Three Mile Island
Accident of 1979
Knowledge
Management Digest
Recovery and Cleanup
NUREG/KM-0001, Supplement 1
Office of Nuclear Regulatory ResearchTMI-2 radioactive material location map (see GEND-057).

Contentsx

Digital Versatile Disc (DVD) Contents ............................................. v
1

Introduction ............................................................................... 1

2

Management and Oversight ...................................................... 7

3
4

Plant Stabilization .................................................................... 41
Worker Protection.................................................................... 69

5

Data Acquisition and Analysis ................................................. 81

6

Waste Management .............................................................. 101

78

Decontamination ................................................................... 131
Defueling ............................................................................... 147

9After Defueling ...................................................................... 179

10 Postscript .............................................................................. 195
11 DVD Navigation and Interpretation ........................................ 199
12 Contributions to Supplement 1 .............................................. 203
13 Endnotes ............................................................................... 205

On April 15, 1990, the last rail shipment of fuel debris packaged inthree Model 125-B shipping casks departs Three Mile Island for the
Department of Energy’s Idaho National Engineering Laboratory.

iii

Above: First entry into the reactor building (shown) occurred on
July 23, 1980. Heavy duty outerwear provided high-energy beta ray protection.
Below: Water level in the basement of the reactor buidling reached the firststairwell landing indicating about 8 feet of water accumulation. One-third ofthis highly radioactive water came from leakage of non-radioactive river water
coolant from a leaking reactor building air-handling cooler.

iv

Digital Versatile Disc (DVD) Contents
Supplement 1, Disc 1 (NUREG/KM-0001, DVD-2) Folder
Status and Summary Reports
TMI Program Office Weekly Status Reports
NRC Preliminary Notifications
Licensee Event Reports
NRC Status Reports
NRC Annual Reports
DOE/National Laboratory Status Reports
DOE/National Laboratory Conference Papers
GPU Status Reports
Other Status Reports

4a-TMIPO
4b-PN
4c-LER
4d-NRC
4e-Annual
4f-DOE
4g-DOE-CONF
4h-GPU
4i-Other

Licensing Actions
Operating License Amendments
OrdersOrder Amendments
Regulation Exemptions
Recovery Operations Plan Changes
Organization Plan Changes

5a-OL Amend
5b-Order5c-Order-Amend
5d-Exemption
5e-ROP-Change
5f-OP-Rev

Management and OversightGeneral Management and OversightCommission/Licensee Correspondence
Commission SECY Papers/Staff Requirements
Advisory Groups
Guidance-PEIS
General Project Design Criteria

6a-Plans
6b-Commission
6c-SECY
6d-Advisory
6e-Guide-PEIS
6f-Design

One of many attachments to long-handled defueling tools used to move coredebris in the reactor vessel.

vSupplement 1, Disc 2 (NUREG/KM-0001, DVD-3) Folder
Plant Stabilization
General Plant Stabilization
Planning Meetings
Recovery Procedures
Industry Advisory Group Reports
Reactor Building Entries

7a-Stabilization
7b-Meeting
7c-Procedure
7d-IAG-Report
7e-RB-Entries

Worker Protection

8a-Protection

Data Acquisition and Analyses
Recovery Data Analyses
Accident Data Analyses
Instrumentation and Electrical Evaluations
Criticality Analyses

9a-Data-Recovery
9b-Data-Accident
9c-Data-Equipmt
9d-Criticality

Decontamination

11a Decon

Supplement 1, Disc 3 (NUREG/KM-0001, DVD-4) Folder
Waste Management
General Waste Management
Water Processing: EPICOR
Water Processing: SDS
Solid Waste

10a-Waste
10b-EPICOR
10c-SDS
10d-Solid

Defueling

12a-DefuelAfter Defueling

13a-Post-Defuel

Pre-Accident: Plant References
Final Safety Analysis Report
System Descriptions
Plant Procedures
Other Plant References

14a-FSAR
14b-Sys-Desc
14c-Procedure
14d-OtherviAbove: Close up view of the upper grid rib section where fuel bundles wereonce held in place. Indication of melting of the stainless steel rids wasobserved during an inspection of the underside of the upper plenum as it
was lifted from the reactor vessel. Below: Broken fuel rods in the void areaat the top of the damaged reactor core.

vii

Oyster Creek Unit 1 (left) and the proposed Unit 2. In 1968, Jersey Central Power and Light Company applied for a reactor construction
permit for the proposed unit. Instead, the plant was built on Three Mile Island adjacent to Unit 1. The new application included
Metropolitan Edison and Pennsylvania Electric as additional owners.

viii

1 Introduction
The safe, expeditious recovery and cleanup of Three Mile Island Unit 2(TMI-2), including removal of the fuel from the accident-damaged reactor,
were necessary for the long-term protection of public health and safety andthe environment. The cleanup campaign ensured that the TMI site did not
become a long-term or permanent waste repository.1 The recovery activities
that unfolded at TMI-2 in the weeks and months (and then years) after the
March 28, 1979 accident were the result of a multi-organizational effort that
included hundreds of dedicated and highly-skilled individuals.
Implementation of recovery and cleanup activities was the responsibility of
the licensee with support from their many contractors. Organizations that
supported the licensee included the original architect engineers of TMI
Units 1 and 2 (Gilbert Associates and Burns and Roe, respectively2); theTMI-2 nuclear steam supply system vendor (Babcock & Wilcox); many
volunteers from other nuclear power plants; the U.S. nuclear industry; andseveral international organizations.3
The U.S. Nuclear Regulatory Commission (NRC) was responsible for the
regulation of TMI-2 cleanup operations to ensure the health and safety ofthe public, and the TMI-2 occupational workforce, as well as the protectionof the environment.4 The U.S. Department of Energy (DOE) was
responsible for supporting an extensive research program, as directed by the
U.S. Congress. In addition, DOE agreed to the removal and disposition of
the entire reactor core for research, as well as certain solid nuclear wastes
generated during the cleanup of TMI-2.5, 6, 7, 8 The DOE and its national
laboratories provided much-needed technical support to the licensee and the
NRC in almost every aspect of the many TMI-2 research and recovery
programs.9 The environmental agencies and nearby communities from thestates of Pennsylvania and Maryland were responsible for ensuring thatwater qualities of the Susquehanna River and the downstream Chesapeake
Bay were not adversely impacted by the damaged plant, or by the cleanup
activities.10
Need for Cleanup. The NRC’s Programmatic Environmental Impact
Statement (PEIS) related to decontamination and disposal of radioactive
waste resulting from the TMI accident (NUREG-0683) concluded that thedecontamination of the TMI-2 facility, including the removal of the nuclear
fuel and radioactive waste from the TMI site, was necessary for the
long-term protection of public health and safety. The PEIS also concluded
that methods existed, and could be suitably adapted to perform the cleanup
operations with minimal releases of radioactivity to the environment.11
1

The cleanup operations removed sources of potential radiation exposure that
posed risks to the health and safety of the workers and the public.
Accident-generated radiation sources were present in the form of airborne
contamination; contaminated waste water; absorption of radioactive material
on building and equipment surfaces; contaminated sludge in tanks, sumps,
and building basement floors; contaminated filter cartridges anddemineralizer resins; and damaged fuel and reactor components. As long asradioactive water occupied sumps and tanks, there existed a possibility ofleakage into the groundwater, and subsequently, into the SusquehannaRiver. The contaminated water was also a source of direct radiation toworkers requiring access to buildings in order to perform critical
maintenance and repairs needed to keep the reactor in a safe-shutdown
condition.12
The PEIS categorized cleanup into four fundamental activities: building andequipment decontamination; fuel removal and decontamination of theTMI-2 reactor building: a comparison (approximately to scale). The TMI-2
reactor building is 130 feet in diameter and rises 191 feet from the floor to thedome, with a volume of about 2.4 million cubic feet. The U.S. Capitol Rotunda
is 96 feet in diameter and rises 180 feet from the floor to the canopy, with avolume of about 1.3 million cubic feet.

2primary coolant system; treatment of radioactive liquids; and packaging,handling, storage, and transportation of radioactive solid wastes.13
About This Supplement. The main objective of this supplement is toprovide key historical documents in electronic format that were issued
during the recovery and cleanup efforts. Brief overviews of various
structures, systems, equipment, and activities that were associated with therecovery and cleanup of TMI-2 are provided in the written portion of thisNUREG/KM to describe the contents of the many document collections in
the accompanying DVDs.
Thorough overviews of the TMI-2 recovery and cleanup, including lessons
learned, can be found in the Electric Power Research Institute (EPRI) report,“The Cleanup of Three Mile Island Unit 2, A Technical History:
1979 to 1990” and the special volume of the Nuclear Technology journal of
the American Nuclear Society documenting 138 papers presented at the
TMI-2 topical meeting in 1988.a
The seven major aspects of the recovery and cleanup, as presented in the
EPRI report, were used to organize the contents in this supplement into thefollowing sections: management and oversight; plant stabilization; worker
protection; data acquisition and analysis; radioactive waste management;
decontamination; and defueling; an additional section on after-defuelingactivities follows. This supplement chronicles those activities, which begana week following the accident, and ended with the completion of disposal of
accident-generated water, and entry into post-defueling monitored storage
in 1993.
The document collections in this supplement were mainly derived frompublicly available correspondence, including attached reports, between thelicensee and the NRC, and the results of research activities sponsored by the
NRC and DOE. In all, the accompanying DVDs contain about
4,000 documents. Although an attempt was made to find and include a wide
range of key documents, the collections provided on the DVDs are notcomplete. As such, a document collection might not provide a complete
chronology of recovery, cleanup, and regulatory actions. A listing of
documents in each document-collection folder is provided in spreadsheet
format on each DVD (see the DVD folder, Common). Also, a list of more
a EPRI-NP-6931 and many of EPRI’s historical reports on the TMI-2 accident and cleanup

are currently (at the time of this publication) available from the EPRI’s website. Individual
papers from Nuclear Technology, Vol. 87, Nos. 1 through 4, are currently available from theAmerican Nuclear Society’s website.

3

than 25,000 TMI-2 records indexed in the Public Legacy Library, datingfrom 1979 to 1999, is included in the spreadsheet.b
How to Use this Supplement. A few suggestions for navigating throughthis supplement, and the many documents on the enclosed DVDs, areprovided at the end of this NUREG/KM (see the section on DVD
Navigation and Interpretation). Documents in the DVD folders, Status andSummary Reports, Licensing Actions, and Management and Oversight,
might be applicable to all sections in this supplement. The historical
documents provided on the DVDs are for reference only, and are not official
NRC records. End notes to this supplement provide filenames of the cited
documents on the DVDs. The units of measure used in this NUREG/KM
reflect those used in the original source reports. In some cases, conversions
were provided in the original reports. Refer to the back cover for conversion
factors and formulas.

TMI-2 reactor building under construction. Shown is the top dome of thereactor building. The Susquehanna River can be seen in the background.

b Public Legacy Library of the NRC Agencywide Documents Access and Management

System (ADAMS) provides bibliographic citations for earlier documents during the period
from 1979 through 1999, which are available in microfiche formats. This library can beviewed from the NRC public website.

4Installation of a reactor vessel and steam generators inside containment
similar to the TMI-2 design. Concrete shielding structures will be build
around these components. Notice the ladder next to the reactor vessel. (Froma training visual aid from the Babcock & Wilcox Company, the fabricator of
the primary reactor components at TMI-2.)

5

Above: NRC’s Harold Denton (left) and Victor Stello briefing President Jimmy Carter,
Governor Richard Thronburgh, and others at the Air National Guard Facility in
Middletown, Pennsylvania. Below: Senior managers from the Office of Nuclear
Reactor Regulation in the NRC office trailer at Three Mile Island. Left-to-right: Roger
Mattson, Harold Denton, Denwood Ross (on telephone), and Victor Stello.

6

2 Management and Oversight
The NRC was responsible for the regulation and oversight of TMI-2 cleanup
operations to ensure the health and safety of the public, and the TMI-2
occupational workforce, as well as the protection of the environment.
Implementation of recovery and cleanup activities was the responsibility of
the licensee with support from their many contractors.14 NRC’s involvement
covered two major areas: approving the recovery methods employed by thelicensee and responding to public concerns over radiation exposure resulting
from the accident and cleanup activities.15 In addition, the TMI-2 license
remained under the regulatory requirements of an operating reactor (Title
10, “Energy,” of the Code of Federal Regulations (10 CFR) Part 50,“Domestic Licensing of Production and Utilization Facilities”); many
formal licensing actions were required to modify the pre-accident operating
license to reflect the safety requirements of the plant in its unique shutdown
condition.16
Initially, the NRC did not have any specific guidelines or criteria pertaining
to the recovery effort but followed its traditional regulatory practices ofreacting to licensee proposals. This process sometimes required a time
consuming and iterative exchange of written correspondence for questions
and answers. Some actions required the majority approval of the NRC
Commissioners, even for minor technical and radiological problems.17 At
the end of the first year after the accident, it became apparent that a new
approach was needed for maintaining the TMI-2 license during the cleanup
and NRC oversight of the licensee. This approach consisted of an increased
onsite presence, as well as a new decision-making process for NRC
approval of cleanup activities.18 The PEIS provided the environmental
impact bases for all cleanup activities; the Commission policy statement thatendorsed the PEIS provided staff with the authority to approve most cleanup
activities.19
The NRC accomplished its regulatory responsibilities for all post-accident
operations at TMI 2 through licensing actions; safety evaluations ofrecovery and cleanup activities; inspections; daily interactions with thelicensee and their contractors; communications with State and local
governments and the public; coordination with other Federal agencies
involved in the cleanup; and sometimes direction from the NRC
Commissioners by majority vote. Formal licensing actions were required for
proposed changes to NRC orders, the facility license, technicalspecifications, the recovery operations plan, the organization plan, andexemptions to regulations.20
7

The licensee’s organizational structure and functions changed as the cleanup
progressed through various stages of the recovery effort. Plans, schedules,
resources, and cost estimates changed as more information became known
about the condition of the damaged reactor core; reliability of plant systems
and structures; and radiological characterization of accident-generated
water, building structures, reactor components, and systems. Variousorganizations formed working groups to provide guidance on tackling
specific issues, problems, and research activities. Independent oversightgroups reviewed, monitored, and advised on the overall direction ofrecovery and cleanup plans and activities.21 NRC staff attended meetings
conducted by the licensee’s working groups and independent oversightgroups.22
The licensee and its contractors made changes to the facility in order tostabilize the reactor and clean up the damaged plant. New facilities andsystems were built, and existing ones were used in different capacities thanSide view of reactor coolant system components. Shown are the reactor vessel(center), both once-through steam generators (left and right), one of four
reactor coolant pumps (upper left), both hot-legs (or “candy canes”) from thereactor vessel to top of the steam generators, and one of four cold-legs (or “Jlegs”) from the bottom of the steam generators to the reactor coolant pumps.

8

those for which they were originally designed and approved. The changing
configuration and operation of the damaged plant during the cleanup stages
required changes to licensing documents. These changes imposed new
requirements to ensure safety and eliminated pre-accident requirements thatno longer applied to the shut-down plant.23 To implement these changes, thelicensee developed formal documents such as plans, system description
reports, safety analysis reports, technical evaluation reports, environmental
assessment reports, experiments, and procedures. Formal documentation
was required for most, if not all, data-collection and cleanup activities.
Report of the Special Task Force on TMI Cleanup. Eight weeks before
the first anniversary of the accident, a special NRC task force reported avariety of resource, policy, legal, and technical factors that could adversely
influence, if left unchanged, the forward motion of the cleanup process andschedule. The task force observed frustration with the pace of the cleanup;
the lack of interim criteria for the conduct of daily activities pending
completion of the Programmatic Environmental Impact Statement (PEIS);
the tedious NRC decision-making process; erosion of what was once ahigh-priority program; and strong public opinion on cleanup activities. In
their report to the NRC Executive Director of Operations (EDO) on
February 28, 1980, the task force made 15 recommendations stemming fromtheir quick (9-day) evaluation of cleanup operations.
The next day, the report was endorsed by the EDO and forwarded to theCommissioners with commitments to: (1) establish conservative interim
criteria and a proper level of authority for staff decisions to permit small
radiological releases associated with plant maintenance and data-gathering
activities pending PEIS completion; (2) expedite the completion ofassessments supporting the purging of radioactive gases from the reactor
building and the completion of the PEIS; (3) increase permanent staffing of
the NRC’s Middletown office and onsite support group, including a fulltime spokesperson to keep State and local government officials and the
public informed on a continuing basis about the progress and the status ofcleanup operations and about future plans; (4) transfer assignments to
prepare environmental assessments to headquarters; and (5) prepare a plan
and a schedule of activities for proceeding with the cleanup.24
The NRC Commissioners took actions that addressed other concerns
expressed by the Special Task Force. The Commissioners approved bymajority vote the interim criteria for radiological releases from data
gathering and maintenance activities (SECY-80-175 and NUREG-0681), a
contingency plan for cleanup in case of the financial failure of the licensee(SECY-80-416A and NUREG-0689), and the formation of the NRC
9Advisory Panel for the Decontamination of TMI-2. In their April 27, 1981
policy statement that endorsed the PEIS, the Commissioners stated how andby whom major cleanup decisions would be made after the PEIS wascomplete, as well as the expected role of the PEIS in making thosedecisions. The Commissioners also stated that the licensee should accelerate
the pace of the cleanup and that they expected to receive continuous advice
from the TMI-2 Advisory Panel on major activities required to accomplish
expeditious and safe cleanup of the TMI-2 facility.25
Document collections provided under the Management and Oversightcategory include Commission Actions, Licensing Actions, Planning andGuidance, and collections from the Advisory Groups and from the NRC’s
TMI Project Office. Key documents in these document collections are
summarized below.

Isometric view of the reactor coolant system components.

10Commission Actions. The five NRC Commissioners (also known
collectively as the “Commission”) sometimes took actions to direct NRC
staff to take certain licensing actions relating to policy issues. Issues before
the Commission are decided by majority vote. The Commission’s decisionmaking tools used during the TMI-2 recovery and cleanup included policy
statements, Commission orders, SECY papers, staff requirement
memoranda, and general correspondence.
•

Policy statements. A “Statement of Policy” is not a regulation and does
not impose specific regulatory requirements, but rather, provides theCommission’s rationale and motivation for future regulatory positions.
Several policy statements that the Commission issued were directly
applied to TMI accident investigation and recovery activities. These
Commission policy statements (provided in the DVD folder, Policy
Statements) are summarized below:o

On May 25, 1979, the Commission directed staff to prepare an
environmental assessment, with opportunity for public comment,
regarding proposals to decontaminate and dispose of radioactive
contaminated waste water. The Commission’s statement required
assessments on decontamination of intermediate-level waste waterusing the EPICOR II system; alternatives to discharge of waste
water into the Susquehanna River; and decontamination anddisposal of high-level waste water. The statement permitted
discharge of pre-accident waste water decontaminated by the

The five NRC Commissioners participated in an all-day public hearing inHarrisburg, PA on November 9, 1982. Residents and community groups in theThree Mile Island area were invited to express their views and concerns
regarding the future of TMI-2. Shown (left to right), Commissioners Roberts
and Ahearne, Chairman Palladino, and Commissioners Gilinsky andAsselstine.

11

existing EPICOR I decontamination system and discharge ofindustrial waste water (water slightly contaminated because ofleakage from secondary plant service support systems), as
consistent with the facility operating license and NRC regulations.
However, restrictions were imposed on the allowed discharges of
EPICOR I and industrial waste water, the discharge of other(accident) waste water, and the operation of EPICOR II.26
oOn November 21, 1979, the Commission directed staff to prepare aprogrammatic environmental impact statement (PEIS) on thedecontamination and disposal of radioactive waste resulting fromthe accident.27 There were no identifiable legal requirements of theAtomic Energy Act of 1954, as amended, that required different
radiological release criteria be applied to the cleanup of TMI-2 thanwere applied to an operating plant. In keeping with the purposes of
the National Environmental Policy Act (NEPA), the Commission
decided to prepare a PEIS on the decontamination and disposal of
TMI-2 radioactive wastes.28 Under the terms of the policy
statement, the Commission stated that development of the

A computer generated cut-away drawing of the reactor building showing
reactor coolant system components and “D-Ring” concrete radiation shields.

12

programmatic statement would not preclude prompt Commission
action when needed. Such prompt actions would requireconsideration of the advice of the Council on Environmental
Quality about the Commission’s NEPA responsibilities and would
require an environmental review with opportunity for publiccomment (in accordance with the May 25, 1979 policy statement).
The policy statement also allowed rapid action for an emergency
situation. In this situation, the Commission would consult theCouncil to the extent practical.29oOn September 26, 1980, the Commission issued a policy statement
concerning the Pennsylvania Public Utility Commission’s order tothe licensee to cease and desist from using any operating revenues
for cleanup and restoration costs at TMI-2 which were not coveredby insurance. The NRC Commissioners emphasized that all NRC
health, safety, and environmental requirements applicable to TMI-2
must be fully complied with by the licensee, regardless of whether
or not these requirements appeared to conflict with the Utility
Commission’s order.30, 31oOn April 27, 1981, the Commission issued a policy statement
endorsing the final Programmatic Environmental Impact Statement(NUREG-0683) related to the decontamination and disposal ofradioactive wastes resulting from the accident. The Commission
stated that the licensee should accelerate the pace of the cleanup tocomplete expeditiously all decontamination activities, consistent
with ensuring protection of public health and safety, and theenvironment. The policy statement also indicated that, as thelicensee proposed specific major decontamination activities, the
NRC staff would determine if these proposals, and associatedimpacts that were predicted to occur, fell within the scope of thosealready assessed in the PEIS. With the exception of the dispositionof processed accident-generated water, (which the Commissioners
wanted to decide on later), the staff was allowed to act on eachmajor cleanup activity without the Commission’s approval if theactivity and associated impacts fell within the scope of those
assessed in the PEIS. The policy statement required the staff to keep
the Commission informed of staff actions before staff approval ofmajor activities. Further, the statement indicated that the cleanup
should be carried out in accordance with the criteria in Appendix Rof the PEIS, “Proposed Additions to Technical Specifications for
TMI-2 Cleanup Program,” as well as in conformance to the existing
operating license, and previously-imposed orders. The Commission
13expected to receive continuous advice from the Advisory Panel on
the Decontamination of TMI-2 regarding major activities required
to accomplish expeditious and safe cleanup of the TMI-2 facility.32
Commission orders. Commission orders address appeals or motions
before the Commission in such matters as amendments to nuclear
facility licenses, license transfers, license renewals, and enforcement•

Components inside the reactor building.

14matters.33 Two Commission orders resulted in the modification of theoperating license at TMI-2 to require prompt operation of the
EPICOR II system to decontaminate intermediate-level radioactive
waste water in the auxiliary building, and to release krypton-85 from thereactor building’s atmosphere by controlled purging (see discussions
below). 34, 35 These orders also resulted in amendments to the technicalspecifications. These orders are summarized later in this section andprovided on the DVD (see the DVD folder, Orders).
•

SECY papers and staff requirements memoranda. The primarydecision-making tool for the Commission is a written issue paper
submitted by NRC staff to the Commission. Policy, security,
rulemaking, adjudicatory matters, and general information are provided
in a stylized document referred to as a “SECY Paper.” After theCommissioners vote on a SECY Paper, the Office of the Secretary(SECY) records the decision in a memorandum to the staff called a“Staff Requirements Memorandum” (SRM), and also issues a“Commission Voting Record” which includes the record of votes andindividual views of the Commissioners.36 SRMs may be issued
following Commission meetings to document any discussion or requests
made at the meeting. Many of the SECY papers and SRMs related to the
TMI-2 recovery and cleanup efforts and found in the public record areprovided on the DVD (see the DVD folder, Commission SECY
Papers/Staff Requirements).

•

General correspondence. The Chairman is the official spokesperson for
the agency. On a few occasions, the Chairman exchanged
correspondence with licensee corporate executives. The examples ofsuch correspondence that were found in the public record are provided
on the DVD (see the DVD folder, Commission/Licensee
Correspondence).
One notable letter from the NRC Chairman to the President of General
Public Utilities (GPU) Corporation, dated January 12, 1981, responded
to an earlier letter from GPU concerning near-term planning of thecleanup of TMI-2 in light of an order issued by the Pennsylvania Public
Utility Commission. The State agency ordered the licensee to cease anddesist from using any operating revenues for cleanup and restoration
costs which were not covered by insurance. The Commission responded
by issuing a Statement of Policy on the matter (see above) which was
discussed in the Chairman’s response. In addition, the response
provided a list of activities required to be performed during the periodof ongoing discussions between the licensee and the State agency. The
15

list provided a “road map” of minimum activities required in the nearterm to keep the TMI-2 reactor in a safe condition, and activities
required to reduce potential threats over a longer term.37
Licensing Actions. Changes in the facility’s post-accident mode ofoperations required unique regulatory and licensing actions. In order toproperly reflect evolving plant status, the NRC issued orders, modified
those orders, approved license amendment requests, and granted relief fromcertain regulatory requirements. Most of the NRC correspondence
approving each licensing action (including the correspondence supporting

Internal view of the reactor pressure vessel at TMI-2. (From a training visual
aid from the Babcock & Wilcox Company, the fabricator of the TMI-2 reactor
vessel. This diagram was used to brief President Carter during his visit at TMIon April 1, 1979, see photo at the beginning of this section).

16

NRC safety evaluations and environmental reviews, where applicable) is
provided on the DVD as indicated below:•

Orders. The Commissioners (by majority vote), or designatedsenior-level officials, may issue an order that directs the licensee to takea prescribed action. Key orders related to the recovery effort aresummarized below. These orders are provided on the DVD (see theDVD folder, Orders).
o

On July 20, 1979, an Order for Modification of License was issued
to suspend the power operation of TMI-2 and require that thefacility be kept in a shutdown condition, in accordance with
NRC-approved operating and contingency procedures.38oOn October 18, 1979, an Order for Modification of License wasissued to require the licensee to promptly operate the EPICOR II
filtration and ion-exchange decontamination system todecontaminate intermediate-level radioactive waste water held intanks in the TMI-2 auxiliary building. The order further required thelicensee to monitor EPICOR II discharge paths, maintain suitabletankage in TMI-1 as a contingency, and solidify spent resin before
its shipment offsite.39, 40 The requirement to solidify resin was later
removed in a modification to that order.41o

On February 11, 1980, an Order for Modification of License
established new technical specifications governing operation of the
facility in what was called a “recovery mode.” The proposed
technical specifications (also known as “recovery technicalspecifications”) took into account the present condition of plantsystems to ensure that the unit would remain in a safe and stableposture during the recovery mode.42 The order explicitly prohibited
venting or purging, or other treatment of the reactor building’s
atmosphere; the discharge of water decontaminated by the
EPICOR II system; and the treatment and disposal of high-level,
radioactively-contaminated water in the reactor building, until eachof these activities had been approved by the NRC, in accordancewith the Commission’s November 21, 1979 policy statement. The
attached safety evaluation and environmental assessment explainedthat these activities could have been allowed under the same
effluent limitations as would apply in the case of a normallyoperating facility, had the Commission not determined that public
interest warranted prohibiting these undertakings pending
completion of an environmental review. The order provided that the17existing pre-accident technical specifications imposed for theprotection of the environment (Appendix B to the facility operating
license), including the established limitations on effluent releases
and discharges, were unchanged and would remain in effect except
as provided in the order. These prohibitions effectively precluded
the planned release of any radioactive liquid and gaseous material
from TMI-2 without prior Commission approval (by majority vote).
Low-level solid waste, such as rags and clothing, generated during
cleanup operations in the auxiliary building, was permitted to betransported offsite to a commercial, low-level radioactive waste
burial facility.43
The order did not change existing pre-accident Appendix B
technical specifications, except that the licensee’s pre-accident
management organization for activities addressed by thosespecifications was replaced with sections in the proposed
Appendix A technical specifications that addressed current
requirements imposed on the licensee’s management organization
for all post-accident licensed activities.44 The order proposed that

The reactor pressure vessel on its way for installation at TMI-2. (This photo
was taken either during the loading operation at the Babcock and Wilcox
fabrication facility in Mount Vernon, Indiana, or during the unloading near
Havre de Grace, Maryland.)18

the facility operating license would be formally amended to include
the proposed technical specifications. As discussed later, this did
not happen until many years later. The technical specifications
originally proposed were documented in NUREG-0432.45 The
related safety evaluation report and environment assessment were
documented in NUREG-0647.46
o

On June 12, 1980, an Order for Temporary Modification of License
was issued to require the licensee to release krypton-85 from thereactor building’s atmosphere by controlled purge. The order set
offsite dose limits for the purge. The related final environmental
assessment for the reactor building purge was documented in
NUREG-0662.47

o

On June 18, 1981, an order was issued to require the licensee topromptly operate the submerged demineralizer system with effluent
polishing by the EPICOR II system to process theintermediate-level contaminated water in the auxiliary building’s
tanks, and the highly-contaminated water in the reactor building’s
sump, and in the reactor coolant system. The related safety
evaluation report was documented in NUREG-0796.48•

Modification of orders. From time to time, some orders required
modification to clarify language, or reflect changes in plant status, as aresult of ongoing decontamination and maintenance efforts. The most
overarching order that required numerous amendments was related tothe proposed technical specifications. Although the NRC aimed toincorporate the proposed recovery technical specifications in facility
operating license (Appendix A technical specifications to the facility
operating license), several requests for a hearing were filed inconnection with the order. After that time, discussions with therecognized parties led to mutual agreement in the areas of concern, andresulted in the resolution of all outstanding issues and the withdrawal of
all petitions. On November 8, 1985, the NRC Atomic Safety Licensing
Board issued an order terminating the proceeding. During the pendencyof this matter, a number of changes in the proposed technicalspecifications were necessary, each requiring a Modification of Order.49
Most modifications and related safety evaluations are provided on the
DVD (see the DVD folder, Order Amendments).

•

Changes to the recovery operations plan. The recovery operations plan
defined the surveillance requirements to be performed to ensure
equipment operability as required by the plant’s technical specifications.
19

The recovery operations plan was included as Section 4 of the proposed
technical specifications. However, the plan was not considered a part ofthe proposed technical specifications. As such, changes made tosurveillance requirements were approved by NRC staff outside of a
Modification of Order.50 Some changes were issued concurrently with a
Modification of Order, instituting corresponding changes to the
proposed technical specifications. Approvals of most changes to therecovery operations plan are provided on the DVD (see the DVD folder,
Recovery Operations Plan Changes).
Changes to the organization plan. The licensee’s organization plan
provided the organizational structure (e.g., charts) for management of
TMI-2 recovery operations, including the support functions ofengineering and administration. The plan was cited in the organization
section of the proposed technical specifications. The NRC approved thelicensee’s concept of providing charts of the TMI-2 recovery
management in the organization plan, instead of the proposed technicalspecifications, so that future changes could be made effective in atimely manner. Changes made to the organization plan required NRC
approval, but certain changes did not require a Modification of Order.51
The approvals of most changes to the organization plan are provided on
the DVD (see the DVD folder, Organization Plan Changes).

•

View of the cold-water basin inside a natural draft cooling tower at Three
Mile Island. The cross columns support the curved concrete wall of the tower.

20•

Amendments of license. The TMI-2 facility operating license, which
included the technical specifications, was amended, modified, extended,and transferred during the three decades following the accident.
A chronology of key amendments is summarized below. Most
amendment actions are provided on the DVD (see the DVD folder,Operating License Amendments).
o

On March 12, 1980, Amendment No. 10 was issued to revise the
Appendix B technical specifications relating to the operation of the
EPICOR II filtration and ion-exchange decontamination system todecontaminate intermediate-level waste water held in tanks in the
TMI-2 auxiliary building. This amendment was approved under the
NRC Commissioners’ Order of October 18, 1979.52

o

On June 24, 1980, Amendment No. 11 was issued to approve a
temporary change to Appendix B technical specifications relating to
the bypassing of interlocks between the reactor building exhaust
monitors and dampers during purging of the reactor building’s
atmosphere. This amendment was approved under the NRC
Commissioners’ Order of June 12, 1980.53

o

On December 30, 1981, Amendment No. 18 was issued to reflectthat GPU Nuclear Corporation was added as the licensee for TMI-2
and replaced Metropolitan Edison Company as the licensee
authorized to operate (maintain) TMI-2.54o

On January 27, 1986, Amendment No. 26 was issued to formallyincorporate the proposed “recovery mode” technical specifications
that were established by the February 11, 1980, Order andsubsequently modified numerous times. The amendment wasallowed by the NRC Atomic Safety Licensing Board following theresolution of all outstanding issues and the withdrawal of all
petitions related to the order. Changes to the recovery technicalspecifications now required a license amendment (except forchanges to the recovery operations plan and organization plan,
which continued to be approved through a separate process).55o

On May 27, 1988, Amendment No. 30 was issued to extensively
revise the technical specifications by aligning licensing
requirements with appropriate plant conditions, through theremainder of the cleanup operations. Three facility modes were
defined in the revision, which allowed transition from the defuelingphase by incorporating technical specifications that were applicable21

during specific phases or modes of the cleanup. These modes wereas follows:56
Mode 1: The reactor coolant shall be subcritical, with an averagereactor coolant temperature of less than 200 degrees Fahrenheit.
Mode 2: This mode shall exist when the following conditions are
met: (a) the reactor vessel and reactor coolant system are defueled
to the extent reasonably achievable; (b) the possibility of criticality
in the reactor building is precluded; and (c) there are no defuelingcanisters containing core material in the reactor building.
Mode 3: This mode shall exist when the conditions for Mode 2 aremet, and no defueling canisters containing core material are stored
on the TMI-2 site.
o

On September 11, 1989, Amendment No. 35 was issued to modify
the technical specifications, by deleting the prohibition for
disposing of accident-generated water. The amendment retained therequirement for prior NRC approval of procedures associated withthe disposal of the accident-generated water. The associated NRC
safety evaluation approved disposal of the accident-generated waterby evaporation, subject to restrictions.57

East view of TMI-2 buildings (left of TMI-1).

22

•

o

On September 14, 1993, Amendment No. 45 was issued to modify
the facility operating license to be a possession-only license. The
NRC planned to issue the post-defueling monitored storage (PDMS)
technical specifications after the licensee substantially satisfied the
PDMS commitments and requirements.58

o

On December 28, 1993, Amendment No. 48 was issued toextensively modify the technical specifications in ways consistent
with the licensee’s plans for post-defueling monitoring storage ofthe facility.59oOn June 21, 1995, Amendment No. 49 was issued to extend thelicense until April 19, 2014, to allow both units at TMI to be
decommissioned at the same time.60

Exemptions. The NRC granted exemptions from certain requirementsof the regulations for nuclear power plants, but only under special
circumstances as permitted in the regulation (see Title 10, “Energy,” ofthe Code of Federal Regulations (10 CFR) 50.12, “Specific
Exemptions”). Exemptions were necessary at TMI-2 because of theplant’s damaged configuration and changing status during cleanup. The
approvals of most exemptions are provided on the DVD (see the DVD
folder, Regulation Exemptions).

Planning and Guidance. Planning and guidance documents that were
essential to the formulation of recovery and cleanup plans and activities
included the NRC’s Programmatic Environmental Impact Statement (PEIS);
the NRC policy statement endorsing the PEIS; the licensee’s planningstudies; the recovery quality assurance plan; and the general project design
criteria document. These and other planning and guidance documents are
summarized below. These documents and others, except as noted in thesections that follow, are provided in the DVD folder, General Management
and Oversight.
•

Planning Study for Containment Entry and Decontamination. The
licensee contracted Bechtel Power Corporation to develop a conceptual
plan for reentry and decontamination of the reactor building. The
primary objective of the July 2, 1979 report was to develop a plan for
placing the reactor building in a configuration for removal of the reactor
vessel head. The report provided an assessment of the reactor building’sradiation environment and the physical condition of (and degree ofdamage to) it; evaluated alternatives for reactor building23decontamination and reentry; and provided conceptual designs for new
systems that may be needed to support reentry and decontamination.61, 62•

Summary Technical Plan for TMI-2 Decontamination and Defueling.
The licensee’s plan for decontamination and defueling at TMI-2 wascontained in their December 12, 1979 report, “Summary Technical Plan
for TMI-2 Decontamination and Defueling.” The technical plan
identified the major steps to clean up the plant, which are as follows:(1) decontamination of the auxiliary and fuel handling building,including removal of contaminated water held in storage tanks andsumps; (2) decontamination of the reactor building, including removalof the radioactive, gaseous atmosphere and contaminated water in thesump; (3) reactor examination and defueling; (4) decontamination of thereactor coolant system; (5) radioactive waste processing, including
construction of the EPICOR II system, submerged demineralizer
system, and evaporator and solidification system; and (6) solidradioactive waste management.63•

Interim criteria for radiological effluents from TMI-2 data gatheringand maintenance operations (SECY-80-175). On April 14, 1980, theCommission approved radiological effluent criteria for the interim
period before the issuance of the programmatic environmental impact
statement for the purpose of data-gathering and maintenance operations.
Releases which were specifically not covered by these criteria were
purging of the reactor building’s atmosphere, disposal of EPICOR II
decontaminated water, and treatment and disposal of high-level,
radioactively-contaminated water in the reactor building. The interim
criteria provided a mechanism by which the licensee may request to
make small radioactive releases resulting in data collection andmaintenance operations. These criteria described the information thatthe licensee must submit to the NRC for approval before performing
these operations, and the type of review that the staff will perform toapprove each request.64, 65
A lack of definitive release criteria hampered planning and engineering
activities, resulted in a dilution of personnel resources to obtain specificCommission approval for activities that have insignificant impacts, andcaused public concern over minor unplanned releases that would beinconsequential in a normal operating plant. An example of the former
was the perceived need to obtain Commission approval to initially open
the outer personnel air lock door which would release about 0.05 curieof radioactive krypton gas (see SECY-80-10566). An example of thelatter was the occurrence on February 11, 1980, in which the incidental

24

off-gassing of about 0.3 curie of radioactive krypton gas from a leaked
primary water system caused considerable public concern (see NRC
Preliminary Notification 80-0367). To put these releases in perspective,
TMI-2 had been releasing between 65 to 80 curies of krypton gas permonth; at that time, a normal operating facility of this type may releaseover a 1,000 curies of radioactive gasses per month. In 1978, TMI-1
released an average of 1,300 curies of radioactive gasses per month.68•

Report of the Governor's Commission on Three Mile Island. On
May 14, 1979, Governor Richard Thornburgh of Pennsylvania
established a special commission under the chairmanship of Lieutenant
Governor William Scranton III to study and evaluate the consequencesof the accident. The results of that commission's work were released on
February 26, 1980, in a report entitled, "Report of the Governor'sCommission on Three Mile Island." It contained a number ofrecommendations and findings aimed at protecting public health andsafety in the wake of the TMI-2 accident. The seven-monthinvestigation assessed the environmental, economic, health, legal, andsocial effects of the accident, and made recommendations for action or
further study. One of the recommendations submitted to Governor
Thornburgh was that Unit 2 be promptly decontaminated, under proper
safety controls, in order to avoid possibly serious and uncontrolled
releases of radiation.69, 70•

Programmatic Environmental Impact Statement (PEIS). The NRC’s
PEIS related to the decontamination and disposal of radioactive wastes
resulting from the accident (NUREG-0683), and three supplements to
the PEIS were an important set of guidance documents for the NRC andlicensee. The PEIS discussed the options and associated environmental
impacts of four fundamental activities necessary to the cleanup:
treatment of radioactive liquids; decontamination of the building andequipment; removal of fuel and decontamination of the coolant system;and packaging, handling, storing and transporting nuclear waste. The
draft PEIS and supplements underwent comment periods by thelicensee, Federal, State and local government agencies, and the public.
The final PEIS was issued in March 1981.71
In terms of radiological health and safety, there was no known technicalreason for the radiological release criteria to be more restrictive than
had been acceptable at normal operating facilities. However, because of
the unique characteristics of the cleanup operation that were notconsidered and evaluated in the pre-accident safety review of the plant,
there was a need to define what keeping radiation exposure “as low as
25reasonably achievable” or “ALARA” meant with respect to offsite
releases and occupational exposures. The PEIS provided the basis formaking that determination.72
The PEIS had three supplements that were considered part of theoriginal PEIS. (See the DVD folder, Guidance-PEIS.)

26

oSupplement 1, “Final Supplement Dealing with Occupational
Radiation Dose.” The earlier PEIS stated that the most significant
environmental impact of cleanup activities at TMI-2 would result
from the radiation dose to the cleanup work force. This supplement
was issued in October 1984 to reevaluate the occupational radiationdose and resulting health effects from cleanup and to address
additional alternative cleanup approaches using information
gathered since the PEIS was prepared in 1980.73

oSupplement 2, “Final Supplement Dealing with Disposal of
Accident-Generated Water.” This supplement was issued inJune 1987 to update the environmental evaluation of
accident-generated water disposal alternatives published in theoriginal PEIS, using more complete and current information. Also,
the supplement included a specific environmental evaluation of thelicensee's proposal for water disposition.74
oSupplement 3, “Final Supplement Dealing with Post-Defueling
Monitored Storage and Subsequent Cleanup.” This supplement wasissued in August 1989 to evaluate the licensee's proposal tocomplete the current cleanup effort and place the facility into
monitored storage for an unspecified period of time. The
supplement provided an environmental evaluation of the licensee’s
proposal, and a number of alternative courses of action from the endof current defueling efforts, to the beginning of decommissioning.
However, it did not provide an evaluation of the environmental
impacts associated with decommission.75

Because these reports were programmatic in-nature, the reports were notintended to provide a step-by-step work plan. However, the mostprobable sequences and methods for cleanup had been assumed in orderto predict the resulting environmental impacts. The best available
information had been used and documented in these impact analyses.
Where uncertainties existed, conservative assumptions had been madeand documented in the main text and appendixes as appropriate.76•

NRC policy statement endorsing PEIS. On April 27, 1981, the NRC
Commissioners issued by majority vote a policy statement endorsing thefinal PEIS. The policy statement concluded that the PEIS satisfied the
NRC’s obligations under the National Environmental Policy Act(NEPA). The policy statement also indicated that, as the licensee
proposed specific major decontamination activities, the NRC staff
would determine whether these proposals, and associated impacts thatwere predicted to occur, fell within the scope of those already assessedin the PEIS. With the exception of the disposition of processed
accident-generated water (the Commissioners wanted to decide thisissue later), the staff was allowed to act on each major cleanup activity
without Commissioners’ approval if the activity and associated impacts
fell within the scope of those assessed in the PEIS. The policy statement
required the staff to keep the Commissioners informed of staff actions
before staff approval of major activities.77•

TMI-2 Program Strategy. In June 1984, the licensee issued an internal
technical plan, “TMI-2 Program Strategy,” that provided an overview of
the recovery program, established program priorities, established policy
and technical guidance, and provided the means for communicating the27

program to external people and organizations. The report provided
program policies that addressed the following issues: (1) generic issues,
such as recovery program objectives; definition of program phases,
priorities, and end points; use of the ALARA concept in decisions about
radiation exposure; use of remote technology; application of regulations
and regulatory guides; preservation of plant equipment and structures;
permanence of recovery facilities; sharing of facilities and systems with
TMI-1; and opening containment; (2) characterization of plantconditions, including data-gathering; (3) fuel control, such as criticality
and reactivity control, accountability of fuel, definition of core waste,
and fuel storage and disposal; (4) methods and end points for defueling;(5) disassembly and storage of large radioactive components, including
primary system integrity; (6) decontamination and dose reduction, suchas decontamination during the fuel-removal stage, criticality prevention
during decontamination, worker efficiency, and re-flooding of thereactor building’s basement; and (7) waste management, such as storageof waste and closure of commercial disposal sites, abnormal waste,
reuse of processed water, and segregation of water at TMI-2.78 (Note:
The program strategy is not available on the DVDs.)•

Post-defueling monitored storage (PDMS) plan. This licensee report
was submitted to the NRC on December 2, 1986, to provide a plan forplant conditions following completion of the cleanup program.79 Thisinitial plan was used to develop the proposed license amendment,
proposed technical specifications, and safety analysis report for
implementation of the proposed PDMS plant configuration. The request
was submitted to NRC on August 16, 1988.80 In response to thesubmittal, the NRC developed and issued Supplement 3 to theProgrammatic Environmental Impact Statement (NUREG-0683), which
dealt with PDMS and subsequent cleanup.81 The NRC issued thepossession-only license on September 14, 1993, and issued the PDMS
technical specifications on December 28, 1993.82 See further
discussions on PDMS in the section on After Defueling.

•

Recovery quality assurance plan. This licensee plan ensured regulatorycompliance for recovery and cleanup activities such as decontamination;assessment of damage; design; procurement; fabrication; handling;shipping; storage; cleaning; construction; installation; inspection; test;
operation; maintenance; repair; and modification. This plan was
periodically revised.83 (See the DVD folder, Guidance-PEIS.)•

TMI-2 radiation protection plan. This plan provided the philosophies,
basic policies, and objectives of the licensee’s program for radiological28

controls in accordance with NRC regulations and guidance. The
objectives of the radiological controls program were to control radiation
hazards in order to avoid accidental radiation exposures, to keep
exposures within their regulatory required limits, and to keep exposuresof workers and the general population at ALARA levels. The plan was
implemented within radiological controls procedures. The TMI-2
radiological controls program was fully-integrated into each phase of
the recovery effort at TMI Unit 2. This plan was periodically revised.84(See the DVD folder, Worker Protection.)•

General project design criteria. The general project design criteria
document was developed by the licensee to provide an adequate basis
for the design of recovery facilities and systems. The criteria included ageneral section that provided information common to all engineering
disciplines. The general section defined the regulatory requirements;
operating conditions; dose-reduction considerations; environmental
considerations; sharing of existing facilities and services; and protection
from severe natural phenomena and human-caused events. The
discipline-specific sections provided detailed, generic criteria forarchitectural design; civil-structural design; control-system design;
non-safety-related electrical design; mechanical design; and shielding
design and access control. To ease the review of future submittals, thelicensee requested that the NRC review the general section andsubsequent revisions.85 (See the DVD folder, Guidance-PEIS. Early
revisions of the discipline-specific sections were provided to the NRC
for information only in the licensee’s submittal dated
December 22, 1981.)•

Special nuclear material (SNM) accountability plan. This licensee
plan identified the methods and sequence of SNM accountability; thequality assurance program that was built into SNM measurement
activities; and the areas, systems, and components that had undergone
formal SNM measurement and those that did not require SNM
assessment. SNM accountability is required of all licensee holders ofreactor fuel and other SNM. As the result of the accident, the damaged
fuel debris was dispersed throughout the plant, and the origin of thedebris could not be traced to specific fuel assemblies. The NRC andDOE (receiver of the fuel debris) allowed the final SNM accountability
for TMI-2 to be performed after defueling was completed.86 Further, the
NRC granted the licensee exemptions regarding regulatory requirements
for record keeping, inventorying, and reporting of special nuclear,source, and byproduct materials.87 Accountability was based on athorough post-defueling survey of areas, systems, and components. The
29

results of materials accountability surveys and analyses were importantin evaluating the potential for recriticality of the remaining fuel debris
under postulated conditions during the plant’s long-term, post-defuelingmonitored storage.88 (See the DVD folder, Guidance-PEIS; see also the
DVD folder, After Defueling for post-defueling survey reports andpost-defueling completion reports.)•

Licensed operator qualification and training procedures. Theselicensee procedures provided the processes and requirements for the
training and certification of licensed operator candidates, and for the
renewals of their licenses. Licensed operators included reactor
operators, senior reactor operators, and fuel-handling-only (defueling)
senior reactor operators. These NRC-approved training and qualification
procedures reflected the unique and rapidly-changing plant conditions
as recovery and cleanup progressed. As such, licensed operators weretrained on new systems and defueling operations as part of their
ongoing requalification training.89, 90, 91 Training of fuel-handling-only
senior reactor operators began in December 1984. NRC granted fuelhandling-only licenses to the first five senior reactor operators in
October 1985.92 (See the DVD folder, Guidance-PEIS.)

Advisory Groups. Advisory and working groups were formed by GPU, the
NRC, and the DOE to provide advice, and sometimes direction, on
important recovery activities. Documents included in this collection (see theDVD folder, Advisory Groups) provide insights into the technical issues andproblems encountered during plant recovery and cleanup. Some important
advisory and working groups are summarized below:•

30

Industry Advisory Group (March 30 to May 6, 1979). The Industry
Advisory Group of outside organizations and individuals was formed by
the licensee three days after the accident to help determine plantconditions, and evaluate approaches to achieving a stable condition. The
scope of the group was expanded to evaluate operations andmodifications proposed by the recovery organization; to independently
assimilate, integrate, and interpret plant status information and data; andto review detailed procedures for plant recovery operations. Once theformal TMI-2 recovery organization was established on April 4, 1979,
the information flow between the Industry Advisory Group andelements of that organization was accommodated through the Technical
Working Group (described below). Activities for the Industry Advisory
Group were assigned by the Technical Working Group.93 Some reportsby the Industry Advisory Group are provided in the DVD folder,
Industry Advisory Group Reports.

•Technical Working Group (1979). The Technical Working Group was
an executive-level committee established by the licensee a week after
the accident to propose and discuss actions associated with daily
operations. The managers of the newly-established TMI-2 recovery
organization attended the initial twice-daily meetings to coordinate
activities, assign task responsibilities, and obtain general agreement on
activities for the 24-hour period following each meeting.
Representatives from outside organizations, such as the TMI-2 nuclear
steam supply system vendor (Babcock & Wilcox); the original TMI-2
architect engineer (Burns and Roe); the NRC; and the Industry Advisory
Group, also attended these meetings. Licensee personnel chaired themeetings and were responsible for making the decisions, anda consensus was achieved in most cases. The NRC, by virtue of its legal
authority to issue orders, did have the authority and therefore, theimplicit responsibility, to override a decision whenever that seemednecessary.94 The Technical Working Group’s daily meeting minutes forthe first four months after the accident are provided on the DVD (seethe DVD folder, Planning Meetings).

•

Advisory Panel for the Decontamination of TMI-2 (1980 to 1993).
In October 1980, the Commission established a 12-member TMI-2
advisory panel to consult with and provide advice to the NRC
Commissioners and staff on major activities related to thedecontamination and cleanup of TMI-2. The panel consisted ofmembers from the Commonwealth of Pennsylvania, local government,
and the scientific community, as well as residents in the vicinity of TMI.
The NRC TMI Program Office (discussed below) acted as a liaison
between the NRC and the TMI-2 advisory panel, and also provided
information to the panel on the status of the cleanup. Panel meetings
were open for the general public to attend and transcriptions were
produced for the public record. The panel provided an outlet for the
public, a way for the NRC and the utility to report on the progress of thecleanup and to gauge the public’s reaction to various alternative actions.
The most crucial panel influence on cleanup activities was the increased
public scrutiny of both NRC and licensee decisions and activities. The
panel facilitated communication with the public for both the NRC andthe licensee. This communication helped sensitize the agency and thelicensee to public concerns. It also kept the cleanup before the NRC
Commissioners through periodic meetings. Panel members traveled to
Washington, DC at least once each year to meet with theCommissioners and provide a report on current panel activities. The last
panel meeting, the 78th overall, was held in September 1993.95 The“Lessons Learned from the Three Mile Island Unit 2 Advisory Panel”
31

report (NUREG/CR-6252) provided lessons learned based on interviewsof former members and reviews of transcriptions of public meetings(see the DVD folder, Advisory Groups). Appendix A, “Panel Meeting
Dates and Transcript Microfiche Addresses,” to the report listed dates ofpanel meetings. Panel members were listed in an appendix of the NRC
annual reports (see DVD folder, NRC Annual Reports).

The Advisory Panel for the Decontamination of Three Mile Island Unit 2 held
its last meeting on September 23, 1993, in Harrisburg, Pennsylvania (Pa).
Panel members attending the final meeting are pictured. They are, left-toright, front row: Ann Trunk, Resident of Middletown, Pa.; Arthur E. Morris(Panel Chairman), Resident and former Mayor of Lancaster, Pa.; Joel Roth(Panel Vice Chairman), Resident of Harrisburg, Pa.; Elizabeth Marshall,
Resident of York, Pa. In the back row, left-to-right, are: Kenneth L. Miller,
Director of the Division of Health Physics and Professor of Radiology, Milton
S. Hershey Medical Center, Hershey, Pa.; Thomas Smithgall, Resident of
Lancaster, Pa.; Lee H. Thonus, Alternate Designated Federal Official, NRC
Office of Nuclear Reactor Regulation (Region I); John Leutzelschwab,
Professor of Physics, Dickinson College, Carlisle, Pa.; Niel Wald, Professor,
Department of Environmental and Occupational Health, University ofPittsburgh, Pittsburgh, Pa.; Michael T. Masnik, Designated Federal Official,
NRC Office of Nuclear Reactor Regulation; Frederick S. Rice, Resident ofHarrisburg, Pa.; and Gordon Robinson, Associate Professor of Nuclear
Engineering, Pennsylvania State University, University Park.

32

•

TMI-2 Safety Advisory Board (1981 to 1989). The TMI-2 Safety
Advisory Board was established by the licensee on March 16, 1981, toprovide the licensee with an independent appraisal of the recovery
program that gave particular emphasis to the assurance of public andworker health and safety. The advisory board met every three months,
and reviewed many aspects of recovery activities, including regulations;
nuclear criticality; safety; risk assessment; project organization; project
financing; project procedures; technical planning; and public
communications. Additionally, the advisory board regularly expressed
its views to the NRC, and to the Board of Directors of GPU.
Periodically, the advisory board participated in public hearings,
especially those of the NRC’s Advisory Panel for the Decontaminationof TMI-2. The first advisory board chairman was Dr. James C. Fletcher,former administrator of the National Aeronautics and Space
Administration (NASA). He later resigned as chair of the advisoryboard at the time of his second appointment as NASA Administrator,
following the Challenger space shuttle accident in 1986.96 The final
report of the Safety Advisory Board summarized its activities during its
8-year period of existence (see the DVD folder, Advisory Groups).
Appendix A, “Possible Research Opportunities as a Result of the TMI-2
Accident and Cleanup,” to that report provided a set ofrecommendations on possible research opportunities.

TMI-2 Safety Advisory Board. Front row, left-to-right: Bruce Lundin,
Robert Marston (Chairman 1986-1989), Jacob Fabrikant, Howard Raiffa.
Back row, left-to-right: Lombard Squires, Ronald Fillnow, Norman
Rasmussen, Merril Eisenbud, Robert Friedman, William Stratton, John
Auxier. Not shown: James Fletcher (Chairman 1981-1986).

33

•Technical Assistance and Advisory Group (1981 to 1989). The TMI-2
Technical Advisory and Assistance Group (TAAG) was established by
the licensee, with the cooperation of the DOE and NRC, to provide
independent technical assessment and advice on the decontaminationand defueling of TMI-2. This group ensured that approaches to thevarious cleanup and defueling operations were technically adequate andthat consideration was given to keeping radiation exposures at ALARA
levels. The TAAG consisted of approximately 10 permanent members
and additional ad hoc members when their special expertise was needed.
The group responded to specific requests from the licensee’s recovery
organization, the NRC, and DOE. The group’s work was fundedthrough the Idaho National Engineering Laboratory (INEL) by the
DOE. Representatives from INEL, DOE, and the NRC attended TAAG
meetings as observers. The TAAG reported directly to the president of
GPU and documented its results to INEL to assist in execution of the
DOE program.97, 98 Many of the TAAG reports are provided on the
DVD (see the DVD folder, Advisory Groups).

•

Technical Working Group for the TMI Information and Examination
Program. The TMI Information and Examination Program wasestablished to acquire data to improve current understanding of nuclear
plant accident environments and of the phenomena which contributed to

TASK PLAN FOR THE U.S. DEPARTMENT OF ENERGY
TMI-2 PROGRAMSTechnical Integration Office

Prepared for the
U.S. DEPARTMENT OF ENERGY
Three Mile Island Operations Office
Under DOE Contract No. DE-AC07-76ID01570

Two key planning documents used by NRC and DOE at TMI-2.
NUREG-0698 report defined NRC’s roles in the cleanup operations andassociated regulatory responsibilities. GEND-INF-036 report identified
tasks to be planned and administered by the DOE to retrieve useful
information from the cleanup effort to address key problems and issues
in the areas of plant accident response and recovery.

34

those environments. The licensee (GPU), EPRI, NRC, and DOE (known
as “GEND”) signed a coordination agreement on March 26, 1980, tojointly sponsor and participate in this program. The program wasadministered through DOE’s prime contractor at TMI-2 and was staffedby INEL personnel. In addition to the participation of NRC in thetechnical working group for this program, the NRC (1) reviewed thedata-acquisition tasks to ensure that they were implemented incoordination with ongoing cleanup activities and (2) used the dataacquired for the benefit of the cleanup to the maximum possible
extent.99
Results of this program were summarized in periodic newsletters andannual reports (see the DVD folder, DOE/National Laboratory Status
Reports). Technical details of the results were documented in
GEND-series reports, reports from DOE’s national laboratories, the
NRC’s NUREG-series reports, and EPRI reports. Most DOE-sponsoredreports were digitized by the Idaho National Laboratory (formerly
INEL) and the Oak Ridge National Laboratory and are currently
available from their public websites. Most of these reports and
NUREG-series reports are provided on the DVDs.
NRC’s TMI Program Office. The NRC was responsible for the regulationof TMI-2 cleanup operations. Regulatory objectives were to (1) maintainreactor safety and control of radioactivity; (2) ensure that environmental
impacts were minimized and that the radiation exposure of workers, the

NRC office trailers were relocated in late April 1979 from the TMI Visitors’Center to this location at the south end on Three Mile Island. Another officein nearby Middletown was opened and staffed on a regular basis, including
evenings, to provide the public an opportunity to talk with NRC.

35

public, and the environment was within regulatory limits and were at
ALARA levels; and (3) ensure interim safe storage and disposal ofradioactive wastes from cleanup operations.100
On April 7, 1979, the director of the Office of Nuclear Reactor Regulation(NRR) formalized the NRC operations at TMI with three principalorganizations: NRR Operations, NRR Technical Review, and the Office ofInspection and Enforcement Operations. The NRR operations function
interacted with the NRR staff in the TMI-2 control room and with the GPU
Task and Schedule Managerial Team. The GPU managerial team prioritizedwork and NRR provided liaison. The NRR technical review function
provided safety reviews of plant modifications proposed by GPU’s Plant
Modification Team. Early modifications included the new decay heatremoval system, electrical power distribution, primary system
instrumentation alternatives, radioactive waste systems, secondary coolingsystems, and other plant modifications. The NRR technical review groupalso provided liaisons to GPU’s Industry Advisory Group and the GPU
Technical Working Group. The inspection and enforcement function
provided surveillance of TMI-2 operations and in-plant health physics andcontinued monitoring and analysis of the environment.101
The TMI support organization was renamed the TMI Program Office(TMIPO) within NRR on April 1, 1980; it provided the NRC’s overall
direction of TMI-2 recovery and cleanup operations.102, 103 The TMIPO
established a staff with management and technical expertise in key TMI-2
cleanup activities such as radiation protection, radiological assessment,

NRC staff suited up for an inspection
inside the reactor building. Left:
Lake Barrett (onsite TMIPO
manager) in 1984. Above: Michael
Masnik and Lee Thomus (TMIPO
staff and later licensing project
managers of TMI-2) in 1993.

36

radiological waste treatment, and nuclear safety. Support by experts in otherdisciplines was available from other NRC staff and from contractors, underarrangement with the DOE. The TMIPO coordinated its activities with thelicensee, the DOE, other Federal agencies, State and local government
officials, and the public.104
Information flow was a major responsibility of the site office. A weekly
status report containing pertinent reactor, radiological, and environmental
information was prepared and distributed to all NRC offices. This report
was also distributed to the public, with copies available at the Middletown
office. The Middletown office was open and staffed on a regular basis,including evening hours, to provide the public an opportunity to remain
informed of the cleanup progress. Information was also supplied to the
public through press releases, television and radio interviews, and direct
responses to both written and oral public concerns. Information-exchange
meetings were also held periodically with officials of the DOE and the
U.S. Environmental Protection Agency (EPA).105
The TMIPO had the following regulatory responsibilities: (1) planning andmanaging all NRC involvement in TMI-2 cleanup activities; (2) obtaining
information about and evaluating the current facility status; (3) analyzing
and reviewing the licensee's proposed actions and procedures; (4) preparing
technical review documents on the safety and environmental impacts oflicensee-proposed cleanup actions; (5) approving or disapproving thelicensee's proposed actions and procedures; (6) advising the NRC
Commissioners on major cleanup actions; (7) coordinating the NRC's
TMI-2 cleanup activities with other governmental agencies, as necessary,such as the DOE and EPA; (8) informing State and local governments andthe public on the status and plans for cleanup activities; (9) overseeing
day-to-day licensee activities to ensure that operations were implemented inaccordance with NRC regulations, the facility's operating license, technicalspecifications, NRC orders, recovery plans, and approved procedures;(10) ensuring that activities are carried out in compliance with approved
NRC limits and procedures; and (11) coordinating with the NRC Office ofInspection and Enforcement on its TMI-2 inspection activities.106
The TMIPO report, “NRC Plan for Cleanup Operations at Three Mile Island
Unit 2” (NUREG-0698, as revised) defined the NRC’s role in cleanup
operations at TMI-2, outlined the NRC’s regulatory responsibilities infulfilling this role, and provided the NRC’s review and decision-making
procedures. The TMI-2 Cleanup Project Directorate (formerly TMIPO) was
dissolved on February 1, 1988. The NRC’s TMI resident inspector office
took over the inspection program for TMI-2 and a headquarters project37

Key: Advisory Committee for Reactor Safeguards (ACRS), President’s Council on
Environmental Quality (CEQ), Nuclear Reactor Regulation (NRR), Technical Advisory
and Assistance Group (TAAG)

Major NRC functional roles in TMI-2 cleanup operations in 1984(see NUREG-0698, Revision 2).

38

directorate assumed responsibility for technical review and project
management functions.107
The TMIPO weekly status reports provided a detailed chronology of plantstatus, environmental monitoring results, the licensee’s recovery activities,
NRC actions, and public meetings. Reports during the period from 1980
to 1990 are provided on the DVD (see the DVD folder, TMI Program OfficeWeekly Status Reports). NUREG reports and other TMIPO-related
correspondence are provided on the DVD (see the DVD folder, General
Management and Oversight).
Other Documents. Other documents relating to the management of therecovery and cleanup efforts provided on the DVD (see the DVD folder,
General Management and Oversight) include the following:•

licensee recovery organizations (see also the DVD folder, Organization
Plan Changes)•

defueling and decontamination plans, schedules, and cost estimates•

NRC and DOE work plans and agreements

Two diesel generators (center) were temporary installed at TMI-2 to provide
diverse backup power to the balance-of-plant (BOP) electrical power buses inthe event of a failure of normal offsite power sources. Backup protection was
required for new recovery systems and existing BOP systems that were used tokeep the plant stable. Many of the safety systems used during normal
operations were rendered inoperable or inaccessible due to the accident.

39

TMI-2 control room after the accident. The plant computer alarm station is shown in the foreground.Core exit thermocouple temperature
indication panel which was installed shortly after the accident is shown in the backgound (center left) behind the hanging sign.

40

3 Plant Stabilization
The near-term cooling of the reactor core was considered to be stable within
the first week after the accident. However, many technical issues involving
safety and control were identified in the following weeks and months. In thelicensee’s report to the NRC, “Interim Report on the Three Mile Island
Nuclear Station Unit 2 Accident,” dated May 5, 1979,108 the major
near-term objectives of the recovery organization included the following:(1) keep the plant in a stable shutdown condition; (2) control and manage
the volumes of existing radioactivity; (3) develop an overall
waste-management plan for liquid, gas, and solids; (4) develop a strategy toreach cold shutdown safely and expeditiously; (5) modify the procedures,
facilities, and equipment necessary to accomplish the above; and(6) institute the plan for accomplishing a transition into the organizationnecessary to proceed with the more long-term recovery efforts.
As a result of the degradation of the reactor core and plant equipment,
certain equipment, required to be operable, was no longer available. Other
non-safety-related systems, not generally relied on for safe shutdown, wereused to keep the plant in a stable condition. High radioactivity incontainment, reactor coolant, and certain auxiliary building areas limitedaccess to maintain important components needed for safe shutdown. Harshenvironmental conditions and high-radiation levels limited the ability ofcertain components and instrumentation to survive for long periods.109
Early Plant Status. Maintaining the plant in a stable shutdown condition
required: (1) reliable means to remove decay heat from the reactor core to
the ultimate heat sink; (2) maintenance of reactor core sub-criticality to limit
heat production; and (3) confinement of radioactivity within the reactor
building.110 Decay heat removal required the maintenance of critical
functions; namely, the reactor coolant system (RCS) flow control; RCS
pressure control; RCS water inventory control; and RCS heat removal
control. The control of RCS flow initially required the removal of noncompressible gases, such as accident-generated hydrogen, which could
potentially block flow at the high points of both hot-legs and damage an
operating reactor coolant pump. The removal of gases was successful andone reactor coolant pump provided forced circulation for almost one month,followed by natural circulation.111
The control of RCS pressure was important to maintain an adequate net
positive suction head to prevent reactor coolant pump damage and tomaintain the hot coolant at a subcooled condition to prevent boiling.
Pressure was controlled initially by the pressurizer heaters, spray, and41

instrumentation. Contingency procedures were developed in the anticipationof failures of heater cable installation due to high radiation levels in thereactor building or failures of pressurizer instrument transmitters located in
the flooded basement of the reactor building. Upon the loss of the lastpressurizer level instrument channel on April 27, 1979, pressure wascontrolled with the plant solid for almost a year by manually balancing
makeup injection and letdown flows. Letdown flow was fixed at a reduced
one-third capacity due to blockages in the system. The desirable pressure
was achieved by throttling the manual valves on the reactor coolant pump

Systems and components inside the reactor building (RB). Key: core flood tank
(CFT), decay heat (DH) system, once-through steam generator (OTSG),
pressurizer (PZR), reactor coolant drain tank (RCDT), and reactor vessel (RV).

42

seal injection lines with the makeup injection valve completely closed. The
lack of routine maintenance due to adverse radiological conditions inside
the auxiliary and fuel handling building contributed to makeup pump
failures and valve leaks, which further exuberated radiologicalcontamination and releases inside the building. In March 1980, the makeup
system was replaced with the newly-installed reactor coolant pressure
control system; this new system was in operation until preparations toremove the reactor vessel head in 1984.112

Systems and components inside the auxiliary and fuel handling building.
Key: waste disposal liquid (WDL) system, makeup and purification (MU&P)
system, makeup filter (MU-F), and makeup pump (MU-P).

43

The control of RCS heat removal was through the “A” steam generator
during the periods of forced and natural circulation cooling. The secondaryor steam side of the steam generator was maintained in a partial vacuum by
the main condenser mechanical vacuum pumps. A partial vacuum allows
boiling and steaming at temperatures below 212 degrees Fahrenheit (°F).
Steam flowed from the steam generator to the main condenser through aturbine bypass valve. The condensate system was used to feed water fromthe main condenser hotwell to the steam generator. The circulating water
system removed heat from the condenser to the natural draft cooling
towers.113 The use of secondary systems to achieve and maintain the plant incold shutdown was never performed in commercial pressurized water
reactors, especially for a long period of time. Cold shutdown condition was
initially achieved on April 27th when reactor coolant temperature decreased
to 188°F.114 Plant modifications improved the reliability of electrical power
supplies to non-safety grade secondary system components that provided
long-term cooling. A backup mini-condensate system was installed toreplace the existing condensate pumps, if needed. Other diverse backup
cooling systems were installed, such as the long-term “B” backup coolingsystem and mini decay heat removal system (discussed later).115
The control of the reactor core sub-criticality was accomplished byincreasing the concentration of neutron-absorbing boron solution in thereactor coolant. However, the lack of information on the state of the reactor
core and the condition of control rods, caused difficulty in accurately
calculating the boron concentration required to maintain subcriticalconditions.116 The required minimum boron concentration in the reactor
coolant system was increased to account for the potential (actual) melting ofcontrol rod materials during the accident and to support defueling
operations. Sources of makeup to the reactor vessel were limited andcontrolled to prevent boron dilution. Nuclear source range instrumentation
used to detect re-criticality was down to one operable channel prior to its
refurbishment during early reactor building entries.117 The potential for recriticality of the damaged reactor core, as well as fuel “fines” (very small
fuel debris) consolidated in sludge on the reactor building basement floor,
and in filters of cleanup systems, remained a concern during the first half ofthe new decade.118
Large quantities of radioactive krypton-85 gas (about 44,000 curies)119 andhighly-contaminated water in the reactor building’s basement(approximately 630,000 gallons of water, which contained about one-half ofthe core inventory of the radionuclide cesium-137)120 prevented personnel
access into the reactor building to assess damage and maintain importantequipment. Radioactive gas in waste gas decay tanks leaked into the44auxiliary building, causing airborne contamination problems. About 66 to
80 curies of krypton gas was leaking out of the reactor building every
month. The partial vacuum on the secondary side of the steam generator
caused a pressure difference between the reactor building and turbine
building, which enhanced leakage through the packing of various steam
valves. These gases were subsequently discharged from the secondary
system through the auxiliary building ventilation system to theenvironment.121 Waste-water storage at TMI was already at 60 percent (full)
capacity before the accident, and that water was subsequently contaminatedduring the accident.122 Operational leakage from support systems located in
the auxiliary and fuel handling building added to the mix about 800 to
1,000 gallons a day of mostly uncontaminated water.123
Organizations involved in the early recovery effort tackled many unique
technical issues aimed at ensuring the safe shutdown of the reactor core, andensuring sufficient removal of decay heat from the reactor core to theultimate heat sink. A number of systems and components were modified in
the near-term to maintain a stable plant. Some of these systems were placedin standby readiness and never had to be used. Many new systems andfacilities were also built to improve the diversity and reliability of existing
systems. These new systems provided capabilities to immediately clean upradioactivity hazards that would help future plant recovery operations. New
procedures were developed, (many early ones handwritten), to provide
guidance for operating the plant in its degraded condition. First-of-theirkind procedures were written, reviewed, and approved to handle many
imaginable contingencies (see the DVD folder, Recovery Procedures).
The Industry Advisory Group provided the utility with expert advice on
many technical issues during the first 5 weeks following the accident (seethe DVD folder, Industry Advisory Group Reports). The NRC extensively
studied methods and associated plant modifications for achieving andmaintaining cold shutdown. Other early issues and concerns that werestudied included recriticality of the reactor core; impact of core damage on
decay heat removal; reliability and diversity of decay heat removal,
including electrical power; loss of instrumentation; post-accident hydrogenproduction; mitigation of radiological releases; radioactive waste
characterization; contaminated water storage; potential for groundwatercontamination; fire protection; quality assurance of plant modifications;radiological protection of plant workers; and safety analysis of the damaged
plant (see the DVD folder, General Plant Stabilization).
Near-Term Recovery Actions. Key systems that were used for near-term
plant stabilization are listed below. Some systems were placed in standby
45and never had to be used. Summary descriptions of these systems were
provided in a series of monthly and quarterly licensee status reports thatwere submitted to the NRC over the first 18 months following the accident(see the DVD folder, GPU Status Reports).
•

Sub-criticality control. The reactor core was maintained in a shutdown
condition by boron solution that was injected in the reactor coolant
system. Since the integrity of the control rods and fuel rods was
unknown, the reactor coolant boron concentration was maintained
between 3,000 and 4,500 parts per million (ppm). The maximum boron
concentration and a minimum reactor coolant temperature of 50 °F were
specified in the proposed technical specifications to ensure that boron
precipitation blockage would not occur.124 Criticality analyses of thereactor core to ensure adequate shutdown margin under worst caseconditions had been made by the licensee, NRC, Babcock and Wilcox,
Brookhaven National Laboratory, and Oak Ridge National Laboratory.
A further evaluation of the risk of recriticality by NRC in April 1980
concluded that the most probable mechanism for criticality was boron
dilution and that the process was slow enough to detect and correct theapproach to criticality given appropriate instrumentation andprocedures.125
In April 1984, the maximum allowable boron concentration in the
proposed technical specifications was increased to 6,000 ppm to support
defueling options that could potentially rearrange the core, or portionsof the core, into a more reactive configuration. The increase in boron
concentration, if needed, would be buffered by sodium hydroxide
addition to maintain the pH in the reactor coolant system between 7.6
and 7.8 for corrosion control. The modification to the technicalspecifications included an upper pH limit of 8.4 to insure no boron
precipitation in the reactor coolant system (RCS) and associated samplelines (solubility of boric acid decreases with increasing pH).126 In July
1984, the minimum boron concentration was increased to 3,500 ppm toensure at least 1 percent shutdown margin, based on all creditable coreconfigurations, including those resulting from a reactor vessel head drop
accident. The licensee’s criticality analyses conservatively assumed ahypothetical 100 percent fuel failure; no neutron leakage or absorption
by structural or poison material; no fuel burnup; maximum fuelenrichment; and optimum fuel-moderator ratio. For the out-of-core
criticality model, the analysis assumed that 50 percent of the core
formed a hemisphere in the bottom of the reactor vessel.127, 128, 129 In
April 1985, the technical specifications were modified to increase theminimum boron concentration to 4,350 ppm to ensure that the fuel in46

the reactor coolant system would remain subcritical throughout all
reactor disassembly and defueling operations. Boron concentration
requirements for the spent fuel storage pool “A” and the fuel transfer
canal were added in the technical specifications. The values were thesame as those required for the reactor coolant system.130, 131, 132
Other criticality analyses were performed to support safety evaluationsof defueling systems and equipment, defueling and cleanup operations,and the fuel debris remaining inside and outside the reactor vessel.
Other criticality control methods were evaluated in GEND-026,“Addition of Soluble and insoluble Neutron Absorbers to the Reactor
Coolant System of TMI-2.” Most criticality safety analysis reports areprovided in the DVD folder, Criticality Analysis. Results of criticality
analyses that were performed in support of the post-defueling monitored
storage licensing application are provided in the DVD folder, After
Defueling.
•

Transition to natural circulation core cooling. Since the start of theaccident until 8:00 p.m. on the same day when the “1A” reactor coolant
pump was restarted, decay heat was primarily removed by the release ofreactor coolant to the reactor building through the stuck-openpressurizer pilot-operated relief valve, as the operator opened its block
valve to maintain pressure. For about a month, a reactor coolant pumpprovided forced circulation from the reactor core through the “A” steam
generator. The reactor decay heat was transferred from the reactor
coolant side of the “A” steam generator tubes, and out to the main
turbine condenser in the form of steam, generated in the secondary sideof the steam generator tubes.133
On April 27, the failure of the last remaining pressurizer level indication
prompted the operators and engineers to enact the emergency procedure
to stop the fourth and last remaining reactor coolant pump. That
afternoon, the reactor coolant pump was intentionally shut down by acontrol room operator. The “forward flow” in both of the reactor coolant
system loops decreased exponentially, as indicated on the then-operable
flow instrumentation. Within a minute, TMI-2 entered smooth, natural
circulation without incident. (The driving head for natural circulation
was the difference in density between the hot 12-foot core water
elevation, against a comparable 12-foot cooler elevation of water in the
tubes in either or both of the adjacent steam generators.) The feasibilityof long-term, natural circulation as a viable means for placing thedamaged reactor core into a long-term, stable condition was extensively
studied and evaluated by the licensee, the TMI-2 reactor vendor
47

(Babcock and Wilcox), and the NRC (see reports in DVD folder,
General Plant Stabilization).134
With the passage of time, and the associated reduction of decay heatgeneration rate (about 75 kilowatts 20 months following the accident),
the natural circulation flow slowed and then changed from continuous tocyclic with increasing intervals between the cyclic flow “burps.” Whenthe steam generator and cold-legs gradually cooled until the density washigh enough to initiate natural circulation flow, flow then diminished as
the warmer coolant from the reactor vessel displaced the colder water inthe steam generator and cold-legs.135
Alternate instrumentation. Alternate plant and reactor core instruments
were used and new ones were installed to replace those that weredamaged during the accident, or that failed because of harshenvironmental or high-radiation conditions. Instrumentation andelectrical equipment were exposed to moderately severe accident
conditions, including steam, reactor building spray, high radiation,
burning hydrogen, and the resultant overpressure. The equipment thatsurvived the accident was then exposed to long-term moisture•Core flood tank 1A level monitor CF-2-LT2 showing the condition of theenclosed electronic module. There was severe corrosion and degradation ofinternal components rendering the unit inoperable.

48

conditions, including high humidity, elevated temperature, andcondensation. The TMI-2 Instrumentation and Electrical Program
funded by DOE identified and analyzed a number of pre-accident
installation problems and instrument response characteristics that led tomisleading information and equipment failures. These problems
included faulty component seals and inadequate drains and vents toprotect enclosed equipment against moisture, anomalous response ofradiation monitors, and substantial corrosion of electrical contacts overa period of a few years.136 (See the DVD folder, Instrumentation andEquipment Evaluations.)
Emergency procedures were written by the licensee and approved by the
NRC to respond to losses of important instrumentation (see the DVD
folder, Recovery Procedures). Indications from other systems weresometimes used to replace similar information from failed instruments.
Installation of recovery systems provided alternative means formonitoring conditions of the reactor. Probes were used to measureconditions inside the reactor building through its penetrations before thefirst reactor building entry in July 1980. Once entrance began, importantelectronic components were replaced or repaired.
•

Modified decay heat removal system. Installed to improve the reliabilityof the existing decay heat removal system in order to reduce its

Long-term “B” backup cooling system (horizontal heat exchanger shown)provided a high pressure, closed cooling loop for the water filled “B” steam
generator. This backup system was never used.

49

radiological impact in the auxiliary building and on the environment in
the event that the system was needed to cool the core. This system wasplaced in standby and never had to be used.137•

Alternate decay heat removal system. Intended as a backup to existingand planned decay heat removal capabilities. The new system was to bebuilt outside the fuel handling building and was to include a complete,integral closed-loop system to circulate reactor coolant throughconnections from the existing decay heat removal system. Heat would
be removed through its own intermediate component cooling water
system through connections to the existing nuclear services river water
system; this ambitious project was never completed.138•

Long-term “B” backup cooling system. Installed to provide ahigh-pressure, closed cooling loop for the water-filled “B” steam
generator. This new system was a planned backup to the preferredsteaming of the “A” steam generator to the main condenser. The system
was installed in the turbine building and included a new heat exchanger,
pump, surge tank, piping, and valves. Heat was transferred from thesecondary side of the “B” steam generator to a new heat exchanger. The

Mini decay heat removal system being fabricated at the Babcock and Wilcox
facility. Two heat exchangers on stand in back; two pumps on floor.

50

heat exchanger was cooled by the existing secondary services closed
cooling water system, which was then cooled by water from the existing
nuclear services river water system. The system was installed, but neverrequired use. An identical system for steam generator “A” wasdesigned, but never built.139•

Mini decay heat removal system. Installed to replace the decay heatremoval system and served as a technical specifications required means
for boron injection and makeup to the reactor vessel. It was askid-mounted system with two pumps and two heat exchangers that was
fabricated at the Babcock & Wilcox facility in Lynchburg, Virginia.140
The system was installed in the fuel handling building with connections
to the existing decay heat removal piping. The two heat exchangers
were cooled by water from the existing nuclear services closed coolingsystem. The system was installed but was not used because loss of heatfrom the reactor coolant system to the reactor building’s ambientenvironment was shown to be adequate.141 The system became

Mini decay heat removal system installed at TMI-2. Both heatexchangers are shown. The system became operational in
October 1980.

51

operational on October 29, 1980,142 and could also be used during
defueling operations when the reactor vessel was not configured tomaintain forced cooling. However, loss-to-ambient cooling mode, as
described below, was the preferred method to remove the very low
decay heat levels.143
Reactor coolant pressure control system. Installed to provide long-term
reactor coolant system (RCS) pressure control and inventory control.
This system kept the RCS in a water-solid condition for natural
circulation core cooling; maintained sufficient available net positive
suction head for reactor coolant pump operation, if needed; controlled
the quality of the makeup water; and maintained pressure whileaccommodating thermal and volumetric changes in the RCS inventory.
The system included a makeup subsystem to maintain the reactor
coolant system in a water-solid condition for natural circulation cooling;
the makeup subsystem also injected chemically-treated water.
A letdown subsystem provided RCS overpressurization protection byincreasing letdown. Three surge tanks arranged in series provided RCS
pressure control.144 The system was installed in the fuel handlingbuilding and placed in service in March 1980. The existing makeup
system and letdown through the reactor coolant pump seal return line
were secured to reduce leakage and contamination in the auxiliary andfuel handling building. The new system was no longer needed when the•Reactor coolant pressure control system installed in fuel handling building.
Three 900-gallon surge tanks shown.

52

RCS was depressurized in the summer of 1984 in preparation forremoval of the reactor vessel head.145•

Alternate condensate pumps system. Installed to provide backup to theexisting condensate pumps in order to supply feedwater to the steam
generators for decay heat removal. In addition, the system could provide
feedwater to the new temporary auxiliary boiler. The system included
two 50-gallon-per-minute pumps that took suction from the condenser
hotwell and discharged to the steam generator through either of the two
newly-installed demineralizers.146•

Reliability improvements to balance-of-plant (BOP) electrical
distribution. Modifications to the existing BOP electrical power systemprovided backup protection in the event of a failure of normal offsite
power sources to BOP buses. Backup protection was required for new
recovery systems and existing BOP systems that were used to keep theplant stable, such as decay heat removal, pressure control, plant
instrumentation, and ventilation. Early modifications included two

Reactor coolant pressure control system provided long-term reactor coolant
system pressure and inventory control. The system became operational in
March 1980.

53

independent power buses, each supplied by a new 2,500-kilowatt-rated
diesel generator, and two existing circulating water pump buses
supplied by one new 13.2-kilovolt line from a nearby offsite electricalsubstation.147 The new diesel generators became available on May 16,
1979.148 The 13.2-kilovolt line was powered by an offsite 115-kilovolt
network which was backed by combustion turbine generators capable ofbeing energized independently of the normal offsite power source fromthe 230-kilovolt network.149 Later in 1980, the NRC granted approval
for removing the BOP diesel generators given that the improved
reliability of the 115-kilovolt network with the combustion turbinesallowed the network to be classified as an acceptable backup source ofpower for decay heat removal.150•

Additional hydrogen recombiner. Installed to provide additional
capacity to process hydrogen inside the reactor building following theaccident, due to uncertainty regarding the quantity of hydrogen being
generated. A skid-mounted, thermal-type hydrogen recombiner was
installed next to the existing thermal-type hydrogen recombiner locatedin the fuel handling building.151•

Portable disposable demineralizer system. Installed to remove
radioactive fission products in the “B” steam generator in order tominimize personnel exposure and potential for contamination of theturbine building before the new long-term “B” closed-loop coolingsystem was placed into service. The system included a disposable
demineralizer (18 inches in diameter and 30 inches in height) that was
attached to the discharge and suction of the closed-loop pump. The
portable disposable demineralizer system was used for wet layup of thelong-term “B” cooling system during standby.152•

Groundwater monitoring system. Installed to detect radioactivity
leakage in the ground around the reactor building and auxiliarybuilding. The Reactor Building Integrity Assessment Program wasestablished by the licensee, at the direction of the NRC, to monitor
potential leakage paths from the TMI-2 reactor building sump. The
leakage monitoring points, which were based on engineering
evaluations,153 included groundwater monitoring wells; storm drainage
areas; cork seals (concrete joint seals) in structures surrounding thereactor building; and the tendon access gallery (a passageway
surrounding the reactor building below the basement, approximately
20 feet below the water surface in the reactor building.154) The system
initially included eight monitoring wells located around the TMI-2
reactor building and other nearby locations. Seven observation wells

54were installed early in 1980 to determine the source of tritium found in
the wells. Monitoring wells had pumps; observation wells had grabsampling capability. Additional wells were installed in later years.155
The Programmatic Environmental Impact Statement (NUREG-0683)
concluded that an accidental release of contaminated water in thebasement of the reactor building was highly unlikely. Had this waterleaked through the steel-lined concrete base of the reactor building, thewave fronts of tritium, strontium-90, and cesium-137 in the groundwould reach the Susquehanna River after a minimum of 350 days, 23
years, and 284 years, respectively; and would continue to enter the river
over periods of about 130 days, 8.5 years, and 140 years, respectively.
Accounting for radionuclide decay and total mixing of effluents in theriver, the peak radionuclide concentrations would be orders ofmagnitude below concentration limits from 10 CFR Part 20, “Standards
for Protection Against Radiation,” for liquid effluents released to thegeneral environment. Furthermore, the peaks of these three
radionuclides would occur at different times.156

Groundwater monitoring wells (MW) and observation wells (OW) were
installed in early 1980 to detect potential radioactivity leakage in the
ground around the reactor building and the auxiliary and fuelhandling building.

55

•

Nuclear sampling system. Installed to be used as a temporary
liquid-waste sampling facility in TMI-2 without interfering with theexisting sampling system that was shared with TMI-1. TMI-2 samplelines were rerouted to a new sample sink located in the fuel handlingbuilding. Sampling capabilities were provided for various existing andnew storage tanks, the reactor coolant system, and continuous
monitoring of boron concentration inside the reactor vessel.157

Temporary filtration system for the auxiliary and fuel handlingbuilding installed on the auxiliary building roof. TMI-2 reactor
building in the background; fuel handling building at centerright.

56

•

Auxiliary and fuel handling building supplementary air filtration
system. Installed to temporarily replace the existing charcoal filter trains
in order to reduce offsite releases from the auxiliary and fuel handlingbuilding. Replacements of the existing filters were not practical becauseof the high radiation and contamination levels in the filter areas. The
new system interfaced with existing ventilation systems and consistedof four parallel filter units. Each unit had an exhaust fan and a filter unit.
Each filter unit consisted of a prefilter, a high-efficiency particulate air(HEPA) filter, a charcoal filter, and a second HEPA filter. The
ventilation stack was capped to direct ventilation from the buildings into
the new filtration units. The filtration units were installed on the roof of
the auxiliary building.158 Both trains of the existing auxiliary building
exhaust system were placed back into service by the end of April 1980.
The supplementary units were secured and the ventilation stack was
uncapped for the upcoming reactor building purge of krypton gas.159, 160•

Main condenser air extractor filtration system. Installed within two
weeks of the accident to remove contamination extracted from thecondenser condensate by the mechanical vacuum pumps before
discharging to the unfiltered segment of the auxiliary building
ventilation stack. The system included a pre-filter, a HEPA filter, acharcoal filter, and a second HEPA filter. The system remained inoperation until steam generator cooling was secured in January 1981.161•

Fuel pool waste storage system (also known as the “tank farm”).
Installed to provide temporary storage of 110,000 gallons of radioactive
water from the reactor building sump and miscellaneous waste holdup
tanks. The system consisted of 4 upper tanks (capable of holding
15,000 gallons each) and 2 lower tanks (25,000 gallons each) forming
two separate storage areas located in TMI-2 spent fuel pool “A.” 162 In
order to preserve the integrity of the structure, tanks were not attachedto the spent fuel pool structure. Therefore, the pool was not filled withwater for shielding to prevent the tanks from floating. Installation began
on April 6, 1979, and was completed in July 1979. On May 17, 1984,
the removal of these tanks started following the cleanup of the highlycontaminated water in the reactor building to make room for the transfer
and storage of reactor fuel debris.163•

Temporary auxiliary boiler system. Installed to supply steam to the
TMI-2 main turbine gland seals while the existing auxiliary boilers
shared with TMI-1 were serviced. Turbine gland seals were needed tomaintain condenser vacuum for steam generator heat removal.164
57

•

Waste water cleanup systems. EPICOR I, EPICOR II, and theprocessed water storage and recycling system are discussed in thesection on Waste Management.

•

Solid waste storage facilities. The interim solid waste staging facility
and the solid waste storage facility are discussed in the section on Waste
Management.

Longer-Term Recovery Actions. Key activities that were taken to keep theplant stable in the longer term are summarized below:
Transition to loss-to-ambient core cooling On January 5, 1981, thelicensee stopped cooling (also known as steaming) the "A" steam
generator by shutting a turbine bypass valve to the main condenser. Thisaction put the reactor coolant system (RCS) in a "loss-to-ambient" modeof cooling. Decay heat was then removed by heat losses from thesystem to the air inside the reactor building without the need of any
RCS circulation. A test of the loss-to-ambient cooling mode•

Fuel pool waste storage system or “tank farm” provided temporary storage of
110,000 gallons of radioactive water. Lower tank shown next to inactive fueltransfer upenders (to be used later to move fuel canisters from the reactor
building into the spent fuel pool).

58

commenced on November 6, 1980, for 33 days. Technical specifications
were then modified to recognize loss-to-ambient mode as an acceptablemeans for long-term cooling of the reactor core. Operating procedures
were prepared, and loss-to-ambient cooling was subsequently adopted
as the primary means for cooling the reactor core. By February 1, 1981,
decay heat diminished to about 43 kilowatts of thermal energy.165 The
long-term “B” backup cooling system and the mini decay heat removal
system were available as alternative cooling modes. The discontinued
use of the “A” steam generator to remove decay heat allowed its
removal from service and technical specifications for several majorbalance-of-plant systems and equipment, such as the circulating water
system; main steam system; “A” steam generator; condensate pumps;
condensate and feedwater systems; main condenser; and associatedsupport systems.166•

Waste water storage management. The storage of pre-accident andaccident-generated radioactive water challenged the licensee for many
months and years. Before the accident, radioactive waste tanks werealready 60 percent full with waste water from the recent TMI-1
shutdown outage.167 In addition, leakage from various operating systems
in the auxiliary and fuel handling building added about 800 to
1,000 gallons per day.168 On May 25, 1979, the Commission issued apolicy statement prohibiting the discharge of accident-generated water
into the river. The installation of additional water storage tanks and thestart of the EPICOR I and II water cleanup systems provided
much-needed relief in storage capacity for varying radioactive
concentrations. During the early weeks and months following theaccident, plant operators transferred contaminated water between
existing tanks (see correspondence in DVD folder, General Plant
Stabilization). The licensee considered using railroad tank cars for
onsite storage of low-level radioactive waste water. The licensee asked
the NRC to help them find 100 tank cars.169 Two tank cars werepurchased by the licensee; however, they were never used to store
contaminated or processed water. In July 1979, the fuel pool waste
storage system (“tank farm”) was available with 110,000 gallons ofstorage capacity. In July 1981, two new processed-water storage tanks
were available (each with 500,000 gallons of capacity).170•

Purge of the reactor building’s atmosphere. The controlled purging ofradioactive krypton gas inside the reactor building was carried out underdetailed procedures approved by the NRC staff. The purging began onJune 28, 1980, and continued until the morning of July 11. The removalof radioactive krypton-85 gas from the reactor building’s atmosphere59

allowed workers to begin work to clean up the reactor building, tomaintain instruments and equipment, and to remove the damaged fuelfrom the reactor vessel.171
The approval process for an environmental assessment of a number ofalternatives for the decontamination of the reactor building’s
atmosphere, began with the issuance of a draft for public comment.
Approximately 800 responses were received from various Federal,
State, and local agencies and officials, as well as from nongovernmentalorganizations andprivate citizens.
Based on thecomments received,
the “Final
Environmental
Assessment for
Decontamination of
the Three Mile
Island Unit 2
Reactor Building
Atmosphere”(NUREG-0662) was
issued in May 1980.
Having reviewed thestaff assessment andrecommendations,along with thecomments from thepublic, the Governorof Pennsylvania, andmany others, theCommission issued
a Memorandum andOrder onJune 12, 1980,
which authorized thelicensee to clean thereactor building’s
The TMI-2 ventilation stack being uncapped for theatmosphere byplanned purge of accident-generated radioactive krypton means of agas from the reactor building. The ventilation stack was controlled purge. On
previously capped to redirect ventilation from thethe same day, theauxiliary and fuel handling building to the temporary
Commission issued
supplementary air filtration system.

60

a temporary modification of the TMI operating license setting offsite
dose limits for the purge.172 The order included a one-time waiver of theenvironmental technical specifications (Appendix B) requirements forthe instantaneous and quarterly average limits for the release of noble
gases.173 On July 24, 1980, Amendment No. 11 to the facility’s
operating license was issued to allow the bypassing of the interlocks
from the reactor building’s exhaust radiation monitors to the reactor
building’s exhaust purge dampers for the duration of the purge.174
The vented activity was estimated to range from 38,302 to 50,254 curiesof krypton-85 with a median value of 44,132 curies during the release
period. Environmental monitoring was performed with substantial
instrumentation, with both fixed monitors in the ventilation system, andmobile sampling around the plant.175 Offsite radiation monitoring
programs were conducted by the licensee, the NRC, the United States
Environmental Protection Agency (EPA), the Department ofEnvironmental Resources of the Commonwealth of Pennsylvania, andalso by local citizens working through the Community Radiation
Monitoring Program set up by the DOE, and the Commonwealth ofPennsylvania. The maximum cumulative radiation dose and the
maximum dose rate measured at offsite locations were a fraction of thelimits allowed under NRC regulations.176 There were subsequent purges
to release the krypton gas that was slowly desorbing from the water andwalls. The monthly releases that occurred between September 1980 and
December 1980 decreased in an approximately exponential manner(emitting 27, 15, 12, and 7.5 curies respectively).177•

Reactor building entries. Entries into the reactor building were initiallydesired for damage assessment, data collection, and equipment
maintenance; in support of these entries, a set of experiments was
performed to identify any safety issues. Experiments included weekly
analyses of airborne samples from the reactor building; a gamma scan61

of the equipment hatch; a gamma scan of the basement water throughpenetration R605; an analysis of samples of the basement water throughpenetration R401 with support from the Oak Ridge National Laboratory(ORNL); radiation mapping of the no. 2 personnel air lock; plate out
analysis of the spool piece from the hydrogen recombiner with ORNL
support; analysis of hazardous gas in the reactor building’s atmosphere;
operating level (347-foot elevation) gamma scan and video survey
through penetration R626; and testing of protective clothing,telemetered dosimetry, and dose-rate instrumentation.178
Entries began on July 23, 1980,179 after the reactor building’s
atmosphere was purged of krypton, though initially, the plan was toenter the reactor building prior to purging.180 An attempt was made on
May 20, 1979, but was delayed until after the purging because of amalfunction of the inner door to the reactor building’s personnel
airlock.181 The initial entries used the following equipment: (1) portable
lighting; (2) radios (for communication); (3) respiratory protection, suchas self-contained closed-circuit breathing apparatus for the first entry
and a positive-pressure filtered breathing air mask for subsequent
entries; (4) protective clothing, such as paper overalls, cotton overalls,
and a firefighter’s suit; (5) protective shoes, such as cloth shoe covers,
three pairs of plastic disposable shoe covers, and firefighter’s boots;

View from inside the reactor building of the inner flange to reactor building
penetration R626 with video camera and telemeter dosimetry antennas(center). Initial experiments through this penetration supported the initial
reactor building entry program.

62

(6) protective gloves, such as one pair of latex gloves, two pairs ofneoprene gloves, and one pair of electric power lineman’s gloves;(7) head protection, such as a cotton hood and a rain suit hood;(8) personnel dosimetry, such as telemetered dosimetry, self-reading
digital dosimeters, thermoluminescent dosimeters, and film badges; and(9) handheld radiation instrumentation, such as a telescoping gamma
dose-rate meter for taking measures at a distance and a prototype survey
meter for taking high-range beta-ray measurements. The heavy-duty
outerwear provided high-energy beta-ray protection. During the first
20-minute stay inside the reactor building, each engineer received awhole-body gamma exposure of about 200 millirem with no beta skin
exposure.182 On May 24, 1989, the 2000th reactor building entry was
logged by the licensee and support personnel.183 Photographs,
communication transcripts, and reports of the first six entries areprovided in the DVD folder, Reactor Building Entries.
•

Reactor building sump water removal. By the time cleanup systems
were ready to remove contaminated water from the reactor building,
about 624,000 gallons of contaminated reactor coolant was standing on
the reactor building basement floor. About 264,000 gallons was spilled

Engineers being suited up for the initial entry attempt into the reactor
building. The orange underwater diver’s dry suit was used to prevent inleakage of radioactive material and to provide sufficient density to attenuate
the anticipated high-energy beta radiation. The yellow rain suit was used toprevent direct contamination of the dry suit. After the purge of krypton-85
from the reactor building, the dry suit was no longer required.

63

through the stuck-open pressurizer pilot-operated relief valve during theaccident. The pressurizer continued to leak during cooldown operations
and contributed another 178,000 gallons. The automatic actuation of thereactor building spray pumps during the hydrogen burn contributed an
additional 1,700 gallons of borated water directly into the sump. An
additional 180,000 gallons of non-contaminated river water leaked fromthe river water cooling system through a leaking relief valve on areactor building cooling unit.184
Although the contaminated water was safely contained in the reactor
building sump, its presence there constituted a continuing risk ofleakage to the environment, and prevented or hindered the performanceof the major decontamination activities.185 The potential for thisradioactive water to leak into the Susquehanna River, contaminating
both the river and the downstream Chesapeake Bay, as well as
downstream drinking water supplies, was a major concern for the GPU
leadership, the leaders and residents of communities downstream, andthe State and local governments of both Pennsylvania and Maryland. 186
Further, an early NRC task force concluded that the greatest risk of lossof containment was a confined leak of the reactor building water to thelower parts of the auxiliary building, where water in the reactor building
stood almost 30 feet above the decay heat removal pumps and thereactor building spray pumps. These pumps may have been needed later
during the recovery.187

First entry into the reactor building (shown) occurred on July 23, 1980. Heavy
duty outerwear provided beta ray protection.

64

The reactor building was constructed to prevent water in the sump fromleaking into the environment; however, early concerns existed over the
long-term storage of the highly-contaminated water in the basement,
and the integrity of the containment following the hydrogen burn inside
the reactor building during the first day of the accident.188 The reactor
building was designed and fabricated to withstand an internal pressureof 60 psig as well as major earthquakes. A structural integrity test was
conducted prior to unit startup to pressurize the containment to 69
pounds per square inch,
115 percent of the designpressure.189 The
foundation mat for thereactor building is 11.5feet-thick and rests on
bedrock. The reactor
building has a carbon
steel liner, 3/8-inch-thickon the sides, 1/2-inchthick dome, and 1/4-inchthick base. A concreteslab 2-feet-thick waspoured above the linerbase plate. The exposed
face of the liner wascoated with a prime andfinish coat of an epoxy
lining material. An
additional leakage barrier
was provided by a 40-milthick polyvinyl chloride
Cross-section view of the reactor building andpolymer waterproofing
auxiliary building with water level inside themembrane which wasreactor building as of April 1980. Gound water
elevation shown on the right represented the water installed over the outer
table outside the buildings.
surfaces of the foundation
mat.190
The submerged demineralizer system (discussed later) and the surface
suction system191 were designed and installed to clean up the highlycontaminated water inside the reactor building. A homemade floating
pump, carefully crafted using polystyrene foam, glue, and epoxy from a
Corvette repair kit, was designed and tested to initially float at an angle
and then lie flat on its side as it approaches the bottom. In 60 roentgen
per hour radiation field, a team of four wearing over 100 pounds of anti65

contamination clothing and equipment, dropped the suction pump withrubber hose over the open stairwell into the reactor building basement inless than 60 seconds.192 Details of this experience are provided in theseminar video, “The 35th Anniversary of the Three Mile Island Nuclear
Power Plant Accident of 1979: Working at TMI During and Following
the Accident,” (see the first volume of this NUREG/KM).
Floor expansion joint contamination. On November 26, 1980,
contamination in the cork seam of expansion joints was first discovered
during a routine radiation survey in the control and service building
area. The seam is a cork-filled construction joint located between major
structures to accommodate differential expansion between building
structures and to attenuate vibration and wave motions during a seismic
event. During the time period following the accident, the cork seam
located in the auxiliary building seal injection valve room was saturated
with reactor coolant water due to leaking valves. Initial decontamination
attempts were not successful. Over the years, the radioactive material
had spread along the joint into non-contaminated areas inside the plant.
However, the radioactive contamination is prevented from entering the
groundwater table by a water-stop barrier. Modifications have been
made to the cork seam to allow periodic monitoring of the water levels•

The surface suction system and floating pump. Picture shows a floating pumpduring a test in the outlet flume between the two natural draft cooling towers.

66

in the joint, to permit periodic water removal, and to prevent water andcontamination migration within the cork-filled joint. 193, 194, 195•

Auxiliary and fuel handling building decontamination (discussed in
the section on Decontamination)

Document Collections. Documents relating to plant stabilization efforts aregrouped into five document collections on the DVDs.
•

Correspondence between the NRC and the licensee (i.e., notifications,requests, reviews, and approvals) relating to the installation andoperation of plant systems and documents associated with other plantstabilization operations (see the DVD folder, General Plant
Stabilization). Enclosures (such as reports of system descriptions,
technical evaluations, and safety evaluations) are included withcorrespondence. NUREG reports that supported stabilization activities
are also included.

•

Agenda and minutes of daily meetings of the Technical Working
Group during the first four months after the accident (see the DVD
folder, Planning Meetings). As previously noted, this group was an
executive committee established by the licensee within a week after theaccident to propose and discuss actions associated with daily operations.
Many of the meeting handouts contained notes by NRC staff attending
these meetings.

•

Emergency and special operating procedures that were initiallydeveloped to cope with the plant’s damaged state during the firstmonths after the accident (see the DVD folder, Recovery Procedures).
These procedures were reviewed and approved by NRC staff located at
TMI. Many early procedures that were approved for use were
handwritten.

•

Industry Advisory Group task close-out reports (see the DVD folder,
Industry Advisory Group Reports). As discussed previously, this groupof industry organizations and individuals assisted the licensee todetermine plant conditions and evaluate approaches to stabilize theplant. Tasks assigned to the group were documented in these reports.

•

Photographs, communication transcripts, and reports of the first six
reactor building entries (see DVD folder, Reactor Building Entries).

67

68

Technicians inside the personnel airlock to the reactor building, as seen
through the round window of the outer personnel airlock door.

4 Worker Protection
Post-accident radiological conditions at TMI-2 were substantially different
from those normally encountered at commercial operating nuclear plants
because of the magnitude and specific mix of the radionuclide
contamination. Radiation surveys made shortly after the accident showedthat general area radiation readings ranged from 150 to 500 millirem perhour in the fuel handling building, and 50 to 5,000 millirem per hour in theauxiliary building. Hot spots were measured in the auxiliary building
reaching up to 125 rem per hour, and exceeding 1,000 rem per hour in somecubicles. The high-energy beta component was up to one hundred times the
gamma component. 196
During the first entry into the reactor building in July 1980, dose rates at the
305-foot entry-level elevation ranged from 400 to 600 millirem per hour.
Localized areas of high radiation were measured at 18 rem per hour over the
open stairwell and 2 to 5 rem per hour at floor drains. The general-area floor
and wall beta-radiation readings ranged from 1 to 2 rad per hour.197 Surveys
performed during the second entry at the next-higher level, the 347-footoperating-floor elevation, showed general radiation readings of 100 to
400 millirem per hour.
Below the 305-foot entry level elevation was the 282-foot basement-levelelevation, which was flooded with highly-contaminated water and sludge.
A telescoping radiation detector was inserted down through one of thereactor building stairwells and measured 40 to 45 rem per hour at 5 to 7 feetfrom the surface of the basement water.198 Once the water was drained andprocessed through the submerged demineralizer system, dose rates from theremaining sludge ranged from 1 to 1,000 rem per hour, depending on
location and distance from the floor.199
A unique concern at TMI-2 was high-energy beta contamination fromfission products in the reactor coolant. The generation of activation products
from corrosion such as cobolt-60 was minimal because the new plant had
less than a year of full-power operation. Areas in the auxiliary building that
had experienced coolant leakage were measured in the 10- to
100-rem-per-hour gamma range with associated beta dose rates in the
1,000- to 10,000-rad-per-hour range. Similar gamma-to-beta ratios were
measured on surfaces in the reactor building. The cesium isotopes have beta
energies in the 0.5-million-electron-volt (0.5-MeV) range; strontium-89,
which has the highest concentration of beta emissions, has a maximum
energy of about 1.5 MeV; yttrium-90, the decay daughter of strontium-90,
69

has a 2.3-MeV beta energy.200 Typically, beta radiation from beta-emitting
radionuclides at operating plants is low-energy, so protective clothing
provides sufficient shielding. The high-energy beta emitters present at
TMI-2 required special radiological protection practices, such as monitoring
equipment, personnel dosimetry, heavy protective clothing, procedures, andtraining.201 Access into many areas in the plant with high levels ofhigh-energy beta-emitting yttrium-90 required a combination of
double-respirator face pieces or face shields and safety glasses.202 The needfor these special worker protection practices became clear after several
incidents of skin and extremity exposures exceeding regulatory limits.
Radiation Exposure Events. The licensee reported several accident-related
exposures to the whole body and extremities during the response to theaccident on March 28th and 29th in excess of NRC regulatory limits. During
the very early phases of post-accident recovery, about five months after theaccident, six workers received overexposures to the skin and extremities.
The last extremity overexposure of two workers occurred in 1989. These
events are summarized below:•

70

Whole-body exposure. On the evening of March 28, 1979, an auxiliary
operator received a collective whole-body exposure in excess of the
NRC quarterly limit, while on two tours in the auxiliary building. The
auxiliary operator made two entries into the auxiliary building without
radiation work permits, a high-range self-reading dosimeter, and proper
planning with health physics support. There was no one to challenge theoperator’s entry and no locked access to restrict entry into the building
at that time. Upon exiting the building both times, the operator’s
low-range, (0 to 200 millirem), self-reading pocket dosimeter read
off-scale high. In addition, the operator failed to perform surveys for
personal contamination. Upon entering the TMI-2 control room after the
first job and then again after the second, a count rate meter on the table
alarmed both times. After being told by a radiation protection
technician, who happened to see the alarm the first time, todecontaminate, the operator decided to enter the auxiliary building again
for a second 10-minute job (the operator reasoned that he was alreadycontaminated). On the operator’s return to the control room the second
time, the operator told the shift supervisor that the self-reading
dosimeter had gone off scale. The shift supervisor told the operator todecontaminate. After decontamination, the operator’s personal
dosimeter was read, indicating a whole-body gamma exposure of
3.170 rem. An NRC investigation concluded that this entry exemplified
the lack of good health physics practices.203 (See NUREG-0600,
Section 3.2.4.8.)

(The NRC radiation protection regulations that were applicable during
the first decade of the recovery effort allowed up to 3 rem per calendar
quarter for a total of 12 rem in one year with proper documentation of
the worker’s exposure history. This higher limit is no longer stated incurrent regulations.204)•

Extremity exposures. During the afternoon of March 29, 1979, twoworkers received overexposures to their hands while taking reactor
coolant samples to measure boron concentration, in order to ascertain
whether the reactor core was still critical. The team discussed theassignment for a period of less than one hour. No one wore extremity
monitoring on their hands. Self-reading pocket dosimeters of one
worker were worn inside the protective clothing, making theminaccessible for periodic checks. The other worker wore a high-range
self-reading pocket dosimeter taped to the forearm, but it was knocked
off during collection of the sample. Before taking the sample, the room
was surveyed and found to have an average exposure rate of 8 roentgen
per hour. About 300 milliliters (ml) was drawn in a sample bottle, of
which 100 ml was poured into a graduated cylinder for gross radiation
measurement. Using a telescoping radiation-detection instrument, thecylinder measured off-scale high on contact (greater than 1,000 rem perhour), 400 rem per hour at 1 foot, and 10 to 15 rem per hour at 3 feet.
A second 40-ml sample was drawn and carried into an adjacentlaboratory for boron concentration measurement. The area radiation
monitor in the primary sample laboratory increased from 20 to
800 millirem per hour, as the worker entered the room. Both workers
handled the sample containers unshielded at contact with gloved hands,
while hand-carrying the samples for disposal.205
Both workers and a third worker involved with the boron measurement
received contamination on a finger, wrist, and other parts of theirbodies. The workers attempted decontamination for several hours. All
went home with some areas of their bodies having fixed contamination,such as 50 millirem per hour on a finger and 30 millirem per hour on awrist. Residual contamination remained on small areas of one worker’s
skin for over 30 days, which added to the total dose to the skin. Eachworker supplied urine samples the next day, but these were notpromptly analyzed by the licensee. The doses that the two workers
received were later calculated by an NRC radiation specialist. One
worker received an estimated exposure of 147 rem to the fingers, a skindose to the top of the head in the range of 10 to 32 rem, and awhole-body exposure of 4.25 rem. The other worker received an
estimated extremity dose in the range of 44 to 54 rem to the forearm71

skin and a whole-body exposure of less than 1 rem.206 (SeeNUREG-0600, Section 3.2.4.10.)•

Whole-body exposures. On the evening of March 29, 1979, one of two
engineers received whole-body exposure in excess of the quarterly limit
while he toured the auxiliary building to check for leaks. Before thetour, the engineers were briefed by a radiation protection foreman,dressed appropriately for the degree of hazard, and were provided with
two handheld radiation-detection instruments, one of which was a

NRC inspectors monitoring radiological controls
activities shortly after the accident. On March 30, the
NRC became more actively involved in the radiation
protection program by assigning NRC radiation
specialists to each of the three shifts, providing aroundthe clock coverage.

72

high-range gamma instrument. Shortly after entering the auxiliarybuilding, the high-range gamma instrument failed. They continued on
together noting that the other instrument frequently pegged off-scalehigh (2 rem per hour). Upon exiting the building, the high-range
self-reading pocket dosimeter read about 3 rem for one individual andless than 1 rem for the other. Their personal dosimeters were read andindicated 3.14 rem and 0.170 rem. An NRC investigation concluded that
this entry demonstrated a lack of implementation of basic radiation
protection training that could have resulted in serious consequences.207(See NUREG-0600, Section 3.2.4.11.)•

Whole-body exposures. On August 28, 1979, six workers entered aroom in the TMI-2 fuel handling building to inspect and tighten leaking
valves in preparation for decontamination of the area. Reactor coolant
water, extremely contaminated from the March 28 accident, was leaking
from the valves. The radiation survey instrument used by the workers
showed a gamma dose rate in the room of 10 to 15 rem per hour and, in
one small zone, 25 rem per hour. It was decided that the time limit on
the presence of each worker in the radiation area was four minutes.
What the survey instrument failed to disclose was that the beta radiation
rates in the room were running as high as 2,500 rem per hour. It waslater estimated that the workers had received doses from the beta
radiation in excess of regulatory limits. The doses were as high as
166 rem to the whole body in one instance, and 161 rem in another. No
indication of medically significant effects on the personnel was
identified by medical examination. The causes of the overexposures
were determined to be inadequate instrumentation for radiation
detection and a failure to require adequately-protective clothing for theworkers. Corrective action was taken under NRC direction.208•

Extremity exposures. On September 25, 1989, two workers received
overexposures to their extremities while unknowingly handling a pieceof fuel debris in the TMI-2 reactor building decontamination facility
inside the reactor building. Both workers were performing flushingdecontamination operations when one worker picked up what wasthought to be a nut lying on top of the drain grating in the flush facility.
The worker picked up the material and tossed it toward a trash bag, but
missed. The other worker picked up the material, presumably to place itin the trash bag, and was told by the first worker to leave it alone. The
first worker picked up the debris again and set it down 8 feet away.
Management was not informed of this event until the following day
when the first worker asked a radiation control technician (RCT) about
the implications of handling fuel debris. Fortunately, the RCT realized73

from his questions and further discussions that he had handled a piece ofcorium. The licensee conducted numerous reviews and assessments of
the incident, as well as medical examinations of and consultations withthe two workers.
A subsequent radiation survey performed on the material indicatedcontact dose rates of 1,320 rem per hour of gamma radiation, and11,580 rem per hour of beta radiation. The highly-radioactive material
was placed in the reactor vessel using long-handled tools. The dose
assessment, including re-enactments using a full-scale mockup of thedecontamination facility, estimated that one worker might haveexperienced an extremity exposure to the hands in the range of 75 to
375 rem, and the second worker might have experienced an exposure to
the hands in the range of 18.75 to 75 rem. The wide range of theseestimates was attributed to the uncertainty in the estimates of the timethat the workers were handling and were near the radioactive source.
The licensee instituted a site-wide training program on the appropriate
handling of fuel debris.209 The RCT demonstrated an attitude thatneeded to be fostered for conducting good health physics practices.
Radiological Protection Programs and Activities. Many programs andactivities helped to improve radiation protection practices at TMI-2. New
approaches were needed in a number of basic worker-protection anddose-reduction areas, including protective clothing, respiratory protection,

Technicians taking radiation readings inside the reactor building. Notice thepersonnel dosimetry attached to the back of the technician.

74dosimetry, radiation field and contamination characterization,
exposure-tracking systems, dose-reduction planning, procedures, training,
and robotics. In the summer, when the temperature in the reactor building
approached 33°C (90°F), an ice vest was commonly used by workers tocontrol heat stress and extend work periods.210, 211 Several respiratory
protection breathing apparatuses were developed or adapted to extend staytimes in the reactor building, including a power air-purifying respirator anda power air-purifying hood.212 Some of the radiological protection programs
and activities are summarized below:•

Increased radiation protection oversight. On March 30, 1979, more
control was established by the licensee over in-plant radiation hazards.
Radiation protection technicians were being used onsite to establish
control points. Temporary dose-accountability forms were implemented
that evening at the site access point. The whole-body counter was inoperation and individuals were being counted. On the afternoon ofMarch 30, the NRC became more actively involved in the radiation
protection program. Five NRC radiation specialists were assigned toeach of the three shifts, providing around-the-clock coverage. In
addition, two NRC health physicists provided technical guidance.213
A summary of radiation protection findings that were identified by NRC
inspectors during the early months of the recovery, were documented in
the NRC investigative report, “Investigation into the March 28, 1979
Three Mile Accident by Office of Inspection and Enforcement”(NUREG-0600).214•

NRC Special Panel on TMI-2 Radiation Protection Program. On
September 26, 1979, the director of NRC’s Office of Nuclear Reactor
Regulation, created a special panel to provide an independent review ofthe licensee’s existing and planned radiation protection program thatwas intended to keep personnel’s exposure to radiation at ALARA
levels during the recovery and cleanup of Unit 2. The panel of external
experts recommended the following in its report (NUREG-0640): an
upgraded radiation safety program for major radiological recovery
efforts; independent assessment of the proposed upgraded program; anda management plan and firm schedule for resolution of existing
technical and management deficiencies in the radiation safety
program.215•

Dose-reduction program. Because of the increasing amount of workinside the reactor building, a major effort to reduce dose rates was
initiated by the NRC and the licensee in December 1982. The objectiveof this effort was to identify and eliminate or shield as many sources of75

radiation exposure as possible in occupied portions of the reactor
building before and during reactor disassembly and defueling. The
dose-reduction program initially focused on both the 305-footentry-level elevation and the 347-foot operating-floor elevation, wheremost of the defueling work took place.216 Dose-reduction techniques
included (1) shortening the transit time of workers in the reactor
building by opening both personnel airlocks and modifying theingress/egress paths; (2) decontaminating by water flushing discreteradiation sources, such as the air coolers, elevator shaft, and enclosed
stairwell; (3) eliminating other discrete radiation sources by removal of
trash and contaminated equipment; and (4) placing shielding at the
305-foot elevation, such as lead curtains around the core flood tank,lead sheets on the covered floor hatch, and water columns and bladdershields around the open stairwell, elevator, and enclosed stairwell.217
Some of the more complex dose-reduction activities included extensive
pre-task planning and mockup training for each task, decontaminationof selected surfaces with chemicals, removal of paint, and scabbling (themechanical removal of a thin layer) of concrete floors. These efforts
resulted in significant reductions in the dose rate in the reactor
building.218 In July 1984, workers entered the reactor building without
respiratory protection for the first time since the accident, and, in
accordance with ALARA principles, subsequent entries were madewithout respirators.219 Dose-reduction activities to support defueling
operations were completed in March 1985.220

Protective clothing worn during early entries into the reactor building.

76•

PEIS supplement on occupational radiation dose. In October 1984, the
NRC TMI Program Office issued Supplement 1 to the Programmatic
Environmental Impact Statement or PEIS (NUREG-0683) after a publiccomment period. The purpose of this supplement was to re-evaluate theoccupational radiation dose, and consequent health effects fromcleanup, and address additional alternative cleanup approaches using
information gathered since the original PEIS was prepared in 1980. The
total radiation dose to cleanup workers was estimated to range between
13,000 and 46,000 person-rem as opposed to earlier estimates of
2,000 to 8,000 person-rem. The higher estimates resulted from a moreaccurate characterization of radiation fields in the reactor building basedon numerous worker entries.221

Radiation Detection Instruments and Systems. Special instrumentation,systems, and techniques were developed or modified to measure andcharacterize the unique radiation situation at TMI-2 for ensuring worker
safety and determining the effectiveness of decontamination processes. Key
instruments are summarized below:222•

Thermoluminescent dosimeter (TLD) pseudo cores to take beta
radiation measurements of the building floors. This simple device
included a beta measuring TLD mounted on top of a plastic ring toshield the TLD from beta radiations coming from the floor area outside

A technician lowering a telescoping radiation detector down into the open
stairwell to the reactor building basement. Measurements revealed radiationlevels of 40 to 45 rem per hour at 5 to 7 feet from the surface of water.

77

the ring. TLD pseudo cores replaced the need to manually drill andprocess concrete core samples.
•

Wall and floor sampler to mill the concrete surface and collect the
sample in a filter for offsite analyses.

•

Modified handheld ion chamber detector to provide omnidirectional
detection for gamma measurements and over 180 degrees for beta
measurements. This modification provided a wider field of view for
rapid directional exposure measurement of beta and gamma
contamination.

•

Modified handheld tungsten-shielded, Geiger-Mueller detector with aconical lead collimator on the face of the probe to reduce the field ofview from 140 degrees to 80-90 degrees. This modification provided
rapid and accurate directional exposure measurements.

•

Mobile radiochemistry laboratory to perform transuranic andradionuclide analyses of high-activity (less than 5 R per hour) liquid andsolid samples. Higher activity samples had to be packaged,
characterized, and shipped to offsite laboratories for analyses. This
mobile laboratory was provided by the DOE.

Water level in the basement reached the first stairwell landing indicating
about 8 feet of water accumulation.

78

•

Radiation mapping and ALARA planning system to provide 3dimensional maps of radiation exposures in plant areas and components,
and to train and plan for missions in contaminated areas, and track theirresults.

•

Improved personnel dosimetry system to approximate in-containment
beta source conditions. The system included a modified 4-element
dosimeter and an automated system used at TMI-2 each month toprocess that data from about 6,000 dosimeters.

Cumulative Worker Exposure Over 10 Years. Although worker activities
at TMI-2 have been quite different than those at operating power plants, thecumulative doses at TMI-2 since the accident had been lower than the
average doses experienced at operating reactors. By the end of 1989, withthe cleanup about 99 percent completed, the collective dose to all workers
was 6,180 person-rem. This was comparable to the collective occupationalradiation exposure that was estimated in the original PEIS (2,000 to
8,000 person-rem).223, 224
Document Collections. Documents relating to worker-protection anddose-reduction activities provided on the DVDs include the following:•

an assortment of documents relating to radiation protection, such as
annual reports of personnel exposures; radiation protection plans;dose-reduction plans; radiological training programs; andcorrespondence (see the DVD folder, Worker Protection)•an annual summary of dose-reduction activities provided in the NRC’s
annual reports (see the DVD folder, NRC Annual Reports)•

the report, “TMI-2: Lessons Learned by the U.S. Department of Energy,
A Programmatic Perspective,” which summarizes applications ofresearch and development in the areas of worker protection (see theDVD folder, DOE/National Laboratory Status Reports)

Defueling tool: spade bucket hydraulic attachment.

79

80

5 Data Acquisition and Analysis
The collection and analysis of data contributed to the safe stabilization of
the damaged plant and supported planning of decontamination and defuelingactivities. The TMI-2 accident also provided research opportunities that
added to the knowledge about light-water-reactor behavior following asevere accident. Information that was required to ensure stable plant
operations included measurement of reactor core and plant conditions;characterization of radioactivity inside the plant (for the design of recovery
systems and planning of activities); and verification of recovery system andactivity performance. The characterization of damaged fuel and internal
reactor vessel components supported defueling planning, preparations, andoperations, as well as interim storage, transportation, final storage, anddisposal of fuel debris.
The severe reactor core damage and behavior of fission products within theplant systems at TMI-2 gave government and industry researchers
opportunities to: (1) measure the performance of instrumentation and ofelectrical and mechanical equipment within the reactor building during andafter the accident; (2) determine physical damage to surfaces, components,and equipment caused by radiation exposure; (3) assess metallurgical andphysical behavior of fuel, clad, and core components during and after theaccident; and (4) assess new technologies for decontamination, and thedisposal of radioactive waste.225
The DOE supported an extensive research program as directed by the
U.S. Congress. Four organizations with a common interest in obtaining
valuable information from the TMI-2 accident jointly established the TMI-2
Information and Examination Program. The GPU, EPRI, NRC and DOE(hence the acronym GEND) signed the Joint Coordination Agreement226 on
March 26, 1980, which identified the objectives and defined methods toimplement the TMI-2 Information and Examination Program. The TechnicalIntegration Office was established at TMI according to the terms of the Joint
Coordination Agreement, with responsibility for implementation and daily
management of the DOE programs. The office was staffed by INEL
personnel and reported to DOE’s site office. A technical working group (seeabove section on Advisory Groups) was formed to coordinate research
opportunities and activities.227 In April 1984, the DOE signed a cooperationagreement with 17 Japanese nuclear power organizations to participate inresearch and development for recovery activities. As part of the agreement,
DOE permitted Japanese organizations to participate in DOE research anddevelopment activities. In return, Japanese participants provided funding to81

DOE and research staff (up to 22 engineers at a time) at DOE laboratories
and the TMI site.228, 229
The NRC was responsible for evaluating and approving proposed in-plantresearch activities and implementing procedures to ensure that they wereconducted in a safe manner.230
Reactor Core Inspections. Of particular interest to all organizations
involved in the cleanup and research, was the extent of damage to thereactor core and internal reactor vessel components. Data acquisition andanalysis that were conducted inside the reactor vessel included visual “quicklook” inspections using video cameras; ultrasonic measures; radiation dose
measurement; grab sampling of loose core debris; and core samples using acore bore drilling machine.231 The most revealing examination, with themost far-reaching implications, was the first “quick look” visual inspectionin July 1982, which showed extensive reactor core damage (fuel melting
would not be identified and announced until four years later). Up to the time
just before the quick look, the licensee had held out hope that the plant
would one day be restarted. After quick look, the licensee’s primary focus

The first closed-circuit video inspection of the upper reactor core region was
performed on July 21, 1982. A technician sitting on a platform positioned on
top of the reactor vessel service structure is inserting a camera, 1.5 inches indiameter and 12 inches long, through an empty leadscrew support tube.

82was cleanup.232 A chronology of the reactor core inspection activities issummarized below:•

Insertion test of axial power shaping rods. In June 1982,233 an axial
power shaping rod (APSR) insertion test was performed in an attempt tomove each APSR leadscrew inside the control rod drive mechanism to
its fully inserted position. During normal power operations, eight
APSRs helped shape the power generation (neutron flux) uniformly
across the reactor core to ensure even fuel burnup during the core’s
lifetime. These APSRs do not perform safety functions and do not drop
into the reactor core during an automatic reactor shutdown. During theaccident, the eight APSRs rods remained at a 25 percent withdrawn
position while 75 percent of their length remained inserted in the fuelassemblies. One purpose of the test was thought to provide insight into
the extent of damage to the reactor core and upper plenum. The analysisof the data concluded that the test provided little definitive information
about the physical condition of the reactor core.234 (SeeGEND-INF-038.)•“Quick look” camera inspection. The first closed-circuit video
inspection of the upper reactor core region was performed on
July 21, 1982. A camera 1.5 inches in diameter and 12 inches long wasinserted through the empty leadscrew support tube (inside a control rod
drive mechanism), and then into a central control rod guide tube (insidethe upper plenum). As the camera was lowered into the upper coreregion, it revealed a bed of rubble approximately five feet below thenormal location of the top of the fuel assembly. It was believed that the
rubble bed contained oxidized cladding, fuel fragments and/or pellets,
poison material, and core structural components. No evidence of melted
fuel pellets was found. Another inspection, on August 4, midway
between the periphery and the center of the reactor core, also revealed arubble bed approximately five feet below the top of the core region.
Intact pellets, which might have been fuel or poison material, werevisible on the top of the rubble. During a third inspection, which took
place on August 12, a probe was poked through the rubble and it
penetrated approximately one foot below the surface, indicating that the
rubble in this region was composed of loose material.235 The visualinspections did not find any apparent distortion of the upper plenum.236(See GEND-030.)

83Underhead characterization study. In late summer of 1983, a series ofactivities conducted for the underhead characterization study provided
radiological data to support the radiological protection measures thatwould be needed to conceptualize and plan procedures for removal of
the reactor vessel head. Previously, in December 1982, a simple vertical“quick scan” of the underhead region (using an ionization chamberlowered into an empty control rod drive mechanism opening) foundhigher-than-expected radiation levels, which ranged from 40 to
600 roentgen per hour with an unknown beta-radiation contribution.237
The results from the quick scan prompted a more detailed study.238•

The initial underhead characterization study obtained dose rates aroundthe reactor vessel head and service structure; re-measured dose rates
under the head using a beta-shielded ionization chamber (“quick

On July 21, 1982, the first video camera was inserted through a control rod
drive mechanism motor tube for a “quick look” of the upper reactor coreregion. Shown is the top of the rubble bed of the reactor core cavity.

84

scan-2”), and a string of thermoluminescent dosimeters; performed
remote visual inspections of the top of the upper plenum and undersideof the reactor vessel head; obtained debris samples from the top of theplenum for pyrophoricity tests; examined the lower end of a leadscrew
to characterize the stainless steel in the underhead environment; andevaluated the dose rate increase associated with moving several
radioactive leadscrews from inside the reactor vessel to a position in thecontrol rod drive mechanism’s motor tubes in the reactor vessel head’s
service structure. The study required the depressurization of the reactor
coolant system to atmospheric pressure, a slight draindown of thereactor vessel’s water level to uncover the top one foot of the upper
plenum (10 feet above the reactor core region), and the removal of acontrol rod drive mechanism.239 The NRC established a contract with
DOE’s Pacific Northwest Laboratory to review the radiation
measurements, fission products plate-out (absorption) on the upper
plenum, and other chemical phenomena.240
The results of the data analysis revealed the following: no visual
structural damage on the upper plenum; areas inspected on top of theplenum appeared relatively free of debris (though some sedimentation
appeared to be present on some horizontal surfaces);241 pyrophoricity
tests on two samples of material from the plenum surface proved
negative;242 gamma radiation fields in the range of 300 to 700 roentgen
per hour were measured in the space formed by the underside of thereactor vessel head and the top of the plenum;243 and no significant
dose-rate increase was observed on the service structure platform
following the withdrawal of four leadscrews.244 The analysis of the datafrom the underhead characterization study supported plans to remove
the reactor vessel head without flooding the refueling canal (also called“dry lift”).245•

Reactor core debris sampling program. The safety evaluation report for
the underhead characterization study was amended to include two new
activities that were sponsored by DOE’s TMI Reactor Evaluation
Program: a reactor core debris sampling program and reactor core
topography program. The reactor core debris sampling programprovided data that was essential to prepare for future reactor vessel
defueling activities and for the design of water cleanup systems anddefueling canisters. In September and October 1983, the program
obtained six specimens of reactor core debris by lowering speciallydesigned tools into the reactor. The tool scooped up small samples ofloose debris and transferred them into small shielded casks for offsite
shipment and analysis. The analysis of the samples included85determining their particulate composition, particle size, fission-productcontent, and drying properties, as well as the fission-productleachability from the debris and the pyrophoricity of zirconium hydrideor partially unoxidized zirconium fines.246
The first three samples were taken at various depths in the center of thereactor core. The radiation field at one foot from one sampler (which
had a capacity of one cubic inch) was approximately 3 roentgen perhour.247 The other three samples were taken midway between the centerand the periphery of the reactor core at various depths in the debris
bed.248 The six samples obtained during the grab-sample work wereanalyzed at the INEL and the Babcock & Wilcox (one sample) researchfacilities.249 Gamma-radiation levels of five samples, using a telescoping
radiation instrument from 2.5 cm away, ranged from 3 to 36 roentgen
per hour. Particle sizes ranged from about 0.6 cm to a fine debris.250
Five more samples were obtained in March 1984.251 Results from thesampling program were documented in the INEL report “TMI-2 Core
Debris Grab Samples—Examination and Analysis” (GEND-INF-075).
•

Reactor core topography program. In August and September of 1983,
the core topography system, designed and built by DOE at INEL, wasused to conduct an ultrasonic profile of the void area in the upperCore grab sample tool with fuel debris from the rubble bed obtained forthe reactor core debris sampling program in September 1983.

86

reactor core region. A total of about 500,000 data points was obtained
during system operation.252 The data and information obtained included
the radial and axial extent of the reactor core cavity, the location of
supported and unsupported fuel assembly end-fittings, and the locationof the core cavity boundary with respect to structurally intact fuelassemblies. The analysis of this data supported upper plenum lift anddefueling operations. 253
A clear plastic scale model of the damaged upper reactor core region
was constructed at INEL in late 1983 based on ultrasonicmeasurements. This topographic model provided the most accurate
indication of the extent of reactor core damage at that time. The volumeof the cavity in the damaged area of the reactor core was measured at
330 cubic feet or 26 percent of the original core volume. The bottom of
the cavity ranged from 5 to 6 feet below the top of the core. Of theoriginal 177 fuel assemblies, 42 appeared to contain some full-length
fuel rods, but 23 of those 42 had less than 50 percent of the rods intact.
The sonic mapping also revealed several partial fuel assemblies hanging
from the underside of the upper plenum and indicated some distortion of
the core former wall.254, 255 Each half of the plastic model wastransferred to the NRC by DOE and the Smithsonian Institution. The
complete model now resides at the NRC at Rockville, Maryland.

Topographic model of one-half of the cavity in the upper core region based onan ultrasonic profile from the reactor core topography program during
August and September of 1983. The individual layers of the model represent
2-inch contours within the cavity. The rods represent the locations andapproximate lenghts of the axial power shaping rods within the core at thetime of measurement. This model is currently located at NRC Headquarters.

87•

Reactor core video mapping. In April 1984, a comprehensive video
mapping of the upper reactor core region between the plenum andrubble bed was completed. Video snapshots were assembled into amosaic panoramic view of the rubble bed, core periphery, and theunderside of the upper plenum grid section. Videos of the rubble bed
showed broken fuel rods scattered around, fuel rod internal springs,
intact fuel pellets, control rod assembly end couplings, and partiallyintact fuel assemblies around the periphery of the reactor core region.
The video also showed unsupported partial fuel assemblies hanging
from the underside of the upper plenum grid section, which had to be
removed before plenum lifting. This map supported defueling
operations planning, and eventually the removal of fuel debris.256, 257, 258
A few photographs and partial mosaics from the video mapping activity
are provided in the DVD folder, Photo Gallery.

•

Video inspection of the lower head of the reactor vessel. In
February 1985, the first video inspection of the reactor vessel’s lower
head region was performed by guiding a small video camera and lightthrough a gap between the upper plenum and core support flange during
plenum jacking. The video inspection revealed re-solidified mass in the

Mosaic panorama of the reactor core cavity from comprehensive video
mapping in April 1984. Shown are hanging control rod assemblies, andbroken fuel rods and control rod upper end fittings on top of the rubble bed.

88

reactor vessel lower head on February 20, 1985.259 The camera followed
a path down to the lower head region between the reactor vessel walland the internal thermal shield. The video revealed the accumulation of
a substantial quantity (estimated at 10 to 20 tons) of accident-generated
debris with the appearance of a gravel pile. Over the next two years,
four additional camera inspections of the lower head region were
performed at different quadrants in the lower core support assembly(July 1985, December 1985, July 1986, and February 1987). The
analysis of the video data documented in the INEL report “TMI-2
Lower Plenum Video Data Summary” (EGG-TMI-7429), revealed thatlarge inhomogeneity existed in the physical appearance of the debris
bed, ranging from a very fine dust-like and smooth surface, to a
relatively flat, but coarse surface with large chunks, to a solid wall oflava-like debris.260 A few photographs from the video inspections areprovided in the DVD folder, Photo Gallery.
•

Vertical gamma profiles of the reactor vessel’s lower head region. In
March 1985, an attempt was made to insert a miniature
gamma-radiation sensor into several incore instrument tubes to obtain
vertical gamma profiles for characterizing fuel deposits that had settled
on the bottom of the reactor vessel.261 The incore instrument tubes

In February 1985, the first video inspection of the reactor vessel lower headregion revealed a gravel-like pile of core debris. Shown left is an instrument
tube penetration.

89

penetrate the bottom of the reactor vessel, and extend upward into thereactor core region. A dummy detector wire of the same size andstiffness as the actual probes was inserted into 26 of the 52 incore
instrument tubes; only one of the 26 wire probes reached the reactor
vessel. The remaining 25 wires were blocked at various locations alongthe 120 feet of pipe between the bottom of the vessel and the pipe
access at the incore seal table located in the reactor building. The single
wire probe was inserted 22 inches into the lower reactor core region.
A gamma sensor with a slightly larger diameter was inserted into thesame instrument tube and reached the vicinity of the
5-and-3/8-inch-thick lower head of the reactor vessel.262
Core stratification sample acquisition program. In July 1986, core
boring operations were performed using a special computer-controlled
drilling machine. The core stratification sample acquisition program
was conducted as part of DOE’s TMI-2 Accident Evaluation Program toprovide data on the material properties of the core debris. The core bore
samples provided insights into fission-product release from the fuel,fission-product retention in the core, maximum temperature during theaccident, and reactor core material interactions.263,264 A special
commercially available drilling rig was assembled on top of thedefueling work platform to bore into the hard crust. Ten full-length core•

A damaged fuel assembly being examined at the INEL.

90

bore samples were obtained from all regions of the lower reactor core;these samples (approximately 2.5 inches in diameter and eight feet long)
were analyzed at INEL, along with earlier samples of debris collectedfrom the reactor vessel’s lower head. Video inspections of the reactor
core below the debris bed were performed through several of the bore
holes created by the drilling operations. Initial inspections indicated thatperipheral fuel assemblies were essentially intact below the hard crustlayer, but that the central reactor core region consisted largely of a fusedmass of material.265 The INEL report “TMI-2 Core Bore Examinations”(GEND-INF-092) documented results of the physical, metallurgical,
and radiochemical examinations of the core bore samples and evaluatedthese results as they relate to the progression of core damage andfission-product behavior in the lower region of the reactor core. 266, 267•

TMI-2 vessel investigation project. In July 1989, a video inspection of
the reactor vessel’s lower head revealed several cracks that appeared to
be associated with incore instrument penetration nozzles. Higher-quality
color videos and a mechanical probe were used in August to obtain

Tears or cracks were found in the cladding of the lower reactor vessel headaround nozzle E-7 (instrumentation penetration stub upper right). These
cracks were analyzed by Argonne National Laboratory and were attributed tostresses associated with the thermal gradient in the thick-walled carbon steel
vessel during the heating and cooling phases of the accident.

91better information on the cracks. The cracks appeared to be up to
approximately 6 inches in length, 0.25 inches wide, and about
0.19 inches deep.268 Penetration into the thick reactor vessel steel waslater determined to be superficial.269
In February 1990, an international research effort obtained metallurgicalsamples from the reactor vessel’s lower head after defueling wascompleted. The program, which was sponsored by the NRC’s Office of
Nuclear Regulatory Research and the Organisation for Economic
Co-operation and Development, evaluated the potential modes of failure
and the reactor vessel’s margin of failure during the TMI-2 accident.
The condition and properties of material extracted from the reactor
vessel’s lower head were investigated to determine the extent of damage
to the lower head by chemical and thermal attack, the thermal input tothe reactor vessel, and the margin of structural integrity that remained
during the accident. A total of 15 “boat” samples were obtained fromthe reactor vessel’s lower head. In addition, 14 incore instrument
penetration nozzles were cut off 1 to 2 inches above the reactor vessel’s
lower head, and obtained as samples. Two incore instrument guide tubes
were cut free from the flow distributor head as samples.270 Results from

Investigations from the TMI-2 Vessel Investigation Project identified thelocation of a hot spot in the lower reactor vessel head, but concluded that the
hot spot would unlikely have caused vessel failure due to creep rupture for thetemperatures and pressures that occurred during the accident.

92the reactor vessel investigation project were documented in a series of
NRC reports by the INEL and Argonne National Laboratory. (See theDVD folder, Accident Data Analysis.)
Fuel Detection Techniques. Various fuel detection techniques were used tomeasure fuel debris and deposits internal and external to the reactor vessel.
The licensee was required to document the measurements and calculations
that were performed to ensure that the plant had been defueled to the extent
reasonably achievable, and that the potential for a nuclear criticality had
been precluded during normal and accident conditions.271 Measurements ofresidual fuel were required before the plant could begin post-defuelingmonitored storage (see section on After Defueling). Five general methods
were used for fuel detection (detection of gamma rays, neutrons, alphaparticles; sample and analysis; visual evidence). Key detection methods are
summarized below:272, 273•

Sodium iodide (NaI) gamma spectrometer was used to survey much of
the fuel deposits outside of the reactor building. This detector measured
Ce-144 from fission products.

•

Other gamma spectrometers were used to measure fuel deposits in theauxiliary building. A high-purity germanium (HPGe) detector measured
gamma radiation emitted from both Cerium-144 (Ce-144) andEuropium-154 with improved energy resolution capability. A silicon
lithium Si(Li) Compton recoil gamma ray spectrometer measured Ce144 gamma radiation in the A and B makeup and purification systemdemineralizer cubicles.

•

Solid-state track recorders were used to measure residual fuel indemineralizer cubicles and to profile fuel distribution within the reactor
vessel from the annular gap outside the vessel. This passive detector
used 93 percent enriched Uranium-235 fissile isotope foils to emit
induced fast fission neutrons that created visible tracks in adjacentacrylic sheets.

•

Copper activation coupons that become irradiated in the presence of aneutron flux. This method is insensitive to high gamma radiation fields.

•

Boron trifluoride (BF3) neutron detection system that thermalized fastneutrons from fissions for efficient counting in high gamma fields.

•

Alpha fuel detectors were used to measure thin films of fuel debris on
steam generator tube surfaces and on reactor coolant system surfaces.
93Active neutron detection technique was used to quantify smaller
quantities of residual fuel. This photo neutron interrogation method used
an antimony-beryllium neutron source to induce fission neutrons in thefuel that were measured by a helium-4 fast neutron recoil proportionalcounter. This method was more sensitive than passive counting forquantifying small deposits of fuel.

•Document Collections. Research tasks funded by NRC were documented
in NUREG reports. Research results from the TMI-2 Information andExamination Program were documented in the GEND-series reports and
DOE laboratory reports. GPU and EPRI documented the data acquisition
and analysis tasks they performed in their own reports. Most reports from
Government-sponsored work are provided on the DVDs. The EPRI report“TMI-2 Post-accident Data Acquisition and Analysis Experience” (NP7156s) provides a concise overview of data activities during the TMI-2
cleanup (not provided on the DVDs).

The first sample of sludge from the reactor building basement was taken usinga small scoop in June 1982. Later in 1982, three additional samples weretaken from different locations using solenoid-operated samplers (top right).

94Document collections relating to data acquisition and analysis tasks arelisted below:•

Tasks that generally supported plant recovery and cleanup activities areprovided in the DVD folder, Recovery Data Analysis. Research topics
included the following:opyrophoricity studies of potentially combustible core debris
materialsoexamination of debris on the reactor vessel’s lower headoexamination of fuel assembly componentsocriticality studies of the reactor core, reactor vessel, reactor coolant
system, reactor building sludge, and cleanup system filtersothe reactor building entry programothe dose-rate reduction program (see the section on Worker
Protection above)

A highly-radioactive EPICOR II prefilter liner being examined in a hotcell at the Battelle Columbus Laboratory. This examination and researchat other national laboratories provided information on the processing ofcontaminated ion exchange media and on the degradation of these media
and liners.

95

opolar crane refurbishmentocharacterization of the sludge in the reactor building’s basementooradiation and contamination surveys of building surfaces andinternal to system components
characterization of the reactor building’s atmosphereothe reactor building gross decontamination experimentothe fission-product inventory program, which included radionuclide
mass balance in the plantocharacterization of accident-generated water samples in various
systems and componentsosample acquisition from reactor building concrete surfacesocharacterization of and accountability for fuel deposits outside thereactor vesselothe citizen radiation-monitoring program

A telephone inside the reactor building during the accident was determinedby INEL to sustain high temperatures from the hydrogen burn of 200 to
220 degrees Fahrenheit. The telephone case, numeral ring, and dial weredistorted and the cord coils relaxed.

96

•

o

characterization of radioactive waste formsohydrogen production in spent filter canistersoeffects of radiation on organic ion-exchange resinsocharacterization and testing of EPICOR II waste form, resin, andliner integrity

Tasks that improved the understanding of severe core damage accident
phenomenology and its effects are provided in the DVD folder,
Accident Data Analysis. Research topics included the following:oTMI-2 accident forensic investigationsobenchmark thermal-hydraulic and severe-accident computer
analysis codes with TMI-2 dataoexternal influences affecting the accidentocore relocation and debris-bed coolability

55-gallon drums were partially collapsed by the external pressure caused by
the hydrogen burn inside the reactor building. This drum damage and lack of
air duct damage (upper right) was consistent with a pressure pulse thatdeveloped over seconds by a deflagradion and not by the passage of adetonation wave. (The undamaged drum on the left was open to theatmosphere, thus did not experience a crushing differential pressure.)97

oradionuclide (source term) behavior inside the plantoin-vessel fuel distribution characterization using gamma-ray andneutron dosimetryoresponse of nuclear and non-nuclear instrumentation during theaccidentohydrogen generation and burning inside the reactor building

A sampler, called a water and sludge sampling device, was designed tosimultaneously draw eight samples from the reactor building basement
water.

98

•

Tasks that were associated with the DOE-sponsored TMI-2
Instrumentation and Electrical Program are provided in the DVD folder,
Instrumentation and Electrical Evaluations. The program evaluatedinstrumentation and electrical equipment for the effects of exposure to
steam, spray, radiation, hydrogen burning, and resultant overpressure, as
well as long-term exposure to moisture. Components inside the reactor
building that were studied included radiation monitors, pressuretransmitters, loose-parts monitors, various switches and contacts,
solenoid operators for valves, and various other devices that sufferedmoisture intrusion. Summary reports are also provided in the DVD
folder.

Several electrical components from the reactor building were retrieved andexamined as part of the DOE-sponsored Instrumentation and Electrical
Program. Shown is the removal of an area radiation monitor (HP-R-211)
from the northeast elevator wall.

Thermowell of a “worst case” platinum resistance thermometer removed fromthe reactor coolant system hot-leg showed radioactive surface deposits (right).
This device was exposed to superheated steam during the accident; however,its calibration or its time response was not adversely affected by the accident.

99

100

Submerged demineralizer system inside the spent fuel pool.

6 Waste Management
The TMI-2 accident and subsequent cleanup challenged the management ofvarious forms and concentrations of radioactive waste. The management ofhighly contaminated water, fuel debris, and related solid-waste byproducts
included handling, processing, temporary onsite storage, transportation, andfinal disposal. Decontamination activities resulted in substantial quantities
of contaminated water and organic resins and inorganic zeolites produced
from water-processing systems. Fuel debris that spread throughout the plantcreated unique radiological waste characteristics. Also, some waste did not
fit into established regulatory waste-classification categories fortransportation and disposal, while the possible generation of flammablegases inside sealed radioactive waste containers was a potential hazard.
As part of the TMI-2 Information and Examination Program, DOE and the
NRC sponsored research and development in the areas of volume-reduction
techniques, performance of ion-exchange media, control of combustible
gases, waste-disposal techniques, and radioactive material shipping cask
designs.274 The NRC and DOE signed a memorandum of understanding(MOU) that specified interagency procedures for the removal anddisposition of nuclear waste resulting from the cleanup, including fueldebris. The memorandum and addendum ensured that the TMI site did not
become a long-term waste disposal facility.275
Key Actions. A chronology of key NRC and DOE actions related to waste
management is provided below. Most actions required research,development, safety evaluations, reviews, and approvals. Thoroughoverviews of waste management and transportation at TMI-2 are provided
in the EPRI report, “TMI-2 Waste Management Experience”(EPRI TR-100640) and the DOE-sponsored reports, “Historical Summary of
the Fuel and Waste Handling and Disposition Activities of the TMI-2
Information and Examination Program: 1980–1988” (EGG-2529), and“Historical Summary of the Three Mile Island Unit 2 Core Debris
Transportation Campaign” (DOE-ID-10400). Commission policy statements
and NRC orders relating to waste management are further discussed in theprior section on Management and Oversight.
•

Policy statement on environmental assessments. On May 25, 1979, theCommission issued a policy statement directing NRC staff to prepare an
environmental assessment regarding proposals to decontaminate anddispose of radioactively contaminated waste water. The policy required
assessments on the decontamination of intermediate-level waste water101

using the EPICOR II system, alternatives to discharge of waste water
into the Susquehanna River, and decontamination and disposal ofhigh-level waste water.276•

EPICOR II environmental assessment. On August 14, 1979, in
response to the Commission policy statement of May 25, 1979, NRC
staff prepared and sent out for public comment an environmental
assessment for the use of the EPICOR II filtration and ion-exchange
decontamination system to remove radionuclides fromintermediate-level radioactive waste water held in storage tanks in the
TMI-2 auxiliary building (NUREG-0591). The proposed action was
limited to cleanup and storage of waste water. The action also included
the impacts of wet solid waste generated from EPICOR II operation,such as temporary storage, packaging, handling, transportation, andburial.277•

Order to operate EPICOR II. On October 16, 1979, the Commission
issued a “Memorandum and Order” directing the licensee to operate the
EPICOR II to decontaminate intermediate-level radioactive waste water
held in storage tanks in the TMI-2 auxiliary building. In response to that

The basement in the reactor building contained about 8 feet of highly
contaminated water. The submerged demineralizer system started processing
basement water on September 9, 1981.

102

order, NRC staff issued an “Order for Modification of License” twodays later to permit EPICOR II system operations. EPICOR II began
operation on October 22.278•

City of Lancaster agreement. On February 27, 1980, the NRC and theCity of Lancaster, Pennsylvania, signed a litigation settlementagreement that prohibited the discharge of accident-generated water in
the Susquehanna River until December 31, 1981, or until NRC
completed necessary environmental reviews.279•

GEND research coordination agreement. On March 26, 1980, thecoordination agreement for the TMI-2 Information and Examination
Program was signed by GPU, EPRI, NRC, and DOE (also known as
GEND) to provide research and development coordination on efforts tomanage wastes at TMI-2.280•

Final PEIS. On March 9, 1981, the NRC staff issued the Final
Programmatic Environmental Impact Statement (PEIS) relating to thedecontamination and disposal of radioactive wastes resulting from theTMI-2 accident (NUREG-0683). The PEIS was an overall study of theactivities necessary for decontamination of the facility, defueling, anddisposition of the radioactive wastes.281•

Policy statement endorsing PEIS. On April 27, 1981, the Commission
issued a policy statement endorsing the final PEIS. With the exceptionof the disposition of processed accident-generated water (theCommission wanted to decide this issue later), the Commission directed
the staff to act on each major cleanup activity (without further
Commission direction) if the activity and associated impacts fell within
the scope of those assessed in the PEIS.282•

Order to operate the submerged demineralizer system (SDS). On
June 18, 1981, the NRC staff issued an order to direct the licensee topromptly commence the operation of SDS with effluent polishing by the
EPICOR II system. The order directed the complete processing of theremaining intermediate-level contaminated water in the auxiliarybuilding tanks (100,000 gallons), and the highly contaminated water inthe reactor building sump (700,000 gallons) and the reactor coolant
system (95,000 gallons). The approval to operate SDS did not include
water disposal. All processed water had to be stored in existing onsite
tanks; however, portions were allowed to be cycled for reuse within theplant. The order required that decisions related to the disposition ofprocessed water be made by the Commission at a future date.283 The
103

NRC review of the SDS formally started when the licensee submitted
their technical evaluation report on April 10, 1980. After a number ofdesign changes by the licensee and technical questions from the NRC,
the NRC safety evaluation report (NUREG-0796) was issued inJune 1981.284 In its review, the NRC determined that the potential
environmental impacts from the proposed operation of the SDS were
within the scope of the PEIS. On July 10, 1981, the SDS began
processing water from the reactor coolant bleed tanks. On
September 22, 1981, the water from the reactor building sump andbasement was pumped to the SDS feed tanks located outside the reactor
building. The following day, the SDS began processing this water. The
processing of water from the reactor building’s basement wascompleted in May 1982. In June 1981, the NRC issued the certificate ofcompliance for the shipping cask that was designed and fabricated totransport spent SDS vessels at first to a DOE research facility andsubsequently to a commercial low-level radioactive waste burial
facility.285•

NRC-DOE MOU. On July 15, 1981, the NRC and DOE signed MOU toformalize the working relationship between the two agencies withrespect to the removal and disposition of solid nuclear wastes generated
during the cleanup of TMI-2. The DOE agreed to carry out research andSubmerged demineralizer system started processing highly contaminatedwater in June 1981 from the reactor building basement, reactor coolant
system, and auxiliary building tanks.

104

conduct tests on solid wastes shipped from the plant to DOE facilities
that had generic information value. DOE also agreed to accept otherwaste that were too highly radioactive for disposal in commercial
facilities and provided no research value. The licensee provided
reimbursement to the DOE for the shipment, storage, and disposal ofsuch waste. Most low-level waste associated with decontamination,such as some ion-exchange media, boots, gloves, and trash, wasdisposed of by the utility in licensed commercial low-level radioactive
waste burial facilities. The original MOU provided plans for DOE toaccept some fuel assemblies and samples for analysis characterization
and archiving. Moreover, the original plan was to place the remainingbalance of the reactor core in fuel storage containers and store the fuelin the TMI-2 spent fuel storage pool to await resolution of thecommercial nuclear power spent fuel storage issue.286 However, thisMOU was formulated before the first “quick look” inside the reactor
vessel that revealed a severely damaged reactor core.
•

NRC-DOE MOU addendum. On March 15, 1982, the NRC and the
DOE signed a revision to the MOU. The DOE agreed to accept theentire reactor core for research and development and for temporarystorage at a DOE facility.287 During the first decade after the accident,
the DOE accepted 50 EPICOR II pre-filters during the initial processingof radioactive waste water from the auxiliary building, 19 highly loaded
SDS vessels, 3 SDS vessels with highly loaded cartridge filters, cut
control rod leadscrew segments, and original cartridge filters from the
makeup and purification system that filtered reactor coolant system
letdown during the accident. The disposal by DOE of waste productsthat were not useful for research was funded by the licensee.288•

DOE-GPU agreement. On March 19, 1982, the DOE and GPU signed
an agreement for the acquisition of the damaged TMI-2 reactor core by
DOE. In the agreement, DOE acquired ownership of the damage corefrom GPU and arranged for the shipment of the entire core to a DOE
facility for research. GPU reimbursed DOE for shipping, storage after
the research period, and ultimate disposal.289•

Environmental assessment for processed water disposal. In June 1987,
the NRC issued Final Supplement No. 2 to the Programmatic
Environmental Impact Statement (NUREG-0683) dealing with disposalof accident-generated water. The supplement presented NRC’s
evaluation of the licensee’s July 1986 proposal for disposing ofapproximately 2.3 million gallons of slightly radioactive water. Thiswater was contaminated either during the accident or during subsequent105

cleanup operations. The proposed method involved the forced
evaporation of the water at the TMI site over a period of two andone-half years. The residue from this operation contained small amountsof the radioactive isotopes tritium (hydrogen-3), cesium-137, andstrontium-90 and larger amounts of nonradioactive boric acid andsodium hydroxide. This residue required solidification and disposal as

EPICOR II system originally decontaminated 380,000 gallons ofintermediate-level radioactive water held in the auxiliary building
tanks. Each vessel contained ion-exchange resin. The vessel at the topof the photo was the prefilter demineralizer, the center vessel was a
cation ion-exchanger, and the third vessel was a mixed-bed polishingion-exchanger. Each was fitted with three quick-disconnect hoses: aliquid waste influent line, a processed waste effluent line, and a ventline with attached overflow hose.

106

low-level waste. The NRC evaluated the licensee’s proposal together
with eight alternative approaches, giving consideration to the risk ofradiation exposure to workers and to the general public; the probability
and consequences of potential accidents; the necessary commitment of
resources, including costs; and regulatory constraints. The NRC
concluded in the supplement that the licensee’s proposal to dispose of
the water by forced evaporation to the atmosphere, followed by onsite
solidification of the remaining solids and their disposal at a commercial
low-level radioactive waste burial facility, was an acceptable plan. The
supplement also established that no alternative method of disposing of
the contaminated water was without question, clearly preferable to thelicensee’s proposal.290, 291
Public hearings on the licensee’s proposal to evaporate 2.3 million
gallons of accident-generated water were held by an NRC Atomic
Safety and Licensing Board, and concluded on November 15, 1988. On
February 3, 1989, the Board issued a decision finding in favor of thelicensee on all relevant issues. On April 13, 1989, the Commission
upheld the Board’s decision. Construction of the evaporator systembegan in August 1989.292•

License amendment permitting processed water disposal. On
September 11, 1989, the NRC issued Amendment No. 35 to thefacility’s operating license, modifying the plant’s technicalspecifications by deleting the prohibition for disposing ofaccident-generated water. Disposal was allowed in accordance with
NRC-approved procedures.293 The evaporator system began vaporizing
accident-generated water on January 24, 1991, after a prolonged periodof system testing, modification, and repair.294 On August 12, 1993, thedecontamination and evaporation of 2.23 million gallons ofaccident-generated water was completed.295 The system evaporated
about 99 percent of the initial pre-processing volume of 2.3 million
gallons. The residual volume that remained in various tanks andbuilding sumps was estimated to be about 18,500 gallons.296 The
evaporator system was disassembled and shipped offsite by the ownerand operator of the system.297

Water Processing and Storage Systems. Before defueling operations
could begin, pre-accident and accident-generated waste water required
decontamination of low, intermediate, and high levels of radioactivity.
Additional holding tanks were constructed and existing tanks were managed
to maximize storage capacity. After decontamination, the slightly
radioactive water (still contaminated with tritium, a radioisotope of107hydrogen, and traces of fission products) had to be stored onsite until the
NRC Commissioners decided the ultimate disposal requirements. Until then,the NRC permitted the licensee to reuse this water for surface andequipment decontamination activities inside the auxiliary and fuel handlingbuilding and reactor building. This water was reprocessed through watercleanup systems and stored onsite in large carbon-steel plate-welded tanks.
During defueling operations, several water cleanup systems were designed
and installed to maintain water clarity for the manual removal of fuel debris
inside the reactor vessel and remove fuel debris contamination resulting
from the movement of defueling canisters from the reactor vessel to thespent fuel storage pool. Keeping the water inside the reactor vessel free of
microorganisms was a complex technical problem that hampered visibility
and eventually disrupted defueling operations for a period of time.
More than 2.3 million gallons of processed accident-generated waste waterthat were stored in various tanks at TMI-2 was disposed of by evaporation
over a 31-month period. The systems and facilities that were reviewed andapproved by the NRC to store and process waste water are described below:•

The fuel pool waste storage system (also known as the “tank farm”)
was installed to provide feed staging for the processing of the waterfrom the reactor building’s basement by the EPICOR II system andsubmerged demineralizer system. (See Plant Stabilization section.)

Typical SDS vessel.

108

Typical EPIROR II liner.

•

The EPICOR I system was installed within a week after the accident toprocess low-activity, non-accident-generated liquid waste water, mainly
generated from the TMI-1 outage. The system included one pre-filter
liner, one demineralizer liner, and two receiving tanks. Each
6-foot-by-6-foot liner was made of carbon steel. Over its 19-month
lifetime, EPICOR I processed over 1.3 million gallons at an average
flow rate of 10 gallons per minute.298•

The auxiliary building emergency liquid cleanup system (also known
as “EPICOR II”) was designed and installed after the accident to clean
up about 450,000 gallons of intermediate-level waste water. This water
was held in various storage tanks and sumps inside the auxiliary andfuel handling building. The EPICOR II system was a liquid waste
processing system to decontaminate water by filtration andion-exchange demineralization. This process for treatment ofradioactive water was (and currently is) standard practice in nuclear
power plants. The EPICOR II system was housed in an existing onsite
building that was originally intended for chemical cleaning of the steam
generators for TMI Units 1 and 2.299

EPICOR II system originally decontaminated 380,000 gallons of intermediatelevel radioactive water held in auxiliary building tanks. Shown is theschematic of the flow path in 1980.

109

The EPICOR II system operated in three main configurations during its
service life at TMI-2.
o

The original configuration included one pre-filter liner followed by
two demineralizer liners and cartridge filters to retain any resin
fines and particulates. The 4-foot-by-4-foot carbon-steel pre-filter
liner contained a pre-coat material to remove particulate radioactive
waste, such as activated corrosion products, and other suspended
solids. The pre-filter also contained cation resin which was highly
efficient for the removal of cesium and other cationic radionuclides
from the waste stream. The first demineralizer (a 4-foot-by-4-footcarbon-steel liner) contained cation resins to remove cesium. The
second demineralizer (a 6-foot-by-6-foot carbon-steel liner)
contained mixed cation and anion resins to remove cesium andiodine, respectively. After processing, the water was collected in aclean water receiving tank for sample measurements of radionuclide
concentrations. The water could be reprocessed as necessary toachieve the desired concentration of radionuclides.300, 301 The
processing of intermediate-level waste water inside the auxiliarybuilding began on October 22, 1979, and was completed on
December 2, 1980. During this period, a total of 1,051,047 gallons
of water was processed, with about half of this amount being
recycled processing.302

o

EPICOR II was used to remove residual radioactivity fromsubmerged demineralizer system (SDS) effluents and to process
miscellaneous wastes from September 11, 1981, to April 1987.303
This configuration included two 6-foot-by-6-foot liners in series
followed by one 4-foot-by-4-foot liner. The first-stage
demineralizer generally removed sodium and other non-radioactive
chemicals, although low concentrations of radionuclides were also
removed. (SDS removed most of the gross cesium and strontium.)
The second demineralizer reduced radioactive concentrations
through ion exchange and filtering. The last liner refined theeffluent water and caught resins that broke through the second
liner’s retention screens.304oAfter the SDS was removed from operation in 1988, EPICOR II
was the primary system to clean up waste water that was mainly
generated from building decontamination activities. The system
configuration was the same as before; however, a high-integrity
container (HIC) loaded with zeolite resins was placed in the first
position to act as a roughing filter to remove gross cesium and110strontium radionuclides. The HIC was followed by the two standard
EPICOR II carbon-steel liners. This configuration allowed
EPICOR II to replace the aging SDS. The HIC was similar in size tothe carbon-steel 4-foot-by-4-foot liner, but the HIC was constructedfrom very high-grade stainless steel. The HIC allowed the loadingof higher concentrations of radionuclides for burial at a commercial
low-level radioactive waste burial facility.305
Upon completion of the TMI-2 cleanup activities, the EPICOR II system
was released to general site use and included under the TMI-1 facility
operating license.306•

The submerged demineralizer system (SDS) was designed and installedin the TMI-2 spent fuel pool to clean up high-level radioactive
accident-generated waste water from the reactor building’s basement,
reactor coolant system, and reactor coolant bleed tanks. Processed waterfrom the processed water storage tanks was used to fill the spent fuelpool for shielding of the highly radioactive spent filters. The SDS
consisted of a liquid waste treatment subsystem, a gaseous waste
treatment subsystem, and a solid waste handling subsystem. The liquid
waste treatment subsystem was designed to remove cesium andSubmerged demineralizer system started processing highly contaminated waterin June 1981 from the reactor building basement, reactor coolant system, andauxiliary building tanks.

111strontium from the high-activity waste water by filtration and ion
exchange. The liquid waste treatment subsystem included the following
primary components (in the order of the flow path):307oPre-filter and final filter consisting of sand to remove particulates.

oFeed tank system (known as the “tank farm”) of four 15,000-gallon
storage tanks to hold contaminated water before processing it in theion-exchange vessels.

A typical submerged demineralizer system ion exchange vessel. About
4.5 feet in height and 2 feet in diameter, each vessel included fivenozzles on the top: inlet, outlet, vent, and general vessel access (twonozzles) for loading and unloading the ion exchange media.

112

o

A 30-gallon-per-minute pump.

o

Two parallel trains of identical inorganic zeolite-filled ion-exchange
vessels. The first train included two vessels to process water fromthe reactor coolant system and reactor coolant bleed tanks. The
second train included four vessels to process the highly
contaminated water from the reactor building’s basement. Each ofthese vessels was a stainless steel pressure vessel two feet indiameter and four feet tall. Each vessel also contained a catalyst bed
to recombine the available hydrogen and oxygen (generated byradiolysis of water) back into water.

oPost-filter to retain zeolite fines from new zeolite beds.

o

Two 12,000-gallon monitoring tanks to hold processed water formeasurement. SDS effluents in the tanks could be reprocessed or
sent to the EPICOR II system to remove residual contamination (aprocess known as “polishing”). The final processed water wastransferred from the monitoring tanks to the processed water storage
tanks for reuse or deposition.

oSample connections, sample glove boxes, and a continuous in-line
radiation monitor to estimate curie loadings and monitor vessel
efficiency for process control.

The first vessel in a train upstream was the first to be removed fromservice. A spent vessel was moved to the dewatering station and then to
the storage rack in the spent fuel pool using the fuel handling building’s
crane. Downstream vessels were disconnected and moved upstream one
position. A new vessel was installed at the last open position. Spention-exchange resins and filters were loaded into a shieldedtransportation cask underwater. The cask with the vessel inside was
moved from the spent fuel pool, sealed, decontaminated, and loaded
onto a truck trailer for shipment.308
DOE national laboratories provided technical assistance to the licensee
and their contractors to develop the SDS.309 The SDS started testing on
July 12, 1981, started processing reactor building basement water on
September 9, 1981, and started processing reactor coolant system wateron May 21, 1982.310 During its 7-year service life, SDS processed
4,566,000 gallons of water.311 Details of the SDS design, installation,and testing were documented in the INEL report “Submerged
Demineralizer System Processing of TMI-2 Waste Water” (GEND-31).
113

Following the initial SDS pumping of 50,000 gallons of water from thereactor building sump on September 27, 1981, the incore thermocouple
temperatures increased. On September 28, 1981, the temperatures
stabilized and started a slight downward trend. The licensee reported
that the temperature change appeared to have occurred as a result of
lowering the water level in the reactor building’s basement below thepoint at which the water was in contact with the lower dome of thereactor vessel. The heat-transfer characteristics had changed because
heat might no longer be conducted to the water in contact with thevessel and the temperatures would change until equilibrium conditions
were established. The highest temperature of an incore thermocouple
was approximately 147°F (a 12°F increase) and the calculated averagetemperature was 117°F (a 2°F increase).312•

The processed-water storage and recycling system was installed to holdprocessed water from the EPICOR II and SDS cleanup systems. The
processed water was stored onsite in two 500,000-gallon tanks. The
tanks became operational on July 8, 1981.313 A recycling system waslater installed to transfer the processed water into the reactor building
for decontamination activities. This water was subsequently reprocessedby the SDS and returned back to the processed water storage tanks forfurther use or to await ultimate disposal. The storage system included
two steel plate-welded tanks that were epoxy-lined and insulated.

Processed water storage tanks, each with 500,000-gallon capacity, were
installed to hold processed water from EPICOR-II and SDS cleanup systems.
Insulation and heat tracing were installed for freeze protection (not shown).

114

A heat-traced subsystem prevented freezing in winter months. Twotransfer pumps were used to transfer water from the processed watertanks.314 A decision was made not to build a dike around the tanks forwater retention because it was determined that any liquid spillage froma tank rupture would not represent a significant radiological health,safety, or environmental hazard.315, 316 On completion of the TMI-2
cleanup activities, the processed water storage system was released togeneral site use and included under the TMI-1 facility operating
license. 317•

The internals indexing fixture (IIF) processing system provided
interim reactor coolant system processing capability following reactor
vessel head removal to minimize radiation dose rates around the IIF.
The IIF cylinder was mounted on the reactor vessel’s flange and thereactor coolant level was raised to partially fill the IIF. The IIF
processing system consisted of a submersible pump located inside the
IIF which transferred water from the IIF through the submerged
demineralizer system to a reactor coolant bleed holdup tank. Reactor
coolant-grade water was concurrently returned to the reactor vessel
from a second bleed tank by a waste transfer pump to maintain the IIF
level.318 The system was placed in operation in August 1984 and was
replaced by the defueling water cleanup system in November 1985.319•

The defueling water cleanup system (DWCS) was used to remove
organic carbon, soluble fission products, and particulate matter from thefuel transfer canal (FTC), spent fuel pool “A” (SFP-A), and the reactor
vessel. The DWCS system was composed of two independent systems:
the reactor vessel cleanup system and the FTC/SFP-A cleanup system.
Each system included four modified defueling canisters (filter canisters)with sintered metal filter media to remove debris, mainly fuel fines(uranium oxide) and core debris (zirconium oxide), down to a
0.5-micron rating. Each system had either one (FTC/SFP-A cleanup) or
two (reactor vessel cleanup) 4-foot-by-4-foot demineralizer liners filledwith zeolite resins to remove cesium. The liners were high-integrity
containers, similar to those later used in the EPICOR II system. Eachsystem had two submersible pumps that fed two filter canisters (twotrains of two filters each), followed by one or two demineralizers.320 The
FTC/SFP-A cleanup system was ready when the upper plenum was
moved from the reactor vessel to the deep end of the refueling canal in
May 1985. The startup of the reactor vessel cleanup system followed in
November 1985.321

115

•

Treatment of microorganisms. Problems plagued the cleanup systems
early because of microorganism growth in the reactor coolant. In
January 1986, a filamentous growth was observed on surfaces inside thereactor vessel. A rapid growth occurred when hydraulic fluid from thedefueling tools was inadvertently spilled into the reactor vessel, alongwith the use of lighting in the reactor vessel, use of biologicallycontaminated water for makeup, and aeration of the coolant by the
movement of tools in and out of the water and by startup of watercleanup systems. The resulting loss of visibility halted all defueling
operations. Principal microorganisms found in the reactor coolant werebacteria and fungi, which were common in river water sediment. During
the months following the accident, a large volume of river water enteredinto the reactor building through a leaking air cooler. River water andaccident-generated water on the reactor building’s basement floor were
processed by the submerged demineralizer system and stored in theprocessed water storage tanks. The biologically contaminated,
processed water was reused for reactor coolant makeup. The hydraulicfluid used in defueling tools was also found to be contaminated. Thesemicroorganisms were resistant to high radiation fields, borated reactor
coolant chemistry, and biocides.322
Considerable time was spent attempting to control visibility inside thereactor vessel. Numerous techniques, with varying degrees of success,were evaluated serially to bring microorganism growth under control
and to improve clarity by the end of 1986.323 As the result of extensive
research and testing, the NRC approved 200 parts per million hydrogen
peroxide as the biocide. NRC’s evaluation concluded that the biocideconcentration was compatible with existing water chemistry and water
processing systems, as well as the effects on criticality, defuelingcanister catalytic hydrogen recombiners, and waste disposal. In
addition, the increase in reactor coolant activity levels due to theincrease in rate of cesium leaching from the debris bed would not
impose a significant impact on worker safety with the implementationof normal radiological control practices.324 Inspections, corrosion
studies, and laboratory tests found no microbial-induced corrosion ofcomponents and defueling equipment in the reactor vessel.325•

The modified DWCS. In addition to the microorganisms being filteredby the DWCS, colloidal suspensions of fine particles of silicon and ironoxides caused premature plugging of the sintered metal filter media in
the filter canisters. A temporary reactor vessel water filtration system(see below) was installed until solutions to stop the microorganism
growth and remove the colloidal material could be incorporated in the116

DWCS design. Under normal conditions, the filter efficiency increases
as a cake is built up on the surface of the media. However, thesuspended colloids were plugging the pores of the sintered stainlesssteel filter media before the filter cake could form. Modifications to the
DWCS included a filter-aid feed system that injected a coagulant withbody-feed material into the filter canisters to agglomerate the colloids tofilterable sizes and promote cake buildup on filter media.326, 327 The
modified reactor vessel cleanup system was restarted on
January 8, 1987.328 Later in 1987, a crosstie was installed between the
FTC/SFP-A cleanup system and the “B” train of the reactor vessel
cleanup system that was modified with the coagulant and filter-aid feedsystem. This crosstie permitted the processing of FTC and SFP-A water
without the need to make a similar modification in the FTC/SFP
cleanup system located in the fuel handling building.329 Other
operational improvements were made to the DWCS during its service
lifetime.330•

The temporary reactor vessel water filtration system was used to clean
up reactor vessel water above the rubble bed and to maintain an
adequate visibility for defueling. The system took suction from theinternals indexing fixture and pumped the water back to the fixture
through a filter assembly. The temporary system was installed after
discovering difficulties that the defueling water cleanup system washaving with cleaning up reactor vessel water. The system included afilter, a pump, and a “knockout” canister. (The knockout canister, filter
canister, and fuel canister were specifically designed and fabricated tocontain fuel debris.) The filter was a diatomaceous-earth pre-coat filter.
A loaded filter was backwashed of diatomaceous earth and debris into aknockout canister or filter canister for disposal. The system was restoredby establishing flow and injecting six pounds of clean diatomaceous
earth into the pump suction to coat the filter’s internal leaves.331 The
temporary system operated from February 1986 to May 1987, after
which the improved defueling water cleanup system took over the water
filtration function.332•

The in-vessel filtration system was a modified defueling water cleanup
system that supported vacuuming operations for final cleanup of thedefueled reactor vessel. The in-vessel vacuum system used a knockout
canister and filter canister in series.333

117

•

The processed water disposal system (evaporator) was used to disposeof 2.3 million gallons (8.7 million liters) of processed
accident-generated water. The system processed the water through aclosed-cycle evaporator. The purified distillate was reheated anddischarged into the atmosphere as a vapor, using the 100-foot-high
TMI-2 exhaust stack. The vapor contained essentially all of the water’s
tritium and a small fraction of its particulate contamination. The vapor
was released to the atmosphere in a controlled and monitored manner.
The remaining particulate contamination was concentrated in theevaporator bottoms, collected, and further concentrated to a dry solid.
A packaging system prepared the dry solid waste in containers
acceptable for shipment and burial in a commercial low-level
radioactive waste burial facility.334 The 2.3 million gallons of processedwater (influent) contained about 1,020 curies of tritium and about
2.3 curies of all other radionuclides, such as 42 percent strontium-90,
38 percent carbon-14, and 14 percent cesium-137. The water also
contained nonradioactive contaminants, such as 150 tons
(136,000 kilograms) of boric acid and 11 tons (10,000 kilograms) of
sodium hydroxide.335 About 99.9 percent of the dissolved radioactive
contaminants (other than tritium) contained in the evaporator influent
were collected as dry solid waste.336

Temporary Solid Waste Storage. The NRC stated in its Programmatic
Environmental Impact Statement (PEIS) that the TMI site should notProcessed water disposal system disposed of 2.3 million gallons of processed
accident-generated water by evaporation and disposal of solid residue.

118become a permanent radioactive waste disposal site. This statement was
based on the conclusion that TMI did not meet the criteria for a safe
long-term waste disposal facility for storing damaged fuel and radioactive
waste.337 (The Commission endorsed the PEIS in their April 27, 1981,
policy statement.338) The objective of the memorandum of understanding
between the NRC and DOE, as amended, for the removal and disposition ofradiological waste resulting from the cleanup of TMI-2 was to ensure that
the TMI site did not become a long-term waste disposal facility.339, 340
Several onsite facilities were constructed for temporary storage of solidradioactive waste products from cleanup activities that were being readiedfor transportation. Solid waste included spent EPICOR I and EPICOR II
resin liners; contaminated clothing, tools, and equipment; anddecontamination materials.
•

The interim storage facility was built to temporarily store EPICOR
resin liners and filters until the construction of the solid waste staging
facility was completed. The earthen facility was constructedunderground from corrugated galvanized cylinders with a weldedbottom. A 3-foot-thick concrete cover provided shielding at the top. The
ground provided radial shielding. The loaded liners were placed on a
galvanized drip pan. The design requirements for the facility were
provided to the licensee by the NRC341. The facility was ready for use
on November 5, 1979.342 The facility was abandoned in place after theliners were relocated to the new solid waste staging facility.343•

The interim solid waste staging facility (also known as the “carport”344) was used to collect and temporarily store (stage) low-levelsolid waste packages from both Units 1 and 2. All waste packages
placed in the facility, such as liners, drums, and metal boxes, werealready prepared for shipment. The facility performed passive functionsof protecting waste packages from precipitation and provided the means
to load packages onto trucks. The facility included a truck bay forloading and unloading. The facility was originally sized toaccommodate the waste generated by both units over a 6-month period.
Waste packages could be stored for up to 5 years.345 The facility was
ready for use on December 16, 1982.346•

The solid waste staging facility was used to collect and temporarily
stage radioactive waste, such as dewatered resins, filters, and sludge
from both Units 1 and 2 before shipment. Two of six planned concrete
storage modules were built, each consisting of 60 cells. Each cell wasconstructed of galvanized corrugated steel cylinders with a welded steel
base plate, and was sized to accommodate any combination of waste
119containers, such as resin liners, metal boxes, and drums. A concrete
cover 3-feet-thick and weighing about 14 tons covered each cell. Bothmodules shared a sump compartment to collect drainage. The facility
performed no active function other than storage.347, 348 Storage
Module “A” went into service in January 1980.349•

The waste handling and packing facility was used to process andpackage solid radioactive waste, such as contaminated clothing, tools,
and equipment. Processing of contaminated material consisted ofcompaction, size reduction, and sometimes decontamination for reuse.
No radioactive waste was stored in this facility.350 The facility was
ready for use in January 1987.351•

The TMI-2 spent fuel pools “A” and “B” were used to stage spentresin vessels from the submerged demineralizer system and loaded
defueling canisters. The spent fuel pools were filled with water forshielding. The submerged demineralizer system’s liners were preparedfor shipment and lowered into the shipping cask underwater. The loaded
shipping cask was lifted out of the “B” spent fuel pool, moved by craneover to the truck bay in the fuel handling building, and lowered onto thetruck bed. Defueling canisters had to be loaded individually into theshipping cask located on a special stand in the truck bay. The weight ofSolid waste staging facility was constructed to temporarily store radioactive
wastes, such as dewatered resins, prior to shipment. Two module structures
were built (center). The crane is loading a spent liner in a shipping cask. The
temporary interim liner staging modules shown along the bottom.

120

the shipping cask prohibited loading it inside the “A” spent fuel poolwhere the defueling canisters were stored. Each defueling canister was
lifted out of the pool, raised into the fuel transfer cask, moved to thetruck bay, and lowered into one of seven canister slots in the shippingcask. The loaded shipping cask was then lowered horizontally andmounted onto a railroad car. Each fuel shipment took about
1,000 person-hours to prepare.352
Packaging and Transportation. The NRC regulates packaging for thetransport, storage, and disposal of nuclear materials and waste. In addition,
the NRC regulates the design, fabrication, use, and maintenance ofcontainers for high-level radioactive shipments, including spent nuclear
fuel. The U.S. Department of Transportation (DOT) regulates packaging forthe transport of lower-level radioactive materials and waste. In addition,
DOT regulates shippers of all types of radioactive material and oversees
vehicle safety, routing, shipping papers, emergency response, and shipper
training requirements. The NRC requires its licensees to comply with
DOT’s safety regulations in addition to the NRC’s own requirements. NRC
regulations for the safety of transport and certification of packages for large
quantities of radioactive materials, including spent nuclear fuel, can befound in 10 CFR Part 71, “Packaging and Transport of Radioactive
Material.”353, 354 Also, NRC regulations under 10 CFR Part 61, “Licensing
Requirements for Land Disposal of Radioactive Waste,” provide
requirements for shipments of radioactive waste to commercial landdisposal facilities. These regulations provide limits on radionuclide
concentrations and requirements for the long-term structural stability of thedisposal container.355 State regulatory agencies, responsible for landdisposal facilities, and the DOE, responsible for accepting certain wastes
from TMI-2, also provided additional requirements.
The shipment of radioactive waste from TMI-2 by the licensee was required
to follow general regulatory requirements that applied to all licensees ofnuclear power plants. The NRC certified new shipping casks that were
designed and fabricated for the transportation of unique high-level wastethat came out of TMI-2 during cleanup. The NRC issued a “certificate ofcompliance” for each new shipping cask or reissued an existing certificate
of compliance for a modified cask. The certificate of compliance provides
the terms and conditions under which the cask can be used, such as the type,form, and quantity of material authorized for shipment in the cask, as wellas hardware specifications and internal packaging requirements. A user of acask must be certified by the NRC as a registered user. To become aregistered user, an applicant must have a quality assurance program inaccordance with 10 CFR 71 and a copy of the certificate of compliance121

(from the holder/owner of the cask), and must maintain maintenance records
on its casks. The NRC also reviewed and approved special waste containersthat were required to be shipped inside a shipping cask that provided special
safety features during transportation. In addition, the NRC reviewed
containers that were required to meet regulatory requirements (e.g., those of
10 CFR 61 and the NRC Technical Position Paper on Waste Forms356) for
disposal in commercial low-level radioactive waste burial facilities. The
types of shipping casts used at TMI-2 are summarized below:•

NRC Certificate of Compliance No. 9152/B, Revision 0,
Model CNS 1-13C II Type B Shipping Cask was issued on
April 4, 1982, and subsequently
revised. This new cask wasdeveloped for the transport ofspent SDS vessels withdewatered resins. The cask
consisted of a cylinder made oflead and steel lining, a bolted
lid, and two impact limiters(upper and lower) made ofsteel-lined rigid polyurethane
foam.357 This cask was latermodified and recertified for thetransport of core debris samples
to a DOE laboratory(Revision 8, dated
August 14, 1985). The
modification included asecondary package to fill thevoid of the cask, a DOT
Specification 2R container toprovide secondary containment
for the samples, a limit of total
fissile material to a subcriticalquantity, and a restriction to theModel CNS 1-13C-II shipping cask on the transport of only TMI-2 fueltruck transporter.
debris.358, 359•

NRC Certificate of Compliance No. 5026, Revision 10,
Model CNS 14-190 Shipping Cask was reissued in January 1984. This
existing cask was recertified to transport a high-integrity container thatwas used as an overpack (or outer container) for an EPICOR II pre-filter
liner. The cask was used by the DOE to ship EPICOR II liners from122

INEL to the commercial low-level radioactive waste burial facility near
Richland, Washington. The shipments began in May 1984 and werecompleted in February 1985. 360, 361, 362•

NRC Certificate of Compliance No. 9200, Revision 0,
Model No. 125-B Shipping Cask was issued on April 11, 1986, andsubsequently revised. The new cask was designed specifically totransport the loaded defueling canisters. Given that DOE tookpossession (ownership) of the fuel debris in its agreement with GPU,
DOE had authority under DOT regulations to self-certify its own
radioactive material shipping packages. Further, DOT regulations
required that DOE packages meet the requirements and standards of
10 CFR 71. DOE chose the NRC to be the regulator for approval of theModel 125-B shipping cask. Its design provided two testable levels ofleak-tight containment in accordance with NRC regulations. NRC
regulations also required the cask to provide protection against any
radiological release during normal conditions of transport and duringhypothetical accident conditions. The cask safety analysis report was
independently evaluated by the NRC in accordance with regulatory
requirements of 10 CFR 71. The NRC issued the certificate ofcompliance for the rail cask to DOE on April 21, 1986. Three casks
were fabricated for the exclusive shipment of TMI-2 fuel debris. Two
casks were purchased and owned by DOE. The third cask was owned by
the fabricator and leased to GPU. 363

Model 125-B rail shipping cask used to transport seven defuelingcanisters loaded with core debris. Impact limiters shown with flags.

123

The Model 125-B fuel cask consisted of five major components: theouter containment vessel, inner containment vessel, upper and lower
canister impact limiters, canister shield plugs, and cask impact limiters.
Each cask had its own transportation system of a skid and rail car. Gross
shipping weight of the shipping cask was about 183,000 pounds. The
primary and secondary containment lids weighed 3,000 and
5,200 pounds respectively. The cask impact limiters weighed about
11,700 pounds each and the sunshield weighed about 500 pounds.364 Up

Exploded view of the Model No. 125-B Shipping Cask. Fuel debris wascontained in the seven defueling canisters.

124

to 21 defueling canisters could be sent in a single rail shipment (sevencanisters per cask, three casks per shipment). The fuel shipments began
on July 20, 1986, and were completed on May 9, 1990. There were
22 rail shipments for a total of 342 canisters of core debris transported
to the INEL. The total number of canisters shipped included 286 fuelcanisters (which contained partially intact fuel assemblies and large
debris picked up from the reactor vessel), 12 knockout canisters(containing core debris vacuumed from the reactor vessel and reactor
coolant system), and 62 filter canisters (containing fine debris thatpassed through the knockout canisters).365•

HIC overpacks to dispose of EPICOR II pre-filter liners. A special
high-integrity container (HIC) was designed and fabricated for DOE todispose of the original 45 highly loaded EPICOR II pre-filter liners at
the commercial low-level radioactive waste burial facility near
Richland, Washington. The HIC was an overpack (EPICOR II liner
sealed inside the HIC) that would remain stable below ground for aminimum of 300 years (about 10 half-lives of the predominantisotopes). The HIC consisted of a cylinder made of reinforced concrete
and a permanently sealed lid. A vent system cast in the lid provided
passive venting of the container. This HIC was restricted for use with
only EPICOR II pre-filter liners generated at TMI-2. Certificate ofCompliance No. WN-HIC-01 was issued for this HIC by the State ofWashington on March 23, 1984.366, 367•

HIC overpacks to dispose of SDS vessels. After an unsuccessfulattempt to qualify the design of the SDS pressure vessel as a HIC, SDS
vessels were shipped and buried inside a polyethylene HIC. These poly
HICs were permitted exclusively at the commercial low-level
radioactive waste burial facility in Barnwell, South Carolina. A concrete
liner was inserted between the SDS vessel and the poly HIC to provide
the shielding necessary to ensure the authorized service life of the polymaterial.368•

Commercial metal HIC. A commercially available HIC was used at
TMI-2 to dispose of radioactive waste that contained highconcentrations of strontium-90 and transuranic material. (At highconcentrations, the waste classification system of 10 CFR 61 required
stabilization of the waste either by solidification or HIC.) The metallic
HICs were made of a corrosion-resistant alloy material. In 1984, theState of Washington granted interim authorization for the first shipment
to the commercial low-level radioactive waste burial facility near
Richland before the NRC completed its review of the technical125

evaluation report for these HICs. In April 1988, NRC completed its
review and concluded that the HIC met or exceeded all of therequirements of 10 CFR Part 61 and the recommendations of the NRC
Technical Position Paper on Waste Forms. The metallic HICs weremodified and qualified to function as demineralizer vessels in thedefueling water cleanup system and later in the last EPICOR II
configuration.369
Hydrogen Generation in Waste Containers. Generation of flammablegases inside sealed radioactive waste containers was a potential hazard. Tworeactions contribute to gas generation: (1) the reaction between metals andwater, which oxidizes the metal and releases hydrogen gas; and (2) longterm exposure of water and organic materials to ionizing radiation(radiolysis). Because many TMI-2 wastes—particularly SDS zeolites andcanisters of fuel debris—were loaded with high concentrations ofradioactivity, the latter reaction was a potential source of flammable gaseous
mixtures (oxygen plus hydrogen).370•

EPICOR II liners. A unique gas sampler and vent tool was designed by
DOE to remotely remove the vent plug from the EPICOR II ion
exchange resin liner, sample the gas content, vent the container, purge

Demonstration high integrity container (HIC) being covered in the trench at
the commercial radioactive waste disposal site near Richland, Washington.
The HIC contained the highly loaded EPICOR II prefilter No. 18. The
demonstration set the precedent for disposing of the remaining EPICOR II
prefilters that were temporarily stored at INEL.

126

the container with inert gas, and reinstall the vent plug. The venting of
the liners was performed at the Solid Waste Storage Faculty in theindividual cells where EPICOR liners were stored. All operations weredone remotely under a portable concrete shielding blockhouse over the
storage cell. A portable remote support trailer that was located on top of
the storage modules contained the gas circulator for purging theatmosphere of a storage cell with inert gas, a gas chromatograph for online sampling, controls for the gas sampler including TV monitors.371, 372
Laboratory analysis of several highly loaded EPICOR II pre-filter liners
identified the average hydrogen production rate of a 2000 Curie liner
would generate a 4 percent mixture of hydrogen in a nitrogen andcarbon dioxide atmosphere in about 100 days. The vent tool lowered thecontent that would be generated during shipment to less than the
flammable limit of 4 percent.373•

SDS vessels. Radiolytic gas generation in highly loaded SDS vessels
was calculated to be significant. Measurements of gas buildup
confirmed that this could be a problem during storage andtransportation. A DOE research program developed a number oftechniques to deal with the problem. To reduce buildup of flammablegas during transport and storage, the vessels were drained and vacuum
pumped to remove free water. A catalyst was added to each vessel torecombine the hydrogen and oxygen as it was generated. A pressure
relief system, consisting of a burst diaphragm and micropore graphite
filter, was also added to each vessel to prevent uncontrolled, long-termbuildup of non-recombinable gas mixtures. (A net hydrogen buildup canoccur due to oxygen scavenging by various chemical reactions, such as
the formation of carbon monoxide and carbon dioxide from oxidation oforganic materials trapped within the zeolites.)374, 375•

Defueling canisters. Catalyst beds were used in the TMI-2 core debris
canisters to recombine radiolytic hydrogen and oxygen and prevent
buildup of combustible mixtures of gases. Each canister contained arecombiner catalyst package incorporated into the upper and lower
heads. The catalyst bed was designed with porous metal filters at eachend to contain the panicles while allowing gases to flow to and throughthe catalyst.376, 377, 378, 379

Document Collections. Documents relating to waste management activities
are grouped into the following document collections on the DVDs:•

EPICOR II system (see the DVD folder, Water Processing: EPICOR)
127

•

submerged demineralizer system (see the DVD folder, Water
Processing: SDS)•

solid waste processing and storage (see the DVD folder, Solid Waste)•other waste-related processing systems and activities (see the DVD
folder, General Waste Management)•

status and lessons learned reports (see the DVD folders under the Status
and Summary Reports collection)•

incidents and deficiencies (see the DVD folder, NRC Preliminary
Notifications)

The following types of documents are included in the above document
collections:•system and facility description reports•

technical evaluation reports and safety evaluation reports•

certificates of compliance for shipping casks (current at the time of first
shipments)•

research reports, NRC technical NUREG reports, GEND reports, andtechnical reports from DOE national laboratories•

other correspondence between the NRC and the licensee relating to thewaste management systems and activities, such as notifications,requests, reviews, and approvals

Defueling tool: three point gripper hydraulic attachment.

128

Spent EPICOR II resin liner inside the shielded transfer cask (yellow) being
lowered into a shipping cask.

A highly radioactive spent SDS resin liner (center) being transferred into ashipping cask (lower center) while underwater in the spent fuel pool.

129

Need establishedfor surface
decontamination
HP spray3 – 4000 psi
7 – 12 gpm
ambient temp.

Yes

Are surfaces
equipment
surfaces?
No

Are surfaces
vertical?

Yes

No

HP spin jet
6000 psi
15 gpmambient temp.

YesIs
surfacestainlesssteel?

Yes

LP flush
1 – 2000 psi
20 – 25 gpm
ambient temp.

No

HP spray3 – 4000 psi
7 – 12 gpm
ambient temp.

Issurface diamondplate?
NoIs
surface stainlesssteel?

Yes

LP flush
1 – 2000 psi
20 – 25 gpm
ambient temp.

AbrasivetechniqueNo

LP flush
1 – 2000 psi
20 – 25 gpm
ambient temp.

YesIs
surface covered
with loose debris?
No

HP spin jet
6000 psi
15 gpmambient temp.

Mechanicalscrubbingchemicals

Establish need forcontamination
control measures(Abbreviations: Gallons per minute---gpm, high pressure---HP, low pressure---LP,
pounds per inch---psi, temperature---temp)

Decontamination logic diagram based on results of the gross decontamination
experiment in March 1982 (see GEND-034).

130

7 Decontamination
The 1979 accident involved a loss of reactor coolant and resulted in severe
damage to the reactor fuel. When the reactor coolant pump flow was
restored, radioactive contamination in the form of fuel debris and fission
products was distributed by the coolant throughout the reactor coolant
system. Reactor coolant, carrying fuel debris and fission products as
dissolved and particulate material, flowed out of the reactor coolant systemthrough the stuck-open pressurizer pilot-operated relief valve and into thereactor building’s basement and sump. During the first hour of the accident,
spilled reactor coolant in the reactor building sump was automatically
pumped into the auxiliary building holding tanks, which then overflowed
along with the sumps. Although this radioactive water did not initiallycontain damaged fuel, later, equipment leakage from the makeup andpurification system did contain fuel debris and fission products which mixed
with the water in tanks, sumps, and floor drains. Exposed surfaces in thereactor building and the auxiliary and fuel handling building were
contaminated with radionuclides in the reactor coolant, as well asradionuclides that became airborne, as hot reactor coolant flashed into
steam. Airborne contamination entered the ventilation systems and spreadthroughout the auxiliary and fuel handling building.380 For some unknown
reason, the ventilation systems stopped several times, which diverted
airborne radionuclides to unintended areas.381 After the accident, the waterin the reactor building’s basement was heated by residual heat from thereactor vessel, evaporated, condensed on the cooler walls, and drained down
onto the floors and back into the basement.382
Cleanup of the auxiliary and fuel handling building started shortly after theaccident. Airborne releases contaminated surfaces on the upper-level floor(at the 328-foot elevation) and mid-level floor (at the 305-foot elevation) of
the auxiliary building, and liquid releases to the drain system contaminatedsurfaces on the basement-level floor (at the 280-foot elevation). The interiorof the building, including 26 piping systems, was contaminated byradioactive material, though less severely than the interior of the reactor
building; because most of the interior surfaces, such as walls and floors,
were constructed of uncoated concrete, radioactive materials penetrated thesurfaces to varying depths. Approximately 510,000 square feet of surface inthe auxiliary and fuel handling building required decontamination whencleanup operations began.383, 384
The decontamination experience at TMI-2 differed from past experience inthe nuclear industry in that cleanup of the reactor building did not begin
immediately. During the time between the accident and the start of the131cleanup, the humidity in the reactor building was 100 percent, causing
precipitation in the form of rain (steadily dripping condensation) inside thereactor building. One result of the rain was that dose rates during the initial
personnel entries into the reactor building 14 months after the accident werelower than expected because radionuclides had been rinsed downward.
A second result was that radionuclides permeated into porous surfaces suchas uncoated concrete, were incorporated in corrosion layers as iron surfaces
rusted, and were trapped in paint layers. Re-cleaning of previously cleaned
areas was still required during the cleanup period, with concomitant
exposure of workers.385
Reactor coolant containing core debris and fission products was discharged
on the reactor building’s basement floor through the stuck-open pressurizer
pilot-operated relief valve when the block valve was periodically opened tocontrol pressure. Consequently, these areas were covered with radioactive
sediments, consisting primarily of river water sediment from a leaking aircooler inside the reactor building, concrete dust, and dirt.386
Decontamination Objectives. The overall objectives of the TMI-2
decontamination efforts were to maintain access to and operability of plantsystems, to support defueling preparations and operations, and to permit thetransition of the facility to a long-term storage condition. Shorter-term
decontamination objectives focused on the removal or stabilization of

Decontamination using high-pressure spray.

132

contamination in order to reduce occupational exposure and to prevent
release of contamination to the environment. Longer-term decontamination
objectives ensured that any remaining contamination was stable andsufficiently isolated for long-term storage.387
The initial decontamination objectives in the auxiliary and fuel handlingbuilding were to permit access without restriction because of surface orairborne contamination, to reduce radiation exposure from gamma sources
to ALARA levels, and to prevent recontamination from other cleanup
activities or system leaks. Decontamination of cubicles was required topermit access for inspection and maintenance of plant equipment, which had
been deferred since the accident. In addition, decontamination of systems
and equipment in the auxiliary and fuel handling building were required for
activities to decontaminate the reactor building and defuel the reactor vessel
and primary coolant system.388 The initial decontamination objective in thereactor building working areas (at the 305-foot and 347-foot elevations) was
to reduce radiological conditions (general area radiation, airborne gaseousand particulate activities, and surface contamination levels) to ALARA
levels and to maintain those conditions in a way that would permit defueling
operations.389 The objective in the basement was to remove sludge from allaccessible areas and stabilize contamination during long-term storage.390
The final decontamination objective was to stabilize localized radiologicalconditions in the plant, regardless of whether or not access was required forcleanup activities. The associated decontamination efforts were focused on
meeting the specified goals (such as general-area dose rate, surface
contamination level, and hot-spot dose rate) required to place the facility inlong-term storage after defueling completion. Long-term storage would
allow decay of the radionuclides that remained in the facility so that workers
would be exposed to lower levels of radiation during future decontaminationand decommissioning.391, 392
Decontamination Criteria and Goals. The safety analysis report for
post-defueling monitored storage (PDMS) provided a set of
decontamination objectives (see above), baseline radiological criteria, anddecontamination end-point goals for the decontamination program. Two ofmany prerequisites that were required for entering PDMS included(1) contamination reduction consistent with ALARA principles to meetestablished contamination-level goals in each of the areas of the auxiliary
and fuel handling building and (2) radiation reductions consistent with
ALARA principles, as necessary, to levels which would allow necessaryplant monitoring, maintenance, and inspections.393
133

As discussed previously, decontamination objectives were assigned
according to general area within buildings, and were based on initial
contamination levels, the need for personnel access to perform specificactivities inside the area, and the possibility of release of radioactivity to theenvironment. Baseline radiological criteria were established for eachdecontamination objective while considering the anticipated need for
personnel access during cleanup and post-cleanup activities. Criteria werebased on access frequency (hourly, daily, weekly, monthly, or quarterly)
and radiological conditions (general-area dose rate, maximum hot-spot doserate, and smearable surface contamination level). Criteria were intended as
guidelines rather than absolute requirements.394
Specific decontamination goals were set for each decontamination objective(specific area or cubicle) based on baseline radiological criteria. For
example, the specific decontamination goals for the 347-foot elevation
operating floor in the reactor building that required infrequent (quarterly)
access, was less than 30 mR/hr for general-area dose rate, and less than
50,000 dpm/100 cm2 for surface contamination within 7 feet of the floor.
The specific goals for corridors and access ways in the auxiliary building

Strippable coating used in decontamination of floor surfaces.

134that required routine access (40 hours per week) were less than 2.5 mR/hr
for the general-area dose rate, 10 mR/hr for the maximum hot-spot doserate, and less than 1,000 dpm/100 cm2 for surface contamination within 7
feet of the floor. The specific goals for the seal injection filter room in theauxiliary building that may require weekly access, were less than 500 mR/hr
for the general-area dose rate, less than 1,000 mR/hr for maximum hot-spot
dose rate, and less than 50,000 dpm/100 cm2 for surface contamination.
General decontamination criteria for piping systems, equipment, andcomponents were based solely on their contribution to area dose rates.
Specific decontamination goals are listed in the PDMS safety analysis
report (refer to DVD folder, After Defueling).395
Cleanup Activities. Decontamination of building surfaces, systems, andequipment included multiple activities across the following categories:396•

Loose equipment. Removal of miscellaneous equipment and debris thatwere in the facility at the time of the accident, such as ladders,
scaffolding, tools, and portable equipment.

•

Installed equipment. Decontamination or removal of installed
equipment, such as piping systems, air conditioning and exhaustequipment, cable trays, and electrical and lighting equipment.

•Interior surfaces. Decontamination of interior building surfaces
consisting of metal and concrete materials.

•

Sludge. Removal of contaminated sediment (sludge) from tanks andsumps in the auxiliary and fuel handling building and from the reactor
building’s basement floor and sump. The sediment was transferred tothe spent resin storage tanks for processing. Concentrated solids weretransferred to a disposal container for cement solidification.397•

Resins. Removal of highly contaminated resins from the makeup andpurification system demineralizers located in the auxiliary building.
This system cleaned up the reactor coolant system during normal plant
operations. During the accident and thereafter, about six kilograms offuel and fission products were deposited in system filters anddemineralizers.398•

Recovery and cleanup equipment. Decontamination of systems andequipment used for cleanup and defueling activities. Grossdecontamination of refueling tools and cleanup equipment for reuse ordisposal was performed in two temporary equipment decontamination
135

facilities located in the auxiliary building and reactor building. In 1987,
the waste handling and packaging facility became operational, which
provided a permanent facility for integrated equipment
decontamination. This new facility also processed and packaged
discarded equipment for disposal.399 Following defueling operations,equipment located in the TMI-2 spent fuel pools was removed and pools
were emptied of water and decontaminated. Spent fuel pool and the fueltransfer canal contained fuel debris that was carried out on the surfaceof the defueling canisters.
•

Support activities. Various support activities to ensure worker safety
and to measure the effectiveness of the cleanup.

The decontamination of the reactor building’s atmosphere andaccident-generated waste water are discussed in previous sections.

Workers decontaminating the auxiliary building using the manually appliedscrubbing method.

136

Decontamination Methods. Combinations of well-known methods wereused in the decontamination of building and equipment surfaces in theauxiliary and fuel handling building. Methods used for decontamination of
surfaces inside the reactor building were based on the results of the grossdecontamination experiment and subsequent experience gained indecontamination of the auxiliary and fuel handling building. The following
decontamination methods have been reviewed by the NRC and saw at least
limited use:400, 401, 402•

Abrasive blasting of steel surfaces with particulate driven at highvelocity to remove contamination. This method was especially suited
for small or irregular surfaces that were not compatible with otherdecontamination techniques.

•

Chemical decontamination of external surfaces of pipes, tanks, andsystem internals. The gross use of chemical agents was limited becauseof the potential for drain-off of such agents to damage water processing
systems.

•

Dry vacuuming to remove powdered contaminants and dried residue.
This method was used especially for water-sensitive components that
could not be flushed with water.

•

Low-pressure water flush at levels between 100 to 1,000 pounds per
square inch (psi), flow rates up to 25 gallons per minute (gpm), andwater temperatures up to 170 degrees Fahrenheit. This method was used
to clean equipment and loose surface debris. Examples included thepolar crane, steam generator housing structures (“D-rings”), missile
shields, refueling canal, and refueling bridge.

•

High-pressure water flush to remove unbonded surface coatings atpressures between 2,000 to 10,000 psi and flow rates of 4 to 30 gpm.

•

Ultrahigh-pressure water flush to remove rust, scale, nuclear-grade
coatings, and surface concrete at pressures up to 60,000 psi and flow
rates of 1 to 2 gpm.

•

“Scabbling” of walls and floors to aggressively remove concretesurfaces and surface coatings. The scabblers used pneumaticallyoperated reciprocating pistons equipped with tungsten carbide bits topulverize the concrete surface. A vacuum system with a high-efficiency
particulate air filter was attached. A major scabbling campaign began in
the reactor building in October 1984.403
137

•

A steam and vacuum decontamination system was used throughout the
TMI-2 cleanup to decontaminate painted and uncoated concrete,
ductwork, diamond deck plates, lead bricks, penetration covers, piping,conduit cable trays, and drain covers. The system used a single-headmachine that minimized the spread of contamination and impairment ofvision from sprayback.404 The system was first tested in the auxiliarybuilding on June 14, 1984, and decontamination in the reactor building
began in March 1985.405•

Strippable coatings that involved the application of an organic coating
which contained chemicals to aid in the removal of radioactive
contaminants from the surface. As the coating dried, it cracked andpeeled away from the surface.

•

Scrubbing to remove loosely held contamination on floors and walls
using manually applied or mechanically driven rags, absorbent cloths,
brushes, pads, grit, and chemical agents.

•

Wet vacuum to remove puddles of contaminated cleaning fluids after
flushing.

Remote Robotic Equipment. Remote-controlled robotic vehicles andsupporting control equipment were used extensively to perform work in thereactor building’s basement, the makeup demineralizer room in the auxiliarybuilding, the reactor coolant pump seal injection valve room in the fuelhandling building, and the reactor vessel. These vehicles were both versatile

Remote control robotic high-pressure spray washing system.

138

and productive, and proved useful in many different tasks, including video
camera inspections, radiation monitoring, sediment sampling, acquisition of
concrete core samples, high pressure water flushing, concrete scabbling andscarification, and debris pickup and removal. In addition to the remotevehicles, fixed-position, remotely operated tools were developed for workinside the reactor vessel. The tools included a plasma arc cutting system toremove the stainless steel core support assembly, and several manipulator
arms for handling damaged fuel and structural components.406 The use ofrobots at TMI-2 did not require any NRC licensing actions; however,activities in which robots were used, like most recovery and cleanup
activities, required safety evaluations by the licensee and the NRC. Key
robots used at TMI-2 are summarized below:•

ROVER or remote reconnaissance vehicle (RRV) was used in thereactor building’s basement to perform video and radiation surveys,
collect sludge samples from the floor, collect core samples from thewall surface, flush walls with high-pressure water, remove the surfaceof the walls using an ultrahigh-pressure scarification system, andremove sludge. The RRV was a tether-controlled, six-wheeled work
platform with multiple attachments or modules to perform the various
tasks. ROVER was operated by two operators in a control room locatedoutside the reactor building.407, 408•

LOUIE I remote vehicle was used to measure the radiation profiles of
the two makeup demineralizer vessels, and to remove loose pre-accident
debris and salt deposits on the floor inside the seal injection valve room.
On loan from DOE, LOUIE I was a tether-controlled, tracked work
platform with a telescoping boom-mounted manipulator that had been in
use at DOE’s Hanford Site since the 1950s. The control console wassetup in the accessible hallway near the entrance to the room.409•

LOUIE II remote vehicle was used to perform remote floor scabblingin the seal injection valve room. An attached 3-piston pneumatic
scabbler was used to pulverize the floor surface while vacuuming loose
concrete. LOUIE II was a tether-controlled, six-wheeled work platform
with a heavy frame to withstand the stresses of the scabbler. The control
console was setup in the accessible hallway near the entrance to theroom. 410, 411•

WORKHORSE or remote work vehicle (RWV) was a large, heavy-duty
robot built for decontamination and demolition work in the basement of
the reactor building. WORKHORSE was a tether-controlled, sixwheeled work platform that was 10 times heavier than the RRV and had
139a boom with a seven-meter vertical reach. A two-level control building
was built in the turbine building for three operators and support staff tooperate the RWV and manage the work activities. The RWV wassuccessfully tested in mockups, but never used due to changes incleanup direction.412•

Mini-Rover was a commercial submarine vehicle modified to remove
larger fuel debris inside the pressurizer. The Mini-Rover had a color
camera, pincer arm to break apart debris, and a scope. The submarine
was fitted with a ballast tank to improve mobility.413•

Remote manipulator performed defueling operations in areas of thereactor vessel that were not directly below the defueling work
platform’s working slots. The manipulator could handle defueling tools
and the in-vessel video viewing system, and pickup fuel debris. The
manipulator was mounted to the manual tool positioner on the defuelingwork platform and had a 4-foot reach and six degrees of freedom of
motion. The manipulator could be controlled from inside or outside thereactor building.414•

ACES or automated cutting equipment system was a remote-controlledplasma arc torch installed in the reactor vessel to cut the multi-layered
lower core support assembly following bulk defueling. ACES consistedof a support bridge with a carriage and trolley to provide horizontal
X-Y plane movement. A manipulator arm provided motion in thevertical Z-axis direction, including rotation and bending motions. At thelower end of the manipulator arm was a pneumatically operated gripper
with a plasma torch and effector for cutting. A high-velocity stream ofhigh-temperature ionized nitrogen gas plasma transferred an electric arc
to melt the cut area.415, 416 The two control consoles (one for the
manipulator subsystem and the other for plasma subsystem) werelocated in the turbine building.417•

MANFRED or manipulators for reactor defueling was a robot
manipulator system designed and built for underwater disassembly anddefueling of reactor vessel components. MANFRED consisted of amanipulator with various work attachments and grabber manipulators.
The system was tested but never used due to the success of ACES
(above) and manual handling of component pieces and loose fueldebris.418

Additional information on the uses of remote equipment at TMI-2 areprovided in the following documents: International Atomic Energy Agency
140

report “Catalogue of Methods, Tools and Techniques for Recovery from
Fuel Damage Events” (IAEA-TECDOC-627); the DOE report “TMI 2:
Lessons Learned by the U.S. Department of Energy, A Programmatic
Perspective” (DOE-ID-10276); the American Nuclear Society’s Nuclear
Technology journal (Volume 87; not included in the DVDs); and the EPRI
report “Final TMI-2 Technology Transfer Progress Report”(EPRI-TR-100643; not included on the DVDs).
Key Actions. By the end of the cleanup program, the floor contaminationlevels in most areas of the auxiliary and fuel handling building were reduced
to those typical of pre-accident conditions. In the reactor building, radiationlevels in frequently accessed areas were reduced by 85 percent by scabbling
accessible areas and shielding finite radiation sources. A selection of otherkey actions during the cleanup period is summarized below:•

First decontamination workshop. On November 27–29, 1979, theFacility Decontamination Technology Workshop, sponsored by DOE,
was held in Hershey, Pennsylvania. The workshop provided those
involved with the cleanup at TMI-2 a summary of experience regarding
events and incidents at other facilities that necessitated decontaminationand dose-reduction activities (see GEND-002).419•

Gross decontamination experiment. In March 1982, the grossdecontamination experiment was conducted on various levels andsurfaces inside the reactor building. The objectives of this experimentwere to evaluate several decontamination techniques and todecontaminate the reactor building surfaces. Results of the experimentshowed that gross decontamination could achieve the goal of reducing
smearable contamination (see GEND-034).420 In addition, experiments
showed that surface contamination contributed much less togeneral-area exposure rates, whereas rusty metal surfaces, large
equipment, and bags of garbage contributed far more. Findings also
indicated that recontamination was a problem; high-pressure flushing,mechanical scrubbing, and strip coating provided higher
decontamination factors; residual water from flushing must be removed(such as by squeegeeing or vacuuming) to prevent recontamination fromsuspended contaminants; and complex surfaces and equipment were
harder to clean.421•

Gross reactor building decontamination. On September 17, 1982,
gross decontamination began in the reactor building at the 305-footentry-level elevation, and at the operating floor on the next level above.
The effort was designed to reduce smearable levels of contamination to141the point that workers would be able to remove much of their bulky
protective clothing and their respirators.422•

Worker protection restrictions lifted in areas of the auxiliary building.
In November 1982, upper-level corridors of the auxiliary building
became accessible to workers without the need to wear
anti-contamination clothing. (Respirators had not been required forentry into the upper corridors since October 1979.) The following
month, the auxiliary building’s corridors in the basement (at the281-foot elevation) became accessible without the need to wear
respiratory protection masks.423 Overhead areas, such as ceiling andcable trays, were decontaminated only to the extent that they would not
re-contaminate the floor below. In such cases, a radiation work permit
was required to access ceiling areas.424•

Respiratory protection restrictions lifted in areas of the reactor
building. On June 28, 1984, workers entered the reactor building
without respiratory protection for the first time since the accident, andsubsequent entries were made without respirators, in accordance with
ALARA principles.425, 426

The command center for the gross decontamination experiment in March
1982 contained the control functions for entry into the reactor building withpositions for safety, radiological engineering, operations, entry coordinator,and command center management.

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•

GPU’s Decontamination Task Force Report. On December 18, 1985,
the licensee issued the Decontamination Task Force Report. The reportprovided a review of the effort required to decontaminate TMI-2 and anevaluation of the reduction in occupational exposure during
post-defueling monitored storage (PDMS). The objective of the task
force was to arrive at a consensus technical approach to each of the
major areas of decontamination work: remote-equipment development;
sludge transfer and disposal; D-ring dose reduction anddecontamination; recovery of the reactor building’s basement; auxiliary
and fuel handling building surface decontamination; non-reactor coolant
systems decontamination; reactor building ventilation modifications;reactor building surface decontamination; reactor coolant systemdecontamination, and waste management for the decontamination of thereactor building.

The task force estimated that final decontamination, as part of futuredecommissioning, would result in a total occupational exposure in therange of 2,710 to 5,770 person-rem, assuming that further
decontamination was deferred until after a 30-year period of PDMS.
The task force concluded thatdeferring decontamination for aperiod of 30 years would result in a
potential occupational exposuresavings in the range of 4,500 to
9,800 person-rem. This savings was
based, in part, on reduction inradiation dose rates due to the natural
decay of radioactive materials, andadvances in both remote cleanup
technology, and chemical
decontamination methods.427 Resultsfrom the task force report were
documented in Appendix 5A of the
PDMS safety analysis report,“Potential Reductions in
Occupational Exposure Due to PostA radiation work permit (RWP) was
Defueling Monitored Storage,” (seerequired to access contaminatedthe DVD folder, After Defueling; theceiling areas above the clean areas of task force report itself, however, isthe auxiliary building main corridors.
not available on the DVD).
•

Reactor building sludge removal. On March 31, 1987, the robotic
removal of sediment in the reactor building’s basement began and was143

completed in July 1987.428 Approximately 10,800 pounds(4,900 kilograms) of wet sludge, which contained about 4 kilograms offuel debris, was removed from the reactor building’s basement floor,
pumped into a tank located in the auxiliary building, and solidified for
burial at a low-level radioactive waste disposal site. A robotic
desludging system was used to vacuum about 40 percent of thebasement floor area which was the only area accessible to the robot. The
removal efficiency of desludging was greater than 90 percent.429 The
robotic flushing of the basement floor was completed 3 months later.430•

NRC’s PEIS supplement for completing cleanup. In August 1989, the
NRC issued the final supplement to the Programmatic Environmental
Impact Statement (PEIS), NUREG-0683, Supplement 3, which
evaluated the potential environmental impacts of alternative approaches
to completing the TMI-2 cleanup. The supplement examined theestimated occupational radiation doses associated with the licensee’s
proposal for delayed decommissioning and five NRC-identified
alternatives that were evaluated quantitatively. The dose estimates werebased on a task analysis of the cleanup work to be performed. (SeeSupplement 3 of the PEIS in the DVD folder, Guidance-PEIS.)

Highly radioactive sludge on the basement floor of the reactor building as
viewed from an upper level. A robotic desludging system was used to vacuum
the 40 percent of the basement floor area that was accessible to the robot.

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•

Plant-entered PDMS. On December 28, 1993, the NRC granted anamendment to permit the licensee to place the plant in a long-term
PDMS; the amendment also provided the PDMS technicalspecifications. The NRC safety evaluation concluded that the routine
release of any significant quantity of radioactive material during PDMS
had been minimized, in part, by the decontamination of large sections of
the reactor building, auxiliary and fuel handling building surfaces,
equipment, and piping. In addition, the NRC concluded that radiationlevels within the facility had been reduced to such an extent that plantmonitoring, maintenance, and inspection could be performed.431 (Seesection on After Defueling.)

Document Collections. Documents relating to decontamination activities
are provided in the following DVD folders:•documents relating to technical evaluation reports of decontamination
activities; decontamination schedules; NRC safety evaluation reports;
NRC technical NUREG reports; GEND reports and technical reports
documenting research results from DOE national laboratories; and othercorrespondence between the NRC and the licensee, such as
notifications, requests, reviews, and approvals (see the DVD folder,
Decontamination)•

DOE summary reports, such as annual and lessons-learned reports, andperiodical publications (see the DVD folder, DOE/National Laboratory
Status Reports)•

NRC TMI Program Office (TMIPO) weekly status reports (see theDVD folder, TMI Program Office Weekly Status Reports)•

post-defueling monitored storage safety analysis reports and documents
relating to decontamination requirements (see the DVD folder, After
Defueling)

Project schedules for decontamination activities are provided in the DVD
folder General Management and Oversight.

Defueling tool: hook attachment.

145

Technicians on the defueling work platform.

146

8 Defueling
The TMI-2 accident resulted in severe reactor core damage and migration ofmolten core materials onto the reactor vessel’s lower head. Core damage
and relocation occurred within four hours of the accident initiation, after
which long-term cooling stabilized the damaged reactor core. Various
reactor core inspections (see the section on Data Acquisition and Analysis)
and observations from early defueling activities provided information toestimate the end-state core configuration. The INEL report “TMI-2
Accident Scenario Update” (EGG-TMI-7489) identified four regions within
the original core volume, as shown in its iconic figure of the damaged
reactor core (see Section 5): upper cavity void region, debris bed region,
previously molten region, and partially standing fuel assemblies (or “stubs”)
region. Molten reactor core broke through the hard crust of the melt region
at 224 to 230 minutes into the accident, penetrated some of the baffle plates,
flowed through the complex five-layered lower core support assembly, andsettled onto the reactor vessel’s lower head.432 The results of inspections andearly defueling that were cited in the INEL report (1986) and the earlier
GEND-007 report (1981) provided the technical basis for planningdefueling approaches and necessary equipment. During the period of
exploration of the damaged reactor core, information from new inspections
and observations altered defueling strategies and tool designs.433
The TMI-2 cleanup effort took ten years with a collective manpower effortof over 3.6 million person-hours to complete. The reactor vessel defueling
operations spanned a five-year period from October 1985 throughJanuary 1990 and involved over two million person-hours. A total of about
133,000 kilograms of fuel, cladding, structural, and control materials were
removed from the reactor vessel during the five-year effort. During July andAugust of 1991, the reactor vessel was drained to make final measurementsof the residual fuel remaining in the vessel. An estimated residual fuelquantity that remained in the reactor vessel following defueling wasapproximately 1 percent of the original 94,000 kilograms of uranium oxide
fuel inventory. The total occupational dose resulting from all cleanup
activities was less than 6,500 person-rem over the first 10-year period. The
cumulative occupational dose for defueling and defueling support activities
was below 2,000 person-rem.434 The exposure rate to defueling workers
averaged less than 10 millirem per hour.435
Reactor Core Damage. The final defueling report, which was submitted by
the licensee in 1990 for NRC review, provided the complete picture of theend-state configuration of the core. This report was based on actual
147

Reactor vessel with upper plenum removed andinternals indexing fixture installed.

Lower core support assembly inside reactor vessel. This massive, five-layered
structure was cut into pieces to allow access to the fuel debris in the lower
head. Pieces are stored in the reactor building.

148defueling experience during the five-year defueling period. The final report
indicated that the original core inventory included about 94,000 kilogramsof uranium oxide (fuel) and 35,000 kilograms of cladding, structure, andcontrol (neutron-absorption) material. The total amount of core debris wasestimated to be 133,000 kilograms, accounting for oxidation of corecomponents and melt from part of the upper plenum grid plate. About
50 percent of the original core melted. The defueling program had toconsider how to remove the fuel debris to the extent that inadvertentcriticality was precluded in each confined area.436 The distribution of corematerial as documented in the final defueling report is summarized below:•

Upper core void. The upper core void or cavity consisted of only
42 partially intact fuel assemblies standing at the periphery of the void.
Only 2 fuel assemblies of the original 177 assemblies still had most of
their fuel rods intact. The void was about 26 percent of the original core
volume and measured 1.5 meters deep from the top of the original coreto the debris bed.437•

Upper debris bed. The debris bed consisted of 26,000 kilograms of corematerial, such as whole and fractured fuel pellets, control rod spiders,fuel assembly end fittings, broken fuel rods, and resolidified debris. The
bed rested on top of the resolidified, hardened mass and was about
0.6 meter to 1 meter deep.438•

Resolidified mass. The solid metallic and ceramic mass consisted ofabout 33,000 kilograms of core material. The mass rested on partiallyintact fuel assembly “stubs.” The mass measured about 3 meters indiameter, 1.5 meters deep in the center, and 0.25 meter deep around theedges.439•

Intact assemblies. Partially intact fuel assembly stubs located under the
resolidified mass and the peripheral standing assemblies comprised
about 45,000 kilograms of core material. The standing stubs varied inlength from about 0.2 meters in the center to 1.5 meters at the periphery.
The stubs extended from the lower grid plate to the bottom of theresolidified mass.440•

Upper core support assembly. The upper core support assembly
included vertical baffle plates that formed the peripheral boundary of
the core, horizontal core former plates to which the baffle plates werebolted, the core barrel to which the core formers were attached, and thethermal shield. The assembly retained about 4,000 kilograms of loose
debris and resolidified material. Loose debris ranging from 1.5 meters to149

a few millimeters deep settled behind the baffle plate circumference.
A resolidified crust ranging from 0.5 to 4 centimeters thick was attachedto the bottom of three core former plates.441•

Lower core support assembly. The lower core support assembly
included five layers (top to bottom): the lower grid rib section that
supported fuel assemblies; lower grid flow distributor plate; lower grid
forging (attach point for the complete lower core support assembly tothe reactor vessel wall); incore guide support plate; and elliptical flow
distributor head. The layers were supported by the outer circumferential
shell. The assembly structures retained about 6,000 kilograms ofresolidified material around the circumference of the structures.442•

Lower head region. The reactor vessel’s lower head region contained
about 12,000 kilograms of loose core debris and 7,000 kilograms of
agglomerated core debris. The debris on the lower head was 4 meters indiameter and 0.75 to 1 meter deep. The surface debris had particle sizes
which varied from those of large agglomerated debris (up to 0.20 meter)
to those of granular particles. Resolidified material existed on thereactor vessel’s lower head, underneath the loose debris. This material
was approximately 0.5 meters deep in the center and 1.7 meters indiameter.443

Reactor building polar crane being inspected for use in defuelingpreparations.

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•

Fuel debris distribution outside the reactor vessel. A total of about
228 kilograms of fuel debris were transported through the reactor
coolant system. About 95 percent of the debris settled in the “B” loop
where a reactor coolant pump was initially turned on at 174 minutes into
the accident. Several minutes of pump operation quenched the hot
oxidized reactor fuel and caused it to shatter. Other pumps wereoperated sequentially until the last pump was secured in response to an
emergency procedure on April 27, 1979, when the final pressurizer level
instrument failed. Most of the debris inside the reactor coolant systemsettled on the upper tube sheet of the “B” steam generator (about
125 kilograms). Smaller amounts settled in the decay heat system
suction piping or “drop line” (30 kilograms), reactor coolant pumps (atotal of 20 kilograms), and pressurizer (12 kilograms). Approximately
10 kilograms total were deposited in the reactor building on thebasement floor and sump (5 kilograms) and makeup and purification
system letdown coolers (4 kilograms). A total of 23 kilograms entered
the auxiliary building, mainly deposited in the three reactor coolant
bleed tanks (a total of 15 kilograms) and in the makeup and purification
system (6 kilograms).444

Preparation of the Reactor Vessel. Before core debris could be removed
from the reactor vessel, preparations were required to allow direct access tothe damaged reactor core. While many of these activities were routine
during normal refueling operations, the effects of the severe accident onreactor vessel components, the reactor building’s environment, andoccupational radiation exposures presented complex challenges.
Preparations included consideration of numerous potential safety issues; forexample, occupational exposures; decay heat removal; criticality control;boron dilution; radioactivity releases; hydrogen evolution inside the reactor
coolant system; pyrophoricity (spontaneous ignition in air) of zirconium
fines in the reactor vessel; heavy load drops; polar crane failure; reactor
vessel draining; and fire protection. Technical considerations included
potential distortion; warping or physical dislocation of the reactor vessel’shead or upper plenum; reactor coolant cleanup; reactor coolant systemdepressurization; and lowering of reactor coolant level.445, 446, 447, 448 In
addition, cleanup of the reactor building, which included atmospheric gases,
basement water, and surface contamination, was an important prerequisite toensure lower radiation exposures. Key milestones for reactor vessel
preparation activities are summarized as follows:•

Leadscrew uncoupling. The last of the lead screws attached to the
61 control rod and 8 axial power shaping rod spider assemblies were
uncoupled in November 1982. The 22-foot long lead screw was a151

component of each control rod drive mechanism. Verification was
performed in December 1982 to ensure that no partial fuel or control rod
assemblies were attached to the lead screw.449•

Polar crane refurbishment and testing. The reactor building’s polar
crane was needed to remove the reactor vessel head and upper plenum.
The polar crane was refurbished because of the harsh environmental
conditions inside the reactor building during the accident (such as
steam, burning hydrogen, and containment spray) and following theaccident (such as high humidity and high radiation). The NRC safety
review of the use of the polar crane was delayed several months toinvestigate allegations made by a contractor about the safety of the polar
crane. Six months later, a NRC investigation cited deficiencies in theadministrative and procedural aspects of the polar crane repair. The
NRC concluded that the specific deficiencies cited did not result in asignificant increase in risk to the public health and safety. On November
18, 1983, the staff approved the licensee's safety evaluation for the
refurbishment and use of the reactor building polar crane. Load-testing

Reactor vessel head and service structure resting on its stand surrounded byshielding. Shown are white tubes filled with sand and yellow lead blankets.

152

of the polar crane was successfully completed on February 29, 1984,
when a test assembly weighing 214 tons was lifted and moved alongpredetermined test paths.450 Various configurations and uses of the polar
crane were approved by the NRC throughout the defueling and cleanup
efforts.
•

Canal seal plate installation. A modified canal seal plate was installed
on April 13, 1984, to ensure a long-term positive seal between thereactor vessel and refueling canal. The seal allowed the contingency forflooding the refueling canal during head removal.451•

Reactor coolant system draindown. The reactor coolant system waspartially drained and depressurized to allow for the reactor vessel headlift on June 24, 1984. The water level was then raised to cover thecontrol rod guide tubes after head lift and installation of the internals
indexing fixture (discussed below).452•

Removal of reactor vessel studs. Sixty reactor vessel studs werecleaned, detensioned, removed from the reactor vessel, and stored inracks in the reactor building. This activity was completed on
July 6, 1984. The first of the two detensioning passes had been
performed months earlier to identify and correct stuck studs.453•

Leadscrew parking. The last leadscrew to the control rod drive
mechanism was “parked” on July 21, 1984. Parking refers to raising aleadscrew to its uppermost position using a heavy-duty lifting tool. The
leadscrew was then secured in place with a parking tool so that it would
not extend below the head flange level and interfere with lateral
movement of the reactor vessel head during the lifting operation.454•

Reactor vessel head lift. The reactor vessel head and the attachedservice structure were removed and placed in shielded storage on
July 25, 1984. The reactor building’s polar crane was used to lift thehead and place it on its storage stand located in the reactor building. The
actual lifting operation took over five hours. Sand-filled columns wereplaced around the head stand for shielding. Lead blankets were placedaround the service structure. Before the lift operation, training exercises
were performed using a set of mockups.455•

Internals indexing fixture installation. The modified cylindricalinternals indexing fixture (IIF) with a watertight gasket system wasplaced on the open reactor vessel flange by the polar crane on
July 26, 1984. The IIF was normally used to guide the upper plenum
153

and core support assembly during installation and removal. The IIF was
modified to allow the installation of a shielded work platform above thereactor vessel to support future operations. Before the installation of thework platform, the IIF was filled with five feet of borated water toprovide radiation shielding over the exposed upper plenum in thereactor vessel.456•

Upper plenum assembly transfer. The last major structural obstacle todefueling was removed on May 15, 1985, when the upper plenum
assembly was lifted from its jacked position in the reactor vessel, raised
through the water-filled internals indexing fixture, and transferred to its
storage stand in the deep end of the fuel transfer canal. Before plenumtransfer, a six-foot-high dam was constructed, allowing the deep end of
the canal to be flooded to a level sufficient to provide adequate
shielding for the stored plenum. Major preparatory activities included
in-vessel inspection of the plenum, separation of damaged fuelassemblies hanging from the bottom of the plenum, and the initial liftingof the plenum by a system of four hydraulic jacks. The precautionary
jacking ensured that there was no binding during plenum lift by thereactor building’s polar crane. Before the lift operation, training
exercises were performed using a set of mockups.457

Upper plenum being transferred from the reactor vessel (shown) to thepartially flooded deep end of the fuel transfer canal (not shown).

154

•

Pyrophoric issue. The issue of pyrophoricity was addressed in the
Programmatic Environment Impact Statement (PEIS) for defuelingactivities and in the safety evaluation of the underhead characterization
study. This issue was based on the concern that appreciable amounts ofzirconium hydrides might have been formed during the accident when
hot dry hydrogen reacted with zirconium surface. At high temperatures,
zirconium hydrides react with steam to form zirconium oxide andhydrogen gas. In a finely divided form, zirconium hydrides can bepyrophoric out of water. The PEIS concluded that it was unlikely azirconium hydride ignition would occur given that defueling operations
were planned underwater. However, the underhead characterization
study required lowering the reactor coolant level in the reactor vessel
and uncovering some of the vessel internals, such as leadscrew support
tubes, control rod guide assembly tubes, and upper plenum cover plate.
Given that these internal surfaces and fuel debris samples would be
exposed to air, the issue of pyrophoricity was addressed by the licensee
and NRC in the characterization study’s safety evaluation.458

Internals indexing fixture with its temporary cover mounted on
the reactor vessel flange. This fixture would evenually be matedto the defueling work platform.

155

NRC determined in its safety evaluation that (1) the presence of steam
as an oxidizing agent and the temperature condition during the accident
would not likely produce significant quantities of zirconium hydride in apyrophoric condition; (2) the primary system flow dynamics during theaccident would not likely transport large quantities of pyrophoric
material, if formed, to the top of the plenum; and (3) any pyrophoric
materials in finely divided form would be dispersed and mixed withinert materials of core debris which would prevent the development ofpyrophoric conditions. The results of the underhead characterization
study supported the conclusions in the NRC safety evaluation that there
was little potential for a pyrophoric reaction with the plenum cover
exposed to air. Results were based on visual observations of the reactor
vessel underhead conditions and laboratory analyses of the chemicaland pyrophoric properties of samples obtained from components within
the reactor vessel and from solids filtered from the reactor coolant.459
Defueling Systems and Equipment. Unique systems and equipment were
designed and installed to remove damaged fuel and structural debris fromthe reactor vessel. In the early defueling phase, tools were designed for“pick-and-place” in which debris was picked up and placed into fuelcontainers (baskets) or specially engineered defueling canisters. Somelong-handled tools had various hydraulically actuated fittings to tackle thelarger pieces and smaller bits of debris. The core bore machine was placed

Defueling operations from the rotating defueling work platform with a worker
holding a long-handled defueling tool down through the slit into the reactor
vessel. Two jib cranes at each end of the slit were used to lower larger tools.

156

back into service to bore holes in the resolidified mass to help break apartthe previously molten reactor core. Combinations of tools were used toassist defueling the lower reactor vessel region, such as the core bore
machine and plasma arc torch. Three types of defueling canisters andassociated support equipment were specially designed to remove the fueldebris from the reactor vessel and package it for transportation. Water
cleanup systems were installed to ensure water clarity in the reactor vessel
for removing debris and in the “A” spent fuel pool for processing defuelingcanisters.
The defueling systems were designed before the extent of core damage andradiological conditions were fully understood. The design of the canisters
was decided upon early in the cleanup and dictated the design of thedefueling platform and the shipping cask. The design of the canisters was
based on the removal of intact fuel assemblies. Ultimately, very few if any
intact fuel assemblies were removed. The narrow inside dimensions of thecylindrical canister design necessitated substantial cutting of distorted fueldebris in order to load through the opening of the canister. A radiation
analysis program460 was undertaken to identify and quantify possible
radiation sources and to design defueling equipment to achieve dose-rategoals in the defueling area. Specific safety issues were addressed in NRC
evaluations of defueling equipment, systems, and operations. Safety issues
that were considered in the evaluations were similar to those addressed in
the preparation of the reactor vessel for defueling (as listed above). Lessons
learned from experience using defueling equipment were documented in the
Nuclear Technology journal paper “Fuel Removal Equipment for Three
Mile Island Unit 2” (Vol. 87, No. 3; not available on the DVDs).461
Examples of systems and equipment used for defueling the reactor vessel
are summarized below (see GEND-INF-073):•

Defueling work platform. A shielded work platform and support
structure was installed over the water-filled internals indexing fixture(IIF) in August 1985. The support structure circumscribed the IIF andextended upward from the refueling canal floor to the work platform.
The shielded work platform was located 9 feet above the reactor vessel
flange. The work platform included a defueling work platform with arotatable surface 17 feet in diameter with a 6 inch thick steel plateshield. The platform provided a shielded work area for defueling
operations; a support for manual, hydraulic, and mechanical defuelingtools; and a method for removing defueling canisters. An adjustable slot
and hand rail spanning the diameter provided access to the reactor core.
Various lines for water treatment and air ventilation to control
off-gassing were routed into and out of the reactor vessel through the157

platform support structure. A decontamination spray system flushedradioactive debris from the surface of the canisters, long-handled tools,
and other equipment as each item was removed from the reactor vessel.
Two jib cranes were mounted on the defueling work platform to aid theoperators in manipulating the long-handled tools in the long-handled
tool slot.462
A full-scale mockup duplicated the rotating work platform, the tool
racks, and the overhead crane required to handle the 40-foot-long tools.
The defueling crew developed their sense of feel for the tools, thebalance of the tools, and the effort required to pull or shear fuel whileworking on the mockup.463 Operator training and tool testing on themockup contributed to effective defueling operations.464•

Off-gas system. The off-gas system created airflow through thedefueling work platform, into the internals indexing fixture enclosure,and out to the reactor building’s atmosphere through a filtration unit.
The airflow through the platform prevented radioactive gases thatcollected under the platform from reaching personnel working on theplatform.465•

Defueling water cleanup system (DWCS). The DWCS was used toprocess water in the reactor vessel, in the deep end of the fuel transfer
canal, and in the “A” spent fuel pool. The system was designed toCore bore drilling machine, mounted on a platform over the defueling work
platform, was used to drill holes and break apart the resolidified mass.

158

reduce radioactivity and improve water clarity for defueling operations.
The DWCS was composed of two systems: the reactor vessel cleanup
system and the fuel transfer canal/spent fuel pool cleanup system. The
cleanup of the reactor vessel water proved to be more difficult thanoriginally anticipated. (See further discussion in the prior section on
Waste Management.)•

Long-handled defueling tools. Numerous manual and hydraulicallypowered long-handled tools were used to perform a variety of functions,such as pulling, grappling, cutting, scooping, and breaking up the coredebris. These “pick-and-place” tools were used to load debris into fuelcanisters positioned underwater in the reactor vessel. The “workingend” of the long-handled tool was one of many kinds of end effector
attachments. Powered end effectors were hydraulically operated. The
operators relied on the video viewing system to monitor their work as
they manipulated the tools through the tool slot in the defueling work
platform. Most tools were supported by an overhead service crane toprovide vertical and lateral motion, although some lightweight tools
could be handheld for picking smaller debris. End effectors could be
detached from their handles and stored in a tool rack located under thedefueling work platform, or in a rack outside the reactor vessel.466 A few
of the numerous end effector attachments are summarized below. See
GEND-INF-073 for a catalog of early defueling tools.
oHydraulic cutoff saw. An hydraulically driven hacksaw was used
for sizing structural material or other debris (that is, reducing it topieces of smaller size) to facilitate the loading of debris into fuelcanisters or debris baskets.467, 468

oHydraulic impact chisel. The impact chisel was a hydraulicallyactuated miniature jackhammer-type tool for use in breaking aparthard materials. The chisel’s angle of attack could be varied remotely
to achieve any position from vertical to horizontal. Different bit
types were provided for the various anticipated chiseling operations.
469, 470oHydraulic shredder. The hydraulic shredder was used to reduce fuelpins (with and without fuel pellets) and spacer grids to sizes which
would facilitate their placement into fuel canisters or debris baskets.
The shredder was suspended below the defueling work platformusing the support structure attached to the platform. The fuel debris
was loaded into an inlet hopper and the shredder output wasdischarged to a transfer container which was then emptied into a
159

fuel canister or debris basket; discharge from the shredder could
also be directed to the debris bed. Non-shreddable material lodged
in the shredder was retrieved through manual manipulations using
other long-handled tools.471, 472oRemote manipulator. Refer to the discussion on Remote Robotic
Equipment in the section on Decontamination.

oWater-jet cutting system. The water-jet cutting system performed
remote underwater cutting of hard materials with and without theuse of abrasives. The equipment was designed to be positioned at
the cutting location with the remote manipulator.473•

Core bore machine. The core bore machine, which was used to extract
core stratification samples in 1985, was used during defueling to breakup the solidified monolith (melted-together core materials) in the coreregion.474, 475 The machine was used again to drill through and cut
portions of the lower core support assembly.476, 477, 478 The drill was
reinstalled on its elevated mount over the defueling work platform.479•

Automatic cutting equipment system (ACES) or plasma arc torch.
Refer to the discussion on Remote Robotic Equipment in the section on
Decontamination.

•

Defueling canisters. The defueling canisters were designed to accept
and confine core debris ranging in size from particles (known as fines)of about 0.5 microns in diameter up to partial-length fuel assemblies offull cross-section. The canisters were intended to provide confinementfor offsite transport using a shipping cask and long-term storage of coredebris. Three types of defueling canisters were designed and fabricated:a fuel canister, knockout canister, and filter canister. Each canister
required fixed neutron-absorber material for criticality control; catalytic
recombiners to control the concentration of combustible gas mixturesgenerated from radiolytic decomposition of water; and appropriate
process connections for fillings, closing, dewatering, inerting, andmonitoring. All three canisters were 150 inches long, 14 inches indiameter, and ¼ inch thick.480, 481 As a result of a request from DOE, the
NRC performed inspections of defueling canisters during the fabrication
process, which included inspectors observing welding, non-destructive
examinations, and fitting of components.482
o160

The fuel canister was designed as a receptacle for large pieces ofcore material, which were picked up and placed either directly intothe canisters, or into other containers which would then be inserted
into the canister.483 One lesson learned was reported on the
difficulty of placing distorted debris in the fuel canister due to thenarrow inside dimensions of the canister.484•

oThe “knockout” canister was designed for use in the fuel debris
vacuum system to separate debris particles ranging from about
140 microns up to full pellet size or larger. The process inlet line
entered the top of the canister and bent to direct the flow
tangentially along the inner circumference of the shell, creating aswirling action that caused the entrained debris to settle out in thecanister vessel. The water then exited through an 850-micron screen
to a process connection in the top of the canister. The knockout
canister had a capacity of about 1,800 pounds of fuel debris.485

oThe filter canister was designed for use in the fuel debris vacuum
system, the defueling water cleanup system, and the canister
dewatering system. The filter captured debris fines larger than
0.5 microns on sintered metal filters.486

Debris buckets. Debris buckets were used to configure the debris before
insertion into the fuel canister, to maximize the packing density in acanister, and to eliminate many 10-foot vertical trips to load small
pieces in a canister. Two types of disposal buckets, the top loading

Three types of defueling canisters for containing large fuel debris (fuel),
vacuuming fractured fuel pellets (knockout), and filtering fuel fines (filter).

161debris bucket and side loading debris bucket, were designed to fit into
the fuel canisters. A reusable debris bucket fitted to the inside of a fuelcanister also had a remotely operated trap door on the bottom for
unloading into the canister. Accessories included a funnel, handling
tool, and stand.487•

In-vessel vacuum system. The in-vessel vacuum system was designed
to remove small loose debris as large as the approximate size of a fuelpellet maximally. The vacuum system was supported from the undersideof the defueling work platform and was controlled from the console on
the south auxiliary work platform. The vacuum pickup nozzle wasconnected to a canister by a flexible hose and was manipulated using along-handled tool. Debris was picked up and passed first through a

The canister handling bridge was used to stage
defueling canisters in a submerged storage rack in
the “A” spent fuel pool. Shown is the canistertransfer shield and defueling canisters in the pool.

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knockout canister; any remaining debris larger than 0.5 microns wascollected in a filter canister.488 Vacuum defueling began on
December 31, 1985, in which about 300 pounds of debris was loaded
into the knockout canister.489•

Airlift vacuum system. The airlift vacuum system was used to lift debris
as large as 5 centimeters from the core region rubble bed and the lower
head region of the reactor vessel. The system included an air
compressor, airlift pipe, and a fuel canister. Compressed air was
injected at the bottom or suction end of the pipe to draw water into thepipe along with entrained fuel debris. Debris was deposited into a fuelcanister located at the top of the airlift piping. The airlift increased
packing efficiency in fuel canisters already containing oddly shaped
pieces of debris, such as partial fuel assemblies and end fittings.490•

Pressurizer defueling system. The pressurizer defueling system wasdesigned to remove fine fuel debris from the pressurizer. The system
included a submersible vacuum pump, a knockout canister, a filter
canister, and an agitation nozzle. The defueling water cleanup systeminjected water through an agitation nozzle to suspend debris in thewater-filled pressurizer. Water was pumped from the pressurizer,
through the canisters, and into the reactor vessel. Larger debris was
pickup with the use of a robotic submarine.491•

Once-through steam generator (OTSG) defueling system. The OTSG
defueling system was designed to vacuum loose fuel debris from theupper tube sheets. The dry vacuum system included a vacuum head on ahandling pole, vacuum canister, high-efficiency particulate air filter, avacuum breaker, and vacuum pump. Larger debris was manually
removed using long-handled tongs.492•

Canister positioning system. The canister positioning system was arotating carousel installed in the reactor vessel that could hold up to five
fuel and knockout canisters, including knockout canisters for use withthe vacuum system. The height of canisters in the canister positioning
system could be adjusted to three discrete elevations to allow them to beplaced more closely to the debris bed as the bed got lower.493•

Canister handling bridges. Existing fuel handling bridges located in thereactor building and fuel handling building were modified with canister
handling trolleys and canister transfer shields. The reactor building’s
canister handling bridge was used to lift loaded canisters into a canistertransfer shield, and to move the shielded canister from the reactor vessel
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through the air to the flooded deep end of the fuel transfer canal. The
fuel handling building’s canister handling bridge lifted the canister fromthe fuel transfer system to a submerged storage rack or to the canister
dewatering station in the “A” spent fuel pool.494•

Canister storage racks. The defueling canister storage racks provided
storage for loaded defueling canisters. Storage for a total of
263 canisters was available in the racks located within the deep end of
the fuel transfer canal inside the reactor building (11 canisters) and in
the “A” spent fuel pool inside the fuel handling building(252 canisters).495, 496•

Fuel transfer system. The existing fuel transfer system was modified totransfer defueling canisters from inside the reactor building to theadjoining fuel handling building. Canisters were handled in a way
similar to normal fuel assemblies during refueling operations. A canister
was lowered into a modified “upender” in the flooded deep end of thefuel transfer canal, turned to a horizontal position, and moved throughone of the fuel transfer tubes into the “A” spent fuel pool. A second“upender” at the other end raised the canister to a vertical position.497•

Canister dewatering systems. The canister dewatering systems wereused to purge water from submerged defueling canisters in order tosufficiently expose the recombiner catalysts and prevent the buildup of
combustible gases in the canisters. Inert-gas injection systems displaced
water from the canisters. Dewatering stations were installed in thereactor vessel and the “A” spent fuel pool. Typically, a loaded canister
was partially dewatered in the reactor vessel using bottled inert gas
located at the defueling work platform. Water discharged from thecanister during in-vessel dewatering remained in the reactor vessel. Any
inert gas released by dewatering process was vented through the off-gas
system. The fuel handling building’s dewatering station was also
located underwater to shield workers from radiation.498•

Fuel transfer cask. The fuel transfer cask used to transfer a defuelingcanister from the “A” spent fuel pool to the shipping cask was bothcylindrical and bottom-loaded; the transfer cask weighed about
40,600 pounds when loaded with a defueling canister. The cask wassuspended from the fuel handling building’s overhead crane, whose
integral grapple and hoisting mechanism engaged a defueling canisterand lifted the canister out of the water through a shielded platform into
the shielded transfer cask. The bottom door of the cask was closed andthe entire unit was moved with the overhead crane to the Model 125-B

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shipping cask located in the truck bay where it mated with the shippingcask’s loading tower. 499•

Model 125-B shipping cask. The Model 125-B shipping cask was
designed specifically to transport the loaded defueling canisters. The rail(train) cask could carry up to seven defueling canisters. (See further
discussion in the prior section on Waste Management.)•Core flood tank modifications. The top of the “A” core flood tank wasremoved to allow the storage of cut lower core support assemblycomponent pieces. The piping from the “A” core flood tank to thereactor vessel was cut and capped to prevent the possibility of fueltransport.500•

Containment air control envelope (CACE). The CACE was a newbuilding external to the reactor building that enclosed the equipment
hatch and backup personnel airlock. Access to the CACE from theoutside was provided by a personnel door and a 27-foot-wide roll-up
truck door. The building provided an area where cleanup equipment andmaterials could be assembled and staged prior to transfer into thereactor building, thus reducing worker stay times in radiation areas,
resulting in occupational exposure savings. The CACE also functioned
as a staging area for contaminated material removed from the reactor
building, but was not designed to be a storage area for radioactive
wastes. When both equipment hatch personnel airlock doors wereopened, the reactor building purge system induced air flow from theoutside through the CACE and the CACE aided in controlling andconfining potential airborne releases from the reactor building. The
CACE included a HVAC system, consisting of two filtered exhausttrains and associated radioactive effluent release monitor, which
operated to reduce airborne particulate contamination in the building
and to protect the building from overpressurization. The HVAC system
was operated to maintain net airflow into the CACE when isolated fromthe reactor building. Normally, the CACE was maintained at a slightly
negative pressure, relative to the outside, to limit exfiltration from thebuilding, except when the roll up door was open for staging equipment
and materials into or out of the CACE in support of cleanup activities.
Periodic monitoring of the CACE atmosphere was performed to ensurethat potential airborne releases were within regulatory limits.501 The
CACE was intended to support recovery activities through defueling
only and was not designed to satisfy the criteria for a permanent TMI-2
facility.502 The CACE was available for use in June 1986.503
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Reactor Vessel Defueling. Removal of damaged fuel and structural debris
from the reactor vessel started on November 12, 1985, six and one-halfyears after the accident. Numerous manual and hydraulically powered
long-handled tools were used to perform a variety of functions, such as
pulling, grappling, cutting, scooping and breaking up the core debris.
Eventually, more powerful tools were used to disassemble or cut apartreactor vessel components and break up resolidified core material, such as
the core bore machine, plasma arc torch, and water-jet cutting system. After
breaking up and sectioning of oversized debris, long-handled tools wereused to manually load debris into defueling canisters positioned under waterin the reactor vessel. The larger pieces of vessel internal components, suchas lower core support assembly sections, were lifted out of the vessel by
crane and stored in the modified ”A” core flood tank. Smaller pieces or“fines” were vacuumed into specially designed knockout canisters and filter
canisters. Other defueling activities included transferring the loaded
defueling canisters from the reactor building to the fuel handling building,dewatering the filled canisters, and placing canisters into the canister storage
racks located in the “A” spent fuel pool.504

Problems plagued the cleanup systems early because of microorganism
growth in the reactor coolant. The resulting loss of visibility halted alldefueling operations for a few months. Shown is a view down into thereactor vessel annulus region between the core support assembly andthe vessel wall. (A light hanging at the lower right.)

166

Removal of fuel and structural material from the reactor vessel wascompleted in stages. The plan for each stage incorporated experience gained
from previous stages and activities, including in-vessel examinations. The
NRC site office provided reviews and approvals of the licensee’s technicalevaluation reports and procedures. Defueling included the following stages:•

Preliminary defueling. Preliminary defueling involved therearrangement of core debris material within the reactor vessel to allow
complete installation and rotation of the canister positioning system andto provide access for defueling tools. The preliminary activities also
included identification and positioning of core debris samples in thereactor vessel. General movement of the core debris included loading ofdebris baskets (or containers), but not loading defueling canisters.505, 506
Preliminary defueling operations started on October 31, 1985.507•

Early defueling. Early defueling involved depositing loose core debris
into defueling canisters and removing loaded canisters from the reactor
vessel. The loaded canisters were placed in canister storage racks
located in the fuel transfer canal in the reactor building and the“A” spent fuel pool in the fuel handling building. The debris consistedof partial fuel assemblies, fuel rods, end fittings, structural materials,
and loose granular fuel and structural fines. The debris was packaged
into fuel canisters by defueling operators using long-handled tools andFirst loading of debris into a fuel canister started on January 12, 1986.
Shown is an upper fitting of a control rod cluster being dropped into a167

the in-vessel vacuum system. During canister loading, defuelingpersonnel were supervised by a specially trained and NRC-licensed fuelhandling senior reactor operator. Physical and administrative controls
were implemented to prevent the inadvertent lifting of core debris out of
the reactor vessel.508, 509 On November 12, 1985, the licensee received
permission from the NRC onsite office to start loading the fuel canisters
with fuel debris. Using the vise-grip long-handled tool, the defuelingworkers placed the first piece of debris, a broken piece of fuel rod, into
the fuel canister.510 On January 12, 1986, the first three loaded fuelcanisters of core debris were transferred from the reactor vessel to thesubmerged canister storage rack in the “A” spent fuel pool in the fuelhandling building.511•

Bulk defueling of the core region. Bulk defueling involved
reinstallation of the core boring machine (which had previously beenused to obtain core samples) over the reactor vessel to perform
additional drilling operations into the resolidified mass in order tofacilitate defueling. The drilling operations were permitted on themonolith in the core region above the lower core support assembly.512,
513, 514, 515, 516
A total of 409 closely spaced holes in the resolidified
material were started on October 20, 1986, and completed on
November 14. After completion of the drilling, the drill rig wasremoved and the defueling work platform was reconfigured to support
manual debris-removal activities.517, 518, 519 Equipment used during bulk
defueling included “pick-and-place” long-handled tools, air-operated

Bulk defueling system arrangement.

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chisels, and specially designed fuel-assembly pulling and grasping tools.
Removal of the stub end fuel assemblies from the reactor vessel using
custom tools started in March 1987 and was completed in
September 1987. By September 1988, the entire original core region had
been defueled.520•

Removal of the lower core support assembly. This activity required
three years of planning, inspection, research, and development to accessthe debris on the reactor vessel’s lower head. This massive assembly
included the following layers (from top to bottom): lower grid rib
section, lower grid flow distributor plate, lower grid forging, incore
guide support plate, and elliptical flow distributor head. Defuelingpreparations required the use of the core bore machine, plasma arc
cutting torch (also called the automated cutting equipment system),cavitating water jet, and other equipment to dismantle the multi-layered
lower core support assembly. The core bore machine was used to cut
assembly support posts and incore instrumentation guide tubes. The
plasma torch was used to cut straight vertical and horizontal segments inthe plates.521 Pieces of the assembly plates were stored inside themodified “A” core flood tank. Sections that did not contain incore guide

By September 1988, the entire original core region had been defueled.
Shown is a view down into the water-filled reactor vessel of the lower
grid rib section of the lower core support assembly. Vertical baffle plates
formed the peripheral boundary of the core.

169

tubes were bagged and stored inside the “A” steam generator’s D-ring.
All pieces will remain inside the reactor building until the futuredecommissioning of TMI-2. Equipment used to remove loose andresolidified debris that remained on the sectioned remnants included ahigh-volume low-pressure water flush and a low-volume high-pressure
cavitating water-jet flush. Displaced debris was collected on the reactor
vessel’s lower head for eventual removal. Work was started in
January 1988 and completed by March 1989.522, 523, 524•

Defueling of the upper core support assembly. This defueling activity
involved the removal of core debris from between the baffle plates andcore former plates. Defueling preparations required the use of theplasma arc torch to cut the baffle plates in order to enable access to thecore debris on the core former plates. 525, 526 The baffle plates were cut
into eight pieces. An untorquing tool and a drilling tool were used toFuel debris at the bottom of the reactor vessel. The lower core support
assembly structures originally retained about 6,000 kilograms of resolidified
material around the circumference of the structures. The reactor vessel’s
lower head region below the assembly contained about 12,000 kilograms ofloose core debris and 7,000 kilograms of agglomerated core debris.

170

remove 864 bolts that held the baffle plates to the core barrel.
Equipment used to remove loose debris from within the assembly
included hydraulically powered counter-rotating brushes mounted on apivoting deployment end effector and the in-vessel vacuum system.
Resolidified debris was removed using mechanical methods and thecavitating water jet. Defueling of the assembly was completed in
October 1989.527•

Defueling of the reactor vessel’s lower head. This defueling activity
involved removal of core debris from the reactor vessel’s lower head.
Defueling preparations required the removal of the gusseted incore
guide tubes and sections of the elliptical flow distributor head.
Equipment used to dismantle the lower core support assembly was also
used in this activity.528, 529, 530 Equipment used to remove loose debris on
the lower head included an airlift, long-handled tools, and the in-vessel
vacuum system. The cavitating water jet and an impact hammer with achisel point were used to break up the resolidified debris attached to thelower head. The removal of about 30 tons of core debris from the lower
head was completed in November 1989.531•

Final cleanup of the reactor vessel. Defueling of the reactor vessel wascompleted in December 1989. A video inspection inside the reactor
vessel, in conjunction with sample analysis, was completed in
January 1990 to determine the quantity of residual core debris in thevessel.532 Final reactor vessel re-flushing and re-vacuuming for loose,
dust-like debris was completed in March 1990. The remaining quantityof fuel in the reactor vessel was less than 900 kilograms (or less than
one percent of the original inventory). The residual fuel consisted
primarily of finely divided sediment with small particle sizes ininaccessible holes, crevices, corners, and surface films, as well as
resolidified material either tightly adherent to the reactor vessel’s
components or inaccessible for defueling. The final defueling reportconcluded that the residual fuel was not readily transportable between
locations and, therefore, that criticality would be precluded underpostulated worst-case conditions.533 (See further discussion in thesection on After Defueling.)

Defueling Outside the Reactor Vessel. The licensee estimated that about
260 pounds of fuel debris was transported outside the reactor vessel during
the course of the accident. Additionally, about the same amount was
relocated outside the reactor vessel because of cleanup and defueling
operations, mostly in the reactor coolant system. The remaining fuel in
hard-to-reach locations will be removed during the future decommissioning
171

of TMI-2. An extensive program of fuel measurements was implemented,
which included direct measurement by instrumentation, visual inspection,and sample collection and analysis. The results from this program werereviewed by the NRC to ensure that no amount could achieve criticality
under worst-case conditions. (See further discussion in the section on After
Defueling.) Defueling outside the reactor vessel included the following
activities:•

Reactor coolant system (RCS). The two methods of fuel transport tolocations outside the vessel were from sequential operation of the four
reactor coolant pumps and the “burping” phenomenon during natural
circulation. About 228 kilograms of fuel debris was transported fromthe reactor vessel into the RCS during the accident. An estimated170 kilograms was added to the RCS during defueling operations. The
reactor coolant system defueling activities removed more than90 percent of the debris in the pressurizer, decay heat system suction
piping (also known as the “drop line”), and hot-leg piping; andapproximately 70 percent of the debris on the steam generator’s upper
tube sheets.534 The following techniques were used to defuel reactor
coolant system components:oHot-leg piping. The two hot-legs were initially defueled using a
combination scraper-and-vacuuming tool. Additional residual coredebris in the “B” hot-leg was scraped, flushed, and vacuumed into
defueling canisters as part of reactor vessel defueling.535 Defuelingof both hot-legs was completed on August 10, 1988.536

oPressurizer. Initially, the pressurizer was defueled using asubmersible pump, knockout canister, filter canister, and agitation
nozzle. The second phase of pressurizer defueling used a remotely
operated submersible vehicle equipped with an articulating claw
and a scoop to remove larger pieces of debris located on the bottom
head of the pressurizer.537, 538, 539 Defueling of the pressurizer using
the vacuum system started in November 1987 and was completedusing the submersible vehicle on June 14, 1988.540

oPressurizer spray line. The pressurizer spray line defueling system
was used to flush water from the defueling water cleanup system
into the pressurizer and reactor coolant system “2A” cold-leg.541

oDecay heat drop line. The in-vessel vacuum system was used todefuel the decay heat system suction piping drop line. A tool wasdeveloped to guide the vacuum hose into the vertical portion of the172drop line from the “B” hot-leg. A plumber’s drain-cleaning machine
was used to break apart a hard, compacted region of debris below
the loose debris so vacuuming could continue. The material wasairlifted into the “B” hot-leg and was removed as part of the hot-leg
defueling.542
o•Steam generators’ upper tube sheets. Pick-and-place andvacuuming techniques were principally used to defuel the upper
tube sheets in both steam generators. A vacuum system removed thesmaller debris.543 The defueling of the “A” and “B” steam generator
upper tube sheets was completed in September 1987 and October
1987, respectively.544

Reactor building. A small quantity of fuel was released to the reactor
building as a result of leakage through the pressurizer pilot-operated
relief valve during the course of the accident. The licensee estimatedthat the scarification and desludging activities in the reactor building’s
basement had removed approximately 4,900 kilograms of sediment
which contained approximately 4 kilograms of fuel. Sediment included
mostly river water sediment from a leaking air cooler inside the reactor
building, concrete dust, and dirt. A robotic desludging system removed
about 40 percent of the basement floor area that was accessible. The
removal efficiency of desludging was greater than 90 percent. About
75 kilograms of fuel remain outside the reactor vessel in the reactor
building, with the largest quantity attached to dismantled reactor vessel

Fuel debris on the upper tube sheet of “B” steam generator.

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components that were relocated for storage inside the “A” steam
generator’s D-ring shield structure and the modified “A” core floodtank.545•

Auxiliary and fuel handling building. A small quantity of fuel debris
was transported into the auxiliary building, principally through thereactor coolant bleed tanks and the makeup and purification system
during the accident. Additionally, a smaller amount of fuel might havebeen relocated into the auxiliary building as part of the post-accident
water processing, cleanup, and defueling activities. About 3 kilogramsof fuel was removed from the makeup and purification systemdemineralizers and 370 grams of fuel was removed from the blockorifice assembly positioned upstream of the demineralizers. The total
quantity of fuel material remaining in the auxiliary building wasestimated to be less than 17 kilograms. Removal of resins from the
makeup and purification system demineralizers started in October 1987
and completed in September 1988 after limited success.

Remotely operated submersible vehicle equipped with an articulating claw anda scoop was used to remove larger pieces of debris that were located on thebottom head of the pressurizer.

174

175

Defueling arrangement: workers on the defueling platform loaded fuel canisters; a fuel canister bridge transferred acanister from the reactor vessel to the fuel transfer tube and mechanism inside the reactor building; another fuel canister
bridge in the fuel handling building loaded the canister into the fuel storage rack in the spent fuel pool; the fuel transfer
cask and building crane transferred the processed canister from the pool to the shipping cask loading station; and theModel 125-B shipping cask was lowered from its upright position onto the rail car and prepared for shipment.

Shipments of Core Debris. The first shipment of core debris to INEL
began in July 1986. The core debris transportation campaign consisted of
22 rail shipments by dedicated train which resulted in the transport of a totalof 49 casks loaded with 342 defueling canisters. The final fuel shipmentfrom Three Mile Island to the INEL started on April 15, 1990.546
The canister handling and preparation for shipment program included allactivities necessary to prepare and transfer a loaded defueling canister fromits storage rack in the “A” spent fuel pool to the shipping cask; to insert thecanister into the shipping cask; and to verify that the shipping cask was
prepared for transport in accordance with its NRC-issued certificate ofcompliance. The preparation of canisters for shipment required several
activities using custom-designed equipment. Canister preparations included
dewatering and purging the defueling canister with an inert cover gas,verification of final canister weights, verification that the catalytic
recombiners installed inside thecanister were functioning, andverification that the canister had been
dewatered sufficiently to ensure thatthe catalytic recombiners remained
operable regardless of canisterorientation.547, 548
Following final preparations andchecks, the canister handling bridge
moved the defueling canister to thefuel transfer cask (FTC) loading
station in the “A” spent fuel pool.
The fuel handling building’s crane
lowered the FTC onto the loading
station platform over the pool where
the FTC lifted the canister from thewater by a grapple. As the canister
breached the surface of the poolwater, a spray of borated
demineralized water washed thecanister. The fuel handling building’s
Model 125-B shipping cask loading
tower. A fuel canister was lowered into crane then transferred the loaded
the shipping cask (lower center) from
FTC from the spent fuel pool and to
the fuel transfer cask (not shown). The the fuel handling building’s truck bay
loading tower lowered the shipping
for loading into the shipping cask.
cask to a horizontal position onto a
After verifying its conformance to its
skid that would be attached to the rail
NRC-approved shipping cask
car.

176

certificate of compliance, the loaded shipping cask was transported out of
the fuel handling building and off the island by rail. A dedicated traintransported up to three casks at a time to INEL.549 One shipment carrying
two shipping casks experienced a minor accident as the train engine struck a
car while traveling at slow speed. The engine received minor damage, but
the shipping casks were not damaged.550
Once a rail shipment arrived at the Central Facilities Area at INEL, both
impact limiters were removed from each end of the shipping cask and thecask and skid were lifted from the rail car and placed on a tractor trailer.
The cask was then transported 42 kilometers (26 miles) to wet storage at the
INEL Test Area North. Here each canister was removed from the shippingcask, filled with water, placed into a storage module, and transferred to thestorage pool in a designated location. A vent tube was installed on each
canister for continuous venting. The empty shipping cask was surveyed forcontamination and prepared for rail shipment as regular freight back to
TMI-2.551, 552 A detailed account of fuel shipments, including public
outreach, was provided in the DOE report “Historical Summary of the Three
Mile Island Unit-2 Core Debris Transportation Campaign”(DOE-ID-10400).
The 342 stainless steel fuel canisters of core debris were stored inunderwater storage from 1986 to 2001 at the INEL.553 During the 2000 to2001 period, these canisters were transferred to the TMI-2 Independent
Spent Fuel Storage Installation, also located at INEL, for interim storageof the TMI-2 core debris.554
Document Collections. Those documents discussed above and many otherdocuments relating to defueling activities are provided in the DVD folder,
Defueling. The following types of documents are included in this collection:•system description, technical evaluation, and safety evaluation reports•

research reports, NRC technical NUREG reports, GEND reports, andtechnical reports from DOE national laboratories•

other correspondence between the NRC and the licensee relating todefueling systems, equipment, and activities, such as notifications,requests, reviews, and approvals

Project schedules for defueling activities are provided in the DVD folder
General Management and Oversight.
177

Above: The Model 125-B shipping cask being removed from the rail car at the Central
Facilities Area at INEL. Both impact limiters are shown removed from the ends of thecask. Below: Defueling canisters in the temporary storage pool at INEL. A vent tube
was installed on each canister for continuous venting.

178

9 After Defueling
Defueling activities were considered complete with the shipment of the lastremaining defueling canisters containing core material from the TMI site on
April 15, 1990. The major objective of TMI-2 post-defueling activities
focused on preparing the plant for long-term storage. The licensee called theperiod preceding the ultimate disposition (either refurbishment and restart ordecommissioning) of the plant “post-defueling monitored storage” or“PDMS.” Long-term storage was proposed by the licensee on
December 2, 1986.555 However, the concept of PDMS was first introducedby the NRC Advisory Panel for the Decommissioning of TMI-2 onApril 12, 1984.556 The approach to PDMS was expanded when the licensee
submitted its environmental evaluation of PDMS on March 11, 1987.557
During PDMS, the TMI-2 facility would be in long-term monitored storage,
similar to the decommissioning mode SAFSTOR (mothballing with delayed
dismantling), in which the facility is secured, monitored, and maintained ina manner that ensures the protection of the public health and safety for an
extended period.558
The licensee stated in its December 2, 1986, post-defueling monitored
storage plan that a monitored storage period would be beneficial for thefollowing reasons: (1) occupational dose in the plant would be reduced
during monitored storage because of natural decay of radioactive
contamination; (2) a monitored storage period would allow time forcontinued development of decontamination technology; (3) further
reduction of occupational exposure would be achieved through the use ofadvanced robotic technology, automatic cleaning and chemical cleaning
techniques, and advanced waste treatment methods; and (4) developing
technology for radioactive waste packaging and volume reduction could
result in a reduction in the total volume of radioactive waste generated
following PDMS. In addition, the licensee had stated that placing the TMI-2
facility in storage would eliminate any possible impact of TMI-2
decontamination and decommissioning efforts on the operating TMI-1
facility.559, 560
Prerequisites for PDMS. The requirements for transition to PDMS were
contained in the TMI-2 possession-only license and the NRC-approved listof PDMS requirements and commitments.561 The basic criterion fortransition to PDMS was assurance that the health and safety of the public
was protected by conformance to all applicable NRC regulations.562
Transitioning to PDMS required the following conditions: (1) criticality was
no longer possible; (2) potential for fission-product movement waseliminated; (3) fuel was removed and shipped offsite; (4) radioactive waste
179was shipped or stored; (5) radiation levels were reduced commensurately
with the need for access to permit continued plant monitoring and to support
plant-disposition decisions; (6) water was removed from plant systems andspaces, and the potential for reintroduction of water was precluded; and(7) a safe, monitored plant condition was established.563
PDMS Environmental Protection Systems. The principle safety concern
during PDMS was the inadvertent release of radioactive material into theenvironment. For this reason, the NRC identified structures, systems, andcomponents that provided reasonable assurance that the facility could bemaintained in a defueled condition without undue risk to the health andsafety of the public. These systems, called PDMS environmental protection
systems, included the following:564•

Reactor vessel. Maintained residual debris geometry, precluding thepossibility of an inadvertent criticality. To keep the residual fuel in thereactor vessel in the analyzed geometry during PDMS, the PDMS
technical specifications limited activities that could alter the geometryof the fuel debris in the reactor vessel by controlling excess loads over
the reactor vessel, limiting movement of remaining fuel debris outside
analyzed geometries, and minimizing the potential for water
accumulation in the reactor vessel.565 As a conservative measure, thelicensee added 1,700 pounds of insoluble neutron poison in the form ofborosilicate glass shards to the bottom of the reactor vessel to ensure
long-term subcriticality of the residual fuel in the reactor vessel.566•

Containment structure. Ensured containment of the remaining
radioactive contamination during the PDMS period. To maintain the
integrity of the environmental barrier, inactive penetrations were closed
off with isolation valves or with welded or bolted blind flanges. In
addition, the PDMS technical specifications required routine
surveillance inspections of containment penetration isolation. Isolation
valves on active containment penetrations used by the containment
atmospheric breather and the reactor building’s purge system would
close on a high containment pressure.567•

Purge, breather, ventilation, and filtration systems. Controlled
radioactive effluents from the reactor building and the auxiliary and fuelhandling building. Existing systems used during PDMS included thereactor building’s ventilation and purge system, the auxiliary building’s
ventilation and filtration system, and the fuel handling building’s
ventilation and filtration system. These systems would not be operated
continuously, but on an as-needed basis. A passive containment

180

atmospheric “breather” was installed for PDMS to maintain pressure
equilibrium between the auxiliary and fuel handling building and thereactor building. A differential pressure would develop when theauxiliary building’s ventilation system was operating and the reactor
building’s ventilation system was not operating. The containment
atmospheric breather also provided a HEPA-filtered pathway for any
effluent from the containment.568•

Fire protection system. Detected and mitigated any effects of a firewithin the facility. The original TMI-2 system of fire protection had
been modified to address the reduced functional requirements for fire
protection for PDMS. The PDMS fire protection program consists offire detection and alarm capability, manual suppression by fire brigade,removal of most flammable and combustible liquids and materials,
control of transient combustibles, and deenergization of most electricalcircuits.569•

Flood protection system. Minimized the intrusion of water in thefacility and movement of radioactive contamination to the environment.
The existing unit flood protection capabilities were maintained for
PDMS, such as protective island dikes, flood panels, watertight doors,
and an early warning system of flooding conditions.570•

Support and monitoring systems. Supported PDMS configuration toensure personnel and environmental protection and surveillance.
Systems included (1) electrical systems, such as area lighting, fire
detection, radiation monitoring, PDMS support systems, andcommunications; (2) effluent monitoring systems for the reactor
building’s purge system and ventilation systems; (3) environmental
monitoring systems, such as radiation monitoring, sample collection andanalysis, rodent carcass analysis, and pest control; (4) administrative
systems, such as organizational structure, staff qualifications, records,
independent safety reviews, procedures, occupational radiation
protection, quality assurance plan, and emergency plan; and(5) surveillance programs, such as maintenance of reactor vesselgeometry, reactor building isolation, the reactor building’s breather andventilation system, the auxiliary and fuel handling building’s ventilation
system, fire protection, flood protection, support and monitoring
systems, and radiological surveys.571

Key Reports. Reports that supported the licensee’s request and the NRC’s
approval of the proposed license amendments for PDMS (issuance of thepossession-only license and issuance of the PDMS technical specifications)181

are summarized below. Issue dates and report conclusions are further
summarized in the key actions section that follow.
•

Defueling completion report. This report was developed by the licensee
to document the measurements and calculations that were performed toensure that the plant had been defueled to the extent reasonably
achievable and that the potential for a nuclear criticality had been
precluded during normal and accident conditions.572 The report was
required by the plant technical specifications (shown in Table 1.1 ofAmendment No. 30) to document the basis for the TMI-2 facility
transition to Mode 2. (For transition from Mode 1 to Mode 2, thelicensee was required to demonstrate that the reactor vessel and reactor
coolant system had been defueled to the extent reasonably achievable,that the possibility of criticality in the reactor building was precluded,
and that there were no defueling canisters containing core material
remaining in the reactor building.)573 In addition, this report andsubsequent reports provided the criticality safety analyses to support the
proposed license amendment for the possession-only license andissuance of the PDMS technical specifications.574
The contents of the report included the following sections: overviews ofdefueling objectives and guidelines and residual fuel characterization(Section 1); detailed discussions of the post-accident fuel transport anddispersion conditions (Section 2); survey techniques used for residual
fuel measurements (Section 3); major fuel-removal accomplishments
and methods (Section 4); residual fuel quantification, by location, andcriticality analyses for each fuel location, as appropriate (Section 5); an
assessment of possible alternatives and projected occupational doses
associated with attempts to remove the remaining fuel (Section 6);
cumulative occupational exposures during defueling-related activities(Section 7); licensee conclusions (Section 8); and criticality safety
evaluation for the TMI-2 safe fuel mass limit (Appendix B).575 Someanalyses and data in the report were later revised in the reactor vessel
criticality safety analysis report and the reactor vessel post-defuelingsurvey report.576, 577•

PDMS requirements and commitments list. This document wasdeveloped and revised by the NRC and the licensee to list therequirements and commitments needed to place the facility into PDMS.
The document provided requirements for the final PDMS configuration
to support the proposed license amendment for the possession-only
license, the PDMS technical specifications, and license conditions
contained in the requirements for entry into PDMS. This list was182

generated from the PDMS safety analysis report that was submitted by
the licensee in support of their license amendment request, as amended;
the associated safety evaluation issued by the NRC on
February 20, 1992; and several public meetings at TMI.578 The NRC
approved a process to allow changes to the list of requirements andcommitments in recognition of the difficulties associated with readying
the facility for long-term storage and the dynamic nature of the effort.
Deviation requests by the licensee had to include a safety analysis
evaluating each proposed change.579 The list was incorporated in
Amendment 45 to the facility license that modified the original
operating license to a possession-only license.580•

PDMS safety analysis report. This report was developed and revised by
the licensee to document the facility description and safety analysis of
the PDMS configuration. The report provided the basis for long-term
stability and safety of the proposed facility license amendment for thepossession-only license and the PDMS technical specifications.581 The
report is the current licensing-basis document for PDMS and isperiodically updated to reflect current plant conditions.
The contents of the original report included: description of the current
status of the plant after extensive decontamination (Section 1); sitecharacteristics (Section 2); regulatory review of conformance of theTMI-2 facility to 10 CFR Part 50 (Section 3); description of
fuel-removal activities and special nuclear materials accountability(Section 4); enumeration of the radiological status of the plant andradiological goals to be attained before entry into PDMS (Section 5);
list of deactivated systems and facilities (Section 6); description ofoperational systems and facilities (Section 7); identification andquantification of routine and unanticipated releases during PDMS
(Section 8); description of the proposed changes to the technicalspecifications to permit entry into PDMS (Section 9); and primaryadministrative functions for the management of TMI-2 during PDMS,
such as quality assurance, security, emergency preparedness, radiation
protection, and organizational responsibilities (Section 10). The originalsubmittal reflected the anticipated facility conditions at the beginning ofPDMS. The report was amended several times based on new
information, responses to formal questions from the NRC, and changes
in specifications for the facility. Routine updates are required byregulatory requirements.582•

Post-defueling survey reports. A series of 10 reports were developed by
the licensee to document the assessment of special nuclear material
183

(SNM) remaining in the plant following the completion of the defueling
effort. A report was issued for the following reactor components: upper
plenum, letdown coolers, pressurizer, reactor vessel head, reactor
building basement, both steam generators, auxiliary and fuel handlingbuilding, miscellaneous components in the reactor building, reactor
coolant system, and reactor vessel.583
The geometric modelthat was used to
conduct the criticality
safety analysis of theresidual fuel in thereactor vessel. An
annular ring,
representing the vesselinternals andpostulated debris
accumulations locatedalong the outer
periphery of thevessel, wasconservatively
assumed to go 360
degrees around thevessel. The analysis
used in-vessel
inspections of debris
locations and someconservative estimatesof the remaining fuelto develop a specificthree-dimensional
analytical model of thereactor vessel endstate configuration.
Criticality calculations
were performed by the
Oak Ridge National
Laboratory. A
separate criticality
assessment was
performed for
accident scenarios.
(See defuelingcompletion report for
additional
information.)184

On October 17, 1985, the licensee was granted an exemption fromcertain requirements for periodic inventory and reporting of SNM (e.g.,uranium-235 and plutonium) balance for TMI-2. As a condition of the
exemption, the licensee was required to conduct an assessment of the
SNM remaining at TMI-2 following the completion of the defueling
effort.584 The compilation of the individual survey reports provided thebasis for the final assessment of the quantity of residual SNM foraccountability purposes.585 In addition, the reactor vessel post-defuelingsurvey report provided the basis for satisfying the criticality
determination requirement for entry into PDMS.586
The contents of each report included: (1) a detailed description of thearea, system, or component; (2) its role in the accident and/or cleanup
activities; (3) the methodology used to determine the quantity of SNM;
(4) the rationale supporting a conclusion as to whether the area, system,or component contained residual SNM and, if so, a summary of theappropriate SNM engineering calculations; (5) applicable drawings ofthe area; and (6) an assessment of residual fuel.587 Because of thecomplex structure of the components surveyed, the assessment of theresidual SNM used a combination of direct measurements, sample
analyses, volumetric measurements, and engineering analyses. The
survey reports were supported by detailed engineering calculations,contractor reports, and research data.588•

Programmatic Environmental Impact Statement (PEIS)
Supplement 3. The last supplement to the initial PEIS (NUREG-0683)
was developed by the NRC to document the environmental evaluationof the licensee’s proposal to complete the current cleanup effort andplace the facility into monitored storage for an unspecified period oftime. This supplement was used by the NRC to evaluate the licensee’s
proposed facility license amendment for the possession-only license andthe proposed PDMS technical specifications. The supplement provided
an environmental evaluation of the licensee’s proposal and a number ofalternative courses of action from the end of the current defueling effort
to the beginning of decommissioning. However, the objective of thereport was not to provide an evaluation of the environmental impacts
associated with the decommissioning.
The contents of the report included: status of the cleanup activities andconditions in the plant, including radiation-source characteristics(Section 2); evaluation of the licensee’s proposal and NRC-identified
alternatives for potential environmental impacts, such as the offsite
population’s exposure to radiation from routine and accidental releases,
185

occupational radiation dose, waste management impacts, transportation
impacts, socioeconomic impacts, commitment of resources, andregulatory considerations (Section 3); description of the environment
and population that could be affected by the licensee’s proposed action
and alternatives (Section 4); comparison of environmental impacts ofthe proposal and alternatives (Section 5); discussion of the potential forhuman health effects from both offsite and onsite occupational radiation
exposures as a result of TMI-2 cleanup (Section 5.2); discussion ofnonradiological impacts identified, such as the cost of implementation,
long-term commitment of land and burial-ground space, andsocioeconomic effects (Section 5.3); discussion of postulated accidents,
such as radiological impacts resulting from accidents onsite and offsite
during waste transportation, and nonradiological impacts including
traffic accidents, injuries, and fatalities (Section 5.4); NRC conclusions(Section 6); and NRC responses to public comments on the draft
supplement (Section 7).
•

PDMS technical evaluation report. This report was developed by the
NRC to document the technical evaluation of the licensee’s proposal toplace the TMI-2 facility into PDMS. The report provided the basis forthe requirements and controls to be maintained during PDMS. The
report also provided the basis for approving amendments to the facility
license that issued the possession-only license (Amendment No. 45) and
the PDMS technical specification (Amendment No. 48).589
The contents of the report included a brief regulatory history of theTMI-2 facility (Section 2); description of PDMS (Section 3); status of
the facility before entry into PDMS (Section 4); description of the major
prerequisites for facility configuration at the start of PDMS (Section 5);
discussion of the structures, systems, and components for environmental
protection during PDMS (Section 6); and conclusions (Section 7).590•

PDMS safety evaluation report. This report was developed and revisedby the NRC to document the safety evaluation of the licensee’s proposal
to place the TMI-2 facility in PDMS. The report provided the basis forapproving amendments to the facility license that issued thepossession-only license (Amendment No. 45) and the PDMS technicalspecifications (Amendment No. 48). The initial safety evaluation report
was issued in February 1992 in response to the initial licensee
application in August 1988. This safety evaluation was updated toaccount for revisions to the licensee application.591, 592

186

The contents of the report included: a summary of licensee submittals
and NRC actions associated with the licensee’s initial amendment
request in 1988 (Section 2); review of licensee actions that satisfied thelicense condition for entry into PDMS specified by the possession-only
license (Section 3); and evaluation of the proposed PDMS technicalspecifications, as revised, and comparison to the existing TMI-2
technical specifications, as amended (Section 4). The technicalevaluation report (see above) was issued concurrently with thisdocument to provide additional details. Both documents were preparedby the Pacific Northwest Laboratory (currently Pacific Northwest
National Laboratory) under the direction of the NRC. 593, 594
Key Actions. In August 1988, the licensee submitted a safety analysis
report that documented and supported their proposal to amend the TMI-2
license to a “possession-only” license and to allow the facility to enter
PDMS. In response to the request, the NRC issued Final Supplement 3 to
the Programmatic Environmental Impact Statement for the TMI-2
decontamination and cleanup in August 1989. In February 1992, the NRC
issued a safety evaluation report regarding the PDMS license amendment
and a technical evaluation report regarding PDMS. These three NRC
documents formed the basis for the position on the acceptability of PDMS.
The NRC issued a possession-only license in September 1993 and approved
the PDMS technical specifications three months later. Key post-defuelingactivities and actions are summarized as follows:•

Licensee submitted license amendment request. On August 16, 1988,
the licensee submitted a request to amend the operating license to apossession-only license and to extensively modify the technicalspecifications in ways consistent with the licensee plans for long-term
storage at the facility. The request included the proposed amended
facility license for PDMS, proposed PDMS technical specifications, and
the PDMS safety analysis report.595 The pre-accident and recovery
technical specifications consisted of two parts: Appendix A, whichpertained to the facility and recovery, and Appendix B, which pertained
to the environment. The licensee proposed combining these two
documents into one set of technical specifications for PDMS. Also, thelicensee proposed placing the remaining surveillance requirements for
PDMS, which were in the recovery operations plan, back into thetechnical specifications.596•

NRC issued PEIS Supplement 3. In August 1989, the NRC published
Final Supplement No. 3 to the Programmatic Environmental Impact
Statement (NUREG-0683), which dealt with PDMS and subsequent187

cleanup. The NRC assessed the licensee’s proposal and six alternatives.
The licensee’s proposal and one of the alternatives (continuing andcompleting the cleanup without a storage period) were evaluated indetail. The NRC concluded that both the licensee’s proposed plan and
the NRC staff-identified alternative for completion of cleanup were
within the applicable regulatory limits, and each could be implemented
without significant environmental impact. Neither alternative was foundto be clearly preferable from an environmental impact perspective.597, 598•

Defueling completed. In April 1990, the TMI-2 facility transitioned
from Mode 1 to Modes 2 and 3, in accordance with the transition
requirements set forth in Table 1.1 in the recovery technicalspecification. In Mode 2, it was recognized that defueling wascompleted and, thus, boration of the reactor coolant system and staffingof the control room by licensed operators were no longer required. In
Mode 3, it was recognized that offsite shipment of the fuel wascompleted and boration of the spent fuel storage pools was no longer
required. The three criteria for changing from Mode 1 to Mode 2 wereas follows: (1) the reactor vessel and reactor coolant system weredefueled to the extent reasonably achievable; (2) the possibility of a
criticality in the reactor building was precluded; and (3) there were no
defueling canisters containing core material in the reactor building. The
additional requirement for transition to Mode 3 was that no defuelingcanisters containing core material remained on the TMI site. The NRC
and consultants from the Pacific Northwest Laboratory performed adetailed technical review and inspection to verify that the criteria were
met. The facility made the transition from Mode 1 to Mode 2 onApril 26, 1990, and to Mode 3 the following day.599, 600, 601•

Reactor vessel drained down. During July and August of 1991, thereactor vessel was drained to make final measurements of residual fuelremaining in the vessel. The reactor vessel fuel-measurement program
was the final step in the special nuclear materials accountability
program at TMI-2.602•

Petition to intervene in the PDMS license amendment request. On
April 25, 1991, the NRC published a notice of opportunity for a priorpublic hearing regarding the license change to implement PDMS. One
individual petitioned to intervene. The petitioner, the licensee, and the
NRC staff reached a settlement on September 25, 1992, and the request
to intervene was withdrawn; on October 16, 1992, the NRC Atomic
Safety and Licensing Board dismissed the proceeding.603

188

•

NRC issued safety evaluation of PDMS. On February 20, 1992, the
NRC issued a safety evaluation addressing the license conditions andtechnical specifications necessary to implement PDMS. As part of theevaluation, the staff published a technical evaluation report whichappraised PDMS as an integrated process and assessed licensee
commitments that were not in the technical specifications. The safety
and technical evaluation reports and Supplement 3 to the NRC’s Final
Programmatic Environmental Impact Statement formed the basis for thestaff’s position on the acceptability of PDMS.604, 605•

Decision not to restart. On October 26, 1992, the licensee informed
NRC of its intention to discontinue maintaining the TMI-2 restart listthat included a list of NRC generic communications (bulletins, genericletters, and information notices) that would be reviewed for applicability
and required action if a decision was made to refurbish TMI-2 foroperation. The licensee had publicly acknowledged that TMI-2 would
not be refurbished as a nuclear power generation facility, but instead,
TMI-2 would be kept in a monitored storage condition until
decommissioning of Units 1 and 2 simultaneously.606•

Licensee issued PDMS requirements and commitments. On
January 15, 1993, the licensee provided a proposed list of remaining
PDMS requirements and commitments that had to be completed before
issuance of the possession-only license and PDMS technicalspecifications. This list was generated from the PDMS safety analysis
report; the NRC’s February 20, 1992 safety evaluation report; andseveral public meetings at the nearby TMI training facility. The NRC
approved this list on May 19, 1993, and approved subsequent revisions
to the list, as well.

•

Licensee issued the reactor vessel post-defueling survey report. On
February 1, 1993, the licensee notified the NRC in its last
post-defueling survey report that the current best estimate of the residual
fuel remaining in the reactor vessel was 925 kilograms with anuncertainty of plus or minus 40 percent as one standard deviation. The
estimate of remaining fuel in the reactor vessel was based onunderwater video inspections and passive neutron measurements. Video
inspections were used to divide the reactor vessel into nine zones which
separated the major fuel deposits by elevation. An array of helium-filleddetectors were used to measure fast neutrons produced by the residual
fuel as water was removed from the vessel in stages so that the water
could be used as a shadow shield to separate the fuel deposits by zone.
The estimate was derived from calculations made by onsite staff and an189independent review by an offsite group headed by Dr. Norman
Rasmussen of the Massachusetts Institute of Technology. The total
residual fuel estimate (925 kilograms) from fast-neutron measurements
was about 50 percent larger than the less accurate video estimate.607
For the balance of the facility outside the reactor vessel, earlier licensee
estimates based on measurements, sample analyses, and visualobservations indicated that no more than 385 pounds (174.6 kilograms)of residual fuel remained.608, 609•

NRC issued safety evaluation of reactor vessel criticality analysis. On
July 6, 1993, the NRC issued a safety evaluation confirming earlier
analyses done by the licensee which indicated that the fuel remaining in
the TMI-2 reactor vessel would remain subcritical, with an adequate
margin of safety, during PDMS. The NRC and consultants from PacificNorthwest Laboratory performed independent evaluations and madeindependent measurements of earlier fuel measurements in the auxiliary
and reactor buildings. The NRC concluded that the fuel remaining in thereactor vessel would be subcritical under both steady-state and accident
conditions. The NRC also found that there were insufficient discrete
quantities of fuel in areas other than the reactor vessel to sustain acriticality.610, 611, 612•

NRC issued possession-only license. On September 14, 1993, the NRC
issued a possession-only license in Amendment 45 to the facility
operating license without modification of existing recovery technicalspecifications. The PDMS technical specifications would be issued after
the licensee had completed the final stages of the cleanup effort; the
NRC had verified the implementation of the PDMS commitments andrequirements; and the licensee had satisfied a number of PDMS license
conditions. These license conditions included (1) completion of a
special study of the ventilation in the auxiliary and fuel handlingbuilding that monitored airborne levels for one year before terminationof the continuous operation of the ventilation systems; (2) developmentof an NRC-approved surveillance requirement for the unfilteredleak-rate test in the reactor building; and (3) submittal andimplementation of a site flood protection plan, a site radiation protection
plan, an offsite-dose calculation manual, a PDMS fire protection
program evaluation, a PDMS quality assurance plan, and a radiologicalenvironmental monitoring plan. In addition, the licensee was required tosubmit the results of the completed plant radiation and contamination
surveys before entry into PDMS.613

190

•

PDMS requirements and commitments completed. On
November 12, 1993, the licensee informed the NRC that all of therequirements and commitments for entry into PDMS had been satisfied
and that they would be ready to transition to PDMS within the next
30 days. The licensee’s letter to the NRC provided a list of all PDMS
entry requirements along with references to the licensee letters thatprovided closeout documentation.614, 615 The NRC independently
verified that the licensee had satisfied all the PDMS requirements andcommitments.616•

NRC issued PDMS technical specifications. On December 28, 1993,
the NRC issued Amendment 48 to the facility operating license toincorporate the PDMS technical specifications. The PDMS technicalspecifications replaced the technical specifications that pertained to thefacility and recovery (Appendix A to the facility operating license);
technical specifications that pertained to the environment (Appendix B);
and the recovery operations plan that pertained to surveillance. The
license amendment also included an update of the safety evaluation
report for PDMS that was first issued on February 20, 1992. In thesafety evaluation, the NRC evaluated the potential for the routine andaccidental releases of any significant quantity of radioactive material
during PDMS. The NRC performed independent evaluations of eight
potential accidents: vacuum canister failure, spraying of concentrated
contamination with high-pressure spray, cutting of contaminated pipe,
break of contaminated pipe, elevator/stairwell fire in containment,
D-ring compartment fire in containment, containment penetration
failure, and the rupture and release of resins from the makeup andpurification demineralizers.617
During the safety evaluation, the NRC reviewed the final defuelingcompletion report (1990) and the PDMS safety analysis report, as
amended. The following conclusions of the NRC’s safety evaluation
were based on the information from the licensee’s reports, the NRC’s
PEIS Supplement 3 (1989), and the NRC’s PDMS technical evaluation
report (1992):618oDefueling of the reactor had been accomplished to the extent
reasonably achievable.

o

All fuel and core debris removed from the reactor and associatedsystems had been shipped offsite.

191

o

Results of analyses indicated that there was no potential for
criticality in the fuel remaining in the TMI-2 facility during either
normal or accident conditions.

oRemaining radioactive waste from the major TMI-2
decontamination activities had been shipped offsite or packaged andstaged (prepared) for shipment offsite.

oRadiation levels within the facility had been reduced to the extent
that plant monitoring, maintenance and inspections could be
performed.

oRadiological surveillance of activities during PDMS could be
conducted in accordance with the approved offsite-dose calculation
manual and in compliance with the regulatory requirements of
10 CFR Part 20, “Standards for Protection against Radiation,”
which would, with the approved radiation protection plan, ensure
adequate control of occupational exposure and protection ofworkers.

o

The surveillance program proposed by the licensee would
adequately monitor the PDMS environmental protection systems.

Independent spent fuel storage installation (ISFSI) for the dry storage of
TMI-2 fuel debris located at the INEL. The TMI-2 defueling canisters were
removed from the storage pool, dewatered, dried, and placed in a dry shielded
canister. The dry shielded canister with up to 12 defueling canisters wasplaced inside a reinforced concrete horizontal storage module. The ISFSI was
certified and licensed by the NRC.

192

o

Environmental monitoring activities for TMI-2 during PDMS,
included in the TMI site radiological environmental monitoring
plan, would ensure adequate environmental surveillance andcontrol.

oFire prevention, detection, and control as specified by the approved
fire protection program evaluation would ensure adequate reductionof fire potential, as well as detection and control during PDMS.

oRequirements delineated in the proposed PDMS technicalspecifications provided assurance that the facility would bemaintained in a safe condition that would not negatively impact theenvironment.

Document Collections. Those documents discussed above and many otherdocuments relating to post-defueling activities are provided in the DVD
folder, After Defueling.

TMI-2 independent spent fuel storage installation located at INEL for interim
storage of the TMI-2 core debris.

193

194
(FirstEnergy, 2015)

10 Postscript
In 2001, FirstEnergy acquired TMI-2 from GPU. FirstEnergy contracted themonitoring of TMI-2 to Exelon, the current owner and operator of TMI-1.
The companies plan to keep the TMI-2 facility in long-term, monitored
storage until the operating license for the TMI-1 plant expires, at which time
both plants would be decommissioned. On-site NRC resident inspectors andinspectors from the regional office in King of Prussia, PA continue tomonitor the operations at TMI-1 and the conditions at TMI-2.
This knowledge management volume ends with the licensing of the PDMS
configuration and the issuance of the PDMS technical specifications at theend of 1993. Activities to maintain and oversee the TMI-2 facility in PDMS
will continue until the decommissioning of both units on Three Mile Island.
Documents and correspondence pertaining to the maintenance of PDMS,
plans for decommissioning, and associated regulatory activities can beaccessed electronically from the NRC public website (www.nrc.gov).
The Agencywide Documents Access and Management System (ADAMS) is
the official recordkeeping system through which the NRC provides accessto collections or “libraries” of publicly available documents. The NRC’s
Publicly Available Records System (PARS) Library provides full-text
documents dating from about April 2000 onward (and some earlier
documents). The Public Legacy Library contains bibliographic citations(some with abstracts and full text) for earlier documents, including pre- andpost-accident time periods. Documents relating to TMI-2 can be found in
the ADAMS libraries by searching the TMI-2 Docket Number 0500302.
Most documents prior to April 2000 are only available in microfiche format.
Refer to the NRC Public Document Room webpage for instructions for
obtaining documents from the microfiche collection. NRC staff can accesslegacy documents from microfiche stations located at headquarters andregional offices.
Other sources of documents relating to TMI-2 can be currently found fromthe sources listed below.
•

American Nuclear Society (www.ans.org) maintains an extensive
collection of their journal articles and proceedings on every aspect of
the TMI-2 accident. In particular, the proceedings of the topical meeting“The TMI-2 Accident: Materials Behavior and Plant Recovery
Technology,” held in Washington, DC in 1988 were published in the195special volume of the Nuclear Technology journal. The papers in theproceedings and other papers may be purchased from their website.
•

Department of Energy's Office of Scientific and Technical
Information website (www.osti.gov) contains over 1,000 electronic fulltext reports and papers relating to DOE-funded research in support of
the recovery and cleanup efforts at TMI-2. Most of these documents areprovided on the DVDs to this NUREG/KM.

•

Dickinson College maintains the Three Mile Island website(www.threemileisland.org) that contains a collection of documents
relating to the accident, including transcripts and audio recordings ofinterviews collected at the time, government and industry documents,
photographs, and newspaper coverage.

•

Electric Power Research Institute (www.epri.com) made available on
their website many of their research reports that supported the cleanup
and understanding the accident. Four comprehensive reports on theaccident, recovery, and cleanup include “Analysis of Three Mile Island- Unit 2 Accident” (NSAC-80-1), “The Cleanup of Three Mile Island
Unit 2, A Technical History: 1979 to 1990” (EPRI-NP-6931), “TMI-2

Left: The TMI-2 Investigation Center was established on June 2, 1979, at
the Commission’s Washington, DC office to provide a facility forinvestigators from congressional committees and the president's
commission to access all pre- and post-accident documentation concerning
the plant. Right: A technical information assistant uses the first generation
automated Document Control System (DCS) to search NRC documents
since 1978, and all TMI-related documents. Each document received or
generated by the NRC was indexed by bibliographic elements andmicrofiched for storage and use. The early DCS was replaced with the
Nuclear Documents System (NUDOCS) in 1987 with text search capabilityof abstracts and select full-text documents. The Agencywide Documents
Access and Management System (ADAMS) replaced NUDOCS in April
2000 with full-text storage and retrieval of all NRC records and documentsthat were entered into ADAMS.

196

Waste Management Experience” (EPRI-TR-100640), and “TMI-2 PostAccident Data Acquisition and Analysis Experience” (EPRI-NP-7156).
•

Idaho National Laboratory of the Department of Energy maintains acollection of research reports in their INL Knowledge eRepository(http://keros.inl.gov). Most of these documents are provided on theDVDs to this NUREG/KM.

•

International Atomic Energy Agency (www.iaea.org) made available
on their website technical reports that incorporated lessons from theresponse to the accident and the cleanup. Key reports include“Catalogue of Methods, Tools and Techniques for Recovery from Fuel
Damage Events” (IAEA-TECDOC-627), “Issues and Decisions for
Nuclear Power Plant Management after Fuel Damage Events” (IAEATECDOC-935), “Management of Severely Damaged Nuclear Fuel and
Related Waste” (IAEA-TRS-321), and “Cleanup and Decommissioningof a Nuclear Reactor after a Severe Accident” (IAEA-TRS-346).

•

National Technical Information Service of the U.S. Department ofCommerce serves as the largest central resource for government-fundedscientific, technical, and engineering information. Formal government
reports not available elsewhere can be purchased from their website(www.ntis.gov).

•

Pennsylvania State University Libraries’ maintains the TMI-2
Recovery and Decontamination Collection that contains several
thousand videotapes, reports, and photographs of the recovery andcleanup during the 1979–1990 time period (www.libraries.psu.edu). The
GPU Nuclear, EPRI, NRC, and DOE (GEND) co-sponsored a projectwith the University to catalog and maintain the extensive videotape
library for future researchers.

•

WM Symposia (www.wmsym.org) holds an annual international Waste
Management conference covering the management of radioactive waste
and related topics. Past proceedings of their annual conferences datingback to 1975 are available from their website (www.wmsym.org). Many
papers relating to the TMI-2 cleanup were presented during the 1980s.

•

Books about TMI-2 by NRC staff include “Three Mile Island: A
Nuclear Crisis in Historical Perspective” by J. Samuel Walker (2006)619
and “TMI 25 Years Later: The Three Mile Island Nuclear Power Plant
Accident and Its Impact” by Bonnie A. Osif, Anthony J. Baratta, andThomas W. Conkling (2004)620.
197

198

11 DVD Navigation and Interpretation
This knowledge management guide is an HTML-based collection of
information, documents, videos, and photos related to the 1979 accident at
the Three Mile Island nuclear power plant. This interactive feature isprovided on the DVD to help you navigate through the historical records.
Refer to the “Readme” file located on any DVD for viewing instructions.
Navigation•

Main welcome page. From the welcome page, you can navigate thedifferent sections of the guide using the blue tabs from the menu on theleft side of the page. Each tab is linked to a page that contains one ormore subtabs that appear to the right of the tab. Each tab corresponds toa document collection and each subtab links to a folder of documents.
This interactive feature will prompt you to load the appropriate DVD toretrieve the document, if necessary. Additionally, there is a photogallery, interactive timeline, and a document index (green tab). You
may also click on the text box on the timeline for a short description.

•

Document retrieval. Documents can be accessed from the document
page (see figure below) linked to the subtab. Next to each document is a
hyperlink which will display the associated document, photo, or video.
Because the document collection spans several DVDs, you may beprompted to insert a different DVD as you navigate the guide. A sorting
feature is provide on select columns on the document page.

•

Document searches. A simple keyword search feature is provided on
the welcome page and each document page. A search is applied to a listof all documents on all DVDs in this NUREG/KM. This featuresearches words in the file name.

•

Document lists. Several convenient lists of documents on the DVD canbe viewed from the “Document Folder Index” tab. These lists are also
provided in spreadsheet format in the “Common” folder on the DVD(located in the "Documents" folder). A list of more than 25,000 TMI-2
records in the Public Legacy Library is included in the spreadsheet.

•

Alternative. Documents are titled and arranged in topical folders on the
DVD (located in the "Documents" folder) to provide a usable alternative
to the interactive guide. The document can be accessed directly on the
DVD using a file explorer.
199

Things to Keep in Mind•

Legacy documents. Many of the documents on the DVD are historicalin nature and might contain information that is obsolete or supersededby current regulations and research results. The historical documents
provided on the DVDs are for historical reference only and are notofficial NRC records. Please refer to the NRC’s public website(http://www.nrc.gov) for current information on regulations, policy
statements, regulatory guidelines, regulatory processes, and research
results.

•

Units of measure. The unit of measure (in English units or theInternational System of Units) that was used in the original source
document was used in this digest. A conversion chart is provided on theback cover.

•Abbreviations. A small set of abbreviations was used throughout thisreport in order to improve readability: ALARA, CFR, DOE, DVD,
EPRI, GEND, GPU, INEL, NRC, NUREG, NUREG/CR, NUREG/KM,
PDMS, and TMI-2. Others that were less frequently repeated in thereport were spelled out at the beginning of each subsection or paragraphthat contained them.

•

Recovery vs. cleanup. The term “recovery” is used in this NUREG/KM
to mean actions taken to keep the plant in a stable condition and toprevent the inadvertent release of radioactivity. The term “cleanup” isused to mean actions taken to decontaminate and defuel the plant anddispose of radioactive waste. These two terms are often usedinterchangeably for certain actions.

•

EPRI and GPU documents. Documents generated by EPRI and GPU
are generally not provided on the DVDs unless the documents were
submitted to the NRC or funded by DOE.

•

“GPU” and “licensee.” Unless otherwise noted, the GPU Corporationand its subsidiaries (including GPU Service Corporation, GPU Nuclear
Corporation, and Metropolitan Edison Company) are referred tocollectively in a historical context as “GPU” or the “licensee” in the textof this NUREG/KM and document filenames.

•

Document accession number. Each document and each enclosure of adocument that was cataloged in the Agencywide Documents Access andManagement System (ADAMS) Public Legacy Library was assigned a200unique accession number. This number can be found on the first page ofeach document. Each enclosure to a transmittal letter was typicallyassigned its own accession number.
•

Document filename. The filenames used on the DVDs typically containthe date of the document (generally the date of the transmittal letter);originating organization (e.g., NRC or GPU); document type(e.g., safety evaluation or system description); short title; document
revision, if any; and reference date of previous correspondence, if any.
Technical reports by the NRC (e.g., NUREGs), DOE, and national
laboratories start with the report’s identification number, short title, andyear and month issued. See examples below.
(1990-01-18) GPU, Defueling Completion Report, Rev. 3 (re 08-18, 10-09-1989)
NUREG-0698, Rev. 2, NRC Plan for Cleanup Operations at TMI-2 (1984-03)•

Documents, photographs, and diagrams in this NUREG/KM andDVDs were copied from the best available (surviving) sources.
Photographs were generally taken by GPU and DOE contractors.

A typical document page listing the contents of a document folder. Columns
can be sorted. Keywords in document file names can be searched. The
hyperlink directs you to the document, photo, or video.

201

Computer-aided design cut-away drawing showing the defueling work platform
support structure and internals indexing fixture (purple), reactor vessel (green), uppercore support assembly (red), and lower core support assembly (bottom inside vessel).

202

12 Contributions to Supplement 1(NUREG/KM-0001, Recovery and Cleanup)
Knowledge
Don Marksberry
Gordon Skillman
Michael Masnik
Management
Don Marksberry
Felix Gonzalez
Kenneth Hamburger
Theodore Smith
Anita Aikins-Afful
Robert Norman
Thomas Wellock
Andrew Bates
Justin PooleDavid Aird
Ian Gifford
Tom KardarasMark Henry Salley
andThe NRC Print Shop and Publications Staff
Document Collections

NRC’s Agencywide Documents Access and Management System
Publications Branch of the NRC’s Office of Administration
NRC’s Office of the Secretary
U.S. Department of Energy
Idaho National Laboratory
Electric Power Research Institute
Dickinson College Community Studies Center
Pennsylvania State University Engineering Library
National Archives and Records Administration203

Cross-sectional view of the core bore drilling machine mounted on a platform over thedefueling work platform. The core bore machine was originally used to extract core
stratification samples in 1985 and later to break up the solidified monolith in the coreregion during defueling. The machine was used again to drill through and cut portionsof the lower core support assembly.

204

13 Endnotes
Note: Endnote citations are file names of documents on the DVD, except for
EPRI reports and journal papers.
1 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes

Resulting from TMI-2 (1981-03)

2 NUREG-1350, Vol. 25, Information Digest, 2013–2014

3 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman)
4 NUREG-0698, Rev. 2

5 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03)

6 (1980-03-26) Coordination Agreement, TMI Information and Examination Program (DOENRC-GPU-EPRI)
7 (1982-03-19) DOE-GPU Agreement, Acquisition of Damaged TMI-2 Reactor Core byDOE
8 (1982-04-19) NRC-DOE Revised MOU, Removal and Disposition of Solid Nuclear Waste
from TMI-2 (SECY-82-165)
9 DOE-ID-10276, TMI-2: Lessons Learned by the U.S. Department of Energy, A
Programmatic Perspective, March 1990
10 NUREG/CR-1250, Vol. 2, Part 3 (Rogovin Report)
11 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981-03)
12 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981-03)
13 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981-03)
14 NUREG-0698, Rev. 2
15 (1981-08-26) GAO-EMD-81-106, Greater Commitment Needed to Solve Continuing
Problems at TMI
16 (1986-01-26) Operating License Amendment 26 (Fully incorporates Recovery Mode
Proposed Technical Specifications)
17 (1981-08-26) GAO-EMD-81-106, Greater Commitment Needed to Solve Continuing
Problems at TMI
18 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup
19 NUREG-0698, Rev. 2
20 NUREG-0698, Rev. 2
21 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
22 NUREG-0698, Rev. 2
23 (1980-02-13) NRC Order (2-11-1980), Provides TMI-2 Recovery Technical Specifications
24 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup
25 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup
26 (1979-05-26) NRC Policy, NRC to Advise Use of EPICOR-II After Public Comment
27 (1979-11-21) NRC Policy, Notice of Intent to Prepare Programmatic EIS (TMI-2)
28 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup
29 (1979-11-21) NRC Policy, Notice of Intent to Prepare Programmatic EIS (TMI-2)
30 (1980-09-26) NRC Policy, State (Pennsylvania) Requirements at TMI
31 1980 NRC Annual Report (NUREG-0774)
32 (1981-04-27) NRC Policy, Programmatic Environmental Impact Statement of the Cleanupof TMI-2 (46 FR 24764)

205

33 http://www.nrc.gov/about-nrc/policymaking.html (accessed July 28, 2014)34 (1979-10-22) NRC Order (10-18-1979), Operation of EPICOR II to Decontaminate Rad
Waste Water in Auxiliary Building
35 (1980-06-13) NRC Order (6-12-1980), Orders Release Krypton-85 from the Reactor
Building Atmosphere by Controlled Purge
36 http://www.nrc.gov/about-nrc/policymaking.html (accessed July 28, 2014)37 (1981-01-12) NRC Chairman to GPU, NRC Activities Required to be Performed by the
TMI Licensee (re 09-12-1980)
38 (1979-07-20) NRC Order, Suspends Facility Operating License No. DPR-73 to Operate
TMI-2
39 (1979-10-22) NRC Order (10-18-1979), Operation of EPICOR II to Decontaminate Rad
Waste Water in Auxiliary Building
40 (1979-10-26) NRC Order Amend, Clarifies Order dated 10-26-1979 Concerning the
Operation of EPICOR-II
41 (1981-03-25) OL Amend 13
42 (1980-02-13) NRC Order (2-11-1980), Provides TMI-2 Recovery Technical Specifications
43 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup
44 (1980-02-13) NRC Order (2-11-1980), Provides TMI-2 Recovery Technical Specifications
45 NUREG-0432, TMI-2 Recovery Technical Specifications (1980-02-13)
46 NUREG-0647, Safety Evaluation and Environmental Assessment, TMI-2 (1980-02)
47 (1980-06-13) NRC Order (6-12-1980), Orders Release Krypton-85 from the Reactor
Building Atmosphere by Controlled Purge
48 (1981-06-18) NRC Order, Operation of SDS and EPICOR-II to Process Reactor Building
Sump and RCS Water
49 (1986-01-26) OL Amend 26 (Fully incorporates Recovery Mode Proposed Technical
Specifications)
50 (1980-02-13) NRC Order (2-11-1980), Provides TMI-2 Recovery Technical Specifications
51 (1980-01-14) NRC Review, Proposed Organization Plan (re 12-11-1979)
52 (1980-03-12) OL Amend 10
53 (1980-06-24) OL Amend 11
54 (1981-12-30) OL Amend 18 (concurrent with modification of Order)
55 (1986-01-26) OL Amend 26 (Fully incorporates Recovery Mode Proposed Technical
Specifications)
56 (1988-05-25) OL Amend 30
57 (1989-09-11) OL Amend 35
58 (1993-09-14) OL Amend 45 (Possession Only License)
59 (1993-12-28) OL Amend 48 (TS for PDMS)
60 (1995-06-21) OL Amend 49
61 (1979-07-02) GPU, Planning Study for Containment Entry and Decontamination
62 (1979-08-13) GPU, Entry and Decontamination of Reactor Building, Highlights of a
Planning Study
63 (1979-12-12) GPU, Summary Technical Plan for TMI-2 Decontamination and Defueling
64 SECY-80-175 (Interim criteria for radiological effluents from TMI-2 data gathering and
Maintenance operations, April 4, 1980)
65 SECY-80-175-SRM (Staff Requirements – Discussion of Interim Criteria for Radiation
Release at TMI, April 14, 1980)
66 SECY-80-105 (Release of Krypton Gas in Reactor Building Personnel Air Lock at TMI-2,
February 20, 1980)
67 PNO-TMI-80-03 (Primary Water Leak, February 11, 1980)
68 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup206

69 (1980-02-26) Report of The Governor's Commission on TMI

70 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup

71 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981-03)
72 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup
73 NUREG-0683, SUPP. 1, PEIS, Occupational Radiation Dose (1984-10)
74 NUREG-0683, SUPP. 2, PEIS, Disposal of Accident-Generated Water (1987-06)
75 NUREG-0683, SUPP. 3, PEIS, Post-Defueling Storage and Cleanup (1989-08)
76 NUREG-0683, SUPP. 3, PEIS, Post-Defueling Storage and Cleanup (1989-08)
77 1981 NRC Annual Report (NUREG-0920)
78 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
79 (1986-12-02) GPU, Post-Defueling Monitored Storage Plan
80 (1988-08-16) GPU, PDMS Proposed License Amendment and Safety Analysis Report (woSAR) (re 12-02-1986)
81 NUREG-0683, SUPP. 3, PEIS, Post-Defueling Storage and Cleanup (1989-08)
82 (1993-12-28) Operating License Amendment 48 (TS for PDMS)
83 (1985-10-14) GPU, Recovery Quality Assurance Plan, Rev. 4
84 (1980-01-24) GPU, TMI-2 Radiation Protection Plan, Rev. 1
85 (1984-10-30) GPU, General Project Design Criteria, Rev. 5
86 (1987-04-14) GPU, Special Nuclear Material Accountability Plan
87 (1985-10-17) NRC Exempt, 30.51, 40.61, 70.51(d), 70.53 Regarding Reporting Core
Special Material
88 (1987-04-14) GPU, Special Nuclear Material Accountability Plan
89 (1985-06-06) GPU Procedure, Licensed Operator Requalification Program Unit 2, Rev. 3
90 (1985-05-29) GPU Procedure, Fuel Handling Senior Reactor Operator Training Program
Unit 2, Rev. 1
91 (1984-02-10) GPU Procedure, Requirements for Certification of Candidates for NRC
Operator License, Rev. 0
92 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
93 NUREGCR-1250V2PT3, TMI, A Report To The Commissioners And To The Public(Rogovin Report), Vol. II, Part 3
94 NUREGCR-1250V2PT3, TMI, A Report To The Commissioners And To The Public(Rogovin Report), Vol. II, Part 3
95 NUREGCR-6252, Lessons learned from the Three Mile Island Unit 2 Advisory Panel(1994)
96 (1990-03-00) Final Report, Three Mile Island Unit 2 Safety Advisory Board
97 (1981-12-09) NRC Review, Endorses Concept of Technical Assessment and AdvisoryGroup
98 GEND-INF-036, Task Plan for the U.S. DOE TMI-2 Programs (1982)
99 GEND-INF-036, Task Plan for the U.S. DOE TMI-2 Programs (1982)
100 NUREG-0698, Rev. 0, NRC Plan for Cleanup Operations at TMI-2 (1980)
101 (1979-04-06) NRC (Internal), Staffing Report and Reorganization at TMI
102 (1980-03-31) NRC (Internal), Name Change of NRC TMI Organization to TMI Program
Office103 (1980-04-01) NRC, Change of NRC-TMI Organization
104 NUREG-0698, Rev. 0, NRC Plan for Cleanup Operations at TMI-2 (1980-03)
105 1981 NRC Annual Report (NUREG-0920)
106 NUREG-0698, Rev. 2, NRC Plan for Cleanup Operations at TMI-2 (1984-03)

207

107 1988 NRC Annual Report (NUREG-1145, Vol. 5)108 (1979-05-15) GPU, Interim Report on TMI-2 Accident (re 04-11-1979)
109 NUREG-0432, Proposed Technical Specifications

110 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes

Resulting from TMI-2 (1981-03))

111 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

112 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

113 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes

Resulting from TMI-2 (1981-03))
114 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
115 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981-03))
116 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981-03))
117 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981-03))
118 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman)
119 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981-03)
120 DOE-ID-10276
121 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981-03))
122 NUREG/CR-1250, Vol. 2, Part 2 (Rogovin Report) p.348
123 (1979-10-01) GPU, Water Storage Assessment
124 (1980-02-13) NRC Order (2-11-1980), Provides TMI-2 Recovery Technical
Specifications
125 (1980-04-04) NRC Review, A Further Evaluation of Risk of Recriticality at TMI-2 (re 0306-1980)
126 (1984-04-24) Order Amendment (also ROP-19)
127 (1984-07-17) Order Amendment (also ROP-20)
128 (1984-03-09) GPU Safety Evaluation, Head Removal, Rev. 5
129 (1984-07-17) NRC Safety Evaluation, Reactor Pressure Vessel Head Lift (re -3-09-1984)
130 (1985-04-23) Order Amendment (see ROP-27)
131 (1984-11-08) GPU, Criticality Report, Reactor Coolant System
132 (1985-03-15) NRC Safety Evaluation, Reactor Coolant System Criticality Report (re 1108-1984)
133 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981-03)
134 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman)
135 (1980-11-14) Order Amendment (also ROP-4)
136 GEND-056, TMI-2 Instrumentation and Electrical Program Final Evaluation Report(1986)
137 (1980-10-15) GPU, TMI-2 Quarterly Report No. 5 (Followup Report No. 8), PeriodEnding 09-30-1980
138 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman)
139 (1979-05-15) GPU, Interim Report on TMI-2 Accident (re 04-11-1979)
140 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman)208

141 (1980-01-15) GPU, TMI-2 Quarterly Report No. 2 (Followup Report No. 5), PeriodEnding 01-15-1980

142 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

143 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes

Resulting from TMI-2 (1981-03)

144 (1979-05-15) GPU, Interim Report on TMI-2 Accident

145 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

146 (1979-10-15) GPU, TMI-2 Quarterly Report No. 1 (Followup Report No. 4), PeriodEnding 10-15-1979

147 (1979-05-15) GPU, Interim Report on TMI-2 Accident

148 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

149 (1979-05-15) GPU, Interim Report on TMI-2 Accident
150 (1980-03-04) GPU, BOP Diesel Generators

151 (1979-05-15) GPU, Interim Report on TMI-2 Accident

152 (1980-07-15) GPU, TMI-2 Quarterly Report No. 4 (Followup Report No. 7), PeriodEnding 06-30-1980

153 (1979-04-19) NRC, TMI-2 Ground Water Study.pdf
154 (1981-09-28) TMIPO Weekly

155 (2013-04-00) Exelon, TMI Annual Radiological Groundwater Protection Program Reportfor 2012

156 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes

Resulting from TMI-2 (1981-03)

157 (1980-01-15) GPU, TMI-2 Quarterly Report No. 2 (Followup Report No. 5), PeriodEnding 01-15-1980

158 (1980-01-15) GPU, TMI-2 Quarterly Report No. 2 (Followup Report No. 5), PeriodEnding 01-15-1980

159 (1980-07-15) GPU, TMI-2 Quarterly Report No. 4 (Followup Report No. 7), PeriodEnding 06-30-1980

160 (1980-06-04) GPU, Plans to Discontinue Use of Supplemental Ventilation System Atopof Aux. Building

161 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

162 (1979-05-15) GPU, Interim Report on TMI-2 Accident

163 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

164 (1980-01-15) GPU, TMI-2 Quarterly Report No. 2 (Followup Report No. 5), PeriodEnding 01-15-1980

165 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes

Resulting from TMI-2 (1981-03)

166 (1980-08-11) Order Amendment (also ROP-3)

167 NUREG/CR-1250, Vol. 2, Part 2 (Rogovin Report)
168 (1979-10-01) GPU, Water Storage Assessment

169 (1979-04-11) NRC to FDAA, Request to Locate Railway Cars to Store Low Level Waste

from TMI-2

170 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

171 NUREG-0774, U.S. NRC Annual Report: 1980

209

172 NUREG-0774, U.S. NRC Annual Report: 1980

173 (1980-06-13) NRC Order, Orders Release Krypton-85 from the Reactor Building

Atmosphere by Controlled Purge

174 (1980-06-24) OL Amend 11

175 (1980-10-15) GPU, TMI-2 Quarterly Report No. 5 (Followup Report No. 8), PeriodEnding 09-30-1980

176 NUREG-0774, U.S. NRC Annual Report: 1980

177 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes

Resulting from TMI-2 (1981-03)

178 Langenbach, J. W., Three Mile Island Unit 2 Reactor Building Entry Program,” Nuclear

Technology, Vol. 87, No. 2, pp. 368-382, October 1989

179 1980 NRC Annual Report (NUREG-0774)

180 (1980-03-21) GPU, Initial Reactor Building Entry Program

181 (1980-07-15) GPU, Planned Reactor Building Entry (re 05-16-1980)

182 Langenbach, J. W., Three Mile Island Unit 2 Reactor Building Entry Program,” Nuclear

Technology, Vol. 87, No. 2, pp. 368-382, October 1989

183 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990
184 GEND-INF-011-VOL-3, REACTOR BUILDING BASEMENT RADIONUCLIDE
DISTRIBUTION STUDIES (1983-06)
185 (1981-06-18) NRC Order, Operation of SDS and EPICOR-II to Process Reactor Building
Sump and RCS Water
186 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman)
187 (1980-02-28) NRC, Report of Special Task Force on TMI Cleanup
188 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman)
189 (1979-07-02) GPU, Planning Study for Containment Entry and Decontamination
190 (1981-06-18) NRC Order, Operation of SDS and EPICOR-II to Process Reactor Building
Sump and RCS Water
191 (1981-05-11) GPU System Description, SDS, Rev. 2
192 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman)
193 (1981-02-27) GPU, Cork Seal Contamination Report (re 01-19-1981)
194 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety
Analysis Report
195 (1994-06-02) GPU Response to NRC (Discussions), Actions Taken to Modify Cork Joint
at TMI-2
196 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990
197 GEND-INF-001, QUICK LOOK REPORT ENTRY 1 (1981)
198 GEND-INF-002, QUICK LOOK REPORT ENTRY 2 (1981)
199 NUREG-0683, SUPP. 1, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2,
Occupational Radiation Dose (1984)
200 GEND-002, Vol. 1, Facility Decontamination Technology Workshop, Presentations(1979)
201 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990
202 Merchant, D. J., “Worker Exposures During the Three Mile Island Unit 2 Recovery,”
Nuclear Technology, Vol. 87, No. 4, pp. 1099-1105, December 1989
203 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE(1979-08)
204 1990 NRC Annual Report (NUREG-1145, Vol. 7)
205 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE(1979-08)210

206 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE(1979-08)
207 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE(1979-08)
208 NUREG-0774, U.S. NRC Annual Report: 1980
209 Licensee Event Report LER-89-006; NRC Preliminary Notification PNO-I-89-81
210 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990
211 International Atomic Energy Agency, “Catalogue of Methods, Tools and Techniques for
Recovery from Fuel Damage Events,” IAEA-TECDOC-627, October 1991
212 International Atomic Energy Agency, “Catalogue of Methods, Tools and Techniques for
Recovery from Fuel Damage Events,” IAEA-TECDOC-627, October 1991
213 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE(1979-08)
214 NUREG-0600, Investigation into the 03-28-1979 TMI Accident by the Office of IE(1979-08)
215 NUREG-0640, TMI-2, Radiation Protection Program Report of the Special Panel (197912)
216 NUREG-0683, SUPP. 1, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2,
Occupational Radiation Dose (1984)
217 NUREG-1090, U.S. NRC Annual Report: 1983
218 NUREG-1145, Vol. 2, U.S. NRC Annual Report: 1985
219 NUREG-1145, Vol. 1, U.S. NRC Annual Report: 1984
220 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
221 NUREG-0683, SUPP. 1, PEIS, Occupational Radiation Dose (1984-10)
222 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for
Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991
223 (1989-02-16) NRC IR 88-20 (See Encl. 2, TMI-2 Clean-Up Program Status Update)
224 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
225 EGG, Vol. 1, No. 1, TMI Unit-2 Technical Information and Examination Program,Update (4-15-1980)
226 (1980-03-26) Coordination Agreement, TMI Information and Examination Program(DOE-NRC-GPU-EPRI)
227 GEND-INF-036, Task Plan for the U.S. DOE TMI-2 Programs (1982)
228 (1984-04-16) DOE, Agreement with DOE for Japanese Participation in the TMI-2
Research and Development
229 Tabata, N., F. Masuda, “Utilization of Three Mile Island Unit 2 Research andDevelopment Information in Japan,” Nuclear Technology, Vol. 87, No. 4, pp. 814-823,
December 1989.
230 NUREG-0698, Rev. 0, NRC Plan for Cleanup Operations at TMI-2 (1980)
231 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
232 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman)
233 NRC Preliminary Notification PNO-TMI-82-015
234 GEND-INF-038, Assessment of the TMI-2 Axial Power Shaping Rod Dynamic Test
Results (1983-04)
235 NUREG-0998, U.S. NRC Annual Report: 1982
236 GEND-030, Quick Look Inspection Report on the Insertion of a Camera into the TMI-2
Reactor Vessel (1983)

211

237 (1983-06-27) TMIPO Weekly

238 (1983-05-25) GPU Safety Analysis, Radiation Characterization Under the Reactor Vessel
Head, Rev. 0
239 (1983-05-25) GPU Safety Analysis, Radiation Characterization Under the Reactor Vessel
Head, Rev. 0
240 1983 NRC Annual Report (NUREG-1090)
241 (1983-08-15) TMIPO Weekly
242 (1983-08-22) TMIPO Weekly
243 (1983-09-12) TMIPO Weekly
244 (1983-10-24) TMIPO Weekly
245 (1983-09-12) TMIPO Weekly
246 (1983-07-20) GPU Safety Evaluation, Underhead Characterization, Core Sampling
Addendum
247 (1983-09-19) TMIPO Weekly
248 (1983-10-03) TMIPO Weekly
249 EGG, Vol. 4, No. 1, TMI Unit-2 Technical Information and Examination Program,Update (12-15-1983)
250 EGG, Vol. 4, No. 2, TMI Unit-2 Technical Information and Examination Program,Update (06-15-1984)
251 Marley, A. W., “Sampling and Examination Methods Used for the Three Mile Island Unit
2,” Nuclear Technology, Vol. 87, No. 4, pp. 845-856, December 1989
252 EGG, Vol. 4, No. 1, TMI Unit-2 Technical Information and Examination Program,Update (12-15-1983)
253 (1983-05-26) GPU Safety Evaluation, Underhead Characterization, Core Topography
Addendum
254 NUREG-1145, Vol. 1, U.S. NRC Annual Report: 1984
255 EGG, Vol. 4, No. 2, TMI Unit-2 Technical Information and Examination Program,Update (06-15-1984)
256 NUREG-1145, Vol. 1, U.S. NRC Annual Report: 1984257 (1984-04-02) TMIPO Weekly
258 EGG, Vol. 4, No. 2, TMI Unit-2 Technical Information and Examination Program,Update (06-15-1984)
259 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
260 EGG-TMI-7429, TMI-2 Lower Plenum Video Data Summary (1987)
261 (1985-01-03) GPU Safety Analysis, Measuring Uranium Fuel Material in Lower Reactor
Vessel Region, Rev. 0262 (1985-03-25) TMIPO Weekly
263 (1986-06-11) GPU Safety Evaluation, Core Stratification Sample Acquisition, Rev. 4
264 GEND-INF-092-VOL-1, TMI-2 CORE BORE EXAMINATIONS (1990)
265 NUREG-1145, Vol. 3, U.S. NRC Annual Report: 1986
266 GEND-INF-092-VOL-1, TMI-2 CORE BORE EXAMINATIONS (1990)
267 Marley, A. W., “Sampling and Examination Methods Used for the Three Mile Island Unit
2,” Nuclear Technology, Vol. 87, No. 4, pp. 845-856, December 1989
268 NUREG-1145, Vol. 6, U.S. NRC Annual Report: 1989
269 NUREGCR-6197, TMI-2 Vessel Investigation Project Integration Report (1994-03)
270 NUREG-1145, Vol. 7, U.S. NRC Annual Report: 1990
271 (1992-02-20) NRC Safety Evaluation, Post Defueling Monitored Storage (re 02-20-1992)

272 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for
Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991

212

273 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
274 GEND-001, GEND Planning Report (1980)

275 (1982-03-15) NRC-DOE Revised MOU, Removal and Disposition of Solid Nuclear

Waste from TMI-2 (Signed)

276 (1979-11-21) NRC Policy, Notice of Intent to Prepare Programmatic EIS
277 NUREG-0591, Environmental Assessment, Use of EPICOR-II at TMI-2

278 (1979-10-22) NRC Order (10-18-1979), Operation of EPICOR II to Decontaminate Rad
Waste Water in Auxiliary Building
279 (1980-02-29) City of Lancaster Settlement Agreement
280 (1980-03-26) Coordination Agreement, TMI Information and Examination Program(DOE-NRC-GPU-EPRI)
281 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981)
282 (1981-04-27) NRC Policy, Programmatic Environmental Impact Statement of the Cleanupof TMI-2 (46 FR 24764)
283 (1981-06-18) NRC Order, Operation of SDS and EPICOR-II to Process Reactor Building
Sump and RCS Water
284 NUREG-0796, Operation of the Submerged Demineralizer System at TMI-2 (1981)
285 1981 NRC Annual Report (NUREG-0920)
286 (1980-03-26) Coordination Agreement, TMI Information and Examination Program(DOE-NRC-GPU-EPRI)
287 (1982-03-15) NRC-DOE Revised MOU, Removal and Disposition of Solid Nuclear
Waste from TMI-2 (Signed)
288 EPRI, “TMI-2 Waste Management Experience,” EPRI TR-100640, 1992
289 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation Campaign(1993)
2901987 NRC Annual Report (NUREG-1145, Vol. 4)291 NUREG-0683, SUPP. 2, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2,
Post-Defueling Storage and Cleanup (1987)
292 1988 NRC Annual Report (NUREG-1145, Vol. 5)
293 (1989-09-11) OL Amend 35
294 1991 NRC Annual Report (NUREG-1145, Vol. 8)295 1993 NRC Annual Report (NUREG-1145, Vol. 10)296 (1993-08-05) GPU, Accident Generated Water Disposal Completion Report297 (1993-08-17) GPU, Completion of Accident Generated Water Disposal
298 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992
299 NUREG-0591, Environmental Assessment, Use of EPICOR-II at TMI-2
300 NUREG-0591, Environmental Assessment, Use of EPICOR-II at TMI-2
301 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992
302 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992
303 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992
304 (1987-05-14) GPU System Description, EPICOR II, Rev. 5305 (1989-07-24) GPU System Description, EPICOR II, Rev. 7306 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety
Analysis Report
307 GEND-031, Submerged Demineralizer System Processing of TMI-2 Waste Water (1983-02)

308 GEND-031, Submerged Demineralizer System Processing of TMI-2 Waste Water (1983-02)

309 GEND-031, Submerged Demineralizer System Processing of TMI-2 Waste Water (1983)
310 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

213

311 1988 NRC Annual Report (NUREG-1145, Vol. 5)
312 (1981-10-05) TMIPO Weekly

313 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

314 (1983-10-31) GPU System Description, Processed Water Storage and Recycle System,

Rev. 0315 (1980-01-24) GPU, Processed Water Storage Tank Dikes (re 12-13-1979)
316 (1980-02-01) NRC Review, Design Criteria for Processed Water Storage Tanks (re 01-241980)
317 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety
Analysis Report
318 (1985-05-17) GPU Safety Evaluation, Operation of Internal Indexing Fixture Processing
System, Rev. 3
319 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
320 (1989-05-12) GPU Technical Evaluation, Defueling Water Cleanup System, Rev. 12
321 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
322 Hofstetter, K. J., and B. S. Ausmus, “The Identification and Control of Microorganisms at
Three Mile Island,” Nuclear Technology, Vol. 87, No. 4, pp. 814-823, December 1989.
323 1987 NRC Annual Report (NUREG-1145, Vol. 4)324 (1986-05-06) NRC Review, Addition of a Biocide to the TMI-2 Reactor Coolant System(re 05-05-1986)
325 Hofstetter, K. J., and B. S. Ausmus, “The Identification and Control of Microorganisms at
Three Mile Island,” Nuclear Technology, Vol. 87, No. 4, pp. 814-823, December 1989
326 (1986-07-09) NRC Review, Operation of Filter-Aid Feed System, Use of Diatomaceous
Earth as Feed (re various letter)
327 (1987-04-29) GPU Technical Evaluation, Defueling Water Cleanup System, Rev. 10
328 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
329 (1987-06-11) NRC Review, DWCS Cross-Connect to Reactor Vessel Cleanup System (re05-21-1987)
330 (1989-05-12) GPU Technical Evaluation, Defueling Water Cleanup System, Rev. 12
331 (1986-11-17) GPU Safety Evaluation, TMI-2 Temporary Reactor Vessel Filtration
System, Rev. 3
332 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
333 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
334 (1992-05-12) GPU Technical Evaluation, Processed Water Disposal System, Rev. 4335 NUREG-0683, SUPP. 2, PEIS, Disposal of Accident-Generated Water (1987-06)
336 (1993-08-05) GPU, Accident Generated Water Disposal Completion Report
337 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981)
338 (1981-04-27) NRC Policy, Programmatic Environmental Impact Statement of the Cleanupof TMI-2 (46 FR 24764)
339 (1981-07-15) NRC-DOE MOU, Removal and Disposition of Solid Nuclear Waste from
TMI-2 (Signed)
340 (1982-03-15) NRC-DOE Revised MOU, Removal and Disposition of Solid Nuclear
Waste from TMI-2 (Signed)
341 (1979-05-11) NRC, Design Criteria for Interim Storage or Solid Waste

214

342 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

343 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992
344 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992

345 (1993-06-29) GPU Technical Evaluation, Interim Solid Waste Staging Facility, Rev. 10
346 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

347 (1993-09-13) GPU System Description, Solid Waste Staging Facility, Rev. 6

348 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992
349 (1981-08-17) TMIPO Weekly

350 (1992-07-29) GPU Technical Evaluation, Waste Handling and Packaging Facility, Rev. 8
351 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

352 EPRI-TR-100640, TMI-2 Waste Management Experience (1992)

353 NRC Backgrounder, Transportation of Spent Fuel and Radioactive Materials (2014)
354 NUREGBR-0216, Rev. 2, Radioactive Waste Production, Storage, Disposal (2002)

355 U.S. Code of Federal Regulations, “Licensing Requirements for Land Disposal ofRadioactive Waste,” Part 61, Chapter I, Title 10, “Energy” (10 CFR 61)
356 (1983-05-11) NRC, Final Waste Classification and Waste Form Technical Position Papers
357 (1982-05-21) NRC, Certificate of Compliance 9152, Rev. 1, for Model CNS-1-13C II
Shipping Cask
358 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992
359 (1985-08-14) NRC, Certificate of Compliance 9152, Rev. 8, for Model CNS-1-13C II
Shipping Cask
360 EGG-TMI-6304, Design Analysis, High-Integrity Container for Disposal of EPICOR-II
Prefilter Liners (1983)
361 GEND-045, Disposal Demo of High Integrity Container with an EPICOR II Prefilter from
TMI (1985)
362 (1984-01-24) NRC, Certificate of Compliance 5026, Rev. 10, for Model CNS-14-190
Shipping Cask
363 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation Campaign(1993)
364 (1986-06-11) GPU Safety Evaluation, Canister Handling and Preparation for Shipment
SER
365 INL-CON-12-26246 (CONF), Lessons from TMI Packaging, Transport and Disposition
that Apply to Fukushima (2012)
366 EGG-TMI-6304, Design Analysis, High-Integrity Container for Disposal of EPICOR-II
Prefilter Liners (1983)
367 GEND-045, Disposal Demo of High Integrity Container with an EPICOR II Prefilter from
TMI (1985)
368 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992
369 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992
370 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03)
371 GEND-032, TMI-2 Technical Information and Examination Program Annual Report 1982
(1983-04)
372 GEND-INF-025, Development of a Prototype Gas Sampler for EPICOR II Prefilter Liners(1982-09)
373 GEND-029, Preparations to Ship EPICOR Liners (1983-06)
374 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03)215

375 GEND-052, Hydrogen Control in Handling, Shipping, and Storage of Wet Radioactive

Waste (1986-02)

376 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03)

377 GEND-062, Catalyst Tests for Hydrogen Control in Canisters of Wet Rad Waste (1987-08).pdf

378 (1985-11-05) NRC Safety Evaluation, Defueling Canisters (re various letters)

379 GEND-051, Evaluation of Special Safety Issues Associated with Handling the TMI-2
Core Debris (1985-06).pdf
380 NUREG-0683, SUPP. 3, PEIS-Decontamination and Disposal of Rad Wastes Resulting
from TMI-2 (1989)
381 Ruhter, P. E., W. G. Zurliene, “Radiological Conditions and Experiences in the Three
Mile Island Unit 2 Auxiliary Building,” Nuclear Technology, Vol. 87, No. 2, pp. 361-366,
October 1989
382 NUREG-0683, SUPP. 3, PEIS-Decontamination and Disposal of Rad Wastes Resulting
from TMI-2 (1989)
383 NUREG-0683, SUPP. 1, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2,
Occupational Radiation Dose (1984)
384 NUREG-0683, SUPP. 3, PEIS-Decontamination and Disposal of Rad Wastes Resulting
from TMI-2 (1989)
385 NUREG-0683, SUPP. 1, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2,
Occupational Radiation Dose (1984)
386 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
387 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety
Analysis Report
388 NUREG-0683, SUPP. 3, PEIS-Decontamination and Disposal of Rad Wastes Resulting
from TMI-2 (1989)
389 NUREG-0683, Vol. 1, PEIS-Decontamination and Disposal of Radioactive Wastes
Resulting from TMI-2 (1981)
390 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety
Analysis Report
391 NUREG-0683, SUPP. 3, PEIS-Decontamination and Disposal of Rad Wastes Resulting
from TMI-2 (1989)
392 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety
Analysis Report393 (1993-11-12) GPU, Transition to PDMS
394 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety
Analysis Report395 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety
Analysis Report
396 NUREG-0683, SUPP. 1, PEIS-Decontamination and Disposal of Rad Wastes at TMI-2,
Occupational Radiation Dose (1984)
397 (1989-01-10) GPU Safety Evaluation, Sediment Transfer and Processing Operations, Rev.
4398 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
399 EPRI, “TMI-2 Waste Management Experience,” EPRI-TR-100640, April 1992
400 (1985-03-25) GPU Safety Evaluation, Reactor Building Decontamination and Dose
Reduction Activities for 1985, Supp. 1
401 (1985-09-25) NRC Safety Evaluation, Reactor Building Decontamination and Dose
Reduction for 1985 (re 02-08, 03-25-1985)
402 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990

216

403 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990

404 International Atomic Energy Agency, “Catalogue of methods, tools and techniques forrecovery from fuel damage events,” IAEA-TECDOC-627, October 1991

405 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”

EPRI-NP-6931, September 1990
406 DOE-ID-10276, TMI-2 Lessons Learned by the DOE 1979-1990 (1990-03)
407 Pavelek II, M.D., “Operations and Achievements of Remote Equipment at Three Mile
Island,” Nuclear Technology, Vol. 87, No. 4, pp. 1122-1133, December 1989
408 Giefer, D. L., A. B. Jeffries, “Implementation of Remote Equipment at Three Mile Island
Unit 2,” Nuclear Technology, Vol. 87, No. 3, pp. 641-700, November 1989
409 Giefer, D. L., A. B. Jeffries, “Implementation of Remote Equipment at Three Mile Island
Unit 2,” Nuclear Technology, Vol. 87, No. 3, pp. 641-700, November 1989
410 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for
Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991
411 Pavelek II, M.D., “Operations and Achievements of Remote Equipment at Three Mile
Island,” Nuclear Technology, Vol. 87, No. 4, pp. 1122-1133, December 1989
412 Giefer, D. L., A. B. Jeffries, “Implementation of Remote Equipment at Three Mile Island
Unit 2,” Nuclear Technology, Vol. 87, No. 3, pp. 641-700, November 1989
413 EPRI, “Final TMI-2 Technology Transfer Progress Report,” EPRI-TR-100643, May 1992
414 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10
415 GEND-INF-093, Lower Core Support Assembly Defueling Planning and Tools (1988)
416 (1989-01-31) NRC Safety Evaluation, Criticality Assessment for Plasma Arc Torch to Cut
Vessel Internals (08-11, 12-27-1988)
417 Giefer, D. L., A. B. Jeffries, “Implementation of Remote Equipment at Three Mile Island
Unit 2,” Nuclear Technology, Vol. 87, No. 3, pp. 641-700, November 1989
418 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for
Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991
419 GEND-002, Vol. 1, Facility Decontamination Technology Workshop, Presentations(1979)
420 GEND-034, Gross Decontamination Experiment Report (1983)
421 Lazo, E. N., The Three Mile Island Unit 2 Reactor Building Gross Decontamination
Experiment: Effects on Loose Surface Contamination Levels,” Nuclear Technology, Vol. 87,
No. 2, pp. 407-420, October 1989.
422 NUREGCR-3381, Evaluation of the TMI-2 Reactor Building Decontamination Process(1983)
423 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
424 Ruhter, P.E., and W.G. Zurliene, “Radiological Conditions and Experiences in the Three
Mile Island Unit 2 Auxiliary Building,” Nuclear Technology, Vol. 87, No. 2, pp. 361-366,
October 1989
425 1984 NRC Annual Report (NUREG-1145, Vol. 1)
426 (1984-07-02) TMIPO Weekly
427 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety
Analysis Report
428 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
429 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
430 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990

217

431 (1993-12-28) OL Amend 48 (TS for PDMS)

432 EGG-TMI-7489, TMI-2 Accident Scenario Update (1986)

433 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
434 (1990-02-22) GPU, Defueling Completion Report, Final Submittal

435 Calhoun, G. L., “Fuel Removal Equipment for Three Mile Island,” Nuclear Technology,

Vol. 87, No. 3, pp. 587-594, November 1989

436 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
437 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
438 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
439 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
440 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
441 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
442 (1990-02-22) GPU, Defueling Completion Report, Final Submittal

443 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
444 (1990-02-22) GPU, Defueling Completion Report, Final Submittal

445 (1985-05-07) NRC Safety Evaluation, Lifting, Transfer, and Storage of Plenum Assembly(re various letters)

446 (1985-05-01) GPU Safety Evaluation, Plenum Removal and Transfer, Rev. 3447 (1984-11-27) NRC Safety Evaluation, Initial Plenum Lift (re various letters)
448 (1984-03-09) GPU Safety Evaluation, Head Removal, Rev. 5
449 GEND-044, TMI-2 Reactor Vessel Head Removal (1985)
450 1984 NRC Annual Report (NUREG-1145, Vol. 1)

451 GEND-044, TMI-2 Reactor Vessel Head Removal (1985)

452 (1984-03-09) GPU Safety Evaluation, Head Removal, Rev. 5453 (1984-07-09) TMIPO Weekly

454 GEND-044, TMI-2 Reactor Vessel Head Removal (1985)

455 GEND-044, TMI-2 Reactor Vessel Head Removal (1985)
456 GEND-044, TMI-2 Reactor Vessel Head Removal (1985)

457 GEND-054, TMI-2 Reactor Vessel Plenum Final Lift (1986)

458 (1984-02-17) NRC, Directors Decision Under 10 CFR 2.206 Regarding Pyrophoric

Reaction of Zirconium Fines

459 (1984-02-17) NRC, Directors Decision Under 10 CFR 2.206 Regarding Pyrophoric

Reaction of Zirconium Fines

460 Osgood, N.L., et al., “Review of Radiation Shielding Concerns Associated with the Three

Mile Island Unit 2 Defueling Systems,” Nuclear Technology, Vol. 87, No. 2, pp. 556-56,
October 1989
461 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island,” Nuclear Technology,
Vol. 87, No. 3, pp. 587-594, November 1989
462 GEND-INF-065, TMI-2 Defueling System Design Description (1985)
463 (2014-03-25) Text, TMI-2 35th Anniversary Presentation (GSkillman)
464 Calhoun, G. L., “Fuel Removal Equipment for Three Mile Island,” Nuclear Technology,
Vol. 87, No. 3, pp. 587-594, November 1989
465 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10
466 GEND-INF-073, TMI-2 Defueling Tools Engineering Report (1986-02)
467 (1986-02-06) NRC Review, Use of Hydraulic Cutoff Saw (re 02-03-1986)
468 (1986-02-03) GPU, Plan to Use Hydraulic Cutoff Saw for Sizing Operations in theReactor Vessel
469 (1986-06-04) NRC Safety Evaluation, Heavy Duty Tong Tool, Spade Bucket, HydraulicImpact Chisel (re various letters)
470 (1986-01-22) GPU, Plan to Use Hydraulic Impact Chisel to Separate Fused Material218

471 (1986-04-16) NRC Review, Use of Hydraulic Shredder

472 (1986-03-26) GPU, Plan to Use Hydraulic Shredder to Reduce Core Debris
473 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10474 (1986-10-16) NRC Safety Evaluation, Core Bore Operations (re 09-18-1986)
475 (1986-09-19) GPU, Core Bore Operations

476 GEND-INF-093, Lower Core Support Assembly Defueling Planning and Tools (1988)
477 (1988-03-25) NRC Review, Use of Core Bore Machine for Dismantling Lower Core

Support Assembly (re various letters)

478 (1988-03-25) GPU, Use of Core Bore Machine for Dismantling the Lower Core Support

Assembly Defueling479 (1985-12-31) GPU Safety Evaluation, Core Stratification Sample Acquisition, Rev. 3
480 (1985-09-10) GPU Technical Evaluation, Defueling Canister TER (re 08-15-1985)
481 (1985-11-05) NRC Safety Evaluation, Defueling Canisters (re various letters)

482 Abbate, C.M., and J.W. Craig, “U.S. Nuclear Regulatory Commission Inspection ofTransportation Casks,” Nuclear Technology, Vol. 87, No. 4, pp. 755-758, December 1989.
483 (1985-09-10) GPU Technical Evaluation, Defueling Canister TER (re 08-15-1985)
484 Calhoun, G.L., “Fuel Removal Equipment for Three Mile Island,” Nuclear Technology,
Vol. 87, No. 3, pp. 587-594, November 1989
485 (1985-09-10) GPU Technical Evaluation, Defueling Canister TER (re 08-15-1985)
486 (1985-09-10) GPU Technical Evaluation, Defueling Canister TER (re 08-15-1985)
487 GEND-INF-065, TMI-2 Defueling System Design Description (1985-03)
488 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10489 (1986-01-06) TMIPO Weekly
490 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for
Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991
491 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for
Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991
492 International Atomic Energy Agency, Catalogue of Methods, Tools and Techniques for
Recovery from Fuel Damage Events, IAEA-TECDOC-627, October 1991
493 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10494 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10495 (1986-05-29) GPU Technical Evaluation, Fuel Canister Storage Racks, Rev. 1
496 (1986-09-16) NRC Review, Fuel Canister Storage Racks TER (re various letters)
497 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10498 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10499 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10500 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
501 (1984-12-04) NRC Safety Evaluation, Containment Air Control Envelope Design (re 0722-1983, 10-30-1984)
502 (1984-10-30) GPU Technical Evaluation, Containment Air Control Envelope Design
Criteria (Rev. 4), TER (Rev. 2)
503 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
504 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
505 (1985-10-29) NRC Safety Evaluation, Commencement of TMI-2 Preliminary Defueling
Operations (re 10-24-1985)
506 (1985-10-24) GPU, Commencement of Preliminary Defueling Operations
507 (1985-11-04) TMIPO Weekly
508 (1985-11-12) NRC Safety Evaluation, Early Defueling of the TMI-2 Reactor Vessel (1010-1985)

219

509 (1985-10-10) GPU Safety Evaluation, Early Defueling of the TMI-2 Reactor Vessel, Rev.

4510 (1985-11-18) TMIPO Weekly
511 (1986-01-13) TMIPO Weekly

512 (1986-05-15) GPU Safety Evaluation, Defueling SER, Rev. 10513 (1986-05-20) GPU Safety Evaluation, Defueling SER, Rev. 10 (Correction)

514 (1986-08-14) NRC Review, Defueling Safety Evaluation Report, Rev. 10 (re 05-15, 07-

24, 02-21-1986)

515 (1986-07-23) GPU, Plan to Use Core Bore Tooling as Defueling Tool

516 (1986-07-24) NRC Review, Core Region Bulk Defueling Operations Limited to the Core

Region (re 05-15, 07-23-1986)

517 (1986-10-20) TMIPO Weekly
518 (1986-11-17) TMIPO Weekly

519 (1990-02-22) GPU, Defueling Completion Report, Final Submittal

520 (1990-02-22) GPU, Defueling Completion Report, Final Submittal

521 Porter, L.H., and E.E. Austin, “Disassembly and Defueling of the Three Mile Island

Unit 2 Reactor Vessel Lower Core Support Assembly,” Nuclear Technology, Vol. 87, No. 3,
pp. 587-594, November 1989
522 (1988-04-01) NRC Safety Evaluation, Lower Core Support Assembly Defueling (re 0118-1988)
523 (1988-01-18) GPU Safety Evaluation, Lower Core Support Assembly Defueling, Rev. 2
524 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
525 (1989-04-04) NRC Safety Evaluation, Completion of Upper Core Support Assembly
Defueling (re 9-15-1988)
526 (1988-09-15) GPU Safety Evaluation, Completion of Upper Support Core Assembly
Defueling, Rev. 0527 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
528 (1988-12-01) NRC Safety Evaluation, Lower Core Support and Lower Head Defueling (re06-06-1988)
529 (1988-06-06) GPU Safety Evaluation, Completion of Lower Core Support and Lower
Head Defueling, Rev. 0
530 (1988-06-27) GPU Safety Evaluation, Completion of Lower Core Support and Lower
Head Defueling, Rev. 1
531 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
532 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
533 1990 NRC Annual Report (NUREG-1145, Vol. 7)534 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
535 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
536 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
537 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
538 (1987-08-28) GPU Safety Analysis, Pressurizer Defueling System, Rev. 4539 (1987-10-21) NRC Review, Pressurizer Defueling System, Rev. 4 (re 08-28, 10-13-1987)
540 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990
541 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
542 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
543 (1990-02-22) GPU, Defueling Completion Report, Final Submittal
544 EPRI, “The Cleanup of Three Mile Island Unit 2, A Technical History: 1979 to 1990,”
EPRI-NP-6931, September 1990

220

545 (1990-02-22) GPU, Defueling Completion Report, Final Submittal

546 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation Campaign(1993)

547 (1986-06-20) NRC Safety Evaluation, Canister Handling and Preparation for Shipment

Program (re 02-17, 06-11-1986)

548 (1986-06-11) GPU Safety Evaluation, Canister Handling and Preparation for Shipment SER
549 (1986-06-11) GPU Safety Evaluation, Canister Handling and Preparation for Shipment SER
550 PNO-TMI-87-02

551 DOE-ID-10400, Historical Summary of the TMI-2 Core Debris Transportation Campaign(1993-03)

552 INL-CON-12-26246 (CONF), Lessons from TMI Packaging, Transport and Disposition

that Apply to Fukushima (2012)
553 INL-CON-12-26246 (CONF), Lessons from TMI Packaging, Transport and Disposition
that Apply to Fukushima (2012-07)
554 DOE-ID-10739, Annual Radiological Environmental Monitoring Program Report for the
TMI-2, ISFS Installation (2002-02)555 (1986-12-02) GPU, Post-Defueling Monitored Storage Plan
556 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-201992)
557 (1987-03-11) GPU, Environmental Evaluation for TMI-2 PDMS
558 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
559 (1986-12-02) GPU, Post-Defueling Monitored Storage Plan

560 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
561 (1993-09-14) Operating License Amendment 45 (Possession Only License)

562 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
563 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety

Analysis Report564 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-201992)

565 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
566 (1992-01-23) GPU, Addition of Borosilicate Glass to the Reactor Vessel

567 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
568 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
569 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)

570 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
571 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
572 (1992-02-20) NRC Safety Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
573 (1990-02-22) GPU, Defueling Completion Report, Final Submittal

574 (1988-08-16) GPU, PDMS Proposed License Amendment and Safety Analysis Report(wo SAR) (re 12-02-1986)

575 (1990-02-22) GPU, Defueling Completion Report, Final Submittal

576 (1992-12-18) GPU, Criticality Safety Analysis for the Reactor Vessel
577 (1993-02-01) GPU, Post-Defueling Survey Report, Reactor Vessel

578 (1993-01-15) GPU, Proposed List of PDMS Requirements and Commitments (re 12-07-1992)
579 (1993-05-19) NRC Review, List of PDMS Requirements and Commitments (re 01-15-1993)
580 (1993-09-14) Operating License Amendment 45 (Possession Only License)

581 (1988-08-16) GPU, PDMS Proposed License Amendment and Safety Analysis Report(wo SAR) (re 12-02-1986)
582 (1992-02-20) NRC Safety Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
583 (1994-11-04) NRC Review, Post-Defueling Survey Reports (PNL)

221

584 (1985-10-17) NRC Exempt, 30.51, 40.61, 70.51(d), 70.53 Regarding Reporting Core

Special Material

585 (1993-02-01) GPU, Post-Defueling Survey Report, Reactor Vessel

586 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety

Analysis Report587 (2013-03-22) GPU, TMI-2 Update 9 of the Post-Defueling Monitored Storage Safety
Analysis Report588 (1988-09-30) GPU, Post-Defueling Survey Report, Plenum and Letdown Coolers (letter)
589 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
590 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
591 (1992-02-20) NRC Technical Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
592 (1993-12-28) Operating License Amendment 48 (TS for PDMS)

593 (1992-02-20) NRC Safety Evaluation, Post Defueling Monitored Storage (re 02-20-1992)
594 (1993-12-28) Operating License Amendment 48 (TS for PDMS)

595 (1993-09-14) Operating License Amendment 45 (Possession Only License)
596 (1993-12-28) Operating License Amendment 48 (TS for PDMS)
597 1988 NRC Annual Report (NUREG-1145, Vol. 5)

598 NUREG-0683, SUPP. 3, PEIS, Post-Defueling Storage and Cleanup (1989-08)
599 1990 NRC Annual Report (NUREG-1145, Vol. 7)
600 (1990-04-20) GPU, Transition to Facility Mode 3

601 (1990-04-26) NRC Review, Transitioning Plant from Mode 2 to Mode 3 on 4-27-1990 (re04-12-1990)
602 1991 NRC Annual Report (NUREG-1145, Vol. 8)603 1992 NRC Annual Report (NUREG-1145, Vol. 9)604 1993 NRC Annual Report (NUREG-1145, Vol. 10)605 (1993-12-28) Operating License Amendment 48 (TS for PDMS)
606 (1992-10-26) GPU, Discontinuation of TMI-2 Restart Punchlist
607 (1993-02-01) GPU, Post-Defueling Survey Report, Reactor Vessel
608 1993 NRC Annual Report (NUREG-1145, Vol. 10)609 (1993-09-14) Operating License Amendment 45 (Possession Only License)
610 1993 NRC Annual Report (NUREG-1145, Vol. 10)611 (1993-07-06) NRC Safety Analysis, Reactor Vessel Criticality (re various letter)
612 (1993-09-14) Operating License Amendment 45 (Possession Only License)
613 (1993-09-14) Operating License Amendment 45 (Possession Only License)

614 (1993-11-12) GPU, Completed PDMS Requirements and Commitments (re various letters)
615 (1993-11-12) GPU, Transition to PDMS

616 (1993-12-28) Operating License Amendment 48 (TS for PDMS)

617 (1993-12-28) Operating License Amendment 48 (TS for PDMS)
618 (1993-12-28) Operating License Amendment 48 (TS for PDMS)

619 Berkeley CA: University of California Press, 2004. Dr. Walker joined the NRC History

Program in 1979 and retired as the NRC historian in 2010.

620 University Park, PA: Penn State Press, 2004. Dr. Baratta was appointed the Associate

Chief Administrative Judge (Technical) on the NRC Atomic Safety and Licensing Board
Panel in 2003 and retired in 2014.

222

A loaded defueling canister being remotely lifted from the shipping
cask at the INEL receiving facility. The 12.5-foot long canister was
filled with water, placed in a storage module, and transferred to thestorage pool for temporary storage at the INEL.

CONVERSIONS
Radiation Dose
1 mrem (1 millirem, 10-3) = *10 microsieverts (10 µSv, 10-5)
100 mrem = *1 millisievert (1 MSv)
1 rem = *10 mSv
100 rem = *1 Sv
Radioactive Concentration
27 picocuries (27 pCi, 2.7 × 10-11) = *1 becquerel (1 Bq)
1 millicurie (1 mCi, 0.001) = *37 megabecquerels (37 MBq, 3.7 × 107)
1 curie (1 Ci) = *37 gigabecquerels (37 GBq, 3.7 × 1010)
Radiation Absorbed Energy
1 roentgen = *0.877 rad = *0.00877 Gy
100 rad = *1 gray (Gy)
Length1 inch (in) = *2.54 centimeters (cm)
1 foot (ft) = 0.3048 meter (m)Volume and Weight
1 gallon (gal) = 3.7854 liters (l)1 pound (lb) = 0.4536 kilograms (kg)
1 ton (U.S.) = *2000 lb = 907.1847 kg
Pressure
1 pound per square inch (psi) = 6.8948 kilopascals (kPa)
1 atmosphere (atm) = *101.325 kPa
Temperature
Degrees Celsius (°C) = 5/9 × (°F - 32)
Degrees Fahrenheit (°F) = (9/5 × °C) + 32
* Exact conversion factors

NUREG/KM-0001, Supplement 1
June 2016