between 0.04 and 0.12.
Assuming use of three 8-hour shifts a day, and crew sizes as indicated in Table 6.5, the operations associated with removal of the RPVH and internals will require a total calendar time ofthree to ten months.
The removal of the RPVH and internals above the core will require about two
to six months, excluding removal of the core support structure.
After defueling, an additional
one to five months will be spent removing the core support structure and handling the radioactive
wastes generated.
6.3.5.2

Offsite Doses

The dose estimates presented here for removal of the RPV head and internals are based on thesource terms developed in Section 6.3.4.1.
The source terms are for removal of the RPV head(100 Ci of Kr-85) and evaporation of water from the reactor building and fuel handling building(500 Ci of H-3 per year).
The calculational models used-to make these estimates and the interpretation of their results are described in Appendix W. The significance of these doses andtheir human health and environmental consequences are discussed in Section 10.3.
The total bodydose estimate from the 100-Ci Kr-85 release and the tritium releases are listed in Table 6.6.
The total-body population dose received by the human population within 50 miles due to these
releases was estimated to be 4 person-rem.
The offsite doses due to the particulate releases mentioned in Section 6.3.4.1 are expected to beless than those described in Section 5.2.5.2. The estimates in Section 5.2.5.2 are significantly
less than those due to the 500 Ci/year source term of H-3; hence, the values in Table 6.6 encompass the cumulative dose from these activities.
6.3.5.3

Postulated Accident Effects

The type of accidents for which source terms are developed in Section 6.3.4.2 involve leakage byfuel pin failure of Kr-85 from fuel elements and release of a bubble containing Kr-85 gas trappedin the core. The amount of krypton released in these accident scenarios is so small (1.5 Ci perfuel element accident and 35 Ci from a trapped bubble) that there would be negligible offsite
doses.
Other fission products that might be released would either be in a particulate or watersoluble form and would be entrained in the RCS water for subsequent water processing.
6.3.5.4

Psychological-Socioeconomic

Psychological-socioeconomic
cussed in Section 6.5.5.4.
6.3.6

Effects

effects relative to removal of the RPV head and internals are dis-

Economic Costs

Estimates of the economic costs for the best- and worst-case alternatives are given in Table 6.7.
These costs are estimated on a consistent relative basis, and do not necessarily reflect all
costs on an absolute basis.
Thus, these costs should be considered only as relative costs forthe two scenarios (best case vs. worst case).
About 85 percent to 90 percent of the labor costs
are expected to be incurred in preparation for work inside the containment building; the majority

Table 6.5.

Work Effort, Crew Size, and Calendar Time Required to Remove TMI-2
RPV Head and Internals

Operation

Estimated
Work Effort(person-hours)
Best Case
Worst Case

Estimated
Crew Size

Estimated Calendar TimeaTo Complete (months)
Best Case

Worst Case

15

0.6

0.6

5
5

0.1
0.3

0.1
0.7

0.7

1.7

1. Preparation

6,000

2.
3.
4.

Installation of seal plate
Removal of RPVH studs & nuts
Uncoupling of APSR & Cr drives

400
1,200

2,400

RPVH lift

5,000
1,200

4

5.

2,000
900

6.
7.

Removal of upper plenum
Removal of core support structure

2,000

13,000

6

0.5

3.0

2,500

17,000

6

0.6

3.9

45,000(Average of 6)

3.1

10.3

TOTAL
15,000aCalendar time is given to nearest tenthof a month.
bone-quarter month.
COne-third month.

6,000
400

5

0.3 b

0.3 c

0I

6-17

Table 6.6. Estimates of Offsite Doses to the Maximum Exposed
Individual due to Tritium and Krypton Released
during Removal of RPV Head and Internals

Location

Pathway

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

Dose (mrema
Total-Body
3.4 x 10-20
1.2 x 10-1
1.5 x 10-1

Total
Nearest milk
goat

Nearest cow milk

and garden

Inhalation
Ground Shine
Goat Milk Use
Total

1.8 x 10-20
1.4 x 10-1

Inhalation

3.8 x 10-2

Ground Shine
Vegetable Use0
1.3 x 10-1

1.6 x 10-1

5.4 x 10-2

Cow Milk Use

2.2 x 10-1
Total
aDoses were calculated for GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The dose estimates for all organs
except bone were the same as that for total-body. The estimates for bone were zero. Doses were calculated for four age
groups: adults, teenagers, children, and infants. The largestof each group are listed in this table. Infant doses are listed
for the nearest milk goat location and child doses are listed
for the other two.
bThe basis for selecting the special locations is described inAppendix W. The actual locations are: nearest garden = 1.05
mile east-northeast, nearest milk goat = 1.02 mile north, andnearest cow and garden = 1.05 mile east.

Table 6.7. Economic Costs for Removal of
RPV Head and Internals(thousands of dollars)
ItemBest Case

Worst Case

Labor
Equipment
Facilities

2180

4550

1120
250

1750
250

TOTAL

3550

6550

6-18

of this effort consists of training and procedure preparation.
The equipment costs consist ofsuch things as special tooling for seal plate installation, control rod drive removal, and cutting tools for plenum removal.
The differences between the best and worst cases are not great
since much of this tooling must be developed, checked out, and be available before the actual
working conditions have been determined, even though some of the tooling may never be used. The
facility needs are not great and consist only of a large water tank with a suitable work platform
for tooling mock-up tests and operator training. It has been assumed that suitable onsite indoor
locations, such as the turbine hall, exist for this facility.
The methods used to make these cost estimates are described in Appendix K.
6.4

CORE EXAMINATION AND DEFUELINGThe objective of core examination and defueling operations is to determine the extent of damage
to the reactor core and then remove the fuel assemblies and fuel debris.
6.4.1

Status and Specific Considerations

Information regarding the present status of the core is given in Section 6.2.1.
Because the actual condition of the reactor core is unknown, the staff has developed a framework
within which reasonable bounding conditions and alternative courses of action can be developed.
As stated previously, the spent fuel pool (SFP) and fuel transfer canal (FTC) will be filled withdecontaminated water (ý 0.01 pCi/mL)* at the start of defueling operations (see Sec. 6.3). A
suitable decay-heat-removal system will be in operation as described in Section 4.3. Reasonably
clear water is needed to conduct detailed core examination and subsequent defueling operations.
Thus, reactor water clarity should be maintained throughout the core examination and defuelingphase. Fuel assembly debris may be inadvertently rearranged over the top of the core as a resultof removal of the upper plenum. Some trapped fission product gas may be released during defueling; thus, monitoring instruments may be needed to detect any such release. Maintaining thereactor in a shutdown condition is necessary throughout defueling.
Additional incore nuclear
instrumentation sources may be necessary to monitor the effective neutron multiplication of thecore. The conditions of the fuel are as given in Section 6.2.1.
Fuel and fuel debris canning (placement in transfer containers) and transfer to the spent fuelcanal (SFP) has been based upon the Allied-General Nuclear Services study 4 of September 1980.
6.4.2

Alternative Methods Considered

The staff's consideration of the environmental impacts of core examination and defueling of thereactor vessel is based upon the core conditions described in Section 6.2.1. The differences in
estimates of possible core damage and their resulting impacts on recovery operations have led tothe development of two different postulated sets of conditions--a best case based on estimates ofthe most probable condition of the reactor core and a worst case that reflects the impact of more
severe damage conditions (see Sec. 6.2).
The reactor vessel defueling sequence described in this section will involve removing only that
fuel material actually within the reactor vessel--not material that may be lodged in other locations within the reactor system, such as coolant piping external to the reactor vessel.
Fuel andparticulate removal from other portions of the reactor coolant system are discussed in Section 6.5.
The general defueling sequence consists of two series of activities that must be sequenced withactivities described in Section 6.3 for removal of the RPV head and internals.
The first phaseof the reactor defueling will begin with the conclusion of operations 1 to 6 described in Section 6.3 that provide access to the top surface of the core. In this first phase of defuelingremoval of all the fuel assemblies is accomplished. At that point the defueling activities will
be stopped until the lower grid and core support assembly are removed in those activities describedin Section 6.3.2, step 7.
The second and final phase of the defueling of the reactor vessel (to scavenge and hydraulically
vacuum the remaining fuel and cladding material from the bottom area of the reactor vessel) thenwill be initiated.
*Exclusive of tritium.

6-19
Television cameras and periscopes (for visual inspections) and underwater lighting and positioning
equipment will be needed for defueling operations.
6.4.3

Details of Methods and Facilities

Preparations such as design, construction, and demonstration of special tooling may take about
two years. Selection of procedure for some tasks may even have to wait until defueling activities are started. Thus, it may
necessary to plan, design, and fabricate special equipment
before a final decision is made on what procedures to use.
Various core-evaluation activities will be necessary to aid in making operational decisions atseveral points in the defueling process.
The removal of the upper plenum assembly from thereactor vessel will permit the use of remotely operated equipment and underwater optical viewing
apparatus. A full examination of the core top surface will be required (see Fig. 6.2 for a crosssection of core).
It would involve an attempt to assess the extent of damage to the central
portion of the core, and to the peripheral fuel assemblies. The peripheral fuel assemblies would
be checked to determine if they can be removed as entire assemblies with the use of special
handling and removal apparatus.
As described later, the ability to remove at least one fuelassembly intact from the core periphery will significantly affect the activities.
Examinations of the condition of the internal guide tube openings through the axial length of thefuel assemblies also may be performed to determine whether these passages are sufficiently open
to allow for insertion of lifting apparatus into the fuel assemblies. Some of these guide tubes
may be structurally deformed or contain debris that cannot be removed. However, it is expectedthat some guide tube openings will remain clear, especially in peripheral assemblies.
Under the best-case conditions, some fuel cladding and material are expected to be deposited on
the top surfaces of the upper flat surfaces of the fuel assemblies (see Fig. 6.3). Also, structural material is expected to be deposited on the upper surfaces of the fuel assemblies as aresult of the removal of the reactor vessel upper internals.
In the best-case conditions, theparticulate matter is assumed to be fuel and cladding material that can be hydraulically vacuumed
from the various surfaces. The worst-case conditions would necessitate the removal of additional,
larger pieces of structural material and fuel sections, possibly including control and axial
power shaping rods as well as parts of fuel assemblies that fused to the bottom of the upper
plenum., These larger pieces would have to be knocked or scraped off prior to completing theremoval of the plenum structure.
These pieces may fall onto the top surface of the fuel assemblies and must be removed before the core examination can be completed. Removal of loose fuelmaterial during the core examination phase is considered an important preparatory requirement
proceding the actual defueling.
6.4.3.2

Best-Case Conditions

Under even the best-case conditions anticipated by the staff, significant damage exists within
the central portion of the core, with many of the central fuel assemblies bound or fused together
at the spacer grid elevations (see Fig. 6.3). Therefore these fuel assemblies cannot be individually removed. However, the staff assumes that under best-case conditions the peripheral fuelassemblies have not been damaged to an extent that prevents extraction of at least one complete
fuel assembly using specially designed removal apparatus. The normal fuel handling machines lift
the fuel assemblies by the top end fitting. Because successful fuel assembly removal depends
upon a structurally sound fuel assembly throughout its length, it is likely that other fuelremoval equipment will be needed. The cavity created by removal of one peripheral fuel assembly
would permit a sequential extraction of adjacent fuel assemblies radially toward the center ofthe core by use of the same equipment. At some point on the radial removal path around the core
perimeter, the sequential removal activity would reach those fuel assemblies that have sustained
the most damage. The removal of these fuel assemblies will require considerably more effort andmore complex removal equipment.
Any remaining fuel debris and structural material would be removed after the core support structure was removed using articulated grapple equipment and hydraulic suction apparatus.
6.4.3.3

Worst-Case Conditions

For the worst-case conditions the staff acknowledges that core damage could be so extensive that
all of the fuel assemblies have been at least partially damaged and that none of the peripheral

6-20

2

y

Figure 6.2.

\-CORE SUPPORTSTRUCTURE

Reactor Vessel Internals and Core Cross Section.

6-21

TOP .VIEW

S PAM

GRIC

SE
• q0

CROSS SECTION

LOWERENO
FITTINGINSTRUMENTATION TUBE
CONNECTION

Figure 6.3.

TMI Unit 2 Reactor Fuel Assembly.

6-22
assemblies can be removed as a complete fuel assembly. As under the best-case conditions, thefuel assemblies in the upper central portion of the core are bound or fused and cannot be removed
without separation into smaller segments.
Under the worst-case conditions, additional procedures
and special equipment would be required to open a full-length cavity on the periphery of the corein order to begin the removal of the first fuel assembly. This initial cavity would be formed by
cutting and removing the baffle plates in a segment of the core support structure that provides
the most access to the selected peripheral fuel assembly. The removal of adjacent fuel assemblies could then progress by use of the special separation and packaging equipment.
Removal of
the remaining fuel assemblies would progress radially inward from this location, as under bestcase conditions.
Operations under worst-case conditions would require the use of specially adapted equipment.
Additional tools, such as hydraulic jacks, may be required to aid in releasing the fuel assemblies from their pockets in the lower grid plate supporting the core assembly. The necessity foruse of various items of removal equipment will depend on the actual condition of the core and theprocedures chosen.
Among the methods reviewed by the staff was that of using cutting tools to
cut and shear segments of the core. Since this method would have impacts that lie between thebounding conditions discussed above, this and other techniques are not discussed further.
6.4.3.4

Fuel Debris Handling

The TMI-2 core, as loaded, contained 177 fuel assemblies.
It is uncertain what the condition of
the fuel will be when removed from the reactor vessel; the staff assumes that the fuel will be in
any one or a combination of the following three configurations:Intact--intact, but weakened.
Fused Sections--portions of fuel assemblies fused to each other such that they will
have to be physically separated.
Core Debris--consists of two types:
relatively large pieces that can be mechanicallyhandled, and smaller pieces that will have to be hydraulically vacuumed and filtered.
In each of the above configurations, it is likely that the fuel will require canning to provide
containment and structural integrity and to prevent spread of contamination during thesteps
involved in transfer from the reactor building to the spent fuel pool.
Transfer of the fuel from the reactor building to the SFP is via transfer carriage/upender andthe fuel transfer tube with the SFP as the end-point.
Interim wet storage in the spent fuel storage pool pending packaging and offsite disposal isanticipated.
Dry vault or caisson storage in a new onsite facility is another alternative forinterim storage. These options are described in more detail in Section 9.
The procedures and equipment used to handle and package the fuel will be intended to ensure that(1) site personnel exposure is maintained at ALARA levels and (2) the probability of a handling
accident is minimized.
6.4.3.5

Final Cleanup

Final cleanup of the reactor vessel would be the same under both sets of conditions and would
begin after removal of the core support structure. Any remaining large debris would be removed
with remote grappling equipment and placed in transfer containers. All fuel pellets, pieces ofcladding, and other small debris would be removed from the bottom surface of the reactor vesselby use of the hydraulic suction equipment and scoop-type pickup apparatus. All material would beplaced into transfer containers and moved to a temporary storage location.
6.4.3.6

Special Tools and Equipment

Some of the special remotely operated removal tools and equipment that may be needed will requiresignificant lead times for procurement or for design and fabrication.
Design and fabrication
effort also will be required to provide mockup equipment for training of those individuals involved

6-23
in the fuel removal operations.
The equipment development and checkout time may differ significantly between best and worst cases, with the equipment for the worst case being more sophisticated and thus requiring more development and operator-training time.
Sufficient lead time must
be included to permit fabrication and checkout of this equipment prior to the start of the training.
6.4.3.7

Logistics,
Facilities

Crew Size,Work Effort,

Radiation Levels,

Waste Generated,

and AssociatedWhatever method of defueling is eventually selected, the releases and impacts from that method
can be expected to fall within the values given in Sections 6.4.4 and 6.4.5.
Crew size, work
effort (person-hours), radiation fields, waste generated, and associated facilities are discussed
below for the two cases.
Crew Size
Core examination and fuel removal were judged by the staff to require specially skilled personnel
for certain tasks.
The work crew involved in the best-case activities is expected to averageten people per shift on a six-shift-per-day basis.
The worst-case situation may require up
to 14 people per shift.
Once initiated, each activity of the defueling operation will be acontinuous effort until completed.
Work Effort
The total direct person-hours inside the reactor building for the best and worst cases are estimated to be 58,000 and 135,000 person-hours, respectively.
Under nonaccident, optimum conditions, an undamaged core can be loaded or unloaded in about 700 person-hours.
Radiation Fields
The staff has estimated an average gamma radiation field of 10 mR/hr for work associated with thedefueling activities (see Appendix I).
This is based on the assumption that necessary defuelingwork areas in the reactor building will be decontaminated before defueling begins.
Backgroundradiation from building contamination will then be quite Tow and the major radiation source will
be from the water surface of the fuel transfer canal.
Hot spots may be encountered during changeout, repair, and modification of equipment.
These hot spots can be handled effectively byshielding and by careful design of equipment.
Waste Generated
The waste materials
tion 8.

expected to be produced from the two sets of conditions are listed in Sec-

Associated Facilities
Alternatives previously discussed in Section 5.2 will require supporting facilities that also canbe used for defueling.
The only new facility that might be needed would be a water-filled poolfor development, checkout, and training for any special equipment.
6.4.4

Effluents and Releases to the Environment

The defueling and inspection operations will be conducted through the transfer canal and spent
fuel pool water.
The only releases expected during normal operations would escape of trappedfission gas bubbles as pieces of the damaged core are pried apart and the evaporation of tritiated water in the transfer canal and spent fuel pools.
Soluble and particulate material
released by these operations will be picked up by the transfer canal water and removed by theprocessing systems as discussed in Section 7.
6.4.4.1

Normal Releases

The concentration of Kr-85 in the reactor building which might result from the release of 35 Ci
of fission gas was discussed in Section 6.3.4.2.
Any airborne particulate activity in the effluent from the reactor building during the defueling and inspection operations is expected to bemuch less (less than 20 percent) than during the initial decontamination of the reactor building

6-24(see Sec. 5.2).
Tritium releases from evaporation of the transfer canal and spent fuel storage
pool water are discussed in Section 6.3.
6.4.4.2

Accident Scenarios

During the defueling activities, the main accident concern is related to loss of water from thetransfer canal.
This could occur if the canal seal plate were damaged (e.g., with the vesselhead).
Failure of the seal plate would cause a loss of water in the portion of the transfer
canal above the reactor vessel and in the adjacent shallow region. Water would not be lost fromthe deep part next to the transfer tubes, the spent fuel storage pool, or the reactor vessel.
An
accident that causes failure of the seal plate and the subsequent loss of water from the fueltransfer canal could involve the overheating of a subassembly if one were in the shallow part of
the fuel transfer canal at the time of the accident. Although it is likely that a sizable leakin the seal plate would be immediately detected, the rate of water loss could be so rapid that afuel transfer could not be completed before significant loss of shielding occurred.
In thiscase, an uncooled fuel assembly could possibly be situated in the canal for an extended period oftime.
Fuel would not be present in the shallow region except during a transfer or while being placed ina container (can) for a transfer. For the purpose of this accident analysis, a fuel assembly is
assumed to be present in the shallow region when water is lost. At the time of defueling, thedecay heat generation rate in a single fuel assembly will be on the order of 500 watts or less;hence, not much cooling is needed.
However, depending on the time it remained uncovered and on
the geometry, the fuel might heat up and release some fraction of the gaseous fission products(Kr-85) trapped in the fuel matrix or in the fuel-clad gap. It is estimated that a maximum of
10 percent of the gaseous activity would be released in this manner.
A fuel assembly could
contain a maximum of 320 Ci of Kr-85. The resultant concentration in the reactor building would
be comparable to the release discussed in Section 6.3.4.2.
A recriticality associated with the movement of fuel or control rods during the defueling cannot
occur as long as the boric acid concentration of the water is maintained above 3000 ppm. The
consequences of an accident that causes the water to drain from the core are discussed in
Section 10.
6.4.5

Environmental Impacts

6.4.5.1

Occupational Doses

The average dose rate to workers in work areas is expected to be about 10 mrem/hr (Appendix I)
for both best-case and worst-case conditions.
Under best-case conditions, completion of inspection and fuel removal will require 58,000 personhours and result in a cumulative occupational dose of 580 person-rem. Under worst-case conditions, completion of fuel inspection and removal will require 135,000 person-hours and result ina cumulative occupational dose of 1350 person-rem.
The occupational dose includes an estimated 20 to 40 person-rem from inhalation of tritiatedwater vapor that evaporates from the fuel transfer canal and spent fuel pool during core examination and defueling.
The fuel canal and pool will' be filled with processed water having a
tritium concentration of about 0.5 pCi/mL, which is expected to produce an average tritium
concentration in the atmosphere of about 2 x 10-6 pCi/mL in areas were workers will spend about
80 percent of their time.
The staff expects defueling to be performed on a six-shift-per-day basis, with ten-member crews
for the best-case conditions and 14-member crews for the worst-case conditions. For the best
case, each shift could be eight hours long, and crew members would work in the reactor building
for four hours and spend four hours performing related activities such as suiting up and desuiting,personal radioactivity monitoring and decontaminaton, and training. Thus, there could be six
overlapping 8-hour shifts over a 24-hour period. Each crew member could be limited by the licensee's administrative controls to no more than 1 rem per quarter; thus, assuming an average dose
rate of 10 mrem per hour, each crew member could be limited to 100 hours per quarter in thereactor building.
The reasons for the staff's selection of this value (for calculation of manpower needs) rather than limits as high as 3 rem per quarter given in 10 CFR Part 20 are provided
in Appendix L. However, individuals may receive doses up to 3 rem per quarter if necessary.
For

6-25
a dose of 1 rem per quarter, 2.5 times as many crew members would be needed as would be the caseif staffing requirementswere not controlled by the need to limit radiation doses. If radiation
exposure were not a consideration, each crew member could work at least 250 hours per quarter.
The daily work force requirement would be a ten-member crew for each of six shifts per day, or
60 crew members.
For the quarter, therefore, 2.5 times as many workers would be needed, or
150 crew members.
For the worst-case estimate, six 10-hour shifts a day (with four of those hours per shift spentin the reactor building) would be needed. The extra two hours per shift would be for the extra
training and planning that would be needed beyond that required for the best case. If 14-member
crews were used under worst-case conditions, 210 crew members would be needed. The exact size of
the work force may vary, depending on whether unforeseen difficulties develop and on applicationof the licensee's administrative quarterly dose check point procedures (see Appendix L). Given
these variables, the size of the work force could range from 150 to 250 individuals.
For the best-case conditions, six 10-member crews would spend a total. of 24 hours (four hours per
shift) each day in the reactor building. This would be 240 person-hours per day. For a calendar
quarter of 90 days, the work effort in the reactor building would, therefore, be 21,600 personhours.
Since 58,000 person-hours are expected to be needed, about 2.7 quarters, or 8 months,
would elapse before defueling would be completed.
For the worst-case estimate, using 14-member
crews, 30,240 person-hours would be expended per quarter. The work effort of 135,000 person-hours
would thus require about 4.5 quarters, or 13 months. The duration of defueling operations thus
is estimated to be 8 to 13 months.
Based on the above values of occupational dose and work force, the expected number of additional
cancer mortalities in the work force exposed to this cumulative occupational dose would range
from 0.08 for the best case to 0.18 for the worst case. The added probability that the average
individual worker would die of cancer varies from 1 in 1400 to 1 in 1900. The expected number of
additional genetic effects in the offspring of the work force exposed would range from 0.15
to 0.35.
6.4.5.2

Offsite Doses

The offsite dose estimates to the maximum exposed individual resulting from core examination anddefueling operations are based on source terms described in Section 6.4.4.1 (35 Ci of Kr-85 and20 percent of the source term listed in Table 5.10, and assuming no tritium is released) and arelisted in Table 6.8. The calculational models used to make these estimates and the interpretationof their results are described in Appendix W. The significance of these doses and their humanhealth and environmental consequences are discussed in Section 10.3. The total-body populationdose received by the human population within a 50-mile radius from these activities was estimatedto be 2 x 10-3 person-rem.
6.4.5.3

Postulated Accident Effects

The type of accident postulated in Section 6.4.4.2, which would result in offsite doses, is thatof uncovering a fuel assembly and subsequently releasing 32 Ci of Kr-85. The doses offsite
resulting from such a release are insignificant, and they are negligible in comparison to thosefrom the routine decontamination operations.
Consequently, no numeric values for this accident
scenario are presented here.
6.4.5.4

Psychological-Socioeconomic Effects

Psychological-socioeconomic
in Section 6.5.5.4.
6.4.6

effects relative to core examination and fuel removal are discussed

Economic Costs

To evaluate all the economic costs directly associated with a particular alternative, all labor,facility, and equipment costs need to be included. Estimates of the economic costs for the bestand worst-case alternatives are given in Table 6.9.
These costs are estimated on a consistent
relative basis and do not necessarily reflect all costs on an absolute basis. These costs should
be considered only as relative costs for the two scenarios (best case versus worst case).

6-26

Table 6.8. Dose Estimates for the Maximum Exposed Individualfrom Core Examination and Defueling Operations

Pathway

Total-Body

Dose (mrem)a
Bone

Inhalation
Ground Shine

4.8 x 10-7
1.9 x 10-6

6.3 x 10-6
1.9 x 10-6

Vegetable Use

3.7 x 10-5

1.6 x 10-4

4.2 x 10-5

Total

3.9 x 10-5

1.7 x 10-4

4.5 x 10-5

Location
Nearest b
gardenNearest milk

goat

Nearest cow milk
and garden

Liver
8.6 x 10-7
1.9 X 10-6

Inhalation

7.3 x 10-7

2.5 x 10-6

5.7 x 10-7

Ground Shine

1.8 x 10-6

1.8 x 10-6

3.1 x 10-5

2.2 x 10-4

1.8 x 10-6

Goat Milk Use
Total

3.4 x 10-5

2.2 x 10-4

2.5 x 10-4

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

9.0 X 10-7
2.8 x 10-6
3.6 x 10-5
1.2 x 10-5

6.9 x 10-6
2.8 x 10-6
2.4 x 10-4
5.5 x 10-5

9.5 x 10-7
2.8 x 10-6
6.2 x 10-5
4.8 x 10-5Total

5.2 x 10-5

3.0 x 10-4

1.1 x 10-4

2.5 x 1 0 -4

aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three organ doses are listed in thistable. Doses were calculated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and nearest cow and garden locations are for children. The dose estimates for the nearest goat
location are for adults for total-body, and for infants are for bone andliver.
bThe basis for selecting the special locations is described in Appendix W.
The actual loactions are: Nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden = 1.05
miles east.

Table 6.9. Economic Costs for Defueling(thousands of dollars)
ItemBest Case

Worst Case

Labor
Equipment

6,620
5,860

9,910
5,860

TOTAL

12,48015,7706-27
Because of uncertainties as to the condition of the core, plus lead time required for tooling,
the staff has assumed that regardless of the core condition, costs for preplanning, tooling, andtraining will be the same for both the.best and worst cases. Only the cost of labor required toremove and support removal of the fuel will be different for the best and worst cases.
Compared to the labor and equipment costs, the facility needs are small.
In addition to thewater-filled tank mentioned inSection 6.3, a fairly large mockup area will be required to set up
the special tooling for operator training. The staff has assumed that suitable mockup locations
will be available onsite and has made no allowance for facility costs.
6.5

DECONTAMINATION OF PRIMARY SYSTEM COMPONENTS

Postaccident analyses at TMI suggest that fuel debris and perhaps other particulates were scattered throughout the primary system.1' 2 Fission products are thought to have been carried fromthe exposed fuel and subsequently plated out on all primary system component inner surfaces. The
objective of decontamination of the primary system components is to remove the fuel debris andother particulate matter from the system and reduce the fission product plateout on internalsurfaces. to a level similar to that of operating reactors.
For purposes of this discussion,
decontamination of the primary system components is considered as being principally two distinct
activities--draining and flushing and chemical decontamination.
6.5.1

Status and Specific Considerations

Before decontamination of the primary system components begins, reactor core defueling and removalof the RPVH and reactor internals will have been completed. The components to be decontaminatedare the reactor vessel, the steam generators, pressurizer, fuel transfer canal and reactor draintank, drain tank transfer pumps, reactor coolant pumps and motors, letdown coolers, and associatedpiping.
A number of special factors must be taken into consideration in the planning and execution ofcomponent decontamination.
Attention must be given to flushing and chemically decontaminating
the dead legs, low points, drain and vent lines, flushing connections, and injection lines.
Chemical decontamination solutions should be compatible with component materials. Spot application of chemicals may be necessary.
The inventory of contaminated solutions generated byflushing and decontamination should be kept as low as practical. Acquisition of additional equipment or systems needed to flush and chemically decontaminate the pressurizer because of its
location in the system, must be planned.
Decontamination of the reactor drain tank may present special problems. The tank (and its associated piping) is partially submerged in highly contaminated water in the reactor building basement and is believed to contain high levels of radioactivity because of the presence of contaminated reactor cooling water and core debris. The exact condition of the tank is not known, but
it is believed to be structurally sound and to contain reactor cooling water. The reactor draintank transfer pumps and coolers also are submerged in water. Repair of some components may benecessary.
The reactor coolant pumps and motors are not in operation and it is not known if they are operable. It is believed that the pumps are mechanically sound. The pumps were shut down during theaccident when they were vibrating because of being vapor bound, rather than because of pumpmalfunction. Two of the pumps were operated (one at a time) for a total of about one month after
the accident. to provide forced cooling through the primary loop. The condition of the motors at
this time is not known.
However, pump motor and cabling damage, both mechanical and electrical,
could have occurred as a result of reactor building conditions; e.g., the building pressure
increase during the accident and the high humidity of the atmosphere in the building.
Pumpshafts, shaft seals, and impellers may have been damaged by vibration caused by the pump's being
vapor bound. A laydown area for pump and motor dismantling and decontamination may be needed.
The letdown piping and coolers also are submerged in the highly contaminated water in the reactor
building basement. Both coolers were online during the accident and, therefore, have been exposed
to highly contaminated reactor coolant.
It also is expected that these coolers will contain aconsiderable amount of highly radioactive plateout, estimated to range from 10 to 100 pCi/cm2 .

6-28
6.5.2

Alternative Methods Considered

The decontamination of the primary system components can be considered principally as two distinct
activities--removal of loose particulates and removal of more tightly adhering fission-productplateout.
It is likely that particulates can be removed by draining, flushing, and filtration,*but that plateout removal will require use of chemical reagents.
Reactor system inspections(Sec. 6.2) will indicate the sequence of operations, but the staff anticipates that particulateremoval will be carried out first.
The first activity considered in this section is removal of particulates that may have been
scattered throughout the primary system and that cannot be readily removed by use of chemical
reagents (Sec. 6.5.2.1).
Although the exact quantities, locations, and composition of theseparticulates are unknown, the staff believes that most of the primary system components arecontaminated with particulates consisting primarily of irradiated fuel and oxidized cladding(ZrO2 ).

The second activity considered is removal of fission product plateout material adhering to various internal surfaces of primary system piping and components (Sec. 6.5.2.2). Although theactual concentration of plateout materials is unknown, higher concentrations of plateout could be
expected in areas of heat transfer than in other parts of the system. Plateout
removal using
5chemical solutions is a well-developed and effective decontamination technique.
6.5.2.1

Particulate Removal Using Drain and Flush Techniques

Reactor Coolant System
Two alternative methods have been considered for flushing the reactor coolant system:
the existing reactor coolant pumps and (2) using system pressurization.

(1) using

Reactor Coolant Drain Tank
Because of the system configuration,
the rest of the system.

the reactor coolant drain tank is discussed separately from

Two alternative methods have been considered for the draining of the reactor coolant drain tank.
The only difference between them is the method of initial'draining. Under the first alternative,
the tank contents would be drained to the reactor building sump; under the second, the contents
would be drained to the reactor coolant bleed holdup tanks. Installation of drain line filters
would be required for either alternative.
Drain tank flushing would be the final step for both of the alternatives. Flushing would consist
basically of filling the drain tank-with water and draining the water to the reactor coolant
bleed holdup tank, followed by manual flushing with hoses and water jets. Flushing also would
include the 14-inch-diameter pressurizer relief valve header. Small lines of 2 inches or less indiameter would be flushed by other means, removed and capped, or replaced.
6.5.2.2

Plateout Removal Using Chemical Decontamination Techniques

Major considerations involved in chemical cleaning of the reactor coolant system are (1) the
chemical reagents to be used and the method by which they are prepared for use, and (2) theprocedures used to conduct decontamination--more specifically, the method used to transport thechemical cleaning agents through the coolant system. The reagent chemicals used for decontamination normally would be shipped in concentrated form and then diluted withwater for use. To
prepare the solutions, large tanks with mixing capabi'lity would be needed. The existing reactor
bleed tanks, or the borated water storage tank could provide this capability.
Contamination
remaining after drain/flush activities would probably be in the form of activation and fission
product plateout in the corrosion product film and some particulate uranium oxide not removed byflushing.
Films on stainless steel and other high nickel alloys are not easily removed andrequire either corrosive treatments or some conditioning step to change the film characteristics.
Satisfactory procedures have been developed for removing such films on surfaces of primary coolant*Design of particulate filters will require consideration of potential criticality hazards.

6-29
systems of nuclear reactors and for decontaminating the surfaces. 5 Six alternative processes
have been considered for the TMI-2 cooling system:(1) OPG process should fuel particulatedissolution be necessary, (2) the oxalic-citrate-peroxide process, (3) the CAN-DECON process,(4) the alkaline permanganate-citrox (AP-Citrox) process, (5) the alkaline permanganate-ammonium
citrate process (APAC),
and (6) the Dow Chemical NS-1 process.
Details of each are given in
Appendix P.
Circulation of Chemical Solutions through Principal Reactor Coolant System Components
Two alternative methods of circulating the decontamination reagents through the principal reactor
coolant system components (Fig. 6.4) have been considered:use of the existing reactor coolant
pumps or use of new pumping equipment.
Circulation of Chemical

Solutions to the Reactor Coolant Drain Tank

Two alternative methods have been considered for circulation of chemical decontamination solution
to remove fission product plateout from the internal surfaces of the reactor coolant drain tank.
One method would be to feed and bleed (drain) the chemicals through the tank; the other method
would be to fill the tank and then recirculate the contents.
For both alternatives, decontamination solutions could be provided from a makeup system during the decontamination process.
Chemical Decontamination of Reactor Coolant Pumps and Motors
The procedure for reactor coolant pump decontamination would depend on the condition of the pumps
and pump motors, which could range from insignificant contamination of motor internals and nodamage to either the motor or the wetted pump components, to a condition of high levels of contamination and severe damage to motor and wetted pump components.
Two alternatives have been considered by the staff for decontamination of the external surfacesof the four reactor coolant pumps and motors.
The first alternative would involve inplace decontamination; the second would involve removal and out-of-containment decontamination, and/or
disposal.
Inplace decontamination would be carried out in the reactor building, and for out-ofcontainment decontamination, the pumps would be moved out of the building.
Inplace decontamination would be carried out in cases when only minor contamination of the motor internals and no
significant damage to the motor and other pump components had occurred.
Out-of-containment
decontamination would be necessary if there was contamination or severe damage to the motor orpump components.
These alternatives do not cover those pump components that are normally exposedto the primary coolant.
Those components would be decontaminated with the reactor coolant system
as discussed previously.
6.5.3
6.5.3.1

Details of Methods and Facilities
Particle Removal Using Drain and Flush Techniques

The draining and flushing alternatives considered differ only in the equipment used and work
effort (person-hours) required to perform a given activity. Processed and demineralized waterprobably would be used for flushing, as practical.
The use of "new" water would be minimized.
Tanks and piping used to receive and/or store flush solutions most likely would have been alreadydesludged and decontaminated.
Reactor Coolant System
Flushing Using the Reactor Coolant Pumps.
This alternative would involve the use of the reactor
coolant pumps, a large, specially built core filter installed in the reactor vessel, and inline
filters in reactor coolant system drains.
After the reactor defueling is completed and the fuel transfer canal has been drained, the corefilter would be installed in the reactor vessel and the RPVH would be reinstalled.
This filter
would be designed to permit backflow and to allow flow bypass if the pressure drop across it
exceeded a prescribed value.
Instrumentation would be included to monitor filter performance.
Additional filters would be installed on the reactor coolant system drains to trap particulates
during the drain and flush activities.

6-30

REACTOR COOLANT
PUMP MOTOR(TYPICAL OF FOUR)

PRESSURIZER

REACTORCOOLANT PUMP(TYPICAL OF FOUR)-

SURGE
LINE

-STEAM
GENERATOR N"A

Figure 6.4.

Schematic Drawing of Primary System.

6-31
To permit use of the four reactor coolant pumps for flushing operations, the reactor system would
be filled with water, borated as necessary, and then pressurized so that the coolant pumps could
be operated.
Water flow through the four pump legs would be used to wash the particulates to thecore filter.
Drain and sample lines on the various flow loops would be opened and the lines
flushed to in-line filters.
Particulates collected in the filters would be placed in storage
containers.
The water would be circulated throughout the system, with the particulates being captured in acore filter. During circulation, drain valves would be opened to flush particulates to cartridgetype, inline filters.
The advantages of this alternative are that flushing would be largely a remote operation andwould not require removal of major components.
The disadvantage is that the design, construction, installation, and disposal of the specially built core filter would be costly and time
consuming.
For the radiation level estimate for flushing the reactor coolant system, it was assumed that thereactor building and component external surfaces will have been decontaminated to essentiallyclean conditions and that local hot-spots will have been shielded before the flushing operation.
The radiation exposures estimated here apply only to direct effort associated with this flush/drain activity.
To take into account the particulates that could exist in the reactor coolant
system, the staff has assumed that a core debris volume of 2 x 105 cm3 , equivalent to about two
fuel assemblies, is distributed throughout the system, with major collection points at each of
the steam generator inlets and outlets and at the pressurizer bottom.
These particulates would
be collected in several filters during the flush.
Particulates captured in the incore filter
would be removed and placed in smaller containers for storage.
The inline filters containing theaccumulated particulates would be stored.
Changing of filters would constitute the major sourceof exposure for the workers.
Flushing Using System Pressure.
With this alternative, flushing by use of system pressure would
be used.
This procedure would not require the use of reactor coolant pumps or support systems.
The reactor head (or any other suitable cover) would be reinstalled, and the reactor coolant
system then would be filled with water and pressurized with nitrogen in the pressurizer.
This
would provide the necessary pressure for a high-velocity flush of the drain lines and the samplelines through inline filters to the reactor bleed holdup tanks.
The advantages of this alternative are that the design, construction, installation, and disposalof the special core filter would not be necessary.
The disadvantages are that the operation
would require more hands-on operations (with greater worker exposure) and, depending on theparticulate size and location, flushing might not be as effective as forced circulation using thereactor coolant pumps.
Special equipment might have to be designed for use with either alternative to permit use of ahigh-pressure water jet to remove remaining particles from the piping. The equipment could be
pulled, pushed, or self-propelled through the pipe.
This operation would require removal of the
primary pumps and component access covers to gain access to certain portions of the piping.
Spray water would be used to flush particulates to the nearest low-point drain line available.
Any small piping that could not be effectively flushed would be removed and capped.
Reactor Coolant Drain Tank
Flushing and Draining to Reactor Building Sump.
The reactor coolant drain tank now contains
primary coolant and could contain both particulates and plateout.
To implement this alternative,
personnel would have to be able to gain access to a drain valve on the bottom of the reactor
coolant drain tank that would direct the liquid to the reactor building sump.
Preliminary calculations that are based on the soluble fission product concentration in the water and that do nottake into account the presence of particulates, plateout, or other nearby sources of radioactivity, indicate that the radiation levels may be low, perhaps on the order of 50 mR/hr at the tanksurface.
If primary system inspection (Sec. 6.2) indicates that these calculations are accurate,
access would not be a problem; however, if particulates and plateout are present, there may be
hot spots that could limit personnel access without use of shielding and special tools.
The advantages of this alternative are that the operation would be simple, there would be little
impact on schedule,'and no refurbishment of equipment would be necessary.
The disadvantages are6-32
that local shielding could be required and additional radioactive water would be released to thereactor building sump.
Flushing and Draining to Reactor Coolant Bleed Holdup Tanks.
Draining of the reactor coolant
drain tank into the reactor coolant bleed holdup tanks would require replacement or repair of thedrain tank leakage transfer pumps and possibly of the reactor coolant drain pump. In addition,
filtration would be required within the reactor building to remove particulates from the water.
Since the pumps are located outside of the drain tank cubicle, radiation exposure from the tankmay not be a problem; therefore, pumps could be replaced and the tank drained without special
shielding or tooling.
The advantage of this alternative is that no additional radioactive water would be released to
the reactor building. The disadvantage is that system modification and equipment replacement or
repair would be required.
6.5.3.2

Particulate Removal by Dissolution

It is possible that flushing may not adequately remove fuel debris from theRCS. It is very
important to either a rehabilitation program or a decommissioning program that fuel debris becompletely removed since each particle is extremely radioactive.
Should decommissioning becontemplated, this would adversely effect primary system sectioning and disposal efforts. Should
rehabilitation efforts be contemplated, similar problems would occur with required maintenance.
Also, a clean (fuel debris free) RCS would be required to prevent recontamination of the RCS atstartup. One proven solution for dissolution of fuel element debris is oxalic-peroxide-gluconic(OPG). solution. 6 The corrosion resulting from use of this solution is generally minor for mostreactor materials of construction.
If an inventory of materials of construction in the primarysystem revealed potential corrosion problems, appropriate counter measures would be required.
Since the amount of fuel element debris that will be left in the system following defueling andflushing cannot be determined until the RPVH has been removed and the core damage assessed,
development of dissolution equipment schemes are counterproductive at this time. If the amountof fuel element debris in the system is large, several system volumes of dissolving solution may
be required to assure that the solution does not become too radioactive to handle with existing
decontamination solution disposal equipment.
Nuclear criticality would be considered in thedesign of the cleanup equipment.
The OPG solution is only partially effective for removal ofplateout material and, of course, the RCS film will have aged so no forecast can be made regarding
the need for further RCS decontamination until fuel debris removal is complete. In any event, up
to this point in the TMI decontamination, there would be no difference in the approach for either
decommissioning or rehabilitation.
6.5.3.3

Plateout Removal Using Chemical Decontamination Techniques

The chemical decontamination alternatives considered differ in the type of solutions used, methodsof preparation, methods of introducing the chemical solution into the system, and the equipmentused. Decontamination process alternatives are detailed in Appendix P.
Reactor Coolant System
C11 culation Using Reactor Coolant Pumps. This alternative would involve use of the normal reactor
coolant pumps and other inplant equipment to circulate decontamination solutions through thesystem.
The solution would be allowed to flow through the letdown-makeup system and the minidecay-heat-removal system both for decontamination and cooling. Additional heat may be removed
by filling the secondary side of the steam generators with water and maintaining secondary flow.
While the decontamination solution was being circulated, drain lines and sample line valves would
be opened as necessary and solution allowed to flow through the piping to decontaminate it.
Thereactor head and internals also could be decontaminated at this time if they were reinstalled in
the RPV.
At completion of decontamination, the reactor coolant pumps would be stopped and thesolution allowed to drain to the reactor bleed tanks. The advantage of this alternative are thatno major pieces of new equipment would be needed, and systems for solution mixing and transport
already exist. The disadvantage is that existing equipment (pumps) must be operable before
decontamination can proceed.
Circulation Using New Pumping Equipment.
If new pumping equipment, heat exchangers, and tanks
are used to circulate chemical decontamination solution through the reactor coolant system, tanks

6-33
mounted on trucks or inplant tanks would be used to mix the chemical solution and then it would
be pumped to the reactor coolant system. As an example, the mini-decay-heat-removal system could
be used to fill the primary coolant system. The solution could be piped to the intake of themini-decay-heat-removal system pump.
The heat exchanger could be used to heat the solution, oran auxiliary heat exchanger could be installed. The chemical solution would be circulated bysequencing the opening of various drain valves. Piping would be provided from the drain valves
back to the intake of the mini-decay-heat-removal system pump. Flow would be established throughindependent components by opening the drain valves on these components.
The spray, line to thepressurizer would be connected to the outlet of the pump and then the pressurizer drain valve
would be opened to promote circulation.
At the completion of circulation,, the decontamination
solution would be drained from the reactor coolant system into the reactor bleed holdup tanks forfurther processing.
The advantage of this alternative is operation of reactor coolant pumps and other reactor support
systems would 'not be required. The disadvantages are that procurement and installation of additional equipment would be necessary and that decontamination would require more time because of
the lower flow capacity of the new chemical' solution pumping system.
Reactor Coolant Drain Tank
Feed and Bleed.
For this alternative, decontamination solutions would be introduced' into the
14-inch-diameter pressurizer relief valve header near the relief valves to fill the.tank andheader.
Once the tank and header were filled, the solution would be drained to the reactor
coolant bleed holdup tank, while at the same time more solution was being fed into theheader.
The''advantage of this alternative' is that' fewer system modifications would be..required. The
disadvantages are that positive circulation would not be ensured, and more l.iquid wastes would be
generated than for the next alternative..
Recirculation.
For this alternative,, a recirculation 'line and pump would be installed with aheat exchanger to circulate the decontamination solution from the reactor coolant drain tank to
the 14-inch pressurizer relief valve header ata point adjacent to the relief valves and back to
the tank via the header. While the solution was circulating, it would be heated to 180-250OF toincrease the effectiveness of decontamination. After decontamination was completed, the solution
would be drained to the reactor coolant bleed tank for storage and processing.The advantages of this alternative are ensured circulation and minimumliquid waste. The disadvantage is that additional equipment and modification to existing systems would be needed.
Chemical Decontamination of Reactor Coolant Pumps and Motors
Inplace Decontamination.
Inplace decontamination' basically would consist of cleaning accessible
surfaces within the motor housing; draining and flushing the oil lift system; cleaning, inspecting, and replacing pump seals; and making electrical and mechanical checks. This is the procedure that would be used if the pumps were to be operable (perhaps minor repairs might be needed)
and could be used for flushing and decontamination of the reactor coolant system after pumpdecontamination was completed.
Out-of-Containment Decontamination.
Out-of-containment decontamination can be divided into twoareas of effort--(1) motor decontamination and repair and (2) pump shaft and impeller decontamination and repair. Since there are four pumps, it may be found that both types of,,effort or only
partial effort is required for.a given pump.
Under this alternative, the pump motor and pumpshaft' and impeller assembly would be removed from the reactor building to.a laydown area fordecontamination and repair.
The pump motors would be decontaminated and repaired or replaced as needed. In cases of severe
damage, the motors might have to' be shipped offsite for repair (such as for rewinding of the
motor).
The pump shaft, impeller, seals, bearings, and other wetted components also would be
decontaminated, repaired, or' replaced as necessary to the extent that the pumps could be used.

6-34
6.5.3.4

Logistics, Crew Size, Work Effort, Radiation Levels,
Associated Facilities

Waste Generated,

andSpecific methods for treatment of the various groups of reactor coolant system components will
not be selected from among the alternatives until decontamination work is underway.
The staff
has made estimates of crew sizes, work effort, radiation levels, wastes generated, and special
facilities needed for each alternative.
These estimates are summarized in Table 6.10 and discussed below.
Crew Size.

The work crews required for each activity are shown in Table 6.11.,

Work Effort.
The estimates of work effort (person-hours) include a nominal estimate, based ongood working conditions (low radiation levels, clean atmosphere) and a maximum estimate, based on
adverse working conditions (high radiation levels, use of breathing equipment).
The estimates
for the various activities are shown in Table 6.10.
Considerably more effort is required forflushing the reactor coolant system using system pressure than for flushing using the reactor
coolant pumps.
This is because of the greater number of cartridge-type filter changeouts neededin the former alternative.
The alternative of out-of-containment reactor coolant pump decontamination has several possible
subalternatives.
The estimated work effort for each subalternative is given in Table 6.12.
Equipment movement and handling is a significant portion of the work effort.
Radiation Levels.
The average radiation levels estimated for each activity are given in
Table 6.10.
The activity of draining and flushing reactor coolant drain tanks has somewhat
higher radiation levels than for subsequent activities because this work effort could occur
during reactor building decontamination.
The average radiation levels for draining the reactor
coolant drain tanks to the reactor coolant bleed holdup tanks are somewhat lower than draining to
the reactor building sump because the pumps for the former alternative are located outside of thedrain tank cubicle.
Since the chemical decontamination of the reactor coolant system and thereactor coolant drain tanks would be one of the last activities performed,. radiation levels would
be low.
Waste Generated.
The waste expected to be generated by each activity is given in Table 6.13.
The flushing water will probably be processed water, minimizing the overall volume of radioactive
waste water generated.
Special Facility Needs.
A filter backflush facility that would allow particulate removal andsubsequent reuse of cartridge-type filters might be needed for the flushing of the reactor coolant
system using system pressure alternatives.
This would substantially reduce the total number offilters needed.
The facility would be a well-shielded room with a high-pressure water supply andsuitably shielded equipment for containing the backflushed material.
No other special facilities
have been identified.
The out-of-containment reactor coolant pump decontamination alternativeneeds those facilities used for reactor building decontamination,"such as the containment service
building.
6.5.4
6.5.4.1

Effluents and Releases to the Environment
Normal Releases

The decontamination operations will generate solid and liquid wastes.
The solid wastes resulting
from these operations are described in Section 8. The solutions resulting from decontamination
operations are of two types--relatively clear water from simple flushing operations and chemical
reagents containing substantial solids. The treatment of these wastes are described and resulting releases discussed in Section 7.
The staff expects decontaminating reagents to be made up
from processed water; thus the inventories of liquid in the system will not be increased.
Although there will be fission products present in the decontamination solutions, the potential
airborne particulates will be negligible; however, tritium will be released by the evaporationof water from the spent fuel pool.
These releases will be less than those listed for the defueling operation described in Section 6.3.4.1 (< 300 Ci/year).

6-35

Table 6.10

Work Effort, Crew Size, and Radiation Field for Activities
in Decontamination of Reactor Primary System
Work Efforta(person-hours)

Activity

Crew

Radiation Field(mR/hr)Nominal b

Maximumc

Size

NominalMaximum

Using reactor coolant pumps

8,200

13,6008

1030

Using system pressure

34,00040,00081030Drain to reactor
building sump

160

460

3

30

100

Drain to reactor coolant
bleed holdup tanks

500

1,500

3

20

50

Using reactor coolant pumps

5,000

9,000

10

3

10

Using new pumps

7,000

11,00010

310

By feed and bleed

1,100

2,600

8

3

10

By recirculation

2,700

6,600

8

3

10

1,200

2,900

3

10

30

1 , 3 0 0 -4 , 5 0 0 d

2 ,600-10 ,600d

6

20

50

Flushing Reactor Coolant System

Draining and Flushing Reactor
Coolant Drain Tanks

Chemical Decontamination of
Reactor Coolant System

Chemical Decontamination of
Reactor Coolant Drain Tanks

Reactor Coolant Pump
Decontamination
In-placeOut-of-containment

aDirect work in a radiation environment.
bGood working conditions - low radiation levels,clean atmosphere.

cAdverse working conditions - high radiation levels, use of breathing equipment.
dSee Table 6.12.

6-36

Table 6.11.

Work Crews Required for Activities in Primary
System Decontamination
Crew Required

Act Ivity

Flushing reactor coolant system

Five persons for handling and installation ofequipment (two technicians, a supervisor, one
health physics technician, craftsman), four
persons for operation (two technicians, asupervisor, one health physics technician).

Draining and flushing reactor
coolant drain tanks

Two technicians or craftsmen, one health physics
technician.

Chemical decontamination ofreactor coolant system
Handling and installation of equipment

Two technicians, one supervisor, and one
health physics technician, and two craftsmen.
Two technicians, one supervisor, and one
health physics technician.

Operation
Chemical decontamination ofreactor coolant drain tanks
Handling and installation ofequipment
Operation

Three craftsmen, one technician,health physics technician.

and oneTwo technicians, and one health physics
technician.

Reactor coolant pump decontamination
Inplace
Out of containment

Two technicians and one health physics
technician (multiple crews).
One supervisor three craftsmen, one technician, and one health physics technician.

Table 6.12. Work Effort Estimates for Subalternatives
for Out-of-Containment Decontamination of Reactor
Coolant Pump Components
Work Effort(person-hours)
Subalternative

NominalMaximum

Onsite motor repair (motor rewound incontainment service building)

4,500

10,600

Offsite motor repair (motor rewoundoffsite after decontamination)

4,500

9,000

Motor replacement

1,300

2,600

Pump shaft and impeller repair (onsite incontainment service building)

2,800

6,000

6-37

Table 6.13.

Activity

Wastes Generated from Activities in Primary
System Decontamination

.

Waste

,Flushing reactor coolant system
Using reactor coolant pumps

Using .system pressure

Draining and flushing reactor
coolant drain tanks

Chemical decontamination ofreactor coolant system
Chemical decontamination ofreactor coolant drain tanks

Reactor coolant pump decontamination
In-placeOut-of-containment

Two large core filters, one core filter
support structure, about 20 cartridge-type
inline filters. Miscellaneous tools andequipment, contaminated clothing, contaminated cleanup items (rags, brushes, etc.),
and about 235,000 gallons of liquid.
200 cartridge-type, inline filters, miscellaneous tools and equipment, contaminatedclothing, contaminated cleanup items (rags,
brushes, etc.), and about 235,000 gallons ofliquid.

Six cartridge-type inline filters, miscellaneous tools-and equipment, contaminatedclothing, contaminated cleanup items,
two leakage-transfer pumps, and about)
15,000 gallons of liquid.
100,000 to 500,000 gallons of liquid(see Appendix P).
Miscellaneous tools, equipment, piping, contaminated clothing; contaminated cleanup
items; 30,000 to 45,000 gallons of liquid
for feed and bleed (see Appendix P); and15,000 to 25,000 gallons of liquid forrecirculation (see Appendix P).

Miscellaneous tools and equipment, contaminated clothing, and contaminated cleaning
items (see Appendix P). Small quantities
of liquids.
Miscellaneous tools and equipment, contaminated clothing, contaminated cleaning
items (rags, brushes, etc.), pump motors,
pump shafts, pump impellers, and pump seals.
Small quantities of liquids.

6-38
6.5.4.2

Accident Scenarios

To eliminate the possibility of recriticality, it will be necessary to carefully inspect thereactor coolant system to ensure that any remaining fuel particle deposits represent much lessthan a critical mass before the boric acid concentration is reduced below 3000 ppm. Anotherpotential accident involves the spill of decontamination liquid from the RCS into the reactor
It is assumed that 10 percent (2000 Ci)
building while the primary system pumps are operating.
of the maximum activity content in the untreated liquids (see Appendix G, Table G.2) is spilled
before corrective action occurs. Because the solution will be at a temperature of about 200 0 F,
the fraction of activity that would become airborne is somewhat higher thant that at ambient
temperature.
Accordingly, it is estimated that 0.1 percent (2 Ci) of the activity is transmitted
to the building HEPA filters, which release 0.1 percent (0.002 Ci) to the atmosphere in the formof cesium and strontium, as shown in Table 6.14.

Table 6.14. Airborne Releases ofPrincipal Radionuclides due toa Spill of Decontamination Liquids
from the RCS into the Reactor Building
Radionuclide

Total Release(Ci)

Cs-137

1.0 x 10-3

Cs-134

1.6 x 10-4

Sr-90

7.8 x 10-4

Sr-89

8.0 x 10-5

6.5.5

Environmental Impacts

6.5.5.1

Occupational Doses

The cumulative occupational dose, the expected calendar time, and the total work force required
for each activity are listed in Table 6.15.
Occupational dose, calendar time, and personnel required are based upon the person-hours and crew
size given for each activity in Table 6.10. Occupation dose (person-rem) is calculated as the
product of person-hours and radiation field. The calendar time is based upon continuous operaIn all cases the worktion, three shifts per day, with the crew sizes indicated in Table 6.10.
force requirements were determined using a dose to individual workers of 1 rem per quarter,
rather than by the available work hours in a quarter of a year.
The occupational dose associated with these activities ranges from about 108 to 1740 person-rem.
The 108 person-rem value was obtained by adding all the lowest numbers in the "Radiation Dose,
Similarly, the 1740 person-rem value is based on the highest
Nominal" column in Table 6.15.
numbers.
The decontamination operations could be accomplished with a work force of from 85 to
Of the five operations given in Table 6.15, only the first two can be done
500 individuals.
Thus, 85 persons would be needed for these two operations performed concurrently.
concurrently.
The remaining three operations cannot be done concurrently, so that no additional persons need to
be included.
The occupational dose estimates include about 2 to 4 person-rem from inhalation of tritiatedwater vapor. The tritiated water will be produced by ebaporation from the fuel transfer canaland spent fuel pool, which will be filled with processed water having a tritium concentration of
Evaporation from these open pools of water is expected to produce an average
about 0.5 pCi/mL.
tritium concentration of about 2 x 10-6 pCi/mL in the atmosphere in work areas in the vicinity ofTable 6.15.

Occupational Dose, Calendar Time, and Personnel Required for Activities in Decontaminationof the Reactor Primary System

Activity

Radiation Dose a(person-rem)NominalMaximum

Flushing reactor coolant system
Using reactor coolant pumps
Using system pressure

80
300

400
1000

40
200

90
200

100b

400
330b

Drain to containment building sump

5

50

2

6

5

50

Drain to reactor coolant bleed holdup tanks

10

70

7

20

10

70

Calendar Time(days)
NominalMaximum

Personnel Required
NominalMaximum

80

Draining and flushing reactor coolant drain tankC.

Chemical decontamination of reactor coolant system
Using reactor coolant pumps
Using new pumps

10
20

90
100

20
30

40
50

10
20

90
100

Chemical decontamination of reactor coolant drain tank
By feed and bleed
By recirculation

3
8

30

10

70

6
10

30

8
8

30
70

Reactor coolant pump decontamination
In-place

10

90

20

40

10

90

30

500

20

70

30

500

Out-of-containmentaRounded to one significant figure.
bWork performed over three calendar quarters by same personnel.

6-40
these pools (see Sec. 6.3.5.1).
The staff estimates that primary system cleanup tasks will
require workers to spend 5000 to 10,000 person-hours of work effort in these areas. For theestimate of maximum personnel required, 470 persons may be needed for the first two operations,but 500 could be necessary for the last.
Based on the above values of occupational dose and work force required, the expected number of
additional cancer mortalities in the work force would range from 0.014 to 0.23. The added probability that the average individual worker would die of cancer would range from 1 in 2200 to
1 in 6000.
The expected number of additional genetic effects in the offspring of the work force
exposed would range from 0.028 to 0.45.
6.5.5.2

Offsite Doses

The offsite dose estimates to the maximum exposed individual for the decontamination of primarysystem components are listed in Table 6.16. These estimates are based on the source term describedin Section 6.5.4.1 (300 Ci/year of H-3).
The calculational models used to make these estimates
and the interpretation of their results are described in Appendix W. The significance of thesedoses and their human health and environmental consequences are discussed in Section 10.3. The
total-body population dose received by the human population within a 50-mile radius from these
activities was estimated to be 3 person-rem.

Table 6.16. Dose Estimates to the Maximum Exposed
Individual due to Tritium during Decontaminationof Primary System Components

Location

Pathway

Dose (mreml
Total-Body

Nearest b
gardenInhalation
Ground Shine

2.1 x 10-20

Nearest milk

goat

Nearest cow milk
and garden

Vegetable Use

7.2 x 10-2

Total

9.3 x 10-2

Inhalation

1.1 x 10-2

Ground Shine
Goat Milk Use

0
8.2 x 10-2

Total

9.3 x 10-2-

Inhalation
Ground Shine
Vegetable Use

2.3 x 10-20
7.9 x 10-2

Cow Milk Use

3.2 x 10-2

Total
1.3 x 10-1
aDoses were calculated for GI-tract, bone, liver,kidney, thyroid, lung, and skin. The dose estimates
for all organs except bone were the same as that for
total-body. The estimates for bone were zero. Doses
were calculated for four age groups: adults, teenagers,
children, and infants. The largest of each group are
listed in this table. Infant doses are listed for thenearest milk goat location and child doses are listed
for the other two.
bThe basis for selecting the special locations is
described in Appendix W. The actual locations are:nearest garden = 1.05 miles east-northeast, nearest
milk goat = 1.02 miles north, and nearest cow andgarden = 1.05 miles east.

6-41
6.5.5.3

Postulated Accidents

The type of accident for which dose estimates are presented here is the spill of decontamination
liquid from the RCS into the reactor building. The accident scenario and source terms are discussed in Section 6.5.4.2. The calculational models used to make these estimates and the interpretation of their results are described in Appendix W. The significance of these doses is discussed
in Section 10.4. Table 6.17 lists the dose estimates for the maximum exposed individual.
6.5.5..4

Psychological-Socioeconomic Effects

The activities considered in this section include reactor cooling system inspection, removal of
the reactor pressure vessel head and internal components, and decontamination of primary systemcomponents.
The staff concludes that although these activities pose negligible offsite healtheffects, people living in the vicinity of the reactor may view the removal of fuel elements(defueling) and primary system decontamination as threatening.
This option is based on theexisting low level of anxiety generated by the uncertainties surrounding the impacts of theoriginal accident and subsequent decontamination actions.
The level and proportion of the public's anxiety related to the removal of fuel elements is
associated with several factors, including:(1) unforseen delay, danger or controversy surrounding core examination, (2) the condition of the core, (3) availability of technology and skilled
personnel to safely defuel the reactor, and (4) the capability of Met-Ed to manage, and NRC tooversee, the decontamination process.
As additional information becomes available from Met-Ed and the news media, and as the certainty
and scope of the task is formally defined, stressors should shift from uncertainty to believed
threat associated with actual decontamination procedures.
Thus, the local public should haveless concern with the activities as they safely proceed. However, descriptions of the defuelingprocess are highly technical and open to diverse public interpretation. Differing opinions havein the past lead to controversy and to a belief that hidden danger exists. The degree to which
actions remain on schedule and have consistent interpretation will ultimately determine thepsychological impact of decontamination procedures.
In addition, the factuality and balance ofthe media's coverage of decontamination will strongly affect psychological consequences.
The hypothesized accidents, although having negligible offsite effects, are expected to aggravate
existing public concerns and produce increased levels of anxiety. The level and duration ofanxiety would be linked to the length of the accident period, the initiating threat, the credibility of Met-Ed and NRC, the level of controversy or uncertainty generated, and/or the qualityof media coverage.
6.5.6

Economic Costs

The work effort (person-hours) estimated for the various alternatives in Section 6.5.3 include
only the direct work effort, mostly conducted in the reactor building in a radiation environment,
for the described alternatives. To evaluate the economic costs of these alternatives, all associated costs need to be included--labor, equipment, and facilities. Estimates of the economic
costs for the best- and worst-case alternatives are given in Table 6.18. These costs are estimated on a consistent relative basis, and do not necessarily reflect all costs on an absolutebasis. Thus, these costs should be considered only as relative costs for the two scenarios (bestcase vs. worst case). The basis for the costs shown here are presented in Appendix K.
About 45 percent of the labor costs shown in the table are estimated to be incurred in preparation for the direct work effort. Most of this effort consists of training and procedure preparation. Estimated equipment costs consist primarily of special filters, pumps, heat exchangers,
and system piping modifications needed to accomplish the described cleanup activities. Facility
needs in addition to those mentioned in other sections of this document would consist of a filter
backflush facility. This facility would permit removal of particulates from and subsequent reuseof cartridge-type filters.

6-42

Table 6.17. Dose Estimates for the Maximum Exposed IndividualDue to Spilling RCS Water in the Reactor Building

Location
Nearest b

gardengoat

Nearest cow milk
and garden

Liver

Pathway

Total-Body

Inhalation

4.8 x 10-3

7.4 x 10-2

8.8 x 10-4

Ground Shine
Vegetable Use

6.6 x 10-3
1.4

6.6 x 10-3
5.6

6.6 x 10-3
1.5 x 10-2

1.4

5.7

2.2 x 10-2

Inhalation

1.9 X I0-3

3.0 x 10-2

6.6 x 10-4

Ground Shine
Goat Milk Use

6.6 x 10-3
3.0 x 10-'

6.6 x 10-3
1.6

6.6 x 10-3
9.2 x 10-'Total

3.0 x 10-1

1.6

9.2 x 10-1

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

3.2 x I0-3
6.6 x 10-3
1.4
8.8 x 10,2

5.0 x 10-2
6.6 x 10-3
5.6
3.8 x 10-'

6.2 x 10-4
6.6 x 10-3
1.5 x 10-1
1.1 x 10-'

Total
Nearest milk

Dose (mrem)a
Bone

Total
1.5
6.0
2.6 x 10-1
aDoses were calculated for total-body, GI-tract, bone, liver, kidney, thyroid,
lung, and skin. The maximum three organ doses are listed in this table.
Doses were calculated for four age groups: adults, teenagers, children, andinfants. The highest dose estimates for each age group are listed. The dose
estimates for the nearest garden and nearest cow and garden locations are forchildren. The dose estimates for the nearest goat location are for infants.
bThe basis for selecting the special locations is described in Appendix W. The
actual locations are: Nearest garden = 1.05 miles east-northeast, nearest
milk goat = 1.02 miles north, and nearest cow and garden = 1.05 miles east.

Table 6.18. Economic Costs ofDecontamination of Primary
System Components(thousands of dollars)
Item

Low

High

Labor

1210

2410

Equipment

1240

4390

Facilities

100

100

TOTAL

2550

6900

6-43
References--Section 6
1.

"Analysis of Three Mile Island - Unit 2 Accident," Nuclear Safety Analysis Center, NSAC-1,
Palo Alto, CA, July 1979.

2..

"Three Mile Island, A Report to the Commissioners and to the Public," U.S. Nuclear Regulatory Commission Special Inquiry Group, Mitchell Rogovin, Program Director, January 24, 1980.

3.

Memorandum from D.J. Malloy to J.B. Heineman, Argonne National Laboratory,
Limiting Constraints for TMI-2 Recriticality," April 14, 1980.

4.

"Scoping Studies of the Alternative Options for Defueling, Packaging, Shipping, and Disposingof the TMI-2 Spent Fuel Core,"t Allied-General Nuclear Services, AGNS-35900-1.5-79, September
1980.

5.

J.A. Ayres, "Equipment Decontamination with Special Attention to Solid Treatment," Battelle
Pacific Northwest Laboratories, BNWL-B-90, Richland, WA, June 1971.

6.

J.A. Ayres,"Some Theoretical"Decontamination of Nuclear Reactors and Equipment," Ronald Press, 1970.

7.

WATER TREATMENT PROCESSES AND PROCESSED WATER DISPOSAL

Large volumes of water were contaminated as a result of the accident.
Prior to the onset ofprocessing, about 370,000 gallons of this water were present in the Auxiliary and Fuel Handling
Building (AFHB) tanks.
An additional 700,000 gallons of this water collected and is located in
the reactor building (RB) sump and about 96,000 gallons is in the reactor coolant system (RCS).
The water in these three locations is referred to as accident generated water, which is specifically defined as:'(a)

Water that existed in the TMI-2 auxiliary, fuel handling, and containment buildings
including the primary system as of October 16, 1979, with the exception of water which
as a result of decontamination operations becomes commingled with nonaccident-generated
water such that the commingled water has a tritium content of 0.025 pCi/mL or lessbefore processing.

(b)

Water that has a total activity of greater than one pCi/mL prior to processing except
where such water is originally nonaccident water and becomes contaminated by use incleanup.

(c)

Water that contains greater than 0.025 pCi/mL of tritium before processing.

Per the February 27, 1980 agreement between the City of Lancaster, Metropolitan Edison Company,
and the NRC, accident generated water will not be discharged into the Susquehanna River (barring
unforseen emergencies) until the completion of this Programmatic Environmental Impact Statement
or January 1, 1982, whichever is earlier. In accordance with the agreement, there have been nodischarges of accident generated water from the site. There have, however, been releases ofwater from the TMI site which contained minute amounts of radioactivity which originated from
Unit 2.
These releases are made in accordance with the agreement and pertain to the effluentsfrom the Industrial Waste Treatment System (IWTS).
This system processes water from normally
non-radioactive areas of the TMI Units 1 and 2 such as the Turbine Buildings and Control andService Buildings. The water from these areas contains trace amounts of contamination.
The totalreleases from IWTS are estimated in excess of ten million gallons per year. During the last halfof 1980, water containing less than one thousandth (0.001) of a curie of tritium and one ten
thousandth (0.0001) of a curie of other radioisotopes (e.g., cesium) was released from Unit 2
sources. These minor releases are expected to continue throughout the cleanup but have virtuallyno significance on the impacts discussed in this section.
The water in the AFHB has been decontaminated by processing through an ion exchange system referred
to as EPICOR II and the processed water is stored in various tanks onsite. The remaining RB and
RCS water must be either removed from its present location or processed for the cleanup to proceed.
This section discusses the alternatives considered for disposition of RB ahd RCS accident water.
Where these alternatives lead to liquid effluents or processed water, thealternatives considered
for ultimate disposition of this water are also described. Processed water reuse is discussed inAppendix F.
Large volumes of water will be needed for flushing, draining, and decontamination of piping
systems, equipment, and interior building surfaces. These decontamination activities also involve
mixing this water with strong chemical agents and detergents to enhance decontamination effectiveness leading to chemical-based decontamination solutions during cleanup activities. Thesewater-based and chemical decontamination solutions will become contaminated. The alternatives
considered for treatment of these decontamination solutions are also discussed in this section.
7.1

LIQUID WASTE TREATMENT

Liquids involved in the TMI-2 decontamination will require further processing to permit theirsafe disposal in accordance with the staff's proposed use of the effluent criterion in Appendix I

7-1

7-2
as discussed in Section 1.6.3.2; these liquids include those directly generated during the
March 28, 1979 accident (accident water) as well as liquids contaminated during the cleanup
operations. The sources of TMI-2 liquids that will have to be treated and their estimated radioactivity inventories are shown in Table 7.1.

Table 7.1

Summary of Estimated TMI-2 Liquid Waste
Curie Inventory In
Untreated Liquid

Source ofLiquid Waste

Volume(gallons)

Minimum

Maximum

1.
2.

AFHB Chemical Decon Solutions
RB Sump Water

7,000
700,000

60
500,000

60
500,000

3.
4.

RCS Water
RCS Flush and Drain

96,000250,000a

20,00020,00020,000100,000

5.

RB Decon Solutions
150,000a
40,000

90
10

90

6.

(a) Water Based(b) Chemical
RCS Decon Solutionsb(a) Water based(b) Chemical

100,000a
500,000a

2,000
2,000

20,00010

20,000

aprocessed water could be used for the cleanup activities resulting in thegeneration of this liquid waste.
bThe RCS water-based and chemical decontamination processes are mutually
exclusive. Either the water-based or chemical process will be used in
the decontamination of the RCS.

To compare alternative treatment systems on a relative basis, the performance of the systems
considered was inferred from (1) the performance of the EPICOR II system used to clean up AFHB
liquids (2) laboratory-scale tests on zeolite-based ion-exchange systems and (3) the performanceof similar systems used to treat liquid wastes containing radionuclides and chemical contaminantsof the same or similar species to those present in TMI-2 liquids. The assumptions used to characterize treatment system performance are conservative relative to the decontamination factors that
could be achieved.
Moreover, the actual performance of any treatment system can be adjusted byvarying operating criteria to achieve the decontamination factors desired for a particular sourceof liquid waste. Therefore, while the performance of the systems will vary, the staff believes
that the performance characteristics considered are representative and provide a consistent basis
for comparing alternative treatment systems. The performance parameters which characterize eachof the systems considered are discussed in detail in Appendix G.
7.1.1
7.1.1.1

Status and Specific Considerations
Efforts to Date

Processing of the accident water in the Auxiliary and Fuel Handling Buildings began in the fallof 1979 using the EPICOR II system and has been essentially completed. In addition, flushing and7-3
draining of the activity in the AFHB piping and tanks and the subsequent processing of the contaminated water has largely been completed. There will be continued inleakage collected in the AFHB
sump, however, the activity associated with this water is assumed to derive from primary systemleakage to the AFHB.
Impacts related to disposition of this activity are included in the discussions that follow regarding treatment of the activity in the RCS water.
One objective of processing AFHB accident water was to make the AFHB tankage available for contingencies such as the storage of reactor building sump water. The initial composition of thewater in the AFHB before processing, as well as the initial location of the water, is given in
Table 7.2.2 The volume shown in this table is about 370,000 gallons.
Due to in-leakage andflushing operations, the volume actually processed through the EPICOR II system is about 570,000
gallons--501,000 gallons through September 30, 1980 plus about 68,000 gallons thereafter. This
processed accident water is stored in tanks (e.g., the Unit 2 Borated Water Storage Tank) which
in some cases contained other slightly contaminated water. The commingling of these various
liquids has led to a current inventory of about 743,000 gallons. The storage locations of thiswater and its major constituents are presented in Table 7.3.3
About 2200 gallons of chemical decontamination solutions were generated as of September 30, 1980
from AFHB cleanup activities. These liquids were immobilized with vinyl ester styrene (VES) andpackaged for offsite disposal (see Section 8).
7.1.1.2

Projected Requirements

Table 7.4 shows the major radionuclides present in the liquid waste sources identified in Table 7.1.
The characteristics of each liquid waste source in this table are described below.
Reactor Building Sump Water
About 700,000 gallons of contaminated water (sump water) floods the reactor building basement toa level of about 8 ft above the basement floor. This water is primarily from the reactor coolant
drain tank that overflowed when the reactor pressurizer relief valve stuck open early in theaccident. Concentrations of dissolved radionuclides and important chemical species in the reactor
building sump water are given in Table 7.5. Estimates of the filterable solids in the sump water
are given in Table 7.6.4,5 Oils and grease also are likely components of the reactor building
sump water, and the presence of these materials and any colloidal matter must be considered forthe various process alternatives.
Reactor Building Decontamination Liquids (Water Based)
The initial decontamination of the reactor building interior and equipment surfaces could involvethe use principally of water sprays or steam cleaning methods. The volume of liquid wastes that
could be generated in these operations is estimated to be about 150,000 gallons. Processed water
could be used for this activity. The concentrations of radioactvity in decontamination liquids
that could be generated through the use of remote washing techniques have been estimated to beabout 0.2 pCi/mL.
These wash liquids would drain down into the reactor building sump and may be
added to the existing reactor building sump water for processing. This would depend on theradiation fields at the operating levels in the reactor building and on the schedules developedin the various water processing operations.
If necessary, these liquids could also be treatedseparately.
AFHB and Reactor Building Chemical Decontamination Liquids
Chemical decontamination operations in the AFHB and the reactor building are expected to generatechemical waste solutions that have a full range of detergents and complexing agents. The volumeof these chemical decontamination liquids to be processed is expected to be on the order of
40,000 gallons for the reactor building and 7,000 gallons for the AFHB. All the chemical decontamination liquids could be blended, resulting in an average concentration of radioactivity of
about 0.4 pCi/mL.

7-4

Table 7.2.

Composition of AFHB Water Prior to Start of Processinga'b

Location

Vol ume(gallons)

Radionuclide
Concentrations (pCi/mL)
Cs-137
Cs-134
H-3

Reactor coolant bleed tank A

77,250

28

6.5

0.23

Reactor coolant bleed tank B
Reactor coolant bleed tank C
Neutralizer tank A

77,25077,250
8,780

35
35
2.5

7.6
8.7
0.56

0.27
0.29

Neutralizer tank B
Miscellaneous waste holdup
tank, auxiliary building sumpand sumptank, miscellaneous
sump

8,780

3.3

0.72

13,500

10.1

2.4

Tank farm B

93,000

-

--

-

0.03

aThese values pertain to the situation following the accident but prior towater processing.
bBased on memo from W.K. Lehto to J.A. Opelka, Argonne National Laboratory, Subject: Compilation of Information and Data from Visit to TMI-2
on May 6, 1980; May 12, 1980.

Table 7.3 Storage Locations, Volume, and Radionuclide
Concentrations of Processed AFHB Accident Water

Radionuclide Concentrations (pCi/mL)
H-3
Cs-137
Cs-134Storage Tank

Volumeb(gallons)

BWST

422,000

1.1 x 10-'

3.6 x 10-4

1.6 x 10-4

COT-1A

240,000

1.8 x 10-1

4.6 x 10-6

7 x 10-7CCT-1

26,0006.6 x 10-2CCT-2

45,000

6.4 x 10-2

6.8 x 10-6
1.9 x 10-6

1 x 10-6
3 x 10-7

T9Bc

10,0006.6 x 10-2

6.8 x 10-6

1 x 10-6

TOTALS743,000

1.3 x 10-1

2 x 10-4< 9 x 10-s

aBased on Letter (with enclosures entitled,

1.

Radionuclide Distribu-

tion of Auxiliary Building Water, 2. Agenda for Dec. 18, 1980 meetingof Task Group SC-38) from B.J. Snyder, U.S. Nuclear Regulatory Commission, to F.P. Parker, Vanderbuilt University, December 15, 1980.
bRounded.
cNo measured data--values shown are assumed based on CCT-1 data.

7-5

Table 7.4.

Source ofLiquid Waste
1.
AFHB Chemical
Decon Solutions
2.
RB Sump Water
3.
RCS Water

Radioactivity in TMI-2 Untreated Liquid Waste

Average
Radionuclide
Concentrations(pCi/mL)

H-3

Cs-137

Cs-134

2

_e

51

8

1

190
60

2,500

430,000
9,000

66,000
1,500

7,000
7,800

190
820

75
110

4,100

550

820

110

Isotope Inventory (Ci) a

30

Sr-90

Sr-89

4..

RCS Flush & Drainb

100-

49,0007,500

39,0005.

RB Decon Solutions(a) Water based
Chemical(b)
Solutionsb
Decon
RCS

0.2
0.1

-

12
2

1

-

77
8
9,900

1,500

7,800

9,900

1,500

6.

(a)

Water Basedc

50

-

(b)

Chemicalc

10

-

--

7,800
820
aRounded.
bThe curie content corresponds to the maximum estimated values in Table 7.1.
cThese two solutions are mutually exclusive. Either the water-based or chemical decon
solution will be generated during decontamination of the RCS.
dSee Table G.8 for detailed distribution of other radionuclides.
el_,, denotes less than one curie.

Table 7.5. Estimated Radionuclide Concentrations of
Dissolved Constituents of Reactor Builging Sump
Water as of September 30, 1980

Constituent
H-3
Cs-134
Cs-137

Concentration(pCi/mL)

Constituent

0.95

U

0.028

Pu
Na

3.3 x 10-5
1.2 x 103

24
160

Concentration(pg/mL)

2.0 x 103
15
3
10

Sr-90
Sr-89
Nb-95
Zr-95

2.6
0.07
2 x 10-5
2 x 10-5

B
Cl
Al
Ca

Ru-1063 x 10-3

Cu10

Sb-125
Te-125m
Te-127m
Te-129m

0.02
5 x 10- 4
5 x 10-4
2 x 10-4

Fe
K
Li
Ni

1.8
4
1.6
3

Ce-1441-129

2 x 10-3
1.2 x 10-5

Othersdd

P
0.3
S
9
0.5
Zn
aBased on R.E. Brooksbank and W.J. Armento, "Post Accident Cleanupof Radioactivity at the Three Mile Island Nuclear Power Station,"
Oak Ridge National Laboratory, ORNL/TM-7091, February 1980; andD.O. Campbell, "Hot Cell Studies,". Oak Ridge National Laboratory,
presentation to General Public Utilities and U.S. Dept. of Energy
staff members, January 31, 1980, but corrected for radioactive
decay to September 30, 1980.

110

7-6

Table 7.6. Concentration ofRadionuclides in Filterable
Solids in Reactor Building
Sump Water as ofa b

September 30, 1980

NuclideConcentration(pCi/mL)

Cs-134
Cs-137

0.11
0.71

Nb-95

1.7 x 10-3

Zr-95
Ru-106

0.037
0.18

Ru-103

1.4 x 10-4

Ce-141

3.8 x 10-6

Ce-144

0.083

Co-58

7.7 x 10-4

Co-60

0.011

Sr-89+Sr-90

8.6

Ag-110m

5.9 x 10-3

Te-127m

0.27

aTotal solids = 13 x 103 kg (based on
0.5% solids content).
bR.E. Brooksbank and W.J. Armento, "Post
Accident cleanup of Radioactivity at
the Three Mile Island Nuclear Power
Station," Oak Ridge National Laboratory, ORNL/TM-7091, February 1980;
and D.O. Campbell, "Hot Cell Studies,"
Oak Ridge National' Laboratory, presentation to General Public Utilities andU.S. Dept. of Energy staff members,
January 31, 1980, but corrected forradioactive decay to September 30, 1980.

Reactor Coolant System Water (Primary Water)
The reactor coolant system (RCS) contains about 96,000 gallons of water contaminated with fission
products and reactor core debris from the accident.
The processing of this water is subject tosomewhat different constraints than processing AFHB or reactor building sump water.
Althoughremoval of contamination from the primary water is necessary to permit safe access to the reactor
vessel, the decontamination processes and subsequent makeup must still maintain adequate boron toprovide neutron-absorption capability during subsequent reactor defueling.
Estimated concentrations of important constituents in the primary water are given in Table 7.7.6

7-7

Table 7.7. Estimated Concentrations of the Principal
Constituents of the Reactor Coolant Jystem Water
as of September 30, 1980

Constituent
H-3

Cs-134
Cs-137
Sr-89
Sr-90
Nb-95
Zr-95
Ru-106
Sb-125Cr,)centration,pCi/mL)

Constituent

0.08

Na

4.5

B

Concentration(pg/mL)
1.1 x 103
3.7 x 103

29
2.4
23
8 x 10-4
5 x 10-4
0.1
4 x 10-3

Te-125m
Te-127m
Te-129m

1.6 x 10-3
0.03
6 x 10-5

Ce-144

0.03

Co-58

2 x 10-4

aBased on "TMI-Data Base," compiled by Argonne National Laboratory - W.K. Lehto, November 21, 1980.

Reactor Coolant System Flush and Drain Water

The flushing and draining of the RCS will commence following defueling of the core. About
250,000 gallons of water will be generated from the draining and flushing of the RCS and will
require processing.
Processed water could be used for this activity.
The concentrations ofcontaminants expected in the drain and flush water will depend on at least two variables--(1) theamount of radioactivity released to the water during the flush and drain operation and (2) theamount of water that will be required to flush and drain the system. It is estimated that between
20,000 Ci and 100,000 Ci could be present in this water with resulting concentrations ranging
from about 20 pCi/mL to 100 pCi/mL. 7 The choice of processing alternatives depends on the actual
concentration of radioactivity in the water and the volume of contaminated water that must betreated.
Reactor Coolant System Chemical Decontamination Liquids
Chemical decontamination of the reactor coolant system following drain and flush operations may
involve use of relatively mild reagents such as in the CAN DECON proprietary process or morechemically aggressive chemicals such as alkaline permanganate. These decontamination techniques
are described in detail in Appendix P. It is estimated that from 2,000 to 20,000 Ci of radioactivity could be contained in the reactor coolant system decontamination liquids, depending on
the extent of contamination within the coolant system surfaces being treated. If the CAN DECON
process is employed, about 100,000 gallons of decontamination liquids would result. If a moreaggressive chemical decontamination is pursued, as much as 500,000 gallons of liquid might require
eventual processing. This combination of radioactivity and volumes leads to liquid waste concentrations between 1 pCi/mL and 50 pCi/mL. Processed water could be used for these activities.
Some of these liquid wastes in the form of accident water and AFHB chemical decontamination
solutions already have been generated. Other liquids will be generated during the project. The

7-8
relative time periods over which these liquids could be generated and be available for treatment
are shown in Figure 7.1.
7.1.2

Alternatives Considered

The accident generated water in the reactor building sump and the primary system cannot be leftin its present condition and location if the cleanup effort is to proceed. Some of the alternatives considered for disposition of this water involve its cleanup through the use of filtration,ion exchange, evaporation and bitumenization techniques. Others include transfer from its present
location to onsite storage facilities or processing the water for transport and disposal at a
low-level radioactive waste disposal facility. As decontamination solutions are generated, they
too must be either cleaned up, stored, or processed and shipped offsite. The alternatives considered for accident-generated water and decontamination solutions are discussed below.
7.1.2.1

Long-Term On-Site Storage

The onsite storage alternative involves transfer of the accident water from its present locationsin the reactor building and primary system to storage tanks. This water could be transferred totanks within the plant, if available, or to newly constructed exterior tanks.
In either case,the storage tanks would have to be heavily shielded to reduce radiation levels in areas near thetanks.
7.1.2.2

Direct Immobilization

This alternative involves direct immobilization of the accident water using a binder material
such as cement or VES for either temporary onsite storage or offsite shipment to a commercial
shallow land burial facility. The processes that use these binder materials to immobilize liquidradioactive waste to make it acceptable for transport and disposal are described in Appendix H.
Implementation of this alternative would involve construction of a waste immobilization andhandling facility. The activity level of the accident water would require that this facility be
remotely operated.
7.1.2.3, Treatment Processes and Systems
The treatment alternatives considered involve processing the liquid through treatment systems
which remove the contaminants.
One or more of the following processes can be used to achieve thedesired cleanup objective; (1) filtration, (2) ion exchange, (3) evaporation, and (4) bitumenization. The principles involved in each process and their applicability to cleanup of TMI-2 liquid
wastes are discussed below.
Filtration
Filtration is a physical process whereby particles suspended in a liquid are separated from it by
forcing the liquid through a porous medium. As the liquid passes through the porous medium, thesuspended particles are trapped on the surface of the medium or within the body of the medium
itself. The mechanical device which contains or supports the filter medium is referred to as afilter cartridge.
Loading of the filter medium to its capacity requires removal of the particulates from the medium (e.g. backwashable filter) or replacement of the medium.
Replaceablefilters will be used in processing most TMI-2 liquid wastes. Filtration is applicable to TMI-2
liquid wastes as an initial step in a process. It is not an appropriate treatment process in andby itself, because much of the radioactivity is in solution and thus is not removed by filtering.
Ion Exchange
Ion exchange involves the removal of ionic species from an aqueous phase. The earliest applications of ion exchange were "water softening"--the substitution of sodium ions for calcium andmagnesium ions in water to reduce its hardness. These applications were initiated in the late1800's and early 1900's using natural and synthetic zeolites (aluminosilicate minerals). Synthetic ion exchange resins were discovered in the late 1930's and were developed rapidly. At
present, the synthetic organic resins are used in most ion exchange applications.
In practice, the liquid containing the ionic species is placed in contact with the ion exchange
media selected to preferentially remove specific ions. To provide this contact, the ion exchange
media is placed in demineralizer vessels or lines through which the liquid flows. Vessels containing the ion exchange media for treatment of TMI-2 liquid wastes are referred to as (1) prefilter

LU

RB
DECON SOLUTIONSAFHBCHEMICAL DECON SOLUTIONSU.

RBSUMP WATER

o
LU
o

z

"-

Or

ILl

I

PRIMARYSYSTEMWATER

RCS FLUSH& DRAINWATERRCS DECONSOLUTIONS

TIME

Figure 7.1

Typical Sequence for Liquid Waste Processing

7-10
and zeolite liners, (2) cation liners, and (3) mixed bed liners (liners containing anion andcation media).
These designations refer to the type of ion exchange media in the vessel. Once asystem is designed, the performance of the media in each vessel can be adjusted to suit a particular liquid waste treatment need, as specific as individual batches of water to be processed.
Ion exchange is appropriate for accident water and water based decontamination solutions.
It isnot appropriate for treatment of, chemical decontamination solutions because the chemical natureof these liquids would lead to rapid breakdown and plugging of ion exchange media.
Evaporation
Evaporation is the removal of a volatile solvent from a nonvolatile solution by boiling thesolution.
Evaporation is an ancient art. Open, direct-fired pans containing salt water wereused for making salt in the Middle Ages.
Solar energy has long been used for concentrating brinein open ponds.
The solution to be evaporated at TMI-2 is the liquid waste and the volatile solvent to be vaporizedis water, essentially free of contaminants.
The contaminants are retained in the concentrated
solution (or bottoms) while the relatively clean volatiles are vaporized and then condensed asprocessed water or distillates. The major parameters which describe evaporator performance are(1) the decontamination factor, which is defined as the ratio of the contaminant concentrationsin the feed solution and the concentrations in the distillate, and (2) the volume reduction
factor which is the ratio of feed solution volume to the concentrated "bottoms volume." Evaporation is only appropriate for treatment of TMI-2 liquid wastes with low to moderate concentrationsof dissolved solids.
Bitumenization
This is a process which combines evaporation and immobilization in one step. The evaporator
bottoms are immobilized in an asphalt-like material (bitumen) and the vaporized water removed
from the liquid waste is condensed for further treatment as processed water.
Bitumenization is
only appropriate for TMI-2 liquids of low to moderate radioactivity concentration with at least
5 weight percent solids content.
Treatment Systems
The four basic treatment processes can be combined to form treatment systems for TMI-2 liquid
waste.
Figure 7.2 shows a simplified diagram for these systems. The specific treatment systems
considered for TMI-2 liquid waste include the following:
1.

Submerged Demineralizer System (SDS) - This system combines filtration with three stages(zeolite, cation, and mixed bed media) of ion exchange.

2.

Modified Submerged Demineralizer System - This system is similar to the SDS above with the
exception that it employs a larger cation ion exchanger.

3.

SDS/EPICOR II System - This system combines the SDS (alternative number 1) with polishing(or final cleanup) in EPICOR II (alternative number 6).

4.

Zeolite/Evaporator System - This system combines filtration with ion exchange (zeolite
media), evaporation, and ion exchange (mixed bed media).

5.

Zeolite/EPICOR II System - This system combines filtration with ion exchange (zeolite media)
and polishing by the EPICOR II system (alternative number 6).

6.

EPICOR II System - This system combines three stages of ion exchange (layered organic andinorganic, cation, and mixed bed media).

7.

Modified EPICOR II - This system is similar to EPICOR II (alternative number 6) with the
exception that only inorganic (zeolite or zeolite/titanate) media is used in the first stage
ion exchanger.
The titanate is inorganic material which has strong removal specificity
for strontium.

ALTERNATIVESYSTEMFILTERINGSYTMREMOVALGROSS CS/SR

STAGE 2CS/SR
POLISHING

STAGE3FINAL
POLISHINGZEOLITE BEDCATION BED

MIXED BEDEXNDDFLRSDEMIERALZEREXPENDED
FILTER•SUBMERGED
DEMINERALIZEREFFLUENTS

PROCESSED FILTERS,WATER,INORGANIC MEDIAND

SYSTEM (SDS)

MODIFIEDSDS
SYSTEMLREPROCESSEDCATION
BED

BED

SDS/PICOPROCESSED WATER,FILTERS,IfEXPENDED
PREFLTECATONZEOLITEEVAPORATORSYSTEM

WATER,
EXPENDED FILTERS,INORGANIC AND
ORGANIC MEDIAMIXED BEDFZEOLITE

BED

ED

I

rI

ORGANIC MEDIAIXEDREDAND

EXPENDED FILTERS,INORGANIC AND
ORGANIC MEDIA,
EVAPORATOR BOTTOMSEVAPORATORMIXED BEDCATION BED

ED
BED

CATION BED

MIXED BEDPROCESSED WATER,
EXPENDED PREFILTERSAND ORGANIC MEDIAZELTRPROCESSEDZEOLITE AND
TITANATE BEDCATION BED

MIXED BED

WATER,
EXPENDED INORGANICAND ORGANIC MEDIA

EVAPORATORCATION BED

MIXED BED

WATER,FILTERS,
EXPENDED
ORGANIC MEDIA,
EVAPORATOR BOTTOMSEXTRUDERCO

MIXED BEDPROCESSED WATER.
EXPENDED ORGANIC
BITUMENIZEDMEDIA,SOLID
WASTE

ZEOLLEBE

EPICOR 11
SYSTEM-

PROCESSED WATER,
EXPENDED FILTERS.
PREFILTERS, INORGANICAND ORGANIC MEDIAI-,

EPICOR I'SYSTEM

PREFILTER

MODIFIEDEPICOR IISYSTEM

H-1

EVAPOATORPROCESSEDRESINSYSTEMSYSTEMAND

BITUMENRESIN
SYSTEM

Figure 7.2

B

Alternative Liquid Waste Treatment Systems

7-12
8.

Evaporator/Resin System - This system combines filtration with evaporation and two stages(cation and mixed bed media) of ion exchange.

9.

Bitumen/Resin System - This system combines evaporation and immobilization with two stages(cation and mixed bed media) of ion exchange.

The functions of each stage of the systems shown in Figure 7.2 and their applicability to TMI-2
liquid wastes are as follows:
Filtration Units
Filters are required upstream of ion exchange vessels to prevent solids from clogging the ion
exchange media.
First Stage Processing Units
The function of the first stage is to remove large quantities of cesium and strontium from theliquid. In ion exchange vessels, zeolites and mixtures of zeolites and titanates are most effective for the higher concentration liquids. Organic resins can also be used but the radioactivity
loading on these media is limited by the stability of the organic resin media. Thus, first stageorganic resin ion exchange media cannot be used for RB sump water and may not be suitable for RCS(primary system) water and RCS flush and drain water.
The use of evaporators for first stage cleanup is also activity level limited. Bitumenization
has the same limitations as an evaporator for gross removal of cesium and strontrium from highconcentration activity liquids.
Second Stage Processing Units
Once gross removal of cesium and strontium is accomplished, the influent to the second stage isrelatively low in concentration.
Therefore, ion exchange units containing organic resins or anevaporator are both suited for further removal of cesium and strontium and other ionic contaminants.
The vapors produced from first stage evaporation and bitumenization are condensed before
processing through ion exchange vessels.
Third Stage Processing Units
The function of this stage is to polish the liquid through removal of trace contaminants. The
mixed bed designation indicates that both anion and cation organic resins are present in thedemineralizer vessel.
The quality of the effluent from the second stage unit could be such thata mixed bed demineralizer is not necessary.
The details of the alternatives considered for the handling and treatment of TMI-2 radioactive
liquid waste are presented in the following section (7.1.3). Appendix G presents detailed performance characteristics for the treatment systems.
7.1.3

Details of Liquid Waste Treatment Alternatives

Each alternative considered above was evaluated in detail to assess its suitability for treatmentof TMI-2 liquid waste sources.
This section presents the results of these evaluations, provides
the basis for eliminating alternatives from further consideration, and compares the alternatives
judged suitable for treatment of each waste type. Some details of system hardware and facilities
are also presented.
7.1.3.1

Long-Term Storage

The accident water in the RB sump contains about 500,000 curies of radioactivity or about 0.7
curies per gallon.
If this water were placed, unprocessed, in 55-gallon drums, each drum would
contain about 35 curies and exhibit radiation levels of about 30 R/hr on its surface. About
three feet of concrete shielding would have to be placed around this drum to reduce radiationlevels to acceptable levels.
Comparable shielding would have to be placed around storage tanks
that would be needed to store this unprocessed liquid on site.
Tankage could be constructed and appropriate shielding could be installed. However, the majorcontaminants in this liquid have relatively long-half lives and any substantial reduction in the7-13
activity level in the water to render it innocuous would take over 300 years.
Removal of thewater from the storage tanks any sooner would require that it be treated prior to disposition.
Several years would be required to license, design, and construct long-term high activity liquid
waste storage facilities at the site.
Even when constructed, these facilities could not, withadequate assurance, protect the public health and safety by providing isolation of the radioactive
water from the environment for at least 300 years.
This facility would require maintenance, andmonitoring continuously for 300 years.
The staff does not consider this option to be feasible,
considering the length of time required for decay and the inability to ensure adequate institutional controls throughout this period.
Furthermore, it is anticipated that the cleanup will not
be considered complete without processing or removal of the accident water.
7.1.3.2

Direct Immobilization of Accident Water

Under this alternative, accident water (RB sump and RCS) would be immobilized in cement or VES
and either stored onsite or transported offsite for disposal at a low-level radioactive waste
disposal facility.
Table 7.8 summarizes the impact of implementing this alternative based on the approximately
800,000 gallons of accident water in the RB sump and RCS.
The highlights of this summary include:-

Cumulative public radiation dose will be about 75 person-rem-

Cumulative occupational-

Implementation will take 5 years-

Work in the Reactor Building would be deferred for about 4.5 years-

Total cost in 1980 dollars will be about $40 million-

3900 cubic yards of cement would be required.

radiation dose will be about 825 person-rem

Based on these impacts, this alternative is not considered practicable for this high activity
water, principally because of the high radiation doses to workers and the long delay in completingthe cleanup.
Immobilization of AFHB chemical decontamintion solutions is currently being carried out using the
VES technique described in Appendix H. Thus, direct immobilization without evaporation to reducevolume is still being considered for relatively low volume chemical decontamination solutions.
A separate discussion of untreated chemical
8.2.
7.1.3.3

decontamination

solutions ispresented in Section

Zeolite-Based Treatment Systems

This section provides a detailed discussion of the zeolite-based treatment systems which include
the Submerged Demineralizer System (SDS), the modified SDS, the SDS/EPICOR-II system, the zeolite/
evaporator system, and the zeolite/EPICOR-Il system.
The licensee has designed and is presently
constructing an ion exchange system for cleanup of high specific activity liquids.
This system
is designed to operate under water in Spent Fuel Pool B and is referred to as the Submerged
Demineralizer System or SDS.8
A process flow diagram for the SDS system is presented in Figure 7.3.
Filtration is performed by
two filters, one rough and one fine.
The filters are disposable cartridges about 1.7 ft in diameter and 4.5 ft long, with a volume of about 10 ft 3 . Following filtration, the liquid is transferred to a batching tank.
The next step involves processing through two parallel trains of three
ion exchange vessels each.
These vessels are also 1.7 ft ft in diameter and 4.5 ft long and constructed of stainless steel.
The internal volume of each vessel is about 10 ft 3 and contains
8 ft 3 of zeolite (or zeolite/titanate) ion exchange media. This step removes over 99 percent of
the cesium and strontium in the water.
Further removal of cesium and strontium, plus removal ofother ions, results from processing through two downstream parallel cation vessels, each containing
8 ft 3 of organic resins.
These vessels are essentially the same size as the zeolite vessels but
are constructed of carbon steel.
Final polishing is performed by processing through a mixed 3 bedofThis vessel contains about 135 ft
ion exchange vessel about 6 ft in diameter by 6½ ft high.

7-14

Table 7.8.

A.

Impacts of Immobilizationa and Offsite
Disposal of Accident Water

Waste Characteristics
1.
2.
3.

Liquid Volume - 796,000 gal
Average Radionuclide Concentration - 170 pCi/mL
Liquid Content per Drum - 30 gallons

4.
B.

C.

D.

E.

F.

Radioactivity per Drum - 19 Ci
5. Drum Radiation Level - 15 R/hr
Solidification Operations
1.

Facility Design and Construction - 2 years

2.

Number of Drums to be Processed - 26,300

3.
4.

Average Facility Throughput - 29 drums/day
Time Required to Process all Drums - 907 days

5.

Total Time to Implement - 5 years

Packaging and Transportation
1. Drums per Cask Shipment - 14
2. Number of Cask Shipments - 1,800
3. Cask Miles Travelled (Round Trip) - 9.7 million
Resource Use
1.

55-Gallon Drums - 26,300

2.
3.

Cement - 5,300 tons, or 3900 cubic yards
Gasoline/Diesel Fuel - 1.64 million gallons

4.

Burial Site Land Use - 395,000 ft 3

Cumulative Radiation Dose1.
2.

Occupational (handling) - 79 person-rem
Occupational (transport) - 617 person-rem

3.

Occupational (solidification) - 130 person-rem

4.

Public - 75 person-rem due to passing on route, 24 person-rem
due to onlookers near truck

Costs
1.

Design and Construction - $3.5 million

2.

Operation - $3.2 million

3.

Transport and Disposal - $33 million

aBased on cement immobilization.

FILTERING

INFLUENT

SOURCE1I

GROSS CS/SRREMOVALFINAL
POLISHING

CS/SR
POLISHINGI
FILTERS

SDS

I

I

TANK1

MIXED BEDFARZEOLITE BEDS

CATION DEMINSIFLOW RATERESIN VOLUME

1

30 GPMI

u1

I
30 GPM

30 GPMI

1

5 GPM/TRAIN

10 GPM

3

3

8 FT

IL

Figure 7.3/BED

SDS Process Diagram

8 FT /BED

10 GPM
135 FT3

7-16

anion and cation organic resin. As filters or ion exchange media are expended, the filter cartridge or the vessel containing the media are removed and replaced. These expended filter cartridges and ion exchange vessels are then handled as solid waste.
The design flow rates shown on Figure 7.3 show a throughput of 10 gpm. Based on EPICOR II system
9experience, the average (including downtime) expected from an SDS type system is 2-3 gpm.
RB Sump Liquid
Due to the high concentration of radionuclides in RB sump water, about 190 pCi/ml, systems which
use zeolites (or zeolites and titanates) are best suited for cleanup of this liquid waste becauseof the radiation stability of inorganic ion exchange media. 10 Figure 7.4 shows the configurationsof the zeolite systems considered and evaluated.
The performance of the SDS has been independently evaluated using small samples of RB sump waterand laboratory scale ion-exchange columns.' 1
These evaluations indicate that the current design
may not adequately remove colloidal Cs and Sr, Ru, and Sb from the sump water. These limited
evaluations also indicate that process changes may enhance system performance.
First, the size of the cation ion-exchange beds, two at 8 ft 3 each, is not operationally adequate
to remove sodium ions in the RB sump water and the effective decontamination factor across theseion-exchange vessels could be one.
The excess sodium ions in the cation bed effluents also
hamper the performance of the mixed bed ion-exchange vessel reducing the decontamination factors
for the above radionuclides that would otherwise be achieved. The effect of these factors is to
reduce overall performance of the system to the point where effluents need further polishing if
the processed water is to be disposed of by discharge to the Susquehanna River (see Section 7.2).
The SDS cation and mixed bed vessels do provide the benefit of removing sodium, making the taskof subsequent polishing easier. Second, colloidal forms of Cs and Sr that are not easily removed
by once-through ion-exchange treatment appear to be present. Processing modifications such asaddition of heat and holdup time may improve the removal of this colloidal Cs and Sr.
In order to evaluate the performance of SDS relative to alternative treatment systems, it wasconservatively assumed that the worst case conditions described above for Cs, Sr, Ru and Sb
removal occur in the full scale system.
Actual system performance could yield improved decontamination factors in excess of those predicted by the laboratory experiments for these radionuclides.
Additionally, processed water could be recycled through SDS or another polishingsystem to improve water quality.
The alternative referred to as the modified SDS system includes the first suggested change replacing the cation demineralizers with larger resin beds. Two parallel cation beds containing 40 ft 3of organic resin each will enhance sodium removal and lead to substantial improvements in system
performance.
the second modification noted above has not been considered in detail in thissection or Appendix G; however, it might add to the efficacy of the system without increased
waste generation and is considered to be covered by this alternative.
In addition, this alternative is conceived to cover any other small modifications that might be incorporated in the SDS
process as a result of knowledge gained from initial operations, etc. that do not significantly
change the types or amounts of generated wastes.
Other alternatives involve use of the SDS type filters and zeolite beds in series with othertreatment systems. One alternative would process SDS zeolite bed effluent through an evaporator
and the other would polish this effluent by processing through a second system, EPICOR II.
Another variation would merely take SDS effluent after the mixed bed ion exchanger and process it
through the EPICOR II system. The SDS followed by EPICOR II polishing, as well as the otherzeolite-based systems, are considered adequate for treatment of RB sump water.
Primary System Water
The radionuclide concentration in this water is about a factor 3 lower than that of RB sump
water, 60 pCi/mL, because of dilution by makeup water since the accident.
The contaminants
include almost equal quantities of cesium and strontium so particular care must be taken regarding
strontium removal.
Moreover, since this water provides cooling for the damaged fuel, a portionof it must be processed on a feed-and-bleed basis and kept adequately borated.

ZEOLITE I EPICOR II

SDSMODIFIED SDS

SDS I EPICOR IIZEOLITE I EVAPORATOREPICOR II

PREFILTER

Figure 7.4

iCATIONBEDMIXED BED

Alternative Treatment System Configuration for RB Sump Water.

7-18This water contains sodium in concentrations comparable to those in RB sump water. Therefore SDS
treatment of primary system water could be impeded for the same reasons as those described above
under RB sump water.
However, the modified SDS and other zeolite based systems are consideredadequate for treatment of this water.
RCS Flush and Drain Water
This water is expected to have contaminants similar to those present in primary system water;
however, the estimated radionuclide concentrations could range from 20 pCi/mL to 100 pCi/mL.
Since the need to treat this water will not arise until after defueling when there is no longer aneed to maintain boron levels, the system used to treat RB sump water or primary system water
could also be used for treatment of these liquids.
RCS Decon Solutions (Water Based)
Use of the CAN-DECON Technique (see Appendix D) to decontaminate the reactor coolant system could
generate about 100,000 gallons of liquid waste.
This liquid is expected to contain between
2000 Ci and 20,000 Ci which results in radionuclide concentrations between 1 pCi/mL and 50 pCi/mL.
A zeolite bed could be used to strip radioactive ions from the decontamination solution during
the decontaminating phase. During the liquid cleanup phase, a zeolite bed in series with a mixed
resin bed could be used to treat these liquids.
RCS and AFHB/RB Chemical Decontamination Solutions
These liquid wastes will contain relatively high concentrations of chemicals and detergents.
Treatment of these liquids using ion exchange media will lead to chemical breakdown or plugging
of the ion exchange media. Therefore, none of the ion-exchange treatment systems described above
can be used to treat this waste.
This waste is normally treated by an evaporative process or
immobilized directly in a solidification facility.
Processing Rates
All the above zeolite-based treatment systems have comparable throughputs, about 100,000 gallons
per month.
The time required to process this water will, however, depend on how the selected
system is operated.
For once-through treatment of the various liquids, the RB sump water could
be processed in 7 months, the RCS flush and drain water in about 2½ months, and the RCS water
based decon solutions in about 1 month.
For the primary system water, only a portion of the
96,000-gallon inventory can be processed on a once-through basis because the reactor vessel andcooling circuits must remain filled and borated. Two alternatives are available for treatment of
the primary system water. The minimum elapsed time results when about 70,000 gallons are drained
from the primary system and processed on a once-through basis and then the remaining 26,000
gallons used for fuel cooling and boration are processed on a feed-and-bleed basis. This approachleads to processing about 300,000 gallons over a period of about 3 months. If the entire inventory is processed on a feed-and-bleed basis, about 800,000 gallons would have to be processedover a period of about 8 months.
Facilities
The equipment that comprises the zeolite-based system used for cleanup of RB sump water and otherhigh activity water that can be processed prior to defueling would be located in Spent Fuel Pool
B and operated under water for shielding. Figure 7.5 shows a schematic of the SDS system when
installed in the spent fuel pool. The layout shown is equally applicable to the other alternatives. As shown, filters, zeolite vessels, and cation vessels are submerged while the relatively
large mixed bed vessel is operated within a shield cask above the spent fuel pool water surface.
Both spent fuel pools will have to be emptied of equipment prior to the defueling;
If a zeolitebased system is selected for cleanup of RB sump water and is also used for cleanup of other watersuch as RCS flush and drain or RCS water-based decon solutions, the equipment will have to be
relocated to existing onsite space or a new building.

7-19

OTHER
SOURCES
Figure 7.5.

7.1.3.4

Schematic Diagram of TMI-2 SDS Radwaste Treatment System.

EPICOR-II and Modified EPICOR II Treatment Systems

This section provides a detailed discussion of the EPICOR II and modified EPICOR II treatment
systems.
The licensee has successfully operated the EPICOR II ion exchange system for cleanup of
AFHB accident water.
This system was designed to process liquid wastes with specific activities
below 100 pCi/mL.
A process flow diagram for the EPICOR II system is presented in Figure 7.6. As shown, the influent
is first transferred to a batch tank.
For gross cesium and strontium removal, the waste isprocessed through an ion exchange vessel referred to as a prefilter liner. This vessel is about
4 ft in diameter and 4 ft high and contains about 30 ft 3 of inorganic and organic resin in layers
designed to best clean up a particular batch of influent.
Experience to date with processing of
AFHB water shows that over 99 percent of the cesium and strontium in the influent is removed in
the prefilter ion exchange vessel.
The liquids are then processed through a cation exchange
vessel for further removal of cesium, strontium, and other ions.
This vessel is the same size as
the prefilter and contains about 30 ft 3 of cation organic resin.
Final polishing is accomplishedby processing through a relatively large mixed bed ion exchange vessel.
This vessel is 6 ft
diameter by-6 ft high and contains about 115 ft 3 of anion and cation organic resins. When theion exchange media are expended, each of the three vessels are removed from the system and replaced.
Performance of the EPICOR II system during cleanup of the AFHB accident water has exceeded design
expectations.
System performance for a typical batch of AFHB water is shown on Table G.11 inAppendix G. It is reasonable to expect comparable performance if the system is used to process

INFLUENTFILTERINGGROSS CS/SR

CS/SR

REMOVALI]SOURCEI(I CBT IK-1•
T ANK

FINAL
POLISHING

POLISHING

l.

I-1
I

--

I

PREFILTER

I

DEMIN

DEMIN

10 GPM

10 GPMr

10 GPMI

10 GPM
30 FT

Figure 7.6.

EPICOR II

3

Process Diagram.

I

30 FT

3

115 FT3

7-21
other liquids.
has operated.

The average throughput of the system has been about 2 gpm over the 16 months it

A modified version of EPICOR II is also being considered.1 2 ' 13 The modification would consist of
replacing the first stage prefilter with a zeolite ion exchange vessel which would be loaded tohigher radioactivity levels than the prefilters, resulting in corresponding reductions in solid
waste generated.
This vessel would have the same water processing characteristics as the firststage vessels used in the SDS.
Since zeolites would be loaded to higher radioactivity levels
than the expended prefilters, the zeolite bed would be operated within its own shield.
The
modified version of EPICOR II may bypass the mixed bed ion exchanger.
RB Sump Water
The radionuclide concentrations in this water are a factor of 2 greater than the design basis fortreatment with EPICOR II or the modified EPICOR II systems. Therefore, neither system is suitable
for treatment of this water.
RCS Water
Primary system water, RCS flush and drain water, and RCS water-based decon water have, or are
expected to have, specific activities below the 100 pCi/mL design limit for EPICOR II. Therefore, EPICOR II and the modified EPICOR II systems are suitable for treatment of these liquids.
The EPICOR II system generates relatively large volumes of organic resins compared to the othersystems which use higher loaded zeolites for gross removal of contaminants in the first stage.
Moreover, these first stage organic resins would be highly loaded which could lead to potential
solid waste storage and disposition problems as discussed in Section 8.1. The modified EPICOR II
system eliminates the first stage organic resins, reduces the volume of generated solid waste,
and alleviates the concerns associated with the extended storage of higher specific activity
organic resins (i.e., greater than 10 Ci/ft 3 ).
RCS and AFHB/RB Chemical Decontamination Solutions
The chemicals and detergents in these liquids would lead to chemical decomposition of the ion
exchange media.
Therefore, neither EPICOR II nor the modified EPICOR II system are considered
suitable for treatment of these wastes.
These wastes are normally treated with an evaporative
process or immobilized directly in a solidification facility.
Processing Rates
The EPICOR II system was designed for a throughput of 10 gpm but the time averaged throughput(including outages) has been about 2 gpm. Similar performance can be expected for the modified
EPICOR II system. This throughput corresponds to a processing rate of about 80,000 gallons permonth. Once-through treatment of RCS flush and drain water will take about 3 months. Treatmentof the RCS water-based decon water will require an additional 1½ months. As indicated in Section 7.1.3.3 above, the entire primary system inventory cannot be processed on a once-throughbasis.
The minimum time required to process this water would be about 4 months. Feed-and-bleed
processing of the entire inventory would take about 10 months.
Facilities
If either the EPICOR II system or a modifed version of it is selected for treatment of RCS water,
the equipment will remain in its present location. A schematic of the system in its present
locale is presented on Figure 7.7. As shown, each ion exchange vessel is operated within ashield and, when expended, is removed using a transfer shield. If the prefilter is replaced by azeolite ion exchange vessel, a second shield will be placed in the existing prefilter shield tomaintain radiation levels at their current levels during operation of a modified EPICOR II system.
Additional shielding will also be required for the transfer shield.
7.1.3.5

Evaporator/Resin-System

The licensee initiated the process of designing an evaporator/solidification facility to treat
liquid wastes and immobilize evaporator bottoms.. This activity has been deferred but is still
considered an alternative.

7-22

AUX ILIARY/FUELHANDLING
BUILDINGSCONTAMINATED
STORAGE TANKSIJTRANSFER/,•'MIXED*CA7,I

TEMPORARYSTORAGE TANKS*Referred

to as prefilter in text.

CATIONDEMIN••.

oMIXEDMIXE
ATION
DEMIN

OFF SPEC
BATCH TANKI
OFF SPEC
WATER

90,000 GAL.

IMONITOR
TN

TANKI

I

ICLEAN

130,000 GAL.

WATER

TMI-2 EPICOR II

I< I

SHIELDING

RETURN FOR
REPROCESSING

Figure 7.7.

BEDI
NDE

I

Radwaste System.

A process flow diagram for a typical evaporator/resin system is shown on Figure 7.8. As shown,
influent is transferred to a batch tank for pretreatment (e.g., pH adjustment).
Following thisconditioning step, it is transferred to the evaporator. The feed rate can vary from 10 to 30 gpm
for the applications considered for TMI liquids.
The contaminants are concentrated in the evaporator bottoms which are transferred from the evaporator for immobilization and disposed of as
solid waste.
The Volume Reduction Factor (VRF) can vary between 10 and 100, dependent on thecharacteristics of the influent.
A VRF of 100 is representative of water based liquids while therelatively strong chemicals that could be generated from decontamination activities would have a
VRF of 10 to 30.
Thus, the radioactivity concentration in the bottoms could be 10 to 100 times
greater than that in the influent.
The vaporized water referred to as a distillate is condensed.
The condensed water is then processed through demineralizers for removal of soluble contaminants.
A cation demineralizer vessel
and a mixed bed demineralizer may be required.
Use of a 30-gpm evaporator/resin system fortreatment of TMI-2 liquids is discussed below.
RB Sump Water
This water has radionuclide concentrations of about 190 pCi/mL which when processed will lead toevaporator bottoms with concentrations in the 2,000 pCi/mL to 20,000 pCi/mL range.
These highconcentrations will lead to extremely high radiation fields in and around the evaporator facility.
Industry experience with processing even relatively low concentration liquid waste, 0.1 to
10 pCi/mL, in similar evaporators has indicated potential problems.1 4 Most systems have not beenreliable and outage maintenance has been excessive relative to other types of treatment systems(e.g., ion exchange).' 5
Some nuclear facilities have started up their evaporators, encounteredI
I
I

INFLUENTFILTERINGGROSS CS/SRREMOVALCONDENSATE

VAPOR

I

CONDITIONTANKKBEDOEMNFINAL
POLISHINGCONDENSERFILTER

SOURCE

CS/SR
POLISHING__________________________________________________EVAPORATOR

'

NMI

E

DEMIN

BOTTOMSFLOW RATEI

VRF INFLUENT/BOTTOMSRESIN VOLUME:

10-30 QPM

10-30 OPM
10 TO 100
30 FT

IFigure 7.8.

Evaporator/Resin Process Diagram.

3

/BED

110 FT

3

/BED

7-24
problems with their operation and maintenance, and then shut them down permanently. 16 Whenevaporator experience to date is considered in conjunction with the highly concentrated evaporator
bottoms that would arise from processing RB sump water, personnel exposure during outage maintenance would be excessive.
Additionally, evaporator bottoms with radionuclide concentrations in the 2,000 pCi/mL to
20,000 pCi/mL range when immobilized in a solidification facility, -are well above the levels
considered suitable for shallow land burial disposal.
Thus, these materials would have to behandled as high specific activity wastes. The presence of borates and other precipitates in thebottoms and their liquid form also reduce the options available for their disposition as highspecific activity waste.
Neither government nor commercial facilities have produced or treatedhigh specific activity wastes, from evaporators, with similar characteristics.
Therefore, for all of the above reasons,treatment of RB sump water.

an evaporator is not considered suitable for first stage

RCS Water
Primary system water (N 60 pCi/mL), RCS flush and drain water (20 pCi/mL to 100 pCi/mL) and RCS
water-based decon solutions (5 pCi/mL to 50 pCi/mL) all have relatively high radionuclide concentrations compared to the liquids that have been treated to date using evaporators. Thus theabove discussion for RB sump water is applicable to this water and evaporators are not considered
suitable for treatment of this waste.
Chemical Decontamination Solutions
The chemical solutions that could be generated from AFHB/RB cleanup and RCS decontamination are
expected to have relatively low radionuclide concentrations, less than 1 pCi/mL to 10 pCi/mL.
Wastes with similar concentrations and characteristics have been treated with evaporators.
Therefore, evaporation is considered a suitable alternative for treatment of these wastes.
Processing Rates
A 30-gpm capacity. unit is expected to have a time averaged throughput (which includes outages) of
3 to 6 gpm. The design to average throughput ratio (6 to 10) is higher than that considered forion exchange systems (3 to 5) because of the more frequent maintenance outages expected with theevaporator. The anticipated volume reduction factor for the solutions that could be processed
would range from about 10 to 30. Thus the radionuclide concentrations in the evaporator bottoms
could range from 10 pCi/mL to 300 pCi/mL.
The time required' to process 500,000 gallons of decon solutions at 3 gpm would be about fourmonths. The higher throughput of 6 gpm would reduce this to about 2 months.
7.1.3.6

Bitumen/Resin System

The bitumen/resin system described in this section is based on a design which employs one step
evaporation/soli'dification to process TMI-2 radiactive liquid waste.1' 7 "The basic component of
the system is.a :twin screw extruder/evaporator.
The screws rotate inside heavy metal, steam
jacketed barrels.
The system combines a volume reduction and solidification process by simultaneously feeding measured quantities of liquid waste and hot asphalt to the extruder/evaporator
which evaporates the free water, mixes-the remaining radioactive solids with asphalt, and dumps
the mixture into a waiting 55 gallon drum on a turntable.
The mixture typically consists of
50 wt% radioactive solids in asphalt. The extruder/evaporator is equipped with steam domes withintegral condensers and the distillate is directed to a downstream distillate collection and oil
filtration system prior to polishing in cation and mixed bed demineralizer vessels. A process
flow diagram of the bitumen/resin system is shown in Figure 7.9.
The bitumen/resin system is a relatively low capacity system, capable of processing liquid waste
at approximately Q.5 gpm. This evaluation assumes that two systems would be installed to provide
a combined flow capacity of 1 gpm. Additionally, the radiation stability of the bitumen/radwaste*product is adequate to only about 109 rads. Thus, administrative controls would have to be
imposed on waste influent to the system, taking into account the concentration of the waste andanticipated onsite storage to ensure the integrity of the product and its container prior todisposal.

I

CS/SR
GROSS
POLISHINGGROSS CS/SRFILTERING

INFLUENTIREMOVALFITEOILCONDENSER
SOURCEI

CONDITIONINGTANK

DEMINERALIZER

_

_

_

I

I

VOLUME
REDUCTION FACTORIRESIN VOLUMEI_--- MIXEDRBED -DEMINERALIZER]

EXTRUDER
EVAPORATOR

BITUMENIZEDI
FLOW RATE-CAN-CONDENSATEFINALPOLISHING

N4Ln

SOLID WASTE
1 GPM

1GPM

1GPM

3 TO 20
30 FT

Figure 7.9.

Bitumen/Resin Process Diagram.

3

110 FT3

7-26
The staff is well aware that given the low capacity of the system, the application of a bitumen/
resin system in a commercial nuclear power plant is usually associated with the use of a preconcentrator (e.g., crystallizer evaporator).
However, given the higher than normal activity concentrations and/or anticipated solids content of the TMI-2 liquid waste (see Table 7.1), the use ofa preconcentrator was judged to be neither necessary nor desirable. The use of a redundant(i.e., 1 gpm capacity) bitumen/resin system for treatment of TMI-2 liquid wastes is discussed
below.
RB Sump Water
This water has a concentration of approximately 190 pCi/mL (5.3 Ci/ft 3 ) and an estimated solids
content considerably less than 10%. Volume reduction of this water in the bitumen/ resin system
would result in a waste product with an activity concentration considerably in excess of thatwhich is considered suitable for organic material (i.e., up to 10 Ci/ft 3 ).
Therefore, thissystem is considered unsuitable for RB sump water given the uncertainties associated with ultimate
disposition and near-term (on the order of several years) stability of the waste product. Radiation damage to bitumen wastes occurs at approximately 109 rads exposure and is characterized byproduct swelling as a result of the generation of gases such as hydrogen, methane, carbon dioxide
and carbon monoxide.
RCS Water
Primary system water has a concentration of approximately 60 pCi/mL (1.8 Ci/ft 3 ) and a solids
content considerably less than 10%. Volume reduction of this water in the bitumen/ resin system
would be unsuitable for the same reasons as for the RB sump water.
AFHB/RB Chemical Decontamination Solutions
The chemical solutions that are expected to be generated from decontamination activities in the
AFHB and RB are expected to have relatively low activity concentrations, less than 1 pCi/mL to
10 pCi/mL, and relatively high solids content, well in excess of 10%. This waste is consideredparticularly well suited for processing in a bitumen/resin system. Additionally, the total
estimated waste volume (47,000 gallons) is not an unreasonable quantity for a 1 gpm system. The
bitumen/resin system is a high reliability system and it should not take more than several months
to process this waste. Given the high solids content of the waste, the staff estimates that thevolume reduction factor for this waste would be in the range of 3 to 20.
RB Water Based Decontamination Solutions
The water-based decontamination solutions that are expected to be generated from activities in
the reactor building are expected to be mixed with RB sump water prior to processing. Thus, forthe reasons cited above, regarding RB sump water processing, the bitumen/resin system is considered unsuitable for processing this water.
In the event that these decontamination solutions are
not mixed with sump water but are collected separately, the bitumen/resin system would be adequate
for processing this waste.
RCS Flush and Drain Water
The radionuclide concentrations in the RCS flush and drain water is expected to range from
20 pCi/mL to 100 pCi/mL. This is comparable to the RCS water concentration.
Therefore, thebitumen/resin system is not considered suitable for treatment of this water.
RCS Decon Solutions
If the CAN DECON technique is used, the radionuclide concentrations are expected to be in the
5 pCi/mL to 50 pCi/mL range.
At 50 pCi/mL, these concentrations are too high for treatment ofthese liquids with a bitumen/resin system. If one of the other chemical techniques is used the
expected concentrations are about a factor of 5 lower i.e., 1 pCi/mL to 10 pCi/mL.
Chemicalsolutions with concentrations in this lower range are suitable for treatment with a bitumen/resin
system.
7.1.3.7

Summary of Liquid Waste Treatment Alternatives

Table 7.9 shows the alternatives considered for treatment of liquid wastes. A "*" on the tableindicates the alternative is considered suitable for treatment of the liquid waste shown. The

Table 7.9.
(A)

Treatment Alternatives Suitable for TMI Liquidsa
(B)

AFHB
Che l

ChemicalDecontamination
Solutions

(C)

(D)
(E)
Type of Liquid to be Processed
Reactor
BReactor
Reactor Reactor Building

Reactor
Building
Sump Water

Building
Decontamination Water

resin (Mod SDS)*(c) Zeolite/evaporator(d)(e)

Decontamination
Chemicalsc

(F)

(G)

Coolant System

Cooligg
Water

Flush andDraindWater

Decontaminatiop
Water

*

*

*

*

*

*

*

*

*

Zeolite/EPICOR Il

*

*

*

SDS/EPICOR II

*

*

*

*

*

*

Treatment System
Alternative
1. Zeolite Alternatives(a) Zeolite/resin (SDS)(b) Zeolite modified(f) Modified EPICOR II
2.

EPICOR II

3.

Evaporator/resin

4.

Bitumen/resin

(H)

(f)

aNote: * indicates system to be discussed as an alternative method.
bAlternative in the event this water is not processed along with reactor building sump water.
cChemical properties of these liquids are not compatible with purely ion-exchange alternatives.
dIon-exchange processes provide adequate alternatives.
eprocessing of CAN DECON decontamination solutions will probably require system alignment changes.
fOnly if reactor building decon water collected separately from the reactor building sump water.

**

Decontamination
Chemicals

7-28

alternatives indicated by a "* have advantages and disadvantages when compared to each other ona relative basis. The major parameters considered to compare alternatives are listed in Table 7.10
and discussed below.
Time to Install a System
The time required to place a system into operation depends on many factors. Some systems require
the design, construction and checkout of equipment and facilities (including new buildings) ,while others do not. The relative time spans among alternatives are indicative of when cleanupof accident water could begin.
Time to Implement
Once the facilities and equipment needed to treat liquid wastes are operational, the performance
characteristics of the process determines how long it will take to treat the liquid waste. The
processing time is also an indication of the level of manpower that would be required to process
the water.
Liquid Effluents Produced
Some systems produce a liquid effluent. The inventory of radionuclides in these effluents is ameasure of relative treatment system performance.
After processing, these effluents must be
stored onsite for reuse or ultimate disposition. Whether liquid effluents satisfy the criteria
for outside storage is another measure of relative treatment system performance.
Solid Wastes Produced
Each system also produces solid wastes, some of which require further processing and handling
before disposal.
The characteristics of these solids determine the extent of additional processing
and handling needed and in some cases limit the options available for storage and ultimate disposal.
Thus, the characteristics of the resultant solid wastes and their ability to be handled
and readily disposed of is another measure of relative system performance.
7.1.4
7.1.4.1

Effluents and Releases to the Environment
Normal Operations

Effluents and releases during treatment of the liquid wastes described above for the processes
considered are presented in this section. Releases attributable to airborne material (aerosol)
resulting from liquid processing are shown in Table 7.11.
Releases are shown in two ways:(1) the total release for an operation, expressed in curies, and (2) the concentration of therelease, expressed as microcuries per millileter (pCi/mL), of gaseous effluent. The bases forthe calculation of releases are given below.
As a result of various processing operations involving liquid materials, small quantities of theliquids become airborne and enter the vapor spaces above the liquid. The composition of suchaerosols is identical (at the time of formation) to that of the liquid from which it was derived.
The liquid in the aerosol evaporates, leaving suspended residual solids. The airborne material
is carried by subsequent movement of air in process vent systems, then ultimately combined withbuilding ventilation air, treated by the air cleaning system for the building, monitored byradiation detectors, and finally released to the environment by controlled discharge. If necessary, releases to the environment can be stopped by shutting down the ventilation system. Whilein transit larger particles settle out rapidly, while very small particles remain suspended andtend to agglomerate with each other to form larger particles. The air cleaning system includes
one or more stages of HEPA filters which remove the bulk of the solid airborne material.
Volatile
materials, such as tritiated water vapor are not affected by the air cleaning system.
The gaseous effluents and releases from these operations attributable to the formation of aerosols
are dependent on four principal factors:(1) the concentration of the radionuclides in theliquid being treated, (2) the processing rate, (3) the fraction of the material processed that
becomes aerosolized, and (4) the fraction of such materials that passes through the air cleaningsystem uncontrolled.
To calculate effluents and releases, the staff used a throughput of 10 gallons per minute, thedesign basis. The time averaged throughput, i.e., gallons processed over several months, will beTable 7.10.

Comparison of Treatment Systems
Parameter

Time

Liquid Effluents
Suitable for Relative Quality
OutsideBased on Curie
Storage
Content

High-Specific-Activity Waste
Suitable forStableDisposable atDuringRoutine Shallow
ExtendedLand Burial
Produced Storage
Facilities

To Install
SystemTo Process
100,000
Gallons

Produced

Direct immobilizationa

3 yrs

3 mos

No

NA

NA

-No

NA

NA

Long term storage
Evaporator/resin systema
Bitumen/resin systema

3 yrs
3yrs
.3 yrs

1 wk
1 mo
4 mo

NA
Yes
Yes

NA
Yes
Yes

NA
High
High

NA
No
No

NA
NA
NA

NA
NA
NA

3 mos
6 mos

Imo
1 mo

Yes
Yes

Yese
Yes

Low
Low

Yes
Yes

Yes
Yes

No"
No

Zeolite/Evaporator

3 yrs

1 mo

Yes

Yes

High

Yes

Yes

No

Zeolite/EPICOR II
SDS/EPICOR II

6 mos
3 mos

1.5 mos
1.5 mos

Yes
Yes

Yes
Yes

High
High

Yes
Yes

Yes
Yes

No
No

Modified EPICOR IIc
EPICOR II Systemc

3 mos
0

1.5 mos
1.5 mos

Yes
Yes

Yes
Yes

High
High

Yes
Yes

Yes
No

No
Nod

Alternative

Zeolite based systems
SDS
Modified SDS

b

IN.

aConsidered suitable only for treatment of chemical decontamination solutions. Table entries apply only to this waste-type.
bConsidered suitable only for treatment of waste water, not chemical decontamination solutions. Table entries apply only tothese waste types.
cConsidered suitable for all sources of waste water except reactor building sump water.
waste types.
dRefer to discussions in Sections 8 and 9.
eDependent upon radionuclide distribution, see Section 7.2.4.2.

Table entries apply only to theseTable 7.11.

Gaseous Effluents and Releases Attributable to Aerosol Formationfrom Processing of Various Liquids (Normal Operations)
Identification of Process Liquid

NuclideRB Sump

RCS Decon
Water

RCS Decona
Chemicals

RB Decon
Water

Decon
Chemicalsa

RCS Primary
Water

RCS Flush
and Drain

b

*

*

*

*

4.9 x 10-3

9.9 x 10-4

9.9 x 10-4

Normal Releases, Total Curies
Cs-137

4.1 x 10-2

6.4 x 10-6

1.2 x 10-5

2.7 x I0-3
9.9 X 10-4

Cs-1346.4 x 10-3

9.8 x 10-7

1.8 x 10-6

1.5 x 10-4

7.7 x 10-4

1.5 x 10-4

1.5 x 10-4

Sr-90

6.9 x 10-4

7.6 x 10-8

1.4 x 10-7

7.8 x 10-4

3.9 x 10-3

7.8 x 10-4

7.8 x 10-4

Sr-89

1.9 x 10-5

3.0 x 10-9

5.6 x 10-9

8.2 x 10-5

4.1 x 10-4

8.2 x 10-s

8.2 x 10-5

H-3

2.5 x 10-i

Normal Release, Concentration, pCi/mL
H-3

1.1 x 10-8**

1.1 x 10-9

*

*

*

Cs-137

1.9 x 10-92.8 x 10-10

2.2 x 10-12

1.7 x 10-11

3.9 x 10-10

3.5 x 10-10

7.0 x 10-11

3.4 x 10-13
2.7 x 10-14
1.0 x I0-15

2.7 x 10-12
2.0 x 10-13
8.3 x 10-15.6.0 x 10-11
3.1 x 10-10
3.2 x 10-11

7.0 x 10-10
1.1 X 10-1°

5.4 x 10-11
2.8 x 1 0 -10
2.9 x 10-11

1.0 x 10-11

Cs-134
Sr-90
Sr-89

3.0 x 10-11
8.2 x 10-13

5.5 x 10-10
5.8 x 10-"

5.5 x i0-11
5.8 x 10-12

aChemical-solutions are processed through evaporator/resin or bitumen/resin system.
bNote: * indicates that if processed water is recycled for these applications, tritium concentrations will approximate that.in processed water used.

7-31
about one-fourth of this value. This rate is typical of that expected to be used for the operation considered and is based on operating experience to date with EPICOR 11.19 A conservativevalue of 1 x 10-4 has been used as the fraction of material processed that is aerosolized andultimately reaches the air cleaning system. For the processing operations considered, the actual
value is expected to be much less. These statements are based on the operating experience withEPICOR-II and the experience accumulated in the processing of irradiated nuclear fuels. The
processing of nuclear fuels consists of a.large number of complex operations that include fueldissolution, solvent extraction, and evaporation. The systems considered for treatment of TMI-2
liquids are, by contrast, less complex and involve only a few process steps that are simple innature and thus should result in considerably less aerosolized material.
The efficiency of the air cleaning system for removing solid airborne materials (aerosols) is
assumed to be 99.9% (stated alternatively, this corresponds to a penetration of aerosols throughthe'system of 0.1%, or 0.001). Actual penetration values are expected to be much lower based on
the theory of operation of air cleaning systems and based on operating experience.
The key
element in the cleaning system for removal of airborne materials is the HEPA filter. At TMI, twostages of HEPAs in series are present in the cleaning-system. The principal characteristic of
the HEPA filter is that it is very efficient for collecting very small and very large particles.
The specification for the removal efficiency of each installed HEPA filter for 0.3 pm particle
sizes is 99.97% (corresponding to a penetration fraction of 0.0003).
This efficiency is demonstrated to be within specification by suitable in-place testing. A second HEPA filter in series
with the first will remove 99.97% of the material passing through the first HEPA. Thus, it is 20
evident that the assumed 99.9% efficiency for the system is conservative and readily achievable.
To calculate the release concentrations, expressed as pCi/mL of gaseous effluent, the ventilation
flow rate passing through the air cleaning system must be known. Because the locations of allprocessing operations considered in this section are not known at this time, the specific appropriate ventilation rates for these calculations also are not known. Therefore, all calculations
have been normalized to a ventilation flow rate of 10,000 cfm. If, for example, actual ventilation rates are doubled (i.e., 20,000 cfm), the calculated concentrations would be one-half of
those shown in Table 7.11.
This assumption has no effect on the total releases, expressed ascuries.
7.1.4.2

Accident Conditions

The failure of a HEPA filter is considered a credible 'accident event (i.e., an accident that
could occur).
The operation of the air cleaning system is constantly monitored, and a HEPA
filter failure would be detectable in a matter of minutes. However, it is assumed here that theaccident conditions would exist for a period of .15 minutes beforeremedial action is taken.
Experience has shown that when a HEPA filter fails, a fraction of its inventory of previously
filtered materials would be released.
A representative value of 0.001 has been used for thisfraction.
The releases calculated to result from such an accident are shown in Table 7.12.
7.1.5

Environmental Impacts

7.1.5.1

Occupational Doses

AFHB Water Processing (EPICORII) Experience 9

The EPICOR II system is designed for use in decontaminating radioactive Water with minimal personnel radiation exposure.
The EPICOR II operations which cause occupational radiation exposure
are: operation and sampling (about 20 percent of the occupational dose), liner changeout (about
40 percent), and maintenance and outages (about 40 percent).
Liners have been placed in aninterim onsite storage facility in a dewatered condition; none of the liner contents have been
immobilized to date.
In general, there has been a decreasing trend in the occupational doses incurred during the processing of AFHB water through the EPICOR II system. For the first batches handled, only about
14,000 gallons of water were processed per person-rem of occupational dose. As of May 5, 1980,
225,000 gallons had been processed through EPICOR II, with-a cumulative occupational dose of
8.4 person-rem, equivalent to about 30,000 gallons of water processed per person-rem of occupational dose. 7 As of June 8, 1980, about 331,000 gallons had been processed with a cumulative
occupational dose of 11.4 person-rem, 8 corresponding to 29,000 gallons processed per person-rem,
mated unit maintenance dose of about 0.003 person-mrem per gallon processed.
On
this basis, the total maintenance cumulative dose is about 2 person-rem.

Table 7.12.

Gaseous Effluents and Releases from Failure of a HEPA Filter
during Processing of Various Liquids
Identification of Process Liquid

NuclideRB Sump

Accident Releases,

Total Curies

RB Decon
Water

Decon
Chemicals

RCS Primary
Water

RCS Flush
and Drain

RCS Decon
Water

RCS Decon
Chemicals

Cs-137

4.3 x 10-2

6.4 x 10-6

1.2 x I0-5

9.9 x 10-4

4.9 x 10-3

9.9 x 10-4

9.9 x.10- 4

Cs-1346.4 x 10-3

9.8 x 10-7

1.8 x 10-6

1.5 x 10-.4

7.7 x 10-4

1.5 x 10-4

1.5 x 10-4

Sr-90

6.9 x 10-4

7.6 x 10-8

1.4 x 10-7

7.8 x 10-4

3.9 x 10-3

8.0 x 10-4

7.8 x 10-4

Sr-89

1.9. x 10-5

3.0 x 10-9

5.6 x 10-9

8.2 x 10-s

4.1 x 10-4

8.4 x 10-5

8.2 x 10-5

Accident Release, Concentration, pCi/mL
1.5 x 10-9
1.0 x 10-5
Cs-1372.8 x 10-9

2.3 x 10-7

1.2 x 10-6

2.3 x 10-7

2.3 x 10-7

3.6 x 10-8
1.8 x 10-7

1.8 x 10-7
9.2 x 10-7

3.6 x 10-8
1.9 X 10- 7

3.5 x 10-8
1.8 X 10-7

1.9 x 10-8

9.6 x 10-8

2.0 x 10-8

-1.9 x 10-8

Cs-134

1.5 x 10-6

2.3 x 10-1o

Sr-90

1.6 x 10-7

1.8 x 10-11

4.3 x 10-10
3.2 x 10-11

4.4 x 10-9

7.1 x 10-13

1.3 x 10-12

Sr-89

L
F\)

7-33
still a marked improvement over the dose incurred for initial batch processed.
As of August
1980, the bulk of the AFHB water had been processed. About 55,000 Ci had been removed from about
500,000 gallons of water.
The cumulative occupational dose incurred was a little less than
15 person-rem.
The expected number of additional cancer mortalities in the work force exposed to a cumulative
dose of 15 person-rem would be 0.002. The expected number of additional genetic effects in theoffspring of the work force exposed to this cumulative dose of radiation would be 0.004.
This EPICOR II experience is relevant to the exposure that could be received by personnel operating
similar systems.
Table 7.13 summarizes this experience in terms of various unit exposures.

Table 7.13.

Summary of EPICOR II System Occupational Exposure Experience

Operational Authority
1.

Total Cumulative
Dose Received(person-rem)

Operation and Sampling

Unit Dose

6.0 x 10-6 persdn-rem per gallon
processed.
2.
Liner Changeout
6
8.5 x 10-2 person-rem per liner
removed.
3. Maintenance and6
1.2 x 10-s person-rem per gallon
Outages
15a
processed.
aOverall average performance shows 3 x 10-5 person-rem per gallon processed and about3

2.7 x 10-4 person rem per Ci removed.

Reactor Building Sump Water
This water is assumed treated with ion exchange systems which may be submerged in the spent fuelpool.
These systems include SDS, modified SDS, zeolite/evaporator, zeolite/EPICOR II, and SDS/EPICOR II (see Section 7.1.3.1).
The system will be operated from a remote control console
similar to the EPICOR II arrangement.
Thus, the occupational exposure incident to sampling andoperation will be comparable to EPICOR II. The average dose for these activities was about
0.006 person-mrem per gallon.
Thus, cleanup of RB sump water could lead to cumulative doses ofabout 4.2 person-rem.
The dose due to liner changeout for an SOS type system should be below that experienced withEPICOR II since liners will be changed out underwater.
The number of liners to be handled isabout twice the number handled for EPICOR II because zeolite liners have to be sequenced. Basedon fields a factor of 10 below those during EPICOR II liner changeout and twice the handling, theliner changeout cumulative dose should be a factor of 5 below the EPICOR II experience or about
1.2 person-rem.
Maintenance on the submerged systems will also be performed through the water shielding. However,
the time required to perform this maintenance remotely will increase. Combining these two factors
leads to an estimated unit maintenance dose of about 0.003 person-mrem per gallon processed. On
this basis, the total maintenance cumulative dose is about 2 person-rem.
Based on the above assumptions, the total estimated cumulated dose for treatment of 700,000
gallons of RB sump water using a submerged ion exchange treatment system is about 7.4 person-rem.
If this water is further treated by polishing with a second system such as EPICOR II, the estimated additional dose would be about 3 person-rem for operation, about 0.5 person-rem for liner

7-34changeout and about 0.5 person-rem for outages and maintenance for a total of 4 person-rem.
These lower doses are due to the much lower activity water that would be processed through asecond polishing system.
The sump water is expected to be processed over four quarters of the year. The cumulative dose
range of 7.4 to 11.4 person-rem incurred over four quarters is equivalent to 1.9 to 2.9 person-rem
per quarter.
Based on an individual dose of 1 rem per quarter, a work force of 2 to 3 persons
would be needed.
The expected number of additional cancer mortalities in the work force exposed to this cumulative
radiation dose is less than 0.001. This means that the added probability that the average individual worker would die of cancer is about one in 1400. The expected number of additional genetic
effects in the offspring of the work force exposed to this cumulative dose is about 0.002.
Reactor Coolant System and Flush/Drain Water
Since these two sources of liquid waste will probably be treated by the same system, they havebeen combined in this section. The total volume of liquid to be cleaned up is about 350,000 gallons containing between 40,000 Ci and 120,000 Ci.
However, feed and bleed processing of thecoolant system water will result in about 530,000 gallons being processed through the selected
treatment system.
EPICOR II as well as a modified version of EPICOR II is being considered for treatment of thiswater.
The most recent experience with EPICOR II indicates an average cumulative dose of
1 person-rem per 30,000 gallons processed. Based on this experience, treatment of the 530,000 gallons could result in about 18 person-rem.
At the higher curie content, about three times thenumber of liners will have to be changed out, leading to about 14 person-rem of additional dose.
Thus, use of these systems could lead to a cumulative dose between 18 and 32 person-rem.
If an SDS type ion exchange system is used for cleanup of this water, the system could be operated
submerged for cleanup of coolant water but would have to be moved to a different locale forcleanup of flush and drain water.
Using assumptions similar to those described above for RB sump water, the estimated cumulative
dose for processing 280,000 gallons of water containing 20,000 Ci is 3 person-rem.
Estimates for
subsequent cleanup of 250,000 gallons of RCS flush and drain water can be based on the EPICOR II
summary in Table 7.13. To remove 20,000 Ci, the cumulative dose estimate is about 5 person-rem.
Removal of 100,000 Ci will increase this to about 7 person-rem due to the increased number ofzeolite liners to be changed out.
Therefore, the cumulative dose can range from a minimum of about 8 person-rem (40,000 Ci removed
by an SDS type system) to a maximum of 32 person-rem (120,000 Ci removed by an EPICOR II system).
The staff assumes that the crew size for the processing of this water will be the same as thatused for processing the water in the AFHB.
At a rate of 23,000 gallons per week (based upon
EPICOR II experience), processing of 530,000 gallons will take about 23 weeks.
The expected number of additional cancer mortalities in the work force of 16 persons receiving
this cumulative dose of radiation (8 to 32 person-rem) ranges between 0.001 and 0.004. Thismeans that the added probability that the average individual worker would die of cancer ranges
from one in 4000 to one in 7600. The expected number of additional genetic effects in theoffspring of the work force receiving this cumulative radiation dose ranges between 0.002
and 0.008.
Chemical Decontamination Liquids
The relatively low volume of AFHB/RB chemical decontamination' solutions could be directly immobilized. The dose estimates for immobilizing these liquids are presented in Section 8.2.
The relatively large volume of chemical solutions that could be generated from RCS decontamination
with one of the chemical techniques could be treated with either an evaporator/ resin or a bitumen/
resin system.
This 500,000 gallons of chemical solution could contain between 2,000 Ci and
20,000 Ci and the estimated doses for removing 20,000 Ci are presented below.

7-35
Evaporator Resin Alternative.
The average throughput of an evaporator that could be used would
range from 3 gpm to 5 gpm. The unit would be operated from a remote control console so that the
EPICOR II experience is relevant.
This experience indicated that 0.006 person-rem was received
per gallon. Thus, for the lower throughput, cumulative personnel dose for these operations would
be about 2.5 person-rem.
For the higher throughput of 5 gpm, this would be reduced to about
1.5 person-rem. The dose arising from resin vessel changeout will be lower than that experienced
with EPICOR II since the ion exchange media is only used for polishing. For EPICOR II, about
0.085 person-rem was received per liner changed-out. Processing of 500,000 gallons would requirechangeout of about 5 resin liners. Conservatively using the EPICOR II experience leads to a
cumulative dose of about 0.5 person-rem.
On the other hand, maintenance on an evaporator will
result in higher dose than EPICOR II experience.
For EPICOR II, this cumulative dose was about
1.2 x 10-s person-rem per gallon. For the evaporator, maintenance exposure is estimated to be afactor of 5 to 10 higher than this, or 6.0 x 10-5 to 1.2 x 15-4 person-rem per gallon. At 500,000
gallons, the maintenance dose could range from 30 person-rem to 60 person-rem. Thus, the total
cumulative dose for processing 500,000 gallons could range from a minimum of about 32 person-rem
to a maximum of about 63 person-rem. If the activity of the water is 1 pCi/mL instead of 10 pCi/mL,
this dose could be reduced by as much as 50 percent.
Bitumen/Resin Alternative.
The bitumen/resin system is a high reliability unit and the time
averaged throughput will approximate the design throughput of 1 gpm. The unit would be operated
from a remote control console so the EPICOR II experience of 0.006 person-rem per gallon isrelevant.
For a throughput-,of 1 gpm, the average cumulative dose would be twice the EPICOR II
figure or 0.012 per gallon--6 person-rem for 500,000 gallons. Since this unit uses ion exchange
vessels to polish distillate, the number of vessels to be removed will be comparable to thoserequired with the evaporator resin unit described above.
Conservatively using the EPICOR II
experience figure of 0.085 person-rem per liner, the dose incident to changeout will be about
0.5 person-rem.
The extruder-evaporator units require a minimum amount of maintenance compared to an evaporator/
resin system. The lower throughput of two units, I gpm, is compensated for by less outage so themaintenance requirements for these units would be comparable of those experienced by the 2-gpm
EPICOR II-system, 1.2 x 10-5 person-rem per gallon. At this dose rate, processing approximately
500,000 gallons will result in a maintenance outage dose of about 6 person-rem. Thus, the total
estimated cumulative dose to process 500,000 gallons is about 12.5 person-rem.
Summary.
If RCS chemical decontamination solutions are treated with an evaporator/resin system,
the estimated occupational cumulative dose could range from 32 to 63 person-rem.
Processing thesame liquid waste through two bitumen/resin systems could result in an occupational cumulative
dose of about 12.5 person-rem.
Thus, the estimated cumulative dose incident to processing RCS
chemical decontamination solutions could vary from 12.5 person-rem over about 11 months to a
maximum of 63 person-rem over about 4 months.
The estimated cumulative dose for cleanup of RCS chemical solutions, if they are generated,
ranges from a minimum of 13 person-rem to a maximum of 63 person-rem. For the evaporator/resin,
64 person-rem will be received over about 5 months. For the bitumen/resin system, 13 person-rem
will be received over about 25 months.
The expected number of additional cancer mortalities in the work force receiving to this cumulative radiation dose would range from 0.002 to 0.008. The expected number of additional genetic
effects in the offspring of the work force receiving this cumulative dose would range from 0.0034
to 0.016.
7.1.5.2

Offsite Doses

The dose estimates presented here for processing of accident and decontamination water are basedon the source terms developed in Section 7.1.4.1 and listed in Table 7. 11. The calculationalmodels used to make these estimates and the interpretation of their results are described inAppendix W. The significance of these doses and their human health and environmental consequences
are discussed in Section 10.3. The dose estimates to the maximum exposed individual for processing the water are listed in Table 7.14. The 50-mile total body population dose associated withthe processing was estimated to be about 2 person-rems.,

7-36

Table 7.14. Dose Estimates for the Maximum Exposed Individual due
to Processing of Accident and Decontamination Water
Dose (mrem)a
Location
Nearest b

gardenNearestmilk goat

Pathway

Total-Body

Bone

Liver

Inhalation

2.9 x 10-3

5.5 x 10-3
2.3 x 10-1

4.0 x 10-2
5.5 x 10-3

2.5 x 10-3

Ground Shine
Vegetable Use

5.5 x 10-3
1.2 x 10-1

Total

2.4 x 10-'1.0

1.3 x 10-1

Inhalation

3.3 x 10-3

1.5 x 10-2

1.7 x 10-3

Ground Shine

5.3 x I0-3

5.3 x 10-3

5.3 x 10-3

Goat Milk Use

9.9 x 10-2

7.2 x 10-1

7.3 x 10-1

1.1 x 10-1

7.4 x 10-1-

7.4 x 10-1

Inhalation
Ground Shine

3.1 x 10-3
8.1 x 10-3

4.5 x 10-2
8.1 x 10-3

2.8 x 10-3
8.1 x 10-3

Vegetable Use
Cow Milk Use

3.4 x 10-1
3.8.x 10-2

1.4.
2.0 x 10-'

1.8 x 10-1
1.4 x 10-i

3.9 x 10-1

1.7

3.3 x 10-1

Total
Nearest cow
and garden

9.6 x 10-1

Total

aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung and skin. The maximum three organ doses are listed in thistable. Doses were calculated for four age groups; adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and nearest cow andgarden locations are for children and for the nearest goat location are
for adults for total-body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

7.1.5.3

Postulated Accident Effects

The type of accident for which dose estimates are presented is the rupture of a HEPA filter while
processing accident or decontamination water.
The accident scenarios are described in
Section 7.1.4.3 along with their source terms. The calculational models used to make theseestimates and the interpretation of their results are described in Appendix W. The significanceof these doses is discussed in Section 10.4. The estimated doses to the maximum exposed individualfor each accident scenario are listed in the following tables:
Scenario
Liquid Processing:RB Sump Accident Water
RB Decon. Water
AFHB & RB Chemical Decon. Water,
RCS Accident Water
RCS Flush and Drain Water
RCS Decon. Water and RCS Chemical Decon. Water

Table Number
7.15
.7.16
7.17
7.18
7.19
7.20

7-37

Table 7.15. Dose Estimates for the Maximum Exposed Individual
Caused by Failure of a HEPA Filter during Processingof Reactor Building Sump Accident Water
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenb

Inhalation
Ground Shine
Vegetable Use

1.1 x 10-2
2.8 x 10-1
2.2

1.0 x 10-1
2.8 x 10-1
11

3.9 x 10-2
2.8 x 10-1
6.5

2.5

11

6.8

2.5 x 10-2
2.8 x 10-1
4.7

5.0 x 10-2
2.8 x 10-'33

2.8 x 10-2
2.8 x 10-1
40

5.0

33

40

7.2.x 10-3
2.8 x 10-1
2.2
8.4 x 10-1

7.2 x 10-2
2.8 x 10-1
115.1

2.7 x 10-2
2.8 x 10-•
6.5
4.9

3.3

16

12

Total
Nearestmilk goat

Inhalation
Ground Shine
Goat Milk Use
Total

Nearest cow
and garden

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use
Total

aDoses were calculated for total-body, GI-tract, bone,

liver, kidney,

thyroid, lung and skin. The maximum three organ doses are listed in thistable.
Doses were calculated for four age groups; adults, teenagers,
children, and infants.
The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and nearest cow andgarden locations are for children and for the nearest goat location are
for adults for total body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are:nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north,'and nearest cow and garden= 1.05 miles east.

7-38

Table 7.16.. Dose Estimates for the Maximum Exposed Individual
Caused by Failure of a HEPA Filter during Processingof Reactor Building Decontamination Water

Location

Pathway

Total-Body

Dose (mrem)a
Bone

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

3.6 x 10-6
4.3 x 10-s
3.0 x 10-4

6.4 x 10-6
4.3 x iO-S
1.5 X 10-2

4.2 x 10-6
4.3 x 10-5
9.7 X 10-4

3.5 x 10-4

1.5 x 10-3

1.0 x 10-3

3.6 x 10-6
4.3 x 10-5
7.1 x 10-4

6.4 x 10-6
4.3 x 10-s
4.9 x 10-3

4.2 x 10-6
4.3 x 10-5
5.9 x 10-3

Total

7.6 x 10-4

4.9 x 10-3

5.9 x 10-3

Inhalation

2.5 x 10-6

9.0 x 10-6

4.0 x 10-6

Total

Nearestmilk goat

Nearest cow

and garden

Inhalation
Ground Shine
Goat Milk Use

Ground Shine

4.3 x 10-s

Vegetable Use
Cow Milk Use

3.0 x 10-4
1.7 x 10-4

4.3 x 10-s
1.5 x 10-3
7.6 x 10-4

5.2 x 10-4

2.3 x 10-3

Total

Liver

4.3 x 10-s

9.7 x 10-4
7.4 x 10-4

1.8 x 10-3

aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung and skin. The maximum three organ doses are listed in thistable. Doses were calculated for four age groups; adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and nearest.cow andgarden locations are for children and for the nearest goat location are,for adults for total-body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

7-39

Table 7.17. Dose Estimates for the Maximum Exposed Individual
Caused by Failure of a HEPA Filter during Processingof Reactor Building Decontamination Water
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

6.7 x 10-6
7.9 x 10-s
5.7 x 10-4

2.4 x 10-5
7.9 x 10-s
2.7 x 10-3

1.1 X 10-s
7.9 x 10-5
1.7 x 10-3

6.6 x 10-4

2.8 x 10-3

1.8 x 10-3

6.7 x 10-6
7.9 x 10-5
1.3 x 10-3

1.2 x 10-5
7.9 x 10-5
9.2 x 10-3

7.9 x 10-6
7.9 x 10-5
1.1 x 10-2

1.4 x 10-3

9.3 x I0-3

1.1 x 10-2

4.6 x 10-6
7.9 x 10-5
5.7 x 10-4
3.2 x 10-4

1.7 x 10-5
7.9 x 1 0 -s
2.7 x 10-3
1.4 x 10-3

7.4 x
7.9 x
1.7 x
1.4 x

9.7 x 10-4

4.2 x 10-3

3.2 x 10-3

Total
Nearestmilk goat

Inhalation
Ground Shine
Goat Milk Use
Total

Nearest cow
and garden

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use
Total

aDoses were calculated for total-body, GI-tract, bone,

liver,

10-6
10-s
10-3
10-3

kidney,thyroid, lung and skin.
The maximum three organ doses are listed in thistable.
Doses were calculated for four age groups; adults, teenagers,
children, and infants.
The highest dose estimates for each age group are
listed.
The dose estimates for the nearest garden and nearest cow andgarden locations are for children and for the nearest goat location are
for adults for total-body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

7-40

Table 7.18. Dose Estimates for the Maximum Exposed Individual
Caused by Failure of a HEPA Filter during Processingof RCS Accident Water

Location

Pathway

Total-Body

Nearest b

Inhalation

4.8 x 10-3

Ground Shine
Vegetable Use

6.5 x 10-3
1.4

gardenNearestmilk goat

Nearest cow

and garden

Dose (mrem)aBone
7.3 x 10-2
6.5 x 10-3

Liver
8.9 x 10-4

5.5

6.5 x 10-3
1.5 x 10-1

Total

1.4

'5.6

1.6 x 10-1

Inhalation

1.9 x 10-3
6.5 x 10-3

3.0 x 10-2

6.5 x 10-4

2.8 x 10-1

6.5 x 10-3
1.6

6.5 x 10-3
9.2 x 10-1

Total

2.9 x 10-1

1.6

9.3 x 10-1

Inhalation

3.3 x 10-3
6.5 x 10-3

5.1 x 10-2
6.3 x 10-3

6.1 x 10-4

1.4
8.9 x 10-2

5.5
4.0 x 10-1

6.5 x 10-3
1.5 x 10-1
1.2 x 10-1

1.5

6.0

2.8 x 10-1

Ground Shine
Goat Milk Use

Ground Shine
Vegetable Use
Cow Milk Use
Total

afDoses were calculated for total-body, GI-tract,'bone,

liver, kidney,

thyroid, lung and skin. The maximum three organ doses are listed in thistable. Doses were calculated for four age groups; adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and nearest cow andgarden locations are for children and for the nearest goat location are
for adults for total-body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

7-41

Table 7.19. Dose Estimates for the Maximum Exposed Individual
Caused by Failure of a HEPA Filter during Processingof RCS Flush and Drain Water
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

2.4 x 10-2
3.3 x 10-2
7.0

3.7 x 1 0 -1
3.3 x 10-2
28

4.4 x 10-3
3.3 x 10-2
7.4 x 10-1

7.1

28

7.8 x 10-1

9.6 x 10-3
3.3 x 10-2
1.5

1.5 x 10-1
3.3 x 10-2
8.1

3.3 x 10-3
3.3 x 10-2
4.6

1.5

8.3

4.6

1.6 x 10-2
3.3 x 10-2
7.0
4.4 x 10-1

2.5 x 10-'
3.3 x 10-2
28
1.9

7.5

30

Total
Nearestmilk goat

Inhalation
Ground Shine
Goat Milk Use
Total

Nearest cow
and garden

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use
Total

3.1 x
3.3 x
7.4 x
5.7 x

10-3
10-2
10-1
10-1

1.3

Doses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung and skin.
The maximum three organ doses are listed in thistable.
Doses were calculated for four age groups; adults, teenagers,
children, and infants.
The highest dose estimates for each-age group are
listed. The dose estimates for-the nearest garden and nearest cow andgarden locations are for children and for the nearest goat location are
for adults for total-body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appi-ndix W.
The actual locations are:nearest garden.
1.05 miles east-northeast,nearest milk goat.= 1.02 miles north, and nearest cow and garden= 1.05 miles east.

7-42

Table 7.20.
Dose Estimates for the Maximum Exposed Individual
Caused by Failure of a HEPA Filter during Processing of RCS
Decontamination Water or RCS Chemical Decontamination Watera
Dose (mrem)
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

4.9 x 10-l
6.5 x 10-3
1.4

7.5 x 10-2
6.5 x 10-3
5.7

8.9 x 10-4
6.5 x 10-3
1.5x 10-1

1.4

5.8

1.6 x 10-1

2.0 x 10-3
6.5 x 10-3
3.0 x 10-1

3.1 x 10-2
6.5 x 10-3
1.6

6.5 x 10-4
6.5 x 10-3
9.2 x 10-1

3.1 x 10-1

1.6

9.3 x 10-1

3.4 x 10-3
6.5 x I0-3
1.4
9.0 x 10-2

5.2 x 10-2
6.5 x 10-3
5.7
4.0ýx 10-1

1.5

6.2

Total
Nearestmilk goat

Inhalation
Ground Shine
Goat Milk Use
Total

Nearest cow
and garden

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use
Total

aDoses were calculated for total-body, GI-tract, bone,

6.1 x
6.5 x
1.5 x
1.2 x

10-4
10-3

10-1
10-1

2.8 x 10-1

liver, kidney,

thyroid, lung and skin. The maximum three organ doses are listed in thistable.
Doses were calculated for four age groups; adults, teenagers,
children, and-infants.
The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and nearest cow andgarden locations are for children and for the nearest goat location are
for adults for total-body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
nearest garden = 1.05 miles east-northeast,
The actual locations are:nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

7.1.6

Liquid Processing Costs

The alternative water processing facilities and associated operating costs are detailed in Appendix K. Based upon the details contained in Appendix K, the staff has bounded the costs for thevarious water processing alternatives as a function of the contaminated waters to be processed.
The staff believes that the costs for water processing and facility construction will lie somewhereThe
between $21 million and. $29 million regardless of the combination of alternatives selected.
staff has made no attempt to quantify the costs of licensing or qualification testing of thevarious alternatives, therefore these costs should only be used as relative comparison.

7-43
7.1.7

Comparison of Liquid Waste Treatment Alternatives

The alternatives discussed above were compared on a relative basis considering time to initiate
processing, time to complete processing, worker exposure, quality of liquid effluent (if any) andcosts.
Table 7.21 summarizes these parameters as a function of alternative.
The conclusions
that can be drawn from this table are as follows:Time
The times to initiate processing vary significantly due to facility and equipment
availability. Those alternatives which use systems and equipment currently installed
or being installed can be initiated in 0 to 6 months, while those alternatives requiring
new facilities or hardware take a minimum of 3 years.
The implementation times to. process 100,000 gallons vary from 1 week to 4 months withmost alternatives in the 1- to 3-month range. The ion exchange techniques require theleast time for implementation.
Worker Exposure
For comparison purposes, it was assumed that worker exposure was based on processing
the RB sump water. As shown, direct immobilization leads to maximum exposure, 16 personrem per 100,000 gallons of water processed (see Table 7.21).
The estimated exposures for the other alternatives are lower than the exposure fordirect solidification and roughly comparable to one another.
Number of MPCs in Effluent
The allowable Maximum Permissible Concentrations (MPC) in 10 CFR Part 20 are based on
the relative toxicity of radionuclides released to either air or water and provide ameasure of the processing performance for a particular alternative. Thus, the numberof MPC's in liquid effluents are a measure of relative toxicity among the various
alternatives.
As shown, systems which process RB sump and primary system water on a
once through basis and do not recycle or polish the effluent with a second system or
technique, have MPC values as much as two orders of magnitude higher than other treatment alternatives.
The highest MPC values correspond to those alternatives which
exhibit the poorest relative processing performance.
As shown in Table 7.21, theestimated processing performance of EPICOR II for primary system water compares favorablywith the most effective zeolite-based treatment systems (i.e., zeolite/EPICOR II andSDS/EPICOR II) and exhibits better performance than the remaining zeolite-based treatment
systems, including the modified EPICOR II. However, it should be noted that the predicted performance of EPICOR II is based on the operating performance of the system todate. The predicted performance of the zeolite-based systems is based, in part, on
conservatively estimated decontamination factors for those system components for which
there is no site operating experience.
Costs
Treatment costs were summarized in Section 7.1.6 and waste disposal costs are summarized
in Section 9, with the basis for all estimates presented in Appendix K.
7.2

DISPOSAL OF PROCESSED WATER

Processed water is the liquid effluent arising from treatment of AFHB accident water, RB sump
water, and RCS primary system water.
It is distinquished from other contaminated water by its
tritium content. About 743,000 gallons of this water are in storage, and another 796,000 gallons
will arise from future treatment of RB sump and primary system water. This section discusses thealternatives considered for disposition of processed water and their impacts.
In general, where
disposal alternatives result in liquid and gaseous effluents to the environment, the radionuclide
distributions provided in the source term tables in this section account for greater than 95% of
the calculated offsite doses.

Table 7.21.

Parameters for Comparison of Liquid Waste Treatment Alternatives
Parameter
MPC's in Liquidb
Effluent
Primary
RB Sump
SystemWater
Water

Other Wastesc
Generated
High-Specific
Low-Level
Activity
ResinsResinsEstimated
Implementatiog
DateTime to
Implement
per 100,000
gallons

Estimated
Worker Dose
person-rem
per 100,000
gallonsbilization

1/84

3 mos

16

NA

NA

Long-Term
Storage

1/84

1 wk

NA

NA

NA

1/84

1 mo

NA

NA

NA

+

1/84

4 mo

NA

NA

NA

+

Alternative

Direct imm8-

Evaporator/ dResin System
Bitumen/Resin
SystemImmobilized
SolidsConcentrated
Liquids

High-Specific
Activity
Zeolites+

Zeolite Based
Systems
-

SDS

2/81

1 mo

1.1

6.2 x 103

5.4 x 104++-

Modified
SDS

3/81

1 mo

1.1

2.2 x 103

2.2 x 104

+

+

1/84

1 mo

2.0

3 x 102

2.7 x 102

+

+

3/81

1.5 mo1.6

3 x 102

27

++

2/81

1.5 mo1.6

3 x 102

27

+

2/81

1.5 moNANA

2.2 x 104+

1/81

1.5 mo

NA

NA

35

+

-

Zeolite/Evaporator-

Zeolite/EPICOR II-SDS/EPICOR II
Modified
EPICOR II-

EPICOR II
System-aBased on time zero of .1st quarter 1981. Table shows quarter/year.
bA measure of the relative processing performance of the alternatives suitable for pr ocessing RB sump water and primary system water.
cA + indicates waste form is generated.
dAlternative considered suitable for treatment of chemical decontamination solutions.

+

7-45
7.2.1

Current Status and Projected Water Disposal

Requirements

The current status of processed water and the amounts that may be generated for disposal aredescribed below.
7.2.1.1

Current Status

As described in Section 7.1.1, 743,000 gallons of processed AFHB accident water are in storage on
the site.
This water contains about 360 Ci of tritium (H-3) plus about 0.8 Ci of cesium; characteristics of the water were given in Table 7.3.3
No decision has been made as to which treatment system(s) will be used to process the accident
water in the reactor building sump or the water in the reactor coolant system (RCS).
However,
the tritium concentration will not be changed by whatever water treatment alternative is selected
unless clean water is added.
The current tritium inventory in the plant is as follows:
Processed Water in Storage (mostly AFHB).
About 360 Ci of tritium in 743,000 gallons of waterwith an average tritium concentration of 0.13 pCi/mL.
Untreated Reactor Building Sump Water.
About 2520 Ci of tritium in 700,000 gallons of accident
water with an average tritium concentration of 0.95 pCi/mL.
Untreated RCS Water.
About 30 Ci
concentration of 0.08 pCi/mL.

of tritium in 96,000 gallons of water with an average tritium

The processed water available from these three sources is about 1,540,000 gallons containing
about 2,910 curies of tritium plus trace quantities of other radionuclides.
This amount oftritium is the maximum that will be available in processed water.
Unavoidable in-leakage andcross contamination of water sources might increase the volume of water available for disposal
and thereby reduce concentrations of tritium by dilution.
7.2.1.2

Projected Water.Disposal

Requirements

As described above, a minimum of 1,540,000 gallons of processed water will be available for
disposal.
The characteristics of the processed AFHB water are known since this water has alreadybeen processed through EPICOR-II.
The characteristics of the processed water arising from treatment of RB sump water and RCS primary system water are not known since they will depend on thetreatment system(s) used.
However, the tritium content of this untreated water will not change
during treatment, so the current tritium inventory of 2,910 curies represents a maximum.
Reuse
of processed water and other activities could reduce this tritium inventory due to evaporation
losses.
Table 7.22 summarizes these potential, losses and the activities that lead to them.
However, for the purposes of evaluating alternative methods of processed water disposition andtheir impacts, the maximum tritium inventory of 2,910 curies was used.
If the losses shown in
Table 7.22 are experienced, then the impacts due to tritium releases will be proportionately
lower than those presented in succeeding sections.
Reuse of processed water is discussed indetail in Appendix F.
The volume of 1,540,000 gallons of processed water is a minimum and could increase by 10 to20 percent.
This increase could be due to in-leakage into the reactor building, addition ofclean water used for decontamination to the reactor building sump and commingling of clean waterwith processed water.
NRC has directed the licensee to minimize the volume of accident related
water. 2 1 Any such volume increases would merely dilute the concentrations of tritium and othercontaminants in the AFHB water and the water to be processed.
To ensure that release impacts areassessed under worst-case conditions, the discussions that follow are based on this minimum
volume of 1,540,000 gallons..
The characteristics of treated reactor building sump water and RCS primary system water will
depend on the treatment system(s) selected to process this water.
The alternatives being considered were discussed in Section 7.1.3.
Staff estimates of the quality of processed water that

7-46

Table 7.22.

Tritium Evapgration Losses from Potential Reuse
Applications and Other Facilities

Application

Water Volume

Tritium IniTial
Loss Rate

1.

Shielding of ion
exchange systemin spent fuel pool

230,000 gallons
of processed AFHB

50 Ci/yr2.

Use in SFPs and FTC
during defueling

1,040,000 gallons
of processed water

500 Ci/yr

3.

Use to shield spent fueland waste in SFPs after
defueling
Loss from reactor building sump prior toprocessed

690,000 gallons
of processed water

300 Ci/yr

700,000 gallons
of unprocessed
reactor building
water

500 Ci/yrb

4.

asee Appendix F for discussion of reuse applications.
bConservative estimate see Section 5.2.4.1.
cRate will decrease exponentially with time.

would arise from implementation of various alternatives for treatment of RB sump water are presented in Table G.19 of Appendix G and those for RCS (primary system) water are presented inTable G.20 of Appendix G.
To bound the concentrations of radionuclides in the processed water to be disposed of, the bestand worst-case treatment systems, were identified from these tables and used to characterize
processed reactor building and RCS water.
The system which produced effluent with the lowest
concentrations of contaminants without significantly delaying cleanup, was the SDS/EPICOR II
alternative.
The other system which produced effluent with the highest concentrations of contaminants, was the SDS alternative. The characteristics of the effluents arising from the use of
these treatment systems were combined with those of the stored AFHB water to provide boundingcases for processed water. Table 7.23 shows the best-case conditions with SDS/EPICOR II systemtreatment.
The worst-case conditions, based on the SDS alternative, are shown in Table 7.24.
For purposes of this analysis, it has been conservatively assumed that, for these two conditions,
all processed water will eventually be disposed of and will not be reused regardless of the
future disposition of the facility. In the event water is retained in the spent fuel pools after
defueling and this water is left in place, the volumes would be reduced by about 690,000 gallons.
7.2.2

Alternative Methods Considered

Disposal alternatives were evaluated on the basis of the bounding conditions for radionuclide
concentrations given in Section 7.2.1. Ten major alternatives have been considered for disposition of processed water. As shown in Figure 7.10, these alternatives can be grouped as follows:
Lonq-Term, Onsite Storaqe-

In liquid storage tanks
Solidified as cement slabs

Onsite Disposal
As solidified low-level waste in shallow land burial trenches
As bulk liquid using an underground injection well.

7-47

Table 7.23.
Radionuclide Inventory - Processed Water
Best Case - SDS/EPICOR II Treatment

(Ci)

a

Reactor bu ildiWng

Primaryc

AFHB Water(743,000 gallons)

Sump Water(700,000 gallons)

System Water(96,000 gallons)

Totald(1,540,000 gallons)H-3

3.6 x 102

2.5 x 103

2.7 x 101

2.9 x 103

Cs-1375.7 x 10-1

4.7 x 10-7

9.9 x 10-9

5.7 x 10-1

Cs-134

2.5 x 10-1

6.6 x 10-8

1.5 x 10-9

2.5 x 10-1

Sr-90

NDe

4.1 x 10-s

4.6 x 10-s

8.7 x 10-s

Sr-89

ND

1.1 x 10-6

Ru-106<2.1 x 10-2

3.9 x 10-3

4.8 x 10-6
1.7 x 10-2

5.9 x 10-6
4.2 x 10-2

Sb-125<1.7 x 10-2

5.3 x 10-2

1.4 x 10-3

7.1 x 10-2

Te-127m

ND

1.3 x 10-3

1

x 10-1

Ce-144<1.8 x 10-2

2.5 x 10-4

5

x 10-31
x 10-1
2.3 x 10-2

RadionuclideaFrom Table F.I.
bFrom Table G.19.
CFrom Table G.20.
dRounded to two significant figures.
e"ND" means not detectable.

Table 7.24.

Radionuclide Inventory - Processed Water
Worst Case - SDS Treatment(Ci)a
AFHB Water(743,000 gallons)

Reactorb Bui dieng
Sump Water(700,000 gallons)Primaryc
System Water(96,000 gallons)

Totald(1,540,000 gallons)H-3

3.6 x 102

2.5 x 103

2.7 x 101

2.9 x 103

Cs-1375.7 x 10-1

4.3

9.9 x 10-2

5.0

Cs-134,

2.5 x 10-1

6.6 x 10-1

1.5 x 10-2

9.3 x 10-1

Sr-90

NDe

4.1

4.6

8.7

Sr-89

ND

1.1 x I0-1

0.48

0.6

Ru-106<2.1 x 10-2

3.9

17

21

Sb-125<1.7 x 10-2

53

1.4

54

Te-127m

ND

0.7

50

51

Ce-144<1.8 x 10-2

0.25

5.0

5.3

RadionuclideaFrom Table F.1.
bFrom Table G.19.
cFrom Table G.20.
dRounded toe"NDV means

two significant figures.

not detectable.

NO

EVENTUAL DISPOSAL

00

Figure 7.10.

Decision Tree for Disposition of Processed Water.

7-49
Shipment Offsite for Disposal- As bulk liquid for deep-well-injection disposal- As bulk liquid for ocean disposal- Solidified and packaged for disposal at a low-level radioactive waste diposal
facility.
Discharge to the Environment- Controlled release to the Susquehanna River- Controlled release to air via natural evaporation- Controlled release to air via forced evaporation.
Evaluations for these alternatives are presented below.
7.2.2.1

Long-Term,

Onsite Storage

Under this alternative, processed water would be retained at the plant site for periods in excessof the plant's projected operating lifetime of 25 or more years.
The water could be stored inliquid form in large storage tanks or it could be solidified with cement and stored in solidform.
As shown in Tables 7.23 and 7.24, the principal contaminant is tritium, a beta emitter.
Since beta radiation does not penetrate the tanks or the cement, direct radiation exposure is not
a problem during storage in either liquid or solid form.
However, immobilization of the processed water in cement would not prevent the loss of tritium to the environment by diffusion.
Storage as Bulk LiquidLong-term storage (at least 10 half lives) to reduce concentrations of tritium either to innocuous
levelsor to primary drinking water standards is not practicable because of tritium's 12.8-year
half-life.
Given a maximum tritium concentration of 1.0 pCi/mL in storage tanks, attainment of
EPA primary drinking water standard concentrations (2 x 10-5 pCi/mL) would require about 200 years.
Retention for these periods of time is not considered practicable because of the inability toguarantee institutional controls for periods in excess of 100 years.
In effect, the TMI sitewould become a low-level waste disposal site for radioactivity in a mobile form.
The staff does
not consider this a satisfactory alternative.
Two outside processed water storage tanks with a capaci'ty of 500,000 gallons each have been constructed.
A criterion for storage of processed accident water in these tanks is that the content of radioactivity stored in each outside processed water storage tank should be limited such that a tankfailure would not result in greater than 10 CFR Part 20 (Table II, Col. 2) concentrations at thenearest drinking water intake for combined radionuclides as a function of actual tank volume.
Inorder to assure conformance with this criterion, several assumptions and calculations are required.
Details of the assumptions used by the NRC staff for this storage criterion are discussed in
Section 7.2.4.2.
In equation form this criterion requires that the storage of radioactivity in an outside storagetank be limited such that in the event of a rupture:

*

Ci
PCi

<1

where:
Ci = concentration (pCi/mL) of the ith nuclide at the nearest downstream drinking
water intake (Brunner Island)

7-50
MPCi = maximum permissible concentration (10 CFR Part 20, Table II, Col. 2 - pCi/mL)
The concentration at the nearest drinking water intake (Ci,Ci = (0.5) Ai

QAt_

pCi/mL) is determined by:

Ai

6.4x 10where:
Q = minimum river flow rate to overtop Red Hill Dam (16,000 ft 3 /sec)
At = release period (2 hours)
Ai = tank activity prior to rupture (Curies)
0.5 = fraction of the tank volume that discharges to the river
6.4 x 106 is a constant (Ci/pCi/mL)
Combining these two equations yields the limiting storage requirement expressed in the above
criterion:Ai< 6.4 x 106

RPC=

Thus, the quantity (Curies) of each radionuclide in the storage tank will be considered in assuring that a tank failure would not result in greater than 10 CFR Part 20 concentrations at thenearest drinking water supply intake (see Section 7.2.4.2).
At present, 743,000 gallons of AFHB water are stored in tanks within the plant. When available,
the two PWSTs will provide an additional 1 million gallons of storage capacity for a total minimum
onsite storage capacity of about 1.75 million gallons.
This capacity exceeds the minimum volumeof about 1.54 million gallons.
This excludes storage of some of this water in the spent fuelpools.
If processed AFHB water is used to shield an SDS type system in the spent fuel pool, thestorage capacity will increase to about 2 million gallons.
These storage tanks can be used for relatively long-term storage, 15 to 25 years or until adecision on disposal of the processed water is made.
Following any such interim storage, theprocessed water will have to be permanently disposed of, and some of the alternatives discussed
in this section will have to be considered. Therefore, while bulk liquid storage for periods up
to about 25 years is practicable, it merely defers the ultimate disposition of the water andthus, is not considered further as a disposal option.
Storage As Cement Blocks
Under this alternative, the processed water would be mixed with cement to form relatively large
blocks.
To characterize this alternative, it was assumed that the blocks would be rectangles(6' x 6' x 10') containing 1600 gallons of processed water and weighing 40,000 pounds.
About 960
blocks would be required to immobilize 1,540,000 gallons of water.
Each block would require2
about 190 ft
of surface area for storage with inspection aisles between blocks.
Thus, theseblocks would occupy a maximum of about 4 acres of land if stored one layer only.
These cement blocks would immobilize the water, and when coated with asphalt or other weatherresistant material, they could be stored outside for relatively long periods.
Following suchstorage, they could be shipped offsite to a radioactive waste disposal facility.
A portion of the tritium would be released to the atmosphere during immobilization.
Additionally,
the use of cement does not permanently immobilize the tritium. Recent work by Monsanto indicatesthat with similar mixtures of cement and tritiates, diffusion through the concrete leads to lossof tritium in relatively short time periods.
Storage of processed accident water for relatively long periods is a practicable alternative
which would permit defer.ral of a decision on ultimate disposition of the water.
Since immobilization of the processed water into cement blocks would minimize flexibility relative to ultimate
disposition, interim storage as bulk liquid seems preferable.

7-51
7.2.2.2

Onsite Disposal

Under this alternative, the site would be used as a permanent disposal facility for processedwater in either liquid or solid form.
For disposal in solid form, shallow land burial trenches would be constructed to dispose ofeither cement blocks (see Section 7.2.2.1) or processed water immobilized in cement and packaged
in drums.
Onsite disposal in liquid form would involve construction of a Class I deep injection
well and injection of the processed water into the well as a means of permanent disposal.
There
are regulatory obstacles to implementation of both these alternatives. The deep injection wellwould require an EPA permit and may also be subjected to NRC licensing. An onsite shallow land
burial facility would also be subject to NRC licensing. Moreover, implementation of these alternatives would convert the site into a permanent repository for low-level radioactive waste--an
unacceptable condition since the staff has consistently committed to not converting the site to apermanent repository. Therefore, these alternatives are not considered further.
7.2.2.3

Shipment Offsite for Disposal

Under this alternative, processed water would be packaged in solid or bulk liquid form and transported off the plant site. If transported offsite in solid form, the material could be disposedof at a licensed, low-level radioactive waste disposal facility.
If transported offsite as abulk liquid, the water could be processed at another facility, and then disposed of in a deep
injection well or in the ocean.
Any of the other alternatives described below could also be
implemented offsite.
7.2.2.4

Discharges to the Environment.

Three of the alternatives considered would discharge processed water to the environs under controlled conditions:(1) release of liquids to the river, (2) release of vapor to the atmosphere
via natural evaporation, and (3) release of vapor to the atmosphere via forced evaporation combinedwith a partial release of liquids to the river.
Discharge to the Susquehannna River
Under this alternative, processed water would be sampled, analyzed, and discharged under controlled conditions to the Susquehanna River in accordance with the plant Technical Specifications
which implement 10 CFR Part 20 and Appendix I criteria.
Natural Evaporation Pond
Under this alternative, processed water would be placed in a lined pond.
Local climatologicalconditions will evaporate the water, and vapor containing tritium will be released to the atmosphere reducing the tritium concentration in the pond. Water losses due to evaporation are assumed
to be offset by rainfall so there will be no net loss of pond water volume.
Forced Evaporation
The natural pond evaporation technique described above could be improved by heating the water
entering the ponds.
A spray system would also improve evaporation rates from the ponds. These
procedures may result in a fog over the pond and river channels adjacent to the island. They
would also have to be controlled to prevent worker exposures in areas contiguous to the ponds
above 10 CFR Part 20 guidelines. Thus, this alternative was not evaluated in depth.
Another method of forced evaporation involves the use of the Unit 2 service water cooling tower.
The mechanical draft service water cooling tower is a three-cell unit with a recirculation rateof about 19,000 gpm.
Its use would provide a controlled method of simultaneously releasing
processed water to the air via the cooling tower vapor and to the river via cooling tower blowdown.
Use of this approach would produce fog and a small amount of drift, both containing tritium.
This could be minimized by operating only during favorable meteorological conditions.
7.2.3

Details of Alternative Methods

Details of the alternatives considered practicable from a technical and regulatory standpoint arepresented below. No decision has been made concerning the method that will be used to dispose of

7-52
processed water,
each other.
7.2.3.1

and the alternatives

described below may be used singly or in combination withTransport Offsite as Solid

Under this alternative, the water would be immobilized with cement, packaged, and transported
offsite to a low-level waste disposal facility.
To characterize this alternative, the staff
assumed that the water would be packaged in 55-gallon drums, with each drum containing 30 gallons
of immobilized water.
About 51,300 drums would have to be transported offsite in 640 shipments.
From 12 to 18 months
would be required to design and construct the necessary solidification facilities.
At an averagethroughput of 4 drums per hour, or 96 drums per day, about 18 months would be required to package
the 51,300 drums.
This is based on operation 24 hours a day, seven days a week.
There are no technical or Federal regulatory obstacles to implementing this alternative.
There
are, however, potential State restrictions on the use of low-level waste disposal facilities todispose of this material.
There is a shortage of shallow land burial facilities, and it isdoubtful that relatively innocuous tritiated waste will be given space that could otherwise be
used for other radioactive waste.
7.2.3.2

Transport Offsite as Bulk Liquid

Under this alternative, processed accident water would be packaged as bulk liquid in tank trucks
and transported away from the site.
It could then be disposed of in a deep injection well,packaged for ocean disposal, or released at another location.
Current Federal regulations do not prohibit bulk liquid shipments of tritiated water.
Under theLSA definitions in 10 CFR Part 71, tritium oxide in aqueous solution with concentrations up to
5 mCi/mL qualifies as LSA material.
Under 49 CFR Part 173.392, bulk liquids may be transported
in tank trucks provided the concentrations are less than 10'percent of the LSA concentrations.
Thus, water with tritium concentrations up to 500 pCi/mL could be transported in bulk.
Themaximum tritium concentration expected from any source at TMI-2 is 0.95 pCi/mL.
These shipments
also would have to conform to other transportation criteria in 49 CFR Part 173.
Tritium is a
Type IV radionuclide, so each shipment would be limited to 20 curies to ensure that Type A quantity limits were not exceeded.
Transport of water with the maximum tritium level of 0.95 pCi/mL
in a 4,500-gallon tank truck would result in about 16 curies of tritium per truck.
For theaverage concentration of 0.5 pCi/mL, the tritium inventory per truck will be about 8 Ci.
About
340 shipments would be required to transport 1.54 million gallons.
Two alternatives were considered for disposition of the water after removal from the site.
One
alternative involves disposal in a deep injection well.
Commercial injection wells used for
disposal of hazardous waste are in operation in 25 states.
These hazardous chemical wastes
include solutions.containing heavy metals, organic compounds, and pesticides.
Shipment of bulk liquids to a deep injection well or to a facility which prepares the processedwater for ocean disposal is practical.
Moreover, since immobilization with cement and packaging
in drums would negate the potential for use of a deep injection well and possibly for ocean
disposal, alternatives which involve bulk transport of liquids are preferred.
However, transportation from the site is merely the initial step toward ultimate disposition of the processed water.
Disposal through use of a deep injection well or burial at sea both involve solving potential
regulatory problems in relatively new fields of regulated activities.
Disposal of hazardous and radioactive waste in deep wells is regulated by the EPA through its
Underground Injection Control (UIC) Program under the Safe Drinking Water Act.
This program isstill evolving.
Technical criteria and standards are contained in 40 CFR Part 146; proposed
procedural requirements for State UIC permit programs are contained in 40 CFR Part 122; procedures
for approving State programs in 40 CFR Part 123; and procedures for issuing State permits in
40 CFR Part 124.
Within these regulations, Class I wells are those used to inject industrial,
nuclear, and municipal wastes beneath the deepest stratum containing an underground drinking
water source.
These wells would be regulated by the State in which they are located and would be
regulated by the State authority responsible for groundwater quality.
Historically these wells
have not been used for disposal of nuclear waste, and a special permit would have to be obtained
to use such a well for the processed water from TMI-2.
There is no certainty that a cognizant

7-53
State authority would grant a permit to dispose of processed TMI-2 accident water through aClass I well within its borders.
Similar regulatory obstacles would also have to be overcome to dispose of the processed water inthe ocean.
The regulations governing ocean dumping were promulgated under Title I of the Marine
Protection Research and Sanctuaries Act (1972) and are currently found in 40 CFR Part 220 (EPA
1980).
Under these regulations, a statement of the need for the proposed dumping and evaluationof alternative means of disposal, treatment, or recycle of the material must be included. Also,an assessment of the anticipated environmental impacts must be part of the application. This
could require several years of effort.
In addition, under EPA's proposed regulations the tritiated water would have to be packaged incontainers which would retain their integrity until radiodecay reduced the activity to "environmentally innocuous" levels. Satisfying this container criteria could also mitigate against thisalternative dependent on what level of tritium activity is deemed innocuous in an ocean environment.
Previous ocean disposal practices would permit immobilizing the liquid and packaging it in drums.
Since these practices have been banned, some higher standard of packaging, as yet undefined is
anticipated by the proposed regulations.
Therefore, although there are no regulatory obstacles to transporting the tritiated water off thesite, there are potential regulatory obstacles to its disposition in either a deep injection wellor the ocean, which make this option undesirable.
7.2.3.3

Discharge to the Susquehanna River

The release of water to the river is presently prohibited by the Commission's statement of May 25,
1979, the City of Lancaster Litigation Settlement Agreement, and the NRC Staff Order of February 11,
1980.
Criteria governing potential discharge of processed water to the river include; (1) 10 CFR Part 50
Appendix I criteria for offsite radiological exposure, (2) Clean Water Act criteria related to
EPA's Primary Drinking Water Standards, (3) the plant's NPDES permit limitations for non-nuclear
discharges to surface waters, (4) State and local ordinances governing poi'ot source discharges tothe river, and (5) the requirements of 10 CFR Part 20.
In the event this alternative is pursued, specifications for radionuclide levels and for nonradioactive chemicals would have to be considered.
o Nonradioactive Chemical Constraints

Releases to the river are controlled by the plant's National Pollution Disharge Elimination
System (NPDES) permit. This permit limits the pH to between 6 and 9 and the average boron concentration at the discharge point to 0.7 ppm.
o Radionuclide Concentration and Dose Constraints

For release to the river, the radionuclide concentrations in liquid effluents must be diluted tolevels at the discharge point which are below 10 CFR Part 20 release limits for the specificradionuclides in the processed water. Another set of criteria are EPA's Primary Drinking Water
Standards, which must not be exceeded at a drinking water intake downstream from the discharge
point. Additionally, the dose design objectives of Appendix I of 10 CFR Part 50 must be met.
Processed water discharged to the river is mixed with service water cooling tower blowdown.
The
cooling tower blowdown flow is 38,000 gpm, but current technical specifications limit the flow
available for dilution to 90 percent of this value, or about 34,000 gpm. Thus, processed water
discharged at 1 gpm can be diluted by a factor of 34,000,. A change in the Technical Specifications could increase this dilution flow to 140,000 gpm.
To consider discharges of the processed water shown in Table 7.25 a discharge rate of 10 gpm wasselected. This discharge rate leads to a dilution factor of 3,400. At a continuous discharge of
10 gpm, the inventory of 1,540,000 gallons would require a release period of about 110 days. If
the processed water is borated for defueling and is discharged in the borated condition, theboron concentration limitation in the NPDES permit will reduce this discharge rate. When
1,040,000 gallons of water borated to 3,650 ppm is mixed with the other processed water, a dilu-'
tion factor of about 3,500 is required to satisfy the NPDES limit.

7-54
Since up to 140,000 gpm of dilution water is available, discharge rates can be varied to satisfy
10 CFR Part 20 release limits at the discharge point, Primary Drinking Water Standards at adownstream uptake for drinking water, and the NPDES limits. The dilution factor that satisfies
the most limiting of these criteria will be used if this alternative is implemented.
7.2.3.4

Natural Evaporation from Ponds

The climatological conditions at the TMI-2 site are not conducive to natural evaporation as ameans of reducing processed water volumes because the evaporation rates are equivalent to theamount of rainfall the area receives.
Natural evaporation can, however, reduce the tritiumcontent of processed water which is ponded due to the losses of tritiated water vapor and the
addition of rainfall.
To characterize this alternative, the staff assumed an evaporation pond with a capacity of 2.5 million gallons and a surface area of 60,000 ft 2 . To assess this alternative the following criteria
were established:
0

Criterion 1 - Offsite doses due to airborne tritium releases from the pond shall conform tolevels consistent with Appendix I guidelines.

0

Criterion 2 - The airborne concentrations for tritium above the pond shall be less thanlevels in 10 CFR Part 20 Appendix B for occupational personnel (5 x 10-6 pCi/mL).

0

Criterion 3 - The processed water shall be retained in the pond at least until the concentrations of tritium are equivalent to those which naturally occur in the Susquehanna River 200 pCi/L to 400 pCi/L.

The methodology used to consider the natural evaporation alternative is presented in Appendix S.
The conditions which satisfy the above criteria and also bound (1) tritium release rates to theatmosphere, (2) exposure to onsite workers, and (3) time frames to achieve tritium concentrations
comparable to those in the river are defined in that appendix. These conditions are summarized
below.
Under the assumed worst case conditions for offsite atmospheric releases of tritium (maximum
evaporation rates - see S.2 of Appendix S), 80 percent of the tritium is released in 3 to 6 months
after the processed water is placed in the pond. Maximum release rates are attained when ponding
is started in July. Minimum release rates are attained when ponding is started in January.
Inboth cases, essentially the entire inventory of about 2900 Ci of tritium is released to theatmosphere during the first year.
Under the assumed worst-case conditions for occupational exposure (static conditions above thepond, see S.3 of Appendix S) ponding of 0.5 pCi/mL water in July may not be acceptable. For the
average relative humidity during this month, the tritium concentrations in the ponded water must
be maintained below 0.36 pCi/mL to insure that airborne concentrations of tritium above the pond
are acceptable for restricted use. An extreme worst-case climatological condition is represented
by several 90'F days with 90 percent relative humidity. Under these conditions, the ponded water
tritium concentration must be maintained below 0.17 pCi/mL.
Operating practices can avoid these conditions. The 1,540,000 gallons of processed water can beponded during the winter months and the exchange between tritiated pond water and untritiatedprecipitation will reduce the level below 0.17 pCi/mL before the worst-case summer month of July.
If this is not practicable, processed water can be diluted to reduce the average concentration
during the first July below 0.17 pCi/mL. In both cases,. tritium concentrations in the pond canbe controlled to maintain airborne concentrations below 10 CFR Part 20 Appendix B occupational
exposure limits for restricted use under poor climatological conditions.
The time frames to achieve tritium concentrations in the 200 to 400 pCi/L range were determined
under maximum and minimum evaporation rate conditions. Under maximum evaporation rate conditions(case one in Appendix S), tritium concentrations approach these levels in 1½ to 2 years. Whenthe pond is filled to capacity (2.5 million gallons) these levels are attained in about 3 to
5 years.
While the tritium concentration reduction that takes place due to evaporation of tritiated waterand dilution with precipitation is rather rapid, the concentrations of other radionuclides in thepond water would remain essentially unchanged.
This does not present a problem for processed

7-55
accident water with characteristics similar to those represented by the SDS/EPICOR II treatment
case.
However, the relatively large inventory of cesium, strontium, and antimony present inprocessed accident water with characteristics similar to those represented by the SDS treatmentcase could be a problem because these radionuclides would settle to the bottom of the pond and beretained in the sediment.
Therefore, the type and quality of radionuclides other than tritium
should be considered if the evaporation alternative is implemented.
In summary, evaporation of accident water from a natural evaporation pond will substantiallyreduce tritium concentrations over relatively short time periods and is considered a practicable
alternative when coupled with retention of the ponded water until tritium concentrations approach
natural levels.
7.2.3.5

Forced Evaporation Through Cooling Tower

Under this alternative, processed water would be evaporated using the mechanical draft service
water cooling tower.
Figure 7.11 presents a simplified flow diagram for disposition of theprocessed water using this alternative.
As shown, the action of the cooling tower will lead tothe formation of water vapor containing tritium. This tritiated vapor will be released to theatmosphere.
Most of the fission product contaminants in the processed water will remain in theconcentrated cooling tower blowdown, but a small fraction of these contaminants in the cooling
tower blowdown would be diluted to the extent required to satisfy 10 CFR Part 20 criteria prior
to discharge to the river. Typically 90 to 95 percent of the tritium in the processed water isreleased to the atmosphere, while the remainder is discharged with the blowdown. The reverse istrue for fission product contaminants, in which 20 to 30 percent of the radionuclides are released
to the atmosphere via the drift while 70 to 80 percent of the radionuclides are discharged withthe blowdown.
To evaluate this alternative, the following assumptions relative to cooling tower performance
were made.
o

The processed water feed rate to the tower would be 100 gpm.

o

The evaporation rate would be 93 gpm and the drift rate would be 2 gpm.

o

The air flow would be 2.25 x 106 cfm.

0

The blowdown rate would be 5 gpm.

Based on these assumptions, 95 percent of the processed water will be released to the atmospherein the form of vapor. This vapor will contain about 95 percent of the tritium, originally within
the processed water feed and about 30 percent of the fission product contaminants in the processedwater feed. The radionuclides not released to the atmosphere will be present in the cooling
tower blowdown in concentrated form. For the stated assumptions, the concentration factor acrossthe cooling tower is 14.3. Therefore the concentrations of radionuclides in the cooling tower
blowdown will be 14.3 times these concentrations in the processed water feed. Based on current
technical specifications, up to 34,000 gallons of clean water would be available to dilute theblowdown.
At a blowdown rate of 5 gpm, discharges to the river can be diluted by a factor of 6800.
With a change in technical specifications, the water available for dilution could be increased toabout 140,000 gpm without additional facilities.
The cooling tower would be operated intermittently under favorable meteorological conditions. At
the assumed feed rate of 100 gpm, about 260 hours of cooling tower operation are needed to process
the 1,540,000 gallon inventory. Releases would be made in accordance with the dose design objectives of Appendix I to 10 CFR Part 50 and the concentration levels of 10 CFR Part 20.
7.2.4

Effluents and Releases to the Environment

The nature and impacts of the releases that could arise from implementation of the processed-waterdisposal alternatives discussed above are summarized in this section. Releases under both normal
and accident conditions are discussed.

7-56
EVAPORATEDTRITIUM

VAPOR

ttI

OTHER CONTAMINANT
DRIFT

PROCESSED
WATER
FEED

DISCHARGE
TO RIVER

Figure 7.11.

7.2.4.1

Simplified Process Diagram Forced Evaporation through Cooling Tower.

Effluents and Releases from Normal Operations

The alternatives described above which involve bulk liquid storage in the processed water storage
tanks, immobilization, discharge to the Susquehanna River, natural evaporation in a lined pond,
and forced evaporation through the cooling tower lead to releases under normal conditions ofoperations.
Storage in Processed Water Storage Tanks

During the transfer of processed accident water to and between processed water storage tanks,
gaseous effluents could be released through the tank vents. This can occur during a tank filling
operation because the air is displaced through the tank vent. Assuming that-the air in the tankis in equilibrium with the water at 100 percent relative humidity and the tritium concentrationin the water is 0.95 pCi/mL (the maximum), the air leaving the vent would contain 34 x 10-6 pCi/mLof tritium. For a filling rate of 30 gpm, the tank venting rate would be about 4 cfm or 3.8 pCi
of tritium per minute, which would result in a total discharge of 0.13 Ci of tritium.
Discharges to the Susquehanna River
The characteristics of the effluents released to the river depend on the quality of the processed
accident water and the dilution factor. Section 7.3.3.2 assumed the maximum volume of water
available for dilution would be 34,000 gpm of cooling water. This quantity of dilution water anda discharge rate of 10 gpm were used to determine the characteristics of effluents at the discharge point. These characteristics are shown on Table 7.25, the assumed best-case conditions.
Similar information on the assumed worst-case condition is presented in Table 7.26.

7-57

Table 7.25.

Releases to the Susquehannj River for the,. Best-Case--SDS/EPICOR II

Averageb
Concentration(pCi/mL)Radionuclide
Inventory(curies)

Average
Concentration
After Dilutionc(pCi/mL)0.50

2.9 x 103

1.5 x 10-4

Cs-1379.8 x 10-'

5.7 x 10-1

2.9 x 10-8

Cs-134Sr-90

4.3 x 10-5
1.5 x 10-8

2.5 x 10-'8.7 x 10-5

1.3 x 10-8,
4.4 x 10-12

Sr-89

1.0 X i0-9

5.9 x 10-6

2.9 x 10-1-

Ru-106

7.2 x 10-6

4.2 x 10-2

Sb-125l x 10-5

7 x 10-2

2.1 x 10-9
2.9 x 10-9

Te-127m

1.7 x 10-5

1 X 10-i

5 x 10-9

Ce-144

3.9 x 10-6

2.3 x 10-2

1.1 x 10-9

Radionuclide
H-3

aprocessed water with average characteristics shown in Table 7.23.
b~ased on volume of 1,540,000 gallons.
cDilution factor of 3400 based on 10 gpm discharge.

Table 7.26.

Release to the Susque anna River for the
Worst-Case SDS

Averageb
Concentration(pCi/mL)Radionuclide
Inventory(curies)

Average
Concentration
After Dilutionc(pCi/mL)0.50

2.9 x 103

1.5 x 10-4

Cs-1378.6 x 10-45

2.5 x 10-7

Cs-134

1.6 x I0-4

9.3 x 10-1

4.7 x 10-8

Sr-90

1.5 x 10-3

8.7

4.4 x 10-7

6 x 10-'3 x 10-8

Radonuclide
H-3

Sr-89

1 x

10-4

Ru-106

3.6 x 10-3

21

1.1 X 10-6

Sb-125
Te-127m

9.3 x 10-3
8.8 x 10-35451

2.7 x 10-6
2.6 x 10-6

Ce-144

9.1 x 10-4

5.3

2.7 x I0-7

aWater treated with SDS on once-through basis will have to be recycled toachieve water quality acceptable for discharge in accordance with
Appendix I effluent technical specifications.
bBased on volume of 1,540,000 gallons.
cDilution factor of 3400.

7-58
Evaporation from the Lined Ponds
The only effluent of concern released during normal operations of the lined evaporation ponds
would be tritiated water vapor.
The amount of tritiated effluent released depends on varying
meteorological conditions, i.e. time of ponding, and initial concentration of the water.' For thecase of July ponding of 0.50 pCi/mL water, approximately 1300 Ci could evaporate during the firstmonth, 700 Ci during August and about 300 Ci during September.
The remaining inventory would
continue to be diluted by evaporation and precipitation.
If January ponding of the water isimplemented, it would take about six months to release this same 2300 Ci.
In either case, mostof the tritium evaporates within the first year after ponding. The total tritium inventory of
2910 curies would ultimately be evaporated over a period of several years, to a concentrationlevel comparable to that in the Susquehanna River (200 to 400 pCi/L).
Forced Evaporation in Coolinq Tower
The assumptions presented in Section 7.2.3.5 for cooling tower performance were used to estimate
releases.
Table 7.27 shows estimated radionuclide releases to the atmosphere and the liquid
releases via cooling tower blowdown discharges.
The liquid releases would have concentrations
14.3 times the effluent concentrations, and would be diluted with up to 34,000 gpm of water
before discharge to the river.
Table 7.27 presents these releases for best- and worst-caseprocessed water effluents to the cooling tower.
The quantities of radionuclides shown would be released intermittently over 260 hours of cooling
tower operation during a 12- to 15-month period. The actual release rates would be dictated bycompliance with levels which satisfy 10 CFR Part 20 criteria.

Table 7.27.

Radionuclide

Normal Releases from Cooling Tower Evaporationof Process Water

In Influent

Best-Case Influent SDS/EPICOR II
H-3
2910
Cs-137
0.57
Cs-134< 0.25
Sr-90
8.7 x 10-5
Sr-89
Sb-125
Ru-106
Te-127m
Ce-144

5.9 x 10-6
7.1 x 10-2
4.2 x 10-2
1 x 10-1
2.3 x 10-2

Curies Released
To Atmosphere

To Blowdownb

2710
0.16
< 0.07
2.5 x 10-5

200
0.41
< 0.186.2 x 10-5

1.7 x
2 x
1.2 x
2.9 x
7 x

10-6
10-2
10-2
10-2

10-3

4.2 x
5.1 x
3.0 x
7.1 x
1.6 x

10-6
10-2
10-2
10-2
10-2

Worst-Case Influent SDS a
H-3
2910
2710
200
Cs-137
5.0
1.4
3.6
Cs-134
9.3 x 10-i
2.7 x 10-1
6.6 x 10-1
Sr-90
8.7
2.5
6.2
Sr-89
6 x 10-1
1.7 x 10-1
4.3 x 10-1
Sb-125
54
15.4
38.6
Ru-10621
6.1
14.9
Te-127m
51
14.6
36.4
1.5
3.8
5.3
Ce-144aWater treated with SDS on once-through basis will have to be recycled
achieve water quality acceptable for discharge in accordance with
Appendix I effluent technical specifications.
bCan be diluted as needed prior to release.

7-59
7.2.4.2

Accident Releases

The worst-case accident arises due to rupture of an outside processed water storage tank containing quantities of residual radioactivity at the upper limit of the storage criteria discussed
in Section 7.2.2.1.
For this evaluation a 500,000-gallon water storage tank containing a composite of processed AFHB water and RCS and reactor building sump water processed by the SDS
system is assumed to rupture and release its contents instantaneously.
Water from the tank would be partly absorbed into the ground and partly carried over several
hundred yards (approximately 250 meters) via the site surface drainage system to the flood control
facility located along the east channel of the island.
The water reaching the collection basinof the flood control facility would be prevented from entering the east channel of the river by amechanical sluice gate which is normally closed.
Even if the sluice gate was inadvertently left
in the open position, it could be closed relatively quickly following a tank rupture.
For thepurpose of this evaluation, however, it was conservatively assumed that half of the water fromthe ruptured tank would enter the river through an open sluice gate and half would be absorbed
into the ground.*Based on qualitative consideration of the following factors the processedwater (250,000 gallons) is assumed to enter the river over a two hour period.
The rate of spillage from the ruptured tank,Physical obstructions between the tank and collection basin,
The volume of water in the tank at the time of the break,The quantity of fresh water drainage from the site (e.g.,rainstorm).

if

rupture occurred during a

A two-hour period is considered to be a lower, and therefore conservative,
period over which the river release would be expected to occur.

limit of the time

Once the processed water entered the east channel it would be either carried downstream or retained
behind Red Hill Dam.
The NRC staff estimates that Red Hill Dam would overtop at SusquehannaRiver flow rates in excess of 16,000 ft 3 /sec.
For the situation where the dam was not being
overtopped, the radioactive contamination would mix over a period of days with the bulk of thewater in the east channel.
(Drinking water is not taken from this source.)
When the dam was
finally overtopped, the water would be carried downstream, but at a very low concentration as aresult of mixing with the river flow. For the situation where the east channel was flowing andovertopping the dam, contaminated water would be carried downstream as soon as it enters thechannel.
The two nearest downstream drinking water intakes are the Brunner Island electric station, approximately 5 miles from the site, and the Columbia intake, approximately 16 miles from the site. The
Brunner Island intake, which is not a public source of drinking water, is on the west side of theriver, whereas the contaminated water would be released along the east side of the river. Therefore, the upper bound estimate of the concentrations of radionuclides at this intake is determinedbased on the assumption that the radionuclides become uniformly mixed over the cross section of
the channel.
Longitudinal mixing, however, would be unlikely.
In order to estimate the maximum amount of radioactivity which could be released to the river asa result of tank rupture it is necessary to apply the following criterion discussed in Section 7.2.2.1.

*The radionuclides in the processed water which is absorbed into the ground will not enter
the river for a year or more (see Appendix V).
The gradual release of these nuclides will
result in insignificant environmental impacts (i.e., doses which are a small fraction ofAppendix I to 10 CFR Part 50, and 10 CFR Part 20).

7-60
Specifically this criterion requires the licensee to limit processed water storage tank content
such that if the tank is breached, the following equation is satisfied.

Ci
MPCi

< 1

where
Ci = concentration (pCi/mL) of the ith nuclide at the nearest downstream drinking water
intake (Brunner Island)
MPCi = maximum permissable concentration (10 CFR Part 20, Table II, Col. 2 - pCi/mL).
The concentration at the nearest drinking water intake (Ci,

Ci

= Ai

(0.5) Ai-pCi/mL) is determined by:

QAt

6.4

10

where
Q = minimum river flow (16,000 ft 3 /sec, cfs) to overtop Red Hill Dam
At = release period (2 hours)
Ai = tank activity prior to rupture (curies)
0.5 = fraction of the tank volume that discharges to the river
6.4 x 106 is a constant (Ci/pCi/mL)
Combining these two equations yields the limiting storage requirement expressed by the criterion
listed above.
Ai

P i < 6.4 * 106

Thus, this expression defines the maximum release of radioactivity to the river which could
result from a rupture of an outside process water storage tank.
Based on this, Table 7.28 presents a maximum source term for this accident.
These values were
calculated using the relative radionuclide spectrum and quantities listed in Table 7.24 for acomposite of processed AFHB water and RCS and reactor building sump water processed by the SDS
system.
Table 7.28 represents an upper limit estimate of radionuclide inventory which could be
released to the east channel as a result of a tank rupture.
It should be noted that, in accordance with the tank failure model (i.e., 50 percent of the tank inventory is discharged to theriver), the radionuclide inventory in Table 7.28 is only 50 percent of the inventory that could
be stored in the tank.
7.2.5
7.2.5.1

Environmental

Impacts

Occupational Doses

The principal radionuclide in the processed water is tritium, a soft beta emitter, which does notconstitute a direct radiation source.
Tritium is, however, an internal source of radiation if
ingested by workers.
Since it is not contemplated that the workers will ingest processed water,
there are no occupational doses attributable to processed water disposal.

7-61

Table 7.28. Maximum Activity
Released to the SusquehannaRiver due to a Rupture of aProcessed Water Storage Tank

Radionuclide

Activity(Ci)

H-3
Cs-137280
0.49

Cs-134

0.09

Sr-90
Sr-89

0.84
0.058
2.0

Ru-106
Sb-125
Te-127m

5.2
5.0

Ce-1440.51
aThese numbers represent one-halfthe maximum activity which could
be stored in an outside processedwater storage\tank in accordancewith the model described in thissection and in Section 7.2.2.1.

7.225.2

Offsite Doses

The dose estimates presented here for disposal of the processed water are based on the source
terms of Section 7.2.4.
The calculational models used to make these estimates are described inAppendix W. The environmental impacts of the four options for disposing of the processed water
are considered here. The four options are (1) discharging directly to the river, (2) discharging
to the atmosphere via natural evaporation from a holding pond, (3) combined discharge to theatmosphere and river by forced evaporation and blowdown, and (4) ship offsite to a waste disposalsite. The significance of the doses described below and their human health and environmental
consequences are discussed in Section 10.3.
The option of shipping the processed water offsite would result in no exposure to the generalpublic as the only nuclide of significant concentration is tritium. No exposure to the publicwould occur because tritium emits beta radiation which would not penetrate the shipping containers.
The dose estimates for the controlled release to the river option for the maximum exposed individual are listed in Tables 7.29 and 7.30 for the SDS/EPICOR II and SDS alternatives, respectively.
The downstream population dose for the SDS/EPICOR II processing option was estimated to be
30 person-rems and for the SDS processing option was estimated to be 900 person-rems.
The dose estimates for releasing the processed water to the atmosphere via natural evaporation
are based on the assumption that 2910 curies of tritium are released uniformly over the period of
one year. The dose received by the maximum exposed individual is listed in Table 7.31. Notethat for tritium the dose to each body organ is the same as that for the total-body. The 50-mile
total population dose was estimated to be 30 person-rems.

7-62

Table 7.29. Dose Estimates to the Maximum Exposed Individualfor Relea~ing Processed Water to the River SDS/EPICOR-II Option
Dose (mrem)a
Pathwaya

Total-Body

Bone

Liver

Fish consumption

1.1

1.2

Drinking water

2.4 x 10-2

1.1 x 10-2

1.5
2.0 x 10-2

Shoreline exposure

8.5 x 10-4

9.9 x 10-4

4.8 x 10-3

1.1

1.2

1.6

Total

aDoses were calculated for total-body, GI-tract,bone,

liver,

kidney, thyroid, lung and skin.
The maximum three organ doses
are listed in this table.
Doses were calculated for four age
groups; adults, teenagers, children and infants.
The highest
dose estimates for each age group are listed.
The total-bodyvalues are for adults, the bone values are for children, andthe liver values are for teenagers.
bThe basis for selecting the special locations is described inAppendix W. The assumption is that drinking water is from thenearest downstream municipal intake, and shoreline exposure andfish consumption is from the center channel in the river.

Table 7.30. Dose Estimates to the Maximum Exposed Individualfor Releasing Processed Water to the River - SDS Option

Pathwaya

Total-Body

Dose (mrem)aBoneFish consumption
Drinking water

9.1
6.4 x 10-1

18
4.7

1
6.4 x 10-2

Shoreline exposure

2.1 x 10-2

2.4 x 10-2

1.1 x 10-1

9.8

23

11

Total

aDoses were calculated for total-body, GI-tract, bone,Liverliver,

kidney, thyroid, lung and skin. The maximum three organ doses
are listed in this table. Doses were calculated for four age
groups; adults, teenagers, children and infants. The highest
dose estimates for each age group are listed. The total-bodyvalues are for adults, the bone values are for children, andthe liver values are for teenagers.
bThe basis for selecting the special locations is described inAppendix W. The assumption is that drinking water is from thenearest downstream municipal intake, and shoreline exposure andfish consumption is from the center channel in the river.

7-63

Table 7.31. Dose Estimates to the Maximum Exposed
Individual for Releasing Processed Water
to the Atmosphere via Natural Evaporation

Location
Nearest b

gardenPathway

Dose (mrem)a
Total-Body

Inhalation

2.0 x 10-1

Ground Shine
Vegetable Use0
7.0 x 10-1

Total
Nearest milkgoat

Inhalation
Ground Shine
Goat Milk Use
Total

Nearest cow
and garden

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

9.0 x 10-1
1.0 x 10-10
7.9 x 10-1
8.9 x 10-1
2.2 x 10-10
7.7 x 10-1
3.1 x 10-1

Total
1.3
aDoses were calculated for Gi-tract, bone, liver,kidney, thyroid, lung and skin. The dose estimates
for all organs except bone were the same as that for
total-body. The estimates for bone were zero. Doses
were calculated for four age groups; adults, teenagers, children, and infants. The largest of eachgroup are listed in this table. Infant doses arelisted for the nearest milk goat location and childdoses are listed for the other two.
bThe basis for selecting the special locations is
described in Appendix W. The actual locations
are: nearest garden = 1.05 mile east-northeast,nearest milk goat = 1.02 mile north, and nearest
cow and garden = 1.05 mile east.

The dose estimates for the forced evaporation option for the maximum exposed individual are
listed in Tables 7.32 and 7.33.
The combined 50-mile population dose and downstream population dose were estimated to be
30 person-rems for the SDS/EPICOR II processing option and 1000 person-rems for the SDS processing option, respectively.**The population dose estimates have been rounded to one significant figure. The combined
50-mile population dose and downstream population dose of 30 person-rems for the forced
evaporation SDS/EPICOR II processing option was based on a dose of 7 person-rems from theliquid pathways and a dose of 26 person-rems from the atmospheric pathways. The combined
50-mile population dose and downstream population dose of 1000 person-rems for the forced
evaporation SDS processing option was based on a dose of 590 person-rems from the liquid
pathways and 440 person-rems from the atmospheric pathways.

7-64

Table 7.32.
Dose Estimates for the Maximum Exposed Individual for Releasing
Processed Water Via the Forced Evaporation Option - SDS/EPICOR II
Processing Option
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Inhalation
Ground Shine
Vegetable Use

1.9 x 10-1
2.7 x 10-2
7.4 x 10-1

1.0 x 10-2
2.7 x 10-2
4.7 x 10-1

2.0 x 10-1
2.7 x 10-2
1.2

9.6 x 10-1

5.1 x 10-1

1.4

2.0 x 10-1
2.6 x 10-2
6.3 x 10-1

5.6 x 10-3
2.6 x 10-2
2.3

1.0 x 10-1
2.6 x 10-2
3.8

8.6 x 10-1

2.3

3.9

2.1 x 10-1
4.0 x 10-2
8.5 x 10-1
3.9 x 10-1

1.1 X 10-2
4.0 x 10-2
6.9 x 10-1
5.4 x 10-1

2.2 x 10-1
4.0 x 10-2
1.5
8.9 x 10-1

1.5

1.3

2.7

7.6 x 10-1
4.7 x 10-3
6.1 x 10-4

8.7 x 10-1
8.1 x 10-3
7.1 x 10-4

1.1
5.3 x 10-3
3.4 x 10-3

7.7 x 10-1

8.8 x 10-1

1.1

kidney,lung

Atmospheric Releases:
Nearest gardenb

Total
Nearest milkgoat

Inhalation
Ground Shine
Goat Milk Use
Total

Nearest cow
and garden

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use
Total

Liquid Releases:
Near TMI

Fish Consumption
Drinking Water
Shoreline Exposure
Total

a Doses were calculated for total-body, GI-tract,bone,

liver,

thyroid,and skin.
The maximum three organ doses are listed in this table.
Doses were calculated for four age groups; adults, teenagers, children, and infants.
The highest
dose estimates for each age group are listed.
For the atmospheric releases, thedoses are for children for the nearest garden and nearest cow and garden locations,and for the nearest milk goat location are for adults for total-body and for infants
for bone and liver.
For liquid releases, the doses for total-body are for adults,
for bone are for children, and for liver are for teenagers.
bThe basis for selecting the special locations is described in Appendix W. The actual
locations for atmospheric releases are:nearest garden = 1.05 mile east-north-east,nearest milk goat = 1.02 mile north, and nearest cow and garden = 1.05 mile east.
For liquid releases the assumption was made that drinking water is from the nearest
downstream municipal intake, and shoreline exposure and fish consumption is from thecenter channel in the river.

7-65

Table 7.33. Dose Estimates for the Maximum Exposed Individual for Releasing
Processed Water Via the Forced Evaporation Option - SDS Processing Option
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Inhalation
Ground Shine
Vegetable Use

1.1
1.8 x 10-177

14
1.8 x 10-1300

4.3 x 10-1
1.8 x 10-1
4.4

Atmospheric Releases:
Nearest garden

Total
Nearest milkgoat

Inhalation
Ground Shine
Goat Milk Use

78

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

5.0

4.3 x 10-'
1.7 x 10-1
145.2
1.7 x 10-164

2.4 x 10-1
1.7 x 10-1
23

15

69

23

1.2
2.6 x 10-1
110
6.7

16
2.6 x 10-'44027

4.8 x 10-1
2.6 x 10-1
6.2
4.5

480

11

Total
Nearest cow
and garden

310

Total

120

Liquid Releases:
Near TMI

Fish Consumption
Drinking Water
Shoreline Exposure

6.5
4.4 x 10-1
1.5 x 10-2

13
3.3
1.7 x 10-2

7.9
3.7 x 10-2
8.2 x 10-2

7.0

16

8.0

Total
aDoses were calculated for total-body, GI-tract,bone, liver, kidney, thyroid, lungand skin. The maximum three organ doses are listed in this table. Doses were calculated for four age groups; adults, teenagers, children, and infants.
The highest
dose estimates for each age group are listed.
For the atmospheric releases, thedoses are for children for the nearest garden and nearest cow and garden locations,and for the nearest milk goat location are for adults for total-body and for infants
for bone and liver.
For liquid releases, the doses for total-body are for adults,
for bone are for children, and for liver are for teenagers.
bThe basis for selecting the special locations is described in Appendix W. The actual
locations for atmospheric releases are:nearest garden = 1.05 mile east-north-east,nearest milk goat = 1.02 mile north, and nearest cow and garden = 1.05 mile east*
For liquid releases the assumption was made that drinking water is from the nearest
downstream municipal intake, and shoreline exposure and fish consumption is from thecenter channel in the river.

7-66
7.2.5.3

Postulated Accident Effects

The accident described in Section 7.3.4.2 that involves releases of processed water to the riveris breaking of a processed water storage tank.
For this accident scenario, 50 percent of thereleased water will run over the surface to the flood control facility and drain into the east
channel of the Susquehanna River.
The potential offsite dose consequences to humans is dependent
upon whether or not Red Hill Dam is overtopping.
If Red Hill Dam is overtopping, the released
water will be diluted with the flow of the Susquehanna River resulting in doses to humans thatare fairly low and within annual limits for routine operation.
However, if Red Hill Dam is notovertopping, the released radioactivity could reside in the east channel for an extended periodof time at typically fairly high concentrations.
Dose calculations are presented here for bothriver flow situations.
Flowing East Channel
The source terms that were used for the tank break accident are listed in Table 7.28. The calculations are based on the assumption that the SDS processing option is used for reactor building
sump and RCS accident water.
This is the least effective processing option, hence, if a more
effective processing option is chosen, an accidental release would probably result in lower doses
than presented here.
Table7.34 lists the offsite dose estimates to the maximum exposed individual resulting from the tank rupture accident scenario.
These dose estimates were based on thecalculational methods described in Appendix W. The significance of these doses related to humans
are discussed in Section 10.4.
The ecological significance of these accidents is discussed in
Section 7.2.5.4.
Stagnant East Channel-The dose estimates for the accident situation where it is assumed Red Hill Dam is not overtopping
were calculated based on the estimates for the situation where it was overtopping, except that acorrection factor was employed to take into consideration the decreased dilution of the east
channel, and an adjustment was made to present dose estimates on a unit consumption basis.
Table 7.35 lists dose estimates for consumption of 1 liter of water or 1 kg of fish flesh fromthe east channel for the and SDS processing option. The fish consumption doses are fairly high,even for the lowest case.
However, should an accident occur, consumption of fish caught fromthis pond could certainly be avoided.
The significance of these doses related to humans andpossible mitigative measures are discussed in Section 10.4.
The ecological significance of these
accidents is discussed in Section 7.2.5.4.
7.2.5.4

Radioecological Consequences of Discharge to the Susquehanna River

Additional factors which must be considered in the evaluation of the consequences of liquidradioactive discharges from the cleanup include the transport pathways, environmental removal
mechanisms, accumulation and persistance of the radionuclides and consequences of radiation
exposure to organisms other than man.
The radionuclides of primary concern in these analyses are
Cs-137, Sr-90 and H-3.
Cesium-134, Sr-89, Sb-125 and traces of other radionuclides are contaminants present in radioactive waste streams but are of lesser concern, either because they are in
low concentrations or have low dose factors.
Further details on the significance of the nuclide
mix are given in Appendix J.
Release Scenarios
The staff has postulated several mechanisms for the release of treated radioactive water from TMI
and evaluated the radioecological consequences to the biota of the Susquehanna River and Chesapeake Bay.
Controlled Release
Effluent discharges to the Susquehanna River from TMI during planned controlled releases oftreated water would occur at the combined two-unit discharge structure on the west shore of Three
Mile Island (Figure 1.3).
This river discharge structure, therefore,,would release effluents
into the center channel of the Susquehanna River as it flows south toward the York Haven Dam and7-67

Table 7.34.
Dose Estimates to the Maximum Exposed Individual
Caused by Rupture of a Processed Water Storage Task during
High River Flows - SDS Processing Option
Dose (mrem)b
Pathwayc

Total-Body

Fish consumption

8.9 x 10-1

Drinking water

6.2 x

Shoreline exposure
Total

Bone

Liver

1.7

1.1

2.0 x 10-3

4.6 x 10-'
2.3 x 10-2

6.2 x 10-3
1.1 X 10-2

9.5 x 10-1

2.2

1.1

10-2

aThese doses are based on the assumption that the flow in theriver during the accident is sufficient to overtop Red Hill Dam.
bDoses were calculated for total-body, GI-tract, bone, liver,kidney, thyroid, lung and skin.
The maximum three organ doses
are listed in this table. Doses were calculated for four age
groups; adults, teenagers, children and infants.
The highest
dose estimates for each age group are listed. The dose estimates
for total-body are for adults, for bone are for children, and forliver are for teenagers.
cThe basis for selecting the special locations is described inAppendix W. The assumption is that drinking water is from thenearest downstream municipal intake, and shoreline exposure andfish consumption is from an area downstream of TMI.

Table 7.35.
Dose Estimates for Consumption of Drinking Water
or Fish Flesh from the East Channel \due to an Accidental
Processed Water, Storage Tanka upture during
Low River Flows
Adult Total-Body
Dose from Consuming
1 liter of Water from
East Channel (mrem)

Adult Total-Body
Dose from Consuming
1 kg of Fish from
East Channel (mrem)

SDS/EPICOR IIc

0.14

56

SDSb

3.8

470

Processing
OptionaThe dose estimates here are based on the assumption that the tankruptures during an extended period of low-river flow such thatRed Hill Dam is not being overtopped.
bValues determined by multiplying appropriate values in Table 7.34
by the increased concentration factor (44,900 for drinking
water and 11,200 for fish consumption), divided by the annual
usage rate (730 1/yr for drinking water and 21 kg/yr for fishconsumption).
CThese values were estimated by scaling the values for the SDS
processing option in this table by the ratio of the appropriate
values in Table 7.29 to the appropriate values in Table 7.30.
The scaling factor for drinking water was determined by dividing
the total-body dose for drinking water in Table 7.29 by the
total-body dose in Table 7.30 (2.4 x 10-2/6.4 xlO-1 = 0.038).
In a similar fashion the scaling factor for fish consumption
was estimated (1.1/9.1 = .12).

7-68
the York Haven Generating Station (hydroelectric).
Studies of river flow and effluent plume
behavior at TMI have shown that the plume stays very close to the shoreline as it moves southwith the center channel river flow. The wide shallow nature of the river apparently results in
low initial mixing of tributary waters with river flow.
This type of flow phenomena restricts the effluent from TMI to very near the west shore of TMI17
until it either passes over the York Haven Dam or is discharged from York Haven Pond through thehydroelectric station. After exiting the Pond, the effluent receives a considerable amount of
dilution with the River flow. The most productive spawning and nursery areas of important fishes
in the Pond are in the east and west river channels (Appendix E) and the areas most frequently
fished by anglers are upstream of TMI between Hill, Shelley, and Fall Islands. Therefore, during
planned releases, the effluent from the nuclear station discharge would not intercept or directly
contact those important areas of York Haven Pond.
To illustrate, controlled releases, 10 gpm of treated water was assumed to be diluted with 34,000
gpm of cooling tower blowdown before release to the center channel of the river through the plantdischarge structure. The river flow during this release was assumed to be 12,600 CFS, which is aweighted average flow for dilution, whose derivation is described in Section 3.4.1. The centerchannel of the Susquehanna River is presumed to be carrying 25 percent of the total river flow.
Exposure to biota living above York Haven Dam are calculated on the basis of the flowrate in thecenter channel, while exposure to organisms downstream of the dam are based on the total riverflow.
The expected activities of the radionuclides of concern in the processed water are listed inTable 7.25 and 7.26 for the SDS/EPICOR II and SDS processing options, respectively. Table 7.36
lists the background concentrations of H-3, Cs-137 and Sr-90 in the water of the SusquehannaRiver and the detection capability of the Licensee's monitoring program.
Also listed inTable 7.36 are the annual average and instantaneous concentrations in the river below York Haven
Dam resulting from the release of water treated by the two options based on the 10 gpm release
rate and 12,600 CFS river flow.
Instantaneous concentrations due to the releases of treatedeffluent from the plant would, of course, depend on the rate of release and the river flow rate
at that time.
Table 7.36 reveals that only H-3 and possibly Sr-90 (SDS processing) could be detected in levels
above background for either treatment option.
Since doses and exposures to man and otherorganisms are calculated on a yearly average basis, the yearly average concentrations in Table 7.36
are more useful from the standpoint of impact evaluation than the instantaneous values.
Accidental Release
The other scenario considered for the release of the treated water involves the instantaneous
rupture and spillage of a 500,000-gallon holding tank. Water from the tank would run over surface
to the flood control facility and be routed to the east channel of the river above Red Hill Dam,an important spawning and nursery area of the river. The concentration of radionuclides in Red
Hill Pond would depend on whether or not the flow rate in the Susquehanna River was sufficient toovertop Red Hill Dam.
For flows below 16,000 CFS, the dam does not overtop and there would be
only limited exchange of water between the pond and the Susquehanna River at the upstream end ofThree Mile Island. The concentration in the pond for this case was based on the assumption that
one-half the total contents of the 500,000-gallon holding tank spilled into and was diluted by
the approximately 65-million-gallon volume of Red Hill Pond. The concentrations of radionuclides
were assumed to remain at these levels until the river stage was sufficient to overtop Red Hill
Dam. For flows greater than 16,000 CFS, the dam would be overtopped and the radionuclides would
be carried downstream.

Table 7.36.

Background and Added Concentrations of Radionuclides in Susquehanna River B~low York Haven Dam
due to Controlled Releases from SDS/EPICOR II and SDS Treatment Systems

River Concentrations (pCi/mL)
SDS/EPICOR II

SDS

Licensee's

Background b
Concentration

InstantaneousRiver ConEentration

Annual AverageRiver Con8entration

InstantaneousRiver ConEentration

Annual AverageRiver Con8entration

Lower Limitof Detection(pCi/mL)

H-3

1.8 X0-7

8.9 x 10-7

2.6 x i0-7

8.9 x 10-7

2.6 x I0-7

3.3 x I0-7

Cs-1372.8 x 10-9

1.7 x 10-1o

5.1 x 10-11

1.5 x 10-9

4.5 x 10-l0

6.0 x 10-9

Sr-90

2.3 x 10-9

2.7 x 10-14

7.8 x 10-'s

2.7 x 10-9

7.8 x 10-1o

1.0 x 10-9

RadionuclideaThe annual average concentration and doses also apply to the accidental

release scenarios downstream of Red

Hill Dam.
bFrom Section 3 of this document.
cBased on source terms of Tables 7.25 and 7.26 and the volumes of Tables 7.23 and 7.24 released at a rate of
10 gpm into a river flow of 12,600 cfs.
dBased on all activity of Tables 7.25 and 7.26 released over a one-year period into a river flow of 12,600 cfs.
eFrom "Three Mile Island Nuclear Station Radiological Environmental Monitoring Program, Annual Report, 1979,"
Metropolitan Edison Company,

April 1980,

p.

12.

0&

7-70
Behavior of Radionuclides in Susquehanna River
Radionuclides entering the Susquehanna River at TMI-2 will move in the downstream direction withthe river flow. Some radioactivity will deposit in bottom sediments and some will be taken up by
aquatic organisms.
Tritium discharged to the river will be in the form of tritiated water (HTO
as compared to normal water, H2 0).
The chemical and physical behavior of tritiated water isbasically that of ordinary water with regard to mobility, interaction with suspended particles,
and general dilution.
In contrast, the isotopes Cs-134 and Cs-137 have an appreciable tendency to be adsorbed onto
suspended sediments and to concentrate in aquatic organisms.
Dissolved cesium introduced to theriver will partition itself between the water and the suspended sediment.
Contaminated sediments
would collect to a certain degree behind the York Haven, Safe Harbor, Holtwood and Conowingo
Dams, but would not reside there indefinitely since much of it would be flushed into the Chesapeake Bay during major floods. Although radioactively contaminated sediments would be resuspended
during floods, they would simultaneously be highly diluted by the great quantity of flowing
water.
As suggested in Table 7.36, it is highly unlikely that any of the radiocesium released from the
TMI cleanup could be detected in the Susquehanna River, except possibly in the immediate area ofthe site.
Radiocesium levels would be masked by background levels caused by nuclear explosive
testing fallout, previous normal operating releases from Three Mile Island Units 1 and 2, andcurrent releases from Peach Bottom Units 2 and 3.
The isotopes of strontium have a much smaller tendency than cesium to adsorb onto sediments, andthus will remain almost completely in the water. The effects of sediment on the radioecology ofSusquehanna River and Chesapeake Bay are expected to be inconsequential.
Details of the importance of sediment are given in Appendix T.
Behavior of Radionuclides in the Chesapeake Bay
The upper Bay area, including the Susquehanna Flats and its tributaries, are major spawning andnursery grounds for many important species of fishes and other aquatic biota. Some of the radionuclides discharged to the Susquehanna River at Three Mile Island will ultimately be carrieddownstream to the Chesapeake Bay either in dissolved form or on suspended sediment. The Chesapeake Bay (Figure 7.12) is an estuary where fresh waters from the rivers and salt water from theocean mix. Salinity varies from practically zero at the mouths of the rivers up to about 35 parts
per thousand (PPT) in the open ocean.
Substances which enter Chesapeake Bay mostly or entirely in the dissolved form such as all of thetritium, most of the cesium, and practically all of the strontium will eventually be transported
to the sea. The transport would be caused by the combined effects of advection by fresh waterand dispersion caused by the astronomical tides and wind wave activity. The "flushing time," inthe Bay is approximately one year. The flushing time is indicative of the rate at which a
dissolved pollutant would be purged from the Bay, but because of the effects of sediments and thenonideal mixing and entrapment in the complicated backwater areas of the Bay, small traces ofradioactive contamination would probably linger for several years.
The bulk of the sediments transported into the Chesapeake Bay will be deposited in the upper Bay
in the region known as the "turbidity maximum." 2 2 Eaton suggests that some SusquehannaRiver23
derived sediment would be deposited as far down-Bay as the mouth of the Potomac River.
Direct measurements of radionuclides released from the Peach Bottom Nuclear Plant in Conowingo
Pond suggests that the maximum sediment deposition of cesium is at the mouth of the SusquehannaRiver, and has decreased by two orders of magnitude by the Sassafras River. Radiocesium levels
are undetectable beyond that point. 2 4 The measurements may not have included the periods ofmajor sediment transport during floods, which would both dilute and disperse contaminated sediment further into the Bay to levels below detection limits.
Once the contaminated sediments enter the brackish waters of the Bay, a portion of the Cs-134 2 5which was previously adsorbed will be released back into the water column in a dissolved state.
Estimates of radiation doses made below for the Susquehanna River represent maximum expectedvalues.
Even with desorption of radionuclides into the water column, radiocesium concentrations
are expected to be lower than in the river.

7-71

CAPE HENRY

Figure 7.12.

Chesapeake Bay Area.

7-72
Chesapeake Bay is habitat for a great variety of aquatic life, some of which is consumed by man.
Maximum concentrations of radionuclides in the Bay water would not exceed those in theSusquehanna River because the river water undergoes further dilution with salt water and othersources of freshwater in the Bay.
The Susquehanna River discharges approximately 52 percent of the total fresh water input to the
Chesapeake Bay, and about 80 percent of freshwater input above the Potomac River. 2 2 For releasesof long duration it is possible to relate the steady state dilution in Chesapeake Bay of
dissolved substances contained in Susquehanna River water directly to the observed salinity.
Table 7.37 gives the additional dilution factors for Susquehanna River water as a function ofsalinity in the Bay.
These factors are underestimates of dilution because they are based on theassumption that all fresh water in the bay comes from the Susquehanna River, while there actually
are other substantial sources of freshwater input.
Figures 7.13a and 7.13b show the salinity
along the axis of the Chesapeake Bay for periods of high flow and low to moderate flow respectively. 2 6 Since doses are calculated on an average annual basis, these dilution factors may also
be applied to radioactive releases of short duration.
It should also be noted, while it appears
that the additional dilution in the Bay would be larger for the lower river flow of Figure 7.13a,the initial dilution of TMI effluent by the river water flowing by the site would initially belower.

Table 7.37.
Conservative Dilution of Sgsquehanna River Water in Chesapeake Bay

Salinity - Parts
per Thousands

Additional Dilutionof SusquehanngRiver Water

0 (Fresh water)

1

2

1.06

5

1.17

10

1.4

15

1.75

20

2.33

25

3.5

30

7

32

11.7

35 (sea water)

Infinity

aFor a steady release of fresh water.

Conser-vatively assumes all fresh water input to Bay
due to Susquehanna River.
bExample:If the Susquehanna River Concentration of Cs-137 were 1.0 pCi/L, the calculated
Chesapeake Bay concentration where thesalinity were 10 parts per thousand would
be 1.0/1.4 = 0.714 pCi/L.

(A)
1229
16
a 20
24

BAY

-

BOTTOM

28
32
40

SALINITY (0/a)
II APRIL 1968

-A
300

280

260

240

220

200

180

160
140
120
0o0
Distance from Entrance of Bay 1km)

80

60

40

20

0

-20

0

-20

(B)

o0•

I',ll 13A

is

16

2 22l

80a
12

9

2102
6
_20

-

BAY

BOTTOM

28. -

'

32 -

SALINITY (0/)
28 FEB 1969

:•

36 -

40 -B
300

280

260

240

220

200

180

160

140

20

t00

80

60

40

20

Distance from Entrance of Bay (k m)Figure 7.13.

Longitudinal Salinity Distribution along Axis of Chesapeake Bay during(A) Period of High River Flow and (B) Period of Low to Moderate River Flow.

7-74
Water exchange and transport in the northern bay area also occurs through the Chesapeake andDelaware Canal, which is the sole direct link between the upper Chesapeake Bay and theDelaware River/Bay system. 2 7 Water flow in the canal can be erratic, with water transported
from either bay through the entire length of the canal and discharged into the other bay. A
net difference in mean tidal level between the Chesapeake and Delaware ends of the canalresults in net transport of water to the eastward at an estimated average of 2450 cfs (-.69 m3 /sec),
in effect making the canal a tributary of the Delaware River. 2 8 ' 2 9 Therefore, it seems probablethat some radionculides of TMI-2 origin that enter the Chesapeake Bay could be expected to betransported into the canal and thereby into the Delaware River, where they could be accumulated
by aquatic species present there (Appendix E).
Doses to Fishes and Other Biota in the Susquehanna River and Chesapeake Bay
Controlled Releases
The dose to fishes and other biota in the Susquehanna River is directly proportional to theconcentration of radionuclides in the river water and to the time period over which the organisms
are exposed.
The concentration of radionuclide in the river is inversely related to the river
flow during discharge of the radionuclide material.
If it is assumed that the flow in the riveris the same for a short-term release as for a long-term release, the annual average doses tofishes below York Haven Dam would be those listed in Table 7.38.
For a controlled release of treated effluent, the maximum nearsite dose to fish presented inTable 7.38 was calculated considering the dilution in the center channel, which is assumed to becarrying only 25 percent of the total river flow. As described in Appendix E, fishes importantto the sport fishery in the Susquehanna River exhibit movement upstream, downstream and acrossthe channel, so they would be exposed to an average concentration rather than the concentration
at a particular point.
The dose to fishes downstream of York Haven Dam assumes total mixing in
all the waters of the Susquehanna River.
The calculational methods used to make these dose
estimates are described in WASH-1258. 30
Bioaccumulation factors are taken from Regulatory
Guide 1.109.

Table 7.38. Average Annual Doses to Fishes in theSusquehanna River Downstream of York Haven Dam
Resulting from Normal or Accidental Releasesof Processed Water
Dose (mrad)a
Scenario

Near Site

Downstream

SDS/EPICOR IIbprocessing option

6.0

1.5

SDS processing
option

99

25

9.6

2.4

Normal Releases

Accidental Releases
Tank failure - SD•processing optionaA river dilution flow of 3150 cfs was used for nearsite
estimates and 12,600 cfs for downstream estimates.
bSee Table 7.25 for radionuclide source term.
csee Table 7.26 for radionuclide source term.
dSee Table 7.28 for radionuclide source term.

7-75
Limited data are available from water and fish samples in the Susquehanna River for estimating
the bioaccumulation factors of cesium and strontium.
Bioaccumulation factors were estimated frombackground levels of Cs-137 and Sr-90 taken at control locations upstream from the site. Bioaccumulation factors for cesium were also independently estimated from the concentrations of Cs-134
in fish caught in Conowingo Pond resulting from the releases of Peach Bottom Units 2 and 3.
Water concentrations in Conowingo Pond are estimated from known Cs-134 releases and an assumed
river flow rate of 12,600 cfs.
The results are tabulated in Table 7.39, and demonstrate that thebioaccumulation factors from Regulatory Guide 1.109, used in the staff's dose assessments, overestimate site-specific values.

Table 7.39.

Radionuclide
Sr-90

Cs-134
Cs-137

Bioaccumulation Factors Estimated from Measured Concentrationsin Fish and Waters of the Susquehanna River

Fish
Concentration(pCi/g)Water
Concentration(pCi/mL)

Bioaccumulation Factor(pCi/g)/(pCi/mL)

0 .0217b(range 0.00040.075)

0 . 0 0 2 3c(range 0 . 0 0 0 3 -b
0.0006)9 .4d(range 1.3-250)

0 .23e0.00058f

0 . 0 5 4 b'(range 0.01- b
0.145)

aRegulatory Guide 1.109,

Rev.

bValues for control locations,

1,

400

0 .0028 c(range 0.0026
0.02)

p.

19
(range 0.5-73)

Regulatory
Guide 1.109
Value(pCi/g)/(pCi/mL)
30

2000
2000

13.

taken from "Three Mile Island Nuclear Station, Radiological Envi-ronmental Monitoring Program, Annual Report 1979," Metropolitan Edison Company, April 1980,
p. 13.
CValues reported in Section 3.4.1 of this document.
dFish concentration divided by water concentration.
eHighest value reported in Conowingo Pond (Reference

24).

fEstimated average concentration based on 1.62 Ci released from Peach.Bottom in Second Quarter,
1979 and 12,600 cfs average river flow.

Effects of Accidental

Releases

The highest doses to aquatic organisms living in the Susquehanna River would be caused by the
postulated accidental release of the contents of the processed water storage tank during an
extended period of low river flow for which there would be no flow over Red Hill Dam.
Table 3.1
demonstrates that statistically, river flows lower than that required to overtop Red Hill Dam andlasting for a half year could have a return period of less than two years.
An extended period of
non-flowing conditions in Red Hill Pond is therefore a reasonable assumption.
In order to demonstrate the upper limit of possible environmental damage to the fisheries of Red
Hill Pond, the staff postulated a one-year period for which Susquehanna River flow was lower than
16,000 cfs. Neglecting exchange of water at the upstream end, the concentration of radionuclides
in the pond was calculated by assuming that one-half the total contents of one of the 500,000-gallontanks spilled into the approximately 65-million-gallon volume of the pond. The concentration ofradionuclides in the pond was assumed to remain at this level for one year.
The doses received
by organisms (fish and fish eggs and larvae) living in the pond were calculated assuming the same
bioaccumulation factors used for the controlled release scenario, and are presented in Table 7.40.
For adult fish, average annual doses are presented in Table 7.40, and for eggs and larvae, doses
over the 60-day developmental period from eggs to juviniles. are present in the table.
The 60-day
developmental period from eggs to juviniles is considered to be the most radiosensitive period
for fishes.

7-76

Table 7.40. Doses to Adult Fishes, Eggs, Embryos and Larvae
in Red Hill Pond Caused by Failurg of a Processed Water
Storage Tank
Accidentb
Type
Failure ofprocessed waWer
storage tankDose
to Adultc
Fishes- Rad

to Eggsd
Dose
Larvae- Rad or

110

18aBased on dilution of one-half contents of 500,000 gallon holdingtank or evaporation pond with 65 x 106 gallon content of Red
Hill Pond, stagnant for 1 year, no interchange with otherSusquehanna River water.
(See Table 7.28 for radionuclide
inventory.)bAll dose estimates based on SDS processing option. Other options
would result in lower doses.
Cone-year exposure based on concentration in pond and Regulatory
Guide 1.109 bioaccumulation factors.
d60-day exposure as above.
eDoses were estimated by multiplying the downstream accidental
release value of Table 7.38 (2.4 mrad) by a factor to take
into consideration the decreased dilution in the east channel
under stagnant flow conditions (decreased dilution = flow inriver in one year/volume in east channel = (12,600 cfs x
3.1 x 107 sec/yr x 7.48 gal/ft 3 )/(6.5 x 107 gal) = (4.5 x 104).

Value for eggs and larvae was estimated to be 1/6 that for
mature fishes.

Doses to Fishes in Chesapeake Bay
Doses to fishes in the Chesapeake Bay would be greatest at the mouth of the Susquehanna River andwould diminish with distance due to the combined effects of increased dilution and the markedly
decreased biological accumulation factors for most radionuclides in saline water. 25 Annual
average doses to fishes would be lower in the Chesapeake Bay than to those fishes residing in theSusquehanna River below York Haven Dam shown in Table 7.38. This conclusion is confirmed byradiological monitoring by the State of Maryland in Conowingo Pond which shows that the maximum
concentration in fish due to radiocesium which has been released from the Peach Bottom nuclear
plant is found in the pond itself, and that concentrations decrease markedly with distance down
the bay. 3'
32Radiosensitivity of Fish

Figure 7.14 and Table 7.41 show that, as a group, fish are the most radiosensitive of the poikilothermic (coldblooded) aquatic animals, although some crustaceans exhibit similar sensitivities.
The data available indicate that the average median lethal dose within 30 days of an acute radiation exposure (LDso/3o) is about 2,400 ± 1,500 rads. For larvae and embryos of fish the LD50 is
considerably lower, about 200 rads.
32
Radiosensitivity of Crustaceans

Crustaceans exhibit a range of radiation sensitivities covering approximately two orders ofmagnitude (Figure 7.14, Table 7.41). Various amphipods and shrimp, exhibit LD50 /30 s of about
2,300 ± 1,400 rads--not significantly different from those of typical fishes.
Some speciesof crabs exhibit LD50 o3oS of about 23,000 ± 17,000 rads.
Because the crabs live significantly
longer and may not shdw lethal effects as soon as smaller crustaceans, a conservative estimate of
the radiation dose inducing somatic effects on adult crustacea would be approximately the same as
for fishes.
Thus, an LD1 0o of 1500 rads is assumed for adults. This estimate considerably
overestimates the impact on certain commercially and recreationally important species, such as
the blue crab (Callinectes sapidus).

7-77

102

106

105

104

10 3

RADs
Figure 7.14.

Tolerance of Aquatic Organisms to Radiation.
(Relative tolerances of different groups ofaquatic organisms to radiation in terms ofdosages of radiation required to kill 50%of the exposed individuals in a given periodof time.)Table 7.41. Comparison of Acute Lethal
Radiation Doses for Various Groupsof Organisms
DoseType of Organism(Kilorads)

Ranges of LDs-

735(LD9 0 )

<400

-

>1200

(LD9 o)Other algae

3

-

120

(LD50 )

Protozoa?

600(LD50 )

Mollusks

20

109(LDso/3o)

Crustaceans

1.5 -

56.6

(L050 /30 )

Fish

1.1 -

5.6

(LD50 /30 )Bacteria
Blue green algae

al.L. Ophel et al.,

4.5-

"Effects of Ionizing Radiation on Aquatic
Organisms," in: "Effects of Ionizing Radiation on Aquatic
Organisms and Ecosystem," Technical Report 172, IAEA, Vienna,
1976.

7-78
3
Radiosensitivity of Mollusks

2

Mollusks generally exhibit a much lower radiosensitivity than fish, although there is comparability with some crustaceans, such as the blue crab (Figure 7.14, Table 7.41).
The LDso /30 for
clams and oysters is about 100,000 rads.
Like crustaceans, they have a much slower respbnse toradiation than mammals.
This results in the LD5 0 for long periods of time being much lower.
For
observation periods ranging from 50 to 90 days, the LD5 0 for adult mollusks is about
28,000 ± 23,000 rads.
It is assumed that a reasonable LD10 0 /1 80 for adult mollusks would be
about 50,000 rads.
Evaluation of Radiation Effects
Controlled Releases
The annual average downstream doses listed in Table 7.38 do not present a significant hazard tobiota of the Susquehanna River and Chesapeake Bay, and are within levels generally established tobe safe for humans.
Although guidelines have not been established for acceptable limits forradiation exposure to species other than man, it is generally agreed that the limits established.
for humans are also conservative for other species.
Experience has shown that it is the maintenance of population stability that is crucial to the survival of a species, and species in mostecosystems suffer rather high mortality rates from natural causes.
While the existence of
extremely radiosensitive biota is possible, and whereas increased radiosensitivity in organisms
may result from environmental interactions with other stresses (e.g., heat, biocides, etc.), nobiota have have yet been discovered that show a sensitivity (in terms of increased morbidity or
mortality) to radiation exposures as low as those expected from normal operations in the areasurrounding and downstream from the Three Mile Island nuclear power plant.
Furthermore, in all
the plants for which an analysis of radiation exposure to biota other than man has been made,there have been no cases of exposures that can be considered significant in terms of harm to thespecies, or that approach the exposure limits to members of the public permitted by 10 CFRPart 20.33
Since the BEIR-III Report 3 4 concluded that the evidence to date indicates that noother living organisms are very much more radiosensitive than man, no significantly radiologicalimpact on populations of biota is expected as a result of the routine decontamination of thisplant.
Accidental

Releases

Doses to biota residing in Red Hill Pond for the postulated holding tank rupture or scenerio
would be much higher than doses expected due to the controlled release scenerios.
Comparison of
the doses estimated for the SDS system (the system which gave the highest doses) in Table 7.40 to
the ranges of senitivities estimated in Table 7.41 suggests that the accidental doses are well
below levels that would significantly affect the adult fish, eggs, embryos or larvae in Red Hill
Pond.
Annual average doses to fishes downstream of Red Hill Dam for the case where the dam isovertopped would be 10 times less than the controlled release scenerio listed in Table 7.38, as
for the SDS processing option, only about 20 percent of the total activity can be stored in thetank at a given time (one-half is assumed to immediately reach the river in an accident).
7.2.5.5

Psychological - Socioeconomic

The staff considers four methods for disposing of processed water to be practicable; these methods
are:onsite storage, evaporation, offsite shipment, and discharge to the Susquehanna River.
Although each would have negligible offsite radiological impacts under normal operating considerations, they can have varying psychological and socioeconomic impacts for different groups ofpeople.
Onsite storage of the processed water either in tanks or mixed with cement to form
concrete blocks for a long period of time may generate chronic psychological distress and socioeconomic impacts among people living in communities near the station.
In fact, the extent ofpotential distress and impact can be related to the length of time required for storage.
Thepossible association between time of storage and distress reflects some individual's belief that
there is an increasing probability of accidents over time and that long-term storage represents
neither decontamination nor disposal.
This latter point is especially important because someresidents believe that storage, even long-term storage, is only a stage preparatory to the ultimate
disposition of the radioactive waste.
A strategy of long-term, onsite storage has the potential
to prolong anxiety.

7-79
Release to the atmosphere through either forced or natural evaporation from the island are possible
alternatives.
As with the long-term storage option discussed above, the extent of anxiety may bepositively related to the length of time required for evaporation.
Moreover, the population
impacted by the evaporation option is expected to be similar to that discussed above.
Dependingon processing scenarios, shipment of processed water offsite in solidified form would involve
640 truck shipments of 51,300 drums, or 340 shipments would be required if the processed water
were shipped offsite in bulk liquid form.
Shipping the processed water to an offsite disposal area can be generally less stressful locallythan either the long-term storage or evaporation options because it removes the believed sourceof threat from the TMI area.
However, those living along or in proximately to the truck route
may be subjected to stress.
Such potential stress would be related to increased awareness ofresidents, and the perceived possibility of on-the-road accidents.
The transportation of waste,particularly through Middletown where population density is higher and people live closer to theroadway than other locations in the local impact area, can stimulate greater stress.
It is thestaff's judgment that the marketability of residential property abutting the route through Middletown conceivably could be adversely impacted during the extended period of shipments.
The final disposal alternative considered practicable by the staff is the controlled release ofprocessed water to the Susquehanna River.
The staff estimates that a minimum of 110 days would
be required to release the entire inventory of processed water.
Although disposal to the Susquehanna would present negligible public health hazards, some residents of local and downriver(including Chesapeake Bay) communities find this alternative both threatening and distressful.
In short, some members of the public do not believe that even the low concentration of tritiumand even lower quantities of other radionuclides remaining after dilution of the processed water
are safe.
The staff concludes that these concerns would be concentrated among those who use theSusquehanna River/Chesapeake Bay system as a water supply, as a source of income and enjoyment,
and those who consume the Bay's resources.
Although models to predict behavior from attitudes are difficult to specify under the best of
circumstances, the staff has concluded that disposal of processed water into the SusquehannaRiver has the potential for producing widespread socioeconomic impacts.
These impacts may include
the following:(1)

temporary avoidance of drinking water originating from either the Susquehanna River or the
Chesapeake Bay and the substitution of other liquids;(2)

losses to recreation as people cease their use of surface water for swimming and boating;(3)

losses to river and bay recreational fishing and hunting of waterfowl;(4)

avoidance of game products;(5)

consumer avoidance of river, bay,

(6)

economic losses to watermen, onshore support activities, seafood processors, restaurants,
and retail and service outlets associated with Bay products and activities (bait and tackle
stores, motels, charter boat captains, etc.).

and ocean shellfish and finfish; andPotentially, the most consequential economic impacts of the Susquehanna River disposal options
for processed water can result if widespread consumer avoidance of Chesapeake Bay resources
occurs.*Consumer avoidance would be related to a number of factors including general attitudes
toward radiation.
There is some evidence to suggest that the public perceives nuclear power tobe more dangerous than many other technologies or activities including alcoholic beverages,
general aviation, motor vehicles, handguns, and smoking.
This belief is said to be derived fromthe expectation that nuclear power activities can produce an enormous number of deaths and threaten
the survival of the human race (pp. 25-34 of Reference 35).

*"Consumer" in this context refers not only to customers of retail establishments but also
to to watermen, wholesalers, hunters.

7-80
A second factor which would underlie consumer avoidance behavior is the confusion and distrust ina segment of the public that believes the risk of nuclear power to be considerably greater thanthe experts' assessments indicate (p. 15 of Ref. 35). A survey of residents within 20 miles of
TMI to determine levels of trust in authorities indicated a leaning toward distrust. 3 6 Although
the tendency toward distrust indicated in April 1979 had declined by August 1979, the tendency
was still for opinions to lean on the average toward distrust and appears to have been above thenational average. Although data are not available for a larger area around TMI, the staff has
assumed that the population, particularly those downriver of TMI who might be indirectly impacted
by the disposal of processed water in the Susquehanna, would share in levels of distrust characteristic of those living within 20 miles of TMI.
The potential for consumer avoidance and economic loss would also be related to two characteristics of seafood and the seafood industry. As indicated in Section 3.6.1.3, shellfish represents
approximately 90% of the landed value from the Chesapeake fishery. Although accurate data are
not available, a substantial portion of the total shellfish harvest is believed to be consumed inrestaurants. Those who have a long familiarity with seafood marketing consider fresh shellfish a"luxury" item of food consumption for which substitutions are easily made. 3 7 Secondly, theindustry is subjected to competitive pressures not only from fresh products originating in Gulf
of Mexico and other waters but also from frozen, processed products. Those familiar with the
Chesapeake seafood industry
believe that sales losses are difficult, if not impossible, to recap37
ture from competition.
Finally, historical evidence strongly points to a link between water quality and consumer demand
for fish and shellfish.
Consumer scepticism concerning the safety of seafood which is served
uncooked rises with outbreaks of hepatitis following consumption of these products. 3 8 Sportfisheries, in particular, are adversely affected by highly publicized examples of "contaminated"waters; such examples include mercury and PCB incidents in Japan; Kepone in the James River andlower Chesapeake Bay, Virginia; PCB and DOT in the Hudson River, New York and throughout theGreat Lakes; Mirex in Lake Ontario, New York; mercury in Lake St. Claire, Lake Erie and mainstream
TVA lakes 3 7 and natural phenomena such as red tides.
In the months following the accident at
TMI, local fishing activity in the vicinity of the plant declined in response to actual andbelieved changes in water quality levels (see Section 3.6.2.9).
Several experts in the field of marine economics and marketing suggested that the media plays animportant role in determining the public's perception of water quality changes and, ultimately,
consumer behavior. 40 If the media lacks adequate knowledge and context within which to evaluate
misleading statements, such statements could be passed on to the public as fact. The end results
are the generation of fears and consumer avoidance of seafood products.
A concern for media
reporting and its relationship to general public fears underlies the concern of the State of
Maryland.41
The most recent and well-publicized incident of water contamination involves the insecticide
kepone.
During the 1970s, traces of kepone in amounts exceeding Federal limits were found inblue crabs, oysters, bluefish, and in several other species of finfish in the James, York, andRappahannock Rivers in Virginia.
Rather than risk the distribution of contaminated products,
Federal and State officials closed the James River south of Richmond and adjacent portions of thelower Chesapeake Bay starting in December 1979 to the commercial harvesting of all seafood;recreational fishing was only permittted on a catch-and-release basis. 40 This ban was lifted formost species in November 1980.
Prior to and following fishing restrictions, numerous newspaper
articles emphasized the possibility of seafood contamination.41 Efforts were made by Maryland
officials to publicly draw a distinction between Maryland seafood
and fish--which contained42levels of kepone below Federal limits--and Virginia products.
Although the direct and indirect economic effects of kepone contamination apparently have notbeen systematically analyzed, several sources of information indicate widespread impact.
In
Virginia, consumers refused to buy bluefish, restaurants refused to serve them, and fishermen had
no market for the fish.
Purchases of bluefish were made only when the restaurant or retailer
could assure his customer that the fish was caught in the Atlantic Ocean or in North Carolina
waters. 4 3 As a result of incomplete information and extensive news coverage, the sale of oysters
harvested from areas which were completely isolated from James River was adversely affected in
Baltimore's wholesale fish market, some 200 miles north of the James River.* During 1976, sport*Researchers (see Ref. 44) calculated a 5% loss of oyster sales in Baltimore during a period whenmedia attention was focused on kepone contamination in Virginia. This loss occurred in a market
200 miles from the James River and to a product which is unrelated to the James River oyster.

7-81
fishery charter boat captains from the lower Maryland portion of the Chesapeake Bay, including
Chesapeake Beach (100 miles north of the James River), reported many trip cancellations due toangler concern with contaminated bluefish. 45 Although trip cancellations among charter boats
based around the Annapolis area were minimal, a small number of captains operating off the New
Jersey coast implicated kepone
contamination of fish as the reason that some fishermen were46
reluctant to charter trips.

The influence of contamination and media coverage on recreation has been demonstrated in a study
which has relevance for waterfowl hunting in the Chesapeake Bay. 4 7 In examining the effects ofmercury contamination on pheasant hunting, the researchers concluded that knowledge of the presence
and dangers of mercury in pheasants helped to explain much of the loss in Oregon pheasant hunters
during a two-year period.
Although the staff is unable to specify the quantitative or temporal extent of potential economic
losses to Chesapeake Bay activities as a result of disposing tritiated water in the SusquehannaRiver, three conclusions appear to be justified at this time. First, changes in the perceived
quality of water, whether or not such changes represent an actual health hazard to either thefauna or human consumers, is linked to altered patterns of consumption, most notably consumer
avoidance.
Second, the accuracy, tone, and quantity of media coverage will exert a powerful
influence on how the public perceives changes in water quality and, ultimately, on human behavior
and consumption patterns. Finally, because imperfect or limited information creates uncertainty,
the demand for products harvested from areas not affected, even remotely, could also decline;that is, consumer avoidance could impact the demand for fish and waterfowl from areas abutting or
beyond the Chesapeake region.
The State of Maryland has informed the NRC that it will be undertaking a comprehensive study ofpotential economic losses to Chesapeake Bay activities which would result from the alternative ofdisposing of TMI-2 processed water by dilution and controlled release to the Susquehanna River
which is expected to be completed in one year. The NRC staff is of the opinion that at least
until such a study is completed and the State of Maryland provides the results to the public and
the NRC staff, no decision should be made regarding disposition of the TMI-2 processed accident
water. This assumes no emergency arises which would require disposition of this water in theinterim. The NRC staff believes any such disposal actions taken in advance of completing thisstudy would be premature and without the benefit of information on societal and socioeconomic
factors important to the decision-making process.* However, processing of the water to remove
and immobilize the hazardous radionuclides should, in the staff's opinion, proceed as expeditiously as safety permits.
No options for disposition of the processed water (as well as theseparated radionuclides) are foreclosed by the expeditious treatment of the highly radioactive
accident water in the reactor building sump and reactor coolant system.
As discussed elsewhere in this chapter, adequate onsite storage capacity already exists.
In
consideration of this, the licensee has recently announced that no proposal for processed waterdisposal will be made to the NRC prior to 1982.
7.2.6

Liquid Disposal Costs

Table 7.42 provides the estimated liquid disposal costs for the alternatives presented.
Thelong-term onsite storage alternatives include the costs for up to 25 years of storage. Ultimate
disposal after the 25-year storage period have not been estimated.
The costs for the offsite disposal alternatives include transportation costs to the final disposal
site.

*The basic regulatory criteria for discharge of radioactive liquids is provided in 10 CFR Part 20.
Specifically, paragraph 20.1.c states that licensees should maintain releases of radioactive
materials to "as low as is reasonably achievable" (ALARA) levels and further defines the term
ALRA as meaning " ... .as low as reasonably achievable taking into account the state of technology
and the economics of improvements in relation to benefits to the public health and safety andother societal and socioeconomic considerations, and in relation to the utilization of atomic
energy in the public interest."

7-82

Table 7.42.

Disposal Costs for TMI-2 Liquids(thousands of dollars)

Disposal Alternative Considered
1. Long-term onsite storagea(a) Stored in tanks(b) Immobilized as concrete slabs
2.
Discharge to the environs(a) River discharge(b) Evaporation ponds(c) Forced evaporation
3.

Offsite disposal(a) Immobilized and shippedb(b) Immobilized and shippedc(c)4.

Deep well injection(d) Ocean disposal
Onsite disposal(a)

SLB trenches(b) Underground injection
aAssumes 25-year storage.

High

Low
5,600

2,800

2,300

150

100
500
250

7,400
5,600
5,100

5,400
4,100
3,700

10,500

7,700

1,900

1,400
250

bRichland, Washington, destination.
CWest Valley, New York, destination.

While a range of cost estimates is presented, the range is intended to include the uncertainty inthe actual volume of processed water to be disposed of. Where a single cost value is presented,
the costs have been considered to be relatively insensitive to the actual volume.
7.2,7Comparison of Processed Water Disposal Alternatives

The alternatives discussed above were compared on a relative basis considering time to complete
water disposal, radionuclide release pathways, potential regulatory obstacles, offsite doses andcosts. Table 7.43 summarizes these parameters.
This table numbers alternatives from 1 to 10 andthese designations are referred to in the discussions presented below. The conclusions that canbe drawn from this summary are as follows:. Time to Complete Water Disposal

To compare onsite and offsite alternatives, time to complete is the length of time required
to either (1) reduce concentrations to innocuous levels while onsite or (2) perform the work
necessary to either discharge or transport processed water from the site, including any
necessary construction work required to implement the alternative.
As discussed above, alternatives 1 and 2 are not practicable due to the long time periods
involved. Four of the other alternatives (3, 4, 5, and 6) could take 5 years to completedue to regulatory obstacles and, thus, may not be suitable.
Three of the remaining four alternatives (7, 8 and 10) can be readily completed while thefourth (alternative 9) is completed over a period of several years.

Table 7.43.

Comparison of Alternatives for Disposal of Processed Water
Potential Regulatory
Obstacle

Release Pathways
Years toComplete

To AtmosphereLong-Term Onsite
Storage
1. In liquid tanksa

200+

2. As concrete slabs

200+

Disposal
Alternatives

To
River

To
Land

To Subsurface Water

To
Ocean

NRC
Licensing

EPA
Permitting

Permanent

Offsitc

State/
Local

Disposition

Doses
Person-rem

Costc($103)

NA

5600

No

3 0e

2300

- No

+

NA

1400

Yes

+

Onsite Disposal
3. SLB trenches

4. Underground

5b

+

+

b
5

+

+

+

+

NA

250

Yes

+

+

+
+

+

NA
NA

3700
7700

Yes
Yes

+

NA

4100

Yes,

+
< 1
+
+
1
+
< 1
+
10. Forced evaporation
aAfter storage alternatives 2 through 10 are applicable.
bBased on potential licensing and permitting delays.
CBased on the low cost values in Table 7.42.
dBased on the SDS/EPICOR II process effluent.
eBased on loss of all tritium in the concrete slab.

30

100

Yes

30

500

Yes

30

250

Yes

injection

Offsite Disposal
5. Deep well injection
6. Ocean disposal

7. SLB facility
Discharge toEnvirons
8. Release to river
9.

Natural evaporation

5b
5b

1

7-84
Release Pathways
All alternatives lead to releases to the environs and-the major difference among them is thepathway(s) through which radionuclides are released. As shown in Table 7.43, alternatives(3, 4 5 6, 7 and 8) have a single pathway, while others (2, 9 and 10) have multiple pathways.
The onsite concrete slab alternative (2) releases tritium to the atmosphere by diffusion
through the concrete for several years after placement in the storage areas. The slabs arelater placed in a shallow land burial facility. The forced evaporation alternative (10)simultaneously releases radionuclides to the atmosphere and the river, while the naturalevaporation alternative (9) releases tritium to the atmosphere.
The impacts of these
releases depend on the pathway and the time period over which the release is made.
Potential Regulatory Obstacles
As discussed previously, the onsite disposal alternatives (3 and 4) are not considered
suitable because they would convert TMI-2 into a permanent waste repository and present
unresolved regulatory issues. The offsite bulk liquid alternatives (5 and 6) could present
different regulatory obstacles since new fields of regulation would be involved. It may not
be practicable to proceed with these alternatives due to these regulatory obstacles.
The other alternatives (1,regulations.

2, 7, 8, 9,

and 10) can be implemented in the framework of existing

Offsite Doses
Offsite doses, based on the SDS/EPICOR II process effluent, were only determined for alternatives which affect the population within 50 miles of the TMI-2 site. Therefore, the doses
arising from offsite disposal alternatives were not determined and are not shown on thetable. However, if the processed water is transported to another location and either evaporated or discharged to a river, the doses from these operations would be comparable to thoseshown for similar onsite activities.
Cost
The costs shown on Table 7.42 include the cost of any new facilities that are needed and theoperating cost to perform the work required to complete' the alternative. None of the costs
shown include the cost of licensing for the various alternatives.

References--Section 7
1.

Settlement Agreement in the matter of City of Lancaster vs. United States Nuclear Regulatory
Commission between City of Lancaster, City of Lancaster Authority, Albert B. Wohlsen, Jr.
the United States Nuclear Regulatory Commission, and the Metropolitan Edison Company, Jersey
Central Power and Light Company, and Pennsylvania Electric Company, Civil Action No. 79-1368,
U.S. District Court, District of Columbia, dated February 27, 1980.

2.

Memorandum from W.K. Lehto to J.A. Opelka, Argonne National Laboratory, Subject: Compilationof Information and Data from Visit to TMI-2 on May 6, 1980, dated May 12, 1980.

3.

Letter (with enclosures entitled, 1. Radionuclide Distribution of Auxiliary Building Water,
2. Agenda for December 18, 1980 Meeting of Task Group SC-38) from B.J. Snyder, U.S. Nuclear
Regulatory Commission, to F.P. Parker,Vanderbuilt University, December 15, 1980.

4.

R.E. Brooksbank and W.J. Armento, "Post Accident Cleanup of Radioactivity at the Three Mile
Island Nuclear Power Station," Oak Ridge National Laboratory, ORNL/TM-7091, February 1980.

5.

D.O. Campbell, "Hot Cell Studies," Oak Ridge National Laboratory, presentation to General
Public Utilities-and U.S. Department of Energy staff members, January 31, 1980.

7-85
6.

"TMI - Data Base," released by Argonne National Laboratory - W.K Lehto, November 21, 1980.

7.

Memorandum from R.J. McConnell, Argonne National Laboratory, to J. Opelka, M. Chasanov,
C. Johnson, P. Tuite, "Additional Revisions to Section 7.0," December 19, 1980.

8.

Letter (TLL-160), from G. Hovey, GPU, to J. Collins,"Submerged Demineralizer System," April 10 1980.

9.

"Data Handout - TMI-2," as available to the TMI Working Group meeting held at TMI, Middletown PA, September 23, 1980,

10.

Letter from R.E. Brooksbank, Oak Ridge National Laboratory, to S.W. Ahrends, U.S. Departmentof Energy, "Survey of the Stability of Zeolites in Plant-Scale Application," December 12, 1980.

11.

D.O. Campblell, E.D. Collins, L.J. King, and J.B. Knaver, "Evaluation of Submerged Demineralizer System (SDS) Flowsheet for Decontamination of High Activity-Level Water at the Three
Mile Island Unit 2 Nuclear Power Station," Oak Ridge National Laboratory, ORNL/TM-7448,
July 1980.

12.

"Future EPICOR II Operation," Letter from G.K. Hovey, GPU,sion TMI Program Office, January 13, 1981.

13.

H. Lowenberg, U.S. Nuclear Regulatory Commission, "Trip Report for December 16-18, 1980;"
December 1980.
H.W. Goodbee, "Use of Evaporation for the Treatment of Liquids in the Nuclear Industry," Oak
Ridge National Laboratory, ORNL-4790, September 1973..

14.

U.S. Nuclear Regulatory Commission,

to U.S. Nuclear Regulatory Commis-

15.

U. Yang, "Radwaste Evaporator Improvement," Power Engineering, p.

16.

"State of the Art Review of Radioactive Waste Volume Reduction Techniques for Commercial
Nuclear Power Plants," Oak Ridge National Laboratory, ORNL/Sub-79/13837/2, April 1980.

17.

G. Lefillatre, "Progress on the Techniques of Bituminizing Liquid Effluents of Pressurized
Water Nuclear Power Plants," Battelle Northwest Laboratory, BNWL-TR-196, August 1976.

18.

J.E. Stewart, "Volume Reduction and Solidification of Low-Level Radioactive Wastes," PowerEngineering, p. 64, February 1980.

19.

W.B. Seefeldt et al., "Characterization of Particulate Plutonium Released in Fuel Cycle
Operations," Argonne National Laboratory, ANL-75-78, May 1976.

20.

Note from R. Weller, to 0. Lynch, U.S. Nuclear Regulatory Commission, "Decontamination
Factors for High Efficiency Particulate Air (HEPA) Filters in Ventilation Exhaust Systems at
TMI-2," January 12, 1981.

21.

Letter from J.F. Ahearn, U.S. Nuclear Regulatory Commission, to H. Dieckamp,
Utilities Corporation, January 12, 1981.

22.

J.R. Shubel, "Suspended Sediment Discharge of the Susquehanna River at Conowingo Maryland
During 1969," Chesapeake Science 13:53-58, March 1972.

23.

A. Eaton, V. Grant, M. G. Gross, "Chemical Tracers for Particle Transport on the Chesapeake
Bay," Estuarine and Coastal Marine Science, 10 - 75-83, 1980.

24.

R. McLean, Appendix C to letter to Dr. B. Snyder, Program Director, TMI Program Office from
J. Long dated November 20, 1980 (Included in Comments, Appendix A).

25.

S.M. Jinks and M.E. Wrenn, Radiocesium Transport in the Hudson River Estuary," Ch. II in:
Environmental Toxicity of Aquatic Radionuclides - Models and Mechanisms,
M.W. Miller and
J.N. Stannard (eds.), Ann Arbor Science Publisher Inc., Ann Arbor, Michigan 1975.

26.

M.G. Wolman, C.D. Meyers, C.R. Flynn and D.P. Sheer, "Use of the Chesapeake Bay for Cooling
Thermal Electric Power Plants, Parts I and 2," Atomic Energy Commission, NYO 4260-1, NYO
4260-2, prepared by the Department of Geography and Environmental Engineering, Johns Hopkins
University Baltimore, MD, May 1972.

90, April 1980.

General Public

7-86
27.

"Final Environmental Statement Related to the Proposed Summit Power Station Units 1 and 2,
U.S. Atomic Energy Commission, July 1974.

28.

R.K. Johnson and T.S.Y. Koo, "Production and Distribution of Striped Bass (Morone saxatilis)
Eggs in the Chesapeake and Delaware Canal," Chesapeake Science 16 (1):39-55, 1975.

29.

L.E. Cronin, D.W. Pritchard, T.S.Y. Koo, and V. Lotrich, "Effects of Enlargement of the
Chesapeake and Delaware Canal," pp. 18-32, in: Estuarine Processes, Vol. II, Academic
Press, New York, 1977.

30.

Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet theCriterion As Low As Practicable for Radioactive Material in Light-Water-Cooled Nuclear Power
Reactor Effluents, WASH-1258," Vol. 2, pp. F-35, thru F-40, U.S. Atomic Energy Commission,
July 1973.

31.

Letter from S. M. Long to Dr. Bernard Snyder, Program Director, TMI Program Office,U.S. Nuclear Regulatory Commission, November 20, 1980 (included in comments, Appendix A).

32.

"Liquid Pathway Generic Study, NUREG-0440,"
February 1978.

33.

B. G. Blaylock and J. P. Witherspoon "Radiation Doses and Estimations for Aquatic Biota
Exposed to Radioactive Releases from LWR Fuel-Cycle Facilities," Nuclear Safety, Vol. 17,
No. 3, May-June 1976.

34.

"The Effect on Populations of Exposure to Low Levels of Ionizing Radiation," National Academyof Sciences, Committee on the Biological Effects of Ionizing Radiation, 1980 Report.

35.

P. Slovic, S. Lichtenstein, and B. Bishhoff, "Images of Disaster:
Perception and Acceptanceof Risks from Nuclear Power," Electric Perspectives, Vol. 79, No. 3, pp. 8-20; andC. Hohenemser, R. Kasperson, and R. Kates, "Distrust of Nuclear Power," Science, Vol. 196,
pp. 25-34, April 1977.

36.

B.P. Dohrenwend et al., "Technical Staff Analysis Report on Behavioral Effects to President's
Commission on the Accident at Three Mile Island," October 1979.

37.

Interviews with W.R. Prier, Office of Seafood Marketing, Maryland Department of Economic andCommunity Development, on October 30, 1980; and T.J. Billy, National Marine Fisheries Service,
U.S. Department of Commerce, on November 5, 1980.

38.

D.L. Long and C.W. Coale, Jr., "Seafood Consumption Patterns," Research Report MB 182,
Department of Agricultural Economics, Virginia Polytechnic Institute and State University,
Blacksburg, Virginia, March 1972.

39.

"Impact of Contaminants on Angling," The Sport Fishing Institute, SFI Bulletin No. 289,
October 1977;,"Impacts of Aquatic Contaminants on Angling," SFI Bulletin No. 293, April 1978,
Washington, DC; and J. Kauffman, "Effect of a Mercury-Induced Consumption Ban on Angling
Pressure:
South Fork of the Shenandoah River, Virginia," Fisheries Vol. 5, No. 1, JanuaryFebruary 1980.
"California Anchovy Reduction Fishery," The Sport Fishing Institute, SFI Bulletin, No. 280,

40.

U.S. Nuclear Regulatory Commission, Section 5.1,

November-December 1976, and SF1 Bulletin No. 293, April 1978, Washington, D.C.
41.

J. Beck, "Alarm Grows Over Kepone in Bluefish," Washington Star, July 27, 1976.

42.

"Bluefish Reflect Kepone Presence," Queen Anne's Record Observer, July 28,
B. Peterson, "Md. Bluefish Safer: Mandel," Washington Post, July 30, 1976.

43.

J. Frye,44.

D.G. Swartz and I.E. Strand, Jr., "Avoidance Costs Associated with Imperfect Information:
The Case of Kepone," Scientific Article No. A2817, Contribution No. 5867, Maryland Agricultural Experiment Station.

"Virginia Industry Likely Victim of Kepone,"

1976,

and

National Fisherman, October 1976.

7-87
45.

H. J. Speir, D. R. Weinrich, R. S. Early, "1976 Maryland Chesapeake Bay Sport Fishery Survey,"
Maryland Department of Natural Resources, Fisheries Administration, June 1977, pp. 26 and29; and G. Fisbee, "Bay's Skippers Hit Publicity About Kepone," The Maryland Gazette,
August 26, 1976.

46.

R.L. Swanson and G.J. Sindermann (eds.), "Oxygen Depletion and Associated Benthic Mortalities in New York Bight, 1976," U.S. Department of Commerce, National Oceanic and Atmospheric
Administration, NOAA Professional Paper II, Chapter 14. Socioeconomic Impacts, Rockville,
MD, December 1979.

47.

R.N. Shulstad and H.H. Stoevener, "The Effects of Mercury Contamination in Pheasants on the
Value of Pheasant Hunting in Oregon," Land Economics, Vol. 54, No. 1, February 1978.

8.

SOLID WASTE PACKAGING AND HANDLINGThe activities described in Sections 5, 6, and 7 result in the generation of radioactive materials that must be packaged and disposed of as radioactive waste.
Most of this radioactive
material will be low-activity radioactive waste that can be disposed of at licensed shallow land
burial facilities used for disposal of wastes from other nuclear power plants. Some of thesematerials may be transuranic (TRU) radioactive wastes, and others will have characteristics, suchas high specific activity, that warrant their handling and disposal as high-level waste on a
special case-by-case basis.
In the staff's judgment, the TRU and high-specific-activity wastes
will need to be transported to federal government (U.S. Department of Energy) facilities fortreatment and disposal.
In this section, wastes are discussed in terms of their characteristics,
the alternatives considered for their packaging and handling, and the impact of these operations.
The wastes that could be generated have been divided into four major categories that are related
to their source or physical form:
Process Solids - The radioactive wastes generated from treatment of liquid wastes. These materials consist of accident sludge, spent filter cartridges, expended ion-exchange media in the formof zeolites* and organic resins, and evaporator bottoms. Bituminized materials generated fromonce-through treatment of liquid wastes also are included in this category.
Chemical Decontamination Solutions - The liquid radioactive wastes generated from "hands-on"
decontamination activities and decontamination of systems that have not been treated to generate
process solids.
Solid Materials - The dry radioactive waste generated during decontamination and defueling operations.
These materials consist of trash, contaminated equipment and material, and irradiated,
activated hardware.
Damaged Fuel Assemblies and Core Debris - The radioactive materials removed from the reactor
vessel during defueling operations.
The waste types within each of these categories are discussed separately in Sections 8.1
through 8.4.
8.1

PROCESS SOLID WASTES

Packaging and handling of contaminated sludges generated by the accident and of process-solid
wastes that could arise from treatment of accident water and decontamination liquids are discussed
in this section. These solid wastes would include accident sludges and, depending on the liquid
processing alternatives implemented (see Sec. 7.1), filter cartridge assemblies, ion-exchange
media (inorganic zeolites and organic resins), evaporator bottoms, and bituminized materials.
The waste forms that could be generated from treatment of the various liquid waste sources arelisted in Table 8.1 and discussed in detail below.
8.1.1

Status and Specific Considerations

The status of the process solid wastes generated through September 22, 1980, and the practices
used to package, handle, and store these wastes are summarized in this subsection. The amountsof the various types of such wastes that could be generated by processing of liquid wastes anddecontamination of various systems are estimated.
*Other inorganic ion-exchange media may be used with or in place of zeolites.

8-1

Table 8.1.

Process Solid Waste Summary

Curie Inventory an
Source of TreatedLiquid Waste

Untreated Liquid
Minimum
Maximum

1.

AFHB accident water

55,000d2.

Reactor building sumpwater

500,000

500,000

3.

RCS water

20,00020,000

4.

RCS flush & drain water

20,000100,000

5.

AFHB/reactor building
decontamination solutions

6.

Process Solid Waste Formsb'c
Filters

Zeolites

Organic
Resins

X

Xe

Xe

X

X

X

X

X

X

X

X

X

55 , 0 0 0 d(a)Aqueous

90

90

(b)

Chemical

70

70

X

Evaporator
Bottoms

BitumenX
X

X

RCS decontamination
solutions

Total

Sludge70
N!

(a)Aqueousf

2,000

20,000(b)

Chemicalf

2,000

20,000600,000

700,000

X

X

X

aExclusive of H-3 and noble gases--rounded to two significant figures.
bWaste form combinations are alternative-dependent--see

Section 7.1.3.

CX indicates process solid waste form could be generated and is considered.
dCuries removed by system through September 22,

X

1980.

eSome liners contain zeolites mixed with organic resins.
fMutually exclusive alternatives; one waste form will be produced,

not both.

X

8-3
8.1.1.1

Efforts to Date

The work performed through September 22, 1980, involved processing of AFHB accident water with
the EPICOR II system and resulted in the generation of expended ion-exchange media.
The characteristics of these media are summarized in Table 8.2.
All the liners* listed in the table arebeing stored onsite in the EPICOR II storage facility (see Appendix Q for description).
In
compliance with the Commission's order of October 1979 to solidify these wastes prior to shipment, the licensee has been investigating the use of cement.'
In addition,
the licensee has been2
investigating alternatives to solidification of these wastes onsite.
As contaminated liquids are passed through the EPICOR II system and the ion-exchange media become
loaded with radioactive contaminants removed from the liquids, the expended media are replacedwith new liners.
The steps involved in the removal of expended liners from the EPICOR II systemand their subsequent handling are as follows:
Disconnecting the connections between the resin liners and the EPICOR II system
Capping liner inlet, outlet and vent connections
Transfer of the expended resin liner to a bottom-loaded, bell-type transfer cask
mounted on an overhead monorail (mixed-bed third-stage liners are handled unshielded)
Placement of the transfer cask liner into a top-loaded, reinforced-concrete shield
mounted on a flatbed truck
Transport (within the site boundaries) of the doubly shielded liner to the onsite
storage facility
Removal of the bell transfer cask from the truck-mounted shield with a crane
Movement of the bell transfer cask to a position above the storage module
Placement of the liner within a shielded storage module
Closure of the storage module with a concrete shield plug.
The radiation exposure of personnel
been about 6 person-rem.
8.1.1.2

3

involved in transfer of these liners to the storage area hasThis corresponds to about 85 person-mrem per liner transferred.

Projected Requirements

The different types of process solid wastes that could be generated and the curies of radioactivity that would be removed from various liquid waste streams and deposited on these solids
were listed in Table 8.1.
To estimate the quantities of process solid wastes that could be generated,
best- and worst-case conditions that bound waste quantities as follows:

the staff defined-Best-Case Conditions.
Liquid wastes are assumed to be treated using the methods thatgenerate minimum waste volumes.
Where specific activity or curie content per unit of
the process solid waste type could be variable, the maximum curie content is assumed.
(The greater the amount of radioactive contaminants allowed to be accumulated on asingle batch of resins, for example, the smaller the total volume of resins that will
be required to remove a given amount of contaminants from the liquids.)
These minimum
volume projections for process solid waste generation are shown in Table 8.3.

-

Worst-Case Conditions.
Liquids are assumed to be treated using the methods that generate maximum waste volumes.
Where specific activity or curie content per unit of theprocess solid waste type could be variable, the minimum curie content was assumed.
(The smaller the amount of radioactive contaminants allowed to accumulate on a single
batch of resins, for example, the greater the total volume of resins that will be
required to remove a given amount of contaminants from the liquids.)
These maximum
volume projections for process solid waste generation are shown in Table 8.4.

The process solid waste volumes shown in Tables 8.3 and 8.4 were used to evaluate alternatives
and characterize process solid wastes throughout this section.

*The "liners" are the containers housing the ion-exchange material.

8-4
Table 8.2.

Package

Total
Ion-Exchange
Media Volume(ft 3 )

Summary of EPICOR II

LinerbSize

Numberof Liners

Waste Packagesa

Range of Curie
Loadings per Liner

Prefilter liner(30 ft 3 /liner)
1470
4'D x 4'H
49
100 to 1 3 0 0 d
Second-stage liner
1 to 60f
4'D x 4'H
14
(30 ft 3 /liner)
420
Third-stage liner
1 to 6
6'D x 6'H
6(130 ft /liner)
780
aThese media are currently being stored in the dewatered condition.
bLiners are carbon steel.

Maximum Liner
Radiation Level(R/hr)
1500
50
5

cMay be mixture of organic and inorganic.
d30 liners are in 1200 to 1300 Ci range, 14 liners are in the 800 to 1200 Ci range, 2 liners
are in the 400 to 800 Ci range, 3 liners are in the 100 to 200 Ci range. These loadings do
not include the daughter products of Cs and Sr which would increase the total Ci's by 50%.
eMaximum hot-spot radiation level on liners in 1200 to 1300 Ci range. Average radiationlevels
in R/hr are about 0.8 times curie content.
fFive liners are in the 10 to 60 Ci range, remainder are below 10 Ci.
Source: "Data Handout-TMI-2," as available to the TMI Working Group Meeting held at TMI,
Middletown, PA, September 23, 1980.

8.1.2

Alternative Methods Considered

The techniques used to treat, package, and handle each of the various types of process solid
waste materials that could be generated depend on the physical characteristics of the waste form
and its specific activity or curie content.
8.1.2.1

Accident Sludge

The volume of accident sludge that could be generated from cleanup of the AFHB and reactor building is estimated to be about 250 ft 3 . The estimated specific activity of this sludge could range
from 1.6 Ci/ft 3 in the reactor building to 60 Ci/ft 3 in the AFHB. 4 Therefore, the packaging andhandling operations for sludge will require the use of remote techniques.
Alternatives considered for sludge are dewatering or immobilization with cement or with vinyl
ester styrene.
In each case, the sludge could be packaged in 55-gallon drums or large liners.
The maximum waste package production would involve immobilization with cement or vinyl ester
styrene and packaging in 55-gallon drums.
The minimum waste package production would involve
packaging dewatered sludge in 4-ft-diameter by 4-ft-high liners with a capacity of 45 ft 3 . The
quantities, container characteristics, and maximum radiation levels of packaged sludge underthese maximum and minimum conditions are given in Table 8.5. To date, no proposal.has been madeby the licensee on whether to dewater or immobilize the sludge.
The high specific activity of the AFHB sludge will lead to container radiation levels up to2500 R/hr. Since these levels are in excess of those experienced to date with EPICOR II prefilter liners of the same size, they could present handling and personnel radiation exposure problems. Therefore, drums are the preferred container for AFHB sludge packaging.

Table 8.3.

Source of TreatedLiquid Waste
1.

Minimum Generation of Process Solid Waste

Treatment b
Alternative

Process Solid Waste Forms and Unpackaged Volume (ft3)a
Organic
Evaporator
Bottoms
BitumenSludgec
Filters
ZeolitesResinsEPICOR II system

NA

NAd

2600d

NA

NA

150

2.

AFHB accident water
Reactor building water

SDS/Modified SDS

110

50e

200

NA

NA

100

3.

RCS water

SDS/Modified SDS

NA

20

200

NA

NA

NA

4.

RCS flush and drain waterf

SDS/Modified SS

360

20

200

NA

NA

NA

5.

AFHB/reactor building decontamination solutions(Chemical)

Evaporator/Resin

NA

NA

100

200

NA

NA

6.

RCS decontamination
solutions (Aqueous)gCo

CAN DECON/SDS Type

20

10

100

NA

NA

avolumes rounded to nearest 10 ft 3 .
bAlternatives discussed in Section 7.1 and defined in Appendix G.
CSludge generation is independent of treatment alternative.
dMay be mixture of organic and inorganic.
eVolume shown based on specific activity up to 15,000 Ci/ft 3 , themaximum loading considered for zeolites
used in reactor building accident water treatment.
fBased on removal of 20,000 Ci.
gBased on removal of 2000 Ci.

NA1!

Table 8.4.

Maximum Generation of Process Solid Waste

Process Solid Waste Forms and Unpackaged Volume (ft3)a
Source of TreatedLiquid Waste

Treatment b
Alternative

Filtersc
Zeolites

Organic
ResinsEvaporator
Bottoms

BitumenSludged1.

AFHB accident water

EPICOR II

NA

NAe

2 600e

NA

NA

150

2.

Reactor building sump water

SDS/EPICOR II

110

430

390

NA

NA

100

3.

RCS water

Modified EPICOR II

NA

50

540

NA

NA

NA

4.

RCS flush and drain waterf

Modified EPICOR II

450

100

1980

NA

NA

NA

5.

AFHB/reactor building
chemical decontamination
solutions

Bituminization

NA

NA

150

NA

50

NA

Evaporator/Resin

NA

NA

250

7000

NA

NA

6.

RCS chemical decontamination
solutionsg

aVolumes are rounded to nearest 10 ft 3 .
bAlternatives discussed in Section 7.1 and defined in Appendix G.
cVolume based on specific activity of up3

to 1250 Ci/ft , the minimum considered for zeolites.

dSludge quantities are independent of tre
atment alternatives.
eMay be mixture of organic and inorganic.
fBased on removal of 100,000 Ci.
gBased on removal of 20,000 Ci.

8-7
Table 8.5.

Accident Sludge Packages--Minimum and Maximum Waste Generation

Reactor Building
Sludge

AFHB Sludge
Factor

Minimum

Maximum

Minimum

Maximum

Original volume (ft 3 )Process used
Volume increase factor-Container

150
Dewater

150
Immobilize

1
50 ft3a

1.65
Drum

100
Dewater
1
50 ft3a100
Immobilize
1.65
Drum

Maximum Ci/package

2700

250

72

7

Maximum surface radiation level (R/hr)

2200

200

60

6

• 3

36

2,'
2

24

Number of packages
33

a50-ft liner with 45-ft capacity.
bBased on 0.8 R/hr per curie.

8.1.2.2

Zeolites

Zeolites are effective ion-exchange media for removal of cesium. For the alternative treatment
systems under consideration, zeolite ion-exchange vessels are used as the first stage in a multistage treatment system. Each vessel is made of stainless steel, has an internal volume of 10 ft 3
and is loaded with between 8 ft 3 and 10 ft 3 of zeolite. The amount of radioactivity that can beremoved by zeolites is extremely high; specific activities in excess of 15,000 Ci/ft 3 of zeolite
can be achieved without radiation damage to the ion-exchange material.
Thus, the limits on theradioactivity in the zeolite vessels are based on shielding requirements and remote handlingtechniques rather than on ion-exchange media stability. It is expected that these zeolite ionexchange vessels will be loaded with between 10,000 and 120,000 Ci each. The higher loadings
will minimize container handling and resultant exposure, as well as the number of shipments andvolume to be disposed.
Even with the lower loadings of radioactive materials (predominated by
30-year half-life radioisotopes), the zeolite containers cannot be handled or disposed of as
low-level radioactive waste. Present intentions are for interim storage of zeolite ion-exchange
vessels at the plant site followed by transportation to a federal government facility for treatment and eventual disposal. 5
Currently, the primary alternative for managing similar zeolites produced at federal government
facilities, which also could be applied to zeolites from TMI, is to comingle the material withhigh-level waste sludge being stored in tanks pending the availability of facilities to immobilize the waste.
Alternatively, the TMI-2 materials could be processed separately.
Eventual
immobilization in a borosilicate glass is now considered to be the preferred approach, althoughdevelopmental work on ceramics continues. After immobilization and packaging, the waste would be
stored until a geologic repository is developed.
Some developmental work has been done on elution of Cs-137 from zeolite resins, with subsequentstorage of the eluate for eventual immobilization.
In the event that waste forms containing
zeolites are unacceptable for disposal in a high-level-waste repository, elution of the activity
from zeolites may be required.
However, applicability of this procedure to specific conditionsof the TMI zeolite would have to be verified before this limited treatment option could beconsidered.

8-8
The volume of zeolites that could be generated depends on the treatment system selected and theamount of radioactivity placed on zeolites in each ion-exchange vessel.
The higher the radioactivity loading per vessel, the lower the volume of packaged waste. The quantities and characteristics of zeolites that could be generated under best- and worst-case conditions are summarized in Table 8.6 and the conditions of maximum and minimum waste production are outlined below:
Maximum Waste Production.

Systems using zeolite ion-exchange vessels are assumed to be

used for treatment of reactor building sump water, primary system water, reactor coolingsystem flush and drain liquid, and CAN DECON liquid waste. Each vessel would be loadedwith 10,000 Ci of radioactive contaminants or about 1250 Ci/ft 3, the minimum loading
desirable.
Minimum Waste Production.

Systems using zeolite ion-exchange vessels would be used fortreatment of reactor building sump liquid, primary system water, and RCS flush anddrain liquid.
Each vessel used for reactor building sump water would be loaded with
120,000 Ci, or about 15,000 Ci/ft 3 , the maximum loading considered.
Vessels used toprocess other water would be loaded with 10,000 Ci, or 1250 Ci/ft 3 .
Since no proposal has been made on the use of zeolites rather than other treatment methods, thequantities and characteristics shown in Table 8.6 bound the volumes of zeolite wastes that could
be generated under the stated assumptions.

Table 8.6.

Generation of Zeolite Waste--Minimum and Maximum Alternatives
Minimum Generation

Source of TreatedLiquid Waste

Maximum Generation

Volume (ft3)aCuriesb

Volume (ftW)aCuriesb

48

500,000

432

500,000

2.

Reactor cbuilding sumpwater
RCS waterd

48

20,00016

20,0003.

SDSe
Modified EPICOR II
RCS flush and draind

96

100,000

1.

SDSe
4.

Modified EPICOR II

16

20,000

RCS decontarination
solutions

8

2,000

16

20,00088

542,000

592

640,000

Total

aBased on gross curies removed and estimated zeolite loading per cubic foot.
bTotal estimated curie content in source. Zeolites will remove 98+ percent.
CRange based on minimum zeolite loading of 1250 Ci/ft 3 and maximum zeolite loading of
15,000 Ci/ft 3 .
dBased on zeolite loading of 1250 Ci/ft 3 .

eAll versions of zeolite-based systems considered for these liquids have the samezeolite loading of 1250 Ci/ft 3 .

fCAN DECON alternative only.

8-9
8.1.2.3

Organic Resins

Organic resins are used as the primary ion-exchange media in the EPICOR II system. These materials also are used in second- and third-stage ion-exchange vessels for some of the other alternative treatment methods being considered.
Generally, these second- and third-stage ion-exchange
vessels remove relatively small quantities of radionuclides, and the specific activity of theorganic resin wastes is well below that of first-stage ion-exchange vessels, such as the EPICOR II
prefilter liners and the SDS zeolite liners.
High-Specific-Activity Organic Resins
The quantities and characteristics of the organic resins generated from EPICOR II system treatment of AFHB accident water are described in Table 8.2. As shown in that table, first-stage
prefilter liners contain about 30 ft 3 of organic resin with a specific activity of 10 to
40 Ci/ft 3 . This activity level, coupled with the cesium and strontium content of these ionexchange materials, precludes their being handled and disposed of as routine low-level radioactive waste.
Forty-six of these 49 prefilter liners fall into this category of high-specificactivity waste.
The characteristics of the 49 prefilter liners and of additional prefilter
liners that could be generated if the EPICOR II system is used to treat primary system water are
given in Table 8.7. The EPICOR II resins currently are being stored in carbon steel vessels inthe dewatered condition in shielded modules at the plant.
As indicated in Section 7.1, theEPICOR II system is being considered for treatment of RCS water and RCS flush and drain water.
Since the gross curie content of these liners would be comparable to that of liners generated bytreatment of AFHB liquids, the liners would be handled and disposed of the same way.

Table 8.7.

Generation of High-Specific-Activity Organic Resins-Minimum and Maximum Alternatives

Source of TreatedLiquid Waste
1.
2.

AFHB accident waterb
RCS water

3.

RCS flush and drain water
Total

Minimum Generation
Volume (ft 3 )Curiesa

Maximum Generation
Volume (ft 3 )Curiesa

1,380
540

54,50019,9001,380
540

54,50019,900540

19,9002,690

99,5002,460

94,000c

4,610

174,000c

aDetailed information on EPICOR II is proprietary. Curies were estimated from actual
performance with AFHB liquids extrapolated to other sources.
b4 6 high-specific-activity prefilter liners in storage as shown in footnote c on
Table 8.2.
CRounded to nearest thousand.

Because of the relatively high radiation levels, 500 R/hr to 1500 R/hr, there is a potential for
breakdown of the organic resin materials.
Similar organic resins have been shown to exhibit
instability with absorbed doses of 108 rads. Some of the EPICOR II prefilter liners have been instorage for 18 months and could remain in storage until waste handling options have been selected
and the appropriate facilities become available. These storage times could lead to accumulated
doses greater than 108 rads.
Under these conditions, the organic resins could begin to deteriorate. Gases could be generated and corrosive liquids could be formed in the liners. The licensee currently stores the EPICOR II prefilter liners in the interim waste storage facility modules(see Appendix Q), which are constructed of reinforced concrete and include provisions for sampling8-10
the module sump.
In the event the liners develop leaks, the storage module would prevent the
migration of the small quantities of liquids which could be present in the dewatered liners.
Thelicensee is planning to perform tests on liquids pumped from a representative prefilter liner toassess the corrosion potential of the liners.
The DOE will also perform an investigation of theresins and the corrosion effects of an actual liner to determine if resin degradation is occurringat the rates projected from small-scale laboratory testing.
In the event these investigations or
actual experi.ence confirm the concerns regarding liner integrity, the licensee will be required
to take appropriate corrective actions such as use of overpacks.
The staff requested the licensee
to develop contingency plans for possible corrective actions.
The licensee has informed the NRC
staff that future operations of EPICOR II will not use organic resins in the first-stage prefilter
liners. 6 ' 7 Only inorganic ion-exchange media will be used in these liners in the future.
Until these high-specific-activity wastes can be shipped offsite, they will be placed in theconcrete interim waste storage facility.
Because of their characteristics, these materials
cannot be disposed of using routine methods at low-level disposal facilities.
The most practicable alternative is to transfer these wastes to an existing federal government facility for
future processing and eventual disposal.
Other less desirable alternatives include processing
onsite and/or packaging in a high-integrity container for special disposal at commercial LLW
burial sites.
The offsite alternative treatment techniques at federal government facilities,
which will have to be demonstrated for specific applicability to the condition of the TMI resins,
include:Storage as a sludge in tanks at government facilities pending eventual
when facilities are available.

immobilization

Incineration of the combustible material to reduce volume and mass, followed by immobilization of the residue.
A number of incineration facilities being developed for use
with transuranic wastes might be capable of modification for incineration of highspecific-activity wastes.
These include:
Controlled air incinerator developmental units-

Cyclone incinerator-

Rotary-kiln production unit
Slagging pyrolysis incinerator.

-

Acid digestion of combustible waste.
This is a potential alternative for treating theorganic resins that would result in substantial volume reductions (>60).
Elution of selected radionuclides (Cs-137, Sr-90) from the resins onto more stable
inorganic ion-exchange media (zeolites).
The loaded inorganic ion-exchange media could
be stored for eventual immobilization.
It should be noted that although significant
development work has been done at federal government facilities on elution of organic
resins, the processes have been developed for highly alkaline systems and would require
adaptations for use on TMI resins.
This factor and the question of cost effectivenessfor such a process requires careful evaluation of the feasibility of this approach.
For each of the alternatives described/above, the DOE facilities and staff have experience withsimilar operations.
The various federal government facilities at which these treatment alternatives could be performed and the status of their availability are described in Section 9.1.3.3.
Because of the existence of these DOE facilities with experienced staffs, it is the NRC staff'sview that these are logical places (as opposed to the TMI site) for further processing and interim
storage of the higher specific activity materials, pending ultimate disposal.
Low-Activity Organic Resins
The organic resins that have been or would be generated from second- and third-stage ion-exchange
vessels have activity levels and contents that appear to be similar to organic resins from othernuclear facilities and could likely be handled and disposed of as low-level radioactive wastes
provided the isotopic concentrations are similar to other resins.
Some of these materials,
generated by the EPICOR II system, are already in storage at TMI.
Additional quantities could be
generated through the continued use of EPICOR II for AFHB water and use of the alternative treatment systems, such as the SDS and modified EPICOR II treatment of primary system water.
As

8-11
low-level radioactive wastes, these resins could be immobilized prior to offsite transport for
disposal.
Immobilization techniques are discussed in Appendix H, and both cement and vinyl ester
styrene binders were considered for immobilization of organic resins.
The quantities and characteristics of organic resins that could be generated depend on the treatment method(s) selected and the binder material used for immobilization.
To bound these quantities, best- and worst-case conditions were considered.
-Maximum Package Production.
Systems using evaporation followed by resin polishing areassumed to be used for treatment of reactor building sump liquids, and reactor coolant
system decontamination liquids.
The modified EPICOR II system is assumed to be used
for treatment of primary system water and reactor coolant system flush and drain liquids.
All resins would be immobilized.

-

Minimum Waste Production.
Systems using zeolites followed by organic resins polishin'g
are assumed to be used for treatment of reactor building sump water, primary systemwater, and flush and drain liquid.
All resins would be immobilized.

The quantities and characteristics of the resins already generated by EPICOR II and those that
could be generated under the assumed bounding conditions are summarized in Table 8.8.

Table 8.8.

Generation of Low-Activity Organic Resins-Minimum and Maximum Alternatives

Source of TreatedLiquid Waste

Minimum Generation

Maximum Generation

Volume (ft 3 )CiVolume (ft

3

)

Ci

1.

AFHB accident watera

1200

260

1200

260

2.

RB sump water
SDS/Modified SDS

200

75

-

-

SDS/EPICOR II

-

-

390

80

RCS accident water
SDS/Modified SDS

200

30

-

-

-

-

540

100

SDS/Modified SOS

200

60

Modified EPICOR II

-

-

1970

500

Water Based RCS Decontamination

140

140

5

3.

Modified EPICOR IIb

4.

5.

RCS flush and drainc

Total

1

1940

aEPICOR II system resins in storage.
bWaste volumes based on staff estimate.
CBest case removes 20,000 Ci; worst case removes 100,000 Ci.

4240

8-12
8.1.2.4

Filters

Most of the treatment alternatives being considered involve use of filters to remove particulates
upstream of the main processing system. These filters can be either the cartridge or the precoat
type.
The volume of waste represented by these filters would be small in comparison with otherwaste volumes generated by a treatment process. However, the radioactivity level of these filters would be relatively high in comparison with the radioactivity levels of second- and thirdstage organic resins, and they may also contain fuel particles and uranium and plutonium.
To date, most of the filter configurations considered have been of the cartridge type, and thestaff has based its characterization of these process solid wastes on use of this type of filter.
Three different size cartridges have been considered. Cartridge assembiles (6 inches diameter by
26 inches high) with an exterior volume of about 0.5 ft3 were considered for removal of fuel andcore debris from the reactor cooling system water during defueling. When systems include zeolites, a 10-ft 3 filter is used upstream of the zeolite vessel.
A limited number of relatively
large (96 inches diameter by 36 inches high) core filters could also be generated. The quantityof cartridge-type filters generated depends on the particulate content of the influent and theradioactivity level to which the filters can be loaded.
The estimated quantities of filter
cartridge assemblies that could be generated are presented in Table 8.9.
The smaller, 0.5-ft 3 filters used for liquid waste pretreatment that contain no fuel particles
can be disposed of as low-level radioactive waste inthe dewatered condition. They can be readily
packaged in 55-gallon drums, each containing seven assemblies. When these same or similar-sized
cartridges contain core debris or fuel particles, such cartridges may have to be handled aseither transuranic or high-specific-activity waste.
In either case, they will have to be stored
onsite until suitable facilities are provided by the federal government (Department of Energy).
They will have to be packaged in a manner that satisfies U.S. Department of Energy requirements.
The zeolite system prefilters are self contained in that the cartridge assembly also can serve as
the disposable container. Repackaging of these filters is not necessary.
In the event the larger core filters are generated,to accommodate their unusual configurations.

Table 8.9.

Source of TreatedLiquid Waste
1.
2.

Generation of Spent Filter Cartridges
Filter Cartridge Type and Loading
Core Typec
Inline TypebSDS Typea

Quantity

Curie/Unitsump water

11

RCS flush and drain water
RCSe

20

Quantity

Curie/Unit

Quantity

Curie/Unit

<100

-

-

-

-

<100

20 - 200d

40

2

<5

6

20

-

-

-

-

Reactor building

Reactor coolant
drain tank3.

special containers will have to be fabricatedd

RCS aqueous decontamination solutions

2<50

alO-ft3 stainless steel assemblies used with SDS-type systems.
0.5-ft 3 cartridges used for pre- and postremoval of particulates and ion-exchange media
fines.
c 8 -ft diameter x 3-ft high cartridge used to remove particulates during defueling.
dCould contain fuel debris and would be handled with spent fuel.
Range of units depends
on extent of particulates mobilized during defueling.
eRange of spent filters depends on activity in system (i.e., 20,000 Ci, or 100,000 Ci).

8-13
8.1.2.5

Evaporator Bottoms

Evaporator bottoms could be generated through the. use of an evaporator to reduce the volume of
AFHB/reactor building and RCS decontamination solutions.
Current practice is to immobilize AFHB and reactor building chemical decontamination solutions
with vinyl ester styrene, and the staff is not aware of any reason to discontinue this practice(see Sec. 8.2).
In contrast, the relatively large volume of chemical decontamination solutions
that could be generated from decontamination of the primary system (500,000 gallons) and the timeframe in which these liquids could be generated (after defueling) make the use of evaporation an
attractive alternative.
The characteristics of packaged waste that could be generated if 500,000 gallons of chemical
decontamination solutions are processed through an evaporator are presented in Table 8.10. The
minimum and maximums shown are based on varying the volume reduction factor from 10 to 30. As
shown in the table, it has been assumed that these wastes will be immobilized either with vinyl
ester styrene or cement and packaged in 55-gallon drums. The use of 50-ft 3 liners would reduce
the number of packages shown by a factor of 6.4 and increase the curie content and radiationlevels shown by a factor of 6.8. If the CAN DECON technique is used for reactor coolant systemdecontamination (treated by ion-exchange techniques) or if these chemical decontamination solutions are not treated by evaporation (see Sec. 8.2), the minimum packaged waste volume would be
zero.

Table 8.10.

Characteristics of Packaged Evaporator Bottoms-Minimum and Maximum Generation
Minimum Generation

Maximum Generation

Original volume (gallons)
Volume afte 9 evaporation(gallons)

500,000

500,000

16,70050,000

Package type
Volume increase factor

55-gallon drum
1.65

55-gallon drum
1.65

Number of packagesb

560

1,670

Average Ci per packagec

36

12

30 R/hr

10 R/hrAveraye surface radiation
level

avolume reduction factor of 10 for maximum and 30 for minimum.
bBased on 30 gallons per package for vinyl ester styrene or cement.'
cBased on 20,000 Ci inventory. Could be as low as 2000 Ci which would
reduce activity levels and curie content to 10 percent of values shown.
dBased on 0.8 R/hr per curie.

8.1.2.6

Bituminized Materials

Bituminized materials could be generated through the use of a bitumen/resin system to reduce thevolume and immobilize AFHB/reactor building and RCS decontamination solutions. This one-stepvolume reduction and immobilization system directly immobilizes the waste in bitumen. The otherwaste forms generated include organic resins. For the reasons stated above in Section 8.1.2.5,
it is doubtful that this system would be used for the AFHB/reactor building solutions; however,it could be used for RCS decontamination solutions.

8-14
The characteristics of the waste that could be generated if 500,000 gallons of chemical decontamination liquids are processed through a bitumen/resin system are presented in Table 8.11. The
system considered is designed to package bituminized waste in drums, and there is no reason to
Organic resins could be packaged and immobilized with cement or vinyl esterchange this package.
Alternatively, the organic resins could be removed fromstyrene within the demineralizer vessel.
the demineralizer vessel, combined with liquid waste, and processed through the extruder/evaporator
for packaging in drums.

Table 8.11.

Characteristics of Waste Generated from Bitumen/Resin System

Characteristics

Bituminized
Liquids

Organic
ResinsOriginal volume (gallons)

500,000

NA

Volume after evaporatAonin extruder (gallons)

25,000NA55-gallon drum

50-ft 3 liner

Volume increase factorb

2

1.5

Number of packages

1,000

2

Average Ci/packagec

20

5

Average surface radiationlevel (R/hr)

16

4

Package typeaVolume reduction factor of 20.
bBased on 50/50 weight percent bitumen/waste salts.
CBased on 20,000 Ci inventory.

Could be as low as 2,000 Ci,

which would

reduce activity levels and curie content to 10 percent of values shown.
dBased on 0.8 R/hr per curie.

8.1.3

Details of Methods and Facilities

Process solid wastes arise from treatment of liquid wastes, and, as discussed in Section 7.1,
alternative techniques for treatment of liquid wastes from all sources except the AFHB are still
The characteristics of the packaged process solid waste thatbeing considered by the licensee.
Packaging andbound the alternatives still being considered are described in this section.
handling of the EPICOR II organic resins which have already been generated also are discussed.
8.1.3.1 Accident Sludge
The bounding conditions for generation of packaged accident sludge were presented in Table 8.5
The relatively high specific activity ofbased on the conditions described in Section 8.1.2.1.
No decision has been made on whether
AFHB sludge precludes the use of large, 50-ft 3 containers.
these materials would be packaged in a dewatered or an immobilized state. However, so as not toforeclose future treatment of these materials because of their high specific activity, packaging
Thus, based on the information currently available on thesein the dewatered state is preferred.
materials, it is assumed for this analysis that AFHB sludge will be packaged in 55-gallon drums
in the dewatered condition.
The radioactivity level of reactor building sludge permits the use of either drums or large
Under current requirements forcontainers and disposition as low-level radioactive waste.

8-15
disposal of low-level radioactive waste, sludges can be packaged in the dewatered condition.
Ifthese sludges are contaminated with transuranic elements with concentrations in excess of
10 nCi/g, they would have to be handled as transuranic (TRU) wastes.
Under these conditions,retention of 'the sludges in the dewatered form would provide maximum flexibility for subsequenttransfer of TRU sludges to a federal government facility. Based on these considerations, reactor
building accident sludge also would be packaged in the dewatered condition either in drums orlarge containers.
The characteristics of the packaged accident sludge that could be generated are shown in
Table 8.12.
If the AFHB or reactor building sludge is disposed of as low-level radioactive waste
and shallow land burial site criteria are changed to require immobilization of these sludges, thenumber of packages shown in Table 8.12 would be increased by 65 percent and the curie content andradiation levels shown would be reduced by 40 percent.

Table 8.12.

Characteristics of Packaged Accident Sludge

Package Characteristics
Total Vo ume

Surface
Radiation bevel(R/hr)(ft3)1Type

Numberb

Curie
Content

150

55-gallon drum

22

435

350

Minimum

100

50-ft 3 liner

2

72

60

Maximum

100

55-gallon drum

15

11

9

AFHB sludge
Reactor building sludge

aSlud e volume; packaged volume greater than this due to freeboard in container.
3

7 ft,;
50 ft
liners contain 45 ft
bRounded to nearest package.

3

Drums contain.

cBased on 0.8 R/hr per Ci in package.

8.1.3.2

Zeolites

The bounding conditions for generation of 10-ft 3 stainless-steel liners containing 8 ft 3 ofzeolite in the dewatered condition were given in Table 8.6. As discussed in Section 8.1.2.2,
these zeolites will be handled as high-specific-activity waste and with the intention of beingtransferred to federal government facilities for disposition.
The alternatives related to treatment at such facilities discussed in Section 8.1.2.2 are still being considered.
Therefore, toensure compatibility with potential federal government facility requirements, zeolites will be
stored at the TMI site in the dewatered condition until arrangements are made with the U.S. DOE
for their disposition or until a suitable disposal facility becomes available.
In cases when zeolites are selected as the ion-exchange media for treatment of liquid wastes, theliner configurations actually used and the radioactivity loadings of the liners will conform torequirements acceptable to DOE.
However, to characterize the zeolites that could be generated,
the staff made certain assumptions relative to the liner configuration, zeolite loading per
liner, radioactivity loading per liner, and the sources of liquid waste that could readily betreated using zeolites or'zeolite/titanate ion-exchange media. These assumptions include:
Zeolites would be placed in stainless-steel liners of the SDS configuration.
capacity liner would contain 8 ft 3 of zeolite or zeolite/titanate mixture.

Each 10-ft 3

8-16
For treatment of reactor building sump liquids,
10,000 Ci and a maximum of 120,000 Ci.

liners would be loaded with a minimum of

For treatment of other sources of liquid wastes, the zeolite system would not be located in
Under these conditions, zeolite liners would be loaded to 10,000 Ci tothe spent fuel pool.
minimize shielding requirements around the relocated equipment.
Reactor building sump liquids would be treated with a modified SDS system, and RCS liquids
and RCS flush and drain liquids would be treated with a modified EPICOR II system.
The characteristics of the zeolite liners that could be generated under these conditions areshown in Table 8.13.
As shown, the minimum number of liners that could be generated is 11, while
The major factor contributing to this range is the loading of the zeolites
the maximum is 78.
used for reactor building sump water treatment.
For treatment of reactor building sump water and possibly RCS water, the SDS-type system will be
operated submerged in spent fuel pool B. The steps involved in handling expended zeolite andcation vessels are as follows:
Vessel couplings are disconnected with a monorail
vessel requires three disconnects.

lifting device and a special tool.

Each

After a vessel is disconnected from the system, it is transferred in the pool using the fuelhandling crane.
A special yoke and shaft are attached to the crane to perform these transfer
operations.
Transfer operations involve movement of a partially expended zeolite vessel to anotherposition in the system or transfer of a completely expended zeolite or cation vessel to theunderwater storage rack.
Expended filters are also transferred in this manner.
For shipment or transfer from the pool, a shield cask is loaded into the spent fuel pool.
As the cask is raised
The cask is loaded and closed except for the vent and drain plug.
above the pool, it is gravity drained.
The cask is then transferred to a transport vehicle or used to move the expended liner orfilter cartridge assembly to another storage location in the plant.
The mixed bed demineralizer is operated within a shield above the surface of the spent fuel pool.
When the contents of this vessel are expended, the liner and cask are replaced.
Prior to defueling, the system will have to be removed from the spent fuel pool and relocatedwithin the plant.
The handling procedures that would be used in another location have not been
defined, but would be similar to those used for handling EPICOR II system liners.
8.1.3.3

Organic Resins

The organic resins that have been generated through use of EPICOR II and those that could be
generated from treatment of other liquids with ion-exchange techniques can be divided into three(1) high specific acticategories that reflect the radioactive characteristics of the wastes:
Each catevity waste materials, (2) TRU waste materials, and (3) low-activity waste materials.
gory of organic resin wastes is discussed separately below.
High-Specific-Activity Organic Resins
The characteristics of the organic resins generated by EPICOR II which are currently being stored
in the dewatered condition were summarized in Table 8.7. The 46 prefilter liners listed in thattable, representing about 1400 ft 3 of resin, must be handled as high-specific-activity waste.
Because of their characterDetailed information on these liners is presented in Table 8.14.
istics, these materials cannot be disposed of using routine methods at low-level waste disposal
facilities.
Transfer to a federal government facility represents the most practicable alternaThe federal facility/treatment alternatives being considered weretive for these materials.
described in Section 8.1.2.3.

Table 8.13.

Characteristics of Packaged Zeolite Liners
Packaged Liner Characteristicsa

Minimum Generation

Source of TreatedLiquid Waste

Maximum Generation

Number ofLinersMaximum Curies
per Liner

Maximum Surface
Radiation Level
per Liner(R/hr)

Number ofLinersMaximum Curies
per Liner

Maximum Surface
Radiation Level
per Liner(R/hr)

1.

Reactor building sump water

6

120,000

100,000

54

10,0008,000

2.

RCS waterb

2

10,000

8,000

6

5,000

4,-00O3.

RCS flush and drain waterc2

10,0008,000

1210,0008,000

4.

Aqueous RCS d decontamination solutions
Total

1
11

2,000

1,600

6

5,000

4,000

78

aEach liner contains 8 ft 3 of zeolite.
bMinimum based on modified EPICOR II system; maximum based on modified SDS.
cMinimum based on modified EPICOR-II removal of 20,000 Ci; maximum based on modified SDS removal of 100,000 Ci.
dMinimum based on SDS/Modified SDS removal of 2,000 Ci; maximum based on SDS/Modified SDS removal of 20,000 Ci.

8-18
Table 8.14.

Characteristics of Packaged High-Specific-Activity
EPICOR II Resins

Number ofPrefiltsr
LinersTotal Curies
in Group

Maximum Curies b
Content per Liner

Maximum Surface
Radiation L~velof Liners(R/hr)

A

30

38,000

1,300

1,040

B
C
D

7
7
2

8,000
6,500
1,000

1,200
1,000
650

960
800
520

Total

46

53,500Group

aLiners are 4-ft-diameter x 4-ft-high carbon steel with volume of 50 ft 3
and contain 30 ft 3 of dewatered resin.
bMaximum gross curies for any liner in group.
cBased on 0.8 R/hr per Ci for the maximum curie content liner in group.
Hot spots in liners could lead to local radiation readings 1.5 times the maximum
shown.

Low-Activity Organic Resins
The organic resins generated by EPICOR II treatment of AFHB liquids were characterized inTable 8.2.
The organic resins in the three low-activity-level prefilter liners and the secondand third-stage liners may be immobilized. The characteristics of these resins after immobilization with cement are presented in Table 8.15.
Techniques for immobilization of these EPICOR II organic resins with cement are currently being
evaluated by the licensee.
To minimize package handling and the number of cask shipments
required, these evaluations are based on the use of large containers that correspond to the sizesof the demineralizer vessels in the EPICOR II system. Two approaches are being considered:Out-of-Container Mixing--This approach involves transfer of the resins from the demineralizer vessels, mixing the resins and immobilization agent with an in-line mixer, and repackaging the mixture in the demineralizer vessel, which would serve as the disposable container
This approach requires the design and installation of a remc.oly operated solidification
facility and removal of the resins from the demineralizer vesbels.
In-Situ Mixing--This approach involves adding the immobilization agent to the demineralizer
vessel and mixing the resins and immobilization agent within the vessel.
This does not
require an extensive, remotely operated solidification facility, but does require theredesign of the demineralizer vessels to accommodate a mixing device. Thus, resins still
would have to be transferred from the vessels currently being used into new containers.
The mixing approach selected by the licensee will be based on operational considerations and will
not affect the volume of packaged waste. The immobilization agent used, however, will affect thevolumes of packaged waste. The licensee's work to date 7 with cement indicates that a mixture ofabout 33 weight percent resins, 14 weight percent water, 48 weight percent Portland II cement,
and 5 weight percent of a chemical additive could yield an acceptable immobilized product. Thesemixing ratios represent a volume increase factor of about 1.5. The vessels have sufficient void
space to accommodate a volume-increase factor of 1.5, so the planned use of cement will notincrease the number of containers required. Thus, Table 8.15 is based on a volume increase
factor of 1.5.
If vinyl ester styrene is used, the estimated volume increase factor also would
be about 1.5.

8-19
Table 8.15.

Characteristics of Packaged Low-Activity EPICOR II
Organic Resins

Maximum Surface
Radiation Levelof Liners(R/hr)

Number ofLinersTotal Curies
in Group

Maximum Curie
Content per Liner

Prefilterd
Cation Ad

3
4

440
160

155
60

125

Cation B

3

55

20

16

Cation C
Mixed bede

7
6

25
20

6
6

5
5

Total

23

700

Group

50

aOrganic resins are immobilized in cement with a volume increase factor of
1.5 and packaged in EPICOR II system vessels.
bMaximum gross curies for any liner in group.
cBased on 0.8 R/hr per Ci for the maximum curie content liner in group.
Hot spots could lead to local radiation readings 1.5 times the maximum
shown. Mixed bed vessels are 6-ft-diameter x 6-ft-high liners whichcontain 115 ft 3 of resin plus cement.
dPrefilter and cation vessels are 4-ft-diameter x 4-ft-high liners whichcontain 30 ft 3 of resins plus cement.
eMixed-bed vessels are 6-ft-diameter x 6-ft-high liners which contain
115 ft 3 of resin plus cement.
Source: "EPICOR II Resin Solidification Procurement Specification,"
Letter (TLL 545) from G.K. Hovey, Metropolitan Edison Co. - TMI, to
TMI Program Office, November 17, 1980.

Low-activity resins also could be generated from treatment of other liquids with ion-exchange
techniques.
The bounding quantities of these materials that could be generated were given in
Table 8.8.
The characteristics of these packaged materials under minimum and maximum waste
generation conditions are described in Table 8.16.
The table is based on the assumption thatorganic resins will be immobilized and repackaged in the ion-exchange vessel removed* from thetreatment system. The assumed volume increase factor due to immobilization is 1.5 as discussed
above.
As shown in Table 8.16, under best-case conditions,' 1 liners containing about 650 ft 3 ofresins could be generated.
Under worst-case conditions, 62 liners containing about 3100 ft 3 ofresins could be generated.
TRU Organic Resins
The RCS flush and drain liquids could contain dissolved transuranic elements. Treatment of theseliquids with ion exchange technology will lead to disposition of these transuranic elements on
the organic resins used in the treatment system. The concentration of these transuranic elements
could exceed 10 nCi/g, which would require their packaging and handling as TRU wastes. Under
these conditions, the organic resins would have to be retained in the dewatered condition andtransferred to a DOE facility for disposition as TRU waste.
The characteristics of the TRU organic resins that could be generated from treatment of RCS flush
and drain water are presented in item 3 of Table 8.16. These materials would be transferred to
DOE in the dewatered condition. They would be packaged in the ion-exchange vessels removed fromthe treatment system.
Since Table 8.16 is based on an immobilization volume increase factorof 1.5, there is no difference between the estimated number of immobilized and dewatered liners.

Table 8.16.

Characteristics of Packaged Low-Activity Organic Resins
Minimum Generation

Maximum Generation

Number ofLinersMaximum Curies
per Liner

Maximum Surface
Radiation Level
per Liner(R/hr)

Number ofLinersMaximum Curies
per Liner

Maximum Surface
Radiation Level
per Liner(R/hr)

Cation

2

50

40

4

50

40

1

5

4

2

5

4

2.

Mixed bed
RCS waterb

2

50
5

40

7

50

40

4

3

5

4

3.

CationMixed bed
RCS flush and drain waterc

Source of TreatedLiquid Wastes
1.

Reactor building sump watera

OC

2
1

50
5

40
4

35
8

50

40

5

4

Cation

1

5

4

2

5

4

Mixed bed
Total

1

5

4

1

5

4

11

CationMixed bed
4.

1

RCS decootamination
solution

62

aMinimum based on SDS/Modified SDS; maximum on SDS/EPICOR II.
bMinimum based on SDS/Modified SOS; maximum on modified EPICOR II.
CMinimum based on SDS/Modified SOS removal of 20,000 Ci; maximum on modified EPICOR II removal of 100,000 Ci.
dMinimum based on use of evaporator/resin or bitumen/resin system; maximum based on
use of system with first-stage
zeolite liner.

8-21
8.1.3.4

Filters

The estimates for spent filter cartridge generation were presented in Table 8.9. As described in
Section 8.1.2.4, cartridges to be handled as low-level waste can be packaged in 55-gallon drums
or in special large containers.
Cartridges that could require handling as TRU or high-level
waste would be packaged in containers compatible with DOE requirements.
The latter include thefilter cartridges generated during RCS and reactor coolant flushing and draining activities.
Ifthese filters contain fuel particulates, the radiation level for a cartridge could be 500 R/hr.
8.1.3.5

Evaporator Bottoms

The bounding conditions for evaporator bottoms generation were presented in Table 8.10.
The
quantities shown in that table are based on the use of an evaporator/resin system to treat RCS
chemical decontamination solutions.
Implementation will require the design and construction of a separate evaporator facility that
could take up to three to four years to become operational.
Design work on such a facility was
initiated by the licensee but has been deferred.
A 1O-gpm-capacity evaporator compatible withprocessing of chemical decontamination liquids and other liquids was ordered, but the licensee
cancelled the order in 1980, pending selection of a technique for reactor coolant system decontamination.
If RCS decontamination solutions are generated, this project would have to be reactivated to process these liquids through an evaporator.
The characteristics of packaged waste that could be generated if 500,000 gallons of chemical
decontamination solutions are processed through an evaporator were presented in Table 8.10.
No decision has been made concerning the use of an evaporator.
The characteristics of the packaged waste arising from direct immobilization of these chemical decontamination solutions arepresented in Section 8.2.4.
8.1.3.6

Bituminized Materials

The characteristics of the packaged waste arising from use of a bitumen/resin system to process
500,000 gallons of RCS decontamination solutions were presented in Table 8.11.
This system is
described in Appendix H.
Space requirements for a system of the type considered are such that the equipment could not belocated within existing facilities.
Design and construction of facilities to incorporate thisequipment plus procurement and installation of the equipment could take three to four years.
Since no decision has been made on selection of a technique for decontamination of the reactor
coolant system, a decision cannot be made on the use of a bitumen/resin system at this time.
8.1.4

Effluents and Releases to the Environment

The nature and impacts of releases to the environment that could occur during normal and abnormal
conditions or accidents in the packaging and handling of process solid wastes are discussed inthis section.
8.1.4.1

Normal Operations

Under normal conditions of operation, airborne releases could arise during immobilization of
accident sludge, organic resins, and evaporator bottoms.
These releases would be released to theplant vent system.
Estimated releases during immobilization are presented in Table 8.17. These
values are based on a fractional release of 10-5, or 0.001 percent, of the radionuclides in thewaste that will be immobilized.
8.1.4.2

Package-Handling Accidents

Worst-case accidents during packaged-waste-handling operations would arise when a drum or resin
liner is dropped from a crane or monorail during remote handling.
The consequences of such accidents would depend on the waste type and its condition, the radioactivity and radionuclide content
of the breached container, the fractional release of materials in the container, its interaction
with other containers, and the area in the plant where the accident occurred.

Table 8.17.

Source of TreatedLiquid Waste
1.

2.

3.

4.

5.

Amounts Released to Exhaust System (Ci)Cs-134
Ru-106
Sb-125

Cs-137Ce-144

3.7 x 10-4NEG

3.4 x 10-7

2

x 10.6NEG

3.5 x 10-8

1.6 x 10-7NEG

1.4 x 10-8

9.7 x 10-8

3.7 x 10-9NEG

6.6 x 10-9

NEG

NEG

NEG

1.6 x 10-7

x 10-9

NEG

NEG

7

x 10-9

-

NEG

NEG

8

x 10-7

2:5 x 10-8

NEG

NEG

3.5 x 10-8NEG

1.5 x 10-5

Sr-89

Sr-90

Prefilters (440)Cation liners (240)

NEG

NEG

NEG

NEG

7

NEG

NEG

NEG

NEG

Mixed bed liners (20)

NEG

NEG

NEG

NEG

Cation liners (25)NEG

6.7 x 10-8

6.7 x 10-8

Mixed bed liners (55)
RCS primary waterc

NEG

1.4 x 10-8

2.2 x 10-8

Cation liners (95)
Mixed bed liners (5)RCS flush and drain
water
Cation liners (475)

1.9 x 10-8

1.6 x 10-7

6.1 x

5

5.5 x 10-9

3.2 x 10-8

9.5 x 10- 883Mixed bed liners (25)
RCS chemical decontamination solution
Evaporator bottoms

2.8 x 10-9

2.8 x 10-7

1.6 x 10-7

7.6 x 10-6

7.8 x 10-5

NEG

AFHB watera
X

10-7

Reactor building
sump water(20,000)(

Maximum Estimated Airborne Effluents Released to the Environment
during Immobilization of Process Solid Waste

x 10-10x 10-7

2.3 x 10-7

10-7

x 10-6) indicates total Ci in waste to be immobilized, worst case.

5

aResins presently in storage.
bTreated with modified SDS system or other system which removes Ru, Sb and Ce.
cTreated with modified EPICOR II system.

1

X 10-4

Assumed fractional release is 0.001 percent.

NEG

8-23
The fractional release rates used to estimate the amounts of radionuclides released in the formof respirable particulates for the postulated accidents are as follows:
Dewatered zeolites, and resins,
radionuclide content

accident sludges, and filter cartridges--lO- 4 of the package

Resins, accident sludges,
age radionuclide content

and evaporator bottoms immobilized with cement--10- 5 of the pack-Resins,and evaporator bottoms immobilized with vinyl ester styrene--10- 6

accident sludges,of the package radionuclide content
Waste materials immobilized with bitumen--10- 6 of the package radionuclide content.
The worst-case accidents postulated during handling of packaged waste from each alternative are
discussed separately below, and the releases that would result are summarized for all the accidents in separate tables.
Accidents Involving Zeolite-Based System Wastes
Different handling procedures are *used for each type of waste container or vessel used in thezeolite-based systems.
The zeolite and cation vessels and the spent filter cartridges are placed
in top-loaded shipping or transfer casks underwater in the spent fuel pool and then removed fromthe pool using the overhead crane. Mixed-bed demineralizer vessels are not handled in the spent
fuel pool.
The worst-case accident that could arise from these operations would occur during thetransfer of a first-stage zeolite liner in a shipping cask within the reactor building. Whilethe cask is designed to retain its integrity during a 30-ft drop, at one point it could fall a
distance of 60 ft.
For this 60-ft drop accident it was assumed that the zeolite liner in thecask, as well as the cask seal, would be breached.
The estimated releases for a zeolite liner
containing 120,000 curies of dewatered zeolites arising from treatment of reactor building sump
water are shown in Table 8.18.
This curie loading is the maximum expected.

Table 8.18.

Radionuclide

Estimated Releases during SDS-Based System Package Handling Accidents

Zeolite Linerab

Estimated Releases (Ci)Cation Linerb'dFilter Assemblyb'c

Mixed Bed Linere•Sr-891 x 10-2

2

x I0-4

1

x I0-55

x 10-6

Sr-90

0.12

3

x 10-38.5 x 10-4

6

x 10-5

Ru-106f-

3.2 x 10-3

3.2 x 10-4

Sb-125

4

Cs-134

1.6

4
1

Cs-13710.3

1.2 x 10-3

Ce-144x 10x 10-4-

5

8.5 x 10-4

x 10-s

8.5 x 10-5

aone liner containing 120,000 Ci.
bAccidents occur in AFHB; therefore,

releases to atmosphere will be 1/1000 thevalues shown.
cFilter loading of 50 Ci after cleanup of reactor building sump water.
dCation liner loading of 50 Ci after cleanup of primary system water with modified SDS
system; a worst-case isotopic distribution.
eMixed-bed liner loading of 5 Ci after cleanup of primary system water with modified SDS
system; a worst-case isotopic distribution.
Released directly to the atmosphere.
f"-" indicates that radionuclide is less than one percent of release.

8-24
The estimated releases arising from the breach of a spent filter cartridge containing 50 Ci and acation liner containing 50 Ci under similar accident conditions also are shown in Table 8.18.
In all cases,

the estimated releases would be vented to the plant exhaust system.

If the zeolite/resin system were installed in another location where there was no potential for acask drop of more thant 30 ft, there would be no releases because the cask would retain its
integrity under those accident conditions.
Use of a zeolite-based system also generates mixed-bed organic resins.
The worst-case accident
would involve the breach of one of these liners during transfer to a storage location outside of
the reactor building.
The estimated releases from a 190-ft 3 mixed-'bed liner are also shown in
Table 8.18.
The worst-case conditions shown are based on use of a modified SDS system for treatment of primary system water.
Accidents Involving EPICOR II

and Modified EPICOR II System Wastes

The worst-case scenario for EPICOR II wastes involves an accident with a prefilter liner containing dewatered organic resins with a loading of 1300 Ci.
This accident could occur during transferof the liner from within the chemical cleaning building to a storage module in the EPICOR II
waste storage facility.
The accident with the maximum release consequences would occur when adropped liner containing dewatered resins breaches the integrity of a second container already instorage and these releases are directly to the atmosphere.
The estimated releases for thispostulated accident for EPICOR II prefilter liners with a gross activity of 1300 Ci each are
given in Table 8.19.
Use of the EPICOR II system also generates second- and third-stage organic resin liners.
For theliners already generated, the worst-case accident would be when these liners are being removed
from the storage modules for transfer to an immobilization facility.
The resins are in a dewateredcondition, and the releases would be directly to the atmosphere.
For the smaller 50-ft 3 cation
liners, a second liner also could be breached when one liner is dropped on another.
The estimatedreleases from these accidents, based on two 50-ft 3 liners containing 60 Ci each and a 170-ft 3
mixed-bed liner containing 5 Ci, are also shown in Table 8.19.

Table 8.19.

Estimated Releases during EPICOR II and Modified EPICOR II
System Package Handling Accidents

EPICOR II

Radionuclide

Prefilter
Liner(2,600 Ci)

Sr-89
Sr-90
Ru-106

Estimated Releases (Ci)Modified EPICOR II

CationLiner(120 Ci)

Mixed BedLiner
(5 Ci)

ZeoliteLiner(10,000 Ci)--

4.1 x 10-2
0.39

-

--

Sb-125---

Cs-1344 x 10- 2

Cs-1370.22

Ce-144

Systema

1.7 x 10- 3

1 x 10- 4

7.5 x 10- 2

x 10-25 x 10-4

0.50

1

--

Systemb

CationLiner(120 Ci)

Mixed BedLiner
(5 Ci)

2x 10- 42

3-

x 107.5 x 10-3

9

x 10- 55

x 10-4-8

x 10-5---2.5 x 10-3

1.1 x 10-4"-" show presence of isotope or less than 1 percent or no data for EPICOR II.

aBased on isotopic distribution in EPICOR II

liners in storage,

prefilter PF-10,

cation DR-11,

and mixed bed DS-2. ("Scoping Studies of the Alternative Options for Defueling, Packaging,
Shipping, and Disposing of the TMI-2 Spent Fuel Core," Allied-General Nuclear Services,
AGNS-35900-1.5-79, September 1980.)
bBased on cleanup of primary system water.

8-25
A modified version of the EPICOR II system also is being considered for treatment of primarysystem water.
If this alternative is selected, cation and mixed-bed liners will be generated.
The worst-case accident for these liners is identical to that described above for liners alreadygenerated.
The estimated releases from these accidents, based on two 50-ft 3 cation liners containing 60 Ci each and a 170-ft 3 mixed-bed liner containing 5 Ci, are also shown in Table 8.1-19.
The modified EPICOR II system that could be used to treat primary system water also generates a
10-ft 3 zeolite liner.
This liner could contain up to 10,000 Ci and would also be transferredfrom the chemical cleaning building to another area for storage.
During this transfer operation,
the liner could be dropped 15 ft.
The estimated releases for this accident are shown in
Table 8.19.
Accidents Involvinq Other Wastes

Accident releases could also result from the breach of packages containing accident sludge,
immobilized evaporator bottoms, and bituminized chemical decontamination solutions.
The worst-case accident for accident sludges would bethe breach of a 55-gallon drum containing
dewatered AFHB sludge during remote placement of the drum in a shipping cask. This accident
could occur outdoors, and releases would be directly to the atmosphere. The releases for this
accident are shown in Table 8.20 for a drum containing 250 Ci, the maximum estimated curie content
of a drum of dewatered AFHB sludge. The estimated releases from the breach of a 50-ft 3 liner
containing dewatered reactor building sludge also are shown in the table. This liner was assumed
to contain 70 Ci, the maximum curie content estimated for a liner of reactor building accident
sludge.
Cement immobilization would reduce the releases shown by a factor of 165, and vinyl
ester styrene immobilization would reduce them by a factor of 1650.
The worst-case accident for immobilized evaporator bottoms or bituminized decontamination solutions would occur if a drum containing these materials was breached during cask loading.
Theestimated releases for these accidents also are shown in Table 8.20.

Table 8.20.

Estimated Releases during Other Waste Type
Package Handling Accidents

Estimated Releases (Ci)aRadionuclide

AFHB
Accident Sludge(250 Ci)

RCS
Evaporat 8 r
Bottoms(36 Ci)

RCS
Bituminizgd
Liquids(20 Ci)-

1.4 x 10-5

8

x 10-3

1.4 x 10-4

7.8 x 10-6

-

-

-

-

2.9 x 10-5
1.8 x I0-4

1.6 x 10-6

Reactor Building
Accident Sludge(70 Ci)x 10-4

Sr-89

5

Sr-90

2.5 x 10-3

6

.Cs-134

3.5 x 10-3

1.3 x 10-4
1.4 x 10-4
x 10-s8

Cs-137

1.8 x 10-2

5x I0-4

7

5

Ru-106
Sb-125

Ce-144x 10-

x 10-79.8 x 10-6-

aReleases are directly to atmosphere.
If accident occurs in building with
HEPA filter, releases will be a factor of 103 less than the values shown.
bWaste forms arise from different treatment of same liquid waste source.
Therefore releases are mutually exclusive.

8-26
8.1.5

Environmental Impacts

8.1.5.1

Occupational Doses

The estimated occupational doses to immobilize organic resins and evaporator bottoms and to
package and handle these and other process solid wastes are discussed in this section. Theseestimates cover the operations involved in handling the packages through placement of the waste
in onsite storage facilities to await shipment.
Organic Resins
The EPICOR II system exposure experience for system operation and liner handling was used toestimate occupational exposure during immobilization of organic resins and placement of thesepackaged wastes into storage to await shipment.
The operations and sampling exposure was 0.006 person-mrem per gallon for a throughput of 2 gpm.
This corresponds to about 0.7 person-mrem per hour of system operation. For resin immobilization, liners have to be sluiced to a transfer tank at a rate of about 1 gpm, mixed with a bindermaterial and repackaged.
Thus, resins can be transferred at an average rate of about 8 ft 3 perhour. The throughput of a typical immobilization system for large liners is about 40 ft 3 perhour. Combining these two factors results in an average cumulative dose rate of about 0.1 personmrem/ft 3 of resin.
For the maximum volume of resins to be immobilized, about 3300 ft 3 , theestimated cumulative dose is about 0.3 person-rem.
The EPICOR II liner handling experience of 85 mrem per liner is also relevant since resin liners
must be removed from storage, immobilized, and then returned to storage. A maximum of about
100 liners will have to be immobilized, resulting in cumulative occupational dose of about
8.5 person-rem.
For EPICOR II, cumulative maintenance dose was about 1.4 person-mrem per hour ofsystem operation.
Since the maintenance needs of an immobilization system will exceed those of
EPICOR II, the estimated cumulative maintenance dose for the immobilization system was increased
to 4 person-mrem per hour.
For 100 hours of operation, the cumulative maintenance dose is
0.4 person-rem.
Thus, the total estimated cumulative dose is about 9 person-rem. These operations will be performed by a three-person crew over a period of 18 months. Under these conditions, the estimateddose to each crew member is 3 rem, and the quarterly dose is about 0.5 rem.
Based on a maximum cumulative dose of 9 person-rem and a work force of three persons, the expected
number of additional cancer mortalities in the work force would be less than 0.001. The added
probability that the average worker will die of cancer is 1 in 3000.
The expected number of
additional genetic effects in the offspring of the work force exposed to this cumulative dosewould be less than 0.002.
Evaporator Bottoms and Bituminized Decontamination Solutions
The generation of evaporator bottoms or bituminized materials results from treatment of RCS
chemical decontamination solutions. Thus, they are mutually exclusive alternatives.
The dose estimates to process these wastes through a bitumen/resin system were presented in
Section 7.1.4. This cumulative dose was about 12.5 person-rem. The handling of the 1000 drums
that could be generated will lead to additional cumulative dose of about 1 person-rem (see Appendix N).
The dose estimates for treating RCS decontamination liquids with an evaporator/resin system also
were presented in Section 7.1.4. The range of cumulative dose estimates was 32 to 63 person-rem.
Additional occupational dose will arise from immobilization of evaporator bottoms and drum handling.
The typical throughput of an immobilization system packaging waste in drums is about two drums
per hour.
At 0.7 person-mrem per hour (EPICOR II experience), the operations cumulative dose
incident to immobilizing the maximum of 1670 drums is about 0.6 person-rem.
Handling will becompletely remote and the estimated dose is less than 1 person-mrem per drum (see Appendix N).
For 1670 drums, this is about 1.7 person-rem. At a throughput of two drums per hour, the immobilization system will have to be operated for about 850 hours to package 1670 drums. At

8-27
4 person-mrem per hour for cumulative maintenance dose, this results in a cumulative dose ofabout 3.4 person-rem.
Since immobilization and handling of evaporator bottoms drums represents aworst-case exposure condition, the total cumulative dose of about 5.7 person-rem is considered
here.
These operations will be performed by a single three-person crew over a period of 18 months.
Under these conditions, the estimated dose to each crew member would be 1.9 rem, and the quarterly
dose would be about 0.3 rem.
Based on a maximum cumculative dose of 5.7 person-rem and a work force of three persons, theexpected number of additional cancer mortalities in the work force would be less than 0.0007.
The added probability that the average worker will die of cancer is 1 in 4000.
The expected
number of additional genetic effects in the offspring of the work force exposed to this cumulative radiation dose would be less than 0.0015.
Other Process Solids
The estimated doses for packaging and handling other process solids were based on the informationin Appendix N. The cumulative occupational dose estimates for packaged accident sludge and spent
filters are shown in Table 8.21.
These wastes will be generated and packaged intermittently over about 36 months. 'For the cumulative occupational dose of 2.3 person-rem, the expected number of additional cancer mortalities ina work force of three people is 0.0003.
This means that the added probability that the average
individual worker would die of cancer is 1 in 10,000.
The expected number of additional genetic
effects in the offspring of the work force exposed to this cumulative dose of radiation is about
0.0006.

Table 8.21.

Estimated Worker Exposure to Package and Handle
Accident Sludge and Filters

Waste Type

Person-mrem

Cumulative Occupational

per Package

Dose (person-rem)

2

0.01

1.

AFHB accident sludge

2.

Reactor building
accident sludge

2

0.05

3.

Core filtersb

22

0.12

4.

In-line filtersc

11

2.1

Total
aBased on Appendix N estimates for EPICOR II

2.3
liners.

bBased on Appendix N estimates for irradiated hardware.
cBased on maximum number of filters--could be a factor of 10 lower.

8.1.5.2

Offsite Doses

The dose estimates presented here for processing of solid wastes are based on the source terms
developed in Section 8.1.4. The calculational models used to make these estimates and the interpretation of their results are described in Appendix W. The significance of these doses andtheir human health and environmental consequences are discussed in Section 10.3.
The dose estimates are listed in Table 8.22.
The total-body population dose received by the human population
residing within 50 miles during these activities is estimated to be about 2 x 10-2 person-rem.

8-28

Table 8.22.

Location

Dose Estimates for the Maximum Exposed IndividualDue to Processing Solid Wastes
Dose (mrem)a
Bone

Liver

3.1 x 10-5
4.2 x 10-5

4.5 x 10-4

2.2 x 10-5

4.2 x 10-5

4.2 x 10-5

Vegetable Use

2.6 x 10-3

1.1 X 10-2

1.1 x 10-3

Total

2.7 x 10-3

1.1 X 10-2

1.2 x 10-3

Inhalation
Ground Shine
Goat Milk Use

3.3 x 10-5
4.0 x 10-5
8.6 x 10-4

1.7 x 10-4
4.0 x 10-s
6.8 x 10-3

1.5 x 10-5
4.0 x 10-5
6.3 x 10-3

Total

9.3 x 10-4

7.0 x 10-3

6.4 x 10-3
2.4 x 10-5
6.2 x 10-5

Pathway

Total-Body

Nearest b

Inhalation

gardenGround Shine

Nearestmilk goat

Nearest cow

and garden

Inhalation

3.4 x 10-5

5.0 x 10-4

Ground Shine
Vegetable Use
Cow Milk Use

6.2 x 10-5
3.8 x 10-3
3.6 x 10-4

6.2 x 10-5
1.6 x 10-2
2.0 x 10-3

Total

4.3 x 10-3

1.9 x 10-2

aDoses were calculated for total-body,

1.6 x 10-3
1.2 x 10-3
2.9 x 10-3

GI-tract, bone, liver, kidney,
thyroid, lung, and skin.
The maximum three organ doses are listed in thistable. Doses were calcuated for four age groups:
adults, teenagers,
children, and infants.
The highest dcse estimates for each age group are
listed. The dose estimates for the neearest garden and nearest cow
and garden locations are for children and those for the nearest milkgoat location are total-body estimates for adults and bone and liver estimates for infants.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are:nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-29

8.1.5.3

Postulated Accident Effects

The type of accident for which dose estimates are presented here is the breach of a package
containing waste. The accident scenarios are described in Section 8.1.4.2. The calculationalmodels used to make these estimates and the interpretation of their results are described inAppendix W. The significance of these doses is discussed in Section 10.4. The estimated doses
to the maximum exposed individual and the source for each accident scenario are listed in thefollowing tables:

Scenario

Dose Estimate
Table Number

Source Term
Table Number

SDS waste (reactor building sump water-zeolite liner)

8.23

8.18

SDS waste (reactor building sump water-filter assembly)

8.24

8.18

SDS waste (primary water--cation liner)

8.25

8.18

SDS waste (primary water--mixed-bed liner)

8.26

8.18

EPICOR II waste (AFHB water-prefilter liner)

8.27

8.19

EPICOR II waste (AFHB water--cation
liner)

8.28

8.19

EPICOR II waste (AFHB water--mixedbed liner)

8.29

8.19

Modified EPICOR II waste (primarywater--zeolite liner)

8.30

8.19

Modified EPICOR II waste (primarywater--cation liner)

8.31

8.19

Modified EPICOR II waste (primarywater--mixed-bed liner)

8.32

8.19

8-30

Table 8.23. Dose Estimates for the Maximum Exposed IndividualDue to Breaching of a Package Containing SDS Waste-Reactor Building Sump Water - Zeolite Liner

Location

Pathway

Total-Body

Dose (mrem)a
Bone

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

5.7 x 10-6
6.8 x 10-5
4.7 x 10-4

2.0 x 10-5
6.8 x 10-5
2.4 x 10-3

9.1 x 10-6
6.8 x 10-5
1.5 x 10-3

Total

5.4 x 10-4

2.5 x 10-3

1.6 x 10-3

Inhalation

5.7 x 10-6

Ground Shine
Goat Milk Use

1.1 x 10-3

1.0 x 10-5
6.8 x 10-5
7.6 x 10-3

6.7 x 10-6

6.8 x 10-s

Total

.1.2 x 10-3

7.7 x 10-3

9.4 x 10-3

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

3.9 x 10-6
6.8 x 10-5
4.7 x 10-4
2.7 x 10-4

1.4 x 10'6.8 x 10-s
2.4 x 10-3
1.2 x 10-3

6.3 x 10-6
6.8 x 10-5
1.5 x 10-3
1.2 x 10-3

Nearestmilk goat

Nearest cow
and garden

Liver

6.8 x 10-s
9.3 x 10-3

3.7 x 10-3
2.8 x 10-3
8.1 x 10-4
liver,
kidney,aDoses were calculated for total-body, GI-tract, bone,
thyroid, lung, and skin. The maximum three organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
The highest dose estimates for each age group arechildren, and infants.
listed. The dose estimates for the nearest garden and for the nearest cow
and garden locations are for adults for total-body and for children forbone and liver. The dose estimates for the nearest milk goat location are
for adults for total-body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
Total

The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-31

Table 8.24.

Dose Estimates for the Maximum Exposed IndividualDue to Breaching of a Package Containing
SDS Waste--Reactor Building Sump Water Filter Assembly

Location

Pathway

Total-Body

Dose (mrem)a
Bone

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

1.8 x 10-5
7.0 x 10-6
5.4 x 10-3

2.8 x 10-4
7.0 x 10-6
2.1 x 10-2

1.0 x 10-6
7.0 x 10-6
1.7 x 10-4

Total

5.4 x 10-3

2.1 x 10-2

1.8 x 10-4

Inhalation
Ground Shine

1.1 x 10-4

7.0 x 10-6

7.4 x 10-7
7.0 x 10-6

Goat Milk Use

7.2 x 10-6
7.0 x 10-6
9.3 x 10-4

4.3 x 10-3

1.0 x 10-3

Total

9.4 x 10-4

4.4 x 10-3

1.0 x 10-3

Inhalation
Vegetable Use
Cow Milk Use

1.2 x 10-5
7.0 x 10-6
5.4 x 10-3
2.8 x 10-4

1.9 x 10-4
7.0 x 10-6
2.1 x 10-2
1.2 x 10-3

6.9 x 10-7
7.0 x 10-6
1.7 x 10-4
1.3 x 10-4

Total

5.7 x 10-3

2.2 x 10-2

3.1 x 10-4

Nearestmilk goat

Nearest cow

and garden

Ground Shine

LiveraDoses were calculated for total-body, GI-tract, bo)[ne, liver, kidney,
thyroid, lung, and skin. The maximum three organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants.
The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and the nearest cow
and garden locations are for children, and for the nearest milk goat
location are for infants.

bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-32

Table 8.25.

Dose Estimates for the Maximum Exposed IndividualDue to Breaching of a Package Containing
SDS Waste--Primary Water - Cation Liner
Dose (mrem)a

Location

Pathway

Total-Body

Bone

Liver

NearestInhalation

gardenb

Nearestmilk goat

Nearest cow
and garden

5.3 x 10-6

Ground Shine
Vegetable Use

2.1 x 10-7
1.5 x 10-3

8.4 x 10-5
2.1 x 10-7
6.0 x 10-3

2.1 x 10-7
1.9 x 10-7

1.7 x 10-6

Total

1.5 x 10-3

6.1 x 10-3

2.1 x 10-6

Inhalation
Ground Shine
Goat Milk Use

2.2 x 10-6
2.1 x 10-7
2.4 x 10-4

3.4 x 10-s
2.1 x 10-7
9.5 x 10-4

9.5 x 10-7
2.1 x 10-7
2.5 x 10-9

Total

2.4 x 10-4

9.8 x 10-4

1.2 x 10-6

Inhalation
Ground Shine
Vegetable Use

3.7 x 10-6
2.1 x 10-7
1.5 x 10-3

Cow Milk Use

7.8 x 10-5

5.8 x 10-5
2.1 x 10-7
6.0 X 10-3
3.0 x 1 0 -4

1.1 x 10-6
2.1 x 10-7
1.9 X 10-7
7.9 x 10-9

Total

1.6 x 10-3

6.4 x 10-3

1.5 x 10-6

aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and the nearest cow
and garden locations are for children, and for the nearest milk goat
location are for infants.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-33

Table 8.26.
Dose Estimates for the Maximum Exposed Individual
due to Breaching of a Package Containing SDS Waste--Primary
Water - Mixed-Bed Liner
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

3.8 x 10-4
2.1 x 10-5
1.1 x 10-1

6.1 x 10-3
2.1 x 10-s
4.3 x 10-1

1.7 x 10-4
2.1 x 10-5
1.9 x 10-s

Total

1.1 x 10-1

4.4 x 10-1

2.1 x 10-4

Nearestmilk goat

Inhalation
Ground Shine
Goat Milk Use
Total

1.6 x 10-4
2.1 x 10-s
1.7 x 10-2
1.7 x 10-2

2.5 x
2.1 x
6.8 x
7.1 x

10-3
10-5
10-2
10-2

9.5 x 10-5
2.1 x 10-5
2.5 x 10-7
1.2 x10-4

Nearest cow
and garden

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

2.6 x 10-4
2.1 x 10-5
1.1 x 10-1
5.4 x 10-3

.4.2 x
2.1 x
4.3 x
2.2 x

I0-3

1.1 x
2.1 x
1.9 x
7.9 x

Total

1.2 x 10-1

4.6 x 10-1

10-5

10-12
10-

10-4

10-5
10-5
10-7

1.5 x 10-4

afDoses were calculated for total-body, GI-tract, boine, liver, kidney,
thyroid, lung, and skin. The maximum three organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and the nearest cow
and garden locations are for children, and for the nearest milk goat
location are for infants.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-34

Table 8.27.
Dose Estimates for the Maximum Exposed Individual
due to Breaching of a Package Containing EPICOR II
Waste--AFHB Water - Prefilter Liner
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

1.1 x 10-1
1.5
8.7

2.1 x 10-1
1.5
33

2.1 x 10-1
1.5
35

10

35

37

1.1 x 10-1
1.5
25

1.3 x 10-1
1.5
170

1.5 x 10-1
1.5
210

27

170

210

7.8 x 10-2
1.5
8.7
6.0

8.7 x 10-2
1.5
33
41

1.0 X 10-1
1.5
35
51

16

76

88

GI-tract, bone,

liver, kidney,

Total
Nearestmilk goat

Inhalation
Ground Shine
Goat Milk Use
Total

Nearest cow
and garden

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use
Total

afDoses were calculated for total-body,thyroid, lung, and skin.
The maximum three organ doses are listed in thistable.
Doses were calcuated for four age groups:
adults, teenagers,
children, and infants.
The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and the nearest cow
and garden locations are for adults for total-body and for children forbone and liver. The dose estimates for the nearest goat location are for
adults for total-body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are:nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-35

Table 8.28.

Dose Estimates for the Maximum Exposed Individual

due to Breaching of a Package Containing EPICOR II
Waste--AFHB Water - Cation Liner

Pathway

Total-Body

Dose (mrem)a
Bone

Nearest b

Inhalation

5.1 x 10-3

9.4 x 10-3

gardenGround Shine

6.8 x 10-2

6.8 x 10-2

6.8 x 10-2

Vegetable Use

4.0 x 10-1

1.5

1.5

Total

4.7 x 10-1

1.6

1.6

Inhalation

5.1 x 10-3

5.7 x 10-3

6.8 x 10-2

6.8 x 10-2

6.7 x 10-3

Ground Shine

Goat Milk Use

1.1

7.6

9.5

Total

1.2

7.7

9.6

Location

Nearestmilk goat

Nearest cow
and garden

Inhalation

Liver
9.2 x 10-3

6.8 x 10-2

3.5 x 10-3

6.8 x 10-2

6.5 x I0-3

6.3 x I0-3

Ground Shine

Vegetable Use
Cow Milk Use

4.0 x 10-1
1.1

1.5
4.7

1.5
4.9

6.8 x 10-2

6.8 x 10-2

Total
1.6
6.3
6.5
aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and the nearest cow
and garden locations are for adults for total-body and for children forbone and liver. The dose estimates for the nearest goat location are for
adults for total-body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-36

Table 8.29. Dose Estimates for the Maximum Exposed Individual
due to Breaching of a Package Containing EPICOR II
Waste--AFHB Water - Mixed Bed Liner
Dose (mrem)a
Location
Nearest b

gardenNearestmilk goat

Nearest cow

and garden

Pathway

Total-Body

Bone

Liver

Inhalation

2.6 x 10-4

2.9 x 10-4

3.5 x 10-4

2.1 x 10-2

3.6 x 10-3

7.6 x 10-2

3.6 x 10-3

Vegetable Use

Ground Shine

3.6 x 10-3

Total

2.5 x 10-2

8.0 x 10-2

8.3 x 10-2

Inhalation
Ground Shine

2.6 x 10-4
3.6 x 10-3

2.9 x 10-4
3.6 x 10-3

3.5 x 10-4
3.6 x 10-3

Goat Milk Use

5.9 x 10-2

3.8 x 10-1

4.9 x 10-1

Total

6.3 x 10-2

3.8 x 10-1

4.9 x 10-1

7.9 x 10-2

Inhalation

1.8 x 10-4

2.0 x 10-4

2.4 x 10-4

Ground Shine
Vegetable Use

3.6 x 10-3
2.1 x 10-2

3.6 x 10-3
7.6-x 10-2

Cow Milk Use

1.4 x 10-2

9.3 x 10-2

3.6 x 10-3
7.9 x 10-2
1.2 x 10-1

Total
3.9 x 10-2
1.7 x 10-1
2.0 x 10-1
afDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and the nearest cow
and garden locations are for adults for total-body and for children forbone and liver. The dose estimates for the nearest goat location are for
adults for total-body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-37

Table 8.30.
Dose Estimates for the Maximum Exposed Individual
due to Breaching of a Package Containing Modified EPICOR II
Waste--Primary Water - Zeolite Liner
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

2.3
3.3
650

36
3.3
2700

4.6 x 10-1
3.3
730

Total

650

2800

730

9.6 x 10-1
3.3
150

15
3.3
810

3.3 x 10-1
3.3
460

Total

150

820

460

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

1.6
3.3
650
42

2.5
3.3
2700
200

3.1 x 10-1
3.3
730
58

Total

700

2900

790

Nearestmilk goat

Nearest cow
and garden

Inhalation
Ground Shine
Goat Milk UseaDoses were calculated for total-body,GI-tract, bone,

liver, kidney,

thyroid, lung, and skin. The maximum three organ doses are listed in thistable.
Doses were calcuated for four age groups:
adults, teenagers,
children, and infants.
The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and the nearest cow
and garden locations are for adults for total-body and for children forbone and liver.
The dose estimates for the nearest goat location are for
adults for total-body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are:nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-38

Table 8.31. Dose Estimates for the Maximum Exposed Individual
due to Breaching of a Package Containing Modified EPICOR II
Waste--Primary Water - Cation Liner

Location

Pathway

Total-Body

Dose (mrem)a
Bone

Nearestgardenb

Inhalation

1.3 x 10-2

2.0 x 10-1

4.9 x 10-3

Ground Shine

6.2 x 10-4

6.2 x 10-4

Nearestmilk goat

Nearest cow
and garden

Liver

Vegetable Use

3.5

14

6.2 x 10-4
5.5 x 10-4

Total

3.5

14

6.1 x 10-3

8.3 x 10-2
6.2 x 10-4
2.2

2.8 x 10-3

Goat Milk Use

5.2 x 10-3
6.2 x 10-4
5.7 x 10-i

7.6 x 10-6

Total

5.8 x 10-1

2.3

3.4 x 10-3
3.4 x i0-3

Inhalation
Ground Shine

Inhalation

8.7 x I0-3

1.4 x 1 0 -i

Ground Shine

6.2 x 10-4

6.2 x 10-4

Vegetable Use
Cow Milk Use

3.5
1.7 x 10-1

14
7.1 x 10-1

3.7

15

Total

6.2 x 10-4

6.2 x 10-4
5.5 x 10-4

2.4 x 10-5
4.6 x 10-3

aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and the nearest cow
and garden locations are for children, the nearest goat locations are for
infants.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-39

Table 8.32.
Dose Estimates for the Maximum Exposed Individual
due to Breaching of a Package Containing Modified EPICOR II
Waste--Primary Water - Mixed Bed Liner
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

5.7 x 10-4
2.7 x 10-s
1.6 x 10-1

9.0 x I0-3
2.7 x 10-5
6.3 x 10-1

2.1 x 10-4
2.7 x 10-5
2.4 x 10-5Total

1.6 x 10-1

6.4 x 10-1

2.6 x 10-4

Inhalation
Ground Shine
Goat Milk Use

2.3 x 10-4
2.7 x 10-5
2.5 x 10-2

3.7 x 10-3
2.7 x 10-5
1.0 x 10-1

1.2 x 10-4
2.7 x 10-5
3.3 x 10-7

Total

215 x 10-2

1.0 X 10-1

1.5 x 10-4

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

3.9 x
2.7 x
1.6 x
8.2 x

10-4
10-5
10-1
10-3

6.2 x 10-3
2.7 x 10-s
6.3 x 10-1
3.2 x 10-2

1.5 x 10-4
2.7 x 10-5
2.4 x 10-5
1.0 x 10-6

Total

1.7 x 10-1

6.7 x 10-1

2.0 x 10-4

Nearestmilk goat

Nearest cow
and garden

aDoses were calculated for total-body, GI-tract, bo)Fne, liver, kidney,
thyroid, lung, and skin. The maximum three organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and the nearest cow
and garden locations are for children, the nearest goat locations are for
infants.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-40
8.1.5.4

Psychological-Socioeconomic

Effects

See Section 8.3.5.4.
8.1.6 Economic Costs
Solid waste processing is broken down into two basic activities--(l) conditioning and (2) waste
packaging and handling.
Conditioning refers to those operations that transform the concentrates
produced during processing of untreated materials into forms suitable for transportation anddisposal.
Packaging refers to placement of the waste material into disposable containers.
Package handling refers to those operations that involve movement of containers in preparation
for storage and shipment.
Because of thenumerous alternatives available for this activity, minimum and maximum costs havebeen developed based on the waste volumes presented in Tables 8.3 and 8.4. Details supportingthese cost estimates are presented in Appendix K. The staff believes that the costs for solid
waste processing lie somewhere between $11,600,000 and $16,000,000 regardless of the alternatives
selected.
8.2

CHEMICAL DECONTAMINATION SOLUTIONS

Chemical decontamination solutions are liquid wastes with high chemical and detergent content
that are generated during "hands-on" decontamination of building and equipment surfaces and,depending on the techniques used, possibly during decontamination of the reactor coolant system.
Section 8.1 discussed evaporation and bituminization of these liquids. This section discusses
direct immobilization without volume reduction.
Characteristics of these wastes; the alternatives considered for their immobilization, packaging, and handling; and the impacts of these
operations are discussed.
8.2.1

Status and Specific Considerations

The status of the chemical decontamination solutions generated through September 22, 1980, andthe practices used to package, handle, and solidify these wastes are summarized in this subsection; the types and amounts of chemical decontamination solutions that could be generated bycontinued decontamination activities also are projected.
8.2.1.1

Efforts to Date

The characteristics and current disposition of chemical decontamination solutions generated duringdecontamination of the AFHB through September 22, 1980, are shown in Table 8.33.
These wastes
were generated by the use of a commercially available chemical decontamination solution and havebeen immobilized in 55-gallon drums using the vinyl ester styrene (VES) process (see Appendix H
for details).
8.2.1.2

Projected Requirements

Since decontamination of the AFHB is not yet complete, more chemical decontamination solution
waste will be generated there, and similar solutions also will be generated during decontamination of reactor building surfaces and equipment (see Appendix G for details).
Decontamination of
the reactor coolant system also could result in generation of chemical decontamination solutions
if a technique other than the CAN DECON process is used. Estimates of the volumes and characteristics of the chemical decontamination solution wastes that could be generated from these sources
are presented in Table 8.34.
As shown, decontamination of the reactor coolant system represents
the largest potential source of these liquid wastes.
8.2.2

Alternative Methods Considered

Immobilization is the only feasible treatment alternative for AFHB and reactor building chemical
decontamination solutions.
The options within this alternative involve the selection of a bindermaterial and the type of disposable container used for packaging.
The binder materials considered include bitumen, cement, and vinyl ester styrene.
A discussion of these immobilization
techniques is presented in Appendix H. The types of containers considered include 55-gallon
drums and larger containers.

8-41
Table 8.33. Characteristics and Disposition of Chemical
Decontamination Solution Wastes Generated
through September 22, 1980
FactorValue

Waste Volume (ft 3 )290 a

Waste treatment process

Solidification
with vinyl ester
styrenePackaging Characteristics

Drum
7.35
118

Type
Volume (ft3)

NumberMaximum curies
Maximum surface radiation level

0.15
180 mR/hr

Disposition of Containers

1972

Number stored onsite
Number shipped offsite

aoriginal liquid volume prior to solidification.
bVolume increase factor of 2.5 (20 gallons).
Source: Memorandum from T.L. Gilbert to W.K. Lehto,
Argonne National Laboratory, Subject: AFHB Sludge
Volumes and Activities, December 9, 1980.

Table 8.34.

Projected Volumes and Characteristics of Chemical
Decontamination Solutions
Range ofSpecific Activity(pCi/mL)

Estimated
Volume(gallons)

Curie
Content

AFHB decontamination

7,000

60

1 to 2

Reactor building decontamination
Reactor coolant asystemdecontamination

40,00010

0.1 to 1

SourceaIf CAN-DECON technique is used,

500,000b
20,000b

this source of waste will not occur.

bMaximum estimate of activity that could be removed.
is a factor of 10 lower.

1 to 10

Maximum estimate

8-42
Because of the relatively large volume of waste that could be generated during reactor coolant
system decontamination, the use of evaporation was considered for treatment of these liquids (seeSection 7.1).
Evaporator bottoms then would be immobilized and packaged as discussed in Section 8.1.2.
If these liquids are not processed through an evaporator, their characteristics
would be similar to AFHB and reactor building chemical decontamination liquids.
Under these
conditions, the alternatives considered in this section for the immobilization and packaging ofthese chemical decontamination solution wastes from reactor coolant system decontamination are
the same as those alternatives considerd for AFHB and reactor building chemical decontamination
liquids--immobilization with either bitumen, cement, or vinyl ester styrene and packaging ineither drums or large containers.
Consideration of the alternative immobilization techniques and disposable containers in conjunction with the projected volumes bounds the quantities of packaged waste that could be generated.
The assumed bounding conditions are:-Maximum Waste Production.
AFHB and reactor building waste volumes plus use of a technique other than the CAN DECON process to decontaminate the reactor coolant system.
All liquids would be immobilized with vinyl ester styrene or cement using maximum
volume increase factors.

-

Minimum Waste Production.
AFHB and reactor building waste volumes plus use of the
CAN DECON technique to decontaminate the reactor coolant system.
Al1 liquids would be
immobilized with vinyl ester styrene or cement using minimum volume increase factors.

The characteristics of the packaged waste generated under these bounding conditions are summarized in Table 8.35.

Table 8.35.

Chemical Decontamination Solution Waste Generation-Minimum and Maximum Cases

Source
AFHB and Reactor Building
Factor

55-Gallon Drumb

50-ft

3

LinercMinimum Generation
Original volume (gallons)

47,00047,000Volume increase factord

1.67

1.67

Number of packages
Average curies/package

1600
.04

220
0.3

Maximum Generation
Original volume (gallons)

47,000
2.5
2,400
.03

Volume increase factord
Number of packages
Average curies/package

Reactor Coolant System
55-Gallon Drum

50-ft 3 Liner

zero

zero

47,000500,000

500,000

2.5

2.5

1.67

350
0.2

25,000
0 .8e

2,380
8.4e

awhere CAN DECON technique is used, this source of waste will not occur.
bUseable drum volume is 50 gallons.
Cuseable liner volume is 350 gallons.
dVolume increase factors shown are bounding conditions for immobilization of these liquids
with either cement or vinyl ester styrene.
eAverages shown are based on estimated maximum inventory of 20,000 Ci in reactor coolant
system--could be a factor of ten lower.

8-43
The characteristics of the projected AFHB and reactor building chemical decontamination solutions
are similar to those generated to date and described in Table 8.32.
Immobilization with vinyl
ester styrene has been successful and the staff knows of no reason to depart from this technique.
The projected volume for these wastes is about ten times greater than the AFHB solutions packaged
thus far, and the container size warrants consideration.
Use of drums will result in relatively
low-radiation-level packages that can be readily handled and disposed of with contaminated trash.
The use of the larger liners will substantially reduce the number of packages to be handled, but
the radiation levels will be higher and will necessitate the use of remote handling techniquesand shielded transport.
If.chemical decontamination solutions are generated from reactor coolant system decontamination
and are not processed through an evaporator, as discussed in Section 8.1, these liquids will be
the largest source of liquid waste to be packaged.
As shown in Table 8.35, the use of drums andlarge liners will result in radiation levels that could require shielded shipment.
Under these
conditions, the use of larger liners is preferred since the number of packages would be minimized.
8.2.3

Details of Methods and Facilities

The facilities currently being used to immobilize AFHB decontamination solutions with vinyl ester
styrene and package these wastes in drums will continue to be used.
Minor modifications would
have to be made to handle large containers, but this can be readily accomplished.
These facilities are not adequate to handle the relatively large volume of primary system decontamination solutions that could be generated.
These wastes could, however, be immobilized andpackaged in the same facility that is being considered for immobilization of process solid wastes.
The characteristics of the packaged waste that would be generated are summarized in Table 8.36.
As shown, AFHB and reactor building decontamination solutions would be immobilized with vinyl
ester styrene and packaged in drums or large containers.
If chemical decontamination solutions
are generated from reactor coolant system decontamination, large containers are the preferred
package.
These wastes could be immobilized with either vinyl ester styrene, cement, or bitumen.
The selected binder material will depend on the immobilization technique chosen.
8.2.4

Effluents and Releases to the Environment

The nature and impacts of releases to the environment that could occur during packaging andhandling of chemical decontamination solution wastes under normal conditions and under abnormal
or accident conditions-are discussed below.
8.2.4.1

Normal Conditions

Generally, airborne effluents arising from waste-packaging operations are vented to the plant
exhaust system.
The radionuclides released during immobilization of chemical decontamination
solutions were estimated using an assumed fractional release rate of 10-s, or 0.001 percent of
the package radionuclide content. These estimates are presented in Table 8.37.
8.2.4.2

Accident Conditions

The accidents postulated during handling and storage consist of a breach of container integrity
as a result of the container's being punctured or dropped during waste-handling operations.
Worst-case conditions arise when waste is packaged in a large liner with maximum curie content
and the breach occurs on the loading dock or some other outdoor location.
The fractional release
rates used to estimate the amounts of radionuclides released in the form of respirable particulates for postulated accidents are as follows:
Liquids immobilized with cement--1O- 5 of the packaged waste volume
Liquids immobilized with vinyl ester styrene--1O- 6 of the packaged waste volume.
The estimated releases for an accident in which thestyrene are shown in Table 8.38.

liquids were immobilized with vinyl ester

8-44

Table 8.36.

Packaged Decontamination Solution Waste--Minimum
And Maximum Cases
Reactor Coolant

Factor

AFHB and Reactor Building

SystemMinimum Generation
Process used

Immobilize with vinyl
ester styrene

Volume increase factor

1.67

Package type
Number of package
Average curie/package
Average package radiation level

50-ft 3 liner

Maximum Generation
Process used

NA

220
0.3
250 mR/hrd

Immobilize with vinyl
ester styrene

Immobilizeb

Volume increase factor
Package type2.555-gallon drum

2.0
50-ft 3 liner

Number of packages
Average curies/package
Average package radiation level

2400
0.03
10 mR/hr

2860
7c
6 R/hrdawhere CAN DECON technique used, this source of waste will not occur.
bEither cement or vinyl ester styrene.
CBased on maximum inventory of 20,000 Ci in reactor coolant system--could be factor often lower.
dAt this radiation level shielded shipment is required.

Table 8.37. Estimated Releases toEnvironment under Normal
Conditions during Immobilization of AFHB and Reactor
Building Decontamination Solutions
Releasesb
Radionuclide(curies)Cs-1375.5 x 10-7

Cs-1348.4 x 10-8

Sr-90

6.5 x 10-8

Sr-89
almmobilization of the curies inreactor coolant system alreadycovered in Table 8.17.
bBased on immobilization of liquids
with total radionuclide inventoryof 70 Ci.

8-45
Table 8.38. Estimated Accident
Releases to Environment from
Breach of Packaged Immobilized AFHB and Reactor
Building Decontamination Solutions
Releasesb
Radionuclide(Ci/package)

Cs-1378.5 x I0-7

Cs-134

1.3 x 10-7

Sr-90

1

x 10- 8

Sr-89
aAccidents assumed to occur outsidebuildings; releases are directly toatmosphere.
b5 0-fta liner containing I Ci immobilized with vinyl ester styrene.

8.2.5

Environmental Impacts

8.2.5.1

Occupational Doses

Occupational radiation doses incurred by workers involved in packaging and handling of chemical
decontamination solution wastes (through the stage of placing the packaged wastes in onsite
storage facilities to await shipment) are summarized in this section. The basis for the estimates are given in Appendix N.
Estimated occupational doses are presented in Table 8.39. As shown, the cumulative dose from
AFHB and reactor building solutions varies from 2.9 to 4.2 person-rem, depending on package type.
These wastes would be packaged and handled by two-person crews, and three crews would be usedover the 18-month period required to perform this activity. The average dose to each crew member
would range from about 0.5 rem to 0.7 rem. The average quarterly dose to each crew member would
be about 0.1 rem. The expected number of additional cancer mortalities in the work force of six
persons exposed to the cumulative dose of radiation would be between 0.0004 and 0.0006. Thismeans that the added probability that the average individual worker would die of cancer would be
between I in 20,000 and I in 10,000. The expected number of additional genetic effects in theoffspring of the work force exposed to the cumulative dose of radiation would be between 0.0008
and 0.001.
If reactor coolant system chemical decontamination solutions are generated, these wastes would bepackaged and handled over a 12-month period. Two crews of two persons each would be used, leading
to an average dose per crew member of 1.5 and a cumulative dose to the group of 6 person-rem.
The average quarterly dose to each crew member would be about 0.4 rem. The expected number of
additional cancer mortalities in the work force of four persons exposed to the cumulative dose ofradiation would be 0.0008. This means that the added probability that the average individual
worker would die of cancer would be 1 in 1200. The expected number of additional genetic effects
in the offspring of the work force exposed to the cumulative dose of radiation would be 0.002.
8.2.5.2

Offsite Doses

The dose estimates presented here for processing of chemical decontamination solutions are basedon the source terms developed in Section 8.2.4. The calculational models used to make theseestimates and the interpretation of their results are described in Appendix W. The significanceof these doses and their human health and environmental consequences are discussed in Section 10.3.
The estimates are listed in Table 8.40. The 50-mile total body population dose received by thehuman population during these activities is estimated to be about 2 x 10-5 person-rem.

8-46
Table 8.39.

Estimated Occupational Doses from Handling and Packaging
Chemical Decontamination Solution Wastes

Unit Dose(person-mrem/package)

Source andPackage Type

Cumulative Occupational
Dose (person-rem)
Best Case
Worst Case

AFHB and reactor building
55-gallon drum
Large liner

3.5
21

NA
2.9

4.2
NA

Reactor coolant system
Large liner

21

NA

6.0

2.9

lob

TotalsaQuantities from Table 8.35.
bTotals rounded to two significant figures.

Table 8.40. Dose Estimates for the Maximum Exposed Individual
due to Immobilizing Chemical Decontamination Solutions

Pathway

Total-Body

Dose (mrem)a
Bone

Nearest b

Inhalation

2.7 x 10-8

3.8 x 10-7

2.9 x 10-8

gardenb

Ground Shine

7.0 x 10-8

7.0 x 10- 8

7.0 x 10-8

Vegetable Use

2.2 x 10-6

9.0 x 1O-6

1.4 x 10-6

Total

2.3 x 10-6

1.0 x 10-5

1.5 x 10-6

Location

Nearestmilk goat

Nearest cow
and garden

Liver

Inhalation

3.4 x 10-8

1.5 x 10-7

1.9 x 10-8

Ground Shine

6.0 x 10-8

6.0 x 10-8

6.0 x 10-8

Goat Milk Use

1.1 x 10-6

8.0 x 10-6

9.0 X 10-6

Total

1.2 x 10-6

8.0 x 10-6

9.0 x 10-6

Inhalation

3.0 x 10-8

4.2 x 10-1

3.2 x 10-8

Ground shine

9.0 x 10-8

9.0 x 10-8

9.0 x 10-8

Vegetable Use

3.2 x 10-6

1.4 x 10-5

2.1 x 10-6

Cow Milk Use

4.0 x 10-7

2.2 x 10-6

1.6 x 10-6

3.8 x 10-6
1.6 x 10-5
3.7 x 10-6
Total
aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and nearest cow/garden
locations are for children, and for the nearest goat location are for
adults for total body and for infants for bone and liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-47
8.2.5.3

Postulated Accident Effects

The type of accident for which dose estimates are made here is the breach of a package containing
waste from AFHB and reactor building decontamination. The accident is described in Section 8.2.4.2
and the source term is listed in Table 8.38. Calculational models used to make these estimates
and the interpretation of their results are described in Appendix W. The significance of thesedoses is discussed in Section 10.4. The dose estimates for the maximum exposed individual due tobreaching a package containing AFHB or reactor building wastes are listed in Table 8.41.

Table 8.41.
Estimates of Offsite Doses to the Maximum Exposed Individual
Caused by Breach of Package Containing AFHB and Reactor
Building Immobilized Decontamination Solution Waste
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

4.8 x 10-7

1.7 x 10-6
5.7 x 10-6
1.9 x 10-4

7.7 x 10-7

5.7 x 10-6
4.0 x 10-5

Total

4.6 x 10-5

2.0 x 10-4

1.4 x 10-4

Inhalation
Ground Shine
Goat Milk Use

4.8 x I0-7
5.7 x 10-6

8.5 x 10-7

9.3 x 10-5

5.7 x 10-6
6.5 x 10-4

5.6 x 10-7
5.7 x 10-6
7.9 x 10-4

Total

9.9 x 10-5

6.6 x 10-4

8.0 x 10-4

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

3.3 X 10-7
5.7 X 10-6
4.0 X 10-5
2.2 X 10-5

1.2 X 10-6
5.7 X 10-6
1.9 X 10-4
1.0 X

5'3 X 10-7
5.7 X 10-6
1.3 X I0-4
9.8 X 10-s

Total

6.8 x 10-5

3.0 x 10-4

2.3 x 10-4

Nearestmil k goat

Nearest cow
and garden

5.7 x 10-6
1.3 x 10-4

aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each group are
listed. The dose estimates for total body exposure are for adults for all
locations.
For the bone and liver doses the estimates are for children
for the nearest garden and nearest cow and garden locations and for infants
for the nearest goat location.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8.2.5.4

Psychological-Socioeconomic

Effects

See Section 8.3.5.4.
8.2.6

Economic Costs

Chemical-based decontamination solutions are broken down into two categories--treated and untreated.
Treated solutions are those which are processed through an evaporator and then immobilized.
Untreated solutions are those which are immoblized directly in vinyl ester styrene or cement and8-48

require appropriate immobilizing facilities.
The costs for packaging and handling these solutions are highly variable because of the uncertainty in the quantities to be handled. The quantities may vary from 47,000 gallons to as much as 550,000 gallons.
The staff believes that the costs for chemical decontamination solution immobilization lie somewhere between $2,000,000 and $13,000,000.
Details supporting these cost estimates are presented
in Appendix K.
8.3

SOLID MATERIALS

Decontamination of the auxiliary and fuel handling buildings (AFHB) began shortly after theaccident, and decontamination of the reactor building is scheduled to begin in early 1983.
These
activities and defueling of the reactor, followed by primary system decontamination, will generate
solid waste materials in the form of trash, contaminated equipment, and irradiated hardware.
The
characteristics of these wastes, the alternatives considered for their -treatment, packaging, andhandling, and the environmental impacts of these operations are discused in this section.
8.3.1

Status and Specific Considerations

The work performed to date in decontamination of the AFHB has not resulted in the generation ofsolid materials other than trash.
The characteristics and disposition of the trash generated by
AFHB decontamination activities through September 22, 1980, are given in Table 8.42.
This trash consisted of compactible and noncompactible solid material, some of which also is
combustible.
The compactible and combustible solids consist of disposable clothing, rags, plastic covers, laydown pads, and miscellaneous trash. The noncompactible solids consist of tools,
hoses, safety goggles, miscellaneous construction materials, and other small items of equipmentused by decontamination personnel.
The compactible trash has been processed through a compactor
and packaged in 55-gallon drums.
This method of treatment reduces trash volume by about a factorof 5.
Noncompactible trash has been packaged in 3-ft x 4-ft x 6.5-ft wooden low-specific-activity(LSA) boxes with a capacity of 80 ft 3 each.
8.3.1.2

Projected Requirements

Additional work in the AFHB and the other decontamination and defueling activities will lead tothe generation of additional trash plus other solid waste materials.
The estimated amounts ofthese other solid wastes that will be generated, their sources, and projected, radioactivity
levels are summarized in Table 8.43.
As shown, relatively large volumes of trash are projected to be generated during decontaminationof the reactor building and during defueling of the core.
Prior to August and September 1980
entries into the reactor building, it was assumed that this trash would have radioactivity levels
about a factor of 10 higher than the material generated during AFHB decontamination (characterized in Table 8.42).
However, the swipe samples taken during these entries did not substantiate
this assumption, and it now appears that this additional trash will have gross radioactivity
levels comparable to AFHB trash.
The gross activities shown in Table 8.43 have been revised from those presented in the draft PEIS
to reflect containment entry data.
The volume of trash that could be generated depends on thenumber of personnel involved in decontamination operations and on the time required to decontaminate areas to acceptable levels.
The minimum and maximum volumes projected in Table 8.43
are bounding conditions for trash generation.
The condition of equipment in the reactor building which could be contaminated cannot be defined
until decontamination begins.
The estimated volumes shown in Table 8.43 were derived by identifying the equipment that could have been contaminated during the accident and then estimating
best- and worst-6ase conditions for the portion of this equipment that would have to be disposedof as waste.
Since the surface contamination level data obtained during reactor building entry
shows levels below those assumed for the draft PEIS, the projected radioactivity levels shown in
Table 8.43 have been reduced from earlier estimates.
The contaminated equipment generated during
defueling will consist of tooling and equipment placed in the spent fuel pool to assist with
defueling; the volumes shown in the table are based on the staff's estimate of needs.

Table 8.42.

Waste

Waste Forma

Volume (ft 3 )Compactible
Noncompactible

Solid Waste Generated through September 22, 1980

VolumeNumberMaximum
CuriesMaximum Surface
Radiation Level

7.35

1961

0.02

20 mR/hr

338

1623

91

234

Neg.

15 mR/hr62

172

Process

Type(ft

72,300Compactionb

Drum

21,300None

LSA BoxaAll solid waste generated through September
bVolume reduction factor of 5.

Disposition

Packaging

Waste
Treatment

22,

3

)

1980, consisted of trash and rubbish.

OnsiteStorage

OffsiteDisposal

Table 8.43.

Solid Waste Projections

AFHB
Volume(ft 3 )

Reactor Building
Cleanup Volume(ft 3 )

Defueling
Volume(ft 3 )

Total
Volume(ft 3 )

Average
Radioactivity(Ci/ft 3 )Estimated
Inventory(Ci)

Compactible trash
Noncompactible trash
Contaminated equipment
Mirror insulation
Irradiated hardware

115,000
49,000NA
NA
NA
164,000

50,00013,000450NA3,200
66,650225,000
82,0003,450
15,0003,200
328,650

0.006
0.001
0.01
0.003
0.01

Total

60,00020,0003,000
15,000NA
98,000

1,350
80
35
60
32
1,550bCompactible trash
Noncompactible trash
Contaminated equipment
Mirror insulation
Irradiated hardware

115,000
49,000
NA
NA
NA

Total

164,000

Waste Type
Minimums

00

Maximums
200,000
54,00024,00015,000NA
293,000

140,000
40,00012,000NA
3,500
195,500

455,000
143,000
36,000.
15,0003,500
652,500

0.006
0.001
0.003
0.003
0.01

2,730
140
110
60
35
3,080baThe estimated curie content of irradiated hardware was assumed to be the same as the highest levelcontaminated equipment.
bRounded to three significant figures.

8-51

The reactor vessel is surrounded by mirror insulation that was wetted during the accident.
Thiscontaminated solid material, which is neither trash or equipment, will have to be removed andpackaged as solid waste.
The estimated volume is based on the volume installed, and the radioactivity level is based on the residual radioactivity in this material after it has been dried.
The condition of the reactor vessel internals, which when removed may be handled as irradiatedhardware, will not be known until the reactor vessel head is removed and the core internals areinspected.
Therefore, the values given in this section for radiation levels for this hardware
areestimates and could vary~by an order of magnitude.
Ithas been assumed that the upper plenum
assembly and the core support structure will have to be removed and disposed of as radioactive
waste.
8.3.2

Alternative Methods Considered

The alternatives considered for treatment,in Table 8.43 are discussed below.
8.3.2.1

packaging,and handling of the solid wastes projected

Trash

Currently, trash is processed through a compactor to achieve a volume reduction factor of about 5.
Volume reduction also could be achieved by burning combustible trash in an incinerator.
A numberof incinerators for radioactive waste have been operated.
This technique would reduce the trash
volume by a factor of 80 to 100; however, immobilization of the resultant ash will result in an
effective volume reduction factor of 40 to 50.
Consideration of these two alternatives and therange of volumes shown in Table 8.43 bound the volumes of trash that could be generated.
These
best- and worst-case conditions for packaged trash are as follows:-

Maximum Waste Package Production.
The maximum volumes from Table 8.43, compaction of compactible trash to reduce volumes by a factor of 5, and packaging of noncompactible trash in
80-ft 3 LSA boxes.

-

Minimum Waste Package Production.
The minimum waste volumes from Table 8.43, incinerationof 75 percent of the compactible trash to reduce effective volume by a factor of 50, Compaction of the remaining 25 percent of the compactible trash to reduce volumes by a factorof 5, and packaging of noncompactible trash in 80-ft 3 LSA boxes.

The amounts of packaged trash that would be generated under these two conditions are summarizedin Table 8.44.

Table 8.44.

Minimum and Maximum Alternatives for Trash Generation
Noncompactible,
CoCompactible,oa
Combustible Trash

Noncombustible Trash

Noncombustible
Trash

Minimum Generation
Original volume (ft 3 )Process used
Volume reduction factor
Package type
Number of packages

168,000
Incineation
50
Drum
480

56,000Compaction
5
Drum
1,600

82,000None
None
LSA box
1,025

Maximum Generation
Original volume (ft 3 )Process used
Volume reduction f:actor
Package type
Number of packages

341,000
Compaction
5
Drum
9,740

114,000
Compaction
5
Drum
3,260

143,000
None
None
LSA box
1,790

FactoraAssumes 25 percent of compactible trash is noncombustible.
bIncinerator ash is immobilized with volume increase factor of 2.

8-52

8.3.2.2

Contaminated Equipment

The decontamination of the reactor building, as discussed in Section 5.2, will require the removal
and disposal of materials and equipment that cannot be decontaminated and refurbished to an
acceptable operational condition.
A partial listing of materials and equipment that may beremoved for packaging and disposal is given in Table 8.45. Defueling and decontamination of the
primary system will generate similar waste materials.
A partial listing of this equipment andmaterials is given in Table 8.46.
The alternatives considered to package and handle these wastes involve volume reduction, but themethod of volume reduction depends on the physical characteristics of the waste. The methods
considered include:
Disassembly - Large equipment items, such as pumps and motors, will be disassembled toreduce bulk and packaged volume.
Sectioning - Large metal structures, such as the polar crane cab and ventilation fans,could be dismantled and cut up, or sectioned.
Baling -Sheet metal and the mirror insulation could be baled in a press after the
havy structural fittings are removed.
Baling will reduce the volume by a factor of 2
to 3.
In general, these techniques will reduce volumes by a factor of 2 but are not suitable for all
equipment.
For example, the reactor coolant pumps present some unique problems if they cannot be
decontaminated in place for return to service. There are four pumps and four motors. Each pumpis 8 ft in diameter, 15 ft long, and weighs about 75,000 pounds. The volume of each, excluding
the casing, is about 750 ft 3 , and the maximum expected radiation level is 5 R/hr. Each pump
motor is 13 ft in diamater, 17 ft long, and weighs about 103,000 pounds. Each has a volume ofabout 2300 ft 3 ; the maximum expected radiation level is 10 mR/hr.
The packaging methods considered depend on the radioactivity level of the waste. Most of thematerial will be of low radioactivity and can be packaged in 80-ft 3 -capacity LSA wooden boxes.
Other material, with radioactivity levels which require shielded transport, would be packaged insteel liners that can be placed in shielded overpacks. These liners would have a waste capacityof 70 ft 3 .

When these alternatives are considered relative to the volumes shown in Table 8.43, bounding
conditions for packaged contaminated equipment can be defined. The bounding conditions, which
represent minimum and maximum volumes of packaged waste, are as follows:-

Maximum Waste Package Production.
The maximum waste volumes and activity levels in
Table 8.43, packaging in LSA boxes or liners, and no volume reduction for equipment or
mirror insulation.

-

Minimum Waste Package Production.
The minimum waste volumes and activity levels in
Table 8.43, decontamination and reuse of the reactor coolant pumps, volume reduction of
50 percent of the remaining equipment by a factor of 2, and volume reduction of mirror
insulation by a factor of 2.5.

The quantities and characteristics of waste packages arising from these bounding conditions areshown in Table 8.47.
8.3.2.3

Irradiated Hardware

The material that would be removed from within and around the reactor vessel as irradiated hardware is characterized in Table 8.48. The alternatives for packaging and handling these wastes
depend on their physical characteristics and radiation levels. Disassembly and sectioning could
reduce volumes by a factor of 2.
Low-activity material could be packaged in 80-ft 3 -capacity LSA
boxes, while higher activity material could be packaged in 70-ft 3 -capacity steel liners compatible
with shipment in a shielded overpack.

8-53

Table 8.45.

Partial Listing of Reactor Building Contaminated Equipment for Disposal
Air Supply System

Ancillary Equipment
Instrument packs
Control stations

Purge system:
Vertical mounted fans

Drain pumps
Letdown coolers

Samplers
Coolers
Air cooler portion of the main system

Crane

Main system ductwork on the 305-ft level

ValvesWire trays
CablesMain system ductwork above the 305-ft levelbut not including that within thesecondary shield system
Main system ductwork below the 305-ft levelbut not including that within thesecondary shield system
Registers,

dampers,

relief valves,etc.

Chiller plenum

Small piping
Instrument piping and connectors
Light fixtures
Mirror insulation
Small tools and fixtures

Fans and coolers
Ductwork within the secondary shield system

Polar Crane
MotorsWiringElevator
Shaft block walls
Drive motors
Electric cables
Switches, relays
Hoist cables
El-evator car

Hoist rope
Contacts and controllers

Fuel Handling Bridges
Steam Generator
Mirror insulation
Cables,wire trays,

conduit

Instrumentation
Main coolant pump motors
Pump coolers
Lubrication pump systems
Electrical cabinet

Bridge assemblyElectricals
Mechanisms
Missile shields

8-54

Table 8.46.

Primary System Contaminated Equipment

Quantity
4

Estimated Surface
Radiation Level
5 R/hr a

Estimated
Volume(ft 3 )
3000

Reactor coolant pump motors

4

10 mR/hr

9200

Underwater vacuum

1

300 mR/hr

30

Underwater cutting machines

2

300 mR/hr

10-

50 mR/hr

100

8000 lb

100 mR/hr

100

Material
Reactor coolant pump impellers

Tools and grapples
Core filter support structureaIn seal area.

Table 8.47.

Contaminated Equipment--Minimum and Maximum Waste Generation

Minimum Generation

Maximum Generation

FactorEquipment

Mirror Insulation

Equipment

Mirror Insulation

Original volumea (ft 3 )Process used

3,450
Sectioning

15,000
Baling36,000None

15,000
None

2

2.5

None

None

Package type

LSA box3

LSA box c

LSA boxNumber of packages

33

450

188

Volume reduction factorb

70-ft

linerd86
3

3

aFrom Table 8.43; minimum equipment consists of 3000 ft plus 450 ft for defueling; maximum
equipment consists of 24000 ft 3 from RB plus 12000 ft 3 from defueling.
bAssumes 50 percent of volume can be sectioned with volume reduction factor of 2.
cLSA box assumed to have 80-ft 3 capacity. If 120-ft 3 -capacity LSA liners are used, number ofpackages is reduced by 33 percent.
dEach liner contains 0.7 Ci; could require shielded shipment.

8-55

Table 8.48.

Reactor Vessel--Irradiated Hardware

Quantity

Estimated
Volume(ft 3 )

Estimated Surface
Radiation Level(mR/hr)16 sections

120

300

Diaphragm over seal plate

1

8

50

Stud and nut fragments

15

24

50

Upper plenum assembly
fragments

150

1200

50

Core support assembly

225

1300

50

Electric cables and coolant
lines

200

300

300

Material
Thermal insulation fromreactor head

Orifice rods
Control rods
Burnable poison rods
Axial power shaping rods

Assumed to be
removed with fuel177

Control & shaping roddrive mechanisms

69

52

50

Control & shaping rodlead screws

69

67

50

Control & shaping rodlead stators

69

100

300

The alternatives considered provide a basis for bounding the quantities of packaged irradiatedhardware that could be generated. The bounding conditions considered were as follows:-

Maximum Packaged Waste.
The items listed in Table 8.48, with radiation levels requiring
shipment in a shielded overpack and no volume reduction.

-

Minimum Packaged Waste.
The items listed in Table 8.48 with the radiation levels shown plus
volume reduction by a factor of 2 through sectioning and disassembly.

The quantities and characteristics of waste packages under these bounding conditions are shown in
Table 8.49.
8.3.3

Details of Methods and Facilities

The methods used to handle and package solid materials will be selected and performed using
procedures and techniques that (1) maintain site personnel exposure at "as low as reasonably
achievable" (ALARA) levels, (2) minimize the volumes of packaged waste, and (3) minimize theprobability of a handling accident which could breach the integrity of the disposable container
in which the waste is packaged.
The methods currently used to package and handle AFHB trashwould be continued.
Drums containing compacted trash have been handled on 16-ft 2 pallets that hold four drums eachand are moved with forklifts. Palletization of drums reduces the handling time and provides astable means of stacking drums in an interim storage area to await shipment. The larger LSA
boxes also have been handled with forklifts.
Palletized drums and LSA boxes are stacked threehigh in an interim storage area to await transfer to a transport vehicle.

8-56

Table 8.49.

Irradiated Hardware--Minimum and Maximum
Waste Generation

Factor

Minimum
Generation

Maximum
Generation

2700

NAb

Dissassembly

NA

2

NA

LSA box

NA

Radiation levels less than
200 mR/hra
3
Original volume (ft )Process used
Volume reduction factor
Package type
Number of packages

31

Radiation levels greater than
200 mR/hra

510

3300

Original volume

None

None

Process used

None

None

Volume reduction factor

None

None

Package type

70-ft

Number of cask shipmentsc

3

15

cask

70-ft 3 cask
105

aAt surface of equipment.
bNA = not applicable.

(For the maximum generation conditions

it is assumed that all irradiated hardware would have radioactivity levels in excess of 200 mR/hr.)
c 7 0-ft3 capacity cask and 50 percent packaging efficiency.

8.3.3.1

Trash

The practices and techniques used to collect, package, and handle trash from decontamination of
the reactor building and from defueling will be essentially the same as those currently used for
AFHB trash.
Personnel anticontamination clothing will be accumulated in the containment service building.
Other dry waste, both compactible and noncompactible, will be accumulated at several locations in
the reactor building.
The accumulated material will be wrapped in polyethylene and removed fromthe reactor building through the equipment hatch.
This bagged waste then will be transferred tothe packaging area.
It will be segregated, i.e., compactible versus noncompactible, and thenpackaged.
Compactible trash will continue to be processed through the existing drum compactor toachieve a volume reduction factor of 5 and placed in 55-gallon drums.
AFHB noncompactible trash
has been packaged in 80-ft 3 -capacity wooden LSA boxes.
In the future, these 80-ft 3 wooden boxes
and 120-ft 3 -capacity metal LSA boxes will be used to package noncompactible trash and contaminatedequipment.
After packaging, materials will be moved to the low-level storage area via forklift
truck to await shipment.
Drums will be moved and stored on four-drum pallets.
The minimum and maximum quantities of trash that could be generated are listed in Table 8.50.
Asshown, incineration of combustible trash could reduce the number of drums containing trash by afactor of 10.
No decision has been made on the use of incineration, and a comparison of thistechnique relative to compaction is summarized in Table 8.51.
As illustrated in this table, bothalternatives have advantages and disadvantages.

Table 8.50.

Packaged Trash Summary

Compactible,
Combustible Trash
Without Incineration
With Incineration

FactorCompactible,
Noncombustible
Trash

Noncompactible,
Noncombustible

56,000Drum
1,600

82,000

Trash

Minimum Case
Original volume (ft 3 )Package type
Number of packages
Average Ci/package
Maximum surface radiationlevel/packagee

168,000
Drum
480

168,000
Drum
4,800

2

0.2

0.2

1.6 R/hr

160 mR/hr

160 mR/hr

LSA box
1,025a0.1
80 mR/hr

Maximum Case

oD

u-I

3

Original volume (ft )341,000

341,000

114,000

143,000

Package type
Number of packages

Drum

Drum
9,740

Drum
3,260

LSA box
1 , 7 9 0a

975

0.1d
0.2
0 . 2cAverage Ci/package
2b
Maximum surface radiation
80 mR/hr
160 mR/hr
160 mR/hr
1.6 R/hrlevel/packagee
3
3
would
be reduced
If 120-ft -capacity LSA liners are used, number of packages
aBased on 80-ft capacity.
by 33 percent.
bFrom Table 8.43, trash contains 0.006 Ci/ft 3 ; at 35 ft 3 /drum compacted, each drum contains 0.2 Ci.
cIncinerator ash drums contain equivalent of 350 ft 3 of trash, or 10 times the amount in drums containing
compacted trash.
dFrom Table 8.43, this trash contains 0.001 Ci/ft 3 ; at 80 ft 3 per LSA box, each box contains about
0.1 Ci.
eRadiation level estimates based on 0.8 R/hr per Ci.

8-58

Table 8.51.

Comparison of Incineration Versus Compaction of Combustible Trash
Minimum
Trash Volumes
Incineration
Compaction

FactorMaximum
Trash Volumes
Incineration
Compaction

Number of drums
Number of shipments
Plant storage space (ft2)a

480

4,800

975

9,740

34
1,440

40
14,400

70
2,900

81
29,000

Disposal site use (ft3)b

7,200

72,00014,600146,000

35

45

67

84

Occupational exposure (person-rem)
Packaging, handling,transportation

1 x I 0 -4 c

3 x 10-4

2 x 10_4C

2
5.6

0
0

2
5.6

0
0

0.43-0.65

0.49-0.61

0.88-1.3

0.98-1.2

1.5 x 10-4

Releases to atmosphere (Ci)Time to implement (years)
Capital cost ($ millions)Operational cost ($ millions)
2

aDrums stacked two high; 3 ft /drum.
b1 5 ft 3 /drum.
cBased on building vent system with conservative HEPA decontamination factor of 101.

8.3.3.2

Contaminated Equipment

If contaminated equipment cannot be decontaminated in-place, it will be disassembled and moved
with forklifts and cranes to a package-preparation area in the containment service building.
Equipment that can be decontaminated using special techniques will be moved to separate decontamination stations.
Equipment that must be disposed of as waste will be separated by size forcompatibility with LSA boxes and radioactivity level.
The smaller equipment that will fit in LSA boxes will be packaged using "hands-on" contact methods.
Some equipment may be wrapped in polyethylene prior to placement in the LSA box or while awaiting
packaging in the laydown area.
Individual boxes will be loaded within shielded enclosures as
required to minimize exposure to site personnel.
Administrative controls also will be placed on
the surface dose for individual boxes.
Larger items of equipment and materials that can be readily reduced in volume will also be handled
and packaged using "hands-on" methods such as use of power saws.
The volume reduction techniques described in Section 8.3.2.2, (disassembly,
baling) will be used to the extent practicable on a case-by-case basis.

sectioning,andFor large items like the reactor coolant pumps and motors, emphasis will be on in-place decontamination.
If in-place decontamination is not successful, these units will be disassembled to theextent practicable and removed from the reactor building through the equipment hatch to an onsite
decontamination area.
If further decontamination is not possible, they will be packaged in
special containers for offsite shipment and disposal.
Large crates would be built for equipment that could not be readily sectioned for placement in
LSA boxes.
After packaging, the LSA boxes would be transferred to storage to await shipment.
Large crates would be shielded and left in place to await shipment.
The minimum and maximum quantities of packaged contaminated equipment that could be generated areshown in Table 8.52.
In this table, the minimum and maximum unpackaged volumes given in
Table 8.43 are used, and it has been assumed that 50 percent of this equipment is compatible withtechniques that will reduce packaged volume by a factor of two.

8-59

Table 8.52.

Packaged Contaminated Equipment Summary

FactorIntactEquipment

Sectioned
Equipmenta

MirrorInsulation

1,725

1,725

15,000c33Minimum Generation
Original Volume (ft3)

Package type

70-ft

liner

70-ft

liner

70-ft 3 liner

Number of packages

25

13

86

Average Ci/package

0.7

1.3

0.7

500 mR/hr

1000 mR/hrd

500 mR/hrd

18,00018,00015,000

LSA boxb

LSA box

LSA box

225

113

188

Package surface
radiation leveld

Maximum Generation
Original volume (ft3)

Package type
Number of packages
Average Ci/package
Package surface
radiation level

0.2

0.4

0.3

160 mR/hr

320 mR/hr

240 mR/hr

If 120-ft 3 -capacity

LSA liners

aSectioning reduces volume by factor of 2.
bLSA box assumed to have 80-ft 3 capacity.

are used, number of packages is reduced by 33 percent and curie content andradiation level are increased by 50 percent.
cMinimum generation based on baling with volume reduction factor of 2.5.
dPackages with these radiation levels could require shielded shipment.

8.3.3.3

Irradiated Hardware

The irradiated hardware listed in Table 8.48 could be packaged in steel containers within thereactor building and transferred through the equipment hatch.
The characteristics of some ofthis irradiated hardware warrant special handling considerations as discussed below.
Control Rods and Control Rod Lead Screws

The exact extent of core damage is unknown, but some of these components may be fused to fuelassemblies to the extent they cannot be readily separated. Such components may have to be handled
together with the damaged fuel assembly to which they are attached.
Control Rod Drive Mechanisms
Each control rod drive mechanism is about 4½ inches in diameter and 17 ft long, and the stator
sections are 9½ inches in diameter and 21 inches long. These drive mechanisms are normally
attached to the reactor pressure vessel head. Because of the expected high radiation dose rates
from internal surfaces of these mechanisms, the whole assembly motor and pressure boundary will
be cut off from each mechanism and handled underwater or dismantled and packaged for shipment if
radiation levels permit.
The other items of irradiated hardware are expected to have relatively low radiation levels,
below 1 R/hr.
This hardware will be sectioned in the containment service building and packaged
in LSA boxes or in containers compatible with shielded shipment as required.
The bounding conditions shown in Table 8.49 are the staff's estimate of the best- and worst-caserequirements for packaged irradiated hardware.

8-60

8.3.4

Effluents and Releases to the Environment

The nature and impacts of releases to the environment that could occur during solid waste packaging and handling under normal conditions and under abnormal conditions or accidents are discussed
below.
8.3.4.1

Normal Operations

Generally, airborne effluents arising from waste packaging operations are vented to the plant
exhaust system.
Under normal conditions, the operations involved in packaging and handling ofcontaminated equipment and irradiated hardware would not result in effluent releases to theenvironment.
Compaction of trash could result in release of radioactive effluents to the plant
exhaust system, and the releases that could arise from these operations are discussed below.
Trash Compaction
The radionuclide content of trash consists of surface contamination, and a portion of this material in the form of particulates could be released to the plant vent system during compaction.
The gross activity of each drum is conservatively estimated to be about 0.2 Ci.
The amounts(curies) of major radionuclides that could be released during compaction of trash for a fractional
release of 10-4, or 0.01 percent of drum radionuclide to the building atmosphere content, areshown in Table 8.53.

Table 8.53.

Estimated Radionuclide Releases to Building
Atmosphere during Trash Compaction

Amount Released to Building
Atmosphere (Ci/drum)

Total Amount Released
to Building Atmosphere (Ci)aRadionuclide

AFHB Trash

Reactor Building
Trash

Cs-137

1.6 x I0-5

1.8 x 10-5

Cs-134

1.6 x 10-6

5.3 x 10-2
9.2 x 10-3

1.7 x 10-1

2.8 x 10-667

3

1.9 x 10- 3

Sr-90
Sr-89

1.2 x 104.0 x 10-7

2.0 x 10-

AFHB Trash

3.9 x 101.3 x 10-3

Reactor BuildingTrash
1.6 x 10-2

aReleases before HEPA filters. The HEPA filter on the building exhaustsystem would also reduce these values by a factor of 103.

Trash Incineration
If an incinerator is used for combustible trash, the effluents released during normal operations
depend on the method of combustion, the combustion chamber design, and the operational characteristics of the off-gas cleanup system. For well-designed commercial incinerators, up to 98 percent of the ash containing nonvolatile radionuclides can be retained in the combustion chamber.
The off-gas cleanup system typically consists of a wet scrubber, followed by a HEPA filter,
followed by a volatile radionuclide absorption system. This cleanup train typically has an
off-gas decontamination

factor of 106 to 107.

To quantify reactor building trash incinerator

particulate effluents, it is assumed by the staff that 85 percent of the nonvolatile radionuclides would be retained in the ash in the combustion chamber and that the off-gas cleanup system
had a decontamination factor of 106.
Based on these assumptions, the estimated radionuclide
releases per hour of incinerator operation with a 33-ft 3 -per-hour trash feed are presented inTable 8.54. The amounts shown would be released directly to the environment.

8-61

Table 8.54. Estimated Radionuclide Releases
to the Environment during Trash Incineration

Radionuclide
Cs-137
Cs-134

Amount Released(Ci/hr)Total AmountReleased (Ci)a

2.3 x 10-8

2.4 x 10-4

4.2 x 10-9

4.3 x 10-s

Sr-90
1.8 x 10-9
1.9 x 10-5
Sr-89
5.9 x 10-lo
6.1 x 10-s
aBased on the fractional release of a maximum
341,000 ft 3 of combustible trash from Table 8.50,
using a 33-ft 3 -per-hour trash feed.

8.3.4.2

Accident Conditions

The accidents postulated during packaged-waste handling and storage can be divided into two
categories:(1) storage-area fires and (2) breach of container integrity as a result of thecontainer's being punctured or dropped during waste handling operations. The accident conditions
postulated and their consequences in terms of effluent releases are discussed below.
Storage-Area FireA worst-case accident would arise from a fire in the storage area for waste with low levels ofradioactivity. The conservative conditions postulated for this accident are as follows:
The low-level packaged waste in the storage area consists of 200 drums containing
compacted trash and 40 LSA boxes of uncompacted trash.
The fractional radionuclide release rate, in the form of respirable ash, for the combustible trash in the storage area is 0.1 volume percent.
The principal radionuclide releases arising from this postulated accident are given in Table 8.55.
The gross radioactivity in compacted trash drums and LSA boxes was assumed to be 0.2 and 0.1 Ci,
respectively.

Table 8.55.
Estimated Radionuclide Releases
to the Environment during Low-Level
Waste Storage Area FireRadionuclide

Fraction (%)
in Storage Area

AmountReleased (Ci)x 10- 2

Cs-137

78

3

Cs-13414

6.2 x 10-3

Sr-90

6

2.6 x 10-3

Sr-89

2

9

x 10-4

8-62

Packaging/Handlinq Accidents
The consequences of these accidents would depend on the waste type and its condition, the radioactivity and radionuclide content of the breached container, the fractional release of materials
in the container, the interaction with other containers, and the area in the plant where theaccident occurred.
The worst-case conditions considered for each type of packaged solid are summarized in Table 8.56.
To estimate the amounts of radionuclides released in the form of respirable particulates for eachof these accidents, it was assumed that I0-3 of the packaged waste volume would be released fortrash and mirror insulation packages.
The fractional release rate assumed for immobilized incinerator ash was 10-s of the package contents.
The estimated releases for each of the accidents
shown in Table 8.56 are presented in Table 8.57.
The releases shown are based on worst-case(maximum) radionuclide content for each package type.

Table 8.56.
Postulated Worst-Case Package Handling
Accidents Involving Solid Waste
Waste TypeContainer

Accident

LSA compactible
trash

Drum

Forklift penetration

Incinerator ash

Drum

Drop from crane ormonorail

LSA Noncompactible
trash

LSA boxForklift penetration

Mirror insulation

70-ft 3 liner

Drop from crane ormonorailaAll accidents assumed to occur on the loading dock, withreleases directly to the environs.

Table 8.57.

Estimated Accident Releases to the Environment

Amount Released (Ci)Radionuclide

Compactiyle
Trash

Noncompact~ble
Trash

MirrorInsulation

Cs-137

1.6 x 10-4

7.8 x 10-5

2.3 x 10-4

1.6 x 10-5

Cs-134
Sr-90

2.8 x 10-s
1.2 x 10-s

1.4 x 10-5
x 10-66

4.2 x 10-s
1.8 x 10-s

2.8 x 10-6
1.2 x 10-6

Sr-89

4

x 10-6ao. 2 Ci/drum.
b0.1 Ci/LSA box.
c 0 . 3 Ci/70-ft 3 liner.
d2 Ci/drum.

2

x 10-6

6

x 10-6

IncineatorAsh4

X 10-7

8-63

8.3.5

Environmental Impacts

8.3.5.1

Occupational Doses

Occupational doses for waste handling and packaging, through placement of packaged waste in
onsite storage facilities to await shipment, are summarized in this section. Based on information in Appendix N, doses were projected for best- and worst-case conditions based on minimum andmaximum waste generation volumes.
These exposure estimates for AFHB, reactor building, anddefueling waste are presented in Tables 8.58, 8.59, and 8.60, respectively, and are summarized inTable 8.61.
AFHB Wastes
AFHB waste (see Table 8.58) would be packaged and handled by two-person crews, and four crews
would be used over the 15-month period required to perform this activity. The average exposure
to each crew member over a 15-month period is estimated to be 2.8 rem, or an average of 0.6 rem
per quarter.
If the work were conducted over an 18-month period, the average dose would be
0.5 rem per crew member per quarter.
The expected number of additional cancer mortalities in the work force of eight persons receiving
this cumulative dose of radiation would be 0.003. This means that the added probability that the
average individual worker would die of cancer would be 1 in 3000. The expected number of additional genetic effects in the offspring of the work force receiving to this cumulative dose ofradiation would be 0.005.
Reactor Building Wastes
Reactor building wastes (see Table 8.59) would be packaged and handled by two-member crews, andfour crews would be used over the period required to complete this activity. Under minimum waste
generation conditions, this activity could be performed over an 18-month period, and crew members
would receive an average radiation dose of about 1.6 rem, or about 0.3 rem per quarter. Undermaximum waste generation conditions, this activity could be performed over a 30-month period, andeach crew member would receive a total of about 5 rem, or 0.5 rem per quarter.
For the cumulative occupational dose range of 12 to 43 person-rem, the expected number of additional cancer mortalities in the work force of eight people ranges between 0.002 and 0.006. Thismeans that the added probability that the average individual worker would die of cancer ranges
from 1 in 5000 to 1 in 1500. The expected number of additional genetic effects in the offspringof the work force exposed to this cumulative dose of radiation ranges between 0.003 and 0.01.
Defueling Wastes
Defueling wastes (Table 8.60) would be packaged and handled by two-member crews, and three crews
would be used over an 18- to 36-month period. Under minimum waste generation conditions, theactivity could be completed in 18 months, and each crew member would receive a total dose ofabout 1.4 rem, or about 0.24 rem per quarter. Under maximum waste generation conditions, theactivity could be completed in 30 months, and each crew member would receive a total of about
4.7 rem, or about 0.47 rem per quarter.
Based on a maximum cumulative dose of 30 person-rem for handling and packaging of solid wastes,
the expected number of additional cancer mortalities in the work force exposed to this maximum
cumulative dose of radiation would be 0.004. The added probability that the average individual
worker would die of cancer would be 1 in 1500. The expected number of additional genetic effects
in the offspring of the work force exposed to this maximum cumulative dose of radiation would be
0.008.

8-64

Table 8.58.

Estimated Occupational Radiation Doses from Handling and Packagingof AFHB Solid Wastes
Unit Dose

Waste Form

PackageType(person-mrem/
package)

Cumulative Occupational
Dose (person-rem)

Trash--without incinerationa
Compactible
Noncompactible

Drum
LSA box

3.5
21

11.5
13
25b

Drum
Drum
LSA box

17
3.5
21

4.2
2.8
13

Total
Trash--with incinerationa
Combustible
Compactible
Noncompactible
Total

20

aAlternatives being considered.
bRounded to two significant figures.

Table 8.59.

Estimated Occupational Radiation Doses from Handling and Packaging ofReactor Building Solid Wastes

Waste Form

Package
Type

Unit Dose(person-mrem/
package)

Trash--without incinerationa
Compactible
Noncompactible

Drum
LSA box3.5
21

Subtotal
Trash--with incinerationa
Combustible
Compactible
Noncompactible
Subtotal
Contaminated equipment
Mirror insulation
Low activity
High-specific activityc

Cumulative Occupational
Dose (person-rem)
Best Case

Worst Case

6
5.2

20
14

1 1b

34

Drum
Drum
LSA box

17
3.5
21

2.2
1.5
5.2
8.9

7.5
5
14
27b

LSA box

21

0.8

4.7

LSA boxLiners21
22

2.2
12 - 14 d

3.9

TotalsbaAlternatives being considered.

bRounded to two significant figures.
cSame original volume reduced by compaction, which increases specific activity.
dRange reflects effect of incineration.

36-43d8-65

Table 8.60.
Estimated Occupational Radiation Doses from Handling and Packagingof Solid Wastes from Defueling and Primary System Decontamination

Waste Form

Package
Type

Unit Dose(person-mrem/
package)Drum
LSA box3.5
21

Cumulative Occupational
Dose (person-rem)
Best Case
Worst Case

Trash--without incinerationa
Compactible
Noncompactible
Subtotal

5
3.5

14
11

8.5

25

1.8
1.2
3.5

6.1
3.5
11

6.5

20

Trash--with incinerationa
Combustible
Compactible
Noncompactible

Drum
Drum
LSA box

17
3.5
21

Subtotal
Contaminated equipment

LSA box

21

0.1

3.1

LSA boxLiners

21
22

0.6
0.3

2.3

Irradiated hardware
Low activity
High-specific activity

7 .5-9.5c

Totals b
aAlternatives being considered.
bRounded to two significant figures.
cRange reflects effect of incineration.

Table 8.61.
Summary of Estimated Occupational
Radiation Doses from Handling andPackaging of Solid Wastes

Waste Form

Cumulative Occupational
Dose (person-rem)
Best Case
Worst Case

Trash--without incineration
Trash--with incineration
Contaminated equipment

45.0
35.0
1.0

84.0
67.0
9.0

Mirror insulation

2.2

3.9

Irradiated hardware

1.0

2.3

82- 99b
39- 49b
Total
aCombines estimates from Tables 8.59, 8.60, and 8.61.

bRange in totals reflects impact of trash incineration,rounded to two significant figures.

26-30c

8-66

8.3.5.2

Offsite Doses

The dose estimates presented here are for normal releases from trash compaction and incineration.
The source terms are presented in Section 8.3.4.1, Tables 8.53 and 8.54.
The calculationalmodels used to make these estimates and the interpretation of their results are described inAppendix W. The significance of these doses and their human health and environmental consequences
are discussed in Section 10.3.
The dose estimates to the maximum exposed individual for trashcompaction of AFHB wastes are listed in Table 8.62, and the 50-mile population dose was estimatedto be 1 x 10-3 person-rem. The dose estimates to the maximum exposed individual for trash compaction of reactor building wastes are listed in Table 8.63, and the 50-mile population dose wasestimated to be 2 x 10-9 person-rem. The dose estimates to the maximum exposed individual for
trash incineration are listed in Table 8.64, and the 50-mile population dose was estimated to be
6 x 10-3 person-rem.
8.3.5.3

Postulated Accident Effects

The types of accidents for which dose estimates are made here are the following: (1) fire in
low-level storage area; (2) breach of a package containing compactable trash; (3) breach of a
package containing noncompactible trash; (4) breach of a package containing mirror insulation;and (5) breach of a package containing incinerator ash. The accidents are described in Section 8.3.4.2 and their source terms are listed in Tables 8.55 and 8.57. The calculational models
used to make these estimates and the interpretation of their results are described in Appendix W.
The significance of these doses is discussed in Section 10.4. The dose estimates for the maximum
exposed individual are in Table 8.65 for the low-level trash fire, Table 8.66 for the breach of a
package containing compactable trash, Table 8.67 for the breach of a package containing noncompactable trash, Table 8.68 for the breach of a package containing mirror insulation trash, andTable 8.69 for the breach of a package containing incinerator ash.
8.3.5.4

Psychological-Socioeconomic Effects

Decontamination of the AFHB and reactor building, primary system processing, reactor defueling,
and primary system decontamination involve activities that generate solid waste. Such waste canbe classified as either solid materials (trash, contaminated equipment, irradiated hardware, anddamaged fuel assemblies) or process solids (exchange resins, accident sludge, and evaporator
bottoms).
The staff concludes that although airborne releases will pose a negligible health threat toindividuals living in the vicinity of TMI, the packaging and handling of solid waste could arouse
additional distress for some members of the public because of uncertainties surrounding theultimate disposal of the waste.
For local people, this concern is focused on the possibility ofusing the station site as a long-term storage facility and is a factor exacerbating existing
uncertainty and anxiety.
Although they would have negligible offsite health consequences, the accidents depicted in the
scenarios considered by the staff would aggravate existing public uncertainty, especially in
local communities. The level and duration of psychological distress would be associated with thecharacter of the initiating threat, the level of controversy, and/or the type and length of media
coverage.
8.3.6

Economic Costs

Solid waste materials consist of trash, contaminated equipment, and irradiated hardware. The
costs associated with managing these waste forms consist of those for treatment, conditioning,and packaging and handling.
Not all waste forms require consideration of all three areas ofcost. As indicated previously, some waste materials can be packaged directly without treatment
or conditioning and therefore can be processed with less expense. In other cases, both treatment
and conditioning are required and higher costs result. Using this basis, the staff has developed
bounding costs for each of the three waste categories as shown in Table 8.70. Details supportingthese cost estimates are presented in Appendix K.

8-67

Dose Estimates for the Maximum Exposed Individual
for Trash Compaction of AFHB Wastes

Table 8.62.

Location
Nearestgardenb

Nearestmilk goat

Nearest cow
and garden

Pathway

Total-Body

Dose (mrem)a
Bone

Liver

Inhalation

1.8 x 10-6

2.4 x 10-5

6.2 x 10-6
1.4 x 10-4

6.2 x 10-6
6.0 x 10-4

2.8 x 10-6

Ground Shine
Vegetable Use
Total

1.5 x 10-4

6.3 x 10-4

1.5 x 10-4

Inhalation

2.6 x 10-6

Ground Shine

6.0 x 10-6

Goat Milk Use

1.0 x 10-4

9.3 x 10-6
6.0 x 10-6
7.3 x 10-4

1.9 x 10-6
6.0 x 10-6
8.2 x 10-4

Total

1.1 x 10-4

7.5 x 10-4

8.3 x 10-4

Inhalation
Ground Shine

2.0 x 10-6
9.2 x 10-6

2.6 x 10-s
9.2 x 10-6

3.1 x 10-6
9.2 x 10-6

Vegetable Use
Cow Milk Use

2.1 x 10-4
3.4 x 10-5

8.9 x 10-4
1.9 x 10-4

2.0 x 10-4
1.6 x 10-4

Total

2.6 x 10-4

1.1 x 10-3

3.7 x 10-4

6.2 x 10-6
1.4 x 10-4

aDoses were calculated for total-body, GI-tract, bone, liver-, kidney,
thyroid, lung, and skin. The maximum three-organ doses are listed in thistable. Doses were calcuated for four age groups:adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and for the nearest cow
and garden locations are for children. The dose estimates for the nearest
goat location are for adults for total-body and for infants for bone andliver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-68

Table 8.63.
Dose Estimates for the Maximum Exposed Individual
for Trash Compaction of Reactor Building Wastes

Pathway

Total-Body

Dose (mrem)a
Bone

Inhalation

5.1 x 10-6

1.9 x 10-s

GroundShine
Vegetable Use

1.7
1.3 xx 10-s
10-4

1.7 x 10-5
6.4 x 10-4

8.4 x 10-6
1.7 x 10-4
4.1 x 10-5Total

1.5 x 10-4

6.8 x 10-4

4.4 x 10-4

Inhalation

4.6 x 10-6

8.6 x 10-6

5.6 x 10-6

Ground Shine
Goat Milk Use

1.7 x 10-5
2.8 x 10-4

1.7 x 10-5
2.0 x 10-3

1.7 x 10-5
2.4 x 10-3

Total

3.0 x 10-4

2.0 x 10-3

2.4 x 10-3

Inhalation

5.7 x 10-6

Ground Shine

2.5 x 10-5

.2.5 x 10-5

Vegetable Use
Cow Milk Use

1.9 x 10-4
1.1 x 10-4

2.1 x 10-5
2.5 x 10-5
9.4 x 10-4
4.8 x 10-4

Location
Nearest b

gardenNearestmilk goat

Nearest cow

and garden

Liver

9.2 x 10-6
6.1 x 10-4
4.7 x 10-4

Total
3.3 x 10-4
1.5 x 10-3
1.1 x 10-3
aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three-organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers.,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and for the nearest cow
and garden locations are for children. The dose estimates for the nearest
milk goat location are for adults for total-body and for infants for boneand liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-69

Table 8.64.

Dose Estimates for the Maximum Exposed Individualfor Releases Made due to Trash Incineration

Location

Pathway

Total-Body

Dose (mrem)a
Bone

Liver

Nearest b
gardenInhalation
Ground Shine

8.7 x 10-6
2:9 x 10-5

1.2 x 10-4
2.9 x 10-5

Vegetable Use

6.8 x 10-4

2.9 x 10-3

1.3 x 10-5
2.9 x 10-5

Total

7.2 x 10-4

3.0 x 10-3

6.7 x 10-4

Inhalation
Ground Shine

1.2 x 10-5
2.7 x 10-5

4.5 x 10-s
2.7 x 10-5

8.6 x 10-6
2.7 x 10-5

Goat Milk Use

4.8 x 10-4

3.4 x 10-3

3.7 x 10-3

Total

5.2 x 10-4

3.5 x 10-3

3.7 x 10-3

Nearestmilk goat

Nearest cow

and garden

6.3 x 10-4

Inhalation

9.6 x 10-6

1.3 x 10-4

1.4 x 10-s

Ground Shine

4.2 x 10-5

4.2 x 10-5

Vegetable Use
Cow Milk Use

9.9 x 10-4
1.6 x 10-4

4.3 x 10-3
8.7 x 10-4

4.2 x 10-5

Total

1.2 x 10-3

5.3 x 10-3

9.3 x 10-4
7.2 x 10-4
1.7 x 10-3

aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three-organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and for the nearest cow
and garden locations are for children. The dose estimates for the nearest
milk goat location are for adults for total-body and for infants for boneand liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-70

Table 8.65.

Dose Estimates to the Maximum Exposed Individual
Caused by a Low-Level Storage Area Fire
Dose (mrem)a

Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

2.0 x 10-2
2.2 x 10-1
5.4

2.7 x 10-1
2.2 x 10-1
24

2.9 x 10-2
2.2 x 10-1
4.7

5.6

24

4.9

3.1 x 10-2
2.2 x 10-1
3.8

1.2 x 10-1
2.2 x 10-127

2.1 x 10-2
2.2 x 10-1
30

4.1

27

30

1.4 x 10-2
2.2 x 10-'
5.4
8.2 x 10-1

1.9 x 10-1
2.2 x 10-1
24
4.4

2.0 x 10-2
2.2 x 10-1
4.7
3.6

6.5

29

8.5

Total
Nearestmilk goat

Inhalation
Ground Shine
Goat Milk Use
Total

Nearest cow
and garden

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use
Total

aDoses were calculated for total-body, GI-tract, bone,

liver,

kidney,thyroid, lung, and skin. The maximum three-organ doses are listed in thistable.
Doses were calcuated for four age groups: adults, teenagers,
children, and infants.
The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and for the nearest cow
and garden locations are for children.
The dose estimates for the nearest
milk goat location are for adults for total-body and for infants for boneand liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are:nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-71

Table 8.66.
Dose Estimates to the Maximum Exposed Individual
Caused by Breaching a Package Containing Compactible Trash
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

9.6 x 10-5
1.1 x 10-3
2.5 x 10-2

1.3 x 10-3
1.1 x 10-3
1.1 x 10-1

1.5 x 10-4
1.1 x 10-3
2.5 x 10-2Total

2.6 x 10-2

1.1 x 10-1

2.6 x 10-2

Inhalation
Ground Shine
Goat Milk Use

1.5 x 10-4
1.1 x 10-3
1.9 x 10-2

5.4 x 10-4
1.1 X 10-3
1.3 x 10-1

1.1 x 10-4
1.1 X 10-3
1.5 x 10-1

Total

2.0 x 10-2

1.3 x 10-1

1.5 x 10-1

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

6.6 x 10-5
1.1 x 10-3
2.5 x 10-2
4.1 x 10-3

8.7 x
1.1 x
1.1 x
2.2 x

10-3
10-1
10-2

1.0 x 10-4
1.1 x 10-3
2.5 x 10-2
1.9 x 10-2

Total

3.0 x 10-2

1.3 x 10-1

4.5 x 10-2

Nearestmilk goat

Nearest cow
and garden

10-4

aDoses were calculated for total-body, GI-tract, bone , liver, kidney,
thyroid, lung, and skin.
The maximum three-organ doses are listed in thistable.
Doses were calcuated for four age groups:
adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed.
The dose estimates for the nearest garden and for the nearest cow
and garden locations are for children.
The dose estimates for the nearest
milk goat location are for adults for total-body and for infants for boneand liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are:nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-72

Table 8.67. Dose Estimates to the Maximum Exposed Individual
Caused by Breaching a Package Containing Noncompactible Trash

Location

Pathway

Total-Body

Dose (mrem)a
Bone

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

4.8 x 10-s
5.4 x 10-4
1.3 x 10-2

6.3 x 10-4
5.4 x 10-4
5.5 x 10-2

7.2 x 10-5
5.4 x 10-4
1.3 x 10-2

Total

1.4 x 10-2

5.5 x 10-2

.1.3

Inhalation

7.4 x 10-5

2.7 x 10-4

5.4 x 10-4
9.5 x 10-3

5.4 x 10-4
6.6 x 10-2

5.3 x 10-5

Ground Shine
Goat Milk Use
Total

1.0 x 10-2

6.7 x 10-2

7.5 x 10-2

Inhalation

3.3 x 10-5
5.4 x 10-4
1.3 x 10-2
2.1 x 10-3

4.4 x 10-4

Ground Shine
Vegetable Use
Cow Milk Use
Total

1.6 x 10-2

Nearestmilk goat

Nearest cow
and garden

Liverx 10-25.4 x 10-4
7.4 x 10-2

5.0 x 10-5

5.4 x 10-4
5.4 x 10-2
1.1 x 10-2

.5.4 x 10-4
1.2 x 10-2
9.3 x 10-3

6.6 x 10-2

2.2 x 10-2

aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three-organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants.
The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and for the nearest cow
and garden locations are for children. The dose estimates for the nearest
milk goat location are for adults for total-body and for infants for boneand liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are:nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-73

Table 8.68.
Dose Estimates to the Maximum Exposed Individual
Caused by Breaching a Package Containing
Trash Mirror Insulation
Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearest b
gardenInhalation
Ground Shine
Vegetable Use

1.4 x 10-4
1.6 x 10-3
3.8 x 10-2

1.9 x 10-3
1.6 x 10-3
1.6 x 10-1

2.1 x 10-4
1.6 x 10-3
3.6 x 10-2

Total.

4.0 x 10-2

1.6 x 10-1

3.8 x 10-2

Inhalation
Ground Shine
Goat Milk Use

2.2 x 10-4
1.6 x 10-2
2.8 x 10-2

8.1 x 10-4
1.6 x 10-3
1.9 x 10-1

1.6 x 10-4
1.6 x 10-3
2.2 x 10-1

Total

3.0 x 10-2

1.9 x 10-1

2.2 x 10-1

Inhalation
Ground Shine
Vegetable Use
Cow Milk Use

9.8 x
1.6 x
3.8 x
6.0 x

10-5
10-3
10-2
10-3

1.3 x 10-3
1.6 x 10-3
1.6 x 10-'
3.3 x 10-2

1.5 x 10-4
1.6 x 10-3
3.6 x 10-2
2.7 x 10-2

Total

4.6 x 10-2

2.0 x 10-1

6.5 x 10-2

Nearestmilk goat

Nearest cow
and garden

aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three-organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and for the nearest cow
and garden locations are for children. The dose estimates for the nearest
milk goat location are for adults for total-body and for infants for boneand liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-74

Table 8.69.

Dose Estimates for the Maximum Exposed Individual
Caused by Breaching a Package Containing Incinerator Ash

Dose (mrem)a
Location

Pathway

Total-Body

Bone

Liver

Nearestgardenb

Inhalation
Ground Shine
Vegetable Use

9.6 x 10-6
1.1 x 10-4
2.5 x 10-3

1.3 x 10-4
1.1 x 10-4
1.1 x 10-2

1.5 x 10-5
1.1 x 10-4
2.5 x 10-3

Total

2.6 x 10-3

1.1 X 10-2

2.6 x 10-3

Nearestmilk goat

Nearest cow
and garden

Inhalation

1.5 x I0-5

5.4 x 10-s

1.1 x 10-4
1.9 x 10-3

1.1 x I0-4
1.3 x 10-2

1.1 x 10-5

Ground Shine
Goat Milk Use
Total

2.0 x 10-3

1.3 x 10-2

1.5 x 10-2

8.7 x 10-5

1.0 x 10-5

4.5 x 10-3

Inhalation

6.6 x 10-6

Ground Shine
Vegetable Use
Cow Milk Use

1.1 x 10-4

2.5 x 10-3
4.1 x 10-4

1.1 x 10-4
1.1 X 10-2
2.2 x 10-3

Total

3.0 x 10-3

1.3 x 10-2

1.1 x 10-4
1.5 x 10-2

1.1 x 10-4
2.5 x 10-3
1.9 x 10-3

aDoses were calculated for total-body, GI-tract, bone, liver, kidney,
thyroid, lung, and skin. The maximum three-organ doses are listed in thistable. Doses were calcuated for four age groups: adults, teenagers,
children, and infants. The highest dose estimates for each age group are
listed. The dose estimates for the nearest garden and for the nearest cow
and garden locations are for children. The dose estimates for the nearest
milk goat location are for adults for total-body and for infants for boneand liver.
bThe basis for selecting the special locations is described in Appendix W.
The actual locations are: nearest garden = 1.05 miles east-northeast,nearest milk goat = 1.02 miles north, and nearest cow and garden= 1.05 miles east.

8-75
Table 8.70.

Cost Estimates for Solid Materials Processing(thousands of dollars)

Waste TypeBest Case

Worst Case

Trash
Contaminated equipment

650a

6900b

50

150

Irradiated hardware

20

80

720

7130

Total

alncludes major capital cost for compactor of $180,000.
blncludes major capital cost for compaction and incineration
facilities of $5,780,000.

8.4

FUEL ASSEMBLIES AND CORE DEBRISAs the intact and damaged fuel assemblies and core debris are removed from the reactor vessel(see Sec. 6.4 for description of this process), they will be placed in transfer containers andmoved from the reactor building to the AFHB for temporary storage.
The characteristics of the removed fuel and debris, the alternatives considered for their packaging and handling, and the environmental impacts of these operations are considered in this section.
8.4.1

Status and Specific Considerations

8.4.1.1

Efforts and Practices to Date

As was previously discussed, reactor defueling, and thus fuel canning (placement in transfer
containers) and transfer, is not expected to begin until 1983.
The primary effort to date has consisted of scoping studies assessing the potential status of thefuel and available options for packaging and handling the fuel after removal from the reactor.
The Allied-General Nuclear Services study of September. 1980 is the basis
for a number of the8
options discussed herein and should be referred to for specific details.
8.4.1.2

Projected Requirements

The TMI-2 core, as loaded, contained 177 fuel assemblies.
Each assembly was about 8.5 inches
square and 170 inches long. It is uncertain what the condition of the fuel will be when.removed
from the reactor vessel; the staff assumes that the fuel will be in any one of three configurations as follows:
Intact--intact,
the assemblies.

but in some cases weakened,

and probably bowed in the upper regions ofFused Sections--portions of adjacent fuel assemblies fused to each other such that they
will have to be physically separated prior to placement in transfer containers (cans).
Core Debris--consists of two types: relatively large pieces that can be mechanicallyhandled, and smaller pieces that will have to be vacuumed and filtered prior to canning.
It is necessary that alternative plans be made to package and handle the range of above described
conditions. In each case, it is likely that the fuel will require canning to provide containmentand structural integrity and to prevent spread of contamination during the steps involved inonsite storage and handling.

8-76
The potential packaging and handling problems resulting from the physical shape of the fuel may
be offset to some extent, by the low burnup of the fuel and the aging of the fuel that will haveoccurred prior to removal from the reactor. This fuel will have low thermal decay heat levels.
Based on the range' of fuel conditions anticipated, the best-case condition is estimated to consist of 50 units (cans) of damaged fuel assemblies (both intact but weakened and separated sections), and six cans of debris, with the remainder of the fuel assemblies undamaged and capableof being handled as ordinary spent fuel. The worst case is estimated to be 177 cans of damaged
fuel assemblies plus six additional cans of debris. It is likely that all fuel assemblies will
have to be conservatively handled, even if they appear to be undamaged, and placed in a transfer
container upon defueling.
8.4.2

Alternative Methods Considered

The alternatives considered for packaging and handling (transfer to storage) of the fuel,on its condition, are described below.
8.4.2.1

based

Intact Fuel Assemblies

Fuel elements that are structurally sound, can bear their own weight, are not distorted, andretain the integrity of their cladding can be handled as normal spent fuel and transferred directly
to pool storage without prior canning.
Intact elements that are structurally weakened, and possibly perforated in the upper region of
the assembly, will require canning immediately upon defueling. The can would provide both vertical and lateral support of the fuel assembly as it is being moved to the AFHB and prevent
spread of radioactive contamination.
Because it may not be possible to determine structural soundness of the assembly upon removalfrom the reactor vessel, it may be necessary to can all units.
There are two alternative routes for transfer of the fuel from the reactor building to the AFHB.
The more direct and simpler route is movement directly between the buildings using the fueltransfer carriage/upender and the fuel transfer tube with the AFHB storage pool as the end-point.
Because of the dimensions of the tube, transfer by this alternative would limit the size of the
package to 11.48 square inches cross section (or 15.25 inches in diameter for cylindrical packages by 15.1 ft long).'
The second transfer route involves moving the packages through the containment hatch, external to
the reactor, and back into the spent fuel pool for interim storage. This routing involves designing specialized transfer packages and equipment and entails greater operational complexity.
It
therefore would be appropriate only if canned fuel assemblies (or sections) were of larger dimensions than could traverse the fuel transfer tube.
Interim wet storage in the spent fuel storage pool pending packaging and offsite disposal isanticipated.
Dry vault or caisson storage in a new onsite facility is another alternative forinterim storage. These options are described in more detail in Section 9.
8.4.2.2-Fused Sections

Where the fuel assemblies (or sections) are fused together in a configuration that does not fit
into the can used for intact assemblies, the following alternatives are available upon removal of
the fuel from the reactor:
Use of a larger transporter can or "bucket" for the fused sections and route the canthrough the containment hatch (see Sec. 8.4.2.1).
Mechanically separate the larger pieces into sections small enough to be loaded into
the fuel cans that can be moved via the upender through the fuel transfer tube.
Interim storage alternatives are the same as for intact assemblies.

8-77
8.4.2.3. Core Debris
The handling and packaging of the core debris potentially will require the greatest use of auxilliary equipment and the maximum number of operational steps. The remote removal of fuel debris
from the core and loading into a transporter can is expected to be done in two steps as follows:'
Mechanical tongs will be used to remove the larger pieces and place them in a container
acting as a receiving "bucket".
A vacuuming system will be used to remove the small pieces and fuel fines, filter, andsegregate the pieces by size. These pieces will then be canned in the same manner as
the larger ones.
8.4.3

Details of Methods and Facilities

The procedures and equipment used to handle and package the fuel will be intended to ensure that(1) site personnel exposure is maintained at ALARA levels and (2) the probability of a handling
accident is minimized.
The following design and operational considerations will govern the implementation of any packaging and handling alternatives:
Ideally, a single can (transfer container) design will be developed for all projected
uses. This can would be sized to handle a range of fuel component sizes and also fit
in the interim storage racks and potential shipping casks. The can will be designed toprevent the occurrence of criticality.
The fuel can will be designed to be compatible with the storage pool water, if thisinterim storage alternative is selected.
Containment to prevent dispersion of radioactivity will be ensured.
Special procedures will be developed for fuel handling and packaging because of theuniqueness of the TMI-2 situation.
Paramount will be the need to ensure that contamination does not spread during handling and canning operations.
Regulatory review
and approval of these procedures will be obtained.
8.4.3.1

Intact Fuel Assemblies

The likely approach for packaging and handling the intact fuel assemblies would be to place theassemblies directly into a fuel can upon removal from the reactor vessel, route the canned fuelthrough the fuel transfer tube, and provide interim storage in the spent fuel pool.
Following interim storage in racks in the spent fuel pool, the fuel cans will be loaded into a
LWR spent fuel shipping cask for offsite transport for storage and/or reprocessing and, ultimately, disposal.
These options are discussed in Section 9. A separate shipping container may
be required as an additional containment barrier within the cask for offsite shipment.
8..4.3.2

Fused Sections

In the case of the fused sections, both (1) the use of larger fuel cans that would be routed
outside of the reactor building and (2) the alternative of mechanically separating the f sed
sections to permit use of the smaller fuel cans that can be routed through the fuel transfer tube
are feasible alternatives that must be considered until design and operational studies arc completed.
After removal to the interim storage area,

the handling procedures for cans containing fusedsections will be the same as for the intact assemblies.
8.4.3.3

Core Debris

Remote handling of the core debris, either by use of mechanical tongs for larger pieces or a"vacuum" system for the small pieces, is the likely approach. After the debris are loaded into
the fuel can, the can will be routed through the fuel transfer tube for interim storage in the
AFHB.

8-78
8.4.4

Effluents and Releases to the Environment

The nature and impacts of the releases to the environment that could occur during the packaging
and handling operations have been covered in the discussion under core defueling (Sec. 6.4).
8.4.5

Environmental Impacts

Occupational doses to workers involved in the packaging and handling operations and offsite doses
to the population are treated as components of the core defueling impacts (Sec. 6.4.5).
8.4.6

Economic Costs

Capital and operating costs for the packaging and handling operations are included as part of thecore defueling costs (Sec. 6.4.6).

Reference--Section 8
1.

Amendment No.

10 to License No. DPR-73, U.S. Nuclear Regulatory Commission,

2.

Letter from J.F. Ahearn, U.S. Nuclear Regulatory Commission to H. Dieckamp, General Public
Utilities Corporation, January 12, 1981.

3.

"Data Handout - TMI-2" as available to the TMI Working Group meeting held at TMI,
PA, September 23, 1980.

4.

Memorandum from T.L. Gilbert to W.K. Lehto, Argonne National Laboratory,
Sludge Volumes and Activities, December 9, 1980.

5.

Letter from W.J. Dircks,
Energy, January 7, 1981.

6.

"Future EPICOR II Operation," Letter from G.K. Hovey of Metropolitan Edison Company - TMI-2
to TMI Program Office, January 13, 1981.

7.

"EPICOR II Resin Solidification Procurement Specification," Letter (TLL 545) from G.K. Hoveyof Metropolitan Edison Company - TMI-2 to TMI Program Office, November 17, 1980.

8.

"Scoping Studies of the Alternative Options for Defueling, Packaging, Shipping, and Disposing of the TMI-2 Spent Fuel Core," Allied-General Nuclear Services, AGNS-35900-1.5-79,
September 1980.

March 12, 1980.

Middletown,

Subject:

AFHB

U.S. Nuclear Regulatory Commission to C.W. Bateman, U.S. Dept. of9.

STORAGE,

TRANSPORTATION,

AND DISPOSAL OF FUEL AND SOLID WASTE

The waste management activities of onsite storage of the packages of TMI-2 waste and spent fueland the alternative offsite transportation of the waste and fuel to storage or disposal facilities are discussed in this section. The purposes of this section are to:
Describe the waste management activities conducted to date at the TMI site subsequent
to the accident.
Provide estimates of the quantities of each type of waste package and numbers of waste
shipments (shielded and unshielded).
Discuss the regulatory and technical constraints on the waste storage,
and disposal operations.

transportation,

Define the range of alternative approaches for the waste storage, transportation, anddisposal options; identify the viable alternatives; and provide the details of theselected alternatives.
Describe the effluents and releases to the environment from storage,
and disposal activities.

transportation,

Determine the environmental impacts and occupational radiation doses under both normal
and accident conditions for these activities.
9.1

CURRENT STATUS AND APPLICABLE CONSTRAINTS

9.1.1

Waste Management Activities to Date

Current waste management activities involve handling radioactive waste that has been produced indecontamination and other cleanup activities at TMI-2.
No significant environmental impacts have.
been identified for the current phase of the operations.
Shipment of wastes for disposal has
been by truck to the commercial low-level-waste disposal facility near Richland, Washington.
Wastes shipped to date, consisting of compacted waste in drums and cleanup materials in wooden
boxes, have been contaminated with low levels of radioactivity.
As of February 5, 1981,
2013 drums and 273 LSA boxes of low-level waste had been transferred off the island in 36 truck
shipments.
An interim storage facility at the Unit 2 cooling tower desilting basin has been constructed tostore wastes such as the resin bed liners from the EPICOR II system until these wastes can beprocessed, as necessary, and shipped for disposal.
The interim storage facility presently
consists of two modules, with space available for up to six modules.
Each module has 60 cells,
and each cell can hold two of the 4-ft x 4-ft resin bed liners or one 6-ft x 6-ft liner.
In addition,
9.1.2

a temporary radwaste surface facility is located in this general area (see Fig. 1.3).

Waste Package and Shipment Parameters

In Sections 5 through 8, the volumes of each type of waste generated from the accident cleanup
and decontamination operations have been estimated.
In addition, best-case (minimum number ofpackages) and worst-case (maximum number of packages) estimates have been made of the number ofpackages that will be required for each waste form. 'In this section the totals of each type of
package are provided for all the waste to be shipped from TMI-2 under best-case and worst-caseconditions.
The number of shipments of each type of package are then projected.

9-1

9-2
The regulatory and technical constraints that influence the shipment parameters are discussed in
Section 9.1.3.
Shipment of any radioactive material must comply with regulations on external
dose rate.
Shielding is generally not necessary to meet these regulations for truckloads ofdrums or boxes of low-activity waste material, but it is necessary for higher-activity wastes
such as ion-exchange materials (dewatered or solidified) and damaged irradiated fuel.
For purposes of estimating numbers of shipments when shielded casks or overpacks are used, the shipments
will be called shielded shipments; when such shielded casks or overpacks are not used, the shipments will be called unshielded shipments.
The number of unshielded packages (in the form of
55-gallon drums and low-specific-activity, LSA, boxes) and the number of unshielded waste shipments for each of these categories of waste are listed in Table 9.1 for best and worst cases.
The summary in Table 9.1 was compiled from the information in Section 8.
The waste in drums
includes compacted trash and immobilized decontamination liquids of low specific activity.
The
waste packaged in LSA boxes includes noncompactible trash, contaminated equipment, and low-activity irradiated hardware.
The large quantity of packages containing EPICOR II and zeolite system liners, drums containing
other high-specific-activity waste, and the fuel casks will, be shielded shipments.
The total
number of packages of each type and the number of shielded shipments for the best- and worst-caseconditions are given in Tables 9.2 through 9.5.

Table 9.1.

Estimated Number of Unshielded Waste Shipments
Best-Case Conditions

Type of Waste

Worst-Case Conditions

Number of
Packages

Number of
Shipments

Number of
Packages

Number of
Shipments

6 ,4 0 0 b

53

13,000c

108

1,600
8,000

14
67

2,400
15,400

20
128

1,025
17
1,042

86
2
88

1,790
338
2,128

149
28
177

55-Gallon Drumsa
Trash
Decontamination solutions (AFHB
and reactor building)Drum totals
LSA Boxesd
Trash
Contaminated equipment and hardware
Box totals
Totals155

aBased on surface radiation level distribution with 25 percent of drums at 50 mR/hr,

305
50 percent

at 100 mR/hr, and 25 percent at 200 mR/hr. Average load is about 120 drums per shipment.
bUnder best-case conditions, trash incineration will reduce number of trash drums to 1600.
This will reduce number of low-level trash shipments from 53 to 13.
Incineration will,
however, produce an additional 480 drums requiring 34 shielded drum shipments.
cUnder worst-case conditions, trash incineration will reduce number of trash drums from 13,000
to 3,260.
This will reduce number of low-level trash shipments from 108 to 27.
Incineration
will, however, produce an additional 974 drums requiring 70 shielded drum shipments.
dBased on surface radiation level distribution with 25 percent of boxes at 50 mR/hr, 50 percent
at 100 mR/hr, and 25 percent at 200 mR/hr. Average load is 12 boxes per shipment.
boxes are assumed, but metal LSA boxes are a feasible alternative.

WoodenEstimated Number of Shielded Ion-Exchange Material Shipments

Table 9.2.

Worst-Case Conditions

Best-Case Conditions
Prefilters

ZeoliteLinersCationLinersMixed Bed
Liners

Total

Prefilters

ZeoliteLinersCationLinersMixed Bed
LinersTotal

AFHB water

49a

-

14a

6

69

4 9a

-

14a

6a

69

Reactor building sump waterb

-

6

1

1

8

-

27

4

2

33

RCS accident waterc

-

1

1

1

3

-

3

7

3

13

1

1

1

3

-

6

35

8

49

Source of TreatedLiquid Waste

RCS flush and drain waterc
RCS decontamination solutionsd

-

-

1

1

2

-

3

2

1

6

Total number of shipments

49

8

18

10

85

49

39

62

20

170

aEPICOR II

system wastes currently in storage.

bBest case assumes zeolite liners loaded to 120,000 Ci; worst case assumes 10,000 Ci loading.
cBest case assumes SDS processing; worst case assumes modified EPICOR II
dBest case is evaporation or bituminization; worst case assumes SDS.

system is used.

(!

9-4

Table 9.3.

Waste Type andDrum Surface
Radiation Level

Estimated Number of Shielded Drum Shipments

Best-Case Conditions
Number of
Number ofDrums perShipments
Drums
Shipment

Worst-Case Conditions
Drums perNumber of
Shipment
Shipments

Number of
Drums

Sludge> 5 R < 20 R/hr
> 20 R < 500 R/hr
Incinerator Asha......
22

4

6

15
22

14
4

1
6

> 1 R < 2 R/hr

480

14

34

974

14

70

4

7

1

29

7

5

1670

14

119

Spent Filters> 10 R < 100 R/hr
Immobilized
Evaporator Bottoms> 2 R < 20 R/hr

b/

201

41

Totalsalncinerator ash drums produced if

trash incineration alternative is implemented.

Seefootnotes b and c in Table 9.1.
bEvaporator bottoms are not generated under best-case conditions when the CAN DECON
technique is used for primary water system decontamination.

Table 9.4.

Estimated Number of Miscellaneous
Best-Case Conditions

Number of
Shipments

-

86

86

6

6

6

15

15

105

105

17

6

33

11

Number of
Shipments

38

13

Mirror insulation-

Core filter

6

Irradiated hardware
Zeolite system filters

Contaminated equipment

208

40

TotalsTable 9.5.

Worst-Case Conditions
Number of
Packages

Number of
PackagesType of Waste

Shielded Shipments

Estimated Number of Shielded Shipments in Fuel Casks
Best-Case Conditions

Worst-Case Conditions

Numberof Casks

Number of
Shipments

Numberof Casks

Number of
Shipments

Damaged fuel assemblies

50

50

177

177

Core particulates (debris)

6

6

6

6

Totals

56

56

183

183

56

56

183

183

Type of Shipment

9-5
The number of ion-exchange liner shipments under best- and worst-case conditions are summarized
Ion-exchange material liners (10-ft 3 to 195-ft 3 ) would be shipped in licensed
in Table 9.2.
The choice of alternatives to process reactor
shipping casks with a capacity of one liner each.
building sump and primary system water discussed in Section 7.1 will have a significant impact
on the number of liners to be shipped offsite.
Waste packaged in 55-gallon drums with surface radiation levels above 200 mR/hr is assumed to be
The number of drums that could be generated undertransported in a shielded shipping overpack.
best- and worst-case conditions and the maximum surface radiation levels for these drums are
Two of the waste forms--accident sludge and spent filter cartridge
given in Table 9.3.
The other waste forms
assemblies--will be generated regardless of the alternatives selected.
will be generated only if the treatment alternative which leads to their generation is implemented.
If an incinerator is used for combustible trash, the number of drums containing compacted trash
shown in Table 9.1 will be reduced by a factor of 10, but between 480 and 974 drums containing
immobilized ash with surface radiation levels of up to 2 R/hr will have to be shipped offsite inshielded casks.
If the alkaline permanganate solution is used to decontaminate the primary system, 1670 drums ofimmobilized evaporator bottoms with surface radiation levels of up to 16 R/hr could be generated.
As shown in Table 9.6, the estimated minimum number of total shipments is 353 and the maximum
is 997.
Regulatory and Technical Constraints

9.1.3

There are four basic safety
transported:

requirementsthat must be

met when radioactive materials

1.

Adequate containment of the radioactive material.

2.

Adequate control of the radiation emitted by the material.

3.

Safe dissipation of heat generated in the process of absorbing the radiation.

4.

Prevention of nuclear criticality, i.e., prevention of the accumulation
fissile material in one location to result in a nuclear chain reaction.

areof enoughThe transportation of radioactive materials within the United States is regulated by the Nuclear
Part 71 of Title 10 ofRegulatory Commission (NRC) and the Department of Transportation (DOT).
NRC regulations
the Code of Federal Regulations contains applicable NRC rules and regulations.
provide the standards which must be met, rather than attempt to specify how they are to be met.
An example of the application of this basic concept is the fact that the regulations do notprohibit the shipment of any specific radioisotope as long as the basic safety standards are met.
The Department of Transportation (DOT), under the Department of Transportation Act of-1966, theTransportation of Explosives Act, the Dangerous Cargo Act, the Federal Aviation Act of 1958, andthe Transportation Safety Act of 1974, has regulatory responsibility for safety in transportation.
The DOT regulations governing carriage of radioactive materials by rail and by common,
contract, or private carriers by public highway (e.g., truck) are found in 49 CFR Parts 170-189.
The DOT regulations governing packaging of radioactive materials, which are consistent with the
NRC regulations, are found in 49 CFR Parts 173 and 178.
9.1.3.1

Packaging

Applicable Regulations
Packaging for radioactive materials isradioactive material to be transported.

determined by the amount,kind,

and physical form of the

In defining the packaging standards and the package content limits, the consequences of loss ofcontainment must be considered.
In the event of radioactive material release, a hazard to transport workers and to the general public exists because of the external radiation emitted from theexposed radionuclides and the often more serious problem of intake into the body, particularly
through ingestion or inhalation. The radiotoxicity hazards of radionuclides vary by eight orders

9-6

Table 9.6.

Summary of Estimated Number of Shipments

Type of Waste

Best-Case
Conditions

Worst-Case
Conditions

13a

108

86

149

2
16c

28

Low-level solids
Drums

-

trash

LSA boxes - trash
LSA boxes - equipment andhardware
LSA boxes - mirror insulation
Immobilized decontamination
liquids
Unshielded drums

14

20

None

119

AFHB water

69

69

Reactor building sump water

8

33

RCS accident water

3

13

RCS flush and drain water

3

49

RCS decontamination solutions

2

6

6

7

Spent filters

1

Incinerator ash

34a

5
-

Shielded drums (evap.

bottoms)

Shielded ion-exchange materials

Shielded drums
Accident sludge

Miscellaneous shielded
shipments
Mirror insulation-

86c

Core filters

66

Irradiated hardware

15

105

Zeolite system filters

611

Damaged fuel assemblies (andcore debris)

56

183

353

997

Contaminated equipment

Totals

13b

aBest case for trash drums includes generation of 34 shielded incinerator
ash drums.
bContaminated equipment can be packaged in unshielded 80 ft 3 LSA boxes(worst-case conditions) or shielded 70 ft 3 liners (best-case conditions).
cMirror insulation can be packaged in unshielded 80 ft 3 LSA boxes (best-caseconditions) or shielded 70 ft 3 liners (worst-case conditions).

9-7
of magnitude.
Standards have been developed that take into account the toxicity of each radioisotope that is being transported.
For this reason, each radioisotope is classified, for transport
purposes, into one of the seven transport groups, according to its relative toxicity.
A listofthe radionuclides and their respective transport groups is found in 10 CFR Part 71, Appendix C,
and in 49 CFR Part 173.390.
Radioisotope quantity limits are established for each transport group, in order of increasing
quantity, as limited quantity, Type A, Type B, and large quantity.
These quantity limits thenestablish the use of either Type A, Type B, or low-specific activity (LSA) packaging.
Thesecategories are further defined as:
Type A and Type B--The distinction between Type A and Type B packaging is significant.
Since Type B packages carry larger quantities of radioactive materials than Type A,
they are designed to more stringent standards and are considerably more accident resistant than Type A packages.
Type A and Type B packages are both NRC-certified based on
testing and design data submitted to the NRC.
Type A packages, in addition to having adequate radiation shielding, are designed towithstand normal transportation conditions.
Limitations on Type A package contents are
such that an intake of one-millionth of the maximum allowable package contents would
not result in a radiation dose to any organ of the body exceeding internationallyaccepted limits nor a radiation dose greater than I rem/hr at 10 ft from the unshielded
contents.
Type B packages often must be heavily shielded and are designed, based on testing andengineering analysis, to withstand severe accident conditions as well as normal conditions.
These packages would be expected to survive a severe accident without any
significant release of their contents.
In spite of the demonstrated integrity ofType B packages, to be conservative, some of the accident scenarios considered in thisstatement include releases from Type B packages.
LSA Materials--Frequently, large volume shipments (e.g., compacted trash) are transported in Type A packages. Strong, tight industrial packaging is permitted for Type A
quantities of LSA materials moved in exclusive use vehicles.
Available Packages
Shipments of wastes requiring shielded Type A or Type B packages will be made using available
certified shielded casks. Waste containers have been designed to fit into these existing casks.
The limited number of available casks, however, may constrain waste shipment schedules.
The
purchase or lease of additional casks, which is being considered, would alleviate this situation.
9.1.3.2

Transportation

Applicable Regulations
Adequate control of the radiation is required when transporting radioactive material.
To meetthe radiation control limits, the shipper must often provide necessary shielding as an integral
component of the packaging-of the material.
Because TMI-2 Waste and fuel shipments will be consigned for sole use, the following dose limits
specified in 49 CFR Part 173.393(j) apply:
1.

1,000 mrem/hr at 3 ft from the external surface of the package (closed transport vehicle only);2.

200 mrem/hr at any point on the external surface of the car or vehicle (closed transport vehicle only);

3.

10 mrem/hr at any point 6 ft from the vehicle planes projected by the outer lateralsurface of the car or vehicle; or if the load is transported in an open transport
vehicle, at any point 6 ft from the vertical planes projected from the outer edges ofthe vehicle.

9-8
4.

2 mrem/hr in any normally occupied position in the car or vehicle, except that thisprovision does not apply to private motor carriers.

Based on these constraints, certain portions of the TMI-2 waste material can be shipped inunshielded packages; others will require shielded packages.
Vehicles carrying the waste packages
will be placarded and marked in conformance with DOT requirements.
Shipping Configurations
The shipping configurations used to transport packaged radioactive waste from TMI-2 to anothertreatment, storage, or disposal facility depend on the types of radionuclides in the waste, theirgross activities, the type of disposal container and its radiation level.
All shipments must
conform to radioactive material regulations, as well as transportation regulations.
To date, radioactive waste shipments from TMI-2, with a few special exceptions, have been limited
to legal-weight truck transport.
Most states require that the total weight of the tractor,
trailer, and package be less than about 73,000 pounds.
Within this weight constraint, two categories of shipments (i.e., shielded and unshielded) were considered for low-level radioactive
waste.
Most containers for low-activity waste are metal drums or wooden boxes.
Materials such as
dewatered ion-exchange materials or damaged fuel may contain considerable radioactivity.
Shipment of any radioactive material must comply with regulations on external dose rate.
Shieldingis generally not necessary to meet these regulations for truckloads of drums or boxes of lowactivity waste material, but is necessary for high-specific-activity wastes such as ion-exchange
resins (dewatered or solidified) and irradiated fuel.
In certain cases, aggregates of drums orboxes of low-level waste must be shielded, and they may be loaded into shielded overpacks forshipment.
The shielded casks or overpacks supply additional mechanical integrity to the truckcontents, for prevention of uncontrolled releases to the environment, in both normal transportation and in transportation accidents.
The legal weight constraint limits the total payload (the amount of packaged waste) plus shielding (if required) to about 42,000 pounds.
The shipment payload is further limited by the shipping cask internal dimensions and its compatibility with the type of disposable container used.
The characteristics of available containers for the TMI-2 waste are presented in Table 9.7.
The shipping configurations considered for unshielded and shielded TMI-2 packaged waste are shownin Tables 9.8 and 9.9 as a function of the containers shown in Table 9.7, the radiation level at
the surface of the disposable container, and available shipping casks.

Table 9.7.

Characteristics of Disposable Containers

Approximate
Type ofContainer

Dimensionsa (ft)Diameter Height

55-gallon drum
10-ft 3 liner
50-ft 3 liner
75-ft 3 liner
80-ft 3 liner
180-ft 3 liner
195-ft 3 liner
LSA boxb

2
2
4
4.5
5.5
6
6.5

x
x
x
x
x
x
x
c/

3
4.5
4
5.5
5.5
6
6.5

Internal
Volume (ft 3 )Waste
Capacity (ft 3 )

7.35
10
44
75
80
175
195
81

7
8-10
40
70
75
165
180
80

aD = diameter; H = height.
bUnsh'ielded configurations only, all others are compatible with shielded shipment.
cLSA box is 3 ft x 4 ft x 6.5 ft.

9-9Table 9.8.

Unshielded Shipping Configurations

Shipment Waste Capacity (ft3)

LSA Boxes

Radiation Level
At Container Surface

55-Gallon Drums
VolumeNumberVolumeNumber

50 mR/hr

840

120

960

12

50 to 100 mR/hr

630

90

720

9

100 to 200 mR/hr

350

50

480

6

Table 9.9.

Radiation Level
At Container Surface

Shielded Shipping Configurations
Shipment Waste Capacity (ft3)a55-Gallon Drumsb

Large Containers

To 1 R/hr

130

200

To 10 R/hr

98

170

To 20 R/hr

98

170

To 100 R/hr
To 1,000 R/hr

49
28

40 to 7 5 c
40 to 75c

To 10,000 R/hr

14

20

apartial listof containers used.

bDivide by 7 to obtain number of drums.
cTwo different-sized disposable containers.

Shipping Methods
Shipments by truck and intermodal rail and truck are alternatives for the transport of TMI
wastes.
Final selection may depend on the choice of the disposal site. Trucks can depart from
TMI and go directly to any of the potential storage or disposal locations.
While railroad facilities are available on the TMI site, there are no rail facilities at a number of the potential
storage and disposal sites.
For example, the nearest railroad to the Beatty, Nevada, disposalsite is about 115 miles away in Las Vegas, Nevada.
There is a rail siding on the Hanford site
about one mile from the commercial disposal site. Unless this siding is extended to the disposal
site, a transfer by truck would be necessary.
Special arrangements for unloading rail cars using
Hanford site facilities would have to be made with DOE.
In addition, availability of shielded rail casks is limited; there are constraints imposed onrail shipments that create logistics problems and increase shipment duration; and intermodal
shipments may involve higher exposure levels to handlers than for single mode shipments becauseof the transfers required.
Therefore, truck shipment from TMI to the treatment, storage, or disposal locations is considered
to be the most likely mode of shipment for the majority of the TMI-2 waste.
The projected transport routes are discussed in Section 3.2, and the impact analyses in Section 9.4 are based on theuse of truck shipments for the waste and damaged fuel.

9-10
Routing Considerations
Until recently, the primary safety measures in regulation of radioactive materials transportation
were controls on packaging and related transportation parameters.
However, the DOT has now
enacted regulations (January 19, 1981; effective February 1, 1982) that focus on routing andrelated operational controls for highway transportation of certain radioactive materials, including waste.
These regulations are revisions to 49 CFR Parts 173 and 177.
NRC approval of routing
for spent fuel shipments is also now required (10 CFR Section 73.37)'.
In brief, the regulations require the following:
A general rule would require a motor vehicle carrying radioactive material that is
placarded to be operated on a route that presents a risk to the fewest persons, unlessthere is not any practicable alternative highway route or it is operated on a "preferred" highway.
The motor vehicle would thus have to be operated on a route which
minimizes transit times, so as to minimize exposure.
A more specific rule would require that any motor vehicle transporting a package containing a "large quantity" of radioactive materials be operated on "preferred" highways
in accordance with a written route plan prepared by the carrier before departure.
Preferred highways would be designated by state agencies based on a policy of an overall minimization of impacts from normal transportation and from transportation accidents.
This rule would require use of an interstate urban circumferential, or bypass,route to avoid cities if available, instead of an interstate through route.
Notification to states of Type B shipments.
9.1.3.3

Treatment (Offsite),

Storage,and/or Disposal

Availability of Facilities
The potential alternatives for management of TMI waste and fuel include use of offsite treatment,
storage, and disposal facilities. The use of these facilities is dependent on the characteristicsof the waste, the costs and benefits of each alternative, and the capabilities and availabilityof the facilities for their handling.
Other factors, as discussed in the following sections alsoneed to be considered regarding disposition of these waste forms.
Low-Level Waste (LLW)
The only offsite alternative considered for LLW is burial in a shallow land burial site. Consistent with current LLW management practices, no offsite treatment or storage alternatives areevaluated for LLW management.
Only three of the six commercial LLW burial sites currently are operative.
These three are thesites at Beatty, Nevada; Richland, Washington; and Barnwell, South Carolina. Of these three, theBarnwell facility is currently excluded from receiving TMI-2 accident cleanup wastes by order of
the Governor, and without modification of that exclusion leaves the Beatty and Hanford facilities
as the only potential sites currently available for receipt of LLW from TMI-2.
As a result of an
initiative passed in the State of Washington in November 1980, the Richland facility will not be
permitted to accept out-of-state shipments of reactor plant wastes after July 1981 unless a
specific "compact" is negotiated on an individual case basis. The governors of South Carolina
and Washington have urged the development of regional LLW disposal sites in other parts of thecountry to reduce the need for continuing long-range shipments of high volumes of wastes to thesites in their states.
In December 1980, the Low-Level Radioactive Waste Policy Act was enacted (Congressional Record,
516545, December 13, 1980).
This act stated that each state is responsible for providing for theavailability of disposal capacity for commercial low-level wastes generated within its borders.
The act also allowed states to enter into compacts necessary to provide for the establishment andoperation of regional disposal facilities.
These compacts would require Congressional consent
prior to taking effect and could restrict the use of the regional facilities to states within theregion after January 1, 1986, but not before that time.

9-11
While the Low-Level Radioactive Waste Policy Act assigns responsibility to each state for providing low-level radioactive waste disposal capacity, the federal government recognizes a need toassist states in dealing with unique wastes, such as those wastes which will result from theTMI-2 cleanup, which may not be suitable for normal shallow land disposal.
The act, however,emphasizes a need for state authorities to site new disposal facilities or enter into agreements
needed to provide for the adequate disposal capacity for the low-level wastes generated withinthat state.
Congress also has requested that the DOE, as the lead federal agency for nuclear waste management, prepare a preliminary assessment pertaining to the development of regionally distributed
disposal sites for LLW.
The DOE has estimated that six to eight additional sites will be needed
within the next ten years and that the Northeast and Midwest sectors of the country are theregions with the greatest near-term need for disposal facilities. The establishment of a new LLW
Such aburial site in the state of Pennsylvania to accommodate TMI waste also is an option.
facility would provide a minimal transport distance for the waste shipped from TMI, 'and also
would permit disposal to be accomplished in the state where the waste was generated.
However, no
new land burial sites are in the development stage, and it is likely that five to ten years will
be required before new sites are available.
Thus, early direct shipment to a new LLW disposalsite is not a near-term option.
Another option for the disposal of the TMI-2 LLW is to reopen existing commercial sites currently
There are three sites in this category--West Valley in New York; Maxey Flats
not receiving waste.
in Kentucky; and Sheffield in Illinois. The Sheffield site has limited capacity, but the capacity at the other two sites would be sufficient for the TMI-2 LLW.
The Maxey Flats site currently
is constrained from receiving any waste shipments by state legislative order, and plans are being
developed for decommissioning of this facility. The West Valley site is the closest LLW burial
facility to TMI.
It would, thus, appear to be a technically feasible alternative.
However, thefuture disposition of this facility, as well as other waste management facilities at West Valley,cannot now be projected.
Recent Congressional legislation (West Valley Demonstration Project
Act-Public Law 96-368, dated October 1, 1980) has committed the DOE to participate with New YorkState in a program to solidify and dispose of the HLW stored at the site, but the legislation has
not resolved the future utilization of the buri-a! locations.
A number of the federally (DOE) owned LLW disposal facilities also are alternative backup locations for disposal of TMI-2 LLW waste.
The DOE facilities at Hanford, Washington; the SavannahRiver Plant in South Carolina; and the Idaho National Engineering Laboratory are suitable locaCurrent DOE policy does not allow fortions with sufficient capacity for additional LLW burial.
this option, and the DOE has not indicated that any exception would be made for TMI-2 waste at
this time.
Interim onsite storage, of LLW packages prior to shipment is anticipated,

and facilities are now

under construction or in use for temporary storage of certain of the packages.
High-Specific-Activity Waste
In the case of the management of the high-specific-activity waste (HSAW), offsite alternatives
include the use of facilities for special packaging or treating the waste, interim or long-term
storage, and disposal in a range of potential facilities, bounded by disposal in a geologic
repository at one extreme and intermediate depth burial at the other.
The capability of both commercial and governmental organizations to undertake the special
handling, processing, long-term storage and disposal of the HSA first-stage ion-exchange wastes
has been considered.
There are several commercial organizations in the fuel fabrication fieldthat have basic hot-cell units that have been used for spent fuel examination, and have theequipment and staff for handling these types of radioactive waste materials.
However, none ofthese commercial organizations has routinely handled the immobilizaton, long-term storage, or
disposal of such large quantities (thousands of curies) of long-lived (30-year half-life) radionuclides, nor do they have experience or facilities for processing such wastes to achieve a form
On the other hand, there are commercial firms in thesuitable for long-term storage or disposal.
waste handling field that have experience in handling and immobilization of essentially normal
reactor wastes that do not have the special facilities and capabilities to handle the HSA firststage wastes.

9-12
In overview, while some of the firms might have facilities and staff along with some interest in
developing processes for such work, and others might have business interests in actual waste
immobilization, none has the required combination of extended availability of special facilities,
specialized staff, and waste management business interest that would have to be applied to carryout such a waste management task.
Thus, due to its special nature and limited application,adequate commercial capability is not available (and commercial interest to develop that capability is not evident) for the handling of high-specific-activity special TMI-2 wastes.
On the other hand, the DOE, through its contractors and laboratories, as a result of efforts todevelop suitable methods for immobilization and disposal of a wide variety of government andcommercial radioactive wastes, appears to have the only suitable combination of establishedpersonnel, technological capabilities, and interests for developing and carrying'out such special
operations for a one-time application.
Accordingly, in analyzing the alternatives for the handling, processing, long-term storage or disposal of high-specific-activity TMI-2 wastes, consideration of generally capable DOE sites or facilities has been included.
Treatment alternatives include high-integrity packaging, processing and immobilization, incineration, chemical digestion, and elution and immobilization of the eluate as described in Section 8.1.
The only currently available offsite treatment facilities for special processing arelocated at federally owned (DOE) sites, where the TMI waste could be either treated separately or
possibly comingled with the DOE and DOD waste.
However, these facilities currently have not been
made available for treatment of TMI-2 waste because the DOE is concerned that NRC's regulatory
authority would be extended to the entire DOE facility. The DOE has acknowledged' that only its
facilities have the unique technical capabilities (both staff and equipment) to handle thesewastes and that nontechnical considerations have dictated their policies on this question to.
date.
Requests have been made by the NRC to the DOE 2 to accept these waste materials, and clarification provided of the NRC's position that no licensing would be required if the DOE facility's
primary functions were other than handling commercial wastes in accordance with the Energy
Reorganization Act, Section 202.3
The DOE sites where these treatment processes could be conducted, if permitted, are as follows:-

Processing and Immobilization.
Zeolites and resins could be immobilized and packaged
for eventual repository disposal.
DOE solidification facilities exist at the waste
calcining plant at the Idaho National Engineering Laboratory and are planned for construction at the Savannah River plant in the late 1980s.
A commercial pilot-plant
high-level waste processing operation also is located on the nonmilitary portion of the
Hanford Reservation.

-

Incineration.
Effective incineration of the combustible component of the HSAW (resins)to reduce volume and mass could be possible at a number of facilities being developed
for treating higher specific activity.waste.
These facilities include a controlled air
incinerator (CAI) at the Los Alamos Scientific Laboratory used for transuranic waste; a
CAI unit being developed at the Savannah River Plant for operation in 1981; a cyclone
incinerator being adapted for TRU waste burning at the Mound Laboratory; a rotary kiln
production unit at the Rocky Flats Plant scheduled for operation in 1981; and aslagging pyrolysis unit proposed for operation in 1986 to treat transuranic waste at
the Idaho National Engineering Laboratory.
In order for these incineration facilities
to be used for TMI-2 HSAW, systems modifications would be required.
Acid Digestion.
Acid digestion of combustible organic-based waste is accomplished at
the Radioactive Acid Digestion Test Unit (RADTU) at the Hanford Engineering Development
Laboratory (HEDL).
The unit has been used for low-level waste and is being modified toincrease its capacity. In addition, other modifications might be required to permit
processing of the HSAW from TMI-2.

-

Elution and Immobilization.
Elution of resins and zeolites to obtain the separated
fission products has been developed at the Savannah River Plant and the Hanford facilities of the DOE.
The work has been done in processing waste streams currently atthese facilities.

Either short-term or long-term storage of the HSAW could be carried out at the TMI site or an
offsite location.
However, further processing at TMI to put these materials in a form suitablefor eventual disposal in a future repository is not, in the staff's opinion, an appropriate
activity to be performed at TMI since the licensee's staff and industry in general have no9-13
experience in this area.
On the other hand, the existence of DOE facilities, with experienced
personnel, argues against construction of duplicate equipment at TMI, for a one-time application,with the attendant increase in contaminated equipment and risk to workers and the public.
Onsitestorage at TMI will require construction of a facility designed to accommodate the required
packaging and handling equipment.
Handling and space constraints attendant to use of a range of
packaging configurations, and the package surface radiation levels of the HSAW, will have to betaken into account.
Offsite short-term storage at DOE storage facilities, at one of the aforementioned treatment
facilities (which would minimize transportation and handling) or at a final disposal site (pending its completion), is a potential alternative for the HSAW.
The potential SDS first-stage wastes will be very special in nature and should most logically behandled like spent fuel materials.
Accordingly, long-term storage options for the SDS firststage wastes include nuclear power plant spent fuel pools if excess capacity is available, DOE
facility spent fuel pools or other suitable areas; storage at the shutdown commercial reprocessing facilities, or an away-from-reactor (AFR) storage pool facility.
However, AFR facilities do
not currently exist and future locations have not been designated.
Under current policy, DOE storage facilities would not be available for TMI waste.
Other institutional constraints on the use of specific storage facilities include the commitment on the partof the Federal Government to the State of Idaho to remove existing transuranic waste stored at
the site (which would likely mitigate against any additional HSAW being stored there); the factthat the DOE facilities at the Savannah River Plant and Hanford are in states (South Carolina andWashington) which have commercial LLW disposal sites, and any constraints imposed by these states
on shipping might also apply to waste shipments to a DOE facility in the same state.
The alternatives for disposal of the HSAW range from disposal in a geologic repository when one
is available, which will require an extended storage period in a stable form (e.g., vitrified),to the possible use of intermediate depth burial incorporating an intruder barrier to prevent
violation of the cover integrity for some of the lower end of the HSAW.
Although a national program is in progress to identify and qualify sites for geologic disposal
facilities in a variety of media, specific locations have not yet been identified.
Intermediate
depth burial as an alternative to shallow land burial is currently being developed.
This technique may be a feasible alternative for some of the HSAW.
The DOE plans to demonstrateintermediate-depth burial.
This demonstration could possibly use some of the HSAW, such as the
EPICOR-II first-stage waste.
Fuel
Two independent assessments 4 's have been made on the condition of the TMI-2 core resulting fromthe accident.
Although most probable conclusions reached in the two assessments differ as to theextent of the projected damage, both conclude that the temperatures and pressures experienced by
the fuel assemblies during the accident have significantly altered the fuel components to theextent that they no longer meet specifications for reuse.
Lacking specific observational information, it must be presumed for planning purposes that the core will include intact fuel assemblies that may have undergone some damage and distortion, fused formations of fuel larger incross section than an individual assembly, and loose debris.
As previously discussed, all the fuel will be retrieved from the reactor. Some portion all of
the core will be sent to hot cell areas at a designated national laboratory for diagnostic
evaluation.
Upon completion of the diagnostic evaluation, the residue of the examined fuel could
be either added to the waste stream at the laboratory for processing as high-level waste, or
treated as described below for the remainder of the fuel.
The remainder of the fuel would bepackaged for interim storage pending final disposition.
The alternatives available and theconstraints affecting their availability are as follows:-

Wet (Pool) Storage - Damaged fuel can be stored either onsite in the TMI storage pool,offsite at another reactor pool, or at a regional AFR-storage pool-;-.in-pool spaces at
the national laboratories; or at the storage pools at the shutdown commercial reprocessing facilities.
As previously noted, AFR facilities may not be available until
1983-1984 at the earliest; excess capacity at other reactor locations is extremely
limited; and DOE facilities are not available at this time for commercial use (see,however, discussion provided above).
In addition, this alternative represents a temporary solution for the fuel disposition, not a final one.

9-14
Dry Storage - Another potential interim alternative involves placing the canned TMI-2
fuel in either a storage vault or hot cell, or in a specially designed spent fuelcaisson.
The constraints here are, again, the current unavailability of DOE or otherfacilities of this nature, and the temporary nature of this option pending final disposition.
Chemical Processing at a Commercial or DOE Facility - This alternative involves the
reprocessing of the fuel to recover the useful uranium and plutonium and subsequentimmobilization and disposal of the high-level waste.
However, current national policy
precludes the implementation of the reprocessing alternative and, even if this policy
were to be changed, commercial reprocessing would probably not be available until 1990
at the earliest.
For each of the above-described alternatives, final disposal will be in a geological repository
which will be sited, constructed, and operated by the DOE.
9.2

ALTERNATIVE METHODS CONSIDERED

The alternatives considered for storage,
packages are discussed in this section.
9.2.1
9.2.1.1

transportation,and disposal of the TMI-2 waste and fuelStorage
OnsiteOnsite storage of the waste for an interim period prior to shipment is a feasible and necessary
option.
Interim storage facilities are now under construction or in use for temporary storage of
certain waste types (see Fig. 9.1).
Interim storage with suitable monitoring for continued waste
containment integrity permits timely removal of wastes from shipment staging areas, allows fordecay of radionuclides with short half-lives, permits continuing cleanup operations while ultimate disposal strategies are formulated, and provides a buffer for contingencies beyond controlof the licensee, such as shipping strikes, embargoes on certain waste types, or unavailability oftransport equipment (casks), offsite storage, or disposal facilities.
Licensee conceptual plans now exist for construction of an interim storage and staging area forlow-level radioactive waste including certain of the high-specific-activity wastes.
The facility
will be sized to accommodate about 800 55-gallon drums, 150 wooden boxes, each 4 ft x 3 ft
3
x 6½ ft, and sixty 50-ft
evaporator bottom liners.
Radiation levels from these waste packages
will range up to about 500 mR/hr.
Maximum radiation levels at the fence surrounding the storage
facility will be less than 0.6 mR/hr.
As a buffer against shipping delays or other problems,
this facility will accommodate the waste expected to be generated during a six-month period.
As an alternative to the interim storage and staging area described above, a longer term storage
facility for the various forms of LLW and HSAW could be considered.
Based on current projectionsof total waste volume to be generated in decontamination activities, the facilities would have to
be considerably larger than needed for the six-month storage area previously described.
Evaluation of the feasibility of this facility would include consideration of the potential for failureof container integrity during longer term storage.
In addition, the means for adequately handling, shipping, and disposing or storing some of the involved waste types are available offsite.
Another potential interim waste storage alternative is to place the packages in a designated areain the reactor building.
While providing a readily available shielded storage facility of significant capacity, this option has a number of drawbacks.
It would necessitate storing the waste
packages in the reactor building while both cleanup and fuel removal operations were being conducted, requiring revisions to existing plans, potential delays in completion of these operations, and additional costs.
In addition, package-handling capabilities are limited at the lower
levels of the building.
Alternatives for handling the damaged irradiated fuel removed from the core would be to adapt theexisting pool for long-term storage or to construct new onsite facilities for long-term storage.
These options will not be necessary in the time frame being considered if long-term storage
facilities become available offsite and if research and development plans being formulated forthe core components are carried out for the entire core.

9-15

41

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0)el

0(a

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4-)U
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0)
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MR,=U

9-16
9.2.1.2

OffsiteOffsite storage of the waste for periods pending a decision as to final disposal is a feasible
option for the HSAW that will be disposed of by other than routine shallow land burial techniques.
Offsite storage may be utilized either before or after immobilization of the waste.
Storage at the same location where processing to immobilize the wastes is carried out will minimize exposure, due to reduced handling and transportation.
Offsite storage at designated storage sites, at a treatment facility, or at a site selected
because of its proximity to a final disposal facility is a potential alternative for the HSAW.
The following are those facility categories and specific locations that currently are considered
to be prime candidates for retrievable storage facilities for TMI-2 waste because appropriate
storage capacity is available or is contemplated:-

Commercial nuclear power plant spent fuel pools where excess capacity is available.
It
may be difficult to identify such facilities since the majority of power plants arefacing a storage capacity problem.

-

An AFR pool facility.
As previously noted, AFR facilities do not currently exist andfuture locations have not been designated at this time.

-DOE-operated storage facilities, including those at the Hanford Reservation,
River Plant, or Idaho National Engineering Laboratory.

-

The storage pools at the commercial reprocessing facilities at West Valley, New York;and Morris, Illinois.
In addition, the Barnwell, South Carolina, facility is a potential site, but has not been licensed to store nuclear material.

SavannahLong-term storage of that portion of the fuel not sent to DOE laborato ries for diagnostic evaluation is a likely alternative until it is determined whether the fuel will be placed in a repository without first undergoing reprocessing or whether the fuel will be reprocessed and the HLW
produced, then immobilized for disposal.
The storage alternatives can be categorized as either
wet storage or dry storage, as follows:-

Wet (Pool) Storage -- The available alternativecannisters are the same as for the HSAW.

-Dry Storage -,- Involves placing the canned TMI-2 fuel in a storage vault,
specially designed spent fuel caisson.

locationsfor storageof the fuelhot cell,or

The duration of storage of the fuel and HSAW for final disposition in a geologic repository will
depend on the program to select and to qualify a repository, which currently is not expected tobe accomplished until the 1990s.
Under present laws these facilities will be federally owned and
NRC licensed.
9.2.2

Disposal

Different alternatives are available for final disposal of the low-level waste and high-specificactivity waste generated at TMI-2.
The regulatory and institutional constraints on the availability of disposal sites are discussed in Section 9.1.3.3.
9.2.2.1

Low-Level-Waste Disposal Alternatives

The feasible low-level-waste disposal alternative for TMI-2.waste generated in the near term islimited to burial in a shallow land burial site (at a maximum depth of about 30 ft below thetrench cover surface) under present practice for routine wastes, the current technologically
acceptable disposal approach.
The potential burial sites for wastes that are similar to routinely
generated wastes are as follows:-

Operating commercial LLW burial sites - The sites at Beatty,
ton; and Barnwell, South Carolina.

-

A new regional LLW disposal site in the Northeast or in Pennsylvania.
(A period ofinterim storage of the LLW packages would be required until such a site is developed.)

Nevada,Richland, Washing-9-17
Currently shutdown commercial burial sites having available capacity--the sites at West
Valley in New York and at Maxey Flats in Kentucky.
DOE-owned LLW burial sites having, available capacity--among the major DOE LLW disposalsites, the facilities at Hanford, Washington; the Savannah River Plant in South
Carolina; and the Idaho National Engineering Laboratory could be suitable locations.
9.2.2.2

Alternatives for Disposal of High-Specific-Activity Waste

The alternatives for disposal of the HSAW range from disposal in a geologic repository when one
is available (which will require that the waste be stored for an extended period in a stableform) to the use of intermediate depth burial incorporating an intruder barrier (i.e., layer ofartificial material) to prevent violation of the cover integrity, depending on the radionuclide
inventory and specific activity involved.
In the case of disposal in a geologic repository, although a national program is in progress to
identify and qualify such facilities in a variety of media, specific locations have not yet been
identified.
Intermediate depth burial (at a level 50 to 60 ft below the trench cover surface) represents analternative for disposal of the lower range of the HSAW because it provides significantly better
isolation from intrusion and disturbance from natural disasters (e.g., flooding) than shallowlevel burial at sites where hydrogeology ensures at least comparable isolation from groundwater.
A number of the existing commercial and DOE shallow land burial sites provide suitable hydrogeology for intermediate depth burial.
Among the potential suitable alternatives are:-

The commercial burial sites at Richland, Washington, and Beatty, Nevada, would permit
intermediate depth burial with a remaining distance in excess of 200 ft to the nearest
aquifer.

-

The currently shutdown burial site at West Valley, New York.
The "hulls" burial areaat this site has been previously used for burial of TRU and other HSAW at depths comparable to intermediate depth burial.

-

The DOE burial sites at Hanford, Washington; Idaho National Engineering Laboratory; andLos Alamos are suitable 'for intermediate depth burial because of their hydrogeologic
characteristics.

9.2.2.3

Fuel and High-Level Waste

Final disposal of the fuel and the high-level waste generated from reprocessing the fuel (ifimplemented) will be in a geologic repository which will be sited, constructed, and operated by
the DOE.
9.2.3
9.2.3.1

Transportation
Transportation Mode

Truck shipment and combination of truck and rail are potential options for shipment of the TMI-2
waste and spent fuel packages.
TMI-2 wastes shipped to date have been sent exclusively by truck.
When waste packages are shipped primarily by unshielded vehicle directly to a commercial shallow
land burial site, truck transport is the preferred mode.
There are no rail facilities at thelikely disposal sites, and intermodal rail and truck shipment would be required if rail shipmentfrom TMI is used.
Truck shipment also would be the likely choice for shipment of the majority of
the intermediate-level waste for either immobilization at a treatment facility or storage pending
final disposal.
In this instance, shielded vehicle shipments may be required.
Rail transport may be preferable in those cases when shielding needs for certain intermediatelevel waste packages and for-irradiated fuel cannisters necessitate the use of large, heavy casks
and when off-loading rail spurs are available near the storage or disposal location.
Another factor that may influence the selection of rail or truck transport is the availability of
the appropriate types of shielded casks at the time shipments are being scheduled.

9-18
9.2.3.2

Transportation Routes to Potential Treatment,

Storage,and Disposal Facilities

The truck transportation route to the alternative treatment, storage, and disposal facilities
These routes, which are the currently
that have been identified are described in this section.
established "most likely" routes used by nuclear waste shippers, make use of the federal interstate highway system except for short distances near the starting and termination points, where
They are the routes that the waste shippers use with thelocal and state roads are used.
greatest frequency, but are not necessarily the only routes traversed. The final selection of
the routes at the time of shipment will involve consideration of the requirements of applicable
DOT regulations (Sec. 9.1) on routing and specific constraints imposed by the states and municipalities through which the shipments will pass.
Routes to Treatment Facilities
The potential locations for treatment of HSAW have been identified as the DOE facilities at the
Hanford Reservation, the Savannah River Plant, and the Idaho National Engineering Laboratory.
Local routing of truck shipments from TMI to the interstate highway system, which is applicablefor shipments to all potential locations, is shown in Figure 9.2.
The complete interstate routing to Hanford, Savannah River, and INEL is shown in Figure 9.3.
Local routing at each of theselocations is shown in Figures 9.4, 9.5, and 9.6, respectively.
Routes to Offsite Storage Facilities
The specific locations identified in Section 9.2.1.2 as potential interim storage facilities for
the HSAW or fuel include the three DOE facilities discussed above as possible treatment facilities.
The routing to the Hanford Reservation, Savannah River Plant, and Idaho National Engineering Laboratory are as shown in Figures 9.3 (interstate routing), 9.4, 9.5, and 9.6, respectively(local routing).
In addition, the routes to the pool storage at the reprocessing facilities at West Valley,
New York; Barnwell, South Carolina; and Morris, Illinois, which are also possible storage locations, are shown in Figure 9.7.
Routes to Disposal Facilities
For the low level waste, the routes to the commercial sites at Richland, Washington; Beatty,
Nevada; and Barnwell, South Carolina, are shown in Figure 9.8. The routes to the shutdown commercial burial sites at West Valley, New York, and Maxey Flats, Kentucky, also are shown on thismap.
Figures 9.9 through 9.12, respectively, show the local routing at Beatty, Barnwell, West
Valley, and Maxey Flats. The routes to the DOE burial sites at Hanford, Savannah River, and INEL
are the same as shown in Figures 9.3 through 9.6.
Intermediate depth burial of some of the lower range of HSAW can be considered at the commercial
burial sites at Richland and Beatty; the shutdown site at West Valley; and the DOE sites at
Hanford, INEL, and Los Alamos.
The routes to these facilities, with the exception of Los Alamos,
are as shown previously.
The interstate routing to Los Alamos is shown in Figure 9.3, and the
local routing in Figure 9.13.
9.3
9.3.1

DETAILS OF METHODS AND FACILITIESStorage

The onsite storage facilities currently in use or under construction are needed as an interim
storage and staging area prior to shipment of wastes.
The use of offsite storage for certain of
the HSAW or the fuel, while technically and operationally feasible, depends on decisions yet to
be made as to final disposition of the material.
The availability of specific government facilities also needs to be determined (see discussion in Sec. 9.1.3.3).
The onsite concrete storage facilities for interim storage
EPICOR II spent liners) are described in Appendix Q.

of certain of the HSAW (e.g.,

All onsite package-handling and loading operations would be conducted in a manner that ensuresthat occupational radiation doses are kept as low as reasonably achievable.
Remote operations
will be employed and shielding will be used to minimize the radiation fields to which handlers
are exposed.

Penno.

Penno Turnpike

GOLDSBORO
TMIj..,
1•'•

0

I

Mi

I

2

3

Miles

Figure 9.2.

Route from TMI through Harrisburg Area.

-Turnpike-,,-8
ELIZA BE THTOWN

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Figue
o DO 9..
TretmenRotes and

StorgeIFciliies

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9-21

Figure 9.4.

Routing in Washington to Hanford/Richland

Figure 9.5.

Routing in South Carolina
to Savannah River Plant.

Facilities.

9-22

Figure 9.6.

Figure 9.7.

Routing in Idaho to Idaho National
Engineering Laboratory.

Routes to Potential Spent Fuel
Storage Facilities.

CA N AD A

0 50 100

MOIV'FAIVA

Miles

0

CGoA,

"

..

/

'DA ,0WYOMING/\NEBRASKA

A' 4 14COLOR ADOLI?.

KANSAS/KENTUCKYj

I

ARIZONA
:'

I

/

NEW

OKLAHOMAMEXICOI

TX

TXA

Figure 9.8.

Routes to LLW Disposal Facilities.

9-24

Figure 9.9.

Figure 9.10.

Routing in Nevada to Beatty
LLW Burial Site.

Routing in South Carolina to Barnwell
LLW Disposal Site.

9-25

Figure 9.11.

Routing in New York to West Valley
LLW Burial Site.

OHO

-..

PORTSMO .UIP,F

r .o
I
I

ASHLAND

HI

I'
(V6

-"\LEXINGTON--

HUTNGO-0KENTUCKY

Figure 9.12.

1 40% RýOýUTE

Routing in Kentucky to Maxey Flats LLW Burial Site.

p 10 5

9-26

Figure 9.13.

9.3.2

Routing in New Mexico to Los Alamos Site.

Disposal

LLW currently is being shipped from TMI-2 to the commercial burial site near Richland, Washington.
Shipment of the remainder of the LLW to this site or one of the other commercial sites
remains the preferred alternative, assuming current institutional constraints do not pose impediments.
The alternatives of use of shutdown commercial sites or DOE sites, although operationallyfeasible, are less desirable because existing prohibitions would have to be lifted and newThis could be a lengthy and complex process.
arrangements negotiated.
In the case of HSAW, the use of intermediate depth burial, while a feasible disposal alternative
currently being developed for this type of waste, is not a disposal technique in current practiceat any burial site.
The other option of disposal is a geologic repository.
Since a geologic
interim storage is a required step.
repository will not be in operation for 20 to 30 years,
9.3.3

Transportation

Once the specific offsite storage, treatment, and disposal facilities have been identified, theappropriate truck transportation routes of those alternatives presented in Section 9.2.3 will beselected.
For the purpose of evaluating the potential bounds of the transportation impacts under normal andaccident conditions, two feasible routes having extremes in distance traveled have been selected.
These are:-

The current transportation route from TMI-2 to the commercial burial site at Richland,
Washington, a distance of about 2750 miles.

9-27
The route to the shutdown waste management facilities at West Valley, New York, whichis the closest feasible location for all the treatment, storage, and disposal alternatives.
This distance is about 370 miles.
The rule proposed by DOT regarding the highway routing of radioactive materials (45 FR 7140,
January 31, 1980) would make uniform laws by state and local governments on shipments of radioactive materials by highway.
If adopted as proposed, existing state and local laws that prohibit
In effect, the proposed regulations
the transportation of nuclear materials would be preempted.
establish the interstate highways as the preferred routes for movement of radioactive materials.
For the shipment of irradiated fuel, consideration must be given to this proposed DOT regulation,
the NRC requirement for physical protection of shipments of irradiated fuel (10 CFR Part 73.37),
and the NRC interim guidance for physical protection of such shipments (NUREG-0561, Rev. 1,
June 1980).
In consideration of these factors, the NRC has provided public disclosure of theselected shipment routes for irradiated fuel as well as information regarding safety and safeguards regulations (NUREG-0725, November 1980).
9.4

EFFLUENTS AND RELEASES TO THE ENVIRONMENT

Under normal storage, transportation, and disposal conditions, no routine effluents or releases
from the waste packages or transport vehicles are expected.
"Normal" transport is the situation
when transport occurs without unusual delay, loss or damage to the package, or an accident
involving the transporting vehicle.
The various types of packages that may be used to ship the TMI wastes are designed to prevent any
releases during storage, handling, normal transportation, or disposal operations if their integrity is maintained. Thus, no impact on the environment from this source will occur.
The impact of the direct radiation from the packages under normal transport conditions is discussed in Section 9.5.1. Nonroutine releases under abnormal or accident conditions during transport are discussed in Section 9.5.2.
9.5

ENVIRONMENTAL IMPACTS

The principal environmental impacts from normal storage, transport, and disposal conditions areradiological, resulting from direct radiation exposure to package handlers, crew, and bystanders.
Radiological impacts are discussed in the following sections for truck transportation from TMI tothe storage or disposal site.
In the case of accidents, direct radiological impacts on man are
considered the most important potential component of postulated environmental impacts.
For this evaluation, the impacts are categorized under normal and accident conditions, and arefurther differentiated as occupational exposures to transport workers, environmental exposures tothe population (bystanders), and nonradiological impacts.
9.5.1

Normal Conditions

For evaluation of the radiological impact in the case of normal storage, transport, and disposal,
radionuclide content of the packaged waste and most characteristics of the package are not significant.
The important characteristic is the radiation dose rate as a function of distance fromthe package surface.
Federal regulations (DOT and NRC) impose constraints on the dose rate at
specified distances from the package surface to protect package handlers, transport workers, andbystanders.
Therefore, if required, storage areas and shipments of waste packages will beshielded or fenced 'to reduce the dose rate to acceptable levels.
In addition, vehicles transporting the waste will be placarded to indicate to the public that they are transporting radioactive materials.
For this analysis, conservative dose estimates have been made by using the regulatory dose-rate
limits for the TMI-2 shipments.
In addition, where available, dose rates recorded during actual
shipments ("experience" values) are used to obtain more realistic values.
The material presented
in NUREG-0170 6 and
WASH-1238
has
been
the
basis
for
much
of
the
analysispresented
in the succeed7
ing sections. 'Estimated doses to handlers during storage, package movement, staging, and loading have been
calculated by using experience values for exposure time for each step in the process in conjunction with expected exposure rates for the packages handled.

9-28
9.5.1.1

Occupational Doses

The occupational doses resulting from the storage, onsite movement,
wastes are received by package handlers and vehicle crew members.

loading,

and transport of theOnsite Storage and Handling
The onsite handling of unshielded packages from storage to truck loading involves the following
steps:
storage, pickup at storage area, transfer to loading area, placement in loading area,
truck loading, and truck radiation survey.
In the case of shielded shipments, the packages will be placed in the shipping cask at the loading area and the shipping cask will be closed before the cask and truck are surveyed for radiation.
The analysis of the occupational doses for each different type of waste handling operation at TMI
is presented in Appendix N. Using the relevant data from that appendix, the staff has determined
the occupational radiation dose to a crew on a per-shipment basis and for all the waste shipments(best and worst case), including both shielded and unshielded packages.
These data are summarized in Table 9.10.
Packaged waste will be removed from storage and loaded by two-member crews, and the loaded vehicle will be monitored by a radiation technician.
A total of four two-member loading crews plus
two technicians would be used for the shipments listed in Table 9.10.
The average dose to the loading crews was estimated based on an average of two shipments per day
and time spans of 8 months for the minimum and 23 months for the maximum number of shipments.
It
was assumed by the staff that the radiation technicians would receive 10 percent of the dose andthe loading crews 90 percent of the dose from'each shipment.
The resultant occupational doses
are as follows:
Radiation Technicians - For the minimum number of shipments each radiation technician
would receive a total of 0.73 rem, or about 0.24 rem per quarter.
For the maximum
number of shipments, the total dose would be about 1.7 rem, or about 0.21 rem perquarter.
Loading Crew - For the minimum number of shipments, each crew member would receive a
total of 1.6 rem, or about 0.53 rem per quarter.
For the maximum number of shipments,
each crew member would receive a total of 4.6 rem, or about 0.58 rem per quarter.
Transportation
Occupational doses from truck transport of the wastes and spent fuel are received by the truck
crew. The following parameters are assumed to apply to the shipments of the TMI waste:
Shipments are made by exclusive-use vehicle.
Dose rate is a maximum of 2 mrem/hr in any normally occupied position in the truck, andup to 10 mrem/hr at 6 ft from either an unshielded package or shielded cask.
Two crew members per truck.
Crew occupies cab of truck only during period of actual travel.
The transport and shipment parameters presented in Table 9.11 apply to the TMI-2 waste
shipments.6
Two transport distances are used to bound the extremes of the dose analysis.
longest distance used is 2750 miles, the shortest is 370 miles.

The

The crew doses for the two routes considered are determined as follows:
Longest Route.
Since the duration of crew exposure equals distance traveled divided byaverage speed, the total exposure duration is 60 hours.
Assuming that a crew member spends 58 ofTable 9.10.

*
eOccupational Radiation Doses during Transfer of Waste Packages
from Storage and Truck Loading

Cumulativs
Type of Shipment
and Package

Best-Case Conditions
Cumulative

Worst-Case Conditions
Cumulative

Crew Doseper Shipment(person-rem)

Number of
Shipments

Crew Dose(person-rem)

Number ofb
Shipments

Total
Crew Dose(person-rem)

0.15
0.046

27
104

4.1
4.8

128
177

19.2
8.1

0.017-0.06
0.003-0.006
0'017
0.018

6
1
0
34

0.4
Negligible.
0.6

7
5
119

0.4
Negligible
2.0

Unshielded shipments
Drumsc
LSA boxesd
Shielded drum shipments
SludgeFilter cartridges
Evaporator bottoms
Incinerator ash-

Shielded ion-exchange material
Liner shipments
EPICOR II liners
Zeolite/resin system liners(reactor building sump)

N)

0.016

69

1.1

69

1.1

0.024

8

0.3

33

0.8

Primary system cleanup resins(RCS)

0.016-0.024

8

0.2

68

1.6

Miscellaneous shielded.shipments

0.022

40

0.9

208

4.6

297

11.3

814

37.8

TotalseaCrew dose based on number of packages per shipment and exposure estimates in Appendix N.
bFrom Table 9.6, excluding damaged fuel assemblies.
CUnshielded trash drums include AFHB and RB decontamination solutions and compactable trash.
dUnshielded LSA boxes include noncompactable trash, contaminated equipment, irradiated hardware
and mirror insulation.
eExclusive of fuel packages.
The occupational exposure during handling of fuel packages is
included in the discussion of environmental
Section 6.4.

impacts during core examination and defueling,9-30

Table 9.11.

Transport Parameters for Calculation of Occupitional Doses
for Waste and Spent Fuel Shipments from TMI-2
High
Population
Areas

MediumPopulation
Areas

Low
Population
Areas

Average speed (mph)

15

25

55

Fraction of travel distance

0.05

0.05

0.9

Parameters
Transport Parameters

Distance for TMI shipments (miles)- Longest route- Shortest route
Duration of exposure time (hrs)- Longest route- Shortest route

138
19
9.2
1.3

138
19
5.5
0.8

2475
333
45.0
6.1

aThe number of shipments to all locations ranges from a minimum of 353 to a maximumof 997 (see Table 9.6).

those hours in the cab and 2 hours at a distance of 3 ft from the package (or cask), his/her
If themaximum possible dose per trip is 182 mrem (2 mrem/hr* x 58 hr + 33 mrem/hr** x 2 hr).
Thesame crew member made 25 such trips per year, his/her annual dose would be 4.6 rem.
comparable annual cumulative crew dose would be 9.2 person-rem, and range from 130 person-rem to360 person-rem for the range of shipments of all TMI wastes for the time interval required.
Maintaining
In this case, the total exposure duration would be 8.2 hours.
Shortest Route.
the same proportions of the total hours in the cab and at 3 ft from the package results in a
maximum individual dose per trip of 25.7 mrem and an annual individual dose of 0.8 rem for 30 trips.
The crew cumulative dose would be 1.3 person-rem, and range from 18 person-rem to 51 person-rem
for shipment of all TMI waste.
However, experience indicates 6 that dose rates in the cab of an exclusive-use truck are usually
On this basis the crewless than 0.2 mrem/hr, and about 25 mrem/hr at 3 ft from the package.
doses would be as follows:
The truck crew member would receive about 62 mrem per trip, and a maximum
Longest Route.
For all shipments of TMI-2 waste the total cumulative occupaannual individual dose of 1.6 rem.
tional radiation dose would range from about 44 person-rem to about 124 person-rem.
In this case, the truck crew member would receive about 9.1 mrem per trip,
Shortest Route.
For all shipments of TMI-2 waste the total
and a maximum annual individual dose of 0.2 rem.
cumulative occupational radiation dose would range from about 6 person-rem to about 18 person-rem.
Offsite Storage and Disposal
The occupational doses to workers at the offsite storage sites are not treated here as additional
impacts from the TMI-2 waste shipments; instead, such doses represent a component of the normal
occupational doses associated with handling of all waste shipments at those facilities.

*Dose rate in the cab.
**Dose rate at 3 ft from package.

9-31
9.5.1.2

Environmental(offsite) Exposures

Onsite Storage
The environmental exposures to the offsite population in the vicinity of TMI-2 resulting fromonsite storage and handling operations are considered to be negligible.
As previously noted,
maximum radiation levels at the fence surrounding the facility from the interim storage andstaging facility will be less than 0.6 mR/hr.
The nearest offsite location where members of the public can be expected to gather is the
Visitors Center.
It is conservatively postulated that there would be an average of ten people inthe center for ten hours each day every day of the year.
Under these conditions, the populationdose would be about 0.0004 person-rem per year.
Transportation
Environmental exposures to the general population for truck transportation of the TMI waste
involve exposure to people residing along the shipping route, exposure to persons in other vehicles on the route, and exposure to bystanders while the transport vehicle is stopped.
People Residing Along the Shipping RouteLongest Route.
An estimated 700,000 persons who reside along a 2750-mile route from TMI to thedisposal site might receive a cumulative population dose of about 0.05 person-rem for one shipment and a range of from 17 person-rem to 50 person-rem for all TMI waste and fuel shipments.
These doses were calculated for persons in an area between 100 ft and 1/2 mile on either side of
the shipping route, assuming 330 persons per square mile, 10 mrem/hr dose at 6 ft from the vehicle and each shipment traveling 200 miles per day.
Shortest Route.
The estimated .125,000 persons residing along the 370-mile shortest route from
TMI to the disposal site might receive a cumulative population dose of about 0.007 person-rem forone shipment and a range of from 2 person-rem to 7 person-rem for all TMI wastes and fuel shipments.
The assumptions made are the same as for the longest route case.
The maximum dose received by a person living along any transport route, as determined in
NUREG-0170, 6 would probably be received by an individual living adjacent to a highway whereradioactive material was frequently shipped.
The dose for one shipment received by a person
living 100 ft from a roadway on which irradiated fuel shipments (worst case) would pass at an
average speed of 30 mph would be 3.6 x 10-5 mrem.
The total dose received by the same person for
all TMI fuel shipments would range from 0.001 mrem to 0.004 mrem and, conservatively'assuming thedose from a waste shipment to be the same as for the fuel, would range from 0.013 to 0.036 forall waste shipments.
Persons In Other Vehicles on the Route.
The dose received by an individual driving 100 ft behind
the truck (the location of probable maximum dose for personssharing the transport route) for one
hour would be 0.1 mrem.
However, the staff believes that it is highly unlikely any individual
driver of the general population would follow the truck for even such a long time, and insteadwould likely pass the truck or turn off after a short while.
Onlookers While Vehicle is Stopped.
Members of the general public might be exposed to radiationfrom shipments of waste at points where the trucks stop along the route.
A member of the generalpublic who spends 3 minutes at an average distance of 3 ft from a loaded truck might receive adose of up to 1.3 mrem.
If ten people were so exposed during a shipment, the population dose foreach shipment would be 0.013 person-rem, and for the total number of waste shipments from TMI, a
population dose of from 5 person-rem to 13 person-rem might result.
Offsite Storage and Disposal
The environmental exposures to the offsite population adjacent to licensed storage and disposalsites are not treated as. additional impacts from the TMI waste shipments.
Instead, they are
considered as a component of the normal doses associated with the handling of all waste shipments
at that facility, regardless of origin.
As such, these exposures are covered in the environmental impact statements for the specific sites.

9-32
9.5.1.3

Nonradiological

Impacts

Nonradiological impacts on the environment from the normal storage, transport, and disposal of
In the case of onsite storage andthe TMI waste are primarily in the form of resource use.
transport, the resource use takes the form of a commitment of packages and shipping containers to
The impact of such a commitment on the radioactive materials transportation
the TMI waste.
industry is considered to be insignificant, except in those instances where specialized casks are
required for the shipment of the irradiated fuel or other high-activity material requiring Type B
packaging.
In those instances, prior scheduling and coordination within the industry will be
required to ensure availability of these packages on as-needed basis.
The major nonradiological impact will occur in the offsite storage and disposal of the waste
The packages that are disposed of by burial will use part of the limited capacity inpackages.
If a burial efficiency of 50 percent isthe existing commercial low-level-waste disposal sites.
assumed for "random" placement of packages, based on current burial practices, burying the TMI
For apackages will require from 243,000 ft 3 to 645,000 ft 3 of trench volume (see Table 9.12).
typical trench 300 ft long by 30 ft wide by 20 ft deep, from two to four trenches will be needed
for the TMI waste.
Since the total remaining commercial waste disposal capacity is approximately
75 million cubic feet this will reduce the total commercial capacity available for disposal ofwastes from other sources by 0.3 to 0.9 percent.
In the case of the intermediate level wastes that cannot be disposed of by normal shallow land
The form of
burial, interim offsite storage may be used until final disposition is determined.
these wastes would be such that subsequent additional processing options for conversion into
Assumwaste forms acceptable for the method used for final disposition would not be foreclosed.
ing a storage efficiency of 75 percent based on current technology, storing the HSAW packages
will require from 4000 ft 3 to 5100 ft 3 of storage capacity (see Table 9.13).
The irradiated fuel, if shipped to a reactor or AFR facility for storage, will utilize available
For a range of 56 to 183 fuel elements, the cross-sectional area ofcapacity in a storage pool.
storage pool space required will be from 66 ft 2 to 216 ft 2 . The need for additional pool storage
capacity will at some point exist for all AFR shipments for the nuclear power industry, with the
TMI fuel elements representing only a very small increment of the total requirement.
9.5.2

Accident Effects

The other than "normal" occurrences associated with storage, transportation, and disposal can be
Abnormal events include occurrences thatcategorized as either abnormal events or accidents.
(1) compromise package integrity, such as dropping of packages by material handlers, packages
being run over and crushed by a vehicle, and skewering of packages by a forklift, and (2) relate
Accidents are considered, in this context,
to packaging and handling procedures or package loss.
Thus, abnormal occurrences are related to
to involve vehicles carrying the radioactive waste.
The packaging
packaging and handling activities and accidents during transportation activities.
and handling activities have been previously covered in Section 8; transportation accidents arecovered in this section.
9.5.2.1

Transport Accident Probability

The probability of the occurrence of a transport accident that results in release of radioactive
material can be described in terms of the expected number of accidents for the transport mode,
together with the package response to those accidents and the dispersal that is expected.
Accident Rate
The accident rate used in this assessment for truck transport is 1.7 x 10-6 accident/mile (andfor rail is 1.5 x 10-6 railcar accident/railcar mile). 6 Therefore, there is a probability of one
accident occurring for every 314 truck shipments along the longest route (2750 miles) from TMI toa disposal site (worst case) and one for every 1590 truck shipments along the shortest route
If all the waste is shipped on the long(370 miles) from TMI to the disposal site (best case).
route, it is estimated that 1 to 3 accidents can occur for the range of waste shipments from
TMI-2.
On the other hand if all the waste is shipped on the short route, 0.2 to 0.6 accident canoccur.

9-33

Table 9.12.

Volumes of Packaged Solid Waste to Be Disposed of at a Commercial
Low-Level Waste Disposal Site
Best-Case Conditions

Type of Package

Package
Volume(ft3)

Worst-Case Conditions

Numberof
Packages

ShippedVolume(ft3)

BuriedVolumea(ft3)

Numberof
Packages

ShippedVolume(ft3)

BuriedaVolume
(ft3)

7.5
7.5

3,200
502

24,0003,765

48,0007,530

15,400
1,707

115,500
12,800231,000
25,60080

1,042

83,360167,7202,128

170,240

340,480

70
80

86
53

6,020
4,240

12,0408,480

29320,51041,020

-

-

-

50
50
175

49
14
6

2,450
700
1,050

4,900
1,400
2,100

49
14
6

2,450
700
1,050

4,900
1,400
2,100

10
50
190

11
2
1

110
100
190

220
200
380

.11
4
2

110
200
380

220
400
760

10/7.5/150
50
190

16
4
3

990
200
570

1,980
400
1,140

57
44
12

1,340
2,200
2,280

2,680
4,400
4,560

55-Gallon Drums
Low activity
Intermediate activity
LSA Boxes
Low activity
Contaminated Equipment
and Hardware, Mirror
Insulation
EPICOR II

Resins

1st stageb
2nd stage
3rd stage
RB Sump Cleanup
Filtersc
2nd stage
3rd stage
Primary System Cleanupc
Filtersd
2nd stage
3rd stage
Totals

256,520

659,520

aAssumes 50 percent efficiency.
bWill require special disposal procedures (e.g., deeper burial) if disposed of at a commercial
disposal site.
cIf any of these wastes contain fuel debris or greater than 10 nCi/gm transuranic materials,
they would not be accepted at a commercial LLW facility.
dprimary system cleanup generates 3 filter types.

9-34

Table 9.13.

Volumes of Waste Packages to Be Storeda

Total Volume of
Number of
Packages

VolumeShipped (ft3)

Storage Space
Used (ft3)

Best
Case

Worst
Case

Best
Case

Type of Package

Best
Case

Worst
Case

Volumepe
perPackage(ft3)

SDS 1st stage

11

78

10

110

780

150

1,040

EPICOR II

46

46

50

2,300

2,300

3,070

3,070

3,220

4,110

1st stagec

Total

Worst
Case

apackages containing wastes from RCS cleanup that contain fuel debris or greater than
10 nCi/g transuranic materials also will require storage.
b75 percent storage efficiency.
cTo be stored if

disposal at commercial burial sites with special precautions is not

feasible.

Airborne Release Fraction for Accidental Release
To arrive at a realistic worst-case scenario for airborne releases under accident conditions, asevere truck accident sufficient to breach a Type B container with an attendant fire is postulated.
As previously discussed, this accident is highly unlikely because the integrity of theType B container has been demonstrated by engineering analysis and testing. Airborne releases
would occur if the container ruptured and there was a fire or explosion. To determine airborne
release fractions for the TMI waste forms, it is necessary to extrapolate data from othersources.
In the absence of specific test data on TMI-2 wastes, release fractions for the TMI
waste forms are assumed to be similar to fractions obtained from tests of simulated high-level
waste immobilized in glass.8 It was found that after being subjected to container-rupture forces
and fire, the fraction of the vitrified waste sufficiently small (less than 10 mm)* to become
respirable was 5 v 10-5 for a container velocity of 55 mph on impact.
It can be further assumed
that only 20 percent of this fraction would actually be released from a Type B container,9 giving
an overall respirable release fraction of 1 v 10-5.
The staff believes this would be a bounding
worst-case situation.
The potential for airborne releases from representative Type B waste
packages under combined impact (at a velocity of at least 50 mph) and fire is given in
Table 9.14.
The frequency of this specific type of accident has been estimated to be 6 v 10-13
accidents per truck mile, or 1.6 v 10-9 for shipments on the longest route, and 2.2 v 10-10 for
shipments on the shortest route.
This is the minimum accident that must occur to sufficiently
breach a Type B package to cause a release of radioactivity.
Waterborne Release Fraction for Accidental Release
The accident scenario resulting in waterborne releases requires that a truck fall into a water
body, the containers be punctured, and the waste forms be solubilized and carried out of thecontainer and vehicle.
The solidified waste forms will be relatively insoluble in water, and it will take some time forwater penetration and radionuclide mobilization t6 occur, permitting mitigating measures to beundertaken to minimize or prevent dispersion. However, if conservative release assumptions are
made, a waterborne release fraction of about 1 v 10-6 is obtained for a breached package under
prolonged leaking conditions.
These assumptions include use of a solidified waste surface area

*Particle size that becomes airborne and can be inhaled.

9-35

Table 9.14.

Estimated Airborne Releases from Selected Type-B
Waste Packages under Accident Conditionsfor 1 x 10-5 Fractional Release

Type of
Packagea

Package Content

LinersResinsCurie
Content
per
PackageReleaseper
Package(Ci)

NumberPackagesper
Shipment

Releaseper
Shipment
(Ci)

0.013

1

0.013

1 .2b

1

1.2

EPICOR II

1st stage

1,300

SOS zeolite

1st stage

120,000

Drum

Filter40<0.00114

0.006

Evaporator bottoms

36<0.001

14

0.005

Accident sludge

435<0.004

4

0.017

Ash

2.0

<0.00114<0.001aAll shielded.
bRelease of individual radionuclides per package for this worst caseCs-137, 1.008 Ci; Sr-89,

are:Cs-134,0.168 Ci;

0.0034 Ci; and Sr-90, 0.0144 Ci.

of 25 ft 2 before the accident, an area increase factor of 5 because of the accident, an average
waste weight of about 1300 pounds per drum, a solubility of 1 x 10-6 cm2 -day, and five days ofleaching.
Applying this release fraction to an accident involving Type B packages results in thepotential waterborne releases shown in Table 9.15.
The frequency of such an accident would beless than one that would involve airborne releases, and would be dependent on the frequency andduration of proximity to water bodies during transport.
9.5.2.2

Transport Accident Radiological Risk

Once release of the radioactive material has occurred, radiological impacts would result fromdispersion along a number of exposure pathways to man, including inhalation, ingestion, or direct
radiation.
Models have been developed to assess these impacts along the viable exposure pathways. 6
In addition, an accident may result in environmental contamination of land or structures,
necessitating subsequent abandonment or cleanup operations.
The meteorological conditions that would exist if a transportation accident were to occur are
difficult to predict since they would change with location and time.
Relative to the severe
truck accident described earlier in this section, the dose from inhalation was calculated assuming meteorological conditions equivalent to x/Q = 5 x 10-1 sec/m3 .
The total body inhalationdose that would be received by an adult breathing at a rate of 8000 m3 /yr would be 100 mrem.
This dose is for an individual assumed to be within several hundred feet for the duration of theaccident.
If the accident occurs near a vegetable garden, it may be necessary to take precautions to prevent large doses from vegetable consumption.
The inhalation dose calculated represents a conservative value and may be lower depending upon meteorological conditions at thetime of the accident.
9.5.2.3

Severe Accidents in Very High Population Density Urban Areas

Accidents occurring in urban areas of very high population density (> 10 4 /km 2 ) may produce consequences more serious than discussed.
This type of accident has a very small probability ofoccurrence which would be further diminished by the proposed new rules that would require thatshipments of radioactive material be made on circumferential or bypass routes around urban centers.

9-36

Table 9.15.

Estimated Waterborne Releases from Selected Type-B
Waste Packages under Accident Conditionsfor 1 x 10-6 Fractional Release

Type of
Packagea

Package Content

LinersResinsCurie
Content
per
PackageReleaseper
Package(Ci)

NumberPackagesper
Shipment

Releaseper
Shipment(Ci)

EPICOR II

1st stage

1,300

0.001310.0013

SOS zeolite

1st stage

120,000

0 . 12b

1

0.12

40

<0.0001140.0006

Evaporator bottoms

36<0.0001140.0005

Accident sludge

435<0.000440.0011Ash

2.0<0.000114<0.0001-

Drum

FilteraAll shielded.
bRelease of individual radionuclides per package for this worst case are:Cs-137,

9.5.3

Cs-134, 0.0168 Ci;

0.100 Ci; Sr-89, 0.00034 Ci; and Sr-90, 0.0014 Ci.

Psychological-Socioeconomic Effects

Assuming two shipments per day, the transportation of solid waste would involve either 353 or
997 truck shipments over 14 or 34 months, depending on best- or worst-case conditions.
The
worst-case scenario for shipping solid waste would be modified further if a decision were made totransport the tritiated water offsite by truck; under this scenario, an additional average of
1.2 truck shipments per day would be made over a 16-month period.
Incident-free transportation of waste from the site is expected to produce little
impact. 1 0 The
current level of public sensitivity to and awareness of TMI-2 activities leads the staff to
expect that the psychological and socioeconomic impact generated would be concentrated in thepeople living near the shipping route within the local impact area.
The psychological impact is
expected to be characterized as low-level anxiety, and would reflect apprehension of the possibility of accidents.
The staff also believes that the transportation of waste through Middletown, where population density is higher and people live closer to the roadway than at otherlocations, in the vicinity of TMI, could result in the decreased marketability of residential
property during the period of shipments.
Shipments through Middletown also could result inadjustments in daily schedules and activities as households attempt to avoid the shipping route
through Middletown.
The staff considered three accident scenarios involving offsite transportation:(1) incidents
not involving environmental releases, (2) incidents resulting in releases to the air, and(3) incidents resulting in waterborne releases.
Accidents involving airborne or waterborne
releases are considered to be remote, although accidents not accompanied by releases to theenvironment are considered probable.
If an accident does occur in the local area of impact, it
would have the potential for producing a notable public response' 0 because of current concerns
focused on nuclear waste disposal.
The nature and extent of any psychological or socioeconomic
consequences are expected to be determined by the event, the factuality and duration of media
coverage, the number of people impacted, and the danger posed to public health.
9.6

ECONOMIC COSTS

The incremental and total costs involved in the transportation of the solid waste and irradiatedfuel from TMI, and the subsequent offsite storage, treatment (if applicable),, and disposal
charges are summarized in this section.
In general the cost elements are (1) shipping cask use

9-37
or rental, (2) transportation, and (3) commercial burial.
used to quantify these cost elements and the incremental

The basic approach and methodologies
values are provided in Appendix K.

For the purpose of bounding disposal costs, the costs associated with both the maximum transit
distance to Hanford (Richland), Washington, and the minimum distance to West Valley, New York,are determined for both the least and greatest number of shipments for each type of waste.
The range of estimated costs for transportation and storage of LLW and HSAW are between$2,610,000 and $6,680,000 for disposal at West Valley, New York, and $3,900,000 and $11,700,000
for disposal at Hanford, Washington.
Although it is recognized that there are cost factors associated with the processing and storageof the damaged fuel, sufficient data are not currently available to assess these considerations.

References--Section 9
1.

G.W.

2.

Correspondence,
1980.

J.F. Ahearn,

3.

Correspondence,
1980.

W. Dircks to C.W. Bateman,

4.

M. Rogovin et al., "Three Mile Island - A Report to the Commissioners and to the Public,"
TMI 18:52, Volume I, April 5, 1979.

5.

"Supplement to Analysis of Three Mile Island - Unit 2 Accident,"
Center-I, Palo Alto, CA, October 1979.

6.

"Final Environmental Statement on the Transportation of Radioactive Material
Other Modes," U.S. Nuclear Regulatory Commission, NUREG-0170, December 1977.

7.

"Environmental Survey of Transportation of Radioactive Materials To and From Nuclear Power
Plants," U.S. Atomic Energy Commission, WASH-1238, December 1972.

8.

T.H. Smith and W.A. Ross, "Impact Testing of 'Vitreous Simulated High-Level Waste in Canisters," Battelle Pacific Northwest Laboratories, BNWL-1903, 1975.

9.

W.A.

10.

N.C. Finley et al., "Transporation of Radionuclides in Urban Environs:
Assessment," NUREG/CR-0743, Appendix L, Social Impacts.

Cunningham, Testimony to TMI Advisory Panel,
to C.W. Duncan,

Brobst, "Transportation Accidents,

December 18, 1980.

U.S. Nuclear Regulatory Commission,

October 20,

U.S.

January 7,

Nuclear Regulatory Commission,

How Probable?"

Nuclear Safety Analysis

Nuclear News,by Air andMay 1973.
Draft Environmental

10.

10.1

SUMMARY OF ENVIRONMENTAL IMPACTS OF THE PROPOSED ACTIVITIESSUMMARY OF EFFLUENTS AND RELEASES TO THE ENVIRONMENTIn Sections 5 through 8 of this document, various alternative processes are described for each of
the major stages in decontaminating TMI-2, for processing the radioactive wastes, and for preparing the wastes for shipment to offsite waste repositories.
In Section 9 the procedures fortransportation and terminal disposal of these wastes are described.
For each of the cleanup
activities, estimates were made of the potential releases of radionuclides for each alternative
process.
A detailed description of the effluents from each alternative process can be found in
the appropriate sections.
In some cases, a range of values representing the uncertainty of theestimates is indicated.
The summary presented here is the summation of the amounts released fromthe process alternative that produces the greatest quantity of effluent of all processes considered for a particular activity. In cases where a range of values is given, the largest quantity is included in the summation.
Essentially all of the Kr-85 that was contained in the reactor building air (about 44,000 Ci) was
vented to the atmosphere over the two-week period prior to July 11, 1980.
There have been subsequent purges to release the Kr-85 that has been slowly desorbing from the water and walls.
The
monthly releases during September 1980 through December 1980 decreased in an approximately
exponential manner (27, 15, 12 and 7.5 Ci).
Future releases of radioactivity into the atmosphere
will be a direct consequence of the normal decontamination processes, solidification and packaging of the wastes, and disposal of the fuel* During the normal decontamination operations,gases and very small particles of solid (dust) radionuclides will continually escape to thebuilding atmosphere.
The building ventilation systems contain HEPA filters that will trap mostof the dust; however, some of the particles and all of the gases will pass through the filtersand escape to the atmosphere.
These effluents will accompany the operations throughout theentire period of the cleanup.
The maximum estimates for the principal radionuclides which could
be expected to be released during the course of the cleanup are summarized in Table 10.1.
During the course of the TMI-2 accident, a considerable quantity of water was contaminated withradionuclides as both dissolved and suspended solids.
About 372,000 gallons of water flowed fromthe reactor system into the auxiliary and fuel handling buildings.
This water has been processedby the EPICOR II system.
About 700,000 gallons of contaminated water are in the bottom of thereactor building and will be processed during cleanup by one of three alternative means.
The
process chosen will also likely be used to process the nearly 96,000 gallons of water in thereactor coolant system.
An additional source of contaminated water will result from the use of
aqueous solutions or water sprays for decontamination of reactor building surfaces.
This water
also will need to be processed.
The purpose of processing the water is to remove the suspended
and dissolved radionuclides for subsequent immobilization and shipment to an offsite repository.
However, the processed water will still contain some radionuclides, mainly tritium as HTO, which
is not removed by the processing.
Table 10.2 is a summation of the total radionuclide inventory
remaining in all the processed water sources.
The "best-case" and "worst-case" values represent
the expected range of the residual radionuclide inventory in the processed water.
This processedwater must ultimately be utilized or disposed of by one of the means previously discussed in thisdocument.
The discharge of processed water (tritium is the major remaining contaminant) to the SusquehannaRiver would occur only as a deliberate choice (and only after approval by the NRC) of that meansof disposal.
The processed water would be diluted with blowdown or river water and then discharged to the river at controlled rates such that the concentration of radionuclides in theriver would be well below the threshold level for deleterious effects in aquatic species orhumans.
Table 10.2 is a summary of the total amount of the principal radionuclides that would bedischarged under the above conditions.

10-1

Table 10.1.

Maximum Estimated Amounts of the Principal Radionuclides Released to the Atmosphereas a Consequence of Normal Cleanup Operations at TMI-2a
Radionuclide Releases (Ci)bDocument
Document
SectionOperation

H-3

Kr-85

Sr-89

Sr-90

Ru-106-

1.5x1O-5

9x10-s

-5.1.4.1

Decontamination of the AFHBc

5.2.1.1,
5.2.4.1

Decontamination of the reactor
building

750

44,0006.3.4.1,
6.4.4.1

Removal of RPV head and internals, core examination, anddefueling

500

140

6.5.4.1,

Decontamination of primary systemcomponents

300

7.1.4.1

Liquid waste treatmentd

7.2.4.1

Disposal of processed watere0.1

Sb-125

Te-127m

5xlO-6

Cs-134

Cs-137Ce-1441.4x10-4

7x10-4-1.4x1O-5

8x10- 5-C

Storage tank venting
Evaporation from lined ponds
Forced evaporation in cooling towerf-

7x10_4

7x1O-3

0.1

2900
2

2700

-

1.7x10_6

3x1O-5

1. 2xl0-2

2x10_

Worst case

2700

-

1.7x10- I

3

6

15

-

8X10-6

8x1O-s4xlO-6

3x10_7

Immobilization of process solid wasteg

8.2.4.1

Immobilization oA chemical decontamination solutions
Packaging-and handling solidmaterials

5x10_2-8x10-3

Best case

8.1.4.1

8.3.4.1

0.3--

1.3x10- 3

3x10152<7x 10_2

1.6x10-7x10-33x 10-1

1.4

1.5

1.6x 10- 5

1. IxlO-4

1. Ox 10-6

8

7xlO-8-8x10-

6x10-7

6x1O-3-3x1O_2

2x10-1

Table 10.1.

Radionuclide Releases (Ci)bDocument
Section
9.4

Continued

Operation

H-3

Kr-85

Sr-89

Sr-90

Ru-106

Sb-125

Total (for water disposal to river)k1600

44,000

0.002

0.013

4x10-6

3x10-

Total (for water disposal by naturalevaporation)

2900

44,000

0.002

0.013

4x10-62900

44,0000.17

3

6

Te-127m

Cs-134

Cs-137Ce-144

-

0.04

0.3

1X10-6

3x10-7

-

0.04

0.3

1X10- 6

15

15

0.3

1.7

1.5

Storage, transportation, anddisposal of fuel and solid wasted
7

Total (for water disposal by forced
evaporation)aThe releases for all radionuclides except H-3 and Kr-85 were calculated on the basis of an overall penetration factor of 0.001 for two
HEPA filters in series (see Sec. 5.1.4.1).
bValues and totals are rounded to one, or in some cases two,cEffluents and releases prior to September 1,

significant digits.

1980, which were all below technical specification

limits,are not included.

dSee Table 7.11.
eThe three entries under water disposal correspond to the three alternatives of:(2) natural evaporation;
fSee Table 7.27.

and (3)(1) storage and controlled release to the river;forced evaporation.

gsee Table 8.17.
hSee Table 8.37.
iSee Tables 8.53 and 8.54.
jUnder normal storage, transportation, and disposal conditions the waste and fuel remains sealed in containers so that no effluents or releases occur.
kThis total corresponds to best-case water disposal conditions, i.e., disposal by controlled release to the river (for which
airborne releases would be negligible) and worst-case conditions for all other operations.
This is the total amount of H-3 in the accident water from the AFHB (370 Ci), reactor building (2500 Ci)

1and RCS (30 Ci).
The individual releases for different operations add up to more than this total because mutually exclusive worst-case conditions wereconsidered (e.g., tritiated water that evaporates from the spent fuel pool during defueling cannot also be released during liquid waste
processing).
mThe worst case (processing the water with the SOS alone) is assumed.
Note, however, that release of effluent to the atmospheremust not result in offsite doses exceeding those proposed in Appendix R and discussed in Section 1.6.3.2.

W-h

10-4
Table 10.2.

Radionuclides

Amounts of Principal Radionuclides that Will Be
Present in All of the Stored Watera
Total Radioactivity in Processed Water (Ci)b
Worst Case (SDS)cBest Case (SDS/EPICOR II)c

H-3

2900

2900

Sr-89

6 x 10-6

0.6

Sr-90

9 X 10-5

9

Ru-106

0.04

21

Sb-1250.07

54

Te-127m

0.1

51

Cs-134<0.3

0.9

Cs-137

0.6

5

Ce-1440.02

5aThe total volume of stored processed water would be slightly
over 1.5 million gallons if no clean water were added and none
was lost by evaporation.
The origins of this water are:743,000 gallons from the AFHB that has already been processedby EPICOR II, 700,000 gallons of contaminated water in thereactor building basement that has not yet been processed, and96,000 gallons of water in the primary system of the reactor
that also remains to be processed (see Tables 7.23 and 7.24).
If the processed water were released to the river, the rate
and the mixing with uncontaminated water would be adjusted sothat the concentration of radionuclides in the river would bewell below the threshold level for deleterious effects inaquatic species or humans.
bValues are rounded to one or two significant digits.
cSee Section 7.1.3.3 for a discussion of these systems.

Except for the small releases to air and water (if approved), by far the greatest part of theradionuclides from the accident will be immobilized as solids and shipped offsite to authorized
waste disposal facilities.
It is very likely that the bulk of these shipments will be by truck
as is presently the case.
Although shielding is generally not necessary for truckloads of drums
or low-specific-activity (LSA) boxes, high-specific-activity ion-exchange materials and spent
nuclear fuel must be packaged in shielded casks for shipment. A summary of the estimated numberof solid waste shipments, broken down by type of packaging, is presented in Table 10.3.
The potential accidents that could occur during cleanup operations are described in Sections 5
through 9 of this document.
These accidents are postulated on very conservative assumptions, andthe releases to the environment are similarly computed.
The estimated amounts of the principal
radionuclides that would be released to the atmosphere for each accident are listed in Table 10.4.
For those accidents that would result in releases of radioactive materials into the river, theconcentrations of radionuclides at the nearest intake for potable water would be a small fractionof the limits set forth in 10 CFR Part 20 for unrestricted releases.
Details of each postulated
accident and the effluent estimates can be found in the document section listed in the firstcolumn of Table 10.4. -It
is assumed by the staff that two or more accidents would not happen
simultaneously because the cleanup activities with which the accidents are associated will takeplace at different times or at different places on the site, or both.

10-5

Table 10.3.

Summary of Estimated Number of Waste Shipmentsa

Number of Shipments
Type of Waste

Best Case

Worst Case

Drums

13

108

LSA boxes

104

177

Unshielded drums

14

20

Shielded drums

0119

Shielded ion-exchange materials

85

170

Accident sludge, spent
filters, ash

41

12Contaminated equipment,materials, hardware

40

208

Damaged fuel assemblies

56

183

353

997

Low-level solids

Decontamination liquids

Total

aSee Table 9.6 and accompanying discussion in Section 9.

10.2

SUMMARY OF OCCUPATIONAL DOSES AND HEALTH EFFECTS

Estimates for the occupational doses are presented and discussed individually for each of the
major cleanup operations in appropriate subsections in Sections 5 through 9.
Estimates of theduration of the work effort are included in the discussion.
These estimates are summarized
below.
10.2.1

Occupational DoseThe cumulative occupational radiation dose from all decontamination, defueling and waste disposal
operations is estimated to be between 2000 and 8000 person-rem.
Contributions from the different
cleanup operations are summarized in Table 10.5. Defueling, primary system c-leanup, and related
operations contribute almost half of the total; decontamination of the reactor building contributes about one-third, and the remaining operations, which include decontamination of the auxiliary
and fuel handling building, processing the contaminated water, and packaging and disposing of thewaste, contribute the remainder.
10.2.2

Duration of Effort

The total time that will be needed to complete decontamination and defueling is estimated by thestaff to be 5 to 9 years from the time of the accident on March 28, 1979.
This estimate does nottake into account delays due to financial problems, litigation, or other unpredictable events;
the actual time could well be longer.
The licensee's schedule calls for completion of cleanup
activities in 1986, approximately 7 years from the accident (see Fig. 1.4).

Table 10.4. Amounts of the Principal Radionuclides Released to the Atmosphereas a Consequence of Postulated Accidents during Cleanup of TMI-2

Document
Section
5.1.4.2

5.2.4.2

6.3.4.2
6.4.4.2

6.5.4.2

7.1.4.2
8.1.4.2

Radionuclide Releases (Ci)aAccident

Kr-85

Sr-89

Sr-90

HEPA filter failure duringdecootamination of theAFHB-

1.5 x I0-s

HEPA filter failure duringdecontamination of thereactor building
Release of trapped fission
products during removal of
the RPV head and internals,
core examination, anddefuelingRu-106

Sb-125

Cs-134

Cs-1379 x 10-5

1.4 x 10-4

7 x I0-4

5 x 10-6

1.4 x 10-s

8 x I0-5

Cs-14470

Spill of decontamination
liquids from reactor
coolant system

8 x 10-5

8 x 10- 4

1.6 x 10-4

1.0 x 10-3

HEPA filter failure duringliquid waste treatment

77x10-4

7 x10-3

8 x 10-3

5 x 10-224104
x

2 x10- 1-

5 x 10-1

3 x 10-3

Package handling accidents
during immobilization ofprocess solid wastes from:EPICOR II---

44x 10-2

Modified EPICOR II
SDS system
Other wastesc-

4 x 10-1

8 x 10-3

8 x 10-5

8x 10-2

10-2

1.2 x 10-1

4 x 10- 3

4 x 10- 4

1.6

55x 10-4

9 X 10-3

1 Xi0-4

1 X 10-4Ix

4 x 10-3

104
9 x 10-2 x 10-2

7 x 10-5

Table 10.4.

Radionuclide Releases (Ci)aDocument
Section
8.2.4.2

Accident

8.3.4.2

9.5.2

Kr-85

Sr-89

Sr-90

Ru-106

Sb-125

Cs-134

Cs-137

Cs-144

Package handling accidents
during immobilization ofchemical gecontamination

solutions
8.3.4.2

Continued

x 10-8-Storage area fire during
packaging and handling-

1.3x i0-7

9xx 10-6 x 10-3

3 x 10-2

2 x 10-4

of solid materials-

9 x 1 0 -4

3 x 10-3

Package handling accident
while packaging, handlingand storing solid waste-

6 x 10-6

2 x 10--4 x 10-5

Transportations accidentsf-

3 x 10-3

1.4 x 10-2-

1.7 x 10-1

1.0

avalues are rounded to one or two significant digits.
blncludes releases from surface decontamination and desludging (see Table 5.2).
cOther wastes include accident sludge,

immobilized evaporator bottoms and bituminized decontamination liquids (seeTable 8.20).
dEntries are the releases for an accident with a single package (see Table 8.38).
be handled (see Table 8.36).
eThis is the release from a single package containing mirror insulation.

A maximum of about 5000 packages will

The release from an accident with one ofthe other package types, which include drums of compactible trash or incinerator ash, or LSA boxes containing noncompactible trash, would be less (see Table 8.57).
The number of packages handled would be in the range of 3000 to 15,000(see Table 8.50).
fReleases are those expected for a transportation accident involving a type-B package containing first-stage SDS
zeoliteliners which would have a radioactivity content of 120,000 Ci per package (see Table 9.14).
The number of packagesof this kind that will be shipped should not exceed 90 (see Table 9.14).

10-8
Table 10.5.

Summary of Cumulative Dose and Health Effects
to Workers from Cleanup of TMI-2
Health Effectsa

Operation

Cumulative
Occupational
Dose(person-rem)

Maintenance of the Reactor
in Safe Condition

8

0.001

0.002

Decontamination of the
Auxiliary and Fuel Handling
Buildings

375 - 550

0.05 - 0.07

0.10 - 0.14

5.2.5.1

Decontamination of theReactor Building

660 - 3000

0.09 - 0.4

0.2 - 0.8

6.2.5.1

Reactor Coolant System
Inspection

52 - 580

0.007 - 0.08

0.014 - 0.15

Removal of RPV Head andInternals

150 - 450

0.02 - 0.06

0.04 - 0.12

6.4.5.1

Core Examination andDefueling

580 - 1350

0.08 - 0.2

0.15 - 0.4

6.5.5.1

Decontamination of Primary
System Components

108 - 1740

0.014 - 0.2

0.03 - 0.5

7.1.5.1

Liquid Waste Treatment

43 - 121

0.006 - 0.016

0.01 - 0.03

8.1.5.1

Handling and Packaging ofProcess Solid Wastes

17

0.002

0.004

Handling and Packaging
of Chemical Decontamination
Solution Wastes

3 - 100.0004 - 0.001

0.0008 - 0.003

Handling and Packaging ofSolid Wastes

39 - 99

0.005 - 0.013

0.01 - 0.03

Transfer from Storage andTruck Loading

11 - 380.001 - 0.005

0.003 - 0.009

Transportationb

6 - 360

0.001 - 0.05

0.002 - 0.09

0.3 - 1

0.5 - 2Document
Section
4.5.1
5.1.5.1

6.3.5.1

8.2.5.1

8.3.5.1
9.5.1.1

9.5.1.1
Totals2 000

Additional
CancerDeaths in
Work Force- 8000a

Additional
Genetic Effects
Among Offspringof Work Forceavalues have been rounded to one or two significant digits; totals have been rounded to one
significant digit.
bDifferent routes and different estimates for the expected exposure during transit lead to alarge range in the transportation estimates;

see Sec.

9.5.1.1.

10-9
The basis for the staff estimate is (starting from the date of issuance of this PEIS, approximately 2 years from the time of the accident):
18 to 48 months to decontaminate the reactor
building to the point that defueling activities can begin; 3 to 10 months to remove the reactor
pressure vessel head, upper internals and core support structure; 8 to 13 months for core examination and defueling; 3 to 12 months for primary system decontamination; and a final 6 months tocomplete interim waste storage or waste disposal activities.
The large spread in the reactor
building decontamination estimates reflects the bounding alternatives of:(1) proceeding simultaneously with sump water cleanup and building decontamination (and using lower bounds for thetask duration estimates) or (2) deferring building decontamination until the sump water has been
removed (and using upper bounds for task duration estimates).
The large spread in the primarysystem decontamination estimates reflects the uncertainty in the condition of the primary systemand the extent to which fuel debtis has been distributed throughout the system.
10.2.3

Health Effects

The work force for the TMI-2 cleanup will be exposed predominantly to penetrating radiation
distributed over the whole body, so that any consequences will not be restricted to a particulararea or organ of the body.
A great deal of data on the biological (health) effects of radiation
has been accumulated on a worldwide basis over the past several decades.
These data have been
2
analyzed by international and national organizations responsible for radiation protection.1'
The up-to-date findings of these organizations are the basis for estimating radiation-related
human health effects in this document.
The occupational doses from routine operations during thecourse of the TMI-2 cleanup may result in somatic and genetic effects. The somatic effect (tothe body of the worker) of greatest concern is the possibility of inducing a fatal cancer; thegenetic effects include a variety of inheritable changes which may affect future generations.
The risk factors utilized in the estimates of health effects are:131 fatal cancers in the exposed workers per one million person-rem.
260 genetic effects among the offspring of the work force per one million person-rem.
More detailed information on the health effect risk estimators used by the staff is contained inAppendix Z.
It should be stressed that these risks, or probabilities, are increments above or additions tothose risks to which the entire population currently is exposed.
Current public health statistics show that for the entire U.S. population there is a 1 in 5 probability that death will be
due to some form of cancer.
The normal occurrence of hereditary disease in the offspring of thepresent U.S. population is about 1 in 17.
It is expected that the occupational dose to the workforce cleaning up TMI-2 will increase the workers' risk of death from cancer, but this added risk
is relatively small in comparison with the existing risk.
In addition, the risk of genetic
changes can be expected to increase for the offspring of the work force, but this increment isalso very small compared to existing risks.
The health effects from occupational exposure to radiation were calculated for the work force on
the basis of radiation doses ranging between 2000 and 8000 person-rem.
For the minimum cumulative dose case (2000 person-rem), it is expected that 0.3 additional fatal cancer would becaused.
For the maximum dose case (8000 person-rem), 1 additional cancer fatality would result.
Although it is possible to compute a range of probabilities for cancer induction among average
individual workers based on the above figures, the results of such a calculation may not bear aclose relationship to actual risks since the work force size and cumulative dose associated withthe various tasks can differ by large factors, rendering inapplicable the concept of an average
individual worker.
The licensee applies administrative controls for doses to its employees in order to ensure compliance with the regulations given in 10 CFR Part 20. These controls result in keeping most doses
to less than 1 rem per quarter.
Most of the workers involved in the cleanup can be expected to
be in this category.
10 CFR Part 20 regulations limit the highest quarterly dose that an individual worker may receive to 3 rem per quarter. Individuals are not allowed to receive exposures inexcess of I rem per quarter unless there are special circumstances.
For example, a complex taskthat would normally be done by a single worker might require several workers if the 1 rem perquarter administrative control were imposed.
In such situations, the total exposure to the workforce can often be reduced if one worker is allowed to exceed 1 rem per quarter (but not the
10 CFR Part 20 limits) in order to complete the task.

10-10For an individual worker who gets 1 rem per quarter throughout an assumed nine-year cleanup
period, the total dose would be 36 rem.
For a person of age 30 (the expected average age ofcleanup workers), the probability of dying of cancer would normally be 1 in 5. The added probability of a premature death from cancer as a result of receiving a radiation dose of 36 rem
would be 1 in 210.
Thus, for the decontamination workers, the overall probability of death fromcancer would be 1 in 4.9.
The equivalent decrease in life expectancy from a 36-rem dose would be
about 23 days.
The risk for a younger worker would be greater, and for an older worker it would
be less.
For the minimum cumulative dose case, the expected number of genetic effects among the offspringof the work force would be 0.5. For the maximum cumulative dose case, the expected number wouldbe 2.
The normal (exclusive of occupational dose) incidence rate for a hypothetical work forceof 1000 persons would be about 60.
10.3

OFFSITE DOSES AND HEALTH EFFECTS FROM NORMAL OPERATION

For estimating dose and health effects, the quantities of radioactive material that may bereleased from the plant as a result of cleanup alternatives are based on the description of thesource terms and the radioactive waste treatment system alternatives given in this statement.
Site-specific and environmental data provided during licensing in the environmental report and in
subsequent Metropolitan Edison answers to NRC staff questions relating to this PEIS were used
extensively in performing dose calculations.
Using estimated quantities of radioactive materials
anticipated to be released and exposure pathway information, the dose commitments to individuals
and to the population were estimated.
Population doses from atmospheric releases are based on
the projected population distribution within 50 miles in the year 2010 (3.2 million people).
Population doses from liquid releases for consumption of drinking water are conservatively basedon an assumed consumer population of 2.2 million people.
The calculational methods used for
estimating offsite doses given in this statement were those in Regulatory Guide 1.109.3 Since
the dose models in Regulatory Guide 1.109 assume a normally operating reactor with a 30-year
lifetime, certain minor modifications were necessary to account for the relatively shorter time
period of the decontamination program.
These modifications and site-specific parameters are
discussed in Appendix W.
The internal dose commitments in this statement represent the total dose received over a periodof 50 years following the intake of radioactivity for one year under the environmental conditions
existing during the cleanup period.
This calculational approach, 4 which is consistent with therecommendations of the ICRP, is described in detail in NUREG-0172.
10.3.1

Dose Commitments from Radioactive Releases to the AtmosphereRadioactive materials released to the atmosphere during the cleanup operation will result insmall radiation doses to individuals and populations.
NRC staff estimates of the expected gaseousand particulate releases (summarized in Table 10.1) and site meteorological considerations(discussed in Appendix W of this statement and summarized in Tables W-2 through W-6) were used toestimate radiation doses to individuals and populations.
The results of the calculations are
summarized below.
10.3.1.1

Radiation Dose Commitments to Individuals

The basis for selection of maximum individual receptor location (1.05 miles east of the site) andpathways considered for the maximum individual are described in Appendix W. The estimated dose
commitments to the maximum exposed individual from tritium and particulate releases at the selected
offsite location resulting in highest doses are listed in Table 10.6.
Pathways from which radiation dose at this location resulted include inhalation, ground shine, cow milk ingestion, andvegetable consumption.
10.3.1.2

Radiation Dose Commitments to Populations

The estimated radiation dose commitments to the population within a 50-mile radius of the Three
Mile Island nuclear plant from tritium and particulate releases to the atmosphere are shown in
Table 10.7. Annual natural background radiation doses to the population within 50 miles would be
370,000 person-rem.
The population dose commitments from atmospheric releases resulting from thecleanup operation represent an extremely small increase in the normal population dose due tobackground radiation sources.

10-11

Table 10.6.
Dose Estimates for the Maximum Exposed Individual for Normal
Radioactive Releases to the Atmosphere during Decommissioning
and Decontamination of the Facility

Document Section
or Table Number

Dose (mrem)a
Total-Body

Bone

Liver

1.0 x 10-3

1.0 x 10-3

1.0 x I0-3

5.2 x 10-3

2.3 x 10-2

5.3 x 10-3

1.2 x 10-3
3.4 x 10-1

1.5 x i0-3
1.5 x 10-3

1.5 x 10-3
3.4 x 10-1

Removal of RPV Headand Internals

2.2 x 10-10

2.2 x 10-1

Core Examination andDefueling

5.2 x 10-5

3.0 x I0-4

1.1 x 10-4

Decon. of Primary
System Components

1.3 x 10-10

1.3 x 10-1

Processing Accident
and Decon. Water

3.9 x 10-1

1.7

3.3 x 10-1

Option - Natural
Evaporation ofProcessed Waterc(1.3)b(0)(1.3)

Option - Forced
Evaporation of SDS/EPICOR II Processed
Waterc

(1.5)

(1.3)

(2.7)

Description

Decontamination
Section 4.5.2

Kr-85 Releases

Table 5.4

AFHB Decontamination
and Desludging
RB Decontamination
Lowest tritium estimate
Highest tritium estimate

Table 5.12
Table 5.13
Table 6.6
Table 6.8
Table 6.16
Table 7.14
Table 7.31

Table 7.32

Table 7.33

Option - Forced
Evaporation of
SDS Processed Waterc'f(120.)(480.)(11.)

Table 8.22

Processing Solid Wastes

4.3 x 10-3

1.9 x 10-2

2.9 x 10-3

Table 8.40

Processing Chemical
Decon. Solutions

3.7 x 10-6

1.6 x 10-5

3.8 x 10-6

Trash Compaction of
AFHB Wastes

2.6 x 10-4

1. 1 x I0-3

3.7 x 10-4

Trash Compaction of
RB Wastes

3.3 x 10-4

1.5 x I0-3

1.1 x 10-3

Trash Incineration

1.2 x 10-3

5.3 x 10-3

1.7 x 10-3

120.

480.

12.

0.8

1.8

0.7

Table 8.62
Table 8.63
Table 8.64
Decontamination
Totals

Options that restult in highest dosed'f
Options that restult in lowest dosee(continued)

10-12

Table 10.6.

Continued

Dose (mrem)aDocument Section
or Table Number

Description

Total-Body

Bone

Liver

U.14

DECON. activities

9.0 x 10-6

4.5 x 10-s

4.8 x 10-s

U.23

SAFSTOR activities

1.0 x 10-s

4.3 x 10-5

4.0 x 10-5

8.9 X 10-6

4.5 x 10-5

4.8 x 10-s

2.8 x 10-5

1.3 x 10-4

1.4 x 10-4

Decommissioning

U.35

ENTOMB activities

Decommissioning
Totals:aThe special location for which dose estimates are presented here is the nearest cow/
garden location.
The doses that are listed here are for the age group (adults, teenagers, children, infants) for which it was calculated to be highest. For this location,
children's doses were highest.
bValues in parentheses represent values for ocFour methods were evaluated for disposal ofprocessed water:
ship offsite, natural evaporation, forced evaporation, and release to river. Atmospheric releases result from thenatural and forced evaporation options only.
Values appear in this table for both.
dBased on forced evaporation option for disposal of processed water, SDS processing
assumed.
eBased on disposal of processed water by shipping offsite or release to river, and lower
tritium range.
fThis alternative results in doses which exceed the numerical criteria of 10 CFR Part 50,
Appendix I, with only one pass through the water treatment system.

10-13
Table 10.7.
Dose Estimates for the Projected Population in Year 2010 Residing
within 50-Mile Radius of TMI Resulting from Atmospheric Radioactive
Releases during Decontamination and Decommissioning of the Facilitya

Document Section

50-Mile Total-Body Population
Dose (person-rem)

Description

Decontamination
4.5.2
5.1.5.2

Maintain Reactor in Safe Condition

3 x 10-2

AFHB Decontamination and Desludging

2 x

5.2.5.2

Decontamination of Reactor Building

10-2

Lowest tritium range
Highest tritium range

2 x 10-2

6.3.5.2

Removal of RPV Head and Internals

4.

6.4.5.2
6.5.5.2
7.1.5.2
7.2.5.2

Core Examination and Defueling2 x 10-3

Decontamination of Primary System

3.

Processing of Water
Option - Natural Evaporation Proc. Waterc

2.
( 3 0)b

7.2.5.2

Option - Forced Evap.
Option - Forced Evap.

.(30)

7.2.5.2
8.1.5.2

6.

-

SDS/EPICOR II
SDSf

(400)
2 x 10-2

Processing Solid Wastes

8.2.5.2
8.3.5.2

Processing Chemical Decon.

Trash Compaction of AFHB Wastes

2 x 10-5
1 x 10-3

8.3.5.2
8.3.5.2

Trash Compaction of RB Wastes

2 x 10-3

Trash Incineration

6 x 10-3

Options resulting in highest dosed
Options resulting in lowest dosee

400.

U.3.4.2

DECON activities

6 x 10-5

U.4.6.2
U.5.4.2

SAFSTOR activities

7 x 10-5

ENTOMB activities

6 x 10-4

Decontamination
TotalsSolutions10.

Decommissioning

Decommissioning1 x 10-3
TotalsaAll values rounded to one significant figure. Pathways considered were external shine, groundshine, inhalation, meat consumption, vegetable consumption and milk consumption.
bValues in parentheses represent values for optional programs.
CFour methods were evaluated for disposal of processed water: ship offsite, natural evaporation, forced evaporation, and release to river. Atmospheric releases result from the naturaland forced evaporation options only. Values appear in this table for both.
dBased on forced evaporation option for disposing of processed water, modified SDS processing
assumed, and higher tritium range of Section 5.2.5.2.
eBased on disposal of processed water by shipping offsite or release to river, and lower tritium
range of Section 5.2.5.2.
fThis alternative results in doses which exceed the numerical criteria of 10 CFR Part 50,
Appendix I, with only one pass through the water treatment system.

10-1410.3.2

Dose Commitments from Radioactive Liquid Releases to the Hydrosphere

Liquid radioactive effluents released to the hydrosphere as a result of the Three Mile Islandcleanup operation have been estimated to result in small radiation doses to both individuals andthe population.
NRC staff estimates of the expected liquid releases and the site hydrologicalconsiderations (discussed in Appendix W of this statement and summarized in Table W.7) were used
to calculate anticipated radiation dose commitments to individuals and populations.
The resultsof these calculations are discussed below.
10.3.2.1

Radiation Dose Commitments to Individuals

The basis for selection of individual receptor locations where fish were caught, drinking water
was taken, and shoreline exposure was received is described in Appendix W. The estimated cumulative dose commitments to the maximum exposed individual from liquid releases at selected offsite
locations are listed in Table 10.8.

Table 10.8. Dose Estimates for the Maximum Exposed Individual
for Normal Radioactive Releases to the River
Resulting from Disposal of Processed Water

Document
Table No.
Table 7.29

Table 7.30

Dose (mrem)a
Description

Total-Body

Bone

Liver

Option - Release Processed
Water to River b
SDS/EPICOR II Processing

1.1

1.2

1.6

Option - Release Processed
Water to R ver SDS
Processing

9.8

23.

11.

7.0

16.

8.

Optiop Resulting in Highest
Doses e

9.8

23.

11.

Optiogs resulting in Lowest
Doses

1.5 x 10-3

Table 7.32

Option - Release of Blowdown
from Forced Evaporation ofProcessed Water - SDS/EPICOR II
Processing

Table 7.33

Option - Release of Blowdown
from Forced Evaporation ofProcessed Water - SDS
Processinge

Totals:2.1 x 10-3

2.3 x 10-3

aThe doses that are listed here are for the age group (adults, teenagers, children, infants)
calculated to receive highest dose. The total-body doses are for adults, the bone doses arefor children, and the liver doses are for teenagers.
bFour methods were evaluated for disposal of processed water:

ship offsite, forced evaporation,

natural evaporation, and release to river.
Liquid releases result from forced evaporation
method (blowdown) and from direct releases to the river. Values in this table appear for both.
cBased on direct release to river option, SDS processing assumed.
dBased on discharge of treated routine operational liquids (Sec.

4.5.2) and disposal of pro-

cessed water by shipping offsite or by natural evaporation.
eThis option results in doses that exceed the numerical criteria of 10 CFR Part 50,
with only one pass through the water treatment system.

Appendix I,

10-150.3.2.2

Radiation Dose Commitments to Populations

he estimated annual radiation dose commitments to an assumed population of 2.2 million drinkinglater consumers downstream of the Three Mile Island nuclear plant are shown in Table 10.9.
;ackground radiation doses to the same population would be 255,000 person-rem.
The dose commitients from liquid releases from the Three Mile Island cleanup operation represent an extremely;mall increase in the normal population dose due to background radiation sources.
-0.3.3

Transportation of Radioactive Material-he transportation of TMI waste from the reactor to burial grounds is within the scope of the NRC'eport entitled, "Environmental Survey of Transportation of Radioactive Materials to and from
Juclear Power Plants."'
The estimated population dose commitments associated with transportation)f fuels and wastes are listed in Table 10.10.

Table 10.9. Dose Estimates to the Population Residing Downstreamof TMI Resulting from Normal Radioactive Releases ofProcessed Water to the River
Document
Section
7.3.5.2
7.3.5.2
7.3.5.2

7.3.5.2

Totals:Description

Downstream Populatign
Dose (person-rem)

Option - Release Processsd Water to River SDS/EPICOR II Processing

30

Option - Releaae Processed Water to River SDS Processing

900

Option - Release of Blowdown from Forced
Evaporation of Processed Water SDS/EPICOR II Processing

7

Option - Release of Blowdown from Forced
Evaporation of Processed Water SDS Processinge

600

Options resulting in highest doseqce
Options resulting in lowest doses

900
0.02

aAll values rounded to one significant figure.

Pathways considered were drinking

water consumption and sport fish consumption.
bFour methods were evaluated for disposal of processed water:
ship offsite, naturalevaporation, forced evaporation, and release to river.
Liquid releases result fromthe forced evaporation method (blowdown) and from direct releases to the river.
Values in this table appear for both.
Based on the option of direct release to river, SDS processing assumed.
dBased on discharge of treated routine operationalliquids (Section 4.5.2) and disposalof processed water by shipping offsite or by natural evaporation.
eThis option results in doses that exceed the numerical criteria of 10 CFR Part 50,
Appendix I, with only one pass through the water treatment system.

10-16

Dose Estimates to the Population Residing Along Transport Routes
Table 10.10.
Enroute to Waste Burial Grounds - Decontamination and Decommissioning
Document Section
or Table No.

Population Doge
(person-rem)

Description

Decontamination:
Section 9.5.1.2

Section 9.5.1.2
Decommissioning:
Section 2.2
Totals:

Dose to the public residing alongtransportation routes:
Longest route
Shortest route20-50b

Dose to onlookers while vehicle is stoppedc

5-10

Dose to public residing along transportationroutes (longest route)

100

Range

2-7

100-200

aAll estimates rounded off to one significant figure.
bRange reflects different assumptions on length of route:the longest route estimate

is based on 700,000 persons residing along the 2750 mile route and calculated for the
persons in an area between 100 ft and ½ mile on either side of the route, assuming
330 persons per square mile, 10 mrem/hr dose at 6 ft from the vehicle and each shipment traveling 200 miles per day.
The shortest route is based on the same assumptions as above except that 125,000 persons are assumed to reside along a 370-mileroute.
cAssumes an individual spends 3 minutes at an average of 3 ft from the loaded truck
and that 10 persons are exposed during each shipment.

10.3.4

Radiological Impact on Man

Based on the NRC staff's evaluation of the potential performance of cleanup alternatives, it isconcluded that the cleanup can be accomplished within the dose design objectives of 10 CFR Part 50,
Appendix I.*
In Table 10.11, the calculated maximum individual doses for the entire cleanup
program for each option are compared to the dose design objectives for atmospheric and liquid
releases.
The estimated doses in Table 10.11 represent the total dose commitment over the entire
decontamination program, which will last several years, whereas the Appendix I values in thetable are annual dose values.
Upon comparing the estimated values in Table 10.11 to the design
objective values, all options, except for forced evaporation of the SDS processed water anddirect release of water that has received only one pass through the SDS system to the river, areestimated to meet the annual Appendix I design criteria and 10 CFR Part 20.
In Appendix I dose design objectives are defined as the as low as reasonably achievable (ALARA)
requirements for normally operating reactors and were not intended to apply to the post-accident
Nevertheless, these objectives do serve as a basis for comparing the magnitudesituation at TMI-2.
of potential cleanup releases to the potential radiological environmental impact of a normallyoperating reactor.
Furthermore, as described in Appendix R, the staff proposed that the dose
design objectives be applied as technical specification limits for the purpose of the deconEven if the offsite doses are as high as the Appendix I design objectives,tamination operation.
the maximum individual doses will still be relatively small when compared with either natural
background doses ("116 mrem/yr) or of the dose limits specified in 10 CFR Part 20 (500 mrem/yr).
As a result, the staff has concluded that there will be no significant radiological impact on manfrom the decontamination operation.

10-17
Table 10.11.
Dose Estimates for the Maximum Exposed Individualfor Entire Decontamination Program for Each
Processed Accident Water Disposition Alternative, andComparison with 10 CFR 50 Requirements
Dose (mrem)a(Total-body/max.
Optionborgan)

Estimated
Value

Appendix I,
10 CFR 50 Value

Atmospheric Releases:
1.

Ship processed water offsite

0.8/1.8-/15

2.

Natural

Evaporation to Atmosphere2.1/2,0-/15

3.

Forced Evaporation to Atmosphere SDS/EPICOR II Processing

2.3/3,4-/15

4.

Forced Evaporation to Atmosphere
SDS Processing

120./480.

-/15

5.

Release Water to River - SDS/EPICOR II Processing

0.8/1.8-/15

Release Water to RiverSDS Processing

0.8/1.8-/15

6.

Liquid Releases:
1.

Ship processed water offsite.0015/.0023

3/10

2.

Natural Evaporation to Atmosphere.0015/.0023

3/10

3.

Forced Evaporation to Atmosphere
SDS/EPICOR II Processing

0.8/1.1

3/10

Forced Evaporation to Atmosphere
SDS Processing

7.0/16.

3/10

Release Water to River - SDS
EPICOR II Processing

1.1/1.6

3/10

Release Water to RiverSDS Processing

9.8/23.

3/10

4.
5.
6.

aThe dose estimates represent the contribution of all decontamination programsfor the entire cleanup.
bDose estimates are listed separately for atmospheric releases and liquid
releases, rather than adding them together, because different individuals are
generally considered to be involved, and because the 10 CFR 50 dose objectives are different for each.
cA small fraction of water vapor may precipitate as rain over the Chesapeake
Bay watershed.

10-18
Radiological doses to the general public from normal operation of the decontamination program may
result in:a.

Late somatic effects in the form of fatal and non-fatal cancer in various body organs,following age- and organ-specific latency periods within the exposed population, andb.

Fatal and non-fatal genetic disorders in future generations of the exposed population.

Estimates of these health effects, which could occur randomly in an exposed population, arenormally based on estimates of cumulative population dose expressed as person-rems (average dose
X number of people receiving dose).
Population health effect estimates presented in this statement reflect the total effect incurred by the population from all cleanup activities at TMI.
In
order to quantify individual risks, calculations are also made here for the maximum exposed
individual.
Absolute risk estimators of 135** deaths from latent cancer per 106 total-bodyperson-rems in the exposed population and 258 cases of genetic disorders per 106 total-bodyperson-rems in the future generations of the exposed population were derived from the 1972 BEIR
report and the Reactor Safety Study (WASH-1400, October 1975).
This derivation assumes alinear, nonthreshold dose-effect relationship at all sublethal dose levels.t Total body cancer
risk estimators are used because they are larger than those for bone or liver.tt
Using the above risk estimators for cancer deaths and genetic disorders, health risks as a resultof releases from TMI were calculated for the population residing around TMI and for the maximum
exposed individual.
Health risks also were calculated for the population residing along thetransport route to waste disposal grounds.
The results of these calculations are described
below.
Table 10.12 lists the expected number of cancer deaths or genetic abnormalities, designated as
rates, for the 50-mile population of 2.2 million people around TMI as a result of decontamination
activities.
As these values are much less than 1 they suggest that it is very unlikely that a
future cancer death occurs in the exposed population over the remaining lifetime of the population, or that a genetic abnormality occurs in the next 5 generations of the exposed populationthat could be associated with the clean-up operation.
A better appreciation of the meaning of
the numeric values in Table 10.12 can be gleaned by comparison of the rates of Table 10.12 to
expected cancer death rates and genetic abnormality rates from causes other than TMI releases inthe same population.
For example, in 1976, about 20 percent of all deaths in the United States
were due to cancer. 6
If those statistics are applied to the 2.2 million people living within
50-miles of TMI, the number of people in this population expected to die of cancer is 440,000.
This number can be directly compared to the numeric rates in Table 10.12.
This comparison
indicates that the incremental chance of fatal cancer to the population or to an average individual in the population due to the decontamination activities is in the range of 1 chance in 400
million to 1 chance in 4 million, depending on the decontamination option that is chosen.
The same type of estimates for the population residing along the transport route (700,000 people)
are listed in Table 10.13.
A similar comparison of the numeric rates in Table 10.13 to thenumber of individuals in the population along the route who are'expected to die of cancer (20 percent of 700,000 = 140,000) indicates that the incremental chance of fatal cancer to the population along the route or to an average individual along the route is 1 chance in 20 million due toshipments of decontamination wastes.
BEIR-I describes the normal incidence of diseases in which there is some evidence as being associated with a genetic component abnormality as 6 in 100, or 6 percent (p. 57, Table 4, of Ref.
2).
In a similar fashion as was done for cancer death rates, the genetic abnormality rate due to*Appendix R describes proposed modifications to the plant's operating technical specification
which will impose the 10 CFR Part 50 Appendix I design objectives as operational limits.
**This value, 135, is slightly greater than the risk estimator (131) used for occupational dose
effects because of age distribution differences in the two populations.
The occupational
population excludes individuals less than 18 years of age.
tDetails of the derivation of radiation-induced health impacts on man are provided in Appendix Z.
ttThe risk estimates are 7 deaths and less than 22 deaths per 106 person-rem for bone and liver,respectively.

10-19

Table 10.12.
Estimates of Cancer Death Rate and Genetic
Abnormality Rate for Exposure of the Population
around TMI due to Releases from Decontamination fSr
the Range of Processing and Disposal Alternatives

Rateb(deaths or abnormalities
per 2.2 million people)
Estimated
Dose(person-rem
total body)

Cancer Fatality overRemaining
Lifetime

Ship processed water
offsite-3

3 x 10- 3

10

1 x 10Forced evaporationof - SDS
processed water1000

1 X 10-1

3 x 10-1

Decontamination
OptionGenetic
Abnormality
over Next 5
Generations

aAll estimates rounded off to one significant figure.
bPopulation death or abnormality rate estimates are expected

to beslightly smaller than the values presented here because the year
2010 population estimate of 3.2 million people was used for dose
estimates from atmospheric pathways.

Table 10.13.
Estimate of Cancer Death Rate and Genetic
Abnormality Rate for Exposure of the Population
Residing along Transport Routes

Ratea(deaths or abnormality
per 700,000 people)
Range ofDoseb(person-rem)CancerFatality
over Remaining Lifetime

Genetric
Abnormality
over Next 5
Generations

Decontamination

20-50

7 x 10-3

1 x 10-2

Decommissioning100

1 x 10-2

3 x 10-2

Activity

aRate estimates based on upper limit of range.
bAll values in table have been rounded off to one significant
figure.

10-20the decontamination and decommissioning activities at TMI can be compared to the incidence ofdiseases related to genetic abnormalities from causes other than TMI releases.
For the 2.2 million
people in the 50-mile radius around TMI the expected incidence of non-TMI related genetic abnormalities is 132,000 (0.06 x 2.2 x 106 = 132,000).
Comparing this to the rates of Table 10.12
indicates that the incremental chance of a genetic abnormalities to descendants of this populationor to an average exposed individual in it over the next five generations ranges from 1 chance in
40 million to 1 chance in 400 thousand, depending upon the decontamination options which areused.
In a similar fashion, for the 700,000 people residing along the transportation route it is estimated that the expected incidence of diseases related to genetic abnormalities from causes otherthan transport of TMI waste is 42,000 (0.06 x 700,000 = 42,000).
Comparing this to the rates ofTable 10.13 indicates that the incremental chance of a genetic abnormality in descendants of the
population or in descendents of an average exposed individual over the next five generations is
1 chance in 4 million due to radiation exposure along the transport route.
The total-body dose to the maximum exposed individual near TMI for the option involving shipping
processed water offsite was estimated to be 0.8 mrem.
The maximum individual dose is about 600
times larger than the average individual 50-mile population dose for the-option involving shippingof water offsite. The incremental chance of cancer death to the maximum exposed individual
was estimated to be 1 chance in 2 million, and the incremental chance of a genetic abnormality to
the descendants of the maximum exposed individual in the next five generations was 1 chance in
300,000.
If some other option is used for processing and disposal of the water the resulting
maximum individual dose will be limited to the requirements of 10 CFR Part 50, Appendix I, on anannual basis, which would result in an increased chance of cancer death of 1 chance in 100,000,
and an increased chance of a genetic abnormality in the next five generations of 1 chance in 20,000.

On the basis of the small likelihood of increased incidence of cancer deaths or genetic abnormalities from decontamination and decommissioning activites, the staff has concluded that therisk to the public health and safety will be insignificant.
10.4

DOSES FROM POSTULATED ACCIDENTS

10.4.1

Dose Commitments from Accidental Radioactive Releases to the Atmosphere

The quantities of radioactive material that may be released from postulated accidents are derived
based on the systems described in this document and assumed failure scenarios.
The licensee'ssite and environmental data provided in the environmental report and in subsequent answers to NRC
staff questions relating to this PEIS are used extensively in the dose calculations.
These
quantities of radioactive materials released and the exposure pathway information are used toestimate the dose commitments to individuals.
The calculational methods that were used throughout
this statement for estimating offsite doses were those of Regulatory Guide 1.109.3 Since Regulatory Guide 1.109 is for a normal operating reactor, certain minor modifications were made to take
into account the short time period of accidental release.
These modifications are discussed inAppendix W.
The dose commitments from accidents postulated in this statement represent the total dose received
over a period of 50 years following the intake of radioactivity for one year under the environmental conditions existing during the cleanup period.
For the younger age groups, changes inorgan mass with age after the initial intake of radioactivity are accounted for in a stepwisemanner.
10.4.1.1

HEPA Filter Failure during Decontamination Operations

During decontamination operations, dust and radioactive particles can become airborne and enter
the building ventilation systems.
To minimize releases of radioactive particles to the environment, the building air is drawn through the plant ventilation system.
The ventilation exhaustsystem consists of a prefilter, HEPA filter, adsorber (charcoal), and final HEPA filter.
The
prefilter provides a measure of protection for the HEPA filters installed in the exhaust ventilation system.
The HEPA filters are highly efficient for all particle sizes.
A minimum efficiencyof 99.97 percent for installed HEPA filters is based on the most penetrating particles (0.3 mm).
The first HEPA filter in the system collects most of the radioactive material.
If the first10-21
HEPA filter failed, the second filter would collect most of the materials released from the first
filter.
The dose estimates summarized in Table 10.14 are based on the assumption that both
For most operations, even if both filters failed simultaneously,
filters fail simultaneously.
doses would be within decontamination operation requirements (Appendix R).
If only the first
HEPA filter fails, the expected dose would be much smaller than those listed in the table andwould be within regulatory criteria of 10 CFR Part 50, Appendix I, for normal operating reactors.
Even if both filters failed, the resulting offsite doses for all accidents would be withinrequirements of 10 CFR Part 20.

Table 10.14. Dose Estimates for the Maximum Exposed Individual
Caused by a HEPA Filter Failure during Specific
Decontamination Operations
Decontamination
Operation

Table
Number
5.5
5.6

Surface Decontamination
Operation in AFHB
Desludging Operationsin AFHB

Dose (mrem)a
Total-Body

Bone

Liver

4.2 x 10-3

1.8 x 10-2

1.4 x 10-2

2.0 x 10-1

8.5 x 10-1

1.9 x 10-1

Decontamination of Reactor
Building

1.3 x 10-2

5.9 x 10-2

2.2 x 10-2

Processing Reactor Building
Sump Water

3.3

16.

12.

7.16

Processing Reactor Building
Decontamination Water

5.2 x 10-4

2.3 x 10-3

1.8 x 10-3

7.17

Processing Chemical Decontamination Water

9.7 x 10-4

4.2 x 10-3

3.2 x 10-3

Processing RCS Accident
Water

1.5

6.0

2.8 x 10-1

7.19

Processing RCS Flush andDrain Water

7.5

30.

1.3

7.20

Processing RCS Decontamination
Water and RCS Chemical Decontamination Water

1.5

6.2

2.8 x 10-1

5.14
7.15

7.18

aDoses were calculated for total-body, GI-tract, bone,and skin. The
calculated for
dose estimatesdose estimates

liver,

kidney,thyroid,lungmaximum three organ doses are listed in this table.
Doses werefour age groups; adults, teenagers, children, and infants.
The
are for the nearest cow/garden location (1.05 miles east).
The
presented in this table are for children.

For the following reasons, the probability of both HEPA filters failing simultaneously is verysmall.
The most likely cause of HEPA filter failure would be overpressurization.
During operation of.the system, overpressurization can occur due to heavy buildup of particulate material on
the filter. However, each HEPA filter in the ventilation exhaust system is provided with pressure
indication.
High pressure differential across a filter will set off local and remote alarms
which would alert the operators to take action, such as shutting the system down.
Even if no action were taken to alleviate a HEPA filter overpressurization and the filter failed,
the particulate matter that would come off the failed filter would collect on the second downFor there to be a significant release to
stream HEPA with a removal efficiency of 99.97 percent.
the environment, both HEPA filters in the ventilation system would have to fail simultaneously.

10-22Since a failure of a single filter in the series will not result in doses in excess of regulatory
criteria for normal operating reactors (10 CFR Part 50, Appendix I) or in excess of those describedin Appendix R of this statement, and since the chance of a multiple failure is highly unlikely,
the staff concludes that the HEPA filter failure accident scenario does not pose a significant
level of risk to the public health and safety.
10.4.1.2

Contaminated Material

Fire

Most of the contaminated material onsite is water, sludge, ion-exchange media, and building andequipment surface materials that will not support combustion.
However, there are large quantities of rags, blotter paper, plastic sheets, etc., used for decontamination activities. Whenthese materials become contaminated, they are usually collected in plastic bags, compacted insteel drums, and placed in temporary storage prior to offsite shipment for disposal.
There arealso significant quantities of non-compactible contaminated materials such as lumber, tools, pumpcasings, etc., which are collected in wooden boxes and placed in temporary storage prior to
offsite shipment for disposal.
A fire in the storage area containing packaged drums and boxes is
considered a credible accident. An analysis of fires in barrels of contaminated trash in a fuelreprocessing plant indicates that the fraction released from such fires is 8 x 10-7 of the total
activity present.
Using this model, the largest release analyzed (Section 8.3.4.2) is approximately 3 x 10-8 pCi with the radionuclide distribution listed in Table 8.53. The estimated dose
resulting from this release to the maximum exposed individual is 6.5 mrem for total-body, 29 mrem
for bone, and 8.5 mrem for liver.
This accident involving a fire in a low level waste storage
area does not result in offsite doses in excess of the requirements of 10 CFR Part 20.
On thisbasis, the staff concludes that this type of accident does not pose a significant risk to the
public health and safety.
10.4.1.3

Breach of a Waste-Containing Package

The staff has evaluated the consequences of breaching a package containing spent ion exchange
media, spent filters, compactible trash, noncompactible trash, mirror insulation, or incinerator
ash from a postulated drop of the package and its contents.
The dropping of a waste package andthe resulting breach and release of a portion of the contents of the package, is considered acredible accident.
The estimated doses to the maximum exposed individual resulting from thistype of accident for a variety of waste-containing packages are listed in Table 10.15. With the
exception of the accident involving the postulated drop of a spent liner generated from theprocessing of primary system water through modified EPICOR II, none of the breach accidents
listed in Table 10.15 results in offsite doses in excess of the requirements of 10 CFR Part 20,
and only a few are above normal requirements described in Appendix R. In order to mitigate theconsequences of a postulated drop of a modified EPICOR II zeolite liner used for processing
primary system water, the licensee will be required to either administratively control the curie
inventory on the zeolite liner or design and test, prior to actual use, a liner which will becapable of withstanding, without breaching, the worst-case accident (i.e., the highest drop onto
an unyeilding surface) in the event modified EPICOR II is approved for processing primary systemwater.
On this basis, the staff concludes that none of the breaching accidents listed inTable 10.15 poses a significant risk to the public health and safety.
10.4.1.4

Vehicle Accident Along Transport RouteTo arrive at a realistic "worst-case" scenario for airborne releases resulting from a truck
transport accident, the staff assumed that a Type B container would break preceding a fire of the
package.
Data from other sources suggest that a release fraction of 10-s could be expected fromsuch an accident.
The total-body dose from such an accident under assumed meteorological conditions was estimated to be 100 mrem.
The estimated dose indicates that this accident will not
exceed the requirements of 10 CFR Part 20 and probably not exceed those of 10 CFR Part 50, Appendix I, as the assumed meteorological conditions are considered upper bound conditions.
On thebasis of these results, the staff concluded that this accident scenario will not result in significant impact to the environment or risk to the public.
10.4.1.5

Spill of Reactor Coolant System Liquid in Reactor Building

The staff has evaluated the consequences of spilling reactor coolant system (RCS) water in thereactor building while the primary system pumps are operating.
For the purpose of estimatingconsequences, 10 percent of the liquid was assumed to spill in the reactor building (2000 Ci).
About 0.1 percent of this would become airborne in the reactor building atmosphere and about
0.1 percent of the airborne fraction could be expected to pass through the HEPA filters. The
resulting dose to the maximum exposed offsite individual was estimated to be 1.5 mrem to the10-23
These values are well below thetotal body, 6 mrem to the bone, and 0.26 mrem to the liver.
On thisrequirements of 10 CFR Part 50, Appendix I, for normal operation and of 10 CFR Part 20.
basis, the staff concludes that this type of accident does not pose a significant risk to the
public health and safety.
10.4.2

Dose Commitments from Accidental

Radioactive Releases to the Hydrosphere

Accidental radioactive effluent releases to the Susquehanna River that may occur during theSite hydrological considercleanup operation are described in Section 7.2 in this statement.
ation for accidental releases are similar to those for normal decontamination releases which arepresented in Appendix W. The differences are discussed in the sections of this document where
the actual calculations are presented.
The results of the postulated accident are summarized inTable 10.16.
The type of accident that was considered in this document involves the failure of aprocessed water storage tank and subsequent leakage of its contents into the east channel of theSusquehanna River.
This accident and its effects are discussed below.

Table 10.15.
Dose Estimates for the Maximum Exposed Individual
Caused by Breaching a Package Containing Radioactive Waste

Table
Number

Dose (mrem)
Total-Body

Bone

Liver

8.24

SDS-Filter Assembly

8.1 x 10-4
5.7 x 10-3

3.7 x 10-3
2.2 x 10-2

2.8 x 10-3
3.1 x 10-4

8.25

SDS Cation Liner

1.6 x 10-3

6.4 x 10-3

1.5 x 10-6

8.26

SDS Mixed-Bed Liner

1.2 x 10-1

4.6 x 10-1

1.5 x 10-4

8.27

EPICOR II-AFHB; Prefilter Liner

16.

76.

88.

8.28

EPICOR II-AFHB; Cation Liner

1.6

6.3

6.5

8.29

EPICOR-II-AFHB; Mixed-Bed Liner

3.9 x 10-2

1.7 x 10-1

2.0 x 10-1

8.30

Modified EPICOR Il-Primary Water;
Zeolite Liner

700

2900

790

Modified EPICOR lI-Primary Water;
Cation Liner

3.7

15.

4.6 x 10-3

Modified EPICOR-II Primary Water,
Mixed-Bed Liner

1.7 x 10-1

6.7 x 10-1

2.0 x 10-4

AFHB & Reactor Building
Chemical Decontamination

6.8 x 10-5

3.0 x 10-4

2.3 x 10-4

8.66

Compactible Trash

3.0 x 10-2

1.3 x 10-1

4.5 x 10-2

8.67

Noncompactible Trash

1.6 x 10-2

6.6 x 10-2

2.2 x 10-2

8.68

Mirror Insulation Trash

4.6 x 10-2

2.0 x 10-1

6.5 x 10-2

8.69

Compactible Ash

3.0 x 10-3

1.3 x 10-2

4.5 x 10-3

8.23

Type of Waste
SDS-Reactor Building Sump
Water Zeolite

8.31
8.32
8.41

10-24

Table 10.16. Dose Estimates for the Maximum
Exposed Individual Caused by Breachinga Processed Water Storage Tank
and Releasing Contents into theEast Channel of the Susquehanna River

Table
Number

Processing
OptionDose (mrem)
Total-Body

Bone

Liver

SDS

0.95

2.2

1.1

SDS/EPICOR II

56/kg fishconsumed--High River Flowb:
7.34
Low River Flow:
7.35
7.35

SOS

470/kg fishconsumedaTotal-body dose estimates are for adults, bone dose
estimates are for children, and liver dose estimates
are for teenagers.
bl"High river flow" for the purposes of this table isdefined as that river flow which causes overtopping of
Red Hill Dam.
"Low river flow" is defined as thatriver flow which does not cause Red Hill Dam
to overtop.

10.4.2.1

Failure of Processed Water Storage Tank

During water processing operations, processed water will be temporarily stored in two holding
tanks located outdoors, each of 500,000 gallon capacity.
The water in the holding tanks willthen be disposed of by one of the methods described in Section 7.2.
If one of these tanks ruptured and its entire contents were released, storm drains would transport the water to the east
channel of the river. The potential offsite dose to humans from this accident is highly dependent
upon whether or not Red Hill Dam is overtopping.
If Red Hill Dam is overtopping, the released
water will be diluted with the flow of the Susquehanna River resulting in doses to humans thatare fairly low and within annual limits for routine operation (see Appendix R).
However, if Red
Hill Dam is not overtopping, the released radioactivity could remain in the east channel for an
extended period of time at fairly high concentrations.
Dose calculations are presented inTable 10.16 for both river flow situations.
For the high river flow situation, the resulting offsite doses are estimated to be below therequirements of 10 CFR Part 20, and below the dose design objectives for normal operating reactorsof 10 CFR Part 50, Appendix I.
Thus, the staff concluded that if this postulated accident wereto occur during high river flow, the resultant environmental impact would be insignificant.
For the low river flow situation, the resulting offsite doses due to consumption of drinking
water or fish from the east channel would be large enough to warrant that action be taken to
avoid such consumption.
Consequently, the staff recommends that mitigative action be taken to
avoid consumption of fish or drinking water from the east channel if the accident occurs during
low river flow.
Since there is no municipal use or known private use of water from the east
channel for consumption purposes, doses are not expected to occur through the drinking water

10-25The main concern is to prohibit the catching and consumption of fish from the channel.
pathway.
As the bioaccumulation of radionuclides in fish occurs over periods of days to weeks, the staff
concluded that there would be ample time to take preventive measures to ensure that fishing is
stopped in the east channel area.
It is conceivable that a fish could reside in the channel for a long period, bioaccumulate radioDepending upon the water
nuclides, and then move to some other area and be caught and consumed.
processing option, 1 to 10 kilograms (kg) of fish would need to be consumed before the 10 CFR
Assuming that the average weight of a fish harvested
Part 20 protective dose limits are exceeded.
from the river near TMI is about 0.5 kg of whole body weight (thus yielding < 0.2 kg of ediblemeat), an angler would have to harvest between 5 and 50 fish to obtain 1 to 10 kg of edible fishmeat.
Six years of studies have shown that the mean harvest from the York Haven Pond during theIt seems
summer-fall months is less than one fish per angler (or per fishing trip by an angler).
unlikely, therefore, that any given angler would harvest enough fish (all of which had resided in
the east river channel following a tank rupture) to permit consumption of enough meat to resultin a dose that exceeds the protective limits. Additionally, studies of the post-accident (1979)
river fishery showed that anglers released their catches in greater than normal proportions andate fewer fish due to their concerns that the fish might have been radioactively contaminated by
the accident.
Similar angler behavior could be expected following a tank rupture, with adequate
public notification, thus reducing the likelihood of any anglers receiving unacceptable doses
from consuming river fishes.
Hence, with proper mitigative action, the public health and safety will be protected in the event
Such mitigative actions could include fishing advisoof an accident during low flow conditions.
ries or consumption bans; or physically blocking the movements of fish into and out of the shallow
east river channel by placing a fine-mesh net across the channel near Sand Beach Island or thenorth access bridge.
10.4.2.2

Leakage of Reactor Building Sump Water

The largest amount of contaminated water presently on the site is the 700,000 gallons of water inthe bottom of the reactor building.
This water contains an estimated 500,000 Ci of radionuclides.
It is postulated that if this water should leak through the thick steel-lined concrete base of
In Appenthe reactor building into the ground, it would ultimately reach the Susquehanna River.
For purposes of this analysis,dix V the movement of this water through the ground was analyzed.
it was assumed that the volume of water above the water table could leak into the groundwater in
one to two days.
It was calculated that the water would not begin to reach the river for about
one year, and while the water was in the ground, processes of adsorption, filtration, and ion
exchange would remove a large fraction of the dissolved and particulate radionuclides except H-3.
Dilution of the water reaching the river would further reduce the concentration of radionuclides
so that the peak concentrations of Cs-137 in the river would be 5.1 x 10-1o pCi/mL, of.Sr-90
These values are orders
would be 5.1 x 10-8 pCi/mL, and that of H-3 would be 5.2 x I0-7 pCi/mL.
of magnitude below the MPC limits of 10 CFR Part 20.
Monitoring wells around the reactor building
would provide indication of the increase in radionuclides in the groundwater long before theywould reach the river; the remaining water could be transferred to storage tanks or other mitigating measures when a grout curtain could be instituted.
10.4.2.3

Accidents Associated with Dropping Heavy Components on the Vessel Seal Ring

At this stage of the operation, a drop accident that caused a failure of the seal plate would
result in drainage of the transfer canal to the reactor building basement, but would have noradiosignificant radiological consequences since the canal water would contain very little
activity (0.01 pCi/mL, exclusive of tritium), and there would be no fuel in the transfer canal.
10.5
10.5.1POTENTIAL RELEASES DUE TO EXTERNAL EVENTS
Potential Releases due to Flooding

An evaluation has been made to determine the potential impact of Susquehanna River floods on the
nuclear waste storage facilities at Three Mile Island. Of primary concern are the interim storage facility (which is to be decommissioned shortly) and the concrete storage facility located in
These facilities were designed so as not to be affected by thethe Unit 2 desilting basin.
A descripdesign basis flood, but both would be inundated by the probable maximum flood (PMF).
The design of the faciltion of the design basis flood and PMF is provided in Section 10.5.1.2.
ities has been reviewed to evaluate the potential for release of radioactivity to the environment
during a PMF.

10-2610.5.1.1

Temporary Radwaste Facility

The temporary radwaste storage facility was designed to provide shielded storage for 28 spent
EPICOR I and II liners (16 cells 4.5 ft in diameter by 8 ft high, and 12 cells 7 ft in diameterby 8 ft high).
The cells are galvanized, corrugated metal cylinders that have steel plates
welded to one end to provide a bottom.
The cells are placed on compacted earth fill in the
Unit 2 desilting basin and backfilled with compacted earth. Each cell is provided with a 16-ton
concrete shield cover (Appendix D).
The elevation of the top of the cells is below the flood
control dike elevation of 304 ft Maximum Sea Level (MSL) at the downstream side of the island.
Any flood level exceeding the elevation of the dike would result in water levels greater than the
top of the cells, but the cells are sealed against leakage by the welded base plate and themassive, sealed concrete top.
10.5.1.2

Interim Radwaste Storage Facility

The interim radwaste storage facility is a modular structure, each module consisting of a 57 x
91 x 19-ft-high, reinforced-concrete box with 3-ft-thick concrete base and 4-ft-thick walls(Appendix D).
The top of the walls is at an elevation of 305.33 ft MSL.
Each cell is 7 ft indiameter by 13 ft high.
The module cells are made of galvanized, corrugated steel cylinders withwelded steel base plates.
The cells are encased in concrete for shielding and have a 3-ft-thickconcrete cap with a seal.
Each cell is equipped with a drain on the base plate to allow collection of washdown liquid into
the facility sump.
The sump is equipped with a high-level alarm and must be manually set up fortransfer of fluid.
The sump pump discharge is capped, and sump water analysis is to be done
before any uncapping and transfer of liquid. The sump and sump pump are contained in a concrete
structure having two massive, gasketed covers--one 3-ft-thick concrete and one 1-ft-thick concrete.
The concrete cells, their covers, and the sump and its covers create a barrier to release ofradionuclides.
10.5.1.3

Potential Impacts of Floods

The Three Mile Island Nuclear Station site is protected by a rock-armored levee system thatcompletely surrounds the station structures.
The levees range in elevation from 304 ft MSL at
the southern (downstream) end to about 310 ft MSL at the northern (upstream) end.
The levees
were designed and constructed to protect the plant from a flood having a peak discharge or runoff
rate of 1,100,000 cubic feet per second (cfs).
The height includes a margin of safety to allow
for settlement, wind wave activity, and larger floods.
The design of safety-related plant facilities inside the levee system will accommodate a PMF having a peak discharge rate of about
1,600,000 cfs.
The Hurricane Agnes flood of June 1972 had a peak discharge of 1,020,000 cfs(less than the levee design flood).
Although portions of the plant site were flooded during
Agnes, construction of the levee system had not been completed at that time, and flood waters
inundated portions of the island through unfinished sections of the levee.
Had the levees beenin place, no flooding would have occurred inside the levee system, as indicated in Table 10.17,
in which levee and flood elevations at Three Mile Island are summarized.

Table 10.17.

Summary of Levee and Flood Elevations at Three Mile Island
Elevations (ft MSL)
Probable

Levee
Elevation

1936
aFlood Level

Agnesb
Flood Level

Design c
Flood Level

Maximum
Flood Leveld

South end of levee

304

296.4

299.4

303.0

308.5

Unit 2 intake

305

296.5

300.4

303.5

309.2

North end of levee

310

298.0

301.3

304.5

-309.8Location

apeak discharge of 740,000 cfs--flood of record prior to 1972.
bPeak discharge of 1,020,000 cfs--flood of record.
CPeak discharge of 1,100,000 cfs--minimum freeboard is about one foot at south end.
dPeak discharge of 1,600,000 cfs--levee is overtopped at downstream end.

10-27
The PMF is derived in a very conservative manner and represents the upper limit of flood potential
at a given location, such that there is virtually no chance of the PMF being exceeded.
Based onflow frequency data developed by the NRC staff, the Agnes flood of 1972 has a probability ofoccurrence of less than once in 1000 years; the design flood has a recurrence interval of lessthan once in 2000 years; and the PMF has an undefined probability, but the occurrence of such aflood~is less likely than the design flood.
However, if the worst-case flood (PMF) were tooccur, the flood level would be about 308 ft MSL.
In this case, the top of interim storage
facility cells and the concrete storage facility walls, having respective elevations of 304.0 and
305.33 ft MSL, would be flooded.
The prefilter material, resins, and evaporator bottoms are contained, however, in sealed steel
containers, which are stored in the cells of the storage facilities. The cells provide a second
barrier to release of radioactive material.
In order for radioactive waste to escape, bothbarriers would have to develop an undetected leak path coincident with the PMF.
In view of the very low likelihood of the occurrence and short duration of any flood that would
overtop the protective dikes and the low likelihood of simultaneous breaching of the double
containment afforded by the storage facilities, it is considered highly unlikely that flooding
could cause release of radioactive material.
10.5.2

Potential Releases Due to Tornado

The principal factors that must be considered in evaluating the potential impact of a tornado are
the high winds, the pressure drop in the center of a tornado, and the danger of tornado-generated
missiles.
The design basis tornado for the TMI-2 site has wind velocities up to 360 mph, a pressure drop of 3 psi in 3 seconds, and potential tornado missiles as given in Table 10.18. The
facilities and procedures at TMI-2 have been reviewed with these potential factors in mind, andhave been found to be adequate to protect the environment and the health and safety of the publicfrom potential effects of tornados.

Table 10.18. Significant Design Basis
Tornado Missiles

Impact10.5.2.1

MissileWeight(lb)

Impact(ft 2 )

Velocity(mph)

Utility pole

1200

1.0

200

Passenger auto

4000

30

100

Concrete fragment

4500

30

60

Wood plank

110

0.33

360

Tornado Probabilities

The probability of a particular point being affected by a tornado is a function of the number of
tornados occurring, on the average, in the vicinity and the average path area affected by each.
For a 40-year period of record, an average of one tornado per year was reported for the two-degreesquare (8000 square miles) containing the TMI-2 site.
From 1955 through 1967 an average of one
tornado per year in the one-degree square containing the site (4000 square miles) was reported.
This higher apparent frequency is due to more complete observation and reporting and is used forthese evaluations.
Typical tornados are a quarter-mile in diameter.
The average path length of tornados in thevicinity of the site is 8.5 miles, but for conservatism 10 miles is used, giving a path area of
The probability of a tornado affecting the TMI-2 site is therefore 6.3 x 10-4
2.5 square miles.
per year, or once in 1600 years.

10-28The potential for release of radioactive material is not just a function of the probability of
any tornado striking the site, but must also include consideration of severity. The design basis
tornado is a very severe one, and one that is not likely to occur in Pennsylvania.
Based onobservations mainly in the Midwest and South Central United States, the combined probability of a
tornado having maximum winds above 275 mph striking the site is 10-7 per year.
The design basis
tornado, with wind speeds of 360 mph, would have a significantly lower probability of occurance.
10.5.2.2

Potential

For Water-Processing-Related

Releases

The procedures used and protection provided during water processing preclude tornado-induced
releases that could endanger the environment or public health and safety. The protection describedin Section 10.5.2 is sufficient to mitigate the consequences of high tornado winds, pressure
drop, and missiles.
Tornado missiles cannot penetrate the safety-related structures (reactor
building and the auxiliary and fuel handling buildings) and cannot compromise the integrity of
the casks housing the EPICOR II prefilters and liners in the chemical cleaning building.
The exposed processed water storage tanks are susceptible to tornado damage from high winds or
missiles.
These tanks could release processed water containing an average of less than
0.6 pCi/mL tritium.
This tritiated water would likely be drawn into the tornado and then bedispersed by the violent mixing action of the air flow inside the tornado.
Even if all theavailable processed water (about 1.5 x 106 gallons) were removed from tanks by the tornado andblown into the Susquehanna River, and if the river were at its 50 percentile flow rate of
20,000 cfs (9 x 106 gpm), maximum permissible concentration (MPC) limits (10 CFR Part 20) for
tritium are not likely to be exceeded.
The York Haven Dam and the Red Hill Dam form a 10,000 acre-foot (3 x 109 gallon) water impoundment in the vicinity of the TMI-2 site. Based on downstream channel volume, it is assumed conservatively that 20 percent of the water impounded is effective in diluting the 1.5 x 106 gallons of processed water, resulting in a downstream tritium concentration of 2.5 x 10-3 pCi/mL,
which is 83 percent of the MPC limit for continuous release of tritium (3 x 10-3 pCi/mL).
Since
all users of surface water downstream of Three Mile Island are also downstream of York Haven Dam,the sudden, one-time release of processed water would pose a very minimal risk to the environment.
10.5.2.3

Potential

For Releases From Waste Storage Areas

Both the interim waste storage facility and the concrete waste storage facility are fully protected from the effects of tornados.
The cells of both facilities are covered with 3-ft-thickconcrete tops and both are essentially at grade elevation.
Tornado winds, pressure drop, andmissiles cannot compromise the integrity of the cells.
10.5.3

Potential for Releases Due to Aircraft Impact

The potential risk to public health and the environment has been evaluated for the impact oflarge aircraft on liquid and solid radioactive waste processing and storage facilities.
Criticalsafety-related structures (Category 1) on the site, such as the reactor building and the auxiliary and fuel handling buildings, originally were designed to withstand the impact of large
aircraft (300,000 pounds) at 200 knots.
Therefore, the sources of relatively high-level radioactivity, such as the nuclear fuel and the reactor building sump water, are protected from theeffects of aircraft impact.
Likewise, any cleanup operations, that may occur inside these buildings also are protected.
There are, however, some sources of radioactivity outside of these buildings designed to withstand aircraft impact, and the potential vulnerability of these sources has been evaluated.
The
facilities or components that have been considered are the chemical cleaning building housing
EPICOR II, the concrete waste storage facility, the interim waste storage facility, the borated
water storage tank (which is used to store processed water containing tritium), and the processedwater storage tanks.
10.5.3.1

Possibility of Aircraft Impact

An evaluation has been done by the staff to determine the probability of impact of large aircrafton radioactive material storage facilites at TMI-2.
On page 31 of Reference 7, the staff determined the impact probability for each potential target to be 1.6 x 10-8 per year.
Therefore, theaircraft impact probability is a factor of approximately 6 smaller than the 10- 7 /yr probability
that is often used as a limit beyond which events are considered highly unlikely or incredible,and not seriously considered in design.

10-2910.5.3.2

Potential For Water-Processing-Related

ReleasesDuring processing of liquids, contaminated water is not stored in tanks susceptible to aircraft
impact.
It is either pumped from a protected building directly to a water treatment system, suchas EPICOR II, or it is decontaminated inside a protected building, as would be the case with thesubmerged demineralizer system.
Processed Water Storage
The processed water resulting from cleanup operations contains tritium and may be stored outside
in tanks vulnerable to aircraft impact, such as the borated water storage tank or the processedwater storage tanks.
However, the potential risk associated with this storage is minimal becauseof the very low probability of aircraft impact (on the order of 4 x 10-10) and the low consequence of release of the tritiated water.
As was described in Section 10.5.1, there is a flood
protection dike around the site, and this dike would serve to contain processed water that might
be released by aircraft impact.
The expected average tritium concentration in processed water
after cleanup of reactor building sump water and reactor coolant system water is less than
0.6 pCi/mL, and the slow leakage or evaporation of this processed water does not present a publichealth hazard.
Chemical Cleaning Building (EPICORII)The chemical cleaning building tanks used for interim storage of processed water are susceptible
to failure caused by an aircraft impact, but the water would be contained by the flood protection
dike.
The highly radioactive EPICOR II prefilters and liners are protected by 12-inch-thickreinforced-concrete walls surrounding the processing area. The prefilter/demineralizer is installed inside a 12-inch-thick cylindrical concrete cask, and the cask is surrounded by a lead brickwall 5 inches thick.
The top of the prefilter/demineralizer is covered with a lead shield and5-inch-thick steel lid. The second- and third-stage demineralizers are similarly protected.
Theroom containing the demineralizer stages is further protected by 24-inch-thick concrete walls
inside the 12-inch-thick exterior walls.
Considering the substantial amount of impact resistanceand the very low probability of aircraft impact, it is considered highly unlikely that radioactive
releases would occur from the chemical cleaning building as a result of aircraft impact.
10.5.3.3

Potential for Releases from Waste Storage Areas

Relatively high-level wastes such as water processing prefilters and liners will be stored in theinterim radwaste storage facility on the site.
The potential for releases occurring as a resultof aircraft impact on these waste storage areas was evaluated as being very small.
The temporary
radwaste storage facility, described in Section 10.5.1.1, had prefilter and resin material protected from aircraft impact by the 3-ft-thick concrete cell top and the primary steel container.
These have been removed to the interim radwaste storage facility.
The interim radwaste storage
facility described in Section 10.5.1.2 also has a 3-ft-thick concrete cell top and is well protected from aircraft impact.
In view of the substantial concrete protection afforded the waste
and the very low probability of aircraft impact, it is highly unlikely that such an accident
would result in any release that would endanger public health or safety.
10.6

PSYCHOLOGICAL-SOCIOECONOMIC EFFECTS

The staff concludes that low-level chronic distress could continue throughout the cleanup period,and some groups will be more affected by stress than others.
Most of the impacted public currently demonstrate no psychological effects that have detrimental health consequences.
Exact
estimates of long-term psychological consequences associated with the accident or the cleanup are
not available.
However, one study suggests that even among the most vulnerable of the impacted
groups, the distress levels measured on the first anniversary of the TMI accident are below thethreshold associated with health problems and are not much higher than control group scores.
The staff believes that the potential level of distress generated by the decontamination process
could be related to the present psychological setting and factors surrounding decontamination.
The present psychological setting for those who are distressed is characterized by uncertainty
over the effect of released radiation on health and future generations, NRC's and Met-Ed's lackof credibility, and the believed inability of Met-Ed to safely conduct and NRC to review thevarious decontamination procedures.
Those factors which could contribute to the depth and breadthof future distress levels include the length of time required for overall decontamination and forspecific decontamination procedures.
In general, it is the staff's belief that as time spent on
decontamination activities increases, the consequences of distress could increase.
Other factors

10-30that could influence future levels of potential distress include the believed probability ofaccidents, the type and duration of media coverage, the credibility of those managing the decontamination process, and believed levels of safety and health impacts.
Considerable public concern has been expressed regarding the prospect of processed water being
released into the Susquehanna River.
Much of this concern centers on possible disruption ofrecreational and commercial fishing activity on the Bay due to a perceived threat to health,whether justified or not.
The staff concluded (Section 7.2.5.4) that the disposal of tritiatedwater in the Susquehanna has the potential to produce dispersed socioeconomic impacts as a consequence of distress and uncertainty.
These impacts could include the temporary avoidance ofdrinking water, of recreational opportunities, and of seafood and waterfowl supplied by the riveror the Bay.
Potential economic losses to watermen, processors, restaurants, and other retail andservice outlets might be significant.
As discussed in Section 7.2.5.5, the State of Maryland
will be undertaking a study of the potential economic losses to persons who depend upon Chesapeake
Bay activities.
There is also the potential for public uncertainty over radiation effects andwidespread economic loss. This condition could require mitigative measures to help reduce publicanxiety if the Susquehanna disposal option is selected.
An appropriate mitigation program shouldseek to relieve several problems including (1) the NRC's and Met-Ed's lack of credibility, (2) the
limited knowledge reflected in some of the media and general public perceptions of radiationrelated impacts, (3) the inadequate dissemination of consistent information by public agencies,and (4) the inadequate incorporation of socioeconomic and ecological considerations in the decision on the timing of disposal.
The transportation of waste was also indicated as having the potential for producing socioeconomic impacts.
The expected range of number of truck shipments is 353 to 997.
These shipments
would be made over a period of 5 to 9 years; hence, the average rate of shipments during thisperiod would range from about one truckload every three days to about one truckload every fivedays.
Those living along or in proximity to the truck route could be subjected to stress.
The
staff judgment is that the marketability of residential property abutting the route throughMiddletown could be adversely impacted during the period of shipment.
Also, the staff would
expect that some residents will alter their daily schedules to avoid the shipping route at certain
times.
The staff believes that the potential increase in stress and annoyance can be mitigatedthrough a process that considers the apprehensions of those living in the Middletown transportation corridor.
To accomplish this end, the staff considers it necessary to continue the ongoing
public dialogue to discuss the structural integrity of shipping containers, the low probabilityof radiation leakage under accident conditions, and the optimal timing for shipping waste.
Although they would have negligible offsite effects, airborne releases as d result of postulated
accidents could create public uncertainty over health and safety with consequent impacts to theagricultural sector.
As noted in Section 3.6.2.3, "milk juggers"--local dairy farmers who sellmilk directly to retail customers--experienced sales declines subsequent to the TMI accident.
The staff believes that because of the small quantity of radioactive material potentially involved
and the existing monitoring programs, such losses, should they occur, would be of short duration.
10.7

IRREVERSIBLE AND IRRETRIEVABLE COMMITMENTS OF RESOURCES

Many resources will be used in the cleanup of TMI Unit 2.
Some of those uses involve irreversible and irretrievable commitments.
Irreversible commitments concern changes set in motion by
the proposed action that at some later time could not be altered so as to restore the present
order of environmental resources.
Irretrievable commitments are the use or consumption of
resources that are neither renewable nor recoverable for subsequent utilization.
10.7.1

Commitments Considered

The types of resources of concern in this case can be identified as:(1) material resources-materials of construction, renewable resource material, and depletable resources consumed, and(2) nonmaterial resources--including a range of beneficial uses of the environment.
Resources that generally may be irreversibly committed by the construction and cleanup are:(1) construction materials that cannot be recovered and recycled with present technology,(2) materials that are rendered radioactive but cannot be decontaminated, (3) materials consumed
or reduced to unrecoverable forms of waste, (4) the atmosphere and water bodies for disposal ofwaste effluents, to the extent that other beneficial uses are curtailed, and (5) land areas
rendered unfit for their original uses.

10-31
10.7.2

Material Resources

10.7.2.1

Construction Materials

Materials of construction are almost entirely of the depletable category of resources.
Concrete
No commitments have been made on whether theseand steel constitute the bulk of these materials.
materials will be recycled when their present use terminates.
The amounts to be used for thecleanup are not known at this time because detailed procedures have not been determined; however,quantities of construction materials used will be very small compared to annual U.S. productionof these materials.
10.7.2.2

Replaceable Components and Consumable Materials

Materials consumed or reduced to unrecoverable status are chemicals such as detergents used forthe cleanup and gasoline and diesel fuel used in vehicles bringing materials and the workforce toand from the site, for haulage onsite, and for transportation of waste material offsite. The
amounts used in local transport are unpredictable; however, the transportation of wastes tolicensed disposal facilities will require from 1/2 to 1 1/2 million gallons of diesel fuel,depending on the distance to the burial sites and the waste forms selected.
The materials used for processing, immobilizing, and packaging the wastes have not yet beenchosen, but the candidates are ion-exchange resins, vinyl ester styrene, portland cement, andbitumen.
The organic material, resins, vinyl ester styrene, and bitumen are derived from petroleum, but the quantities required are small in comparison to the total world production of petroleum.
Portland cement may be used for immobilizing the wastes or as an alternative to processing
the wastes.
Here again, the quantities expected to be used are small compared to U.S. production.
10.7.3

Water and Air Resources

Because of the rapidly renewable nature of the Susquehanna River and the regenerative powers andvast dispersive capacity of the atmosphere, the use of these resources to dilute and disperse theeffluents of chemicals and radioactive materials from the cleanup of TMI-2 is not considered torepresent irreversible or irretrievable commitments of these resources.
10.7.4

Land Resources

Land required for the burial of low-level radioactive wastes is estimated to range from 1/2 to
1 acre, depending upon the amount and volume of the wastes.
High activity wastes could be accommodated within these estimates, assuming land, burial is permitted.
Nuclear fuel from the reactor
would require from 66 ft 2 to 216 ft 2 in a storage pool at a presently undesignated location.
10.8

SUMMARY OF ECONOMIC COSTS OF THE TMI-2 CLEANUP

Numerous alternatives have been presented in each of the previous sections of this statement.
Because each section of this statement represents a particular evolution in the cleanup of TMI,
the combinations of alternatives and evolutions produce an enormous number of reasonable paths
that could be selected for the actual cleanup.
Therefore, the staff has developed bounding costestimates for each of the major sections of this statement. Table 10.19 summarizes these costs.
The costs stated in Table 10.19 have been developed based upon constant 1980 dollars and do notinclude the time value of money nor the inflationary conditions that may exist in the future.
Additionally, costs have been excluded unless they can be directly associated with a given alternaCosts that have been excluded from Table 10.19 are:maintaining the plant in a safe
tive.
shutdown condition, site security costs, licensing costs, housekeeping costs, receipt/ inspection
costs, planning and scheduling costs, warehousing costs, material handling costs, general technicalsurveillance, including quality assurance, health physics and engineering.
In addition, costs
have not been included for construction of general support facilities, such as office and generallaboratory buildings.

10-32Thus, the economic data presented here, and elsewhere in the PEIS, are only for the purpose ofcomparison among alternatives of direct costs. The costs given here can not be summed to give a
total cost estimate for the cleanup.
Some generalities can, however, be stated:
The selection of alternatives will be an important factor in determining the total
cleanup cost, but the timing of the cleanup will most likely dominate the total cost.

There is a large uncertainty in the relative costs for most of the alternatives. The
uncertainty in individual cost estimates could, in some cases, be greater than the costdifferences between alternatives.
Because of the uncertainty in estimated costs among the alternatives, and because public andworker safety, as well as protection of the environment, are the paramount concerns, review ofalternatives based on costs is a secondary consideration.

Table 10.19.
Summary of Alternative Egonomic
Costs for the Cleanup of TMI-2(thousands of dollars)

Estimated Cost Range

PEIS
Section

Low

High

5.1

AFHB cleanup$16,000$ 22,000

5.2

Reactor building cleanup

25,00063,000

6.2

Cooling system inspection

1,400

5,900

6.3

RPVH and internals removal

3,600

6,600

6.4

Core examination and defueling

12,00016,0006.5

RCS decontamination

2,600

6,900

7.1

Liquid waste treatment

21,00029,000

7.2

Disposal of processed accident water100

11,000

8.1

Process solid wastes

23,00027,0008.2

Chemical decontamination solutions

2,000

10,000

700

7,100

2,600

11,0008.3

Solid materials

9.0

Storage, transportation,of solid waste

and disposal

aThis summary of economic costs is for the purpose of making comparisonsamong alternatives.
The relative costs can not be summed to arrive at a
total cost estimate for the entire cleanup.

References--Section 10
1.

"Ionizing Radiation:
Levels and Effects," Volume II:
Effects, a Report of the United
Nations Scientific Committee on the Effects of Atomic Radiation to the General Assembly,
1972.

2.

"Effects on Populations of Exposure to Low Levels of Ionizing Radiation," National Academyof Sciences Advisory Committee on the Biological Effects of Ionizing Radiations, November
1972.

3.

NRC Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases ofReactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"
Office of Standards Development, U.S. Nuclear Regulatory Commission, October 1977, Rev. 1.

10-33
Battelle

4.

"Age-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake,"
Pacific Northwest Laboratory-NRC, NUREG-0172, November 1977.

5.

"Environmental Survey of Transportation of Radioactive Material to and from Nuclear Power
Plants." U.S. Nuclear Regulatory Commission, WASH 1238, December 1972, and Supplement,
NUREG-75/038, April, 1975.

6.

L. Garfinkel and E. Silverberg, "Cancer Statistics,
Professional Education Publication, p. 7, 1979.

7.

L.J. Chandler and S.A. Treby, "NRC Staff Posthearing Memorandum Regarding Aircraft Crash
Probability Issue," Before the ALAB in the Matter of Metropolitan Edison Company, et al.
(Three Mile Island Nuclear Station, Unit 2). Docket No. 50-320, April 30, 1980.

1979,"

American Cancer Society

11.

11.1

ENVIRONMENTAL RADIOLOGICAL MONITORINGINTRODUCTION

The radiological environmental monitoring around the TMI site and nearby communities during thedecontamination of Unit 2 would be performed by (1) Metropolitan Edison Company (the licensee),(2) the U.S. Environmental Protection Agency (EPA), (3) The Commonwealth of Pennsylvania, (4) the
U.S. Department of Energy (DOE),(5) the Nuclear Regulatory Commission (NRC), and (6) the Stateof Maryland.
Each program is summarized in the following subparagraphs; a more complete description is given in the EPA report, "Long-Term Environmental Radiation Surveillance Plan for Three
Mile Island," 1981, which is provided as Appendix M to this statement.
11.2

METROPOLITAN EDISON COMPANY RADIOLOGICAL MONITORING PROGRAM

The Met-Ed radiological environmental monitoring program which will be in effect during thedecontamination of Unit 2 is a combination of the TMI-1 and TMI-2 Environmental Technical Specification required programs and increased monitoring activities which were initiated after March 28,
1979.
This monitoring program is subject to change based upon review of the results and any
requests for additional monitoring.
The licensee's radiological monitoring program is comprehensive, covering sampling of air, milk,
water, fish, aquatic plants, sediments, miscellaneous food products, and exposure rates in theenvirons in and around the TMI facility to a distance of about 21 miles.
The licensee's air sampler network consists of eight stations which are sampled weekly using both
air particulate filters and charcoal cartridges.
Air particulate samples are analyzed weekly forgross beta activity, and gamma spectral analysis is also performed monthly. A radioiodine analysis is performed on the charcoal cartridge.
For a quarterly composite of the air particulate
samples, analyses for Sr-89 and Sr-90, gross alphas, and gamma spectra are made.
Met-Ed's milk network samples semimonthly from five farms in the offsite area.
Radioiodine andgamma spectra determinations are performed on these samples.
A Sr-89, Sr-90 analysis is performed on quarterly composites of these samples.
Water samples from Met-Ed's offsite water sampling network are collected from eight stations.
These samples are composited hourly over a two-week period utilizing automatic water samplers.
These semimonthly samples are analyzed for iodine (semimonthly), gamma scan and gross beta
analyses on monthly composite, tritium on a monthly and quarterly composite, and Sr-89 and Sr-90
on a quarterly composite.
In addition, grab samples are taken weekly at two surface water sta
tions.
These are composited and the above analyses are performed.
Daily grab samples are also
taken from the plant discharge and composited for the above analyses.
Fish, aquatic plants, and aquatic sediments are sampled periodically,
food products as they become available.

as well as miscellaneous

Met-Ed operates an extensive thermolumenescent dosimeter (TLD) network for monitoring environmental exposure rates in the area around the site. These dosimeters are exchanged on a monthly(20 locations) and a quarterly (53 stations) basis.
Met-Ed has a groundwater monitoring program (see Fig. 11.1) that presently samples from 15 observation and monitoring wells.
Tritium analysis and gamma scans are performed on the samples
taken.
11.3

U.S.

ENVIRONMENTAL PROTECTION AGENCY RADIOLOGICAL MONITORING PROGRAM

EPA has been designated by the Executive Office of the President as the lead Federal agency for
conducting a comprehensive, long-term environmental radiation surveillance program as a followup
to the accident at TMI-2.
As of December 31, 1980, EPA operates a network of 13 continuous

11-1

11-2

MW-2

0)

MW-4
DW
STORAGE
TANK

JMW-5

1nT
LegendBWST- Borated Water Storage Tank
DW -Decontaminated Water

OW -Observation Well

MW -Monitor Well

0

OW-.13

OW-14

MW-6

10 MW-7
COMMENTS:
N 301,460.04
E 2,286,538.94
N292,985.44
2.0W-15 LOCATED ON SOUTH END OF ISLAND AT COORDINATES E 2,287,765.09
I. MW-I LOCATED IN THE NORTH PARKING LOT AT COORDINATES

Figure 11.1.

Monitor Well Locations at TMI-2.

11-3
air-monitoring stations at radial distances ranging from 0.5 mile to 5 miles from TMI.
Five
miles was established as the point well beyond that which EPA expects to detect any emissions
from TMI-2.
Each station includes an air sampler, a gamma rate recorder, and three TLDs.
The
air sampler units operate at approximately 2 cfm, and particulate samples are collected from eachstation and analyzed typically three times per week.
Charcoal filters are collected and analyzed
weekly.
All samples are analyzed by gamma spectroscopy at EPA's TMI Field Station, Middletown,
using a Ge(Li) detector with a lower limit of detection for Cs-137 or 1-131 of about 25 pCi(0.15 pCi/m 3 for a 48-hour sample).
Each monitoring station is equipped with a gamma-rate recorder for measuring and recording external exposure.
Recorder charts are read on the same schedule used for air sample collection, andthe charts are removed weekly for review and storage at EPA's laboratory in Las Vegas, Nevada.
EPA is presently planning to install telemetered gamma monitors at these stations. When thischange is effected, EPA will revise its collection and analysis of charcoal and particulate
filters to once a week.
Thermoluminescent dosimeters have been placed at each monitoring station as well as at a representative number of population centers surrounding TMI.
Locations are shown in Appendix M ofthis programmatic statement (Appendix D of EPA surveillance plan).
These dosimeters are readquarterly.
In addition to the above, weekly compressed gas samples are taken at Yorkhaven, Goldsboro, Middletown, and at the TMI Observation Center opposite the plant and sent to EPA Las Vegas for a determination of krypton-84.
Tritium in air samples will also be collected from these stations.
The EPA's base long-term program discussed above will continue and can be augmented in the future
if particular decontaminations make it necessary.
Such augmentation of the monitoring program
for the venting of the Kr-85 from the reactor building consists of survey meter and ion chamber
measurements, collection of compressed air samples for Kr-85 analysis, and intensified collectionof samples from routine air monitoring stations.
EPA also collects and analyzes water samples as follows:(EPA does gamma spectroscopy, DER
analyzes for tritium, gross alpha and gross beta; weekly composites are analyzed for strontium-89
and 90 at the Eastern Environmental Radiation Facility, EPA, Montgomery, Ala.)(1)

TMI outfall(2)

Lancaster Water Works intake - daily,(3)

City Island - (upstream river water) - weekly,(4)

Sediment pond, TMI (runoff water) behind Unit 2 cooling tower.

(all plant discharge,

both units) - daily,

andThere is a continuous gamma monitor on the 001 TMI outfall with a high-level alarm that automatically alerts EPA and DER to the presence of gamma activity in the water in excess of1000 pCi/L Cs-137 (1/20 of permissible level).
EPA reports all results of monitoring measurements
week to the public and news media.
11.4

from their baseline program on Friday each

COMMONWEALTH OF PENNSYLVANIA RADIOLOGICAL MONITORING PROGRAM

The Department of Environmental Resources of the Commonwealth of Pennsylvania operates threecontinuous air-sampling stations; one at the Evangelical Press Building in Harrisburg, one at the
TMI Observation Building, and one in Goldsboro near the boat dock.
Each air-sampling stationconsists of a particulate filter followed by a charcoal cartridge. The filters and cartridges
are changed weekly; the particulate air samples are gamma scanned and beta counted for reactorrelated radionuclides.
The particulate air samples are composited quarterly and analyzed for
Sr-89 and Sr-90.
The charcoal samples are gamma scanned for reactor-related radionuclides.
The
Commonwealth, however, does not have the capability to sample or analyze for Kr-85.
The Commonwealth's milk sampling has reverted to its routine surveillance program, which consistsof monthly milk sampling at two dairy farms near the site.
The milk samples will be gamma scanned
for all reactor-related gamma-emitting radionuclides.
The Commonwealth has TLDs at ten locations
which are collected and read monthly.
The Commonwealth also will collect local produce and fish11-4
in season.
The produce and fish samples will be analyzed by gamma spectroscopy for any reactorrelated radionuclides.
The Commonwealth also participates with EPA in monitoring the principal
aqueous outfalls of TMI-2.
11.5
11.5.1U.S.

DEPARTMENT OF ENERGY RADIOLOGICAL MONITORING PROGRAM

Radiological Monitoring Program

In-situ gamma
DOE will provide soil and vegetation analyses from seven sites semiannually.
spectrometry analyses will be conducted at these seven plus one additional site. TLDs are also
If levels of radionuclides demonin place at these sites plus four state monitoring locations.
strate any increase above background levels, the samples will be subjected to detailed radiochemical analyses.
11.5.2

Atmospheric Release Advisory Capacity

DOE has available for significant radiological releases to the atmosphere its Atmospheric Release
Advisory Capacity (ARAC).
This ARAC system provides independent predictions of the dispersion
patterns for such releases based on local meteorological data and National Weather Service reports.
These predictions use atmospheric dispersion models which have been verified during many years offield experience and tests in government programs.
The predicted dispersion patterns would be
provided to EPA, the utility, and NRC to serve as a basis for their positioning of ground-levelmonitoring teams.
11.5.3

Community Monitoring Program

DOE and the Commonwealth of Pennsylvania are sponsoring a community radiation monitoring program.
This program has as its purpose to:(a) provide independent verification of radiation levels in
the TMI area by trained local community people, and (b) increase public understanding of radiation and its effects.
The approach to achieving this purpose has involved the selection ofindividuals by local officials from the following 12 communities within about five miles around
TMI:
East Manchester Twp.
Londonberry Twp.
York Haven
Lower Swatara Twp.
Conoy Twp.
Goldsboro

Fairview Twp.
RoyaltonWest Donegal Twp.
Middletown
Newberry Twp.
Elizabethtown

About 50 individuals participated in training classes conducted by members of the Nuclear Engineering Department of the Pennsylvania State University.
About 15 training sessions were conducted involving classroom instructions, laboratory training, and actual radiation monitoring in
the field. The teams utilized EPA gamma rate recording devices which are in place around TMI andwill be supplemented by gamma/beta-sensitive devices which are being furnished by DOE throughEG&E Idaho, Inc.
The training sessions provided basic information on radiation, its effects and detection techniques, and included hands-on experience with monitoring equipment in the field.
Following thecompletion of training in the third week of April, team representatives in each of the 12 selected areas began data acquisition from the gamma and gamma/beta-sensitive instruments on a routine
basis.
Detailed procedures were developed to consolidate the information being obtained into acentral point of contact in the Commonwealth of Pennsylvania for dissemination to the press,
local officials, and other interested parties on a routine basis. Maintenance and calibration
procedures also were developed and were in effect prior to the initiation of routine field monitoring.
The community monitoring program was initiated on May 21, 1980, and the results ofmeasurements from this program were reported daily to the public.
Presently only the units at
Fairview, West Donegal and Newberry continue to be active.
11.6

U.S.

NUCLEAR REGULATORY COMMISSION RADIOLOGICAL MONITORING PROGRAM

The NRC operates one air-sampling station located in the middle of the reactor complex.
The air
sample is changed weekly and will be analyzed by gamma spectrometry.
The NRC places two sets of
TLDs at 59 locations in the area around TMI-2.
Each set contains two lithium borate and two
calcium sulfate phosphors.
The lithium borate phosphor has the ability to detect beta radiation

11-5
from Kr-85.
Both sets will be read on a monthly basis; however,set at more frequent intervals should conditions warrant.
11.7

flexibility exists to read one

STATE OF MARYLAND RADIOLOGICAL MONITORING PROGRAM

In early April 1979, the State of Maryland, Department of National Resources (Radiation Section,
Division of Environmental Chemistry) began monitoring the levels of radionuclide activity infishes and sediments of the Susquehanna River at Falmouth (about 1 mile downstream of Three Mile
Island), at several locations near and downstream of Safe Harbor Dam and Holtwood Dam, and in theConowingo Pond.
The Chesapeake Bay also has been monitored at several locations from the mouthof the Susquehanna River to about Worton Point (south of the Sassafrus River).
Fishes, oysters,
blue crab, and bay sediments have been studied.
Monitoring of both the river and bay will continue during the cleanup of Three Mile Island.
Since the accident, the Maryland Department of Health and-Mental Hygiene (Division of Radiation
Control, Environmental Health Administration) has been testing Susquehanna River water for radionuclide content using weekly composite samples collected at Conowingo Dam.
This sampling willcontinue.
For discharging tritiated water at Three Mile Island (should that be authorized),strong consideration will be given to accelerating the program, as appropriate, to monitor radionuclides in the water.
Consideration also will be given to sampling upstream of the Conowingo
Dam sampling point in order to differentiate more accurately the radionuclides originating at
Three Mile Island from those originating at Peach Bottom.
Users of Susquehanna River water
within Maryland will be advised, as appropriate, on actions to be taken, should that become
necessary.
11.8

CONTINGENCY SURVEILLANCE PROCEDURES

Contingency planning for the protection of the public must address the possibility of unplanned
releases of airborne radioactivity to the general environment, as well as liquid releases to theSusquehanna River.
In the event of a release of airborne radioactivity in excess of the licensee's Technical Specifications limits, the EPA On-Site Coordinator will be notified by the NRC.
NRC health physics
personnel would be supported by radiation monitoring equipment and analytical capabilities,
including the NRC Region I mobile laboratory.
Additional NRC personnel would be onsite within
two hours; the location of the mobile laboratory at the time of the occurrence would dictate its
response time.
The Senior NRC Site Representative will ensure that the EPA On-Site Coordinatorhas access to current release data and meteorological information.
In addition, the DOE Emergency Coordination Center will be notified by NRC and may be requested to provide aerial, measurements and plume tracking.
The response time for an aircraft to reach TMI can be expected to befrom two to three hours under normal conditions, with a six-hour maximum under virtually any
condition.
Air sampling will be a measure of inhalation exposure and potential contamination of milk andfood crops.
Should a prolonged airborne release occur, supplemental air monitoring stations willbe established.
Ten air samplers will be kept available by EPA for this purpose.
EPA also will
have three compressed air samplers available for krypton gas sampling.
Apparatus to sample airfor subsequent tritium analysis also will be available for prompt use by EPA.
Releases of contaminated water that are above the licensee's permitted level for discharge to theSusquehanna River should not occur.
The contingency plan for releases above the licensee's
permitted level involves prompt confirmation of the released activity by composite sample analysis, followed by notification of the impact of the release to downstream users.
EPA's Region III
Office will be responsible for notifying adjoining states.
11.9

REPORTING PROCEDURES

There are two types of data-reporting procedures.
The first type is designed to distributeinformation upon which immediate action might be taken and consists of informal reporting methods;
the second procedure is designed to provide a verified data base.
In addition, there will be aprocedure for reporting information to the news media.

11-6
11.9.1

Immediate Reporting Procedures

Each of the monitoring agencies will promptly inform the other monitoring participants of confirmed, positive levels of reactor-related radionuclides through the EPA onsite representative.
He will promptly relay the information to the other organizations by telephone or in person toeach Federal agency and the Commonwealth of Pennsylvania, followed in either case by written
documentation of the event.
Periodic meetings are called by EPA at TMI to discuss, proposed and ongoing operations which could
impact the offsite agencies and to exchange information.
11.9.2

Reporting Data into the Data Base

All data are reported in the format previously specified by EPA.
Data from NRC, DOE, and theCommonwealth of Pennsylvania are submitted to EPA monthly for inclusion in the data base.
EPA
data also are placed in the data base monthly.
On a monthly basis, EPA will place data obtained from Metropolitan Edison and the Commonwealth ofPennsylvania, as well as relevant data from other organizations, into the data base.
EPA thenwill use computer transfers to transmit monthly updates to the data base to the originatingorganizations for verification.
All data will be verified by the originating organization within
15 days of receipt.
Any errors will be referenced by sample number for correction.
Periodic
updates will be made available to all participants.
11.9.3

Reporting Information to the News MediaThe EPA is the lead Federal agency responsible for distribution of environmental data to the news
media.
Each participant will keep each of the other participants in this plan advised in advanceof pending media releases concerning TMI.
As appropriate, releases will be coordinated with
Metropolitan Edison Company.
11.10

MONITORING OF AQUATIC ECOLOGY AND RECREATIONAL FISHERIES OF THE SUSQUEHANNA RIVER

Appendix B of the operating license for TMI-2 contains Environmental Technical Specifications(ETS) that require Metropolitan Edison Company to monitor station effluents and the SusquehannaRiver for nonradiological impact from operation.
The ETS for Unit 2 were issued by NRC on February 8, 1978, and require three years of operational studies of:(1) physical and chemical
characteristics of the York Haven Pond of the Susquehanna River (see Fig. 11.2); (2) thermal andchemical characteristics of the cooling water discharge effluent; (3-) general ecology of thepond, including benthic macroinvertebrates, fish eggs and larvae, fish populations, fish impingement, and fish egg and larvae entrainment; and (4) recreational fishery creel survey of fishingeffort, catches, and harvests in the pond and other nearby fishing areas.
Additionally, the ETS
require the licensee to make a prompt report to NRC of any unusual or important events, such as
fish kills, near or downstream of the station.
These studies are a continuation of those that have been ongoing since Unit 1 commenced operationin 1974.
Results have been presented in a series of annual reports as required by the ETS.
Annual reports for the years 1974 through 1978 are complete and constitute a five-year preaccident record of environmental conditions in York Haven Pond.
The annual report for the firstpostaccident year of 1979 is complete.
These reports have formed the bases for impact assessments in the 1976 NRC Final Supplement to the Final Environmental Statement on the operation of
TMI-2 (NUREG-0112) and assessments presented as testimony during the 1977 environmental hearings
in Harrisburg.
Following the accident, those annual reports plus other independent data sourceswere used in assessing the status of the aquatic biota and the recreational fishery of the rivernear TMI, with results contained in NUREG-0596.
The annual reports and the postaccident biological assessment are used as informational sources in this PEIS and are contained in References I through 7 of Appendix E.
These nonradiological studies do not directly monitor either environmental radiation levels or
their dose effects, but they do provide information on ecological status of the river upstream,
near, and downstream of the nuclear station, and in all three channels of the York Haven Pond(see Appendix E, Fig. E.1 for reference).
Fish disease, parasite, abnormality, and mortality
conditions are routinely monitored during the river studies.
These are useful for short-term(mortalities or fish kills) and long-term (disease) effect studies after an accident or radiological release event.
The studies have defined (and continue to monitor) those areas of the11-7

Figure 11.2.

Sampling of Susquehanna River. Environmental techniciansroutinely sample water upstream and downstream from TMI.

The samples are analyzed for several materials, althoughno accident-generated water may be discharged from thestation. (Official TMI Photo)11-8
river near TMI that are most important to the fishes as spawning and nursery areas.
Radiologicalreleases (planned or accidental) therefore can be evaluated in relation to their presence orabsence in important areas of the river.
Monitoring of the recreational fishery permits an
examination of TMI-2 operation on the fishery resource at the point of exploitation.
The studies
define those areas of the river most important to fishermen and provide information on the foodchain that can be used for assessing the liquid radiological pathway to man via finfish consumption.
These programs monitor the aquatic biota and sport fishery in that segment of the Susquehanna River where the TMI effluent first enters, where it is the least dilute, and where effects(if any) would be seen first.
The Commonwealth of Pennsylvania National Pollutant Discharge Elimination System (NPDES) Permit
requires in-plant monitoring of thermal and chemical effluents, as well as entrainment and impingement.
The combination of the studies required by the NPDES and the ETS provide a spectrum ofdata needs for analysis of the environmental status of the aquatic resources of the river. Bothgroups of studies are continuing, and data summaries are provided to the respective agencies on amonthly basis.
The annual report required by the ETS is on a calendar year basis.

12.

CONCLUSIONSIn this programmatic environmental impact statement, the NRC staff has evaluated the environmental impacts and other costs and benefits associated with the proposed cleanup of Three Mile
Island Unit 2.
As a result of its evaluation, the staff has made the following findings andconclusions:12.1 Conclusions on Environmental
1.

Impacts and Other Costs

The cumulative radiation dose received by the entire work force would be in the range of
It is predicted that less than one
2000 to 8000 person-rem for the whole cleanup program.
additional cancer death attributable to exposure to radiation will occur among the workforce (the death rate from cancer due to other causes among the U.S. population averages
approximately 200 deaths per 1000 people).
Not more than two additional genetic effects
in descendents of the workers are expected to occur (among the U.S. population, approximately 60 genetic defects can be expected per 1000 people).
This is the most significant(See Sec. 10.2)
radiological impact expected from the cleanup activities.
The occupational dose to an individual worker will be limited to 3 rem/quarter in accordance with 10 CFR Part 20; however, the exact dose below 3 rem/quarter to any one individual
cannot be determined due to lack of information about specific work assignments.

2.

Throughout the cleanup, any anticipated releases to the environment must be controlled by
the licensee in accordance with the staff's proposed effluent criteria to conform to theindividual dose design objectives listed in 10 CFR Part 50, Appendix I, as mandatory
limits.
The total-body dose design objectives are 15 mrem/year from airborne particulate(See Sec. 10.3.4)
releases and 3 mrem/year from liquid releases.
Decontamination methods and technology are available which can be used to complete thecleanup in accordance with the offsite dose limits stated above.
If the cleanup is conducted in accordance with the staff's proposed effluent criteria, the staff estimates
that, for the entire cleanup, the total body dose to the maximum exposed individual offsite will range from 0.8 to 2.3 mrem for gaseous effluents and from 0.0015 to 1.1 for
liquid effluents.
The cleanup is expected to take from 5 to 9 years to complete.

3.

An individual offsite receiving the maximum estimated dose resulting from atmosphericreleases during the entire cleanup (0.8 to 2.3 mrem) would incur an estimated increased
risk of dying from cancer of between 1 in 2 million and 1 in 600,000, and an increased
risk of a genetic effect to offspring over the next 5 generations of between 1 in 300,000
and 1 in 100,000.
As a result of liquid releases which may occur over the entire cleanup
period, an individual receiving the maximum estimated dose (0.0015 to 1.1 mrem) would
incur an estimated increased risk of dying from cancer of between 1 in 1 billion and 1 in
1 million, and an increased risk of genetic effect to offspring over the next 5 generations of between 1 in 200 million and 1 in 200,000.
(See Sec. 10.3)
If an offsite dose equal to the staff's proposed (10 CFR Part 50, Appendix I) atmosphericannual limit (15 mrem/yr., total body) were received, that individual would incur an estimated increased risk of dying from cancer of 1 in 100,000 and an increased risk of agenetic effect to offspring over the next 5 generations of about 1 in 20,000.
An offsite
dose equal to the staff's proposed liquid annual limit (3 mrem/yr., total body) would
result in an individual incurring an estimated increased risk of dying from cancer of 1 in
500,000 and an increased risk of a genetic effect to offspring over the next 5 generationsof about 1 in 80,000.
(The average risk to members of the U.S. population of dying fromcancer is approximately 1 in 5 and the risk of genetic effects is about 1 in 17.)

4.

If the cleanup is conducted in accordance with the staff's proposed effluent criteria, thetotal cumulative dose received by the entire population within a 50-mile radius of TMI-2
due to both gaseous and liquid releases would range from 10 to 30 person-rem for theentire cleanup.
This is a small fraction (about .01%) of the background radiation dose
received by the population from causes other than releases from TMI (population backgroundradiation dose = 116 mrem/yr x 2.2 x 106 people= 25•5,000 person-rems).
(See Sec. 10.3)

TMI12 Job G Argonne

12-2

21201815.

Although the number of truck shipments necessary to carry solid radioactive wastes todisposal sites will be large (ranging from about 350 to 1000), the shipments will be madeover a long period and should cause little traffic congestion. Adherence to Federal
packaging and shipping regulations will result in small radiation doses to those along theshipping route. If all TMI wastes are shipped to the furthest potential storage site, theestimated 700,000 persons who reside along the 2750-mile route might receive a cumulative
population dose within the range of 20 to 50 person-rem.
(See Sec. 10.3.3)

6.

An individual onlooker who spends three minutes at an average distance of 3 ft from atruck loaded with radioactive waste might receive a dose of up to 1.3 mrem. This dosewould increase the individual risk of dying from cancer by 1 in 1 million. The increased
risk of genetic effects from the dose to offspring of the exposed individual is about 1 in
200,000.
(See Sec. 9.5.1.2)

7.

Radioactive fuel and high-specific-activity wastes from TMI-2 must be packaged andwill have to be stored at the site temporarily until a suitable disposal site is established elsewhere. No significant environmental effects are expected from these activities.

8.

From 1/2 to 1 acre of land at authorized disposal sites will be required for the low-level
wastes from TMI-2.
(See Sec. 9.5.1.3)

9.

The expected consequences of credible accidents are small (below the requirement of 10 CFRPart 20 for normal operation).
(See Sec. 10.4)

10.

Resources that will necessarily be committed to the cleanup are materials of construction
such as steel and cement, chemicals, organic resins, and other materials, none of which isin short supply in comparison to the annual U.S. production.
(See Sec. 10.7)

11.

Psychological distress caused by the accident and operations necessary to proceed with thecleanup has declined but there is a potential for temporary increases in distress as
various cleanup activities are undertaken. (See Sec. 10.6)

12.

Socioeconomic impacts include potential consumer avoidance of Chesapeake Bay seafood
products that the public may believe are contaminated if processed water is released to
the Susquehanna River, potential consumer avoidance of milk products sold directly to
consumers following airborne releases of radioactivity, and the potential adverse market
effect on residential property close to the transport route through Middletown,
Pennsylvania, during the period of waste shipments from TMI.
(See Sec. 10.6)

13.

The differential monetary costs among suitable cleanup methods are small compared to the
expected total costs of the entire cleanup and therefore do not constitute sufficientconcern to affect a decision as to which alternatives should be chosen to accomplish thecleanup activities. The overriding considerations should be ensuring the public's healthand safety and protection of the environment. (See Sec. 10.8)

12.2 Additional Conclusions
1.

Existing methods are adequate, or can be suitably modified, to perform virtually all of
the necessary operations without incurring significant environmental impacts.
Where
special tools or methods are found necessary for operations such as defueling, engineeringexpertise is available to cope with such requirements.
(See Sec. 6.4)

2.

An early decision to decommission TMI-2 would have very little effect on the choices ofalternatives for the cleanup tasks because most of the same tasks must be performed inorder to remove and dispose of the damaged fuel. (See Sec. 2.2)

3.

If the damaged fuel and radioactive wastes are not removed, the Island would, in effect,
become a permanent waste disposal site. The location, geology, and hydrology of Three
Mile Island are among the factors that do not meet current criteria for a'safe long-term
waste disposal facility. Removing the damaged fuel and radioactive waste to storage sitesthat do meet all of the relevant criteria is the only reliable means for eliminating therisk of widespread uncontrolled contamination of the environment by the accident wastes.
The staff has concluded that TMI should not become a permanent waste disposal site.
(See Sec. 2.1)

TMI12 Job G Argonne

12-3

2/20/814.

Procedures have not yet been established for processing and disposal of high-specificactivity waste.
Therefore, the staff regards the transfer of damaged fuel andhigh-specific-activity waste to facilities operated by the Department of Energy (where thenational expertise exists) to be the most appropriate course of action for process.ing andfinal disposal of this material.
(See Sec. 9.1.3.3)

5.

The contaminated accident-generated water in the reactor building basement (sump) and in
the reactor primary system cannot be left in its present condition and location if thecleanup effort is to proceed. Removal of this contaminated accident water will reduce theairborne and direct radiation levels in the building sufficiently to permit other cleanup
operations to be accomplished with greater safety. (See Sec. 2.1.2)

6.

Treatment of the contaminated accident water will transform the entrained radioactivity
from its current mobile state to a more manageable form by concentrating and immobilizing
the activity by an appropriate process.
The cleanup activity will eliminate the risks
associated with leaving the contaminated accident water radionuclide inventory in themobile unprocessed state.
(See Sec. 7.1)

7.

A decision on the ultimate disposal of the processed water can be deferred until after thewater has been processed. Then, the concentration of radionuclides remaining in the water
will be low enough for the water to be stored safely onsite until the disposal decision is
made.
Processing the water to immobilize most of the radionuclides and storage of theprocessed water will not foreclose any reasonable options for disposition of the water orconcentrated wastes. (See Sec 7.2.5.5)

8.

The alternatives adopted for the various cleanup tasks will be those that keep the occupational doses as low as reasonably achievable.
Delaying full cleanup will not appreciably
lower the radiation fields (as a result of radioactive decay) or occupational doses.
However, full and prompt cleanup would reduce the risks of uncontrolled radiation releases
and would keep the doses to workers involved in cleanup tasks and to the public at aminimum.
(See Sec. 10.2 & 10.3)

12.3 Benefits
1.

The major benefit of the cleanup will be the elimination of the continuing risk of potential uncontrolled releases of radioactivity to the environment from damaged fuel or fromthe radioactive materials which are distributed throughout the primary system, the reactor
containment building, and the auxiliary and fuel handling buildings. The radionuclides
are also in the contaminated accident water in the reactor building basement and in theradioactive waste in temporary storage on the Island. These sources are a hazard becauseof the potential for uncontrolled radiation exposure to workers on the Island and to the
local population.
Removal of this hazard will relieve anxiety in some members of thelocal population and those dependent on the Susquehanna River and the Chesapeake Bay for a
livelihood, for drinking water, or for recreation.
The only way to eliminate this continuing hazard and anxiety is to clean up the facilities and remove the radioactive waste
and damaged fuel to suitable storage sites. (See Chapters 2 & 4)

2.

An incidential benefit that would accrue from the cleanup (and the ongoing studies thatwill be needed for planning and implementation) is additional knowledge that would be
useful for reducing the risks and consequences of possible future accidents at nuclear
power plants.

12.4 Cost-Benefit Balance
The staff concludes that on balance the above benefits and other considerations relative
to the full decontamination, core removal, and disposal of the radioactive wastes from the
March 28, 1979 accident at TMI-2 greatly outweigh the environmental costs of the cleanup
activities.
Until TMI-2 is largely decontaminated, there is a small possibility (which
increases with time) of uncontrolled releases of radioactivity to the environment.
Decontamination of the plant and disposal of the wastes will eliminate this possibility
for potential harm to the public and workers at TMI, and will alleviate the attendantanxiety concerning radioactive releases from the plant.
The staff therefore concludes
that the full cleanup of the facility must proceed as expeditiously as is reasonably
feasible, consistent with ensuring public health and safety and protecting theenvironment.

13.

DISCUSSION OF COMMENTS ON THE DRAFT PROGRAMMATIC ENVIRONMENTAL IMPACT STATEMENT

Pursuant to 10 CFR Part 51, the Draft Programmatic Environmental Impact Statement (PEIS) related
to the decontamination and disposal of radioactive wastes as a result of the March 28, 1979,
accident at Three Mile Island Nuclear Station, Unit 2, was transmitted in Auqust 1980 with arequest for comments to the following federal, state and local government agencies:
Advisory Council on Historic Preservation
Council on Environmental Quality
Department of Agriculture
Department of Army,

Corps of Engineers

Department of Commerce
Department of Energy
Department of Health and Human Services
Department of Housing and Urban Development
Department of the Interior
Department of Transportation.
Environmental

Protection Agency

Delaware Department of Natural Resources and Environmental

Control

Maryland Department of Natural Resources
Maryland Department of State Planning
New Jersey Department of Energy
New Jersey Department of Environmental

Protection

New Jersey Department of Public Utilities
Pennsylvania Department of Environmental Resources
Pennsylvania State Clearinghouse
Pennsylvania Public Utility Commission
Virginia Council on the Environment
Virginia State Corporations Commission
Susquehanna River Basin Commission
Dauphin County,

Pennsylvania,

Londonderry Township,

Office of Emergency Preparedness

Pennsylvania,

Board of Supervisors

Tri-County Regional Planning Commission
Commissioners of Dauphin County, Lancaster County,

and York County Pennsylvania

Mayors of Middletown, Harrisburg, Goldsboro, Lancaster, York, Hershey, Lebanon, Hummelstown,
Highspire, Camp Hill, Paxtang, Lewisberry, York Haven, Elizabethtown, Columbia,
New Cumberland, Manchester, and Steelton, Pennsylvania
Commissioners of Baltimore and Hartford Counties,
Mayors of Baltimore and Havre de Grace, Maryland

13-1

Maryland13-2
In addition, a notice requesting comments from interested members of the public was published in
the Federal Register on August 15, 1980, and about 2000 copies of the PEIS were subsequently
mailed to individuals and organizations at their request.
The comments received are reproduced inAppendix A of this Final PEIS, which is reserved solely for them.
The staff's consideration of the comments received and its disposition of the issues involved arereflected in part by revisions in the pertinent sections of this PEIS and in part by the following
discussions.
Where data corrections suggested in the comments have been adopted by the staff,
these changes have usually been made without discussion here.
The organization of this section
corresponds generally to the ordering of the chapters of the statement; however, the discussionsof comments on similar topics are grouped together.
The comment letters to which these discussions apply are referenced by the numbers following the titleof each response; these numbers are
keyed to the letters in the Table of Contents in Appendix A.
13.1

GENERAL,

13.1.113.1.1.1

Purpose and Scope of PEIS
The Purpose of Cleanup Operations(125)

The purpose of cleanup at TMI-2 is to remove from TMI-2 the potential for uncontrolled releases ofradioactive material to the environment and the attendant risks to the health and safety of thosemembers of the public residing in nearby communities.
The cleanup would require the accomplishment of: building and equipment decontamination, reactor defueling and primary system decontamination, and processing, packaging, transportation and disposal of radioactive wastes.
13.1.1.2

Purpose of the PEIS (2,33,84, 86,

100,

107)

It is the purpose of the PEIS to present an overall study of the cleanup alternatives, including
decontamination and disposal options for TMI-2, to assist the Commission in its decision-making on
cleanup activities.
The PEIS indicates viable choices for each of the cleanup steps.
It is not,however, the purpose of the PEIS to predetermine which cleanup action should be chosen.
The scope and purpose of the PEIS were discussed with representatives of the President's Council
on Environmental Ouality prior to the drafting of the PEIS.
It was agreed that in keeping withthe purposes of NEPA to engage the public in the Commission's decision-making processes, and to
focus on environmental issues and alternatives before commitments are made, no cleanup preferences
should be predetermined.
The PEIS is written for the general public and government agencies with the purpose of involving
the public in the NRC's decision making process regarding the cleanup of TMI-2 in accordance with
the National Environmental Policy Act.
Copies of this PEIS have been provided to government agencies and to all individuals who requested copies as well as to those who commented on the draft
PEIS.
In addition, copies of the draft PEIS were distributed to all who requested a copy during
the numerous public meetings with interested individuals.
Copies of this PEIS are available uponrequest.
All comments and suggestions on the draft PEIS were evaluated by the NRC staff and allcomments are addressed in this PEIS.
In order to encourage public involvement, efforts have been made to write this PEIS in laymen's
terms such that the public reading the document can understand it.
To this end, a glossary has
been included at the beginning of this statement.
In addition, a document, "Answers to Frequently
Asked Questions About Cleanup Activities at Three Mile Island, Unit-2" (NUREG-0732) was preparedand made available to the public to explain the draft PEIS and to assist in preparing comments.
The NRC staff also welcomes any questions from individuals referring to any item in the PEIS for
explanation or clarification.
Inquiries should be referred to the individuals indicated in theforeword of this statement.

13-3
13.1.1.3

PEIS Is Not a "Blueprint" for Cleanup (20,

121,

130)

It is not a requirement nor is it prudent that the PEIS should 'serve as a blueprint" or be a"complete guide" to the cleanup activities.
Many details of the decontamination process are notknown because of limited access to the facility and its systems.
The PEIS is intended to provide
an overall evaluation of the environmental impacts that could result from the cleanup alternatives.
As stated in the Commission's Statement of Policy and Notice of Intent to Prepare a PEIS
(Appendix B), an overall study of the decontamination and disposal process will be "in keepingwith the purpose of the National Environmental Policy Act to engage the public in the Commissions'decision-making processes, and to focus on environmental issues and alternatives before commitments to specific cleanup choices are made".
13.1.1.4

Supplements to the PEIS (20,

59,

85,

100, 115)

To avoid as much as possible any segmentation of the evaluation of environmental impacts, the PEIS
presents an overall description of all of the main cleanup activities along with a discussion ofalternatives and the environmental impacts of these alternatives.
However, there are many uncertainties regarding the cleanup operations.
For example, the precise condition of the reactor core
will not be known until the reactor vessel has been opened and the core inspected.
If, when more
information becomes available, the effects of any proposed activities are found to be significantly beyond the scope of the assessments made in this statement, appropriate supplements will be
issued.
These supplements will not be isolated from the PEIS, but instead, they will take into
account any additional effects on alternatives previously discussed in the PEIS (e.g., change ofschedule) and the cumulative environmental impact of the cleanup operations (e.g., total occupational dose, total radioactivity releases).
Furthermore, the public will be informed of any supplements and they will be made available for public comment.
13.1.1.5

Ultimate Disposition of Unit-2 (20,

31,

60,

100,

124)

Whether or not Unit 2 will be decommissioned or restored to a condition acceptable for licensed
operation is not within the scope of the PEIS.
Discussions concerning the ultimate disposition of
Unit 2 are presently premature since the licensee has not made any proposals to reactivate ordecommission the reactor.
In either case, the facility will have to be cleaned-up.
However, todetermine if there would be any effects on the way the cleanup could be accomplished, the staff
did evaluate the effects on an early decision to decommission TMI-2.
This is discussed in Section 2.2 and detailed in Appendix U of this PEIS.
13.1.1.6

NRC's Role in Prevention of the TMI-2 Accident (67)Discussion of NRC's role in preventing the accident at TMI-2 or at any operating reactors isbeyond the scope of this Statement.
This topic is discussed in several of the various reports ofgroups which investigated the accident, in particular those of the President's Commission (Kemeny)and the NRC's Special Inquiry directed by Mitchell Rogovin.
13.1.1.7

Restart of Unit-1 (27,

31,

102,

124)

Issues concerning the restart of TMI-1 are addressed in a separate environmental review.
The
staff considers the restart of TMI-1, if authorized, to be wholly independent of the TMI-2 decontamination process.
13.1.1.8

Cleanup Alternatives Involving Other Agencies Included (124)

Cleanup alternatives that may involve another agency other than the NRC (e.g.,cluded in the discussion presented in this Statement.

EPA,

DOE)

are in-13-4
13.1.213.1.2.1

Content and Organization
Format of the Final PEIS (30,33, 85, 86,

100, 110, 32,

115, 116,

123)

While the draft statement had the benefit of enabling the reader to follow the chronologicalsequence of the cleanup activities, it also had the disadvantage of scattering significant information on a particular issue, such as water processing, throughout various parts of the draft. As
a result, Sections 5 through 8 are reorganized to group like activities together.
Further efforts have been made to write the PEIS in laymen's terms so that the document can beeasily understood. Toward this end, the glossary has been extended.
13.1.2.2

Inadequacy of Summaries (52)

While some additional technical information has been added to appropriate parts of the summary, in
general, the summary has been simplified so that it is more readily understandable by the public.
13.1.2.3

Numbers and Calculations in Draft PEIS are not "traceable" (32,35, 37, 79)

To the extent practicable, sources of numerical calculations have been identified and citedthroughout this Statement.
References which point to such items as international reports, reviews, etc. are generally available at specialist libraries maintained at nuclear energy facilities or at major universities.
Page numbers have also been added to many of the reference citations so that specific points ofinterest can be located easily.
"Handouts to presentation of GPU personnel at staff site visit, January 1980" is available in the
An update of the cost and schedule information, "TMI-2 Recovery ProNRC public document rooms.
gram Estimate, August 1, 1980" has been transmitted to the NRC by the licensee and is also available for public inspection at the NRC public document rooms.
13.1.2.4

Organization of Page Numbers (100)Individual secThe page numbers are organized by sections and are preceded by a section number.
tions are grouped according to major topics, making it easier to locate the topic of interest.
13.1.2.5

Basis for Sludge Volumes (32)The volume of sludge varies somewhat at each location when fluids are circulated. The movement ofIn areas such as the Auxiliary andfluid transports sludge to the filters where it is trapped.
Fuel Handling Building (AFHB) Sump, sludge levels may increase due to decontamination efforts thatflush solids to the sump, and they may decrease as water and some sludge are pumped out of thesump.
The AFHB Sump was tested by licensee personnel with long rods to determine the depth of the water
above the sludge. Using that water depth and the as-constructed dimensions of the Sump, thesludge volume was calculated to be 200 ft3. The measurement and calculations gave the sludge
volume at one point in time and it could increase or decrease somewhat due to operations. For
this reason sludge volume is usually an estimated value.
In those tanks that have not been tested by personnel, the sludge volumes were estimated conservatively using the standpipe height and the tank dimensions. This was done for the Miscellaneous
Experience has shown that tanks involved in numerous fluid
Waste Holdup Tank and the Sump Tank.
transfers have little or no sludge buildup. The estimates for the Reactor Coolant Bleed Tanks andothers used in the management of liquid wastes were based on this experience. The larger sourcesof sludge, such as the AFHB Sump, have a better defined sludge volume than most of the tanks.
Therefore, the overall uncertainty in total sludge volumes given is not large, probably at most
5%.

13-5
13.1.3

Licensee's Plans and NRC Actions

13.1.3.1

Licensee's Obligation to Cleanup TMI-2 (100)The Commission, in its Statement of Policy of September 26, 1980, has clearly stated the need forthe licensee to continue with its cleanup activities and maintenance of the facility to protect
the health and safety of the public. Although the NRC does not have the authority to determine
how the licensee should allocate its resources, the licensee has the obligation to complete thecleanup of TMI-2.
13.1.3.2

NRC's Responsibilities Regarding the Cleanup Activities (12,

38,

66,

67,

101)

By the Atomic Energy Act of 1954 as amended, the NRC is mandated to protect the health and safetyof the public from activities under its license.
The NRC is the independent agency established by
the Congress, staffed with knowledgeable experts to monitor and oversee the licensee's cleanup
activities to ensure the safety of the public.
All cleanup operations at TMI-2 have to receive
the authorization of the NRC.
In addition, through its TMI Program Office onsite, the NRC oversees all licensee cleanup activities to ensure that operations are implemented according to NRCapproved plans and in compliance with NRC-approved limits, procedures and Orders.
13.1.3.3

Authorization of Cleanup Activities (30,67,

70, 125)

When the NRC receives specific cleanup proposals from the licensee, the staff will review themvis-a-vis the environmental evaluations in the PEIS and with respect to safety requirements.
After these reviews, the NRC will either proceed to act upon the proposal or conduct additional
environmental and safety reviews, if necessary, before making a decision on approval.
13.1.3.4

PEIS Not to Predetermine Cleanup Actions (64,75,

80, 90,

100,

107,

124)

Viable alternatives for cleanup are presented and evaluated in this statement.
However, it is not
the purpose of this Statement to predetermine which of the viable alternatives should be accepted.
Only when the license submits an actual cleanup proposal will the NRC staff review the proposal
and act accordingly or make recommendations to the Commission, if appropriate.
Based on the staff
evaluation and recommendation, the comments from the public and other agencies on the PEIS, theCommission will make decisions on the authorization and acceptability of the proposed activity.
13.1.3.5

Need to Prepare PEIS (116)

The NRC recognizes that the cleanup must proceed expeditiously.
However, it is also importantthat potential environmental consequences of cleanup alternatives as well as public comments andsuggestions be considered prior to cleanup decisions being made.
13.1.3.6

Cleanup Decisions prior to the PEIS (16,20)

NRC decisions on previous cleanup operations were made following extensive reviews, including
evaluation of feasible alternatives.
For each of the major cleanup operations (e.g.
decontamination of the reactor-building atmosphere and processing of contaminated wastewater by the EPICOR-II
system), the NRC staff issued separate environmental assessments which were circulated for publiccomment.
In each case, all feasible alternatives were evaluated and comments and suggestions fromthe public and interested groups were considered before the decisions were made.
These actions
were consistent with the requirements of the National Environmental Policy Act; however, it was
not prudent to wait for the completion of the PEIS to carry out these cleanup operations becauseof safety and environmental considerations.
The construction of the Submerged Demineralizer System to clean up the contaminated water in thereactor building is proceeding at the risk of the licensee, without NRC authorization of its
usage.

13-6
13.1.3.7

Safety Evaluation of Cleanup Proposals (59,91)

When the licensee submits a proposed cleanup action, the staff will then review it and perform asafety evaluation in addition to evaluating environmental considerations.
For example, safety
precautions and backup systems will be reviewed for adequacy to ensure that the proposed action
will be safe to implement.
Appropriate technical specifications will be required for the implementation of the approved activities, along with approved procedure for implementation.
The NRC
staff also will maintain an inspection program to ensure that the approved procedures are followed
during cleanup operations.
13.1.3.8

Decision Procedure Prior to Authorization of Processed Water Disposition (32,42, 62, 77, 125, 130)

39,

40,

Pending completion of the final PEIS, the licensee has not made any proposal on the action to be
implemented for the dispostion of processed water.
In fact the licensee indicated that no proposal will be made before 1982.
When such a proposal is received by the NRC for approval, the NRC
staff will evaluate the proposed action against the alternatives already assessed in the PEIS andmake recommendations to the Commission.
The decision of the Commission will be based on theenvironmental assessment of the alternatives contained in the PETS, the staff's recommendation,
the comments and suggestions from the public and from other government officials, and the Commission's own assessment.
Potential environmental impacts, such as the quality of public water andpotential impacts on fish and seafood in the Susquehanna River and the Cheasapeake Bay, are certainly major considerations to be included in any decisions.
13.1.3.9

The Pace of Cleanup Activities (50,

70,

124)

While the need to expeditiously proceed with the cleanup is desirable, it is also important thatthe cleanup proceed in such a manner as to minimize any risks to the health of workers and nearby
residents i.e., only after careful assessments of the potential environmental impacts from thecleanup activities have been made.
13.1.3.10

Cleanup Delays (67)

It is unlikely that short delays would appreciably affect the environmental impact assessment.
Iflong delays that may significantly impact on the environment occur because of the financial condition of the licensee, the NRC will act to ensure that the health and safety of the public is
ensured and that the cleanup proceeds without further adverse impact.
If the delay is because of
new conditions which have an environmental impact significantly beyond the scope discussed in thisPEIS, supplement(s) to the PEIS will be prepared and issued for public review and comment.
13.1.3.11

Cleanup Activities Planned for 1981 (114,124)

Major progress on the Unit-2 cleanup is planned for 1981.
Some examples of cleanup *activities thelicensee plans for 1981 include:(1) The establishment of long-term decay-heat removal for maintaining safe reactor shut down; (2) continue cleanup of auxiliary buildinq individual areas andequipment; (3) continue reactor building entries for data acquisition; (4) cleanup of reactor
building sump water; (5) engineering and construction of support facilities required for cleanup
and waste dispostion.
13.1.3.12

NRC Policy Requires Cleanup to Proceed (51,52, 59)On September 29, 1980, the NRC issued a Statement of Policy with regard to the requirement of thelicensee to proceed with the cleanup.
It states that "The Commission will not excuse Met Ed fromcompliance with any order, regulation or other requirement imposed by this Commission for purposesof protecting public health and safety or the environment."
Should the licensee fail to meet its
obligation for financial reasons, the NRC has, under existing laws, the authority to step in andto act to ensure that the cleanup proceeds in a timely manner.

13-7

13.1.3.13

Uncertainties on Future Cleanup Proposals (100)To the extent known, the cleanup conditions have been evaluated. Where conditions are not yet
known, the environmental review was performed on the basis of the best information available with
due acknowledgment of uncertainties.
In some instances this necessitated reliance on "best-case"and "worst-case" evaluations.
Upon submission of specific proposals by the licensee, any necessary further environmental review undertaken by the staff will include all information thenavailable.
13.1.3.14

Systems Shared with Unit 1 (51,101)

There is no Unit 1 system presently planned to be shared with Unit 2 that may affect the safety of
Unit-1 operation or Unit-2 cleanup. This separation includes the facilities to store waste water.
13.1.3.15

Quality Assurance Regarding Cleanup Personnel (66)The licensee's quality assurance program provides that all cleanup workers should be properly
trained for the particular cleaning tasks assigned. The radiation protection plan also requires
workers entering the radiation area to have a Radiation Work Permit (RWP).
Prior to receiving the
RWP, the worker must be trained in the basics of radiation hazards and protection techniques. The
NRC also reviews all cleanup plans, equipment, methods and procedures prior to their implementation to ensure personnel safety and adequate protection of the environment.
13.1.3.16

Cost of Cleanup Not a Major Factor for NRC Decisions (32, 64, 66, 67)

The licensee's proposed actions will be independently evaluated by the NRC staff according to the
mandated function of the NRC to protect the health and safety of the public.. Estimated differential costs between the alternatives are provided in the PEIS. However, these cost estimates will
not be a major factor in the evaluation of the licensee's proposed cleanup activities.
13.1.413.1.4.1

Public Concerns and Participation
Public Participation prior to the Draft PEIS (11)The NRC staff, as directed by a Commission Order dated May 25, 1979, prepared an "Environmental
Assessment--Use of EPICOR-II at Three Mile Island, Unit 2" (NUREG-0591) dated August 14, 1979.
This document was circulated to appropriate government agencies and the general public for a 30day comment period. The staff provided a discussion of the public comments on the EPICOR-II
environmental assessment, as well as its recommendation to the Commission on October 4, 1979. The
Commission, after considering NUREG-0591, public comments and the staff's recommendation, authorized operation of EPICOR-II on October 16, 1980.
The public also took part in the decision-making process which led to the decontamination of the
TMI 2 reactor building atmosphere.
Approximately 800 comments were received from members of the
public, state and federal agencies in response to the circulation of the draft environmental
assessment (NUREG-0662) on the subject in March, 1980. At the request of the Governor of Pennsylvania and members of the public, the NRC extended the public comment period on draft NUREG-0662
for 30 days beyond the original deadline of April 17, 1980. Nearly 50 meetings were held withlocal officials, organizations and members of the public to discuss the issue. The NRC staff considered and responded to public comments on the draft in its final environmental assessment dated
May 1980, and the Commission was briefed in two public meetings.
The Commission's decision wasgiven in its June 12, 1980 order to make a controlled purge of the reactor building to theatmosphere.
Public scoping meetings on the proposed content of this PEIS were held in the Harrisburg, Middletown and Baltimore areas during January, February and March of 1980 prior to completion of thedraft PEIS. These scoping meetings are discussed in Section 1.4.

13-8
13.1.4.2

16,

Opportunities for Public to Comment on PEIS (11,

20,

124)

Members of the public have had the opportunity to participate in commenting on the PEIS in severalways:
An extended, 90-day comment period, which ended on November 20, 1980, was provided for interested
The NRC staff also
individuals and groups to submit their comments on the draft PEIS in writing.
met with numerous groups in Pennsylvania, in the area around Three Mile Island, and in Maryland to
Individuals were notified of these meetings, which were pubsolicit comments on the draft PEIS.
Any portion of the PEIS could also have been dislicized and were open to public participation.
cussed with staff members at the NRC-TMI Middletown office.
13.1.4.3

Public Hearing on PEIS (16,20,

60, 64,

100,

121)

The NRC staff does not believe that a formal, adjudicatory hearing is required in connection with
When thethe PEIS, nor has the Commission itself deemed it otherwise appropriate at this time.
licensee submits a specific proposal regarding a particular cleanup activity to the NRC whichAt
requires the issuance of a license amendment, appropriate notice will be given to the public.
that time, a hearing may be requested in accordance with Section 189 of the Atomic Energy Act andThis process is in accordance with the NRC's statutory
the Commission's implementing regulations.
obligations and is consistent with Section 1506.6(c) of The President's Council on Environmental
Quality (CEO) regulations.
13.1.4.4

Recirculation of PEIS Because of Cost Data (64,79,

85, 94,

100,

104,

130)

A number of commenters requested that the draft PEIS be supplemented and recirculated in light of
the absence of cost data and the development of newer or additional information for inclusion in
The guidelines of the Council on Environmental Quality (CEQ) state that a supplethe final PEIS.
(i) the agency makes substantial changes in the proposed
ment must be prepared by an agency if:
action that are relevant to environmental concerns; or (ii) there are significant new circumstances, or information relevant to environmental concerns and bearing on the proposed action or
Cost data does not, in the context of the PEIS, constitute an(40 CFR § 1502.9(1)).
its impacts.
"environmental concern" so as to give rise to the requirement to supplement and recirculate theCost data is relevant primarily for purposes of comparing the overall desirability ofdraft PEIS.
Economic considerations are nota chosen course of action with reasonable alternatives thereto.
of foremost importance in terms of protecting the public health and safety. The CEQ guidelines
further provide that the weighing of the merits and drawbacks of the various alternatives need not
be displayed in a monetary cost-benefit analysis and should not be when there are important quali(40 CFR § 1502.23.)
Similarly, no "new" information in the final PEIS istative considerations.
Rather, it con"significant" in terms of its impact on the evaluation of the proposed cleanup.
sists of revisions in the textual material as the result of comments received on the draft PEIS,
The National Environminor corrections and reflects the acquisition of more timely information.
mental Policy Act expressly contemplates that changes in a draft environmental statement will
This fact, coupled with the overriding
That is its purpose.
result from the comment process.
importance of proceeding with the cleanup, mitigates against the provision of an additional circuThe Staff believes that the draft PEIS has afforded the public with a fair opporlation process.
tunity to comment on aspects of the cleanup process to the extent presently known.
13.1.4.5

Influence of Comments on Decisions (66,

69)

Each
The comments from the public form a substantial factor in the NRC's decision-making process.
comment receives careful review by the NRC staff and all pertinent comments are considered in its
preparation of this Statement.
13.1.4.6

Independent Scientific Review (16,30,

66, 67, 99,

101)

The NRC is staffed with scientists, engineers and experts to evaluate all phases of cleanup operations.
Other government agencies, e.g., the EPA and the DOE, also participate in the review of13-9
specialized areas of the cleanup, e.g., monitoring the environment and disposal of nuclear fuelIn addition, the Commission has established andmaterials and hiqh-activity radioactive wastes.
funded an Advisory Panel to consult with, and provide advice to the Nuclear Regulatory Commission
on major activities required to accomplish the safe cleanup of the TMI-2 facility.
The NRC
Advisory Committee on Reactor Safeguards (ACRS) is available to provide the Panel technical assistance, as necessary, related to the cleanup.
As originally established, the Advisory Panel consists of twelve members:
three members of the public residing in the vicinity of TMI; threemembers from the independent scientific community; three members of the Pennsylvania state officials and three members from local government officials.
13.1.4.7

Future Public Participation (20,39, 40,

72,

102,

114)

If future proposed actions or data related to conditions in the reactor are within the scope of
potential environmental impact discussed in the PEIS, then public participation in the assessment
would have already taken place by means of the public meetings and comments received on the PEIS.
But, if future information, or the proposed cleanup actions, are significantly different fromthose alternatives addressed in the PEIS, appropriate supplements to the PEIS will be prepared andthe public will be provided opportunities to participate in these new environmental assessments.
Public meetings may be scheduled.
For licensee actions that would require an amendment to thelicense, an opportunity to request and participate in a formal adjudicatory hearing will be provided, consistent with the Atomic Energy Act of 1954 as amended, and the NRC's implementing
regulations.
13.1.4.8

Comments on the Final PEIS (41,92)

Public comments on the final PEIS would be considered by the NRC staff if the comments are on new
information in the final PEIS not previously presented in the draft PEIS and (2) such new information is significant in terms of environmental impact.
13.1.4.9

Future Information on Cleanup (88)

Relevant information on the cleanup will be provided to the public throughout the cleanup process
in several ways; 1)status reports will continue to be issued by the NRC; 2) safety evaluations by
the NRC staff on specific licensee cleanups proposals will be available to the public and 3) decisions by the Commission will also be published documents available to the public.
13.1.4.10

Local-Referendum on Waste Disposal Alternatives (67)

A local referendum on waste disposal alternatives has not been planned.
13.1.513.1.5.1

Decontamination Experience at Other Nuclear Facilities
Experience with Decommissioning Reactors (69,100)

A number of nuclear reactors have been decontaminated and decommissioned in the U. S., althoughnone were identical to TMI-2.
Of these, perhaps the most notable is the Experimental Breeder
Reactor-I (EBR-I) which operated at the Idaho National Engineering Laboratory from 1951 to 1964.
EBR-I was decommissioned and turned over to the U. S. Department of Interior.
(It was subsequently declared a National Historical Landmark and has been open to the public for tours since1975.)
Other examples of decommissioned reactors include reactors at Elk River, Montana, several
at the Hanford Reservation in the State of Washington, and the Enrico Fermi I Plant near Detroit,
Michigan.
The power reactor at Shippingport, Pennsylvania is currently undergoing
decommissioning.
Significant experience exists in surface decontamination and cleanup of contaminated water.
Most
proven decontamination methods are easily applied at TMI-2.
Even though there is more surface
area and contaminated water volume at TMI-2 than has been experienced at other facilities, thetechnology and methods are basically the same.

13-10Because of the uncertainties with respect to the condition of the damaged core, of the majorcleanup steps, only fuel removal may require techniques that are relatively new, although this is
doubtful.
Prior experiences with damaged core removal with the Canadian NRX and NRU reactors(Section 1.5.3) and the Enrico Fermi I reactor indicate that the operation can be successfullycompleted without major delays or the need to develop radical new technologies.
The fact that large scale decontamination is not factored into designs of commercial nuclear power
plants does not mean that the decontamination of TMI-2 would be an experiment requiring untested
technology.
For example, the amount of krypton-85 at TMI-2 is a major difference between TMI andother decontamination situations.
The krypton problem was solved by purging with no significant
environmental impact.
The amount of water and its contamination level in the containment sump at
TMI-2 is, relatively high.
However, decontamination can be accomplished without significant
departure from existing technology.
The decontamination will result, however, in a larger quantity of radioactive waste.
Methods for handling and disposing of this waste are also within
existing technology.
13.1.5.2

Information on Dresden-I Decontamination (81,100)

The NRC staff has reviewed the information in the Environmental Impact Statement prepared for theDresden decontamination.
Information germane to the TMI-2 decontamination has already been considered in this Statement.
13.1.5.3

Decontamination Experience at Chalk River (100,115)

In each of the two incidents of overexposure at Chalk River, the over-exposures (in excess of 5rem) were incurred by individuals who were not normally radiation workers. 1 , 2 At TMI-2, allcleanup workers entering radiation areas are required to receive basic training in radiation
protection.
In addition, the health physics program and the radiation protection plan procedure
at TMI-2 would provide added assurance for preventing incident of over-exposure for cleanup
workers.
With respect to the contaminated water in the NRX incident in 1952, the water was pumped
from the reactor basement for about 1-1/4 miles to trenches which had been cut into the sandy
soil.
These trenches were about 20 feet wide, 10 feet deep, and about 1000 feet long. 3 Follow-up
studies concerning the disposition of the radioactiity have been conducted by personnel of theEnvironmental Research Branch of the Chalk River installation. 3 - 6 Initially, most of theradionuclides were trapped near the surface of the sand.3,4 After several years, some radioactive
material migrated underground from the disposal area toward a swamp area.
These radionuclides had
been transported down to the water table by leaching and percolation and subsequently migrated
with the ground water. 6 Further investigations by the Environmental Research Branch have

1

A. J. Cipraini, "Health and Safety Activities in Reactor Operations and Chemical Processing
Plants," International Conference on Peaceful Uses of Atomic Energy, Vol. 13, 263-265, UN Publications, Geneva 1956.
2
E. 0. Hughs and J. W. Greenwood, "Contamination and Cleanup of NRU," Nucleonics, Vol. 18, No. 1,
76-80, January 1960.
C. A. Mawson, "Waste Disposal in the Ground," International Conference on Peaceful Uses of Atomic
Energy, Vol. 9, 676-678, 696, UN Publications, Geneva 1956.
4
G. C. Butler et al., "Health and Safety in Canadian Operations," 2nd UN Conference on PeacefulUses of Atomic Energy, Vol. 21, 19-24, UN Publications, Geneva 1958.
5P. J. Parsons, "Movement of Radioactive Waste Through Soil," AECL-1038, June 1960.
6p. J. Parsons, "Movement of Radioactive Waste Through Soil," Second Ground Disposal of Radio3

active Wastes Conference, TID-7628,

March 1962.

13-11attempted to characterize the migration of activity. 5 , 6 Strontium-90 was the most prominent
radionuclide identified and its transport in soil was such that it would require about 150 years
before releases to the environment, at which time the amount will have sufficiently decayed topresent a minimal risk to the public. 6 Review of the experience gained and estimates of the7
population exposure due to the waste disposal operations at Chalk River have been published. ,8With respect to the TMI-2 cleanup, the treatment of the contaminated water will differ significantly from that at Chalk River.
Regulations applicable to TMI-2 do not allow contaminated water
to be released.
The water will be extensively decontaminated prior to disposal from the site.
13.1.5.4

Limited Cleanup Experience (100)The apparently contradictory nature of some statements in the draft PEIS can be easily resolved.
The staff concludes in the PEIS that "the basic technologies for decontamination are well established and that available techniques can be modified to suit the conditions at TMI-2."
The key inthe above statement is the term "basic technologies" which refers to the general level of scientific knowledge about decontamination, and the industry's ability to apply that knowledge to thepractical purpose of decontaminating TMI-2.
In Section 1.5.2, a statement is made regarding limited experience with high level decontaminationof large areas of interior building surfaces.
This is not contradictory to the statement above
because the basic technology to accomplish high level decontamination of large areas is wellestablished, but has not been extensively required before TMI-2.
Limited experience also does notmean no experience.
Three examples are given in Section 1.5.2, with the Canadian NRX Reactor amajor one.
While TMI-2 is a larger decontamination effort, "available techniques can be modified
to suit conditions at TMI-2."
The statement on the limited experience with removal of damaged fuel and core components, and theneed for specific techniques for TMI-2 is also not contradictory.
The basic technology is wellestablished and we are confident that the TMI-2 core can be cleaned up by using that basic technology to develop TMI-2 specific techniques, methods, and tools.
Each decontamination effort isin a sense one-of-a-kind, having its own specific problems that must be overcome.
It is wellestablished basic technology that allows industry to accomplish these cleanup activities.
The statement regarding chemical decontamination experience to remove fuel failure debris(Sec. 1.5.4), is also not really contradictory.
Again, very limited does not mean non-existent,
and examples are given where chemical means have been employed.
In fact, Section 1.5.4 concludes
in the third paragraph that chemical "fuel removal technology is available for application at
nuclear plants." This again refers to the existence of the basic technology needed to accomplish
the decontamination.
13.1.6

Regulatory Requirements and Authorization

13.1.6.1

Current TMI-2 Licensing Conditions (20,38, 79,116)

Since the accident, the Unit-2 license has been amended by order of the Office of Nuclear Reactor
Regulations (NRR).
On July 20, 1979, NRR issued an order for Modification of License which suspended the licensee's authority to operate the facility except in its shutdown condition and
5p.

J. Parsons,
J. Parsons,"Movement of Radioactive Waste Through Soil," AECL-1038, June 1960.
"Movement of Radioactive Waste Through Soil," Second Ground Disposal of Radioactive Wastes Conference, TID-7628, March 1962.
71. L. Ophel, "Environmental Consequences of Radioactive Waste Disposal," Pollution and Our
Environment, Vol. 1, Background Paper A4-4-2, Canadian Council of Resource Ministers, 1966.
8
C. A. Mawson and A. E. Russel, "Canadian Experience with a National Waste Management Facility,"
Management of Low- and Intermediate-Level Radioactive Wastes, IAEA, Vienna, 1970.
6p.

13-12require that pending further amendment of the license (DPR-73), the licensee shall maintain thefacility in a shutdown condition in accordance with the approved operating and contingency procedures for the facility. On February 11, 1980, by Order of the Director, NRR, a new set of formal
license requirements was imposed to reflect the post-accident condition of the facility. These
requirements ensure the continued maintenance of the current safe, stable, long-term cooling condition of the facility. These requirements were set forth in a new set of proposed Technical
Specifications contained in an attachment to the Order.
Future license amendments on specific proposed cleanup activities will be based on the environmental assessments contained in the PEIS or any supplements thereto, together with an appropriate
safety evaluation of the proposed activity.
13.1.6.2

Technical Specification Unrelated to Cleanup (100)

The Technical Specifications issued on February 11, 1980 which are subject to the pending litigation address maintenance of the reactor in a safe shutdown condition.
These Technical Specifications are unrelated to the decontamination process and are not included in this statement.
However, please note that appropriate regulatory criteria and proposed Technical Specifications
pertaining to the decontamination process are included in Section 1.6 and Appendix K.
13.1.6.3

As Low As Reasonably Achievable (ALARA)

Principle and Cleanup Procedures

(52,

125)

In carrying out the cleanup operations, the licensee is required to comply with regulations set
forth in 10 CFR Parts 20 and 50 in maintaining radioactivity releases to the environment andworker exposure as low as reasonably achievable (ALARA).
Specific numerical limits to implement
the ALARA principle will be developed for the cleanup programs.
A more detailed discussion of the
proposed criteria for radiological effluents is in Section 1.6.3.2.
When cleanup activities arebeing authorized, the technical specifications and operating procedures will be reviewed to assurethat the ALARA principle will be met during implementation of cleanup operations.
13.1.6.4

Cleanup Criteria (13,50, 100,

124)

Discussions covering the criteria currently in effect at TMI-2 as well as proposals for future
criteria are presented in Section 1.6.
The interim criteria established for radiological effluents are based on data-gathering and maintenance operations.
These criteria were approved by the
NRC on April 7, 1980, and are presented in Section 1.6.1.6.
Some criteria affecting the TMI-2 are
expected to be developed over a period of time as a result of the issuance of this Statement.
These are discussed in Section 1.6.3.
These criteria consist of:(1) future radioactive waste
disposal standards (Section 1.6.3.1); (2) proposed criteria for radiological effluents from decontamination activities (Section 1.6.3.2); and (3) acceptable removable surface contamination levels(Section 1.6.3.3) for unrestricted access.
13.1.6.5

Amendments to Technical Specifications(100)

Proposed additions to Technical Specifications for the TMI-2 cleanup program are included inAppendix R. Any additions or modifications to the existing TMI-2 Licensing conditions (Technical
Specifications) implementing TMI-2 cleanup operation will be added to the TMI-2 operating license
by means of the license-amendment process.
This process includes NRC review and evaluation forsafety and environmental concerns, notification of the amendment and provides an opportunity forinterested persons to request a hearing.
13.1.6.6

Redundancy of Technical Specifications on Dose Estimates and Reporting (50)

Proposed Technical Specification R.1.3 has been revised to eliminate the redundancy in dose estimate and reporting requirements.
13.1.6.7

The Need to Perform Dose Estimates During Cleanup(50)

The bases for the proposed technical specifications are described in Appendix R. Dose calculations will provide NRC with necessary information as the cleanup progresses to assess the environmental impact due to cleanup activities, to maintain radiation doses to as low as reasonably13-13
achievable. As more experience is gained with each program, calculations for later programs canmore accurately be made than the dose projected in this Statement.
Dose estimates accompanying proposals of cleanup activities are necessary for therstaff to ascertain the potential environmental impact, both during normal operation and under unplanned conditions. Dose estimates are also required for the review of on-going and future cleanup programs.
The requirements for dose estimates are not different from current requirements for operating
reactors to assure compliance with the As low As Reasonably Achievable (ALARA) principle of Appendix I to 10 CFR 50. The staff is aware that the estimated doses may be very low for some of thealternatives. The estimates for a particular cleanup proposal would still be necessary to form
the basis of comparison with other alternatives and to ensure that the ALARA principle is met.
13.1.6.8

The Application of Appendix I Criteria (100)Appendix I is not developed on cost-benefit consideration based on electricity generation. Rather
the numerical values are based on the requirement of radioactive effluent treatment systems meeting the objective for maintaining radiation doses to the public at levels as low as reasonably
By the same analogy, the benefit gained from cleanup is not measured byachievable (ALARA).
generation of electricity but rather by the need to ensure protection of public health and safety
and the environment. The same ALARA principle should therefore be applicable to cleanup
operations.
13.1.6.9

Potential Conflict Between NRC Requirements and Other Constraints (80)

A list of constraints (settlements with local government agencies, discharge permits of theCommonwealth of Pennsylvania, etc.) are presented in Section 1.6.2. With respect to the potentialof future conflicts with requirements of other agencies, a Statement of Policy issued by theCommission on September 29, 1980, illustrates the general principle under which the NRC written injurisdiction . . . creates an irreconcilable conflict with NRC requirements which have been
imposed on Met Ed or which may be imposed in the future. We wish to state clearly, however, thatin the event of any. such conflict, NRC health, safety and environmental requirements must supersede state agency requirements that result in a lesser degree of protection to the public."
13.1.6.10

Emergency Preparedness Plan (20, 32, 51, 46, 99, 115)

The scope of the PEIS does not include discussion on emergency planning. However, the following
information is available. Nuclear powerplant licenses are required to submit their emergency
preparedness plans for NRC staff review and evaluation against criteria of NUREG-0654, "Criteria
for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants." On June 10, 1980, the licensee submitted to the NRC for review the
emergency preparedness plan for TMI Unit 1 ("GPU Corporation Emergency Plan for TMI Nuclear
Station, Unit 1"). A Safety Evaluation Report (SER) to address the licensee's proposed plan for
Unit 1 will be issued by the NRC staff prior to the conclusion of public hearings currently being
held on the proposal to restart Unit 1. The Emergency Preparedness Plan for Unit 2 is being prepared by the licensee and is expected to be submitted to the NRC in the near future. After
receiving the proposed plan for Unit 2, the NRC staff will review and evaluate the proposal anddiscuss its findings in SER.
13.1.6.11

Violations of the Clean Water Act and NEPA (20, 52, 67, 91, 114)

The Federal Water Pollution Control Act authorizes discharge of pollutants only in conformancewith its provisions and those of the National Pollutant Discharge Elimination System (NPDES) Permit issued by the U. S. Environmental Protection Agency or its delegate (i.e., the Commonwealth ofPennsylvania).
NPDES Permit No. PA0009920 (dated July 12, 1977) was issued to Metropolitan Edison
Company by the USEPA for Three Mile Island Nuclear Station. Special conditions numbers 7 and 8 inPart III of that NPDES Permit require radiological effluents to meet NRC requirements as follows:13-14"All limitations and monitoring requirements for liquid radioactive waste discharges shall be inaccordance with the Nuclear Regulatory Commission regulations as set forth in 10 CFR Part 20 and
10 CFR Part 50" [Special Condition No. 7]. "The conditions of this permit shall in no way super*sede the mandatory requirements for operation of Nuclear Power Plants imposed by the Nuclear
Regulatory Commission" [Special Condition No. 8].
If the licensee proposes to release the processed water into the Susquehanna River and if the NRC
authorizes such a proposal, neither the intent nor any of the provisions of the Act would beviolated, since the NPDES Permit would be followed.
The National Environmental Policy Act of 1969 (NEPA) (PL 91-190), established national policy forconsidering environmental consequences in decision making. NEPA also established procedural
requirements for the consideration of environmental matters in decision making. A major procedural requirement is the preparation of an environmental statement for Federal actions which havepotentially significant environmental impacts, with a consideration of alternatives to a proposed
action. This Statement was developed in conformance with NEPA, as stated on page 1-1: (the PEISis) ". . .in keeping with the purposes of the National Environmental Policy Act to engage thepublic in the Commission's decision-making processes, and to focus on environmental issues andalternatives before commitments to specific cleanup choices are made."
NEPA does not specifically prohibit discharging nor does it establish effluent limitations forradioactive wastes. Section 104 of NEPA states that "Nothing in Section 102 or 103 [procedural
sections] shall in any way affect the specific statutory obligations of any Federal agency (1) tocomply with criteria or standards of environmental -quality. . .or (3) to act, or refrain fromacting contingent upon the recommendations or certification of any Federal or State agency." With
respect to this Statement, NRC is: involving the public in its decision-making process;considering alternative courses of action and the environmental consequences of each; and, isrequiring any radioactive discharges under consideration or examination to meet all effluent
standards for protection of the public health and safety in accordance with the FWPCA and NPDES
permit.

13-15
13.2

MAJOR PROGRAMMATIC ALTERNATIVES

13.2.113.2.1.1

General Alternatives"Entombment" of Fuel and Radioactive Wastes in the Reactor Building (4,116, 121, 124, 125)

104,. 107,114,

As pointed out in Sections 2.1 and 2.2, entombment of fuel and radioactive wastes in the reactor
building would require changes in current national policies and federal regulations, for it would
convert the TMI-2 site to a high-level waste repository. TMI-2 does not qualify as a repository
under current policies and regulations.
The advantages of such entombment must be weighed against the disadvantage of creating a longterm hazard and source of public stress and anxiety that would continue for several hundredyears.
The staff regards this disadvantage to be so serious that it more than offsets any gain
sealing
up the reactor building with the fuel and contamination inside until the radioactivfrom
Decommissioning alternatives for TMI-2 are evaluated in Secity has decayed to "safe" levels.
tion 2.2 and discussed in detail in Appendix U.
13.2.1.2

A Wider Look at Clean-up Alternatives?

(64,

100)

The staff has made every effort to address in sufficient depth credible alternatives for those
Where there is significant
aspects of the decontamination which are reasonably well defined.
uncertainty, such as the condition of the core, the most probably or bounding situations were
A number of options are listed for each operation and attention is focused on theconsidered.
Reasons for deletion of certain alternatives from detailed consideration
most feasible options.
Because the time and risk associated with dedepth in the final PEIS.
ingreaterare discussed
veloping new equipment, methods has been given considerable weight in selecting procedures, theThis fact,
options chosen tend to use proven methods for which considerable experience exists.
coupled with the generally conservative approach taken in calculating doses to workers and the
public, leads the staff to believe that the depth of analysis provided is appropriate for this
Detailed event tree analysis of all the options would significantly increase the size
document.
of the PEIS and would detract from its major objective; the overall evaluation of the environmenIf, as the work progresses, it appears that the decontamination
tal impacts of decontamination.
operations actually used will differ significantly from those described in the PEIS, supplements
will be issued which evaluate those operations.
13.2.1.3

Reuse of Contaminated Equipment (55,100)

Equipment which is required
The impacts of refurbishing TMI-2 are beyond the scope of the PEIS.
during the cleanup would be decontaminated and restored to operation if necessary as part of thecleanup.
13.2.213.2.2.1
4

Decommissioning
Decommissioning Alternatives and Impacts (11, 16,
99, 100, 114, 115, 121, 122, 124, 125)

32,

61,

64,

66,

67,

69,

78,

84,

92,

The public has expressed strong interest in (1) the impacts of decommissioning, and (2) the effect that various decommissioning alternatives would have on the choice of cleanup alternatives.
In addition an assessment of the impacts of various disposition alternatives will be needed in
Therefore, an analysis of the impacts of an early decision to decommisthe foreseeable future.
sioning has been added to the PEIS (see Sec. 2.2 and Appendix U).
13.2.2.2

Information Source on Decommissioning (32)The staff drew upon a previous report (Technology, Safety, and Costs of Decommissioning a reference PWR, NUREG/CRO130) in its analysis of decommissioning the damaged reactor TMI-2.

13-1613.2.2.3

Duration of Cleanup (32,71)

Four basic factors will determine the length of time required to complete cleanup operations:the time required for individual operations; the need to carry out many of the operations in sequence rather than simultaneously; the time required to resolve differences of opinion on whatshould be done and how it should be done; and financial restrictions.
The first two provided thebasis for the initial estimate of 5 to 9 years.
The third and fourth factors may increase theduration beyond this estimate.
The fourth factor has been considered in the licensee's revised
estimate that the duration of the cleanup will be seven years or longer.
Estimates of the times required to carry out the individual cleanup operations are based on
analysis of different alternatives presented in detail in Sections 5 through 9, taking into consideration the licensee's estimates (Fig. 1.3.1).
The various tasks involved in.the cleanup operations must be carefully planned and executed to keep worker exposure to radiation as low asreasonably achievable (ALARA) and to ensure that the health and safety of the public are adequately protected.
This approach is basic for the entire cleanup, and increases the required
time over that which would be necessary for a less careful operation.
Considerable uncertainty exists in the estimates of the worker effort, occupational exposure andduration of the cleanup operations because there is very little
information available on thecondition of some parts of the buildings (e.g., the reactor building basement) and equipment(e.g., the fuel elements inside the reactor pressure vessel).
However, it is possible to obtainreasonable bounding estimates by using knowledge of the design and operation of the equipment,
well-known physical principles, the limited amount of data that is available, and experience
gained in previous situations (in which similar problems occurred, even though the number andscale of problems at TMI is unique (see Sec. 1.5).
An understanding of the important factors involved and sound engineering judgement are necesssary
for such estimates.
One would, for example, obtain misleading results if unique or anomalous
situations were used as a general basis for estimating task durations.
(The long delay in opening the malfunctioning entry door is an example of an inapplicable basis.)
The delay was primarily a consequence of a decision to delay further entry attempts until the NRC issued a decisionon purging the Kr-85 from the reactor building.
Many of the individual estimates of upper bounds are probably conservative.
For example, initial
estimates of the radiation levels inside the reactor building, which were based on measurements
utilizing penetrations through the reactor building exterior walls that allowed insertion of instruments at very few locations, turned out to be larger than the levels measured during the entries by a factor of 2 to 3.
The sequence of operations, and also the licensee's estimate of the duration of the operations,
are summarized in Figure 1.3.1.
As an example of the necessity for sequencing (and of an application of the ALARA principle), the primary coolant system cannot be cleaned up until the reactor
has been defueled.
Defueling will not start until decontamination of the reactor building has
progressed to the point where exposure of defueling workers to radiation has been substantiallyreduced by decontamination operations.
Decontamination cannot be started until plans based on
information gained during entries have been prepared, and entries were delayed until the Kr-85
was purged from the reactor building atmosphere.
The need to resolve conflicting views on how the operations should be carried out introduces delays of variable duration that are difficult to estimate. An example is the delay in the purging
of the Kr-85 from the reactor building atmosphere in order to provide time for consideration of
the objections of those opposed to a direct purge to the atmosphere and to assess alternatives(see Sec. 5.2).
There may be other delays in order to give full and fair consideration to different views on other issues, such as the means that should be used for disposing of the processed water.
Publication of this document should help minimize further delays of this kind.

13-17
There must be
Delays due to financial constraints are also difficult to predict at this time.
Restrictions on themoney available to pay workers and purchase clean-up equipment and supplies.
amount of money available and on how it could be spent have severely curtailed the rate ofSome of the money has and will come from insurance; the source ofcleanup since September 1980.
the balance of the money needed is not yet determined.
13.2.313.2.3.1

Disposal of Radioactive Waste
On Site Landfill(70)

No radioactive wastes have been or will be disposed of at the sanitary landfill at TMI.
13.2.3.2

TMI Not To Be Site for. Radioactive Waste Disposal (31,

70,

125)

It is not intended that TMI be a site for the permanent disposal of either spent nuclear fuel or
To use TMI as a permanent waste repository is neither compatible with curradioactive wastes.
rent national policies nor NRC regulatory guidelines for radioactive waste disposal.
13.2.3.3

No Regulation to Prohibit Temporary Waste Storage Onsite (114)There is no regulation that specifically prohibits the operation of a nuclear power plant with
However, there are rehigh activity radioactive wastes stored outside of the spent fuel pool.
quirements that the storage of these wastes has to be safe and radiation exposures due to thestorage be within regulatory limits and be maintained to as low as reasonably achievable levels.
13.2.3.4

Alternative to Dispose Processed Water by Controlled Release into River (12)The alternative to dispose processed water by controlled release into the Susquehanna River isSecone of several alternatives evaluated in the statement for the disposal of processed water.
tion 7.3 contains a detailed discussion on the potential environmental impacts of this alternative.
13.2.413.2.4.1

Current Status
Radioactivity Releases Since the Accident (60)Since the accident, the only significant release of radioactivity was the controlled purging of
Kr-85 from the reactor building. The total amount of Kr-85 released has been about 43,000 Ci.
The NRC has not authorized the release of liquid radioactive waste from the site.
13.2.4.2

Reactor Building Sump Water Level

(130)

Water
Between 650,000 and 700,000 gallons of water are currently in the reactor building sump.
The waterlevel in the building increases at a small rate, on the order of an inch per month.
However, should transfer of this waterlevel in the sump presents no immediate safety hazard.
become necessary, storage tanks onsite have sufficient capacity to accommodate this water.

13-18
13.3

TMI VICINITY AND DOWNSTREAM AREA

13.3.113.3.1.1

GeologyTMI Under Water (52)Comparison of "water's edge" elevation of 280 ft. MSL on p. 3-1 of the draft PEIS with the"average elevation" of the island of 277 ft. MSL on p. 3-5 led the commentator to the conclusion
A more careful reading indicates that the statement beginning withthat TMI is 3 ft. under water.
the last sentence on page 3-1 and continuing on page 3-5 reads "Bedrock beneath the site is essenThe sentence is referring totially flat, with an approximate average elevation of +277 ft. MSL."
Ground surface ranges from +280 at
the average surface elevation of bedrock, not ground surface.
the water's edge to more than +300 ft MSL in the north-central portion of the island, as stated in
the last sentence of the next to last paragraph on page 3-1.
13.3.1.2

References on Geology (51)

The NRC is guided in its evaluation of nuclear sites by Federal Regulation 10 CFR Part 100
Appendix A to that document, which defines the seismic and geologic criteria for nuclear plant
siting, and requires that in the absence of specifically identified geologic structures that havethe potential of generating earthquakes in the site region, the site analysis must include adesignation of tectonic province within which it must be assumed that the largest credible earthquake unique to each tectonic province can occur anywhere in that province, including its closest
approach to the site.
In defining tectonic provinces, the NRC staff utilizes many standard, wellaccepted references including the cited, "Structural Geology of North America" by A. J. Eardley.
"SeismotecOther references that are used routinely for this purpose by NRC Geologists include:
tonic Map of the Eastern United States" USGS MF 620 by J. B. Hadley and J. F. Devine, "The
Tectonics of North America-A Discussion to Accompany the Tectonic Map of North America" Prof.
Paper 628, USGS, by P. B. King; and "Tectonic Map of the U. S." USGS and AAPG, by G. V. Cohee andothers.
These references, although published several years ago, are still widely accepted as
accurately describing the general structural geology of the eastern United States for the purpose
Eardley was cited because his was the main descriptionof broad tectonic province definition.
used in the report.
The staff is not aware of any state, federal, and contracted services on TMI related studies in
the areas of geology and seismology other than the site investigations by consultants to theutilities.
In regard to using "original sources," the NRC geological-seismological staff has only the capability to review data from investigations conducted by applicants for nuclear power plant
licenses, information in the open literature, and when available, U. S. Geological Survey work.
Other than exerting some control over the investigative methods and analysis techniques used toassess a site and conducting brief geological reconnaissances to sites, NRC geoscientists do not
generate original site information.
The geologic and seismic information generated
Reference 11 was not used in writing Section 3.1.
Most of the data utiby the utility was taken from Reference 42 and the published literature.
lized from Reference 4 were results of site investigations such as topographic surveys, borings,
Except for changes in the topography during site
geophysical surveys, and field geologic mapping.
preparation and construction, this information is' still considered valid.

1"Final Environmental Impact Statement Related to Operation of Three Mile Island Nuclear Station
Units 1 and 2," US Atomic Energy Commission, Docket Nos. 50-289 and 50-320, 1972.
2"Final Safety Analysis Report, Three Mile Island Nuclear Station, Unit 2," Metropolitan
Edison Co., 1974.

13-19The staff concentrated on the upper strata at the site because the properties of that soil androck, and the orientations, attitudes and nature of bedding, faults, and joints strongly influence
the activity of groundwater.
Significant faults, joints and other geologic features that arepresent in rocks at 1000 feet extend up through the 100 million-year-old bedrock and are detectable in the upper rock strata which has been mapped or penetrated by core borings.
13.3.1.3

Stability of the Onsite Storage Facility (75)

The NRC is guided in its evaluation of nuclear sites by Appendix A to 10 Code of Federal Regulations, Part 100, "Seismic and Geologic Siting Criteria for Nuclear Power Plants."
As noted in the
PEIS, all category I structures and tanks are founded in bedrock, except the Category I storage
tanks, which are founded in compacted backfill.
The information used to develop this input was
based on information provided in the PSAR, FSAR and relevant amendments.
Faults in the site area
were also assessed and found to be within the meaning of Appendix A to 10 CFR 100.
Based on ourreview of the available information in the geoscience area, we continue to conclude that the site
is safe for a nuclear power plant; therefore any onsite storage facility founded in the same
foundation material should be stable.
13.3.213.3.2.1

Meteorology
Sources of Meteorological

Information (51)Reference 91 for Section 3.1.3.1 is intended to provide a definite source of information regarding
the TMI region's climate based on 40 years of meteorological data collection.
General climatic
information regarding the area could also be determined from any number of other climatology publications; however, the reference provides greater detail for the TMI region than any othergenerally available publication.
Onsite meteorological measurements are made continuously on a
tower at the north end of the island.
The measurement program is described in Section 3.1.3.2.
Upper air wind information is routinely available through the national weather service measurements program at Philadelphia, Pittsburgh, and Washington, D.C.
This information is available on
facsimile maps received by the NRC.
13.3.2.2"Appropriate" Meteorological

Conditions for Water Evaporation(99)

As described in Section 3.1.3.2, forced evaporation may produce or enhance fog in the TMI site
vicinity.
Evaporation of water is a function of the water temperature, the speed of the air flow
across the water surface and the amount of moisture in the air over the local area.
Since evaporation is most dependent on the above conditions, it is possible to limit the forced evaporation
process to times when the meteorological conditions are: 1) winds 5 mph or greater, regardless of
direction, 2) turbulent atmospheric conditions, 3) relatively "dry air" over the region, and 4) no
precipitation.
Although natural evaporation will take place from the water surface, using the"forcing" process under the four conditions given above should reduce the environmental impact of
this evaporation procedure.
13.3.2.3"Northwest Anomaly"

The statement on page 5-7, Sect. 5.1.5.2 of the draft PEIS, relative to the WNW sector at
0.37 mile from the plant was incorrect.
See Appendix W of the final PEIS, particularly Table W.2,
for directions and distances used for offsite close calculations.

1"Local Climatological Data, Annual Summary with Comparative Data, Harrisburg,of Commerce, Environmental Data Service, published annually.

Pa," US Department

13-2013.3.3

Hydrology and River Use

13.3.3.1 Susquehanna River Hydrology (21,79, 99,100, 102,130)

The minimum daily flow of record for the Susquehanna River at Harrisburg is 1600 cfs; the average
flow is 34,000 cfs.
Approximately 25% of the flow at the site passes the discharge point in the
middle channel between TMI and Shelley Island.
Below York Haven Dam, complete mixing is assumed
and full flow of the river is used in determining dilution factors.
Further discussion of themix.ing characteristics may be found in Section 3.4 of this Statement.
The nearest potable water supply is at Brunner Island,
proposed effluent release.
13.3.3.2

Hydrology Data Corrections(35,

approximately 5 miles downstream from the51)

Section 3.1.4.1 (Surface Water Hydrology) has been revised, where appropriate, to reflect thecorrections and comments provided by the U. S. Geological Survey.
The last two sentences on
page 3-6 of the draft PEIS have also been deleted due to the recent cancellation of the Stony
Creek Project.
13.3.3.3

Lowest River Flow (32,

66,

102)

The Susquehanna River has never been known to run dry. As indicated in Section 3.2.4.1, thelowest flow of record at Harrisburg was 1600 cfs on November 29, 1930, and was caused by an ice
jam.
Due to the nature of hydropower operation, the flow out of Conowingo Reservoir is essentially zeroat times when power is not being generated.
This assurance, however, has little
or no effect on
downstream dilution factors and net flushing effects of the reservoir.
The lack of outward flow
does not stop incoming flow.
With no outflow, the water can only rise upstream of the dam.
Thewater in the reservoir has essentially the same physical characteristics as the incoming water.
The water which is eventually released when the turbines are operating is the same water thatoriginally would have gone through the turbines but was delayed a day or so by not being released.
13.3.3.4

Flood Forecast

(51,

120)

The flood forecast or predicted rate of discharge for large floods will not change significantly
due to the cancellation of the Stony Creek Project.
The drainage area controlled by this projectis minor in relation to the total drainage area of the Susquehanna River.
Thus, the amount offlood control afforded by that facility is minor at TMI, and the flow rate of the river at TMI
will change insignificantly.
It is true that portions of the site and the South access bridge were flooded during the Tropical
Storm Agnes flood of 1972.
The reason that the site was flooded was that the dikes which now
protect the island were not completed at that time and water entered through the unfinished portions.
The north access bridge is.above the elevation of the design flood and would not be
flooded by a recurrence of Tropical Storm Agnes.
Section 10.5 discusses the designs of the facilities and dikes and provides pertinent elevations and flood levels.
13.3.3.5

Flood Protection (32,

67,

96,

120, 123)

Section 10.5 provides discussion of floods and flood probabilities, and of the flood protection
for the island.
The auxiliary and fuel handling building would not be affected by the probablemaximum flood, since the building is closed off from the entry of flood waters.
13.3.3.6

Use of Susquehanna River Water for Water Supply (73,100)

The statement on page 3-12 that "the river is not an attractive source of public water supply" hasbeen modified.
According to information provided to the NRC by the Baltimore Department of Public

13-21
Works, the city of Baltimore does not currently withdraw, and does not plan to withdraw, water
The last significant withfrom the Susquehanna River at the rate of 250 million gallons per day.
drawal of water from the Susquehanna River for the city of Baltimore occurred in 1968 and once
again in the early 1970s.
The city plans to withdraw water only during prolonged drought periods
when the levels in its water supply reservoirs drop very low.
13.3.3.7

River Flow Rates Considered for Dose Estimates

(55,

66)

The flow that was used in the draft PEIS for estimating dose to humans due to consumption of water
On the average, the flow is approxior fish downstream was the mean daily low flow of 1700 cfs.
mately twenty times greater than this so that it is very unlikely that controlled discharges would
be made at a still lower flow. To provide more realistic assessments in the final PEIS, a river
flow of 12,600 cfs has been adopted for most calculations; the basis for this number is given in
It should also be noted that the licensee is required to conform to the techniSection 3.1.4.1.
cal specifications.
If the release of processed water is permitted, the technical specifications
would require that the release be controlled to ensure that the public and the environment areprotected from harmful doses of radiation.
13.3.4Aquatic Ecology Studies (123)Section 3.1.5 of the PEIS is a summary description of the ecology of the Three Mile Island vicinIn-depth specific accounts of
ity and other areas potentially affected during cleanup operations.
the aquatic ecology and fisheries of the Susquehanna River and Chesapeake Bay are included inAppendix E was prepared from an extensive review of the literature, with
Appendix E of the PEIS.
the most pertinent sources cited and included in the references.
No experimental research per se on the ecosystems of the TMI vicinity has been done specificallyfor the PEIS.
However, studies of the aquatic ecology and fisheries of the Susquehanna River near
These summaries are mentioned in Section 11.10 of
TMI have been ongoing continuously since 1974.
Additionally, investigations of the consequences of the accident to aquatic biota,
the PEIS.
fisheries, plants and animals of the TMI area have been completed and are contained in two NRC
technical reports: NUREG-0596, "Non-radiological consequences to the aquatic biota and fisheriesof the Susquehanna River from the 1979 accident at Three Mile Island Nuclear Station"; and NUREG0738, "Investigations of reported plant and animal health effects in the Three Mile Island
Areas."

13-2213.4
13.4.1

REACTOR SAFETY PRIOR TO DEFUELING
Delays in the Decontamination Process (80)

The NRC staff recognizes the potential risks to public health and safety and to the environment
from delays in decontamination of TMI-2.
There are references to the effects of delay in decontamination actions in several places in the PEIS, e.g., in Sections 2.1.1.1 and 2.1.1.2, and in
Appendix C. Of primary concern is the potential degradation of instrumentation and systems required for the continued safe shutdown of the reactor.
The staff considers it imperative that thecleanup proceed in a timely manner, and has indicated this position to the licensee.
13.4.2

Maintenance of Fan Coolers (100,107)

The fan coolers in the TMI-2 reactor building have operated continuously for more than 21 months
without failure since the March 28, 1979 accident; eventual failure of the fan coolers can still
be expected if maintenance is not performed.
The controlled releases of krypton gas from thereactor building during 1980 eliminated the possibility of uncontrolled releases of the gas in theevent the fan coolers failed.
The urgency of maintaining them has therefore declined and is probably not necessary until some gross decontamination of these highly radioactive components has
been accomplished.
13.4.3

Mini Decay Heat Removal System (32,55)

The MDHRS is described in Section 4; it was installed as an additional method of removing decay
heat from the core.
However, the loss-to-ambient mode of cooling has been in effect since
January 5, 1981.
The MDHRS is connected to the reactor coolant system at two points by electric-motor operated
valves located in the basement.
For a time there was some concern that the rising water level inthe basement would cover the valve motors, incapacitate them, and prevent the valves from being
opened.
However, the valves have now been opened and the rising water level in the basement is nolonger a threat to operation of the MDHRS.
13.4.4

Boron Analysis (32)

The comment questioned the lack of sampling of boron for several weeks last summer and the implications to maintenance of the reactor in a safe condition.
The reactor is being kept in a
subcritical state by the boron that has been added to the reactor cooling system (RCS) water;
dilution of the boron could obviously lead to recriticality and must be avoided.
To ensure thatthe boron concentration is adequate at all times, steps have been taken to limit sources of water
to the RCS and minimize the need to make any increases and reductions in the amount of water inthe RCS.
Also, regular analysis of boron concentration is being performed, the frequency of whichis largely dependent on the frequency of and extent to which changes in RCS water inventory are
occurring.
Prior to the inspection of the RCS and defueling, there is no need to move large volumes of waterin and out of the RCS.
Analysis of boron concentration is then necessary less frequently and has
been scheduled to be done on a weekly basis. During the summer of 1980, for a two-week period,
analyses were not done by the laboratory usually doing them.
However, since no large amounts ofwater were to be introduced into the RCS during this period, rapid dilution of boron due to accidental introduction of underborated water into the RCS was not possible.
It should be noted thata neutron monitor was and is available that can detect any increases in neutron "levels" in thecore that would have occurred if boron was being slowly diluted and the reactor had begun to respond to the reduced neutron absorption.
The staff considers that the boron analyses were adequate and that at no time was the health and safety of the public compromised.
During reactor
inspection and defueling boron analyses will be done more frequently, and an additional neutron
monitor will be put into use.

13-2313.4.5

Recriticality (91,99, 100, 115,

125)

A full description of the criticality control status of the core is given in Section 4.4.2.1.
While the introduction of under borated water into the core could lead to recriticality, such an
event is very unlikely and is guarded against by both engineering design and administrative
controls.
To ensure subcriticality, the boron concentration in the RCS water must be maintained at more than3000 ppm.
This minimum limit was established on the basis of calculations that assumed grossredistribution of the reactor core, so that even in the unlikely event that the core melted intosuch a geometry, the 3000 ppm minimum limit would be sufficient to maintain subcriticality. Boronconcentration in the RCS water is being maintained at more than 3500 ppm by the licensee.
Thislimit was established on the basis of calculations that assumed the most reactive configuration of
the reactor core. Much higher concentrations of boron in the water would have to occur before
precipitation would present a possible problem of constricting water flow through the core; suchhigh concentrations are improbable.
Prior to the beginning of reactor cooling system (RCS) inspection and defueling, there will bevery little
need for large quantities of water to be removed from and added to the RCS.
Rapidboron dilution of the RCS liquid is thus all but impossible at this stage and boron sampling andanalysis is necessary only on a weekly basis.
When defueling is in progress, it will be necessary to carry out some decontamination of RCS water
on a regular basis because manipulation of the damaged fuel can be expected to cause release of
more fission products from the damaged fuel pins into the RCS water.
RCS water would then have tobe processed and larger quantities of water would need to be removed from and added to the RCS.
In order to ensure that under borated water is not added to the system, many procedural controls
have been and would be instituted.
These controls are described in Section 4.4.2.1 and they include requirements that major sources of water to the RCS be maintained at sufficiently high boron
concentrations.
For example, addition of water can be required to be performed in batches rather
than continuously, and analysis of boron concentration can be made more frequently, perhaps asoften as several times a day.
13.4.6

Dry Core (55)The statement on page 2-8 to the effect that criticality could result from a dry core was incorrect.
A dry core would not become critical, although it could become physically unstable.
This
has been clarified in the PEIS.
13.4.7

Fission Products from Recriticality(55)

There was a discrepancy between the summary and Chapter 4 regarding the quantity of fission products that could be released due to recriticality.
The text has been clarified to indicate that a
recriticality could release fission products into the reactor building but that such a releasewould be small and not likely to escape from the reactor building.
13.4.8

Reactor Building Sump Water Level (32,99, 130)

Between 650,000 and 700,000 gallons of water are currently in the reactor building sump.
Waterlevel in the building increases on the order of a fraction of an inch per day.
The water level inthe sump currently presents no safety hazard.
However, should transfer become necessary, thecontingency plan calls for pumping to storage tanks which have sufficient capacity to accommodate
this water.

13-24
13.5

DECONTAMINATION OF BUILDINGS AND EQUIPMENT

13.5.1PEIS Unnecessarily Restrictive (50,75)

The staff agrees that the draft PEIS seemed too restrictive with respect to certain operations orsequences in which the decontamination operations might be performed and with respect to the useof robots for radiation measurements. The PEIS is not intended to provide detailed cleanup plansor to recommend or preclude specific methods or procedures. The responsibility for developing
It is the responsibility of theand implementing specific cleanup plans rests with the licensee.
NRC to review the licensee's plans and monitor the implementation of those which are approved.
The analyses in the PEIS could, of course, lead to the rejection of some procedures.
The licensee's detailed plans must be prepared and then implemented in a step-by-step manner.
The first steps have already been taken. They started with initial decontamination of the auxiliary and fuel handling building and processing of the water from these buildings in order toremove dissolved and suspended radionuclides. Subsequent steps included construction of facilities for packaging and temporary storage of radioactive waste, and preparation for cleanup of thereactor building and reactor by purging the Kr-85 from the reactor building atmosphere and surveying the radiation levels and damage inside the building. The foregoing activities proceeded,with appropriate review and monitoring by the NRC, prior to publication of the PEIS. The next
steps will be to prepare and then implement detailed plans for decontaminating the reactor building, defueling, cleaning up the primary system, and disposing of the wastes, including the processed water. These plans will be prepared by the licensee and then reviewed and monitored by
the NRC, taking into consideration the guidelines provided by the PEIS.
13.5.2

Radioactivity within the Buildings (50,107)

Early.estimates of the amount of radioactivity in the reactor building and the auxiliary building
were based on remote sensors and indirect measurements. However, several entries have been madeinto the reactor building and the actual measurements show that most of the radiation levels areless than the original estimates. Also, a large portion of the cleanup in the auxiliary buildingis already completed.
13.5.3

Personnel Training (79)

Although the draft PEIS makes no specific mention of training, the decontamination effort will
require extensive training of all types of personnel involved in the cleanup operations. Maintenance personnel, equipment operators, engineers, reactor operators, and health physics will be
included. Formal lecture and discussion programs dealing with the radiation hazards and practical approaches to keep exposures as low as reasonably achievable (ALARA) have already been
instituted by the licensee. These programs are designed to prepare cleanup personnel for thesituations they will encounter. Demonstrations, trial dry runs, and work on mock-ups of reactor
systems will also be needed in order to maximize the efficiency of the workers and minimize thetime spent in the radiation fields. Planning sessions will be utilized to develop the best way
to perform each task and to employ experience gained in previous tasks to best advantage
In addition, certain decontamination tasks will require specialists with proven experience toperform them. Outside organizations can and have been contracted by the licensee for some ofthese specialized jobs.
13.5.4

Condition of Equipment in the AFHB (100)The condition of the equipment in the Auxiliary and Fuel Handling Building has been questioned,
particularly whether this equipment will meet safety requirements during the cleanup.
The Auxiliary and Fuel Handling Buildings and their equipment were designed to meet all applicable safety regulations. They were also designed to prevent radioactive material released to the13-25environment from exceeding the limits established by applicable regulations.
The equipment damage is primarily due to flooding and atmospheric conditions which caused widespread contamination.
After the buildings and equipment have been decontaminated, and repaired, they should be
The NRC wil inspect them in order to make sure that this is
as safe as before the accident.
actually the case.
13.5.5

Concrete Surface Removal(71)

Most of the radioactive contamination deposited on building surfaces can be removed by rinsing,washing with a high-pressure water-jet, or scrubbing with detergent.
However, in some cases theradioactive contamination has diffused into the surface and a surface layer of concrete must be
removed in order to get rid of it.
A prior situation involving an accident at a reactor in which concrete removal was necessary fordecontamination occurred at Chalk River in Canada. 1 , 2 The Chalk River experience involved removal of concrete "down to the aggregate." Then concrete was added to a depth of about 6 inches
greater than the original wall to act as additional shielding.
At TMI-2, most of the concrete surfaces were coated with materials to limit the absorption ofliquid-containing contaminants.
However, concrete removal may be required in two areas.
Below
the 305-ft level in the reactor building, the walls were coated only up to about 5-1/2 feet above
the floor.
The sump water has been in contact with about 2-1/2 feet of noncoated, but very
dense, concrete.
As a result, surface removal is anticipated down to a depth of about 0.5 inch.
This estimate is based on the experience in the AFHB, where sump water was in contact with noncoated concrete for about a year.
The maximum surface removal required was 3/16 inch.
The second area that may require some concrete removal is the elevator shaft in the reactor
building. The concrete blocks of the elevator shaft were not coated and some surface removal isanticipated.
The blocks will be cored prior to decontamination to determine how far the contamination may have penetrated.
13.5.6

Alternatives for Removing the Sump Water from the Reactor Building (50,75)

The list of sump-water removal alternatives has been expanded in the final PEIS (seeSec. 5.2.2.1).
The removal procedures involve a number of tasks that must be coordinated withother cleanup activities. There are many alternatives for each task and coordination possibility. The total number of overall alternatives is very large, making it impossible to carrythrough a detailed analysis and selection of an optimum plan prior to the start of cleanup.
This
would be true even if the occupational dose could be accurately predicted for each task.
Many uncertainties (including radiation sources remaining in the basement after the sump water
has been pumped out or the amount of water that will be needed for decontaminating the reactor
building) cannot be resolved until cleanup operations have progressed to where detailed, accurate
data become available.
At the present time, we can only make approximate bounding estimates ofoccupational exposures and environmental impacts.
These are based on reasonable assumptions fora sampling of alternatives.
An overall approach can be sketched out; but detailed plans anddecisions regarding individual task alternatives must be made on a task-by-task basis.

'Harry2Collins,

John Logie; "Canadian Decontamination Experience:"
Chalk River Nuclear Lab.
J. W. Logie; "Three Vessel Replacements at Chalk River;" Atomic Energy of Canada Limited; Decontamination and Decommissioning of Nuclear Facilities, Marilyn M. Osterhout, ed.; 1980.

13-2613.5.7Installation of Large Scale Cleanup Systems (20)

Decontamination and cleanup systems in operating nuclear power plants are designed and licensed
for safe operation of the facility and reduction of radioactivity releases during normal operation and anticipated transients. These systems are not intended for the large-scale cleanup
operation necessary following a major accident. Installation of large-scale cleanup systems,
like those considered for the post accident cleanup at TMI-2, will not improve the safety ofnormal operation nor will they significantly reduce normal releases from operating reactors.
13.5.8

Effects of Prior Spill (32)The spill "prior to the accident" mentioned on page 5-25 of the draft PEIS consisted of minor
amounts of resin from flush bags used during the servicing of the condensate polishing unit andflushing of the demineralized water system. Reportedly, less than 5 gallons of resin were involved. The incident is recorded in the TIMI-operation records. Flushing the small amount of
spilled resin into the sump is an effective "housekeeping" measure and is representative of thetype of approved housekeeping that takes place in nuclear power plants. The extent to which thesmall amounts of resin in the auxiliary and fuel handling building complicated the accident
cleanup is negligible. The changed description of the spill in the final PEIS will betterreflect its true importance.

13-27
13.6

REACTOR DEFUELING AND PRIMARY SYSTEM CLEANUP

13.6.1
13.6.1.1

Uncertainties

in Plant Cond.itions

Best-case/Worst-case Analysis (75)The staff has stated in the draft PEIS (Sec. 7.1) that because of unknown conditions within thereactor pressure vessel and because of the difficulty in performing an accurate, detailed analysis of vessel conditions, the condition of these components will not be known until inspections
and examinations can be performed.
For these reasons, the staff has presented an evaluation in
the document of the impacts associated with best-case/worst-case conditions.
Actual conditions
should be within those bounding conditions presented.
The best-case/worst-case approach to TMI-2 cleanup activities was applied to provide limiting
conditions.
The purpose of the analysis is not to specify cleanup activities for the licensee
but to provide an environmental review of the several bounding alternatives.
The weakness ofsuch an approach is that it does not help in predicting realistic budgets.
Several studies have been performed concerning the accident which, based on conditions experienced during the accident, describe possible reactor building conditions.
"Analysis of Three
Mile Island-Unit 2," a NSAC report, and "Three Mile Island, a Report to the Commissioners and to
the Public," by the Special Inquiry Group of the U. S. Nuclear. Regulatory Commission, were used
extensively to establish the most-probable-and worst-case conditions that were used in the draft
PEIS.
The data in these reports indicate that there will not be any overwhelming problems forhealth and safety arising from the uncertain conditions in the reactor pressure vessel or in thecore.
13.6.1.2

Need for New Technological

Developments (4)The staff does not agree with the assertion that "safe operations will require new technologicaldevelopments that are beyond the present state of the art." As stated in the PEIS (Chapter 12
and elsewhere), the staff concluded from its analyses of the cleanup that existing methods areadequate, or can be suitably modified, to perform all of the necessary operations with only minimal releases of radioactivity.
13.6.1.3

Effects of Unknown Reactor Conditions on River Water (13)Concerning the effect of unknown reactor core conditions on effluents that may be released to theriver, it must be stressed that release of liquid effluents (after dilution) is only one of thedisposal alternatives considered in this document.
The accident water in the AFHB and RB and theexisting primary coolant have been analyzed for radionuclide content (Sect. 7.1);
these liquids
represent the greatest part of the radioactivity to be processed.
The condition of the core will
not affect these values.
The core condition will affect the RCS decontamination and flushing liquids.
Here the uncertainty is much greater, but under the worst-case conditions the total radioactivity sources during the RCS decontamination and defueling will be a small fraction of the sump inventory.
Further, a significant part of the activity released may be in the form of suspended particulates
that will be removed in the first filter stage of any processing alternative and disposed of as
solid waste.
Tritium, however, cannot be so removed.
Tritium is formed almost entirely in the primary coolant
water during reactor operation; very little tritium is formed in the fuel at the present time.
Thus, the condition of the core will not significantly affect the tritium content of the water tobe processed or the tritium concentration in the effluents.

13-2813.6.1.4

How Much Kr-85?

Where Is It? (50, 55,

72)

The total inventory of Kr-85 just before the accident was 100,000 Ci, as determined essentiallyfrom reactor burnup information. Its location was all in the uranium oxide pellets inside thezirconium cladding of the fuel rods. After the accident, about 45,000 Ci was trapped in the reactor building and subsequently was vented to the atmosphere. The remaining 55,000 Ci can bedivided into (a) a portion still in the fuel, (b) a solute in the primary cooling water, and (c)releases to the atmosphere during the accident. The solubility of krypton in water, especiallywhen the water is at high temperature, is small enough so that even 700,000 gallons would not
hold a significant inventory. The nuclide Cs-137 is known to have about 60% of its inventory
outside the fuel elements. This suggests that some 40,000 Ci of Kr-85 remains in the fuel, but
it might be less.
13.6.1.5

1.5 Curie of Kr-85 per Fuel Element? (55)

The 45,000 Ci of Kr-85 in the fuel is divided into about 30,000 fuel pins, or 1.5 Ci per pin, not
per fuel element (there are about 200 fuel pins per element).
This is an error in the draft
PEIS, page 8-13, and has been corrected.
13.6.1.6

Leach of Radionuclides from Spent Fuel Rods (73)

It has been suggested that all radionuclides in the spent fuel may leach out through the destroyed cladding into the primary coolant water. Radiochemical analysis of the primary coolant
water has been conducted weekly since shortly after the accident and no further radionuclides
have been found in the primary system water. The staff cannot be absolutely sure that significant leaching will not occur; therefore, the alternatives in the final PEIS include cases where
there is little or no leaching, and cases where significant leaching occurs.
13.6.213.6.2.1

Reactor Vessel
Reactor Vessel Inspection (75)

Reactor inspection consists mainly of examination of the reactor core and internals to assess thedamage and plan for cleanup. Recent entries into the reactor building were conducted mainly forradiation surveys and assessments of contamination. No significant reactor inspection work has
yet been performed.
13.6.2.2

Reactor Vessel Integrity (50, 100)

The exact structural integrity of the reactor vessel is unknown at this time. We do know positively, however, that the reactor vessel integrity is completely satisfactory for current andcontemplated future cleanup operations. The facts are: after the core severely dried out during
the accident, pressure in the pressure vessel was repeatedly cycled from approximately 1200 to1100 psi. During these pressure transients, there was no indication of loss of integrity. The
reactor pressure vessel is now operating at less than 100 psi and will continue to be maintainedat this pressure or lower until after removal of the fuel. Because the reactor vessel is essentially new (very little neutron exposure), nulductility problems do not now exist nor will theyin the foreseeable future.
The condition of the reactor vessel is pertinent to cleanup activities only to the point that thecore maintains its integrity to holdcoolant water. No cleanup activities are envisioned thatwould require that the core be subjected to conditions beyond those experienced so far. Structural conditions would require evaluation prior to recommissioning, but this is not within thescope of the PEIS.
13.6.2.3

Radiation Levels during Defueling (50)The licensee stated that the staff was unrealistic in saying the residual reactivity of the reactor building (after decontamination of the reactor sump water and hot spot shielding) would make

13-29no contribution to the radiation levels around the defueling area. Based upon radiation levels
measured during the entry on July 23, 1980, the text has been modified, in Appendix I of the
PEIS, to reflect a value of 5 mR/hr, which is the criteria of the initial building decontamination prior to defueling and is felt to be readily achievable.
13.6.2.4

Unsuitable Vessel Diagram (50)

Figure 8.1.1 is not precisely applicable to TMI-2. However,The caption has been changed.
current generation LWR(S).
13.6.2.5

it is a typical representation ofContamination of Reactor Vessel Surfaces (32)

A statement in Section 5.1.4.1 that the distribution of different nuclides in the surface contamination would be similar to the distribution in the reactor building sump water was criticized. Although the surface contamination came from the water now in the sump, we agree thatdifferent rates of precipitation and settling for different elements would change the relative
concentrations of different elements, and therefore nuclides. What is important is that measurement must be made during cleanup to guide the detailed choices required during that step-by-stepprocedure.
13.6.313.6.3.1

Working Time and Productivity Estimates
Criteria for Estimation of Working Time (50)The licensee suggests that the best-case working time estimate described in Section 8.1.3 wasconservative when compared to "normal" conditions, and that some operations could be carried out
with less working time than estimated by the staff.
Several studies have been performed concerning the accident which, based on conditions experienced during the accident, discuss possible reactor building conditions. "Analysis of Three Mile
Island-Unit 2 Accident," a NSAC report and "Three Mile Island, A Report to the Commissioners andto the Public," by the Special Inquiry Group of the U. S. Nuclear Regulatory Commission, were extensively used to establish the best- and worst-case conditions which have been applied in the
PEIS. Because of the data in those reports, the staff's judgement was that even in the bestcase, "normal" conditions as known in the past will not exist in the activities involved in thereactor defueling.
13.6.3.2

Time Estimate for Core Removal (50)

It was suggested that the staff's estimate of 10 months for head and plenum removal and defuelingis too conservative, even for worst-case conditions. The NRC staff believes that worst-case conditions are rather probable, so that the hope that this part of the cleanup can indeed be done inless than 12 months is unrealistic.
13.6.3.3

Worker Productivity (50,75)

The "productivity factor" is used to take into account the decrease in efficiency of personnel
working in contaminated areas. This factor (expressed in percent) gives the reduction in theamount of work accomplished per unit time due to encumbrances from protective clothing, facemasks, respirators, etc. Factors in the range of 40 to 67% were used for different tasks andsituations. These factors apply only to the time spent working in contaminated areas; they do
not include time spent in putting on and removing protective clothing, planning, training, receiving instructions, etc. The fraction of time that decontamination personnel spend working incontaminated areas ranges from 30 to 60%; the staff assumed 50% as an average figure for decontamination work. The remaining time would be spent in the preparatory activities noted above,13-30and may be regarded as part of the overhead.
(The remainder of the overhead would be for support
workers with assignments that do not require entry into contaminated areas.) The numbers used
for productivity factors and "preparatory overhead" are based on experience in the industry forsimilar decontamination and cleanup activities. They represent the best estimates that can beprovided at the present time of worker efficiency in environments and under conditions that, formany tasks, have yet to be determined.
13.6.3.4

Workers Volunteer for Cleanup Tasks (46)Workers for the cleanup are volunteers in that they are informed of the natue of the cleanuptasks and have chosen to undertake the assignments.
These cleanup workers are provided with appropriate training including information on the potential health effects of radiation exposure.
They are notbeing paid extra for particular cleanup functions; i.e., workers are paid on thesame scale for the same work functions at TMI, whether the work is related to cleanup or not.
13.6.3.5

Extrapolation of Data from Work Already Done (32)

As indicated in Section 5, information from the-work effort already expended in decontaminating
the AFHB has been utilized in making estimates for the cleanup work yet to be done on building
and equipment surfaces and processing contaminated water. This experience has been applied in
the staff's estimates regarding decontamination work in the reactor building, and indicates thatboth time and person-rem can be saved through the use of careful planning. Unfortunately, theremoval of the reactor vessel head and subsequent activities including removal of the fuel anddebris involve different operations than simple building decontamination. Even so, it is obviousthat some experience from the building decontamination efforts will be beneficial to the defueling operations.
13.6.4

Need for Special Tools (20,125)

In Section 8.2.2.2 of the draft PEIS, the staff mentioned the clamshell as an idea for fuel removal under worst-case conditions. Although this approach is not very probable, it is a possible
alternative. Regardless of the defueling means employed, the possibility of releasing krypton
gas exists.
Section 8.2.3.3 provides a list of special tools and equipment that may be needed.
Specific tool
requirements are not entirely identifiable at this time due to lack of detailed knowledge of core
conditions. However, engineering capability exists to develop whatever tools may not be
available.
13.6.5.'Corridor Concept" (100)The discussion of this concept in Section 2.1.2.2 of the draft PEIS indicates the need for a considerable amount of space for laydown areas and the movement of fuel handling machines, etc.
This is the reason why the staff does not regard the corridor concept as workable.

13-31
13.7

WATER PROCESSING AND PROCESSED WATER DISPOSAL

13.7.1
13.7.1.1

Water Processing
Airborne Releases of Radioactive Material
116, 120)

from Water Processing (32,

55,

70,

73,

100,

In order to estimate airborne releases of radioactive material from water processing, a distinction between the behavior of tritium as a fission product as opposed to most others present in
the various waters to be processed should be noted.
Tritium is an isosope of hydrogen and it
combines with oxygen and hydrogen to form a molecule (HTO) with the same chemical properties aswater.
When water is evaporated, all water molecules, including those with tritium (HTO), exist
in a vapor state and would pass through a filter system uncollected.
Other fission products (andactinides, such as plutonium or uranium) exist in the process water as dissolved constituents.
When the water is evaporated, the dissolved constituents stay behind in the remaining liquid.
Evaporation is thus an effective method for separating bulk water from a solution containing
dissolved radioactive constituents, but is totally ineffective for separating tritium.
The action of evaporation also creates turbulence at the surface which results in the formationof small droplets of water that are physically entrained in the rising vapor.
The vapor contains
only tritium, no dissolved radioactive materials; the entrained droplets contain both tritium anddissolved radioactive materials.
In a well-designed evaporator, the fraction of the total material that exists in the form of entrained liquid droplets seldom exceeds 1 x 10-5 of that in theoriginal liquid.
This then forms the basis for estimating how much dissolved radioactivity appears in the vapor stream.
Following condensation, most of the entrained radioactivity appears
About 1 x 10-3 ofin the condensate with a small quantity remaining with the uncondensed vapors.
Thus, less than 1 x 10-8
the radioactivity in the condensate remains in the uncondensed vapors.
of the initial radioactive material except tritium remains in uncondensed vapor.
Other processing operations are generally less energetic and less turbulent than evaporation.
In
ion exchange operations, process liquids are pumped in closed pipes through an ion exchange column.
The fluids are relatively free of substantial turbulence.
However, even in these operations, a fraction of the process liquids may form small droplets carrying with it the dissolved
radioactive material and that are physically entrained into surrounding air spaces, and subsequently transported by the process off-gas system to the air cleaning system.
As some of theliquid droplets evaporates en-route, the dissolved radioactivity is left as a solid aerosol in
the air stream.
Entrainment for this kind of operation seldom exceeds a fraction of 1 x 10-6 ofthe process liquids.
For the purposes of estimating airborne releases for the processing of TMI liquids, it was assumed that a fraction of I x 10-4 of the processed liquid becomes airborne and enters the process
off-gas streams.
This value is conservative when compared with the information presented above
and is compatible with data obtained from the processing of nuclear fuels.
The air cleaningsystem for the removal of particulate matter consists of the HEPA filter. Regulatory Guide 1.140
gives guidance to the testing procedure and requirements for installed HEPA filters. The required efficiency for each stage tested is 99.95%, which corresponds to a penetration fraction of
5 x 10-4; the penetration through two stages of HEPA filters (properly tested and qualified)would be (5 x 10-4)2 or 2.5 x 10-7.
In this statement, a penetration fraction of 1 x 10-3 wasused for calculating effluents and releases for a filter train of two HEPA filters in series, avalue that is very conservative to achievable values.
The failure of a HEPA filter was considered as a credible accident in this statement.
The releases of the facility are constantly monitored with appropriate instrumentation.
A failure of a
HEPA filter would be evidenced by an increase in releases by the monitoring system and would
alert operating personnel to take corrective action such as securing the ventilation system. The
failure would likely occur only in one of the HEPA filters in the filter train, thus, containment
integrity for the system would be maintained but at a reduced level.
The detection of a significant.failure within 15 minutes of its occurrence is considered reasonable.
No failures of the
HEPA filters occured during EPICOR-II processing of contaminated waters.

13-3213.7.1.2

Processing of Accident Water in AFHB (11,

72,

74)

Soon after the accident, it was recognized that the containment water in the AFHB would pose an
imminent problem and had to be expeditiously cleaned up.
The approval of the EPICOR-II System,
however, received extensive review by the NRC staff and the public commented on the environmental
assessment issued for the evaluation.
The licensee is evaluating methods for solidification of
the EPICOR-II resins. The solidified wastes will probably be shipped to a DOE waste disposal
facility. The DOE is currently studying suitable sites for disposal.
13.7.1.3

Processing of Accident Water from Reactor Building (27,28)

The reactor building sump (RBS) water differs from the auxiliary and fuel handling building(AFHB) water chiefly in having about double the volume and ten times the total initial radioactivity. Because of the higher radioactivity concentration, the RBS water may be processed by asystem using the Submerged Demineralized System (SDS), if approved.
The activity level of theprocessed RBS water will be of the same order of magnitude as that of the processed AFHB waterand the same disposal options will be considered (as discussed in Sec. 7).
13.7.1.4

Processing System Optimization (75,81)

Optimization of processing systems for TMI liquids is not within the scope of this PEIS; rather,
the discussion here considers existing state-of-the-art water-processing methods and evaluates
the degrees of decontamination that are achievable through their use either alone or in combinations.
Design optimization would occur at a later date when specific decontamination processes
are chosen for application at TMI.
This statement does, however, utilize revised decontamination
factors for SDS and evaporators that are based on more recent laboratory or operational
experience.
13.7.1.5

Chemical Interference with Water Processing (21,

79,

99)

Regarding materials which may interfere with water processing, oil and grease will be removed as
far as possible prior to processing, and residual amounts will be trapped in prefilters.
In
considering processing alternatives, solutions containing detergents or chelating agents, andparticularly those containing more agressive chemical reagents, are regarded as separate processing streams, distinct from the major water sources (accident water, primary coolant, RCS
decontamination and flushing liquids).
The decontamination of these streams is discussed indetail in Appendix G.
13.7.1.6

State-of-the-Art Decontamination and Volume Reduction Factors (50)

Although state-of-the-art radwaste processing equipment can provide improved decontamination
factors (DF) and volume reduction factors, those discussed in Section 7.1 and Appendix G wereselected based on operating experience at nuclear facilities. These factors also provide a measure of conservatism to account for abnormal operating conditions.
The selection of these parameters provide realistic bounding estimates for processing liquid wastes.
13.7.1.7

Use of Other Ion Exchange Materials(26)

The ion-exchange systems selected for discussion in this statement have been proven, throughlaboratory and field applications, to be applicable for the specific radiochemical conditions andvolumes of liquid to be processed.
Should other proven systems become available that provide
significant operational, impact, and/or cost improvements over existing systems, they will beconsidered if they are proposed by the licensee.
13.7.1.8

Processing of Decontamination Liquids (50)

Ion exchange techniques, among other options, are being considered for processing of decontamination liquids.
In general, however, decontamination liquids have detergents and other cleaning13-33chemicals that are not compatible with ion-exchange systems.
These processing alternatives are
Corresponding flow diagrams are provided in Appendix G.
discussed in Section 7.1 of the PEIS.
13.7.1.9

Consistency of Inventory of Processed Water (79,85)

The volumes and concentrations of radionuclides in the processed water are dependent on the volThese parameters
ume and concentration of the input streams and the water processing system.
They areestablish the bounds for the processed water inventory for the best and worst cases.
discussed in Section 7.1.
13.7.1.10

EPICOR-II System Uses (50,

55,

115)

Variations of the EPICOR-II and the Submerged Demineralizer System (SDS) have been considered forboth Reactor Building (RB) sump and primary system liquid processing, e.g., the EPICOR-II System
is considered for polishing RB sump liquids in Section 7.1.
13.7.1.11

EPICOR-I System Uses (55)

As for whether the EPICOR-I system would require an environmental assessment if used for Reactor
Coolant System (RCS) water cleanup, it should be noted that the EPICOR-I system is used for lowlevel contaminated water (<1 pCi/mL) and is unsuitable for processing primary system (RCS) water.
There are no plans to use EPICOR-I for TMI-2 decontamination and the system was not evaluated in
the PEIS.
13.7.1.12

Proprietary Information on EPICOR-II System (84)

A "Confidential Disclosure Agreement" exists between the NRC and EPICOR, Inc. regarding thedisclosures of proprietary information to enable the NRC staff to review the safety of the
EPICOR-II System.
13.7.1.13

Processing of Dissolved Gases in Liquids (55)Process equipment is designed with vent lines which are directed to the plant ventilation sysThese ventilation systems are monitored prior to discharge to assure releases meet release
tems.
limits.
In the evaluation of liquid processing alternatives, considerations were given to the
However, it is expected that the quantities of dissolved gases
effluents of dissolved gases.
which could be released during liquid processing would not be significant.
13.7.1.14

Reactor Building Sump Liquid Processing Systems (50,81)

Current design considerations and decontamination factor (DF) for the zeolite/resin systems to be
Possibleused for processing Reactor Building Sump liquids are discussed in Section 7.1.
modifications to process parameters needed to achieve the design DF's are also taken into
account.
13.7.1.15

Reactor Coolant System Liquid Waste Estimates (32)The value of 200,000 gallons maximum of liquid waste generated in decontamination of the reactor
coolant system was in fact obtained from experience at the Dresden decontamination operations.
13.7.1.16

EPICOR-1I

and Boron Removal (50)The interaction between the boron content of primary water and the action on ion-exchange system,
It has been assumed that using EPICOR-II to remove
EPICOR-II, has been taken into consideration.
radioactive ions from the water would also remove the boron needed to keep the core subcritical.
In fact, EPICOR-II can, and has, been used in such a way that leaves the boron in solution, andthis consideration is factored into the discussion of this statement.

13-3413.7.1.17

Usage of RCS Makeup Purification System (50)

The RCS makeup and purification system could be used if required.
However, other factors have tobe evaluated as cleanup progresses, for example the condition of the system integrity such that
it would not become a source of reactor coolant leakage.
The staff does expect that this system
will be cleaned up and that the basic piping, valves, pumps, and other components can be used tohandle the RCS water before its usage can be considered.
13.7.213.7.2.1

Reuse and Disposal of Processed Water
Disposal Alternatives for Processed Water (13,84, 85, 91, 100, 112, 126, 130)

27,

28,

31,

55,

66, 67,

72,

75,

76,

79,

The alternatives for disposal of processed water is discussed in detail in Section 7.2.
Section 7.2 provides detailed assessments on the alternative for the disposal of the processedwater, including facilities, effluents and environmental impacts under normal operation and under
accident conditions.
While the actual method of disposal has not been selected, the proposed
method of processed water disposal to be submitted by the licensee will be evaluated based on thealternatives discussed.
13.7.2.2

Radioactivity Levels in Water for Processed Water Release Alternative (12)At present, the NRC has not authorized the release of processed accident-2 contaminated waterfrom TMI.
The PEIS does contain an environmental assessment of the alternative for the controlled release of processed accident-2 contaminated water into the Susquehanna River.
If the
NRC decides to authorize the release, the NRC will act to ensure that the licensee will implement, according to the conditions (technical specifications) and criteria determined by the NRC,
to ensure the health and safety of the public including people who may drink the water and consume the seafood.
It should be noted that the PEIS concludes that the concentration of radioactivity in the water at the nearest drinking water intake downstream of TMI will be below the EPA
drinking water standards (which must be met by the operators of the drinking water distribution
systems) and safe for drinking during any controlled release of the processed water from TMI.
The PEIS also concludes that the absorbed radiation dose in fish and seafood in the river andCheasapeake Bay area would be an insignificant fraction of the normally occurring natural background radiation dose and should cause no detectable biological effect.
Consequently, consumption of fish and seafood should cause no harmful health effect because of radioactivity
contamination.
13.7.2.3

Water Discharged and EPA and State Requirements

(20,

67,

114)

The present Clean Water Act does not prohibit discharge of low level radioactive waste into navigable rivers.
The prohibition refers to the discharge of radiological warfare agents or highlevel radioactive waste. 1 Low level waste discharge limits are either promulgated by an individual state (agreement states) or are governed by NRC regulations (nonagreement states).
Effluent
limits from commercial reactors, however, are under the jurisdiction of the NRC.
The staff will evaluate the impacts of any discharge, if authorized, at the level permitted by
EPA or the state, as appropriate.
Actual compliance with the Clean Water Act and its implementing regulations is subject to the jurisdiction of EPA and/or an authorized permitting state.
13.7.2.4

Long-term On-site Storage Would Require Special License(52)

Unless the solidified processed water was determined to be a "nonradioactive" waste, licensing of
TMI as a disposal site would be required for long-term storage of wastes.
A disposal license is
not required for waste storage.

13-3513.7.2.5

Solidification and Storage of Processed Water (50,55, 72,76, 84, 114)

It is not intended that the TMI site to become a de facto long-term depository for radioactive
waste.
Intentional temporary storage of processed water and some solid wastes is inevitable,
prior to approval of methods and sites for their disposal.
Solidification of the processed water would not reduce the volume of the water or the mass andvolume of the resultant bulk. Prior processing of water to reduce the radioactivity level would
still leave tritium content undiminished.
Although concrete made with decontaminated water would
present an insignificant radiation hazard, there is currently no definition of a lower limit for
the radioactive concentration of waste regarded as radioactive and subject to regulation. Technically, therefore, storage of this concrete would make the island a low-level waste disposal
site.
Vitrification, which is essentially the encapsulation in a glass media, is not feasible for thewater per se. However, vitrification of the resins and ziolites is a possible alternative discussed in the Sections 8.1.2.2 and 8.1.2.3.
13.7.2.6

Direct Solidification of Water in Reactor Building Water (50,52, 72)

Alternatives considered for cleanup of reactor building and primary system accident water are
discussed in Section 7.1. One of the alternatives considered is direct immobilization of unprocessed accident water. This scenario is summarized in Section 7.1. For a number of reasons,
including implementation time and occupational exposures direct solidification of unprocessed
accident water is considered impractical.
13.7.2.7

Methods for Concentrating Tritium (72)There is no large-scale tritium-concentration technique for the amount present in the more than
one million gallons of wastewater at TMI-2.
13.7.2.8

Reuse of Processed Accident Water (102,Processed water availability, reuse applications,
Appendix F.
13.7.2.9

126)
and limitations are identified and discussed in

Dilution of Processed Water During Disposal (58)

The feasibility of using low tritium concentration water to dilute the processed water such thatthe tritium concentration of the diluted process water would be below that of the river water is
not likely since the tritium levels in the river upstream of the plant are significantly different than those of other nearby water sources. For example, the concentration upstream in theriver is 1.8 x 10-7 VCi/mL and even if the low tritium source had 90% less tritium (1.8 x
10-8 tiCi/mL), it would take about 13 x 106 acre-feet of water to dilute 1000 Ci of tritium toriver concentration.

'Section 301 subparagraph F, "The Clean Water Act showing changes made by the 1977 Amendments,"
Serial No. 95-12, U. S. Government Printing Office, Washington, 1977.

13-36
13.8

SOLID WASTE MANAGEMENT

13.8.113.8.1.1

NRC Regulations and Criteria
Definitions of Low-Level and High-Level Wastes (64,70)

High-level radioactive wastes have been defined in 10 CFR 60 as "those aqueous wastes resulting
from the operation of the first cycle solvent extraction system, or equivalent, and the concentrated-wastes from subsequent extraction cycles, or equivalent, in a facility for reprocessing
Irradiated nuclear fuel is also considered to be high-level waste.
irradiated reactor fuels."
Based onLow-level wastes are considered to be all other wastes not defined as high-level wastes.
these definitions, EPICOR-II wastes are considered to be low-level waste even though they will
require shielding for handling and shipment.
Within the category of low-level wastes, various disposal approaches may be warranted; e.g. someof the high-specific-activity wastes resulting from TMI cleanup will not be acceptable for routine
For those wastes which are unacceptable for routine shallow land burial, ashallow land burial.
case-by-case evaluation will be performed to determine which offsite treatment and/or disposal
methods will be required.
13.8.1.2

Waste Management Regulations (32)

Regulations regarding the disposal of low-level radioactive wastes are currently in 10 CFR 20.
However, 10 CFR 20 provides only broad requirements for waste disposal and does not contain specific requirements for such wastes as those which may be generated during the TMI cleanup.
The NRC is currently preparing specific regulations for high-level and low-level waste management
The procedural part of 10 CFR 60, the high-level waste
in 10 CFR 60 and 10 CFR 61, respectively.
The
regulation, was published in the Federal Reaister as a proposed rule on December 6, 1979.
technical part of 10 CFR 60 was published as an Advanced Notice of Proposed Rulemaking in theThe
A proposed rule is scheduled to be published in late 1980.
Federal Register on May 13, 1980.
high-level waste management regulation is scheduled to be published as a final rule in end of
1981.
The low-level waste management regulation, 10 CFR 61, is available for public comment as apreliminary draft. A notice of availability of 10 CFR 61 was published in the Federal Register on
February 28, 1980.
Disposal of wastes generated at TMI will be carried out consistent with these regulations even if
final rulemaking proceedings may not have been completed.
13.8.1.3

Standards for Waste Disposal

(50,

100)

The final waste forms and contents of the wastes generated to date (including EPICOR-II wastes)
However, the NRC has stated that wastes which
are not yet completely established by the licensee.
have the same characteristics as those routinely generated from nuclear power plants would be
In addition, for those wastes
acceptable for disposal at commercial shallow land burial sites.
which are clearly unique to TMI (e g., first stage EPICOR-II wastes), NRC has provided qualitative
The NRC staff has indicated that high-specific-activity EPICOR-II ion-exchange
disposal criteria.
wastes could be disposed of at a commercial land burial site at an arid location if these wastes
are solidified, or placed in a high integrity container, and special disposal procedures are inHowever, the most practical alternative iscluded to minimize exposure to inadvertent intruders.
to transfer these wastes to an existing federal government facility for future processing and
In addition, NRC staff has also indicated that any first stage zeolite wastes
eventual disposal.
from the proposed Submerged Demineralizer System (SDS) having Cs-137 activities on the order of1000 Ci/ft 3 from processing reactor building sump liquids are more like high-level wastes thanmaterial that is normally disposed of at commercial land burial sites and the assumption thatthese wastes can be disposed of in a commercial shallow land burial system is not valid.

13-37For unique wastes which have not been encountered in the commercial nuclear industry and are yet
to be identified, specific disposal criteria cannot be developed since the criteria would dependon the characteristics of radionuclides present in the wastes, e.g., the specific activity, physical and chemical form of the waste product as well as the available disposal options.
NRC will
assign a high priority to making decisions regarding the management of wastes following the receipt of data which provide the characteristics of the proposed wastes and the proposed disposal
or storage options.
13.8.1.4

Case-by-Case Evaluation of Unique, Non-Routinely Generated Wastes (22,32,

50, 53, 76,100)
Some of the waste forms which will be generated during the TMI-2 cleanup will have different characteristics than wastes routinely generated and disposed of at commercial shallow land burial
facilities.
These special wastes include first stage EPICOR-II resins which have specific activities of approximately 40 Ci/ft 3 . Because of their characteristics, these materials cannot be
disposed of using routine methods at low-level disposal facilities.
The most practicable alternative is to transfer these wastes to an existing federal government facility for future processing
and eventual disposal.
Other less desirable alternatives include processing on-site and/or
packaging in high-integrity containers for special disposal at commercial LLW burial sites.
Evaluations of waste characteristics will be performed by the NRC after the licensee identifies
that a special waste form will be generated but prior to generation of the special form. Organic
ion-exchange material will be limited to 10 Ci/ft 3 . The evaluations will include considerationsof the commercial and non-commercial disposal options available for the wastes, transportation
modes and on- site and off-site storage provisions.
Confirmatory testing by NRC consultants will
be performed where necessary to verify licensee-submitted data.
Physical, chemical, and radiologic conditions in wastes can be simulated by using nonradioactive
materials, tracer isotopes, gamma irradiators and research reactors.
These simulated conditions
can be used to obtain realistic data on the characteristics of wastes to be generated at TMI.
The performance objectives for waste disposal evaluations will be (1) doses to inadvertent disposal site intruders will not exceed 500 mrem/yr and (2) doses from groundwater pathways will not
exceed 25 mrem/yr at the disposal site boundary.
For transportation, the criteria are specified
in the DOT regulations 49 CFR 171 to 179 and the NRC regulations 10 CFR 71.
For storage of wastes
personnel exposures will be as low as reasonably achievable (ALARA) and migration pathways into
the environment will be minimized under both normal and accident conditions.
13.8.1.5

Criteria for Waste Packaging (75)

The choice of packaging for the wastes generated at TMI is dependent on the physical and chemical
characteristics, the specific radioactivities and radiation levels of the wastes.
NRC regulations
govern the choice of packaging and extent of shielding specified during the packaging, handling,storage, and transportation operations.
DOT regulations govern shipping mode and route selection.
The regulatory constraints and feasible alternatives for these waste management operations aredescribed in Sections 9.1 and 9.2.
13.8.1.6

Acceptability for Disposal of TMI Wastes (114)

Low-activity wastes having characteristics similar to those being routinely disposed of by commercial shallow land burial are being disposed of at the Hanford, WA, commercial disposal site.
Another commercial disposal site which would be acceptable for disposing of low-activity wastes
exists in Beatty, NV.
EPICOR-II first-stage wastes are unlike routinely-generated wastes due tohigh Cs-137 and Sr-90 concentrations, but if they are solidified or placed in a high integrity
container, they could be disposed of at an arid site provided emplacement is such that future13-38
However, the most practical alternaaccess to the wastes by inadvertent reclaimers is minimized.
tive is to transfer these wastes to an existing federal government facility for future processing
High-activity wastes such as the first-stage zeolite liners in theand eventual disposal.
These high-activity wastes
licensee-proposed SDS will be unacceptable for near surface disposal.
will require on-site or off-site storage until suitable disposal facilities (such as a high-level
Spent fuel wastes will require disposal in a
waste repository) are constructed and licensed.
A high-level waste repository is scheduled to be available for dishigh-level waste repository.
The DOE has the legislative jurisdiction for siting, constructing,posal between 1997 and 2006.
and operating high-level waste repositories.
13.8.213.8.2.1

Volume Reduction/Solidification Processes
Incineration of Wastes (50,

75,

100,

114)

3
On-site incineration of wastes is currently only being considered for low-activity (0.006 Ci/ft )The effluents and releases
compactible trash to reduce final volumes by a factor of at least 50.
to the outside environment for incineration of combustible compactible trash are detailed in Section 8.3.

Two types of wastes are generated by incineration; one is the resulting ash generated by burning,
Resulting ash could be packaged in 55 gallon
and the other is off-gases released from combustion.
It is estimated that 85% of thedrums and sent in shielded shipments to a low level burial site.
Thus, 15% of the total trash non-volatilenon-volatile radionuclides would remain in this ash.
Treatment consists of a wet scrubber,radionuclide content would be subject to off-gas treatment.
followed by a HEPA (high efficiency particulate filter) and a volatile radionuclide absorption
Thus, in theThis cleanup train will result in a decontamination factor of 106 to 107.
system.
most conservative case, the cleanup system will reduce volatile and non-volatile off-gas concenEstimated releases
trations by a factor of 1 million prior to their release to the environment.
during trash incineration for the four major radionuclides present in combustible trash are given
in Section 8.3.
Off-site incineration at a DOE facility is a potential treatment alternative for the high-specificactivity organic resins.
Such a facility would be equipped with appropriate effluent controls to
reduce releases to levels that are as low as are reasonably achievable and below releasestandards.
The alternative treatment options for the high-specific-activity wastes are discussed
in Section 8.1, and specific DOE incineration facilities are identified in Section 9.1.
13.8.2.2

Evaporation/Bituminization

System and Liquid Volume in the Condensate

The value of 1.44 is the increase in liquid volume required to be handled in the overhead condenThe value is based on data from BNWLsate system for the Evaporation/Bituminization alternative.
TR-196 referenced in Appendix G.
13.8.2.3

Vinyl Ester Styrene System Characteristics(50)

It was assumed that the Vinyl Ester Styrene (VES) solidification process would include aThis accounts for the increase in volume factor of only
dewatering step prior to solidification.
Dow Chemical Company has tested VES up to 2 x 109 rads accumulated dose without degrading
1.5.
If VES is proposed for use, demonstration of. waste form stability under radiathe VES polymer.
tion would be required for the specific wastes to be solidified.

13-3913.8.2.4

Volume Increase Factor for Cement Solidification (50)The volume increase factor for cement was conservatively selected to assure that the solidified
resin product would remain a stable solid monolith without crumbing or spalling.
It is recognizedthat a lower volume increase factor could be used provided the licensee demonstrates the proposed
formulation will produce a stable monolithic waste form.
13.8.2.5

Solidification and Immobilization of Ion-Exchange Media (20,50)

Solidification of the ion-exchange media will immobilize radionuclides to a greater extent thanwill routine packaging of dewatered resins.
Brookhaven National Laboratory studies indicate thatradionuclides will quickly leach from unsolidified resin beads in ground or salt water leachates.
Solidified products, however, would be expected to retain radionuclides better than dewateredresin beads, since the surface area available for leaching has been greatly reduced.
13.8.2.6

Chemical Reactions Between Waste Constituents and Immobilization Agents (78)The agents and process'alternatives discussed for solidifying the waste generated at TMI have been
proven through laboratory evaluation and field applications.
Chemical compatibility under a variety of conditions has been verified. Additional testing of small batches of TMI wastes, solidified with the selected immobilization agents, will be performed prior to actual large scale
processing.
13.8.2.7

Occupational

Exposures during Waste Solidification (50)

Occupational exposures, as provided in Chapter 8.0, include the exposure to workers during solidification operations conducted on-site prior to shipment.
Where the wastes are shipped off-site
for subsequent treatment, the exposures at the off-site treatment facility are not included.
13.8.2.8

Low-Level Waste Act (125)

The Low-level Radioactive Waste Policy Act passed by the Congress in December 1980 and how it
could affect TMI is discussed in Section 9.1.
13.8.313.8.3.1

EPICOR-I Wastes
EPICOR-I Waste'Storage(52)

EPICOR-I liners which were stored in the temporary storage facility were backlogged from the first
several weeks of the accident.
EPICOR-I wastes have been subsequently removed from storage andshipped.
EPICOR-I resins have significantly less activity than do the first stage EPICOR-II
wastes.
13.8.3.2

EPICOR-II Waste (76,

100,

121,

123,

130)

EPICOR-II first-stage resin liners contain approximately 30 ft 3 of ion-exchange material and areloaded to a maximum of approximately 1300 Ci.
These liners contain bulk specific activities ofapproximately 45 Ci/ft 3 , as opposed to 0.1 to I Ci/ft 3 for resins routinely generated by operational
reactors.
In addition, the EPICOR-II first-stage activity is essentially all Cs-137 and Sr-90,
isotopes that have 30-year half-lives.
Routinely generated resins, on the other hand, primarily
contain Co-60 with a 5.3-year half-life.
The Commission order of October 16, 1979 authorizing EPICOR-II operation states, "The licensee
shall not ship spent resins off-site unless they have been solidified, and only then with theprior approval of the Director of NRR (the Director of the Office of Nuclear Reactor Regulation),provided however, that the licensee may ship non-solidified but dewatered spent resins off-site if

13-40it determines, and the Director of NRR concurs, that such shipment is required to assure continued
operation of EPICOR-II or otherwise required to protect public health and safety." This order isdiscussed in Section 1.6.1.5.
The NRC has stated that the first stage EPICOR-II resins would be acceptable for disposal at anarid commercial disposal site if they are solidified, or placed in a high integrity container, and
if disposal procedures include provisions to minimize inadvertent future contact with the waste.
However, the most practical alternative is to transfer these wastes to an existing federal government facility for future processing and eventual disposal.
13.8.3.3

Storage of EPICOR-II Wastes (51,74, 121)

Depleted resins from the EPICOR-II system are being temporarily stored in a facility referred toThis facility is a modular structure with each module
as the Interim Radwaste Storage Facility.
Each cell can hold one 6 x 6 liner or two 4 x
consisting of approximately 60 storage cells.
However, in the event that corrosion
Radiation dose can cause the resins to degrade.
4 liners.
caused the liners to rupture, the contents of the liner would be contained in the storage cells
The NRC staff,
consisting of galvanized corregated steel cylinders with welded steel base plates.
along with the Department of Energy, has anticipated the effects of radiation damage to the resins
and is in the process of developing tests to ascertain the condition of the resins currently instorage.
The NRC staff has not considered the incineration of the EPICOR-II resins as an alternative process prior to storage because of the high-specific activities which would result from theprocess.
13.8.3.4

Degradation of EPICOR-II Resins and Container Integrity (20,93, 64, 100, 114, 115, 120, 121)

53, 59,

75,

76,

74,

84,

Brookhaven National Laboratory (BNL) has indicated that degradation of the ion-exchanae material
in stored EPICOR-II liners may accelerate liner corrosion and produce a material which agglomerates, making further handling and processing of the material difficult. 1- 3 Due to this concern,
NRC has contracted with BNL to perform additional tests to evaluate radiation degradation effects
To evaluate the current condition of the stored EPICOR-II resins, the licensee plans to obtain
In addition, the DOE plans toliquid and gas samples from actual EPICOR-II first-stage liners.
perform detailed tests on EPICOR-II first-stage liners which will include examination of resins
The licensee and DOE tests are designed tofor degradation and carbon steel liner for corrosion.
determine the extent of radiation damage on the ion-exchange material and to compare the actual
The condition of theresin conditions to those projected by testing performed by NRC consultants.
liner, as determined from these tests, would establish the handling, packaging, and disposal conIn addition to requiring solidification prior to disposal, thestraints imposed on each shipment.
The resins
EPICOR-II first-stage resins will require special handling and disposal procedures.
are being stored in specially designed concrete storage modules that contain monitored sumps, sothat in the event of any radioactivity leakage, mitigating actions could be taken.
13.8.3.5

EPICOR-II Waste Volume Projections (74)

The cumulative number of liners from EPICOR-II operation is projected to be higher than indicatedin the October 1979, Environmental Assessment for EPICOR-II System (NUREG-0591).
This is in partIR. Barletta et al., "Status Report on Leachability, Structural Integrity, and Radiation Stabilityof Organic Ion Exchange Resins Solidified in Cement and Cement with Additives,' BNL-NUREG-28286,
May 1980;
2
K. K. S. Pilay, "Radiation Effects on Ion-Exchangers Used in Radioactive Waste Management," NE/
RWM-80-3, Pennsylvania State University and Brookhaven National Laboratory, October 1980;
3
K. Swyler et al., "Review of Recent Studies of the Radiation Induced Behavior of Ioh Exchange
Media," BNL-NUREG-28682, November 1980.

13-41due to the processing of some flushing water used in the decontamination of tanks and equipment in
the AFHB.
However, the increased number of liners does not have a significantly different environmental impact, in terms of either worker exposures or dose to the public.
13.8.3.6

Special Processing Facilities for EPICOR-II Wastes (76)The off-site treatment alternatives for EPICOR-II wastes, which include immobilization and elution
and reuse for the zeolites and resins and, additionally, incineration and acid digestion forresins, are described in Section 8.1.
The facilities at which these treatment options can be conducted, i.e., DOE facilities, are described in Section 9.1.
The transportation routes to thesepotential facilities are provided in Section 9.2.
Off-site storage is a potential alternative that may be considered both in combination with atreatment alternative, or without treatment.
The storage alternatives, and constraints on theiravailability, are described in Sections 9.1 and 9.2.
13.8.3.7

EPICOR-II Resin Packaging and Disposal (64)The discussion of the alternatives for handling, treatment, storage, and disposal of the wastes
generated by the decontamination and cleanup of TMI-2 has been expanded in Sections 8.1, 9.2 and
9.3.
In addition, the costs of these waste management alternatives are provided in Section 9.6.
Alternatives of various disposal sites have been examined and are discussed in Section 9.2. These
alternatives include use of DOE facilities, certain shutdown commercial sites, or a new regional
or intrastate LLW disposal site.
It is likely that interim storage of the TMI-2 wastes either onsite or off-site would then be required until an alternative government or commercial site became
available.
13.8.3.8

EPICOR-II Resins Relative to 10 CFR 61 Guidelines (74)

The proposed limits for shallow land burial are given in the radionuclide concentration guidelines
presented in "Draft Technical Basis for Supporting Additional Technical Criteria and Regulatory
Guides to Implement 10 CFR 61 for Land Burial of Low-level Wastes."(This document is a sectionof the preliminary draft of 10 CFR 61, "Disposal of Low-Level Radioactive Waste and Low-Activity
Bulk Solid Waste.")Wastes containing radionuclides exceeding the guideline values presented in the above document
would generally not be acceptable for disposal with a minimum of three feet of cover material,
although higher concentrations may be acceptable provided there are sufficient barriers to reclaimer intrusion.
Higher concentrations may also be acceptable depending on the form of thewastes, for example, if the waste is solidified.
13.8.413.8.4.1

Waste Storage
Storage of Accident Sludges (52,55)

The treatment, packaging and handling of sludge generated from the cleanup of the Auxiliary and
Fuel Handling Building (AFHB) and the Reactor Building (RB) are discussed in Section 8.1.
Two
alternatives are presented for the packaging of AFHB and RB accident sludge:"

Immobilization of the sludge with cement or vinyl ester styrene in 55-gallon drums
yielding a maximum number of packages.
Activities would range from 17 Ci/package (RBsludge) to 250 Ci/package (AFHB sludge).

"

Dewatering of the sludge and placement in 4' x 4' 45-ft 3-high integrity steel liners yield
ing the-minimum number of packages.
Activities would range from 72 Ci/package (RB sludge)
to 2,700 Ci/package (AFHB sludge).

13-42
Interim storage of dewatered accident sludge liners prior to further off-site treatment and/or
final disposal will take place for as short a period of time as possible.
Liners will be shieldedand handled remotely, and periodic inspection of liner integrity will be performed.
Immobilized
accident sludge drums will be handled and stored in a similar manner prior to transport to adisposal site.
13.8.4.2

On-site Storage Facilities (31,

70,

71,

75,

125)

The on-site storage facilities (Interim Radwaste Storage Facility and Spent Fuel Pool), are intended for temporary storage of radioactive wastes prior to shipment and disposal off-site.
Allradioactive wastes, including fuel from the reactor, will not remain or be disposed of on-siteover the long-term.
It is not intended that, Three Mile Island become a radioactive waste disposal site. The NRC staff has determined that these interim storage facilities are safe alternatives for the purpose of temporary storage.
13.8.4.3

Special Considerations for Temporary On-site Storage (20, 31, 40, 50, 52,
69, 71, 72, 75, 78, 80, 84, 100, 102, 103, 114, 120, 121, 123, 124, 130)

56,

60,

67,

Section 9 includes specific discussions of alternatives for storage of both low-level and highlevel wastes.
"Storage of wastes" connotes a temporary condition where wastes remain retrievable
until final disposition or "disposal" can be performed.
"Storage" does not imply the permanentand ultimate disposition of waste.
Temporary storage of wastes at TMI can be acceptable provided that appropriate engineered facilities are constructed.
The storage modules for EPICOR-II liners and the spent fuel pool are examples of facilities that have been approved for storage of specific radioactive materials.
These
modules would be expected to have at least a 20- to 30-year lifetime.
Temporary storage of radioactive wastes at a nuclear facility is not a violation of NRC regulations.
Any storage facility for TMI wastes will be designed to minimize environmental effects and will be
provided with monitoring systems to detect leaks in waste containers should they occur.
Monitoring systems would provide the licensee with early warning of leaking containers so that appropriate measures can be taken to prevent migration of radionuclides.
The geologic and hydrologic characteristics of the TMI site make it unacceptable as a permanent
repository for low-level wastes and also for high-level wastes that require isolation for thousands of years. On-site storage of wastes could be required for periods of 20-30 years if adequate off-site facilities are not constructed.
NRC is aware that the designation of disposal sites should be performed as soon as possible.
In
addition, the shipment and disposal of TMI wastes should also be performed in a timely manner
provided that these operations are not inimical to the health and safety of workers and the public. The removal of wastes from TMI depends on the availability of suitable off-site storage
facilities or disposal sites. Therefore, a specific cutoff date, after which all wastes would be
required to be shipped off-site, cannot be selected.
13.8.4.4

No Regulation to Prohibit Temporary Waste Storage On-site (114)There is no regulation that specifically prohibits the operation of a nuclear power plant withhigh-activity radioactive wastes stored outside of the spent fuel pool.
However, there are requirements that the storage of these wastes be safe and resulting radiation exposures due to thestorage be within regulatory limits and maintained to as low as reasonably achievable levels.

13-4313.8.4.5

Availability of Off-site Storage Facilities for Fuel and TRU Wastes (32,

55,

103)

There are currently transuranic waste storage facilities operated by DOE at the Los Alamos Scientific Laboratory and at the Idaho National Engineering Laboratory. While the DOE has not agreed
to use these sites for commercial wastes, the use of these facilities is a technically feasible
option.
Stipulations imposed by DOE on the licensee would include meeting specific requirements
for the packaging of wastes to be stored. Off-site storage alternatives, both for the highactivity solid waste and damaged fuel, are described in Sections 9.1 and 9.2.
13.8.4.6

Spent Fuel Storage Alternatives and U. S. Reprocessing Policy (20, 50,

75, 84)The damaged and undamaged spent fuel can be placed in storage either on-site or off-site.
The
spent fuel would be stored until packaging options were developed for disposal in a high-level
waste repository or the fuel was reprocessed.
It is expected that a portion o•f the spent fuelwill be shipped to various laboratories for research and examination.
While reprocessing of spent
fuel is a technically feasible alternative, current national policy precludes reprocessing.
Should this policy be changed in the future, the reprocessing of the TMI spent fuel would bereconsidered.
The existing spent fuel pool, located in the Auxiliary and Fuel Handling Building (AFHB), could be
used for the temporary storage of damaged and undamaged spent fuel removed from the reactor core.
Damaged fuel would be stored in special containers to prevent contamination of the water in the
spent fuel pool.
High activity wastes, such as the zeolite ion-exchange media liners from thezeolite/resin processes used to clean up the liquids in the containment sump, could also be stored
in the existing spent fuel pool.
It is estimated that space for up to 70 zeolite liners is available in the spent fuel pool.
It is not expected that EPICOR-II liners would be stored in the
spent fuel pool.
Since these liners, which have lower maximum total activities than the zeolite
liners, would not require as much shielding.
The components of the core suport structure could be
cut up into smaller sections, and stored in the fuel transfer canal after transfer of the fuel.
In this location, the components should not interfere with other aspects of the cleanup
operations.
Dry storage of spent fuel is discussed as a potential storage option in Section 9.1 and 9.2.
The
use of storage vaults, hot cells, or specially designed spent fuel caissons are possible dry storage options.
13.8.4.7

Storage of Wastes in the Reactor Building (104)The use of the reactor building for storage of wastes is an alternative discussed in Section 9.2.
13.8.4.9

Storage of EPICOR-II Processed Waste (20)The EPICOR-II system does not itself remove radioactive waste from the plant site. EPICOR-II
decontaminates the water and concentrates the radioactivity in ion-exchange liners.
The EPICOR-II process has proved highly successful in removing radioactivity from the accident
water in the AFHB, giving decontamination factors of approximately 107 for the two most importantfission product elements, cesium and strontium. About 55,000 Ci of radioactivity have been removed from 500,000 gallons of water.
The decontaminated water is stored on the site, and the bulkof the radioactivity is contained in spent ion-exchange liners which are stored in the Interim

13-44
Radwaste Storage Facility. Disposal of the processed water (containing tritium) and the liners
Since the site has not been approved
must await final selection and approval of disposal methods.
as a waste-storage facility, and probably cannot be so approved, long-term storage of these materials is not contemplated.
13.8.4.10

On-site Storage Facility Shielding Requirements(50)

The on-site interim storage facilities currently in use at TMI, and those under construction, arediscussed in Section 9.2.
Several facilities will ultimately be used on-site for both storage andstaging areas for different types of waste packages.
Each would be designed for specific characteristics of the wastes to be stored.
Construction of an unshielded interim storage and staging
area for the low-level waste packages is contemplated.
However, all storage facilities will be
required to meet NRC regulatory standards for doses to workers and at the facility boundary, as
well as the principle to maintain radiation doses to levels as low as reasonably achievable.
13.8.513.8.5.1

Transportation of Wastes
Adherence to Transportation Regulations (25,75)

Inherent in the approaches to be selected for packaging and transporting the wastes from TMI isadherence to all DOT and NRC regulations that have been developed to minimize the risks associatedwith transportation of nuclear material.
This applies to the design of the selected shipping
packages and casks, transportation modes and routes, and operational procedures to be followed
under both normal and accident conditions.
Should a vehicle accident occur enroute, the carrier will follow stipulated notification and operational procedures, as defined in an approved emergency plan developed in accordance with NRC andDOT requirements.
13.8.5.2

Transportation Regulations and Procedures (25,40, 75,85)

Transportation of radioactive materials is regulated primarily by the Department of Transportation(DOT) and by the Nuclear Regulatory Commission (NRC).
The DOT regulations are set forth in Title 49 of the Code of Federal Regulations, primarily in 49 CFR Parts 170-189.
The NRC safety
regulations are set forth in Title 10 of the Code of Federal Regulations, primarily in 10 CFRPart 71.
NRC regulations apply both to persons who ship radioactive materials or who offer themfor transport and to carriers who load and transport these materials, providing for protection ofboth transport workers and transport of hazardous materials, including radioactive materials.
In a recent assessment of regulations (NUREG-0170), the NRC staff found that the risks to the
public health and safety from transportation of radioactive materials are very small, concludingthat the environmental effects of normal transportation and the attendant risks allow continued
shipments by all modes and that no changes to the safety regulations are necessary.
The NRC also has an inspection and enforcement program to assure that licensees do in fact comply
with the safety and safeguards requirements during the shipment process.
Current inspection planscall for inspection at the point of origin or the point of destination to determine the licensee's
compliance with applicable requirements.
The NRC is empowered to take enforcement action where
licensees are not satisfying NRC requirements.
The NRC requires licensees to give the NRC advance notice for shipments of spent fuel and passes
this information on to appropriate state agencies on request.
The NRC has set up a special procedure for waste shipped from the Three Mile Island power station
whereby the utility notifies designated persons in each state through which a shipment will pass.

13-45
If requested, the utility also informs the state police of pending shipments,
identification, estimated time of arrival, and route data.

giving them vehicle

As required in Section 301 of Public Law 96-295, the NRC is drafting regulations to require timely
notice for governors of states in which will pass shipments of radioactive waste and spent fuel.
Such notice will not apply, however, to radioactive waste shipments that the NRC determines do notpose a significant hazard to the health and safety of the public.
These regulations cover
The DOT regulations address qualifications of drivers at 49 CFR Part 391.
familiarity with rules, prohibition of aiding or abetting violations, driving ability, age,fluency in communicating with the public and understanding highway signs and signals, driving
experience or training, driver disqualifications, driver background and character, road tests andwritten examinations, physical qualifications and examinations, and files and records.
In addition to training required by the DOT, state transportation agencies, and individual employers, the NRC requires for spent fuel shipments that each driver and escort know under whatcircumstances the cargo vehicle should be immobilized and how to use communication equipment(10 CFR 73.37).
In addition, the escort(s) must have completed training in accordance with Appendix D to 10 CFR Part 73.
The licensee is responsible for establishing the length and frequency ofcourses necessary to meet the requirements of Appendix D, 10 CFR Part 73.
Section 73.73 of 10 CFRPart 73 does not specifically require fresher courses; however, the licensee is responsible for
conducting necessary training to develop and maintain the competency of escorts to accomplish
effective emergency response procedures.
13.8.5.3

Proposed Changes in Transportation Regulations (130)The proposed DOT regulations will not change the special procedure set up by the NRC for advancenotice of TMI shipments.
Under this procedure, the utility notifies designated persons in eachstate through which a shipment will pass.
The notice may include vehicle identification, estimated time of arrival, and route data, depending on what information each state requests of theutility. The NRC has proposed more general advance notice rules for shipments of radioactive
waste materials (45 Fed. Reg. 81058).
The basic effect of the proposed DOT regulations is not to deregulate truck transportation ofradioactive materials but to make uniform the routing requirements of such transportation.
Withthese regulations, the DOT intends to reduce the possibility of exposure and inadvertent releases
in normal and accident situations in transportation and to clarify the scope of permissible stateand local action (45 Fed. Reg. 7140).
The NRC has studied the effect of highway routing controls on accidents in its environmental
statement of transportation of radioactive materials (NUREG-0170).
By routing trucks on turnpikes
or interstate highways, the accident rate would decrease by about ten percent (NUREG-0170, p. 612).
The number of accidents cited would not change significantly.
13.8.5.4

Possible Illegal

Waste Shipments (64)The wastes generated at TMI-2 are carefully accounted for on-site, and manifested for off-site
shipment.
The NRC, utility, and carrier maintain detailed records for each shipment, as does thedisposal site. Records are available for tracking the waste from its generation through disposal.
Illegal shipments from TMI-2 are extremely unlikely, and none have ever been detected.
13.8.5.5

Commercial Shipment of Wastes (130)

Shipment of wastes to the commercial disposal sites will be performed by commercial shippers inaccordance with DOT and NRC regulations.
Some shipments of radioactive material to DOE research13-46
facilities may be performed for the aovernment by commercial shippers or in special cases ingovernment-owned and operated vehicles and equipment.
13.8.5.6

Shipment of Small Ouantities of Contaminated Liquids (50)

Shipments of small test quantities of contaminated liquids and ion-exchange materials for laboratory examination, packaged and handled in accordance with NRC and DOT regulations, arepermissible.
13.8.5.7

Risk Associated with Cross-Country Shipment of Wastes (45)The risk in transporting waste, is linearly proportional with distance traveled.
Although therisk of transport for a short trip would be less than that for a longer trip, however, in either
case, the risk would be small.
The current "most likely" truck route from TMI to Hanford, Washington, is 2750 miles long.
Anevaluation has been made of the exposures that individuals along the transportation route receivefrom the waste shipments from TMI.
The results of this evaluation, which are provided in
Section 9.5.1.2, show that individuals living along the route are being exposed to very small(< 0.01%) increases in radiation levels over natural background for the maximum potential numberof waste shipments.
13.8.5.8

Modes of Waste Transport (43)

Truck shipment is the current transportation mode for the TMI-2 wastes, and it is anticipated that
it will be used for future waste and fuel shipments.
Rail and intermodal transportation are mentioned in the context of available alternatives but, because of their relative constraints anddisadvantages, (Section 9.2) they are not presently being considered for shipment of any waste or
fuel.. Should this situation change, the environmental effects of these modes would be evaluatedseparately.
13.8.5.9

Transportation Routes (3,

9,

32,

50, 51,70, 86)

Only one shipment of radioactive solid wastes has traveled via either Route 11 or Route 15.
two roads will not be used for future shipments.

These

Section 9.2 provides a discussion of transportation alternatives and routing which includes thepotential routes to all viable off-site storage, treatment, and disposal facilities, not only to
the Hanford facility.
The routes for transporting the radioactive waste to an off-site waste storage, treatment, or
disposal facility are selected on the basis of current DOT and NRC guidelines.
The interstate
highways, in conjunction with urban bypasses, represent "preferred" routes.
The transport routes
are selected after considerable study, and in consulation with state and municipal officials.
Prior to individual shipments, when shipping plans are confirmed, factors such as temporary hazards enroute are considered and, if necessary, modifications are made in the plan to avoid them.
Section 9.2 shows the "most likely" transportation routes to all the possible waste treatment,
storage, and disposal facilities.
Since a number of interstate, U. S., and state highways arepotential routes, it would not be appropriate to single out one route, such as 1-80, for a separate accident study.
Accident statistics are a factor in selection of all the routes.
The selected routes are subject to review by state agencies and to revision to assure that impacts
from normal transportation and potential accidents are minimized.
Therefore, the selected13-47transportation routes will incorporate any requirements imposed by the governmental bodies of thestate being traversed.
The routes to all possible waste treatments, storage, and disposal
facilities are presented in Figure 9.2-9.13.
13.P.5.10

Radiation Exposure during Waste Transport (25,60,

66, 70,

72,

75,

98,

107,

125)

The occupational exposures calculated for the waste truck drivers are presented in Section 9.5.
The analysis is based on two drivers making 30 trips per year on each of two potential routes--thelongest route of 2,750 miles to Hanford, Washington; and the shortest of 370 miles to West Valley,
New York.
The annual crew dose is 11.0 person-rem for shipments on the longest route, and1.6 person-rem for shipments on the shortest route. The occupational exposures truck drivers
receive are regulated, and each individual is monitored to assure that the established limits are
not exceeded.
In determining the 1.3-mrem exposure of members of the public standing three feet from a stopped
loaded truck, a three-minute exposure duration was assumed.
During the course of a shipment, it
has been conservatively assumed that ten people would be so exposed, resulting in a total population dose of 0.013 person-rem.
As an upper bound the transportation doses to people residing along the shipping route from TMI to
the Hanford disposal site for the minimum and maximum number of shipments have been evaluated torange from 16 person-rem to 50 per-rem (Section 9.5.1.2).
The approach used for determining such
exposures is based on the analysis provided in NUREG-0170, "Final Environmental Statement on theTransportation of Radioactive Material by Air and Other Modes."Occupational exposures have been determined for all the workers involved in the waste management
steps, including those handling and loading the waste packages onto the trucks and the healthphysicists monitoring these activities.
These exposures are provided in Section 9.5.
The exposure of transport workers during loading and shipment is not comparable to exposure ofindividuals in the general public.
Occupational radiation exposure limits are higher because of
several reasons; for example, radiation workers are closely monitored for radiation exposures,
records of cumulative radiation doses are maintained, and the work they perform are under theprovisions of the licensee's radiation protection plan procedures.
The dose to people living near the disposal sites has not been calculated for specific TMI waste
shipments since the TMI waste does not add to the waste disposed of at the facility, but replaces
other waste shipments.
Thus, the TMI waste does not add to the exposure to be received by the
population in the vicinity of those facilities.
The potential for transportation accidents and the impacts associated with both airborne andwaterborne releases from such accidents have also been evaluated (Sect. 9.5).
Release fractionsof radionuclides are calculated based on conservative assumptions of severe accident conditions,and a representative inhalation dose determined for this type of accident.
Under this "worst-caseaccident scenario, whose potential frequency of occurrence is in the order 10-9 accidents pertruck mile for the longest route (2,750 miles), the inhalation dose is less than 10% of the annual
dose due to background radiation.
Thus, in the unlikely case an accident does occur, the effectof radiation release is not considered to be a significant impact.
The statistics for accident rates used in Section 9.5 were obtained from accident rate statistics
in "Transportation of Radioactive Material by Air and Other Modes," NUREG-0170.

13-4813.8.5.11

Responses to Transporation Accidents (25)Currently, if an accident occurs, state and local governments are primarily responsible for overseeing the response of carrier, shipper, and others and for taking any actions deemed necessary toprotect health and safety. To assist state and local governments, the federal government has a
program called the Federal Radiological Monitoring and Assessment Plan (formerly called the Interagency Radiological Assistance Plan) which could be used to assist state and local authorities.
The DOE charges eight regional
The program is coordinated by the Department of Energy (DOE).
toconveneradiological assistanceandauthority
withtheresponsibility
coordinating offices
When called, a team reports to the scene of an accident or other radiological emergency
teams.
and assists the emergency response personnel already on the scene.
The subject of emergency preparedness in transportation of radioactive materials is under active
A joint NRC/DOT study group (NUREG-0535) recommends
consideration by both the NRC and the DOT.
several actions, including federal rulemaking, on response planning by shippers, carriers, andThe group recommends that state and local agencies develop plans tostate and local agencies.
both advise and assist the carrier and to take appropriate control actions at the scene of an
accident to protect public health and safety.
To assist state and local governments in planning emergency responses to radiological incidents atfixed sites or in transportation, the federal government has an interagency program to coordinate
In this program, the DOT supplies guidanceplanning, guidance, and training (44 Fed. Reg. 69904).
The Federal
on emergency response planning related to transportation of radioactive materials.
Emergency Management Agency (FEMA) is the lead agency in this program.
13.8.5.12

Transportation Accident Parameters(120)

An analogy drawn between the behavior of solidified TMI waste and the waste immobilized in glassin Section 9.5 indicates that the fractional release and dispersion of airborne particles would be
similar to those under the conditions of a severe truck accident causing container rupture andattendant fire explosion.
Two accident cases described in Section 9.5, the release of airborne respirable particles from aruptured container with a fire and the release of waterborne material from a ruptured containerThe
underwater, are considered to be boundary cases for maximun impact of potential accidents.
B
since
Type
assumptions,conservative
using
highlyanalyzed
been
have
used
release fractionstransportation packages, which would be used for the high-activity-wastes, are designed to withA waterstand the effects of a 30-minute diesel-fuel fire without damaging the contents.
immersion test is also included for this package.
13.8.5.13

Evaluation of Transportation Accidents (66,Transportation accidents are addressed

25,

107,

125)

in Section 9.5.

The environment of the core during the TMI-2 accident was significantly different from the enviThe TMI-2 core was apparently
ronment of the fuel in a transportation accident would be.
uncovered for over half its length for a time long enough to allow extensive damage from overThe high temperatures that occurred caused fuel clad failure and the high fractional
heating.
On the other hand, in a transportation accident,ofthe more volatile fission products.
releaseeven with loss of coolant for spent fuel, both temperatures and the release of fission products
would be much lower.
13.8.5.14

Accidental Releases from a Breached Cask (75)

Although the Type B casks are subjected to a test program inclusive of a 30' drop test, they are
In the highly unlikely
designed to withstand stresses associated with greater drop distances.

13-49event that a cask is breached as a result of the postulated 60' drop from within the building thebuilding effluent control systems are designed to handle the airborne effluents, maintaining
outside releases within regulatory limits.
13.8.5.15

Shipping Radioactive Wastes in Lower Concentrations(47)

Shipping radioactive wastes in less concentrated forms would not necessarily lessen the total
Rather, an increased number of shipments, increased radioactive
potential environmental impact.
An increased
waste handling, and a potential increase in total worker exposures would result.
number of shipments may also result in a higher potential for accidents and subsequent releasesof radioactivity to the environment.
The current practice of radioactive waste shipments withadequate safeguard features (i.e., waste immobilization, radiation shielding and container
integrity requirements) has a smaller potential for environmental impact.
13.8.5.16

Vehicle Driver Exposure (130)

Routine exposure to both truckdrivers and members of the general public from transportation of
A truck crew member would
radioactive waste materials from TMI are discussed in Section 9.5.
Since the TMI
receive about 60 mrem per trip for the TMI wastes for the longest shipping route.
wastes are to be shipped under exclusive-use arrangements (a regulatory term for shipment with
only one consignor and only one consignee), the radiation level inside the cab is limited by bothAssuming a crew made round-trip shipments to Hanford,
NRC and DOT regulations to 2 mrem per hour.
WA, every ten days, the maximum number of trips which could be made in a year is 36.5. At 60 mremThis value is less than the occupational
per trip, the dose to each crew member would be 2.2 rem.
exposure limit in 10 CFR Part 20.
13.8.5.17

Availability of Shielded Shipping Casks (75)Based on the current cleanup, decontamination, and waste disposal schedules and estimates to packaging waste generation, it does not appear that the availability of casks will pose severe conHowever, as noted in Section 9.1 the purchase or lease ofstraint on the waste shipment schedule.
additional shielded casks is being considered to alleviate any possible constraints on theschedule.
13.8.613.8.6.1

Waste Disposal
DOE to Engage in Radwaste Disposal Evaluations (78)The question of where commercial radioactive wastes, including those from TMI, will be sent forlong term storage/disposal has been studied by both the Department of Energy (DOE) and the NRC.
The general conclusions of these studies are that radioactive waste can be safely disposed of,
The details of these studies
that present plans for establishing disposal sites are established.
can be found in several documents, among which are the "Statement of Position of the United States
Department of Energy" in regard to "Proposed Rulemaking on the Storage and Disposal of Nuclear
Waste" dated April 15, 1980 (DOE/NE-0700) and the Final Environmental Impact Statement on"Management of Commercially Generated Radioactive Waste (DOE/EIS-0466F, 3 volumes).
DOE is also actively engaged in the evaluation of several aspects of the alternatives for theThis includes the testing
ultimate disposal of the radioactive wastes and nuclear fuel material.
of waste forms, selection of suitable sites, and methods of disposal of both the solid radioactive
wastes and the fuel from the reactor.
13.8.6.2

Determination of Disposal

Impacts (24,64)

The specific disposal methods and sites for TMN-generated wastes will be selected based on the
Even though not all the specific disposal
characteristics of the individual waste materials.

13-50methods and sites have been selected for TMI wastes, it is possible to identify alternatives andestablish bounding estimates for overall differential disposal costs and potential health effects
to workers and the public.
Inflation costs have not been considered by escalating base year costs
using an estimated inflation factor.
13.8.6.3

Waste Disposal Should Not Be a Prerequisite to Cleanup (50,98)

Although questions regarding offsite disposal still exist, especially with regard to highspecific-activity wastes and spent fuel.
However, the NRC staff has indicated that resolution ofoff-site disposal questions should not be a prerequisite to proceeding with on-site cleanup
activities.
13.8.6.4

Low-Level Waste Disposal Site Availability (11, 16, 32,
75, 79, 84, 85, 92, 93, 98, 101, 107, 114, 115, 123)

50,

51,

53,

55,

61,

64, 67,70,

The commercial disposal site at Hanford, WA, is currently accepting certain wastes generated at
TMI.
The wastes which have been shipped to Hanford include contaminated trash, some solidified
decontamination solutions, and EPICOR-I resins.
In November 1980, an initiative was passed in theState of Washington which would prohibit the Hanford commercial disposal site from accepting outof-state non-medical wastes after July 1, 1981.
Agreements between other states and the State ofWashington, however, could be formulated which would allow the acceptance of wastes from thoseother states.
In the event that existing commercial low-level waste disposal sites are unavailable for TMI
wastes, storage of these wastes either on-site or off-site would be required until an alternative
disposal site becomes available.
On-site or off-site storage would be required for wastes requiring disposal in a high-level waste repository until such a facility becomes available.
It isintended that the cleanup of TMI would proceed and not be predicated by selection of ultimate
disposal methods.
13.8.6.5

West Valley (125)

The West Valley disposal site is considered as an alternative for disposing of TMI generated lowlevel wastes.
However, there are several institutional and technical issues which must be resolved prior to reopening West Valley for TMI wastes.
13.8.6.6

Quantities of Waste Generated (32,50, 75)The volume of each waste form generated from various sources during the cleanup and decontamin'ation operations was determined from an evaluation of the treatment or conditioning process characteristics, in conjunction with the quantities of input material to be processed.
Experience atother facilities undergoing decontamination, results of development testing at industrial andgovernment facilities, and actual processing experience have provided the process parameters usedin each case.
The waste volume generated are tabulated for each waste form in Section 8.1.
13.8.6.7

Reactor Systems Wastes (50)The evaluation of waste types and volumes generated in the reactor system cleanup or core removal
considers that all of the system components, both major parts and miscellaneous hardware, thatwill be disposed of as waste.
The type and quantities of waste are described in Section 8.1.

13-5113.8.713.8.7.1

High Level Waste
Final Disposition of High-Specific-Activity Wastes (52, 64, 75, 76, 85)Section 9 includes specific discussions regarding the alternatives for the disposition of the SDS
wastes. NRC agrees that the first stage zeolite wastes, which may have Cs-137 at specific activities on the order of 1000 Ci/ft, are unacceptable for routing shallow land burial methods.
Wastes which are unacceptable at the commercial low-level waste disposal sites due to very highspecific activities would need to be disposed of using methods which provide greater isolationfrom the environment than does commercial land burial. The only option which is currently being
developed and that would provide the required degree of isolation, is the high-level waste repository. Therefore, high-specific-activity wastes are being considered as candidates for the highlevel waste repository.
13.8.7.2

Disposal of High-Level Wastes and Spent Fuel (11, 13, 16, 20, 51, 52, 60, 64, 67, 71,
76, 79, 85, 92, 93, 100, 107, 109, 114, 115)

The DOE has the legislative responsibility for siting, building and operating high-level waste
respositories. The DOE currently is performing field studies in several locations throughout theU. S. to determine potential sites for a high-level waste repository e.g., salt domes in Louisianaand Mississippi, in basalt formations in Hanford, Washington, in bedded salt in New Mexico, andtuff formations at the Nevada test site. Based on current plans, it is expected that Final Environmental Impact Statements will be completed for a potential repository for salt domes in July
1983, basalt formations by February 1983, bedded salt in September 1984, and for the Nevada testsite tuff formations in November 1984. Following a review of the characteristics of the potentialsites in various media, a site will be selected for the first repository. The current schedules
include availability of the first repository for commercial high-level wastes between 1997 and2006.
On-site or off-site storage would be required for wastes requiring disposal in a high-level waste
repository until such a facility becomes available. Storage facilities for high-activity wastes
would be designed to minimize environmental impacts over a 20- to 30-year storage period to account for possible delays in repository operation.
Currently, spent fuel from nuclear power plants is stored in fuel pools similar to the TMI fuelpool. This spent fuel also requires continued storage until a high-level repository is sited,
constructed, and licensed. Therefore, the spent fuel at TMI will be disposed of in a similarmanner to the spent fuel which exists at other nuclear power plants. The damaged spent fuel will,
however, require special packaging for transportation and possibly special processing to meet thewaste form requirements for the high-level waste repository.
13.8.7.3

High-Level Waste Repository Site Selection (51)The distinction between site selection for a high-level waste disposal facility and site selection
for a nuclear power plant, is found in the requirements for storage of the waste on or near thesurface in a retrievable manner versus the ultimate deep geologic disposal in a high-level waste
repository. Deep geologic disposal requires that the site, located thousands of feet below thesurface, isolate the waste from the environment for thousands of years. The basic site requirements for a geologic repository are geologic stability, limited flow of ground water and isolationfrom the activities of future generations of human beings.
13.8.7.4

DOE Acceptance of High-Activity Wastes (48, 52, 59, 80, 105)

DOE has the facilities, capabilities, and expertise in the processing, storage, and disposal of
the high specific activity wastes. The issue of DOE acceptance of these wastes involves the combining of defense and commercial wastes and the possible NRC licensing of these activities.

13-52In this regard NRC Chairman Ahearne, in a letter to DOE, has written to Secretary Duncan on
October 20, 1980 and has requested DOE assistance for the processing, storage, and disposal of
In a follow-up
some of the high-activity waste forms which will be generated in the TMI cleanup.
letter, the NRC Executive Director for Operations wrote to the Under-Secretary of DOE and indicated that NRC licensing jurisdiction would not include DOE facilities whose primary function
remained the handling of defense and research waste materials.
13.8.7.5

Future Status of DOE (102)The DOE facilities which are identified as options for waste processing and disposal are involved
These facilities would continue to be operated whether under thein national defense activities.
jurisdiction of the DOE or under another government agency and TMI would not become a radioactive
waste disposal site.
13.8.7.6

Submerged Demineralizer System (SDS)

Waste Disposal(70)

The licensee has proposed to use the submerged demineralizer system (SDS) for processing the
The SDS(The NRC has not authorized the use of the SDS.)
liquids in the reactor building sump.
The SDS and EPICOR-II, however, utilize ionuses different components than the EPICOR-II.
exchange as the mechanism for cleanup of the radionuclides in the liquids.
The SDS first stage will generate an inorganic ion-exchange material (zeolite) having a specificactivity much higher than wastes which are routinely generated at other nuclear power plants.
Because these wastes will be unacceptable for disposal at commercial shallow land burial sites,
storage on-site will be required until either the DOE accepts these wastes for processing with
their own similar wastes or until acceptable disposal options, such as a high-level waste repository, is licensed and becomes operational.
13.8.7.7

High Specific Activity Waste (53,

76,

98)

For the ion-exchange wastes whose specific activity and inventory of radionuclide contamination
requires that they be handled as High Specific Activity Waste (HSAW) e.g., the first stage of the
SDS system, off-site storage, treatment, and disposal options will be different from those forlow-level waste (LLW).
The off-site treatment alternatives for this waste, which include immobilization and elution fromthe zeolites and resins and, additionally, incineration and acid digestion for the resins, are
The DOE facilities at which these treatment options can be conducted,
described in Section 8.1.
The transportation routes to these facilities are provided inare described in Section 9.1.
Section 9.2.
Interim storage of HSAW and fuel, on-site and probably off-site, are viable and necessary options.
The cleanup of TMI-2 is not considered to be dependent on the selection of an ultimate disposal
The cleanup can proceed and, as required, thesite (i.e., geologic repository) for this material.
Offtreatment/disposal option is available.
until
the
selectedcanbestoredwaste
forms
varioussite storage may be considered for HSAW both in combination with a treatment alternative, or
The storage alternatives, and constraints on their availability, are describedwithout treatment.
in Sections 9.1 and 9.2. The routes to these facilities are provided in Section 9.2.
The alternatives and existing constraints for disposal of HSAW are discussed in Sections 9.1 andSince the prime option for disposal of the HSAW is in a geologic repository, specific loca9.2.
tions cannot be identified pending selection of potential repositories.
13.8.7.8

Vitrification of Ion-Exchange Media (50,

55,

76,

121)

The discussion of a treatment alternative for the high-specific-activity ion-exchange material isAmong the treatment alternatives discussed is immobilization by vitrififound in Section 8.1.
cation in glass, which would be conducted at off-site facilities designed to process high-level13-53Specific facilities at which these immobilization capabilities exist are idenwaste streams.
tified in Section 9.1.
Facilities available for immobilizing the HSAW in a glass or ceramic matrix are located at DOE sites.
Representative costs for each of the viable treatment, packaging, and disposal alternatives are
discussed in Section 9.6 and Appendix G. Thus, it will be possible to weigh costs along withtechnical feasibility and environmental impacts to evaluate an alternative.
13.8.7.9

Acceptability of High-Level

Wastes at Hanford (92)High-level wastes would be unacceptable for disposal at the Hanford commercial disposal site,
notwithstanding the initiative passed by the Washington State Electorate in November 1980.
Highlevel wastes would require disposal in a high-level waste repository which is to be developed by
the DOE.
DOE is currently investigating suitable geologic media at several sites for locating ahigh-level waste repository.
The first repository is scheduled to be available for disposing ofwastes between 1997 and 2006.
13.8.7.10

Special Nuclear Material

In Wastes (50,100)

Special nuclear material (SNM) is defined in 10 CFR 70 as: "(1) plutonium, uranium-233, uranium
enriched in the isotope 233 or in the isotope 235, and many other materials which the Commission,
pursuant to the provisions of Section 51 of the [Atomic Energy Act], determines to be special
nuclear material, but does not include source material; or (2) any material artificially enrichedby any of the foregoing but does not include source material."
The regulations in 10 CFR 70 do
not provide numerical concentration or mass limitation exemptions for SNM.
The commercial disposal sites are currently licensed to accept for disposal waste materials containing less than 10 nCi/gm of TRU materials including plutonium.
Wastes containing greater than
10 nCi/gm would require storage until either a high-level commercial waste repository or a TRU
disposal facility becomes available.
Disposal site license conditions also include specific possession limits for uranium-233.
These
possession limits vary at each disposal site and are based on criticality considerations.
Since
most uranium-233 is derived from the thorium fuel cycle, which is not applicable to TMI, uranium233 will not limit disposal considerations.
The source of plutonium in the wastes would be fuel debris,which could be removed as particulates

in the processing of the reactor building water or the primary system liquids.
13.8.813.8.8.1

Other Comments
Comparison of Requirements for Processing and Disposal of Wastes at TMI and at Other
Nuclear Power Facilities (130)Those TMI wastes which are similar to routinely generated wastes are acceptable for disposal atcommercial shallow land burial sites. These wastes include contaminated trash, some solidified
decontamination solutions, and some of the EPICOR-II resins.
EPICOR-II first stage resins, however, contain bulk activities of approximately 40 Ci/ft 3 as opposed to 0.1 to 1 Ci/ft 3 for routinely generated resins.
Also, the EPICOR-II first-stage activity
is essentially all Cs-137 and Sr-90, isotopes that have 30-year half-lives.
Routinely generated
resins, on the other hand, primarily contain Co-60 with a 5.3-year half-life. Because of thehigher concentrations and longer half-lives, EPICOR-II first-stage resins would require solidification, or placement in a high integrity container, prior to disposal as well as special handling
operations and disposal procedures at commercial disposal sites.
The most practicable alternativeis transfer these wastes to an existing federal government facility for future processing andeventual disposal.
First-stage ion-exchange material from the licensee proposed system to process
the reactor building sump liquids will have Cs-137 concentrations of approximately 1000 Ci/ft 3 .
Wastes having thesecharacteristics will be unacceptable for routine shallow land burial.

13-5413.8.8.2

Foreknowledge of the Characteristics of Unique Accident Waste (32)The staff has reviewed material, including documents and testimony of individuals, as well as proand anti-nuclear groups pertaining to the probabilities of severe accidents at nuclear power facilities. None of this material provides specific characteristics of waste products that might begenerated during the post-accident cleanup to any great detail useful for the cleanup.
13.8.8.3

Possible Use of Salem Nuclear Power Facility as a Disposal Site (48)

Radioactive solid wastes from the evaporation and the zeolite/resin alternatives for processing
liquids at TMI will have to be disposed of at either licensed commercial or government disposal
facilities, which have acceptable geologic hydrologic characteristics for the disposal method
used.
Currently, the only licensed commercial disposal facilities are located in Barnwell, SC,
The geologic/hydrologic characteristics of the Salem site (shallow
Beatty, NV, and Hanford, WA.
ground water table and oceanside location) would likely prevent it from being acceptable as adisposal site for TMI wastes.
13.8.8.4

Access to Proprietary Data (53)

Under NRC regulations, NRC can obtain and use proprietary data affecting safety reviews and proIn most cases, commercial firms freely releasevide protection of a firm's proprietary interests.
For firms that do not release proprietary data freely, NRC has
the information requested by NRC.
subpoena power to obtain the information necessary to prepare safety reviews.
13.8.8.5

Decontamination and Disposal of Reactor Coolant Pumps and Motors (75)The hazards of handling, transporting, and disposing of all wastes, including the reactor coolant
pumps and motors, will be considered to assure compliance with NRC and DOT regulations.
13.8.8.6

Use of Metal LSA Boxes (50)

Metal LSA boxes are a viable alternative to the use of wooden boxes, cost and availability being
considerations.
13.8.8.7

Clarifications (75)

The term "remedial activities" as used in the PEIS intended to apply to the various cleanup activities (other than decontamination) currently in progress at TMI.
The term "current phase of operations" used in the PEIS relates to the work performed to date
(January 1981) at TMI-2 in decontamination and other cleanup operations as far as the managementof the generated waste is concerned.

13-55
13.9

RADIOLOGICAL EFFECTS

13.9.113.9.1.1

Releases & Effluents
Prudent vs Zero-release Levels (125)zero-release objective is unattainable and does not provide useful guidance for establishingguidelines or choosing between alternatives.
It is more appropriate to choose those alternatives
that will minimize all of the environmental impacts (not just those that result from releases)
and hazards for all of the cleanup operations over the entire cleanup period and beyond. The
best method is to: (1) analyze and divide the overall cleanup project into a series of operations and tasks that are not too strongly interdependent, (2) identify reasonable alternative
approaches for each, and (3) choose those alternatives for which the environmental impacts andhazards are "as low as reasonably achievable" (ALARA) as the cleanup proceeds.
In many cases thedata needed for an informed judgement for a particular cleanup task will not be available until
prior cleanup tasks have been completed.
However, the complex situation and the limited amountof information available require an approach in which the experienced personel responsible forthe cleanup operations make many specific decisions. The NRC will, of course, retain the rightto review and approve specific plans as they are developed.
If the dose any individual receives from a particular release were very small compared to thedoses for natural sources of radiation, it would not be reasonable to undertake a major expenditure to further reduce the release. The "ALARA" concept used in the PEIS recognizes and deals
with such situations; the "zero-release" concept does not.
13.9.1.2

Present and Future Abundances and Locations of Radionuclides (64,74, 75, 79, 84, 85)

The total quantities of fission products and actinides in the TMI-2 reactor at the time of theaccident, and for any subsequent time thereafter, are accurately known from detailed calculationsof the buildup and decay history of the individual isotopes. Tritium, cesium and strontium arepresent in 3 places: namely 1) the primary cooling system, 2) the Reactor Containment Building(RCB) Sump and 3) the EPICOR II resins (which removed the radionuclides from the AFHB water).
In
addition, there is still some tritium left in the processed AFHB water. Radionuclides are present in one of the following four forms: 1) a solute in water; 2) "plateout" on surfaces of thecooling system or RCB sump; 3) sludge in the AFHB and the RCB sump and 4) suspension of fineparticles (about 5% of the sump concentrations).
In chapters 7 and 8 a detailed mass balance is given showing the source and final disposition(waste form) of the radioisotopes in the RCS and the sump water, and also those radioisotopes
which entered the RCS during defueling. The quantity of sludge, filterable solids and"plateout", and the way in which it is treated are discussed in Sections 5 and 8.
13.9.1.3

Best and Worst Cases of Processed Water Activity (13)Estimates of the concentrations of the radionuclides in the processed water are dependent on thewater treatment system (Section 7.3.1.2).
The best and worst case treatment systems were identified and used to characterize processed
Reactor Building and Reactor Coolant System water. The best case, which led to the effluent withthe lowest concentrations of contaminants, was the SDS/EPICOR II treatment system. The worst
case, which led to the effluent with the highest concentrations of contaminants, was the SDS
treatment system. These systems and other alternatives are discussed in Section 7 andAppendix G.

13-5613.9.1.4

Results of Reactor Building Purge Samples (73)

Airborne particulate samples, released during the purge of the TMI-2 reactor building atmosphere,
All results were less than thewere analyzed for gross beta activity by EPA and the licensee.
NRC also estimated the Sr-90
minimum detectable levels (0.1 pCi/m 3 of air sampled for the EPA).
and Cs-137 airborne particulate radioactivities to be less than 0.06 pCi each inthe TMI-2 reacFollowing the purge, EPA performed radiochemical analysis for the
tor building prior to purging.
combined filters for strontium-90, but those results are not yet available.
13.9.1.5

Discharge of Krypton (20)Information regarding the purge of krypton from the TMI-2 reactor building atmosphere is available in a separate comprehensive environmental review NUREG-0664.
13.9.1.6

Alternatives for Disposing of Processed Water (85)Section 7 provides a detailed assessment of the alternatives for the controlled release of theprocessed water, including facility, effluent and radiological doses under normal operation andOther potential impacts such as psychological stress and socioeconomic
accident conditions.
impacts are also addressed.
13.9.1.7

Radioactivity Levels for Processed Water Release Alternative (12,

72,

79)

At present, the NRC has not authorized any alternative for the release of processed accident
contaminated water from TMI-2, including the alternative for the controlled release of processed
Should the NRC decide to authorize such release, it
accident water into the Susquehanna River.
will ensure that the licensee implement the release in a controlled manner, according to theconditions (technical specifications) and criteria determined by the NRC, to protect the public
In particular, the concentration of radioactivity in thehealth and safety and the environment.
water at the nearest drinking water intake downstream of TMI would be diluted to levels below EPA
Drinking Water Standards.
Also, the absorbed radiation dose in fish and seafood in the river andChesapeake Bay area would be an insignificant fraction of the normally occurring natural background radiation dose and should cause no detectable biological effect.
(Section 7.2.)
13.9.1.8

Additional Processing of Decontaminated Water (52)Processing of radioactive waste liquids would be required to follow the principle of maintaining
Thus, when theradioactive material releases to as low as reasonably achievable (ALARA) levels.
processing alternative meets the numerical design objectives of Appendix I to 10 CFR 50, as proposed by the Staff in Section 1.6.3.2 and Appendix R, further processing of the processed water
One alternative discussed for the disposal of processed
would not be considered to be necessary.
A suggestion to consider such an evaporationwater is by forced evaporation via a heated pond.
"in a container, with filtered vapor venting, (that) would allow release of essentially nothing
but tritium, and would allow non-volatiles in the bottom water to be solidified" is analogous toconsidering an additional processing step for the processed water which already meets the ALARA
principle.
Such additional processing is not considered to be necessary and has not been included in the discussion.
13.9.1.9

Factors for River Dilution Considerations(52)

For releases of long duration, it is possible to relate the steady state dilution of Susquehanna
See Section 7.2.5.4 for a moreRiver water into the Chesapeake Bay to the observed salinity.
detailed discussion.
13.9.1.10

Interaction of Tritium and Other Radionuclides with Sediment (46)The behavior of tritium in the form of tritiated water (HTO) is practically identical to that ofordinary water (H 2 0).
Tritium does not settle and remain in sediment any longer than does the13-57hydrogen in ordinary water. Of the environmentally significant radionuclides which could be
In any event, thereleased from TMI, only Cs-134 and Cs-137 interact strongly with sediments.
radiocesium would be present in much lower concentrations in the sediments of the Chesapeake Baythan deposits due to other sources such as nuclear fallout or routine releases from operating
nuclear plants. The behavior of these nuclides is discussed in Section 7 and Appendix T.
13.9.1.11

Distribution of Radionuclides in River (13)The calculations of radionuclides in the Susquehanna River and Chesapeake Bay have been revised.
Calculations of radionuclide concentrations in fish flesh near TMI consider the limited flow of
the center channel of the Susquehanna River. Radionuclides will be completely mixed at the
points in the river where drinking water supplies are taken, for the reasons given in Section 3.
An analysis of radionuclide concentration in fish near TMI appears in Section 7.2.5.4.
13.9.1.12

Effect of Sediment in Radionuclide Distribution (21,64, 74, 75,

79, 85)The importance of sediment on the radioecology of the Susquehanna River and Chesapeak Bay is
discussed in Section 7 and Appendix T. Of those radionuclides interacting with sediment, themost radiologically important are Cs-134 and Cs-137. A conservative calculation of the effects
of sediment on the transport of radiocesium released from the Peach Bottom Nuclear Plant was
performed in Appendix T. This calculation is based on measurements of Cs-134 and Cs-137 in sediment and fish of Conowingo Pond and the upper Chesapeake Bay. The staff estimates that no morethan 12 percent of the radiocesium released from the Peach Bottom Plant became associated withthe sediment.
The effects of sediment on the radioecology of the Susquehanna River andChesapeake Bay will therefore be minor. Gross (1978) estimates that sediment is eroded from theSusquehanna River for flows of 400,000 cfs or greater. Flows of this magnitude would have arecurrence interval of about six years. Dilution of resuspended sediment would be great for suchlarge flows.
13.9.1.13

Location in River/Bay Where Radioactivity Can Be Detected (42, 52)

Except in the immediate vicinity of TMI-2, it is unlikely that any radioactivity released from
TMI will be detectable in the Susquehanna River or in the Chesapeake Bay above naturally occurring levels caused by nuclear fallout or normal releases from other nuclear plants. See Appendix T.
13.9.1.14

Release Estimates fromnTrash Incineration Alternative (83)

Release estimates from trash incineration are provided in Section 8.
13.9.1.15

Source of Tritium in Sample Wells (69)

Tritium has been detected in sample wells at the TMI-2 facility. The licensee and NRC believe,
from the isotopic analysis results and from the location of the samples, that the leaks were fromthe valves of the outdoor Borated Water Storage Tank, and not the reactor building. The licensee
has a continuing program to monitor and rectify the source of leaks.
13.9.1.16

Radioactivity Releases from TMI Prior .to the Accident (107)

The radioactivity effluents from TMI were monitored and reported to the NRC in compliance withthe requirements of the TMI technical specifications.
The offsite radiation doses during theperiod prior to the accident were within the regulatory limits of 10 CFR Part 20 and met thenumerical design objectives of Appendix I to 10 CFR Part 50.
13.9.1.17

Unplanned Releases Prior to the TMI-2 Accident (99)There were no unplanned releases reported to the NRC by the TMI-2 facility prior to the March 28,
1979, accident.

13-5813.9.1.18

Radioactivity Releases Since the Accident

(60,

100)

Since the accident, the only significant release of radioactivity was the controlled purging of
Kr-85 from the reactor building.
The total amount of Kr-85 released was about 43,000 Ci.
"Miniventings" may be required as long as radioactive waste water remains in the reactor buildingThis requirement is adressed in Sections 4 and 6.
unprocessed.
13.9.1.19

Releases of Radioactivity to River from Other Facilities (107,

112,

20)

The EPA Drinking Water Standards apply at the point of drinking water distribution regardless ofthe number of sources of discharge.
Therefore, safety of the drinking water is assured.
In
addition, the NRC subscribes to the EPA 40 CFR 190 requirement on radiation resulting from thenuclear fuel cycle, which includes the radiation dose received from all pathways, i.e., airborne,
liquid, and direct radiation, and which includes cumulative radioactive releases.
13.9.1.20

Precedents for Radioactive Liquid Releases from Operating Reactors(115)

Precedents have been established for the release of radioactive water into rivers.
Normal operating effluents are currently being released into the Susquehanna River and other bodies of water.
For example, the average amount of tritium released from a normal generating unit of TMI-2
size is 400-500 curies/year.
These planned releases are limited by the conditions of the operating license and must be reported to the NRC on a periodic basis.
The guidelines on these releases are such that the effects on fish, wildlife and, ultimately, man
are in compliance with the NRC design objective of limiting exposure to "as low as is reasonably
achievable."In the case of liquid effluents, the total annual quantity of liquid effluent mustbe such that a calculated dose to an individual in an unrestricted area is less than 3 millirems
to the whole body or 10 millirems to any organ.
13.9.1.21

Tritium Effluents and Releases(83)

The discussion of tritium evaporation during defueling is given in Section 6.
Tritium in theform of water vapor would be released into the reactor building from the spent fuel pool andtransfer canal.
The evaporation rate would depend on the surface area of the pool and transfer
canal (both of which are known) as well as on the humidity in the reactor building.
The expected
tritium concentration in the reactor building is not a health hazard to workers in the reactor
building.
Any release by purging to the atmosphere would'be carefully monitored and kept within
regulatory limits.
13.9.1.22

Data on Tritium Effluent Estimates (13)The alternative of discharge to the river along with the radionuclide inventory of the processedwater including tritum is presented in Section 7.2.
Dilution factors of 3400 are required for the radiological constraints and 3500 under the NPDES
criteria. Since 34,000 gpm of dilution water from cooling tower blowdown is available and a total of 140,000 gpm of dilution capability is available from other equipment on site, discharge
rates can be varied to satisfy Appendix I release limits at the discharge point, Primary Drinking
Water Standards at a downstream uptake for drinking water and the NPDES limits.
The dilutionthat satisfies the most limiting of these criteria will be used if
this alternative is
implemented.
13.9.1.23

Water Treatment at Intake (100)The municipal water treatment processes at the point of potable water intake do not of themselves
necessarily change the radionuclide concentrations in the river water.
Should the alternative ofcontrolled release of the processed accident water to the river be chosen, the control at the13-59discharge is the most effective way to keep radionuclide concentrations at intakes below EPA
Drinking Water Standards.
This alternative along with others are fully described in Section 7.2.
The staff concludes that processing alternatives exist such that if the processed water is permitted to be dispersed by controlled release into the river, the change in the radionuclide contents in potable water derived from downstream river water would be negligible.
13.9.1.24

Total Radionuclide and Concentration(100)

The PEIS presents concentrations of radioactivity measured in pCi/mL as well as the total radioactivity in curies.
Both units are used, depending on which are most appropriate for the topic
under discussion.
For example, when discussing the efficiency of a cleanup alternative or environmental pathways to man, it may be more important to give the radioactivity concentrations.
Onthe other hand, when discussing the radioactive waste liners and containers that have to be disposed of, the total radioactivity (Ci) may be a significant concern.
13.9.1.25

Cumulative Doses from Events at TMI-2 (100)The impact of the accident can be found in numerous other reports assessing the accident, forexample, NUREG-0558, "Population Dose and Health Impact of the Accident at the Three Mile Island
Nuclear Station" published by the NRC in May, 1979.
The radioactivity releases prior to theaccident from TMI were well within the technical specification limits.
When the licensee makes acleanup proposal, the review of the proposal will include the potential cumulative health effects
to the worker and the public which would result from that proposal.
This PEIS has bounded thesedoses for the cleanup, which are summarized in Section 10.2 and 10.3.
13.9.1.26

Airborne Effluents (32,55, 73,

83,

100,

116,

120)

The staff's estimate of the release of radioactive materials from cleanup activities to the environment considered several things, including:1) the fraction of processed material that becomesairborne, and 2) the fraction of airborne material that passes through an air cleaning system
without being collected.
A brief discussion of the air cleaning system, including HEPA filters,
is presented here to explain the staff's assumptions.
Tritium is an isotope of hydrogen that has chemical properties like hydrogen.
Thus it combineswith oxygen to form a molecule similar to that of H20, but that retains the radioactivity of thetritium. When the water is evaporated, all water molecules, including those with tritium, exist
in a gaseous state. And, like other gases, they pass through a filter system.
Other fission
products (and actinides such as plutonium or uranium) exist in the process water as dissolved
constituents.
When the water is evaporated, the dissolved constituents stay behind in the remaining liquid. The evaporated water in the gaseous state contains none of the radioactive constituents and passes through a filter system unchanged.
Evaporation is thus an effective method
for separating bulk water from a solution containing dissolved radioactive constituents, but it
is totally ineffective for separating tritium.
This difference in behavior must be recognized inestimating radioactive releases.
However, evaporation, as it is conducted in plant operation, is not a perfect operation.
Evaporation action creates turbulence at the surface which results in the formation of small dropletsof water that are physically entrained in the rising vapor.
The rising vapor contains only tritium, no dissolved radioactive materials; the entrained droplets contain both tritium and dissolved radioactive materials.
In a well-designed evaporator, the fraction of the total vaporizedmaterial that exists in the form of entrained liquid droplets seldom exceeds I x 10-5.
This
figure then forms the base for estimating how much dissolved radioactivity appears in the vapor
stream.
Following condensation, most of the entrained radioactivity appears in the condensate
with a small quantity remaining with. the uncondensed vapors.
About 1 x 10-3 of the radioactivity
in the condensate remains in the uncondensed vapors.
Thus, less than I x 10-8 of the initialradioactive material remains in uncondensed vapor.

13-60
Other processing operations are generally less energetic and less turbulent than evaporation.
In
ion-exchange operations, process liquids are pumped in closed pipes through an ion exchange column.
The fluids are relatively free of substantial turbulence.
However, even in these operations, a fraction of the process liquids (less than for evaporation) may form small droplets thatare physically entrained into surrounding airspaces, and are subsequently transported by theprocess off-gas system to the air-cleaning system.
En route, the liquid droplets evaporate,leaving the dissolved radioactivity as a solid aerosol in the air stream.
Entrainment for thiskind of operation seldom exceeds 1 x 1 0-4 of the processed liquid and becomes airborne and enters
the process off-gas streams.
This value is conservative when compared with the information presented-above and is compatible with data obtained from the processing of nuclear fuels.
Thelatter consists of a series of many individual operations, some of which are far more complex
than the few relatively simple operations envisioned for Three Mile Island.
The heart of the air cleaning system that pertains to the removal of particulate matter is the
HEPA filter.
In the PEIS, a penetration fraction of 1 x 10-3 was used for calculating effluentsand releases thru the HEPA filters, a value that is conservative relative to achievable values.
The failure of a single HEPA filter was considered as a credible accident in the PEIS, that is
why the staff requires two in series. The releases of a facility are constantly monitored withappropriate instrumentation.
Failure of a HEPA filter would be evidenced by an increase in releases by the monitoring sytem and would alert operating personnel to take corrective action.
The failure would have no effect on the functioning of the second filter in series with thefirst; thus containment integrity for the system would be maintained, but at a reduced level.
The design of the monitoring sytem detectors is such that the failure of a single filter is detected within 15 minutes.
No failures occurred during EPICOR-II processing of contaminatedwaters.
13.9.1.27

Potential for Airborne Releases from Fire (73)

The potential impact of fires considered to be even remotely credible has been addressed inappropriate sections of the PEIS.
An accidental fire involving resins or spent fuel is not considered to be a credible event.
The resins are contained in sealed steel liners, which aretransferred to the onsite waste storage facility in a steel and lead cask.
The storage facility
consists of modules with thick concrete walls, and a 16-ton sealed concrete cover over each cell.
The water is drained from the resinsprior to storage, but they remain wet (like wet sand).
In
such a condition ignition and combustion is highly unlikely.
The nuclear fuel cannot sustain combustion, and the spent fuel is stored in water, with double
containment used during transportation.
Fire involving spent nuclear fuel is considered incredible under the conditions at TMI.
Combustible materials and fluids are controlled administratively to assure that nuclear fuel orresins will not be involved in a fire.
13.9.213.9.2.1

In addition,

fire control equipment is available on site.

Occupational Doses and Health Effects
Doses to Workers (4,

32,

78,

115,

116,

120)

Occupational doses for cleanup operations were-estimated by the staff based on experience gained
from work done in the past which include parameters such as:* The manpower required to complete each task," The radiation field in which the work islikely to be done," The effectiveness of shielding and other radiations protection technique,* Optimal use of personnel to minimize overall personnel

exposure.

13-61The estimates have been conservative, mainly due to conservative estimates of radiation field andmanpower required to complete tasks.
The estimating process has proven to be reasonably reliable
for work done in the auxiliary and fuel handling building.
The auxiliary and final handling building exposure experience has been factored into the estimates for the reactor building.
The total radiation exposure through September 1, 1980, for the work effort in decontaminating
the AFHB is given as 146 person-rem in Section 5.1.5.1.
These values were obtained from analysis*of personnel dosimetry records.
Current radiation protection standards for occupational exposure have been developed over the
past 50 years. Most of the radiation standards were first proposed by the National Council on
Radiation Protection and Measurements (NCRP).
Other organizations which have undertaken evaluations of radiation exposure standards, particularly with respect to the cancer and genetic effects, include the National Academy of Sciences (NAS), the Medical Research Council (MRC) and theInternational Commission on Radiation Protection (ICRP).
In January 1957, the NCRP recommended amaximum permissible dose of 5 rem per year for radiation workers.
There is much information toshow no effect on humans exposed to 1-15 rem per year.
Therefore, it has been the judgement of
the NCRP that the potential risk from a 5-rem per year occupational exposure is small.
However,the NRC recognizes that the standards are based upon currently available information and should
not be regarded as permanent.
Accordingly, it has required that exposure to radiation be kept at
the lowest practicable level when and wherever possible.
The licensee is required to ensure that the occupational doses that its radiation workers incur
are as far below those specified in radiation protection regulations as reasonably achievable.
Whole body exposures are monitored and recorded, and in situations where extremity (e.g., hand)doses can be higher than whole body doses, these extremity doses are also monitored and recorded.
The licensee's radiation protection plan includes engineered safeguards and personnel access
control to ensure protection from unnecessary and/or excessive inhalation of radionuclides.
Respiratory protection is implemented whenever necessary.
In this regard, air sampling is also
performed to determine the level of any airborne contamination and document the adequacy of theengineered safeguards.
In addition, provision is made for routine bioassay samples on an appropriate schedule for the radiation workers and prompt sampling and evaluation in the case of asuspected radionuclide intake.
Blood cell counts are not generally taken because changes in cellcounts occur only following relatively large over-exposures.
The exposure rates associated withcleanup work at TMI-2 have not and would not be of this nature.
13.9.2.2

Radiation Levels for Workers After Shielding (50)The estimated radiation levels are for worker exposure after precautions such as shielding andremote operations have been incorporated.
However, these levels do not represent surface readings for filter cartridges or other components which may have localized areas of higher radiation.
13.9.2.3

Average Radiation Fields for Workers (50)The 10 mR/hr dose rate is not an average used for all PEIS cleanup activities.
It is, however,applied to all activities associated with reactor vessel head and internals removal and defueling.
Based on conditions found in the containment building during the reactor building entries
and decontamination achieved at other nuclear facilities, when the reactor building sump andbasement have been drained and limited cleanup completed, average exposure levels of 10 mR/hr are
believed to be achievable.
Average radiation fields are difficult to estimate; however, there isa general tendency to over-estimate.
13.9.2.4

Radiation Protection Plan (115)

A radiation protection plan to ensure radiation exposure to cleanup workers are below regulatory
limits and to keep those exposures as low as reasonably achievable (ALARA) has been submitted by
the licensee to the NRC staff for review and approval.

13-6213.9.2.5

Radiation Dose Estimates to Workers for Cleanup Alternatives(108)

Potential occupational radiation exposures to cleanup workers are evaluated for every cleanup
The potential health effects of worker exposures are summaalternative discussed in the PEIS.
rized in Section 10 of the statement.
13.9.2.6

Airborne Radioactivity Levels During Water Processing (55)

Although the airborne effluents prior
Water processing would take place inside enclosed systems.
to treatment by HEPA filters may exceed MPC levels, these effluents will not be exposed to building air and should not cause the airborne radioactivity levels outside of the system to exceed
MPC levels.
13.9.2.7

Worker Exposure Estimates for Reactor Coolant Sampling (32)Worker exposure is estimated to be 20-30 mrem per sample while taking reactor coolant samples.
It is recognized that during the accident, taking coolant samples resulted in some overexposures.
However, the radioactivity levels of reactor coolant are now orders of magnitude
lower than those during the accident because of substantial decays of shorter half-life radioisotopes.
Other factors also contributed to the high exposures when the samples were taken during the accident, e.g., urgency of the situation, high area radiation levels, lack of radiation
These conditions which existed during theshielding and proper radiation protection procedures.
emergency sampling have been corrected and would not be expected to recur during the cleanup.
13.9.2.8

Occupational

Dose Records (69,114, 115)

Questions were raised regarding records of occupational doses. The NRC requires all its licenWhenever an employee
sees to maintain complete radiation exposure records of its employees.
joins a licensee, the licensee is required to obtain a record of the new employee's exposure
history. Such a record is usually obtained by requesting exposure records from the employee's
former work places.
At TMI, the cumulative occupational dose information is entered into the
The licensee has a
Radiation Work Permits of each cleanup worker for entry to radiation areas.
This information is updated by thecomputer program to record the cumulative occupational doses.
TLD dosimeter readings of each cleanup worker and is maintained for both regular and-transient
Nonoccupational doses, e.g., medical exposures, are not recorded.
workers on a permanent basis.
13.9.2.9

Cancer Rates and Exposures for Workers (125)The
The staff uses well documented and accepted risk estimators in evaluating health effects.
staff estimates of the health effects occurring to the occupational work force, (the largest
environmental impact identified) would indicate a very small increase in the cancer rate thatwould normally be experienced due to causes other than occupational radiation exposure.
13.9.2.10

Doses to Workers and Effect on Gene Pool (67)

Radiation protection standards for radiation workers are higher than for the general public.
However, radiation workers are a small fraction of the population, so their contribution to theRadiation standards are designed to limit any intotal genetic pool is proportionately small.
crease in overall mutation rate in the total genetic pool to very small fractions of the normallyoccurring rate. Since radiation workers make up only a small fraction of the general population,
radiation doses for them can be somewhat higher than for the general population without causing a
It should be noted that many industrial processes
large increase in overall mutation rates.
other than work involving radiation also expose workers to mutagens.
13.9.2.11

Health Effects from Doses to Workers (34,50)

In section 10.2.2, the estimate of 131 fatal cancers per one million person-rem is for people in
The cleanup workers involved constitute the assessed population.
The
the worker age group.
135 fatal cancers per one million person-rem is based on a lifetime in the general population.

13-6313.9.2.12

Occupational Doses Due to Tritium (79, 85)Processed water containing tritium has been used in decontamination at TMI-2 to reduce the volume
There is an occupational exposureof tritiated water that must be stored for eventual disposal.
associated with the use of the tritiated water.
Prior to the use of processed (tritiated) water in the water jet for the auxiliary and fuel handling building decontamination, tests were conducted by the licensee to determine the tritium
level in the atmosphere. These tests were done with little ventilation and indicated that thetritium level in the atmosphere during the use of the water jet was approximately 10% of the
The water jet tests are expected to be bounding
allowable airborne concentration (10 CFR 20).
on airborne tritium concentrations; other activities, such as wet mopping or hose washing, areDuring future activities, monitoring of airborne
expected to have lower airborne concentrations.
tritium will be done as appropriate.
When compared to the estimates of the occupational dose due to the radiation fields produced by
the contamination on the surfaces, the dose from tritium would be a small increment in the estimate of the occupational dose incurred during the decontamination efforts.
13.9.2.13

Radiation Exposure from Waste Storage Facility (100)The exposure rate indicated in the PEIS for the waste storage facility is less then 0.5 mR/hr at
the facility boundary, not the site boundary. The site boundary and personnel working areas arewell removed from the waste storage facility boundary. The radiation levels at the site boundary
will be maintained within regulatory limits.
13.9.2.14

Effect of Schedule on Occupational Exposures (115)

An extension of schedule in itself does not necessarily imply that the level of effort has increased, or that personnel doses or total dose will increase. An extension in schedule may have
many causes and those that have occurred at TMI have generally been because of the need for licensing approvals and extra time spent to assure that the best approaches are used. Problems
have been encountered which have caused delays, but these have not generally increased total or
In fact, even though manpower estimates have increased, experience
individual worker exposures.
at TMI has shown that exposure values have generally compared well with predictions.
13.9.3

Offsite Doses & Health Effects

13.9.3.1

Offsite Dose Criteria (20,85, 52, 80, 64)

10 CFR Part 50, Appendix I requires that reactors be designed so that the annual dose commitment
to the maximum exposed individual offsite does not exceed 5 mrem total body or 15 mrem skin dose
Dose predictions made prior to the Kr-85 purge indicated that the Appenfor gaseous releases.
dix I design objectives would not be exceeded during the purge. Monitoring during the purge
10 CFR Part 50, Appendix I also
process confirmed that the dose predictions were conservative.
requires that reactors be designed so that the annual dose commitment to the maximum exposed
individual offsite not exceed 3 mrem for liquid pathways. Dose estimates made in this PEIS indicate that over the course of the entire decontamination process, including the discharge of processed water to the river, if proposed and authorized, feasible alternatives exist such that theCriteria regarding offAppendix I dose commitment through liquid pathways will not be exceeded.
site doses recommended for the cleanup operation are described in Section 1.6. These criteria
will assure that the releases are made in such a manner to conform to the Appendix I design
objectives.

13-6413.9.3.2

Offsite Dose to be Kept as Low as Reasonably Achievable (ALARA)

(52,

109)

The doses that will occur as a result of the cleanup operation are summarized in Table 10.3-1.
The numeric values in this table are believed to be conservative and represent upper bound values.
The basis for this conservation is described in Appendix W of the PEIS.
As described in the PEIS,
the basis used in approving programs will depend partly on the success of earlier programs inmeeting the ALARA principle (10 CFR 20).
Further, the licensee will be required to periodicallyreport to NRC estimates of offsite doses resulting from the decontamination operations.
If aprogram is resulting in offsite doses significantly larger than originally estimated, the NRC canrequire modification as appropriate.
13.9.3.3

Dose Estimates for Future Cleanup Alternatives are Projections(20)

The dose estimates in the PEIS for future cleanup alternatives are projections because the releases have not yet been made.
Actual measurements will be made when releases occur.
A summaryof the projected estimates and the associated potential risk probabilities can be found in Section 10 of the statement.
13.9.3.4

Comparison of Cleanup Impacts with Other Sources (116)

Section 10.3 provides additional information on the comparison of risks of the cleanup with othertypes of human activity.
13.9.3.5

Comparison of Cleanup Doses to Background Radiation (99,100)

The potential offsite radiation doses resulting from cleanup activities and the naturally occurring background radiation were compared to illustrate the relative significance of radiation doses
resulting from the cleanup.
It was suggested that the comparison should be made because of thedifferent exposure pathways and types of radiation involved.
Potential health effects from wholebody dose or dose to certain organs are independent of the pathway.
The dose conversion factors
used in the assessment were based on the different radioisotopes and types of radiation involved.
The result was an equivalent effective dose value.
13.9.3.6

Dose to Maximumly Exposed Individual Considered(60)

The dose estimates discussed in the PEIS represent the potential dose to the maximumly exposed
individual, as well as the average dose to the population from the cleanup alternatives.
A discussion of the models and the methods used to calculate these values are provided in Appendix W
to the PEIS.
13.9.3.7

Offsite Locations and Radiation Doses (67)The offsite dose calculation of a few millirem as a result of all cleanup activities is estimatedfor the individual residing at the most critical location.
It has been suggested that the farther
away from the TMI site, the better off the individual may be.
This, however, needs to be qualified, because it would depend on where the individual goes.
Natural background radiation levelsin the U. S. vary considerably (from 70-310 mrem/yr), and in many locations in the U. S., thebackground radiation would be higher by more than the few millirems which would result from thecleanup plus the natural background radiation (116 mrem/yr) at the critical location near TMI.
13.9.3.8

Offsite Dose Fluctuation Ranges (32)The actual offsite monthly dose rate could be different from the values calculated because thecalculations are based on annual average conditions.
The differences would not be expected to begreater than a factor of "four" which is acceptable for these low doses.

13-65The background radiation levels reported in the PEIS do not include man-made radiation exposures,such as radiation exposures due to medical diagnosis.
In addition, it has been found that theradioactivity releases due to nuclear power generation are not significant enough to impact on the
average range of background radiation levels reported.
13.9.3.9

Offsite Contamination from Cleanup (37)

The NRC appreciates the concerns of people who live in communities around TMI, and agrees that thebenefits of nuclear power would be questionable if nuclear waste did result in widespread contamination of farm lands or cities.
However, water, air, and vegetation are being monitored and there
is no evidence of contamination which would cause radiation doses which are more than a very small
fraction of the dose from natural background radiation (due to natural radioactivity in soil,
water, air, the human body and from cosmic radiation).
Also, instrumentation is available withwhich radioactivity can be traced and measured, so contamination in the environment would be
detected and mitigative actions taken long before it could become a serious problem.
13.9.3.10 Drinking Water Standards and Immersion Dose (67,

72,

79)

In addition to the reference to Drinking Water Standard, the cleanup operation also has to meetthe numerical design objectives in Appendix I to 10 CFR Part 50, which state that the dose to anindividual offsite should not exceed 3 mrem total body through all liquid pathways.
This would
include the dose from bathing and other use of the Susquehanna River for recreational purposes.
In the PEIS, the dose estimates consider all the major pathways.
The assessment in the PEIS is based on calculations of radioactivity concentrations in the waterand the potential radiation dose to the maximumly exposed offsite individual.
The conclusions in
the PEIS are based on these dose pathway calculations, and not on the EPA Drinking Water Standard,
although meeting the standard would be a condition to be considered for processed water release.
13.9.3.11

Calculation of Dose from Drinking Water (50,52, 55,85)

Doses are calculated, for the nearest downstream drinking water intake at PP&L's Brunner Island
Station approximately five miles downstream.
An average river flow of 12,600 cfs is used for
dilution. Calculation models and parameters used in estimating these doses are discussed in Appendix W. The assumption that the effluents will be well mixed at the drinking water intake isconsidered to be realistic.
13.9.3.12

Dose Rate at Site Boundary (52)

Proposed criteria to adopt Appendix I would require offsite doses due to airborne releases to beless than 5 mrem/yr whole body dose and less than 15 mrem/yr skin dose.. Doses due to liquid pathways would have to be below 3 mrem total body dose.
The direct radiation level at the fence
surrounding the interim storage and staging facility will be less than 0.5 mrem/h.
The direct
radiation level at the site boundary will be substantially smaller because of decrease due todistance.
13.9.3.13

Tritium Dose through Liquid Pathways (66)

Tritium has its primary importance in the drinking water pathway as opposed to the fish consumption pathway because the fish only equilibriate with the tritium concentrations in the water and*do not concentrate this radionuclide.
The concentrations of tritium in the fish would be the sameas in the drinking water.
13.9.3.14

Errors in Drinking Water Calculations in Draft PEIS (76)The staff has completely revised the calculations on drinking water doses, incorporating new
information made available by the licensee and other sources.
These calculations are indicated inAppendix W and the results are indicated in Section 7.2.

13-6613.9.3.15

Environmental

Impact During Normal Power Operation (46)The staff concludes in this PEIS that the cleanup at TMI Unit 2 could be accomplished with offsite
radiation doses to the public no greater than those assessed for the normal operation of TMI-2
prior to the accident.
The environmental impacts associated with the normal operation of the TMI
Unit 2 are described in "Final Supplement to the Final Environmental Statement Related to the
Operation of Three Mile Island Nuclear Station, Unit 2," (NRC report NUREG-OII2, December 1976).
13.9.3.16

Tritium Release Due to Reactor Pressure Vessel Head Removal(76)

The offsite dose from tritium released from the reactor vessel during head removal would be extremely small, much smaller than that now occuring whenever the reactor building is vented, because the sump water constitutes a much larger source of tritium.
Special precautions against
tritium release during head removal would not be warranted.
13.9.3.17

Offsite Dose from TMI-2 Accident Taken Into Consideration (70,

93,

109,

114,

67,

128,

20, 85, 84)
The dose and potential health impacts that occurred as a result of the accident and the methodology behind these dose calculations are contained in NUREG-0558, "Population Dose and HealthImpact of the Accident at the Three Mile Island Nuclear Station." The offsite doses from thecleanup operation will be negligible (few mrem) in comparison to that from the accident -(less than
100 mrem).
13.9.3.18

Basis for Dose Estimates to Base on Major Isotopes (4,

73,

84)

Appendix J presents the basis for considering only a few of the radionuclides present (is H-3, Cs137, Cs-134, Sr-90, and Sr=89) in the source spectrum for offsite dose calculations.
By use ofthe existing concentration in the facility and MPC values of 10 CFR Part 20 (which incorporate
usage factors and ingestion dose conversion factors for water, and inhalation rates and inhalationdose conversion factors for air) the relative significance from the standpoint of dose of eachnuclide in the entire spectrum is estimated.
13.9.3.19

Justification of Dispersion Factors in Dose Calculations (100)For the application to operate the Unit-2 facility, the licensee was required to establish both
long-term and short-term meteorological dispersion data. These data were reviewed by the NRC
staff and described in the Final Environmental Report and Safety Evaluation Reports related to theoperation of the TMI-2 facility. The NRC staff has also determined that meteorological conditions
presently existing in the vicinity of TMI-2 do not differ significantly from those described inthose reports.
13.9.3.20

Conservatism in Dose Rate Estimates (52,55, 84,107)

The decontamination program is scheduled over a period of several years, hence the approach usedin the Draft PEIS of evaluating the total release as if it occurred over a period of one year
actually overestimates release rates and is overly conservative.
Meteorological parameters specific to accidents are now used for these calculations in the PEIS (see Appendix W).
13.9.3.21

Radionuclide Pathways to Humans (72,68)

The nuclides which could be released would accumulate in the body through inhalation of the plume,and by ingestion (drinking water, eating fish, meat and farm produce, and drinking milk, etc.)
These pathways are considered in the dose calculations.

13-6713.9.3.22

Dose Commitment to Humans Through Aquatic Food Chain (67,

13, 55)All the doses that were calculated in the PEIS include accumulation in the food chain in the final
receptor next to man, e.g., fish and shellfish. Also, all of the internal doses that were calculated for human beings includes the "50 year dose commitment effect." This means that the onetime intake includes the dose that is received from the one-time intake out to a period of 50
years.
13.9.3.23

Meteorological Dispersion Factor (55)The value of meteorological dispersion factor (X/O) which was used throughout the text for estimating maximum individual doses due to gaseous releases during normal cleanup activities was
6.7 x 10-6. This X/Q value represents the highest annual average X/Q value for all the 16 sectors
at the site boundary.
13.9.3.24

Estimates Made for Organ Dose (32, 55,

121)

The PEIS contains estimates of critical organ doses and their associated health effects.
13.9.3.25

Offsite Radiation Doses by Inhalation Considered (120)People will absorb radioactive materials such as Sr-90 and Cs-137 by breathing if they are downwind of the plant and these materials are being released to the atmosphere. The dose calculationsof the PEIS include this pathway.
13.9.3.26

Dose Factor per Unit Activity Released (73)

The population dose estimates indicate that for every curie of Sr-90 released uniformly over one,year to the atmosphere, the resulting 50-mile population dose is about 53 person-rem. For eachcurie of Cs-137 and Cs-134, the 50-mile population dose is 33 person-rem and 40-person rem,
respectively.
13.9.3.27

Tritium Dose Considerations (66,114, 115,

117, 130)

Tritium behaves differently from other isotopes in that it becomes distributed in the body in amanner that is proportional to the water content for each organ. As a result, except for bone,
the dose that each type of organ tissue receives is about the same for a given amount of activity
taken in by the body. Other isotopes generally have varying affinities for the different bodyorgans. Regarding its effect on tissues, tritium is a weak (18 Kev) beta emmitter, therefore, its
effect will be fairly local and inter-organ effects are small to negligible.
Tritium can often be a problem from the engineering standpoint because it is chemically indistinguishable from the hydrogen in water. Therefore, it is not removed in filtering or ion-exchangesystems. For this reason it is usually given special consideration, rather than because of any
special behavior within the body. Except for its uniform distribution within the body, it is notexpected to behave any differently from the standpoint of dose effects than any other beta
emitter.
13.9.3.28

Exposure to the Public from Waste Disposal Operations (107)The dose to people living near the disposal sites has not been calculated for specific TMI waste
shipments since the TMI waste does not add to the waste disposed of at the facility; rather it
replaces other waste shipments. Thus, the TMI waste does not add to the exposure received by the
population in the vicinity of those facilities.

13-68The performance objectives for a waste disposal facility will be (1) doses to inadvertent intruders will not exceed 500 mrem/yr and (2) doses from groundwater pathways will not exceed
25 mrem/yr at the disposal site boundary.
13.9.3.29

Global Dose Commitments (120)This is generally conservative as the
In the PEIS the 50-year dose commitments were estimated.
average age of the human population is over 20 years and the life span of a person is 70 years.
Estimating dose commitments beyond this does not realistically reflect the age structure of theExtrapolating population doses beyond 50 miles is meaningless because they are so
population.
small that not only are they greatly exceeded by natural background doses, but they are exceededby fluctuations in natural background doses.
Calculations of global doses due to releases at TMI serve no useful purpose as the all-time backIf it is assumed that the 13 person-rem dose isground dose to the qlobal population is infinite.
received by the 4 x 109 people who are now alive, each person would receive an average dose of
A dose of this value has no meaning.
3.2 x 10-6 mrem.
13.9.3.30

Radionuclide Concentrations and Population Doses (73)

In fact, individual doses are
Population doses are directly related to activity that is released.
Nevertheless, both doses are inversely related to factors
also directly related to activity.
The calculational models used by NRC work
related to dilution; for example, the river flow.
The assumptions described in the PEIS indicate that both activity andexactly in this manner.
There are at least three reasons for discussing dilution
river flow are used in the calculations.
First, there are federal regulations requiring that certain concentrations
and concentrations:
Low dose rates are
Second, the greater the dilution, the lower the dose rate.
not be exceeded.
desirable because basic health physics principles suggest that, for the same cumulative dose,Third, the closer that
biological tissue is less affected by lower dose rates than higher ones.
concentrations and dose rates are to background concentrations and dose rates, the less chance the
The ideal situation is to keep release levels
release can perturb anything in the ecosystem.
The approach that is used in discussing these
within the natural fluctuations of the ecosystem.
matters in the PEIS is not intended to create false impressions, rather, it is to present theComparing doses toreleases from a perspective that the majority of the public will understand.
background, and concentrations to federal standards, seems a reasonable approach.
13,9.3.31

No Basis to Declare Region Around Unsafe (73)Based onThe potential environmental impacts of cleanup alternatives were evaluated in the PEIS.
the potential environmental impacts discussed in the PEIS, there is no basis for declaring theregion around TMI unsafe for human habitation nor that agricultural products from the area bedeclared unfit for human or livestock consumption.
13.9.3.32

Too Many Units Used for Radioactivity (121)

It is necessary to use several different units in the PEIS when referring to radiation since difThe units which describe these aspects have beenferent aspects of radiation are being treated.
adopted by the International Commission on Radiation Units and Measurements (ICRU) Report 33 Radiation Ouantities and Units.

13-6913.9.413.9.4.1

Health Effects from Radiation Exposure
Health Risks Associated with Radiation Doses During Cleanup (2,20, 59,60, 67,

69,

82,

91)
Several studies analyzed the data relating to the influence of the magnitude and temporal distribution of dose on the biological effectiveness of low linear energy transfer (LET) radiation perunit absorbed dose.
All of the isotopes that could be released as a result of cleanup activities
are of this category.
The National Council on Radiation Protection and Measurements (NCRP) concluded in Report No. 52,1 "that there has been no direct demonstration of deleterious effects
from worldwide fallout and therefore from environmental Cs-137.
While this lack of evidence does
not prove a total absence of damage associated with internal exposure to Cs-137, it does indicatethat risks due to present and past levels of environmental contamination of Cs-137 have been atlevels so low that harmful effects, if they exist, are not readily apparent."
Also, in the summary and conclusion of BEIR-III Report 2 the committee concluded that:"It is unlikely that carcinogenic and teratogenic effects of doses of low LET radiation administered at this dose rate
(100 mrad/yr) will be demonstrable in the foreseeable future." Furthermore, it is likely thatthe risk models from low-level radiation are conservative, in fact, certain studies with animals
suggest the existence of a threshold below which there are no adverse health risks. Hence, it
may be that there are no health effects, detrimental or otherwise associated with these releases.
The PEIS describes health risks to the plant workers and to offsite individuals and population.
The regulations, 10 CFR Part 20, require that an individual plant worker receive no more than an
average of 5000 mrem per year occupational exposure.
This is equivalent to a lifetime chance of
7 x 10-4 for the worker to die from cancer as a result of the radiation exposure based on calculational methods described in the PEIS.
The risk of mortality from lung cancer due to smoking
one pack of cigarettes per day is 5 x 10-3 per year and the risk of mortality from driving an
automobile 10,000 miles per year is 7 x 10-5 per year.
There are other examples that could be
used but these are sufficient to illustrate that the greatest threat to plant workers is not fromradiation exposure.
13.9.4.2

Up-to-Date Information on Health Effects (99)

At the time the draft PEIS was written, the Committee on Biological Effects of Radiation had notreleased its 1980 draft report (BEIR-III) 2 . However, in the Summary and Conclusions of BEIR-III,
the committee concluded that low-level radiation derived from the earlier BEIR reports are probably conservative.
13.9.4.3

EpidemiologicalStudy on Low-level Radiation (10,60, 64,72, 107)

In a study funded by the NRC and reported in NUREG/CR-1728, the feasibility of epidemiologic investigations of the health effects of low-level ionizing radiation was investigated.
In the summary to NUREG/CR-1728, it was concluded that occupational groups are the most practical to study
and that it is not feasible to study the dose-response effect for cumulative doses that are lessthan 10,000 mrems.
The maximum offsite dose resulting from the accident was about 100 mrem.
Theannual maximum dose from TMI-i, should it be restarted, will be about 10 to 20 mrem, and the
maximum dose from the decontamination program will be a few millirems.
If TMI-1 operates for

INational Council on Radiation Protection and Measurements, "Cesium-137 from the Environment to
Man: Metabolism and Dose," Report No. 52 (1977).
2
Committee on the Biological Effects of Ionizing Radiation.
"The Effect on Populations of Exposure to Low Levels of Ionizing Radiation," National Academy of Sciences, 1980.

13-7030 years, the cumulative offsite maximum dose will be about 300 to 600 mrem. The cumulative
30-year dose from the accident, operation of Unit 1, and the decontamination program is 400 to
700 mrem, a number much smaller than the 10,000 mrem lower limit study feasibility number.
Hence, a detailed epidemiological study around TMI would not likely detect any health effects
that could be related to doses from TMI.
NRC is not requiring the licensee to do these types ofstudies.
13.9.4.4

Detection of Radiation Health Effects (67,

72, 128)

The amount of radioactivity that will be released will be so low that no effects will be observable. The radioactivity released will result in doses of a few millirems to ten millirems overthe entire decontamination period (several years in length). The average individual doses forthe 50 mile human population will be much smaller than these maximum individual calculations.
Natural background radiation is about 116 mrem per year in the Harrisburg area. One study suggests that doses of 10,000 mrem would be required before epidemiological effects could be detected in humans.1
13.9.4.5

Approach to Estimate Health Effects (67,121)

To estimate the health effect due to cleanup alternatives the PEIS adopts the "linear nothreshold" approach. Thus the health effect is estimated to be linearly proportional to the dose
received. That is, doubling the dose would double the probability of potentially harmful healtheffects. Also, there is no threshold dose below which there is zero probability of potential
health effect. Therefore, although the radiation dose to the public from cleanup alternatives isvery small, in the order of a few millirems, a correspondingly small probability of health effectstill exists. This is in the order of less than one in one million.
13.9.4.6

Risk Estimators for Radiation Doses (73, 98, 114, 125)

Just as there are studies suggesting that risk estimators should be higher than the ones thatwere used in the PEIS calculations, there are also studies suggesting that they should be lower.
For example, people who live in Colorado, Wyoming, and New Mexico receive considerably more background radiation than the average U. S. citizen. Nevertheless, their leukemia rates are lower
than that of the average U. S. citizen. 2 The risk estimators that were used in the PEIS werebased on the BEIR-I report. 3 The latest BEIR report (BEIR-III) suggests that the risk estimators
should be even smaller than those of BEIR-I.
13.9.4.7

Accuracy of Dose-Risk Estimates (50)Cancer risk estimates are not accurate to 20%, therefore rounding off is justified. The meaningof cancer risk estimates was described in BEIR-II in the following manner: "As suggested by the
ICRP, the expression of risk estimates in absolute terms--for example, 2 cases per one million
exposed people per year per rem--might be misinterpreted as implying considerably greater accuracy than the facts justify. For this reason, estimates are sometimes expressed in terms of'orders of risk,' e.g., 1 to 10 cases/106/year/rad is a 6th order risk."

IThe Feasibility of Epidemiological Investigations of Health Effects of Low Level Ionizing
Radiation," NRC Report NUREG/CR 1728, Nov. 1980.
2
Cohen, Bernard L., "The Cancer Risk from Low-Level Radiation," J. Health Physics, Vol. 39,
pp. 659-678, 1980.
3
Report of the Advisory Committee on the Biological Effects of Ionizing Radiations, "The Effects on Population of Exposure to Low Levels of Ionizing Radiation," National Academy of
Sciences, p. 99, November 1972.

13-7113.9.4.8

Lifetime Health Effect Estimated (67)The dose estimates given in the PEIS are the total doses that the maximum individual could receive during the entire decontamination process. This means that if the decontamination lasted
longer than one year, the total dose received would be the same as if the decontamination lasted
only a year. The risk estimates stated in the PEIS are the lifetime risks based on a one-year
exposure dose.
With regard to the data base used to determine risk estimates in the PEIS, most of the data camefrom human beings. These estimates are based on studies of the survivors of the Hiroshima andNagasaki bombings, and also from other individuals exposed occupationally, e.g., in uranium
mining or as a result of medical diagnosis or therapy. With the Hiroshima and Nagasaki populations we are able to observe a wide spectrum of ages, from exposed fetuses to older people. In
this manner, the dose effect on all ages are taken into consideration.
13.9.4.9

Offsite Health Effects To Be Essentially Non-existent (120)

Risk comparisons in the PEIS are made in an attempt to quantify the impact. What the PEIS suggests is that the risks associated with this decontamination program are within levels considered
normal.
The statement that health effects are essentially non-existent is based on a comparisonof the calculated doses to background. This is not inconsistent with the quantitative estimates
made farther on, because the quantitative estimates result in negligible risk.
13.9.4.10

Controversy of Health Effects at Low Radiation Levels (64)The low levels of radiation which may be received by individuals living close to the plant were
estimated to be from one to a few millirem. There are no risk models suggesting that this levelof radiation is harmful.
The controversy exists at much higher levels of radiation and in calculating population risks from very low radiation levels. Nevertheless, the individual risk isinsignificant.
The following table, which lists the sources of radiation and associated doses for a typical individual living on the east coast, illustrates the insignificance of a dose of a few millirem.
Sources of the Average East Coast American's
Annual Intake of Radiation
SourceDose(mrem)NaturalSky
Housing
Food
Ground
Air

35
34
25
11
5

Total Natural Sources

110

Man-made
Medical and Dental
Weapons Fallout

41
4

Total Man-made Sources
Total All Sources45
155"Nuclear Radiation and Health" by Roger E. Linnemann, M.D.;
The Environmental Impact of Electrical Power Generation:
Nuclear and Fossil" by Pennsylvania Department of Education.

13-72
The quantities in the table can vary significantly from one individual to the next.
For example,
the sky dose depends on the amount of time spent outdoors, the housing dose depends on the typeof construction materials used for the house, the food dose depends upon the quantity of food ingested, and the medical dose depends on the amount and type of treatments an individual receives.
The variations are far above the doses that could occur as a result of the cleanup operation.
13.9.4.11

Synergistic Effects and Health Effect Estimates (121)The potential increases in fatal cancer assessed in the PEIS does take into account synergistic
effects in the same way as the normal cancer death rate of 1 in 5. We cannot state exactly what
the synergistic effects are in either case, but since the risk estimators used are based on real
population groups, the synergistic effects, if any, are correctly factored in.
13.9.4.12

Potential Health Risks Estimated (111)

The PEIS includes estimates of potential doses and associated health effects and environmental
impacts.
From these analyses, the staff concludes that the cleanup can be accomplished by using
alternatives having such radioactivity release levels that potential effects on the health of the
public and the safety of the environment would be negligibly small, if not zero.
13.9.4.13

Risk Factors for Non-fatal Cancer (73,121)

The risk estimators used in PEIS represents rate of lifetime cancer mortality per unit radiation
dose.
These risk estimates were developed in WASH 1400, Appendix VI.
Appendix III of WASH-1400
provides an estimate of non-fatal cancer risks. The risk for non-fatal cancers is about
1.5 times that of fatal cancers.
The PEIS concludes that no significant health effects and risks
were expected as a result of decontamination of TMI-2.
This conclusion would have been the same
if non-fatal cancer risks were included.
13.9.4.14

Hospitalin Vicinity to Treat Radiation Patients (99)

It is current NRC policy that there be at least one hospital in the vicinity of a nuclear power
plant that is able to treat patients either contaminated with radiation or suffering from overexposure to radiation.
In the case of TMI it is the Hershey Medical Center in Hershey, Pa.
13.9.4.15

Natural Background Radiation(22,'52)

The natural background radiation includes sources from cosmic and terrestrial radiations.
Itdoes not include man-made sources such as radiation from fallout, medical and dental X-rays.
Nuclear power operation radioactivity releases are not significant enough to impact on the natural background radiation used in the PEIS.
13.9.513.9.5.1

Radioecological Effects
Radiation Doses and Effects on Ecological Community (46,59, 68,75, 117,130)

Doses to animals in the vicinity of Three Mile Island are not expected to be significantly
greater than that received by humans.
Although guidelines have not been established for acceptable limits for radiation exposure to species other than man, it is generally agreed that thelimits established for humans are also conservative for other species.
Experience has shown thatthe maintenance of population stability is crucial to the survival of a species; species in mostecosystems suffer rather high mortality rates from natural causes.
While the existence of extremely radiosensitive biota is possible, and while increased radiosensitivity in organisms may
result from environmental interactions with other stresses (e.g., heat, biocides, etc.), no biota
have yet been discovered that show a sensitivity (in terms of increased morbidity or mortality)
to radiation exposures as low as those expected in the area surrounding the Three Mile Island nuclear power plant. -Furthermore, in all the plants for which an analysis of radiation exposure to13-73

biota other than man has been made, there have been no cases of exposures that can be consideredsignificant in terms of harm to the species, or that approach the exposure limits to members of
the public permitted by 10 CFR 20.1 The BEIR Report 2 (which is still being reviewed by the NRC
staff) concluded that the evidence to date indicates that no other living organisms are very much
more radiosensitive than man. Therefore, no measurable radiological impact on populations ofbiota is expected as a result of the decontamination operation of this plant.
Long-term bioaccumulation is generally not considered as important in animals as in humans because humans generally have the greater longevity. Thus long-term problems are most significant
in humans. The dose calculations for humans represent the 50-year cumulative dose based on a
one-year intake of radioactivity.
13.9.5.2

Studies of Plants and Animals (64,88, 123)

In Section 3.1.5 in the PEIS is a summary description of the ecology of the Three Mile Island
vicinity and other areas potentially affected during cleanup operations. In-depth specific accounts of the aquatic ecology and fisheries of the Susquehanna River and Chesapeake Bay are included in Appendix E. Appendix E was prepared from an extensive review of the literature, withthe most pertinent sources cited and included in the references.
No experimental research per se on the ecosystems of the TMI vicinity has been done specificallyfor the PEIS. However, studies of the aquatic ecology and fisheries of the Susquehanna Rivernear TMI have been ongoing continuously since 1974. These studies give us the description before
the accident. These are summarized in PEIS Section 11.10.
The NRC staff has carried out post-accident examinations of plant and animal health effects in
the TMI area. The results have been reported in two NRC technical reports. NUREG-0596 examineddisease, parasites, abnormalities, and mortality of fish in the Susquehanna River following theaccident and compared the figures with data for corresponding time periods during the preaccident years of 1975-1978, and with data from other upstream and downstream areas of the river.
NUREG-0738 investigated many reported health problems with plants, animals, and livestock following the accident. That report summarizes the investigations of terrestrial biologists, veterinarians, and a radiobiologist.
Studies of aquatic and terrestrial populations near TMI are continuing under the NRC Environmental Technical Specifications. The aquatic biological and fisheries studies are summarized in
Section 11.10 and in Appendix E. The detailed studies of the river biota occur in the York Haven
Pond near TMI. Sampling points in the pond are upstream and downstream of TMI and in all threeriver channels.
The studies have documented no ecological or animal health impacts from the accident. The studies do not include areas "indisputably" beyond the range of accident effects (had there beenany).
To include such locations could mean sampling in areas far removed (i.e., several miles
upstream) from TMI where the biotic community and its habitat are different from those control
stations. However, data from the post-accident period and from the ongoing cleanup period havebeen and will continue to be compared with comparable data collected during the 5 year preaccident period of 1974-1978. Those pre-accident ecological and animal health data serve asbaseline "control" information against which to compare similar data during cleanup.

iBlaylock, B. G. and Witherspoon, J. P., "Radiation Doses and Effects Estimated for Aquatic Biota
Exposed to Radiative Releases from LWR Fuel-Cycle Facilities, Nuclear Safety, Vol. 17, No. 3,
May-June 1976.
2
Committee on the Biological Effects of Ionizing Radiation.
"The Effect on Populations of Exposure to Low Levels of Ionizing Radiation," National Academy of Sciences, 1980.

13-7413.9.5.3

Effects on Ecology of River and Bay (11,20, 46,79, 91, 99,

103, 121,

123)

The potential impacts to the ecology of the Susquehanna River and Chesapeake Bay are discussed indetail in Section 7.2.5.4 of the PEIS. The staff carefully considered potential biological concentration factors as well as dilution and dispersal of the radionuclides, bioaccumulation factors, and the radionuclides in sediments when evaluating the potential impacts to the river ofthe radionuclides released in the processed water. The staff's original conclusion that the release of processed water is not expected to have adverse environmental impacts is still valid.
It is not true that any input to an ecosystem must have an irreversible effect. Indeed, it is
one of the basic tenets of ecology that ecosystems exhibit homeostasis and are capable of selfmaintenance and self-regulation if perturbations (such as input of pollutants) are not too
severe (E. P. Odum, 1971, Fundamentals of Ecology, 3rd ed., W. B. Saunders Co., Philadelphia).
The discharge of processed water to the Susquehanna River is not "assumed" but is discussed as
one alternative means of disposing of the water containing tritium, which is the only radionuclide of consequence. There are discussions of accidents which, if they should occur, would
result in processed or unprocessed water being released to the river. Concentrations of radionuclides in fish and shellfish as a result of both controlled releases of processed water are
discussed in Section 7.2.5.4 of the PEIS.
The biota of the Susquehanna River and the Chesapeake
Bay would receive a small fraction of the radiation dose received from natural background, andthe incremental effect would be unmeasurable.
Marketability of seafood products from the Chesapeake Bay is also discussed in Section 7.7.5.5.
13.9.5.4

Bioaccumulation Factors in Fish Considered (130)The bioaccumulation factors are factored into the calculations of radioactivity in fish flesh andthe doses to fish. The approach that is used in calculations is one of conservative averaging.
For example, the bioaccumulation factors selected are the ones believed to be conservative.
In
addition, conservative flows were used and no credit is taken for sedimentation.
The overall modeling approach has been used at many reactors. Follow-up field studies confirm
that the approach is conservative as a whole. This is not to say that a given individual fish
may not have a greater nuclide burden than the calculated value, but only that it is very unlikely that the average value exceeds the calculated value. Estimates of bioaccumulation of strontium and cesium in fishes of the Susquehanna River based on actual field data are given in
Section 7.2.5.4.
13.9.5.5

Bases for Bioaccumulation Factors (13,72, 91, 100)

The rationale for bioaccumulation factors used by NRC is given in a report by Stanley Thompson
et al., "Concentration Factors of Chemical Elements in Edible Aquatic Organisms," Lawrence
Livermore Laboratory, October 10, 1972. For strontium uptake in freshwater fishes two references
are cited. One reference indicated an average concentration factor of 3-.0 and another indicatedone of 16. For cesium uptake in freshwater fishes six references are cited with values ranging
from a low of 390 to a high of 4700 with a weighted average slightly greater than 2000.
The bioaccumulation factors for Sr and Cs have been observed to be greater for freshwater fishes
than for saltwater fishes. The freshwater fish values are used throughout the PEIS. Temperature
and water quality effects are only indirectly considered as the original sources are based on experiments over a range of conditions.
Backup studies at nuclear plant sites indicate that themodels in which these factors are used are conservative.

13-7513.9.5.6

River Dilution and Bioaccumulation of Radionuclides in Fishes (13,76)

Severe bioaccumulation of radionuclides resulting from fish living for a long period in an undiluted release plume is highly unlikely for several reasons. First, the flow that was used inthese calculations in the draft PEIS was the mean daily low flow. The licensee will be aware of
the river flow prior to releasing radionuclides and will be required to limit actual releases totime periods when the river flow is adequate to disperse the nuclides. Even if the releases weremade randomly it would be fairly unlikely for the flow to be only at the low level used in thecalculations. Secondly, the plume is expected to mix with the river water to some extent. The
amount of mixing that occurs depends on several factors, including the flow in the river, therelative temperature difference between the discharge water and the river water, and the exit
velocity of the discharge. Thirdly, and most importantly, organisms do not instantaneously attain equilibrium with the radioactivity in the water. Equilibrium is attained only after several
days or tens of days for most cases.
Fish are known to move some distance during a similar time
period; the fish would therefore be exposed to some "average" concentration rather than the concentration at a single point. Finally, it is highly unlikely, even if there were fish which
lived solely in the plume, that an individual would catch and consume 21 kg of these fish in ayear. Maximum allowable dose calculations for fish consumption have been revised in the PEIS andare now based on an average flow rate in the center channel of the Susquehanna River.
13.9.5.7

Spawning Ground for Fish (130)The Susquehanna River is not the principal spawning ground of the Chesapeake Bay. The system ofdams on the river prohibits movement of even anadromous fish up and down the river. For a detailed discussion of the aquatic ,ecology of the Susquehanna River and Chesapeake Bay, see Secttion 7.2.5.4 and Appendix E of the PEIS.
13.9.5.8

Aquatic Ecological Population (30,123)

The PEIS recognizes the importance of Upper Chesapeake Bay as a spawning and nursery area for
finfish, shellfish, and crustaceans, and as an important area for recreational and commercial
fishing (see Appendix E.3).
Fish considered to be endangered (shortnose sturgeon, Maryland
darter) that occur in the Upper Bay watershed are recognized, as are fish that have been severely
reduced in number during recent years - American shad, hickory shad, blueback herring, alewife,and striped bass (rockfish).
The Question and Answer document (NUREG-0732) on the TMI-2 cleanup
activities also recognizes that several bay fishes have become reduced in numbers (question
no. 69).
The PEIS indicates that the level of radiation exposure to aquatic biota and fish of the bay
would be extremely low, if any processed water releases were to occur at TMI-2.
At such lowlevels, no biological or ecological impacts are expected.
In the absence of any such effects,
the present or future status of fish now reduced in numbers should not be jeopardized as a resultof releases at TlI.
13.9.5.9

Effect of Sediment on Radioecology (13,72)

The effects of sediment on the radioecology of the Susquehanna River and Chesapeake Bay are included in Section 7.2.5.4 and Appendix T. The staff disagrees with the estimate that virtually
all of the cesium will have dropped out with the sediment within four days as suggested by acomment. The staff has estimated the effect of sediment on the transport of radionuclides in theSusquehanna River and Chesapeake Bay based on measurements of radiocesium released from Peach
Bottom in sediments of Cowingo Pond and the upper portions of Chesapeake Bay. We conclude that,conservatively, less than 12 percent of the radiocesium released from the Peach Bottom Nuclear
Plant is associated with sediments, and that the effects of sediment on the radioecology of theriver and bay are minor.

13-7613.9.5.10

Radiation Doses in the Aquatic Food Chain (11,64, 66, 68, 85)

14,20, 21,

30,

37,

42,

45,

46, 52,63,

The concentration of radioactive material in fishes in the Susquehanna River and the Chesapeak
Bay should the discharge of processed water be permitted is estimated in Section 7.2. The bioaccumulation factors for fishes have taken into consideration for bioaccumulation in lower trophic
levels. Actual bioaccumulation factors for lower trophic levels (invertebrates) are smaller thanfor fishes.
Hence, the radioactive concentrations in fishes represent the maximum concentrationthat would be expected throughout the food chain.
The assessment in the PEIS indicates that theconcentrations of radioactive material and the radiation doses to fishes and other aquatic biota
are very low. These doses to Susquehanna River fishes and Chesapeak Bay fishes will not result
in detrimental effects to those fish populations nor should have an effect on the entire aquatic
food chain or commercial fishery.
13.9.5.11

Effect of Radiation on Livestock(68)

The impacts from TMI on livestock and the results of the NRC analysis are published in NUREG0378, "Investigations of Reported Plant and Animal Health Effects in the Three Mile Island Area."
While many of the symptoms reported are characteristic of radiation sickness and of many otherdiseases, the necessary spectrum of symptoms that would establish a causal link between the reported problems and TMI was not in evidence.
Considering the lack of systematic geographic
pattern of reported problems related to the power plant, and that many of the problems werediagnosed as common occurrences in domestic and wild animals, the staff believes that no
relationship can be established between the operation of TMI, the accidental releases of radioactivity and the reported health effects.
The major source of contamination to milk is the possibility that radioiodine will be taken up by the cows and then delivered to the milk.
Thesource of radioiodine has not existed since the latter part of May of 1979.
Therefore, we do notbelieve that sampling farm milk along the transportation routes will be required.
13.9.5.12

Behavior of Sorbed Radionuclides in River and Bay (72)

Appendix T has been added to the PEIS.
The behavior of radionuclides in the Susquehanna Riverand Chesapeake Bay is discussed in Sect. 7.2.
The staff expects that the sorption of radioactivity by sediment will have only a minor effect on the ecology of the river and bay.
13.9.5.13

Concentration Factors in Water (72)There is an aquatic concentration factor for radionuclides in fish. However, there is no concentration factor in water.
The concentration factors for fish used by the NRC staff are describedin Regulatory Guide 1.109 (bioaccumulation factor).
The concentration factor for cesium is 2000
and for strontium is 30.
Concentration of radionuclides in fish was estimated in the PEIS for
the Susquehanna River and the Chesapeake Bay. Estimates of bioaccumulation factors for cesium
and strontium based on measured values in water and fish of the Susquehanna River are presented
in Section 7.2.5.4.
These comparisons indicate that the Regulatory Guide 1.109 values are conservatively high.
13.9.613.9.6.1

Environmental Monitoring
Environmental Monitoring During the TMI-2 Accident (70)

Extensive environmental monitoring programs exited during the accident, including those conductedby the licensee, the NRC, the EPA and the DOE.
Refer to NUREG-0558, "Population Dose and HealthImpact of the Accident at the Three Mile Island Nuclear Station," for a detailed discussion of
the monitoring program that was in existence around TMI at the time of the accident and the analysis that was used to determinethe dose to the population surrounding TMI.

13-7713.9.6.2

Environmental Surveillance Program (73,74, 75)The Executive Office of the President has designated the EPA the lead federal agency for conducting a comprehensive, long-term environmental radiation surveillance program. In addition toits own extensive program, EPA also coordinates all other environmental monitoring programs conducted by the licensee, the Department of Energy, the Commonwealth of Pennsylvania, the State of
Maryland and the NRC. The diverse programs provide verification of the results and should beconsidered to be superior to one single program. This coordinated effort provides the most extensive monitoring around any nuclear power facility in the United States. In addition to theenvironmental monitoring programs, the licensee must continuously monitor all radioactive efluents from the facility. The inplant radioactive effluent monitoring program should detect any
short-duration releases, and, coupled with the onsite meteorological monitoring program, enable
the assessment of offsite dose consequences. The radiological and aquatic monitory programs are
discussed in Section 11.
13.9.6.3

Effluents Monitoring and Offsite Dose Assessments Required (73)The environmental monitoring programs are not the only programs keeping track of airborne releases into the environment during the cleanup. They are augmented by the plant effluent monitoring program required by Appendix A of 10 CFR Part 50, General Design Criterion 64, for themonitoring of radioactive releases. Appendix R of the PEIS, Proposed Additions to Technical
Specifications for TMI-2 Cleanup Program, requires estimates of amounts and types of radioactivity to be released and reporting of amounts and types of radioactivity that were released.
The plant effluent releases are monitored prior to atmospheric dispersion or aquatic dilution.
In addition to its effluent monitoring program, the licensee is required to estimate populationand maximum individual doses to ensure compliance with offsite dose limits. Therefore, the environmental monitoring sensitivity limits need not allow for the detection of the very low environmental concentrations estimated in the PEIS for the cleanup operations.
13.9.6.4

Independent Environmental Monitoring (73)

In addition to the licensee's environmental monitoring program, there are several other environmental monitoring programs, including those conducted by the EPA and Commonwealth of Pennsylvania,
which are independent of the interests of the licensee. The NRC issues a weekly status reportthat contains data on the radionuclide effluents and the results of environmental monitoring by
the EPA environmental monitoring program.
13.9.6.5

Technical Specification Requirements on Offsite Dosimeters (3,27)

A standard monitoring format is used for all new technical specifications. The new technicalspecifications require that there be two rings of dosimeters around a plant, one at the siteboundary, and one at a distance of three to five miles. Each ring is comprised of 16 locations
where dosimeters are placed, one in each of 16 compass sectors. This system has undergone extensive'technical review, and we have concluded that, in conjunction with other monitoring systems
which monitor releases of radioactivity at the plant, the public health is protected. The offsite dosimetry system at TMI has about twice the number of dosimeter locations as the standard
one described here. The background radiation level in the vicinity of TMI has not changed significantly since the radiological survey of 1976 prior to the operation of Unit 2. The NRC issues a weekly TMI-2 status report. Any significant increases in environmental dosimeter results
will be reported to the public.

13-7813.9.6.6

Technical Specification Requirements on Offsite Water Monitoring (28,74)

The technical specifications for TMI-2 require that Susquehanna River water samples be taken
above and below the plant at a monthly and quarterly frequency.
The licensee also has a samplingThis sampling, in addition to the samprogram for fish, aquatic plants, and aquatic sediments.
pling of all processed water, will assure that the public is protected from planned or unplanned
discharges of radioactivity.
For surface water samples, monthly gamma spectra analyses are reFor drinking water samples, monthly gammaquired and quarterly tritium analyses are required.
spectra analyses, monthly gross beta analyses, monthly 1-131 analyses, and quarterly tritium
analyses are required.
13.9.6.7

Environmental

Sampling Program (64,68, 74,105)

The licensee has an extensive program to appropriately sample and analyze materials related to
the food pathways, such as agricultural products, milk, water, and soil, as well as fish, aquatic
Details of thisplants and aquatic sediments as part of the Environmental Monitoring Program.
program can be found in Section 11 of the PEIS.
13.9.6.8

Monitoring of River Sediments (64)The technical specifications for TMI-2 require that the Susquehanna River sediments and water be
The sediments are required to be sampled semi-annuallyperiodically sampled for radionuclides.
at upstream and downstream locations.
A gamma scan analysis is done on the sediment samples andthe results are reported to NRC.
13.9.6.9

Duration of Monitoring Programs (56)The environmental monitoring will be long-range and will last as long as necessary, certainly forthe five- to seven-year duration of the cleanup., These programs are designed to monitor theInhuman environment and are capable of detecting the first order impacts on the environment.
addition, cleanup workers are covered by a personnel dosimetry program to ensure that regulatory
limits are not exceeded and that radiation protection programs to maintain radiation doses as low
as reasonably achievable are effective.
13.9.6.10

Contingency-Surveillance Program (64)For
The contingency-surveillance program is augmented by the onsite effluent monitoring program.
example, should an unplanned airborne release occur, the effluent monitors and the onsite meteorological monitors will be able to rapidly assess the offsite radiation levels prior to activaIn the event of a significant release into
tion of the offsite contingency-surveillance program.
The
the river, the contingency program will sample and analyze the water on a timely basis.
routine weekly composite sample analysis is performed during normal cleanup operations.
13.9.6.11

Monitoring of Unplanned Releases

(74,

75,)

The licensee is required to establish and implement a continuous onsite radioactive effluentsIn the event of amonitoring program such that any release from the facility will be quantified.
short-term unplanned release, the onsite effluents monitors would detect the release and enable(1) mitigation actions to stop the release, e.g., shut off
the following actions to be taken:
exhaust or discharge, stop the source of leakage, halt cleanup operations, etc., (2) notificationof NRC and EPA to enable activation of contingency environmental monitoring, if necessary, and(3) quantification of the release to the environment for the calculation of offsite radioactivity
levels and dose rates. The existing environmental monitors would continuously monitor the radiation levels (e.g., TLD dosimeters, continuous air samplers, composite water samplers) suchthat subsequent measurements can be used to confirm the offsite dose level calculations.

13-7913.9.6.12

Onsite Monitors to Augment Environmental Monitoring for Estimation of Maximum Offsite
Doses (74)

Estimates of the maximum offsite dose from airborne releases will be based on readings of monitors at the point of effluent release at the facility and on meteorological data, such as windspeed and direction, at the onsite monitors.
The stationary monitors placed offsite will serve
to confirm those dose calculations.
Mobile monitors for release tracking will be available in
the event of any significant airborne release.
13.9.6.13

Expansion of the Federal Monitoring Program (34)

No proposal for release of the processed water has been made by the licensee.
If such a proposal
is made, and if the NRC decides to permit the release, the environmental monitoring program will
receive additional review.
The EPA, when notified of the decision, could institute and coordinate additional, appropriate environmental monitoring programs.
The Baltimore Regional Planning Council suggests that, if the processed water is released from
TMI-2, the current Federal monitoring program should be expanded along the Susquehanna River andupper Chesapeake Bay.
The Baltimore City Health Department also recommends that Baltimore andthe State of Maryland be represented on the official monitoring team.
NRC will forward thesesuggestions to the EPA, which has the responsibility for coordinating the offsite monitoring
programs relative to TMI.
13.9.6.14

Monitoring Programs for Specific Cleanup Activities (79)The environmental monitoring programs are comprehensive and extensive.
In the event of any significant anticipated release, the programs would be implemented to concentrate on the specific
event, if appropriate.
For example, during the controlled Kr-85 release from the reactor building, mobile monitors were employed to track the airborne releases.
13.9.6.15

Tritium Monitoring if

Processed Water Is Disposed by Evaporation (28)The method for disposing of processed water from the cleanup operation has not yet been decided.
If forced or natural evaporation of water is used, NRC will consider whether or not tritium airmonitoring will be required and determine the requirements of such a monitoring program to beincorporated into the technical specifications.
In the event that planned and unplanned airborne
tritium release occurs, an airborne tritium sampling and analysis program will be instituted,
e.g., during the krypton purging operation, a molecular sieve sampler was operated at the Observation Center for collection of atmospheric moisture for tritium analysis.
13.9.6.16

Updating the Aerial Radiological Survey of 1976 (3)The aerial radiological survey of 1976 was conducted by the Energy Research and Development
Administration (ERDA) for the purpose of measuring regional radiation levels.
This activity was
unrelated to surveillance of radiological impacts from TMI.
Present environmental monitoring
programs around TMI are extensive and were specifically designed for the purpose of monitoring
environmental impacts of operations at TMI.
In addition, releases from the accident were essentially composed of noble gases and did not result in any deposition which could be detected by afurther aerial survey at this time. Therefore, updating the ERDA survey of 1976 is unnecessary..

13-8013.10

ACCIDENTS AND SEVERE NATURAL PHENOMENA

13.10.1
13.10.1.1

Accidents
Accidents Involving Human Error (22,

32,

50,

55,

59,

91,

92,

102,

120,

122)

It is recognized that during the cleanup operation accidents could occur due to either equipment
For each of the processing alternatives and operations, the possibility
failure or human error.
of accidents occurring has been considered and the potential effects of these accidents on thehealth and safety of the public and workers have been examined.
13.10.1.2

Accident Effects during Unit 1 Operation (32,60)

The staff has reviewed the ramifications of a TMI-2 type of accident at Unit 1 during the cleanup
at Unit 2, and concluded that there should be no significant additional environmental impact in(1) the only significant reThe following factors were taken into consideration:
this regard.
The environmental impact would
leases from the plant during the. TMI-2 accident were noble gases.
be no more severe either because of the cleanup operations or radioactive wastes temporarily
stored onsite; (2) the licensee, in order to operate Unit 1, must provide demonstration of separation and/or isolation of the TMI-1 and -2 systems such as radioactive liquid transfer lines,
fuel handling areas ventilation systems, and sample lines. Therefore, the cleanup should notaffect the emergency operation during a TMI-1 accident; (3) TMI-I, if restarted, will be staffedwith sufficient personnel such that there would be enough personnel available to backup and aid
in managing the accident independent of TMI-2 cleanup personnel and (4) if necessary, the cleanup
operation can be halted and delayed in the event of an accident at Unit 1. Since the cleanup of
TMI-2 began soon after the accident of TMI-2, such a delay in cleanup of TMI-2 is not expected to
The environmental impact of such a delay should not be significant.
be prolonged.
Conversely, because of the separation and isolation of the TMI Units 1 and 2, there should be no
For example, if there is ansignificant impact on Unit 1 if a cleanup accident occurs at Unit 2.
accident in the Unit 2 reactor building during defueling that results in any significant airborne
radioactivity in the building, the building ventilation will be shut down to prevent airborne reHowever, even if there is a significant airborne release, the Unit 1 control room ventileases.
lation could go into emergency mode and the safety of Unit 1 would not be compromised.
13.10.1.3

Potential Formation of Zirconium Hydride and Fire Hazards (1, 70,

73)

The NRC staff has reviewed the potential for fire hazards due to any zirconium hydride that may
Thehave formed and concluded that it is unlikely that a zirconium hydride ignition would occur.
(1) Review of data on core condition
staff bases its conclusion on the following considerations:indicates that there should not be significant amounts of zirconium hydride present in the reactor vessel.
(2) Only finely divided zirconium hydride, in a powder form, when exposed to air
The presence of hydrided Zircaloy cladding in the powdered state(oxygen) would be pyrophoric.
would be readily identified by visual inspection with underwater TV montior during core inspecPrecautionary actions will be taken if such presence istion (prior to any defueling operation).
(3) Defueling operation will be performed
identified to ensure that no fire hazard would occur.
Zirconium hydride would not ignite under water.
with water coverage.
13.10.1.4

Basis for River Flow Rate in Storage Tank.Rupture Accident(21)

The flow rate of 10,000 cfs was estimated from the rating curve at Goldsboro gage for an elevation of 279 ft MSL and refers to the East channel.1 This assumed flow rate was used to estimate
the consequences of a storage tank rupture and subsequent release of the processed water into theEast Channel with subsequent dilution.

1

Letter (TLL 029) from R. F. Wilson, Met-Ed Co., to J. T. Collins, U. S. Nuclear Regulatory
TMI-2 Processed Water Storage Tanks, January 24, 1980.
Commission, Subject:

13-8113.10.1.5

Analysis of Potential Leakage of Reactor Building Sump (50,

52,

54,

75,

100,

103,

120)

Details of the analyses of leakage of reactor building sump water are presented in Appendix V
which deals strictly with the consequences of a non-mechanistic accidental leakage to the groundwater.
The appendix does not however, attempt to quantify the probability of such an event because it is not considered to be credible.
Numerical values of peak concentrations in the river as a result of the postulated leakage ofreactor building sump water are also provided.
The outcome of theing concentrations
stricted uses.
In
loss of water from
13.10.1.6

analysis demonstrates that even for the severe assumptions chosen, the resultin the Susquehanna River are well below 10 CFR Part 20 standards for unreaddition, mitigative measures could be taken in the event of a significant
the sump.

Response of NRC to Incidents(102)

In the event of an incident, the licensee
The NRC maintains a 24-hour incident response center.
will notify the incident response center and appropriate action by the NRC staff will be taken.
13.10.1.7

Evacuation Routes

(86,

92)

In determining the potential need for rapid evacuation of the region around TMI in the event ofan accident during cleanup, bounding case credible accidents were considered for all cleanup alternatives.
Estimates of the offsite dose consequences of these postulated accidents indicatethat the need for rapid evacuation would not arise. Without evacuation, the exposure to the maxbe a small fraction
imum exposed individual offsite, for the worst credible accident, would stillof existing criteria for potential radiation exposures that would require evacuation.
13.10.1.8

Population Dose Estimation in Accident Analysis (103)For each of the cleaning alternatives, bounding case accidents were postulated and analyzed for
The results calculated for the maximalexposed person
their potential dose impact to the public.
indicate that the potential radiation exposure dose is relatively small in comparison with theUnder this circumstance, and in view of theaccident dose calculated for reactor operations.
fact that the probabilities for the bounding case postulated accidents are very small, the calculation of population exposures (person-rem) is not warranted.
13.10.1.9

Accidents Involving Impact of Airplane Crash (32,

50,

51,

120)

The reactor building (RB) and the auxiliary and fuel handling building (AFHB) are designed to(The types of aircraft using the Harrisburg Interwithstand the impact of an airplane crash.
The consequences of an airplane crash withnational Airport have been taken into consideration.)
respect to the radioactive wastes that must be stored temporarily on-site outside of the RB andIn any event, the probability of an aircraft striking the site is
AFHB, were also considered.
less than 10-7 per year, which is generally accepted as the limit beyond which events are not
likely to occur.
13.10.2
13.10.2.1

Severe Natural Phenomena
Potential

Impact of Flooding (32,

67, 96,

103,

109, 120,

123)

As noted in Section 10.5, the potential for releases due to flooding is small for the following
reasons:
First, flood-control dike protection is provided for radioactive waste storage cells.
The river-water elevation has to exceed 304 ft MSL to flood over the cells.. This is an unlikely
event; the probability of recurrence interval was estimated by the staff to be greater than
2,000 years.
(The flood following Hurricane Agnes in 1972 resulted in an elevation of the riverto 300.5 ft MSL.)
Second, the stored radioactive wastes are protected by double containment in13-82sealed steel containers and sealed storage cells.
Finally, the severe flooding that would top
the station dike would most likely last less than 4 days.
Mitigating actions could be taken to
reduce the water level during that time.
It is unlikely that this time interval would cause significant leaks.
In addition, the auxiliary and fuel handling building and the reactor building
would not be affected by the probable maximum flood since those buildings are fully protectedagainst entry of floodwaters.
13.10.2.2

Basis for the Probable Maximum Flood (22,55)

Flooding is discussed for the case of the Probable Maximum Flood (PMF), which is larger than the
Design Basis Flood (DBF), which in turn is larger than the largest historical flood. The PMF is
the largest flood for which safety-related equipment is protected.
The DBF is the maximum flood
during which a plant may continue normal operation.
The analytical basis for the PMF is given indetail in the Three Mile Island Unit 2 Final Safety Analysis Report.
The PMF is based on very
conservative estimates by the Corps of Engineers of Probable Maximum Precipitation' and runoff
characteristics. 2 The flow at Three Mile Island for the PMF is 1,650,000 CFS with an elevation
at the north end of the island of somewhat less than 310 ft, and 305 ft at the south end of theisland.
It should be emphasized that this is a very conservative flood, having a recurrence interval well in excess of 1000 years.
This very low recurrence interval, combined with the double
containment, make it extremely unlikely that any releases will occur due to flooding of theSusquehanna River.
The 1972 flood associated with Hurricane Agnes was the worst recorded flood since 1786, when suchrecord keeping began.
The U. S. Corps of Engineers has determined that the recurrence interval ofthat type of flood is between 400 and 500 years. The Corps of Engineers also estimated that the1972 flood had a flow of about 1,000,000 CFS.
The DBF for the TMI plant is 1,100,000 CFS andwould have a crest elevation of about 304 ft MSL at the north end of the island.
The grade elevation on the site is approximately 304 ft.
Dikes have been constructed at a 310-ft elevation on
the north end of the island where flood elevation is highest, and at 304 ft at the south end.
13.10.2.3

Tornado Effects

(50,

55)

An assessment of tornado effects and their consequences is given in Section 10.5.2.
The probability of a design basis tornado is less than 10-7 per year, and the cleanup facilities and waste
storage areas are sufficiently protected so that releases are highly unlikely.
13.10.2.4

Effects of Natural Disaster on Private Property (67)

As for the effects on private property (i.e., "homes") from a severe natural disaster (e.g.,
floods, earthquakes), any property destruction would probably only occur as a direct consequenceof the natural phenomena itself. For natural disasters up to certain levels of severity (e.g.,
Design Basis Flood, Tornado or Earthquake), the leakage of radioactivity would be prevented by
the design features reviewed in the Safety Evaluation for the TMI-2 facility. As noted in Section 10.5, potential consequences of a more severe disaster (e.g., Probable Maximum Flood) werealso found to be minimal.
13.10.3

Safeguards Against Sabotage (100)The relationship between sabatoge and accidents is difficult to assess in all but probabilistic
terms.
Given the required safeguards imposed pursuant to Commission Regulation, 10 CFR 73, its
likelihood is regarded too remote to be properly assessed in evaluative terms.

lCorps of Engineers,
Corps of Engineers,2

Hydrometeorological Report No. 40.
Susquehanna River Basin Study, Appendix D, June 1970.

13-8313.11

SOCIOECONOMIC AND PSYCHOLOGICAL EFFECTS

13.11.1
13.11.1.1

Social and Economic Impacts
Scope of the Staff's Socioeconomic Analysis (5,

32,

51,

64,

75,

86, 100)

The staff agrees that Cumberland County should have been considered in Section 3.1.6 because ofThis has been done in revising the section for
its proximity to and heightened awareness of TMI.
Army Depot and the Mechanicsburg NavyNew
Cumberlandthetoimpacts
potential
However,the PEIS.
Depot are not discussed because the staff was unable to develop a credible model linking postulated accidents during decontamination of TMI-2 with continued operation of these military
Similarly, the only accidents which the staff has been able to postulate would not
facilities.
result in significant radiation doses requirino hospitalization; thus, the discussion in the draft
statement on hospital care has been omitted from this statement.
In developing its environmental impact analysis the staff must reach a balance between the socioClearly, an analysis of all
economic costs and benefits of analyzing all, even trivial, issues.
issues, because of its length, would be unreadable and thus counterproductive to public disclosure.
The balance in the staff's analysis is achieved by considering the scoping process required
The scoping prounder current CEQ regulations and by the application of professional judgment.
cess permits the staff to subjectively rank those concerns expressed by individuals in public
Professional judgment augments the expression of public
meetings and written correspondence.
The
concerns by adding or focusing issues which may not be fully appreciated by the public.
staff's balancing process results in substantive, temporal, and spatial limits which describe theSection 3.1.6.1 provides the rationale for theframework for the environmental impact analysis.
We believe this discussion, which has been rewritten for the
spatial limitation of the analysis.
PEIS, is adequate.
13.11.1.2

The Susquehanna River as a Community Water Supply (20,130)

The statement on page 3-19 of the draft PEIS that "use of the Susquehanna River in community waterThis has been corrected in Section 3.1.6, and addisupply systems is very limited" was in error.
tional users are now listed in Section 3.1.4.2.
13.11.1.3

Impact of Increased Construction Workforce (50)Section 5.2.5.4 of this final Statement explains the reasons for not evaluatino the impact of inmoving construction workers on the communities and the economy.
13.11.1.4

Effects of Decontamination Activities on Land and Inhabitants (75)

The comment was made that Section 3.1.6.2 (Impact Study Area) does not include effects of the
The writer apparently did not recognize.that thissection is intended to becleanup operation.
Analyses
Analysis of the effects is in Section 10.6 of the draft statement.
purely descriptive.
of socioeconomic impacts are found in each section of this final Statement where such discussions
are warranted.
13.11.1.5

Property Value Effects (50,131)

In general, the staff agrees with the Met-Ed comment that recent data indicating increases inproperty values between 1979 and 1980 are contrary to the statement in item 6 of Chapter 12 of thedraft Statement, which says i'reduced property values ... may occur." Section 3.1.7.2 of thisPEIS contains information on this subject which was not available when the draft PEIS was
written.

13-8413.11.1.6

Potential Impact on the Marketability of Chesapeake Bay Shellfish and Finfish (11,16, 29, 30, 34, 41, 42, 48, 52, 61, 64, 66, 72, 100, 130)

13,

The potential for adverse impact to Chesapeake Bay activities--commercial fishing, seafood consumption, recreation fishing, and waterfowl hunting--as a result of controlled or uncontrolled
The staff
releases of tritiated water to the Susquehenna River is addressed in Section 7.3.5.4.
concluded that an in-depth study of the potential economic losses to Chesapeake Bay activities
being undertaken by the State of Maryland should precede the decision regarding disposition ofSection 3.1.6.1 contains descriptive information on the Chesapeake Bay
TMI-2 processed water.
economy.
13.11.2
13.11.2.1

Psychological Concerns
Terms Used (2,

15,

19)

"Phobia" has
Misuse of terms, particularly the word "phobia", was cited by several authorities.
Other technicalbeen removed from the discussions of psychological concerns in the final PEIS.
terms have also been examined for appropriateness and consistency by the staff and its
consultants.
13.11.2.2

Analysis Incomplete (15,32,

36, 44,

51,

61,

84,

72,

96,

99,

100, 106,

120,

124)

In response to comments suggesting that more detailed analysis be made of the psychological consequences of decontamination, the staff has expanded these discussions and placed them in the
However, suggested topics not within the scope of theappropriate chapters of the final PEIS.
PEIS, such as the issues relative to restarting Unit 1, have not been addressed.
13.11.2.3

PsychologicalConsequences (59,69, 85,

67,

93,

100,

114, 115,

122,

123)

Many comments allege that the psychological consequences of the cleanup process are understated in
In expanding the assessments of these consequences throughout chapters 5 through 9,
the draft PEIS.
the staff has reexamined its assessments and changed them where appropriate.
13.11.2.4

Impact of Accident Minimized (99,

109,

114,

128)

The psychological impact of the March 1979 accident at TMI-2 was minimized in Section 3.1.7, acThe expanded discussion in this Final PEIS (now in Section 3.1.6)
cording to several comments.
examines the existing stress on the community, and includes recent research by Bromet (1980) and
Hout et al., (1980).
13.11.2.5

Public Reaction Against Releases to the River (12,52,

72, 112)The staff acknowledges the many comments opposing the possible discharge of processed water to theriver.
Further analysis of the psychological effects associated with all the feasible water disposal alternatives is included in Section 7.2.
13.11.2.6

Long-Term Psychological

Effects (67)The staff's discussion of psychological impacts attributed to the TMI-2 accident has been revised
to acknowledge two recent studies which suggest that measurable long-term psychological effects
may occur (Sec. 3.6.3).
However, no conclusive evidence will exist for many years.

13-85
13.12

FINANCIAL ASPECTS OF TMI-2 CLEANUP

13.12.1
13.12.1.1

Costs
Costs Not Available (59,124,.125)

RelaCost figures were not included in the draft PEIS because they were not available in time.
tive costs estimates for the alternatives are given in the PEIS.
Unlike an environmental
statement on a proposed nuclear power plant, cost factors are secondary issues;
the cleanup has
It is appropriate,
to proceed for protection of public health and safety and of the environment.
however, from the standpoint of the National Environmental Policy Act, to consider the costs ofproposed actions in relationship with available alternatives.
Even in this context, however,consideration of costs are not dispositive.
13.12.1.2

Relative :Costs of Alternatives (4,112, 117)

11,

13,

16,

50,

52, 61,

64,

66,

70,

84, 92,

Relative cost estimates of the cleanup alternatives have been included in the PEIS.
Financialcosts, however, will not be a major factor in the Commission's decision on specific cleanup proposals as the mandated responsibilities of the NRC are to ensure public health and safety and ensurethat the environmental impacts will be acceptable.
Expenditures on the cleanup are not regulatedby the NRC but are the responsibility of the licensee.
Cost factors may well influence thelicensee's cleanup proposals the licensee has the responsibility to make.
13.12.1.3

Decommissioning Cost (115)

The comment refers to a cost figure originated in Volume II of NUREG-0662, "Final Environmental
Assessment for Decontamination of the Three Mile Island Unit 2 Reactor Building Atmosphere."The
figure quoted originated from a study by Stanford Research Institute.
It was included inVolume II of the NUREG for the purpose of reproducing all the inputs that were received on the
The discussion of decommissioning costs was actually irrelevant to the topic ofventing question.
krypton venting.
See Section 1.3 for an updated discussion of cleanup costs.
13.12.1.4

Cost-Benefit Analysis (32,59,

64, 66,

105,

123)

Cost-benefit analysis, in the usual sense for nuclear power plant applications, would include thecost and benefit comparisons of the application versus alternative sources of power.
In the caseof the cleanup operations, the costs are discussed in terms of the relative worker radiation exposures associated with the alternatives, and the benefits would be in terms of the reduction or
elimination of potential sources of radiation hazards to the public through cleanup and disposition of radioactive wastes.
13.12.2
13.12.2.1

Financial Responsibility
NRC Financial Responsibility (7,20, 59, 67,

102, 115,

126,

130)

The NRC, as an independent regulatory body, is mandated to ensure public health and safety and theprotection of the environment for licensed activities; it has no responsibility for the financial
well-being of nuclear power, generally, and is not authorized to provide loans to utilities forthe purpose of preventing bankruptcy.
Additionally, it is not the function of the NRC to regulate
the expenditures of Metropolitan Edison Company.
The construction of the Submerged Demineralizer
System (SDS) by the licensee is proceeding at his own risk, since the operation of this system
will require the NRC approval and such decision would not be made pending evaluation of the alternatives and their environmental impacts in this final PEIS.
Construction of the SDS has not prejudiced any NRC decision on which method is most appropriate.
The NRC will evaluate the licensee's
proposed cleanup activities in accordance with the mandate to ensure the health and safety of the
public and to protect the environment.

13-8613.12.2.2

Licensee Financial Responsibility (12,

38,

46,

66,

100,

101)

The licensee, Metropolitan Edison Company and its parent company, General Public Utilities are
responsible for the expenses of the cleanup.
However, the licensee is presently constrained by
the Pennsylvania Public Utilities Commission from spending money from the rate payers on thecleanup of TMI-2.
Should the licensee be financially unable to carry out the required cleanup,
the NRC has the authority under existing laws to act on behalf of the federal government such thatthe cleanup can be continued without hazard to the general public.
13.12.2.3

Licensee Default (11, 27,

66,

67,

100,

103,

114,

115,

125)

The financial aspects of the licensee are not within the regulatory responsibilities of the NRC
and it is not within the scope of the PEIS to discuss the financial impacts of cleanup on the
The NRC Staff, however, has developed contingency plans in the event the licensee
licensee.
should go bankrupt and fail to carry out the cleanup responsibilities.
Details of this contingency planning can be found in a report prepared by the NRC staff, "Potential Impact of Licensee
Default on Cleanup of TMI-2," (NUREG-0689).
A number of options are possible, including NRC's
licensing another organization or another federal agency to clean up the facility or NRC's taking
over the facility itself.
However, additional funding, authorized and appropriated by theCongress, would be necessary if such circumstances arise. The question of who is the plant operator for the cleanup, however, does not affect the evaluations in the PEIS.
The cleanup alternatives will remain the same even in the event the licensee should default on the cleanup
obligations.

GLOSSARY
Absorbed dose--The energy imparted to matter by ionizing radiation.
Accident sludge--Sludge consisting of fine solid material which includes cement dust, dirt, resin
beads, etc., which have settled out from liquid contaminated during the TMI accident.
Accident water:(a)
Water that existed in the TMI-2 auxiliary, fuel handling, and containment buildings,
including the primary system, as of October 16, 1979, with the exception of water which
as a result of decontamination operations becomes commingled with non-accident-generated
water such that the commingled water has a tritium content of 0.025 pCi/mL or less before
processing.
(b)
Water that has a total activity of greater than 1 pCi/mL prior to processing except wheresuch water is originally non-accident water and becomes contaminated by use in cleanup.
(c)Water that contains greater than 0.025 pCi/mL of tritium before processing.
Activation--The induction of radioactivity in material by irradiation with nuclear particles,
usually neutrons produced by a nuclear reactor.
Activity--A measure of the rate at which a material is undergoing spontaneous nuclear transformations, giving off radiation. The special unit of activity is the curie (Ci).
AEC--Atomic Energy Commission,
ment of Energy.

predecessor to the Nuclear Regulatory Commission and the Depart-

AFHB--Auxiliary and fuel handling buildings.
AFR storage--Away-from-reactor storage; refers to spent nuclear fuel storage facilities locatedelsewhere than at a nuclear reactor site.
ALARA--As low as reasonably achievable.
Alpha radiation--An emission of particles (helium nuclei) from a material undergoing nuclear
transformation; the particles have a nuclear mass number of four and a charge of plus two.
Alpha waste--Waste material contaminated by radionuclides that emit alpha particles; inparticular, transuranic elements.
Anadromous fish--Fish that ascend freshwater streams from the sea to spawn.
Anions--Ions that are negatively charged.
Anticipated operational occurrence--Miscellaneous conditions or actions such as equipment failure, operator error, administrative error, that are expected to occur at a nuclear plant thatare not of magnitude great enough to be considered an accident.
AP-citrox process--A two-step decontamination process consisting of a pretreatment step using
an alkaline permanganate solution followed by a cleaning step using a solution of citric andoxalic acids.
APAC process--Decontamination process using an alkaline permanganate solution and an ammonium
citrate solution.

GL- 1

GL-2

APSR--See axial power shaping rods.
Aquifer--A subsurface geological formation containing sufficient saturated permeable material totransmit groundwater and to yield economically significant quantities of water to wells andsprings.
Atomic number--The number of protons in the nucleus of an atom which is also equal to the numberof electrons outside the nucleus. The collective term for atoms of the same atomic number
is "element."
Axial power shapin 9 rods (APSR)--Rods inserted into the core of a reactor to control the rate ofenergy released in localized regions of the core.
Background radiation--The level of radiation in an area which is produced by sources (mostly
natural) other than the one of specific interest, e.g., cosmic radiation and radioactive elementsin the atmosphere, building materials, the human body, and from the crust of earth. In the
Harrisburg area, the background radiation level is about 116 mrem per year, not including
any contribution from medical practice.
Benthic--Generally living on the bottom of a water body.
Beta particles--An electron or a positron (a particle with the same mass as an electron but witha positive charge rather than a negative one). Usually used to refer to a particle moving at avelocity high enough to produce ions. Beta particles are commonly emitted from the nuclei ofatoms undergoing nuclear transformation. Also referred to as beta radiation.
Beta radiation--See beta particles.
Bioconcentration (bioaccumulation)--The process whereby an organic system selectively removes
an element from its environment and accumulates that element in a higher concentration.
Biomass--The total weight (mass) of living and dead organisms present in an area, volume, or
ecological system.
Biota--Plant and animal life.
Biotic--Related to living organisms.
Bitumen--An asphalt material (usually obtained from petroleum or coal-tar refining)that can be used in immobilization of radioactive wastes.
Bituminization--Immobilization of wastes by addition of bitumen.
Body burden--The amount of a specified radioactive material or the summation of the amountsof various radioactive materials present in an animal or human body at the time of interest.
Borated water storage tank (BWST)--A tank at TMI-2 intended for use in storing water containing
boron. This tank now contains water that has been processed through the EPICOR II.
Boron--A neutron-absorbing element used in nuclear reactor systems to control criticality.
Breakthrough volume--The processing volume passed through an absorbent (ion-exchange system) thatnecessitates the replacement or regeneration of the ion-exchange resins.
Burial ground--An area specifically designed for the disposal of solid waste.
Burnable poison rods--Rods containing materials with high neutron-absorbing capability which areused in the core of a reactor to control reactivity. As the rods are irradiated, the amount of
neutron-absorbing material is reduced. This reduction in neutron-absorbing material compensates
for the buildup of fission products in the fuel which absorb neutrons, and a general reductionin the amount of fissionable material as the fuel is irradiated.

GL-3

Burnup--The process of fissioning (consuming) heavy metal isotopes in nuclear fuels.
BWR--Boiling water reactor.
BWST--See borated water storage tank.
Calandria--A sealed vessel containing cooling tubes used in the core of some types of reactors.
CAN-DECON process--A proprietary dilute chemical decontamination process developed by London
Nuclear Services, Ltd., and Ontario Hydro.
Capable fault--A fault in the earth's crust determined to be capable of causing an earthquake ofa specific intensity on the basis of the geologic history of the fault.
Carnivorous--Meat eating.
Catadromous fish--Fish living in fresh water and migrating to salt water to spawn.
Cation--An ion with a positive charge.
CCST--Chemical cleaning solution tank.
CCTV--Closed-circuit television.
cfm--Cubic feet per minute.
CFR--Code of Federal Regulations.
cfs--cubic feet per second.
Chelating agent--An organic compound used to complex some metal ions to prevent them fromprecipitating in neutral or alkaline wash solutions.
Chemical decontamination--Use of solvents to dissolve or suspend radioactive contaminants.
Ci--See curie.
Cladding--See fuel rod cladding.
Cain-- In reference to TMI-2, this term is used to mean decontamination, core removal (defueland waste disposal.

Cold leg--The section of PWR reactor cooling system piping through which the coolant flows fromthe heat exchanger to the core.
Coliform bacteria--Intestinal bacteria that are considered as indicators~of possible waterpollution by human or animal wastes.
Condensate--The liquid product formed by condensation of a vapor.
Containment (or reactor) building--The structure housing the nuclear reactor. The-building is
specially constructed of reinforced concrete and is designed to withstand internal pressureand external collisions. It also is fitted with gas-tight seals designed to contain radioactivity within the building and permit release of radioactive materials only under controlled
conditions.
Containment sump--See reactor building sump.
Contamination--In this document generally used to mean the deposition, solution, or infiltrationof radionuclides on or into an object, material, or area.
Control rods (CR)--An array of tubes that contain material that absorbs neutrons and is inserted
into the core of a nuclear reactor to control or halt nuclear fission.

GL-4
Controlled area--Any specific region of a nuclear plant into which entry by personnel is regulated
by a physical barrier or administrative procedure.
Core support structure--A large basket-like component that fits into the reactor vessel and isused to hold the fuel elements and direct the reactor coolant to the bottom of the reactor
vessel.
Core--The central portion of a nuclear reactor containing the fuel elements.
Corrosion products--Materials formed by chemical reaction of metals with the coolant in the reactor.
CP review--Construction permit review; the review proceedings conducted by the NRC prior toissuance of a permit to construct a nuclear power plant.
CR--See control rods.
CRDM--Control rod drive mechanism.
Critical--The condition in which an arrangement of fissionable material undergoes nuclear
fission at a self-sustaining rate.
Criticality safety--The handling and storage of fissionable materials in such a manner that they
will become critical only under desired, controlled conditions.
Crud--Corrosion products (principally oxides of iron, chromium, and nickel) circulating in orloosely deposited on the surfaces of the primary system.
Cryogenic processing--Low-temperature separation of gases.
CSB--Containment service building.
Cumulative occupational dose--The total radiation dose to workers; determined by multiplying thedose rate times the number of workers exposed times the length of exposure. This is expressedin terms of person-rem.
Curie (Ci)--The special unit of radioactivity. Activity is defined as the number of nuclear
transformations occurring in a given quantity of material per unit time. One curie ofradioactivity is 37 billion transformations per second.
D-rings--The shield enclosures around the steam generator compartments; they are so named
because of their shape.
Daughter products--The nuclides formed by the radioactive disintegration of a first nuclide(parent).
Dead leg--A section of reactor cooling system piping connected in a manner such that it does not
drain with the rest of the system or such that water does not circulate through it.
Decay chain--The sequence of radioactive disintegrations in succession from one nuclide toanother until a stable daughter is reached.
Decay heat--Thermal energy produced by the decay of radioactive materials.
Decommissioning--The planned, orderly execution of steps to place a facility in a permanently
nonoperable, safe condition.
Decontamination--The removal of radioactive material from a surface or from within anothermaterial.
Deionized water--Water from which ionized impurities have been removed.

GL-5
Demineralizer systems--Processing systems in which synthetic ion-exchange materials are used toremove impurities from water.
Design basis earthquake--See operating basis earthquake.
Design response spectra--Plots of the intensities (amplitude) versus the frequencies of groundmovement. Once a spectrum is determined, a nuclear reactor structure's response to that motion
is computed by determining the extent to which it amplifies movement of the ground. Response
spectra are determined from geologic histories of ground motion.
Desorb--To remove materials that have been adsorped on another material.
Detritus--Loose organic material formed from decomposition of organisms.
Disintegrations per minute (dpm)--Obtained by multiplying number of disintegrations per second(dps) by 60.
Disintegrations per second (dps)--The number of radioactive decay events occurring per secondin a given amount of material.
DOE--U.S. Department of Energy.
Dose commitment--The integrated dose that results unavoidably from an intake of radioactive
material starting at the time of intake and continuing (at a decreasing dose rate) to latertime (usually specified to be 50 years from intake).
Dose--A general term indicating the amount of energy absorbed from incident radiation by aunit mass of any material.
Dosimeter--Dose meter; an instrument that measures radiation dose.
dpm--See disintegrations per minute.
dps--See disintegrations per second.
Earthquake intensity--See modified Mercalli scale.
Eductor--An ejectorlike device used to mix or agitate fluids.
pump fluids.

Eductors are frequently used to

EPICOR I--The liquid radioactive waste processing system designed for use during routine operationof TMI to clean up liquids containing activity less than 1 microcurie per milliliter of water.
EPICOR II--A filtration and demineralizer system designed to process some of the liquid radioactive waste resulting from the TMI accident. The system can be used on liquid waste containing between 1 and 100 microcuries of radioactivity per milliliter of water.
Estuary--A semienclosed coastal body of water that has a free connection with the open sea andwithin which sea water is measurably diluted with fresh water.
Eutrophic--Pertaining to a shallow lake containing a high concentration of dissolved nutrientsand having periods of oxygen deficiency.
Evaporator bottoms--The residue in an evaporator following evaporation of liquids.
Exclusive-use vehicles--Refers to vehicles used only to transport radioactive waste from a single
shipper.
Exposure--The condition of being made subject to the action of radiation; also, frequently, thequantity of radiation received.

GL-6
FES--Final Environmental Statement.
Filtrate--Material that has passed through a filter.
Fissile material--Material capable of undergoing fission following the absorption of thermal(essentially zero-energy) neutrons, e.g., U-238 and Pu-239.
Fission products--The nuclides formed by the division of a heavier nucleus, typically in a
nuclear reactor. Isotopes of essentially all elements are produced by fission of fissile
materials. Fission products are the main radioactive components of high-level radioactive
wastes.
Fission--The spontaneous or induced disintegration of a heavy atom into two or more lighter atoms
With an accompanying loss of mass which is converted into nuclear energy.
Fissionable--Material capable of undergoing fission when struck by neutrons of sufficiently highenergy, e.g., U-238.
Freon--Tradename for any of several liquid or gaseous fluorinated hydrocarbons.
FSAR--Final Safety Analysis Report.
FTC--See fuel transfer canal.
Fuel accountability--A system for determining the location of all nuclear fuel to prevent
diversion for non-intended purposes.
Fuel assemblies--Metal grid structures containing arrays of nuclear fuel rods.
Fuel processing--Chemical and physical reprocessing of spent uranium- or thorium-based fuels
for separation and recovery of uranium, thorium, and plutonium from the fission productwastes.
Fuel rod cladding--Metal material forming the exterior of a nuclear reactor fuel rod.
Fuel rod--One of many metal tubes containing uranium fuel for a nuclear reactor.
Fuel transfer canal (FTC)--A water-filled channel that connects the reactor core and the spent
fuel storage pool.
Fuel--See nuclear fuel.
Gamma radiation--Electromagnetic radiation of high energy (and short wavelength), emitted by
nuclei undergoing internal changes. Gamma (rays) have the highest energy and shortest wave
length in the electromagnetic spectrum and are capable of penetrating several inches of asolid such as concrete.
Gamma scan--Use of radiation-detection equipment designed specifically to detect gamma radiationand measure its energy. This process can be used to evaluate the integrity of irradiated fuelelements.
(pm--gallons per minute.
GPU--General Public Utilities Co.
GroUndwater--Water that exists or flows below the ground's surface (within the zone ofsaturation).
Half-life--The time required for half of a given radioactive substance to decay.

GL-7

Hands-on work--Work requiring the presence of workers for physical manipulation of contaminatedequipment or for actual decontamination activities.
HEPA filter--High efficiency particulate air filter.
High-density fixative--An adhesive plastic similar to that used for caulking compounds or roof
patching compounds, but with a high-density filler added to increase radiation shielding.
High-frequency anchor point--A geologic term meaning the maximum intensity (amplitude) determinedfrom the ground motion spectrum; used to compute the design response.
High-level waste (HLW)7-Spent nuclear fuel or the radioactive materials extracted from spent
fuel during reprocessing.
High-pressure water jet--A high-pressure,in removing surface contamination.

low-flow-rate water-jet spray system designed for use

High-specific-activity wastes (HSAW)--Wastes having higher activities than wastes which areroutinely generated at nuclear power plants and which'are disposed of by routine shallow land
burial techniques.
Holdback-carrier solutions--Solution containing inactive atoms which exchange with the radioactive species adhering to surfaces or equipment.
Hot cell--A heavily shielded work area designed to contain highly radioactive materials andinc7uding provisions for remote manipulations of such materials.
Hot leg--The section of PWR reactor cooling system piping through which the coolant flows fromthe reactor core to the heat exchanger.
Hot spots--Specific locations where radiation dose rates are significantly higher than in thegeneral surroundings.
HPI--High-pressure injection system.
This system is used to inject water into the reactor vesselin the event of a loss-of-coolant-accident.
HSAW--See High-Specific-Activity waste.
HTO--Tritiated water in which one of the two hydrogen atoms has been replaced by a tritium atom
7see tritiated water).
Hydrogen control subsystem--A portion of the reactor building ventilation system designed to.control the hydrogen concentration of the building atmosphere to within specified limits.
Ichthyoplankton--Fishlarvae and eggs.

Immobilization--In this document, usually meant to refer to the fixing or solidification of radioactive wastes by any of several possible means (e.g., solidification in cement).
In-situ--In place.
Induced-service employment--Jobs created to fulfill the service needs of workers moving into
the area. These jobs could be in the public sector (e.g., hiring of more firemen) or theprivate sector (e.g., hiring of more waitresses at a local cafe).
In some cases, the jobs
could be with local firms providing goods or services directly for the cleanup effort.
Internals--See reactor internals.
Invertebrates--Animals without backbones (such as insects and worms).
Ion exchange media--Resins or zeolite materials used in ion exchange processes.

GL-8
Ion exchange--Here, a process for selectively removing a constituent from a waste stream byreversibly transferring ions between an insoluble solid and the waste stream.
Ion--An atom or molecule from which an electron has been removed (positive charged ion) or towhich an electron has become attached (negative charged ion).
Ionization--The process by which a neutral atom or molecule acquires a positive or a negative
charge by removal or attachment of an electron.
Ionizing Radiation--Any form of radiation that generates ions.
Irradiation--Exposure to radiation, e.g.,beam.

by being placed near a radioactive source or in an X-ray

Isotopes--Nuclides with the same atomic number but with different atomic masses therefore having
the same chemical properties but different physical properties.
Krypton-85 (Kr-85)--A radioactive noble gas with a half-life of 10.7 years.
This was theprincipal radioactive contaminant in the TMI-2 reactor building atmosphere.
LCF--Latent cancer fatalities.
Lead screws--Elements that connect the reactor control rods and axial power shaping rods totheir respective drive mechanisms (stators).
LET--See linear energy transfer.
Letdown coolers--Heat exchangers used to cool reactor primary water before routine purification.
Linear energy transfer--A measure of the capacity of biological material to absorb ionizing
radiation.
Specifically, this is a measure of the amount of energy deposited per unit
length along the track of a charged particle (a beta ray, for example) as it passes
through an absorber.
LLW--See low-level waste.
Long-lived isotope--A radioactive nuclide that decays at such a slow rate that a quantity of it
will exist for an extended period.
Low-activity wastes--Radioactive wastes that are similar to routinely generated wastes which aredisposed of by routine shallow land burial operations.
Low-level waste (LLW)--All radioactive waste materials that are not high-level or transuranic
waste.
Most TMI-2 wastes will be of this type.
LSA boxes--Boxes designed to hold radioactive wastes classified as low specific activity by
Department of Transportation Regulations.
Macrophytes--Macroscopic plants,

especially in aquatic habitats.

Maximum credible earthquake--The greatest intensity earthquake reasonably expected in a given
region based on current and historical seismicity and geologic structure.
Maximumpermissible concentration (MPC)--The annual average concentration of a radionuclide in
air or water to which an individual may be continuously exposed without exceeding an established limit of radiation dose.
Intermittent exposures at greater than the MPC can occur
without exceeding the dose standard.

MDA--See minimum detectable activity.

GL-9
MDHRS--See mini decay heat removal system.
Mechanical decontamination--Use of mechanical means, such as sandblasting, to dislodge radioactive contaminants.
Meltdown--A state in the core of a reactor in which melting of fuel and internal reactor
components occurs.
Mesotrophic--Refers to a lake characterized by a moderate supply of nutrients in the water.
Meteorological dispersion factor -- A factor (seconds per cubic meter) that takes account ofsite-specific meteorological data in relating the concentration (Ci per cubic meter) ofradioactive materials at a given location to a release rate (Ci/sec) of radioactive
material at another location.
mgd--million gallons per day.
Microcurie (pCi)--Unit for measuring radioactivity.
curie (1/1,000,000).

One microcurie is one one-millionth of a

Milk juggers--Farmers who sell milk directly to consumers.
Millicurie (mCi)--Unit for radioactivity, one.millicurie equals one one-thousandth (1/1000) of a curie.
Millirem (mrem)--One one-thousandth (1/1,000) of a rem (see rem).
Mini decay heat removal system (MDHRS)--A special forced-circulation cooling system installedsince the TMI accident to cool the Unit 2 core.
Minimum detectable activity (MDA)--Minimum level of radioactivity detectable above background.
Moderator--A material such as ordinary water or graphite used in a reactor to slow down highvelocity neutrons, thus increasing the likelihood of fission.
Modified Mercalli (MM) scale--An arbitrary scale of earthquake intensity, ranging from I (detectable only by instruments) to XII (causing almost total destruction).
Molecular sieve--A chemical compound with a latticelike formation used to absorb or separate
molecules.
Usually a synthetic zeolite.
mrem--See millirem.
MSL--Mean sea level'.
MW--Megawatts:

unit of power equal to 1,000,000 watts.

MWe--Megawatts electric: ;unit of power denoting output of an electric power plant equal to one
million watts of electricity.
MWHT--Miscellaneous waste holdup tank.
Neutron activation--The process of irradiating a material with neutrons so that stable atoms
in the material are transformed into radioactive nuclides.
Neutron--A particle in or emitted from the nucleus of an atom which is electrically neutral andhas a mass approximately equal to that of a proton (the nucleus of a stable hydrogen atom).
Neutron-absorber--A material, such as boron, that will readily absorb neutrons emitted during a
nuclear chain reaction and thus make the neutrons unavailable to contribute to continuation of
the chain reaction.

GL-1O
Noble gases--Inert gases that do not readily react chemically with other elements.
include helium, neon, krypton, xenon, and radon.

These gases

NRC--See Nuclear Regulatory Commission.
NRTS--National Reactor Test Station, now Idaho National Engineering Laboratory (INEL),-Tdaho Falls, Idaho.

near

NS-1 process--A proprietary decontamination process-developed by Dow Chemical Company.
NSSS--Nuclear steam supply system.
Nuclear fuel--Fissionable material inserted into a nuclear reactor. The basic fuel for most of
the commercial nuclear reactors in the United States is uranium oxide.
Nuclear power plant--A facility that uses energy from a nuclear reactor to produce electricity.
Nuclear radiation--Particles and electromagnetic energy given off due to the transformation occurring in the nucleus of an atom.
Nuclear reactor--A device containing fissionable material in which a chain of fission events
can be maintained and controlled to meet a particular purpose.
Nuclear Regulatory Commission (NRC)--U.S. agency responsible for the licensing, regulation, andinspection of commercial, test, and research nuclear reactors, as well as nuclear materials.
Nucleus--A positively charged particle, made up of neutrons and protons, situated in the centerof an atom and surrounded by a cloud of negatively charged electrons.
Nuclide--A species of atom having a specific mass, atomic number, and nuclear energy state.
Occupational radiation exposure--The radiation exposure to which workers at a nuclear facility
are subjected during the course of their work.
OL review--Operating license review; the review proceedings conducted by the NRC prior
to issuance of an operating license for a nuclear power plant.
Operating basis earthquake--An earthquake of an intensity through which a nuclear plant isdesigned to remain functional.
OPG process--Decontamination process using a solution of oxalic acid, hydrogen peroxide, andgluconic acid.
Order of magnitude--A specification indicating a range or factor of about ten.
Oxalic-citrate-peroxide process--Decontamination process using a solution of oxalic acid,ammonium citrate, and hydrogen peroxide.
Oxidation--A chemical reaction that increases the oxygen content of a compound.
Particulates--Small particles.
PEIS--Programmatic Environmental Impact Statement.
Penetrating radiation--Forms of radiant energy that are capable of passing through significant
thicknesses of solid material; these usually include gamma rays, x-rays, and neutrons.
Penetration R-626 Cutout--An existing, normally sealed, penetration through the outer wall of
the TMI-2 reactor building through which instruments can be inserted.

U

GL-11
Percolation--Gravity flow of groundwater through the pore spaces in rock or soil.
Permissible dose--The dose of ionizing radiation that, in the light of present knowledge,
carries negligible probability of causing severe somatic injury or genetic effect.
Person-rem--The sum of the individual radiation doses (collective dose) received by members of acertain group or population.
It may be calculated by multiplying the average dose per personby the number of persons. For example, a thousand people each exposed to one millirem(1/1000 rem) would have a collective dose of I person-rem.
pH--A measure of the relative acidity or alkalinity of a solution; a neutral solution has a
pH of 7, acids have a pH of below 7, bases have a pH above 7.
Piscivorous--Fish eating.
Plankton--Small organisms passively floating in.the water; includes phytoplankton (plants) andzooplankton (animals).
Plateout--The deposition of a substance from suspension orsolution onto the internal surfacesof the vessels (e.g., pipes) containing the liquid. In this document, plateout refers specifically to thin layers of radionuclides that were deposited on all exposed building andequipment surfaces inside the reactor building and on the insides of pipes and tanks during
and after the accident.
Plenum--See upper plenum assembly.
Polymer--A large molecule formed by the union of small or simple molecules.
Population dose--The summation of individual radiation doses received by all those exposedto
the source or event being considered, and expressed as person-rem. The same as collective
dose.
PORV--Pilot operated relief valve, located in a pipe leading out of the top of pressurizer anddesigned to open automatically when the primary system pressure exceeds a safe limit and to
close automatically when the system pressure is back to normal.
Positive displacement pump--A pump in which a measured quantity of liquid is entrapped ina space,its pressure is raised, and then it is delivered.
Pressure vessel--See reactor pressure vessel.
Primary production--The formation of the food chain base through photosynthetic carbon fixationby plants and bacteria in water.
Primary system--See reactor cooling system.
Primary water--Water in (or from) the reactor coolant system.
Process solids--Wet solids in the forms of sludge, high-solids-content slurries, or granularmaterials generated during the accident or during subsequent treatment of accident water anddecontamination liquids.
Processed water storage tanks (PWSTs)--Tanks intended for use in storing TMI water that has
been subjected to decontamination processing.
Processed water--Contaminated water that has been treated to remove radionuclides (exclusive oftritium).
Productivity factor--A factor used (in this document) to indicate the expected efficiency of thedecontamination workers, who must wear heavy protective clothing, respirators and other encumberances when they are working in a radiation field; usually expressed as a fraction of normal
productivity. The productivity factor has been used in calculating the amount of time expectedto accomplish given tasks.

GL-12
PRTR--Plutonium Recycle Test Reactor at Hanford, Washington.
psi--Pounds per square inch; a measure of pressure.
PWR--Pressurized water reactor (TMI-2 is this type of reactor). The primary system is pressurized to about 2200 psi so that it will not boil as it is heated in the reactor core.
R & D--Research and development.
R--See roentgen.
Rad--A unit of absorbed dose of ionizing radiation. A dose of one rad results from the absorption of 100 ergs of energy per gram of absorbing material.
Radiation spectrum--The intensity of radiation expressed as a function of energy.
Radiation survey--The instrumental evaluation of an area or object in order to detect, identify,
and quantify radioactive materials and radiation fields present.
Radiation zone--Area that contains radioactive materials in quantities significant enough torequire control of personnel entry to the area.
Radiation--Energy in the form of electromagnetic rays (radiowaves, light, X-rays, gamma rays) or
particles (electrons, neutrons, helium nuclei) sent out through space from atoms, molecules,
or atomic nuclei as they undergo internal change or resulting from particles and electromagnetic
radiation interactions with matter.
Radioactive decay--The spontaneous natural process by which an unstable radioactive nucleus
releases energy or particles.
Radioactive waste--Waste materials (solids'

liquids, or gases), contaminated with radionuclides.

Radioactivity--Product of radioactive decay of an unstable atom.
Radioiodines--Radioactive isotopes of iodine.
Radioisotopes--Radioactive isotope (see also radionuclide and isotope).
Radionuclide--An unstable nuclide that undergoes radioactive decay.
Radiotoxicity--The toxic or poisonous property released by ionizing radiation during radioactive
decay.
Radwaste--See radioactive waste.
RB--See reactor building.
RCS--See reactor cooling (or coolant) system.
Reactor ýnuclear)--A device in which a fission chain reaction can be initiated, maintained, andcontrolled.
Reactor building encroachment area--A staging area for decontamination operations establishedinside the reactor building, near the entry point.
Reactor building sump--The lowest part of the reactor building, designed to receive and hold, on
a temporary basis, drainage and overflow.
Reactor building--The structure housing the nuclear reactor. The building is specially constructed of reinforced concrete and is designed to withstand internal pressure and external
collisions. It also is fitted with gas-tight seals designed to contain all radioactivity within
the building and permit release of radioactive materials only under controlled conditions.

GL-13
Reactor cooling (or coolant) system (RCS)--This system, also known as the primary system, consistsof the reactor piping, steam generators, reactor coolant pumps, pressurizer, letdown piping,
instrument and sample pipelines, and other components (up to the first containment isolation
valve) that routinely come in contact with the reactor cooling water.
Reactor internals--The various component parts and systems within the reactor pressure vessel.
Reactor pressure vessel (RPR)--The steel vessel containing the reactor core; also referred tosimply as the reactor vessel or the pressure vessel.
Reactor pressure vessel head (RPVH)--The
reactor core and associated equipment.

closure, or lid, on top of the vessel that contains theReagents--Substances or solutions used to produce a chemical reaction.
Recirculating vacuum filter system (RVFS)--A water filtration system that is being used
to aid in the removal of sludge from contaminated water in the AFHB.
Recriticality--Reinitiation of a self-sustaining nuclear chain reaction.
Release fraction--The fraction of the radioactive material in a vessel assumed to be released
if the vessel is ruptured or otherwise breached.
Rem--A unit of dose equivalent which is proportional to the risk of biological injury.
Resin liners--Cylindrical metal vessels used to contain the resins and/or zeolites during purification of contaminated water by ion-exchange processes.
Resins--Solid or semisolid products of synthetic origin used in ion-exchange processes
for purification of liquids.
Riparian--Living or located along the shoreline of a body of water (at the land-water interface).
Roentgen (R)--Unit of gamma or x-ray exposure in air.
Energetic gamma rays which produced an
exposure of 1 R would deliver a dose equivalent of approximately 1 rem to a person.
Routine Operational Liquid Wastes (ROLW)--Water that-was initially nonradioactive, e.g., riverwater, which becomes contaminated by contact with low-levels of radioactive materials.
This
is not accident water as defined by the Aqreement between City of Lancaster, NRC, and Met-Ed.
RPV--See reactor pressure vessel.
RPVH--See reactor pressure vessel head.
RVFS--See recirculating vacuum filter system.
Safe shutdown earthquake (SSE)--The greatest intensity earthquake which a nuclear plant must be
designed to withstand and still

be able to be shutdown and maintained in a safe condition.

SDS--See submerged demineralizer system.
Seal elate assembly--A large ring that is fastened between the reactor vessel flange and thefuel transfer canal floor. When this ring is installed it forms a watertight seal allowing
the fuel transfer canal to be flooded.
Sediment--Solid material in the water that is not in solution, but is either distributed in thewater or settled out of it.
Selective absorbtion process--A separation process whereby a liquid is used to selectively-absorb (separate) a selected material (gas) from a source gas stream (air).

GL-14
Sequestering agent--See chelating agent.
Sere--A temporary biological community that occurs during a successional sequence on a givensTte.
SFP--See spent fuel pool.
Shielding--A barrier of solid or liquid material (e.g., lead, concrete, or water) which reduces
the intensity of radiation as it passes through and which can be used to protect personnel
from the damaging effects of ionizing radiation.
Short-lived isotope--A radioactive nuclide that decays so rapidly that a given quantity istransformed into its daughter products within a short period (usually those with a halflife of days or less).
Sludge--Sludge in the case of TMI-2 refers to a mixture of fine solid material which includes particles of cement dust, dirt, and resin beads, etc. that have settled out from a suspension in thewater.
Slurry--A free-flowing, pumpable suspension of fine solid material in liquid.
Smears--Dry or moistened filter papers that have been rubbed on areas of suspected radioactivity
and then subjected to analysis to determine the type and approximate amount of removable radioactivity.
Solubilize--To make a material soluble or to increase its solubility.
Solvent--A liquid capable of dissolving a solid or another liquid.
Source term--The quantity of radioactive material,

released by an accident or operation.

Spent fuel pool (SFP)--A water-filled pool designed specifically for the storage of spent
nuclear fuel.
Spool piece--A short, flanged section of pipe used to connect two pipelines.
Stators--Elements that provide the motive force in lifting.or lowering the reactor control
rods and axial power shaping rods.
Storage--Accumulation for later retrieval and disposal.
Storage can be performed for periods ofseveral months (interim) up to 10-20 years (long term).
Subcritical--The state in a nuclear reactor when the rate of neutron loss exceeds the rate ofneutron production.
Submerged demineralizer system (SDS)--A demineralizer system, similar to EPICOR II, that thelicensee proposes to construct in the spent fuel pool and use to process accident water.
Sump water--Water that has accumulated in a sump.
Sump--The lowest part of a building designed to receive and hold, on a temporary basis, drainage
and overflow.
Supernate--The clear liquid above a settled solid or precipitate.
Surfactants--Abbreviation for surface-active agent. Detergents or soaps which alter surface
tension or interfacial tension between water and other liquids or solids; used for decontamination purposes.
Technical Specifications--Requirements which are part of the NRC Operating' License for a nuclear
power plant.

GL-15
Temporary contamination control envelope--A barrier, such as a plastic tent, used to enclose aspace in an airtight envelope and prevent airborne contamination from being carried outside
the space.
Thermoluminescent detector (TLD)--A solid-state device used to measure radiation doses (see dosimeter).
TLD--See thermoluminescent detector.
Total-body dose--The radiation dose to the total body,
both external and internal radionuclides.

including the bone and all organs,fromTransuranic (TRU) waste--Wastes which contain or are contaminated by in excess of 10 nanocuries(one nanocurie = one-billionth of a curie) per gram of transuranic elements (with mass
numbers higher than uranium).
Tritiated water--Water in which one or both hydrogen atoms have been replaced by a tritium atom.
Tritium (H-3)--A radioactive isotope of hydrogen, approximately three times heavier than the"normal" (most abundant) form. The half-life is 12.5 years.
Trophic structure--A characteristic feature of any ecosystem measured and described either interms of the standing crop per unit area or energy fixed per unit area per unit time; often usedbroadly to refer to the various levels of a food chain.
TRU waste--See transuranic waste.
Turbidity--A measure of the degree to which sediments and other foreign matter are suspended
in water (cloudiness).
Unrestricted areas--An area to which access is not controlled for purposes of protection ofindividuals from exposure to radiation or radioactive materials.
Upper plenum assembly--A large cylindrical plate that fits into the reactor vessel to guidethe control rods and press down on the fuel assemblies in order to maintain proper alignment.
Viscous--Refers to liquids that are thick and not free flowing.
Vitrified wastes--Radioactive wastes immobilized, or solidified,in glass.

Wake-cavity effect--The region of turbulance immediately to the rear of a solid body,building, that is formed when wind currents flow over and around the object.

like aWhole-body dose--See total-body dose.

x/Q--Relative concentration; a term representing the concentration of a pollutant being emitted
to the atmosphere divided by the emission rate.
Zeolites--Any of various natural or synthesized silicates used to purify water.
Zircaloy--A zirconum-base alloy used as the cladding for fuel rods and for other reactor core
hardware.
Zone of saturation--A subsurface zone in which all the interstices are filled with water underpressure greater than the atmosphere.

NRC FORM 335

1. REPORT NUMBER (Assigned by DDC)

(7 77)

U.S. NUCLEAR REGULATORY COMMISSION

4. TITLE AND SUBTITLE(Add Volume No.,f appropriate)

BIBLIOGRAPHIC DATA SHEETNUREG-0683,

Volume I

2. (Leave blank)

Final Programmatic Environmental Impact Statement related
to decontamination and disposal of radioactive wastes reaccident at Three Mile
7. AUTHOR(S)

3. RECIPIENT'S ACCESSION NO.
5. DATE REPORT COMPLETED
MONTH

YEAR

March
9. PERFORMING ORGANIZATIONNAME AND MAILING ADDRESS (Include Zip Code)Three Mile Island Program Office
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, D. C. 20555

1981

DATE REPORT ISSUED

mRPh

YE181

6. (Leave blank)
8. (Leave blank)

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code)Three Mile Island Program Office

10.PROJECTrTASK/WORK UNIT NO.

Office of Nuclear Reactor Regulation

11. CONTRACT NO.

U.S. Nuclear Regulatory Commission
Washington, D. C. 20555
13. TYPE OF REPORT

Final Programmatic Environmental
Statement

ImpactfPERIOD COVERED (Inclusive dates)
I

15. SUPPLEMENTARY NOTES

14. (Leave blank)

16. ABSTRACT (200 words or less)

A Final Programmatic Environmental Impact Statement (PEIS) related to the decontaminationand disposal of radioactive wastes resulting from the March 28, 1979, accident at Three
Mile Island Nuclear Station, Unit 2 (Docket No. 50-320) has been prepared by the Officeof Nuclear Reactor Regulation of the Nuclear Regulatory Commission in response to a
directive issued by the Commission on November 21, 1979.
This statement is an overall
study of the activities necessary for decontamination of the facility, defueling, anddisposition of the radioactive wastes. The available alternatives considered ranged from
implementation of full cleanup to no action other than continuing to maintain the reactor
in a safe shutdown condition. Also included are comments of governmental agencies, otherorganizations, and the general public on the Draft PEIS on this project, and staff
responses to these comments.

17. KEY WORDS AND DOCUMENT ANALYSIS

17a. DESCRIPTORS

17b. IDENTIFIERSOPEN-ENDED TERMS

18

AVAILABILITY STATEMENT

Unlimited
NRC FORM 335 t7 77)

19, SECURITY CLASS (Tnst

p rn6/

21 NO. OF PA-C;ES

Unclassified
ITY QýfSSd(Thlspaqe)
uric Iassi
Tlea

U20. SýCU

22. PRICE

5UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D. C. 20555
OFFICIAL BUSINESSPENALTY FOR PRIVATE USE, $300

-POSTAGE

AND FEES PAID

U.S. NUCLEAR REGULATORYCOMMISSION