~~'W'~~u~@~u~(Q)~~~u(Q) u[}=[]~ ~ffi\~CC[}=[] ~~17u[}={](R1~~ [MTID[L~D~[L~~[QJ~~!l~~~~D[Q)~~u[gSW(Q)~~DCG~ (Q)[F ~~~[p)~~u~(Q)~ ffi\~[Q) ~~~(Q)~~~uwu~~u<C»'\f'\fo@@ @'\f O[/'jJ~~@@ito@[/'jJ (ID[/'jJtdJ~[/'jJ'\f@Ii'@@IMl@[/'jJit

lUJa i:~a~l\JJ(g~@~1!? ~@@l\JJ~~~@[[W(C@mn)[ITfi)O@~o@!ruAvailable from
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555

Available from
National Technical Information Service
Springfield, Virginia 22161

NUREG-0600

INVESTIGATION INTO.THE MARCH 28, 1979
THREE MILE ISLAND ACCIDENT
BY
OFFICE OF INSPECTION AND ENFORCEMENT

Investigative Report No. 50-320/79-10

Manuscript Completed: July 1979
Date Published: August 1979

Office of Inspection and Enforcement
U. S. Nuclear Regulatory Commission
Washington, D.C. 20555

FOREWORD
On March 28, 1979, the Three Mile Island Unit 2 Nuclear Power Plant
experienced the most severe accident in U.S. commercial nuclear power
plant operating history. This report sets forth the facts concerning
the events of the accident determined as a result of an investigationby the NRC Office of Inspection and Enforcement.
The IE investigation,
which is based on the information available at this time, is limited
to two aspects of the accident:
1. Those related operational actions by the licensee during the period
from before the initiating event until approximately 8:00 p.m.,
March 28, when primary coolant flow was re-established by startinga reactor coolant pump, and2.

Thosesteps taken by the licensee to control the release of radioactive material to the off-site environs, and to implement his
emergency plan during the period from the initiation of the eventto midnight, March 30.

These investigation periods were selected because they include the licensee.actions which most significantly affected the accident sequence and its
results.
The results of the IE investigation supports the reported population dosefrom the accident, developed by an an hoc dose assessment group, which
included representatives of various cognizant Federal agencies.
In itsreport dated May 10, 1979, this group concluded that, "Based on thecurrent assessment . . . the off-site collective dose associated withthe radioactive material released during the period of March 28 to April 7,
1979, represents minimal risks (that is, a very small number) of additional
health effects to the off-site population." At the same time, the IE
investigation identifies several-inadequacies in the inplant radiation
protection activities of the licensee and criticizes the measurements ofoff-site radiation levels made.by the licensee.
In spite of these identified flaws, no glaring inconsistencies have been found which would
significantly alter the conclusions reached by the ad hoc group.
The IE investigation also substantiates earlier conclusions concerning theunderlying-causes of the accident and those factors that contributed toi
its severity.
Inadequacies in six major areas have been confirmed:
1. Equipment performance (failu'res and maloperation).
2.

Transient and accident analyses.

3.

Operator training and performance.

2

4.

Equipment and system design.

5.

Information flow, particularly during the early hours of theaccident.

6.

Implementation of emergency planning.

Perhaps the most disturbing result of the IE investigation is confirmationof earlier conclusions that the Three Mile Island Unit 2 accident could
have been prevented, in spite of the inadequacies listed above. The
design of the plant, the equipment that was installed, the various accident
and transient analyses, and the emergency procedures were adequate to have
prevented the serious consequences of the accident, if they had been permitted
to function or be carried out as planned.
For example,. had the operators
allowed the emergency core cooling system to perform its intended function,
damage to the core would most likely have been prevented.
There are otherexamples set forth in the report where, had a particular operator. action
been taken, the consequences of the accident could have been significantly
mitigated.
On the other hand, had certain equipment been designed differently,
it too, could have prevented or reduced the consequences of the accident.
The results of the investigation make it difficult, to fault only the actionsof the operating staff. There is considerable evidence of a "mind set,"
not only by TMI operators but by operators at other plants as well, that
overfilling the reactor coolant system (making the-system solid) was tobe avoided at almost any cost. Undue attention by the TMI operators to
avoiding a solid system led them to ignore other procedural instructions
and indications that the core was not being properly cooled.
Without this"mind set" they might well have acted to preclude or better mitigate theaccident.
Subsequent actions have been required by NRC to retrain all
licensed operators in an effort to preclude recurrence.
Upgraded procedural
instructions have also been required.
It is clear that substantial effort is needed, by both the NRC and theindustry, to assure that these lessons learned concerning the TMI accident
are implemented at other facilities.
Within the NRC, early action has beentaken to inform other nuclear power plant licensees of the circumstances.
surrounding the Three Mile Island accident and to require immediate implementation of compensatory measures to prevent occurrence of similar accidents
elsewhere.
In addition, a special Lessons Learned Task Force was establishedin the NRC Office of Nuclear Reactor Regulation.
This group has studied theThree Mile Island accident and has issued a report (NUREG-0578) containing
short-term recommendations that will significantly improve continued safe
operation of licensed nuclear power plants.
The IE investigation adds
further emphasis to the need for such plant and procedural modifications.

3

Because they have the benefit of hindsight, most retrospective investigations like this tend to emphasize areas where people and equipment
did not perform as desired. The IE investigation team made a concerted
effort to evaluate the reasoning processes of the people who wereoperating the plant during the course of the accident.
The report contains the team's conclusion as to whether or not the operating staff's
actions were appropriate in light of the training and factual information
available to-them at the time they had to make decisions as to whatcourse of action to follow.
Further study is clearly needed with respect to the contributions ofvarious other organizations that influence the operation of nuclear power
plants, including designers, reviewers, builders, vendors and regulatory
agencies.
These various studies are now underway; most notably the
Presidentially appointed Kemeny Commission, as well as a wide-ranging
internal NRC study under Mr. Mitchell Rogovin.
A full assessment of all
the underlying causes of the Three Mile Island accident must await
completion of these studies.
The findings of this IE investigation will be the Subject of appropriate
enforcement action in accordance with the Commission's regulations(Part 2, Title 10, CFR).

picto

loJr

Director
Office of Inspection
and Enforcement

U.S. NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
Region I
Report No.

50-320/79-10

Docket No.

50-320License No.

DPR-73

Licensee:

Metropolitan Edison Company
P.O. Box 542
Reading, Pennsylvania 19640

Facility Name:

Category C

Three Mile Island Nuclear Station Unit-2

Investigation at:

Middletown,

Investigation conducted:

Pennsylvania

March 28-July 31,

1979

Investigators
Operational Aspects TeamRadiological Aspects Team,S-."

D.M. Collins

ate'

D.E.

ate

CreswellDate

A nv,/ ,ý

D.R. Hunter

D.C.

ick
"
S Kirklfatrick

RýJ

'MlFa
rs h

T.T. Martin

Donaldson/
T.H. Essig

Date

2

Date

72 f7-9

'oa

Dbte'

L'L,(/'Jackson'Dat6

O.C. Shackleton

gate

Da te

G.P.

uas

Vate

Date

A.F. Gibson

Date

Team Leaders"R.D. Martin
Approved by:lion I,A Dputy Director,
Investigation Director

PREFACE
This Report was prepared by the investigation team assembled by the
Office of Inspection and Enforcement of the U.S. Nuclear Regulatory Commission
to investigate certainaspects of the accident at the Three Mile Island
Unit 2 facility on March 28, 1979.
The investigation by the Office of Inspection and Enforcement had two
basic goals:oTo establish,

in a comprehensive manner,the facts concerning theevents of Three Mile Island accident during the period investigated.
oTo evaluate the performance of the Licensee in association with
the Three Mile Island accident as a basis for corrective action
or enforement action as appropriate.

This report contains a factual recounting, to the extent it

has beenpossible to establish at this time, of the significant operational andradiological events that transpired during the early hours and days of this
accident.

The report further provides an analysis of the actions of thelicensee staff in light of those facts when compared to the NRC requirements
to which they are subject.
This report is critical,

in several instances,of the actions taken by

the licensee staff during that accident and in the first few days thereafter.
The perspective we ask the reader to retain is to be aware that reports
yet to be issued by other investigatory bodies may address evaluation ofother organizations associated with the Three Mile Island facility.

i

These

organizations include the designers, reviewers,

builders,

vendors,andregulatory agencies that are or were involved with this facility.
This report is not a definitive study of every facet of the Three Mile
Island accident.

Nor is it

an engineering evaluation of the accident.

Those evaluations are being done by other Offices within this agency aswell as by other organizations.

The reports from other ongoing investigations

and studies will be needed to fully understand the causes of this accident
and the appropriate actions to be taken as a result of that understanding.

ii

ACKNOWLEDGEMENT
The investigation team wishes to express its appreciation to the many
individuals of the Headquarters and Regional staffs who provided assistanceto theteam throughout this investigation.

Moreover,the assistance extended

by other Offices within NRC that provided personnel,

review and other forms

of assistance is also appreciated.

however,

Most noteworthy,

is the support

that was provided by the Word Processing Unit in Region I who, with help
from other regional and headquarters personnel,

were faced with the formidable

task of handling not only this report, but the transcription of over 200
interviews.

J. M. Allan
R. D. Martin

A. F. Gibson

S. Creswell
A. N. Fasano

D. M. Collins

D. R. Hunter

T. H. Essig

D. C. Kirkpatrick

L. L. Jackson

R. J.

0. C. Shackleton

J.

Marsh

T. T. Martin

iii

D. E. Donaldson

G. P. Yuhas

TABLE OF CONTENTS
PrefaceAcknowledgementsiii

Table of Contentsiv

Introduction

V

Summary1

Details I Table of Contents

I-i

Details I - Operational Aspects

I-I

Details II Table of Contents

II-i

Details II

II-1

-

Radiological Aspects

Enclosure 1 - Memoranda Establishing Investigation.

El-lEnclosure 2 - Interview List

El-2

iv

Introductiona.

Location of Plant
The Three Mile Island Nuclear Station is located on Three Mile Islandin the Susquehanna River, approximately 10 miles southwest of Harrisburg,
Pennsylvania.
It is in Londonderry Township of Dauphin County.--Goldsboro is
located 1.2 miles west of Three Mile Island and Middletown is located 3 miles
north of the facility.
Three Mile Island is one of the largest of a group of islands in the Susquehanna River and is situated about 900 feet from the east bank.

Itis elon-gated parallel to the flow of the river, with its longer axis oriented approximately due north and south.
An access bridge connects State Highway Route 441 with the north end of
the island.

A wood access bridge connects the south end of the island with

Route 441.

Route 441 is a two-lane black-topped road which runs north andsouth, parallel to Three Mile Island on the east bank of the Susquehanna River.
Route 441 is approximately 2,000 feet from the reactor buildings at the closest
point.
On the east bank of the river there is a one-track railroad line adjacentand parallel to Route 441.

On the west bank of the Susquehanna River, a dis-

tance of approximately 1 1/4 miles,

there is a multi-tract railroad line anda two-lane, black-topped road.
b.

Facility Description

Two pressurized water reactors (designated as Unit 1 and Unit 2) arelocated on Three Mile Island and, together, constitute the Three Mile Island
Nuclear Station.

The licensee of the two units is the Metropolitan Edison

Company of Reading,Pennsylvania.

Unit 1 was licensed to begin operation on April 19, 1974, at a rated power
not to exceed 2,535 megawatts thermal.

It achieved criticality on June 5, 1974,v

and went into commercial operation on September 2, 1974.

Unit 2 was licensed

to begin operation on February 8, 1978, at a rated power of 2,772 megawatts thermal.

It achieved criticality on March 28, 1978,

tion on December 30, 1978.

and went into commercial opera-

The nuclear steam system supplier for Unit 2 was

Babcock and Wilcox and the architect-engineer was Burns and Roe.
c.

Purpose/Objectives of Investigation
On the morning of March 28, 1979, Unit 2 of the Three Mile Island Nuclear

Station experienced an operational transient which evolved into an emergency
situation requiring activation of the Three Mile Island Site Emergency Plan.
The scope of this investigation was described by the Acting Director,
Office of Inspection and Enforcement,
April 20 and June 8, 1979.
this report.
1.

in memoranda to the Commission dated

These memoranda are included in Enclosure 1 to

The stated objectives of this investigation were:

To gather facts concerning the incident, its cause, effect(s),andthe licensee's response; and2.

To evaluate these facts as a basis for corrective or enforcement

action, as appropriate.
d.

Scope of Investigation
The operational portion of this investigation examined selected preaccident

conditions for a period prior to March 28, 1979, and operational events during
the period 0400 hrs to 2000 hrs on March 28, 1979.

Areas investigated included:

operations staff qualifications and training, operating events,

operator actions

and management actions.
The radiological portion of this investigation examined selected preaccident conditions and emergency response activities during the period 0400 hrs on
March 28, 1979, to 2400 hrs on March 30,

1979.

Areas investigated included:

detection and classification of the emergency; emergency organization activa4tion; notifications; environmental assessment and protective actions; effluenvimonitoring and corrective actions; and in-plant radiological assessment andprotective actions.
This investigation did not include an engineering evaluation of the accident.
e.

Such evaluations are being prepared by others.
Sources of Information
Investigators examined records,

matters with involved personnel,

conducted interviews of and discussed

and made observations of facilities and equip-ment.

In the radiological area, scarcity of records caused greater relianceon other sources of information, especially interviews.
In making conclusions,the investigators have attempted to reconcile conflicts by using the best
available information.

A listing of those interviews is included in Enclosure 2

to this report.
f.

Noncompliance

Several matters which are under consideration as potential items of noncompliance were identified during this investigation and are listed in AppendicesI-B and II-F of this report.

These matters are under review and will be handled
through the enforcement channels of the Office of Inspection and Enforcement.

vii

TMI INVESTIGATIONSUMMARY OF OPERATIONAL ASPECTS
PREACCIDENT CONDITIONSOn March 28, 1979, during the first 5 hours of the 11-7 shift (2300
hours, March 27, 1979 to 0700 hours, March 28, 1979), the Three Mile Island
Unit 2 facility was operating at approximately 97% power with the Integrated
Control System in full automatic. Normal makeup, reactor coolant pump seal
injection, and letdown were in operation. The reactor coolant system (RCS)
boron concentration was 1026 ppm, with the pressurizer spray throttled open
and pressurizer heaters energized to equalize RCS and pressurizer boron
concentration.
All system and core physics surveillance testing required by Technical
Specifications was current,' and the facility was in one identified Limiting
Condition for Operation ACTION statement of those'specifications.
The
borated water storage tank to spent fuel pool isolation valve (DH-V157) was
open to permit BWST recirculation. This ACTION statement time limit would
have expired at 1500 hours on March 29, 1979. The RCS leakage as calculatedby the licensee was within Technical Specification limits, with the identified
leakage being that accumulated in the reactor coolant drain tank (RCDT).
This leakage was identified as being from the electromatic relief valve(EMOV) and/or one or both pressurizer code safety valves. A review by theinvestigators of the RCS leakage procedure showed the procedure to be inerror, and the facility was actually operating with an unidentified leakage
in excess of Technical Specification limits.

The assumption-by the licensee staff of EMOV and safety valve leakage
appears reasonable based on EMOV and safety valve discharge pipe temperatures.
The leakage was sufficient to cause the temperature of the discharge pipes(%2000F)

to be in excess of specified limits (1300F)

in plant procedures,and the facility had been operating contrary to these limits for an extended
period of time.
The operating staff on duty in the control room during the 11-7 shiftof March 28, 1979, was in accordance with TechnicalSpecifications.

Eachlicensed staff member was satisfactorily current with regard to the requirements of the licensee's requalification program for licensed operators.
Shift conditions as regards personnel behavior and activities were normal,
and the only unusual log book entries indicated an increase in the amountof water being added to the makeup tank when compared to that of previous
shifts.

No evidence was found showing any maintenance on safety related

components was in progress at.that time.
The shift foreman and two auxiliary operators were engaged in transferring resin from condensate polisher tank No.
tank.

7 to the resin regeneration

This activity was a carryover from the previous shift, and a totalof about 11 hours had already been expended attempting to complete thistransfer.

Difficulty was being encountered in this transfer and was attributedby licensee staff to a resin blockage in the transfer line, and.shift
personnel were involved in an attempt to clear it.
TURBINE TRIP AND PLANT RESPONSE
At this time; and probably as a result of actions taken to clear the
resin blockage in the transfer line, the plant experienced a total loss offeedwater initiated by a loss of condensate flow with an almost simultan-

2eous trip of the main turbine at 04:00:37. All emergency feedwater pumps
started as designed, the reactor continued to operate at full power inaccordance with its protection system design, and RCS temperature andpressure increased for approximately 8 seconds. The EMOV opened as designed
at its setpoint of 2255 psig. The reactor automatically tripped when thehigh RCS pressure trip setpoint was reached.
With the trip of the reactor, the RCS experienced an expected coolant
contraction, loss of inventory, cooldown, and the attendant reduction in
RCS pressure. The EMOV failed to close when its closure setpoint wasreached about 13 seconds later. This failure was not recognized by theoperating staff for more than 2 hours. At approximately one minute after
the start of the accident, the pressurizer level stopped decreasing andbegan to rise. This rise continued until approximately 6 minutes after theaccident, when the level went off scale indicating that the pressurizer wascompletely filled with water (a "solid" pressurizer). Operator efforts tocontrolthe level of the pressurizer, included~throttling high-pressure
injection which initiated automatically at 2 minutes, and increasing letdown
flow to the maximum extent possible.

These efforts were largely unsuccessful. RCS pressure began to increase moderately as the pressurizer went.solid. At this time, the RCS temperature was also increasing. This increase
would also contribute to the pressure rise since saturation conditions now
existed in the loops.
This indication of high pressurizer level Was caused by voids, either
discrete or distributed, that formed in the reactor coolant system coupledwith the open EMOV. The open EMOV vented the steam space of the pressurizer
causing a rapid insurge into the pressurizer. At 8 minutes into the accident,
an operator, upon seeing continuing'low once-through-steam-generator (OTSG)
levels and decreasing OTSG pressures, searched his panels for the cause. -3

The block valves on the emergency feedwater headers were found closed.
Upon opening the block valves, which admitted emergency feedwater to the
OTSG, a rapid cooldown of the RCS and corresponding RCS pressure decrease
occurred.
At approximately 14 minutes, the Reactor Coolant Drain Tank (RCDT).
rupture disc burst, discharging water and steam into the reactor building
and causing a further increase in building pressure (which had begun withlifting of the RCDT relief valve) from the continued discharge of reactor
coolant into the open RCDT.
.Reactor coolant inventory loss continued with the RCS under saturation
conditi.ons. This continuing loss was caused by the-discharge of coolant to
the reactor building through the open EMOV, coupled with sustained low
rates of coolant injection. Reactor coolant pump (RCP) apparent output flowrate decreased while they continued to be operated outside pressure operating
limits. The staff secured the RCPs in the B loop at 74 minutes, and the A
loop RCPs at 101 minutes, with the staff expecting that natural circulation
would occur. However, the plant parameters were outside defined pressure/temperature limits for natural convection. After the trip of the B loop
RCPs, the operating staff believed that the B OTSG had developed a secondaryto- containment leak and this generator was isolated.
By this time, after receiving initial early notification-of the trip,
plant management had become aware of the worsening situation and called for
key individuals to come to the site.

RCS pressure continued to decrease

and temperature increased as a result of the failure of natural circulation
to develop because both loops were vapor bound; the lack of any otheradequate heat sink being available to accommodate the core decay heat;.thecontinuing unrecognized reactor coolant loss through the EMOV; and the4throttled high-pressure-injection flow.

This throttled high pressure

injection flow was justified by the operators because of the apparently
satisfactory but actually misunderstood pressurizer level.

The RCS pressure

decreased to a low point of 660 psig at 2 hours and 1i9 minutes when theleaking EMOV was diagnosed,and closed, and RCS pressure began to increase.
The pressure increase appears to be associated with the heating of theremaining RCS contents by core decay heat, a change in RCS inventory madeby modest increases in high pressure injection, and a zirconium-water
reaction.

The operating staff believed up to this time, that no substantive

inventory loss from the RCS had occurred.
misunderstood pressurizer level,

This belief was based on the

without regard for the low RCS pressure.

At approximately 2-1/2 hours into the accident, substantial fractionsof the reactor core were uncovered and had experienced sustained hightemperatures.

This condition would be expected to result in fuel damage,substantial releases of core fission products,

and hydrogen generation.

The magnitude of these conditions were not recognized by the plant staff.
With the arrival of senior management, the.declaration of a General
Emergency,

an emergency command team was established with the Station

Manager asEmergency Director.

Additional unsuccessfulattempts to establish

sustained forced cooling with one or more RCPs were made.

The plant staff

was faced with, the following conditions:oan inability to achieve forced or natural circulation in the RCSohigh incore and loop temperatures,

which were considered to be

too high to be realisticoan apparent inability to collapse the voids in the loops despite
the increased system pressure and the high pressure injection
flow which by now was increased.

5

Unsuccessful efforts to collapse the voids in the loops were continued
for approximately 2 hours at about 2000 psig.

•Because of a growing staff

concern over the ability of the EMOV block valve to continue to remainfunctional under a high use rate, the decision was made to reduce systempressure and float the core flood tanks on the RCS as an assurance ofadequate core coverage and as a preliminary step in initiating the use of
the decay heat removal system.
approximately one hour,

This depressurization was accomplished inusing the EMOV flow path to the reactor building,and the RCS was held in this low-pressure condition for the next 5 hours.
The failure of the core flood tanks to inject a substantial fractionof their volume was interpreted as an indication that the core was covered.
The piping from these tanks contains large loop seals that prevent themfrom being used effectively for the purpose of ensuring satisfactory core
coverage.

The design function of these tanks is to supply water to thevessel in the event of a large break LOCA, which did not occur during this
accident.
The extended period of low pressure appears to have assisted in therelease of hydrogen gas from the RCS.

This hydrogen resulted from a significant

metal-water reaction with the zirconium fuel cladding.

Some of this gas

burned in the reactor building at about 10 hours after the accident producing
a rapid pressure spike (28 psig) in the containment.
received relatively little

This pressure spike

attention from the majority of the plant staff,

with many of them being unaware that it

had taken place.

However,itisalso possible that the release of this noncondensible gas from the RCS
contributed to the later apparent success of the staff in collapsing thevoids in at least one of the reactor loops (A loop,
isconnected).

to which the pressurizer.

This increasing success in establishing what appeared to besome degree of natural circulation,

despite continuing high temperatures in6

portions of the system, led the plant staff to conclude that they had
achieved a reasonably stable set of conditions.
The Station Manager left the Emergency Control Center (Unit 2 controlroom) at approximately 1400 hrs (10 hours after the start of the accident).
He was absent for approximately, 2-1/2 hours to attend a meeting with the
Lt. Governor of Pennsylvania.

During his absence, another staff member

acted as Emergency Director under-additional.'general

guidelines dictated by

the Station Manager.
After the return of the Station Manager,

the plant staff was directedby corporate management to take the RCS to high pressure to collapse theremaining voids.

During this final repressurization,

the decision wasreached to attempt another start of a reactor coolant pump to establish
forced circulation.
28,

This was successfully achieved at 1950 hours,

on March

15 hours 50 minutes after the start of the accident.

SHIFT CREW ACTIONSWhen the original, turbine trip occurred,

the shift crew on-duty took

the appropriate initial response actions indicated for a combined turbine.
trip/reactor trip initiated as a result of a loss of main feedwater.
actions included control manipulations,-verifications

Theseof automatic actions,and notifications of appropriate personnel.
The misunderstood pressurizer level,

and the conditioning instilled in'the operators by their training and experience to avoid a solid pressurizer
condition at all times caused the shift crew, and those who responded earlyin the transient to provide assistance,
were contrary to procedural

to take a series of actions that

requirements and/or to prudent operating practices.

7

These actions led directly to a sufficient loss of reactor coolant inventory
to cause core damage.
For a period of as much as 2 hours, a reversal ofthese actions could have prevented the extensive core damage that occurred,
although some degree of damage may still have been experienced.
Among the actions taken that contributed to the accidentwere:o

The throttling of high pressure injection to a minimum, averaging
only 70 gpm net input to the RCS for the first 3 1/2 hours of theaccident;o-The continued operation of RCPs at RCS pressures below.the procedural
requirement which requires they be tripped which maintained awater supply at the pressurizer surge line and resulted inma
sustained higher mass flow rate through the EMOV;o

The failure to isolate the EMOV after the RCS pressure continued
to fall., the RCDT rupture disc had blown, and the reactor building
sump pump operation indicated a large discharge of water from thebuilding;o

The failure to establish the conditions for natural circulation
when the combined RCS pressure and temperature conditions wereoutside the procedural requirements.

Other actions were taken by the shift crew members during the earlyhours of the accident that did not directly contribute to the accident, butwould have severely impaired the response of safety-related equipment hadother plant conditions developed.

Among these were:8

0Disabling the automatic start features of the emergency diesel
generators making them unavailable for rapid starting in theevent of a power failure during the course of the accident.
(This condition was noted after plant management arrived, but was
only partially corrected by restoring control room start capability.)oIsolating the core flood tanks early in the event so they were
not available to discharge their contents into the vessel.
The
RCS pressure dropped to within 60 psi of the core flood tankpressure just before the EMOV was isolated, and the core floodtanks apparently had been isolated prior to this time based on
the continuing belief of the plant staff that no loss in inventory
had occurred.

EMERGENCY STAFF ACTIONSWith the arrival of plant management and the establishment of an
Emergency Organization,

one of the initial actions taken was to'increasehigh-pressure-injection flow rates to allow ECCS to function as it would if
operators were not present.
of the core.

This apparently resulted in eventual reflooding

The actions taken over the next 13 hours that eventually led

to the successful operation of one of the reactor coolant pumps (RCP-IA)
have been summarized earlier.
The Emergency Director (the Station Manager) formed a management teamfor overall conduct of the emergency by assigning specific individuals
responsibility for different functional areas.

A system of periodic meetings

with that team for status review and decision-making was established.
Decisions were ultimately made by the-Emergency Director following consultati~onwith that team, with input from offsite management.

Team members thenconveyed the decision to the plant staff for implementation.

9

Within hours of its formation, the management team found themselves
between two desired stable states of forced circulation, being-unable touse the RCPs at high pressure or the decay heat removal system at low
pressure.

Natural circulation was similarly unattainable because of *vapor

binding in the loops.
Their efforts throughout the course of the accident were to movetoward one or the'other of these desired conditions before the borated
water storage tanks (BWST)

inventory was exhausted and they would be forced

to use the water on the reactor building fl'oor.
Plant parameter information was utilized by the team in planningcourses of action to move toward either of these desired conditions, withseveral notable exceptions:0

The persistent disbelief of high temperature data from incore
thermocouples and system RTDs.

This was based on the rationalethat the former were not safety-grade equipment, while the latter
were outside the calibrated range of the detectors;o'The failure to recognize the fact that a full pressurizer did notprovide assurance of core coverage;o

The failure to recognize the significance andpursue evidence ofthe pressure spike that occurred in the reactor'building;o

The failure to recognize the fact that small decreases in coreflood tank level did not provide assurance-of core coverage;

10.For each'of the above areas,

the investigation did-not attempt toconclude whether the course of subsequent events would have or could havebeen altered.
In the case of the high temperatures,

acceptance of the temperature

data as valid might'have prompted a higher high-pressure-injection'flowrate and a reluctance to subsequently depressurize the plant to use thecore flood tanks.

However,

had that occurred,itcannot be ascertained

whether RCP operation could ever have been established in light of the thenunrecognized inventory of noncondensibles (hydrogen)that was in the loops

and reactor vessel as a result of the zirconium-water reaction.
Similarly, the general recognition of the pressure spike in the reactor
building might have led the Station Manager to conclude that conditions,were not sufficiently stable to justify leaving the site.

His remaining on

the site might have altered the subsequent actions taken, or the timing of
those actions.
OFFSITETECHNICAL SUPPORTThe provision of substantive'technicalsupport to the management team

directing emergency actions on operational matters suffered primarily as aresult-of communication difficulties.
oThis was evidenced in three ways:

Information (both data and plans) transmitted to offsite support,
which had been hurriedly mobilized,
Thus,

suffered from time delays.

the offsite groups were dealing with historical and limited

data.

11

o

The individuals who had to provide data to offsite groups had
concurrent duties pertaining to the management of the emergency.
The emergency duties always took precedence as would be appropriate.

o

The physical communications facilities were inadequate to handle
the volume of information requests and transmittals that thiskind of accident required.

The investigation has concluded that these communication problems are
related to the misconception that the envelope of the analyzed major accidents
for this facility are the limiting events.

The duration of these analyzed

events are projected to occur in a relatively short time frame.

The provisionof the mechanisms needed to mobilize and communicate with substantial
offsite technical support on a real-time basis as an accident progresses
had, therefore,

not been warranted as a part of emergency planning.

12-

SUMMARY OF RADIOLOGICAL ASPECTS
Health physics operations at TMI Unit 2 were routine prior to 0400 hrs
March 28,

1979.

The normal complement of radiation protection staff was on site.

The emergency plan and implementing procedures had been rehearsed andevaluated during seven drills conducted in the past year.

Most plant

personnel had received training in their emergency plan duties.

However,some workers who would comprise Emergency Repair Party Teams and Radiological Monitoring Teams had not received adequate training in use of emergency
survey instrumentation and in radiation protection procedures.

Routineretraining of radiation/chemistry technicians was not up to date.

Whileradiation protection training of the-plant staff had been sufficient tomaintain personnel radiation exposures within limits during normal operations(when radiation levels were low),it

had not prepared workers to cope withthe high radiation levels that would soon exist inside the Unit*2 auxiliary
and fuel handling buildings.
Less than half'of the portable radiation survey instruments wereoperational.
Several installed area radiation monitors and airborne radioactivity monitors, which were not essential

for normal.operations,would have been useful during the emergency,

were out of service for repair.

but

Fifty self-contained breathing devices and 175 half and full-facerespirators were on site.
available.

Large quantities of protective clothing wereAll essential communications systems were operational..

Threeemergency environmental monitoring kits containing survey and counting
instrumentation and personnel monitoring devices were in place.

One of thethree kits was later found to have an inoperable instrument for fieldmeasurement of radioactive iodine.
Environmental air samplers were operating
at eight offsite locations,and environmental TLDs were in position at 20

locations.

Tanks in the liquid radwaste system were filled to about 60% of

capacity.

Valves were aligned to pump the reactor building sump to theauxiliary building sump tank.

Ventilation exhaust from fuel handling andauxiliary buildings 'was through high-efficiency filters and charcoal adsorbers.

13

At 0400 hrs,

the Unit 2 turbine and reactor tripped.

was a reactor building sump high water level alarm.

At 0411 hrs,

By 0415 hrs,

there

the reactor

coolant pressure had dropped from 2435 psig at the time of the reactor trip toapproximately 1.275 psig.

This pressure was below the setpoinit for emergency

core cooling system initiation (1600 psig).

At 0415 hrs,

rise of 1.4 psig inside the reactor building.

there was a pressure

A site emergency should have'been declared, based on these indications and criteria in the Site Emergency
Plan.

However, because the drop in reactor pressure was believed to beunder control,

and'the reactor building pressure increase was considered tobe slight, and because there was no evidence of.a release of radioactivity
from the station, an emergency was not declared.

Subsequently,

there wereseveral radiation monitor alarms indicative of an emergency situation, but
no emergency was declared.
At 0622 hrs the first radiation monitor response to cladding failure
occurred.

Radiation levels continued to increase and-a site emergency was

declared at 0655 hrs based on these alarms.
The emergency organization was promptly activated followin-g the declaration of a Site Emergency,

The Station Manager arrived in'the Unit 2 control

room at 0705 hrs and relieved the Shift Supervisor-as Emergency Director.
Initially, the emergency organization approximated the planned organization
described in the TMI Emergency Plan.

An exception was that Repair Parties

were assembled and controlled by both the Emergency Control Center (ECC)
the Unit 2 control room and the Emergency Control Station (ECS)
Unit 1 health physics/ chemistry lab area.
Implementing.Procedures,in theAccording to Emergency Plan

the Repair Party was to assemble only at the ECS,

under the direction of the Supervisor of Maintenance,and coordinatedthrough the Supervisor of Radiation Protection.
Offsite consequences were assessed by performing dose rate calculations.

Because of errors in these calculations,the dose rates initiallypredicted (10 and 40 rem/hr at Goldsboro) were-higher than actual dose
rates.

Radiation measurementsby survey teams revealed actual doses were14

in

low (less than 0.001 rem/hr at Goldsboro).

Offsite agencies and support

groups were notified of the Site Emergency by telephone.

At 0724 hrs, a General

Emergency was declared based on radiation levels inside the reactor building.
Again,

offsite agencies and groups were phoned.

Following the turbine trip, about 8000 gallons of reactor coolant werepumped from the reactor buildi.ng sump to the auxiliary building sump tank.
This transfer was terminated at 0438 hrs and was not resumed.

The auxiliarybuilding sump-tank overflowed to.the auxiliary building sump,causing watercontaininga relatively low concentration of. radioactivity to back up
through floor drains onto the fuel handling building and auxiliary building
floors.

Following fuel damage,the concentration of radioactivity in thereactor coolant increased by several orders of magnitude.

A flow of thishighly contaminated reactor coolant was maintained through the makeup and.
purification system for several days following the accident.

This flow was

the principal pathway by which radioactivity was transferred from thedamaged reactor core to the auxiliary and fuel handling buildings, andultimately to the environment.
Gases evolving from reactor coolant Jn the makeup and purification
system were collected in the waste gas system.
were of little radiologicalfollowing fuel damage,

Small leaks in these systems

significance during normal operation.

However,radioactive gas leaks caused very high concentra-

tions of airborne radioactivity inside the auxiliary and fuel handling_buildings and resulted in much higher than normal environmental releases
via ventilation exhausts from these buildings.

Radiation levels inthe

vicinity of some makeup and purification system components exceeded thelimits of the licensee's measurement capability (i.e.,R/hr).

greater than 1000

High radiation levels inside the Unit 2 auxiliary building caused

full scale-readings on several station effluent monitors. A full scalereading for the plant vent gas monitor is equal to 2.8 E-2 pCi/cc of xenon133.

The particulate and iodine monitors were off-scale due to interference

from the large amounts of.radioactive noble gases.

15

As hazards from direct radiation and airborne radioactive material
developed rapidly in the Unit 2 auxiliary and fuel handling buildings,

thelicensee attempted to control the in-plant radiation protection program inaccordance with Emergency Plan Implementing Procedures.
The ECS was established in the Unit 1 chemistry and health physics
area according to Emergency Plan Implementing Procedures.

The activities'of Radiological Monitoring and Repair Party Teams were to be directed by
the Supervisor,

Radiation Protection (ECS Director) from this location.

A Unit 2 reactor coolant sample was collected in the nuclear sample
room at'about 0845 hrs without the knowledge of the ECS Director.

The nuclear

sample room and primary chemistry laboratory are located in Unit 1 near the
ECS.

Collection of this sample resulted in an"immediate increase in radiation

and airborne radioactivity'levels at the ECS,
to the Unit 2 control room.

causing the ECS to be evacuated

The high radiation levels disabled the Unit 1

counting room, which contained the only instrument on site capable ofperforming gamma isotopic analyses.
The individuals who collected and analyzed this sample did not takeappropriate precautions.

Sample containers were handled directly without

use of remote tools or shielding.to reduce hand exposure, extremity dosimetry
was not worn on hands,

and no air sample was collected.

If the samplelines had been properly recirculated or flushed prior to sampling, theindividuals would likely have received significantly greater radiation-exposure.
Shortly after the ECS was established in the Unit 2 control room,
airborne radioactivity began to increase, as measured by the control room
incoming air monitor.

At about 1017 hrs, personnel were requested to put on

respiratory protective devices (particulate filter masks),'based on an
alarm of the control room air monitor and an air sample that indicated highgross beta radioactivity.

Control room personnel remained in respiratory

protective devices for about six hours.

Isotopic analysis of an air sample

16

would have likely shown that respirators were unnecessary; but, the isotopic
analysis capability had been lost.
The ECS was relocated to the Unit 1 control room at 1012 hrs,

maintaining

the responsibility for 'coordination of the onsite and offsite environmental
survey teams but relinquishing control of the inplant radiation protection
program to the Supervisor, Radiation Protection and Chemistry who remainedin the Unit 2 control room.
At .1110 hrs, all nonessential personnel were evacuated from the site.
Evacuees were surveyed for contamination at the assembly areas, exit gates,
and at an area established offsite at the 500 kV substation.

Severalindividuals were found to be contaminated.
During the evacuation, the auxiliary building access control point was
relocated from outside the auxiliary building entrance to the Unit 2 control
room because of increasing airborne radioactivity in the auxiliary building.
This -left no positive control over entries into the auxiliary building.
Although the Supervisor,

Radiation Protection and Chemistry briefed

some individuals and, at times,

directed radiation/chemistry technicians toaccompany Repair Party Teams.into the auxiliary building, several entries
were made without his knowledge.

These entries were made into areas ofhigh airborne radioactivity and whole-body exposure rates in excess of
100 R/hr.

In at least one instance,

survey intruments were not used.

Two

individuals who 'entered the auxiliary building received a whole-body doseof radiation in excess of a regulatory limit; others became contaminatedand received unnecessary doses.

At times, high-range pocket dosimeters

could not be located and were not worn.

Items of protective clothing suchas hoods, when not readily available, were not worn, resulting in several
instances of head contamination.
worn.

Extremity monitoring devices were not

Air sampling was not performed in the auxiliary building,in the ano

where workers were exposed during the period from about 0900 hrs on March 28
through midnight on, March 30.

Appropriate respiratory protective devices

were not always worn.

17

In at least two instances,individuals failed to leave high radiation

areas in the auxiliary building when their radiation survey instruments
failed or deflected full scale.

In one of these instances,

this resulted

in a whole-body exposure in excess of regulatory limits.
An example which indicates that the radiation protection and chemistrystaff was not adequately trained to cope with the hazards which existed
occurred during the sampling of reactor coolant on March 29.
was at the direction of the Supervisor,

The sampling

Radiation Protection and Chemistry,

and was performed by a chemistry foreman with assistance from a radiation
protection foreman.
Although the need for a reactor coolant sample was known for severalhours,
less than one hour was devoted to planning and preparation for taking it.
The two foremen entered the nuclear sample room to survey the area and makethe valve lineup to recirculate the sample at about 1600 hrs March 29.

Theywore protective clothing, full-face respirators with iodine adorbing cartridges,
and high-range pocket dosimeters.

No air samples were taken to evaluate

airborne radioactivity, and no one was assigned to time their exposure.
Neither remote valve operating nor sample handling tools were used.
The exposure rate to operate a sample valve was remembered to be
90 R/hr.

About 300 ml of reactor coolant was collected in a hand-heldpolyethlyene bottle.

A 100 ml aliquot of this sample in a graduated cylinder

produced a radiation exposurp rate of 400 R/hr at a distance of I foot.
A second sample was collected in a beaker.
placed in a small vial.

A portion was removed and

The remainder was titrated with hydrochloric acidin preparation for a boron analysis.

Another chemistry foreman, wearing aparticulate filter respirator and no extremity monitoring,

performed theboron analysis.
After the operation,inated.

the three individuals were found to be contam-

Decontamination was incomplete,

and residual contamination remained18

on-small areas of one individual's skin for over 30 days.

The licensee

reported one chemistry foreman received a whole-body exposure, as measured
by his TLD,

in excess of NRC limits.

The NRC evaluated the handling of

this sample and concluded that, in addition to the reported whole-bodydose,doses to the hands,

forearms,

and a small area on the skin of thehead of the chemistry foreman and to the hands and forearms of the radiation
protection foreman exceeded NRC limits.
Prior to and during the emergency,
onsite personnel dosimetry program.

the licensee performed his own

No one individual was assigned program-

matic responsibility for this program.

During the incident, some radiation/

chemistry technicians processed their own TLD badges.

Beginning March 29,

one radiation/chemistry technician, who had not operated the system in overa year, worked without procedures for over 40 continuous hours.
The Emergency Plan Implementing Procedures did not address sustainedin plant radiation hazards.

The licensee's radiation protection and chemistrystaff was not adequately trained to deal with this degree of hazard, andsupplies of equipment and instruments were not sufficient to minimize doseto the workers.
During March 28-30, the licensee's land-based onsite and offsite
monitoring teams made about 500 direct radiation measurements.

These

measurements weremade primarily to confirm the predicted location of thenoble gas effluent plume and to determine the dose rate produced by theplume.

The rate of release of radioactivity (source term) from the station

was periodically calculated based on dose rate measurements in the plumeand meteorological conditions existing at the time of measurement.

The

calculated source terms were used to predictdose rates in other areas whenmeteorological conditions changed.

Monitoring team survey results werealso used to assess the need for protective actions and to supplement
thermoluminescent dosimeter (TLD)
dose.

results in assessment of accumulated

These dosimeters were in place at 15 locations within 3 miles and at

5 locations ranging from 9 to 15 miles from the site prior to the accident.

19

These TLDs were used to perform an after the fact assessment of direct
radiation doses to the public.
In general,the licensee's onsite and offsite survey teams performed

surveys in appropriate areas at appropriate times.

However, during a fiveand one-half hour period from 1700 hrs to 2238 hrs on March 28 and a two-hour
period from 0340 to 0540 on March 29,
the plume.

no offsite surveys were performed in

Both of these periods of time were within the interval when the

majority of the noble gases were released and when a plume was well defined
because of sufficient wind speed and almost constant direction.
S.Radiaon-level-s-on March28 ,-withat 1548 hrs on Pennsylvania Rt 441,

the exception of-50 mR/hr measured-

about 1500 feet south of the North Gate,

were not above background until 2238 hrs when a radiation level of 13 mR/hr wasmeasured near Kunkel School (5.6 mi NNW).

Several other radiation levels

above background were noted in this general area prior to midnight.

However,the 13 mR/hr value was the highest one measured, until.30.mR/hr was measured
in Goldsboro at 0600 hrs on March 29.

Radiation levels during the remainder ofMarch 29 were generally less than 1 mR/hr, with the maximum noted as 3
mR/hr in Royalton at 2355 hrs.
were also generally below•

Offsite radiation levels measured on March 30

mR/hr, with the maximum noted as 15 mR/hr at

location S-11 (one mile south of the plant) at 0906 hrs.

The highest radiationlevel measured onsite (outside of the plant) during March.28-30 was 365
mR/hr (p, y) at 2325 hrs on March 28 at a location 1,000 feet northwest of theUnit 2 station vent.
Although not a pre-planned consideration in the licensee's Emergency
Plan, helicopter-based survey teams were used to track the noble gas plume.
Up to three helicopters chartered by the licensee were.used during March
28-30, with the majority of surveys taking place on March 30.
measurements were made by the helicopter teams.

Over 300 radiation

The highest measurements

reported were 3000 mR/hr (P,y) at 15 feet above the plant vent at 1410 hrs on
March 29 and 1200 mR/hr (P,y) at 130 feet above the Unit. 2 reactor building
at 0801 hrs on March 30.

20A total of 57 samples were collected on March 28-30 for the purpose of
assessing radioiodine concentrations in the environment.

Air samples

collected on March 28, which were counted in the field with a single channel
analyzer having a sodium iodine detector,

indicated that "radioiodine" waspresent at offsite locations with concentrations ranging up to 2.3 E-7
pCi/cc.

The "radioiodine" was subsequently shown to be xenon-133 andxenon-135 at 1400 hrs on March 28 (the time at which the first gamma spectrometry of one of these samples was completed by the Pennsylvania Bureau ofRadiological

Health).

Forty of the fifty-seven samples collected wereanalyzed by gamma spectrometry,and no radioiodine was detected.

Results of samples from certain portions of the licensee's routine
radiological environmental monitoring program collected on March 29 (TLDs,
radioiodine in water, and radioiodine in air) were available around mid-day
on March 30.

The sample results confirmed that the offsite radiologicalimpact was no worse than earlier estimates made using data gathered by themonitoring teams.

These data supported the conclusion that radioactive

noble gases released to the atmosphere were the principal cause of exposurefor individuals in the plant environs.

21

DETAILS I
OPERATIONAL ASPECTSTABLE OF CONTENTS

Page

1.

TABLES LIST................................................

I-vii

FIGURE LIST .................................................

I-viiiBACKGROUND ...................................................

I-I-I

1.1

General ...............................................

I-I-I

1.2

Plant.Status Prior to Turbine Trip .....................

I-I-I

1.3

1.2.1

General ........................................

1.2.2

RCS Boron Concentration.........................

1-1-2

1.2.3

RCS Leakage .....................................

1-1-2

1.2.4

EMOV Leakage.............................. .....

1-1-5

1.2.5

Condensate System................................

1-1-7

1.2.6

General Plant Parameters ........................

1-1-7

.

Status of Surveillance and Inspection..................
1.3.1

General Status ..................................

1.3.2

Status of Surveillance on Selected Engineered Safety
Features (ES)System Components .................

I-I-.I

1-1-9
1-1-9
I-1-101.3.2.1 Review of Procedures ....................

I-1-101.3.2.2 Emergency Feedwater Valve Test ..........

1-1-11

1.3.2.3 Review of EFW Procedure Change ...........

1-1-14

1.3.3

Status of Core Physics Surveillance.............

1-1-16

1.3.4

Status of Surveillance Results andRecords Review................................

I-i

1-1-19

PAGE

1.3.5
1.4

1.3.4.1 Performance and Review...................

1-1-19

1.3.4.2 Record Retention ........................

1-1-21

Inspection of Surveillance Activities ...........

1-1-21

Operations Staff .......................................
1.4.1

1-1-22

Operations Personnel Training and1-1-22

Qualifications ...............................
1.4.1.1

General Plant Training Program .........

1-1-22

1.4.1.2

Auxiliary Operator Training .............

1-1-23

1.4.1.3

CRO Training ............................

1-1-241.4.1.4

Training for Individuals with Previous
1-1-25Experience ...........................
1.4.1.5

Requalification Program ................

1-1-251.4.1.6

Training On Significant Events ..........

1-1-271.4.1.7

Annual Examinations ....................

1-1-281.4.1.8

Individuals Involved With the Early
1-1-29Part of the Accident ..................
1.4.2

Shift Composition Compared toTechnical Specifications .......................

1-1-32

1.4.3

Shift Staffing During Prior Trips ............

1-1-33

1.4.4

Physical and Emotional Condition
of Operational Staff ............................

1.4.5
1.4.6
1.5

Possibility of "Anniversary" Celebrationon Site ........................................

1-1-34

Possibility of Plant Sabotage ...................

1-1-35..........

1-1-371.5.1

Summary of Selected Licensee Reports. ............

1-1-38

1.5.2

Review of Selected Prior Trips ...............

Significant Operating Events.................

1.5.2.1
1.5.2.2...

1-1-44

Electromatic Relief Valve (EMOV) ......

1-1-44Previous Event Involving thePrevious Loss of FeedwaterTransient............................

1.6

1-1-33

Corrective Maintenance Practices ......................
l-ii

1-1-441-1-47

1.6.1

General .........................................

1.6.2

Selected Component Maintenance History

1.6.3
2.

1-1-47

Immediately Prior to March 28, 1979 ............

1-1-49

1.6.2.1

Main Feedwater System .................

1-1-49

1.6.2.2

Electromatic Relief Valve.............

1-1-49

1.6.2.2.1

Valve Control Revisions ....

1-1-49

1.6.2.2.2

EMOV Locations andHandling ..................

1-1-51

1.6.2.3

EMOV Block Valve......................

1-1-52

1.6.2.4

Emergency Feedwater Pumps .............

1-1-53

1.6.2.5

Pressurizer Heaters...................

1-1-53

1.6.2.6

Waste Tank Rupture Discs..............

1-1-54

1.6.2.7

Emergency Feedwater Valves ............

1-1-54

1.6.2.8

Pressurizer Safety Valves .............

1-1-56

References ......................................

1-1-56OPERATOR ACTIONS DURING ACCIDENT ...........................

1-2-1

2.1

General ...............................................

1-2-1

2.2

Operator Actions Following Loss of Main Feedwater .....

1-2-1

2.3

Operator Actions Following Turbine Trip ...............

1-2-2

2.4

2.3.1

Immediate Operator Actions ......................

1-2-2

2.3.2

Followup Operator Actions............... I.......

1-2-5

2.3.3

Additional Operator Actions .....................

Operator Action Following Reactor Trip.. ...............

.2.5 Operator Actions Following ES Actuation ...............
2.6

.

1-2-8
1-2-101-2-14

Operator Actions Concerning the Operation of theMakeup System (Following Initial ES Actuation).

1-2-21

2.7

Operator Actions Concerning Condensate System .........

1-2-27

2.8

Operator Actions Concerning Reactor
Building Sump Pumps .................................

2.9

Operator Actions Concerning the B OTSG Tube Leak ......

2.10

Operator Actions Concerning Natural Circulation Decay Heat Removal OTSGS .........................

2.11

1-2-29
1-2-301-2-32

Operator Actions Concerning the Auxiliary
Building Sump Pumps .......................... .......

I-iii

1-2-34

2.12

Operator Actions Concerning Logkeeping ................

1-2-36

2.13

Operator Actions Concerning EMOV ......................

1-2-41

2.14

General Location of Operation Personnel ...............

1-2-422.15

Effect of Training on Operator Actions................

1-2-472.15.1Pressurizer Level Versus Pressure Control .....

1-2-472.15.2

Recognition of an Open EMOV. ..................

1-2-482.15.3

Bypassing of the Engineered Safety Features
1-2-49Actuation System ............................
2.15.4
2.16

3.

Nonlicensed Operator Actions During Accident ..........

1-2-502.16.1

General .......................................

1-2-502.16.2

Activities Prior to Trip ......................

1-2-502.16.3

Activities During Accident ....................

1-2-51

MANAGEMENT ACTIONS DURING ACCIDENT ..........
3.1

General ...............................................

3.2

Onsite Operational Activities of..............

Senior Management ....................................
3.3

3.4

4.

Provision of-Technical Support ........................

.

1-3-1
1-3-1
1-3-1
1-3-16

3.3.1

Onsite Technical Support.......................

1-3-163.3.2

Babcock 'and Wilcox............................

1-3-193.3.3

Corporate Technical Staff .......................

1-3-35

3.3.4

Burns and Roe (Architect Engineer) ..............

1-3-38

Licensee/NRC Interface................................

.1-3-39

3.4.1

Notification of NRC.............................

1-3-393.4.2

Communications Between NRC and the Licensee ....

1-3-393.4.3

Impact of NRC on Licensee Actions ..............

1-3-40SIGNIFICANT EVENTS DURING ACCIDENT .........................
4.1

1-2-49

Shutdown of Reactor Coolant Pumps .............

Turbine Trip..........................................

1-4-1
1-4-1

4.1.1

Apparent Cause ..................................

1-4-1

4.1.2

Plant Air Systems ..............................

1-4-1

4.1.3

Post Accident Testing ...........................

1-4-2

4.1.4

Pump Interlocks and Wiring Errors ..............

1-4-4

I-iv

4.1.5
4.2

4.3

Evaluation ..................................

1-4-4

Closed Emergency Feedwater Block Valves ............

1-4-5

4.2.1

Sequence ....................................

1-4-5

4.2.2

Cause ....................

4.2.3

Evaluation ..................................

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

14-6
1-4-10HPI Flow Adjustment.............................

1-4-11

4.3.1

Operator Actions Regarding Adjustment of HPI

1-4-11

4.3.2

Analysis .....................................

1-4-144.3.2.1

HPI Pump History ...................

1-4-144.3.2.2

Effective Flow Rates to the RCS ....

1-4-18

4.4

Shutdown of the Reactor Coolant Pumps ..............

4.5

Attempted and Successful Restarts of theReactor Coolant Pumps.

1-4-241-4-27I 0 0

4.6

Closing of the Core Flood-Tank Isolat ion Valves....

1-4-284.7

Use of Core Flood Tanks ..............

1-4-30

4.8

Reactor Vessel Level Indication ...................

1-4-31

4.9

Sump Pump Operation .......

;.......................

1-4-32

4.9.1

Sequence .........

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

1-4-32

4.9.2

Discharge Point ............................

1-4-32

Electromatic Relief Valve Failure .................

1-4-334.10.1

History ....

1-4-334.10.2

Evaluation ................................

1-4-344.10.3Temperature History .......................

1-4-344.10.3.1

Relief Valve Tailpipe Layout .....

1-4-34

Thermocouple Locations ...........
......
Emergency Director Leaves Site .......

1-4-354.11.1

Sequence ..................................

1-4-434.11.2Precautions Taken .........................

1-4-43

4.10. .

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

.

°. .

..........

4.10.3.2

4.11•. . . . ..

1-4-43

4.12

RCS Average Temperature Indication ................

1-4-44

4.13

Plant Computer Records.............................

1-4-454.13.1

Alarm Status Printout,
05:13:59 to 06:48:08, 3/28/79 .............

4.13.2

1-4-46

Alarm Status Printout,
18:48:59 to 19:10:29,

3/28/79 .............

I-v

1-4-46

4.13.3

Utilities Typer Output,
00:00:00 to 03:24:24, and 20:08:22 to 20:12:*,4.13.44.14

4.15

3/28/79 ........................................

1-4-46

Analog Trend Recorder Number 2, 3/28/79 ........

1-4-47

Pressure Spike in Containment ...........................

1-4-474.14.1

Sequence......................................

1-4-474.14.2

Response of Personnel to Pressure Spike .......

1-4-484.14.3

Notification of NRC ............................

1-4-49High Temperatures and Core Response ...................

1-4-514.15.1

General .......................................

1-4-514.15.2

SRM and IRM Responses .........................

1-4-544.15.3

Incore Thermocouple Response .................

1-4-544.15.4

Incore Flux Detector Response..................

4.15.5OTSG and Primary Temperature-4-55

Recorder Display .............................

.1-4-644.15.5.1

Hot-Leg Temperature Behavior .........

.,-4-644.15.5.2

RCP Suction Temperature Behavior

1--65
-....

4.16

Early Void Formation .................

•......

1-4-72

4.17

Diesel Generator Inoperability...........................

1-4-74

4.18

Reactor Coolant Makeup PumpStarting Problems .........

1-4-76

4.19

Operational Usefulness of Computer Alarm...........

Status Printout.....................................

1-4-784.19.1Timeliness ....................................

1-4-794.19.2Inaccuracies ...................................

1-4-794.19.3Misinformation .................................

1-4-80

References.............................. ...................

Appendix IA

Operational Sequence of Events.................

Appendix IB

Operational Aspects Potential Items of
Noncompliance ...............................

I-vi

1-5-1
IA-l
IB-1

DETAILS I
TABLES LISTTABLESECTIONDESCRIPTION

I.1-1

Recalculation of RCS Leakage Rate

1.2.2

1.2-1

Location of Key Personnel Interviewed

2.14

1.3-1

Notification and Arrival Time of Key Personnel

3.2

1.4-1

HPI Pump History

4.3.2.1

1.4-2

Makeup Flow Through the MU-; V17 Valves and,,RCP Pump Seals

4.3.2.2

1.4-3

Average Makeup Flows Based on BWST Levels

4.3.2.2

1.4-4

EMOV and Safety Valve Tail Pipe Temperatures

4.10.3.2

1.4-5

Steam Generator and Primary System Temperatures

4.15.5.2

1.4-6

Pressurizer Level Response During First 4 Minutes

4.16

1.4-7

Computer Output Errors

4.19.3I -vi.i

DETAILS IFIGURE LIST
FIGUREDESCRIPTIONSECTION

1.4-1

Makeup & Decay Pump Status

4.3.2.1

1.4-2

Water Injected From BWST

4.3.2.2

1.4-3

Emergency Feedwater Pumps - Status

4.3.2.2

1.4-4

Heat Up/Cool Down Curve

4.15.1

1.4-5

Selected Incore T/C Histories from Computer
Printouts

4.15.3

1.4-6

4.15.3

1.4-7

4.15.3

1.4-8

4.15.3

1.4-9

Incore Thermocouple Data from Computer at
0834-0836 hrs on March 28, 1979

4.15.3

1.4-10

Incore Thermocouple Data from Computer at
0846-0847 hrs on March 28, 1979

4.15.3

1.4-11

Readings Obtained from Incore Thermocouples
at Approximately 4-5 hrs After Start of Accident

4.15.3

I.4-12A

Selected Primary System Temperatures Following
Accident

4.15.5.2

I.4-12B

4.15.5.2

1.4-12C

4.15.5.2

I.4-12D

4.15.5.2

I-viii1. BACKGROUND
1.1

GENERAL
During the course of this investigation,

it appeared that several practices

or conditions could have contributed to the initiation or the consequences of
the accident.

This section of the report addresses only these topics and is

not intended to be a complete background description of the Three Mile Island
facility.
1.2

Plant Status Prior to Turbine Trip

1.2.1

GeneralThe Three Mile Island Unit 2 (TMI-2)

was operating at approximately 97%

power (916 MWe) with the the Integrated Control System (ICS)

in full automaticin accordance with operating procedure (2105-1.4, "Integrated Control System";
Ref. 58).
The normal reactor coolant system (RCS) makeup and reactor
coolant pump seal water injection was established with makeup pump (MUP)lB inservice and 70 gpm letdown flow in accordance with normal procedure; (2104-1.2,
Makeup and Purification System; Ref. 41).
1.2.2

RCS Boron Concentration
The RCS boron concentration was 1026 ppm (0330 hrs March 28,1979) witha gross radioactivity concentration of 0.397 uCi/ml in accordance with normal
operating procedure.
(2304-W], Borated Water Source, 2304-3DI, RCS-Chemistryand 2304-3D2,

RCS Specific Activity;

valve (RC-Vl)

control was in manual and the spray throttled open with the pres-

Ref. 59, 60, 61).

The pressurizer spraysurizer heaters energized to equalize the pressurizer and RCS boron concentration in accordance with normal procedure.
Ref. 62).

(2103-1.3,

Pressurizer Operation;

This was done to counteract the boron increase in the pressurizer

caused by leakage from the pressurizer.

I-I-I

RCS Leakage

1.2.3

One or more of the pressurizer relief valves (EMOV and code safety valves)at approximately 6

were leaking into the reactor coolant drain tank (RCDT)
gpm,

as determined by completed surveillance procedure (2301-3DI,

RCS Inventory,

This continuous leakage caused the boron

dated March 28, 1979; Ref. 63).

concentration to continuously increase in the pressurizer.

The relief valve

exhaust continuously indicated approximately 180-200OF due to the leakage.
One RCDT pump was being operated in manual continuously to cool the RCDT water
inventory and transfer the leakage water to the reactor coolant bleed tank
(RCBT).

Interviews conducted and records review revealed that this condition
The effect of this continuing leakage

had existed since the Fall of 1978.

over a prolonged period on the operation of the EMOV during this accident
cannot be evaluated at this time.
151,

2,

14,

37, 49,61,

153,(Int. 5,17, 38,

145,

147,

189; 33,

15,

56,

118)

Approximately 2,600 gallons of water were transferred each shift (8 hours)
from the RCDT to the makeup tank (MUT)

via the RCBT prior to the shift onwhich the accident occurred as determined by review of the control room log
book (2300 hrs March 1979, through 0400 hrs March 28,1979).

first 4½ hours of the shift on which the accident occurred,transferred.

During the
1,800 gallons wereThis review revealed that the transfer of water to the MUT had

increased to approximately 3,600 gallons per shift on March 28, 1979,
a substantial increase in leak rate.

suggesting

The RCS leakage behavior for a period(March 22-28, 1979) was reviewed to establish if any other apparent trends
could be established prior to the accident.

The following data were takenfrom the completed copies of the RCS Inventory Procedure, 2301-3D].

Date

Gross Leak
Rate (gpm)

3/22/79
3/24/79
3/25/79
3/28/79*

6.73
6.55
6.60
6.94

Net Unidentified
Leak Rate (gpm)
0.581
0.160
0.432
0.010

*This measurement performed from 0134 hrs to 0234 hrs on March 28,

1-1-2

1979.

The licensee calculates leak rate interms of equivalent gpm of water at
RCS Operating conditions (i.e.,

580'F and 2150 psigY which should be conserva-

tive when evaluating results against the requirements of the Technical Specifications.

However, the Technical Specifications are silent on the specificbasis for comparison.
(TCN)

A review of the procedure and Temporary Change Notice

2-79-070 to that Procedure revealed that the basic procedure is in error

resulting in miscalculation of the RCS leak rate.

These errors are discussed

below.
Review of TCN 2-79-070 showed that changes to-the RCDT inventory areappropriately adjusted to RCS conditions for the calculation.

However,line

16 of Data Sheet 1 (used when computer is available) and line 29 of Data Sheet
2 (used when hand calculations are performed) incorrectly adds the MUT water
additions without correcting those values to gallons equivalent to RCS conditions.
The above results were recomputed.

To determine if the omission of thiscorrection leads to substantially different RCS leakage rate calculations.
The results are shown on Table 1.1-1 on the following page.
The recalculated results, whether using the unidentified leak rate resultsfor hot or cold water, shows the licensee was operating the facility during
the March 22-28,

1979,

period with the RCS unidentified leakage rate in excess

of 1.0 gpm.

1-1-3

TABLE I.1-1
RECALCULATION OF RCS LEAKAGE RATE
Reported Gross RCS
Leak Rate

Date

Recalculated Gross
RCS Leak Rate

Unidentified
RCS Leakage*

Unidentified
RCS Leakage**

6.73

8.05

1.90

1.36

3/22/79
3/24/79

6.55

7.91

1.52

1.09

6.60

8.60

2.44

1.75

6.94

8.94

2.01

1.44

3/25/79
3/28/79***-

Calculated as gpm at RCS conditions
Calculated as gpm at 120 degrees F c6nditions

Evaluation
Technical Specification 3.4.6.2 and surveillance procedure 2301-3DI
require that RCS unidentified leakage be limited to 1 gpm.

Operation of theunit during the period March 22-28, 1979, with an unidentified leakage rate in.excess of 1 gpm-is under consideration as a potential item of noncompliance.
.1.2.4EMOV Leakage
Emergency Procedure 2202-1.5, Pressurizer System Failure, Section A.2.B.I,

requires that electromatic relief isolation valve RC-V2, be closed when thefollowing symptoms exist:
Relief valve discharge line temperature exceeding the normal 130'F.
NOTE:

The relief valve (RC-R2)

discharge temperature was approximately

180OF at the time of the accident.

A review of licensee records revealed

that no temporary procedure change had been issued to alter this established procedural requirement.
RCDT pressure above normal on the control room radwaste disposal control
panel,and the temperature above normal on the local radwaste disposal control

panel.
The operators were operating the RCDT transfer pump with flow through the
RCDT cooler continuously to maintain the RCDT temperature at ambient conditions with the apparent valve leakage into the tank.

The continuous operationof the drain pump in order to maintain the tank temperature at ambient conditions
indicated an abnormal condition-existed.
RCS makeup flow above normal- for the variable letdown-flow and RC pumpseal in-leakage conditions.

1-1-5

This was indicated by the frequent transfer of reactor coolant between
the RCDT and the MUT via the reactor coolant RCBT.

Any unusual changes in the

MUT level as occurred during the early hours of the shift is the initial
indication of a change in the RCS leakage.
Additionally,

Emergency Procedure 2202-1.5, Section C.3.3, requires thecode relief discharge line temperature to be placed on the analog trend recorderwhen the discharge line temperatures exceed the computer normal (130 'F).

The

valve discharge line temperatures were approximately 180 'F prior to theincident, and the code valves were not being trend-recorded as required.

The

poi-nts- on thetrendrCbde-rs as indic-a-ed-6by-the computer (group 6) Analog
Assignment Summary included and review of the analog trend charts at 0400 hrs
on March 28, 1979 included:
EVALUATIONThe EMOV block valve (RC-V2)

was not closed and the Code Relief Valve

Discharge Temperatures were not placed on the Analog Trend Recorders withleakage from the EMOV (RC-R2)

and/or both the code relief valves (RC-RVIA and

RC-RVIB) on the morning of March 28,

1979.

Technical Specification 6.8.1.a requires written procedures shall be
implemented,

covering the activities referenced as applicable procedures

recommended in Appendix "A" of Regulatory Guide 1.33, November 1972,
tion of Pressure Control System."block valve (RC-V2)"Malfunc-

The failure to close the EMOV (RC-R2)

and place the Code Relief Valve Discharge Temperatures on

the Analog Trend Recorders is under consideration as a possible item of noncompliance,
ATR 1

Condensate Hotwell Level

ATR 2

Condensate Storage Tank IA Level

ATR 3

T-G Bearing No.

ATR 4

SGFP Suction Header Pressure

1-1-6

2 Vibration

1.2.5Condensate System
The condensate system reject line isolation valve (CO-V60)

from thehotwell to the condensate tanks (CO-T-lA/lB) was normally partially closed by
the operators to limit the transient on the condensate booster pump and feedwater pump suction pressures if the reject valve cycles or fails open for any
reason.

The condensate system includes a full-flow polisher (demineralizer)

system to provide continuous demineralization of the condensate water supplied
to the feedwater system and the OTSGs.

A full-flow motor-operated bypass

valve (CO-V12)

is provided around the polishers which can be operated from thecontrol room.

This valve does not open automatically upon polisher system

malfunctions (high differential

pressure to the condensate booster pump suction).

Prior to the accident, operators were working to transfer resin from polisher
tank No.

7 to the resin regeneration tank.

This had been in progress forapproximately 11 hours perior to the trip.
Evaluation
Personnel interviews revealed that the operators were extremely sensitive
to the fact that the normal operating conditions of the condensate/feedwater
systems were very near the design limits at full power.
145,
1.2.6

147,

189, 2,

14,

37, 4, 9, 69,

(Int.

5,

17, 38,56,

153)

General Plant Parameters
Routine operating and surveillance procedures had been completed prior to

the accident establishing specific operatingconditions within the plant in.
accordance with the Plan-of-the-Day schedule dated March 27,

1979,

including:

Surveillance Procedure 2301-SI, Shift/Daily Checks (March 28, 1979)
BWST Level
55 ft
BWST Temperature
680 F
Reactor Building Pressure
0.1 psig
Reactor Building Temperature 114 0 F
Control Room Temperature
70OF
Reactor Building Sump Pump Starts(The sump routinely pumped approximately 800 gallons per day.)

1-1-7

Core Flood Tank Pressure595-600 psig
Core Flood Tank Level
13.1
Condensate Storage Tank Level 21.5 ft
The major plant parameters immediately prior to the unit trip included:
Reactor Coolant System.Pressure
Temperature, Hot Leg
Temperature, Cold Leg
Temperature, Average2155 psig
606OF
5570F

5820 F107.5-108%
137 x 106 lb/hr
220 in.

Flow (4 RCP)
Pressurizer Level
ReactorPower
97%
Core Power2701'MWt
Control Rod Groups % Withdrawn
1-5
100%
6
95%
7
95%
27%
8
Power Imbalance (approximate)
Incore-2.64%-3.64%
ExcoreSecondary Systems
Main Feedwater System(2 pumps)
Flow
Temperature
Pressure

A
5.798
462.7
1045

OTSG
B
5.727 x 106 lb/hr
462.7 0 F
993 psig

Condensate System,2 of 3 condensate pumps and2 of 3 condensate booster pumps)
2 of 3 heater drain pumps
Full-flow demineralizer(7 vessels)

8.303 x 106 lb/hr

Condenser vacuum

28 in Hg.

1-1-8

Main Steam System

OTSG
A
B
595 594 0 F
910 889.6 psig
257 264 inches

Temperature
Pressure
OTSG Level
Circulating Water System
Inlet temperature
Outlet temperature
(Ref.
1.3.

139,

140,

660 F
950 F

141)

STATUS OF SURVEILLANCE AND INSPECTION

1.3.1

General Status
During this investigation, a review was conducted to ascertain if thelicensee was in any ACTION statement'in accordance with the requirements of
their Technical Specifications.

It was found that the licensee had identified,prior to the accident, that one ACTION statement was in effect.
Technical Specification 4.5.2.a requires that DH-VI57,

the borated water

storage tank to spent fuel pool isolation valve, must be closed during Modes
1, 2, and 3.

DH-V157 had been opened on March 26,

recirculation of the BWST contents.

1979,

at 1500 hrs to permit

The ACTION statement (a)for Technical

Specification 3.5.2 requires that the system be made operable in 72.hours.
This 72-hour period would have expired at 1500 hrs on March 29,

1979.

Inter-

views revealed that this valve was closed at about 0800 hrs on March 28,
(Int.

57,

1979.

111, 184)

The investigator selected 82 Technical Specification requirements as asample,and reviewed the surveillance testing for the period of January 1,

1979 to March 28,

1979,that was performed to satisfy those requirements.

All the surveillance had been performed within the schedule required and theresults were satisfactory.

.1-1-9In addition, the following documents were reviewed for evidence of information that would indicate surveillance that was not satisfactorily completed or
conditions that would indicate inoperable equipment.
oShift Foreman's Log BookoSurveillance Task Indexo0Surveillance Result Index for 1/1/79

12/1/78-

3/29/79

3/29/79o

TMI Unit 2 Histogramo

TMI Unit 2 Core I,o

TS Action Statement Summary, Shift Foreman's Logbook

1/1/79 - 3/29/79

Cycle I Burnup History

During this review, there were no findings which would indicate that anyother ACTION statements,

other than the one identified above,(DH-V157),

were

applicable:
1.3.2

Status of Surveillance on Selected Engineered Safety Features(ESF)

1.3.2.1

System Components

Review of Procedures

Selected surveillance procedures performed during the months of January
through March of 1979, were reviewed including:2302-MlA/B,
22,

Makeup Pump and Valve Functional Test, Revision 5, September

1978 performed on January 29, February 14, February 22 and March 23,

I-I-101979.

2303-M2A/B,

Decay.Heat Removal Pump Functional Test and Valve Operability

Test,. Revision 7, June 20,
3,

1978, performed on January 4, January 29,

and March

1979.
2302-M24A/B,

Reactor Building Spray Pump Functional Test and Valve Oper-ability Test, Revision 3, March 13,
7-, 1979- and-March 5,
1.3.2.2

1978, performed on January.5,

1979.

1979,.February-

Emergency Feedwater Valve Test

The review revealed no discrepancies or technical inadequacies with theimplementation of the above procedures or their procedural content.
tionable practice was identified.

One ques-

During the reView of the procedure fortesting the reactor building spray pumps and valves (2303-M24A/B steps 6.3 and6.4),

it was noted that both the manually operated header isolation valves(DH-V133A and V133B) are closed simultaneously during the performance of thetest procedure (Ref.

65).

In order to perform the test and stroke the individual

motor-operated isolation valves (DH-V8A and V8B) for the sodium hydroxide tank
(DH-T-l),it

is not necessary to close the manual isolation valve in the

redundant flowpath.
Technical Specification 3.6.2.2 requires the spray additive system to be
operable in Modes 1, 2, 3,

and 4.

inoperable for 72 hours.

The total removal of the spray additive system fromThe spray additive system is allowed to be

the operable condition during the performance of the surveillance procedure byclosing the manual isolation valves (DH-V133A and 133B)
contrary to the specification requirements.
automatic valves (DH-V8A and V8B)

does not appear to be

However, the provision of two

in parallel in order to meet the single

failure design criteria also allows the testing of one spray additive flow
path while the redundant path is available for automatic initiation if
A more conservative surveillance procedure,

isolating and testing the flow

paths independently, would provide an additional margin of safety.

I-I-II

required.

Selected surveillance performed on the emergency feedwater system during
the months of January through March of 1979, were reviewed,
2303-M 14A/B,

Revision 8, November 21,

1978,

including:

Emergency Feedwater System

Valve Lineup Verification and Operability Test and Turbine Driven Emergency
Feed Pump Operability Test, performed on January 3,

February 26, and March 26,

1979.
2303-M27A/B,

Revision 4, August 30,

1978, Motor Driven Emergency Feedwater

Functional Test and Valve Operability Test, performed on January 3, February
26, and March 26,

1979.

Evaluation
Both of the above surveillance procedures contain specific steps that
result in the routine isolation of the emergency feedwater supplies to the
OTSGs by closing the header isolation valves (EF-VI2A and Vl2B).simultaneously.
This occurs as follows:
Procedure 2303-M14A/B requires:oStep 6.1.10, Close EF-VI2A and 12B.

o

Step 6.1.52,

Reopen EF-VI2A and 12B.

Identica] manipulations are performed in Steps 6.2.9, 6.2.40, 6.3.8, and6.3.27.

Similarly, procedure 2303-M27A/B requires:oStep 6.1.2, Perform Appendix A(B) valve lineup.
(Appendix A(B) includes the closing of EF-VI2A and 12B.)

1-1-12,o

Step 6.1.22,

Insure EF-V8A(B) is open,

EF-VI2A(B)is open,

EF-VlA(B)is closed, and close EF-V-39 (EF-V40).
Identical manipulations are performed in Steps 6.2.2 and 6.2.17.
NOTE:

The above steps which use the designation "VI2A(B)" suggestthat only-one valve was closed during a previous step or steps,
which was not actually the case.
During the routine performance of the Surveillance Procedure 2303-M27A/B,
the OTSG emergency feedwater flow paths were isolated by closing both header
isolation valves (EF-VI2A and 12B)
February 26,

and March 26,

during the test performance on January 3,

1979.

Similarly, during the routine performance of Steps 6.1.10,
6.3.8 of Surveillance Procedure 2303-Ml4A/B/C,

6.2.9, andboth OTSG flow paths were also

isolated by closing both header isolation valves (EF-VI2A and 12B) during the*test performance on January 3, February 26, and March 26, 1979.
Evaluation
Technical Spec'ification 3.7.1.2 requires that three independent steam
generator emergency feedwater pumps and associated flowpaths be operable in'Modes 1, 2, and 3 (OTSG pressure greater than 800-psig).

The specific action

statement does allow one emergency feedwater system to be inoperable,

providedthe inoperable system is restored to OPERABLE status within 72 hours or thereactor is to be placed in HOT'SHUTDOWN within the next 12 hours.
The routine complete isolation of the emergency steam generator feedwaterflow paths on January 3,

February 26,

and March 26,

1979 is contrary to'therequirements of Technical Specification 3.7.1.2 and is being considered as apotential

item of noncompliance.

1-1-13

1.3.2.3

Review of EFW Procedure Change

Since a previous procedure change resulted in the closure of both emergency feedwater valves during performance of the above surveillance procedures,
an investigator reviewed the evaluation of that change.
specific Plant Operating and Review Committee (PORC)
change requests (PCR)

The review of theminutes and the procedureassociated with the procedure revision revealed anapparent inadequate safety evaluation, review, and approval in each instance,
including:
Procedure Change Request PCR 2-78-707; Item II B of"PORC Meeting No.
was entered on August 30, 1978,

281,

This procedure

as Revision 4 of 2303-M27A/B.

change was recommended to provide new pump reference values because the valve
lineup had been changed.

The emergency feedwater header isolation valves(EF-Vl2A and Vl2B) were closed simultaneously to prevent leakage past both the
emergency feedwater level control valves (EF-VIIA and 11B).

(It

appears that

this leakage is an inherent design feature of these valves to prevent thermaltransient to the emergency feedwater nozzles and should have been recognized
during this review.)
The safety evaluation performed for this procedure change indicated thatthe change would enhance safety in that it would ensure thatscold water would
not be fed into the OTSGs and thermally cycle the emergency feedwater nozzles.
The procedure change specifically included the simultaneous closing of both.
the header isolation valves (EF-VI2A and Vl2B) during the performance of thesurveillance activity.

The safety evaluation failed to address that aspect of

the change and the impact as an unreviewed safety question.
Procedure Change Request, PCR-2-78-895 (Included Temporary Change Notice
2-78-666,

October 10, 1978),November 21,

Item II B, PORC Meeting-No.

1978, as Revision 8 of 2303-Ml4A/B/C.

1-1-14(Ref.

293, was entered on
67)

This procedurechange was recommended to specifically close the header isolation valves(EF-Vl2A and Vl2B) during the performance of the procedures,

but the simultaneous

closing of both the header isolation valves (EF-Vl2A and VI2B) was not specifically addressed in the detailed safety evaluation.
Evaluation
The Plant Operations and Review Committee (PORC)

is required to review

all changes to procedures which could affect nuclear safety.
and Review Committee (PORC)(2303-M27A/B,

The Plant Operations

review and approve of the above procedure changes

Revision 8, dated August 30,

1978,

and 2303-Ml4A/B.C,

Revision

8, dated November 21,1978) placed the emergency feedwater system outside thelicense requirements,

and this constituted an unreviewed safety question.

(Ref.

66, 69)The initial review and request for a procedure change by engineering; thereview and approval of the change by the PORC; and the routing performance ofthe procedure by members of the operating shifts (including licensed operatorsand supervisors) without realizing that the performance of the procedure
placed the plant outside the license requirements is indicative of a serious
breakdown in the licensee knowledge level of the facility.

This breakdown

occurred at numerous levels of expertise and training, including engineering,
management,

and operations.

The safety evaluation performed associated with the procedure change
requests (PCR-2-78-707

and 895,

including temporary change notice, TCN-2-78-666)

did not address the Technical Specifications requirements concerning feedwater
system operability.

The safety evaluations were directed towards obtaining

test data and limiting thermal transients on the emergency feedwater nozzle.
(Ref.

67,

142)

1-1-15

Technical Specification 6.5.1.7.b requires that the PORC shall render
determinations in writing with regard to whether or not changes to procedures
required by Technical Specification 6.8 constitute an unreviewed safety question.
The failure to include the EFW valve lineup changes (simultaneous

closing ofboth feedwater header isolation valves) within the safety evaluation as required
and render a written determination,is under consideration as a potential item ofnoncompliance.
1.3.3

Status of Core Physics Surveillance
The status of the reactor core as revealed by the last surveillance testof that core was reviewed to determine if any core conditions existed that
could have contributed to the response of the core to the accident.

No coreconditions that were outside the acceptable limits defined by the Technical
Specifications for the facility were found.
Core performance surveillance was performed at 0810 hrs on March 19,
1979.

Conditions during the surveillance were:Power - 97.5%
Boron Con.

- 1037 ppm

RC Total Flow - 138.7 x 106 lb/hr
Core Burnup - 2662 MWD/MT
Pressure - 2268 psia
ALHR - 5.9696 kw/ft
Core up - -21.315 psiaT
- 584.3 0 F
avg
Inoperable Self-Powered Nuclear Detectors (SPND)
String 4

level 5 (F-8)

String 5

level 2 (F-7)

String 6

level 2, 4-7 (E-9)

String 11

level 5 (K-5)

1-1-16were:3 (N-8)String 14

level 2,String 27

level 2-7 (D-10)

String 38

All levels (L-2)

String 46

level 4,

String 51

level 6, (D-14)(R-10)

Inoperable Incore Thermocouples were:
String 26(E-II)Hot Fuel Maximum Peaking Factors were:
H -8

1.485

B -8

1.361

C -8

1.234

B -6

1.198

Note:

FSAR does not reference thesevalues in Table 4.4-1 of the FSAR.

Channel Minimum DNB Ratio:
Note:

FSAR references 1.75 for maximum

B -8

3.004

B -6

3.356

design conditions at 100% power.

C -8

3.619

(Table 4.4-1,

H -8

2.830

FSAR).

Hot Channel Maximum Linear Heat Rate (kw/ft)
B -8

11.653C -8

10.778B -6

10.276H -8

13.144

Note:

FSAR references 19.03 kw/ft at
design power (Table 4.4-1,

Hot Channel Surface Heat Flux (Btu/ft

2 _

1-1-17-hr)FSAR).

B
C
B
H

8
8
6
8

Note:

353,220

FSAR references 576,885

2 - hr at design power

326,690

Btu/ft311,480

condition (Table 4.4-1).

398,390

Hot Channel Maximum Surface Temp (*F)
B
B
C
H

8
6

657.66657.53surface temperature of 654 0 F.

88653.78(Table 4.4-1).

657.72

Technical Specification require-

Note:

FSAR references a maximum
There is noment for maximum surface temperatures.
Hot Channel Maximum Fuel Temp ('F)
B- 83664.4C-

83551.3B-

63483.7H-

83850.8Note:

FSAR references 4,170'F at design
conditions.

(Table 4.4-1,

FSAR).

(These data were extracted from Performance Data Output Segments 1-6; Ref.

68).
Core surveillance was also performed per Procedure 2311-FT (2),at 1500 hrs on March 22,

1979.

Conditions during the surveillance were:Power - 97.52%
Imbalance - FO.5% all

4 channels

Quadrant Tilt - FO.2% all 4 channels
Exposure-

90.16 EFPD

Maximum Fuel Ass'y to Ave. Power - 1.441
Minimum DNBR - 2.91
Max.

Revision 0,

LHR - 13.02 kw/ft

1-1-18

Max Heat Flux - 394,500 Btu/ft 2
Max Clad Temp --hr

655.2 0 F

FQ - 1.77
EvaluationIt

is not clear that the licensee is monitoring FAH as defined-in theTechnical Specifications.

This item has been discussed with the licensee andRegion I personnel andwill be the subject of further review by the Regional
Inspection staff.
1.3.4

Status of Surveillance Results and Records Review
Through personnel interviews and a review of available surveillance

records it was revealed that inadequacies existed in the area of performance,
review, and retention of surveillance tests.
The following surveillance procedures were reviewed:
2303-M27A/B,
1.3.4.1

2303-MlA/B,

2303-M2A/B,(2303-Ml4A/B/C,

and 2303-24A/B; Ref. 69,

66,

70, 71,65)

Performance and Review

The review and personnel interviews revealed that completed surveillance
procedures-are not routinely reviewed by the shift supervisor/shift foremanexcept for the completed data sheets.
carded.

The remaining sheets are routinely dis-A search of selected surveillance records in storage (2303-Ml4A/B/C,

2303-M27A/B,

2303-MIA/B,

2303-M2A/B,

and 2303-M24A/B), revealed that no completed

surveillance procedures had been maintained with exception of the data sheets,the computer schedule sheet, and the procedure cover sheet.

Discussions withthe licensee representatives indicated that the details of the surveillance
procedures (Sections 1 through 6) were not retained because of a record storage
space problem.

Moreover,it was noted during the interviews that review of,

the completed procedures by the shift supervisor/shift foreman was not routine

1-1-19because the length of some of the procedures and the routine scheduled and unscheduled workload.

Discussions revealed that the licensee schedules surveil-

lance on the day shift (7-3) when extra personnel should be available toreview the completed procedures as required.
66, 69, 70, 71,

(Int.

107,

136,

116,

131; Ref.

65)

The surveillance procedures reviewed contained procedural steps (equipment
manipulations) which, if

not specifically completed and verified to be satisfac-

torily completed, could leave the specific engineered safety feature system in
an inoperable condition.

Because of a lack of documentation,

the possibility

exists that this occurred during the surveillance performed on March 26,

1979,

on the. emergency feedwater system and the header isolation valves (EF-VI2A andVI2B) left closed.

(This matter of the closed valves is discussed in greater

detail in section 4..2 of this report.)

An operator in an interview statedthat the header isolation valves (EF-VI2A and Vl2B) were reopened on March 26,
1979,

at the completion of the surveillance test (2303-M27A/B).

(Ref.

66)

Evaluation
Administrative procedure 1001,(Procedure Usage),

Revision 13,

March 30, 1978,

step 3.8.5

requires the implementation of surveillance procedures

including performing the procedures rigorously, providing documentation,

andproviding shift supervisor/shift foreman review of results and approval of
surveillance data sheets.

(Ref.

94)

The failure to provide a review of the completed surveillance results by
the shift supervisor/shift foreman-as required to insure the procedures are
implemented rigorously, documentation provided, and results reviewed is being
considered as a possible item of noncompliance pursuant to Technical Specification 6.8.1.a.

1-1-201.3.4.2

Record Retention

Administrative procedure 1007,

Revision 4, September 26,-requires record storage including handl'ing,

1978,

storage, and indexing.

(Records)
(Ref.

95)

Technical Specification 6.10.1.d states that records of surveillance
activities,

inspections, and calibrations required by these Technical Specifi-

cations shall be retained for at least five years.
Evaluation
The failure to maintain the complete records of the surveillance activities
as required by Administrative Procedure 1007 is contrary to Technical Specification 6.10.1.d and is under consideration as a potential item-of noncompliance.
1.3.,5Inspection of Surveillance Activities.
The Metropolitan-Edison Three Mile Island Final Safety Analysis Report,

Chapter 17.2.15,

Inspection,and the Operational Quality Assurance Program,

Section X, Inspection (Surveillance),
unit operation or functional

requires,in part, that during.normal

testing, the surveillance program include random

observation of operations and functional testing.
measurements,or tests of materials,

Inspection,

examinations,products, or activities are required tobe performed for each work operation where necessary to assure quality, whether
the work is done by the station staff or an outside contractor.

In the eventa surveillance of processed material or products is impossible or impractical,
indirect control by monitoring processing methods,
required.

equipment,

and personnel isBoth inspection and process monitoring must be provided when control

is inadequate without both.

1-1-21

The Metropolitan Edison Operational Quality Assurance Surveillance Program,
GP 4014,

Rev 0, December 6,

1977,

requires independent observation of activities

affecting quality to verify conformance with established requirements utilizing
both inspection and auditing techniques for compliance with written procedures,and the Technical Specifications.
Evaluation
The review of the selected surveillance activities and the administrative
controls and the personnel interviews revealed that the operations surveillance
and testing activities were not randomly or routinely inspected by, independent
methods as required by Criterion X of Appendix B to 10 CFR Part 50; amplified
in the FSAR,

Chapter 17.2.15, Section X, Inspections (Surveillance);

GP4014, Operational Quality Assurance Surveillance Program.
As of March 28, 1979,

(Int.

and

131)

the failure of management to adequately implement

the inspection requirement pursuant to Criterion X of Appendix B to 10 CFR 50
is being considered as a potential item of noncompliance.
1.4
1.4.1

OPERATIONS STAFF
Operations Personnel Training and Qualifications
1.4.1.1

General Plant Training Program

The normal training progression starts with Auxiliary Operator C andcontinues through Auxiliary Operator B and A.

The most senior qualified A

operator can apply for the position of Control Room Operator (CRO)when one becomes available.
Operator (SRO)

trainee

CROs can compete for available Senior Reactor

positions.

1-1-22

1.4.1.2

Auxiliary Operator Traininq

The requirements for becoming an Auxiliary Operator-C (AO-C)
school diploma and a course in high school algebra.

are a highInitial training as an

AO-C consists of 8 or 9 weeks of classroom training, dealing mainly withsecondary plant systems.
ooOther areas covered during this period include:

Fundamentals of Mechanics and heat transfer("The basic steam cycleoSwitching and tagging proceduresoIntroduction to the safety features actuation system (SFAS)

and thereactor protection systems (RPS)oFirst aidoFire fighting

0Intermediate health physics

This period is followed by a written examination.

The trainee is thenassigned to a crew under a foreman, and works under the supervision of an A
level AO for the remainder of a year.

The trainee must then pass anotherwritten and an oral examination to become an AO-B.

Failure to pass the exam-

ination results in removal from the training program.
Trainees at the AO-B level must have completed courses in high school
trigonometry and physics.

This training period also covers a year and isinitiated by six weeks of full time classroom instruction.
covered deal mainly with the primary system and include:oPrimary coolant makeup systemoDecay heat removal systemo

Reactor Building coolingoCore flood tanksoReactor Building spray systemo

In depth study of the SFAS and RPSoNuclear instrumentation

I-1`23The subjects

oControl rod drive mechanismoIntegrated control systemoAdvanced health physics

A comprehensive written examination is given at the end of this period.
The remainder of the year is spent in on-the-job training, mainly on primarysystems.

This year is also followed by comprehensive oral and written exam-

inations which the trainee must pass in order to remain in the program and to
be promoted to AO-A.

An AO-A is eligible to enter training for the positionof Control Room Operator.
1.4.1.3

CRO Training

The CRO Trainee enters the Category IV (CIV)nine months.

Training Program which lasts

This program is divided into six cycles of 5 to 6 weeks per

cycle, with a written and an oral test at the end of each cycle.
is set up for individual self study.

The program

Each section lists the procedures thatmust be learned and specifies how proficiency in the procedure is to be demonstrated.

This is done by actual performance,by simulation, or by discussion.

At the end of Cycle 6, a cumulative examination over the entire course isgiven.

Successful completion of this examination is followed by a two-weekcourse on the reactor simulator.

The trainee then takes a mock NRC written

exam, and an NRC-type oral exam given by a senior reactor operator.

The

trainee is then scheduled for the actual NRC licensing examination.
Experienced CROs can compete for SRO positions.
self study tailored to the individual's experience,onuclear theoryoradioactive material handlingooperating characteristicsofuel handling

1-1-24

SRO Training is individual
but includes:

ocore parametersoadministrative proceduresotechnical specificationsotransient analysis

The programs outlined above were compared to the training requirements in
ANSI N18.1,"Selection and Training of Nuclear Power Plant Personhel," which

the licensee is required to meet.
1.4.1.4

No discrepancies were identified.

Training for Individuals with Previous Experience

Individuals with previous nuclear power experience are given abbreviatedtraining that takes into accounttheir previous training. The training records
onsite indicate that many TMI operators were graduates of the Navy Nuclear
Power School.

For these individuals, the normal two years of auxiliary operator

training and work experience were replaced by a special 26 week course..

This

course included 40 hours per week of classroom instruction and plant tours.
Itconcentrated on those areas of the auxiliary operator training not coveredby the Navy training, primarily the TMI plant'systems and operation.

This wasfollowed by the CRo training program.
1.4.1.5

Requalification Program

The licensee's requalification program required of all licensed operators
is divided into four areas:a.

Operationalb.

Fundamental and Systems Reviewc.

On the Job Trainingd.

Annual Evaluation Examinations

Review Lectures

1-1-25

The lectures consist of at least 60 hours per cycle and cover the fol~lowing
areas:a.

Reportable Occurrencesb.

Modificationsc.

Operating Problemsd.

Procedure Changese.

Abnormal and Emergency/Procedures Reviewf.

Technical Specificationsg.

Operational Evaluationsh.

Federal Regulationsi.

System Review Materialj..

Other topics as necessary

Periodic quizzes that must be passed with an 80% grade are given.
The fundamentals and systems review program is based on the annual written
examinations and includes topics in the following areas:a.

Operationb.

Facility Designc.

Operating Characteristicsd.

Instrumentation and Controle.

Safety and Emergency Systemsf.

Normal and Emergency Proceduresg.

Radiation Control and Safety

The Senior operator program includes:a.

Reactor Theoryb.

Radioactive Material Handlingc.

Fuel Handlingd.

Core Parameterse.

Administrative Procedures

1-1-26

Removal from duties and accelerated requalification training is mandatory
for failure to score 70% overall on the annual examination.
Quizzes requiring an 80% passing grade are also given in for this program.
The On-the-Job-Training for requalification includes:
1.

10 reactivity manipulations.

2.

Participation in surveillance testing, system checkout,and equipment

operation relevant to the license.
3.

Review of procedure changes, modifications, Technical Specification
changes,reportable occurrences,and incidents not covered in thelecture series.
1.4.1.6

Training on Significant Events

The training department reviews the NRC listing of the Licensee Event
Reports to identify items for inclusion in the operator training program.
However in at least one case this review did not identify an item of significance to the accident..

On September 24,

1977, Davis Besse experienced aloss of feedwater transient which resulted in reactor transient behavior which
was similar in many respects to-the behavior of the transient which initiatedthe events of the TMI accident.

A review of the training records and discus-

sions with the training staff indicate that this event was not brought to theattention of the operators.

This particular item was included in the NRC

publication "Current Events Power Reactors" for September 1, 1977,
October 31,

1977,

published in December 1977.

through

The NRC distribution list

indicates that 10 copies of this report were sent to various Metropolitan
Edison and GPU addressees.
intendent.

However,

These included the TMI site and the Unit 1 super-

the training staff member responsible for this area

stated that this report had not been made available to the training department.

1-1-27

Another significant event occurred at the Rancho Seco Facility on March
20,

1978.

A transient was initiated by a loss of electrical power to part of

the non-nuclear instrumentation.

As a result of this event B&W sent a memo to

the Davis Besse Station Superintendent which included recommendations regarding
operation training and emergency procedures.

These recommendations.discuss

the use of alternate instrumentation to replace-inoperable instrumentation andthe use of RCS pressure as well as pressurizer level to assure that the reactor
coolant system is full.

(SOM #403 dated August 9,

1978; Ref. 143)

Review of TMI training records showed that the Rancho Seco event had been
covered du'ring operator training, but, no record of any specific B&W recommendation,such as those sent to Davis Besse, existed.

Training personnel said

that they had not been advised of these recommendations.
tion Memoranda (SOMs)
through March 9,

to TMI were reviewed for the period covering May 3,

1979.

Rancho Seco event.

The B&W Site Opera1978,

These memoranda did not address the discussion of the

The Unit 1 superintendent stated that TMI did not receive

any recommendations from B&W based on this event, and this was confirmed by
the B&W site representative.

The stated basis for this decision was that TMI

had a cooldown transient similar in nature to the Rancho Seco event,but causedby different factors.

The B&W analysis of the TMI event was covered in a site

operations memo (SOM)

to TMI.

(Ref.

144)

(SOM ll-140, dated May 2,

1978)

As

a result, B&W did not consider it necessary to send TMI the Rancho Seco related
recommendations.

A review of the SOMs to TMI for the period in question indicatethat the recommendations specifically relating to operator training were notincluded in this correspondence.
1.4.1.7

Annual Examinations

An annual written evaluation examination and oral evaluation are administered to all licensed personnel.
examinations given by NRC.

These examinations simulate the licensing

Failure to score higher than 80% in all sections

1-1-28

of the written exam requires participation in the systems review program.
Failure to score 70% may require removal from duties to enter an accelerated
training program.

An unsatisfactory-or-al

evaluati-on may-require similaraction.
1.4.1i.8

Individuals Involved with the Early Part of the Accident

All four of the NRC licensed individuals who were on shift at the start,of the March 28 event were graduates of the Navy nuclear power school.

Allwere initially hired as auxiliary operators and attended the special 26-weekcourse.

In addition, each had attended from 5 to 9 weeks of B&W simulator

training. The specific experience and training of these individuals as indicatedby licensee records and interviews with the individuals are as follows:o

Shift Supervisor E:
He was hired as an Auxiliary Operator on April 14,

1972.

He became a

Shift Foreman (requires a NRC senior reactor operator's license) on October
1, 1972,

and shift supervisor on May 17,

1976.

He completed the abbre-

viated auxiliary operator training and the initial control room operator
training on October 19,

1973.

By March 1979 he had 7 years of experienceat TMI and a total of 13 years of nuclear experience.
o

Shift Foreman C:
He was hired as an auxiliary operator on March 5,

1973.

He became a con-trol room operator (which requires an NRC senior reactor operator's license)
on August 11,

1975, and a Shift Foreman on February 13,

1978.

He completed

the abbreviated auxiliary operator training program on October 19,
and the initial control room operator (CRO)1973,

training on August 11, 1975.

By March 1979 he had 6 years of nuclear experience at TMI and a total of
15 years of nuclear experience.
oControl Room Operator C:

1-1-29

He was hired as an auxiliary operator on December 5,

1973.

He became a

Unit 1 Control Room Operator on October 20, 1975 and worked in thiscapacity for three months.
on September 13,

1976.

He regained his control room operator status

He completed the abbreviated auxiliary operator

training program on August 26,

1974, and the Unit 2 control room operator

training program on October 20,

1977.

By March 1979 he had 5 years of

nuclear experience at TMI and a total of 12 years of nuclear experience.
oControl Room Operator D:
He was hired as an auxiliary operator on November 28,
Control Room Operator on October 6,

1975.

1973, and became a

He completed the abbreviatedauxiliary operator training program on August 26, 1974,

and the initial

Unit 2 control room operator training program on August 30,

1977.

By

March 1979 he had had 5 years of nuclear experience at TMI and a total of
10 years of nuclear experience.
Other individuals involved with the early part of the accident include:oShift Supervisor A:
He had no nuclear training prior to his employment at TMI on March 3,
1969.

He became a Control Room Operator on August 13,

Foreman on October 13,

1975.

1970, and a Shift

He attained his present pos.ition as Shift

Supervisor on October 1, 1977.

His initial nuclear training was a 48

week Reactor Operator Training course which covered the period from
October 20,

1969, to August 18,

1970.

He completed his Unit 2 Senior

Reactor Operator license training on August 1, 1971.
had had, 10 years of nuclear experience,o

As of March 1979,heall at TMI.

Shift Foreman B:
He had a B.S. degree in Chemistry and no previous nuclear experience when
he was hired as an Auxiliary Operator on April 4,

1-1-301973.

He became aControl Room operator on August 11, 1975,

and a Shift-Foreman on August

1, 1978.

He completed his initial auxiliary operator training on October

19,

1973,

and a Unit 2 Control Room operator training program on August

11, 1975.

He completed his senior operator training program for Unit 2

on August 14,

1978.

As of March 1979,

he had 6 years of nuclear experience,

all at TMI.
Management level individuals were also involved in the Unit 2 operations
during the early hours of the accident.

Their~experience and training are as

follows:oUnit 1 Supervisor of Station Operations Nuclear:
He was a graduate of the Navy Nuclear Power School and had 8 years of
nuclear-experience1968.

in the Navy.

He was employed at TMI on December 9,

He became a Shift Supervisor on July 15,

1972, and attained his

present position as Supervisor of Operations on April 1, 1978.

His

training records show a total of 659 hours of Senior Reactor Operator
license review and training, completed on August 30,
weeks of simulator training.

As of March 1979,

1977, and 2 1/2

he had 11 years of nuclear

experience at TMI and a total of 19 years of nuclear experience.
oSuperintendent of Technical Support:
He had BS degree in Mechanical Engineering and no previous nuclear experience when he was employed by Metropolitan Edison on September 1, 1978.
He was assigned to TMI in 1970.
Nuclear on November 1, 1973,

He was placed in the position of Engineer

and Engineer Nuclear III on July 1, 1975.

He became a Supervisor of Station Operation Nuclear on September 1, 1975,
and was assigned hispresent position as Unit Superintendent of TechnicalSupport on December 1. 1977.

His training records show two short college

courses totaling 160 hours of reactor classroom instruction and 2 weeks
of reactor simulator training.

The records did not show completion of

any of the more extensive TMI operator training programs.

1-1-31

He has aSenior Reactor Operator license for Unit 1, but had not yet aquired an
NRC license on Unit 2.
.nuclear experience,
1.4.2

As of March 1979,

he had a total of 9 years of

all at TMI.

Shift Composition Compared to Technical Specifications
Technical Specification 6.2 specifies the Organization, and 6.2.2 esta-

blishes the Unit Staff, including the shift manning requirements.
consist of a Unit 1-2 Shift Supervisor, a Unit 2 Shift Foreman,
control room operators,

The shifts
licensedand "A" auxiliary operators.

Order of Modification of License,

issued October 13,

1978,

required thecommitment to specific operator actions to be completed within lO'minutes fromthe onset of the small break with a single failure.

The licensee designates asmall-break LOCA-CRO (switching and tagging Control Room Operator) and asmall-break LOCA-AO (primary auxiliary operator).
147,

189, 3, .15, 4, 9,

61,'153,(Int. 5,17, 38,56, 145,

151)

Records reviewed and personnel interviews revealed that the shift complement for Unit 2 on the morning of March 28,oUnit 2 Shift Supervisor - SRPOoUnit 2 Shift Foreman - SROoTwo Unit 2 Conttrol Room.

oFour Unit 2 Auxiliary Operators Ao

Two Unit 2 Auxiliary Operators Bo

Two Unit 2 Auxiliary Operators C.
(Int.

5,

9, 61,

153,

17, 38, 56,

145,

147,

1979, consisted of:

189, 2,

14, 37,

189,

3, 15,

151,. 4,etc.)

Two Control Room Operators were present in the control room at 0400 hrs
on March 28,
control room.

1979.
(Int.

The shift supervisor was in the SS office adjacent to the5,17, 38, 56,

145,

147,

I-i;32

189,

3,

15,

4,

9, 61,

153,

151)

Four Rad-Chem technicians were assigned to the 11-7 shift and were present
at the facility at 0400 hrs on March 28,

1979.

Evaluation
The total shift personnel complement present on the 11-7 shift on March 28,
1979,

and present at 0400 hrs at the initiation of the accident, met the require-ments of the Technical Specification.
1.4.3

Shift Staffing During Prior Trips

The available records of past reactor trip events and plant logs andstaffing records for Unit 2 were reviewed in an attempt to identify any trend
in operating staff assignments that may have significance in understanding thecauses of the March 28,

1979, accident.

No trends of specific significancewere identified.
Of interest though, and outside the scope of this investigation,

it wasnoted that Shift Supervisor C and Shift Supervisor D were the duty Shift
Supervisors during eight of the ten reviewed trip incidents.

Further,

it wasnoted that in seven of the ten reviewed incidents, the trip was experienced
within a nominal 90'minute period before or after shift change.
1.4.4

Physical and Emotional Condition of Operational Staff
During the course of interview sessions, limited observations were madeby the investigators to note any readily apparent health or emotional stability(obvious aberrant behavior) problems of the licensee's operating staff.

Theseinterviews were conducted during-the period early April to early July 1979,
when the members of the operational staff were under a high degree of emotional
stress directly as a result of the incident and, secondarily,-as a result of theextensive investigative activity underway and the high level of local antinuclear sentiment that appeared to be impacting their personal activities.

1-1-33

It

must be noted that the investigators are not trained medical observers,

but rather are lay personnel specifically alert to obvious signs of emotional
stress.

During these interviews, while staff personnel recalled the eventsand their actions of March 28 and were subjected to the direct questioning of
the investigative staff, no abnormal behavior was noted and all operating
staff participated with logical and poised responses.
A review of the licensee's screening policies as detailed in their Policy
and Procedural Manual showed that the licensee has a policy of administering a
psychological
24,

1978,

screening test, to all new personnel hired subsequent to Augustin an effort to determine their suitability to work at the facility.

The results of these tests are evaluated by trained medical staff.
A further review of the licensee's records showed four (one operating andthree support) personnel warranted further evaluation.

These individuals

were identified as a result of a review of all personnel files performed in
preparation for the revised NRC regulation on physical security requirementsthat became effective in August 1978.
cleared.

Three individuals were evaluated and

The single concerned operating staff member subsequently resigned

for independent medical-reasons.
No reasons were identified to suspect that the health or psychologicalcondition of operational staff members played a contributing role in theinitiation of or response of the licensee staff to this accident.
1.4.5

Possibility of "Anniversary" Celebration On Site
Since the date of the incident, March 28,

1979,

represented the firstyear anniversary of the date on which Unit 2 attained criticality, the possibility that unofficial"celebrations" may have been underway at the time of

the incident was evaluated.

1-1-34

During the course of this portion of the investigation,no information*was obtai-ned tha-twould support a-contention.that a celebration of any typewas in progress.

Thus any such contention has no bearing, on the initiation or

the conduct of plant staff during this accident.

In only one instance during

the course of the plant staff interviews was there a mention of an anticipatedonsite celebration.
member,Further discussion with that individual,

not one of the operating shift complement)

both on and off tape,revealed that no alcholic beverages were anticipated,"celebration,"(a support staff

and by use of the term

the interviewee had intended to indicate a "coffee and doughnuts"

type of get-together rather than anything more extensive.
There was no knowledge of any planned "celebration" stated by any operational personnel,

and the investigation has concluded that no such celebration

was in progress immediately prior to. the accident.
1.4.6

Possibility of Plant Sabotage
As a result of the high degree of public concern associated with thepossibility of sabotage or adverse human actions having caused or contributed
in some manner to the severity of the March 28 incident at TMI,

all members of

the investigative team were alert to this possibility.
Within their limited authority, as non-criminal investigators,

the assigned

investigators pursued the question of possible sabotage during the course of
the interview program through liaison with the FBI and other governmental
agencies and evaluation of other referred allegations and matters- of concern.
"Limited authority" means that the NRC investigative staff acted in a noncriminal.
investigative manner commensurate with their lack of power to administer oaths(i.e.,

take sworn statements),

to subpoena necessary documents such as phone

toll records and-official business records, or to obtain access to the criminal
intelligence records of other governmental and/or state agencies.

The goal ofthis investigation was to attempt to identify any intelligence that would

1-1-35

indicate whether sabotage did playa role in the event.

At that point, com-mensurate with agency policy, the obtained intelligence would be referred tothe proper Federal Criminal investigative agency.
No information was developed during this investigation indicating' thatsabotage was a contributing factor to the initiation of the accident or to thesubsequent response of plant personnel or equipment to the accident.
This investigation has also addressed several allegations provided to the
NRC by confidential sources that alleged sabotage or predictions of catastrophic
events by various individuals or groups.
be unfounded.

All allegations were determined to(Contact of the investigators with the local FBI agents incharge indicated asimilar disposition by them of matters under their purview.)"Predictions" in certain newspaper articles were found upon examination to
include "March 28" only as a reference to the 1978 date that Unit 2 attained
criticality as opposed to "predicting" a catastrophic occurrence.

This investi-

gation found no basis for further evaluation of these articles or interviews
with their authors.
The interview program addressed the question of potential sabotage on
selected occasions using questions directed to all working levels regardingknowledge of possible sabotage.
Responses ranged from direct negative responses to disbelief that sabotage
could even be considered.

Noticeably lacking during the course of the interviews

was any disloyal animosity on behalf of the plant operating staff toward
licensee management.

The operating staff appeared generally satisfied intheir relationships with licensee management, with only minor dissatisfaction
expressed over the adequacy of training.

A good degree of rapport appeared toexist between all levels of operating staff supervision and workers.
'Vocalantinuclear sentiment appears to have been relatively absent prior

to the incident.

Relationships between the local governments and licensee

management appear to have been of a tolerant nature.

I-1-36

Those conditions commonly associated or viewed as causative factors
precipitating industrial sabotage were not identified by the investigation as
being present at the time of the incident.
This investigation evaluated the effort that would be required to attempt
to identify the party or parties who closed the emergency feedwater system
block valves (EF-V12 A/B) for whatever undefined reason that may have motivatedthem.

The investigation revealed that the two valves in question were capableof being operated from three specific locations:the control room,'the 480V

Substation panels (2-21EA & 2-11EA) at the 305 elevation of the auxiliarybuilding, and the physicallocation of each valve.

Checks of the licensee's security access badging records showed-approximately 470 licensee personnel and 260 contractor/vendor personnel would have
had unescorted access to one or more of these locations on any of the two workdays preceeding the March 28 incident.

Records exist whereby the identity of

the contractor/vendor personnel entering the protected area could be retrieved.
However,the 470 licensee personnel are only logged in at the site perimeterand need only display their photo ID badge (issued only after psychologicalscreening and pre-employment checks are complete)

to secure access to the TMI

protected areas (Units 1 and 2, auxiliary building,turbine building andenvirons). Further investigative effort of the magnitude that would be required to
specifically identify which of the more than 600 personnel did access the protected areas during the period March,26,

1979,

through March 28, 1979, wasdeemed unwarranted at this time in view of the absence of any intelligencethat adverse human activity was involved in the acqident.
1.5

SIGNIFICANT OPERATING EVENTS
The investigation team reviewed records of operating events related toreactor. trips and events significant enough to be reportable to NRC under the
licensee's Technical Specifications 6.9.1.7,

1-1-376.9.1.8, and 6.9.2.a.

In addition,the team reviewed reports to NRC by the licensee required by 10 CFR 50.55(e)that might have significance when examined in light of this accident.

The

review was conducted to identify potential contributions, precursors,andsimilarities to the accident that occurred on March 28,

1979.

this review and the analysis of leads that were followed

The results ofarepresented below.

Summary of Selected Licensee Reports

1.5.1

The following Construction Deficiency Reports (CDRs)
Reports (LERs)

and. Licensee Event

were deemed to address elements related to the accident:

Occurrence
CDR/LER
77-9c

Related ElementsDate
7/1/77

RCP seal injection determined not to meet singlefailure criterion, with resulting potential of RCPseal failure on loss of offsite power.
Corrective action required modification to provide
redundant seal water sources prior to initial fuel load.
NOTE:

The investigation team reviewed this reportfor background information and did not pursue
it

77-11c9/21/77

further.

GE type SBM switches used in Unit 2 safety-related
equipment determined to be subject to potential cracks
and breaks in the cam followers,

which position contacts

within the switches.
Corrective action required replacement of all questionableswitches prior to core loading.

1-1-38NOTE:The investigation team reviewed this report
after the potential problem with starting
reactor coolant makeup pumps was identified.
See Section 4.18, Details I.

78-020-03L 3/29/78

Containment isolation valves found inoperable due toengagement of manual handwheel pins.
Corrective action required revisions to procedures'to
ensure handwheel pins are disengaged after manual
operation.
NOTE:

The investigation team reviewed this reportfor background information on'previous examplesof valves mispositioned and/or left in an
inoperable condition.

This lead was not pur-sued further, since no transient was involved.
78-021-03L 3/29/78

Reactor trip and ES actuation upon loss of a Vital bus.
The RCS depressurized via the EMOV,
on loss of power.

which failed open

The cause of the depressurization

was not obvious to the operators because the EMOV did
not have actual position indication.

Compensated,pressurizer level indication was also lost, since it

ispowered from the same source.
Equipment modifications subsequent to this event included
providing indication that the EMOV actuating solenoid
had energized (but not that the EMOV was actually openor shut) and a change to the EMOV failure mode so that
it would close on loss of power.

1-1-39NOTE:.

The investigation team reviewed records related
to this event for/background information on
earlier EMOV and pressurizer level indication
problems.

See Sections 1.5.2.1,

1.6.2 and 4.10,

Details I.
78-027-OIT 4/13/78

B&W informs licensee a small-break LOCA on the dischargeside of an RCP is more critical than that previously
analyzed.

Solution to problem requires operator action

to cross-connect and throttle HPI flow during a small-break
LOCA.
NOTE:

The investigation team reviewed this reportfor background information and to gain perspective on the conditioning of operators toensure proper response to the small-break
LOCA.

78-033-OIT 4/23/78

Reactor trips at power, resulting in ES actuation-and
RCS depressurization from excessive OTSG cooldown.
Operator failed to decrease speed of steam-driven
feedwater pump, which was in manual,
OTSG feed had occurred.

until excessive

Preliminary calculations.

indicated a steam bubble may have been introduced into
one or both RCS hot legs.
Corrective action required investigation of the initiating
event, additional tuning of contr'ols,a change in Techni-cal Specifications and revision to operating procedures.
NOTE:

The investigation team reviewed this reportfor background information; but did not pursue it

further, since the transient did not

involve a loss of feedwater or failed open
EMOV.

1-1-40

78-055-03L 9/13/78

Control room emergency air handling system dampers,failed to respond to a close signal.
Corrective action required resolution of construction
problems and re-inspection of air-operated dampers for
operability.
NOTE:

The investigation team reviewed this reportfor background information on potential causesof the buildup of radioactivity in the Unit 1
and Unit.2 control rooms which occurred during
the accident.

78-062-OIT 10/19/78

See Details II.

RCS unidentified leakage exceeds Technical Specification
limits, without shutdown,

due to misinterpretation ofrequirements.
Corrective action required reduction in unidentified
leakage and instruction of personnel.
NOTE:

The investigation team reviewed this reportfor background information. 'See
Details I,

78-065-99X 11/7/78

Section 1.2.3,

for additional problems in this area.

Reactor trip at power and ES actuation on loss of one
steam-driven feedwater pump.

This pump tripped following

loss of one condensate booster pump.

That pump tripped

on a low suction pressure trip; resulting from the lossof two -heater drain pumps,

which supply nearly 30% of

the total feedwater flow.

Pressurizer level indication

was lost during the transient, but licensee calculations
indicated the pressurizer was
not emptied.
I

1-1-41

Licensee reported, "No corrective action is required."NOTE:

The investigation team reviewed the recordsof this event for background information.

78-069-99X 12/2/78

Reactor trip at power and ES actuation due to overfeedingof OTSGs during transfer from the startup to main
feedwater valves.

The main feedwater regulating valve,was found full open and disabled in that position.
Corrective action required procedure revision to prevent reoccurrence.
NOTE:

The investigation team reviewed this reportfor background information on previous ES
actuations and examples of mispositioned
valves, but did not pursue it

further, since

the transient did not involve a loss of feedwater-or faildd open EMOV.
78-073-03L 12/15/78

Sample lines to HP-R-227 suspected to be source ofwater accumulation,

which caused air pump to sieze.

Corrective action required pump replacement,

sampleline heat tracing, periodic inspection for condensation and investigation of appropriate long term
fixes..
NOTE:

The investigation team reviewed this reportfor background information on potential sourcesof water in these lines,
dent.

found during the acci-

See Sequence of Events.

1-1-42

79-001-E

3/21/79

Valve mistakenly left closed following maintenance andclearing of tags, resulting in sulfuric acid discharge
to the river during condensate polisher regeneration.
Corrective action required re-instruction of personnel.
NOTE:

The investigation team reviewed this reportfor background information on additional
examples of misaligned valves.

The lead was

not pursued further, since it did not involvean RCS transient.
79-002-03 12/26/78

Adequate documentation was not retained following NI
calibration surveillance testing to demonstrate Technical
Specification compliance..
Corrective action required re-instruction of personnel.
NOTE:

The investigation team reviewed this report forbackground information on surveillance test
record keeping.

See Section 1.3.4, Details I

for additional problems in this area.
As mentioned at the start of this section,, the above LERs and CDRs wereselected to identify potential contributors to this accident.

No attempt wasmade by the investigation team to perform an in-depth analysis of the LER
history of the licensee.
NOTE:

That analysis is being addressed elsewhere.

NRC has negotiated a modification to the Teknekron,

Inc. contract

titled "Licensee Performance Evaluation,"' that extends their study to include
all .operating B&W plants,

including TMI-2; thereby providing independent

in-depth analyses of the performance of these licensees and the differences
between them.

The methodology developed by Teknekron has been tested on 12

1-1-43

plants and.focuses on LER content to identify trends in parameters that reflect
performance of facility, operating personnel and management.

The results ofthis study should be available within six months and the investigation team
did not attempt to duplicate.that effort.
1.5.2

Review of Selected Prior Trips
1.5.2.1

Previous Event Involving the Electromatic Relief
Valve (EMOV)On March 29,
progress.

1978,

a reactor trip occurred while physics testing was in

The power level was much below 1% prior to the event and there was

essentially no exposure on the core and hence no decay heat.

While reactor

building isolation and cooling surveillance were being performed,

a fuse-blew

in a power supply that supplied power to a pressure transmitter feeding abistable associated with the EMOV.

The EMOV opened, and reactor coolant'

system pressure fell from 2188 psi to approximately 1173 psi at which time
power was reestablished to the pressure transmitter and the EMOV'reclosed.
Saturation conditions in the RCS were not reached (Thot and Tcold were isothermal and the maximum temperature reached was 535'F).
open for approximately 4 minutes.

The valve remained

The reactor tripped on the power/flow tripthat was brought about by the loss of power when the fuse blew.

(The reactor

protection system erroneously sensed that a reactor coolant pump had tripped.)
During the decrease in pressure, high-pressure injection was initiated at
approximately 1600 psi.
Evaluation
A review of computer printouts indicates that the HPI was bypassed.
Pressure recorder traces and records indicate that the HPI system was manually
shut off.

(Ref.

145)

This action was apparently based on the concern ofinjecting sodium hydroxide into the RCS.

Pressure, recorder traces indicatethat RCS pressure was below 1700 psi for approximately 1 hour.

1-1-44It was notedduring the review of computer printouts that high-temperature alarms came inon both safety valves 1.5 minutes after a high outlet'temperature was alarmed
on the EMOV.

In addition,

it was noted that the Reactor Building Sump Pump

was actuated during this event probably as a result of RCDT relief valve
operation.

As a result of this event the licensee recognized (Licensee Event

Report 78-22/99X) that without position indication for the EMOV the-cause of
the depressurization was not obvious to the operators.

(Ref.

42)

Corrective

acti-ons taken in this regard are addressed in Section 1.6.2.2 of Details 1.
NRC review of this event is documented in IE Inspection Report 50-320/78-15,
Paragraph 4.
1.5.2.2

Previous Loss of Feedwater Transient

On November 3,
(Ref.

147)

1978 the reactor tripped due to a loss of feedwater event.

The loss of feedwater occurred when an instrument technician

opened the control'power breaker to the condensate polisher control panel by
mistake.

This action resulted in the closing of the condensate polisher

outlet valves.

When the flow of water was interrupted the condensate booster

pumps and the main feed pumps tripped.
trip was 90%.

The power, level prior to the reactor

Upon loss of feedwater to the steam generators the pressure

increased in the primary system eventually reaching the high pressure trip
setpoint.

The EMOV appears to have functioned normally.

RCDT pressure in-creased from approximately 0.8 psi to 2.0 psi in 15 seconds.
pressure was approximately 11 psi.

'Reactor pressure increased to approximately

2360 psi then reached a minimum of approximately 1940 psi.
returned to approximately 2100 psi.

Maximum RCDT
Pressure thenThere was no initiation of high pressure

injection since pressure did not decrease below the initiation setpoint.
Pressurizer level,

initially at approximately 220 inches reached 'aminimum of

approximately 100 inches and then stabilized at approximately 105 inches.

No

fuel damage was indicated from the review of the event.
Steam generator A startup feedwater level was approximately 140 inches
prior to the event.

Within 30 seconds the level was below 18 inches and in 48

seconds was indicating a "dry" condition (10 inches).

1-1-45The dry condition wasmaintained for approximately 3 minutes and level did not increase above 18
inches for approximately 4 1/2 minutes.

Steam generator B startup feedwaterlevel was approximately 147 inches prior to the reactor trip.

A level of

below 18 inches was experienced in approximately 50 seconds and stayed below
this value for approximately 4 minutes.
Evaluation
Technical Specification 3.4.5 requires that each steam generator shall be
operable with a water level between 18 and 440 inches in Modes 1, 2, 3 and 4.
Since the reactor was in Mode 3 during this occurrence and the steam generator
levels were not maintained above 18 inches the above requirements of Technical
Specification 3.4.5 were not maintained.
At the request of an investigator the licensee was requested to review
records to determine the level of review that the event had received.

This

review indicated that with the exception of a'Reactor Trip Report (single,page) and some hand drawn curves performed by a corporate office engineer
there was no detailed review of the event.

The review by the licensee alsojnoted there were no Site Problem Reports or Site Operations Memoranda written
by B&W as a result of the event.
Technical Specification 6.9.1.9.b requires the licensee report in writing
within 30 days>\the occurrence of event of the type leading to operations in adegraded mode permitted by a limiting condition for operation.

Technical

Specification 3.4.5 requires each steam generator shall be operable with awater level-between 18 and 4-40-inches.

Action statement 3.4.5.b requires thatwith one or more steam generators inoperable due to the water level being outside of limits, be in at least Hot Standby within 6 hours and in Cold Shutdown
within the next 30 hours.
Contrary to the above,the licensee failed to report to NRC the occurrenceof steam generator levels below 18 inches for about 4 minutes,
allowed by the Technical Specifications,1978.

a degraded mode

at about 11:48 p.m. on November 3,

This-item is under consideration as a potential item of noncompliance.

1-1-46

1.6

Corrective Maintenance Practices

1.6.1'

GeneralBecause the accident of March 28,

of one or more components,
maintenance practices.

1979,

involved the failure or malfunction

the investigation team performed a review of plant

The review concentrated on the area of corrective

maintenance since this area addresses the repair of failed or malfunctioning
components as well.as the interface between maintenance and operational
activities.
Corrective maintenance is administered through the use of work requests,
also referred to as "job tickets."

The work request -form allows a plant staff
member to request change ormodification work. This system allows for classifying the work with respect to level of control (QC or non-QC),

change ormodification; radiation or nonradiation, and/or corrective maintenance.
The
work request provides for supervisory review for the issuance of work to the
responsible group as well as a.review for use of appropriate procedures tomeet the defined effort. All work requests are numbered,-and recorded in a
maintenance logbook which contains, among other things, a short title, type ofaction required, and the group to perform the effort. The term "cancel" notedin the log indicates either that a duplicate request is on file or that, after
review of the request, the problem no longer exists.
Maintenance Station Administrative Procedure No. 1026 and Station Corrective
Maintenance Procedure No. 1407-1 define the work request system and responsibilities
for the conduct of corrective maintenance.
the Unit Superintendent,
Maintenance Foreman,group.

Supervisor of Maintenance,

Supervisor of Operations,and Supervisor of Quality Control.

of Maintenance~receives
unit log.

Responsibilities are defined for
Each Uni~t Supervisor

the work request,, which is logged into the specific

The Work request is classified and assigned to the proper maintenance

The work is performed by Metropolitan-Edison personnel or by a contractor.

The maintenance contract work is performed by Catalytic, Inc.

1-1-47The contractmaintenance work is performed under terms,

conditions,

General Maintenance Contract, Metropolitan-Edison,P.O.

and provisions ofNo.

38735.

The same

procedures that cover Metropolitan-Edison maintenance are applicable to thecontractor.

(Int.

193) Catalytic,

Incorporated, maintenance personnel can.,perform maintenance on both safety-related and nonsafety-related systems andequipment.

J

Corrective maintenance is coordinated with and controlled by the Operations personnel from the initiation of the maintenance efforts to its completion.

The work is to start after Technical Specification requirements are met.

Operations is to control,by procedure,(Administrative Procedure 1002) thattagging is accomplished, and returning of the repaired system to service isaccomplished after a retest.
returned through QC,

(Ref.

39)

The signed, completed work request is

if the system is QC classified, to the Maintenance Super-visor for final closeout signatures.

The originator or the 'originators

supervisor is not involved unless the need or results of the work requires his
expertise.
The Operations Shift Foreman approves the commencement of work.

The

Maintenance Foreman is responsible for having Operations perform the tagging
and reporting status, as work progresses,

to Operations.

Thus, Maintenance initiates tag control,
for the use of the tags.

and Operations maintains control

The maintenance and documentation of tags and tag

status is kept by Operations to ensure operational

knowledge of the plantcomponent/system conditions at all times.
The maintenance function is carried out by each discipline of maintenanceunder a supervisor who has foremen who ultimately assign the crafts to performthe work.

The day shift consists of supervisor,

foreman,and craftsmen of thesame discipline; the remaining two shifts are composed of a foreman in one
discipline with craftsmen from the three major disciplines to perform thework.

The personnel are assigned specifically to Unit 1 or to Unit 2, but

perform work as-assigned to either unit.

During active periods such as an

1-1-48outage it

is, not unusual for most craftsmen to be assigned to the unit in an

outage status.
Operations maintains an ongoing list, called a "no name outage," for highpriority work to be performed in the event of an unplanned outage of an operating
plant.

At the time of a plant trip, maintenance reviews the listed work andplans for work to be performed.
1.6.2

Selected Component MaintenanceHistory Immediately
Prior to March 28, 1979

1.6.2.1

Main Feedwater System

The inspector reviewed selected files located in the Unit 2 Mechanical
Maintenance Office on May 8, 1979.

Information in the files on the feedwater (system included work performed on feedwater pump FW-P-IB,
WRCOI80,

December 13,

1978,

work request No.

related to a stuck shaft problem, and WR1357 which

involved the feedwater pump lube oil system.

The file contents did not indicatethat the feedwater system was not in acceptable operational status prior to
the March 28,

1979,

event.•

A work request WR,0334,

which addresses FW-P-IB

turning gear, remains open.
1.6.2.2

Electromatic Relief Valve.

1.6.2.2.1 Valve Control Revisions
Information pertaining to the Unit 2 installed Electromatic pressurizer
relief valve was reviewed.

The -relief Valve is an electrically-actuated pilot,operated pressure relief device.
Inc.

It

This valve is manufactured by Dresser Industries,

is installed in conjunction with a pressure-sensitive element having

set pressure values-for electrical signal actuation of a solenoid that actuates
a pilot valve allowing for automatic pressure relief.

1-1-49The valve had leaked inplace during the hot functional test of the facility.

Corrective maintenance

was performed and testing was accomplished.
A QC surVeillance report (77-266,

October 27,

of'the electromatic pressurizer relief valve.
electr'ical checks,of release and,1977) documents the testing(Ref.

43) The bench test included

pressure application to 2300 psig, and electrical actuation.

after release,leak checks.

The original wiring of the installed Unit 2 electromatic pressurizer
relief valve was modified (ECM No.

S-5934,

EMOV),

to allow visual control room

indication of power across the pilot operator solenoid.

This does not provide

the operator with valve position, only demand position.

The loss of power

logic was also changed to have the valve remain in the nonactuated state upona loss of electrical power.

The original installation did not have thesefeatures and had resulted in the actuation of the valve during a loss of poweroccurrence on March 29, 1978. The followup action, to the March 29, 1978, event
included the performance of corrective action as stated in LER 78-21/3L.
Basedon discussions with the licensee, the change included the indication of power
through the pilot valve solenoid, which appears to function properly.
The investigator reviewed B&R Field Questionnaire,

FQ, No.

2529, April 6,

1978, which addresses ECM S-5934 which specified the circuit revisions to be
made.

The FQ,

page 2,out."

Documentation was notIavailable about how and when the retest was done

and the results.
1615,

states that "Unit wired up per the prints and tested

The Shift Foreman's Log Book, May 6,, 1978, contains an entry

page 488, which indicates the "I&C checked out RC-R2 indicating light on

Panel 4."
Evaluation
Station Administrative Procedure 1021,

Plant Modifications,

Revision 5,

June 28, 1978, paragraph 4.1.a, requires that "the Supervisor of Maintenancereceives a work request which represents a change/modification to a system orcomponent."

Documentation of a Work Request was not available.

1-1-50The originalFQ does. not designate the change as either minor or major.

The lack of designa-tion of the type of change and the absence of a work request is contrary to
the requirements of Station Administrative Procedure 1021,
This procedure,

iffollowed,

Plant Modification.

designates if the change/modification must be

included in operator training.

Failure to follow this procedure is being con-sidered as a possible item of in noncompliance with Technical Specification
6.8.1.a.
The technical specification 6.8.1.a requires written procedures shall be
implemented covering activities referenced by applicable procedures recommendedin Appendix A of Regulatory Guide 1.33, November 1972.

Appendix A of Regulatory

Guide 1.33, addresses typical safety related activites to be covered by written
procedures.

Paragraph 9a of RG 1.33 states that "maintenance that can affectthe performance of a safety-related equipment should be properly planned andperformed in accordance with written procedure... appropriate to circumstances".
Station Administrative Procedure 1021,
June 28,

Plant Modification,

Revision 5,

1979, paragraph 4.1a) requires that "the Supervisor of Maintenancereceives a work request which represents a change/modification to a system orcomponent".
Contrary to the above there was no work request for the work performed incompleting ECM S-5934 which specified changes to the pressurizer relief valve
RC-R2 including the addition of an indicating light and a revised logic onloss of power actuation of the valve.
1.6.2.2.2 EMOV Locations and Handling
During discussions with the licensee and contractors about the pressurizer
electromatic relief valve,

it appeared that both the Unit 1 and Unit 2 valves

had been used on Unit 1 pressurizer during the course of earlier Unit 1 operation.
The valves are interchangeable.

Both are Dresser Type 2-1/2 31533VX-30.

1-1-51

The investigator reviewed Drawing No.
shows Serial No.

2841-60501,

BL-08905 for B&W Contract No.

620-005,

March 20,

1979, which(Unit 1).

The valve

was visually inspected in place and the serial number was checked by theinvestigator to be BL-8905.

The serial number for the installed Unit 2 valve

is believed to be BN-4233.

The Unit 2 B&W Contract Number is 620-006.

Engineers and Constructors,

Inc.,

Order No.

022660LS,

Uniteddoes not list the serial

number of the valve in place on Unit 2.

The investigator inspected the third

electromatic valve located in storage.

The Serial Number of this valve is

BSO-3989.

Itis concluded that electromatic valve Serial Number BN-4233 is on

the Unit 2 pressurizer, the Unit for which it was originally ordered.
Documentation reviewed indicates that valve BN-4233 was sent to the
manufacturer for inspection, refurbishing,gasket replacement,

reassembly,

andtesting as documented in Jersey Central Power and Light Purchase Order No.
C-0224, June 8,

1976.

Evaluation
The valve history indicates that both electromatic valves have been
refurbished and each has been installed on Unit 1 pressurizer.

The current

status is that the original valves as purchased under the two B&W contracts
are installed on their respective pressurizers.
1.6.2.3

EMOV Block Valve.

The pressurizer electromatic relief valve upstream block valve, RC-V2,
had work performed on it on September 14,1166; Ref. 45).

1977.

(Machine History File, WR

The motor operator was removed to allow repacking the valves.

The valve was cycled.

Discussions with licensee personnel indicate that there

has been a concern that the block valve could stick shut or open if
often.

1-1-52

used too1.6.2.4

Emergency Feedwater Pumps.

The machine history file on the electrical emergency feedwater pumps,
EF-P-2A and 2B, did not indicate a problem relative to the event of March 28,
1979.

The most recent history indicated that motor windings and connecting

electrical cables were checked after an event involving steam leakage in thepump area.

The steam leakage was due to the atmospheric dump valves (MS-V3A

and B) bellows (MS-U7A and B) failing during turbine trip testing on January
15,

1979,

and subsequent reactor trip and loss of condenser vacuum.

Reactor Trip Report January 15,

1979,

Ref. 6).

(TMI

The steam was reported to havefilled the M-20 area where the pressurizer breaker, silicon controlled rectifier(SCR)

panels are located as well as the steam-driven emergency feedwater pump,

EF-P-l.

The adjacent area where EF-P-2A and 2B are located also contained asteam environment.
Checks and repairs were made to equipment in.these areas.
the pumps were performed by Catalytic,
1979, WR C0711.

(Ref.

7)

Inc.

The checks on

The work was completed on January 23,

Checks were made on electrical reactor building penetra-

tions in the M-20 area for heat and moisture damage.

(WR C0718; Ref. 8).

Thiswork request and the Maintenance Procedure Format and Approval form TMI-84-2-78
were reviewed.

The documentation indicates that visual checks were completed

January 16-17,

1979.

Review of the documentation indicates that no problems

were found during these visual checks.
1.6.2.5

Pressurizer Heaters.

The Licensee performed maintenance (WR 709) involving the repair ofpressurizer heater panels.

This effort resulted in the repair of the pressurizer

heater transformers and repairs -to cabinet 324.
not specific.

The repair description was

Discussions with licensee personnel indicate that there havebeen problems with the pressurizer heater breakers.

Itis believed that the

problem is associated with the high temperature in the M-20 area due to theproximity of the main steam lines in this area.

1-1-53

It was further noted thatcontrol room indication is not available for individual breakers which aresubsets of main breakers indicated in the control room.

The primary control

room indication of individual heater breaker problems is indirect in that itis related to pressurizer pressure recovery rate.
1.6.2.6

Waste Tank Rupture Discs.

The investigator reviewed the Unit 2 maintenance request log for the
past year and discussed with the mechanical maintenance department personnel
what they knew of a work request to correct rupture discs for auxiliary building sump tank (ABST)

WDL-T-5 and miscellaneous waste tank WDL-T-2.

for Apparatus to be taken out of service, control No.
found in the control room active tag log.

Application

2385, May 10,

1978, wasThis document notes the replacementof a rupture disc as work to be done.

The document does not reference a work

request number in the blank provided.

A work request No.

replace ruptured diaphragm," February 8,
ItC1117,"WDL-U224,

1979, is currently an open work item.

was stated by Licensee personnel that there were problems with the rupture

discs for these tanks that may possibly be traced to the preoperational phaseof TMI-2.

(The discs are manufactured by Zook Enterprises and are rated at

19.6,psig ± 2% at 104 0 F.)

An order for spare discs ('P.O.

February 11, 1976) is in existence (Ref.
for the tanks.
9,

JCP 0C211,

The discs are interchangeable

The information indicates that tank WDL-T-5 had a ruptured

disc on March 28, 1979.
Rev.

10).

No.

(The plan drawings,

B&R Drawing No.

2485,

Radwaste Disposal - Miscellaneous Liquid Auxiliary Building Sump Tank

Plan and Section, and B&R Drawing No.
1.6.2.7

2482; Rev.

13)

Emergency Feedwater Valves.

The investigator reviewed work request No.

756, January 16,

1979,requiring

maintenance checks of emergency feedwater valves 12A and B, 32A and B, and 33A
and B.

(The 32 valves bypass the 11 valves and the 33 valves bypass 11 and 12

valves; Burns & Roe Drawing 2005; Ref. 50) The work request was returned to
the Operations Shift Foreman on January 24, 1979.

1-1-54The information contained inthe work request indicated that the valves remained to be cycled to complete
the assigned work.

The information available did not include Meggering of valve

operator cables to validate acceptance criteria contained in the work request
procedure.

(Station Preventative Maintenance Procedure E-2,

of Insulation, Motors and Cables,

Ref. 51).

Because of the apparent incomplete

data and the lack of documented cycling of the valves,
tagging documentation.
Catalytic,

Inc.,

Dielectric Checkthe investigator requested

Licensee personnel did not locate the tagging document.

did locate copies of Application for "Apparatus to be Taken

Out of Service" No.

4030, January 18,

1979, and No. 4044,

Janury 20,

1979.

Both applications were incomplete and would require Operations personnel for
proper completion.

The investigator did not find suitable documentation

as to whether valve cycle, Megger checks, tagging clearance,

and acceptancetests to return the valves to service were performed for the listed valves.
The Unit 2 maintenance log carries this work request in an open status as ofJune 20,

1979.

Evaluation
Station Preventative Maintenance Procedure E-2, paragraph 4, requires theuse of TMI Administrative Procedure No.
Megger acceptance criteria.
acceptable values.

1002, Tagging.

Procedure E-2 presents

The work request documentation failed to validate

In that the 12A and 12B valves are known to have been

operated and documentation is lacking for retest, satisfaction of acceptancecriteria, and tagging clearance requirements,

this item is under consideration

as a potential item of noncompliance.
Procedure 1002, Revision 14, paragraph 4 "Restoration to Service",states:"Restoration to Service:"When work has been completed,grounds removed,-and all--parties that havereceived clearance have reported clear, and the Shift Foreman has deemedthat the equipment is ready for service, the Switching and Tagging CRO,shall so note on the Clearance Control Document.
include the time,

This notation willdate and a statement as to the status and position of

I-1-55the isolated equipment....
application,
1.6.2.8

The Shift Foreman shall then sign the originalindicating that the equipment is approved for operation."

Pressurizer Safety Valves.

The investigator reviewed licensee records pertaining to pressurizer
safety valves RC-RIA and lB.

WR 340, March 25,

lift

These results indicate that the settings are

tests of these valves.

1977, contains results forwithin the acceptance criteria +1% of the 2450 psig setpoint (Int.
1.6.3

139).

References
The references used in this report section are:109, 99, 55,
121,

100,

122, 123,

110, 111, 112,

124,

125,

126,

113,

127,

114,

128,

I-1-56

115,

129,

116,

130.

117,

118, 119,

120,

2.
2.1

OPERATOR ACTIONS DURING ACCIDENT

GeneralIn this section, the use of the term "operator" is being applied liberally

to address actions by those licensee staff members,
personnel,

including-management level

who are assigned routinely to shift operation responsibility.

Sub-

sections 2.2 - 2.13 will identify the major procedural requirements applicableduring different phases of the accident,

and the manner in which these require-ments were addressed by the operators.
2.2

Operator Actions Following Loss of Main Feedwater
Through records review, analysis,

and-peisonnel-interviews,

it was deter-

mined that the immediate and following actions prescribed in the procedure(2202-2.2,

Loss of S/G Feed, Revision 3, October 13,

1978; Ref. 26) for theloss of Main Feedwater flow to both OTSGs were appropriately addressed as discussed below:oWhen loss of FW is due to loss of both feed pumps,

trip the reactor.

The reactor tripped automatically due to high RCS pressure and the fullrod insertion was verified by the operators immediately after the trip.
The manual reactor trip was delayed approximately 20 minutes.

The inter-views revealed that this action was delayed because of the rapid sequenceof events following the turbine/reactor trip.
in Section 2.4 of this report (Int.
o

3,

15,

The reactor trip is treated9).

Verify that turbine trip and stop valves closed.
Turbine trip procedure (2202-2.2; Ref. 28) covers this activity.
treated in detail in Section 2.3 of this report.

1-2-1This isoVerify that EF-P-l,

EF-P-2A,

and EF-P-2B start as evidenced by pump dis-charge pressures.
The operator noted the pumps as operating and "normal"
0(Int. 4, 9, 61).

Verify that emergency feedwater valves (EF-VIIA and VllB) are in automaticand controlling OTSG level at 30 inches on the SU range indication.
The operator actions relative to this requirement are described in detailin Section 2.3 of this report.

Evaluation
The investigation concluded-that the major procedural requirements applicableto this event were appropriately addressed in the actions taken by the operators
immediately following a loss of all main feedwater to the OTSGs.

This conclusion

does not include those items which are covered in Section 2.3 and 2.4 of thisreport.
2.3

Operator Actions Following Turbine Trip
The TMI-2 unit experienced an analyzed turbine trip as a result of theloss of main feedwater and experienced a loss of all feedwater for approximately
8 minutes because the emergency feedwater header isolation valves were apparently
in the closed position at the onset of the turbine trip transient.
Through records review, analysis, and personnel interviews,

it was determinedthat the immediate and followup actions for a turbine trip on (Emergency
Procedure 2203-2.2, Turbine Trip, Revision 7,

October 25,

1978; Ref.

28) were

addressed by the operators as appropriate including:
2.3.1
oImmediate Operator Actions
Verify that the turbine stop valves are closed and generator breakers andfield breakers are open.

Verify that the start of the seal oil backup pump,

1-2-2

the turbine gear oil pump and the bearing lift

pumps and the closure of theextraction steam valves.
The operator noted that the turbine trip was normal with the exception of
one turbine stop valve which did not indicate closed.
control valve was verified closed and, additionally,
tripped manually.
o

(Int.

The associatedthe turbine was4, 9, 61)

Monitor pressurizer level,

RC pressure,and RC temperature.

After approximately 2 minutes, the high pressure injection (HPI)

initiated.

As the pressure decreased below 1600 psig, the pressurizer level continued
to increase to full scale.

The operators were not surprised by the HPI

actuation but they did not understandy the high pressurizer level and thecontinuing low RCS pressure.

The initiation of ES is covered in moredetail in Section 2.5 of this report.

(Int.

5, 6,

17,

38, 3, 15,

4, 9,

61)
o

Notify HP/Chemistry to sample RC letdown for dose equivalent iodine

between 2 and 6 hours after a power change of greater than 15% power in a one
hour power per Surveillance Procedure 2304-302.
Upon the announcement of the turbine reactor trip, the chemists on duty
insure that letdown sample is placed in recycle in preparation for
obtaining a RCS/letdown system sample.
o

(Int.

5,

17, 38,2, 14,In the event that the vacuum is lost in the main condenser,

37)

verifyatmospheric reliefs are controllling header pressure at 1010 psig and reducing
reactor power to zero (hot standby).

Add required makeup to tank as Tave isreduced to 532 0 F (885 psig OTSG pressure).
Initially following the turbine trip the main steam code safeties lifted,
OTSG pressures reached a maximum 1070 psig, and the operator noted themain steam turbine bypass valves functioning to maintain steam header
pressure.

1-2-3

The operator interviews indicated that the main turbine bypass valves(MS-V25A,

25B,

26A, and 26B) were placed in manual and cracked open

(apparently at 1.2 minutes) when the operators perceived the values were
not responding adequately to the increasing OTSG pressures.

(Int. 4, 9,61)o

If turbine trip is due to loss of both feed pumps,

verify that emergency

feed pumps have started and are delivering water to the OTSG.
The operator noted that all three emergency feed pumps had started andappeared normal with the OTSG levels decreasing and the level control
valves EF-VllA and liB traveling (dualindicating lights).

The controllers

for the emergency feedwater control valves (EF-VlIA and VllB) are locatedabove the lights for the EF-Vl2A and V12B valves on the same panel.

The

operator did not note the fact that the EF-Vl2A and V12B valves wereclosed at that time.

(Int. 4, 9, 61)

The operator watched the steam generator levels for a sign of feedwater
and water level increase.

After observing no level change the operator

quickly reviewed the valve lineup, noting the emergency feedwater control
valves (EF-VIIA and VlIB) opening.

The operator placed the controls inmanual and raised the demand signals to full open in the attempt toincrease the rate of emergency feedwater delivery to the OTSGs.
Evaluation
At approximately 5 minutes into the turbine trip transient, the operator
noted that the emergency feedwater level control valves were continuing totravel to the open position.
operate.
It

The valves were not supplied adequate air to

The valves fail "as is"and only respond if air pressure is available.

can be reasonably postulated that the failure of the condensate polisher

outlet valves to the closed position (loss of instrument air pressure) and thefailure of the emergency feedwater valves to respond normally (loss of instrument
air pressure) were related.

Operator interviews revealed that the valves did

1-2-4not appear to operate properly.

The valves were expected to open fully shortly

after the OTSG levels decreased below 30 inches and the valves were continuing
to travel at about 0405 hours.
2.3.2

Followup Operator Actions
The operator noted the continuing low levels (approximately 10 inches on

the startup range, which was assumed to be dry).
At this time in the event (approximately 5 minutes) after the OTSG levels
had remained below 30 inches on the startup range from 30-40 seconds after thestart of the accident, the feedwater control valves had sufficient time tostroke fully open from their normally closed position.
The Unit 1 Shift Supervisor arrived in the area at about 0408 hours,
noted the low steam generator levels,
delivery, of feedwater.

and stated the concern of no apparent

Simultaneously, the operator had noted the emergency

feedwater header isolation valves (EF-VI2A and 12B) closed.

The operator

announced the condition,and simultaneously opened both feedwater header

isolation valves (EF-VI2A

and VI2B).

The operator stated that he had lookedat the panel valve lineup twice before discovering the valves to be closed.
The first scan of the panel was made standing close to the panel and leaningover the console,possibly obstructing his view of the lower valve (EF-VI2B).

The upper valve (EF-VI2A)

indicator light (green for closed) was covered by acaution tag hanging from the main feedwater pump lB miniflow valve (V16B)
control station.
NOTE:

The caution tag indicated that the control for the main feedpump

IB miniflow valve (Vl6B)

should remain in MANUAL due to a flow instrument

problem.

The tag was issued on March 20, 1979,(CO-V53D)

repair (Work Order 1735).

(Int. 4, 9, 61,

A review of Administrative Procedure 1037,
Caution and Do-Not-Operate (DNO).Tags,

for the transmitter valve
5,17,38, 3,

15).

Revision 1, August 23, 1978,

revealed that no formal guidance isgiven concerning the method of hanging tags to prevent inadvertent covering of

1-2-5

other information devices (lights,

identification tags, etc.) or impeding theoperation of other equipment or controls.
Immediately following the opening of the emergency feedwater header isolation valves (EF-VI2A and VI2B),

the operator noted flow noises from a Loose

Parts Monitor channel (apparently from the main steam area and/or A OTSG),

anincrease in the OTSG pressures above the low of approximately 780 psig (8 minutes),
and a decrease in the RCS temperature during the subsequent 12 minutes from
597'F to 555'F.

(Int. 4, 9, 61)

Evaluation
Technical Specification 3.7.1.2 requires that three independent steam
generator emergency feedwater EFW pumps and associated flowpaths be operable
in Modes 1, 2, and 3 (OTSG pressure greater than 800 psig).

The specific.

action does allow one emergency feedwater system to be inoperable,

providedthe inoperable system is restored to OPERABLE status within 72 hours or. thereactor is to be placed in hot shutdown within the next 12 hours.
0400 hours on March 28,

Prior to1979, the licensee did not provide a formal routine

shift check of engineered safeguards equipment,
emergency feedwater pumps and valves.

including the status of theAdditionally no alarms were provided toannunciate the emergency feedwater system in the inoperable condition.
The plant was operated with the emergency feedwater system in the operable
condition as a result of the emergency feedwater isolation valves (EF-VI2A and
12B) being closed for a period of time prior to 0408 hrs on March 28,

1979.

The continued plant operation with the emergency feedwater header isolation
valves (EF-VI2A and 12B)in the closed position is an apparent breakdown in

the licensee controls over the operability of safety related equipment,.and isbeing considered as a potential item of noncompliance pursuant to Technical
Specification 3.7.1.2.
A review of plant records and charts revealed that the recovery of the A
OTSG level to 30 inches at 0420 hrs and the B OTSG level to 30 inches at 0425
hrs was delayed because the emergency feedwater flow was apparently limited or

1-2-6

ineffective in reestablishing the-OTSG levels even with full emergency feedwater flow available.

(All three emergency feedwater pumps operating.)

The steam generator pressures did not recover to the turbine bypass valve
setpoint of approximately 1010 psig (or the safety valve setpoint of 1070) or
provide significant cooling rates as Tavg decreased from 597°F to 547°F.

The

difference between the actual steam generator pressures and the RCS averagetemperature which determined OTSG saturation pressure, indicate that the OTSGs
were "1dry."'
During this period, between 0408 hrs and 0416 hrs, all three emergency
feedwater pumps were operating with discharge pressures greater than OTSG pressure.

Their discharge pressures show distinct step changes simultaneously.

This uniform variation in EFW pump discharge pressures indicated manual throttling or simultaneous stepwise movement of two automatic valves since OTSG
levels were less than 30 inches.
One operator interview revealed that the emergency feedwater flows weremaintained at the maximum amount (not throttled) until levels were responding
and recovering.
water (EFW)

However, another interview indicated that the emergency feed-

was throttled to limit the thermal shock of the OTSGs and minimize

the effect on the RCS pressure and temperature.

(Int. 4, 9, 61)

The investigation concluded that the extended times (13 minutes for the A
OTSG and 17 minutes for the B OTSG)

for the restoration of minimum OTSG levels

minimum to more than 18 inches (30 inches normal low limit setpoint) was theresult of throttling the EFW flow.

The conclusion was based on operator inter-views and an analysis of EFW, OTSG,

and RCS parameters.

At 0532 hrs during the accident, the A OTSG level decreased to 10 inches(minimum level indication) beteen 0532 hrs and 0543 hrs, at which time the A
OTSG level was increased to about 50% on the operating range to support natural
circulation (Ref.

Section 210).

The failure to maintain OTSG A level at 30

inches was due to the feedwater level control valve (EF-12A)

1-2-7

in manual andshrinkage due to a decrease in the RCS temperature 565'F to 535'F.

(Int. 4,9, 61).
Emergency Procedure 2203-2.2, Turbine Trip, Step 3.2 and 3.4 requires theoperator to adjust feedwater flow to control OTSG levels at 30 inches (on thestartup range).
Contrary to this requirement, the A and B OTSG levels decreased below 18
inches (reaching as minimum of 10 and 12 inches) following the trip until
approximately 0420 hrs and 0425 hrs respectively,of 10 inches) between 0532 hrs and 0543

below 18 inches (reaching a minimum,hrs.

and the A OTSG level was

This failure to maintain required steam generator level,

emergency feedwater available,with adequate

is being considered as a potential item ofnoncompliance pursuant to Technical Specification 6.8.1.a.
Additional Operator Actions

2.3.3

After the OTSG levels were established at 30 inches on the startup range,
the operator continued to maintain the level in manual control using the
emergency feedwater (EFW)

system as required with suction for the EFW pumps

from the condensate storage tank between 0400 hrs and 0459 hrs.

The suction

for the pump was then transferred to the condensate pump operating.
condensate pump (CO-PlB)
the event.

The

continued to operate during the initial minutes of(Computer Memory Trip Review and Int. 4,

9,

61)

The turbine-driven emergency feedwater pump was stopped at approximately
0426 hrs after the operator noted the "B" OTSG level increasing slowly above
the low-level

limits.

The shutdown of the turbine driven EFW pump placed the system into ACTION"A" of Technical Specification 3.7.1.2.
14, August 25,

1978 (Unit Heatup),

Operating Procedure 2102-1.1,

Step 4.53, required the EFW system to beplaced in operation at 800 psig in the OTSG.
August 29,

1979 (Unit Cooldown),

Revision

Procedure 2102-3.2,

Revision 10,

Step 4.14a, removes the EFW system fromservice to 600 psig in the OTSG.

1-2-8

The motor driven emergency feedwater pump (EF-P2B)

was similarily stopped

at 0436 hrs.
The emergency feedwater pump (EF-P2A)

was tripped at 0711 hrs after the

OTSG levels were near the 55-65% level and the OTSG pressures were less than
600 psig.
hrs,

During the operation of the RCP (RC-P2B)

for 19 minutes at 0654

it was noted that the "B".OTSG pressure increased to about 750 psig but

not above the Technical Specification of 800 psig, with all of the EFW pumps
secured.
The emergency feedwater pump (EF-P2A)

was started at 0735 hrs and operated

until 0842 hrs to feed the "A" OTSG as needed to maintain 50% level on theoperating range.
The "A" OTSG level was raised at 1000 hrs to 94-97% on the operating
range using the water from the hotwell via the condensate pump discharge/booster
pump suction pressure to enhance decay heat removal.

The "A" OTSG pressureduring that period decreased from 140 psig to 20 psig and then to 0 psig by
1200 hrs.

The emergency feedwater pump (EF-P2A)

was started and operated for

8 minutes at 1109 hrs to assist in feeding the "A" OTSG which was near thedischarge pressure of. the condensate header.
The "B" OTSG level was increased from 60% to 95% on the operating range
at 1600 hrs using the emergency feedwater pump (EF-P2A)

at 1534 hrs for 18

minutes in preparation for the continued plant shutdown.
Evaluation
The stopping of the steam driven emergency feedwater pump (EF-Pl)
hrs and the motor-driven emergency feedwater pump (EF-P2B)

at 0426(secondary emergency

feedwater'pump stopped) by placing the control switches for the steam supply
valve (MS-V207)

and the pump switch (EF-P2B)(nonautomatic)

at 1040 psig in the OTSGs (greater than 600 psig) is a violation

to the PULL-TO-LOCK positionof the operating procedure for unit cooldown, Step 4.14a.
remained above 800 psig until approximately 0545 hrs.

1-2-9

The A OTSG pressure

The emergency feedwater

pumps start automatically upon loss of two main feedwater pumps if the control
switches are in the NORMAL-AFTER-STOP position.
Technical Specification 3.7.1.2 requires that the Emergency Feedwater System
in Modes 1, 2, and 3 above 800 psig OTSG pressure with two flow

be "OPERABLE"

paths, two motor-driven emergency feedwater pumps,

and one steam-driven emergency

The failure to maintain the emergency feedwater system infeedwater pump.

the fully operable status subsequent to 0426 hrs (turbine-driven emergency
feedwater pump-EF-Pl)

stopped, and 0436 hrs (motor-driven emergency feedwater

stopped, by placing the control switches to the non-automatc

pump-EF-P2B)(PULL-TO-LOCK)

position, at 1040 psig in the OTSG (greater than 800 psig) is

being considered as a potential item of noncompliance.
2.4

Operator Action Following Reactor Trip
The TMI-2 unit experienced a routine reactor trip as a result of a lossof the turbine when the main feed pumps tripped on loss of suction pressure.
The reactor continued to operate in accordance with control system design
at approximately full power for 8-9 seconds when the reactor was tripped as
the RCS pressure increased above 2355 psig, the high-pressure reactor trip
setpoint.

(2202-1.1,

Revision 5, October 6, 1978,

Reactor Trip,

step l.l.e;

Ref. 31)
Through records review, analysis,

and personnel interviews,

it was deter-

mined that the immediate and followup actions for a reactor trip (Emergency
Procedure 2202-1.1,

Revision 6, October 25, 1978,

approximately addressed by the operators,oReactor Trip; Ref. 30); were

including:Manually trip the reactor.
The manual reactor trip was delayed approximately 20 minutes, due tothe sequence of events.

(Int.

3,

1-2-1015)

oVerify that all "in-limit" lights are actuated (except group 8).
All "in-limit" lights were actuated.

o

3,

15,

4, 9, 61)

Close letdown isolation valve MU-V376.
The letdown isolation valve (MU-V376)

(nt.
o

(Int.

4,

9,

was closed by the operator.

61)

Start second makeup pump and open MU-V-16B as necessary to maintain 100
inches in the pressurizer.

The operator started the second makeup pump (MU-PIA)
the suction valve (DH-V5A)

and also opened

from the Borated Water Storage Tank (BWST)

for the two operating makeup pumps (MU-PlA

and PlB)in anticipationof the expected low pressurizer level and low RCS pressure transientthat follows a reactor.trip.

The integrated control system normally

returns the RCS average temperature from the operating (greater than
15 percent) value (582'F)

to the initial hot shutdown value (549'F),

which produces a decrease in the pressurizer level from 240 inches
to 85 inches (about 5 inches per 'F) and an RCS pressure at or near
1640 psig (ESF
events,initiation setpoint).

During the rapidsequence ofthe operators did not realize that the pressurizer level

decreased to only 158 inches and RCS pressure was stabilized forabout 60 seconds at approximately 1500 psig.

(Int.

5,

6,

7,

38,

3,

15)
o

Verify that pressurizer heaters are off at 80 inches. in the pressurizer.

[The pressurizer minimum level was 158 inches.]oAnnounce reactor trip on page system.

The Shift Supervisor announced turbine trip-reactor trip.
7,

60,

61,

96)

1-2-11(Int.

TMI 6,

oMonitor makeup tank level and maintain level higher than 55 inches by using
waste transfer pumps (WDL-P5A/B) and feeding from an RCBT with boron concentration equal to or greater than RCS concentration.
The operator opened the suction valve (DH-V5A)

to the BWST to provide

makeup pump suction (MU-PlA and PlB) in anticipation of the expectedpressurizer level transient.
oVerify that the pressurizer heaters and spray have returned RCS pressure
to normal operating pressure of 2155 psig.
The pressurizer heaters and spray were returned to AUTO by the operator
after the reactor trip to provide normal pressure control and responseto the transient.

The spray valve and the heaters were being main-tained in manual (2103-1.3,
1977,

Pressurizer Operation,

Revision 1, November 1,

step 4.2.4) to continuously recirculate the pressurizer.

Boron

concentration was consistently higher than that in the RCS becauseof the leak through the pressurizer relief valves (exhaust tailpipe
temperatures indicating consistently above 180 0 F) and concentrating
the boron in the pressurizer (Ref.

53).

The operators indicated that

this boron concentration in the pressurizer was often as much as200-300 ppm higher than that in the RCS.

The RCS pressure did notrecover to normal (2155 psig) during the event'under normal pressure
control.
o

(Int.

5,

17, 38, 3,

15, 2,

14, 37)

Reduce pressurizer level setpoint to 100 inches (25%).
The pressurizer level increased to full scale or near full scale
within 6 minutes.
(Int. 5,o

6, 17,The operators did not understand the conditions.

38, 3,

15, 2,

Verify normal electrical lineup,

14, 37)i.e.,

no substation or inplant

distribution breakers are open (except generator breakers).

I-2-12

The operators verified normal electrical system conditions.
(Int. 4,o9, 61)

Check that all RMS channels are normal and that no unplanned or uncontrolled
radioactive release is in progress.
The only radiation alarm that annunciated shortly after the reactor
trip [at approximately 0420 (ALERT)

and 0500 (HIGH)] was theintermediate closed cooling system monitor for the letdown heatexchangers.

This alarm was not considered unusual because of

its location near the containment sump, which was filled fromthe RCDT rupture disc discharge.

Previous experience had also

noted that this monitor alarmed following a unit trip due toincreased reactor coolant activity levels (crud or iodine spikes).
The operater did note that the high alarm was annunciated-rather
than the alert alarm as occurred in the past, but the alarm wasstill considered normal.

(Int.

5,

6,

17, 2,

14,37, 3,

15, 4,

9, 61)oCompute shutdown margin calculations per 2103-1.9 (Reactivity Balance
Calculations).

If shutdown margin is less than 1%, boron should be added

to the RCS.
Based on the rod worths and positions at the time of the reactor
trip (see Section 1.33,

Details I),

the shutdown margin was

satisfactory with all rods inserted and 1026 ppm RCS boron
concentration.
o

(Int.

5, 6, 17,

97)

If reactor startup is not iontended within four hours,

raise OTSG level to97-99% on the operating range using the feedwater valve bypass and thetubesheet drains for level control.
A return to power was initially planned following the trip, but
the system conditions quickly deteriorated.

1-2-13The OTSG levels

were increased to the 97-99% level later in the day.
at 1000 hrs and the B OTSG at 1600 hrs).

(Ref.

(A OTSG

Section 2.4 ofthis report)oNotify HP/Chemistry to sample RC letdown for dose equivalent iodine between
2 and 6 hours after power change greater than 15% within one hour period
per Surveillance Procedure 2304-3D2.
Following the announcement of the reactor trip, the onshift chemistry
technicians placed the RCS letdown sample line on recirculation andobtained a sample.

(Int.

5,

6,

17, 38, 3,

15)

Evaluation
The'investigation concluded that the major procedural requirements applicableto this event were appropriately addressed in the actions taken by the operators
immediately following the reactor trip on high pressure.
2.5

Operator Actions Following ES Actuation
The high-pressure safety injection initiated at a reactor coolant systempressure of approximately 1600 psig (decreasing)following the reactor-trip-at

approximately two minutes after the start of the accident.
The automatic initiation of safety injection due to low RCS pressure
following a turbine-reactor trip was not foreign to the operators,
previous plant experience.

based on

The turbine trip transient the TMI-2 unit wascharacterized by the operators as "severe",resulting in a substantial decreasein pressurizer level (expected to decrease below 80 inches and trip the pressurizer heaters) and RCS pressure (expected to decrease below the ESF actuation
trip point of appproximately 1640 psig).
15,(Int. 5, 6,

17, 38, 2,

14, 37,

Through records review, analysis, and personnel interviews,

it was3,4, 9, 61)

determined that, with two major exceptions,

1-2-14the majority of the immediate andfollowup actions for the procedure Loss of Reactor Coolant/Reactor Coolant System
Pressure (2202-1.3,
addressed.

Revision 8, May 12,

1978,

Sections A and B),

were appropriately

These procedure sections addressed actions considering "automatic"

initiation of the engineered safety feature systems, which occurred at 0406
hrs (Ref.

34).

The symptoms of a significant leak or rupture resulting in an automatic
initiation of the ES systems include a rapid continuing decrease of reactor
coolant pressure', with the following indications:a.

Lo alarm 2055 psig.

b.

Lo-lo alarm 1700 psig.

c.

Safety injection actuation at 1640 psig.

d.

Lo alarm 200".

e.

Lo-lo alarm 80" (interlock heater shutoff).

Within two minutes (initiation of safety injection at approximately 1640
psig), the RCS pressure had peaked at approximately 2435 psig and decreased
rapidly to 1900 psig at 15 to 20 seconds.

RCS pressure then decreased more

slowly from 1900 psig to 1600 psig at 20 to 120 seconds.

During this timeinterval RCS pressure stabilized at approximately 1750 psig for about one minute.
This pressure stabilization indicated the formation of voids within the reactor
core, upper vessel head, or upper OTSG and hot leg ("candy cane") and continued
as the OTSGs secondary side boiled dry in the 1-2 minute time frame.

RCS pressure

continued to decrease after ESF Initiation to a low of approximatelyJ400 psig
at 2 to 6 minutes.

Pressure then increased from 1400 psig to 1600 psig at 6

to 8 minutes as the operators throttled the HPI flow to minimum and removed
one (MUP-lC)

HPI pump from operation.

At this point, the bulk saturation

temperature of the coolant determined the RCS pressure.

RCS pressure

subsequently decreased from 1600 psig to 1100 psig from 8 to 18 minutes as the
emergency feedwater was initiated to the OTSG and the RCS average temperature
decreased from 597'F to 555'F.

(Reference:

period described; 0-18 minutes).

1-2-15See Appendix I-A for the time

The operators had noted the pressurizer level increase to 258 inches at
12 seconds followed by a decrease to 158 inches from 12 to 40 seconds,

after

which the level commenced a continuous rise to full scale indication from 1 to
5 minutes.

The HPI flow between 2 and 5 minutes was throttled to minimum andone HPI pump stopped (MUP-IC),based on the increasing pressurizer level and

RCS pressure decreasing very slowly toward 1400 psig (6 minutes).

With thepressurizer level off scale high and the RCS pressure low (1400-1600 psig from
6 minutes through 18 minutes),

the operators concluded that the problem wasthe collapse of the pressurizer bubble during the severe transient.

Substantial

time and effort was expended attempting to regain the bubble and pressurizer
level using the pressurizer heaters and letdown system.
The operators had previously experienced pressurizer heater breaker failures, ascribed by the plant staff to be due to hot and humid local (control
building M-20 area) conditions therefore, the lack of initial response of the
RCS pressure to the heaters was not unexpected by the operators.
Based on previous experience and training the operators expected thepressurizer level to decrease below 80 inches following the transient.

They

did not understand why the level decrease ceased at 158 inches and began torise steadily. (Int.

5,17, 38, 2,

14, 37, 3,

15, 4, 9, 61)

The continuing high pressurizer level caused the operators to conclude
that no RCS inventory problem existed.

This decision was based on a number ofother events that had occurred but were considered normal or not understood.
These items included:oHigh radiation alarm in Reactor Building.
No high radiation alarm was received (the HPR-227 monitor was subsequently
reported flooded, under positive pressure, and manually isolated at 0600
hrs).

(Int.

5,

17,38)

1-2-16

0

Reactor Building (RB)Ambient Temperature Alarm.

An increase in RB ambient temperature was noted (5 minutes),and hightemperature fire alarms were received and evaluated by the Shift Supervisor
who was in the control room at that time.

That Shift Supervisor was awareat that time that the rupture disc had blown on the RCDT following theinitial turbine trip transient and normal opening of the EMOV (RC-R2).
(Int.
o

5,

17, 38)High Reactor Building Sump Level.
The high RB sump level was attributed to the blown rupture disc on the
RCDT after the opening of the EMOV (RC-R2).

oHigh Reactor Building Pressure (RCSWith the higher-than-normal(Int. 5,17,38, 3, 15,94)

or main steam line rupture).

RB pressure and temperature continuing toincrease with no apparent RCS inventory problem (pressurizer level off
scale high), the shift personnel became suspicious of a steam line or
OTSG leak.

Subsequent to the stopping of the'RCPs in the B loop at

74 minutes,

the B OTSG was isolated at 0527 when the B OTSG pressure wasnoted to be 300 psig lower than the A OTSG pressure.

The operators noteda small decrease in the RB pressure increase.at the time the B OTSG was
isolated and felt that the steam (feed)
been isolated.
o

(Int.

5,

17,

38, 2,

leak to the reactor building had

14, 37, 3,

15, 4,

9,61)

Rapidly Decreasing Makeup Tank Level.
The operators were maintaining the MUT level by cycling the makeup pumpsuction valve.(DH-V5A)(MU-V12)from the BWST.

With the normal suction valve

open from the makeup tank, the makeup pump 1A or IB suction was

momentarily shifted to the MUT as needed to control the tank level.

Thismode of operation was familiar to the operators.

15)

1-2-17(Int.

5,17, 38, 3,o

The operators are required to verify that the high-pressure injection
system is operating properly as evidenced by injection flow in all fourlegs (MU-Vl6A/B/C/D).

Flow was indicated on MU23 FEI,

2, 3, 4.

The high-pressure injection system was verified to be operating properlywith 250 gpm per HPI leg as evidenced by flow through all four injection
paths.

(Int. 3,15)

The operators are also required to verify that the safety injection
equipment is in its ESF position as shown in a table attached to the procedure.
The interviews indicated that the equipment was verified in the ESF
position, but no documentation was available for this action.
14,(int. 3,15, 2,

37)
Procedure Section B, step 3.4, calls for bypassing the safety injection

by depressing the group reset pushbuttons (step 3.5) and throttle MU-Vl6A/B/C/D
as required to prevent pump runout.
NOTE:

The emergency procedure is deficient as this step (3.5) alludes to

HPI flows at less than 250 gpm.
The operators throttled the HPI valves in an attempt to control the pressurizer level without regard to the RCS pressure being below 1640 psig.

They

felt that the pressurizer level had been "caught" when the minimum was 158
inches from 40 to 60 seconds and started to increase.

They further concluded

that flow from the HPI system was not required based on the pressurizer level
continuously increasing to full scale in 5 minutes.

The operator throttled

the HPI flow to minimum by approximately 4.5 minutes and immediately establishedmaximum letdown flow (140-160 gpm)

in an attempt to regain and control pres-surizer level using the normal orifice path,and two letdown coolers.

1-2-18the orifice bypass valve (MU-V5),

Evaluation
The emergency procedures require the operators to trip reactor coolant
pumps before pressure decreases below the net positive suction head (NPSH)
requirements and before reaching 1200 psig.

The reactor coolant systempressure decreased to 1200 psig within approximately 15 minutes after thereactor trip and remained between 1000 psig and 1100 psig until the last two
RCPs (RCP-IA and 2A) were tripped (101 minutes) by the operators.

The

operators did not perceive that a RCS inventory problem existed because of thefull pressurizer and felt that the bubble in the pressurizer had been collapsed(lost).

The recovery of the RCS to normal was considered a matter of energizing

the pressurizer heaters and providing letdown to reestablish pressurizer leveland pressure.

(nt.

5,

17,

The operators did not feel they were involved in a LOCA condition.
38, 2,

14, 37, 3,

15, 4, 9, 61)

The failure to-follow procedures (2202-1.3,
Revision 8, May 12,

Loss of RC/RCS Pressure

1978, Section B, Step 2.2.4) and trip the reactor coolant

pumps at 1200 psig as required is under consideration as a potential
noncompliance pursuant to Technical Specification 6.8.1.a (Ref.

item of35).

The rate .of increase of pressuri-zer level slowed at approximately 5 minutes,
and the letdown.1_low was returned to near normal (70 gpm)

shortly afterwards;

however, the pressurizer level continued to increase and went off scale (400 inches).
With a high pressurizer level,

the operators-did-not consider-that a reactor

coolant inventory problem existed, even though the system pressure was decreasing
steadily.

The opinion of the operators was that the-bubble in the pressurizer

had been collapsed during the severe plant transient.

(Int.

2,

14, 37,3, 15)The operator removed one of the makeup pumps MU-PK at 4.5 minutes after
closing the two associated HPI valves (MU-Vl6C and V16D).

(Int.

2,

14, 37,

15)
Section A, step 3.5 of the procedure contains the following statement:

1-2-193,

"CAUTION:

Continued operation of HPI depends upon the capaility tomaintainpressurizer level and RCS pressure above the 1640 PSIG Safety
Injection Actuation setpoint.
"Ifpressurizer level can be maintained above the low-level-alarm point

and the RCS pressure above the Safety Injection Actuation point; theninitiate a plant shutdown and cooldown per normal procedures.
NOTE:
"If

The plant did not reach a stable hot standby RCS condition.

pressurizer level can not be maintained above the low level alarm point

and the RCS pressure above the Safety Injection point then the plant has
suffered a major rupture of a significant size leak such that Engineered
Safety Features Systems are Automatically Initiated."
The operating procedures require operability of the safety features
actuation system (2105-1.3,
Revision 13, March 19,

Safety Features Actuation System, Revision 2,1979) during plant operating conditions in accordancewith Technical Specifications 3.5.2.

This requirement is amplified by theprocedure for unit heatup (2102-1.1,

Unit Heatup, Revision 19,

which addresses ES system operability.

(Ref.

1979)

79, 28)The procedure for unit cooldown (2101-3.2,
March 10, 1979,

March 7,

Unit Cooldown, Revision 13,

step 4.13) provides for blocking the SFAS channels during

normal plant cooldown activities between 1920 psig and 1800 psig; this allows
for normal plant cooldown and depressurization under controlled conditions
without an unwarranted ES initiation.

(Ref.

27)

No procedural steps exist which permit the reduction in the HPI flow With
RCS pressure below 1640 psig until the RCS is placed on the Decay Heat Removal
System (DH).
The shift personnel throttled the high-pressure-injection flow to thereactor coolant system and the HPI system was in the degraded (throttled)

1-2-20condit-ion-until

approx-imately 100-minutes; even though the reactor coolant

system pressure never recovered above 1640 psig.

Moreover,the RCS pressure

did not recover above approximately 1200 psig after 15 minutes into theaccident until the plant was repressurized at 0750 hrs.
The failure to maintain the operation of the high pressure injection systemdelivering the design injection flow to the reactor core of 250 gpm per HPI
flow leg (4),during the accident,

emergency procedure 2202-1.3,

in accordance with the requirements ofRev. 8,

Section B, step 2.2.3 and 3.5 resulted

in apparent serious core damage and onsite and offsite exposures to radioactive
materials.

This failure to maintain the HPI system in accordance with procedure

2202-1.3 is under consderation as a potential item of noncompliance pursuant
to Technical Specification 6.8.1.a.
2.6

Operator Actions Concerning the Operation of the Makeup System(Following Initial ES Actuation)
The operators continued to operate the HPI/makeup system in the throttled

condition between 0405 hours and 0530 hours.

At about 0525 hours,

emergency

borating operations were performed when the RCS boron sample indicated 700 ppm
and the source range monitors had shown an-oscillatory behav-ior with an increased
count level.

The chemistry analysis indicated a decrease in boron concentrationin the RCS,

the neutron flux levels increased, and the pressurizer and/or makeup

tank levels--increased;

The interviews and records review indicated that theoperators appropriately addressed the actions required by the abnormal procedure(2203-1.1,

Revision 5, September 12,

1978,

Loss of Boron-Moderator Dilution,

as follows:o

Attempt to stop the inflow of water into the RCS.
The operators attempted to verify all sources of water into the RCS that
could be causing a high pressurizer level (full-scale after 6 minutes)
and an increase in the neutron flux level.
3,

15,

10)

1-2-21(Int.. 5,

17, 38, 2,

14,

37,

o

Determine the amount of shutdown margin available in accordance with

2103-1.9, Reactivity Balance Calculations.
The calculated shutdown margin for the boron sample taken at 0515 hours(700 ppm) was approximately 6% reactivity.

(Int. 5,17,

38, 2, 14,

37,

97)
o

Initiate additional sampling.
An additional sample for boron was obtained at 0600 hours (404 ppm) andthe shudown margin was calculated to be 2.4% reactivity at 0630 hours.
(Int. 5,

6,

17, 38, 2,

14,

37)

Emergency borating operations were conducted at approximately 0525 hrs
through the normal flowpath to the MU system from the boric acid mix tank (BAMT)and the boric acid pumps (CA-P-4A or 4B) by opening MU-VlO.

Additionally, at

0600 hrs after the neutron flux level had continued to increase,the operators

borated again and personnel were dispatched to the auxiliary building to open
the alternate borating path (MU-V127)

and verify the valve lineup to the makeup

system from the boric acid mix tank (BAMT).

(Int.

5, 6, 17,38, 2,

14, 37, 3,

15)
The suction for the MU system, with one or two pumps operating from shortly
after the reactor trip occurred, had been the borated water storage tank (BWST).
Therefore,

the water being supplied to the reactor coolant system was approximately

2300 ppm boron.

The operators did not understand specifically why the neutron

flux level was increasing or why the indicated RCS boron concentration had
decreased from 1028 ppm to 700 ppm and then to 404 ppm by about 0600 hrs.
(Int.

5, 6,

o

The operator interviews indicated the HPI flow was increased approximately

17, 38, 2,

14,

37, 27, 3,

15)0530 hrs because neutron flux level was oscillating and increasing,and thereactor coolant system flow through the A loops had decreased to approximately
35 million pounds per hour.

Subsequent to initiating the increased high-pressure-

1-2-22

injection flow, the operators noted that the A loop RCS flow continued todecrease rapidly from about 35 million pounds per hour to 25 million pounds
per hour.

(Int. 5, 6,

17,

38, 2,

14, 37,3, 15,

27)

The operators continued to operate the MU system in the high-pressureinjection mode from 0530 hours through 2000 hours with at least one makeup
pump operating.
o

The apparent exceptions to this include:

High pressure safety injection was initiated at 0720 hrs and MU-PIA

started and operated until 0737 hrs, when the operator,
signal at 0727 hrs, stopped the IC MUP.

having reset the ES

The MU-PIA continued to operate until

0756 hrs, when the building isolation occurred and the MU-PIC started.

The

MU-PIA and MU-PIC pumps operated in the injection mode until about 0817 hrs,
when the operator stopped both ]A and 1C pumps.

The operator immediately

attempted unsuccessfully to start the MU-PIA to reestablish RCP seal water
injection.

The MU-PIA control switch was placed in the PULL-TO-LOCK position

and apparently remained in that position until after 2000 hrs due to thefailure of the pump to start manually and the belief the pump had a mechanicalproblem.

The investigation of this area indicates that the pump breaker could

not have closed, which led the operator to misunderstand indications.
(Int.

2,

o

Another ES initiation (Channelpressure.

14,

37, 3,

15, 4, 9, 61)

The makeup pumps (MU-PIA and PIB) failed to start and the IC makeup

pump was not operating or started.
oA) occurred at 0819 hrs on high'building
(Int.

5,

6, 17,

38, 2,

14, 37, 4, 9, 61)

The makeup pumps were off for approximately 4.5 minutes when,at 0822 hrs,

the operator started the IB pump, and this pump remained in operation until
after 2000 hrs.
o(Int. 2, 55, 95; 4,

35, 36,37 and 145)

The operator started the IC makeup pump at 0827 hrs and the lB and 1C makeup

pumps were operated in the injection mode,delivering borated water heavily, to -the loop A injection lines (MU-Vl6A and MU-Vl6C).

The ICpump was operated

continuously in support of the IB pump until 1304 hrs, when the RCS pressure
had been decreased to about 450 psig.

This was during a period of intentional

1-2-23

pressure reduction of the RCS in an attempt to reach the discharge pressure of
the core flood tanks.

It was intended by the plant staff to continue down inpressure to place the decay heat removel system in service at approximately
350 psig.
o(Int.

26, 5,

6,

17,

38)

After the restart of the IC MUP at 1350 hours on high building pressure

ESF actuation,

the operators reviewed the plant status and stopped MU-PIC at

1351 hrs after resetting the ES signal,
the IC pump in standby.
o

(Int.

leaving the IB pump in operation and

119, 26)

At 1432 hrs, the operators restarted the MU-PIC pump and injected additional

HPI water into the B loop through valves (MU-Vl6C and MU-V16D)

as pressurizer

level decreased from offscale high (more than 400 inches) to 375 inches.
1436 hrs, the operator stopped the IC
increased to more than 400 inches.
oAtmakeup pump after pressurizer level(Int.

119, 26)

The operators operated the 1C makeup pump for 9 minutes at 1519 hrs,

3 minutes at 1533 hrs,

forand for 83 minutes at 1723 hrs in order to prevent

pressurizer level from decreasing below the 80-inch heater cutoff point.
(Int.

119)

o

At 1933 hrs, when the RCP was bumped (10 seconds),

the RCS pressure

dropped to 1450 psig and the actuation of ES attempted to start the IC MUPThe IC makeup pump tripped; six seconds later the operator restarted the
IC pump and the pump operated for six minutes to recover RCS pressure tonormal (2250 psig).
to 1300 psig.

The IC pump was operated again as the pressure decreased

The operation of the additional pump (MU-PIC)

in support of theIB pump was to prevent the RCS pressure'from decreasing drastically.
pressure finally stablized at approximately 1000 psig.

(Int.

The RCS

119, 232,

233)

Evaluation
When the ES initiation (train A) occurred at 0819 hrs on high building
pressure; the makeup pumps 1A and IB failed to start and the IC makeup pump
was not operating or manually started.

The failure of the 1B makeup pump to

1-2-24

start indicated that the pump control switch was in the PULL-TO-LOCK position,
the pump was not selected for standby, or the motor breakers failed to close.
The automatic initiation of the ES Train A occurred normally.

The failureof the ESF train B to initiate was due to the manual defeat of two of the threechannels in train B of the Reactor Building Cooling and Isolation Actuation
System (ESFAS)

at 0800 hrs by the operators.

on Train A at 19 hrs,

Subsequent to the ESF initiation

two of the three channels in Train A of the Reactor

Building Cooling and Isolation Actuation System (ESFAS)

were also defeated at

0820 hrs by the operators.
NOTE:

The first of the two channels in the Train A was defeated at 0800

hrs concurrently with the defeat of the channels in Train B by the operators.
The Emergency Procedure (2202-1.3,

Revision 8, May 12,

1978,

Loss of

Reactor Coolant/Reactor Coolant System Pressure, Step 3.4 of Section B) requires
the "DEFEAT"of any two channels of Reactor Building Isolation and Cooling,

then bypass of all three Safety Injection Channels.
NOTE:

Both Train A and B remained defeated until 0724 hrs when thechannels were automatically reset.
Technical Specification 3.3.2,
Instrumentation Procedures,

Engineered Safety Feature Actuation System

section 3.3.2.1 requires that the ESFAS instrumenta-tion channels shall be OPERABLE as shown in Table 3.3-3 of those specifications.
This includes the safety injection and Reactor Building Cooling and Isolationfrom Reactor Building Pressure High with a minimum of two channels OPERABLE in
MODES 1, 2, and 3.
The manual defeat of two of the three channels in Train B at 0800 hrs andtwo of the three channels in Train A at 0820 hrs in MODE 3 is under consideration
as a possible item of noncompliance pursuant to Technical Specification 3.3.2.1.
As noted above in between 0654 hrs and 2000 hrs, the HPI system flow wasroutinely modified during the accident on March 28,

1-2-251979,

by throttling theHPI flow and starting and stopping makeup pumps.

Except during the time periods

listed below when the RCS pressure was above 1640 psig, the continuous operationof the full,available HPI system was required to be OPERABLE to provide coreprotection.
(1)0654 hrs through 0725 hrs(2)0740 hrs through,0755 hrs(3)0930 hrs through 1140 hrs(4)1840 hrs through 1933 hrs(5)1933 hrs through 1937 hrs(6)1950 hrs through 2000 hrs

The failure to maintain the operation of the HPI system delivering thedesign injection flow to the reactor core of 250 gpm per HPI flow leg, including
the period between 0819 hrs and 0822 hrs when no HPI pumps were operating inaccordance with the requirements of the emergency procedure 2202-1.3,
RC/RCS Pressure,

step 2.2.3,

Loss ofis under consideration as an apparent item ofnoncompliance pursuant to Technical Specification 6.8.1.a.
At approximately 0800 hrs following the initiation of the ESFAS from reactor
building high pressure at 0819 hrs the operator blocked the reactor building
engineered feature actuation system channels in order to reestablish manualcontrol of certain components and equipment.

Technical Specifications 3.3.2

requires that the Engineered Safety Feature Actuation System (ESFAS)

channels,

including Reactor Building High Pressure, shall be operable in Mode 1, 2, and3.

Operating procedure 2105-1.3,

Actuation System, Step 4.3,

Revision 0, April 9,

1977,

Safety Features

requires that the SFAS channels will remain in the

PROTECTION FUNCTION FULLY ENABLED condition except during maintenance or..
testing which will be required to demonstrate the systems ability to actuate
when required.
The failure to maintain the- SFAS channels operable as required is underconsideration as a possible item of noncompliance pursuant to Technical
Specification 3.3.2.

1-2-26

2.7

Operator Actions Concerning Condensate.System
The sequence of events that resulted in a turbine trip are discussed. in

Section. 4.1.
The steaming to the main condenser via the turbine bypass valves (MS-V25A,
V25B,

V26A and V26B) to maintain steam header pressure at 1010 psig apparently

caused the hotwell'level control valve-controller to fail (low level)
hotwell was flooded from the condensate storage tank.

and the

The hotwell level

increased from 21.82 'inches to the completely flooded condition (greater than50" indicated level) within 1 minute and the hotwell level was not recovered
until 0653 hrs.
Interviews,

record reviews,

and analysis indicate that attempts were madeto reestablish the condensate system at 0405 hrs as indicated by the start of
the IA condensate pump through 0653 hrs,,when the hotwell level reject valve
was finally recovered and the hot well pumped down to normal.

This was done

to provide normal feedwater system flow to the generators and to prevent
losing condenser vacuum.

The loss of condenser vacuum (approximately 18 inches)

terminates the release of decay heat steam to the condenser and automaticaloly
shifts-the decay heat steam to the atmosphere.

The operators were involved in

the following activities:o

Attempts were made to start a condensate booster pump.
The operators responded quickly to the turbine building basement area toattempt to recover the condensate feedwater system.

(Int.

4, 9,

5,

6,

17, 38)o

The operators discovered a leak in the 2Acondensate booster pump suction

line and isolated the pump locally to prevent spilling condensate water to theturbine-building floor and sump.

(Int..109,

102, 10, 5,

6,

17,

38)

The OTSG B turbine bypass valve isolation valve (MS-Vl5B)

was closed by'

the operator 8 minutes later (Section 2.9 of Details I).

(Int. 5,38)

1-2-27'

6,

17,

o

The condenser vacuum started to decrease at about 0720 hrs,

operator secured the condenser vacuum pumps at 0930 hrs.

and the

The release of decay

heat steam was automatically transferred to the atmospheric relief valve (MS-V3A).
The vacuum decreased rapidly,*were also lost.

and at approximately 1050 hrs,

the auxiliary boilers

The supply of auxiliary steam to the turbine seals in orderto maintain vacuum was a continuing problem until ajproximately 1700 hrs when
(Int. 57,

the auxiliary boilers were recovered.
o111, 5,

6,

17,

38)

The release of decay heat through the atmospheric relief valve (MS-3A)

was stopped at approximately 1315 hrs after the State had requested that steam
release be terminated.

This limited the removal of decay heat from the reactor

to the HPI flow and RCS blowdown into the reactor building.

(Int.

26,

29, li3,

5, 6,

17, 38)o

The condenser vacuum was reestablished at approximately 1700 when auxiliary

steam was available for the main turbine seals.

The release of decay heat wastransferred back to the turbine bypass valves (MS-V25A and 26A) to the maincondenser at 1750 hrs.
-o

(Int.

119, 26)

The condensate booster pump (2B)

was started at 1917 hrs in order to feed

OTSG A, which was at approximately 100 psig at that time.

(Int.

119, 20)

Evaluation
The problems encountered with the condensate system and condenser vacuum
significantly detracted the operator's attention from the accident.
The shift supervisor,

Unit 2, spent approximately 45 minutes from 0415 to

0500 hrs in the area of the condensate polishers and hotwell due to thepersistent high hotwell level condition and lack of reject capability to thestorage tank.

The reject capability was recovered at about 0655 hrs.

The

Unit 1 shift supervisor was in the Unit 2 control room during the absence of
the Unit 2 shift supervisor.

1-2-282.8

Operator Actions Concerning Reactor Building Sump Pumps
The reactor building sump pump (WDL:P-2A)

came on automatically at 0407

and commenced pumping the RB sump to the auxiliary building.

The source of

the water appeared to be the normal water accumulation in the sump supplemented
by the discharge from the relief valve on the reactor coolant drain tank
(WDL-T-3)
NOTE:

lifting at 120 psig.
The level in the miscellaneous waste holdup tank (WDL-T-2)

change during the accident.

(jnt. 109, 40,

The second RB sump pump (WDL-P-2B)

did not

10)

started at 0410 hours and pumped inparallel with the 2A sump pump to the auxiliary building.

An auxiliary operator

noted an excessive in water level in the auxiliary building sump with the RB
sump pumps operating,and the operator notified the control room.

The RB sump

pumps were stopped at that time by the auxiliary operator at the direction of
the control room operator to cease water transfer into the auxiliary building.
At that time (0438 hrs),

the operators did not note any unusual conditions in

the auxiliary building.

(Int.

5,

6,

17,

38, 3,

15,

10)

Reactor building isolation occurred at 0756 hrs (3.58 psig).

This isolated

the RB sump pump discharge line to the auxiliary building by closing the RB
isolation valves (WDL-FHS-3189 and WDL-FHS-'332).

The operators verified

reactor building isolation using the Emergency Procedure 2202-1.3,(Loss of

Reactor Coolant/Reactor Coolant System Pressure, Revision 3, May 12,

1978,

Appendix B, of Section B.)

(Int.

2,

14, 40, 3,

11, 15)

The 480V motor, control centers 2-32A and 2-42A feeder breakers tripped at
approximately 1351 hrs, terminating the power to the RB sump pumps (WDL-P2A
and P2B) and other equipment supplied by the motor control centers.

The power

was not subsequently returned to the RB pumps (WDL-P2A and P2B) because of thepossibility of electrical grounds on the pump motors after the operation of
the reactor building spray system.

1-2-29

Evaluation
The operator actions concerning the reactor building sump pumps was very
general in nature.

The stopping of the RB sump pumps was viewed as needed to'cease water transfer to the auxiliary building (AB)to limit the liquid waste

handling conditions.
2.9

Operator Actions Concerning the'B OTSB Tube LeakThe sequence of events following the turbine/reactor trip led the operators
to conclude that an OTSG leak from the shell-side to the containment atmosphere.
This conclusion developed based on a number of events,o

The persistent increased reactor building (RB)including:pressure and temperatureincluding high-temperature fire alarms, which were noted and acknowledged byoperations personnel.
o(Int. 95,

315,

109)

There were no radiation alarms in the RB with the exception of the inter-mediate closed cooling alarms, which were considered normal and routine.

Thisabsence of alarms led the operators to conclude there was not an RCS leak.
These detectors were.located-near the RB sump and it was known by the staff inthe control room that the rupture disc had blown on the RCDT at 0415 hrs(Int. 3,15,

previously.

30, 95).

The operators also noted that these alarms had occurred

The RB radiation monitor (HP-R-227)

had been found flooded,

positive pressure, and manually isolated by the, operators at 0550 hrs.
review recalled that the monitor had reached 50,000 cpm at 0518 hrs.

underRecords

The

monitor had been flooded previously and reported to the NRC (LER 78-073-036,
Section 1.6 of this report).
o

The operators noted that the B OTSG pressure had dropped 300 psig below

the A OTSG and continued to drop markedly after the IB and 2B RCPs were stopped
at 0514 hrs.
o(Int. 5,6,

17,38, 4, 9,

After isolating the B OTSG,

61)

the RB pressure was noted to have leveledout, reaffirming the conclusion of the'operating staff that the B OTSG shellside was leaking into the containment building.
1-2-30

(Int.

5,

6,1

17,

38, 4,

9,

61)

o

Subsequently,

at about-0620 hrs, the EMOV (RC-R2)

pressure decreased from approximately 2.5 psig tobuilding temperature decreased from 160'F to 120'F.

was closed and the RB.O. psig.

The reactor

The-operators thenconcluded that the B OTSG had not~been leaking earlier as suspected..
oAt about 0656 hrs, after pumping down the hotwell,

two additiona l circulat-

ing water pumps were started.

The B OTSG was momentarily steamed to the main
steam header when the main steam stop valves (MS-V4B and MS-V7B were cycledopen and closed within 7 seconds.
offgas monitor (VA-R-748)

The B OTSG was reisolated.when the condenser

alarmed indicating a leak in the B OTSG.

The operators

were convinced that the B OTSG had a primary-to~secondary leak and completely
isolated the steam generator,

including closing the supply to the turbine

bypass valves by closing the header isolation valve (MS-Vl5B)(Int. 81,o

57,

at 0704 hrs.

111, 184)

Throughout the remainder of the day, the B OTSG remained isolated and the

A OTSG was utilized as necessary for decay heat removal.
Evaluation
The operations staff initially concluded that the B OTSG had a shell side
leak based on. the increasing reactor-building temperatures coincident with anormal RCS pressurizer level.

This conclusion by the staff was also supportedby the lack of any reactor building radiation alarms early. in the accident.
The operating staff did not. relate the continuing heat load in the reactor.
building to'the loss of coolantaccident even though the assumption of a
B OTSG.shell side leak to reactor'building could not.be supported.

The RCS

temperatures were stable and the B OTSG level was increasing with minimal
feedwater required.

Fuirthermore,,the feedwater to the BOTSG was completely

isolated at 0428 hrs by the operators because of difficulties in maintaining
OTSG level.

(Ref.

Appendix IA)

The substantial evidence of a LOCA was indicated by the continuing RB
high temperature with the full reactor building cooling system in operation

1-2-31(10 million to 100 million BTU per hr heat removal capability) and the continuing
low RCS .system pressure.
The operators completed isolation of the B OTSG at about 0527 hrs basedon 300 psi.g difference between the A'OTSG and the B OTSG. The operators did
not recognize that the pressure d.ifferential was, caused by inadequate backflow
through the B loop after the RCPs were stopped'in the B loop at 0514 hrs.
Thebrief unisolating of the B OTSG at about 0656 hrs indicated to the operatorsthat the B OTSG had suffered a tube leak; since the condenser off gas' monitor(UA'R-748)
2.10

alarmed.

Operator Actions Concerning Initial Natural Circulation - Decay Heat
Removal Via OTSGs
At approximately 0541 hrs, the loop A reactor pumps (RC-PlA and P2A) were

stopped because of the decrease to 950 psig in RCS pressure, reduction of theindicated RCS loop A flow from 35 to 25 million pounds per hours,
vibration alarms including indication at maximum displacement.
was discussed between the shift supervisor,
and the operating crew.

1978).

superintendent-technical

Limits) within the operating procedure

This heatup/cooldown curve was also provided in the procedure for
Decay Heat Removal Via OTSG,

1978)

During this period of time,RCPs,

it was decided to go to natural circulation.

Revision 5, March 17,

immediately prior to stopping the'loop A

specific references,.-limitations and precautions,osupport,

Reactor Coolant Pump Operation, Revision 6, August 16,

decay heat removal (2102-3.3,

steps,The situation

The discussion included the review of the Heatup/

Cooldown'Curve (RCS Pressure-Temperature
for RCPs (2103-1.4,

and pump

Procedure 2102-3.3, contains
prerequisites,and proceduralincluding:

Reactor Coolant System Cooldown Limitations.

temperature,

The reactor coolant

pressure, and cooldown rates must be maintained within the limits

specified in Figure 3.4.2 of Technical Specification 3.4.9.1.

1-2-32

The heatup/cooldown curve attached-to the operating procedure include themaximum and the minimum RCS pressure-temperature.limits
RCPs,

fuel clad compression,for the RCS,

and control rod drives during the decay heatremoval operation.
o

The emergency feedwater pumps (EF-PlEF-P-2A,(as indicated by Panel 3 indicating lights),

and EF-P-2B) are running

supplying emergency feedwater to

the OTSGs from the condensate storage tanks through the emergency feedwater
valves (EF-VlIA and VlIB).

OTSG level is increasing to or at 21 feet (50% asindicated on operating range level instrumentation SG-lA/lB-LT2 or LT3).
The motor-driven emergency feedwater pump (EF-P2A)

was operating, taking'suction from the condensate booster pump suction header and the main hotwell
via the operating condensate pump.
o(Int. 4, 9,

61)

All four RC pumps are tripped, the reactor is tripped, and the turbine istripped.

Steam-pressure-is being -maintained at turbine header setpoint (i.e.,

855 psig during normal operation plus 125 psig) dumping steam to the.main condenser through the turbine bypass valves (MS-V25A and V25B and MS-V26A and
V26B) or dumping steam to the atmosphere,through the atmospheric dump valves(MS-V3A and V3B)
main condenser

if a low condition of less than 24 inches.Hg exists in the

6 r lessthan three circulating water pumps are running.

Steamsafety valves and pressurizer electromatic and code safety valves are closed.
Shift personnel had established that the reactor and turbine were tripped
and had secured the four RCPs.

The turbine bypass valves were in manualcontrol with five circulating water pumps operating.

The operatorshad

previously concluded (incorrectly) that the safety valves and the pressurizer,
electromatic relief valve (EMOV)

were closed.

The EMOV (RC-R2)

was still open

* at this time.
The operators stated that they did not feel that natural circulation was.established(Th- 5 3 0 'F,

because'of the differential temperature across the A OTSG
Tc-520'F at 0652 hrs and diverging),

sure (800 psig and decreasing),

the low A OTSG secondary pres-and the minimum feeding and steaming rates for

1-2-33the A OTSG.
At approximately 0610 hrs the A loop Th had increased to 620 0 F,
the recorder off scale high; TL had decreased to about 480'F differential'temperature of 140'F); and the operators expected a. differential temperature of
250 F-.

Some heat was being removed through the steam condensing mode within

the steam generators,

but the continuous decrease in the A OTSG pressure indi-cate that the heat removal from the RCS was limited.

(Int.

5, 17,

38, 4,

9,61)
Evaluation
Technical Specification 6.8.1.a requires that procedures be implemented
covering the applicable safety-related activities as recommended by Regulatory
Guide 1.33,

1972.

The failure to establish the plant conditions as requiredby the operating procedure (2102-3.3,
1978,

step 2.1.3),

Decay Heat Removal,

Revision 5, March 17,

of approximately 1100 psig or 558'F) (saturation pressurefor 558OF - 528 0 F plus 30'F is under consideration as a possible item ofnoncompliance.
Interviews revealed that the shift personnel associated withfthe accident
had not received specific training in the natural circulation aspects on thefacility on site or at the simulator.

This lack of specific training supports

the, questions concerning the establishment of natural circulation.,when thelast two RCPs were secured at 0541 hrs.
Furthermore,

the lack of specific training contributed to the attempt by

the plant operators to place the plant in the natural circulation mode of'decay heat removal,
2.11

when parameters were outside the procedural requirements.

Operator Actions Concerning the Auxiliary Building Sump Pump'
The auxilia'ry building sump pumps (WDL-P-3Aand 3B) are normally aligned

to automatically pump the auxiliary building sump through a set of two parallelfilters to the auxiliary building sump tank (WDL-T-5). as described in theplant. operating procedure (2104-4.1,
Revision 2, March 14,

Miscellaneous L~iquid Rad Waste Disposal,1979) (Ref.

37) and the plant system description number-,,
A4 (Rad Waste - Miscellaneous Liquid, 10/75).

-1-2-34.

1)

Personnel interviews revealed the following information concerning theauxiliary building sump pumps,oincluding:At approximately 0420-0440 hrs, when operators were in the area of the.

rad waste panel in the auxiliary building, the auxiliary building sump tank
(WDL-T-5)

level was. extremely high with the reactor building sump pumps(WDL-P2A and P2B) operating.

The control room was notified and the RB sump

pumps were secured to'stop the discharge of liquid from the RB sump.
time, the operators had noted no abnormal conditions in the area.,
miscellaneous waste drain tank (WDL-T-2)'At thatThelevel was noted to be normal.

(Int.

10,

109)
o

At approximately 0656"hrs,the operator evacuated the area because ofhigh radiation levels in the auxiliary building areas.
rad waste panel,

Prior to leaving thethe auxiliary building sump pumps (WDL-P-3A-3B)

were securedby the operator because water was backing up through the auxiliary building
floor drains (small amounts of water were reportedin-the.low points near thedrains).
o(Int.

10, 109, 3,

15)*The initial water in the auxiliary building sump system was concluded not

to be highly radioactive since the high radiation levels were observed to befrom specific areas such as the makeup tank (MUT)

room and the pipe alley.

This is treated in greater detail in Details II of this report.
NOTE:

It

was concluded by the staff that the relief valves may havelifted because of a high differential pressure across the makeup filters(MU-F-2A and 2B) as a result of deposits of materials from the RCS, ofundetermined origin within the filters.

Pulsating letdown flows noted by

the operators would indicate the lifting of a relief valve, either to thereactor coolant bleed tank/(RCBT) or to the auxiliary building sump
system.
o(Int. 26,

4, 9,

61)

At approximately 1300 hrs, a tour of the-auxiliary building was performied.

It was determined that there was water near the floor drains.'

1-2-35

Because of highradiafion levels in the area (10 rem per hour) the stay was brief.

The reactor

building and auxiliary building sump pumps were verified in the OFF-position.
(Int.

13,

o

Following the trip of 480-volt~motor control centers 2-32A and 2-32B

140,

173)

(immediately after the initiation of the reactor building spray system on highpressure - 30 psig) at approximately 1350 hrs, the operating engineer was dispatched to the auxiliary building to return power to certain of the equipmenton the buses.
energized.

The reactor building sump'pumps (WDL-P-2A and 2B),were not

The buses subsequently remained energized with the auxiliarybuilding sump pumps (WDL-P-3A and 3B) in the OFF position.

(Int.

102)

Evaluation
The operator actions concerning the auxiliary building sump pumps wasroutine.

The stopping of the reactor building sump pumps to prevent continued

transfer of water.from the reactor building tothe auxiliary building until it
could be processed was not unusual.
The water in the auxiliary building sump and floor.drains became a problem
when the, highly radioactive reactor coolant was discharged into the auxiliarybuilding sump system sometime after the reactor core was seriously damaged.
The discussion of leakage of contaminated RCS water to the other areas of the.
facility is treated in Details II of this report.
2.12

Operator Actions Concerning Logkeeping
Administrative Procedure 1012 (Shift Relief and Log Entries,

October 4,

1977,

(Ref.

revision 8,77) establishes the requirements for shift relief andrecording station operatingactivities

in logs or other controlled documents

on a shift basis (section 1.1) and describes the various'shift records andlogs involved and instructions required to maintain these records to conformwith the Technical Specifications and to ensure adherence to the requirementof the FSAR (Section 1.2).

1-2-36

The general

requirementsare described in section 3.1.1_and specify shift

records-as hourly log, control room log, checkoff lists, recorder charts,

andcomputer printouts that describe or record operating information and events.
These records comprise the. information that is necessary for evaluating operations or for analysis of previous operations.
The hourly log. requirements (action 3.2) specify that the log will reflect
plant parameters on an hourly basis.

It will normally be prepared by theplant computer but can be manually prepared by the control room operator in
the event that the computer is not-functioning.

If manual preparation isnecessary, it will be performed by the control room operators and auxiliary
operators.

The inspectors reviewed the hourly computer log for March 28,

1979.
The control room log requirements (section 3.3) specify that the log will
include information concerning reactivity, alarms pertaining to reactor core
conditions with detailed explanation,

any abnormal condition of operation,

releases of radioactive waste, gaseous or liquid.

The administrative procedure(1012) requires that the. control room log contain specific information (steps
3.3.1-3.3.18 inclusive) and be considered an official document required by the
FSAR.
o

The required information includes:

(Ref.

38)

All alarms that involve-reactor core conditions must be recorded by theoperator along with an explanation' or reason for the alarm, e.g.,reactor coolant system pressure,Tave,flow, or power.

oPlant shutdown - Record-the major steps in shutdown and the associatedoEach system startup, significant status changes,

recorded.

and shutdowns must be

Also record major unit status changes such as opening of the

primary system and flooding of the fuel transfer canal and the time of theevent.

1-2-37oEquipment/malfunction - List the equipment and problem and any restriction

placed on the plant.
oAbnormaloperation-

Record any'condition that causes principal primaryor secondary parameters to vary from normal.
oReactor trip & turbine trip - Record the conditions. prior to the'trip,

cause of trip (ifdetermined),corrective action taken, and time of theevents.
oStart and stop of any radioactive gaseous or liquid releases must berecorded in the Control Room Log along with release permit number.
oAny abnormal valve line ups and equipment out of service or returned toservice must be recorded.
oChanges ofjposition of any "defeat",oAccomplishment of testing-or "bypass" switches must be recorded.

Record title and number of the 'test performed

and the start and completiontimes or time of suspension of the test.

The.performance of all periodic tests and inspections required by the Technical
Specifications just be recorded.
The above sections are not meant to be inclusive but merely indicate thetype of entries that should be made.

When doubt exists, an entry should bemade in.the log.
The shift foreman log requirements (Administrative Procedure 1012,
Section 3.5; Ref. 39) require that-the shift foreman'log contain:o

A summary'of the station operation and major events that bccur on eachshift.

Significant abnormalities that occur will be explained in greater

detail than would be expected in the control room log.

'I

I-2-38oWhen equipment covered by Technical Specifications is taken out of service,

the reason,time, Technical Specification requirements,and sample results (ifapplicable) will be noted on the left-hand page of the shift foreman's log.
Additionally,

all requirements for running, sampling,

testing will be noted,

delineating'times,' when the above must be accomplished.
On March 28, 1979,
0400 hrs and 1314 hrs,

the Control Room Log contained two entries between
including:o0400 hrs-

Turbine trip, reactor'trip,o0527 hrs-

Is-olated.OTSG. B

ES

The next entry was at 1315 hrs and was noted as a late entry.
The shift foreman log contained two. entries and a plant status stamp.
between 0400 hrsand 1315 hrs,

including:.

o0400 hrs-

Turbine trip, reactor trip, HP injection ES.

o0527 hrs-

Isolated S/G "B".

Evaluation
Technical Specification 6.8.1 requires that-written procedures be.
establisfhed,implemented, and maintained covering certain safety-related

activities.
Administrative procedure 1012, Shift Relief and Log Entries, specifies
the requirements for documentation,
plant operations,evaluation,',,and analysis' of significant

activities, and abnormalities.

The operators failed to provide adequate shift log entries of thesignificant ev.ents between 0400 hrs and 1315 hrs including:

1-2- 39I

o

HPI, flow rates and pathsoManual initiation of HPI at 0530 hrs

oShutting and opening the core flood tank isolation valves.

oEmergency feedwater valves EF-Vl2A and 12B discovered closed at 01408 hrs.

oBWST level decrease between 0400 hrs and 1315 hrs.

oMethod of decay heat removal (atmospheric,oEmergency borating activitieso

Reactor building isolation at.0756 hrsoDisabling the auto'start capability of the emergency diesel generators.

oIsolation of the EMOV (RC-R2)oThe extent to which HPI was throttled during various periods throughout

condenser, HPI)'

at 0618 hrs by shutting the block valve (RC-V2).

the event.
The above are examples of items having significant safety implications
which were not available-to the licensee or the investigators after theaccident.
The failure to provide the required logs of events and activities as
required by Administrative Procedure 1012, Shift Relief and Log Entries,

isunder consideration as a possible noncompliance.
NOTE:

The examples chosen are those that are not recorded on the alarm

printer or shown on other instrumentation.

The operators are authorized

by procedure to clear the computer memory in order to have the alarm
printer catch up with current events.

1-2- 40

Had that been done,much of theinformation utilized this and other investigations; including that of thelicensee, would have been lost.
2.13

Operator Actions Concerning EMOVThe operators .stated that they realized that the RCDT rupture disc had',blown,

but they were not alarmed by this because of'the anticipated lifting of
the EMOV following the apparent severe transient (routine-following a
Turbine-Reactor Trip).

(Int:

17,

38, 15,95)

The subsequent high temperatures on the EMOV and code-relief valves'were
similarly not considered unusual.

The operators stated that a key issue wasthat so many things were happening within the first hours of the transientthat time passed very quickly.

The awareness of elapsed time after the unit

trip became secondary to the operations group.

(Int:

17,

38, 95)An operator stated that the EMOV discharge temperature indicating 283'F,
60 0 F above the code relief valves within minutes after the trip and the time
wasactually 0521 hrs did not appear abnormal following the transient.

It"would appear the operator did not realize that it was 81 minutes into theaccident due to the continuing sequence of events.

(Int.

95, 196)

Also, the operators indicated .that the EMOV and code relief valve
discharge temperatures were not significantly above the normal operating
temperatures of approximately 200'F, which is just above the alarm setpoint.,(Int:

17, 38,15, 95)The EMOV block valve was closed at about 2 hours and 19 minutes after theaccident (0619 hrs).

The closing of the block valveat that time occurredwith a smaller differential

temperature (26 0 F) between the relief valve

exhaust lines than had been previously observed (0521hrs).

The interviews

also revealed that the check of the relief valve temperatures at 0618 hrs were
performed by another shift supervisor.-(Int:

1-2-41105)-

Evaluation
The operation of the plant with the EMOV'and code relief valve exhaust
temperatures (about 200'F) significantly above the ambient conditions'(lessthan 130'F by procedures)complicated the diagnosis of the EMOV-LOCA event andsubsequent isolation of the EMOV by closing the block valve (RC-V2).
operati.on of the plant-with the high (200'F vs.

130 0 F)The

exhaust temperatures on

the relief valves is addressed in another section of this report (Section 1.2
of Details I).'
Failure to shut the EMOV block valve until about 0619 hrs, (2.3 hours
after the EMOV initially opened) is a second example of failure to implement
Emergency Procedure 2202-1.5, Pressurizer System Failure,
Section 1.2'of Details'-I,

as discussed in

and is under consideration as a possible item ofnoncompliance pursuant to Technical Specification 6.8.1.a.
2.14 General Location of Operation Personnel
As noted in Section 1.4.2, the shift complement on the morning of March 28,
1979, met the Technical Specification requirements-

The location of theoperations personnel during the event shown in Table 1-2-1 was determinedbased on selected activities performed and interviews with the personnel
involved.

The tabulation does not include all operations'personnel

during the accident.

on siteThe listing provides a general guide to the location ofcertain individuals.
The licensed operators performed their functions from the control room
area, with a few exceptions,

including:.

Shift Supervisor E worked with the condensate polisher bypass valve(CO-V12)

between 0420 hrs and 0500 hrs.

Shift Foreman A unlocked and closed the electrical breakers for the decay
heat removal 'suction valves (OH-VlO2A'and

DH-VIO2B) at MCC 2-11EA and 2-21EA

at about 1000 hrs.

1-2- 42

Shift Supervisor C made a limited tour of the auxiliary building rad-waste
panel

and accessible floor drain areas at about 1300 hrs.

The auxiliary operator actions during the first
were directed by the control

16 hours of the accident

The auxiliary operators who were inroom staff.

their assigned areas performed checking activities in their respective areas
following the announcement of the turbine and reactor trip.
operators outside their'assigned areas,
room for standby. -Thein general,

The auxiliary

went to the Unit 2 control

more experienced auxiliary operators made checks onequipment as they proceeded to the controlroom.

1-2- 43TABLE 1.2-1
LOCATION OF KEY PERSONNEL INTERVIEWEDControl Room
Individual
Shift Supervisor E........

..

Shift Supervisor A
Shift Supervisor D
Shift Supervisor B. . . . ....

. . . . . o..

. . . ....

....

..

Elect

Makeup0420-0500. . . . ..

. .

o.......o..

. . . . . . . .°. . . . . . . . .

0500-1600 ..................
0408-0630 ...............
0400-0402

0550-1600 ..................
1200-2000(+)...............
1300-1330(TOUR)

Shift Supervisor. C
!N~

Waste

Turbine Bldg.

0400-0420 ...................

°.........

. ..

Back
PanelsSEC

PRI

Auxiliary Bldg.

Shift Foreman C

0402-1500 .................

Shift Foreman A.................

0545-1000 ..................

Shift Foreman B.................

0700-2000(+). ..............

Control Room Operator D

0400-0600
0600-1500
0400-0600

Control Room Operator C
0600-1500
Control Room Operator I

0615-1930

Control Room Operator, A

0630-1800
1800-1900

1000-1030

TABLE 1.2-1. (Continued)
LOCATION OF KEY PERSONNEL INTERVIEWEDControl Room
Individual

PRI

Control Room Operator F

SEC

Auxiliary Bldg.
Back
Panels

Waste

Elect

Turbine Bldg.
Makeup0600-1600

Control Room Operator H

1600-1800

Control Room Operator G

1800-2000(+)

Control Room Operator K

0700-1300
1300-1700

Auxiliary Operator G

0400-0410
0410-0420.

I.

0420-0500

U,

0500-0600
0600:0700
0700-1500
Auxiliary Operator.B
0410-0420
0430-0500

0400-0410
0420-0430
0500-0530
0620-0630

0630-0650
0700-1500
Auxiliary Operator K

0400-1030

TABLE 1.2.1 (Continued)
LOCATION OF KEY PERSONNEL INTERVIEWEDControl Room
Individual

PRI

SEC

Auxiliary Building

Back
Panels

Waste

Elect

Turbine Building

MakeupAuxiliary Operator E

0400-0500
0500-0600
0600-0620
0620-0640

Auxiliary Operator C

1400-2000+

Auxiliary Operator H

0500-0600
0700-1300

0600-0700

Auxiliary Operator D
Engineer K.....

0400-0800
0550-0600..............

0700-0900
1030-1630..............
...............

0600-0700
1630-1645

2.15 Effect of Training on Operator Actions
Members of the plant trainingstaff were interviewed regarding the effectof the operator's training on their actions during the incident.

(Int. 80)

Among the areas-discussed were:pressurizer level versus pressure control
recognition of an open EMOV
bypassing of emergency safety features actuation system
shutdown of reactor coolant pumps.
2.15.1Pressurizer Level Versus Pressure Control

The plant procedures (2103-1.3)

and the operator training program specify

that the pressurizer shall not be allowed to go solid at any time except forhydrostatic testing of the RCS (Ref.

53).

The training emphasizes this pro-hibition, stressing the possibility of exceeding the high-pressure safety limitof 2750 psig because the pump discharge head is 2900 psig.
For high pressurizer
level,

the procedures (2103-1..3)

(Ref.

53).

require securing makeup and increasing letdown

For low pressurizer pressure,isolate letdown,

increase makeup,

the procedures require the opposite:and, in addition, turn on the heaters.

The training staff was asked what the operators would be expected to do,
based on their training and experience,if

a high pressurizer level indication

called for one set of actions and a low RCS pressure called for another.

Membersof the training staff stated that the operators would definitely have reacted
to the high level to avoid going solid. This is based on both the TMI training
and the B&W operating procedures at the simulator.

The necessity of maintainingpressure is stressed in connection with-the avoidance of departure from nucleateboiling (Int. 80).
The staff was asked if the operators were taught the significance of
saturation pressure.

The training staff stated, that the operators receive

1-2-47this instruction as part of the basic thermodynamics training.

The staff was

asked if the saturated condition would cause the operators to suspect steam
voids i,n the primary system.

They indicated that under the conditions existingat the beginning of the event, the training staff would not expect the operators
to check for the saturation condition immediately.
expect voiding with the pressurizer full.

The operators would not

With the injection of cold auxiliary

feedwater supply and high pressure injection the operators would expect apressure reduction.

(Int. 80)

The training staff was asked if the possibility of a level rise in' thepressurizer caused by steam flashing in another part of the primary system hadever been recognized and brought to the ope'rators' attention. The answer was"never."
The only training in this area is the discussion of the possibilityof flashing in the hot legs if the pressurizer level is not maintained.

(Int.

80)
2.15.2

Recognition of an Open EMOVThe training staff was asked if the operators were trained to verify theclosure of the electromatic relief valve following events that can be expected
to result in its opening.
was open,

They stated that they were trained to check if it

but considering the other events that were occurring during theaccident, they would not have expected the operators to check this.right away.
The means available to check this were the console demand signal,closed, and the discharge line temperatures.

which indicated

The training staff stated that'high discharge line temperatures were not very meaningful because the EMOV had
been leaking prior to the incident, which resultedin temperatures that were
not much lower than those existing with.the valve open.

Moreover,these

temperatures and their status are printed out.by the alarm printer, and thesealarms would not get printed out for 20 to 30 minutes.

This time delay, under

the conditions of the accident is caused. by the large number of alarms to beprinted, and the limits on the typing speed of the Alarm Typer.
The training staff was asked. if the operator training included actions to
be taken if

there was a pressure rise in the reactor coolant drain tank (RCDT).

1-2-48They stated that the training on the "Response to High RCDT Alarm" procedure(No.

2204-301B,

Ref. 26) covered this.

However, the alarm and indicators

for this system are located behind the back panels.

Also,

determining thesource of the leakage requires a process of elimination because the RCDT
receives other leakage such as the RCP seals and valve packing leakoffs.
i.15.3

Bypassing of the Engineered Safety Features
Actuation System

It was noted that the engineered safety features actuation system wasbypassed by the operators promptly after actuation, even though the coolant
injection might not be throttled back until later. The training staff was
asked if the operators were trained to do this.

The staff stated that theoperators were trained to reset as soon as possible.
injection of sodium hydroxide into the reactor.

This is done to prevent

In addition, the operators

were trained to be prepared to maintain a 220-inch level in the pressurizer bythrottling the HPI valves.

The operators also had to be prepared to throttle

the flow to the makeup pumps to prevent exceeding the 550-gpm flow limitation,
as the flow would increase if the RCS pressure decreased.
2.15.4

Shutdown of the Reactor Coolant Pumps

The training staff was questioned as to the training given the operators
regarding shutdown of the RCPs.
Revision 3) and training,

They stated that the procedures (2203-1.4,

required pump shutdown for high vibration, low amperageor low reactor coolant flow, all conditions that existed during the event.
(Ref.

78)

In the training staff's view, the conditions that existed during

the March 28 event did require shutdown of the pump.

They stated that theoperator is trained that failure to trip the pump under these conditions could
lead to pump seal failure or loss of the impeller.
Evaluation
The operator training program had a substantial impact on decisions madeduring the early phase of the accident.

The prohibitions aaainst allowing the

1-2-49

pressurizer to go solid caused the operators to neglect the low RCS pressurein their attempt to control pressurizer level.

In addition the operators wereconditioned to promptly bypass ES without first determining the condition of
the RCS.
However,

it must be recognized that operator experience also played an

important role in these decisions.

The normal course of most ES initiations,those which did not involve a loss of coolant, reqired bypassing of ES andsecuring of HPI to prevent an extended plant outage and/or possible damage tothe pressurizer safety valves.

Past failure to bypass ES promptly had, in fact,

resulted in the injection of sodium hydroxide into the RCS at TMI2.
The operator training did not address the phenomenon which led to thepressurizer level transient.

The investigation has not established whether

operators at other B&W plants had received such training.
2.16

Nonlicensed Operator Actions During Accident

2.16.1

GeneralThe nonlicensed operator (auxiliary operator) actions during the first 16
hours of the March 28,

1979, event were directed from the control room.

The

auxiliary operators who were in their assigned area performed checking activities
in their respective areas following the announcement of the turbine and reactor
trip.

The auxiliary operators outside their assigned areas,

to the Unit 2 control room for standby.

in general,went

The more experienced auxiliary operators

made checks on-equipment as they proceeded to the control room.
2.16.2

Activities Prior to Trip

The major activity just prior to the turbine trip was being performed in
the turbine building at the condensate polisher unit.

There was a problem intransferring resin out of condensate polisher vessel No.
was plugged with resin for about 11 hours (Int:

123).

7.

The transfer line

There were at least

three individuals working on the problem just before 0400 hrs.

1-2-50This included

two auxiliary operators and the operation shift foreman.
was making use of station air and demineralized water.

The unclogging process
Other activities being

performed just prior to the event included adding hydrogen to the main generatorhydrogen system, performing readings on industrial waste treatment systems,
and logging and maintaining the radwaste panel in the auxiliary building.

The'

consensus of opinions from the interviews involving the auxiliary operators
indicate that the plant conditions were,, in general,
36,

38,. 104,

109,

115,

123,

in the auxiliary building.

125 and 128).

fair to good (Int.

10,

There was a large inventory of water

Water was estimated to be within 8 inches of over-flow in the auxiliary building sump.

This situation has existed in the past

because of the way the units are restricted i~n the release of waste water.
The units share release of water in that only one plant can release at a time.
The schedule for release of Unit 2 water and the status of the Unit 1 water
inventory was not known by the auxiliary operator on March 28,
2.16.3

1979 (Int.

10).

Activities During Accident

The following paragraphs are summaries of auxiliary operator interview
transcripts.

They give a general picture of typical activities on March 28,

1979:
Auxiliary Operator D (Int. 61,
on March 27-28,
the plant.

1979 (Int. 61).

128) was on the 2300 hrs to 0700 hrs shift

The assigned area was the secondary side ofOperator D was on the way to and very near the control room at

0400 hrs and apparently just outside the control room at the time of the trip
announcement.

This operator was assigned to go' to the turbine building toturn the main feedwater pump B shaft.
hrs.

This effort continued until about 0800

Independent checks were made on the main condenser vacuum pump systemand the operator aided in checking the hydrogen oi.l seal system.
Auxiliary Operator G was involved in the condensate polisher problem andactively engaged in unblocking the resin transfer line (Int.

36).

He checked

the condensate, polisher panel and the status of the outlet valves at thecondensate polisher tanks immediately after the trip was announced.
condensate polisher valves were reported to be closed.

1-2-51

The

After completing alineup of the condensate polisher for restart, he proceeded to the control roomat about 0415 hrs to 0430 hrs and was sent back to recheck the condensate polisher
valve lineup.

He noticed leakage from condensate booster pump 2A suction line

and proceeded to isolate the leak, with help from others (Int.
the booster pump suction valve.

10), by closing

He associated this leak with a major movementof this pipe that he had observed just after 0400 hrs prior to going to thecontrol room the first time.
bypass valve (CO-VI2)

He then,_ with help, .manually opened the condensate
He also checked the pressurizer level indicator(Int. 10).

inside the auxiliary building.

At about 0530 hrs he was told to open MU-V127

to allow for emergency boration.
Auxiliary Operator B was assigned to the radiation waste panel in theauxiliary building on the 2300 hrs to 0700 hrs shift (Int.
his area has been described above.

10)

The status of

The miscellaneous waste storage tank level

was stated to be at 7.4 feet just before 0400 hrs and-lined to the RB s'ump pumps.
At 0400 hrs he was just outside the Unit 2 control room, by the I&C area.

He

returned to the control room at about 0410 hrs and was sent by the control room
operator to check breakers for condensate booster pump 2B.

He found the breakers

to be in good status.

He was involved in throttling river water to the inter-mediate closed coolers.

He went to the auxiliary building, his assigned area,

to make a check of his area.
sump pumps were on,

He found indication that both reactor building

the local reading of the reactor building sump level was,pegged high (over 6 feet),and the background activity in his assigned area of

the auxiliary building had increased.
hrs to 0500 hrs.

He estimates this time to be about 0430

He communicated his finding back to a control room operator.

He turned off the reactor building sump pumps at about 0438 hrs at the requestof the CRO.

The pumps were said to be lined up to the miscellaneous waste holdup

tank, but he did not notice a level change from the 7.4'feet seen'earlier.
did not see any overflow of the auxiliary sump at that time.

He

He was told by a

CRO to check for makeup valve alignment for demineralized water to the'makeup
system.

He noticed while assisting closure of CO-V12 that efforts were underway

to lower the hotwell level.

He also noticed that the air supply line to thepneumatic actuator to the normal condensate reject line was failed and blowing
air.

He noted that the shift supervisor was also in this area at this time

1-2-52
11-

(TMI 220, 221).

At about 0600 hrs, he was back in the auxiliary building andfound water backing up out of the floor drain. He was at the radwaste panelat the time the site emergency was announced.

He then went back to Unit 2 control

room to report the status of his area to the shift supervisor and CRO.
Auxiliary Operator H was on the 2300 hrs to 0700 hrs shift assigned tothe industrial waste area (Int.

38 and 125).

He reported to the Unit 2 control

room at approximately 0445 hrs after being paged by another auxiliary operator
at about 0420 hrs.

He was directed by the shift supervisor to check the main

turbine and to ensure that it went on turning gear.

He was also told to isolate

the second stage reheat of the moisture separator and reheater units at about
0510.

He was told, by the operating engineer, to isolate the outlet valves on

the condensate polisher system, which he did.

He had found all outlet valvesclosedand proceeded to drop air pressure to the valves,

which would ePsure

closure.
Auxiliary Operator E reported to the control room after hearing the noisefrom the main steam safety relief valves (Int.
tasks by the control room operations personnel.

109).

He was assigned specific

He stated that he suggested

the stopping, of the reactor building sump pumps based on the auxiliary building
high water inventory.

He recalled that one condensate pump remained on line.
He assisted in opening the main condensate reject valve at about 0500 hrs to0530 hrs.

He closed the breakers for the core flood tank isolation valves atabout 0600 hrs.

He searched for the cause of the fire alarm.

valve lineup for boric acid,

He checked thelevel of the miscellaneous waste holdup tank, andvalve lineup for possible demineralized water entering the makeup system.

He

recalled seeing water coming out of the auxiliary building drains (before 0730
hrs).

1-2-53,

3.
3.1

MANAGEMENT ACTIONS DURING ACCIDENT

GeneralThis section of the report deals with the actions and management decisions

undertaken by those members of licensee management who were called to the site
to provide emergency direction to cope with the operational aspects of theaccident.

Moreover, this section will address the additional support that wasprovided through the licensee organization and by other parties to support theonsite operational activities.

Those actions taken to cope with the radiologicalaspects of the accident are-addressed in Details II of this report.
The use of the word management in this section refers to those individuals
who would be summoned to the site for emergency and technical direction as
well as those contacted to organize appropriate support activities.

The

investigation team recognizes that members of the routine shift organization
do represent levels of licensee management.
3.2

Onsite Operational Activities of Senior Management
Management notification was initiated by the Unit 2 Shift Supervisor

shortly after the turbine and reactor trips.

At the request of the Unit 2

Shift Supervisor, the Unit 1 Shift Foreman called the Station Manager (approximately, 0401 hrs) and the Unit 1 Operations Supervi-sor (approxi-mately 0435 -hrs).
Nuclear engineers,

coming to Unit 2 from the Unit.l. refueling startup crew,

called the Unit 2 Superintendent and the Unit 2 Superintendent - TechnicalSupport (approximately 0410 hrs).

These senior managers were informed of theUnit 2 turbine and reactor trips, as a matter of normal site policy and not
as a result of any suspicion that this trip was unusual.
Supervisor was not called, since he was in Lynchburgh,
simulator training (Int.

1, 17, 26, 27,

The Unit 2 Superintendent-

83,

The Unit 2 Operations

Virginia attending B&W

148).

Technical Support, the assigned on-call "dutysection head" at that time, was the first to arrive on site (approximately 0450

1-3-1hrs).

He was informed of the problems being experienced with pressurizer

level being high,-reactor coolant system (RCS)
coolant drain tank (RCDT)
actuation/pressure being low, reactor

rupture disc being blown, and that an automatic ES

high pressure injection (HPI)

had occurred.

He was not informed ofthe earlier failure of the emergency feedwater system to perform as designed,
due to the improper position of the EF-VI2A andl12B valves; this fact was notlearned by onsite management until engineers began reviewing the sequence of
events, a day or so later.
NOTE:

The Unit 2 SuperintendentW-

Technical Support is the chairman of

the Unit 2 Plant Operating Review Committee and holds a senior reactor
operator license on Unit 1..

He had just started an informal training

program for Unit 2 and he doubted his capability to recognize the meaningof alarms or to interpret everything he was being told.

Technical

Specifications do not require this individual to.be licensed.
Lacking familiarity with the plant, but recognizing an unusual situation,
the Superintendent - Technical. Support directed additional technical andoperations personnel be called in.

Those called included the Unit 2 Operations

Engineer (approximately 0501 hrs), the Station Chemistry/Health Physics Supervisor (approximately 0510 hrs), the Maintenance Superintendent (approximately
0515 hrs) and at least ten others.

(Int. 27,

129; Ref. 21)'

As RCS pressure continued to fall, the Unit 2 Shift Supervisor secured
the reactor coolant pumps (RCPs)

in the B loop.

The Unit 2 Superintendent-

Technical Support reviewed net positive suction head requirements for'the
RCPs and concurred with the decision made and action taken by the Shift
Supervisor.

During this period, the Superintendent - Technical Support was

attempting to become better informed relative to the status of the plant,
making suggestions and offering technical advice, and requesting additional
plant personnel report to the site.

The direction of plant operations continued

to rest with the Unit 2 Shift Supervisor.

1-3-2

(Int.

27, 135)

i ý".I .

tJ

Having received no additional information since the post-trip notification, the Station Manager called Unit 2 (approximately 0515 hrs,
approximately 0535 hrs) to ascertain the status of the plant.

lasting to

The Station

Manager was informed by his-management representative of the current plantconditions and the fact that an automatic actuation of HPI had occurred.

The

Station Manager was disturbed by the coincidence of low RCS pressure and highpressurizer level.

The automatic actuation of HPI was not unexpected,

on experience with previous turbine/reactor trips.

based

The Station Manager decided

to initiate a conference call with selected technical people; but first (approximately 0545 hours) called the Unit I Superintendent,

informing him of the pecu-

liar trip conditions and requesting that Unit 1 cooperate in giving Unit 2 heating steam, and called the Maintenance Superintendent,

requesting he report to theplant.
NOTE:

During the period of the Station Manager's call to the Superin-tendent - Technical Support,

the Shift Supervisor had isolated the B once

thru steam generator (OTSG).

The Shift Supervisor had taken this action

on finding reactor building pressure increasing and B OTSG pressure belowthat in A OTSG; which indicated to him a secondary-to-containment leak.
(Int.

1, 27, 7.1, 77,

120)

Following his conversation with the Station Manager,
Technical Support found the RCS pressure dropping again.

the Superintendent-

With indicated flowin the A Loop at about 30% and dropping, the Shift Supervisor tripped the A
loop RCPs.

Everyone just assumed natural circulation would occur; despite thefact that RCS parameters were outside necessary conditions for. subcooled
natural circulation.

(Int. 27,

38, 135)

With all RCPs tripped, the core was being cooled by HPI water steaming
through the still partially or fully open pressurizer electromatic relief
valve (EMOV)..

The Unit 2 Superintendent (senior reactor operator licensed on

Unit 2) arrived onsite (approximately 0545 hrs) and proceeded to the Unit 2
control room, where he was briefed by the Shift-Supervisor and the Superintendent - Technical Support.

1-3-3

NOTE:The Unit 2 Superintendent lives approximately 50 miles from thesite and his initial notification provided no evidence of an abnormal
trip, which might have prompted him to arrive earlier.
At the request of the Unit 2 Shift Supervisor,

the Unit 1 Operations

Supervisor (senior reactor operator cross-licensed on Units 1 and 2),
in Unit 1 since about 0530 hrs,

who wasto assist in its startup, reported to theUnit 2 control room (approximately 0600 hrs) to assist as needed.
pressurizer level versus RCS pressure,the tripped RCPs,

The abnormal

and the results of an

RCS post-trip boron sample showing a drop of 300 ppm from pre-trip conditions,were discussed.

Shortly thereafter, on seeing indication on the s'ource andintermediate range nuclear instrumentation of what appeared to them to be apossiblereactor restart, the Shift Supervisor was directed by the Operations
Supervisor and the Superintendent - Technical Support to high pressure inject
and emergency borate.

(Int.

5, 9, 26, 27, 83)

The Station Chemistry/Health Physics Supervisor had arrived onsite(approximately 0545 hrs),
tions personnel,

following the earlier call for technical and opera-

directed by the Superintendent - Tech-nical Support.

initial assignments,

His

from the Superintendent - Technical Support, included

making preparations for a reactor building (RB)(Int. 20. 102,

boron sample results.

entry and confirmation'of RCS

129, 41; Ref.

21)

With the Unit 2 Superintendent and the Unit 1 Operations Supervisor
following operations,

the Superintendent - Technical Support retired to theUnit 2 Shift Supervisor's office to participate in the conference call establishedby the Station Manager.

The conference call (approximately 0600 hrs,

lasting

to approximately 0635 hrs) additionally included the B&W Site Operation
Manager and the Vice President -- Generation, Metropolitan Edison (the first
corporate individual to be notified).

The status of the plant was discussed

and it was decided that forced circulation must be reestablished.

The fact

that all available indicators of pressurizer level were checked and found inclose agreement was discussed and it was decided they were to be believed.
The condition of the EMOV was questioned and it was reported to be shut,reportedly based on a demand position light indication.

The B&W Site Manager

and the Station Manager were encouraged to report to the site.
1-3-4

NOTE:

During the period of this conference call,the EMOV block-valve

was, closed; finally isolating the loss-of-coolant accident path'that had
existed since 0400 hrs.

Whether this action was keyed by this discussion

has not been established.

(Int.

1, 27,

53', 71)

With the EMOV block valve shut, RCS pressure began to increase and Reactor
Building pressure began to drop.

The results of a second, RCS boron sample

indicated a further decrease in boron concentration; which the Station Chemistry/
Health Physics Supervisor had confirmed'by splitting the sample and observing
the analysis performed by two technicians.

Following the report of theseresults, the Superintendent - Technical Support then directed the Unit 2
Operations Engineer (senior reactor operator, licensed on Uni't 2), who had
arrived on.site (approximately 0545 hrs) following his earlier call out, toinvestigate possible.sources of demineralize.d water that might be diluting
the RCS boron concentration.

The results of this investigation reportedly-.were never made; presumably the investigation was aborted by subsequent events
related to the declaration of the Site Emergency.

(Int.

20, 27, 102, 129)

With'the RCS sample lines on recirculation to ensure a representative
sample,radiation levels in the vicinity of the sample lines began to increase.

'Responding to an alarm,the Station Chemistry/Health Physics Supervisor discovered
radiation levels of 600 mil~lirem per hour (mr/hr)'and reported these results
to the Superintendent - Technical Support.
information to the Unit 2 Superintendent,
experiencing fuel failures.

He,in turn, then reported thealong with his belief'that they were

The Unit 2 Superintendent directed the MaintenanceSuperintendent, who had arrived on site (approximately 0615 hrs) at the earlier
direction of the Station Manager,of this event.
27, 71,120,

to call the Station Manager and'inform him

The calliwas made as directed (approximately 0650 hrs).

135)

(Int.

20,

_

RCS pressure had now increased to the point that RCPs could be operated.
Each manager recognized the need to, re-establish core cooling.

Initial attempts

to start RCPs under the direction of. the Unit 2 Superintendent,

the Superintendent

TechnicalSupport and the Unit 1 Operations Supervisor were unsuccessful.

1-3-5

The'

-

operators finally succeeded in starti'ng RCP-2B (0654 hrs) and the source andintermediate range nuclear instrumentation showed significant drops in flux
levels, convincing the Superintendent-Technicalreactor restart had been terminated.

Support that the apparent

Within minutes radiation monitors through-out the N plant entered alarm status and a Site Emergency was declared (approximately 0656. hrs) by the Unit 2 Shift Supervisor at the, direction of the Superintendent - Technical Support.

(Int. 2.6,

27, 83,

129,

138)

Immediately following the Site Emergency declaration,

the Shift Supervisor

returned his attention to the primary plant'and reinitiated HPI at the direction
B OTSG was again isolated due to radiationof the Unit 1 OPerations Supervisor.

alarms on the condenser vacuum pump exhaust; it

had been unisolated following

discovery that the leaking EMOV was apparently the source of the Reactor
Building pressure increase.

The Unit 1 Operations Supervisor stationed operators

at each Unit 2 control room panel,coordinating their activities and backing

them up with normal shift and relief shift foremen and supervisors.

The

Superintendent - Technical Support directed various engineers (then present
due to his earlier call-out of technical and operations personnel)
required Emergency Plan notifications.

to make theWith the Station Manager in transit to

the sitea,,th.e. Unitt 2 Superintendent became- the- Emergency- Director -(per TMIEmergency Plan,-Procedures 1670.2,
Ref.

Revision 8).

(Int.

17, 26, 27,

71,

83, 38;)The Unit 1 Superintendent (senior reactor operator cross-licensed on Unit

1 and Unit 2)

had arrived on site (approximately 0645 hrs),

heard the announce-

ment of the Site Emergency, proceeded to the Unit 1 control room (per TMI Emergency Plan,

Procedure 1670.2 Rev. 8) and declared himself the Emergency Directorof that unit after being briefed on the situation.

The.Station Manager (notcurrently licensed or required to be licensed on either unit) arrived onsite(approximately 0705 hr's) and proceeded to the Unit 2 control room, where he
was briefed by the Shift Supervisor and his managers.

The Station Manager

then declared he was the Emergency Director and established an emergency
command team.

The Unit 1 Operations Supervisor was put in charge of operations

to direct the Shift Supervisor; the Station Chemistry/Health Physics Supervisor
was put in charge of onsite and offsite radiation and environmental concerns;

I--3-6the Unit 1 Superintendent-was put in overall charge of Unit 1 and the Emergency
Control Station, then being established in Unit 1; the Unit 2 Superintendent
was put in charge of reviewing and verifying personnel complied with proceduresand plans; the Maintenance Superintendent was put in charge of emergency maintenance; the Superintendent.-

Technical Support was put in charge of notifications,-communications and technical support; and-the B&W Site-Manager, who had justarrived onsite (approximately 0715 hrs) was requested to provide technical assistance and communications with B&W.

The Station-Manager further declared these

personnel were to be the funnels through which informationrelative to areas
under their charge would be directed to or from him.

During this period, theUnit 1 Operations Supervisor and the Shift Supervisor secured the running RCP,
since it showed no flow and a running current of about 100 amps,

and establishedthe dominant-mode of core cooling to be utilized for the next 13 hours, with
HPI feeding cool water to the RCS andthe EMOV and/or its blockvalve open to
remove hot steam.

Attempts to notify the Vice President - Generation; Metropoli-tan Edison of plant status at this time were unsuccessful.
53, 71,77,83,(Int. 20', 26, 27,

120, 38)
2

By about 0724 hrs,

conditions had deteriorated to the point where theStation Manager declared a General Emergency..
NOTE:

Management actions relative to radiation protection, environmental

protection and implementation of the Emergency Plan are discussed in
Details II of this report and will not be repeated here.
NOTE:

From an .operational standpoint, TMI Emergency Plan,,Procedure 1004,

Revision 2,states ... "The Operational personnel are responsible *for thesafe operation and recovery of all systems during an emergency situation."
The Emergency Plan also requires a significant effort in initial notificationof and maintaining communications with, outside agencies and offsite personnel; both of which distracted technical people from the'task of grasping
the significance of events, understanding the situation, and containing
the accident.

(Int. 27,

71)

1-3-7

The Station Manager called and directed the Unit 1 Superintendent (approxximately 0745 hrs) to join him in Unit 2.

The Station Manager then requested his

emergency command team caucus with him in the Shift Supervisor's office.,The

meeting was held away from control room activities to allow more consideredthought relative to plant status, long term goals, potential actions and.,theirconsequences.

It was decided to attempt another RCP start.

-

RCP-lA wasfinally

started, but loop flow and motor current failed to respond normally and everyone
was forced to acknowledge the loops were vapor bound.

(Int. 26,

27, 53,,71,'77)
Unable to'resolve what appeared to them to be discrepancies in RCS param'eters, the emergency command team opted to direct emergency core coolingsystems be allowed to function as they would -had the operators not been present.
The Station Manager directed the Unit 1 Operations Supervisor to keep HPI on
and not to secure it without h-is permission.

Instrument technicians weredirected to obtain local readout of core thermocouples (T/Cs)
leg resistance temperaturedetectors(RTDs).

and RCS loop hot

The emergency command team would

meet frequently during the following hours, usually once or twice an hour.
Plant status would be reviewed,
action considered,opinions and fears expressed,

decisions made and orders issued.

courses ofSubsequent meetings

would discuss results of actions taken and/or communication held with offsite
individuals or groups.

These meetings allowed a sharing of knowledge,butreportedly none of the participants doubted, that the Station Manager was, incharge and that, ultimately,'it would be his decision they would carry out.
It should be noted that the character of these meetings reportedly did notchange,71,

even after the arrival of NRC and GPU personnel.

(Int. 26, 27,153,

129.)
By 0900 hrs,

the plant had experienced one manual ES actuation and ltwo

automatic ES actuations/reactor b.uilding (RB)of the Site Emergency.

isolations,since the declaration

Temperatures in the RCS hot legs and the core indicatedsuperheated vapors were present.

The core cooling that did'exist was provided

by HPI water that flashed to steam in the core and was being vented to thereactor building via the EMOV.

The emergency command team thought the core

1-3-8

was covered, but were not sure.

They acknowledged that temperatures werehigh, but didn't believe temperatures were as high as some readi.ngs had
indicated (2620'F on one core thermocouple).

'They acknowledged the RCS loops

were vapor bound, but couldn't figure a way to refill them.* They knew HPI wasbeing injected, but feared it might be bypassing the core.

They recognizedthat without changing their tactics, clean Borated WaterStorage Tank (BWST)
water would ultimately be exhausted and radioactive dirty water from spillage
on the reactor building floor would then have to be utilized, increasing
auxiliary building radiation levels, significantly.

Finally, they were nowhere

nearer the two possible stable RCS conditions of forced circulation,using

either RCPs at high pressure or decay heat removal (DHR). pumps at low pressure.
The Vice President " Generation called (approximately 0910 hrs) and learned thecurrent status of the plant from the Superintendent - Technical Support andpossibly others.

Shortly thereafter, the Station Manager directed the plant

be repressurized using HPI,

maintaining pressure between 2000 and 2100 psig

with the EMOV block valve.

This pressure range was picked to prevent actuationof the code safety valves and the attendent danger of uncontrolled depressurization shouldkthey stick open.
Ref: 21,(Int. 26, 27,

53, 71,70, 83,113, 129;23, 22)

RCS pressure was increased in an attempt to collapse the. voids in the RCS
loops.

The decision to repressurize lacked unaminity,that iftemperatures were anywhere near correct, the code safeties would not.

since it was pointed out.

allow sufficient pressure to condense the steam, of which the voids were assumed
to be .composed.

Further,'the history of poor reliability of the EMOV block

valve raised concerns for.the consequences of its failure in an indeterminate
position.

The Vice President-Generation again called (approximately 1010 hrs)

to obtain an update on plant status from the Station Manager.

(Initial notifi-

cation, communications and activitiesof other Metropolitan Edison offs~ite personnel and those of General Public Utilities are covered el.sewhere in this report
and will not be repeated here.)(Int..26, 27, 31, 53, 59,.71,

Ref: 21)

1-3-9

83,

129, 38,

The B OTSG remained isolated at the direction of the Station Manager
after the Site Emergency was declared.
not contaminated.

Samples from A OTSG confirmed it was

It was decided to attempt to gain natural circulation

cooling by filling the A OTSG to a 90% 60"erating level and steaming it throughthe atmospheric steam dumps.

As a precaution, an operator was stationed on

the roof with a radiation detector as close as practicable to the exhaust,
should this become a radioactivity release path.

(Int.

16, 27,

31,

77,

83,

41)
The concern for the failure of the.EMOV block valve p~rsisted.
1120 hrs,

By about

the emergency command team decided to increase the pressure control

range from-200.0 to-2100 psig to a range of 1900 to 2100 psig, thereby cutting
the frequency for cycling this valve in half.
The Vice President-(Int. 27,

83; Ref. 23)

Generation (not currently licensed or required to belicensed on either unit) arrived at the Observation Center (approximately 1140hrs) and called the Station Manager *to learn the current plant status.

The

Vice President - Generation indicated during interviews that he did not proceed
to the.Unit 2. Control Room and insert himself in the command chain,sincecompetent people were in charge, performi-ng tasks for which they were trained.
About this time itwas decided to attempt to depressurize the RCS to obtaincore flood tank injection and possibly enable DHR.

This decision was reportedly

the result.of concerns related to the lack of positive indication of core
coverage or natural circulation; fears that the EMOV block valve might fail,that boron might be plating out through crystallization,and that HPI might bebypassing the core; and recognition that the plantappeared no closer toforced circulation by RCPs or DHR pumps,
BWST inventory had been experienced.

even though a significant drop in

A restart of RCPs was not considered

here, since RCS loop and pressurizer parameters had not changed significantly.
Consideration of an RCP restart would be delayed until the emergency command
team was convinced the loops were sufficiently filled and subcooled to prevent
RCP cavitation and seal destruction.

(Int. 26, 27, 53., 71,

38, 41)

1-3-1077,

83, 91-,

113,

While maintaining HPI, the RCS was-depressurized using the EMOV,
block valve and the pressurizer vent valves.

EMOV

RCS pressure bottomed out at 440

psig, but-not before some small (approximately 1 foot based on interviews) dropin core flood tank level was experienced, convincing the emergency command
team that the core was indeed covered.
pressure would not drop further.

HPI was reduced'to minimum, but the(Int. 26,

27,

53, 71,

77,

113)

As the afternoon progressed, the Vice President-Generation worked to
improve communications with offsite and onsite groups.

The Superintendent-

Technical Support, recognizing the need for information following recovery
from this event, directed an engineer to begin a chronologicallog of events.

The Station Manager came under increasing pressure to secure steaming A OTSG
through the atmospheric dump.

Without condenser vacuum,

hesitated to secure this system,

the Station Manager

since, in so doing, he thought progress inremoving core heat through natural circulation would be lost.

Ultimately, theVice P-resident

Efforts to-Generation directed it be secured and it was.

reestablish condenser vacuum were then redoubled.

(Int.

16,

26, 31,

71,

91,

113; Ref. 21)
With the secondary heat sink isolated,

it was decided to increaseHPI

flow while at low pressure and attempt to collapse the *voids in the loops.
Core cooling was maintained by opening the EMOV and/or its block valve tostimulate flow by the fuel assemblies.

The Station: Manager was directed by

the Vice President - Generation to prepare for a briefing trip to the Lieutenant
Governor's office.

The Station Manager directed the Superintendent-

Technical

Support to gather the necessary information and prepare to accompany him.

(nt.

27,

71,

91)

In a further attempt to improve circulation and collapse loop voids, theemergency command team decided to maintain HPI flow, but alternate its injection
point to the RCS..

The EMOV block valve was cycied to control pressure andduring one opening (approximately 1350 hrs) a double "thump" was heard.
Reactorbuilding spray initiated, and ES actuation and reactor building isolation
occurred.

The'Station Manager,

then in the control room verifying plant

1-3-11status, reportedly exclaimed:"What was that?"

The connection between !thenoise, the start of reactor. building spray pumps and a hydrogen burn would not
be made until the following day when engineers reviewed plant data.

Thei noise

was attributed to shifting ventilation dampers, which were being worked on at
the time.

The start of the reactor building spray-pumps and the spike ifn thereactor building pressure were attributed to an electrical noise spike by one
supervisor.

Another supervisor observed the pressure spike and recommended to

the Station Manager that the EMOV not be cycled again, because he noted therapid rise in building pressure corresponded to the timing, of the opening ofthat valve.

(Int.

16,

26,,71,102, 111, 113,

83,,

The Vice President - Generation,

Station Manager and Unit 2'Superintendent

Technical Support left the site (approx.
Lieutenant Governor's office.

1.19)

1400 hours),and proceeded to thePrior to leaving, the Station Manager satisfied

himself that conditions were stable; ordered the Unit 1 Operations Supervisor
to attempt to maintain the status quo and to not reinitiate atmospheric steam
dump; appointed the Unit 2 Superintendent the acting Emergency Director; andcalled the Unit 1 Superintendent to~tell him where he was going and how to-reach him.

The Station Manager carried a beeper and had the Superintendent

Technical Support reestablish phone communication,
Lieutenant Governor's Office.

-once he had reached theDuring the absence of the'Station Manager,

theUnit 2 Shift Supervisor began to have some success in changing indicated A
loop hot and, cold leg temperatures and indicated pressurizer level,
HPI preferentially .to the A loop'.
on the B loop,

by directing

It was decided to attempt the same. maneuver'theexecution of which was without success and, all previous-

progress observed on the A loop was lost.

Increasing core flood tank overpres-sure was also discussed, but whether this action was ever taken has not been
determined.

No evidence or statements were obtained which indicate suchaction was taken.

(Int.

16,

26,"27,

31,

53,

71,

77, 83,91,

111, 113)

At the request of the Unit 1 Operations Supervisor, the Unit 1 Superintendent returned, to the Unit 2. control room from Unit 1, where he had been
since about 1000 hrs.

The Unit 1 Superintendent assisted the Unit 2'Superin-tendent in maintaining communications and joined the emergency command

1-3-12-'status, reportedly exclaimed:"What was that?"

The connection between thenoise, the start of reactor building spray pumps and a hydrogen burn would not
be made until the following day when engineers reviewed plant data.'

The noise

was attributed to shifting ventilIation damrpers., which were being worked on 'at
the time.

The, start of the reactor building spray pumps and the spike in thereactor building pressure were attributed to an electrical noise spike by one
supervisor.

Another supervisor observed the pressure spike and recommended to

the Station Manager that the EMOV not be cycled again, because he noted therapid rise in building pressure corresponded, to the timing, of the opening ofthat valve.

(Int.

16,

26,71,

83,, 102,

The Vice President -Generation,

111, 113,

Station Manager and. Unit 2"Superintendent

Technical Support left the site (approx.
Lieutenant Governor's office.

1.19)

1400 hours)'and proceeded to thePrior to leaving,

the Station Manager satisfied

himself that conditions were stable; ordered the Unit 1 Operations Supervisor
to attempt to maintain thestatus quo and to not reinitiate atmospheric steam
dump; appointed the Unit 2 Superintendent the acting Emergency Director; andcalled the Unit 1 Superintendent to tell him where he was going and how to-reach him.

The Station Manager carried a beeper and had the Superintendent

Technical Support reestablish phone communication,
Lieutenant Governors Office.

-once he had reached theDuring' the absence of the'Station Manager, theUnit 2 Shift Supervisor began to have some success in changing indicated A
loop hot and, cold leg temperatures and indicated pressurizer level,
HPI preferentially -to the A loop.

by directing

It was decided to attempt the same maneuver'on the B loop,, the-executi~on of which was without success and,
progress observed on the A loop was lost.

all previous-

Increasing core flood tank overpres-sure was also discussed, but whether this action was ever taken has not been
determined.

No evidence or statements were obtained which indicate suchaction was taken.

(Int.

16,

26,-27, 31,

53,

71,

77,

83, 91,At the request of the Unit 1 Operations Supervisor,

111, 113)

the.Unit 1 Superin-tendent returned, to the Unit 2 control room from Unit 1, where he had been
since about 1000 hrs.

The Unit 1 Superintendent assisted the Unit 2 Superin-tendent in maintaining communications and joined the emergency command

1-3-12

team that would periodically caucus during the period of absence of the Station
Manager.

By 1700 hrs, problems with.the station auxiliary boilers were tempo-

rarily solved and a vacuum was drawn in the condenser.

The A OTSG was thensteamed thru its turbine bypass valves to the condenser in an attempt to establish natural circulation. The A loop temperatures and pressurizer level again
appeared to respond to the preferential

injection of HPI to the A loop.

Theemergency command team believed'they were seeing the first signs of natural
circulation.

(Int.

16, 26', 27, 31,

77,

The Vice President - Generation,

113)

Station Manager and' Unit 2 Superintendent-

Technical Support returned (approximately 1630 hrs) from their trip to the Lieutenant Governor's office and were briefed-on current plant status and trends'.
The Vice President - Generation, Metropolitan Edison communicated this information to the Vice President-Generation,
them,

General Public Utilities,

it was decided the plant must be repres-surized.

and between

The emergency command

team recommended continuation of their 'current policy of attempting to establishing natural circulation, since they believed they were now making progressand earlier attempts to establish stable conditions at high-pressure were believed
unsuccessful.

The Vice President - Generation,

the Station Manager to repressurize.

Metropolitan Edison directed

The decision to repressurize was basedon a recognition that RCS water and metal temperatures were too high to allowpressure to be brought below the interlock for initiating DHR.
Without aneffective means to cool down the RCS, this goal of using'DHR was unreachable,leaving highpressure RCP forced circulation as the only viable option.

As

pressure was.increased using HPI with the EMOV block valve shut, it appeared'
the-loops were filling and the emergency command team then recommended that an
RCP be started once pressure
was stabilized.
I

Everyone,ly

including B&W,

appeared

to concur on this course of action.

26, 27,31,

83, 91,(Int. 16,44,

53, 71,

113,. 38)
RCS pressure was increased to about 2300 psig and maintained there bythrottling HPI flow at greater than or equal to 400 gpm.' B&W calculation results,,
now available to the emergency command team,. indicated this was sufficient flow
for the existing core decay heat'.
The history of.RCP operations was reviewed

1-3-13

with B&W and it was determined that RCP 1A should be bumped.

An RCP in A loop

was chosen.since it would provide the strongest pressurizer spray and previous
indications convinced them A loop had liquid in it.

The decision to initiallybump the pump for 10 seconds was precautionary and followed normal practices
utilized in RCS filling and venting procedures.

Before the pump could be bumped,a number of interlocks had to be bypassed, based largely on the unavailabilityof the a.c.

lube and lift

oil pumps, which had previously lost power-because,of electrical faults on their motor control centers.
run for 10 seconds.

(It.

16,

As the RCP was bumped,increased,

26,

27, 31,53,

At 1933 hrs,

RCP, IA was

71,,83, 91)

RCS pressure and temperature dropped, OTSG pressuresý

loop flow and RCP current acted normally.

as to the appropriate delay for restarting the RCP,

Discussions were helda normal'consideration

relative to motor overheating, and it was decided to delay 15 minutes and thenrun RCP IA,

again.

RCS pressure was increased to about 2200 psig and RCP IA

was started and left running at 1950 hrs.

RCS pressure again dropped,butstabilized at about 1300 psig; hot and cold leg temperatures converged andstabilized at about 340'F; A and B OTSG pressures stabilize at 94 and 50 psig,
respectively; the pressurizer level remained full scale.

Stable plant condi-

tions, as the emergency command team had defined them earlier, had finally beenreached.

(Int.

16, 26,

Subsequently,

27,

31,

53, 71,

83)

pressurizer level would be'regained and a plant cooldown on

A OTSG would occur.

Activities beyond 2000 hrs on March 28,

the scope of this report and, therefore,

1979 are beyond

are not covered here.-

A summary of notification and arrival times of selected individuals during
this accident isincluded on Table 1.3-1.

1-3-14TABLE 1.3-1
NOTIFICATION AND ARRIVAL TIMES (APPROXIMATE)

OF KEY PERSONNELAPPROX.
INDIVIDUALNOTIFICAT'ION(S)

Station Manager

TRAVELDISTANCE (MILES)

0401

ARRIVAL TIME

10

0705

(initial notification;see Section 3.1
for subsequentcontacts & actions)
Unit 2 Superintendent

0410,

50

0545

Unit I Superintendent

0545

15

0645

Maintenance.Superintendent

0515

10

0615

Superintendent

0410

10

0450

Unit 1 Operations Supervisor

0435,

20

0600

Chemistry/HP Supervisor

0510

15

0545

0600-0635

'15

0715

-

Tech. Support(Unit 2)Site Manager (NSSS Vendor)(Conference Call)
Instrument/Control Engineer'

06,00

10

0630

On-Call Operating Engineer

0501

15

0545

1-3-15

3.3

Provision of Technical Support

3.3.1

Onsite Technical Support
Basically four engineering disciplines.(mechanical,

and instrumentation) were represented during the event.

electrical,

nuclear

In addition to theplant staff, there were two individuals employed by B&W and assigned to theplant staff who were present in the control room.

The major discipline.

actively participating in operational related matters was instrumentation.
The-instrumentation engineer (Int.

121,

192) was involved in the.following

areas:oVerifying RCS temperatures and pressures by comparing meter indicationsin the reactor protection system (RPS) cabinets for two or. more
instrument channels.

He informed the Unit 2 Superintendent-TechnicalSupport that he coul'd see no reason not to believe the instruments.
0

Deploying personnel and equipment to measure hot leg RTD temperaturesand evaluating the results.

He could see no reason not to believe

the instruments and informed the Unit 2 Technical Superintendent of
this fact.
0

Deploying personnel and equipment to measure incore thermocouple
temperature and evaluating the results.
He informed the Station
Manager of temperatures greater than 2000'F in the core.

It should

be noted that during the measurement of the incore thermocouple
temperatures,

ostatements were made by the individuals taking themeasurements that they felt the core was uncovered.

They.statedthat they so informed the instrumentation engineer.

(Int.

181,

183)

Evaluating conditions associated with RCP interlocks to-allowa RCP-start.

He initially tried unsuccessfully to manually energi'ze theK-3 relay on an RCP.. He was later involved in jumpering interlocks

to allow a RCP start.

I--3-16

The Lead Mechanical and Electrical Engineers (Int. 70) were involved in
notification and communication as required by the Emergency Plan. ',There issome evidence that the Lead Electrical Engineer was involved in the RCP inte'rlock.jumper.
(Int. 193)
In addition, the Lead Electrical Engineer recommended
that the diesel, generators'

control switches-be placed in the maintenanceposition and the fuel rack be -reset '(Int. 195)
A nuclear engineer (Int:97) was involved in the performance of a shutdown

margin calculation that was performed shortly after the trip.
showed the reactor shutdown margin to be 8-1,0%.,the SRM and IRM showed an. anomalo'us response,
calculation.

This calculation

Later ,during the event, when

he was asked to verify the

He requested boron, samples and performed a calculation based ona boron concentration of 404.5 PPM.

This calculation showed the reactor was

shut down by -2.445% Ak.

This fact was reported to the Unit

1 Operations Superviso,r.

(Int.

26)

The Lead Nuclear Engineer (Int. 48) Was involved-in offsite releasecalculations as required by the Emergency Plan.
duties,After performing the above

he evaluated the following conditions and drew the indica'ted.

conclusions:oThe,'SRM and IRM showed anomalouslindication.

There was increased

radiation levels at the detectors causing the SRM and IRM response
and not a return to criticality.
0

The incore neutron detector responses,

as indicated by the backup

recorders, weregiving erroneous indications.
o

The T

hot

and T

temperatures.

cold

That Tcl

was as expected.

T

hot

was'high,.
There is no evidence that a staff nuclear engineer evaluated the incore
thermocouple readings during March 28,

1979.'-

1-3-17

The B&W Site Operations Manager was consulted by telephone (Sequence ofEvents; Ref. 53)

prior to coming on site.

During this consultation hequestioned whether the block valve on the EMOV relief line had been isolated.
He also participated in the decision to reestablish RCS flow (by the start ofa reactor coolant pump),

At approximately 0745 hrs, the Site Operations

Manager advised the B&W~corporate offices of the event (See Section 3.3.2).

(nt.

53)

(The B&W Site Operations Manager reported to the B&W Manager ofPlant Startup Services and as a result of a contractual agreement with
Metropolitan Edison Company reported to the Station Manager.

The B&W Site

Operations Manager received direction from the Metropolitan Edison Company inthe form of written communication.)
During the day, the Site Operations Manager was in contact'with a B&W
employee offsite who relayed information to and from the corporate offices.
(See also Section 3.3.2.)

During one of these calls a request was made forradiochemical expertise at the request of the licensee.

The personnel requested

did arrive from B&W in Lynchburg in the area in the afternoon of March 28,
1979,

but they did not enter the plant that day.

They were involved inorganizing mobile laboratory assistance offsite.
The Site Operations Manager participated in the following decisions:o

To increase pressure to collapse the steam void (he did not believe
the Th indications and he evidenced.some concern about increasing
the pressure to the point that the safety relief valves would lift.)o

To decrease RCS pressure and "float" the core flood tanks on'thecore.

The Site Operations Manager participated in the evaluation of HPI flow,
paths that would have resulted in the low pressurizer temperatures.
The B&W Physics Test Coordinator (Int. 16) was also present in thecontrol room after approximately 0900 hrs.
with the conclusions as shown:

1-3-18He evaluated the following areaso

The SRM and IRM indication anomaly.

The SRM and IRM anomaly wascaused by a change in neutron leakage sensed by the detectors.
oThe high Th indication.

The high Th temperature indicated super-heated steam in the hot legs and he adyised that the steam could not
be condensed by increasing system pressure.
o

Steam generator performance during natural circulation.
generators were not promoting natural circulation.

The steamHe recommended

that feedwater levels be increased to increase the steam generator
heat removal capability.
0

The incore thermocouple indications.

He recommended that the incore

thermocouples be monitored to indicate core conditions.
oIndications from the backup recorders for the incore neutron
detectors.

No information was derived from the backup recorders.

He stated that these recommendations and conclusions were discussed with
the B&W Site Operations Manager.
3.3.2

Babcock and Wilcox
At 0745 hrs,

the B&W-Site-Operations Manager contacted the B&W Manager ofPlant Startup Services .located in the Lynchburg, Virginia, Corporate Offices.
The following information was transmitted (Int.

185) as indicated by notes*

taken in the corporate'•offices:0

There'was a loss of feedwater caused by condensate polisher
isolation valve malfunction.

oThere was a turbine trip.

*These notes contain preliminary information and as a result may not be
entirely accurate.

1-3-19*oThere was a reactor trip on high pressure.

oThere was initiati.on of high pressure injection.

o

The pressurizer went solid.

0

RCS pressure went to approximately 2500 psi.

0

The reactor coolant drain tank rupture0

There was an indication of fuel failure.

o1

There was a reading of 800 R/hr at the dome of'the reactor bldg.

oThere was a loss of RCS flow indication.

,o

The reactor coolant pumps had been tripped.

oThere was an indication of a.primary-to-secondary leak.

oPresent RCS conditions were ToThe site was in a state of emergency.

cold:

After this information was received,
Lynchburg.

disc had blown.,

300'F, pressure: 1500 psig.

a task force was formed at B&W in

At 0900 hrs a task force information transmittal' meeting was held

in a classroom adjacent to the simulator at the B&W Training Center.

The

Manager of Plant Startup Services presented the information as received fromthe Site Operations Manager.
Notes taken during the day indicate the task force placed a high priority
on ensuring that the core remained covered and that it was being cooled

1-3-20adequately.
site.

During the meeting three engineers were selected to go to the

The personnel were selected on the basis of their knowledge of systems,

analysis capability, experience with similar events at Oconee and Davis-Besse,and,in some instances,having qualifications for entry onsite.

These personnel

arrived in the site area at about 1330 hrs, but they did not gain entrance
until after 2000 hrs, the period covered by this portion of the report.

A

list of data to be requested during the next contact with the site was developed.
An unsuccessful attempt was made to telecopy this request to the site.

(Int.

84, 85, 86)
At approximately 1030 hrs,

information was obtained from the Unit 2

Operations Supervisor, who was in Lynchburg for simulator training.

The

Operations Supervisor had been in contact with the site and via the Unit 1
control room had obtained the following additional information which was given
to the task force:oThere was a primary-to-secondary leak in the B steam generator.

oThere was 60,000 R/hr in the dome of the reactor building.

(This wasnoted' as doubtful).
0

There were 104 counts on radiation monitor 748,o*There was 100 mR/hr at the personnel hatch outside the reactor(200 is background).

building.
o"Iodine was high off-scale."oThe plant was in natural circulation cooldown.

o"Incore temperature was 450'FP (computer output).

oTcold was 250'F.

1-3-21oOn and offsite area radiation was negligible.

oThere was water in the control air lines of the condensate polisher
isolation valves, which caused the valves to shut and cause a lossof feedwater.

oThere was an ESFAS actuation.

o

The makeup pumps were stopped and letdown established (during RCS
high-pressure condition).

o

The EMOV block valve (RC-V2)

was shut after the quench tank rupture

disc blew.
oAuxiliary feedwater initiated, but there was no flow to the steam
generators until approximately 12 minutes after the trip.

(Laterinformation indicates 8 min.)oThe steam generators did not go dry.

o

RCS pressure went as low as 1200 psi; saturation conditions werepossibly reached.

o

The RCS flow decreased by approximately 1/3; RCPs were tripped.

The Unit 2 Operations Supervisor. (Int.

34) performed a rough calibration

based on radiation levels he had been informed of and estimated that 1/8 ofthe cladding had failed.

Attempts were made by him to duplicate the event on

the simulator based on the information he had obtained but he was unsuccessful.
At 1145 hrs, a message was relayed to the task froce from the B&W Site
Operations Manager via a B&W employee offsite.
conveyed:

1-3-22

The following information waso

The RCPs were shut down.

oThe plant personnel were trying to go solid; there were indicationsof steam bubble in B loop.

oLow-level radioactivity was reported in the atmosphere.

o'

The Metropolitan Edison Company had made a public announcement to
the news media.

oRadiation teams were performing surveys on and off site.

o

An NRC team had been sent to the site to investigate.

oThe primary-to-secondary leak in the B steam generator had been
confirmed by sample analysis.

o

The B steam generator was isolated.

oNatural circulation was being used to cool down the plant using the
A steam generator.

oSuction for HPI was being taken from the BWST.

o

The pressurizer heaters were shorted out.

oThe electromatic block valve was being used to control pressure.

oPresent RCS conditions were Tcold:3OO'F; pressure.2100 psig.

oThere were plans to cool down and depressurize.

oThere was speculation of fuel leakage but no further information onradiation levels.

1-3-23oThere was some level increase in the reactor building sump.

oComponent cooling water and seal injection had been maintained.

oThere was high moisture level in the reactor building.

oRadiochemical expertise was requested.

At approximately 1200 hrs,

the B&W task force was involved in determining

a recommended course of action for cases with and without RCPs.

An assignment

was also made-to- a-task force member to determine the-prerequisites

for-startinga reactor coolant pump.
At approximately 1330 hrs,

the following information was relayed from thesite by the B&W site Operations Manager to the B&W task force via the B&W
employee offsite:o

RCS pressure was 495 psig on core flood tank float.

o

The Thot was 700'F as measured by a digital voltmeter.

o

The pressurizer was full.

oPressure was being controlled by the electromatic relief block
valve.

o

The HPI suction was on the BWST in process of switching to the RC
bleed holdup tanks.

0

B steam generator was isolated; the level was approximately 60% of
the operating level.

oThere was minimum cooldown from the A steam generator.

o

The atmospheric dump valves were being used as the heat sink.

1-3-24o

The gland sealing steam was lost as the turbine-condenser vacuum was
broken.

oEmergency feedwater to the A steam generator was being supplied
through the main feedwater nozzles.

oThere was less than 1 mR/hr at the site security fence.

oRadioactivity was reported in the feedwater system.

oThere was airborne activity in the auxiliary building and control
room (reactor building sump was being pumped to the auxiliarybuilding sump before ito

was discovered and isolated).

Attempts had been made to start RCPs (one in each loop) but drew
only about 100 amps (no-load current).

o

It was not clear why the RCPs were originally stopped.

oThat the three people originally sent to the site from B&W had
arrived in the area.

o

The B&W Physics Test Coordinator was on site.

oIt was planned to go to decay heat removal system operation as soon
as possible.

o

The BWST level was last known at about 37 feet.

o

Arrangements had been made for another call from the Site Operations
Manager.

The following recommendations of the B&W task force were then made to be
relayed to the B&W Site Operations Manager via the B&W employee offsite:

1-3-25oObtain cooldown data.

0

Ensure accurate RCS temperature before going to decay heat removal
system operation.
Confirm core outlet temperature by pressurizer temperature sinceothis is now the'flow path.
The information received by the B&W task force regarding the planned
operation to go on the decay heat removal system, and the recommendation toensure accurate RCS temperature before going on to DHRS operation tends toindicate that there was an intent on the part of the licensee to go into DHRS
operation.

In addition, the recommendation regarding inferring the coreoutlet temperature by reading the pressurizer temperature indicates that thetask force had some knowledge about the flow of coolant in the reactor coolant
system at that time.
At approximately 1400 hrs,

the Manager of Plant Systems Design (Int. 87)

recommended establishing at least 400 gpm of HPI flow.

This value had been

provided by the ECCS Systems Manager based on the decay heat at that time andincluded 50% conservatism.
ways.

This recommendation was relayed to the site in two

The B&W Manager of Project Management (Int.

for GPU (Int. 90).

88) contacted-a vice presiden

The message was also communicated to the Unit 1 control roomby the Unit 2 Operations Supervisor for relaying to the Unit 2 Control Room.
At this time, the Manager of Project Management contacted the Manager ofGeneral Engineering at Metropolitan Edison and requested that a directcommunication link be established between the Manager of Nuclear Services andthe site.
During the time period 1400 hrs to 1600 hrs, additional information wasprovided by the Unit 2 Operations Superintendent based on contacts with thesite:

1-3-26o

The EMOV had stuck open (resulting in the rupture of the reactor
coolant drain tank rupture disc).

oThere had been a delay in emergency feedwater because of incorrect
indication (pumps running - valves indicated open but were not).

o

The RCPs were stopped because of an indication of cavitation(indicated drop in RCS flow); condenser offgas alarm first indicatednear the steam generator.

0

The 60,000 R/hr reactor building dome reading was bad because ofmoisture (pegged high).

oThe best estimate of reactor building radiation level was approximately 100 R/hr based on 100 mR/hr at the personnel hatch.

o

The reactor building pressure went initially to approximately 2.5
psig and then to approximately 4 psig about 6 hours into thetransient.

Sometime prior to 1600 hrs, the individual assigned the task of making
recommendations regarding the restart of the reactor coolant pump reported to
the Task Force that:o

There was a concern about moisture being a problem in starting the
RCPs (motors).

o

The electrical circuit couldn't be meggered (checked for faulted
condition).

It was suggested that the component cooling water (Int.

closed cooling water) leak alarm (moisture detector) could bechecked.
oUp to 30 mils vibration could be tolerated for a short period oftime.

In addition, a reactor coolant pump associated with the sprayline should be chosen for operation if possible.

1-3-27oThere is a possibility that the reactor coolant pumps were notconnected (sheared drive pin) because no-load amperage should beapproximately 100 amps, which is the indication received on earlier
attempts to restart, if

connected, RCP starting current should begreater than 100 amps.
At approximately 1600 hrs, it was reported from the B&W Site Operations
Manager to the task force via the B&W employee offsite that:o

It was difficult to establish direct communications with the site.

0

The "bubble" had been collapsed in the "A" leg -- good indication*from temperatures and pressures.

oThere was normal letdown and pressurizer electromatic isolation
valve pressure control.

oThere was an attempt to collapse the bubble in the B loop.

0

The pressurizer level was 190 inches; pressure: 560.psig; temperature: 460'F.

oPlans were to get on DHR via the suction from the BWST.

o

The core flood tanks were still "floating."

Pressure: 400-500 psig;

temperature: 540'F.
o,:

RCP-2A had indications of leakage but unsure.

0

The cooldown data was available on the reactimeter tapes.

0

The level in the A steam generator'was at 80% of the operatinglevel.

1-3-28o

The level in the B steam generator was at 50% of the operatinglevel.

The following recommendations were made to the Site Operations Manager
via the B&W employee offsite by the task force during this telephone call:o

Establish 400 to 500 gpm of HPI flow.

o

The minimum recommendations for RCP start:-

30 mils vibration limit.

-

HPI injection providing greater than saturated conditions forthe loop.

-"Good" amperage reading.

-Component cooling water (int. cooling water) flow.

-

Normal prerequisites for start.

-

No steam flashing at the seals.

During the call,a request from the licensee was relayed to the task

force by the B&W employee offsite concerning what were the considerations forrunning an RCP (at least one) in the B loop (decay heat drop line side)..
At 1640 hrs,

information originating in the Unit 2 control room wasreceived that the Thot temperature was 550OF and the pressure was 450 psig
(superheated conditions).
In the period from about 1630 hrs to about 1700 hrs, the B&W Manager ofProject Management again emphasized to the GPU Vice President that HPI flow cf
400-500 gpm should be initiated.

The GPU Vice President advised that HPI flow

had been initated.

1-3-29

At 1745 hrs, the following information was relayed from the Site
Operations Manager via the B&W employee offsite:oA vacuum was being established in the condenser.

oHigh pressure injection flow was 100 gpm.

oReactor pressure increased from 560 to 670 psig.

o

The core flood tanks were no longer "floating" on the RCS.

oThere was a bubble in the pressurizer.

oA "bubble" was suspected in the A loop.

o

The Tcold was 520*F and Thot was 53 0 'F.

0The A steam generator was at 180 psi,and the level was at 80% of

the operating range.
o

The HPI was taking suction from the BWST.

During this call,it was recommended that the HPI be immediately increased,

letdown be terminated, and subcooled conditions be attained in the RCS.
At 1810 hrs, a telephone call was received from a vice president of GPU
indicating that, since 1620 hrs,

HPI had been maintained at 400 ppm.

At 1835 hrs, a telephone call was received directly from the Site
Operations Manager in the Unit 2 Control Room.
0

He advised that:

Cooling was from the A steam generator via the dump valve to thecondenser.

1-3-30o

The A loop conditions were Tcold: 300'F; Thot: 540 to 550'F.

0

The RCS was going "solid."oPressurizer level was stable at a temperature of approximately
500 0 F.

o

The pressurizer heater capacity was limited because of shorts.

oThere were "bubble" indications in the B loop with Tcold at 200 andThot was off scale high (>650 0 F).

o

RCS pressure wasf1800 psig-and increasingo

HPI flow was 400 gpm.

o

A steam generator pressure was 50 psig (consistent with a Tcold of
3000F).

oA steam generator Tcold was decreasing with timeo

The A steam generator Thot was consistent with pressurizer pressure.

oThere were two incore thermocouple readings higher than 500'F.

The'

computer was printing out "?????" for the others.
oSeal injection had been maintained since 0700 hrs, and before thetransient.

o

The RCPs had good bearing temperatures and seal pressures.

oThere was no indication that the ESFAS had isolated the seal
injection to the RCPs.

I-3-31oThere was 100 R/hr in the auxiliary building (reactor building sump
automatically pumped to the auxiliary building sump; secured later).

oThere was 100 mR/hr in the plant.

oThere was 70 mR/hr outside the gate.

oThere was more than 1 mR/hr on State routes near the site.

oMoisture and high radiation levels were limiting the use of someequipment.

oThere was an indication of a-primary-to-secondary leak from B steam
generator by sampling,

but the magnitude of the leak was not known.

o

The B steam generator was isolated before 0700 hrs.

o

The electrical buses supplying the oil lift

pumps needed for RCP

start were lost.
At 1900 hrs,

the following information was received during the same

telephone call:o

The Thot was 560'F.

o

The oil lift

pumps were now running.

At about this time, the B&W staff member assigned the task of making
recommendations regarding the start of an RCP provided the following
information to the task force:oSeal injection was to be maintained at 12 gpm.

1-3-32o

The maximum seal.return flow was to be less than 1.9 gpm.

oThere should be a supply of nuclear service water.

oThere was no alarm on the component coolant water (CCW)o

The starting current should be greater than 600 amps.

o

The vibration limit was 30 mils (peak to peak).

oThere should be clear motor permissive start interlocks.

oThe seal injection temperature should be less than 150'F.

o

The seal return temperature should be less than 185'F.

leak alarm.

During this same call it was reported that:oThe 2A and 2B RCPs were the pumps used earlier in attempted restart.

0

The A steam generator pressure was 50 psig, and the level was 80% on
the operating level and decreasing.

o

The B steam generator was at 70% level.

The recommendation was made to the Site Operations Manager by the task
force that the RCP-lA be given a five-second start ("bump")

and then that it

be stopped to let the RCS parameters stabilize.
At 1930 hrs,

it was reported to the task force by the Site Operations

Manager that the RCP-lA was given a 10 second "bump"
In addition,

start/stop by the operators.

the steam generators were at 200 psi and the A steam generator

steaming rate increased.

It was reported that the RTD indications wereconsistent with conditions.

1-3-33

At 1940 hrs, the following information was received during the same
telephone call:o

The A loop Thot was less than 520"F and the Tcold was 320'F.

The A

steam generator pressure was 60 psig, and temperature was 325 0 F.
o

The B loop Thot was 620'F and Tcold was 225 0 F.

*

pressure was 160 psig and temperature was 390'F.

o

The RCS pressure was 1850 psig and it had dropped to 1400 psigo

RCS flow went to 40%.

o

The RCP starting current was 1200 amps (normal).

The B steam generator

At this time the task force evaluated conditions for a second "bump"
the RCP or for a start with continued running of the RCP.

of

The recommendation

was made by the task force to the Site Operations Manager to start and run the
1A RCP.

However, the Station Superintendent had already ordered that this bedone.
At 1950 hrs,

it was reported by the Site Operations Manager during thesame telephone call that the 1A RCP had started and was running.
At 1955 hrs,

it was reported by the Site Operations Manager during thesame telephone call that the plant conditions were as follows:oRCP-lA vibration was 18 mils.

oBoth loops had Thoto

The steam generators were at 100 psi.

1-3-34o

The plant was stable.

o

The RCS pressure was 1800 psig.

o

The RCP current was 620 to 580 amps.

o

The RCS flow was 40%.

o

The pressurizer temperature was 500'F.

o

The incore thermocouples indicated approximately 400'F.

The Site Operations Manager stated that it was planned to stay on one RCP
and cool down to get DHR in operation, which was expected in a few hours.
In addition, the Site Operations Manager was asked by the Manager ofPlant Startup Services to call when DHR was in operation or if there were anyother problems or assistance was needed.

The Site Operations Manager statedno further assistance was needed at this time and the Task Force was adjourned.
3.3.3

Corporate Technical Staff
The licensee Emergency Plan requires that notification be made to Division

Headquarters,in particular, the Metropolitan-Edison Vice President - Generation,in the event of a Site or General Emergency.
Corporation isContact with GPU Service

implied as required but is not stated explicitly.

Notification

is based on the requirements stated in Emergency Procedure 1670.2,

Revision 9,

Figure 4, and in Administrative Procedure 1014, Recall or Standby Personnel to
Plant, Revision 3.
The Emergency Plan also requires, as needed, Metropolitan
Edison, Division of Engineering support, and Metropolitan Edison,
Corporation engineering/technicalservices.

The Emergency Plan is silent on

specific requirements to be provided by the Metropolitan Edison,Engineering, and GPUSC engineering/technicalI-3-35GPU Service

services.

Division of

The first call to Metropolitan Edison corporate management was between
appoximately 0600 hrs and 0630* hrs.

The Metropolitan Edison Vice President -Generation was called again at 0730 hrs to be informed of the declaration of aGeneral Emergency but this call was not completed.

The Metropolitan Edison

Vice President - Generation was told to leave Philadelphia by helicopter andreport to Three Mile Island.

He left by helicopter from Philadelphia at approxi-

mately 1100 hrs, arrived at the TMI Observation Center at about 1140 hrs andestablished phone contact with the Station Manager who was located in the
Unit 2 control room, the Emergency Control Center.

The Metropolitan Edison

Vice President -Generation had become aware sometime between 0830 hrs and 1100
hrs that a general emergency had been declared.

(Int. 91)

Emergency assistance was not requested from the Metropolitan Edisonheadquarters staff for direct assistance to operations for the first 16 hours.
The Metropolitan Edison Vice President - Generation, provided consultation,and through joint agreement with the Site Emergency Director, decisions weremade on operational evaluations until 2000 hrs.

(Int.

91)

The Metropolitan Edison Manager of Generation, (Operation)
about 0700 hrs.
(Engineering)

was called at

The call was taken by Metropolitan Edison Manager of Generation

because the Manager of Generation (Operation) was out of town.

The Manager of Generation (Engineering) was informed that a Site Emergency had-been declared for TMI-2.

A second call at approximately 0730 hrs informed theManager of Generation (Engineering) that a General Emergency had been declared
(Int.

159)
The involvement-of the Metropolitan Ed-ison engineering corporate staff

was of a standby nature.

There was no technical feedback to operational

activities nor were decisions being made by the Metropolitan Edison corporate
staff for transmittal to the TMI-2 control room.
(Engineering)

The Manager of Generation

was advised that Unit 2 had a loss of feedwater and a turbine

trip followed by a reactor trip and that the levels of a General Emergency had

Times may not be in agreement with the time as stated in the referenced
interviews but are based on data acquired from various sources and represents
the most probable hour.
1-3-36been reached.

The President of Metropolitan Edison was apprised of thesituation at approximately 0755 hrs.

Contact with the Station Superintendent

at approximately 0940 hrs resulted in additional information being given to
the corporate staff.

(Int.

159)

The Metropolitan Edison Manager of Generation (Engineering) and Staff wascontacted by B&W and was advised that three B&W personnel were being sent to
TMI to provide assistance.

He was also called by the Metropolitan Edison Vice

President - Generation at about 1200 hrs and given a status of Unit 2 that was
similar to the information given to him by the Station Manager.

Specific datawith regard to temperature and pressures were not transmitted, but generalevaluations of the plant were transmitted to the corporate staff.

The support

provided by the Metropolitan Edison corporate staff from late afternoon andinto the evening was that of assisting communications service personnel andanswering calls from the public and the press.

(Int. 159)

The General Public Utility Service Corporation (GPUSC)

was contacted by

the Metropolitan Edison Manager, Operational Quality Assurance at approximately 0759 hrs.

Contact was made by phone to the Vice President.-

GPU Service Corporation.

Generation,

He was informed that a Site Emergency existed at TMI

and that TMI-2 had undergone a turbine and reactor trip, the steam generator
appeared to have a primary to secondary leak, and.there were increased radiationlevels in the reactor building.

(Int.

90)

The Emergency Plan does not identify specific responsibilities on thepart of GPU Service Corporation, but it

is implied that engineering/technicalservices would be available as necessary.

There are no dedicated communication

links between the GPU Service Corporation office and the TMI site.
There was contact with the control room on behalf of B&W to communicate.
with the B&W Site Manager.
Station Manager.
time,

Contact was made with the B&W Site Manager and theThis phone contact was made at about 1000-1030 hr.

At thisthe Vice President - Generation, GPU Service Corporation, did not knowthat a general emergency had been declared.

(Int.

90).

Later (about 1145 hrs),

he became aware that there were indications of offsite releases.

1-3-37

The GPU

Service Corporation staff had general information about the status of theplant, but they did not have specific hard data.
included the following:

The information they had

turbine trip/reactor trip,, instrument air problem,

condensate pump trip, RCP trip, OTSG possible leak.

(Int.

165)

This infor-

mation was known and discussed at a meeting held around 0900-1000 hrs at theNotes from a GPU Service Corporation

GPU Service Corporation headquarters.

staff member indicates that at a metting held at 0915 hrs they also weretold that the condensate polisher isolated, there was low reactor coolant
pressure,

there was high pressure injection, the reactor coolant was solid,

reactor coolant at 350OF and containment dome radiation alarmed.

The GPU

Service Corporation management decided to dispatch five of their technicalstaff to the site to provide assistance related to the information they had
available.
The GPU Service Corporation engineering group arrived, by automobile,the observation center over a period from 1400 to 1730 hrs.

atThey were given abriefing by the Superintendent - Technical Support at the Observation Centerat approximately 1805 hrs.

(Int.

176)

The dispatched group did not provide

assistance or make decisions with respect to plant operation during the first
16 hours of the event.

One individual of the group did enter the TMI-2 Control

Room at about 1900 hrs to obtain data so that the GPU Service Corporationgroup could have specific information to allow an evaluation of the event.

He

was in the control room for about four hours.
The GPU Service Corporation management personnel that remained at their
Headquarters were involved in the recommendation to increase reactor coolant
makeup flow, take the plant to 2000-2300 psig and to restart a reactor coolant
pump.

This was around 1630 hrs.

(Int. 90)

The decision to recommend thisaction was transmitted to the Metropolitan Edison Vice President - Generation.
3.3.4

Burns and Roe (Architect Engineer)
The investigator made contact by telephone with a Burns and Roe (B&R)

site supervisor with respect to his or Burns and Roe's involvement with theevent of March 28, 1979,

at TMI-2.

He stated that he made contact through a

1-3-38

representative of Met Ed Quality Control to inform the Unit 2 Superintendentof B&R willingness to provide assistance if

it were deemed necessary.

contact was at about 0730 hrs at the TMI observation center.

InitialB&R was not

requested nor did they provide operational assistance during the first 16
hours of the March 28,
3.4

1979, accident at TMI-2.

Licensee/NRC Interface

3.4.1

Notification of NRC
A Site Emergency was declared by the Unit 2 shift supervisor at about0656 hrs.

At 0704 hrs, the first call from the site was received by the NRC

Region I answering service.

(Ref.

3,

54)

The answering service first tried

to reach the Region I Duty Officer at his home,
I Office.
beeper,but he had left for the Region

The answering service then tried to signal the Duty Officer's

starting at 0720 hrs,

but the signal was not received until 0738 hrs,

at which time the Duty Officer had almost reached the Region I Office and hedecided to proceed to the office to answer the call.
Meanwhile,
Station Manager.

a General Emergency had been declared at 0724 hrs
(Ref.

2)

by theThe site called the Region I Office regarding theGeneral Emergency at 0740 hrs.

(Ref.

3)

At 0745 hrs, the Region I telephone

operator took over for the answering service and notified the Reactor Operations
and Nuclear Support Branch Chief of the General Emergency.

The Region I

Director was informed of the General Emergency about 5 minutes later andordered activation of the Regional Emergency Center at about 0800 hrs.
3.4.2

Communications Between NRC and the Licensee
Communications were established between the Region.I Incident Response

Center (IRC)and the licensee and between the IRC and the NRC Headquarters

Office of Inspection and Enforcement by 0810 hrs.,

(Ref.

4)

By 0839 hrs, a permanent line had been established between the Operations
Center (OC)in NRC headquarters and the Region I IRC,

1-3-39and itremained open forthe rest of the day (IRACT
the site.

tapes).

The IRC relayed information between OC andThis procedure as well as the time spent in clarifying information

resulted in considerable delay (one example was about 15 minutes) in transmitting information between the OC and the site.

At 1015 hrs,

Unit 2 control

room was evacuated and communications were interrupted until contact could bemade with the Unit 1 control room.

Communications gradually improved as theday progressed.
At about 1030 hrs, contact was established with Region I personnel who
would obtain information and provide it directly to NRC (IRACT tapes).

Later(prior to 0130 hrs), a conference call was established between the HQ OC,
Region I IRC,

and the site.

At times (for example about 1100 hrs),

thecommunications

were hampered by a high noise level and the necessity of using respirators.
About 1500 hrs,

a phone contact was established with the Unit 1 Shift Supervisor's

office, which remained open for the rest of the incident.

(TMI phone bill;Ref.3 )
3.4.'3

Impact of NRC on Licensee Actions
Prior to the arrival on site of NRC Region I personnel,

the NRC's principal

contacts with the licensee were for the purpose of obtaining information.

The

first team of Region I inspectors arrived on site shortly after 1000 hrs.

Twoof them went to the Unit 2 control room (Int. 31,
the control about 1100/1130 hrs.
and relay it

62),

the first arriving in

Their main function was to gather information

to NRC through the direct phone contact that had been established.

They'reported that there were around 20 to 25 people in the control room andthat, for the most part, actions taken were calm and deliberate.

They dealt

mainly with a -relatively smial] -group of p-eop]& in-the middle of the control
room.

These inspectors spent considerable time getting temperature readings

as these were of considerable interest to NRC.

Some were taken from instruments,

but most were obtained directly from the computer, which was manipulated forthis purpose by NRC personnel.

They reported that they were asked by licensee

personnel for their ideas and they made comments in response,
recommendations were made.

but no significant

One inspector recalls the recommendation being

made that the licensee think hard before putting the decay heat removal system

1-3-40into operation.

At one point an inspector pointed out to the Station Manager

that the steam being released to the atmosphere might be contaminated.
As the day progressed NRC headquarters expressed concern over various
areas.

These concerns were made through the inspectors or directly to licensee

representatives.

The following examples were derived from the tape recordings

made of phone conversations in the NRC:HQ Operators Center (OC):o

At 1215 hrs NRC Headquarters requested that licensee be asked ifthey considered blowing down the primary system,. and if
bumping the RCP's.

It

they consideredappeared that this is the first instance of

NRC Headquarters asking questions of a planning nature rathern thana status nature.
o

At 1317 hrs NRC expressed concern that leaving core flood tankisolation valves open may lead to injecting N2 into vessel.
concern was reiterated at approximately 9 hrs. 30 min.
cases,

This

In bothit was responded to by indicating the licensee's conclusionthat N2 injection was not possible, based on system design and plantconditions.
o

At 1400 hrs NRC Headquarters expressed their concern that continued
HPI injection would prevent discharge of the core flood tanks, withthe result that the system will remain above the DHR interlock for along period.

OC asks how licensee is going to get the pressure down

so that DHR can be put into use.
o

At 1559 hrs NRC Headquarters asked the Unit 1 Shift Supervisor(over the telephone) to communicate the NRC concern that the pressurizer level indication did not preclude a bubble in the core, andthat the temperature readings indicating superheat may have beenreal and imply that the core was uncovered.

They would then need tofind a way to put more water into core and to get the core levelback up.

The question was asked if

1-3-41the licensee had talked to B&W,

because B&W had been trying to get in touch with the licensee andhad the same concern.
At 1615 hrs NRC Headquarters asked if the licensee had consideredoblowing the system down,

and the supervisor responded that it

been discussed and rejected.

NRC:HQ requested that it be consideredagain, stressing this was a request to consider it,
do so.

had

not an order toApproximately 5 minutes later the supervisor returned stating

the licensee thought it was a good idea and decided to do it.
NRC:HQ then relayed their concern about valving out the CFT's before
the blowdown to prevent nitrogen from getting into the vessel.
At 1845 hrs NRC Headquarters asked the inspector onsite if.the licensee0

had considered running an RCP.

Inspector stated that those prepara-tions were underway but that they were having trouble with the oil
liftpumps.

NRC:HQ asked if the licensee had-considered that theymight have a gas bubble rather than a steam bubble in the "B" loop.
They were concerned that it could be nitrogen, xenon, or hydrogen.
.Inspector agreed to pass the concern along. This concern apparently
developed from evidence that only the A leg bubble quenched while
the B loop remained superheated.

Shortly thereafter, the inspector

reported back that there was nothing licensee could do-about it.
There was no way to vent that leg, and the only way to seep it out,
whether it be steam or gas,

would be to use the pumps.

Evaluation
NRC's initial impact on the licensee was to monopolize the attention ofcertain key licensee technical personnel in providing notification and communications.

Licensee personnel involved in maintaining these communications

were faced with difficult-problems of nomenclature and spent much of theirtime in training NRC personnel on plant systems.

As time progressed, thenature of the communications shifted from training to information request andtransmittal,

and finally to information exchange.

. 1-3-42

NRC operation inspectors arrived in the Unit 2 Control Room too late tohave an effect on the events that led to the accident.
Subsequently, theirtime was occupied wit.h gathering information and communicating with the NRC.
The inspectors occasionally made suggestions but it appears their presence had
little impact on the decisions made or' actions taken by the licensee.
NRC management began to make recommendations to the licensee after 1200
hrs.

These recommendations were directly at three areas; blowdown of the RCS,

operation of the core flood tanks,
recommendations,and HPI usage.

At the time of the blowdown

the licensee had already initiated blowdown in an attempt tocause Core Flood Tank injection and ultimately to go on the decay heat removal
system.

There is no evidence that the suggestions pertaining to the coreflood tanks were followed.
HPI and repressurize the RCS.

Licensee management did finally decide to increaseIt

is doubtful the NRC recommendation affected

their decision.

1-3-43

4.

4.1
4.1.1

SIGNIFICANT EVENTS DURING ACCIDENT

Turbine Trip
Apparent Cause
The cause of the initial trip of the steam generator feed pumps(FW-P-1A/1B)

that initiated the turbine trip followed by a reactor trip has

been attributed by some members of the plant staff to water in the instrument
air system.

It

has been stated that water in the instrument air system at thecondensate polishing system will cause the condensate polisher air-operated
effluent valves M12, M22,etc.,

to go closed (Int.

102, 123).

While a turbine

trip is an anticipated operational event that occurs relatively frequently
during the operation of a plant, the investigation studied the possible causesof the turbine trip to determine if

a precise initiating cause could bedetermined, thereby eliminating the need for conjecture.
4.1.2

Plant Air Systems
A partial review of the history of the instrument air service air system

was conducted during this investigation.
A licensee concern as to the capacity of the air system was recognized
early in the construction/preoperational

phase of TMI-2.

1-4-1(Problem Report No.

913; Ref. 18)

The solution of the capacity problem was cross-connecting thestation service air system to the instrument air system as a normal mode ofoperating the two systems.

Discussions with licensee personnel indicate that

there is a change pending that would isolate part of the station service air
system.

This change and its status were not pursued for details.

air supply operation on March 28,

The mode of

1979, was the cross connected system.

The investigator reviewed Instrument and Control Log Book I,(designatedas "Mercury" Daily Log), which indicates that there was water in the air
system at the condensate polishers on October 19,

1977.

While the log called

for attempts to determine the cause, discussions with licensee personnel
knowledgeable of this occurrence indicate that the cause was not determined.
The air lines were cleared of water, and the system functioned correctly.
The investigator noted that reference is made in the I&C Shop Log, May
12,

1978, that condensate polisher air lines were again filled with water, ata time when operations staff members were working on the condensate polishingsystem.
The licensee installed local air dryers at various points in the instrument air system to prevent the accumulation of moisture.

In particular,anair/water separator was installed in the condensate polisher instrument airline in series with two pressure regulators.

This arrangement processed all

air to the condensate valve controls and instruments located on the condensate
polisher local control panel.

This water/air separator utilizes a buoyant

valve plug that rises and allows water to exit the line while keeping theair
above the water.

It also contains a fine mesh screen.

Air lines at thecontrol panel are 1/4 and 3/8 inch diameter copper tubing.

Observation ofthis separator in its current status shows an accumulation of iron-rust-coloredparticles at the screen location of the air/water separator.
4.1.3

Post Accident Testing
The licensee has performed tests on the condensate polisher instrument

air system subsequent to March 28,

1979.

Discussions with the licensee staff

1-4-2

working on these tests indicate that, on isolating instrument air from thecondensate system, the condensate outlet valves for each polisher tank go
closed.
Their tests indicated that introduction of water into the air system did
not affect the polisher outlet valves,properly.

in that the air-water separator functioned

They proceeded to change the test procedure and they removed theair/water separator.

In that test, on introduction of water, the outlet

valves fluttered and then stabilized open.

In all testing cases involving

water, the polisher inlet valves remained open.

The testing with water consistedof allowing about 15-gallons of water to enter, with the air, at the condensate
polisher control cabinet.
The licensee, during testing, inspected the service station air supply
connected to the condensate system.
RO-I 51)

He found that the check valve (Tag No.

in the I inch station service air line, approximately 115 feet fromcondensate resin tank No.

7, was "frozen" open.

The shutoff valve in thatline is open when resin transfer is in progress.

The open shutoff valve andopen check valve allowed a path for water to enter the station service
air-instrument air system on March 28,
Ref.

1979 (Procedure 2106-2.2,

Revision 9;).

The licensee report on their testing and investigation effort has notbeen completed.
The actual circumstances described at the time of the trip (Int.

102,143) indicate that significant amounts of water were found in the station
service and instrument air receivers.

The No.

7 condensate polisher was being"flushed" with water having an exit pressure of 160 psig into the condensate
polisher No.

7 tank.

The 160 psig water and the service air, at 80 to 100

psig, were being used simultaneously.

Discussion with the auxiliary operator

who was working on the condensate polisher tank No.

7 on March 28, 1979,

revealed that the air was being used to "fluff" the resin while introducing
high-pressure demineralized water to transfer the resin.

1-4-3Discussion and taped interview (Int. 123) indicate that at the announcement of the turbine trip/reactor trip, the condensate polisher panel indicators
showed condensate polisher isolation, which indicates no condensate flow.
Condensate flow charts have been reviewed by the investigator and confirm an
abrupt termination of flow (Flow Charts for Condensate Polisher Tanks 1, 2, 3,
4,

5, and 8; as well as Polisher Outflow Chart; Ref. 19).

4.1.3

Pump Interlocks and Wiring Errors
Discussion with a licensee engineer indicates that he also found that thesolenoid switch wiring for the polisher valve controls was not in accordancewith drawings in at least two polisher units.
the valves on power failure.

This could affect the status of

The details on this aspect are not available atthis time, but should be available in his pending report.

(nt.

He also stated166) that there was a wiring error related to the condensate/condensate

booster pump auto/manual switch such that on a trip of condensate booster pump
CO-P-2A,

its paired condensate pump CO-P-1A would trip.

This wiring error was

isolated to the A pump pair so that the condensate B pump would remain on line
when its paired booster pump tripped.
Once the condensate polishing system fails to allow condensate water toflow, the condensate booster pumps will trip on low suction pressure andcondensate pump 1A will trip because of the wiring error.

Once the booster

pump trips, the main feedwater pumps will trip on low feedwater pump suction
pressure.

The loss of both feedwater pumps trips the turbine.

4.1.4

Evaluation
The tests performed to date do not demonstrate conclusively that condensate

polisher valve closure will not occur with water in the condensate polisher
valve instrument air lines as reported by the interviewed operators.
To validate the initiating cause of the turbine/reactor trip as thepresence of water in the air system producing condensate polisher flow isolation,

1-4-4

a confirmatory test would be needed.

The test performed by the licensee casts

some doubt on the apparent initiating cause, but the investigators consider
the testing performed thus far to be insufficient to be conclusive.
This investigation was not able to identify the specific initiating cause
which led to the turbine trip on March 28,

1979.

interest for reasons stated in Section 4.1.1.

The initiating cause is of

However,failing to identify a

specific cause is not a concern in that this plant was designed to sustain aturbine trip irritiated by a-loss of main feedwater as a routine operational
event.

The availability of the condensate system is not critical to thesequence of events in this accident.
4.2

Closed Emergency Feedwater Block Valves

4.2.1

Sequence
At approximately 8 minutes after the start of the accident, the operator

found the OTSG level at 10 inches on the startup range.
this level to indicate that the OTSG was "dry."The operator considered(This level indication isappropriate for a steam filled OTSG with no liquid phase present.)
verified that the emergency feedwater (EFW)The operator

pumps were running and examinedthe valve lineup, and found that the EF-Vl2A and 12B valves were shut.

The

position-indicating lights on 12A were obscured by a caution tag hangling fromanother valve controller.

The position indicating lights for 12B may havebeen obscured by operator as he leaned over the panel.

(This was due to thelight positions on the panel and the operator's location,
stature.)The operator drove the valves open, resulting in the dry OTSG being

fed with relatively cool water.
Itnot his physical(Int. 4, 9)

should be noted that on March 23,

Branch (OLB)

1979, an NRR Operator Licensing

examiner conducted operator examinations which included the EFW

System.

The valves were open on that day based on control room indications.

Moreover,a routine scheduled surveillance test was performed on the A and B

1-4-5

electric-motor-driven emergency feedwater pumps on March 26,
mately 1000 hrs,
operator.

1979 at approxi-

by the training shift control room operator and auxiliary

Implementation of this surveillance test procedure results inclosure of both the 12A and 12B valves regardless of which pump is being
tested.

The procedure calls for reopening of the valves along with ensuring

the proper status of at least three other valves.

The procedure for theelectric-motor-driven emergency feedwater pumps is not sufficiently specificto provide documentation of valve opening.

Instead, the procedural require-ment is in sentence form with one signoff signifying proper positioning of thevalves.

The documentation which is retained contains the completed data

sheets, but does not include the check lists which document the appropriate
steps to return the valves to their proper position.
While the surveillance procedure is deficient as to the manner in which
the test is done, and the documentation which is retained by the licensee,

thecontrol room operator who was on duty in the control room when the March 26,
1979 surveillance was performed stated during interviews that he left the 12A
and 12B valves open at the end of the test (Int.

184) For additional details

see Section 1.3.2 of this report.
4.2.2

Cause
The investigation attempted to ascertain the circumstances under whichthese valves were closed.
oThe following possible reasons were considered:

The valves were left closed after the last surveillance test of theemergency, feedwater system.
o

The valves were closed by the overt act of an individual.

oThe valves were left closed after maintenance work on the system.

o

The valves malfunctioned as a result of an improper design change orplant modification.

1-4-6

o

The-valves-ma-functioned because they were expos-ed-to elevatedtemperatures prior to or during the accident.
o

The valves were closed as an operator action prior to or during thetransient.
Review of all these possible causes revealed no reason to believe thatany of them was the specific cause of the closed valves.
The findings are
summarized below.
The operators and supervisors responsible for conducting the surveillance
test on March 26,

1979 were interviewed.

(Int.

107,

136)

The operator who

actually manipulated the valves involved stated that he specifically recalled
opening that valve.

The investigation found no basis for rejecting his assertion.

If his assertion was incorrect and the valves were left closed after the test,
the investigation found no information to explain how the closed valves would
have gone unnoticed during the 42 hours between the test and the accident.
However,licensee.

routine panel inspections are not required of the staff by thisThe matter of inspections of safety related matters is addressed in

Section 1-1.3.5 of this report.
The possibility that the valves were closed by an overt actof anindividual was addressed in Section 1.4.6.

No information was developed

during this investigation indicating that sabotage was a contributing factor
to the initiation of the accident or to the subsequent response-of plant
personnel or equipment to the accident.
The possibility of maintenance work being done on the valves wasaddressed in Section 1.6.2.

No evidence was found of such maintenance after

both record reviews and interviews.
The possibility that the valves were closed as the result of an improper
response of the valve control circuits to the turbine trip was addressed.

Achange was made in the logic circuitry related to the operation of the emergency
feedwater valves(FCR 2329.1; Ref. 9).

The change included defeating the

1-4-7

automatic closure of the emergency valves EF-VI2A/B with a low OSTG pressure
signal.

This feature had been part of the protective circuitry involved withthe plant response to a steam break accident.

If the modification required by

FCR 2329.1 had not been properly performed there would be a possibility of thevalves closing.

Since the accident, the licensee has written and performed atest to determine if the closure demand feature had been removed from theEP-12VA/B valves (SOP No.

R-2-79-31,

June 29,

1979; Ref.

12).

The results

indicate that the valves did not close when the feedwater latching logic wasintroduced, indicating that the changes affecting EF-V12 A/B appear to becorrect.

Moreover,the pressure in the OTSGs during the first 8 minutes did

not reach the initiation point for this control system, even if the change had
not been properly completed.
The possibility that the valves were closed, as a result of temperatureproblems as might occure from system backflow was addressed.

Information was

obtained that suggests at least one of the valves might have undergone athermal transient.
piping.

This was based on observed discoloration of the valve and

The visual inspection by an investigator confirmed that a plastic

instruction tag on valve EF-VllB,

the EFW control valve, was "melted".

The

investigation included a review of possible reverse flow paths to the B OTSG,
a check of maintenance requests,

and interviews with mechanical and electricalmaintenance personnel and operations personnel.
2005, Flow Diagram Feedwater and Condensate,from B OTSG.

Burns and Roe Drawing No.

shows the possible flow paths,

A backflow from inside containment would have to travel throughreactor building penetration R-616B,

check valve EF-Vl3B to EF-VI2B which

would be the most direct path to the EF-VllB valve.
involve the same penetration,of EF-VllB on to EF-Vl2B.
EF-VI3B,

EF-V32B,

EF-Vl3B,

EF-Vl2B,

An alternative path could

EF-V32B,and end at the backside

A third path could include the penetration R-616B,

and back up to EF-VlIB and/or through EF-V32B.

The discolora-tion of the pipe appears to indicate heating along the pipe from penetration
R-616B to the check valve EF-VI3B through EF-Vl2B to EF-VllB,

the most directroute.
The possibility that oil staining might indicate an overheating of thesevalves was addressed.

The EF-VI2B valve appears to have oil leakage from the

1-4-8

limitorque operator motor which stained.the valve body and.piping.
of a Work Request for the EF-V12 valve just prior to March 28,

No evidence

1979 was found.

Operations auxiliary operators who performed the EFW surveillance test that
required them to be ii the Vicinity of the EF-Vl2B**valve were interviewed
regarding the valve oil leak.

Five stated they did not recall seeing an oil

stain, while the sixth did not recall looking at that valve.
the surveillance over a period from January 3,

They did performed

1979, to March 3,

1979.

Additional

information presented to the investigator indicated that the valve EF-VI2B did
not have such an oil stain on March 26, 1979.

The investigator did note thatthe instruction tag on EF-VIIB was deformed and showed signs of being burnt(brown)on the rear side where it

is in contact with the val-ve-housing.

The condition of the EF-Vl2A valve and piping was inspected and foundsound with no similar condition.

On March 28,

1979 both EF-VI2A and EF-Vl2B

were in a closed status.
There was no evidence to cause the investigator to conclude that either
12 valves would be closed because of the condition of EF-Vl2B or the conditionof the B emergency feedwater piping.

All information indicates that both

valves opened, when actuated by the control room operator, on March 28, 1979,at about 0408 hrs.

This review did not conclude how the emergency feedwater

pipe became discolored,

how the oil leaked,

purpose of this study was to determine if

nor how the tag deformed.

The

the condition could have been areason for the EF-Vl2B valve to be in a closed position at 0400 hrs on March 28,
1979.

The findings do not indicate a relationship.

The possibility of acorrelation to the status of the B OSTG emergency feedwater piping after its
isolation during the accident was not pursued.
The possibility that the valves were closed as an operator action during
the transient was addressed.

The operating staff on duty during the period

when the valves were found closed were interviewed to determine whether thesevalves could have been closed as an operator action to prevent an excessive
cooldown rate of the RCS and an attendant'pressurizer level drop.

The investi-

gators pursued the possibility that the action was initially taken and then1-4-9

forgotten by the operator for 8 minutes (Int.

189).

No information was obtained

during this interview'that would indicate that this operator action took place
during the accident.
The investigation has not been able to establish the specific cause of
the EF-Vl2A and EF-Vl2B valves being closed during the initial part of thetransient.

The investigators concluded that the design or mechanical mal-function possibilities reviewed are not probable causes of the valves being
closed.
4.2.3

Evaluation
The delay in automatic initiation of emergency feedwater for 8 minutes

contributed to an early recovery towards normal values of certain RCS parameters
upon which the operators concentrate.

This recovery of key turbine trip/

reactor trip RCS parameters misled the operators into believing that theiractions had been successful in limiting the severity of the transient.

Thiserroneous belief led them to initiate the routine subsequent operator actions
that were normal for the assumed transient.

These actions occupied the operators'

attention and detracted from their opportunity to establish a correct analysisof the plant conditions.
This investigation did not conclude what the ultimate course of events of
the accident would have been, had emergency feedwater been introduced to the
OTSGs as designed.
Based on interviews conducted with the operating staff, it

is concluded

that once the accident was underway and after the valves were open, theknowledge that the valves were previously shut had no substantive impact on
the actions of the operators during the remainder of the accident.

1-4-104.3

HPI Flow Adjustment

4.3.1

Operator Actions Regarding Adjustment of HPI
The transient which followed a routine Turbine/Reactor Trip at the TMI-2

facility, was described as "severe" by the operations personnel.

Normal plant

response to the loss of feedwater transient should result in the following:o

Tav decrease of 33'F from 582 0 F to 549 0 F based on a turbine bypass valve(MS-V25A,o

V26A,

V25B, V26B)

setpoint of 1010 psig.

Rapid pressurizer level decrease from 220 inches to 55 inches (25 inches

below the pressurizer heater cutoff point of 80 inches),

based on a pressurizer

level change of 5 inches/OF change in Tav (33 0 F Tav change for a 165-inchchange in level).
oReactor pressure decrease from 2200 psig to near the ES setpoint pressureof 1640 psig, during the RCS cooldown.

(ES initiation had occurred at TMI-2

during this type of transient.)
The routine pressure transient resulted in a philosophy of operationthat was formulated to prevent the initiation of ES after each reactor trip.
This philosophy specifically developed following the admission of sodiumhydroxide into the RCS twice following a unit trip with an ES initiation'A

further consideration was the time and effort that was subsequently required
to clean-up and deborate the RCS after addition of substantial amounts of
sodium hydroxide and borated water (2300 ppm) into the RCS following ES
initiations.

(LER 78-021-0'3L, April 29,

Ref.

134)

o

The Reactor Trip Procedure (2202-1.1)

1979; 78-033-01T, April 23,

1978;

contained steps that required

starting the second or third makeup pump, based on pressurizer level decrease
to 100 inches or 20 inches,pumps from the BWST,
valves (MU-V-16A,

respectively, supplying suction for the makeup

and injection of borated water to the RCS via the HPI

16B,

16C,

and 16D).

(Ref.

1-4-11

76)

This operating philosophy

required that the makeup pumps and HPI flow be stopped or throttled, as soon
as possible, on indication of the recovery of pressurizer level or RCS pressure
toward normal conditions of 100 inches and 2155 psig, respectively,in orderto limit the addition of borated water (2300 ppm) to the RCS.
o

The requirements for returning the makeup system (HPI system if

ES hasinitiated) to normal required the immediate throttling of the HPI valves andstopping of one or two makeup pumps.

If an ES had been initiated, the operator

quickly blocked and bypassed the ES signal to allow manual control of the HPI
valves (MU-V16 A, 16B,

16C,

or 16D as appropriate) and the control of the

makeup pumps.
EvaluationIt

must be noted that the reactor trip procedure does not recognize theinitiation of ES following an analyzed plant transient.

Furthermore,theautomatic initiation of the ES was recognized by procedure only in the loss ofreactor coolant/reactor coolant system pressure and steam line break accidents.
The procedure stated that the automatic initiation of ES was indicative of a-large break, one beyond the capability of the makeup system.

For further

details see Section 2.6 of this report.
Based on the operating philosophy developed to cope with the associatedsevere RCS parameter transients,(within seconds),

the operators have bypassed the ES signal

when the RCS pressure was continuing to decrease and had notrecovered above 1640 psig as indicated in the loss of RC/RCS pressure procedure.
Additionally,

the operators have limited (throttled) the injection of borated

water into the RCS,

based on the recovery of the pressurizer level.

Additionally, the operators were very sensitive to taking the pressurizer"water solid." This sensitivity was developed through the formal training
program.
Operating Procedure 2104-1.2, Makeup and Purification,

Step 2.2.1,

that a MUP should not be run with the RCS (Pressurizer) in a water solid

1-4-12statescondition.

OperatingProcedure 2103-1.3,

Pressure Operation,

Step 2.1.8,

states that the pressurizer will not be taken to solid conditions except as
required for system hydrostatic tests.
Technical Specification 3.4.4 requires that the pressurizer level bemaintained between 85 and 380 inches in Modes 1, 2, and 3.
Interviews with the operating and training personnel (see Section 2.15)
indicated that the operators would respond specifically to the pressurizer
level in almost any circumstances,because they were trained to believe thatwater in the pressurizer indicated water over the reactor core.
38, 2,

14,37, 3,(Int. 5,17,

15)

On the morning of March 28, 1979,

the operators responded'as previously

during plant trips; after starting the second makeup pump and receiving an ES
actuation,

they noted the pressurizer level increasing from 155 inches andquickly bypassed the ES signal in order to throttle the HPI valves (MU-Vl6A,
16B,

16C,

and 16D)

and stop the additional makeup pump.

The HPI flow wasessentially stopped within 4.5 minutes following the reactor trip in responseto the increasing pressurizer level to the near-full-scale condition, but not
to the continuing low RCS pressure (below the ES setpoint of 1640 psig).
As noted in this report (Section 2.5), the operators had throttled the
makeup pump flow/HPI flow to a minimum within 4.5 minutes after the reactor
trip and were down to the one pump (MU-PIA).

Additionally the letdown system

was also utilized in an attempt to recover and maintain normal pressurizer
level between 0403 hrs and 0407 hrs.
The interviews revealed that-the makeup pump flow and letdown flow wereoperated in a normal manner between 0410 hrs and 0533 hrs when the HPI flow
was increased manually to approximately 300 gpm.

Moreover,, records review

revealed that the borated water delivered from the BWST to the RCS between
0400 hrs and 0730 hrs was 15,000 gallons or approximately 70 gpm average with

1-4-13the major portion of the injection occurring from 0402 hrs to 0405 hrs,
0533 hrs to 0730 hrs.

andDuring the majority of this 3 1/2 hour period, theaverage net addition to the RCS was 25 gpm.
4.3.2

Analysis

4.3.2.1

HPI Pump History

The. status of all three makeup pumps was established by the investigation
for the entire time of the incident using information from the alarm printer,
operator logs and operator interviews.
pumps are shown on Figure 1.4-1.
on,Graphs of the status of the three

These graphs show three pump states:

and the high pressure coolant injection (HPI)

condition.

off,

During HPI,

two

pumps are on and the throttling valves for the pumps automatically go to fullopen, at which position each pump puts at least 500 gpm into the RCS.

When apump is on, but HPI is not actuated, the flow can be throttled back by theoperators.

According to operator statements (Int.

15) the minimum flow thatcan be adequately controlled is about 100 gpm per pump.

Table 1.4-1 lists the

specific on and off times for each makeup pump.
At least one makeup pump was runnihg during the entire time of the incident,with the exception of a 5-minute period starting at 0817 hrs.
MUP-IA was the pump selected.
start this pump at 0818 hrs,

However,remainder of the incident.

after two unsuccessful attempts to

it was taken out of service.

selected and started at 0822 hrs,

Until 0800 hrs,

MUP-lB was thenafter which it continued to run during the

MUP-IC was used from time to time throughout theaccident to supplement the selected A or B pump.

1-4-14

MAKE-UP AND DECAY HEAT PUMP STATUS

c,,
a.

ON
B
OFF

w

ON
A

Lu

a

OFF

cl,

0'1

0-

HPCI
C

OFF
HPCI

c,,

a.
a.
a.

ON

B

ON
OFF
HPCI

A

ON
OFF
0

2

4

6

8
10
ELAPSED TIME - HOURS

FIG 1.4-1

12

14

16

TABLE 1.4-1
HPI PUMP HISTORY

Elapsed Time

MUP A

Pump Status
MUP B

MUP C

Source of Information

Comment

0

Off

On

Off

Interviews (GPU 1, 20;

MUP B acting as normal makeup
pumpInt. 15)
41 seconds

On

Alarm PrinterInterviews (GPU 1I 20;
Int. 15)

2 minutes
2 seconds

Off

4 minutes,
38 seconds

On

Off

ES initiation; control valvesfull open.

Alarm PrinterInterviews (GPU,20)

Alarm PrinterInterviews (Int.

15)

10

minutes,
24 seconds

Off

Alarm Printer11 minutes,
43 seconds

On

Alarm Printer

MUP-lA started by operator;
MU-Vl6B opened to increaseflow.

MUP A throttled.
MUP 1A stopped, restarted
and stopped within 4 seconds.

3:20

On

Alarm PrinterES initiated by operator;
valves full open.

3:37

Off

Alarm PrinterES reset.

3:56

On

4:00

ES on high reactor building
pressure; valves full open.
Alarm PrinterES defeated.

TABLE 1.4-1 (Continued)
HPI PUMP HISTORY

Elapsed Time

MUP A

4:17

Off

Pump Status
MUP B

MUP C

Source of Information

Comment

Off

Alarm PrinterNo pumps running.

Alarm Printer(Int. 15)After two unsuccessful start
attempts, MUP-lA pulled tolock; remains off for restof incident.

4:18

Interviews
4:22

On

Alarm Printer

MUP-lB remains on for restof incident.

Interviews

I

4:27

On

Alarm Printer

9: 04

Off

Alarm Printer

9:50

On

Alarm PrinterES on high reactor building
pressure; valves full open.

9:51

Off

Al arm Printer

MUP-IC was on 1 minute.

10:32

On

Alarm Printer

10:36

Off

Alarm Printer

MUP-IC was on 4 minutes.

11:19

On

Al arm Printer

MUP-IC was on 9 minutes.

11: 28Off

Al arm Printer

11:33

On

Al arm PrinterNTABLE 1.4-1 (Continued)
HPI PUMP HISTORY

Elapsed Time

MUP A

Pump Status
MUP B

MUP C

Source of Information

Comment

11:36

Off

Alarm Printer

MUP-lC was on 3 minutes.

13:23

On

Alarm Printer

14:41

I-

Operating Logs

MU-Vl6C throttled to 105
or 120 gpm.

14:43

Off

Alarm Printer

15:33

On

Operating Logs

MUP-IC started by operation;
ES initiation at this time
was immediately bypassed.

15:39

Off

Alarm Printer

MUP C was on 6 minutes.

15:49

On

Alarm Printer

15:56

Off

Alarm Printer

MUP-lC was on 7 minutes;
reamined off for rest ofincident; MUP-IB stillrunning.

4.3.2.2

Effective Flow Rates to the RCS

Instantaneous makeup flows are recorded hourly by the Computer Log Typer.
These are listed in Table 1.4-2 for the period of the incident.

They varied

from 8 to 128 gpm, but averaged about 80 gpm during the incident.
flows through the 17 valves used for normal makeup.

These are the

They do not include the RCpump seal water injection, which totals about 32 gpm, or the HPI flow throughthe 16 valves, which is known accurately only immediately after ES actuation,
since these valves can be throttled once ES is reset:
The net makeup from the borated water storage tank (BWST),

generally injected

through the 16 valves, could be derived for various time periods from the known
changes in the BWST levels.

Table 1.4-3 lists the average flow rates for various

time periods and the data and data sources,used to derive them.

These rates

and the total quantities of BWST water expended during each period are plottedon Figure 1.4-2.

These rates, the makeup pump status discussed and pressurizer

or core flood tank volume changes, were used to develop net addition flow rates
to the primary system.

These can be divided into five main time periods forconvenience of discussion:

1-4-18TABLE 1.4-2 MAKEUP FLOW THROUGH THE MU-V17
VALVES AND RCP PUMP SEALS (GPM)
ElapsedTotal

MU-V17 Valve

Pump Seal

Time (Hours)

Flows

Flow

0

16

32

48

1

118

32

150

2

97

32

129

3

125

32

157

4

8

32

40

5

58

32

90

6

52

32

84

7

49

32

81

8

49

32

81

9

97

32

129

10

113

32

145

11

95

32

127

12

116

32

148

13

128

32

160

14

43

32

75

15

72

32

104

16

95

32

127

1-4-19TABLE 1.4-3
AVERAGE MAKEUP FLOWSBASED ON BWST LEVELS

Elapsed

Total

Used inAvg

Source

Level

Used

PeriodFlow inof

Hours:Minutes

(ft)

(gal)

(gal)

Period (gpm)Information

3:30

53.04

15,00015,00070

Alarm Printer

6:55

37

147,000

132,000

640

RI Incident Response

Time

Center9:15

31

197,000

50,000360

RI Incident Response
Center

13:20

26.5

234,000

37,000150

RI Incident Response
Center

14:47

22

271,000

37,0001-4-20430

Operator Logs

60
C)

APPROXIMATE VOLUME (GALLONS)INJECTED FROM BWSTOVER TIME INTERVAL-15,000

40

wz
Lu0

0
I

20

0

0
14

--640 GPM

"

600 -

ccZ
-460 GPM
c,,-H

400

-360 GPM

0>
0-

200
"150 GPM"v70 GPM

0

>ca
4L

0
0

I

I

I

2

4

6

8

I

I

I

I

10

12

14

16

ELAPSED TIME AFTER TURBINE TRIP (HOURS)FIG 1.4-2

The first 3 1/2 hours of generally very low flow,
The following 3 1/2 hours of high flow rates,
A 2-hour period of moderate flow rate,
An interim 4-hour period of low flow rates,
The final repressurization.
These periods are detailed below:
Only 15,000 gallons of water were injected into the reactor from the BWST
during the first 3 1/2 hours,

approximately 2/3 of it during two HPI actuations.

As indicated in Section 4.3.1,

the overall average flow for this period was 70

gpm.

However,during most of this time the flow was much less.

During thefirst minute of the accident, the operators started MUP-lA and opened MU-V-16B.
This was followed by ES initiation at 2 minutes, which was throttled back at
4 minutes (Int.

61,

15).

about 3000 gallons.

Thus, the net flow during the first 4 minutes was

At 0720 hrs,

ES was initiated and maintained for 7 minutes,

resulting in a flow rate of 1,000 gpm for a total injection of 7,000 gallons.
The remaining 5,000 gallons injected during the period from 0404 hrs until
0720 hrs, when ES was not actuated, results in an average net flow rate ofabout 25 gpm for that period.
The BWST level at 1055 hrs shows a total expenditure between 0730 hrs and1055 hrs of 132,000 gallons.

This volume and time period can be adjusted forthe known or maximum flows identified below.
oFrom 0727 hrs to 0756 hrs,

the flow to 500 gpm.

one pump (MUP-IA)

was running, which limited

In addition, between 0730 hrs and 0740 hrs, there was a3600-gallons increase in indicated pressurizer water volume.

This occurred

during a period of stable pressure and decreasing cold leg temperatures with
the EMOV closed.

(Plant Strip Charts and Multipoint Recorder; Ref.

see discussion of EMOV status in Section 4.8)

Thus,

period is estimated to be between 360 and 500 gpm.

1-4-2213.

the flow during this

Also

o

ES was initiated at 0800 hrs,

hrs, when ESF was defeated.
itresulting in a l000-gpm flow until 0804

Both pumps continued to run until 0817 hrs, but

is not known whether or not the flows were throttled.

(Alarm Printer; Ref.

14)
oThere was no flow for 5 minutes from 0817 hrs until 0822 hrs, as all of

the makeup pumps were off.
NOTE:

(Alarm Printer; Ref. 14)

MUP-lB was turned on at 0822 hrs and MUP-IC was turned on

5 minutes later.

Both pumps continued to run until 1304 hrs when

MUP-IC was shut off.
These adjustments result in a minimum net average flow rate during
two-pump operation of 680 gpm for the period between 0730 hrs and 1055 hrs.
(4)

Between 1055 hrs and 1315 hrs, 55,000 gallons of BWST water were

expended for an average flow rate of about 360 gpm during this period.

At theend of this period, the RCS was depressurized in order to inject the coreflood tanks, and a partial injection was achieved.

The available data on coreflood tank levels is not sufficiently consistent to be used to establish theamount of water injected by this means.

However,

measured some time after the incident (Int.
400 psi.

a core flood tank pressure114) was reported to be about

The normal CFT gas volume is 370 cubic foot. (Core Flood System

Description,

Index No.

2813; and Technical Specifications; Ref.

15)

The

volume injected by both tanks, assuming isothermal gas expansion from 600 psi,
is about 2,670 gallons.
(5)

During the 4-hour period between 1315 hrs and 1720 hrs,

gallons of BWST water were expended,from the BWST.

37,000for an average net flow rate of 150 gpm

MUP-IC was used intermittently for a total of 17 minutes,

including a 1-minute ES actuation at 1350 hrs.
(6)

Following the decision to repressurize the RCS,again increased.

the flow rates were

MUP-IC was turned on at 1723 hrs (MUP-lB was already on) and1-4-23

run continuously until 1841 hrs.

The BWST level drops during this period

showed a total of 37,000 gallons injected.

Based on the period of a two-pumpoperation, this results in an average injection rate of about 470 gpm.

No

BWST levels were recorded after 1841 hrs.
4.4

Shutdown of the Reactor Coolant Pumps
During normal operation at full power (97% on 0400 hrs on March 28,1979),

all four reactor coolant-pumps (RC-PIA,

P-2A, P-IB, and P-2B) wereoperating and following the routine loss of feedwater turbine/reactor trip at
0400 hrs.

The pumps continued to operate until manually tripped at 0514 hrs(PIB and P2B) and 0541 hrs (PIA and P2A) by the operators.
Subsequent to the unit turbine/reactor trip, resulting in the normal RCS
pressure increase (2250 to 2435 psig), and the apparent opening of the EMOV
(RC-R2)and the subsequent failure to reclose resulted in the continuing lossof mass from the RCS.
Records review and personnel interviews revealed that conditions ofdecreasing coolant flow,

increasing reactor coolant pump vibration, and pump

motor vibration increases began almost immediately following the reactor trip
at 0400 hrs.
At approximately 0406 hrs, the RCS conditions were at saturation
temperature and pressure conditions (585'F and 1050 psig) and within 15 minutes
the RCS pressure also dropped below 1200 psig again and remained below 1200
psig until about 0605 hrs.

The existence of saturation conditions and the RCS

pressure below 1200 psig requires,by procedures,the tripping of the reactor

coolant pumps.
The RCS flow began to decrease steadily following the reactor trip at
0400 hrs and had decreased to approximately 65 million pounds per hour by 0420
hrs as the RCS pressure decreased to less than 1100 psig.

1-4-24

The normal operating procedure (2103-1.4,
Reactor Coolant Pump Operation,

Step 2.1.1.1)

Revision A, April 18,

1978,

specifically addresses RCS

system pressure requirements that must be maintained for required RCP
operation.

(Ref.

91)

The heatup/cooldown curve addresses the RCS pressure-temperature limits for reactor coolant pump operations.
NOTE:

The RCS pressure-temperature limits for fuel pin compression

considerations are more conservative than the limits for RCP operation.
The minimum pressure-temperature limits for continued RCP operation were
not met.
The Emergency Operating Procedure (2202-1.3,
Loss of RC/RCS Pressure, Step A.3.2.7),

Revision E, May 12,

1978,

requires the operator to trip the RCPs

before RCS pressure decreases below pump NPSH conditions (as indicated by theheatup/cooldown curve).

(Ref.-87)

Step B.2.274 of that procedure requiresthat the operator trip the reactor coolant pumps before reaching 1200 psig
decreasing.
Unit 2 Abnormal Procedure 2203-1.4, RCP and Motor Emergencies,
addresses pump and motor vibration.

Revision 3,

This procedure categorizes the Allis-

Chalmers motor installed vibraswitch as an alarm and as a symptom related to a
dropped impeller.
Typer.

The alarm was recorded at 0654 hrs on the Computer Utility

This procedure also addresses shaft vibration sensed by the vibration

sensors (IRD).

The procedure allows for the operation of one pump per loop

for the first four hours at a vibration level greater than 30 mils and greaterthan 26 mils for one or two pumps per loop with no required automatic action.
The procedure also requires the resetting of the IRD vibration alarm and, if
vibration alarms immediately reoccur,

reactor power must be reduced and theaffected RCP tripped per normal procedure.
vibration as a symptom of seal failure.
and no automatic action occurs.

This procedure addresses pump

A vibration level is not specified

The indication for action is based on sealcavity pressure exceeding 2500 psig.

The action criteria related to vibration

includes a bleed off flow and seal leak off flow exceeding 3 gpm which was notevident from records for March 28,

1979.

1-4-25

A second procedure,
and Precautions,Operating Procedure 2101-1.1,

Nuclear Plant Limits

Revision 4, addresses shaft vibration.

The procedure requiresthat B&W engineering be notified when RCP vibration at the pump coupling(shaft vibration) reaches 15 mils peak to peak amplitude.

The procedureallows a shaft vibration of 20 mils for a period not to exceed four hours.
Operating Procedure 2103-1.4 requires the contacting of the pump manufacturer if the RCP vibration at the pump coupling reaches 15 mils peak
amplitude.

This procedure allows shaft vibration of 20 mils for a period of

4 hours on initial start.

The procedure requires a RCP trip if the motor

stand vibration exceeds 3 mils.

The motor stand vibration is noted on control

room panel 10 as Motor Brg. "x" and "y".
of RCP if

This procedure requires the tripping

shaft vibration exceeds 20 mils for the first four hours and must be

tripped if shaft vibration exceeds 30 mils under any condition.
The pump vibration IRD values are read and recorded by the control room
operator as a routine parameter.

The vibration indicated values are notrecorded through any automatic printout.
From the results of reviewing Abnormal Procedure 2203-1.4, Operating
Procedure 2101-1.1 and Operating Procedure 2103-1.4 itis evident that theprocedures have disagreements as to vibration limits and actions to be taken.
This is mentioned as an item to be resolved and is left for appropriate
licensee action.
Evaluation
Based on the records review of the RCS/temperature relationships anddiscussions with licensee personnel,
exceeded early in the event.

the low pressure RCP trip criteria was

Also, the discussion indicated that the pump

vibration indicators had increased to their full scale indicating condition.
Section 2.5 of Details I describes an item being considered as a possible itemof noncompliance related to the failure to follow procedure and trip the RCPs.

1-4-26

4.5

Attempted and Successful Restarts of the Reactor Coolant Pumps
Following the stopping of the RCPs at 0514 hrs and 0541 hrs, there weredoubts that adequate natural circulation was occurring.
forced cooling must be provided to the reactor.

It was decided that

Reactor coolant pump operation

occurred twice in the morning (0654 hrs and 0808 hrs) and twice in the evening(1933 hrs and 1950 hrs).
The first restart of a reactor coolant pump was attempted following theclosing of the RCS EMOV (RC-R2)

at 0612 hrs.

The RCS pressure was increasingfrom a minimum of 660 psig as a result of HPI flow (increased at approximately
0533 hrs) into the RCS.

The attempt to restart the RCP commenced at about0630 hrs and included jumpering the starting interlocks for the pumps,
allow pump start.

to

The specific conditions which prevented the RCP startcapability was not determined.
After attempting to start sequentially RCP-lA,

RCP-2A,and RCP-lB,

RCP-2B

finally started at 0654 hrs and was operated until 0713 hrs in an attempt to
supply cooling to the reactor core.
The reactor coolant pump (RCP-2B)

starting current returned to about

100 amps after approximately 6 seconds,

and the-reactor coolant flow in the B

loop increased (spiked) to approximately 10 million pounds per hour anddecreased quickly to minimum.

It

was concluded that the pump was not operating

properly, was vapor bound and RCP-2B pump was stopped after 19 minutes at 0713
hrs.
The second attempt to run a reactor coolant pump occurred at 0808 hrs,
and the pump was left on for only 37 seconds,

after the pump conditions

indicated unsatisfactory pump performance again,
motor current above about 100 amps.
bound (Int.

121,

including no flow and no pumpIt was concluded that the pump was vapor

192).

The third attempt to run a reactor coolant pump occurred at 1933 hrs.
The system pressure had been increased via the HPI pumps to about 2300 psig at

1-4-27

1930 hrs.

The decision to repressurize was apparently made at approximately

1730 hrs.

The decision was made to bump the RCP for 10 seconds to sweep any

steam/gas from the pump and ensure that the pump would operate normally.

The

RCP-lA pump was started for 10 seconds and indicated normal starting current
and the loop flow increased to about 30 million pounds per hour.
RCP-1A was restarted at 1950 hrs, following the normal waiting period
after starts, and the RCS was returned to stable conditions at approximately
1000 psig and 280'F with the one pump operating.
4.6

Closing of the Core Flood Tank Isolation Valves
During the first two hours of the accident, the RCS pressure decreased

from the initial value of 2200 psig (normal operating condition) to approximately 660 psig prior to closing the EMOV (RC-R2)

at 0618 hrs.

The interviews of operations personnel revealed that the core flood tankvalves (CF-VIA and VIB) were closed by operator actions at approximately 0600
hrs to prevent the addition of additional borated water to the RCS when the
RCS was believed to be in the solid water condition, as indicated by thenearly full pressurizer (level196,

indicated at or near 400 inches).

(Int.

95,

109)
The specific requirements and steps required to close the valves included

unlocking and closing the valve motor operator electrical breakers at the
motor control centers (2-11EB:

CF-VlA and 2-21EB: CF-VIB) and closing thevalves remotely (electrically from the control room).

(Int. 95,

196,
1

109).

The interviews did not determine the specific time the core flood tankvalves (CF-VIA and VIB) were reopened,but it was apparent that the valves

were opened prior to depressurization of the RCS at approximately 1300 hrs
resulting in the discharge of some amount of core flood tank (CF-TlA and TIB)
water into the reactor vessel.

The discharge of water into the RCS from the

CFTs was confirmed through interviews and computer alarms.
184,

119)

I-4-28(Int. 57,

111,

Operating Procedure 3102-1.1,

Unit Heatup, Steps 4.38, Mode 4 to Mode 3

checklist (Appendix C) and Step 4.41,
This is done by Procedure 2104-1.1,

places the core flood system into operation.
Core Flood, at RCS pressure of 750 psig

and puts the core flood tanks into an OPERABLE status.
2102-3.2,

Unit Cooldown,

Steps 4.16 and 4.17,

Operating Procedure

remove the core flood tanks fromthe OPERABLE status condition between 750 psig and 700 psig, decreasing.
Evaluation
The Operating Procedures referenced above, 3102-1.1 and 2102-3.2 address
the isolation (and unisolation) of the core flood tanks during normal plant
evolutions of heat-up and cooldown.

These procedural requirements indicatethat the core flood tank isolation valves would be maintained in the open
positions in Modes 1, 2, and 3.
Technical Specification 3.5.1,

action statement b,

indicates that withany core flood tank inoperable due to the isolation valve being closed,theplant shall be in HOT SHUTDOWN (280'F > T avg > 200 0 F) within the next 12hours.

The normal cooldown procedures would appear to contain this requiredcondition even though the procedure would allow the plant to be maintained at
800 psig or less and higher than 280OF for an indefinite period of time.
Emergency Procedure 2202-1.3,
Pressure, Section B, Step 3.1,

Loss of Reactor Coolant/Reactor Coolant

requires that the verification that all ESF

equipment is in its ESF position, by observing that all equipment statuslights indicate as shown in a table in the procedure;that the core flood valves (CF-VIA and VIB)

including verification

are open following the automatic

initiation of ESAS.
Technical Specification 6.8.1.a requires that written procedures shall be
implemented covering the activities referenced as the applicable procedures
recommended in Appendix "A" of Regulatory Guide 1.33,

November 1972, which

cover procedures for combating emergencies such as "LOSS OF COOLANT".

1-4-29

The closing of the core flood tank valves (CF-VIA and VIB),
accident on March 28, 1979,
Section B, Step 3.1,

during thecontrary to the Emergency Procedure 2202-1.3,

Table B-i,

is under consideration as a potential item ofnoncompliance pursuant to Technicial Specification 6.8.1.a.
4.7

Use of Core Flood Tanks
At approximately 7 hours and 30 minutes after the start of the accident,

the decision was made to depressurize the RCS so that the core flood tanks
would be available to discharge their contents if the vessel required large
quantities of water.

Once having done so, further steps would then be taken

to enable the staff to put the decay heat removal system into service.
(Further detail is available in Section 3.2)
The RCS depressurization took approximately one hour, the RCS pressure
reaching 600 psig at 1241 hrs.
Starting at about 1212 hrs and lasting for the next several hours,
multiple level alarms were received on the A core flood tank.
were received on the B core flood tank.

No such alarms

The investigation has concluded thatthese alarms are probably erroneous since the nitrogen pressure on both tanks
was reported to have decreased to approximately 400 psig.

This would imply alevel decrease well below the alarms that were indicated.
This condition of the core flood tanks "floating" on the RCS remained in
effect until approximately 13-1/2 hours when the Station Manager was directed
to take the plant solid.

The plant staff has stated that, at that time,

theybelieved the behavior of the core flood tanks demonstrated that the core wascovered with water.

The presence of the loop seals in the piping from thecore flood tank to the reactor vessel precludes the tank level change behavior
from assuring that the core was covered at that time.
further in Section 4.8 of this report.

1-4-30This is discussed

EvaluationIt

is not possible for the investigation team to assess the direct impactof the use of the core flood tanks, as they were used during this accident, on
the outcome of the accident.

To do so would require treatment of thefollowing areas:oAnalysis of the vessel water levels during the low-pressure
period from approximately 8 hours to 14 hours after the startof the accident and subsequent analysis of the core thermal
behavior under those conditions.

oAnalysis of the extent of venting of noncondensible gases fromthe RCS during the low-pressure period and the subsequent
impact that venting had on the ability of the staff toapparently refill the A RCS loop and successfully operate
reactor coolant pump PIA.

The analysis of these matters is beyond the scope of this investigation.
Consequently,
4.8

no conclusions have been drawn in this area.

Reactor Vessel Level Indication
Plant operators attempted to use pressurizer level indication, early inthe accident sequence,and core flood tank level indication,

half way through the accident sequence,
conditions.

approximately

to infer reactor vessel water level

The operators knew the pressurizer and both core flood tanks werepositioned well above the core.

They assumed the core had to be covered withwater; if a water level existed in the pressurizer, regardless of systempressure, or in both core flood tanks, at a system pressure below the nominal
600 psig cover gas pressure.

The operators failed to recognize that theexistence of saturated, even superheated conditions within the RCS,

coupledwith the actual piping configuration from these vessels to the RCS,

allowedloop seals to form.

The loop seals for the pressurizer, core flood tank 1A

1-4-31and core flood tank IB are about llfeet 5 inches,feet 10 inches; respectively.

14 feet 4 inches,

Once loop seals formed,

and 17

RCS pressure needed

only to exceed the pressurizer vapor space pressure by about 13 psi and thecore flood tank cover gas pressure by about 8 psi to force both water levels
to full. scale, even with the RCS completely voided.
Admittedly,

the RCS should not normally experience saturated or superheatedconditions and, without those conditions,the piping configuration would nothave mattered.
4.9

Sump Pump Operation

4.9.1

Sequence
There are two reactor building sump pumps,

having a design pumping capacity of 140 gpm.

WLD-P2A and WLD-P2B,

each

The A pump started at 0408 hrs.

and the B pump started at 0411 hrs as indicated on the Alarm Printer.

Bothpumps continued to operate until they were shut-off by the operators at 0438
hrs.

The amount of water transferred by these pumps totaled about 8000 gpm

based on the 140-gpm per pump capacity.

(Int. 5,15)This sump water was a mixture of reactor coolant drain tank (RCDT)

quenchwater and primary coolant vented through the-RCDT relief valve until 15 minutes
into the event.

This valve has a water relieving capacity of 2270 gpm.

137; Section 5.5.11.3)(Ref.

The RCDT pressure reached the relief level about 4

minutes before the A pump started.

This pump starts at a sump level of 32

inches at which point the sump contains less than 1,000 gallons of water.
second pump comes on at a level of about 44 inches.

The

The flow to the sump was

through the rupture disc after it opened at 15 minutes into the event.
4.9.2

Discharge Point
The destination of this water had not been definitely established.

The

sump tank discharge is normally aligned to the miscellaneous waste holdup tank
WDL-T-2.

However,

the tank level records do not show any level change in this

1-4-32

tank.

An alternative discharge point is the auxiliary building sump tank,which has a 3,200-gallon capacity and was already 3/4 full at this time.

This

subject is addressed in greater detail in Details II of this report.
EvaluationIt

is of interest to note that, during the ten-day period-prior to theaccident,

sump pump operation would occur between 2 and 6 times per day, each

operation discharging approximately 200 gallons of water.

The average forthis period was 4.1 transfer operations per day, or approximately 800 gallons
per day.

The starting of a sump pump would not key the operators to a problem -only its continued operation and the start of the second pump.
During the review of the sump pump operation history for the 10-day
period,no trend was noted that would suggest a steadily increasing rate ofwater accumulation in those sumps.
4.10 Electromatic Relief Valve Failure
4.10.1

History

The EMOV is described in Section 1.6.2.2, which provides a selectedmaintenance history on that valve immediately prior to March 28,

1979.

The

section also presents the history of the valve believed to be installed on the
TMI-2 pressurizer.
The calibration data (TMI-2 Instrument Calibration Data Sheet, InstrumentNo.

RC3-858) performed on September 24, 1978, shows the valve setpoint values

to be 2255 psig high and 2205 psig low (Ref.

16).

During the March 28,

1979

event, the high reactor pressure exceeded the high setpoint of 2255 psig inless than 6 seconds after the turbine trip.

The EMOV then did not reset when

the RCS pressure was below the low pressure setpoint of 2205 psig, approximately
13 seconds later.

The EMOV remained open; this was unknown to the operators

at that time.

1-4-33

Alarm Printer data indicated that the EMOV exhaust pipe temperaturereached 239OF which would be expected within approximately 30 seconds following
an EMOV actuation.

The EMOV exhaust pipe temperature was printed out as 283OF

at 81 minutes (Utility Typer),temperature.

and remained above the code safety valve exhaust

The exhaust temperature of the EMOV was printed out (Utility

Typer) again about 2 hours 18 minutes into the event.

The EMOV was isolated

at this time by operator actuation of the EMOV block valve (RC-V2).
Evaluation
The history of this valve included indications of leakage.
the March 28,

Just prior to1979 event it was suspect, along with the code safety valves, ofleaking to the reactor coolant drain tank.

The effect of the extended periodof leakage on the performance of this valve during the accident cannot be
determined at this time.
The evidence clearly supports that the EMOV failed to close at the lowpressure closure setpoint, but a specific cause of that failure has not been
determined at this time.

Final determination will require access to the valve

itself.
4.10.3Temperature History

4.10.3.1

Relief Valve Tailpipe Layout

The effluent from the EMOV RC-R2 and the two code safety valves (RC-RVIA
and RC-RVlB) is piped to the reactor coolant drain tank via a common 14-inch
header.

(Burns and Roe,

P & ID Dwg.

No.

2024; Ref. 72)

A 4 inch exhaust line

having a total run of about 17 feet leads from the EMOV to the 14-inch header.
(M.W.

Kellog Isometric Drawings No.

223-1 and 2; Ref. 73)

Six-inch exhaustlines having runs of about 20 feet lead from each of the code safety valves tothe 14-inch header, entering 3-1/2 and 5 feet below the EMOV exhaust line
entry point.

Thus the exhaust points for the EMOV and two code safety valves

are separated by approximately 40-42 feet.

1-4-34

4.10.3.2

Thermocouple Locations

One chromel constantan thermocouple (T/C)

is strapped to each relief

valve exhaust line.

The T/C on the EMOV line is located about 3 feet down-

stream of the valve.

(Plant Training Photograph No.

443; Ref. 74).

The T/Cson the code safety exhaust lines are located about one pipe diameter fromthese valves.

(Plant Training Photograph No.

81; Ref. 75)

Thus,about 40feet of piping separate any one T/C from the other two.
NOTE:

This location is based on discussions with plant staff anddisagrees with plant drawings which show the T/C located several
feet further downstream of the valves.

These temperatures are normally recorded on a multipoint temperaturerecorder in the control room.
period of the incident.

However, this data is unavailable for the

The chart drive stopped before the accident andentries were printing over each other during the accident.

No information has

been obtained to explain the cause of this recorder behavior.
the chart drive resumed after the accident.

Operation of

No chart paper is missing.

These

three T/Cs are also connected to alarms that are -printed out by the Computer
Alarm Printer.
about

200 0 F.

"High temperature" is printed when one of these points exceeds
The alarm clearing is printed when the temperature decreases

below about 193 0 F.
period of the event.

These status changes are listed in Table 1.4-4 for the
In addition these temperatures can be called up by theoperators through the computer and printed by the Utility Typer.
done several times during the accident,

This wasand these temperatures are also includedin the Table.
During the accident, the opening of the EMOV or the associated block
valve generally caused all three of these T/Cs to alarm.
to the previously described common exhaust header.

They are connected

If the EMOV/block valve

pathway remained shut for a sufficient time, these alarms would clear.

This

generally required 10 to 20 minutes for the code safety exhaust line temperatures

1-4-35

and up to 37 minutes for the EMOV exhaust line.

The increased time required

to cool-the EMOV exhaust line may have been caused by continuing lower level
leakage from this valve.

In general it

is not known whether the relief line

closure was by the EMOV or by the block valve, since partial use of the EMOV
was apparently recovered during the accident.

For this reason the term "EMOV

flowpath" will be used in discussing the open/closed status of this pathway.
The safety valve temperature alarm status was used to establish the EMOV
flowpath opening times.

Closure times were established by changes in the RCS

or reactor building pressure trends.

The open/closed status of the EMOV flow

path and the method used to determine the time of the changes are included inTable 1.4-4.
The EMOV opened within a few seconds of when the RCS pressure reached the2255 psig EMOV opening setpoint.

Itremained open until 0619 hrs when theoperators discovered that the EMOV was stuck open and closed the block valve.
The EMOV flow path was opened twice for brief periods at 0713 hrs and0719 hrs in order to reduce pressure following the rapid rise that occurred at
the time of the restart of reactor coolant pump 2B at 2 hrs 54 minutes.
The EMOV flowpath was opened again at 0741 hrs and apparently left open
until 0918 hrs when the decision was made to repressurize the reactor.
From 0944 hrs until-1135-hrs,

the block valve was apparently cyc-led-open

and closed intermittently to maintain the pressure between 2200 psi and 1940
psi.

(RCS

and Reactor Building Pressure Charts; Ref. 135)

During this period

the block valve was opened and closed on a 3-to-6 minute cycle, with about a
40% open time.

The cycle time was too short for the temperature alarms toclear during this time.,
Following the decision to depressurize the RCS, the EMOV flowpath was
opened at 1135 hrs.

During the depressurization the code safety A and B valve

temperature alarms cleared at 1217 hrs and 1223 hrs respectively,
the RCS pressure had reached its minimum value.

1-4-36

well before

There were no changes in thepressure trends at this time to indicate EMOV pathway closure.

At 1310 hrs,

the beginning of a drop in reactor building pressure indicated that the EMOV
pathway did close at this time and the EMOV temperature alarms cleared 7 minutes
later to confirm this.

However,the RCS pressure was still above the decay

heat removal pressure of 400 psi.

A possible explanation is that the EMOV had

gone shut spontaneously and the previous clearing of the code safety temperature
alarms indicated a gradual partial EMOV closure at that time.

An EMOV high-temperature alarm and a small rise in reactor building pressure at 1320 hrs,
shows that the EMOV pathway opened briefly at that time and closed again at1323 hrs.

The EMOV temperature alarm cleared 9 minutes later.

The operators opened the EMOV manually at 1350 hrs (the time of thebuilding pressure spike).

The operators reported that the block valve wasalready open at this time and that they opened the EMOV in order to increase
the RCS venting rate (Int.
temperatures,

14,

15,

17).

High alarms on all three relief valve

confirmed that the EMOV valve opening at this time.

The reactor

building spray, which actuated at this time, apparently affected the safetyvalve temperatures,

since they cleared at abnormally low temperatures and twoof them alarmed high again in the following several minutes.
cycled twice.)(The EMOV alarm

The EMOV pathway was closed some time between 1401 hrs and1424 hrs when the EMOV temperature alarm cleared.
The EMOV pathway was opened for the final time between 1635 hrs and 1648
hrs.

Approximately a half hour later the final repressurization of the RCS

was started.

Stable reactor building pressures and the absence of temperature

alarms indicate that the EMOV pathway remained closed for the remainder of theincident.

1-4-37

TABLE 1.4-4
EMOV AND SAFETY VALVE TAIL PIPE TEMPERATURESTAIL PIPE STATUS
AND TEMPERATURES
ELAPSEDTIME
6 seconds

EMOV
V2

CODE
SAFETY
VIA

CODE
SAFETY
V11B

SOURCEHigh 203

Alarm Printer

- 264

- 275

Utility Typer

- 283

- 211

- 218

Utility Typer

- 229

- 190

- 194

Utility Typer

High 239

0:25

- 285

1:21
SHUT

COMMENT

Setpqint for opening
reached.

OPEN

30 seconds

2:18
03

BLOCK VALVEOREMOV STATUS

INFORMATIONRx building pressure
drops.

2:20

- 228

Utility Typer

2:21

- 227

Utility Typer

2:24

- 220

Utility Typer(From 1:13 to 2:47 alarm

Alarm Printerprinter missing.)

3:13

OPEN

3:15

SHUT

Rx Bldg Press

Termination of pressure
rise.

3:19

OPEN

Rx Bldg Press

Beginning of pressure
rise.

High 247

High 203

High 206

TABLE 1.4-4
EMOV AND SAFETY VALVE TAIL PIPE TEMPERATURESTAIL PIPE STATUS
AND TEMPERATURES
ELAPSED
TIME
3:20

BLOCK VALVEOREMOV STATUS

CODE
EMOV
V2

SAFETYVIA

INFORMATION
CODE
SAFETY
V11B

SHUT

Rx Bldg Press

3:31

CLR 193

3:35
3:41

OPEN

5:15

SHUT

SOURCEHigh 202

5:35

CLR 192

5:40
5:44

8:O00

OPEN

High 215

Termination of pressure
rise.

Alarm Printer
CLR 193

Alarm PrinterHigh 205

Alarm Printer

May not have been timefor EMOV T/C to cool.

RCS Press. andRx Bldg Press

Beginning of RCS pressure
rise and Rx Bldg Pressure
drop.

I
!
•0

COMMENT

Alarm Printer

CLR 193
CLR 193

Alarm PrinterHigh 205

Alarm Printer206.1

Utility Typer(Cycling block valve
with about 40% open
time until 7:35.)

TABLE 1.4-4
EMOV AND SAFETY VALVE TAIL PIPE TEMPERATURESTAIL PIPE STATUS
AND TEMPERATURES
ELAPSEDTIMEBLOCK VALVEOREMOV STATUS

EMOV
V2

8:17

CODE
SAFETY
VIA

CODE
SAFETY
V11B

SOURCE

Alarm Printer

CLR 193
CLR 193

8:23
9:10

INFORMATION

May have cleared due topartial EMOV closure.

Alarm PrinterRx Bldg Pressure

SHUT

COMMENT

Termination of risein Rx Bldg pressure.

9:17
9:20

OPEN

9:23

SHUT

9:32

CLR 193

Alarm PrinterHigh 220

Alarm PrinterRx Bldg Pressure
Alarm Printer

CLR 193High 226

High 204

9:52

CLR 181

CLR 179

9:55

CLR 177

9: 49-9:50

OPEN

Termination of risein Rx Bldg pressure.

High 205

CLR 190

Operator, Alarm
Printer, andUtility Printer(H burn and ES
ahtuation.)

Alarm Printer(EMOV temperature
alarms again in 20 sec.)
cleared due to Rx
Bldg Spray

Alarm PrinterClears again in two
minutesTABLE 1.4-4
EMOV AND SAFETY VALVE TAIL PIPE TEMPERATURESTAIL PIPE STATUS
AND TEMPERATURES
ELAPSEDTIMEBLOCK VALVEOREMOV STATUS

10: 01

10:+
10:24-10:25

I

EMOV
V2
High 207

CODE
SAFETY
V11B

SOURCEHigh 200

Alarm PrinterSHUTCOMMENT

Finally returns to highstatus, caused by terminationof building spray.
Clearance of EMOV
alarm at 10:24.

--

CLR 192

11:27--

NORM 190.7

12: 35OPEN

High 233

12:48

SHUT,

CLR 193

Alarm Printer

NORM 170.6

NORM 174.7

Utility PrinterHigh 203

High 201

Alarm PrinterRx Bldg. Pressure

12: 59CLR 192

13: 05
13:25

CODE
SAFETYVIA

INFORMATION

Alarm Printer
CLR 193

CLR 193

End of Pressure rise.

Alarm Printer
Alarm Printer

Stable Rx Bldg Pressure
after this time indicate
EMOV remained closed.

EMERGENCY FEEDWATER PUMPS - STATUS

ON
1
STEAM
DRIVEN
OFF

ON
2A

OFF

.ON

2B

OFF

0

2

4

6

10
8
ELAPSED TIME - HOURS

FIG 1.4-3

12

14

16

4.11 Emergency Director Leaves Site
4.11.1

Sequence

According to interviews,

the Lieutenant Governor of Pennsylvania called

the President of the licensee's organization requesting a meeting on thestatus of the plant and related matters.

It appears that an initial meetingtime of approximately noon (8 hours after the start of the accident was established)
but was changed to about 1430 hrs (10-1/2 hours after the start of the accident)
when the Plant Manager objected to the earlier meeting time.
It

(Int.

1, 21,

34)

appears that it was the decision of the Vice President - Generation toincorporate onsite plant management in that meeting.

As a consequence of thatdecision and at some time prior to the meeting, the Unit 2 Superintendent-

Technical Support was directed to start collecting technical materials forthat meeting.
Based on the interviews,

itappears that the Station Manager and theUnit 2 Superintendent - Technical Support left the control room shortly before
1400 hrs.

They subsequently met the Vice President - Generation at the North

Gate (he was previously at the Observation Center).

They then proceeded to

the Lt. Governor's office where the meeting began at approximately 1430 hrs
and lasted 30-45 minutes.
Itfurther appears that the Station Manager and the Unit 2-Superintendent

Technical Support returned to the control room at or about 1630 hrs.
4.11.2Precautions Taken

Prior to leaving for the meeting,

the Station Manager took the following

actions:
reviewed plant status to determine if conditions were stable.

(The

Station Manager stated that he would not have left for the meeting ifthat were not the case.)

1-4-43-

ordered the Unit 1 Operations Supervisor to maintain the status quo as
regards reactor conditions.
appointed the Unit 2 Superintendent Acting Emergency Director.
informed the Unit 1 Superintendent of where he was going.
was provided with a beeper to enable contact while enroute to and fromthe meeting.
arranged for the Unit 2 Superintendent - Technical Support to make contact
with the site from the Lt. Governor's office.

(This was done, and contact

was maintained for approximately 15 minutes while the meeting was inprogress.)
Evaluation
It was not possible for the investigation team to establish the impactthat the absence of the individuals involved had on the course of the accident.
Prudent precautionary steps were taken prior to his departure.

The steps

taken were comparable or in excess of those that would be expected in theevent an Emergency Director were incapacitated in some fashion.
The decision that repressurization and an RCP start should be pursued
occurred after the meeting participants returned to the site and were briefed
on the status of the plant.

Whether this decision would have been reachedearlier, or the attempts to- refill the loops that proceeded during the meeting
would have been abbreviated would be conjecture and thus, outside the scope of
this investigation.
4.12 RCS Average Temperature Indication
Operator interviews indicate they were confused and possibly misled by
the behavior of the RCS Average Temperature (Tav)

1-4-44indications.

A review ofPlant Strip Charts recorded on March 28,

1979,

indicates that Tav "sat for the

longest time" at 570'F; about 11 hours out of the 16 hours reviewed by theinvestigation team.
Tav indication is developed from signals generated by narrow-range RCS
hot-leg and cold-leg temperature instrumentation.

These source instruments

produce minimum to maximum strength signals corresponding to their allowabletemperature range of 520'F to 620'F.

If temperature at the point of measurement

drops below 520'F, the signal delivered to the Tav computation circuit remains
at minimum levels,

implying that actual temperatures stop dropping at 520'F.

If temperature at the point of measurement rises above 620'F, the input to the
Tav computation circuit is effectively clamped at 620'F.

Normally, RCS hot-leg

and cold-leg temperatures will track essentially together, since major changes
in these parameters are usually due to plant heatup or cooldown.

Tav indication

during these normal plant evolutions would not appear anomalous,

although theindication would be in error as soon as either input left the allowable range
for its instrument.

(Int.

56, 61.)

Evaluation
During the accident,

RCPs were tripped and hot-leg and cold-leg

temperatures quickly diverged.

With the hot leg above 620'F and the cold leg

below 520'F, the Tav computation circuit functioned as designed and indicated
Tav at 570'F.

That the operators found comfort in this near normal indication

indicates they were unaware of the limitations of this instrument.
4.13 Plant Computer Records
The investigation team relied heavily on computer generated records
related to this accident to substantiate reported events and their timing, todetermine information of which the operators should have been aware (based oncontrol room annunciators),
NRC.

and to identify plant problems not yet reported to

The records were typed on the Alarm Typer,

the Utility Typer and the Log

Typer and were inked on the four Analog Trend Recorders.

1-4-45

Copies of theserecords were made available to the investigation team following extensive
searches of plant records by NRC and licensee personnel.

They are essentiallycomplete, with the exceptions discussed below.
4.13.1

Alarm Status Printout, 05:13:59 to 06:48:08, 3/28/79

Analysis of Alarm Typer and Utility Typer output prior to and during thisperiod indicates that alarm status printout records for this period,

ifever existed, would cover only the period 05:13:59 to about 05:15:16.

they
The

indication is consistent with an assumption that the operator actuated thealarm suppress function at about 0648 hrs,

thereby dumping the alarm status

printout memory and restarting the output at the current time.

This action ispermissible and would be in the best interest of the operators,

who neededcurrent information,
before.

not the history of nearly one and three quarter hours

Interviews with plant personnel have not determined the actual causeof data loss.

No evidence has been identified that the data was purposely

destroyed or withheld to hamper the investigation.
4.13.2

Alarm Status Printout,

18:48:59 to 19:10:29, 3/28/79

Analysis of Alarm Typer and Utility Typer output prior to, during,andafter this period indicates that alarm status printout was probably generated
within this period,
Alarm Typer,

but may have been so unintelligible as to be useless.

The

on which the alarm status printout was being typed, had a historyof paper jamming, causing overstrikes, tearing, and general havoc with theresulting record.

The record from 19:10:29 to 19:14:24 shows clear evidenceof this condition. Whether the missing records were lost, thrown away as
useless, or otherwise disposed of has not been determined.

Records of alarms

during this period were not critical to the understanding of the accident
sequence and did not significantly impact on this investigation.
4.13.3

Utility Typer Output, 00:00:00 to 03:24:24, and 20:08:22
to 20:12:*, 3/28/79

The records from this typer during the periods of interest, if they everexisted, were not found.
They are not considered critical to this investigation.

1-4-46

*NOTE:4.13.4Paper is torn and time in seconds is missing.
Analog Trend Recorder Number 2,

3/28/79

The strip chart from this recorder has reportedly not been found.
4.14

Pressure Spike in Containment

4.14.1

Sequence

The 28-psig pressure spike in the TMI-2 containment occurred at 13:50:21(approximately 9 hours 50 minutes after the start of the accident) on March
28,

1979 based on data.

(Alarm Printer, Ref. 14)

directed to open the EMOV valve.
controls,

some operators andAn operator had just been

Concurrent with his manipulation of thethe Station Manager heard a "double thump," andthe alarms and automatic equipment functions associated with high pressure in
the containment actuated.
The Alarm Printer shows the following sequence of events:
13:50:21

Channels A&B HPI and building isolation setpoints reached (^-4psig).

13:50:27

Channel A&B high-pressure setpoint (-28 psig) actuated.

13:50:32

High-pressure setpoint (-.28 psig) clears owing to rapid
decrease in pressure after pressure spike.

13:50:46

A channel 4-psig setpoint clears.

13:50:48

B Channel 4-psig setpoint clears.

13:56:07

Spray pumps A&B tripped.

1-4-47

(Itshould be noted that these entries from the Alarm Printer are included

here to establish precise times of events.

The operators would not have beenaware of these entries at the time because the computer was running behind realtime due to the number of alarms being printed.)
The alarm that initiated building spray cleared in 5 seconds.

The alarmthat initiated the other protective functions (lower setpoint) was in thealarm status for a total of approximately 25-27 seconds.
to restore other equipment to operation,

Because of the need

the building spray pumps were not

shutdown until approximately 5-1/2 minutes after they started.
This event occurred shortly after the licensee had reduced reactor pressure
sufficiently to attempt to "float" the core flood tanks on the reactor coolant
system.

The effectiveness of floating the core flood tanks was receiving mostof the licensee staff attention.

Moreover, all personnel

in the control room

were wearing respirators at this time because of airborne radioactivity
problems which caused substantial communications problems.
4.4.2

Response of Personnel to Pressure Spike
Based on interviews conducted during this investigation,

it appears thatthe response of personnel present in the control room to this pressure spike
was varied,

including a total lack of awareness that it occurred.

No statementsthat have been obtained indicate that anyone present postulated that thepressure spike was due to the rapid burning of hydrogen.
One supervisor,

in the area of the controls but not immediately at them,

ascribed the actuation of the building spray pumps to an electrical problem.
Apparently,

the loss of the electrical buses 2-32A and 2-42A, which occurredat 1351 hrs (approximately 30 seconds after the spike) contributed to thisconclusion.

Others attributed the spike to various sources including-electrical interference on the recorder or the switches that actuate building
spray.

1-4-48

Another shift supervisor, who had entered the control room earlier andassumed responsibility for the controls,coincidence with the opening of the EMOV.

saw the pressure spike occur in
(Int.

119)

Based on this latter supervisor's interview, it

appears that he directed

the operators to leave the spray pumps on until he was assured that building
pressure had stabilized and no further pressure spikes were occurring.
he felt satisfied that conditions had reestablished,pumps be shutdown.

When

he directed the spray

He stated that he recommended to the Station Manager that

the EMOV no longer be cycled because the rapidly rising building pressure
corresponded to opening the EMOV_This apparently was brought to the Station

Manager's attention immediately prior to his leaving for the meeting at the
Lt. Governor's office.

Itappears that the incident was not recalled untillate the following day when the building pressure trace was reviewed.
In general,

the investigation has concluded that the staff on duty in thecontrol room did not attach any special significance to the pressure spike.
Those who actually saw the spike related it to opening of the EMOV,
not attach significance to another cause.

It

and didshould also be noted that therecorder displaying the spike displays the trace for approximately 2 hours,
after which the trace must be removed from the recorder or the recorder pulled
out to the extended position in order to be viewed.
NRC personnel were also in the control room during this period.
interviews with those individuals,pressure spike.
4.14.3(Int.

Duringthey stated that they were not aware of the31)

Notification of NRC

It appears based on interviews and log books that the NRC was notified ofthe pressure spike early in the morning of March 30,

1979.

It was referred toin Preliminary Notification PNO-79-67D prepared late that same day.

An inter-view with an NRC inspector indicated he recalled discussing the matter with
the Station Manager on that date,after the licensee investigation team hadreviewed the Plant Strip Charts the evening before (March 29,

1-4-491979).

An interview (Int. 139) of a control room operator indicated that copiesof the RB recorder having the spike were made and in the control room at about
2230*.

The person stated that he arrived at the Observation Center at approxi-

mately 1400 hrs and went to Unit 2 Control Room, wearing a breathing mask, at
about 1445 hrs.

He stated that he was told about a reactor building pressure

spike and after 2000 hrs that night, he stated that there was discussion as towhat the spike was.

He stated that a hydrogen explosion was discussed.

stated that "We pulled itHeout and I don't know who wanted a copy but we made acouple copies of the chart."
He stated that he discussed the spike with a trainee and the person who
asked for the information.

He did not remember if this was an NRC inspector

or a B&W representative.
The investigation looked into this allegation and has concluded that theindividual has confused the events of the evenings of March 28 and 29,

1979.

This conclusion is based on the following information.
The investigator reviewed the original combined wide-range 0-100 psig
(BS-PT-4388-2)

and narrow-range -5-0 +10 psig (BS-PT-4388-1)

shows that, on March 28, 1979,

at approximately 1350 hrs,

chart.

The chart

two peaks occurred.

The narrow range goes off scale and the wide range peaks at about 28 psig.
portion of the charts had been cut out from the original Strip Chart.

AThe

cutout section of the chart was matched to the two adjacent sections andcarefully reviewed for pen disruptions that could be expected if the chart wasremoved on March 28,

1979,

in order to make copies and then returned to therecorder.

The chart has a written notation that indicates "chart removed,

March 29,

1979 at noon."*NOTE:

CRO stated that copies were made and in the control room after

the establishment of a bubble in the pressurizer.

This would establish atime of 2218 hrs or later based on the CFR log, March 28, 1979.
*CRO stated copies were made in the control room after the establishment of a bubble
in the pressurizer.
This establishes a time of 2218 hours or later, based on the CRO Lo
March 28, 1979.

1-4-50

The cut section has an early cut line at 0200 hrs March 28,
late cut line at 2215 hrs.
29,

1979,

The chart runs out about 1200 hrs (noon)and aon March

1979.
The possibility that copies of the reactor building pressure trace weremade on March 28,

1979,

appears possible but unlikely.

had no obvious pen disruption to confirm this.
tions as to whom he talked to and when.

The reviewed originals

The CRO later expressed reserva-

NRC personnel do not recall a xerox

copy in the control room on March 28, 1979.
Further discussions were held with licensee and GPUSC technical staff who
were identified as having knowledge of the copying of the pressure trace.

A

GPU staff member stated in a discussion that he was not in the control room on
March 28,

1979.

He did get copies of the reactor building pressure spikearound the 29th or 30th of March.

He also stated he made copies for NRC

staff.
Based on these findings, itappears the individual who expressed the viewthat the pressure spike was distributed at the site on March 28,
mistaken,

1979 wasand that the actual analysis and review took place on March 29,

1979

as previously mentioned.
4.15

High Temperatures and Core Response

4.15.1

GeneralThere were four indications available to plant personnel to indicate thatthe core was not being adequately cooled:o

SRM and IRM anomalous indications.

oOffscale (greater than 700'F) incore thermocouple readingsoHot-leg RTD readings that indicated superheated conditions of
the coolant

1-4-51oHigh radiation readings in the containment and elsewhereindicating a large-scale release of radioactivity to the RCS.

The onset of significant release of activity from the fuel because of
inadequate cooling cannot be fixed in time.
charts for the hot machine shop,

Review of radiation monitor

incore detector cabinets, and the reactor

building dome monitor indicates fuel failures occurring prior to 0630 hrs.
Basically two failure mechanisms can be postulated for release of activity
from the fuel.
The first failure mechanism is associated with the pressure differential
across the cladding and is related to the pin compression factors.

Pin compres-sion factors appear to have been violated within the first 20 minutes of theaccident.

Technical Specifications do not address pin compression factors.

The temperature-pressure relationship as experienced in the transient areplotted in Figure 1.4-4.

Review of activity levels in the RCS as monitored bychemical analysis and radiation monitors does not indicate untoward releasesof activity prior to core becoming uncovered.

Lower pressures and higher

temperatures were experienced during the period when core became uncovered.
There is a likelihood that there was some mechanical failure of the cladding
during this period.
There is almost virtual certainty that a zirconium-Water reaction took
place based on the release of hydrogen from the RCS.

The temperature design

limit for loss of coolant accidents is 2200°F peak clad temperature (Ref.
138).

Temperatures of this order were indicated by the incore thermocouples

at approximately 0900 hrs after the core was reflooded.

It

is thought that

this temperature was reached earlier when the core was uncovered, although
there is no direct evidence of this.

Significant structural damage to thecore is thought to have occurred when an RCP was started at 0654 hrs.

Whenthe pump was started, a rapid quenching of the core is indicated by the SRM
and IRM response addressed below.

In addition, the core reflood was attended

by a rapid pressure change.

1-4-52HEAT UP/COOL DOWN CURVEHIGH PRESSURE
REACTOR TRIP

2400 -

POINT TEMP OF
A
82
B
157
C
275

2300 E2200

D2100 -E
2000 -

PRESS
PSIG
155
476
476

280

1319

3..89

2225

1. Instrument error 50 psig - 25
assumed on TECH spec - 25
for installed recorder1900 1800

2. Minimum RC pressure to
maintain compression force on
clad (natural circulation) inst.
error +50 psig -12°F

-

1700-

3. Minimum RC pressure to
maintain compression force on
fuel (forced flow) inst. error+60 psig -121F

1600 -

4. Minimum pressure for control•
rod drive operation inst. error
10+50
psig -12'F

1500 1400 --5. Minimum RC pressure forsingle pump in a loop NPSH.

1400

-(1

CD

1300 -/0, 1 /1, 2/1)inst. error +50
psig -l12'F
6. Minimum RC pressure for two
pumps in a loop NPSH. (2/0,0/2) inst. error +50 psig -12'F

10
I

3

2 I

1

D
1100 -

0LuJ

THIS POINT

1000-

0
•

REACHED 17 MINUTESAFTER REACTOR TRIP

900800 700 600 500 -

B

C

4

400300-

200 -

6

A

100RCS AVERAGE TEMPERATURE, 'F.
100

200

300

400

500

600

COMPARISON OF RCS TEMP/PRESS.
BEHAVIOR TO LIMITS (1ST 17 MIN.)

FIG1.4-4
1-4-53

4.15.2

SRM and IRM Responses

Within 45 minutes of the reactor trip, SRM Channel 1 was showing ananomalous response.
0541 hrs,

From this time until the last two RCPs were tripped atincreased radiation transport due to reduced density in the radiation

path from the core to the detectors is evident.

An increase in the indicationof IRM Channel 3 noted during the latter part of this period is due to thesame effect.

When the last two pumps were tripped,itis felt that many of

the steam voids collapsed in the radiation path; this is noted by the increased
attenuation (decreasing signal) of the SRM and IRM.

The signal levels thenincrease, and it

the core became uncovered,is believed that, prior to 0600 hrs,

although the exact time cannot be determined.

The exact level is also unknown.

At 0654 hrs, on RCP was started and, because of a sharp increase in attenuation
observed on the SRM and IRM,
place.

After this time ititis believed that a reflood of the core took

is believed that the core became uncovered again.

This condition was terminated at about 0715 hrs when substantial HPI flow wasestablished and the effect of increased attenuation is noted.
0745 hrs, an increased signal is noted on the SRM trace.
increase is unknown but it

At approximately

The cause for thisis postulated that the change is consistent with acore geometry change.
4.15.3

Incore Thermocouple Response

Chromel-Alumel

thermocouples are located in 52 selected fuel assemblies

and are positioned approximately 6 inches above the active fuel.

At

04:33:13 the following computer alarm was received:
BAD 0538 IM INCORE T/C 10-R TEMP-???
The cause for this alarm is unknown.

This detector is located in aperipheral fuel assembly and thus is at a relatively low power level.
time, core flow is relatively high and it
geometry,is doubtful,

At thisassuming normal corethat the alarm is indicative of an uprange (>700'F) temperature.

1-4-54

Incore thermocouple (T/C) data from the Alarm Typer does not exist for
the time period prior to 06:48:08 except as previously mentioned.

A list ofincore T/C values as output by the process computer are listed in Figures
1.4-5 to 1.4- 8.

These values cover the period from 0655 hrs to 0927 hrs.

This figure indicates values obtained on normal scam of the computer.
From 8:34:07 to 8:36:40,
from the computer,sixteen incore T/C temperatures were requested

probably by the Instrumentation Engineer.

possible to ascertain exactly who requested the data.

It has not been

In addition,from

08:46:44 to 08:47:58, eleven incore T/C temperatures were requested, again
probably by the Instrument Engineer.

These are shown in Figures 1.4-9 and 10.

A comparison of the data requested shows that, for locations that are duplicatedin the requests, there is an indicated temperature drop between the time thedata was first obtained and the second time the data was requested.
0800 hrs and 0900 hrs,
room.

Betweenreadings were taken at the computer input in the cableThis data was obtained using a thermocouple reader and digital voltmetermeasurements with suitable conversion to temperature.
in Figure 1.4-11.

This data is indicatedA review of this data compared with the data output between

08:46:44 and 08:47:58 reveals, an indicated temperature drop for common locations
to the two sets of data.

This fact would indicate the measurements at thecomputer input were taken after 8:47:58.

Review of this data indicates signifi-

cant flow blockage in the center of the core.

Itshould be noted when

examining Figure 1.4-11 that the values are approximate because corrections
were not made for the reference junction (,\,
750F correction).
4.15.4

Incore Flux Detector Response

The reactor core contains 364 self powered nuclear detectors (SPND)

with

7 detectors equally spaced in 52 strings.
A review of backup recorder traces of incore flux detectors (SPNDs)
outputs indicate a normal response to the reactor trip at approximately 0400
hrs.

At approximately 0640 hrs,

measurable output.

some of the detectors were producing a

By 0700 hrs, the SPND's at locations F-3, G-5,

1-4-55and K-ll,level 4 are off-scale.

Shortly after the SPND at location H-8 level 2 was

off-scale.

the SPND at H-8 level 6 is off-scale. By 0830 hrs,

At 0810 hrs,

the SPND's at levels 2 and 4 of H-9 are off-scale.

The output signal of the

SPND's is believed to be caused by thermionic emission by the rhodium wire in
the detector. Measurements taken based on the current ouput of the detectors
prior to the trip indicate an off-scale current of approximate 2000 nanoamperes.
Other detectors exhibiting on-scale outputs are:

Location

Level

H-8

4
2
4
2
6
6

G-11
G-11
C-1l
K- 11
M-9

6
6
6

F-3
L-13
K-5

4
6

F-13
M-7

6
6

K-5
E-7

6
6

F-13
L-13

It should be noted that the SPND at location H-8 level 6 remains upscale
until approximately 1315.

(The SPND at location L-13 level 6 remained upscale

until the afternoon of March 29.)It

is believed that the core was uncovered

to a depth of at least 9 feet below the active fuel height based on the responseof the level 2 detectors previously referenced.

1-4-56

1

2

4

3

,5

10

9

6

t1

13

12

14

15

A

C

= I 1 . 1 1 1 1 1 -1

DI
-4~4~4-~-4 -4.-----/.4-4-4--.1-.4--4--

4-4-4-

NOTES:

<2;

F
,~

,.

(1) "?' indicates value
outside of computer
software range (700°F).

//

Time shown is actual
Turbine trip
time.
occurred at 04:00:37.

(2)
L

H

1-I-

It

N

. I. . I . . II .

0

P
-4

4-E

2-G

LOCATION:
Time

Temperature

06:55:40
07:16:40
07:21:40
08:36:40
08:48:39

623
630.3
518.6

( 0 F)Time

Temperature

07:03:40
07:09:40
07:10:10
07:10:40
07:11:10
08:36:38681.9694. 1
643.9
309.7

(OF)

Time

Temperature

07:43:40
07:44:40
08:05:40
08:31:10
08:36:22
08:42:00
08:45:09

680.2

FIGURE I.4.-5 SELECTED INCORE T/C HISTORIES FROM COMPUTER PRINTOUTS

610.3

699.3

('F)

1

2

3

4

5

a

7

-T-7

A

'7

3

9

10

11

12

13

14

15

-72L.:CID

I

NOTES:

F

(I)

"?"

indicates value

outside of computer
software range (7001F).
K

(2)

L

=U

Time shown is actual
Turbine trip
time.
occurred at 04:00:37.

I I I I I I' I

0
4. -

-

4-

I. -

4-

I- -

.4.-

4- -4

-

4-.

0,
K

4-N

LOCATI ON:
Time
07:23:41
07:27:41
07:29:11
07:30:41
07:33:41
08:03:11
08:06:41

5-D

Temperature

(OF)

693.4
658.2
695.3

FIGURE 1.4-6

23.7

SELECTED

Time
07:03:40
07:05:40
07:12:40
07:13:10
07:23:40
07:44:40
08:36:15
08:47:58

Temperature ('F)
699.5
664.2
666.1
645.0

Time
06:55:10
07:15:40
07: 21:40
08:36:00
08:47:43

INCORE T/C HISTORIES FROM COMPUTER PRINTOUTS

7-B
Temperature
623
627.7
436.3
377.6

(OF)

1

2

5

4

3

7

6
f

I

a
I

9
I-rI

10

11

12

-,

-

13

14

15

A
-

I -

I -t

-

-

I-

-

D
C
D

111(11111111 IZEIJIIItIF

NOTES:

H
C

(1)

''?'

indicates value

outside of computer
software range (700 0 F).

U

(2)
L

-- FillTime shown is actualtime.
Turbine trip
occurred at 04:00:37.

11 V-A, 11

0

T'

o

Time
06:57:11
07:20:41
07:21:11
07:45:41
07:51:41

FIGURE

10-C

9-N

7-R

LOCATION:

Temperature
679
?
630.5
?
641.0

T.4-7

(T'F)

Time
09:11:39
09:12:09
09:16:09
09:25:39
09:26:39

Temperature
611.7
?
639.0
?
675.4

(OF)

Time
06:55:40
07:19:10
07:21:10
08:35:38
08:47:19

SELECTED INCORE T/C HISTORIES FROM COMPUTER PRINTOUTS

Temperature
599
?
637.9
465.1
398.9

(7F)

I

2

3

5

4

6

7

8

9

10

11

12

.13

14

13

A
B

iTI1~

0
K

i Ii~LIIILL

I liii1K

7
*0

-4

-4-4-

+ -

-

-

4 -4-

*-

4-

-

4-

4--

NOTES:(1) "?" indicates value
outside of computer
software range (700 0 F).
(2)

-4-

L
/,.'

H

Time shown is actual
Turbine trip
time.
occurred at 04:00:37.

'7

i E

H
0
-4-

+ -

-

-4-4-4-4-------

4-4-

-

P
Kt

Time
07:32:10
08:15:10

FIGURE

13-G

12-F

10-M

LOCATION:

Temperature
689.0
693.2

1.4-8

( 0 F)

Time
08:06:10
08:15:10
08:15:40
08:34:44
08:46:51

Temperature
694.9
?
588.5
457.1
393.1

()F)

Time
06:57:1Q,
07:06:10
07:06:40
07:19:40
07:21:40
08:01:10
08:34:37

SELECTED INCORE T/C HISTORIES FROM COMPUTER PRINTOUTS

Temperature
670
?
675.6
?
682.1
?
533.0

(OF)

1

2

3

4

5

6

7

8

9

10

11

12

13

14

'A

B

436.4 511iC

7641 .91

465.1

iiI 1 I I1__ 1 693.OJ I_IJI

D

I_ ?

_

2

E

f57.1

309.7
F

518.6

533.0

H
K

653.0

480.9_

M
N

0
4-

I

I

I
R

NOTE:

S(.?

FIGURE I.,4-9;

TEMPERATURES IN 'F

INDICATES OFFSCALE)INCORE THERMOCOUPLE DATAFROM COMPUTER AT 08:34-08:36
ON MARCH 28, 1979

1-4-61

15

2

i

3

4

5

6

7

I

A

__1J ?j

D

10

9

11

377.6 9-418.6
?

561.4

__1

9-

4-

9

4

4

_

_J582.7JIj

?

393.1

?

1

K

L

H
wý,
_____

[Ill~_____ I_____
I_____Ji 1111
_____

j

4-

.1

+

14

398.9

F

H

13

377.6 418.6

E

G

12

I

9

C

8

4

4

4---------I

I

A-

S

4

1-

R

NOTE: TEMPERATURES IN 'F("?" INDICATES OFFSCALE)
FIGURE L.4-10:

INCORE THERMOCOUPLE DATAFROM COMPUTER AT 08:46-08:47
ON MARCH 28, 1979

1-4-6215

" A"'

HOT LEG
1

2

3

4

5

6

7

8

280

300

9

10

960

300

11

12

13

14

15

-'-v-i-,.-

A

NI-1
B

NI-5

0

C

0
470

"L E
D COLD
LEG

1200

j50

600
90

F

H

2277 1774

260

0

370

/

R

550

1852
630

1500

382

1700

300

1780

123271 350

2178

260

I I4101 598 [21671

COLD LEG

0

460

350

0

"'B"S
COLD LEG
1100

0

310

0

300

N 1-4

350
-

i -

475
. -.

I -

"B"
HOT LEG

A -

NI-6

NI-2

V",\

CORE ARRANGEMENTFOR TMI-2

READINGS OBTAINED FROM INCORE
THERMOCOUPLES AT APPROXIMATELY 4-5 HOURSAFTER START OF ACCIDENT

FIG 1.4-11
1-4-63/COLD LEG

230

12952176

320

NI-8
P

1876

1811

L

N

330
1855

2452

K

M

330

2366 2378

380

G

Z580

1980

220

0

4.15.5

OTSG and Primary Temperature Recorder Display

Various primary and steam generator temperatures are recorded on a
24-point dual-range recorder on panel 10 in the west side of the control room,
behind the console facing the center of the room.
ranges are listed in Table 1.4-5.
coolant temperatures,

Of particular interest are the six primary

which comprise one point for each of the hot-legs andone for each pump suction.
range.

The point designations andAll of these are displayed with a 0 F to 800°F

All of the six remained within range,

except for several 3- to 6-minuteperiods, during the entire incident.

A brief description of the behavior offour of the traces is given below.

This is based on a review of the originalrecorder trace and which are shown on Figures I-4-12A, B, C, and D for theperiod 0400 hrs to 2030 hrs on March 28, 1979.

The original chart 'ismulticolored, and traces outside the period given above were used to correlate
the point numbers with the colors.

All times are given in hours and minutesof elapsed time after the start of the accident.
4.15.5.1

Hot-Leg Temperature Behavior

Both of the hot-leg temperature indicEtions were in excess of saturation
temperature a few-minutes after the A loop pumps Were tripped at 101 minutes.
The A loop temperature increased from 540OF to 800OF over the next hour, going
off scale briefly at about 0648 hrs.
The temperature rise was interrupted by a temporary reduction of 40OF at
the time that the EMOV was closed at 0619 hrs.
about 10 minutes later.

A similar reduction occurred

The B loop temperature rose more slowly, going off scale

briefly at about 0700 hrs.

During the rise, the temperature peaked sharply at

0634 hrs then dropped 60OF over the next 10 minutes.
Both A and B loop hot-leg temperatures had sudden reductions (of 60'F and
30'F,

respectively) when the attempt was made to start RCP-2B at 0654 hrs.

Another sharp drop of 90 0 F occurred in A loop temperature when ES was initiated

1-4-64

by the operator at 0720 hrs.
For the next 6 hours, both hot leg temperatures
remained relatively stable, with A loop varying between 680'F and 760'F and B
loop tracking A loop about 60'F higher.
The ES actuation resulting from the pressure spike in the reactor building
at 1350 hrs caused another sharp drop in both hot leg temperatures of about
50 0 F.

This may have been caused by cooling by the building spray, which was

actuated by this ES actuation.

Following this, the A loop began an irregular

downward trend reaching 4500F by 1510 hrs.

Large swings of between 470OF and

600'F occurred during the period of this trend.

A drop of 150'F starting at

1428 hrs was also indicated by the A loop strip chart recorder; it was attributedby the operators to increased makeup water injection through MU-Vl6C.

However,the start of a sudden increase of 85'F coincides with the start of MUP-lC at
1432 hrs, and this increase is terminated at the time that MUP-lC is secured.
A similar rise coincides with the start of MUP-IC at 1519 hrs.
terminates about the time MUP-IC is stopped at 1128 hrs.
(1350 hrs to 1630 hrs),

This rise also

During this time

B loop hot-leg temperature remained generally constant

above 700'F.
Both hot-leg temperatures began a drop of about 120'F when the operator
opened the EMOV block valve at 1635 hrs.
This was followed by a return to theprevious conditions for the next 2 hours,with A loop hot-leg between 550°F

and 590°F and B loop between 720*F and 760 0 F.
The momentary start of RCP-lA at 1933 hrs caused another sudden decreasein both A and B loop hot-leg temperatures of 150*F and 100'F, respectively.
When this pump was finally restarted at 1950 hrs and left running, all of the
primary temperatures rapidly converged at 350 F.
4.15.5.2

RCP Suction Temperature Behavior

The RCP-suction temperatures began a gradual decrease after the RCPs were
tripped. They were less responsive to system changes than the hot-leg temperatures,
but several trends will be noted.

RCP-2B suction temperature began diverging

1-4-65

from the rest of the primary temperatures at about 0540 hrs, either shortly
before, or at the time of, the A loop RCP shutdown.
RCP-2A suction temperature
did not begin diverging until 0600 hrs.
Both RCP-2A and RCP-2B suction temperatures began a rapid decrease intemperature at 0720 hrs, the time that ES was manually initiated by the operator.
There was a rise in both pump suction temperatures when MUP-IC was shut off at
0737 hrs.

However,

RCP-2B suction temperature rose only 300F while RCP-2A
suction temperature increased about 180 0 F. The pump suction temperature had

decreased to around 2000F by 0900 hrs,

and,in general,

remained around thislevel until about 1430 hrs.
At 1430 hrs, the RCP-2A pump suction temperature began a long trend
upward,increasing from 1700F to 4900F by 12 hours 10 minutes.

This trend wasinterrupted at 1510 hrs, by a 200 F rise over a several minute period, that.
corresponds to a simultaneous pressurizer level drop of nearly 200 inches.
Alarge decrease in the RCP-2A suction temperature occurred after 1723 hrs, when
the operators turned on MUP-IC to begin repressurizing the reactor. A final
sharp drop in this temperature occurred when the RCP-2A was bumped at 1933hrs.

1-4-6,6

TABLE 1.4-5
STEAM GENERATOR AND
PRIMARY SYSTEM TEMPERATURES
Point
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
24

Service
SG IA Outlet (Secondary)
SG IA Outlet (Secondary)
SG lB Outlet (Secondary)
SG lB Outlet (Secondary)
RX Coolant to SG lB
RX Coolant to SG 1A
SPARE
RC Pump IB Suction
RC Pump 2B Suction
RC Pump IA Suction
RC Pump 2A Suction
DH Coolant from RX DH Pump A
DH Coolant from RX DH Pump B
DH Coolant from DH Cooler A
DH Coolant from DH Cooler B
FW to SG IA
FW to SG lB
SPARE
Letdown Cooler Outlet A
Letdown Cooler Outlet B
Letdown Cooler Inlet
SPARE
RX Coolant Makeup Tank
Mechanical Draft Cooling Tower

Range
0 0 0 0 0 0 -

800
800
800
800
800
800

0 0 0 0 0 0 0 0 0 0 -

800
800
800
800
400
400
400
400
800
800

0 - 400
0 - 400
0 - 800
0 - 400*
Not Listed**These are currently printing 0 - Operator says that they are not used.

1-4-6 7

800
STEAM
REACTOR COOLANT TOSTEAM GENERATOR 1B
-

-

-

-

I

LL.
0,

I

LUJ

Lu

RCP28

IL
co

Iý-

00L

0

2
TIME AFTER TURBINE TRIP (HOURS)

3

SELECTED PRIMARY SYSTEM TEMPERATURES FOLLOWING ACCIDENT

FIG 1.4-12A

4

-•

--

.

600

---

--

---

" •-- --•REACTORCOOLANT TOSTEAM GENERATOR 1B
REACTOR COOLANT TOSTEAM GENERATOR IA

LL0

LJ

i,-

RCP 2B, SUCTION200
RCP 2A SUCTION

0

4

5

6
TIME AFTER TURBINE TRIP (HOURS)

7

SELECTED PRIMARY SYSTEM TEMPERATURES FOLLOWING ACCIDENT

1.4-12B

8

dPhAAIMINIM.

r

uuu

-.

-.

-2

600

REACTOR COOLANT TOSTEAM GENERATOR 1B

REACTOR COOLANT TOSTEAM GENERATOR 1A

/

LL

0

LLa

4O0

I

I
!

00'4,-i-0

I
I

RCP 2A SUCTION
200

/
m

-

--

-

-

-

=

-

-

-

~

~

~ ~ N

,c-.-

_

1~~

-.--

*~

9

10
TIME AFTER TURBINE TRIP (HOURS)1112SELECTED PRIMARY SYSTEM TEMPERATURES FOLLOWING ACCIDENT

1.4-12C

-"

-._-

,

REACTOR COOLANT TO\STEAM/

GENERATOR 1A

u 400\"'"-

RCP 2A SUCTION200
RCP 2 BSUCTION

0

12

13

14
TIME AFTER TURBINE TRIP (HOURS)

15

16

SELECTED PRIMARY SYSTEM TEMPERATURES FOLLOWING ACCIDENT

1.4-12D

4.16 Early Void Formation
The Operational Sequence of Events,

Appendix I-A to this report, will

identify the conclusion of the investigation team that early formation ofvoids appears to occur in the reactor coolant system during the 1 to 4-minutetime frame.

No assertion is made as to the nature of these voids because

objective evidence is unavailable to distinguish whether they are distributed
voids in the circulating reactor coolant or discrete voids forming in thereactor vessel or upper portions of the reactor hot-legs ("candy canes").
This conclusion was reached by doing a mass balance on the reactor coolant
system as indicated by RCS Temperature,pressure and recorded pressurizer

level at selected times during the 1 to 4 minute time interval.
data at 1 minute as the basis for comparison,

Using thethis mass balance would indicate

the mass that would have to be added to the RCS, assuming no voids other thanin the pressurizer, to give the indicated levels.

This required mass addition

was then compared to the upper estimate of the mass flow rates being added tothe core as a result of operator and automatic actions as determined frominterviews and objective evidence.
The results are shown in Table 1.4-6 and suggest a mass 'discrepancy ofapproximately 4,000 pounds.

Thus the indicated mass is greater than thecalculated maximum possible mass by this amount.
mass loss through the EMOV during this time (ifis indeterminate),

If one includes the probable
the EMOV was fully open which

the calculated mass discrepancy would increase to not morethan 9,200 pounds.
It was therefore concluded,

because of the magnitude of the discrepancy,that the system was not, in fact, solid and was already experiencing significant void formation of an indeterminate nature.
Some additional support for this conclusion comes from the fact that theobserved RCS heatup rate during this period is low when compared to that
expected for the decay heat load at this time after shutdown and the makeup

1-4-7.2

TABLE 1.4-6
PRESSURIZER LEVEL RESPONSE DURING 1-4
MINUTESTime

Makeup Pump
Letdown Status

Press'erLevel

60 secIA &IB on - max

Apparent
Mass
Addition*

Possible
Mass
Addition*

494,000

5,000

3,600

578°F

503,000

9,000

10,1005790F

518,000

15,00011,500Volume (Ft3)

RCS
Pressure

RCS
Tav

RCS
Mass (lb)165"

10,9381800 psig

576 0 F

489,000

lA&lC full on- 1000 gpm;
letdown secured.

210"

11,0951720 psig

578°F

195 sec

IA & IC on
Operator;
Starting tothrottle (lessthan 1000 gpm);
letdown secured

290

11,3461600 psig

278 sec

IC off; IAthrottled(less than
500 gpm)

370"

11,6561450 psig

RCS Liquid

Flow N420 gpm(flow thru 16B& MU Line onlyor letdown
secured)122 sec(ECCS)*This does not include mass loss out of EMOV during this period,
approximately 1500 lb/minute.

1-4-73

which,

if fully open, should beflow rates.

This temperature discrepancy during the same period could be

accounted for by the energy required to achieve a change in state of 4000-9000
pounds of water over a 3-minute time period.
The apparent stabilization of RCS pressure at 1750 psig (although notconfirmed by all the pressure monitoring channels) would also suggest flashingor void formation resulting in momentary maintenance of RCS pressure.
4.17 Diesel Generator Inoperability
At 0402 hrs,

the first automatic ES actuation was experienced.

Dieselgenerators DF-X-lA and DF-X-lB started, but did not load, since the emergency
buses continued to be powered from offsite sources.
At 0430 hrs,

(Int.

3,

15,

189,

195)

the diesel generator fuel racks were manually tripped, by anoperator dispatched for that purpose, after the diesels had operated unloadedfor 28 minutes.
NOTE:

(Int.

189,

195; Ref. 14)

The following is extracted from Unit 2 Operating Procedure 2104-6.2,"Emergency Diesels and Auxiliaries," Revision 9:"2.1.1.4

Do not run the Diesel Generators unloaded for extended periods.
Unload operation causes carbon deposits in the exhaust system
which cause exhaust system overheating as the carbon burns off
when full load is resumed.

Unloaded operation for periods of

one half hour can be considered completely safe if they arefollowed by a period of operation under load.""4.10.6

The Unit can be shutdown after the ES Actuation has been cleared.
Safety Injection - Bypassed.

NOTE:

R.B.

Isolation & Cooling - Defeated."

Following an automatic diesel start on an ESF actuation, diesels

can only be shut down locally by an operator tripping the fuel racks.
(Int.

189,

195; Ref.

148)

1-4-74

At about 0930 hrs,

an engineer noted the energized overspeed trip alarm(indicating the fuel racks were still tripped) and questioned a Shift Foreman
relative to the status of the diesel generators.
In that condition, thediesel generators would not start automatically on ES actuation,

loss ofoffsite power or manual start initiation.

place the Unit

It was agreed to (1)2 Control Room Emergency Standby/Maintenance Exercise switch for each diesel
in the Maintenance Exercise position (thereby defeating its automatic startcapability) and (2)to reset the diesel fuel racks (thereby regaining manual

start capability).

An operator was dispatched to reset these fuel racks,

which occurred at 0949 hrs.

From this point on, the diesel generators could

only be started manually.

These actions reportedly solved the Shift Foreman's

concern for unnecessary exposures of operators to high radiation fields.
These exposures might be necessary to shut down diesels that start unnecessarily
and run unloaded,should subsequent ES actuations occur with offsite power

available.

195,

NOTE:

(Int.

200; Ref.

14,

148.)

The following is extracted from Unit 2 Operating Procedure 2104-6.2,

Revision 9.
"2.2.3

The requirements of Technical Specifications 3.8.1.1 or
3.8.1.2 must be met depending upon the mode of operation.
In modes 1, 2,

3 and 4, two separate and independent

diesel generators each with:

A separate day fuel tankcontaining a minimum volume of 500 gallons of fuel,a fuelstorage system containing a minimum 19,000 gallons of fuelin DF-T-2A and 19,000 gallons of fuel in DF-T-2B,
separate fuel transfer pump must be operable.

and a

In Modes 5

and 6, one diesel generator and its associated fuel oil
system msut (sic) be operable.

If the above requirements

can not be met, proceed as outlined in the applicable
Action Statement.""4.10

Diesel Generator - Automatic Start upon Engineered Safety
Features Actuation.

1-4-75NOTE:

Unit must be in Emergency Standby and reset for

auto start to occur."With the exception of the ES actuation at 0402 hrs,
actuation (0720 hrs, 0756 hrs,

0807 hrs,

0924 hrs,

no subsequent ES

1350 hrs,

1950 hrs) shows evidence of a diesel generator start.

1933 hrs,

andThis evidence would bein the form of indications of diesel generator room air compressor starts andstops, diesel generator oil pressure alarms and indications of diesel generator
shutdowns in accordance with operating procedure 2104-6.2, Revision 9.
Evaluation

9,

Operating Procedure 2104-6.2, "Emergency Diesels and Auxiliaries," Revision
requires that the diesels be operable while in Hot Standby; and that ifthat requirement can not be met, then a degraded mode of operation is allowed.
Licensee operators willfully defeated the automatic start function of theemergency diesel generators, when loss of site power was still possible andpower for engineered safety features was still needed.
Since the diesels
could have been operable, had the operators not taken the actions described,
this is under consideration as a potential item of noncompliance.
4.18.

Reactor Coolant Makeup Pump Starting Problems
During the course of this accident,

the operators reportedly experienced

multiple problems with starting reactor coolant makeup pumps (MUPs).
At 13 seconds into the event, the computer alarm status printout indicatesan operator unsuccessfully attempted a start of MUP-lA.
the event,At 39 seconds into

the alarm status printout indicates an operator again tried tostart MUP-IA and was successful.

This sequence of events is confirmed byoperator interviews, which further indicate that more than one unsuccessfulattempt to start MUP-IA was experienced.
A review of the computer generated
sequence of events for the period of interest appears to show a condition ofcontact bounce related to the control switch disagreement relay.

I-4-7F

Assumingthat the Burns & Roe Drawings for this pump and its related components represent
actual wiring, that computer indications of changes in electrical contact
position are good, and that the memory of two separate operators is correct;
focuses attention on the control room switch for MUP-lA, whose malfunction
alone could defeat a manual or automatic start, fail to indicate additional
operator start attempts, and provide the apparent anomalous indications to thesequence of event program.
At 12 minutes into the event,the alarm status printout and the sequenceof events review present similar indications of a start failure followed by astart success.

Operator interviews *confirm an event like this occurred, but

their memory of timing does not agree with the computer listing.
At 4 hours 18 minutes into the event, the alarm status printout and thesequence of events review present an additional example of these indications;with the exception that, at these times,

the pump does not start.

Operator

interviews confirm these events and indicate the control room switch was thenplaced in the pull-to-lock position, thereby defeating MUP-IA capability tostart automatically and enabling MUP-lB to take its place.

The alarm status

printout and sequence of event review indications on the following automatic
ES actuation, 20 seconds later, supports that testimony.
MUP-IA was left in pull-to-lock for the rest of the 16 hours reviewed by
the investigation team.

MUP-IA has reportedly been run since the event,indicating the problem may be intermittent.
At 15 hours 33 minutes into the event, the alarm status printout indicatesan operator was attempting to start MUP-lC as an automatic ES actuation occurred.
The pump did not start; but the alarm status printout indicates it was manually
started by an operator 6 seconds later.

The failure of MUP-lC to start during

this ES actuation may be indicative of an example of a similar malfunction of
the same type switch on another pump; but could also be related to the malfunctionof several other components,

such as a low lube oil pressure switch or one ofseveral relays, which are not bypassed by the ES actuation signal.

1-4-77

Evaluation
These switches are replacements for switches of the same type andmanufacturer,
failures.

of earlier vintage,that had experienced contact-cam-follower

These switches are standard items used throughout the plant, butwhich are uniquely modified for the MUP application above.

(Int.

61,

195;

Ref: 11, 136)
The information developed by the investigation team relative to thisproblem has been passed to the licensee and NRC management to ensure theultimate resolution of the problem.
During discussion with the licensee, the licensee indicated the problems
with starting MUP-lA had been determined.

The licensee investigation indicatesthe ammeter for MUP-lA was found stuck downscale and that the MUP had, infact, started.

The operator had seen the ammeter downscale and had secured

the MUP.
Analysis results of computer Alarm Typer records from March 28, 1979,
challenge the credibility of the licensee's determination.

The Alarm Typerrecord indicates the MUP-lA power supply breaker did not close at 04:00:05 and
08:18:53 during operator attempts to start the MUP.

Clearly, ifdoes not close, the ammeter will not register current flow.

Itthe breakerappears thelicensee has not completed an adequate investigation of this matter.
4.19 Operational Usefulness of Computer Alarm Status Printout
The usefulness of the alarm status printout to the operators during this
accident transient appears limited.

This resulted from the computer typewriter

combination inability to keep up with real time and the inaccuracies andmisinformation that the output contained.

1-4-78

4.19.1

TimelinessAt the beginning of this accident, the Alarm Typer output was in close
agreement with real time.

By 0646 hrs, the Utility Typer, then outputting the
At 0648 hrs,

alarm status printout, was running 1 hour and 33 minutes behind.

the Utility Typer output of the alarm status printout was brought in synchronization with real time by equipment failure or operator action.

By 1315 hrs,

the Alarm Typer, then outputting the alarm status printout, was running 2
hours and 39 minutes behind.

Had the alarm status printer memory not been

lost at 0648 hrs, the lag in event information would have been in excess of 4hours, since the lags are cumulative.

The information presented in the alarm

status printout is useful as an historical record, but lacks timeliness for
operators to utilize it

for current decisions.

During interviews,

the operatorsstated that they would use the alarm suppress function to wipe out historical
if they needed real-time alarm information for current decisions.

data,

189; Ref: 14,
4.19.2

(Int.

136)

inaccuracies

Analog instruments are scanned by the computers at various frequencies,
varying from once per second to once per minute,and, supposedly,

depending on group assignment

importance or expected rapidity of output change.

of this scan are stored in assigned memory locations.

The results

Computed values arerecalculated using data currently stored in memory locations every 15 or 30
seconds, again,

depending on group assignment.

The resulting digitized analog

values and the host of open or closed contacts are then scanned once persecond for alarm conditions and changes of contact state.
inaccuracies are ones of association.

The resulting

Alarm conditions and corresponding

analog values are recorded at indicated times that are long past the time when
the transient parameter was actually at that point.

Further,the sequential

nature of the alarm status printout implies event sequence, when near concurrent
events may not have actually occurred in the order presented due to the low
frequency of the alarm scan.

These errors are not apparent in slow transientsbut occurred frequently during this accident.

1-4-79

4.19.3

Misinformation

In researching anomalous indications in the alarm status printer output,
it became apparent that certain indicated events could not have happened,
could not have happened in the indicated sequence,
but were not.

or should have been indicatedSome of the problems indicated in Table 1.4-7 were known by

licensee personnel; but the majority were not, having only recently been
confirmed by the licensee,

following inquiries by the investigation team.

1-4-80TABLE 1.4-7
Computer Output Errors
Computer Point
Designation

Point
Descriptor

AssociatedProblem

3092
3093
3094

Condensate Booster Pump CO-P-2A
Condensate Booster Pump CO-P-2B
Condensate Booster Pump CO-P-2C

Pump "TRIP" information
is available only
in automatic or whilestarting in manual.

3156

Diesel DF-X-IA Lubeoil Press. Low"NORM" indications means
Diesel DF-X-lB is tripped.*

3159

Diesel Gen DF-X-lB Fault

Associated with Diesel
DF-X-lA Lube Oil Pressure
Switches.
Why it indicatesat all during this transientis unclear, since its twin,
Diesel DF-X-lB Lube Oil
Pressure low alarm, isnever received.

3172
3173
3174

RC Makeup Pump 2A Tripped
RC Makeup Pump 2B Tripped
RC Makeup Pump 2C Tripped

Contact meaning
reversed. "NORM"
means tripped. Point 3173
is no good.

3213

RC Pump 2A Off

With the pump running,
the contact appears tomake and break at highfrequency. The meaningof this phenomenon has
not been determined.

*NOTE:

Besides the software and hardware errors indicated here, the actual
contact arrangement does not agree with Burns and Roe
drawings, either.

l-4-81

DETAILS IREFERENCE LIST
1.

Plant Charts & Computer Demand Logs for 0300 hrs & 0400 hrs on March 28,
1979

2.

Statement by Plant Manager.
Written testimony prepared by the Plant
Manager for and presented to the Presidents Commission.
The majorityof the testimony is a verbatium copy of notes generated from a meetingof managers shortly after the accident.

3.

TMI Phone Bill, March 28, 1979.

4.

Steam System Operating Procedure 2106-2.2,
Revision 95.

(duplicate - See #44)

6.

TMI Reactor Trip Report January 15,

7.

Maintenance Work Request, (WR) CO 711, Megger Checks for Water/Heat
Damage on Emergency Feedwater Pumps EF-P2A/B (duplicate - See #52)

8.

WR CO 718,

9.

Field Change Request,

10.

PO No. JCP 0C211, February 11, 1976, Purchase Order for Spare Liquid
Waste Tank Rupture Discs (duplicate - See #49)

11.

B&R Drawing No. 2485, Revision 9, Radwaste Disposal-Miscellaneous
Liquid Auxiliary Building Sump Tank & Section & B&R Drawing No. 2482,
Revision 13

12.

Special Operating Procedure,
Latching Functional Test

13.

Plant Strip Charts and Multipoint Recorder.
EMOV status in Section 4.8

14.

Alarm Printer. Same as Alarm Typer.
lists alarm status.

15.

Core Flood System Description,
tions

Condensate Polishing System,

1979 (duplicate - See #48)

Reactor Building penetration checks, January 15,

1979

FCR, 2329.1

SOP,

No.

R-2-79-31,

1979,

Feedwater

Also see discussion of

A computer output terminal which

Index No.

1-5-1

June 29,

2813,

and Technical Specifica-

Instrument No.

RC3-858

16.

TMI-2 Instrument Calibration Data Sheet,

17.

March 29,
Company

18.

Problem Report No.
April 29, 1974

19.

Flow Charts for Condensate Polisher Tanks 1, 2,
well as Polisher Outflow Chart

20.

Operating Log

21.

Transient Logs - A collection of informal handwritten notes, documenting observations and sequential events or actions, which were generated(Titledby licensee personnel during the course of the accident.
LT-O001 in licensee's files of the records of the event.)

22.

T/C list A list of incore Thermocouple readings which were taken at
about 0830 hours on 3/28/79 by technicians using portable equipment.

23.

Various charts of plant analog parameters,
Plant Strip Charts.
automatically and continuously by installed equipment.

24.

Instrument Drawings

25.

Plant Abnormal Procedure 2203-1.4,
Emergency"

26.

Plant Operating Procedure 2204-301B, "C5 Reactor Coolant Drain Tank
Hi/Lo", Response to High RCDT Alarm

27.

Plant Operating Procedure 2101-3.2,
13, March 10, 1979

"Unit Cooldown",

28.

Plant Operating Procedure 2102-1.1,
March 7, 1979

"Unit Heatup",

Revision 19,

29.

Plant Abnormal Procedure 2203-2.2,
1978"Turbine Trip",

Revision 7, October 25,

30.

Plant Emergency Procedure 2202-1.1,
1978"Reactor Trip",

Revision 6, October 25,

31.

Plant Emergency Procedure 2202-1.1,
1978"Reactor Trip",

Revision 5, October 6,

1976,

letter Dresser Industries to Metropolitan Edison
913,

Instrument Air & Service Air, Low Capacity,
3, 4,

5, and 8, asrecorded"Reactor Coolant Pump and Motor

1-5-2

Step 4.13, Revision32.

Plant Abnormal Procedure 2203-1.1,
Revision 5, September 13,-1978"Loss of Boron Moderator Dilution",

33.

Surveillance Procedure 2304-3D2, "Reactor Coolant Specific Activity",
Revision 0, November 8, 1977 (duplicate - See 61)

34.

Plant Emergency Procedure 2202-1.3, "Loss of Reactor Coolant/Reactor
Coolant System Pressure", Sections A & B, Revision 8, October 6, 1978

35.

Plant Emergency Procedure 2202-1.3, "Loss of Reactor Coolant/Reactor
Coolant Pressure", Revision 8, October 6, 1978

36.

Plant Operating Procedure 2103-1.4,
Revision 6, August 16, 1978

37.

Plant Operating Procedure 2104-4.1, Miscellaneous Liquid Rad Waste
Disposal, Revision 2, March 14, 1979

38.

Administrative Procedure 1012,
November 4, 1977

"Shift Relief and Log Entries",

Revision 8,

39.

Administrative Procedure 1012,
Section 3.5"Shift Relief and Log Entries",

Revision 8,

40.

Plant Emergency Procedure 2202-1.5,
Revision 3, September 29, 1979

41.

Operating Procedure 2104-1.2,
March 19, 1979

42.

Licensee Event Report 78-22/99X and 78-21/3L, March 29,
Trip

43.

QC Surveillance Report No. 77-266, October 27, 1977, Mechanical Mainten-.
ance of RC-RV2, Inspection, Repair and Testing, WR 1467

44.

Fourth EMOV Valve on order, Reference Number 897028-5000-1,
239715, August 31, 1978 (duplicate - See #5)

45.

Machine History File, WR 1166,
Operator for Repacking

46.

Temporary Change Notice 3-78-666, October 10, 1978,
8 to Procedure 2303-Ml4A/B/C"Reactor Coolant Pump Operation","Pressurizer System Failure,"Makeup & Purification System, Revision 13,

September 14,

1-5-3

1977,

1978 Reactor

P.O.

No.

Removal of Valveentered a Revision

47.

Performance Data Output Segments 1-6, Process Computer Program

48.

TMI Reactor Trip Report, January 15, 1979, Failure of the Atmosphere
Dump Valve Bellows. (duplicate - See item 6)

49.

Spare Discs,

P.O.

No. JCP OC211,

February 11, 1976 (duplicate - See

#10)
50.

Burns & Roe Drawing 2005,

Feedwater and Condensate,

Revision 30

51.

Station Preventative Maintenance Procedure E-2, "Dielectric Check ofInsulation, Motors & Cables", Revision 1, January 6, 1978 (duplicate
See item 97)-

52.

Machine History Card, January 23,(duplicate - See item 7)

53.

Plant Operating Procedure 2103-1.3, "Pressurizer Operation", Step
4.2.4, Revision 1, November 1, 1977 (duplicate - see item 62)

54.

Region I Notes: A.series of notes compiled by the Reactor Operations
and Nuclear Support Branch Chief of Region I in the early days of theaccident.

55.

Administrative Procedure 1002, "Rules for the Protection of Employees
Working on Electrical & Mechanical Apparatus", Revision 9, February 23,
1977 (duplicate - see #98)

56.

Plant Emergency Procedure 2202-2.2,-"Loss of Steam Generator Feed",
Turbine Trip/Abnormal Procedure, Revision 3, October 13, 1978

57.

Plant Operating Procedure 2102-3.3, "Decay Heat Removal Via OSTG",
Revision 6, April 17, 1978

58.

Plant Operating Procedure 2105-1.4, "Integrated Control System",
Revision 3, September 1, 1978

59.

Surveillance Procedure 2304-Wl, "Borated Water Source Concentration
Verification", Revision 1, March 14, 1979

60.

2304-3DI,

61.

Surveillance Procedure 2304-3D2,-J'RCS,--Specific Activity",
November 8, 1977 (duplicate - see #33)

"RCS Chemistry",

1979 (see WR C0711 Item No.

Revision 2,

1-5-4

February 5,

7)

1979
Revision 0,

62.

Plant Operating Procedure 2103-1.3,
3, July 19, 1978

63.

Surveillance Procedure 2301-3DI,
1978

64.

Plant Emergency Procedure 2202-1.5,
Revision 3, September 29, 1978

65.

Surveillance Procedure 2303-M24A/B, "Reactor Building Spray Pump
Functional Test and Valve Operability Test, Revision 7, October 6,
1978, Steps 6.3 & 6.4

66.

Surveillance Procedure 2303-M27A/B, "Motor Driven Emergency Feedwater
Functional Test & Valve Operability Test", Performed January 3, 1979,
February 26, 1979, and March 26, 1979, Revision 4, August 30, 1978

67.

Temporary Change Notice 2-78-666, October 10, 1978

68.

Same as 47

69.

Surveillance Procedure 2303-M14A/B, "Emergency Feedwater System Valve
Lineup Verification & Operability Test & Turbine Driven Emergency Feed
Pump Operability Test", Performed January 3, 1979, February 26, 1979,
and March 26, 1979, Revision 8, November 21, 197870.

Surveillance Procedure 2303-MIA/B, "Makeup Pump & Valve Functional
Test", Revision 5, September 22, 1978 (duplicate - See item 93)

71.

Surveillance Procedure 2303-M2A/B,
Test", Revision 7, June 20, 1978

72.

Burns & Roe Pipe and Instrument Drawing (P & ID) Drawing No.

"Pressurizer Operation","RCS Inventory",

Revision 3,

Revision
February 5,"Pressurizer System Failure","Decay Heat Removal Pump Functional

Reactor Coolant Make-Up & Purification,

2024,

Revision 25

73.

M. Kellog Isometric Drawings No.

74.

Plant Training Photograph No.
3 feet downstream of valve)

443 (Thermocouple on EMOV line located75.

Plant Training Photograph No.

81

76.

Plant Emergency Procedure 2202 1.1, "Reactor Trip",
6, October 25, 1978

223-1 and 2

1-5-5

Step l.l.e, Revision

77.

Shift Relief & Log Entries,

Revision 8, October 4,

1977

78.

Plant Abnormal Procedure 2203-1.4, "RCP & Motor Emergencies,
3, May 4, 1978

79.

Plant Operating Procedure 2105-1.3,
Revision 2, October 25, 1978

80.

Plant Emergency Procedure 2202-1.3, "Loss of Reactor Coolant/Reactor
Coolant System Pressure, Revision 11, October 6, 1978

81.

Plant Operating Procedure 2103-1.3, '.Pressurizer Operation",
3, July 19, 1978

82.

Administrative Procedure 1670.2, Site Emergency Procedure,
Revision 9, November 22,'1978 (duplicate of 108)

83.

Administrative Procedure 1014, "Recall of Standby Personnel to Plant",
Revision 3, June 21, 1977

84.

Makeupand Purification, Step 2.2.1

85.

Pressure Operation,

86.

Operating Procedure 2103-1.4, "Reactor Coolant Pump Operation, Step
2.1.1.1, Revision A, April 18, 1978

87.

Emergency Procedure 2202-1.3, "Loss of Reactor Coolant/Reactor Coolant
System Pressure", Step A.3.2.7, Revision E, May 12, 1978

88.

Technical Specifications; Appendix A to the Operating License for Unit
2.

89.

Procedure Step 6.1.3,

90.

Operating Procedure 2101-1.1, "Nuclear Plant Limits & Precautions",
page 27, Revision 4, August 8, 197891.

Operating Procedure 2103-1.4, "Reactor Coolant Pump Operation",
6, August 16, 197892.

Operating Procedure 2104-6.2, "Emergency Diesels and Auxiliaries",
Revision 9, March 14, 197993.

Surveillance Procedure 2303-MIA/B, Make-Up Pump & Valve Functional
Test, Revision 5, September 22, 1978 (duplicate - See item 70)

Revision"Safety Features Actuation System",

Revision
Figure 4,

Step 2.1.8"Addresses the RCP Vibraswitch Alarm"

1-5-6

Revision

94.

Administrative Procedure 1001,
1978"Document Control",95.

Administrative Procedure 1007,
July 14, 1978

"Control of Records',

96.

Engineering Change Memo, ECM No. S-5934, Electromatic Valve, RC-RV2,
Signal Light, Result of March 29, 1979 Open Electromatic Valve (EMOV).

97.

Station Preventive Maintenance Procedure E-2,

Revision 18, March 21,
Revision 4,"Dielectric Check ofInsulation, Motors and Cables" (duplicate - See item 51)
98.

SPMP,

Para. 4, "Tagging" (duplicate - See item 55)

99.

Administrative Procedure 1026, "Corrective Maintenance & Machinery
History", Revision 8, November 29, 1978

100. Administrative Procedure 1021,
July 22, 1977

"Plant Modification",

Revision 0,

101.

Corrective Maintenance Procedure 1407-1, "Station Corrective Maintenance Procedure", Revision 0, November 29, 1978

102.

B&R Drawing No. 2006, "Flow Diagram Makeup Waste Treatment & Condensate Polishing", Revision 24

103.

"Operation Quality Control Surveillance Report",
Quality Assurance" October 26, 1977

No.

104.

Plant Abnormal Procedure 2203-2.2 "Turbine Trip",
1978

Revision 7, October 29,

105.

Administrative Procedure 1037,
Revision 1, August 23, 1978

106.

Plant Operating Procedure 2102-1.1,
7, 1979"Unit Heatup",

107.

Plant Operating Procedure 2102-3.2,
March 10, 1979

"Unit Cooldown",

108.

Emergency Plan Procedure 1670.2,

77-266,"Operation"Control of Caution and DNO Tags",

Revision 9,

Revision 19, March
Revision 13,

November 22,

1978 (duplicateof 82)
109.

The Maintenance,"Work Request Log,

Unit 2",

1-5-7

starting with 1979 input.

110. Station Corrective Maintenance Procedure 1407-1,
1979

Revision 1, February 2,

Ill.

May 8,

Unit 2 Machine History Files (selected sections),

1979.

112. Unit 2 Valve List 2075, two volumes.
113. See 102114. Shift Maintenance Log, March 27-28, 197.9
115. "Instructions for Operating and Repairing Consolidated Electromatic
Relief Valve", B&W Instruction Book No. 620-0006:01-00400.
116. See 43
117. Machine History Card, Mechanical,
and September 29, 1975.

RC-RV2 and RC-V2,

118. Machine History Card (Unit 1), RC-RV2 and RC-V2,
September 29, 1975.
119. Letter, B&W to GPU,

Dresser Valve RC-V2,

120.

Letter, GPU,

to UE&C,

121.

Drawing No.

3079, Sheet 14,

October 26,

October 26,

September 2,19741974 and1975.

Concerning Dresser Valve RC V2.

122. Megger Test, January 23,

Revision 13.

1979,

Control Building East,

EF-P-2A.

123. UE&C Equipment Data Sheet, Mechanical, July 28, 1977, Ingersoll Rang,
470 gpm at 2510 feet and 3560 rpm, discharge pressure 1500 psi, 100 F(EF-P-2A).
124. UE&C Equipment Data Sheet, Motors,
1977; EF-P-2B, July 27, 1977.

July 28,125. UE&C Equipment Data Sheet, Mechanical,
1470 psi, EF-P-2B.
126. UE&C Equipment Data Sheet, Motors,

1977,

July 27,

July 27, 1977,

EF-P-2A, July 28,

1977, discharge pressure
EF-P-2B.

127. UE&C Equipment Data Sheet, Vibration Record,

July 27,

128. UE&C Motor Bumping Check List, July 27, 1977,

EF-P-2B.

1-5-8

1977,

EF-P-2B.

129.

Manual 31700, "Consolidated Closed Bonnet Maxiflow Safety Valve",
Instruction Book No. 620-0006, 01-003700 (Dresser Valves).

130. WR No.

340, March 25, 1977, w/Procedure,

B&W"Hot N2 Gas Testing of Pressurizer

Code Relief Valves"
131.

Sequence of Events; Appendices I-A and II-A to this report132.

2103-1.9,"Reactivity Balance Calculations"133. 2302-MlA/B, Makeup Pump & Valve Functional Test, Revision 5,
1978 (duplicate - See #93 & #70)

September 22,

134.

LER 78-021-03L,135.

RCS and Reactor Building Pressure Charts.

136.

Utility Typer.
Computer output terminal for data specifically requested
by operator.
Also output terminal for alarm status printout should
the Alarm Typer fail.

137.

Final Safety Analysis Report138.

10 CFR 50.46

139.

Computer Logs/Daily .Record Storage; March 30,

140.

System Description - Radwaste Disposal Gas System, October 1975

141.

System Description - Nuclear Sampling System,

142.

Procedure Change Request PCR-2-78-707 and 895 (duplicate - See #46)

143.

Site Operations Memorandum #403 oated August 9,

144.

Site Operations Memorandum

145.

Plant Trip Report for March 28,

April 29,

1979; 78-033-OIT,

April 23,

Plant Strip Charts; see 23.

1979

February 1976

1978

1979 Trip

146. Problem Report #2847
147.

Reactor Trip Report - Unit II #12,

148.

Burns & Roe Drawings

November 3,

1-5-9

1978

1978

Appendix I-A

Operational Sequenceof
Events

Note:
This sequence of events was
prepared based on elapsed timefrom the initiating event,the trip of the turbine,which occurred at 04:00:37
on March 28, 1979.

APPENDIX I-A
OPERATIONAL SEQUENCE OF EVENTSITEM
1ELAPSED TIME
3/28/79

EVENT DESCRIPTIONREFERENCES

Prior toThree Mile Island Unit Two (TMI-2) was operating at 97% power with theIntegrated Control System (ICS) in full automatic.
Normal reactor
coolant system (RCS) makeup, seal injection and letdown were establishedwith makeup pump (MUP) "1B" in service.
Pressurizer spray valve andheater controls were in manual to equalize pressurizer and RCS boron
concentrations.
RCS boron concentration was 1026 ppm and radioactivity
concentration was 0.394 uc/ml.
Identified RCS leakage, believed to befrom the electromatic relief or code safety valves on the pressurizer,-was approximately 6 gpm.
RCS pressure was 2148 psig.

I.
2.

Turbine Trip

3.
4.
5.
6.

Interview 58
Operating Logs
3/28/79
Surveillance Procedure 2301-3D]
Reactimeter
Operating Logs
GPU Sequence of
7/16/79

2

On the secondary side, once thru steam generator (OTSG) levels wereat about 160 inches with pressures of about 900 psig.
Steam generator
feedwater pumps 1A & IB, condensate booster pumps 2A & 2B and condensate pumps 1A & IB were in service.

3

Operators had been trying for about eleven hours to transfer an isolatedcondensate system polisher's spent resins to the regeneration receivingtank. The transfer procedure utilizes station compressed air to fluff
the resin and demineralized water to transfer the resin between tanks.
A resin block had developed in the transfer line.

1.

Interviews 2,
5, 14, 17

4

At this point, plant operators hypothesized that the water pressure,then the discharge pressure of two demineralized water pumps or the

1.

Interviews
166, 189

IA-I

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
condensate pumps, either at approximately 160 psig, exceeded the station
air pressure (approx. 100 psig), forcing water into the air system.
Further, the water then made its way through a check valve (since foundstuck open) and through the open cross-connect between station and instrument air systems to the polisher isolation valve controls, causing these(This problem of water entering the air system has
valves shut.
occurred at least twice before.)56-- sec.

Polisher outlet and/or inlet valves go shut.
Note:

1. Analysis (see notes)

Condenser hotwell level dropped below the low alarm within 13
seconds (21.72"); turned and entered the normal range in 28
seconds (26.44"); and passed the high alarm in 73 seconds
Even with this increasing level (which should initiate(37.77").
rejection flow), normal coastdown flow of the condensate booster
pumps, and one condensate pump still running, the condensate
This strongly
flow was reading "0" at 2 minutes into the event.
supports contention that polisher valves were shut.

l.

3.
4.

Alarm printerHourly computer log
Analysis
Computer Analog
Trend Recorder

7

Note:

All polisher inlet and outlet isolation valves were found shutafter the turbine trip.

I.

Interview 36

8

Note:

Polisher bypass valve on TMI-2 does not open automatically onhigh differential pressure.

I.

Interview 8

9

Note:

Polisher outlet pressure was reportedly so close to the condensate 1I.
booster pump suction pressure trip limit that the full capacity of
the condenser level control reject valve could not be used. A valve
in series with the reject valve was throttled to limit rejectflow which came from the polisher outlet to prevent spurious
trips of the condensate booster pumps.

Interviews
5, 17

IA-2

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTION

10-1 sec.

Condensate booster pumps trip.

REFERENCE(See notes below.)I.

Analysis

11

Note:

Each main feed pump is served by its own low-suction-pressureI.
trip switch.
Since these switches are purposely set differently,
2.
to prevent loss of both pumps on minor transients and since the3.
two pumps tripped together, within 152 milliseconds, the suction
pressure transient must have been exceedingly fast. This transient
could only result from the condensate booster pumps tripping at
minus 1 second.

Utility typer
P&ID
InterviewGPU-1712

Note:

Loss of suction pressure will cause a condensate booster pump
trip.

Alarm Procedure
17.E9
Burns & Roe
drawings

1.
2.

13

Note:

1. Interview 109

Condensate booster pumps were subsequently found tripped.

2.

14

15

-1 sec.

Condensate pump 1A trips (one condensate pump remains running).
time 04:00:36, 3/28/79).

Note:(ActualThe computer trip indication received indicates this pump tripped
due to something other than control room switch operation.
Nointerview has indicated relay flags were reset; therefore electrical faults can be ruled out.
Local breaker switches require
handles that were reportedly not present.

IA-3

See restart attempt
5 min. 15 sec.

1. Alarm printer2.
Interview
GPU-2
3.
Interview 195
1.
2.
3.

Burns & Roe
drawings
Inspection
Discussion
with operators

ITEM

DATE AND TIME
3/28/79

16

Note:

17

18
19

Note:

0 sec.

REFERENCEEVENT OESCRIPTION
A trip of an associated condensate booster pump with the control
switch in AUTO will cause that condensate pump to trip. (Procedures require operation in AUTO, but plant problems have trained
operators to normally operate in MANUAL.)

1.

Wiring error in trip circuit for condensate pump ]A bypassed
AUTO switch contacts, making it appear to that trip circuit alonethat paired condensate pump 1A and condensate booster pump 1A
were in AUTO, no matter that position the'switch really was in.

1.
2.

Main feedwater pumps are tripped resulting in an almost simultaneous(Actual time 04:00:37, 3/28/79)
trip of the turbine.
Note:

2.
3.

3.

Alarm Procedure
17. E8Interview GPU-11
Burns & Roe
drawings
Inspection by GPU
Burns & Roe
drawings
Interview 195

1. Alarm printer2.

Interview GPU-20

Low feedwater pump suction pressure or loss of the condensate
booster pumps, while in automatic, will cause the feed pumpturbines to trip.

1.

Alarm Procedure
17.A15 & 17.A]6Burns & Roe
drawings

2.

20

Note:

Both feed pumps tripping will cause the main turbine to trip.

1. Abnormal Procedure
2203-2.2
2.
Burns & Roe
drawings

21

Note:

Tripping both feedwater pump turbines will start all three
emergency feedwater pumps.

1.
2.

IA-4

FSAR Section
7.7.1.2.1.3
Burns & Roe
drawings

ITEM

DATE AND TIME
3/28/7922

EVENT DESCRIPTIONREFERENCE

Note:

Loss of one or more main feedwater pumps will cause the ICS
to run back reactor power, causing the control rods to be driven
into the core.
None of the interviews conducted indicated thatthe operators specifically observed the runback in progress prior
to the trip of the reactor at 8 sec.

1.

FSAR Table 7.7-2

Alarm printerInterview 9230+ sec.

All three emergency feedwater pumps start. Pressurizer level andpressure increasing rapidly. Condenser hotwell level begins rapidtransient resulting in filling the hotwell to a fully flooded condition
in approximately one minute.

I.
2.

24

1 sec.

Turbine stop and intercept valves closed.

1. Alarm printer25

3 sec.

Press'ure in reactor coolant drain tank (RCDT)26

3-6 sec.

RCS reactor pressure reaches electromatic relief valve (EMOV)
setpoint (2255 psig).

27

8 sec.

Reactor trips from reactor high pressure (setpoint = 2355 psig).
Indicated reactor pressure on the wide range RCS pressure strip chart fromthe control room shows an increase to approximately 2435 psig, which
would normally suggest one of the two safety valves may have lifted.
However, this maximum value ýis not confirmed by the reactimeter or by
the narrow range RCS pressurle recorder trace.

28

Note:

500 KV breakers are tripped.
begins to increase.
opening

The reactimeter samplied data every 3 seconds and may have missed
the peak. The narrow range pressure instrument did not indicate
a peak high enough to trip the Reactor, which is known to occur.
The wide range presssure strip chart, when compared to otherpressure sensor outputs, appears to indicate about 35 psig high.

IA-5

1.

Reactimeter data

1.
2.

Plant strip charts
Reactimeter data

1. Alarm printerPlant strip charts
2.
3.
Interviews
4.
ReactimeterITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

29

8 sec.

Pressurizer heater groups 1-5 trip. This event is believed to berelated to the operator resetting the pressurizer spray and heater controls to "Automatic," as an operator action following the turbine trip,
and pressure exceeding 2125 psig. A continuous pressurizer spray had
been in effect prior to the incident to eliminate a difference in boron
concentration between the pressurizer and the rest of the reactor
coolant system.

I. Alarm printer2.
Burns & Roe
drawing

Shift Supervisor announces a Unit 2 turbine trip - reactor trip.

1.

Interview 5

1.
2.
3.
4.

Reactimeter
Interview GPU-I
Plant strip charts
Emergency Procedure 2202-1.1

30
31

8 sec.
9 sec.

32

33

34

Secondary steam pressure peaks at about 1070 psig.
relief valves are reported to have lifted.

Note:

IO
10 sec.

13 sec.

The main steam

With ICS in AUTO, turbine bypass valves (MS-V25A&B, 26A&B)
would attempt to control main steam pressure at 1010 psig +10 psi.

Operator verifies all control rods tripped and bottomed.

Letdown from reactor secured (MU-V376'shut).
attempts to start makeup pump IA.

IA-6

Operator unsuccessfully

1. Interview 9Emergency Pro2.
cedure 2202-1.1
1. Alarm printerInterviews2.
9 & 15Emergency Pro3.
cedure 2202-1.1

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

35

Note:

These actions, according to interviews, are immediately takenby the operators, following a reactor trip, for the purposeof minimizing a large pressurizer level transient.

36

Note:

Alarm Printer output for makeup pump IA, IB, and IC status(norm/trip) found to be reversed due to software error,
potentially misleading operators who read printout.

1. Burns & Roe drawings
2. Analysis of alarm
printer output
Inspection by
3.
Licensee

.37

Note:

Cause of start failure is indeterminate at this time but thisinvestigation indicates that switch contact "bounce" is highly
Troubleshooting switchgear must await reduction inprobable.
Operator may have believed pump has started,
contamination levels.
because of his actions, but panel indications would dispel thisbelief.

1.

Inspection

I.

Reactimeter
Plant strip charts

RCS pressure reaches setpoint for EMOV closure (2205 psig).
However, later events will show that closure did not occur.

2.

Pressurizer heater groups 1-5 restart in AUTO at 2105 psig, decreasing.

1. Alarm printer40

Emergency feedwater pumps reach full pressure.

1.

41

Note:

Emergency feedwater pump discharge pressure indication is availI.
2.
able to operator on panel, but the emergency feedwater flow must
be inferred from once through steam generator (OTSG) level changes.

42

Note:

During the transient, the operators reported they did haveapparent flow indication in the "A" OTSG from the noise

38
39-- 13 sec.

14 sec.

IA-7

Alarm printerInterview
Inspection of panel

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION("hammering") heard from the vibration and loose parts monitor
system, which was on to monitor one of the steam generators.
(Monitor had been left on by operators to detect OTSG relief
valve operation during any future transients. This monitor was
audible in the control room but was not recorded on tape.)43

N, 15 sec.

1.
2.

Reactimeter
Plant strip charts

RCS pressure
The Unit has just experienced a Turbine/Reactor Trip.
and pressurizer level were decreasing rapidly after reaching theirpeaks.
Unknown to the plant operators the electromatic relief valve(RC-RV2) was not shut and was passing reactor coolant from the steam
Based onspace at the top of the reactor coolant system pressurizer.
control room indications, the RCS pressure and pressurizer level weretrending together and decreasing as was expected after a reactor trip.
The OTSG water levels were at about 90 inches and decreasing at about
4 inches/second. The OTSG steam pressures were about 1030 psig and
The turbine bypass valves were open
decreasing at 4 psi per second.
relieving steam. The OTSG water levels had not yet reached the ICS
setpoint of 30 inches for the programmed opening of the emergency
feedwater valves (EF-VllA and EF-VllB) that would admit feedwater to
the OTSGs.

1.

Plant strip charts
Reactimeter

Reactor pressure reaches low-pressure-trip setpoint (1940 psig),pressurizer relief valve exhaust pipe temperature reaches 239 F, and

1. Alarm printer(The RCS
Pressurizer level reaches peak of approximately 255 inches.
pressure, average temperature, and pressurizer level appear to be
tracking normally at this time.)
PLANT STATUS

44

45

30 sec.

IA-8

2.

ITEM',DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

one code safety relief valve exhaust pipe temperature, reaches 203°F.
(These temperature indications would be expected, whether the EMOV
reseats or not.)46

38 sec.

47
48
49

40 sec.
41 sec.

Steam generator "A" level at 23.8 inches and decreasing.

1.
2.

Alarm printerInterview 9Emergency feedwater valves EF-VllA&B start to open as level decreases
below 30 inches and give dual indication on panel.

1.

Interview 9

Steam generator "B" level at 23.7 inches and decreasing.

I . Alarm printerOperator starts makeup pump 1A and opens MU-Vl6B by operator to increase'injection flow to loops.

2.

Interview 9

1.
2.

Alarm printerInterviews
61, 15Emergency Procedure 2202-1.1

3.

50Operator opens DH-V5A to control makeup tank level using BWST headpressure.

51

Note:

52

54 sec.

During the interviews, the operators noted that they had several
problems with keeping makeup pump 1A operating when desired.
This
sequence will note several trips of that pump.
The last trip
occurs at 4 hr. 18 min. after the start of the incident, andthe pump is not restarted again for the remainder of the incident.

Pressurizer level is at minimum of 158 inches (lowest level reached) andstarting to rise.

IA-9

1.

ReactimeterITEM

DATE AND TIME
3/28/79

53
54

58 sec.

55

56

57

60 sec.

EVENT DESCRIPTIONREFERENCE

Note:

1. Interview

Minimum level experienced was significantly higher than expererienced in previous trips of the same nature.

Pressurizer l:ow level alarm at 185.3" received.

1. Alarm printerNote:

1. Discussion with
Computer Supervisor
2. Log typer

Discrepancy between alarm printer and reactimeter data resultsfrom frequency of calculation. The computer takes differential
pressure readings from each level transmitter (updated'every 30
seconds) temperature compensates them and averages the results,
performing this calculation once every 15 seconds.
The alarm
printer routine samples the most recent result of this calculation once per second and compares it to the alarm setpoint.
The reactimeter routine samples a continuously calculatedpressurizer level signal once every 3 seconds and therefore,
can be considered closest to real-time data.
In this instance,
the alarm printer has just recognized level is below setpoint,even though level is rising at this point.
The memory trip
review confirmed the level dropped lower than the 185.3"
reported at this point on the alarm printer.

One pressurizer c 8 de safety valve tail pipe high temperature alarm
received at 204.5 F with RCDT pressure at 12 psig and increasing.
Note:

This alarm would be expected following an EMOV opening due tobackflow of steam from the common tail pipe dischargeheader.
The temperature elements for the EMOV are located at the EMOV
discharge, which is approximately 40 feet from the temperatureelements for the code safety valves.

IA-10

1. Alarm printer2.
3.

Interviews
Reactimeter

1.
2.
3.

System Description
Burns & Roe drawings
Plant training
photographs

ITEM

AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE
RDATE

58

1 min.

Steam generators A and B levels at approximately 10 inches on startup level instrumentation.

1. Reactimeter
2. Plant strip charts

59

Note:

1. Analysis

60

Differential temperature, hot to cold leg,indicating OISG's are going dry.

61

1-4 min.

This level indication is due to weight of steam alone (dry OTSG).
rapidly approaching zero,

1. Reactimeter
2. Analysis

Pressurizer level rising very rapidly.
The change in RCS mass inferred1.
from pressurizer level occurring during this period cannot be accounted
for under the conditions of makeup, temperature changes, and injection
2.
flow. The mass discrepancy is further increased when the mass loss
3.
through the EMOV is also considered.
Moreover, reactor coolant wide4.
range pressure shows, at 1770 psig (saturation temperature approximately
619'F), a reversal in direction and subsequent stabilization of RCS pressure
for what appears to be 2-3 minutes, suggesting void formation. Moreover, thememory trip review printed from the computer shows pressure at approximately 1750 psig from 68 seconds to 113 seconds.
This stabilization, however,is not confirmed by the reactimeter or narrow range pressure instrumentation.

62

Operating staff interviews covering this period disclose that theoperators had felt that they had "caught" the pressurizer level decrease and that they had actually expected it to drop much further
before being "turned around," based upon previous experience.
Further,the Shift Supervisor noted that the rate of pressurizer level increase
appeared to be wrong for the reactor trip that had occurred.

63

Note:

It appears that beginning at approximately one minute and continuing thereafter, the reactor coolant system experienced

IA-11

Interviews 1, 14,
15
Calculations
Plant strip charts
Utility typer

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

either a general reduction in density, as might occur with
distributed voids, or the formation of one or more discrete
steam vapor voids.
PLANT STATUS
64

RCS pressure was still dropping and pressurizer level had just turnedand begun to rise. This behavior is unexpected, since RCS pressure
and pressurizer level should trend together, during a loss-of-feedwaterflow transient.

65

Note:

66

The reactor coolant drain tank (RCDT) pressure and temperature started toincrease, showing the effects of continued discharge or reactor coolant
thru the EMOV.
Plant operators did not associate these indications withleakage past the EMOV, but rather with the initial opening of the valve.

The deviation from expected behavior was due to the formationof voids within the RCS loops and/or reactor vessel as RCS pressure
plummetted.
The inability of the pressurizer to effectively damp
this pressure transient is attributed to a massive loss of its
energy content, first by dilution during the initial power generation/removal mismatch as the reactor trip is delayed 8 seconds
after the turbine trip., and relatively cool hot leg fluid is
injected into the pressurizer, second by bulk energy removal as
the EMOV lifts and fails to reseat, third by boiling and ejectionof hot water as level and pressure drop to compensate for the RCS
cooldown from the steaming OTSGs, and finally by the short failure
to generate additonal energy during the period when pressurizer
heater groups 1-5 trip on high pressure.

IA-12
L

1.
2.

Plant strip charts
ReactimeterITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

67

1 min.

At request of Unit 2 Shift Supervisor, the Unit 1 Shift Foreman called
the Station Manager to inform him of Unit 2 trip, apparently as a matterof policy and not as a result of any suspicion about the nature of thetrip. The Superintendent of Technical Support - Unit 2, the on-call
Duty Officer, was also called at about this time.

1.

Interviews
1, 27

68

1 min.

The hourly plant log typer displayed the following information as being
applicable for this time:

1.

Log typer

I.
2.

Alarm printerUtility typer

1.
2.

Alarm printerPlant strip char

RCS Flow:

Loop "A"Th

135 MLBH
05747F

Tca: 573 0 F
ca
"B"
Loop
5760F

Thb 5730FT
cb: 573 F
16 GPMMU Flows:
OTSG "A" Level: 8 inches
5 inches
OTSG "B" Level:
Condensate Flow: 0 KLBH
69

1 min. 40 sec.

Reactor coolant drain tank (RCDT) indicates 86 0 F and increasing.
temperature prior to event was about 70 F.)

70

2 min.

Reactor coolant pressure reaches average of 1726 psig.

IA-13

(RCDT

ITEM

DATE AND TIME
3/28/79

71
72

2 min. 2 sec.

EVENT DESCRIPTIONREFERENCE

Operator requests computer print RCS loop B inlet temperature (573.6°F).

1.

Utility printerPressure conECCS initiation (HPI - nomimal setpoint = 1600 psig).
firmed by reactor pressure instrumentation.* Decay heat removal pumps
Makeup pump IC starts leaving makeup
start, and makeup pump IB trips.
pumps IA and IC now running as HPI with discharge valves open to the
250 gpm/loop throttled position.

1.

2.
3.

Alarm printerInterview 61
Utility typer

73

Note:

When HPI is initiated, any discharge valves (MU-Vl6A,B,C or D)
which may have been previously throttled, are designed to returnautomatically to their open position. Based on interviews thevalves on the IA pump had not been manually throttled closed
up to this point in the incident.

74

Note:

The trip of makeup pump lB is appropriate action for control
logic at this point.

1.

System Description

75

Steam generators appear "dry" based on zero Hot to Cold leg RCS temperaOTSG pressures have peaked and are dropping radidly..
ature difference.

I.
2.

Reactimeter
Analysis of parameter trend

Pressurizer level at 225.7 (207.7)* inches and increasing.

1.
2.

Reactimeter(Alarm printer)*

Shift Foreman enters control room and obtains the emergency proceduresfor turbine trip, reactor trip, and safeguards actuation, to confirm
that all appropriate actions occurred.

1. Interviews

76

2 min. 28 sec.

77

78

3 min.

13 sec.

ES bypassed by operator actions.
has started normally.

Operator verifies all ECCS equipment

IA-14

14,

15

1. Alarm printerInterviews
2.
14, 15ITEM

DATE: AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

79

Note:

Bypass of the ES does not change any ECCS equipment status,
but is required to allow operator to control plant equipment.

80

Note:

Operating procedure allows ES Bypass if RCS pressure andPressurizer level have recovered.
Pressure had notrecovered at this time.

1. Operation Procedure 2202-1.3

81

Note:

Operator takes control of MU-V16 valves and throttles them aspressure drops, to prevent MUP runout.
(Runout will cause
cavitation and, potentially, pump failure.)

1.

Interview 80

1.
2.
3.
4.
5.
6.

Alarm printerReactimeter
FSAR
System Description
Interview 15
Plant strip charts

1.

Interview 15

1.

Interview 80

82

3 1/2 min.

83

4 min.

Pressurizer level in high alarm at 266.1".
RCDT temperature 127 0 F, tankpressure rapidly increasing and stabilizes at approximately 120 psig
for 3 minutes.
The reactimeter plot of RCDT pressure exhibits an
oscillatory behavior at 120 psig, similar to what would be expected if
the RCDT safety valve had actuated.
The setpoint for the RCDT safetyvalve is listed as 150 psig. The actual safety valve setpoint is notknown at this time, but reactor building pressure does start increasing
at about this time.- The operators indicate during interviews that the
RCDT cooling system (a manually initiated system) was in operation
prior to the transient due to relief valve leakage.
Operator throttles MU-V16 valves in attempt to control pressurizer level
increase.
Note:

84
85

4 min.

38 sec.

Operator training emphasizes avoidance of "solid" operations.

After MU-Vl6C&D are fully closed operator trips makeup pump 1C.
pump 1A still running with MU-Bl6A&B in throttled condition.

IA-15

Makeup1. Alarm printerInterview 15

2.

ITEM

DATE AND TIME
3 28/79

86

EVENT DESCRIPTIONREFERENCE

Note:

1.

Based on interviews, operators stated that when the flow from amakeup pump operating in the High Pressure Injection mode isthrottled, the operators do not reduce flow below approximately
100 gpm in order to protect the pump.
This minimum flow is
accomplished by shutting the valve in one of the two injection
paths and throttling the flow in the remaining path to 100 gpm.
Based on BWST level changes during the early hours of the incident, this assertion cannot be supported.

Interview 15

87

4 min.

52 sec.

Operator starts second intermediate closed cooling pump in preparation for putting a second letdown cooler in service, allowing increased
letdown to halt pressurizer level rise.

1. Alarm printer88

5 min.

15 sec.

Operator restarts condensate pump ]A. Two condensate pumps should now
be running.
Operator tries to restart condensate booster pump 2B, but
it trips; probably on low suction pressure.

1. Alarm printer

89

90

91

Note:

5 min.

15 sec.+Operators attempting to regain normal condensate and feedwaterflow to pump down condenser, preventing loss of vacuum or turbine
damage on high water level, and to enable use of main feedwater
pumps to feed the OTSGs.

Pressurizer level continues to increase.
The operator opens MUPIA
recirculation valve, throttles the HPI flow to minimum and initiates
letdown at a rate in excess of 160 gpm, in attempt to halt thepressurizer level rise. Orifice and bypass valves are utilized together
with a second heat exchanger now in operation.
Note:

At five minutes after the transient, pressurizer level starts
dropping at a rate of approximately 60 inches per minute for

IA-16

1. Alarm printer

1. Reactimeter
2.

Interview 14

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

15 seconds.
At this same time, an apparent decrease in loop
"A" flow and increasing reactor coolant temperatures areobserved.
The pressurizer level rate of change is not in agreement with the actions, system temperature changes, and makeup
pump operation occurring during this period, unless voids exist
in the RCS.
92

5 min. 20 sec.

Operator again tries to start condensate booster pump 2B and it
trips again.

I.

93

6 min.

Pressurizer level appears lost (bubble lost).
RCDT pressure starts
rapid rise above 120 psig to the RCDT safety valve setpoint of approximately 150 psig, most probably the result of the open EMOV now
passing liquid versus steam.
The combination of full Pressurizer*indication and increasing RCS pressure, the result of rising hot
leg temperatures while charging with the one MU Pump, appears toconvince operators that they have a solid plant.

I . Reactimeter

Alarm printer

2.
3.

Plant Charts
Interview 15

The RCS hot leg temperature and pressure have reached saturation conditions, as indicated by the reactimeter and the wide range RCS
pressure stripchart. The RCS indicated flow rate decreased sharply,
suggesting a reduction in reactor coolant density. The OTSGs hadboiled dry as indicated by a continuously decreasing steam pressure,while RCS hot leg and cold leg temperatures increased.

I.

Reactimeter
Plant strip charts

Operator again tries to start condensate booster pump 2B and it
again.

1. Alarm printer

PLANT STATUS
94

95

6 min. 24 sec.

IA-17

trips

2.

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

96

6 min. 58 sec.

Operator throttles letdown flow to 71 gpm following high temperature
alarm.

I.
2.

Alarm printerInterview 15

Operators noted plant conditions did not match Emergency Procedures.

1.

Interview 15

Reactor building sump pump (WDL-P-2A) turns on, presumably pumping 140
gpm from the reactor building sump to the miscellaneous waste holdup
tank (WDL-T-2) via the normally open reactor building isolation valves.
However, tank level records and operator interviews show this tank not
to have changed level significantly during the incident, and there isreason to believe the sump pump discharge was aligned to the auxiliarybuilding sump tank (an alternate path).
(This tank had a blown rupture
disc which was scheduled for later repair.)These isolation valves do
not close on the SFAS signal which initiates HPI.
These valves willclose when 4 psig in the reactor building SFAS setpoint is reached.

I.
2.

Alarm printerRadiologicalsequence of events

Note:

I.
2.

Alarm printerInterviews
5, 15

1.

Interview 95

1.
2.

Interview 61
Reactimeter97
98

7 mi n.
7 min. 29 sec.

99

100

-, 8 min.

102

Second Shift Supervisor, who was onsite because of start-up in progress
on Unit 1, arrives in Unit 2 control room to assist as necessary.
Note:

101

8 min.

The reactor building sump pump had started (WDL-P-2A) and run
for 1 1/2 minutes at 01:23 hours; and at this time the pumpingof the sump was not considered to be abnormal.

The availability of a second Shift Supervisor is not required byTechnical Specifications, but he was available due to the pending
startup of Unit 1.

Operator finds OTSG level at 10 inches on the startup range. Operator
considers this level means OTSG is "dry" per his training. Operator

IA-18

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

verifies emergency feedwater pumps are running and examines valve lineup.
Operator finds and announces that EF-VI2A & 12B valves are shut.
Position indicating lights on 12A were obscured by a caution tag hanging
from another valve controller. Position indicating lights for 12B may
have been obscured by operator's body as he leaned over panel.
Operatordrives valves open, resulting in dry OTSG being fed with relatively coolwater.
Hot and cold leg temperatures drop.
RCS pressure, now undercontrol of loop saturation conditions, follows accordingly.

3.

103

Note:

On 3/23/79, an NRR Operator Licensing Branch (OLB) examiner
conducted examinations and "walked thru" the EFW System.
Thevalves were open on that day.

1.

104

Note:

A routine, scheduled surveillance test was performed on the A &B electric emergency feedwater pumps on 3/26/79, at approximately
10:00 a.m., by training shift CRO and AO.

105

Implementation of this surveillance test procedure results inclosure of both the 12A and 12B valves, regardless of whichpump is being tested. The procedure calls for reopening of
the valves, along with ensuring the proper status of at least
three other valves.
The procedure for the electric driven
emergency feedwater pumps is insufficiently specific to provide documentation of valve opening, in that the proceduredoes not require individual signoffs for each valve; rather,
the procedural requirement is in sentence .form with one signoff signifying proper positioning of the valves.
Additionally,no documentation is available that the ST steps were completed,with the exception of the completed data sheets.

IA-19

4.
5.
6.

Plant strip
chartsSurveillance test
procedureInterviews 5,
9, 14
Memory trip review
Memo from OLB to
Investigation
Team (5/21/79)

ITEM

DATE AND TIME
3/28/79

The investigation has not found any evidence of willful closureof the valves over the period of 3/26/79 to 3/28/79.

106

107

108

REFERENCEEVENT DESCRIPTION

The Control Room Operator (CRo) who was on duty in the control
room when the 3/26/79 surveillance was performed, stated during
interviews his specific recollection that he left the 12A & 12B
valves open at the end of the test.

1.

Interview 131

Rapid rise in OTSG pressure observed indicating feed flow to generators.

1.

Reactimeter

Confirmed by EFW pump discharge pressure decreasing and "hammering" and"crackling" heard from the vibration and loose parts monitor speakeraligned to listen to "A" OTSG.

2.

Interview 9

Reactimeter

Note:

8+ min.

109

110

8 1/2 min.

Hot leg temperature dropping rapidly at a rate of approximately 110 0 F/hr.

1.

ill

9 min.

(Analysis would suggest this could only
Condensate pump IA trips.
occur if operator holds start-stop switch of condensate booster pump 2A
in start for at least 10 seconds or pump is tripped by operators
locally.)
Memory trip review records that pressurizer level reaches
404.9 inches.

1. Alarm printer2.
Burns & Roe drawings
Memory trip review
3.

112

9+ min.

10
Source Range NI's energized as Intermediate Range NI's drop below 5XO0
amps.

1. Alarm printer113

9 1/2 min.

Turbine bypass valves placed in MANUAL OPEN, because OTSG pressure
was increasing and valves were not responding.

1.

Interview GPU-20

114

10 min.

Plant Nuclear Engineer arrives from Unit 1 and commences routine posttrip review. Another engineer calls the Unit 2 Superintendent to

1.

Interviews 97,
1, 83

IA-20

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

notify him of the trip, apparently as a routine action with no mentionof anomalous conditions.
PLANT STATUS
115

116
117

The RCS pressure was at saturation conditions with respect to RCS
hot leg temperatures.
This was the result of the EMOV remaining open,
high letdown flow rate, throttled high pressure injection isolation
valves and emergency feedwater addition to the OTSGs.
Emergency feedwater flow had been established to both OTSG's resulting in increased
steam pressures and rapid RCS cooldown and depressurization.

10 min.
10 min.

19 sec.

118

10 min.

120

10 min. 48 sec.

121

24 sec.

Plant strip charts

On-call TMI Duty Officer (then the Unit 2 Superintendent - TechnicalSupport) called by site personnel.
He arrives 40 minutes later.

1. Interviews

Second reactor building sump pump (WDL-P-2B) starts.
Now pumping
reactor building sump to the auxiliary building at 280 gpm.

1.

Alarm printer

1.

Alarm printer

Note:

119

I. Reactimeter
2.

27,

83

Second reactor building sump pump starts based on rising levelin sump at 4.4'.

Operator stops,
interval.

restarts and stops makeup pump 1A within a 4 second

Reactor building sump alarms high at 4.65 ft.

1. Alarm printer

Note:

1.

A control room operator reported that the sump overflowed (6 feet)sometime after this point.

IA-21

Interview 15

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

122

10 to 11 min.

Pressurizer level indication comes back on scale and drops rapidly(20" in 1 min.), as RCS loop temperatures continue to drop from heatremoved by the OTSGs.
Apparently, the two-phase hot leg water,
received from the surge line, remains separated as it flows throughthe pressurizer, giving the external appearance of a reestablishedlevel indication.

1.

Reactimeter123

11 1/2 min.

RCDT sooler intermediate cooling temperature indicates offscale(>225 F).

1.
2.

Alarm printerReactimeter

1.

Alarm printer

1.

124
125

Note:
11 min. 43 sec.

Makeup pump IA restarted.
Pump ]A remains in throttled condition until second HPI initiation at
3 hrs. 23 min. 16 seconds.

126
127

RCDT at 125 psig and increasing.

8 to 12 min.

Operators attempting to establish 30 inches in each steam generator
per procedure.
Level increase in "A" OTSG leads that in "B" OTSG.
Interview and records review indicates operators throttled EFW before
reaching minimum steam generator levels to limit RCS pressure andtemperature decrease.

2.
3.
4.

Interview 196
Plant strip charts
Reactimeter
Memory Trip
Review
Alarm printerPlant strip charts

128

13 min.

Decay heat (DH) removal pumps 1A and IB turned off with RCS pressure
about 1400 psig.

1.
2.

129

13 min.

The memory trip review for this period shows that the discharge pressureof the EFW pumps dropped after the opening of the EF 12A and B header
isolation valves.
This drop is appropriate and signifies the start of

1. Log typer
2. Interviews
189, 196

IA-22

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

flow to the OTSG's.
However, the discharge pressure increases againreaching a peak (less than the previous shut-off head) suggesting aninterference with discharge flow. Operator statements differ as towhether the control valves (EFllA & B) were changed by them, but the
low level point in the OTSG's had not yet been reached, inferring that
the ICS should not have automatically modified the EFllA & B position.

130

14 min.

RCDT rupture disc blows at a RCDT indicated pressure of 192 psig,
dropping RCDT pressure to approximately 10 psig in 36 seconds.
Reactorbuilding pressure appears to jump one full psi.
At this time, to about
1.2 psig pressure rise seen in reactor building.

131

15 min.

RCS pressurg approximately 1275 psig with c 8 ld leg temperature of approximately 567 F. Saturation pressure for 567 F is 1185 psig.

2.

Plant strip charts

Note:

.1.

Inspection

132

133

1.
2.
3.

Reactimeter
Plant strip charts
Interview 5

1. Reactimeter

Note:

A review of the control room instruments and alarms anddiscussions with operating personnel relative to the RCDT reveals
the following:
1.

RCDT pressure is

not recorded.

2.

RCDT parameters do not alarm on the panels in immediate viewof the operators.

3.

To determine there is an alarm on the RCDT annunciators,
operators must clear all audible alarms on the front
panels.

Rupture Disc located on top of tank.
fill and overflow.

IA-23

Once blown,tank will

1. Construction
DrawingITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

134

15+ min.

(During entire transient
Reactor coolant pump related alarms received.
These alarms were
there was a slow reduction in indicated loop flows.
received at various times after this point until the pumps were shut down

1.
2.
3.
4.

Alarm printerReactimeter
Interview 61
Interview 5135

18 min.

Fuel handling exhaust monitors show small ramp increase in iodine
Reactor building exhaust shows factor of 10 increase inreading.
(Instrument location on the lower part of the vertical backreading.
panel would prevent operator standing at front panel from viewing thesetrends.)1.

Plant strip charts

Note:

136

139

1. System Descriptions

RCDT vent lines should isolate at 10 psig automatically; but after
the rupture disc blew, the vent line valve would reopen (6 psig)
venting the RCDT to the waste gas system. The waste gas system
header is maintained at 2 psig and as long as the reactor
building isolation had not occurred, the waste gas compressors
pumped gas from the reactor building to the gas surge tank.

137

138

RCS was effectively being degassed into the reactor coolant draintank and, when rupture disc blew, gas was released into thereactor building.

22 min.

Increase seen on SRM nuclear instrumentation and reactor manual trip
Radiation level on letdown monitor increases and returns
performed.
RCP full speed and vibration alarms now begin toto previous level.
appear and will continue until the RCPs are tripped.
Note:

SRM count rate increases are believed as reported by others tobe associated with increases in coolant void fraction, reducing

IA-24

1. Alarm printer2.
Plant strip charts

1. TMI Core Assessment,
EG&G, 4/6/79

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTIONThis evaluation is outside the scope ofshielding of detectors.
this investigation and is presented here for information only.
Note:

140

From approximately 20 to 60 minutes, system parameters generally
stablized at a reactor pressure 1015 psig, reactor coolant averagetemperature of 5500 F, and a pressurizer level of 380-395 inches.

I.
2.

Plant strip charts
Reactimeter141

23 min.

Low level alarm on "A" OTSG first to clear at 26.6".

1.

Alarm printer142

26 min.

Steam driven emergency feedwater pump turned off. Operator trying toestablish low level limit in steam generators but the level in "B" OTSG
kept creeping up.

1.
2.
3.

Alarm printerInterview 61
Plant strip
charts143

27 min.

Low level alarm on "B" OTSG clears at 26.6".

1.

Alarm printer144

27 min. +Operator requests computer print pressurizer relief and code safety outlet 1. Utility printertemperatures, hot leg and cold leg temperatures, loop pressuri~zer andpressurizer level.

145

28 min.

Operator secures feeding "B" OTSG by closing EFV-11B and then EFV-12B as(Later operators will shut down one electriclevel continues to rise.
Steam generator pressure being controlled manually through thepump.
turbine bypass valves.)

146

Note:

I.
2.

Operators have been dispatched to the auxiliary building to confirm 1.
pressurizer level indications and/or determine source of water which
The shift personnel were concerned with thehas filled pressurizer.
continuing high pressurizer level.

IA-25

Plant strip charts
Interviews 5, 14Interview 15

REFERENCEITEMDATE AND TIME
3/28/79

EVENT DESCRIPTION

147

30 min.

Operator requests memory trip review printout.
condensate pump running at shutoff head.)148

30 min.

They had been running
Diesels manually tripped locally by operators.
unloaded since original ES actuation and are not needed at this time.
When the diesels start on an HPI signal, they must be shutdown locally.
At this point, the operating staff did not reset the trip. The diesels
would not have started by any automatic or manual start signal.
This
condition lasted until - 5 1/2 hrs.

1. Alarm printer149

31 min.

Operator requests computer print Sequence of Events Review. There will
be several occasions during this sequence when the operator is pulling"Sequence" information from the computer.
In each instance, the information once typed is eliminated from the "Sequence" portion of the computer
memory.
Subsequent printouts then provide information from the time of
the last printout.

1. Utility typer
2.
Discussion withcomputer engineer

Alarm Printer shows Incore Thermocouple R-lO go offscale at 32 1/2
minutes.

1. Alarm printerUnit 1 Shift Foreman called Unit 1 Operations Supervisor to inform himof unit trip.

1.

1. Alarm printer2.
Reactimeter
3.
Interviews 5, 14

150

35 min.

(Printout shows one

151

36 min.

EFW pump "2B"

152

38 min.

Reactor building sump pumps turned off by auxiliary operator.
The twosump pumps had operated for 31 and 28 minutes, respectively, pumping

turned off; both OTSGs approximately 35".

IA-26

1.

Log typer

Interview 26

1. Alarm printer2.
AnalysisITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

approximately 8260 gallons of water to the auxiliary building. Shutdown
3.
of the pumps does not result in isolation of the pump discharge line. The
possibility of continued water transport to the auxiliary building, caused
by elevated containment pressure and considering tank elevations in theauxiliary building, appears unlikely.

Interviews 5, 15

153

Note:

At this point, the criteria for isolating associated with sump
pump operation EMOV were satisfied.

1. Emergency Procedure
2202-1.5

154

Note:

The automatic pumping of the RB sump to the auxiliary building
was not considered abnormal since operators knew RCDT rupture
disc had blown.

1.

Interviews 5,

Utility typer

155

44 min.

Operator requests computer print pressurizer level differential pressuresI.
in attempt to determine if level indication in error.
(Operators statedduring interviews that all level indicator outputs supported they actually
had high pressurizer level.
Operators stated all the sensors couldn't bewrong.)

156

50 min.

Steam generator "A" level starts trending down while generator "B" level
starts trending upward.

1. Interview 5
2.
Reactimeter

Superintendent of technical support, the on-call Duty Officer, reports tothe site. He had been called at approximately 10 min. and left home forthe site approximately 30 minutes later.

1.

Interview 5Operator requests computer print condenser pressure and steam drive emergency feedwater pump discharge pressure.

1.

Utility typer

157

158

52 min.

IA-27

15

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

159

59 mi n.

Condensate booster pump suction header low pressure alarm clears and slug
(The hot slugof hot condensate passes condensate temperature detector.
of water would be created by condensate pump IB pumping against a shutoff(Operators still attempting to establish
head for the last 59 minutes.)
It appears that
normal condensate flow to restore normal hotwell level.
operators have finally been successful in manually opening the polisher
bypass valve.)

1.
2.

Alarm printerInterviews 5, 14

1.
2.

Interview 5Inspection

160

Note:

Hotwell had been flooded when instrument air line for level controller had broken, probably during initial water hammer that
(This degree of movement
reportedly moved some pipes 2 to 3 feet.
unlikely based on inspection of area.)

161

Note:

By design, the polisher bypass valve on Unit 2 does not open auto- 1.
Attempts tomatically to limit polisher differential pressure.
open valve using valve controls in control room had been unsuccessful.
Valve motor reportedly unable to open valve with differential
pressure across it.

Interviews
109, 123

162

Note:

All polisher inlet and outlet valves had been found shut. Manual
handwheel for operating the polisher bypass was missing from the
The Shift Supervisor
valve, but found on top of ventilation duct.
was present and personally involved in manually opening this valve
locally.

Interview 109

163

Note:

1. Burns and Roe
This operation should have enabled normal reject flow to the condrawings
densate storage tanks and did automatically shift the emergency
2. Transient logs
feedwater pump suctions to the discharge of the condensate pumps.
Operators have yet to discover broken air line to AUTO-REJECT valve(it fails shut) and will wonder why hotwell level doesn't go down.

IA-28

1.

ITEM

DATE AND TIME
3/28/79

164

EVENT DESCRIPTIONREFERENCE

Note:

1. Interviews

A leak had developed on the suction piping of condensate booster
pump 2A.

109,

123

PLANT STATUS
The RCS was near a saturation temperature-pressure relationship at 550°F and
1050 psig. The indicated loop flow rates had decreased from about 69
million pounds per hour to approximately 47 million pounds per hour andcontinued to decrease.
The rea~tor building temperature and pressure
had increased from 120°F to 170 F and 0 psig to 2.5 psig, respectively,
as a result of releasing contents of the reactor coolant drain tank to
the reactor building atmosphere.
The EMOV was still open and makeup
pump 1A was operating.
The operator was having difficulty controlling
the level of steam generator B. Emergency feedwater valves (EF-VIIB
and EF-VI2B) were shut and the operator was admitting feedwater tosteam generator B by cycling emergency feedwater crossconnect valve(EF-V5B).

165

166

1 hr.

167

61 min.

168

On-call operating engineer called to come to site.
approximately 45 minutes later.
Circulating water pumps B, C, D, E off.
shows small increase in level.

Note:

He arrived at site

Letdown line radiation monitor

1.

1. Alarm printer2.
Interview GPU-I
3.
Interviews
5, 9

The circulating water pumps were turned off to automatically switch
steam dump to the atmospheric relief valves to prevent flooding of thecondenser or an uncontrolled shift to those same valves on loss of con-

IA-29

Interview 5

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
denser vacuum; the latter concern relating to inability of Unit 1 to
supply adequate gland sealing steam pressure from their auxiliary
boilers.

169

Note:

The intermediate cooling water radiation monitors (1082 and 1083)
alarm were believed to be due to high background radiation levels
near the containment sump, which was being overflowed from thereactor coolant drain tank rupture disk failure.
It was believed
normal RCS coolant activity levels would be sufficient to cause
these alarms.
Further, the operators were aware of a design change
request to reduce the sensitivity of these instruments to reduce
spurious alarms.
The operator did verify no increase in the intermediate closed cooling surge tank level to ensure no primary system
to closed cooling water system leak.

170

62 min.

Alarm printer unavailable starting at this time. Not returned to service I .
until 3 hr. 51 min.
Operator interviews indicate that there was malfunc2.
tion of the paper feed.
Review of other alarm printer outputs for various
other periods, shows evidence of paper feed problems.

Interview 5
Record review

171

71 min.

Temperature transient on cooling water indicates reactor building air
cooling coil B now operating. The reactor building cooling was initiatedwith emergency river water cooling through the reactor building cooler.

2.
2.

Utility typer
Interview 14

172

Note:

Operators were unable to account for increased reactor building
temperature.

173*Note:

This data point based on transient in alarm printout of outlet
temperatures and may be erroneous conclusion.
Further alarms

IA-30

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

will be shown elsewhere in this sequence.
However, the investigation has not determined an explanation for this recurring
alarm. They are shown for the sake of completeness should a later
investigation determine their cause.
It should also be noted thatno similar alarms are received from any of the other cooler units
throughout the accident.
174

72 min.

Operator requests computer print those conditions currently in alarm
related to RCPs.
(This action believed associated with desire for more
information to base:decision on which RCPs to be tripped.)

175

Prior to 74 min.

Reactor coolant pump flow indication has been steadily decreasing andindicate flow in loop "B" at 74 min. approximately 60% of normal value.
Operators report receiving vibration alarms on the reactor coolant pumps(RCP).
Chart recorders for source range nuclear instrumentation show
increasing average count rate with increasing amplitude of oscillation.
Operators reportedly recognized that they were violating 4 RCP pressuretemperature limits and the pin compression curves. The Shift Supervisor
and the Technical Superintendent specifically reviewed these curves.

176

177

74 min.

RCP's in B loop chosen to be tripped to be able to maintain maximum pressurizer spray capability which comes from "A" loop.
RCP lB and 2B
tripped.
(Utility typer indicates that RCP 2A also tripped.
This inconsistency believed to be associated with inductive pickup in alarm wiring
or bouncing contacts.
Operators reported earlier examples of thisindication were experienced with RCP 2A known to be running.)

IA-31

1. Utility typer

I. Reactimeter
2.
3.

Interview GPU-I
Plant strip chart

1.

3.

Interviews
53, 5Operating Procedure
2202-1.3
Transient Logs

I.
2.
3.
4.
5.

Interview GPU-20
P&ID
Utility typer
Record review
Interview 6

2.

ITEM

DATE AND TIME
3/28/79

Note:

178
179

74 mi n.

RCPs were running without adequate net positive suction head,
jecting these pumps to potential damage from cavitation.

Loop "B"
Th:
Tcb
Tcb:
MU Flow
OTSG "A" Level:
OTSG "B" Level:
Condensate Flow:
74 mimn. +sub-The hourly plant log typer displayed the following information as being
applicable for this time:
RCS Flow:
Loop "A"
Th :
Tc :
Tca:
a

180

REFERENCEEVENT DESCRIPTION

1.

Log typer

89 MLBH
543 F
540 0F539 0F
543 OF
540 0F
540 0 F
118 GPM
34 inches
70 inches
513 KLBH

1.
Operators reported reactor coolant flow oscillations decreased immedi2.
ately following pump trip. Moreover, Source Range (SRM) instrumentation
3.
Average SRM
count rate oscillatory behavior significantly decreased.
count rate was approximately three times the value that would be expected
at this time after reactor trip. RCS pressure indicates an upward pressure
trend, and "B" steam generator pressure drops from approximately 960 psig
to 140 psig over the next 18 minutes.

IA-32Interviews
Plant strip charts
ReactimeterITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

181-. 75 min.

RCS sample yields 700 ppm Boron.
Sample required by emergency procedures
within 2 hours of trip. Sample results were questioned by technician andhe started to get additional sample.
(Calculated shutdown margin at
this time for the reactor was -.6% reactivity.)

1. Interviews
2.

58, 5
Emergency Procedure
2202-1.1

182

Note:

183

Station Manager calls Unit 2 (call could have been as late as 95 min.)to ascertain plant status not having heard further since original notification. When informed of status, Station Manager directs Duty Officer
to establish conference call between selected staff.
(This conference'tall occurs at \ 2 hrs.)

1. Interview 1

Apparent reduction in boron concentration believed associated withdilution of liquid phase of sample by condensation of steam insample stream.
However, this was not a conclusion drawn by theoperator staff at that time.

184

80 min.

Factor of 7 increase in radiation level as shown on letdown line
monitor.

1. Plant strip
charts185

81 min.

Operator requested computer print
Pressurizer Surge Line Temper 8 ture (514 F),
EMOV Outlet Temperatures (283 F),
0
Code Safety Outlet Temperatures (211 F ý 219°F),
Pressurizer Spray Line Temperature (497 F), andCondensate Pump Outlet Header Pressure (165 psig).

1. Utility Typer

186

Note:

EMOV outlet at least 64°F hotter than code safety temperatures.
(Why EMOV block valve was not shut at this point is unclear.)

IA-33

ITEM

DATE AND TIME
3/28/79

187

188

1.

Interview 80

The "B" steam generator was isolated at this point. Operators had
observed increases in reactor building pressure and noted that pressure
They believed steam wasin generator "B" was 300 psig lower than "A".
leaking into the reactor building from "B" OTSG secondary side. Afterisolating steam generator "B", they note building pressure leveling off
and "B" generator level trending upward, so they secured the third EFW
valve to the generator, EFV-5B.

1.
2.
3.
4.

Control room logs
Plant strip
chartsInterview 5Transient Logs

1.
2.

Reactimeter
Analysis

1.

Interview 58

Note:

189

190

Operators knew the EMOV and/or the code safeties leaked prior
to the event. They reportedly had to add approximately 2400
gallons of makeup, per shift to the RCS and the tail pipe
Under these conditions,temperatures were always above ambient.
operators may have been hard-pressed to decide what temperatures
constituted abnormal indications given their recognition of an
earlier lifting of the EMOV.

Note:

87 min.

90 min.

REFERENCEEVENT DESCRIPTION

Reverse flow should have occurred in the "B" loop as a result
No subof continued operation of the pumps in the "A" loop.
stantial changes in steam generator level or pressure occurred
over the next hour.
It appears that reverse flow of RCS coolant
at a temperature of approximately 550 0 F was severely curtailed
through the "B" generator, as was indicated by the rate ofSome flow was occurring aspressure drop in that generator.
indicated by loop "B" RTDs tracking those of loop "A".

RCS sample yields 400-500 ppm boron and 4.0 uc/ml; a further drop inboron concentration and a factor of 10 increase in activity.
There is no clear evidence that this increase in activity was related
The possibility of a crud burst or an iodine
to any fuel failure.

IA-34

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

spike cannot be discounted.

191
192

91 min.
93 min.

193

194

94 min.

Operator requests Computer printout Sequence of Events Review.

1.

Utility typer

Operators report increasing HPI flow; RCS pressure peaks and dropsrapidly; SRM & IRM indication spike upward, and injection flow was maintained at 150-200 gpm/loop.

1.

Interviews
5, 9

Note:

1. Plant strip charts

The actual peak in reactor building pressure appears to occur
here, versus immediately following isolation of "B" OTSG, as
the operators believed.

"A" OTSG appears to boil dry, again,
RCS temperature trends.

based on OTSG level,pressure and

1. Reactimeter
Plant strip charts
2.

195

96 min.

Operator again requests computer print pressurizer Level Indication(372.9").

1. Utility typer

196

Prior to 101 min.

Approximately 5-10 minutes after the trip of the "B" loop RCP's, SRM
count rate instability increases again as well as continuing an upward
trend. IRM-current also begins to show upward trend. Operators reportloop flow instability increasing again, and the indicated loop flow
continues to show a decrease.
Operators asserted during interviews
they were concerned about a "seal failure LOCA" and decided to go on
natural circulation.
Operator stated that he started "emergency
boration" during this period, based on SRM increases and fear of arestart accident.

1.

IA-35

2.

Plant strip charts
Interview 9

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

197

Note:

Conditions for. initiation of natural circulation cooling of
the RCS did not exist at this point.

1. Procedure 2102-3.3

198

Note:

Operator training and plant procedures require securing RCPs on
sustained high vibration.

1. Abnormal Procedure
2203-1.4
Interview 80
2.

199

Note:

Tech. Supt. Transient Log indicates tripped last two RCPs due
to low suction pressure.

1. Transient logs

200

Note:

The indicated increase in NIs has been reported elsewhere to be
due to increased neutron leakage from the core as a result ofof increases in void content. This conclusion may be subject tosome question.. Review of the outputs of 2 IRMs suggests a
The twodifference in gamma compensation of the two chambers.
chambers behave differently during this period which implies that
their output may be dependent, to some degree, on changes in the leakage
gamma flux from the core. The chamber which appears undercompensatedexhibited the behavior described while the chamber which appears
overcompensated did not.

201

Note:

Emergency Boration action consistent with procedure requirements.

202

101 min.

"A" loop RCPs tripped. SRM count rate spikes upward to peak at least
one decade over count rate expected following a normal reactor trip.
All radiation monitors exhibiting substantial ramp increase.

IA-36

1.

Procedure 2203-1.1

1.
2.
3.
4.
5.

Utility typer
Reactimeter
Plant strip charts
Interview 9Emergency Procedure
2203-1.3

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

203

Operators stated they did not believe that natural circulation had been
1. Interviews
established due to the differential temperature across the steam genera9, 15
tor and the low steam generator pressure with minumum heat removal, if any.

204

Note:

RCS was at 1020 psig and 5350F; well outside prerequisitepressure/temperature conditions for establishing natural
circulation.

205

101+ min.

RCS Pressure continues to drop.
significantly.

206

102+ min.

The SRM count rate level, followed by intermediate range monitor
output, start upward trend. Over the next 15 minutes, SRM countlevel increases 2 decades without exhibiting previously observed
latory behavior.
Operator reports he "emergency borated" again,
on fear of restart accident.

207
208

209

Note:

103 min.

SRM counts and IRM current drop

(IRM)

rate
oscilbased

Operating Procedure 2102-3.3

1.
2.

Reactimeter
Plant strip charts

1.
2.

Interview 5
Procedure 2203-1.1

1.

Reactimeter
Plant strip charts
Interview 61
Multipoint recorder

Emergency boration action consistent with procedure.

RCS hot and cold leg temperatures begin t 8 diverge widely.
Loop "A"hot leg temperature begins to rise to 620 F (off 8cale within
14 minutes) and cold leg temperature drops to 150 F (apparently
from effect of HPI water).
Loop "B" divergence is much slower, demonstrating little
coupling between loops and reactor.
ýhe operators
reported average RCS temperature "stablilized" at 570 F.
Note:

1.

The average RCS temperature (Tav) recorder trace could not be
relied upon by the operators for the period fromhours to16 hours.
Tav stabilizes at an indication of 570 F as soon as
both narrow range T hot and T cold instruments reached their

g

IA-37

2.
3.
4.

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
display limits (620°F and 520°F,

210
211

Note:
112+ min.

respectively).

RCS Hot leg temperatures already appear superheated.

Per procedure, operator begins feeding "A" steam generator to reachOperator states he is still controlling steam generator
50% level.
Operator intends to remove decay
manually with turbine bypass valves.
heat via the OTSG using natural circulation.

1.

2.
3.
4.

Interview 61
Plant strip charts
Interview 5Operating Procedure 2102-3.3

PLANT STATUS
212

213

114 min.

214

2 hrs.

215
216

2+ hrs.

1.
The RCS had no forced convection cooling. All RCPs were stopped.
2.
The RCS average temperature and pressure were approximately 5200F
and 750 psig. Makeup pump ]A was operating. The operators were attempting
OTSG
to establish natural circulation flow to cool the reactor core.
B was isolated because of a suspected leak to the reactor building.

Reactimeter
Plant strip charts

Operators requests computer printout Sequence of Events Review.

14

Utility typer

Nuclear Engineer called to site to gather information required forSome question then whether the reactor
standard post-trip report.
had experienced a restart based on SRMs.

1.

Interview

Third boron sample indicates approximately 400 ppm.

1.

Interview GPU-19

Conference call established between Unit 2 Technical Supt. (Unit 2
Control Room) and Station Manager, Vice President of Generation and
B&W site representative (at their homes), lasting approximately 38

1.
2.

Interview 53
Statement'Plant
Manager

IA-38

ITEM

DATE ANDTIME
3/28/79

217

218

EVENT DESCRIPTIONREFERENCE

minutes.
Knew trip was abnormal since RCPs were off and unable to drawpressurizer bubble.
Blown rupture disk and water on floor were notsurprising since this had happened before. The condition of the EMOV
block valve was questioned and reported to be shut. Group decided
need existed to restart an RCP and all should report to TMI.
(Conference
call had been initiated by Station Supt. following discussion with Unit 2
Tech. Supt. (on-call Duty Officer) around 1 hour 15 minutes into event.
The Unit 2 Tech. Supt. had been on site since 50 minutes into the event,following his call to the site shortly following the initial trip.)

3.

Note:

2 hrs.

1 min.

B&W representative reports that the fact that EFV-12A & B were shut 1.
with no feeding to the OTSGs for 8 minutes was not discussed thefirst day.
It was discovered by B&W representatives later, while
delogging the reactimeter. However, it appears that B&W in Lynchburg
had this information as it was given to them by Operations Supervisor
after he had made contact with the site.
Information was thatemergency feedwater had been delayed approximately 12 minutes.

The hourly plant log typer displayed the following information as being
applicable for this time:

RCS Flow:
Loop "A"
Th:
Tc
Tc a
Loop "B4Th:
Tcb:

3 MLBH57R OF? F
052g OF
? F

IA-39

1.

Telephone bills

Interview 53

Log typer

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTIONTc
MU Flo:OTSG "A" Level:
OTSG "B" Level:
Consensate Flow:?o0F
97 GPM
173 inches
79 inches
318 KLBH

219

2 hrs.

5 min.

Operator either stops or reduces signif"A" OTSG level reaches 50%.
icantly the feed rate to that OTSG- Appears operator alternately feeds
and steams down "A" OTSG level for next 50 minutes.

1.
2.

Plant strip charts
Reactimeter220

2 hrs.

14 min.

Operator requests computer printout Sequence of Events Review.

1.

Utility typer

221

2 hrs.

18 min.

Operator rgquestsoComputer Brint EMOV and code safety outlet temperatures (229 F, 190 , and 194 , respectively).

1.

Utility typer

222

1.
Electromatic relief valve is now isolated by operator by closing the
Operators had noted tail pipe temperature on relief 2.
block va ve (RC-V2).
valve 35 F higher than others and believed valve to be leaking. Operators
noted a drop in reactor building pressure after closure of the block
the block valve.

Utility typer
Interview 5

223

The pressure change in the reactor building was more marked than when
"B" OTSG was isolated. The plant operations group decided that the "B"OTSG not have a leak from the shell into the reactor building. The "B"OTSG still had a level.

1.
2.

Interview 5
Plant strip
chartsWith closure of EMOV, RCS pressure begins increase from low point of
660 psig and reaches 1300 psig. During this period from 40 minutes,
At some time
no change in pressurizer level (about 290") is observed.
prior to this point, the isolation valves for the core flood tanks had

1.
2.

Reactimeter
Plant strip
chart

224

2 hrs.

18 min.+

IA-40

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

been isolated to prevent what was judged to be an unneeded discharge ofwater from the core flood tanks. Time of isolation and reopening has
not been determined.
225

Prior to this point, standard post-reactor-trip boron samples takenfrom the reactor coolant system had been analyzedwith no problems
indicated or detected relative to radiation levels.

I.

Interview 7

226

Fifteen to twenty people in control room at this time.

1.

Interview
12

Discussions were held considering an entry into containment to manually vent the RCS hot legs.
The occurrence of radiation alarms halted
further consideration of this option. Area radiation monitor (ARM)
alarms were received at the sample station and the letdown sam~le lineup
was secured.
Loop "A" hot leg temperature peaked at about 750 F.

1.

Interviews

2.

58, 105, 97
Multipoint
recorder

227
228

229

230

231

2 hrs.
2 hrs.

2 hrs.

18 min.30 min.

35 min.

2 hrs.

40 min.

Operator begins feeding isolated "B" OTSG to 50% level.
Intent is tounisolate OTSG and improve natural circulation by using both OTSGs toremove core decay heat through natural circulation.
Operator may haveopened the turbine bypass isolation valve (MS-Vl5B) at this time, also.
These actions are consistent with procedure requirements.

1.

Operator initiates emergency boration based on increasing NI indication, low boron sample results report and calculated shutdown marginof only 2.4% reactivity.

1.
2.

Note:

1.

These actions consistent with requirements of emergency procedure.

IA-41

2.
3.

Plant strip
chart
Reactimeter
Procedure2102-3.3
Statement Plant
ManagerInterview 5Emergency Procedure
2203-1.1

ITEM

DATE AND TIME
3/28/79232

2 hrs.

45 min.

235

2 hrs. 46 min.

2 hrs.

48 min.

236
237

238
239
240

REFERENCE

Makeup pump IC tripped by operator action according to GPU sequence.
However, the GPU sequence does not indicate a start time. The last
identified change of status was at 4 min. 38 sec. when the pump was
No other status change is noted up to 1 hr. 13 min. 27 sec.
tripped.
when the Alarm Printer stopped working.

1. GPU sequenceof 7/16/79

NOTE:

233
234

EVENT DESCRIPTION

2 hrs.

2 hrs.

52 min.

2 hrs.
2 hrs.

50 min.

53 min.

54 min.

This information not supported by Sequence of Events Review
printout from the plant computer.
Utility
Typer

Alarm data from 1 hour 13 min.
The utility typer stops and restarts.
to 2 hours 48 min. is lost. Alarm printer had been lost earlier due
to paper feed problem.

1.

RCS loop "A" hot leg at 800 F and increasing.

1. Multipoint recorderHigh radiation readings obtained in hot machine shop. Hot machine shop
high radiation alarms were received in Unit 1 control room.

1.

Plant strip
chartsOperators finally solve condensate reject problem and pump condensor
Had found automatic reject valve closed due tohotwell level down.
broken air line.

1.

Alarm printerOperator places RCS loop "B" cold leg tempature in analog recorder
trend (#2).

1. Alarm printerOperators finally get condensate reject valve open with technicians,
help which allows them to drop hotwell level.

1. Alarm printerElectricians jumper RCP start interlocks ("K" relays) to allow pumpstart. Operators try to start IA, IB & 2A RCPs without success.

1.
2.

IA-42

Alarm printerInterview 61

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

Operators start RCP 2B and it runs for 19 min. Operators state
they received all radiation alarms.
Operators started RCP because
they were not sure that natural circulation had been established and
NI counts were trending up.
During this operation of the pump, RCS
pressure went up to 2200 psig and SRM counts and IRM current dropped
significantly.

3.

241

Groups 1-5 of the pressurizer heaters are tripped and remain in a*tripped state for 1 1/2 hours.
Throughout the course of the incident,
difficulty is experienced with multiple pressurizer heater trips, presumably due to moisture content in the reactor containment; a problem
previously experienced on Unit 1 at this site and other facilities.
As of this time, no more than eight groups of heaters are operable.

I.
2.

Alarm printerInterview 53

242

NOTE:

1.

Interview 107

ITEM

243
244

2 hrs 54 min.
2 hrs.

54 min.

The Shift Supervisor felt that the pressurizer heaters shouldhave been more effective in recovering RCS pressure and dispatched the the Shift Foreman to the pressurizer heater control
cabinet to confirm that their breakers were shut.
The Shift
Foreman verified breakers were closed; noted the area wasextremely hot and humid (believed due to proximity to main
steam relief lines); noted local vent. fans were tripped (dueto fire heat sensor trip); and he defeated the trips and restarted fans.

Operator places RCS loop "A" cold leg temp.
(#2).

in analog recorder trend

Incore thermocouple data collected by computer that will not be printed
for at least 1 hour:

IA-43

4.

Plant strip
chart
Interviews 5,
105

1. Alarm printerI.

Alarm printerITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
Thermocouple Location

Temperature (0F)
624
596
625
654
688
577
584
599
624
671
682
700

TC 7B
TC I-HTC 12-0
TC 14-M
TC 14-D
TC 13-C
TC I1-L
TC 10-C
TC 2-GTC 13G
TC 3F
TC 5D
NOTE:

245

The majority of readings from this system indicate data points havejust failed or are off-scale high. The data point values, displayed
above, indicate these thermocouples have returned on scale andthe computer has just noted their return.

(This automatically shifts "A"
Started circulating water pump IB & IE.
& "B" OTSG pressure controls to the turbine bypass valves.)

1. Alarm printer247

SRM and IRM indications drop significantly.

1.

Plant strip chart

248

Operator opens and then closed "B" OTSG' MSIVs (MS-V4B&7B), within 7(Operator actions in cycling these valves have not been
seconds.
Operators speculate this event may represent inadvertent
explained.
actions associated with attempts to cope with the rapidly deteriorating

I.
2.
3.

Alarm printerReactimeter
Interviews 5,
189, 111

246

2 hrs.

56 min.

IA-44

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

situation, which included offgas radiation monitor alarms.
It should
be noted that operation of these MSIV's does not affect steam flow to
the condenser if the turbine bypass controls are in "Hand," which theyreportedly were at this time.)

4.
5.
6.

Multipoint
recorderBurns and Roe
drawings
Transient logs

249

HP Technician reports over page that letdown sample lines from Unit 2
was reading 600 R. Auxiliary building evacuated.

1.

Interviews
13, 105

250

Unit 2 Shift Supervisor and Technical Superintendent then declared asite emergency.
Declaration based on radiation alarms in more than one
area.
Notification of offsite authorities begins.

1.
2.

Statement Plant
ManagerInterview 48

North Bridge Gates being closed due to Guards responding to the establishment of site emergency.

1.

Interview 8

Secured liquid release from Unit 1 and Unit 2 Operators verified
appropriate valve closed.
Action was taken by Unit 1 operators at
direction of Unit 2.

1.

Interview 15

Radiation level in vicinity of Unit 2 sample area, where lines wereon recirculation for another boron sample, reportedly showed marked
increase.

1.

Interview 7

Fifty to sixty people estimated to be in control room at this time withnoise level very high.

1.

Interviews 7,
12, 105

Operator places RCS Loop "A" cold leg temp.
Both RCS loop hot leg temps. go off scale.

1. Alarm printer

251

252

2 hrs.

2 hrs.

253
254

56 min.

3 hrs.

57 min.

IA-45

in analog recorder trend (#3).

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

255

3 hrs. 1 min.

The hourly plant log typer displayed the following information as
being applicable for this time:

1.

Log typer

Eighteen to twenty-four people reported to be in control room at thistime. Not known if excess people ordered from control room or if
differences previous estimate related to individual estimates.

1.

Interview 48

257

Operators trying to recover pressurizer level indication which was now
off scale high.

1.

Interview 48

258

Guards at North Gate delay, then allow B&W site operations manager
access to site. He was arriving at site at Station Manager's request
after earlier conference call which ended at - 2 hrs. 58 min. He reached
the control room at about 3 hrs. 45 minutes after contacting B&W in
Lynchburg.
Guards were sending other people arriving at site to
Observation Center.

1.

Interview 48

RCS Flow:
Loop "A"
Th:
Tc :
Tc a:

a

Loop "B"Th:

3 MLBH572 OF
? F
? o

F

52

OF

Tc:?2FTc~b. ;O0F
MU Flow:
OTSG "A" Level:
OTSG "B" Level:
Consensate Flow:256

3+ hrs.

125 GMP245 inches
220 inches
0 KLBH

IA-46

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

259

3 hrs.

2 min.

RCS loop "B" hot leg temperature reaches 800 F.

1. Multipoint
recorder260

3 hrs.

4 min.

Region I Answering Service receives call from Licensee.

1. Telephone bills
2.
Statement Plant
Manager3.
Region I notes

Station Manager arrives at site and proceeds to control room to assume
duties as Emergency Director.

1.

Interview 1

OTSG "B" turbine bypass isolation valve (MS-Vl5B) is closed by operator,
completing the isolation of "B" OTSG . IRM current returns on scale,increasing.
Operator secures feeding "B" OTSG.

I.
2.
3.

Alarm printerStrip charts:
Reactimeter

261

3 hrs. 5 min.

262

263

3 hrs.

11 min.

EFW pump (EF-P-2A) tripped by the operator.
At this point, OTSG levels
were in the 55% to 65% range and no further feed appeared necessary.

1.
2.

Alarm printerReactimeter

264

3 hrs.

13 min.

SRM counts and IRM current peaked; both then dropped significantly aspressure peaked.
IRM current drops off scale just prior to the RCP
being tripped.

1.
2.

Alarm printerReactimeter

265

High temperature noted on tail pipe of EMOV.
Rate of pressure increase
in reactor building suggests operator has opened EMOV block valve and
EMOV, in manual.
RCS pressure and pressurizer level both drop (RCS pressure to 1975 psig; pressurizer level - 300").
A pressure spike of 5.5
psig also seen in RCDT.

1. Alarm printer2.
Plant strip
charts3.
Reactimeter

266

Note:

1.

Pressure increase in reactor building stops about 3 minutes later,suggesting the EMOV discharge path is again isolated. A quick

IA-47

Plant strip
chartsITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

bump.upward in reactor building pressure about 4 minutes later,The
indicated another short RCS blowdown using this path.
Fromvalidity of the RCDT pressure spike is very doubtful.
approximately one hour on, the RCDT pressure, as recorded on thereactimeter, approaches zero and starts to record negative values
No correlation can be
for much of the remainder of the accident.
drawn between RCDT pressure and reactor building pressure which it
should now be tracking in some general fashion.
267

RCP 2B tripped by operators after vibration alarms received and motor
Operator notes no change in Tave; still atonlý drawing 100 amps.
570 F (a false indication).

268

269

3 hrs.

270

3 1/4 hrs.

271

14 min.

2.

Reactimeter

1. Alarm printer2.

Utility typer

Operator stated that coincident with tripping of RCP further radiation
alarms received and they started following radiological emergency
procedures.

1.
2.

Interviews
Plant strip
charts

Received high temperature alarms on code safety outlets.

1. Alarm printerControl building evacuated with exception of.control room.
reported to be clear by operator who has just returned.

Building

After RCS pressure holds at 1975 psig for approximately 5 minutes, a
rapid pressure decrease begins and RCS pressure reaches 1500 psig.
A RCDT
Pressurizer level dropped during this period to 225 inches.
pressure spike of 11 psig is seen at this time and reactor building
pressure increases from 1 to 3 psig.

IA-48

I. Interview 7
2.

Radiation surveylog

1. Operating logs
Reactimeter
2.
3.
Plant strip
chartsEVENT DESCRIPTIONREFERENCE

16 min.

Answering Service attempts to reach Region I Duty Officer, but he has
already left for Region I Office.

I.

Region I
notes

18 min.

IRM current ("14) returns on scale, increasing.

1.

Alarm printerITEM

DATE AND TIME
3/28/79272

3 hrs.

273

3 hrs.

PLANT STATUS

274

275
276

1.
After attempts to establish natural circulation failed, the operator
However, based on a no flow indication RCP 2B was stopped 2.
started RCP 2B.
Superheated steam/gas was present in the reactor vessel
after 19 minutes.
head and RCS hot leg. Both reactor coolant system hot leg temperatures
The RCS cold leg temperwere off scale high (.e. greater than 620F).
OTSG "B" wasatures were about 375 F for loop "A" and 330 for loop "B".
isolated due to a suspected RCS to OTSG leak. OTSG "A" pressure control was by means of the power operated emergency main steam dump
An attempt was in progress to control pressurizer
valve "A" (MSV-3A).
pressure and level with the EMOV.
3 hrs.

20 min.

Reactimeter
Plant strip
chartsAnswering Service signals Region I Duty Officer beeper but beeper
reportedly did not ialarm.

1.

Region I
notes(Believed to have
Operator secures borated water recirculation pump.
been in operation prior to event to clean up borated water storage'tank).

1.
2.

Alarm printerInterview I1

IA-49

EVENT DESCRIPTIONREFERENCE

277

ES manually initiated by operator. Makeup pump IC starts. Makeup pumps
Pressure begins to
1A & IC now running with MU-V-16 valves wide open.
drop.

1. Alarm printer278

RCS loop "A" hot leg temperature drops,

279

NOTE:

Makeup pump lB has been off since ES initiation at 2 minutes
after the start of the incident.

280

NOTE:

RCS loop "A" hot leg temperature varies between 680 F to 7600F
Loop "B" will follow loop "A", but aboutover next 6 hours.
60 F hotter.

ITEM

281
282

DATE AND TIME
3/28/79

3 hrs.

21 min.

3 hrs. 24 min.

IRM current (^-14)

1. Multipoint
recorder

800°F to 710°F.

1. Multipoint
recorder1. Alarm printerdrops off scale.

General Emergency declared by Station Manager.

1. Statement Plant
Manager
Region I tapes
2.

283

3 hrs. 26 min.

Pressurizer high level alarm
Pressurizer level and pressure dropping.
Pressurizer pressure has dropped to 1508 psig.
clears indicating 254".
Pressurizer surge line temperature has returned to normal at 581 F.

1.
2.

284

3 hrs. 27 min.

ES reset by operator.

1. Alarm printer285

3 hrs. 30 min.

Borated water storage tank low level alarms received at 53.03'

IA-50

and 53.06'

Alarm printerPlant strip charts

1. Alarm printerITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

286

NOTE:

Calculations based on these BWST levels indicate the average net
flow rate into the RCS from the BWST has only been about 75 GPM
over this 3 1/2 hour period.

287

Operator places RCS loop "B" cold leg temp.

in-analog recorder trend

1.

Alarm printer

(#2).
288

3 hrs. 31 min.

First code safety outlet temperature high alarm returns to normal at
192 F, indicating EMOV had been shut sometime prior to this.

1.

Alarm printer289

3 hrs. 32 min.

Makeup tank radiation level 3 R/hr and auxiliary building basement reported as flooded.
Auxiliary building airborne activity at about IOE-8
uC/cc.
Spent fuel demineralizer area monitor reading 250-900 mr/hr. SRM
count rate found to have increased by about a factor of 3.
(Aboveradiation level readings from hallway of elevation 305', auxiliarybuilding.
Basement referred to above is elevation 281', auxiliarybuilding.

1.

3.

Radiation survey
logs
Plant strip
chartsInterview 7

2.

290

3 hrs. 33 min.

Pressurizer high level alarm received at 267",

I.
2.

Alarm printerPlant strip charts

291

3 hrs. 35 min.

Second code safety outlet temperature alarm returns to normal at 1930F.
I.
Operator starts emergency feedwater pump 2A.
Pump discharge initially2.
rises to shutoff head conditions.
Operator has started pump to reestablish feeding OTSGs.
Main condenser vacuum is seen to begin degrading.

Alarm printerPlant strip charts

292

3 hrs. 37 min.

Makeup pump 1C is tripped; presumably by operator,
surizer level.
Pump IA is still in operation.

IA-51

rapidly increasing.

based on high pres-

1. Alarm printerITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

293

3 hrs.

SignalAnswering Service signals Region I Duty Officer bepper again.
received by Duty Officer who is now in heavy traffic and close to
office.
He decides to proceed to office to answer call.

1.

Site calls Region I Office to inform them of declaration of General
Emergency.

1. Telephone bills

294

38 min.

3 hrs. 40 min.

Region I notes

295

3 hrs. 40 min.

Reactor building air sample line isolated after reported as blowing air
into auxiliary building.

1.

Radiation survey
log

296

3 hrs. 41 min.

Code safety outlet high temperature alarms return at 202 F and 205 F.
Rate of reactor building pressure increase indicates EMOV discharge
path is again open, with EMOV is manual and block valve open.
Increase in RCDT pressure also seen for approximately 20 minutes.

I.
2.

Alarm printerPlant strip charts

297

3 hrs. 42+ min.

Answering Service calls
Not receiving response from Duty Officer.
Region I Deputy Director, in accordance with procedure, and learns
the Deputy Director is enroute to the Regional Office.

1.

Region I notes298

3 hrs. 43 min.

At this time, there is an indication of rapidly increasing SRM countThis is the only
rate concurrent with a rapid pressure increase.
All other increases in SRM
time this occurs during the accident.
count rate appears to correlate to periods of apparent increased
voiding associated with decreasing RCS pressure and/or loss of heatsink.
Further, the alarm printer indicates a proliferation of "BAD"
These indications are conincore flux and temperature indications.
sistent with an assumption that a core geometry shift has occurred.

I.
2.

Alarm printerPlant strip299

3 hrs. 43 min.

Operator begins raising "A" OTSG level back to 50% to 55% range.

1.

Plant strip

IA-52

ITEM

300
301

DATE AND TIME
3/28/79

3 hrs. 44 min.
3 hrs. 45 min.

301A
302

3 hrs.

45 min.

303

3 hrs.

304

3 hrs. 54 min.

305

50 min.

3 hrs. 55 min.

REFERENCEEVENT DESCRIPTION

2.

chart
Reactimeter

Answering Service receives call informing Regional Office that a
General Emergency has been declared.

1.

Region I notes

StationB&W Site Operations Manager is now in Unit 2 control room.
Manager was in control room having arrived about 3 hours, 5 minutes
into the event. Operators reported they had run a RCP as directed with
100 amp indication. Oper3tors report that "many" pressurizer heaters
are "out of commission."

1.

Interview 53

B&W Site Operations Manager establishes first contact with B&W offices,
informing them of plant status.

1.

B&W Telephone
Log

RegionRegion I telephone operator takes over from Answering Service.
I operator is informed of General Emergency and she notifies Reactor
Operations and Nuclear Support Branch Chief.

1. Region I notes

Reactor Operations
Region I Director informed of General Emergency.
(RO&NS) and Fuel Facility and Materials Safety (FF&MS) Branch Chiefs
call site (informed that Site Emergency was declared at 6:45 a.m. andGeneral Emergency at 7:45 a.m.).

1. Region I notes

Operator secures or significantly reduces feed rate to "A" OTSG.

1.

Station Manager directed outside agencies be contacted to report

IA-53

2.

Plant strip
chart
Reactimeter

1.

Interview 53

ITEM

DATE AND TIME
3/28/79

incident.

306

3 hrs. 56 min.

Established open line to NRC:RI.

ES and reactor building isolation initiated by high reactor building
pressure (--3.58 psig). Makeup pump IC starts automatically, joining
makeup pump IA in injecting through now wide open MU-V16 valves.
Intermediate closed cooling pumps ]A & IB trip.
Note:

307

REFERENCEEVENT DESCRIPTION

Reactor building pressure trace suggests EMOV is shut by operator
at this point, but EMOV block valve is left open.

2.

Statement Plant
Manager

1. Alarm printer

1. Plant strip charts

Operator reES and reactor building isolation defeated by operator.
starts intermediate closed cooling pumps to ensure RCP seal and letdown
flow cooling.

1. Alarm printer

By this time, pressurizer level is approximately 380 inches with areactor pressure of 1500 psig.

1.

Reactimeter

310

Detector shielded with 2 inches of lead located in containment dome
reading 200 R/hr.

1.

Region I
Incident
Messageform

311

Further degradation of main condenser vacuum is noted.

1.

Plant strip charts

312

Station Manager requests B&W Site Operations Manager and other senior
It wassupervisors caucus with him in Shift Supervisor's office.
decided to try another RCP start since pressure was high enough tosatisfy NPSH requirements.

I.

Interview 53
Statement Plant
Manager308

309

4 hrs.

4 hrs.

IA-54

2.

ITEM

DATE AND TIME
3/28/79313

EVENT DESCRIPTION
NOTE:

314

4 hrs.

315

4 hrs.

1 min..

REFERENCE

Nature of subsequent discussions were reported to consider current
situation and options available for return, to normal conditions;discussions were not retrospective to determine what actions orconditions resulted in the plant reaching its current status.

NRC Regional Director calls for activation of Regional Emergency Centerand NRC:OIE Director called by Regional Director.-

1.

Region I notes

The hourly plant log typer displayed the following information as being
applicable for this time:

1.

Log typer

1.

Utility typer

1.

Utility typer

RCS Flow: 3 MLBH
Loop "A"Th:
Tc :? ? 'F
0°F
Tca:
a

? 6oF

Loop "B"
Th:
Tcb:
Tcb:

? 0OF
oF
? 0F

MU Flow: 8 GPM
OTSG "A" Level:
228 inches
OTSG "B" Level:
241 inches
Condensate Flow:
711 KLBH
316
317

4 hrs. 4 min.

Operator requests computer print RCS loop "A" hot leg temp.

(offscale).

Operator requests computer print control rod 13-E percent withdrawal.

IA-55

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

318

4 hrs.

Operator requests computer print control rod drive 3-M motor temperature.

1. Utility typer

319

4 hrs. 8 min.

Electricians and technicians had jumpered interlocks to allow RCP restart. Operating staff had attempted to start RCPs in each loop.
Successfully started RCP-IA, but the pump ran with a current less
Operating staff believed
than 100 amps and without flow indication.
RCP was cavitating and tripped RCP-lA 37 seconds later.

1.

Alarm printer is now running nearly 2 hours behind real time.
Information presented here not typed and available to operator
As a result of thisfor his analysis until that later. time.
delay time, the alarm printer is not used as an operational aid
by the operators.

1.

Alarm printerI.

IRACT tapes
Region I notes

5 min.

NOTE:

320

321

4 hrs.

10 min.

Region I Incident Response Center activated.
lished between Region I, IE:HQ and licensee.

Communications estab-

2.
3.

2.

Alarm printerStatement Plant
ManagerInterviews 53,
181

322

4 hrs.

10 min.

Appears reactor building air cooling coil B placed in service by
operator in attempt to halt building pressure increase, based on(This is similar totemperature transient indication on cooling water.
that reported at - 71 minutes.)

1. Alarm printer323

4 hrs.

17 min.

Operator secures makeup pumps ]A & ]C,leaving no makeup pumps running.

1. Alarm printer324

4 hrs.

18 min.

Operator twice attempts unsuccessful starts of makeup pump 1A to reestablish normal RCP seal flow. Operator then shifts the control switch
to the pull-to-lock position and this pump remains off for the remainder of the incident.

IA-56

I,
2.
3.

Alarm printerInterview 61
Utility typer

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

325

NOTE:

From 4 hrs. 17 min. 22 sec. until 4 hrs. 21 min. 53 sec., no
makeup pumps were operating, a period of approximately 4 1/2
minutes.

326

NOTE:

Based on interviews, it appears that the operators were monitoring the rise in building pressure and they placed the makeup
pump selector switches in the "pull-to-lock" position. Thisaction apparently was taken to prevent HPI during a time when
the pressurizer appeared to be "solid".
(RCS pressure about
1500 psig).

327

4 hrs.

19 min.

328

I.
2.
3.

Interview GPU-I
Reactimeter
Plant strip
chartsThird SFAS channel in the "A" train actuates on high reactor building
pressure. With one channel in defeat, the two out of three logic ofthe "A" train is satisfied.
Full ES and building isolation is initiatedon the "A" train.
Decay heat pump ]A starts and will run for 60
minutes.
Intermediate cooling pump IA trips. The fact makeup pumplA does not attempt to start confirms that its control switch is in thepull-to-lock position.

1.

Alarm printerEighteen seconds later, operator defeats ES and building isolation.
Twenty-three seconds later operator restarts intermediate cooling pump
IB to maintain RCP seal and letdown flow cooling.

I.

Alarm printerI.
2.

Operating logs
Region I
Incident
Messageform

1.

Alarm printer329

4 hrs.

20 min.

Containment dome radiation monitor reads 600 R/hr.

330

4 hrs.

22 min.

With the unsuccessful attempt to start makeup pump IA,

IA-57

the operator

ITEM

331

DATE AND TIME
3/28/79

4 hrs. 22 min.

EVENT DESCRIPTIONREFERENCE

starts pump 1B. This pump remains in operation for the remainder ofthe period covered by this sequence.

2.

Interview 61

Technicians lift leads on RCS loop "A" hot leg RTD to take reading
Initially, temperature readings were not believed due to their(7200).
magnitude and the fact that the readings were outside the calibrated
range of the instruments. All involved site personnel felt the core
was covered at this time.

1.

Interviews 53,
121, 181
Statement Plant
Manager

2.

332

4 hrs. 24 min.

Pressurizer heater groups 1-5 are returned to normal.
heaters are operable at this time.

All pressurizer

1. Alarm printer333

4 hrs. 26 min.

Letdown high temperature alarms received. (Appears operator may haveincreased letdown in excess of one cooling pump's capability).

1. Alarm printer

NOTE:

334

335

336

4 hrs. 27 min.

4-5 hrs.

Operator interviews indicate they experienced letdown flow
oscillations of 40 GPM, peak to peak, indicating relief valve
operation.

1. Interviews
9,

I.

21

Alarm printerInterview 61

This pump remains in operation until
Operator starts makeup pump IC.
9 hrs. 4 min. after the start of the incident, feeding 250 gpm thru
valves 16A & C, only, using IB & IC makeup pumps.

2.

Station Manager had requested verification of incore T/C readings,
Technicians take incore T/C readings
which were indicating offscale.
Readings indicate temperaturesat terminal stripg in control building.
from 80 F to 2620 F. Indications were not considered reliable by thesenior staff present, reportedly because the thermocouples were notsafety grade components.

1. Technician
Records
2.
Interviews
121, 181
3.
Statement Plant
Manager

IA-58

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

337

4 hrs.

31 min.

Group 10 of the pressurizer heaters trip. This group does not become
operational again during the period covered by this sequence.

1.

Alarm printer338

4 hrs.

31 min.

Condenser vacuum pumps ]A & IC secured.
Plant staff was experiencing
problems with Unit 1 auxiliary boiler which supplied turbine seals.
Main condenser vacuum decreases, reaching atmospheric pressure within
15 minutes.

I.
2.

Alarm printerPlant strip charts

339

4 hrs.

35 min.

Operator requests computgr print 16 specific incore T/C reading.
put data ranges from 310 F to offscale.

1.

Utility typer

340

4 hrs.

36 min.

Letdown high temperature alarms clear.

1.

Alarm printer341

4 hrs 39 min.

NOTE:

1.

IRACT tapes

342

Out-

At this time, a telephone line between the IE:HQ Operations
Center (OC) and the Region I Incident Response Center (IRC)
was opened and remained open throughout the incident. The
IRC had already opened one or more telephone lines to theThree Mile Island site to receive information, and the IRC
relayed that information and questions between the OC andthe site. Later in the day, between 8 1/2 to 9 hours after
the start of the accident, these lines were tied together
putting the OC, the IRC and the site on a common conferenceline.
Because the IRACT tapes were transcribed in half-hour intervals only approximate times of information exchanges canbe given within this half hour interval.
The initial information and question exchanges were for the purpose ofobtaining operational and radiological status information

IA-59

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
which are redundant to the material in this sequence,
be repeated unnecessarily.

and will not343

4 hrs. 40.min.

Containment dome radiation monitor reads 1000 R/hr.

I.

Region I
Incident
Messageform344

4 hrs. 42 min.

EFW pump 2A turned off. Steam generator levels appear stable in
Level changes over next 90 minutes in both "A" anddesired range.
"B" OTSGs appear to relate to heating of OTSGs thru steam condensing natural circulation. Neither "A" or "B" OTSGs would be steaming at this point.

I.
2.
3.

Alarm printerPlant strip charts
Reactimeter

345

4 hrs. 43 min.

Vacuum dropping in condenser as evidenced by feedwater turbine low
vacuum trip.

1.
2.

Alarm printerPlant strip charts

346

4 hrs. 46

Groups 4 and 5 of the pressurizer heaters trip and do not become
operational for the remainder of the period covered by this sequence.
No more than 10 heater groups are now operational.

1.

Alarm printer347

4 hrs. 47 min.

Operator requegts computer print 11 specific incore T/C readings whichrange from 378 F to off scale; the latter indicated by "????".

1.

Utility typer

This entry made to indiIRC relays 1000 R/hr dose rate to OC.
cate that at this time, certain aspects of key information being
In general, however, responses to quesrelayed were current.
tions asked about plant conditions were not responded to promptly because of the activities in the control room and the needto relay information, both between OC and IRC and the site,

1.

IRACT tapes

348

NOTE:

IA-60

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
as well as the relaying required within the control room.

349

4 hrs.

350

5 hrs

351

352

59 min.

5 hrs.

5 hrs.

Information relayed to NRC:Region I office is that HPI flow is
gpm/leg and RCS pressure 1400 psig.

250

1.

Region I
Incident
Message form

1.
2.
3.

Strip charts
Reactimeter
Multipoint
recorderthis point are:General plant parameters at
1.

RCS pressure 1375 psig and generally stable;

2.

Reactor building pressure approx. 4 psig and stable;

3.

Pressurizer full;4.

Steam generator levels 40-60% and essentially steady;5.

Steam generator pressures - "B" decreasing slowly from 325 psig
and "A" at 50 psig and increasing.

6.

Hot leg temperatures > 620 0 F; 6000F and 660 0 F on multipoint
recorder.

Containment dome radiation monitor reaches 6000 R/hr.

1. Region I
Incident
Messageform

Licensee reported to Region I at 0900 (elapsed time = 5 hours) that theReported that "B" steam
auxiliary building was isolated at 0800.
Unit 2 control
generator leak was noted at 0800 by vacuum pump monitor.
room placed on recirculation air.

1.

IA-61

2.

Region I
Incident
Messageform
RadiologicalITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

sequence of
events
352A

353

B&W attempts to fax list of information,
without success.
5 hrs.

1 min.

required to make recommendations,

The hourly plant log typer displayed the following information as being
applicable for this time:
RCS Flow:

Loop T"A"
h
Th :
T

1.

B&W Telephone
Logs

1.

Log typer

1.

2.

Radiologicalsequence of
eventsIRACT tapes

1.

Utility typer

3 MLBH
?O0F
?O
?F

Tca:
Loop "0

?oF

Tcb.

?0F

MUS FlW 58 GPM
OTSG "A" Level: 216 inches
OTSG "B" Level: 241 inches
Condensate Flow: 586 KLBH
354

355

5 hrs.

5 hrs.

10 min.

12 min.

Emergency Control Station (ECS)

moved to Unit 2 control room.

Operator places incore T/C 3-L computer point in digital cal.

IA-62

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
display.

356

357

5 hrs. 15 min.

5 hrs.

15 min.

Based on first information recorded on Region I Incident Response
Center (IRC) Tapes, the status of the plant reported to Region
I is as follows:

1.

MSL Safety Valves said to be shut.

2.

OTSG "A" Level reported at 44%.

3.

OTSG "B" Level reported at 66%.

4.

OTSG are reported to be isolated.,

5.

RB Spray had not actuated yet.

Operator requests computer print RCS Pressures (1203,
and pressurizer surge line temperature (3030 F).

1164 & 1126 psig)

1.
2.

Region I
IRC tapes
1-1
Plant strip charts

1.

Utility typer

358

5 hrs.

15 min.

First instance of a request from OC thru IRC to site for specificinformation on steam generator levels and meteorological data.

1.

IRACT tapes

359

5 hrs.

15 min.

Convinced that steam was in each loop, the licensee's staff decided toraise RCS pressure and collapse unwanted steam bubbles,. They verified
again that the EMOV was shut.
(Recall that position "indication" shows
valve position demand rather than actual position.)
An increase in highpressure injection flow was directed; however, they were apparently
concerned that if pressure went too high, it could cause code safeties to

1.
2.

Interview 53
Statement Plant
ManagerTech. Supt.'s Log

IA-63

3.

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
They decided to control pressure at 2000-2100 psig using
be opened.
the EMOV block valve, with the EMOV open.

360

5 hrs.

17 min.

Appears reactor building air cooler B started again.

1. Alarm printer361

5 hrs.

18 min.

Building pressure starting to decay and RCS pressure rising. Appears
EMOV block valve has just been shut to allow plant pressurization.

I.
2.

362

5 hrs.

19 min.

Decay heat pump 1A turned off and pulled-to-lock.

1. Alarm printer2.
Analysis of next
ES initiation363

5 hrs.

24 min.

ES and reactor building isolation initiated in train "A", apparently
the result of a contact race in a reactor building 4 psig pressure
switch or a misaligned or dirty contact in a 63XI relay, when thepressure switch is changing state as reactor building pressure decreases.
Intermediate cooling pump ]A trips. The fact that decay
heat pump ]A or makeup pump ]A do not start and makeup pump IB doesn't
trip confirm decay heat pump IA and makeup pump IA are still inpull-to-lock. Makeup pumps IB & IC are now injecting with the MU-V-16
The operator resets the channel, clearing the ES
valves wide open.
and reactor building isolation thirteen seconds later. The operator
then restarts intermediate cooling pump 1A ten seconds later to ensure
RCP seal and letdown flow cooling.

1.

Alarm printer

5 hrs. 29 mins.

At this point an engineer noted the indication of an overspeed trip
condition on the Diesels (tripped at 30 min.) and questioned Diesel
status.
It was agreed to place the Diesel control switch in theThis rendered
MAINTENANCE EXERCISE position and reset the fuel racks.

I.
2.

Interview 195
Alarm printer

364

IA-64

Plant strip chart
ReactimeterITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
the Diesels capable of manual starts; however, they remained incapableof any automatically initiated starts for the remainder of the accident.

365

5 hrs.

30 mins.

RCS pressure \1500 psi

1.

NRC inspector notes

366

5 hrs.

30 mins.

During the OC-IRC information exchanges, some period of time (- 15 min.)is spent clarifying information.
This appears to some extent to be dueto the quality of communications and to clarification of terminology andplant layout and configuration.

1.

IRACT tapes

Group 3 of the pressurizer heaters trips and remains inoperable forthe remainder of the period covered by this sequence.
No more than
9 heater groups are now operational.

I.

Alarm printer1.

Plant strip charts
Reactimeter

367

5 hrs.

31 min.

368

5-6 hrs.

NOTE:

369
370

NOTE:

5 hrs. 35 min.

Reactor building pressure peaked at 4.4 psig and trended downward to 1.6 psig during this period.
SRM count rate showed ageneral decrease with no anomalies.

2.

Appears vapor lock in "A" RCS Loop is significantly larger than in"B" RCS Loop, based on analysis of OTSG level and pressure trends.

1. Plant strip charts
2. Analysis

Licensee reports to Region I that:
1.

Belief expressed that OTSG pressure transient probably lifted
OTSG Safety Reliefs.

2.

No one had been evacuated from site yet.

3.

No consideration of offsite evacuation yet.

IA-65

1.

Region I IRC
Tapes 1-1 & 2

ITEM

DATE AND TIME
3/28/79371

REFERENCEEVENT DESCRIPTION
4.

Licensee reported to have stated to have no significant leakage.
(Apparently referring to radiological released)5.

No significant offsite activity reported yet.

The EMOV and 8ne code safety outlet temperature high alarm clear at
Remaining code safety outlet tempera192°F and 193 F, respectively
ture high alarm clears at 193 F, 4 1/2 minutes later.

1. Alarm printer372

5 hrs. 39 min.

Temperature transient again indicated on reactor building air coolingcoil B, suggesting it has again been placed in service.

1. Alarm printer373

5 hrs. 40 min.

RCS pressures now read 1791 psig,

1781 psig and 1785 psig; increasing.

1. Alarm printer374

5 hrs. 42 min.

RCS Pressure at 1950 psi, pressurizer level at 400 inches.

1.
2.

375

376

5 hrs. 44 min.

5 hr. 45 min.

Plant strip
chartsReactimeter

RCS pressure rises to 2050 psi and exhibits oscillating behavior between
EMOV and safety valve tailpipe temperatures
2050 psig and 2200 Bsig.
Operators are cycling EMOV block valve with thealarm high (205-215 F).
EMOV open to control pressure within limits dictated by plant staff.

1.
2.
3.

Plant strip
chartsAlarm printerTech. Supt.'s Log

The IRACT tapes at this point show that a second line was openedThis
to the IRC in addition to the primary line from the OC.
resulted in a different set of questions being asked to be relayed to the site (radiological matters) in addition to thoseThisbeing transmitted from the OC (operational matters).
produced some conflict in establishing priorities for the two

1.

IRACT tapes

Note:

IA-66

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

requests,

but appears to have been resolved fairly promptly.

377

5 hrs.

47 min.

Operator secures Intermediate Cooling Water Pump lB.

1.

Alarm printer378

5 hrs.

51 min.

Temperature transient on reactor building cooling coil B experienced
again.

1.

Alarm printer379

6-7 hrs.

NOTE:

1.

Plant strip
charts380

6 hrs.

The hourly plant log typer displayed the following information as being
applicable for this time:

1 min.

RCS pressure cycled between approximately 2050 and 2200 psig
throughout the period.
Reactor building pressure remained in
the range of 1.8 to 2.4 psig.
Pressurizer level at 400 inches.

RCS Flow:
3 MLBH
Loop "A"
Th: ? F
Tc :?
OF
Tca:
? oF
a
Loop "B"Th:
Tb:
MU Flow:
OTSG "A" Level:
OTSG "B" Level:
Condensate Flow:

? F
? °F
52 GPM
218 inches
236 inches
0 KLBHIA-67

1. Log typer

EVENT DESCRIPTIONREFERENCE

10+ min.

Airborne levels in Unit 2 control room require evacuation of all but
essential personnel.

1.

Region I IRC
Tapes 1-2

12 min.

ESC moved from Unit 2 control room to Unit 1 control room.

1.

Radiologicalsequence of events

Believed now

1.
2.

Alarm printerPlant strip
chartsReactor trip procedure requires raising OTSG levels to thisrange, if reactor restart is not anticipated within 4 hours.

1.

Emergency
Procedure 2202-1.1

1.

Alarm printerITEM

DATE AND TIME
3/28/79381

6 hrs.

382

6 hrs.

383

6 hrs.

14 min.

Operator begins raising "A" OTSG level to 97% to 99%.
using condensate pumps, only.

384

NOTE:

385

Pressurizer heater groups I and 2 trip, but are reenergized in
27 seconds.
Licensee reports to Region I:386

6 hrs.

15 min.

1.

140 uc/cc gross p-y activity in RCS.

1.

Region I IRC
tapes 2-1

2.

Concerned "B" OTSG primary to secondary leak may lead to burp
Water level high..
of activity from safety reliefs.

2.

Region I IRC
tapes 2-1

1.

Region I
Incident
Messagefori,
Region I IRC
tapes 2-1

"Still injecting" and BWST level dropping,licensee to Region I.

based on report from

2.

IA-68

ITEM

387

388

DATE AND TIME
3/28/79
6 hrs. 15 min.

6 hrs.

17 min.

EVENT DESCRIPTIONREFERENCE

NOTE:

1.

IRACT Tapes

Licensee reports to Region I:

1.

1.

1.

Region I
Incident
Messageform
Region IRC
Tapes 2-1

2.

OC notified by IRC that evacuation of Unit 2 control
room has been accomplished.
Key events continue to be reported promptly but responses to information requests stillrequire longer periods.
At about this time, OC was told thatthe operators had "secured" HPI within 5 minutes after it was
initiated following the turbine trip. Shortly thereafter,
OC is informed by IRC that site communications will be interrupted while they change telephones to Unit 1 Control Roomdue to evacuation of Unit 2 Control Room.

Unit 2 people having to don masks.
Using "A" OTSG for cooling.
Staff believes natural circulation
was working.

389

6 hrs. 18 min.

Operator requests computer printout Sequence of Events Review.

1.

390

6 hrs. 23 min.

Temperature transient again experienced on reactor building coolingcoil B.

1. Alarm printer391

6 hrs. 27 min.

Everyone moved to Unit 1 CR except essential personnel.

1.

Utility typer

Region I IRC
Tapes 2-1 & 2

392

NOTE:

Information at this point will not be typed-by alarm printerand available for operator anlysis for nearly 2 hours 40 minutes.

IA-69

1. Alarm printerITEM

DATE AND TIME
3/28/79393

6 hrs. 30 min.

393A

6 hrs.

30 min.

REFERENCEEVENT DESCRIPTION
NOTE:

1.
OC informed of evacuation of non-essential personnel from site.
OC also informed that first NRC team from Region I has arrived onsite. Within next few minutes, questions from OC are being brought
into Unit 2 control room from the Unit 1 control room by NRC
However, this process is delayed because of the airinspectors.
borne radioactivity in the Unit 2 area. Shortly thereafter, OC
decides to contact telephone operator to attempt to establish 3way conference call between OC, IRC, and the site.

IRACT tapes

B&W offsite personnel receive update on plant status as relayed byMetropolitan Edison employee at simulator who has been able to getthru to site.

I.

B&W Telephone
Logs

394

6 hrs. 39 min.

Operator requests computer print RCS pressures (2010-2063 psig).

1.

Utility typer

395

6 hrs. 40 min.

Level in "A" OTSG 82.5% and increasing. Operator still filling
OSTG to 97% to 99% level.

1.

Alarm printer396

6 hrs.

50 min.

BWST level 37'.

1.

Transient log

397

6 hrs.

53 min.

Low vacuum trip, feedwater turbine "A" alarm received
were steaming thru atmospheric relief from "A" OTSG.

I.

2.

Line printerStatement Plant
Manager

1.

Interview 59

reactor building sump > 6'.

398

NOTE:

Why this alarm was not received earlier is unexplained.
Vacuum has been below setpoint for 2 hours.

399

NOTE:

Not many individuals actually left control room as result ofevacuation order.

IA-70

ITEM

DATE AND TIME
3/28/79

400

401

6 hrs. 55 min.

402

EVENT DESCRIPTIONREFERENCE

NOTE:

Door to Unit 2 control room without automatic recloser.
Peoplefailed to close door on their own, possibly compromising recirculating ventilation system.

1.

Interview 59

Operator requests computer'print OTSG parameters.
"B" OTSG downcomer
temperature -. 430 F. "A" OTSG downcomer temperature -.265 F. "A" OTSG
shell temperatures increase with elevation, 258 0 F at bottom to 441 F
at top.

1.

Utility typer

NRC inspector now on site reports trying to cooldown using "A" OTSG
and atmospheric dumps.
Possible bubble in both RC pump legs; 37 feetleft in BWST. Pressurizer level > 400".
Feeding "A" OSTG with emergency feed.

I.

Region I
Incident
Messageform

2.

Region I IRC
tapes 2-2

1.

Region I IRC

Level in reactor building indicating > 6',403

which is maximum possiblelevel indication.
404
405

406

7 hrs.

Tapes 2-2

Noise level in Unit 2 control room reported to be high with 20+ people.

1.

Interview 59

Station Manager in charge in Unit 2 control room with majordecisions made in Shift Supervisor's office in conference with seniorstaff members.. Once decision was made, operators were then directed
to perform required manipulations.

1.

Interviews 31,
59

Tour of auxiliary building found 10 R/hr at the Rad Waste Panel, water
standing on the floor in areas with floor drains, and the auxiliarybuilding sumps full.

1.

Interviews 13,
59

IA-71

ITEM

DATE AND TIME
3/28/79

407

408

409

7 hrs.

412

REFERENCE

Operators stated that during this period they were reluctant to start
RCP for fear of vibration induced seal failure and resultant LOCA.
Personnel
They recognized they had a steam bubble in both loops.
believed core was covered and considered potential of uncovering coreas each option was reviewed. Operators were experiencing trouble maintaining adequate letdown flow. Concern for failure of EMOV block
valve in open position may have lead to decision to use pressurizer
vent valves for pressure reductions versus EMOV block valve.

1.

Communications in Unit 2 control
Unit 2 control room in respirators.
Communication problems in Unit 2
room are hampered by respirators.
control room lead some personnel to remove respirators for short periods.

1. Interviews 53,
31
2. Region I Tape 3-1

Inspector at site reports licensee still feeding "A" OSTG and using
atmospheric steam dumps.

1.

NOTE:

410

411

EVENT DESCRIPTION

7 hrs.

7 hrs. 1 min.

Had
OTSG samples had confirmed "A" OTSG was not contaminated.
stationed man on roof to monitor for radiation in steam plume
Readings also confirmed no activity in "A"should it develop.
"B" OTSG sample confirms have or had primary to secondary
OTSG.
leak. "B" OTSG remains isolated thru remainder of accident.

Interviews 31,

59

Region I
Incident
Messageform

1. Statement Plant
Manager
2. B&W Telephone
Logs

Operators controlling pressure using EMOV Block Valve (RC-V2) at - 2000
Opening valve 3 min. for every 5-8 min. About 7½ hrs, shifted topsi.
pressurizer vent valve for fear of RC-V2 failure.

1.

NRC Inspector
notes

The hourly plant log typer displayed the following information as being
applicable for this time:

1.

Log typer

IA-72

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCERCS Flow:
3 MLBH
Loop "A"Th = 0.F
T=?OF
Tca =? oF
ca
Loop "B"
Th
Tcb
cb

? OF
? OF
? OF

MU Flows = 49 GPM
OTSG "A" Level = 351 inches
OTSG "B" Level = 233 inches
Condensate Flow = 0 KLBH
413

7 hrs.

4 min.

Unit 2 ventilation system isolated. NRC expresses concern that above
may lead to ground level release.

I.
2.

NOTE:

414

State reportedly had urged shutdown of auxiliary building ventilation to limit further releases.

415

7 hrs. 9 min.

EFW Pump 2A on.

416

7 hrs.

Operator requests computer print 14 separate incore T/C parameters;
all indicate out of range.

11 min.

Operator appears to raise "A" OTSG level to 93%.

IA-73

Region I
Incident
Messageform
Region I IRC
Tapes 3-1

1.

Interview 59

I.
2.

Alarm printerPlant strip
charts1.

Utility printerREFERENCEITEMDATE AND TIME
3/28/79

EVENT DESCRIPTION

417

7 hrs.

14 min.

1.
Operator requests computer print the 3 pressurizer level differential
pressures, And the one compensated pressurizer level all of which indicate
full (> 400 ), and pressurizer temperature, which indicates about 342 0 F.

418

7 hrs.

15+ min.

100 ml RCS sample reads - 200 mr/hr.

419

420

7 hrs. 15 min.

7 hrs.

16 min.

NOTE: OC informed at this point that licensee planned to depressurize to utilize decay heat system.
Then the person making thisstatement (unidentified) indicates he wants to confirm it, and goes
off to do so. A specific confirmation or denial is not relayed backFor the next 2 hours, OC personnel appear to be underto the OC.
impression that DHR will be used until they learn, at approximately
9 hrs. 15 min, that licensee is floating CF tanks on the core.
Operator requests computer print OTSG pressures (1 psig (A) &302 psig (B))

421

7 hrs.

17 min.

Translates to - 72 uc/ml.

Utility printer1.

Region I IRC
Tapes 3-1

1.

IRACT Tapes

1.

Utility typer

and feedwater temperatures (287 0 F(A) & 415'F(B)).

EFW pump 2A off.

1. Alarm printerPlant strip charts
Interview I

422

7 hrs. 20 min.

The EMOV
Operating staff was concerned EMOV block valve might fail.
block valve reportedly had a history of sticking. Appears it was
decided to remove valve cycling frequency by utilizing wider pressure
control band (1900 to 2100 psig).

1.
2.

423

7 hrs.

26 min.

OTSG shell
Operator requests computer print "B" OTSG parameters.
essentially isothermal with slight rise in temperature with elevation.

1. Utility typer

424

7 hrs.

30 min.

Operators indicated that at this time they believed that pressurizer

1.

IA-74

Interviews 31,

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

heaters were unavailable and concerned that EMOV block valve might failopen.
Didn't appear to be getting anywhere so the group decided to droppressure to get core flood tanks (CFT) to float on core.
425

Operators were controlling pressure with the HPI and EMOV.
The staff
was concerned that the HPI flow might be bypassing the core; sinceT ^ 620, Tc ^'g20 and the pressurizer water space RTD was relatively
cAnstant at 350 F. They wanted to make sure the core was covered; sothey decided to depressurize in an attempt to get Core Flood initiation
and ultimately to initiate DHR below 400 psig.

53

1.
2.

Interviews 53,
107
Statement Plant
Manager

426.

NOTE:

Interviewee asserted that Steam Tables were referred to several
times by control room personnel during sequence.

1.

Interview 53

427

NOTE:

BWST level had dropped 20', and at present rate of use, sourcesof clean water would be exhausted in matter of hours, requiring
the use of dirty containment sump water.

1.

Statement Plant
Manager

428

NOTE:

Calculations, based on level changes in the BWST over the last
90 minutes, indicate current volumes would support at least 10
hours of additional makeup at current rates.

I.

Calculations

Shift Foreman unlocks and closes breakers for DHR isolation valves(DH-VIO2A & B).

1.

Interview 107

Operator open both the EMOV and its block valve to depressurize the RCS.
Reactor building pressure shows rise from low point of 0.2 ps'ig to 2.5
psig during this RCS depressurization.

I.
2.

Reactimeter
Plant strip
charts429
430

7 hrs. 30 min.
7 hrs.

30 min.

IA-75

ITEM

DATE AND TIME
3/28/79

NOTE:

431

REFERENCEEVENT DESCRIPTION
Station Management asserted that operators were directed tomaintain HPI during depressurization.
appears that it

may be necessary to

1.

Statement Plant
Manager

1.

Region I IRC
Tapes 4-1

432

7 hrs. 40 min.

"B" OTSG is bottled, but it
to relieve its pressure.

433

7 hrs.

40 min.

Region I inspector reports that licensee thinks they have release
under control.

1.

Region I
Incident
Messageform434

7 hrs.

42 min.

Reactor pressure at that time is 1735 psig. BeESF A&B bypassed.
lieved to be operator action to prevent inadvertent actuation of HPI
during depressurization.

I.
2.

Alarm printerPlant strip
charts435

7 hrs. 44 min.

Pressurizer heater groups 1 and 2 trip, but are reenergized within 2
seconds.

1. Alarm printer436

7 hrs. 45 min.

Reactor coolant drain tank leakage transfer pumps 9A & 9B trip.

1. Alarm printer

437

439

7 hrs.

45 min.

NOTE:

These pumps are started by operators to recirculate RCDT
water for temperature control and/or to transfer water to thein-service bleed tank.

NOTE:

Following initial reactor building isolation, transfer ofwater to the bleed tanks would be impossible unless operators reNo interview has
opened intervening containment isolation valves.
indicated these valves were reopened.

B&W Site Operations Manager relays site status information to B&W

IA-76

1.

System description

1.

Interviews 53,

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

2.

67
Mgr of Nuclear Services phone log

Operational data is relayed to OC.
This is the first operational data relayed to OC in more than 1/2 hour.
At about
this same time there is a discussion in the background (location unknown) concerning injection at maximum rate despite therisk of blowdown.
This view, however, is not relayed to the site.

1.

IRACT tapes

Pressurizer heater groups 1 and 2 trip and remain in a tripped statefor 2 hrs. 16 min. At this time, there are no more than 7 heater groups
operable.

1.

Alarm printer

NOTE:

1.
2.

Reactimeter
Plant strip
chartsOperator requests computer printout Sequence of Events Review.

1.

Utility typer

Using atmospheric dumps now.

1.

Region I IRC
Tapes 4-2

1.

Utility typer

at Lynchburg via home phone of another B&W site representative.

440

441

7 hrs.

7 hrs.

45 min.

50 min.

442

443

7 hrs.

444

7 hrs. 55+ min.

445

52 min.

8 hrs.

NOTE:

Throughout depressurization, pressurizer level did not change
from 400 inch indication.
Reactor pressure decrease also shows
discontinuity at 1400 psig at about this time; this indication,
may be result of operator shutting the EMOV block valve and opening the pressurizer vent valve to continue pressure decrease.
The drop in rate of increase in reactor building pressure further
suggests a change in blowdown technique.

Unit 1 reported to be in Hot Standby.

Operator requests computer print pressurizer parameters.
Surge line
temperature seen to be increasing, demonstrating flow thru thePressurizer.

IA-77

ITEM
446

DATE AND TIME
3/28/79
8 hrs.

EVENT DESCRIPTIONREFERENCE

Station Manager directed, purportedly over his objection, by theVice President-Generation to be prepared to meet with the Lt.
Governor in Harrisburg.
The meeting was apparently establishedafter a telephone call from the Lt. Governor to the Presidentof the company.
The Station Manager directed the SuperintendentTechnical Support to collect technical materials prior to themeeting.

1. Interviews
71, 91447

8+ hrs.

Letdown cooler lB outlet temperature high alarm received.
As theoperators reduced pressure, they had even greater problems maintaining
adequate letdown flow. Personnel reportedly hesitated to take corrective action because of their concerns relative to whether NRC would
tolerate the deliberate overexposure required to manipulate the valvesnecessary to regain adequate flow.

1.
2.

Interview 31
Alarm printer448

8 hrs.

NOTE:

1.

IRACT tapes

149

8 hrs. 1 min.

The hourly plant log typer displayed the following information
as being applicable for this time:

1.

Log typer

OC is notified of arrival of second Region I team at the site.

RCS Flow: 3 MLBH
Loop "A"
T
F
Tca:
ca

? OF

Loop "B"
Tb:b? OF
T
OF

cb

IA-78

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONTREFERENCE? OF

MU Flow•: 49 GPM
OTSG "A" Level: 373 inches
OTSG "B" Level: 228 inches
Condensate Flow: 0 KLBH
450

8 hrs.

12 min.

High level indicated on core flood tank "A" (13.3
pressure at that point was approximately 800 psi.

451

8 hrs.

15 min.

Letdown cooler 1A outlet temperature high alarm received.

1. Alarm printer

NOTE:

Staff recognized letdown was providing some cooling.

1.

NOTE:

OC requests that licensee be asked if they considered blowing
down the primary system, and if they considered bumping the RCP's.
It appears that this is the first instance of the OC asking
questions of a planning nature rather than a status nature. IRC
agreed to relay request to Inspector at site.

1. IRACT Tapes

452
453

8 hrs.

15 min.

feet of water).

RCS

I.
2.

Alarm printerPlant strip
chartsTransient log

454

8 hrs.

17 min.

Over next 6 minutes, both code safety outlet temperature high alarms
clear at 193 F; believed associated with operator shutting EMOV block
valve at some time earlier. Appears operators are continuing to droppressure using the pressurizer vent valves.

1.
2.

Alarm printerPlant strip
charts455

8 hrs.

20 min.

Pressurizer level is still off scale high.
exists in OTSG "candy canes".

Inspector believes bubble

1.

Region I IRC
Tapes 4-2

OC informed of hot leg temperature of 6200 F. (No comment is

1.

IRACT tapes

456

8 hrs. 20 min.

NOTE:

IA-79

ITEM

DATE AND TIME
3728/79

REFERENCEEVENT DESCRIPTION
made that this is the limit of display for this detector) OC requests information be provided on fuel assembly outlet thermocouples.
There appears to be a discussion in the background of
why the licensee throttled back on HPI flow (as understood by OC)
but no question is relayed to site.

457

8 hrs.

25 min.

Operator requests computer print pressurizer temperature (350 F) and
Available heaters appear incapable of
RCS pressure (618-640 psig).
establishing saturation conditions in pressurizer with existing flow
thru it.

1.

Utility typer

458

8 hrs.

30 min.

"A" OTSG shell temOperator requests comput~r print OTSG param~ters.
"B" OTSG shell
top.
Ft
peratures range from 215 F 8t bottom to 404 F
temperatures range from 401 F at bottom to 492 F at top.

1.

Utility typer

1.

IRACT tapes

Alarm printer459

8 hrs.

30 min.

NOTE:

OC is informed that "B&W representative says please do start
We don't know what the status is and wereactor coolant pumps."
may end up blowing the seal...." OC also informed HPI flow to
Minutes later, a direct connection between OC
core is 400 gpm.
and the site is established.

460

8 hrs.

31 min.

Operators start decay heat closed cooling water pumps in preparation
for initiation of DHR.

1.

461

8 hrs.

40 min.

Operator requests computer print 25 separate incore T/C reading. Printed
data ranges from 322 F to 6350 F, with the majority reading offscale.

1. Utility typer

462

8 hrs. 41 min.

RCS pressure reaches 600 psig, which is equivalent to the nominal gas
pressure maintained in the core flood tank (CFT) nitrogen cover gas.

I.
2.

IA-80

Reactimeter
Plant stripITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

3.
NOTE:

463

464

8 hrs. 43 min.

32'

Throughout the entire period when the core flood tanks were
floating on the reactor coolant system, operators report thatthe tank levels did not drop more than a few feet, thus eliminating any possibility that portions of the nitrogen cover gas
used in the tanks entered the reactor coolant system. An
operator reported refilling the CFT nitrogen from a pressureof 400 psig to 600 psig at approximately 17 hours after thestart of the accident.

BWST,

RCS pressure 576 psig.

1.

Interviews
114, 119I.

Region I
Incident
Messageform
Region I IRC
Tapes 5-1

2.
1.

Transient logs

"A" core flood tank high alarm clears at 13.13'.

1.

Alarm printerOperator requests comp 8 ter print RCS pressure (483 psig) and pressurizer temperatures (350 F).

1.

Utility typer

Were feeding "A" OTSG with condensate pumps and still steaming out
atmospheric reliefs. State had made several calls requesting releases
be secured, each time more militant. Inspector pointed out to Station

1.
2.

Interview 26
IRACT tapes

465

8 hrs. 45 min.

Staff decides not to open the EMOV as method of cooldown.
"where will we be when core flood is gone?"466

8 hrs.

55 min.

467

8 hrs.

58 min.

468

9 hrs.

charts
NRC Inspector
notes

IA-81

Concerned

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
StationManager that source of water was probably contaminated.
Manager may have been unaware that OTSG feed had been shifted from thecondensatestorage tanks to the main condenser.

469

9 hrs.

"A" OSIG is
Operator requests computer print OSTG shell temperatures.
"B" OSTG temperatures are holding.
cooling down, all levels.

1. Utility typer

General plant parameter conditions:

470

1.

RCS Pressure 500 psig.

1.

Reactimeter

2.

Reactor Building pressure 2.6 psig.

2.

Plant strip
charts3.

Pressurizer level 400 inches.

3.

Multipoint
recorder4.

Steam Generator "A" - 100% level indicated on operating range
instrument; indicated pressure 0 psig.

5.

Steam Generator "B" - 60% level and 250 psig (pressure slowly decreas-1.

Region I
Messageform
Region I IRC
tapes 5-1

ing).
6.
471

9 hrs.

RCS Hot Leg Temperatures per multipoint recorder are 605°F and 665°F.

Inspector reports from site that: Unit 1 control room going on masks.
Excess personnel moved to Observation Center.

2.

IA-82

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

472

9 hrs. 1 min.

The hourly plant log typer displayed the following information as
being applicable for this time:

1.

Log typer

RCS Flow: 3 MLBH
Loop "A"Th ? FT:? oF
Tca: ?OFca
Loop "B"
Th: ? OF
TCb: ? OFcb" "MU Flows: 97 GPM
OTSG "A" Level: 380 inches
OTSG "B" Level: 226 inches
Condensate Flow:
0 KLBH473

9 hrs. 4 min.

Makeup pump 1C tripped by operator.

474

9 hrs. 8 min.

Have stopped makeup from the BWST and are floating RCS in the CFTs.

1.

Transient log

475

9 hrs. 9 min.

Operator places "A" loop cold leg temperature in analog recorder
trend (#2).

1.

Alarm printer

Pump IB still operating.

Alarm printer

1.

476

9 hrs.

15 min.

VP directs Site Manager to secure atmospheric steam dump.
Repairs to
the instrument air line on the auto-reject valve had allowed operators
to reestablish normal hotwell level. Log indicates "Electromatic relief shut."

IA-83

2.
3.

Interview 61
Statement Plant
Manager
Transient logs

EVENT DESCRIPTIONREFERENCE

477

NOTE:

1.

Plant strip
charts478

Region I informed that:1.

Region I
Incident
Messageform

1.

Region I IRC

ITEMDATE AND TIME
3/28/79

Reactor building pressure trace indicates all pressurizer
blowdown paths were isolated about 8 minutes earlier,1.

Condenser isisolated.

(Thus not available as a heat sink.)2.

Reported Harrisburg TP Exit at 25 mr/hr using uncalibrated

tapes 5-2

instrument,Core flood tanks have partially injected.

4.

BWST now at 31 feet.

5.

Hoping core flood injection will drop temperature allowing use of
decay heat removal (DHR) system.

6.

General plant announcement heard, "SHUTTING

NOTE:

479

480

3.

9 hrs.

15+ min.

600 psig in RCS.

DOWN ATMOSPHERIC DUMPS".

With the termination of the use of the atmospheric steam dump, theonly cooling mechanism available to the licensee was the injectionof relatively low temperature water from the BWST into the RCS.
This condition lasted until approximately 14 hrs. after thestart of the accident when a heat sink was reestablished bydrawing a vacuum in the condenser and steaming the "A" OTSG
through the condenser steam dump valves.
Thus a heat sink was
not available for a period of approximately 4 3/4 hours.

Personnel felt core was demonstrated to be covered when CFT level dropped

IA-84

1.

Interviews

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

only slightly upon dropping RCS pressure below CFT blanket pressure. They
attributed the inability to drop below 450 psig as sign they had reachedsaturation pressure of loops.
They then would be unable to depressurize
further without some mechanism for cooling those loops. Personnel apparently did not realize at the time that both CFT flow paths have large
loop seals which preclude assurance that core is covered.
481

482

9 hrs.

17 min.

9 hrs.

17 min.

483

2.
3.
4.

31 , 53
Transient logs
NRC Inspector notes
Burns & Roe drawings

EMOV outlet temperature high alarm clears at 193 0 F. Core flood tanks indicate slow discharge.
No indications of changes seen on RCS pressure
or pressurizer level instrumentation.
Pressurizer still at 400 inches.
Reactor building pressure at 2.6 psig and decreasing.
Logs indicate"Stopped HPI" (Initiation time not known - not confirmed by line
printer).

1. Operating logs
2.
Plant strip
charts3.
Reactimeter
4. Alarm printer

NRC inspector reports atmospheric steam dump closed.

1.

NRC expresses concern that leaving core flood tank isolation valves open
may lead to injecting N2 into vessel. This concern was reiterated at
approximately 9 hrs. 30 min. In both cases, it was responded to be indicating the licensee's conclusion that N injection was not possible,
based on system design and plant conditioLs.

1. Region I IRC
Tapes 6-1
2.
IRACT tapes

Region I
Incident
Messageform484

9 hrs.

20 min.

Letdown cooler ]A outlet temperature high alarm clears.

1. ;Alarm printer485

9 hrs.

21 min.

Received EMOV outlet temperature high alarm at 220.40F; this alarm
believed to be associated with opening the EMOV block valve at this time.

1. Alarm printer

IA-85

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

486

9 hrs. 24 min.+Operator places "A" & "B" OTSG pressurizer and RCDT parameters in 2 minute trend.

1.
2.

Utility typer
Alarm printer487

9 hrs.

Reactor building pressure 1.6 psig.
slowly increasing level.

1.

Plant strip
chartsOperating logs

30 min.

SRM count rate continuing to show

2.
488

Region I inspector reports from site to Region I that State of Pennsylvania was concerned about steam dump. (At this time, atmospheric steam
dump had been secured for approximately 15 minutes.
This message appears
to be for the purpose of explaining why the dump was secured).

I•

Region I IRC
tapes 6-1

489

9 hrs.

32 min.

EMOV outlet temperature high alarm clears at 193°F; alarm is believed
to be associated with earlier closure of the EMOV block valve by anoperator.

I.

Alarm Printer490

9 hrs.

35 min.

Region I inspector, in response to question,pressurizer relief valve is closed.

1.

Region I
Incident
Messageform
Region I IRC
tapes 6-1

states to Region that

2.
491

9 hrs. 40 min.

Operator starts intermediate closed cooling pump IB,
temperature alarm on letdown cooler IB outlet.

492

9 hrs. 45 min.

Region I inspector reports to Regional Office that DHR will be initiatedwhen RCS pressure gets to about 350 psig.

IA-86

clearing high1.

Alarm printer1.

Region I
Incident
MessageformITEM

DATE AND TIME
3/28/79493

EVENT DESCRIPTIONREFERENCE

Inspector reports general status to Region I as follows:
1. Rx Press --500 psig.

1.
2.

494

495
496

9 hrs. 49+ min.
9 hrs. 50 min.

2.

Temperature • 250°F.

3.

Pressurizer being vented to vent header.

4.

Normal letdown in operation.

5.

Pumping bleed tank to core flood tanks to complete water cycle.

NOTE:

The path described here by the inspector was not confirmed byany interviews conducted.
It appears to be a misunderstanding on thepart of the inspector.

1.

Region I IRC
tapes 6-2
NRC Inspector
notes

Alarm printerUtility typer

EMOV outlet temperature-high alarm received at 226 F. RCDT and pressurizer temperature transient indicate EMOV block valve is now open.

2.

ES actuation on reactor building isolation by high building pressure (4psig).
Building pressure experienced spike to 28 psig (indicated)tripping reactor building spray pressure switches (nominal 30 psig setpoint) some 6 seconds later. Building spray pumps start and the 30 psig
actuation clears within 4 seconds.
RCS pressure indication shows rapid
40 psi dip (probably due to increased pressure on RCS pressure detector
reference leg) and recovery to 500 psig at that same time. Makeup pump
1C starts (lB already operating and 1A in pull-to-lock). Decay heat pumps
1A and 1B start, intermediate closed cooling pumps 1A and 1B trip. Code

1. Alarm printerPlant strip
2.
chartsInterview 61
3.
4. Statement Plant
Manager5.
Utility typer
6.
Interviews 111, 119
Operations log
7.

IA-87

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
safeties and RCP air temperatures alarm high.
Operator defeats building
isolation and restarts intermediate closed cooling pumps IA & IB tomaintain RCP seal water and letdown cooling.

1.

Interview TMI-226,
232

Oxygen analysis days later indicate combustion has occurred. The Station
Manager recalls hearing a "thump", but attributed it to a change inventilation damper posiion. Recognition that hydrogen combustion has
occurred, will not come'until the next day as data is analyzed.

1.

Reactor building
atmospheric sample
results

2.
3.

Interview
Statement Plant
Manager

Review of IRACT *and Region I IRC tapes shows that this event
was not brought to the attention of NRC Management.

1.
2.

IRACT tapes
Region I tapes

497

Operator reports that the alarm occurred-simultaneous with his openingof the EMOV.
Shift Supervisor believed he had a "noise" or electricalcross connection which yielded "false" signal.
Supervisor requested aid
from instrumentation people.
Some question exists as to whether thisinformation was passed to management onsite. Another Shift Supervisor,
then present in the control room and directing console activities, statedthat he saw the spike occur in coincidence with the opening of the EMOV.
He did not know what caused it, but he believed it was a real pressure
spike.
He stated that only after he was convinced that he no longer
needed the spray did he secure the RB Spray Pumps. This would appear to
explain why the pumps are not secured until 6 minutes later.

498

The Supervisor stated that he recommended to the Station Manager that
the EMOV no longer be cycled because of the rapid rising building
pressure when it was opened.

499

500

NOTE:

IA-88

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

501

9 hrs.

Makeup pump IC is tripped; probably by operator, since pressurizer level
480V motor control centers (MCCs) 2-32A & 42A
indicates full scale.
feeder breakers trip along with leakage closed cooling pumps 2A & 2B
supply breakers.

1. Alarm printer

51 min.

502
503

9 hrs.

504

9 hrs. 55 min.

505

1.

Burns & Roe
drawings

EMOV and one code safety outlet temperature high alarms rapidly clear
together.
The EMOV outlet showing a 45 F drop in 3 minutes; probably
The EMOV outlet alarm appears
the result of building spray quenching.
again within 20 seconds Slearing again in 2 minutes, showing a temperature swing of nearly 30 F, further supporting conclusion that rapid
clearing due to actuation of spray.

I.
2.

Alarm printerUtility typer

Unit 1 CR still in respirators.

1.

Region I IRC
tapes 7-1

Station Manager and Superintendent-Technical Support leave Control
Room to meet Vice President-Generation (who left from Observation
Center) at the North Gate to proceed to the Lt. Governor's Office.
Emergency Director Designee (Unit 2 Superintendent) was directed
to maintain status, if possible, during their absence. Station Manager was fitted with a beeper to permit him to be contacted in theUpon arrival at the Lt. Governor's
event of a change in conditions.
Office, the Superintendent-Technical Services called the plant and remained on the phone approximately 15 minutes after the meeting started.
It appears the meeting started at 2:30 p.m. (10.5 hours after the startof the accident) and lasted 30-45 minutes.

1.

Interviews 91, 1
71, 53, 27
Chronological log in
Lt. Governor's Office.
Control Room muster
log

NOTE:
52 min.

9 hrs. 55 min.

Leakage closed cooling pumps 2A and 2B receive power from the
480V MCCs which lost power.

IA-89

2.
3.

EVENT DESCRIPTIONREFERENCE

55 min.

Pressurizer heater group 8 trips and remains tripped for the remainderof the incident. At this time, no more than 6 heater groups are operable.
The remaining code safety outlet temperature high alarm now clears.

1. Alarm printer

9 hrs.

56 min.

I.
Reactor building spray pumps shut down and placed in pull-to-lock by
operator to secure NaOH washdown of building when pressure drop indicates 2.
no longer needed. It appears these switches were returned to normal withinminutes. Operators close NaOH supply valves and the reactor building
spray line valves.

Alarm printerInterviews 61,
III and 119

9 hrs.

57 min.

Core flood tank "A" level alarms indicate varying levels over the next
1 1/2 hours, between a low of 12.3 feet to a high, level at or above 14
feet. During this period RCS pressure slowly i-ncreased to 560 psig.
These alarms continued intermittently over a period of approximately
1 1/2 hours. No core flood tank "B" level alarms were received on thealarm printer during the incident.

1.
2.

Alarm printerPlant strip
chartsPressure and temperature
Operator secures decay heat pumps IA & IB.
were still too high to initiate DHR and pumps were just running onreci rcul ation.

1.

Alarm printet

B&W at Lynchburg relays message licensee to establish at least 400 gpm of
HPI flow. This done in two ways:

1.

Interview 88

ITEM

DATE AND TIME
3F28/79506

9 hrs.

507

508

509

510

10 hrs.

1.

B&W Manager of Project Management contacts GPU Vice President.

2.

Unit 2 Operations Supervisor (still at Lynchburg)
control room to relay message to Unit 2.

IA-90

calls Unit 1

ITEM

DATE AND TIME
3/28/79

511

10 hrs.

512

10 hrs.

513

10 hrs. 2 min.

1 min.

EVENT DESCRIPTIONREFERENCE

NOTE:

1.

OC expresses their concern that continued injection will
prevent discharge of the core flood tanks, with the result that thesystem will remain above the DHR interlock for a long period. OC
asks how licensee is going to get the pressure down so that DHR
can be put into use.

IRACT Tapes

EMOV and two code safety outlet temperature high alarm received
again.
RCDT and pressurizer temperature swings indicate EMOV block
valve reopened.

1. .Alarm printer2.
Utility Typer

The hourly plant log typer displayed the following information as
being applicable for this time:

1.

Log typer

1.

Utility typer

RCS Flow: 3 MLBH
Loop "A"oF
TTh : ?? OF
caLoop "B"Th :? OFT cb :? OF
cb
MU Flows: 113 GPM
OTSG "A" Level: 395 inches
OTSG "B" Level: 225 inches
Condensate Flow: 0 KLBH
514

10 hrs.

3 min.

Operator requests computer printout Sequence of Events Review.

IA-91

REFERENCEITEMDATE AND TIME
3/28/79515

10 hrs.

6 min.

Pressurizer heater groups 1 and 2 return to normal,
within 1 minute 54 seconds.

516

10 hrs.

15 min.

Status of plant as reported to NRC:RCS Pr~ssure - 500 psig.
T 230 F
Suspect Bubble in Loops
Reported to be venting pressurizer

1.
2.

Region I IRC
tapes 7-2
Utility Typer

One code safety outlet temperature high alarm clears.

1.

Alarm printerGeneral status as reported to NRC:
1. Licensee reported to believe release coming from auxiliary
Reading 1-2 R/hr in area of rad wastebuilding's floor drains.
control panels.

1.

Region I IRC
tapes 8-1

517
518

519

10 hrs.

24 min.

10 hrs.

10 hrs.

26 min.

28 min.

EVENT DESCRIPTION
but trip again

1. Alarm Printer2.

All TMI-2 water tanks are full.

3.

Going to dump TMI-2 neutralizing waste tank to TMI-I miscellaneous
waste tank, then dump that water to floor of TMI-I bleed tank room
(apparently this technique of shifting Water had been used year before),then pump TMI-2 auxiliary building sumps to various TMI-I and 2 tanks.

4.

Some places in auxiliary building water was 6-8" deep,of floor drains.

usually in areas

Rapid drop in RCS "A" loop outlet temperature noted and operators
Operators believe bubble collapse resulted
believe bubble has moved.
from their actions of injecting heavily thru MU-VI6C, only.

IA-92

1. Operating logs
Plant strip charts
2.
3.
Utility typer

ITEM

DATE AND) TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

4.
5.
520

NOTE:

Although hot leg temperature came on scale, Tav indication
was still erroneous due to cold leg temperaturd remaining below 520 F.
1. Reactimeter
Pressurizer level showed slight decrease to 380 inches and then2.
Utility typer
This level change coincides
began to rise again to 400 inches.
with hot leg temperature coming on scale, but is a lesser magnitude
than would be expected under normal conditions for the indicatedAppears pressurizer heaters have finally
hot leg temperature change.
established saturation conditions in the pressurizer, blowingpressurizer water to the loop

521

"A" OTSG pressure begins to increase following reported collapse of
bubble in "A" loop.

522

523

Statement Plant
Manager
NRC Inspector notes

10 hrs. 30 min.

NOTE:

1.

During conversations between the OC and IRC, the OC is informed of contaminated water being on the floor of the auxiliarybuilding. This is the first reference to this condition found in
the telephone tapes.
Pumps IB and IC now operating,524

10 hrs. 32 min.

Makeup pump 1C started by operator.
in pull-to-lock.

525

10 hrs. 33 min.

Pressurizer heater groups 1 and 2 return to normal,

IA-93

but trip again

]A

2.
3.

Interviews 31,
127
Reactimeter
Utility typer

1.

IRACT tapes

1.

Alarm printer

1.

Alarm printerITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
They are returned to normal within another minute.
within 6 1/2 min.
There are now not more than 8 operable pressurizer heater groups.

526

10 hrs. 36 min.

Makeup pump IC secured by operator.

1. Alarm printer527

10 hrs. 38 min.

Trend in RCDT and pressurizer temperatures, and failure to receive
the one cleared code safety outlet temperature alarm, suggest operators
were venting pressurizer with vent valves versus the EMOV.

1.

Utility typer

1. All the auxiliary building sumps are full.

1.

Region I IRC
Tape 8-2528

10 hrs.

41+ min.

2.

A Region I HP indicated his tour of auxiliary building identified
no visible leaks, just water on floor.

529

10 hrs. 45 min.

Operator notes that "A" loop hot leg temperature goes off scale highPressurizer level starts steady decrease to 180
and returns again.
inches from 400 inches (minimum level reached at about 11 hours 24
minutes).

I.
2.
3.

Interview
Reactimeter
Utility typer

530

10 hrs.

NOTE:

OC repeats its request to site for status of licensee's plansto reduce pressure and use DHR system.

1.

IRACT tapes

531

45 min.

10 hrs.

57 min.

Region I inspector reports current status to be:

T = 5300F
532

h = 2360Fc
RCS press 418 psig; plan to go into Decay Heat Removal via BWST; will

IA-94

1. Region I IRC
tape 9-1
2.
Utility typer
3.
Interview I1

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
use flow path from BWST to DH Removal Pumps to core.

ll+ hrs.

Licensee plan reported by inspector to Region I is to initiate lowpressure injection followed by normal hot leg cooldown using DHR.

1.

Region I IRC
tape 9-1

534

NRC Inspector notes that pressurizer level being established andventing of pressurizer has been stopped.

1.

NRC Inpsector
Notes

535

NRC:HQ requests clarification of 16-18 hour estimate to get on DHR.
Told (1) cooling down at 30 F/hr; (2) that there is an interlock on
DHR valves at 430 psig; (3) that plant currently at 450 psig; and (4)that to be conservative they had assumed they would wait to shift overto DHR until they arrived at 400 psig. Thus 18 hr. estimate.

1.

Region I IRC
tape 9-2

Telephone contact with the site is lost for approximately 5
minutes.
When contact is established, it will be with theThis will be the first time that
Unit 1 Shift Supervisor.
the OC has had directed contact with a licensee representative.

1.

IRACT tapes

1.

Log typer

533

536

- I I hrs.

Note:

537

11 hrs. I min.

The hourly plant log typer displayed the following information as
being applicable for this time:
RCS Flow: 3 MLBH
Loop "A"
Th: 524 F
Tc a:
? 0F
Ta:
? o
a
Loop "B"

IA-95

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
Th: ? 0 F
Tcb: ? F
Tb:
Tcb:
?0F
95 GPMMU Flow:
370 inches
OTSG "A" Level:
224 inches
OTSG "B" Level:
Condensate Flow: 0 KLBH
into analog recorder trend

538

11 hrs.

3 min.

Operator places RCS loop "B" cold leg temp.
(#2).

539

11 hrs.

10 min.

Operators authorized to remove respirators in control room.
Personnel in control room had to use respirators for approximately 5
hours.

1.

Alarm printer

1.
2.

Operating logs
Region I IRC
tapes 9-1
Interview 36

3.
540

11 hrs.

10 min.

256
toLoop "A" cold leg temperature starts rapid increase from
400 0 F. Loop "A" hot leg temperature starts rapid increase fromPressurizer levelN 520 0 F to - 590 0 F. RCS pressure is increasing.
Change in rate of reactor building
drops nearly 200" in 15 minutes.
pressure drop indicates all blowdown paths from the pressurizer havebeen isolated.

541

11 hrs.

12 min.

Inspector reports to region I that:
1.
2.
3.

Feed coming from condensate storage.
Steaming out of atmospheric reliefs.
Reported no radioactivity at these reliefs.

IA-96

1.
2.
3.

Reactimeter
Utility typer
Plant strip
charts1.

Region I IRC
tape 9-2

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTION
4.

REFERENCE

Using atmospheric reliefs since have no condenser vacuum.
Lostturbine seals after no longer able to get steam from Unit 1.

542

Note:

543

The licensee is not now cooling down OTSGs.
They have stopped because
contamination.
(Note:
In fact, use of the atmospheric dumps had been
terminated for almost 2 hours at the time of this report.)
RCS pressure
begins upward trend from about 415 psig.
"A" OTSG pressure begins todecay due to vapor lock compression.

544

11 hrs.

15 min.

Actually items 2 and 4 are not correct in this report.

Commenced efforts to draw vacuum in condenser.

2.
3.

Region I IRC
tape 9-2
Utility typer
Reactimeter

1.

Interview 31

1.

545

11 hrs.

15 min.

Status report to.NRC:
Suction from BWST, DH-V5A and B now open,
running, supplying water to RCS via 16 valves.

MUP

1.

Region I IRC
tape 9-2546

11 hrs.

15 min.

Pressurizer high level alarm clears at 254" with level dropping.
Pressurizer heaters have again established saturation conditions andare blowing water to the loop.

I.
2.

Alarm printerUtility typer

547

11 hrs.

19 min.

Pressurizer low level alarm received with level dropping rapidly.
Makeup pump 1C started by operator and runs for less than 10 minutes.
Operator reportedly wanted to prevent pressurizer heater trips on
low level.

1. Alarm printer548

11 hrs. 25 min.

EMOV outlet temperature finally returns to normal range.

1. Alarm printer549

11 hrs.

Operator requests computer print EMOV and code safety outlet temperatures; all indicate below alarm point.

27 min.

IA-97

ITEM
550

DATE AND TIME
3/28/79
II hrs. 28 min.

EVENT DESCRIPTIONREFERENCE

Pressurizer level 175" (Per Operator in Unit 1
Only one 16 valve open.
control room speaking to Region I Office - potential time lag).
Worried continued bumping of relief
Periodically bumping EMOV.
Not using OTSG coolingmight cause it to fail; Failure mode uncertain.
at all at this point.

1.

Region I IRC
tape 10-1

I.
2.

Alarm printerPlant strip charts
Reactimeter
Utility typer
Multipoint recorder

Level in the pressurizer had started

551

11 hrs.

28 min.

Operator stops makeup pump IC.
increasing again.

552

11 hrs.

29 min.

Pressurizer level begins steady increase to 400 inches from 180 inches,
RCS Loopreaches 400 inches at 12 hours, 30 m~nutes (59 minutes later).
30
minutes.
during
next
to
600
F
rises
"A" hot leg temperature
Loop "A" Cold Leg increases 144°F in 2 minutes as steam condensing
natural circulation appears to develop.

1.
2.
3.

Pressurizer heater groups 1 and 2 trip, but are returned to normal
within 16 min.

1. Alarm printerOperators indicate method of cooldown at this time is one high(Region I Incident
pressure injection pump and the core flood tanks.
Messageform states pressurizer level 166" and periodically cycling
EMOV.
Information coming from Unit I Control Room with potential
significant time delay.)

I.
2.

553

Operating logs
Region I Incident
Messageform

554

11 hrs. 32 min.

555

11 hrs.

33 min.

Operator starts makeup pump 1C.

1. Alarm printer556

11 hrs.

34 min.

North Gate 30 mr/hr.

1.
2.

IA-98

Region I Incident
Messageform
Region I IRCITEMDATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

tape I0-1
557

558
559

11 hrs.

36 min.

11 hrs. 43 min.

Operator starts emergency feedwater pump 2B and commences raising"B" OTSG level to 97 to 99% range.
Will reach desired level within
20 minutes.
OTSG pressure will drop from about 250 psig to about 150
psig and stabilize.

1.
2.
3.

Operator stops makeup pump 1C.

1. Alarm printer

Reported readings from Unit 2.

1.

RB Pressure
PZR Level
PSR Temp.
PZR Press.

0 psig170" (Incrsasing)*
460 F (463 F, Slowly Decreasing)*
450 psig (534 psig, Slowly Decreasing)*

Alarm printerPlant strip charts
Reactimeter2.

Region I IRC
tape,10-2(Utility typer)*Alarm printer560

11 hrs. 44 min.

Pressurizer low level alarm clears at 206" and increasing.

I.

561

11 hrs. 52 min.

Operator secures emergency feedwater pump 2B after raising "B" OTSG
to 97%.

1. Alarm printerPlant strip charts
2.

562

11 hrs. 54 min.

Pressurizer high level alarm received at 260" with level increasing.

1. Alarm printer

NRC:HQ asks Unit 1 Shift Supervisor (over telephone) to communicate
NRC concern that pressurizer level indication does not preclude bubble
in core, and that temperature readings indicating superheat may be realand imply core is uncovered. Would then need to find a way to put morewater into core and get core level back up.
Question is asked if licensee has talked to B&W, because B&W had been trying to get in touch with

1. Region I IRC
tape 11-1
2.
IRACT tapes

563

11 hrs. 59 min.

IA-99

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
the licensee and had the same concern.

564

12 hrs.

HPI alignments still heavily biased to "C" injection leg. Operators
Rate of BWST
believed natural circulation did exist at this time.
level loss had been reduced, but still concerned would eventually
run out of clean water.

I.
2.
3.

Statement GPM
IRACT tapes
Region I IRC
tapes

565

12 hrs.

Pressurizer level trending up,
RCDT temperature trending down.
"A" loop hot leg temperatures are stableas RCS pressure increases.
"B" OTSG shell and downcomer
and the pressurizer is heating slowly.
"A" OTSG
temperatures are slowly decaying with level now at 95%.
pressure is slowly decaying with level stable at 92%. Appears operator
is compressing pressurizer bubble and loop vapor lock with HPI flow.

1.
2.

Utility typer
Plant strip charts

565A

12 hrs.

B&W requests.licensee establish site to B&Wcommunications link. Recommends 500 gpm (atleast 400 gpm) HPI flow be established immediately.
Whether this information gets passed to the site at this time has notbeen determined.

1.

B&W Telephone
logs

566

12 hrs.

B&W Site Operations Manager relays information to B&W on plant status.
B&W requested to identify consideration for running RCP in the "B" loop
Various recommendations were relayed back to(DH drop line side).
Site Management at this time.

1.

2.

Interviews 53,
B&W Telephonelogs

The hourly plant log typer displayed the following information as
being applicable for this time:

1.

Log typer567

12 hrs.

2 min.

RCS Flow: 3 MLBH
Loop "A"
Th: 590 F
Tca:
? 0F
a

IA-IO067

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTION
Tc:
a

? 0F

Loop "B"
Th:

? 0

Tc

?F

TCb:

REFERENCEMU Flow:
116 GPMOTSG "A" Level:
365 inches
OTSG "B" Level:336 inches
Condensate Flow:
0 KLBH568

12 hrs.

10 min.

r.V
569

12 hrs. 10+ min.

Reported incore temperatures unavailable.
Supervisor reports to NRC
1. Region I Incident
they are all printing question marks which means either the computer
Messageform
point or the sensor is malfunction. The supervisor did not indicate
2.
Region I IRC
that the same result occurs when the temperature exceeds the range oftape 10-1
the software calibration for those points.
This is the first thremocouple
IRACT tapes
data or comments to the NRC.
First request had been made at approximgtely
8 hrs., 20 min. States core flood tanks floating on core, on T 590 F,other "pegged" and that staff is convinced there is no boiling Yn thecore.
Region I inspector reports to Region that:1.

Licensee believes no bubble in core since core flood tanks sit on
top of core and only a little water "slid in." '(Recall piping design
utilizes "loop seal.")

1. Region I IRC
tape 11-1

2.

Licensee NOT planning to use electromatic to blowdown because could
fail open with resulting rapid cooldown that could overstress RCS.

1.

IA-101

Region I IRC
tape 11-1

ITEM

DATE AND TIME
3/28/79

570
571
572

573
573A
574
575
576

Note:
12 hrs. 11+ min.
12 hrs.

15 min.

12 [irs. 22 min.
12 hrs.

25 min.

12 hrs. 30 min.
12 hrs.
12 hrs.

30 min.

30 min.

REFERENCEEVENT DESCRIPTION(Information from Unit 1 control room with attendant time delay)

Operator requests computer print 15 separate Incore T/C readings.
All but two (572 F and 597 F) read off scale.

1.

Utility typer

NRC:HQ asks if licensee has considered blowing the system down, and
NRC:HQ
supervisor responds that it had been discussed and rejected.
requests it be considered again, stressing this was a request to
Approximately 5 minutes later theconsider it, not an order to do so.
supervisor returned stating the licensee throught it was a good idea
NRC:HQ then relayed their concern about
and decided to do it.
valving out the CFT's before the blowdown to prevent nitrogen fromgetting into the vessel.

1.

IRACT tapes
Region I IRC
tapes

Pressurizer level reaches 400 inches.

1. Reactimeter

Licensee will subsequently inform B&W that HPI of 400 GPM had
been established and maintained since this time.

1. B&W Telephonelogs

License representative reports to Region I: "We're going to decrease
What we're going to do is open that electromatic."pressure.

1;

Region I IRC
tape 11-2

NRC inspector reported Unit 1 control room is back in masks.

1.

Region I IRC
tape 12-1

Rate of increase of reactor
Operator opens EMOV to depressurize.
building pressure indicates both the EMOV and its block valve are
open.

1.

Region I Incident
Messageform
Region I IRC
tape 11-2

IA-1022.

ITEM

DATE AND TIME
3/28/79577

12 1/2 hrs.

578

579

580

EVENT DESCRIPTION
Station Manager and Superintendent-Technical
Governor's Office.
Note:

12 hrs.
12 hrs.

30+ min.

30+ min.

REFERENCE
Support return from Lt.

1. Interviews
27,

53

Review of IRACT tapes and Region I IRC tapes shows that NRC
was not aware that the Station Manager and the SuperintendentTechnical Support had been away from the site for the last
2 1/2 hours.

Operator dispatched to manually reset the DHR inlet valve (DH-VIO8A&B)
pressure interlock when RCS pressure drops below 385 psig.

1.

Interview81

NRC inspector on line in Unit 2 Control Room.

1.

Region I IRC
Tape 12-1
NRC Inspector
Notes

Reports following:1.

Originally had bubble in both loops.

2.

Have collapsed bubble in "A" loop.

3.-

Had bubble in pressurizer, but just lost it.

4.

Floating flood tanks on core and got some injection.

5.

Normal letdown at 120 gpm.

6.

Makeup Pumps feeding from BWST.

7.

B&W calculations indicate 120 gpm insufficient to remove current
core decay heat.

8.

Operating electromatic relief and pressurizer vent valves

IA-1032.

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
periodically to remove additional heat.
9.

Plan is to establish vacuum in condenser,
core heat through natural circulation.

NOTE:

581

steam "A" OTSG and remove

Inspector reported to Region I that licensee believes "A" loop;
is solid since: Original Th (A&B loops) -7000 with T \225
Directed MU to "A" loop to collapse steam bubble and its T
dropped quite rapidly to ^575 0 F. At the same time PZR levey
dropped considerably and came back on scale.

582

12 hrs. 35 min.

RCDT temperature begins to rise andPressurizer level tops out.
reactor building pressure begins to rise. Tailpipe temperature alarms
received on EMOV and both pressurizer safety valves; believed associatedwith operator opening EMOV Block Valve and possibly EMOV, also.

I.
2.
3.
4.

Alarm printerUtility typer
Plant strip charts
Reactimeter

583

12 hrs. 40 min.

Rate of reactor building pressure change indicates EMOV block valve wasjust shut.

1.

Plant strip charts

583A

12 hrs. 40 min.

B&W informed by site of plant status, recognized now to represent superheat conditions by B&W.

1. B&W Telephone
logs

584

12 hrs. 45 min.

"A" loop Th reported -580 0 F.

1.

Utility Typer indicates n5720F.

2.
584A

12 hrs. 45 min.

B&W attempts
B&W informed of plant status via B&W employee offsite.
to relay recommendation that HPI must be immediately increased, letdown
stopped and RCS subcooled condition established.

IA-104

1.

Region I IRC
Tape 12-1
Utility Typer
B&W Telephone
Logs

ITEM

585

586

587

DATE AND TIME
3/28/79

12 hrs.

12 hrs.

12 hrs.

45 min.

50 min.

56+ min.

EVENT DESCRIPTIONREFERENCE

Rate of reactor building pressure increase suggests pressurizer ventvalves may have just been opened.

1.

Plant strip charts

Region I inspector reports conditions are: pressure 450-500, (Utility
Typer and Plant Strip Chart indicates RCS pressure is 620 psig.)
letdown (120 gpm), and EMOV opening intermittently.
Lowest pressure
reached during blowdown was 450 psig and it just "hung" there. "Nothing
tried so far could drive it lower."

1.

Region I Incident
Messageform
Region I IRC
tape 12-1

2.

Reported natural circulation cooldown being conducted using "A" OTSG.
Working on getting main condenser vacuum.
Until can get temperatureand pressure reduced further, can't go on decay heat removal.

2.

•NRC inspector reports the following:

1.

1.

Licensee concerned that current cooldown process is too slow andbelieve will be faster steaming OTSG.

2.

Licensee working to get rid of bubbles in loops,
pressurizer and go on natural circulation.

3.

Licensee concerned for further use of electromatic since water
dumps to floor, and with sources of clean water being exhausted,
would be forced to use dirty sump water for recirculation.

4.

Licensee concluded core was covered.
Discussed and rejected further
blowdown since would ultimately entail recirculation of sump water,

IA-105

establish bubble in1.

Region I Incident
Messageform
Region I IRC
Tape 12-1
Region I IRC
tape 12-2

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
greatly increasing magnitude of cleanup.
repressurization to collapse all bubbles.
to be contrary to recommendations of B&W,
is aware.

588

Discussed and rejected
Latter decision appearsof which the licensee

5.

Have regained some pressurizer heaters.

6.

Plan to draw condenser vacuum using TMI-2 mechanical vacuum pumps
and TMI-I steam for turbine gland seal.
Having problems becauseof high radiation and contamination levels, restricting free movement
thru plant.

Note:

IRC notifies OC of this information.
Asks if licensee can
explain hot leg temperature as superheat without core being
uncovered, and IRC relates to OC the licensee discussion andconclusion they were confident the core was covered.

1.

IRACT tapes

589

12 hrs.

59 min.

One code safety outlet temperature high alarm clears.

I.

Alarm printer590

13 hrs.

1 min.

The Hourly Plant Log Typer displayed the following information as
being applicable for this time:I.

Log typer

RCS Flow: 0 MLBH
Loop "A"TH
522 OF

T
Tca

?
OF
? OF

ca
Loop "B"TH? OF

IA-106ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

Tbc = ? OF
cb = 7 OF
128 GPMMU Flow:
OTSG "A" Level: 345 inches
OTSG "B" Level: 321 inches
Condensate Flow: 0 KLBH
591

13 hrs. 2 min.

Operators attempt unsuccessful start of condenser vacuum pump IA.
Operator then starts condenser vacuum pump 1C.

1. Alarm printerPlant strip charts
2.

592

13 hrs. 4 min.

Reactor building air cooling fan AH-E-IiB trips. RCDT temperatureand reactor building pressure start long term trend down, which willcontinue through the rest of this scenario; suggesting neither the
EMOV or pressurizer vent valves are used again.

1. Alarm printer2.
Utility typer
3.
Plant strip charts

593

13 hrs. 5 min.

Staff remains confident no bubble
Operators starting to draw vacuum.
RCS pressure 650 psig and pressurizer level 400"in "A" OTSG loop.
(Pressurizer heaters appear to have raised pressurizer
decreasing.
temperature to the point it is blowing wat'er to the "A" loop Hot Leg.)"A" loop TH 530 F and dropping;Continuing normal letdown and makeup.
"B" loop TH% 7 00F.

1.

594
595

13 hrs. 8 min.

596

13 hrs.

13 min.

3.
4.
5.

Region I Incident
Messageform
Region I IRC
tape 12-2
Interview 44
Utility typer
Plant strip charts

Second code safety outlet temperature high alarm clears.

1.

Alarm printer

Main feedwater pump turbine low vacuum trips clear as vacuum increases.

1.

Alarm printerOperators succeed in starting condenser vacuum pump ]A.

1.

Alarm printerIA-1072.

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

597

13 hrs. 20 mim.

Operators
Reactor building pressure zero and decreasing to vacuum.
Previous small
had noted reactor cooling system going solid again.
Operators now note pressurizer
change in pressurizer level was noted.
level starting to drop and RCS pressure dropping.
(Present value
636 psig).
Seal injection flow 60 gpm.
HPI flow 425 gpm and BWST
level 26½ ft. Inspector at site reports continued operation ofel.ectromatic relief (RB pressure trace does not support) steam "A"
RCS Loop "A"OTSG to condenser.
"A" OTSG pressure begins to drop.
cold leg temperature begins to drop as heat is removed from OTSG bysteaming, and from the RCS by what appears to be steam condensing
natural circulation.

1.
2.

598

- 13 hrs. 23 min.

3.
4.
5.
6.

Status reported by Region I inspector:

1.

I.

15" Vacuum in Condenser.

2.
3.

2.

Just starting to steam now.

3.

RCS •650 psig.

(635 psig)*Thb 650-7000 F
570-580°F
Tha &
225°F
cb
a

(529 0 F)*(Tca 494OF)*

4.

Operating logs
Region I Incident
MessageformStatement Plant
Manager
Utility typer
Multipoint recorder
Plant strip
chartsRegion I IRC
Tape 13-1(Utility typer)*Plant strip
chartsElectromatic relief indicates closed.

Station Manager directed to take plant solid at high pressure.

IA-108

1. Statement
Plant Manager

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

599

13 hrs. 23 min.

Makeup pump IC started by operator.

I.

Alarm printer600

13 hrs. 25 min.

EMOV outlet temperature high alarm clears.

I.

Alarm printer601

13 hrs. 26 min.

Pressurizer heater groups 1 and 2 trip. At this time there are now
no more than 6 operable pressurizer heater groups.
Pressurizer
temperature begins to drop. Pressurizer level stops dropping andstarts to rise.

I.

Alarm printerUtility typer

Staff repressurizing RCS to collapse voids and prepare to start RC pump.
Plant charts confirm pressure increase.

1.
2.
3.

602

13 hrs. 30 min.

NOTE:

603

606

Utility typer
Plant strip charts
Statement
Plant Manager

Regarding decision to repressurize RCS; interviews indicate
1. Interviews
that control room personnel had desired to remain in
31, 44, 53
depressurized state since they had regained pressurizer level,appeared to have collapsed bubble in "A" loop, and were progressing
on doing same to "B" loop.
After decision to pressurize, control room personnel stated they
recommended restart of RCP to sweep vapor from loop.

604
605

2.

13 hrs. 32 min.

13 hrs. 40 min.

RCS pressure 623 psig and increasing.
RCS pressure is increased to about2300 psig over the next 74 minutes, using two makeup pumps, with the EMOV
block valve shut.

2. Utility typer"A" OTSG outlet valve not opening - not steaming.
Operators havestarted investigating, assuming indicator must be in error.
("A"

1. Region I Incident
Messageforin

IA-1091.

Operating logs

ITEM

607
608
609

DATE AND TIME
3/28779

13 hrs. 45 min.
13 hrs.
13 hrs.

45 min.

50 min.

610

EVENT DESCRIPTIONREFERENCEOTSG pressure had dropped 30 psi from 144 psig, but now is slowly
rising.)2.
3.

Region I IRC
tape 13-2
Utility typer

Started steaming "A" OTSG.

I.

Operations logs

Reactor building pressure is -0.2 psig.

1. Region I IRC
Tape 13-2

Pressurizer level 285 inches.
RCS pressure 765 psig.
increases above this point, pressurizer level climbs,
inches in about 32 minutes.

As pressure
reaching 400

1.
2.
3.

Operating logs
Utility typer
Plant strip charts"A" OTSG started steaming using turbine bypass valves to condenser.
Indication of level change in steam generator and indication of
feedwater flow.
("A" OTSG pressure showed marked downward trend
at 13 hrs. 42 min.)

1. Region I Incident
Messageform
2.
Region I IRC
Tape 14-1
3.
Utility typer
Region I IRC
tape 14-1

611

13 hrs. 50+ min.

Not now wearing respirators in TMI-2 control room.

I.

612

13 hrs. 52 min.

"A" OTSG high level alarm clears at 81.3%; evidence of steaming"A" OTSG.

1. Alarm printer613

13 hrs.

"A" OTSG high level alarm received at 83.5%; evidence of feeding"A" OTSG.

1.

614

14 hrs.

59 min.

RCS Pressure 860 psig.

BWST level 24 feet.

IA-110Alarm printer

1. Operating logs

ITEM

DATE AND TIME
3/28/79615
616

14+ hrs.

EVENT DESCRIPTION

2.

Utility typer

Region I inspector reports BWST 23'.

1.

Region I Incident
Messageform

Region I Inspector reports his understanding that:1.

Region I IRC
tape 14-1
Statement Plant
Manager
NRC inspector notes

1.

2.

2.
Either licensee staff or senior management have concern whether core
(NRC notes show VP-Generation has directed repressur- 3.
covered or not.
ization of RCS.)
Licensee plans to sontinue cooldown by
The plan has changed now.
"A" OTSG; Tp now 5480 F, T now 446 F. They increased makeup
steaming.
to 480 gpm with letdow I at 40 gpm, letting pressure increase and plan
No longer
to take plant solid at ^2000 psig to collapse all bubbles.
concerned with remaining volume of clean water since going solid.
(This information was passed from IRC to OC.)

Note:
617

14 hrs 2 min.

REFERENCE

This information appears to lag actual decision about 1/2 hour.

The hourly plant log typer displayed the following information as being
applicable for this time:
RCS Flow:0 MLBH
Loop "A"
Th: 549°F
T
?OF
Tca:
?oF
ca

IA-111

1.

Log typer

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
Loop "B"
Th:
T -

?OF
?°F

T cb:

?0

cb"

"

MU Flow: 43 GPM
338 inches
OTSG "A" Level:
OTSG "B" Level:
321 inches
0 KLBH
Condensate Flow:618

14 hrs. 7 min.

RCS Pressure 945 psig.

BWST 24 ft.

1.
2.

619

14 hrs.

10 min.

RCS Pressure 1012 psig.

Pressurizer level 370 inches.

1. Operating logs
2.
Utility typer

620-- 14 hrs.

10 min.

Note:

621

v14 hrs.

17+ min.

Region I inspector reports:
1.

Inspector onsite reports to OC that "B" OTSG hot leg still ap 8 ears
to have bubble with temperatures in the range of 650 F to 700 F.
States that B&W people are taking the~mocoupleoreadings on theback panel and they have obtained 700 F to 750 F readings.

Continuing cooldown of "A" OTSG.-

The AT for "A" loop is increasing.

Tha
Ta

5550

ca

300°F

(*)

=Old data from last time B&W employee manually reads backpanel RTD parameters.

(5540F)**(271'F)**

IA-112

Thb
Tb

6500700(*)°F
200 F

Operating logs
Utility typer

1.

IRACT tapes

1.

Region I IRC tape
14-2
NRC Region I
Incident
Messageform(Utility typer)**

2.
3.

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

2.

Makeup at 450 gpm.
Pressure 1100 psig increasing, pressurizer full
but pressure increase indicates bubble still exists somewhere insystem.

3.

Using condensate pumps,

Note:

not emergency feed,

for OTSG level control.

Appears time lag in information flow to NRC is much shorter for data
than it is for changes in management plans.

622

14 hrs. 20 min.

RCS Pressure 1200 psig.
23 ft.

1.
2.

Operating logs
Utility typer

623

14 hrs. 26 min.

Prpisurizer heater groups 1 and 2 return to normal.
There are now no more 1.
th'n 8 operable pressurizer heaters.
Pressurizer temperature begins
2.
Increasing.

Alarm printerUtility typer

624-14 hrs.

Region I inspector reports:

1.
2.

R:I IRC tape 15-1
Plant strip chart

29 min.

Pressurizer level 400 inches and BWST level

1.

OTSG level --90%.

2.

27" Vacuum in condenser.

3.

Main steam bypass valve indicates 30% open.

625

14 hrs. 30 min.

RCS pressure 1520 psig (at approximately this time, the SRM count rate
showed a rapid rise and then decayed to a point below its starting value).

1.
2.
3.

Operating logs
Plant strip charts
Utility typer

626

14 hrs. 33 min.

RCS pressure 1700 psig (Line printer shows 1600 psi interlock cleared at

1.

Operating logs

IA-113ITEM

627

DATE AND TIME
3/28/79\14 hrs.

34 min.

EVENT DESCRIPTIONREFERENCE

32 minutes).

2. Alarm printer3. Utility typer

Region I inspector reports:

1.
*2.

1.

Still steaming "A" OTSG, but indications are that they are not getting("A" OTSG pressure essentially zero now).*much natural circulation.

2.

-220 F (273°F), one Incore T/C reads 5700,
TL a570°F (566°F), T
i icating T may be Scurate; other T/Cs can not be read.
(This apparekntly refers to the "??" output for the T/C's
indicating they are outside the range of the computer software.)

0

628

629

14 hrs. 35 min.

0

1.

0
RCS Pressure 1800 psig (increasing), T =570 (567°F) and T =220 (273 F).*
Indication some bubbles may be collapsSg since have seen Hme temperature changes (Tca showing slow upward trend).*

1. Operating logs
2. (Utility typer)*
R:I Incident
3.
Messageform
R:I IRC Tape 15-1
4.

(Will be maintained

14 hrs. 35 min.
through rest of sequence.)630

14 hrs.

39 min.

Interview 53

B&W Site Operations Manager contacts B&W at Lynchburg who indicate that
had also come to conclusion that licensee should repressurize and starta RCP.

Unit 2 Control Room to B&W open line established.
629A

R:I IRC tape 15-1
Utility typer

Operator closes MU-V16B,

reducing HPI flow to 120 gpm at 2080 psig.

IA-114

1.

B&W Telephone
Logs

1. Operations log
Utility typer
2.

REFERENCEITEMDATE AND TIME
3/28/79

EVENT DESCRIPTION

631

14 hrs. 41 min.

Operator throttles MU-Vl6C,

632

14 hrs. 43 min.

RCS pressure 2275 psig. Makeup pump IC tripped and valve 16C closed byoperator.
Makeup pump IB still running.

1.
2.
3.

Note:

1. IRACT tapes

633

634

14 hrs. 45 min.

reducing HPI flow to 105 gpm at 2220 psig. l.

OC asks inspector onsite if licensee has considered running an
RCP.
Inspector states that those preparations are underway but
they are having trouble with the oil lift
pumps.
OC asks if
they considered that they might have a gas bubble rather thana steam bubble in the "B" loop.
Concerned it could be nitrogen,xenon, or hydrogen.
Inspector agrees to pass concern along.
Thisconcern apparently developed from evidence that only the A leg
bubble quenched while the B loop remained superheated.
Shortly
thereafter, the inspector reported back that there was nothing
licensee could do about it.
There was no way to vent that leg,andthe only way to sweep it out, whether it be steam or gas,
would be to use the pumps.

Operations log
2.
Plant strip charts
Operating logs
Alarm printerPlant strip chart

Operators beginning preparation to start a reactor coolant pump.

1.

Operating logs

635

Control room personnel recommended only bumping RCP,
be done following routine loop refill.

1.

Interview 53

636

Note:

1.

Statement Plant
Manager

14 hrs. 45 min.

as would normally

Earlier loss of vital switchgear has disabled RCP lift
and backstop oil pumps.
Attempts are underway to regain these components
to allow RCP starts, should they become desirable.

IA-115ITEM

DATE AND TIME
3/28/79

Note:

637
638

639

14 hrs. 45+ min.

14 hrs. 47 min.

Oil pump starting interlocks (K3 Relays) would still be jumpered
at this point.
1.

Region I inspector reports:
psig.

(2291

2.

psig)*

1.

Press -2300

2.

Appears still have bubble in "B" loop.

3.

3.

Makeup and Letdown at 150 gpm.

4.
5.

4.

Tha 560 (560 0 F)*,Tca 3200 (287°F)*.

5.

Steaming -lO0,O00
feeding.

#/hr based on constant OTSG level and rate of6.

Level in "A" OTSG at -93%.

Increasing steaming rate from "A"
Holding RCS Pressure at 2300 psig.
bypass valves to provide more
opening
steam generator to condenser bycooldown; one makeup pump in operation; 80 gpm letdown; 32 gpm seal
injection; 20 gpm makeup; BWST 22 ft.
Note:

640

REFERENCEEVENT DESCRIPTION"A" OTSG Pressure indicates zero; will begin slow trend upward in

Region I IRC tape15-2
See later
confirmation
at 14:52
NRC Region I
Incident
Messageform(Utility typer)*Plant strip charts

1. Operating logs

1. Utility typer

3 minutes.
641

14 hrs. 48 min.

Alarm printer output for the period 14 hrs. 48 min. 23 sec.

IA-116

to 15 hrs.

1. Alarm printerITEM

DATE AND TIME
3/28/79642

14 hrs.

643

15 hrs.

644

645

52 min.

EVENT DESCRIPTIONREFERENCE

10 min. 29 sec was not available for NRC review.
Printer output available after this period demonstrates gross paper feed problems and uses
black ink only compared to the normal red for alarms and black for"return to normal" seen in previous records.
Utility typer output,. which
may have information gaps, 'does not indicatý the alarm printer became
unavailable during this time.

2.

Region I inspector reports RCS 2300 psig, bubble still in "B" loop, "A"loop Th=560, Tc= 3 2 0, "A" steam generator 50 psi.

1. NRC Region I
Incident
2 Messageform
2.
Utility typer

Region I

1. R:I IRC Tape 15-2
2.
Utility typer

inspector reports:.1.

Believe getting someTh dropping slowly, T rising slowly.
natural circulation fiow, but still very little.

2.

Licensee has considered starting RCP, but unable to do so since has
Having problems restoring power to oillost power to RCP oil pumps.
pump bus because of high radiation and contamination levels.

Inspector reports that B&W in Lynchburg has been contacted and is doing(Transcripts are not clear as to nature of calcucalculations for TMI.
lations being performed.)
Inspector reports Th =56 0 °F, T =3600F, T ,=220°F, Thb off scale high,a RCP.
2300 psig and procee ing in attmpt to stot

IA-117

1.
2.

Utility typer

Region I IRC tape
16-1
IRACT tapes

1.

NRC Region I Incident Messageform

2.

Region I IRC tape

ITEM

646

DATE AND TIME
3/28/79

15 hrs.

2 min.

EVENT DESCRIPTIONREFERENCE

The hourly:plant log typer displayed the following information as being
applicable for this time:

3.

16-1
Utility typer

1.

Log typer

RCS Flow:0 MLBH
Loop "A"
Th
556°F
Th :
F
?F
Tca:
ca
Loop "B"
Thh
T

?OF
?O0F

cb"
MU Flow:
72 GPMOTSG "A" Level:547 inches
OTSG "B" Level: 320 inches
Condensate Flow:
0 KLBH647

15 hrs.

4 min.

NRC:HQ was of opinion might have to bump RCP to clear "candy-canes" ofvoids to allow full natural circulation, since trend of parametersstrongly suggest little or no natural circulation is occurring.

1.

Region I IRC tape,
16-1

648

15 hrs.

10 min.

Inspector reports Tha 557 0 F, Tca 380 0 F, Tcb24 0 0F, 2300 psig.

1.

NRC Region'i
Incident
Messageform
Region I IRC tape
16-1

Steam Generator "B" 180 psig ,Steam Generator "A" 20 psig (10 psig)*.

IA-118

2.

ITEM

649
650

DATE AND TIME
3/28/79

15 hrs.

11 min.

15 hrs.

15 min.

EVENT DESCRIPTIONREFERENCE

Operator requests computer summarize reactor coolant pump and makeup
pump status.
1.

Heaters now on in an attempt to draw a bubble,
surizer level is off scale high.

2.

Since no power to A.C.
pumps.

RCP oil pumps,

but currently pres-trying to start D.C.

3.

(Utility typer)*1.

Utility typer

1.

Region I IRC tape16-1

RCP oilInterview indicates licensee had to send operators to auxiliary building
to assist in starting RCP oil pumps.

1. Interviews651

15 hrs.

16 min.

Re-established normal vacuum in condensor.

1.

Operations log

652

15 hrs.

20 min.

Region I inspector reports BWST 22 ft. Still steaming; "A" loop AT
decreased, RCS Pressure 2300.
Increasing temperature with heaters.
Trying to steam d8 wn "B" OTSG level and have secured feeding it.
Th 560 0 F; T
395 F; T
220 F; Makeup and Letdown -60 gpm
Reactor bui$ing pressbve -062 psig
Pressurizer temperature (505 ) increasing with heaters on

1.

NRC Region I
Incident
Messageform
Region I IRC tape
16-2
Utility typer
Operations log

2.
3.
4.

653

Note:

Reported two NRC inspectors scheduled to go to Governor's Officefor briefing at -7:30 p.m.
(Would be -,15 hrs. 30 min. after
start of accident.)

I.

Region I IRC tape
16-2

654

Note:

Reported that at 1600 hours (12 hrs. after start of accident) the1.

Region I IRC tape

IA-119ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
Gross P3was less than 5 minimum detectable at condensate pumpuc/ml).
discharge (MDA=lOx]O
Note:

655

656

15 hrs.

22 min.

Reported that have been unable to get a primary sample since -. 0900
hours (-5 hrs. after start of accident) this morning.

Started condensate booster pump 2B to fill "B" OTSG.

16-2
1.

Region I IRC tape
16-2

I.

Operating log
Alarm printer

2.
Note:

657

Reported to NRC early morning sample of OTSGs were checked with
"B" sample pegged meter, "A" OTSG sample did nothing
frisker.
Based on results, had isolated "B" OTSG.
to meter.
1TSG level.

658

15 hrs.

32 min.

Secures condensate booster pump 2B after raising "B"659

15 hrs.

33 min.

RCS
Operator starts RCP 1A, runs it for 10 seconds and then trips it.
pressure begins to drop and operator attempts unsuccessful start ofPressure drop actuates ES and starts decay heat pumps
makeup pump IC.
1A & lB. Operator bypasses ES and starts makeup pump IC within 6
Pressure turns at a minimum of -.1409 psig and climbs toseconds.
-. 2200 psig over next 17 minutes.

1. Region I IRC tape
16-2
Section 3.2.3 of2.
Details II
I.
2.

Alarm printerOperations log

1.
2.
3.

6.
7.

Operating log
Plant strip charts
NRC Region I Incident Messageform
Region I IRC tape
16-2
Statement Plant
Manager
IRACT tapes
Interview 119

1.

Region I IRC tape

4.
5.

660

15 hrs.

35 min.

"A" loop - Th now below 520 F, the lowest meter can read.

IA-120Tc did not goITEM

DATE AND TIME
3/28/79661

EVENT DESCRIPTIONREFERENCE

upmuch.
Staff believes they got some flow.
Must wait 15 minutes due
to RCP motor start limitations until can bump RCP again.

2.

16-2
Utility typer

Station Manager consulted with B&W Site Operations Manager on appropriatedelay before pump restart.

1.

Interview 53

662

15 hrs.

37 min.

Operator places RCS loop "B" cold leg temperature in Analog Recorder
Trend (#1).

I.

Alarm printer663

15 hrs.

38 min.

RCS pressure 1783 psig and climbing.

I.
2.

Alarm printerUtility typer

664

15 hrs.

38+ min.

Inspector reports that he believes current plans are to run that RCP
and conduct normal cooldown. When bumped RCP, OTSG pressure jumped
from 20 psig to 200 psig, indicating they did get heat transfer. (,Still
steaming OTSG to condenser.)1.
2.

Region I IRC tape
17-1
Utility typer

1.
2.
3.
4.

Operating logs
(Utility typer)*Alarm printerInterview 125

2.

Alarm printer1.

Region I IRC tape
17-1
Statement Plant

665

15 hrs. 39 min.

Makeup pump IC turned off.

666

15 hrs. 40 min.

Operator places RCS wide range pressure in Analog Recorder Trend (#2).

RCS pressure (Loop A) 1948 psig (1846 psig)*.

Operator places pressurizer temperature in Analog Recorder Trend (#3).
667

15 hrs.

45+ min.

Setting up to start and run an RCP in "A" loop..

2.

IA-121ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTIONManager
Station Manager waited appropriate delay time on RCP restart andthen directed operators to start RCP ]A.

668

1. Operating logs
2.
Alarm printer669

15 hrs. 49 min.

Makeup pump IC started by operator.

670

15 hrs. 50 min.

Bypassed ES
I.
Flow and amps look good.
Started RCP ]A and let it run.
prior to its actuation on train "A", but train "B" actuates and then2.
Reactor coolant pressure drops to 1123 psig.
clears due to contact race.
"B" OTSG indicated level drops below 79% and recovers to above 85 % withi n 3.
4.
"A" OTSG indicated level drops below 81% and recovers to
8 second span.
5.
above 82% within a 9 second span, 40 seconds after "B" OTSGs transient.

1.

"A" Loop Cold Leg:
"B" Loop Cold Leg:
2.

6.
7.
8.

Loop Temperatures:

Operating logs
Region I IRC tape
17-1
Alarm printerPlant strip charts
Statement Plant
Manager(Utility typer)*Interview 119
IRACT tapes

340°F
338 F

Steam Generator Pressures:"A" Generator increases from 0 to approximately 40 psig (94 psig)*.
(Tsat"B" Generator decreases from 140 psig to approximately 50 psig.
approximately 330 F).

671

15 hrs.

52+ min.

Pressurizer level still offscale high.

IA-122

1.

Region I IRC tape
17-1

ITEM

DATE AND TIME
3T28/79672

EVENT DESCRIPTIONREFERENCE

2.

Utility typer

Operator places pressurizer temperatures, RCS temperatures and computed
saturation temperature in computer display group.

1.

Alarm typer

Operator secures decay heat pumps ]A & lB.

1.

Alarm printer

1.

Alarm printer673

15 hrs.

674

15 hrs 56 min.

Operator secures makeup pump IC.
Makeup pump IB is still operating.
is assumed makeup pump ]A is still in pull-to-lock.

6/5

16 hrs.

Note:

55 min.

It

At this point, the time frame for the Operational Sequence of Events
is concluded due to the continued operation of RCP-lA.

PLANT STATUS
676

RCP 1A is running with both RCS loop cold legs temperatures at about
290 F. Pressurizer level is still full scale with RCS pressure at about
1350 psig.
"B" OTSG is isolated at about 97% level and 99 psig, while"A" OTSG is steaming to the main condenser at about 93% level and 76
psig. Makeup pump IB is operating, supplying RCP seals and normal makeup,
the latter at 95 gpm.
Makeup pumps ]A & IC are secured, as are the decay
heat removal pumps IA & lB. Pressurizer temperature is about 1500F
and operators are letting down in attempt to draw a bubble.
The coreflood tank isolation valves are open with breakers racked out. Operators
are utilizing normal cooldown procedures.

IA-123

1. Plant strip charts
2.
Reactimeter3.
Region.I IRC tapes
17-2 & 17A-1

APPENDIX I-B
OPERATIONAL ASPECTS POTENTIAL ITEMS OF NONCOMPLIANCE

The following items are under consideration as potential items of noncompliance.
A.

Technical Specification 3.4.6.2.b requires Reactor Coolant System(RCS)leakage shall be limited to one gallon per minute (GPM)"Unidentified

Leakage" during "Power Operations"; unless leakage rates above this limit
are reduced to within limits in four hours or the plant be placed in "HotStandby" in the next six hours.
Contrary to the above,_RCS "Unidentified Leakage" remained above one
GPM for the period March 22-28,

1979,

during continuous "Power Operations",

as a result of an incorrect calculational procedure (Section 1.2.3, Details

I).
B.

Technical Specification 6.8.l.a requires written procedures shall be

implemented,

covering the activities referenced as applicable procedures

recommended in Appendix "A" of Regulatory Guide 1.33, November,

1972.

Appendix "A" of Regulatory Guide 1.33 recommends procedures for combating
emergencies; such as,"Malfunction of Pressure Control System."

Unit #2 Emergency Procedure 2202-1.5,
Revision 3,(RC-V2)"Pressurizer System Failure,"requires the closure of the Electromatic Relief Isolation Valvefor a leaking or failed open Electromatic Relief Valve (RC-R2)

and the placing of Code Relief Valve (RC-RlA or RC-RlB)

Discharge Linetemperatures on the Analog Trend Recorder for a leaking Code Relief Valve.

IB-1

1.

Contrary to the above,

RC-V2 was not closed and the Code Relief Valve

Discharge temperatures were not placed on the Analog Trend Recorders,with leakage from

RC-R2

the morning-of March 28,

and/or either or both RC-RIA and RC-RIB,
1979, prior to 0400 hours.

on(Section 1.2.4,

Details I).
2.

Contrary to the above,

RC-V2 was not closed on March 28, 1979,

duringthe accident from 0400 hours until 0619 hours; allowing a significant
loss of RCS inventory (Section 2.13, Details I).
C.

Technical Specification 3.7.1.2 requires three independent steam

generator emergency feedwater pumps and associated flow paths shall be
operable during "Power Operations"; with one emergency feedwater systemallowed to be inoperable if the system is restored to operability in 72 hours
or the plant placed in "Hot Shutdown" within the next 12 hours.
Contrary to the above, the associated flow paths from all three emergency
feedwater pumps were made inoperable on January 3,
1979,

February 26 and March 26,

during the routine performance of, and in accordance with an improper

surveillance tests procedure (Section 1.3.2, Details I).
D.

Technical Specification 6.5.1.7.b requires in part that the Plant

Operations Review Committee (P.O.R.C)

shall render determination in writing

with regard to whether or not changes to those procedures required byTechnical Specification 6.8 constitute an unreviewed safety question.
Technical Specification 6.8 requires in part that procedures be establishedcovering surveillance and test activities of safety related equipment.
Contrary to the above,

the P.O.R.C.

did not specifically address intheir safety evaluations of Revision 4 of Unit 2 Surveillance ProcedureIB-2

2303-M27A/B,

dated August 30,

Procedure 2303-Ml4A/

1978, and Revision 8 of Unit 2 Surveillance

B/C, dated November 21,

1978, the valve lineup modifica-

tions which result in complete isolation of all emergency feedwater flow
paths to the steam generators during conduct of the procedures.
the P.O.R.C.

Further,

did not render a written determination elsewhere that thevalve lineup modifications did not constitute an unreviewed safety question(Section 1.3.2, Details I).
E.

Technical Specification 6.10.1.d requires that records of surveillance

activities,

inspections,and calibrations required by these Technical

Specifications shall be retained for at least five years.
Contrary to the above the complete records of the operations surveillance reviewed during this investigation were not maintained as required(Section 1.3.4, Details I).
F.

Technical Specification 6.8.1.a. requires that written procedures

shall be implemented covering the activities referenced as applicable
procedures recommended in Appendix "A" of Regulatory Guide 1.33, November
1972, which recommends procedures for performance,

administration, andschedule of surveillance activities.
Administrative procedure 1001,(Procedure Usage),

Revision 13,

March 30,

1978,

step 3.8.5

requires the implementation of surveillance procedures

including providing shift supervisor/shift foreman review of the results.
The failure to provide a review on a regular basis of the completed
surveillance procedure results by the shift supervisor/shift foreman as
required is an apparent item of non-compliance pursuant to Technical Specification 6.8.1 (Section 1.3.4, Details I).

IB-3

G.

10 CFR 50, Appendix B, Criterion X requires a program for inspectionof activities affecting quality shall be established and executed to verify
conformance with documented instructions, procedures and drawings foraccomplishing the activity.

Three Mile Island Nuclear Station - Unit 2,

Final Safety Analysis Report, Chapter 17.2.15,

Section X requires theinspection program shall include random observation of operations andfunctional testing by individuals independent of the activity being performed.
Metropolitan Edison Company Operational Quality Assurance Program,
GP 4014,

Procedure

Revision 0, requires independent observation of activities affecting

quality to verify conformance with established requirements utilizing bothinspection and auditing techniques... for compliance with written procedures
and the Technical Specifications.
Contrary to the above,

the normal operations surveillance testing

activities were not subject to, nor were they scheduled for, random and/or
routine inspections by independent methods,

as of March 28,

1979,(Section

1.3.5, Details I).
H.

Technical Specification 6.9.1.9.b requires the licensee report inwriting within 30 days the occurrence of events of the type leading tooperations in a degraded mode permitted by a limiting condition for operation.
Technical Specification 3.4.5 requires each steam generator shall be operablewith a water level between 18 and 440 inches.

Action statement 3.4.5.brequires that with one or more steam generators inoperable due to the waterlevel being outside of limits, be in at least Hot Standby within 6 hours
and in Cold Shutdown within the next 30 hours.
Contrary to the above, the licensee failed to report to NRC the occurrenceof steam generator levels below 18 inches for about 4 minutes,
mode allowed by Technical Specification,a degradedat about 2345 hours on November 3,

IB-4

1978, following a turbine trip initiated by a loss of main feedwater(Section 1.5.2.2, Details I).
I.

Technical Specification 6.8.1,

in part, requires in part that written

procedures shall be implemented covering activities referenced as applicable
procedures recommended in Appendix "A" of Regulatory Guide 1.33,
1972.

NovemberAppendix "A" of Regulatory Guide 1.33 recommends general written

procedures for maintenance,

repair,replacement and modification work.

Station Administrative Procedure 1021,

Plant Modifications,

Revision 5,

paragraph 4.l.a requires that "the Supervisor of Maintenance receives awork request which represents a change/modification to a system or component".
Contrary to the above,

no work request was issued for the modificationof the Pressurizer Electromatic Relief Valve control logic and indicating
circuit under ECM S-5934, performed on or prior to May 6,
1.6,

Details I).

J.

10 CFR 50, Appendix B, Criterion XVII,

requires in part:

1978,(Section

Quality Assurance Records,"Sufficient records shall be maintained to furnish

evidence of activities affecting quality."

The Operational Quality Assurance

Plan for Three Mile Island Nuclear Station,

Revision 7,"Quality Assurance Records,"

Section XVII,

defines "Quality Assurance Records" to include

documentation of principal maintenance activities,
required by Technical Specification 6.10,

etc.

results of tests, records

Technical Specification

6.10.1 requires in part that records and logs be retained o'f unit operation;
principal maintenance activities,

inspections,repair and replacement ofprincipal items of equipment related to nuclear safety; etc.
Station Administrative Procedure 1002,

Rules for the Protection of

Employees Working on Electrical and Mechanical Apparatus,

IB-5

Revision 14,

requires that on restoration of equipment to service, removed tags will
have all required information entered thereon and then suitably stored, and,the Shift Foreman shall approve equipment operation by signing the original
tagging application.
Station Corrective Maintenance Procedure 1407-1,
filing of the Job Package,

Revision 0, requires

following completion of work, that documents thecorrective action taken (resolution description) and the satisfactory post
maintenance test certification.
Station Administrative Procedure 1026,"Corrective Maintenance andMachinery History," Revision 0, requires the individual accomplishing thework adhere to the approved maintenance procedure.
Contrary to the above, documentation was either not generated or not
retained which demonstrates that emergency feedwater isolation valves
EF-VI2A,

12B,

32A,

32B,

33A and 33B,

were left is an operable state, following

the performance of a maintenance check on or about January 20, 1979.
Completed Station Preventative Maintenance Procedure E-2,

"Dielectric Checkof Insulation, Motors and Cables," which was attached to the original work
request,did not contain the data that would indicate a "Meggar Check" wasperformed,
request, No.

although the work was certified as complete.
756, dated January 16,

The original work

1979, did not contain certification of

the completion of the required post maintenance valve cycling test byoperations.

The original application for tagging has not been completed

indicating that tags have been removed and that the valves had been approvedfor equipment operation.

The original tags have not been found.

To satisfy

Technical Specification 3.7.1.2 requirements for operability of emergency
feedwater flowpaths,
January 30,

1979,these valves had to be considered operable on or aboutprior to the plants return to "Power Operations" (Section

1.6.2.7, Details I).

IB-6

K.

Technical Specification 6.8.1.a requires written procedures shall be

implemented covering activities referenced by applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, November 1972.

Appendix Aof Regulatory Guide 1.33 recommends written procedures covering "Loss ofFeedwater" and "Turbine Trip".
of Feedwater,
Revision 7,

Revision 3,

October 25,

Unit 2 Emergency Procedures 2202-2.2,

October 13,

Loss1978, and 2203-2.2, Turbine Trip,

1978, require that the operator verify that the

OTSG levels are being automatically controlled at 30 inches on the SU range
indication contrary to the requirement to control level at 30 inches, thelevel control valves were placed in "manual"and throttled to control RCS

temperature and pressure before the 30 inch level was established.

The A

and B OTSG levels remained below 30 inches from 0408 hours through 0420
hours and 0425 hours respectively.

Moreover, the A OTSG level was below

30 inches between 0532 hours and 0543 hours.

In both these examples,theaffected OTSGs were taken to a dry condition as a result of this improper
and/or inadequate manual control
L.

(Section 2.3,

Details I).

Technical Specification 3.7.1.2 and its Action statement require as aminimum at least two operable independent steam generator emergency feedwater pumps and associated flow paths during "Power Operations".
Contrary to the above,

steam generator emergency feedwater pump paths

to both steam generators were in an isolated condition with feedwater
isolation valves (EF-VI2A

and 12B) shut, prior to the Reactor trip at

approximately 0400 on March 28,
M.

1979,

(Section 2.3, Details I).

Technical Specification 3.7.1.2 and its Action statement require as aminimum at least two operable independent steam generator emergency feedwater
pumps and associated flow paths with average coolant temperature greaterthan 2800F and OTSG steam pressure greater than 800 psig.

IB-7

Contrary to the above, only one steam generator emergency feedwater
pump was operable as a result of operator action at 0436 hours on March 28,
1979,with average coolant temperature at about 540°F and OTSG pressures atabout 965 psig (Section 2.3, Details I).
N.

Technical Specification 6.8.1.a requires written procedures shall be

implemented covering activities referenced by applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33,

November 1972.

Appendix"A" of Regulatory Guide 1.33 recommends written procedures covering "Lossof Coolant".

Unit 2 Emergency Procedure 2202-1.3, "Loss of Reactor Coolant/

Reactor Coolant System Pressure",

Revision 11, Section B, Step 2.2.4,

requires reactor coolant pumps (RCPs)
system (RCS)

be tripped before reactor coolant

pressure decreases below 1200 psig.

Contrary to the above, RCPs were not tripped before RCS pressure
decreased below 1200 psig at about 0416 hours on March 28,

1979,(Section

2.5, Details I)
0.

Technical Specification 3.3.2,

System Instrumentation,

Engineered Safety Feature Actuation

Section 3.3.2.1 requires that the ESFAS instrumen-

tation channels shall be OPERABLE as shown in Table 3.3-3,

including the

Safety Injection and Reactor Building Cooling and Isolation initiation from
Reactor Building Pressure High with the minimum of two channels operable,in MODES 1, 2, and 3.
Contrary to the above requirement, at 0800 hours on March 28, 1979,
two of three ESFAS channel in Train B were defeated, preventing the initiationof Reactor Building Cooling and Isolation and Safety Injection from Train B;
and at 0820 hours on March 28,

1979,

two of three ESFAS channels in Train A

were defeated, preventing the initiation of Reactor Building Cooling and

IB-8

Insulation and Safety Injection from Train A.

The Train A and Train B

Reactor Building Cooling and Isolation Actuation Trains remained defeateduntil 0924 hours on March 28,

1979, when the channels were automatically

reset (Section 2.6, Details I).
P.

Technical Specification 6.8.1.a requires written procedures shall be

implemented,

covering the activities referenced as applicable procedures

recommended in Appendix "A" of Regulatory Guide 1.33.

This guide recommends

procedures for combating emergencies; such as "Loss. of Coolant".
Emergency Procedure 2202-1.3,
System Pressure",

Unit 2"Loss of Reactor Coolant/Reactor Coolant

Revision 11, Section B, Step 2.2.3,

requires that theoperator verify the high pressure injection is operating properly as evidenced
by injection flow in all four legs (MU-V16 A/B/ C/D), with flow indicatedon MU 23-FE 1, 2, 3, 4.
Contrary to above procedure requirement, the high pressure injection
flow was throttled to minimum flow conditions by 0405 hours and one of thetwo high pressure injection pumps, was stopped during the continuing reactor
coolant system low pressure conditions.
Moreover,the high pressure injection system was routinely modified(flow throttled and makeup pumps stopped) throughout the day of March 28,
1979, during continuing low pressure conditions within the reactor coolant
system following the period when the reactor coolant pumps were stopped andthe high pressure injection system was the only certain mode available forthe removal of core decay heat (Section 2.6, Details I).
Q.

Technical Specification 6.8.1.a requires that written procedures shallbe implemented,

covering the activities referenced, as applicable, procedures

recommended in Appendix "A" of Regulatory Guide 1.33, November 1972.

IB-9

Appendix "A" of Regulatory Guide 1.33 recommends procedures for normal
plant operations; such as "Operation at Hot Standby".
Operating procedure 2102-3.3,
Removal,

Step 2.1.3,

Revision 5, March 17,

1978, Decay Heat

required that the RCS Th (not leg temperature) be

maintained 30 degrees below the saturation temperature corresponding topressurizer pressure in order to prevent boiling on the hot legs.
Contrary to the above,the system was placed onto the natural circula-tion mode at 950 psig RCS pressure at 528 F.

The procedural requirements

would indicate the pressure requirements to be about 1100 psig (saturationpressure for 558°F) (Section 2.10, Details I).
R.

Technical Specification 6.8.1.a requires written procedures shall be

implemented,

covering the activities referenced as applicable procedures

recommended in Appendix "A" of Regulatory Guide 1.33, November 1972.
Appendix "A" of Regulatory Guide 1.33 recommends administrative procedures
concerning "log entries".

Unit 2 Administrative Procedure 1012, Shift

Relief and Log Entries, Revision 8, October 4,

1977, establishes the require-ments for shift relief and recording station operating activities in logs
or other controlled documents on a shift basis (section 1.1) and describes
the various shift records and logs involved and instructions required tomaintain these records (Section 1.2, Details I).
Contrary to the above requirements numerous significant, safetyrelated log entries were not entered into the shift foreman and control
room logs between 0400 hours and 1315 hours on March 28, 1979,
2.12, Details I).

IB-10(Section

S.

Technical Specification 6.8.1.a requires that written procedures shallbe implemented,

covering the activities referenced as applicable procedures

recommended in Appendix "A" of Regulatory Guide 1.33,

November 1972.

Appendix "A" of Regulatory Guide 1.33 recommends procedures for combating
emergencies; such as "Loss of Coolant".

Unit 2 Emergency Procedure 2202-

1.3, "Loss of Reactor Coolant/Reactor Coolant System Pressure",
11, Section B, Step 3.1,

Revisionrequires the verification that all ES equipment isin its ESF position by observing all equipment status lights indicate asshow in Table B-1; including the core flood tank valves (CF-VIA and VIB).
*Contrary to the above,following the initiation of ES with continuing

low reactor coolant pressure conditions,both core flood tank valves (CF-VlA

and VIB) were closed at about 0600 hours on March 28,

1979,

and remained

closed for an undetermined period of time between about 0600 hours and 1300
hours (Section 4.6, Details I).
T.

Technical Specification 6.8.1.a requires that written procedures shallbe implemented,

covering the activities referenced as applicable procedures,

recommended in Appendix "A" of Regulatory Guide 1.33, November 1972.
Appendix "A" of Regulatory Guide 1.33 recommends procedures for safetyrelated systems,

including the emergency power sources.

Procedure (2104-6.2,
14,

Emergency Diesels and Auxiliaries,

Unit 2 Operating
Revision 9, March

1979) establishes the control of the emergency diesel generators.

Procedure Step 2.2.3 requires that the Limiting Condition for Operation ofTechnical Specification 3.8.1.1 must be met, depending on the MODE ofoperation,

unless these conditions cannot be met.

Procedure Sections 4.10,"Diesel Generator Automatic Start Upon Engineered

Safety Features Actuation"; 4.6, "Diesel Generator Shutdown to Emergency
Standby"; and 4.2, "Placing the Diesel Generator in Emergency Standby

IB-11

Mode",

specifically control the return of the emergency diesel generator to

the EMERGENCY STANDBY mode unless the requirements of the Limiting Condition
for Operation of Technical Specification 3.8.1.1 cannot be met.
Contrary to the above procedural requirements,

the diesel generator

fue-l racks were tripped at about 0430 hours and remained tripped until
about 0949 hours on March 28,

1979,rendering them incapable of being

started by an automatic or manual start signal.

Furthermore,

the Unit 2

control room "Emergency Standby/Maintenance Exercise" switch was placed inthe "Maintenance Exercise" positions on or about 0949 hours on March 28,
1979,enabling only manual starts of the diesel generators.

Following thetripping of the fuel racks of the emergency diesels at about 0430 hours,
the emergency diesel generators failed to respond to seven automatic initiation ESFAS signals between 0430 hours and 1950 hours on March 28,(Section 4.17,

Details I).

IB-12

1979

DETAILS IRADIOLOGICAL ASPECTSDETAILS IITABLE OF CONTENTS
PAGE
1.

PREACCIDENT CONDITIONS
1.1
1.2

ORGANIZATION AND RADIATION PROTECTION/CHEMISTRY
STAFFING ..............................................

II-1-1TRAINING

11-1-51.2.1

Emergency Plan Training ......................

11-1-5

1.2.2

Radiation Protection Training ................

11-1-12

1.3

EMERGENCY DRILLS .......................................

11-1-17

1.4

INSTALLED PLANT EQUIPMENT ..............................

11-1-22

1.4.1

11-1-22

1.4.2

1.4.3

Area,

Process,and Atmospheric Monitors ......

1.4.1.1

Operational Status .................

11-1-221.4.1.2

Calibration....................

11-1-24

1.4.1.3

Setpoints ..........................

11-1-25

Radwaste Systems .............................

11-1-26

1.4.2.1

Liquid Radwaste System .............

11-1-26

1.4.2.2

Gaseous Radwaste System ............

11-1-28

1.4.2.3

Heating and Ventilation Systems ....

11-1-28

Communications Equipment .....................

Il-i

11-1-32

1.5

RADIATION PROTECTION EQUIPMENT AND SUPPLIES ............
1.5.1

Portable Radiation Survey Instruments ........

11-1-34

1.5.2

Radiochemistry Laboratory Equipment ..........

11-1-36

1.5.3

Personnel Thermoluminescent Dosimetry ........

11-1-37

1.5.4

Self-Reading Dosimeters ......................

11-1-38

1.5.5

Internal Dosimetry ...........................

11-1-39

1.5.6

Respiratory Protective Devices ...............

11-1-39

1.5.7

Protective Clothing and Supplies .............

.. 11-1-41

1.6

EMERGENCY EQUIPMENT ....................................

11-1-42

1.7

ENVIRONMENTAL MONITORING PROGRAM .......................

11-1-44

1.7.1

Sampling Equipment ...........................

11-1-44

1.7.1.1

Air Samplers .......................

11-1-44

1.7.1.2

Water Samplers .....................

11-1-45

1.7.1.3

Environmental Thermoluminescent
Dosimetry .........................

1.7.2
1.7.3
2.

11-1-34

Radiological

11-1-46

Environmental Monitoring

Program .....................................

11-1-47

Meteorological Monitoring Program .............

11-1-49

INITIAL EMERGENCY RESPONSE
2.1

DETECTION AND CLASSIFICATION............................

11-2-1

2.2

ORGANIZATION ACTIVATION ................................

11-2-8

2.3

NOTIFICATIONS AND COMMUNICATIONS ........................

11-2-21

ll-ii

3.

2.3.1

Notifications ................................

11-2-212.3.2

Communications ...............................

11-2-22ASSESSMENT,
3.1

3.2

CORRECTIVE,

AND PROTECTIVE ACTIONSEFFLUENT MONITORING AND CORRECTIVE ACTIONS .............

11-3-1

3.1.1

Liquid and Gaseous Pathways ..................

11-3-1

3.1.2

Determination of Release Rates and Amounts...

11-3-16

3.1.2.1

Noble Gases ........................

11-3-16

3.1.2.2

Iodine .............................

11-3-18

3.1.2.3

Liquid Releases ....................

11-3-22

IN-PLANT ASSESSMENT AND PROTECTIVE ACTIONS
3.2.1

Radiation Protection and Chemistry Prior
to Declaration of The Emergency .............

3.2.2

11-3-24

11-3-25

Initial In-Plant Assessment andProtective Actions .........................

11-3-29

3.2.3

Implementation of Emergency Plan .............

11-3-35

3.2.4

In-Plant Radiation Protection ................

11-3-46

3.2.5

Summary of Radiation Protection Findings .....

11-3-67

3.2.5.1

Facilities .........................

11-3-67

3.2.5.2

Equipment ..........................

11-3-68

3.2.5.3

Procedures .........................

.11-3-68

ii-iii3.3

3.2.5.4

Training ...........................

11-3-68

3.2.5.5

Dosimetry ..........................

11-3-69

3.2.5.6

Air Sampling .......................

11-3-69

3.2.5.7

Respiratory Protection .............

11-3-70

3.2.5.8

High Radiation Area Control ........

11-3-70

3.2.5.9

Management Control .................

11-3-70

ENVIRONMENTAL ASSESSMENT AND PROTECTIVE ACTIONS ........

11-3-71

3.3.1

Initial Offsite Dose Calculation .............

11-3-71

3.3.2

Initial Confirmatory Surveys .................

11-3-77

3.3.3

Offsite Radiation Measurements ...............

11-3-79

3.3.3.1

Direct Radiation Measurements ......

3.3.3.2

Airborne Radioactivity
Measurements ......................

3.3.3.3

11-3-90

Radiological Enviornmental Monitoring
Program Data .......................

3.3.3.4

11-3-80

11-3-91

Assessment/Response to Environmental
Measurement Results ..............

11-3-94

TABLES

II-1-1SUMMARY OF RADIATION PROTECTION TRAINING PROGRAM

11-1-12

11-1-2SUMMARY OF RADIATION/CHEMISTRY TECHNICIAN TRAINING

11-1-15

11-1-3

AUXILIARY BUILDING RADWASTE CAPACITY

11-1-27II-iv

11-1-4

THERMOLUMINESCENT DOSIMETER LOCATIONS

11-1-48

11-2-1EMERGENCY CONDITIONS AND TIMES MET

11-2-7

11-3-1SUMMARY OF SURVEY DATA

11-3-3111-3-2OFFSITE DOSE CALCULATION RESULTS (ACTUAL)-

PROCEDURE 1670.4
11-3-3OFFSITE DOSE CALCUATION RESULTS (RECONSTRUCTEDINVESTIGATORS)

11-3-411-3-72
BY- PROCEDURE 1670.4

11-3-76

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM DATAAVAILABLE TO THE LICENSEE BY MARCH 30,

1979

11-3-92

FIGURES

11-2-1

REACTOR COOLANT PRESSURE VS TIME

11-2-211-2-2"NORMAL"

11-2-911-2-3

EMERGENCY ORGANIZATION IN EFFECT 3/28/79,

EMERGENCY ORGANIZATION

0730-0900
11-2-411-2-12EMERGENCY ORGANIZATION IN EFFECT 3/28/79,-0900-1100

11-2-13lI-v

11-2-5

EMERGENCY ORGANIZATION IN EFFECT 3/28/79,
11-2-15

1100-2000
11-3-1

SUMP PUMP DISCHARGES TO THE INDUSTRIAL
WASTE TREATMENT SYSTEM

11-3-2ON-ISLAND (GE SERIES)

11-3-7SURVEY POINTS AND

NEAR-FIELD TLD LOCATIONS

11-3-73

11-3-3

DOME MONITOR,

HP-R-21411-3-75

11-3-4

TMI - GENERAL AREA MAP

11-3-82

11-3-5

ON-ISLAND (GE SERIES)

11-3-6

PREDICTED PLUME LOCATION (1700-2400,

MARCH 28)

11-3-86

11-3-7

PREDICTED PLUME LOCATION (0340-0540,

MARCH 29)

11-3-87

11-3-8

PREDICTED PLUME LOCATION (0540-0910,

MARCH 29)

11-3-88

SURVEY POINTS

11-3-83

REFERENCES
APPENDICES
II-A RADIOLOGICAL SEQUENCE OF EVENTSII-B OCCUPA>tIONAL DOSE CLACULATIONSII-C DIRECT RADIATION MEASUREMENTS MADE BY METROPOLITANEDISON COMPANY ON AND OFF THREE MILE ISLAND

Il-vi

II-D CAUTIONS REGARDING THE USE OF INSTALLEDRADIATION MONITORING SYSTEM DATA
II-E ACRONYMS,

DEFINITIONS AND UNITS OF MEASURE

II-F POTENTIAL ITEMS OF NONCOMPLIANCE

II-vii1. PREACCIDENT CONDITIONS
1.1

ORGANIZATION AND RADIATION PROTECTION/CHEMISTRY STAFFING
The investigators reviewed the licensee's organizational

effect on March 28 to determine if

configuration in

it was consistent with the organizational

configuration shown in Figure 6.2-1 of the licensee's Technical Specifications
as required by Technical Specification 6.2.2.

A comparison of the actual organi-zation with the organization described in Figure 6.2-1 revealed the following
discrepancies:a.

A position titled "Superintendent of Administration and TechnicalSupport" was added to the organization on September 18, 1978 and filled
on March 1, 1979.

This means the Supervisor,

Radiation Protection

and Chemistry reported through this new position rather than directly
to the Station/Senior Unit Superintendent (Int.
b.

On March 5,

1979,

94).

the "Station Superintendent" was promoted and his

title changed to "Station Manager".

In this new position, he reported

directly to the Vice President, Generation rather than through theManager, Generation Operations.
c.

The position of "Chemical

Supervisor" has been vacant since theissuance of the Technical Specification (Int.
d.

20).

There are two "Supervisors of Maintenance" (one for each unit),

notone, as shown in the Technical Specification (Int. 120).
e.

A position titled "Superintendent of Maintenance" has been added sothat the Supervisors of Maintenance no longer report directly to theStation Superintendent (Station Manager)/Senior Unit Superintendent.
The individual designated as the Superintendent of Maintenance also
fills the position of Unit 1 Supervisor of Maintenance (Int. 120)-.

11-1-1

The licensee had discussed the promotion of the Station Superintendent to
Station Manager and the change in reporting to the Vice President, Generation
by telephone with the Region I Office of the NRC on March 5,
letter to the Director of the NRC Region I Office,1979.

In a followup

also dated March. 5, 1979,

the Vice President, Generation reiterated the changes discussed during thereferenced telephone conversation (Ref.

1).

The licensee did not, however,make application for or receive an amendment of the Technical Specifications
prior to implementing the organizational changes described above. The effectof organizational discrepancies c, d, and e on the licensee's response to the
March 28,

1979 incident is discussed further in Sections 2.2 and 3.2.

The organization of the radiation protection and chemistry staff as shownin Figure 6.2-1 of the Technical Specifications is shown as follows:

11-1-2

The actual organization of the radiation protection and chemistry staff
prior to March 28,

1979 is shown as follows (Int. 94,

Superintendent,Administration andTechnical Support

Supervisor
Radiation Protection
and Chemistry

Radiation/Chemistry
Technicians(24)(12) Technicians(12) .Technicians, Junior

11-1-3124, 133):

The resume for each member of the radiation protection staff was reviewed
to determine compliance with the requirements of Technical Specification 6.3,
which states:"Each member of the unit staff shall meet or exceed
the minimum qualifications of ANSI N18.l-1971 forcomparable positions,

except for the Supervisor of

Radiation Protection and Chemistry, who shall meet or
exceed the qualifications of Regulatory Guide 1.8,
September 1975."
ANSI N18.l-1971,

Section 4.5.2 states:"Technicians in responsible positions shall have aminimum of two years of working experience in their

speciality. These personnel should have a minimum*of one year of related technical training in addition
to their experience."
The review and interviews with the technicians indicated five of the twelve
radiation/chemistry technicians did not appear to have 1 year of related technicaltraining in chemistry or radiation protection.

Nine of the twelve radiation/

chemistry technician juniors did not appear to have 1 year of related technicaltraining.
No apparent item of noncompliance was identified since the term "should",
as used in ANSI N18.l-1971,

Section 2.2.1,

denotes a recommendation,

not a require-ment.
The review indicated six radiation/chemistry technician juniors did nothave 2 years of working experience in their speciality.

The phrase "technician

in responsible position" is not defined in any regulatory requirement.

11-1-4Basedon interviews with each technician with less than 2 years of experience,itappears that they acted in responsible positions during this emergency; however,no one of them acted alone or directed the activities of others during the period
covered by this investigation (Int. 45, 49,

51,

75, 78, 96).

Section 6.2.2.d of the Technical Specifications states,"An individual

qualified in radiation protection procedures shall be on site when fuel is in
the reactor."
To meet this requirement at 0400 on March 28, three radiation/
chemistry technicians and one junior technician were on site performing theirroutine assignments.

The distribution of assignments placed two technicians

performing radiation protection functions and two technicians performing chemistryfunctions (Int. 42).
1.2

TRAINING

1.2.1

Emergency Plan Training
The investigators examined training records and interviewed licensee person-

nel to determine if

individuals had been trained in their emergency duties as

required by the Emergency Plan (Ref.

2, 5).

Paragraph 6.1 of the licensee's Emergency Plan requires that:a.

Individuals having emergency responsibilities be trained annually inthese duties,b.

Lesson plans be provided for each category of emergency training,c.

Training include periodic examinations or assignments,d.

Offsite agencies be invited to attend established training programs.

and that

In January 1978, the licensee developed a written description of the detailed
implementation of the emergency organization training program described in Paragraph 6.1 of the Emergency Plan.

This description was incorporated into Procedure

11-1-5

1670.9,"Emergency Training and Emergency Exercise," Revision 4, dated January 16,

1978.

This procedure specified by job title the assignment of emergency duties

and the training to be provided as follows:
Emergency Function

Persons Assigned

Emergency Director

Station Superintendent
Unit Superintendent
Unit Superintendent-TechnicalSupportSupervisor of Operations
Shift Supervisor
Shift Foreman

Accident Assessment

Supervisor of Operations(Group 1, Operational)

Shift Supervisor
Shift Foreman
Control Room Operators
Radiation Protection

Accident Assessment

Supervisor,(Group 2,

and ChemistryRadiological)

Radiation Protection Foremen
Radiation/Chemistry Technicians
Shift Supervisors
Nuclear Engineers
Radiological Monitoring Team

Radiation/Chemistry Technicians
Auxiliary "A" Operators

Repair Party Team

Maintenance Shift Workers

Division Support Personnel

Supervisor,

Radiation Safety& Environmental Engineering
Radiation Protection Specialist

11-1-6

Site Protection Personnel

Supervisor of Plant Security
Site Protection SergeantsSite Protection Officers

Investigators compared Calendar Year 1978 and first quarter 1979 training
attendance records with individual job titles and potential emergency duties
to determine if all individuals were trained as required (Ref.
individuals were interviewed to determine if

5).

Selected

the training had been provided

and was of the scope and nature required by Procedure 1670.9.
This review indicated that representatives from the following offsite
agencies had participated in training provided by the licensee:

(Ref.

5)

a.

Pennsylvania Bureau of Radiologicalb.

Pennsylvania State Council of Civil Defense (now called Pennsylvania
Emergency Management Agency,

HealthPEMA)c.

Dauphin County Civil Defensed.

Pennsylvania State Policee.

Londonderry Township Fire Companyf.

Goldsboro Fire Departmentg.

York Haven Police Department.

This review indicated several discrepancies as discussed below.
During 1978,

the following personnel had not received required training:

two potential emergency directors (Shift Foremen D and A); one potential radiological accident assessment individual(Shift Supervisor E); eight potential

Radiological Monitoring Team members (Auxiliary Operators A, C, G, and K and
Radiation/Chemistry Technicians B, G, Q and V); and 37 of 38 potential Repair
Party Team members (Ref.

6).

11-1-7

Of the eight potential Radiological Monitoring Team members who were nottrained,Radiation/Chemistry Technicians B, G, Q, and V stated that 1977 entries

listing SAM-2 (portable instrument used for measuring radioiodine in the environment) training and 1978 entries listing Radiation Monitoring Team training on
their individual training records were in error because (1) they had not received
training in the operation of the SAM-2 and (2)the Radiological Monitoring Team

training in 1978 was not of the scope and content outlined in Procedure 1670.9(Refs.

2, 4, 7, 8 and Int. 154, 156).

They further stated that they had discussed

their lack of understanding of their emergency duties (and in particular theirinability to operate the SAM-2) with the Radiation Protection Supervisor in
September 1978.

Shortly thereafter, they were given a short briefing on emergency

response actions during a backshift'(the shift that they worked) but they were
not instructed in operation of the SAM-2 until approximately 1 month after the
March 28 incident.

The investigators determined that it was the briefing on

backshift response that had been entered into the radiation/ chemistry technicians'
training records as "Radiation Monitoring Team training."

An interview of the

Radiation Protection Supervisor verified that the training records were in error
and that Radiation/Chemistry Technicians B, G, Q and V did not receive training
in 1977 on SAM-2 operation.

The interview further established that the 1978

training of Radiation/Chemistry Technicians B, G, Q and V did not meet the requirements of Procedure 1670.9 (nt.

172).

The only Repair Party Team training conducted in 1978 was one session heldon October 10,

1978.

Attendees were the Superintendent of Maintenance and fivemaintenance foremen.

This session was intended to enable the attendees to train

the workers they supervised who were assigned as potential memebers of the Repair
Party Team (Refs.

2,

5).

By the end of 1978,

site protection, accident assessment and operations

personnel training were the only Emergency Plan training courses, which included"tests or assignments" as required by Procedure 1670.9 (Ref.

11-1-89).

The investigators noted that Procedure 1670.9 did not contain provisions
for training indivduals who would perform the duties of chemistry supervisor
during emergencies.
On January 18,

1979,

a Repair Party training session was held.

were mechanical maintenance shift workers.

Attendees

The instructor was a maintenanceforeman who had not been trained himself (Ref.

5).

A record of a telephone conversation between the Group Supervisor, TechnicalTraining and the Supervisor of Mechanical Maintenance dated February 1, 1979,
indicated that the Repair Party Team training requirement had been incorrectlyinterpreted to apply only to mechanical maintenance shift workers (Ref.

10).

Electrical and instrument and control shift maintenance workers had therefore
not been scheduled for training.
Four Repair Party training sessions were conducted in March 1979 by Shift
Maintenance Foreman B.

Maintenance Foreman B had not received any Repair Party

Team training prior to conducting these sessions (Ref.

5).

In a discussion

with an investigator, Maintenance Foreman B stated that he did not fully understand the material he presented and that he was not sure that the training heprovided was correct.

Consequently,

he decided not to submit the completed

Training Program Administrative forms to the Training Department.
stated that he had expressed his concerns to the Supervisor,

He further

Radiation Protection

and Chemistry and to the Radiation Protection Supervisor and had requested additional guidance from them.

On the Training Program Administrative Forms forthese classes which he prepared but did not submit, Maintenance Foreman B wrote,"wish to reschedule more specific type training when received from [Supervisor,
Radiation Protection and Chemistry] (Ref.

5)."

An investigator examined records of the Training Department's attempts tocomplete employee training in emergency duties.
1978,

In a memorandum dated April 5,

the Supervisor of Training (Acting) outlined to seven managers and supervisors

11-1-9the Emergency Plan training requirements detailed in Revision 4 of Procedure
1670.9 (Ref.

1978, the Group Supervisor, Technical Training

On September 21,

11).

met with the Radiation Protection Supervisor and the Supervisor, Technical Training
to discuss incomplete training (Ref.

12).

On November 28, 1978, the Group Supervisor, Technical Training sent a memo
requesting a schedule of Repair Party

to the Superintendent of Maintenance,

Team training and reminding him that training had to be completed by the endof 1978 (Ref.

13).

On November 29,

1978, the Group Supervisor, Technical Train-

ing sent a memo to Engineer F, who was assigned lead responsibility for emergency planning.

The memorandum reiterated that lesson plans in nine areas wereincomplete (Ref.

14).

Training,

On December 20,

1978, the Group Supervisor, Technicalinformed Engineer F that, if the lesson plans and training were notcomplete by the end of 1978,
latory requirements (Ref.

the licensee would be in noncompliance with regu-

15).

Nuclear Engineer F stated to investigators

during an interview that the lesson plans had been completed but copies had
not been sent to the Training Department (Int. 66).
telephone conversation on February 2,

A written record of a1979 between the Supervisor,

Protection and Chemistry, and the Group Supervisor,

Radiation

Technical Training,indi-cated that, once again, the subject of emergency plan training was discussed
(Ref.

16).

In this conversation,

the Group Supervisor, Technical Training,

outlined the training requirements of Procedure 1670.9 to the Supervisor,
Radiation Protection and Chemistry and pointed out that the facility was innoncompliance with regulatory requirements in several areas related to thetraining program.
A record of a telephone conversation dated February 9,

1979 between theGroup Supervisor, Technical Training and the Superintendent of Maintenance reiterated that all maintenance shift workers,

not just mechanical maintenance workers,

had to receive Repair Party Team training and that the licensee had been innoncompliance with regulatory requirements for over a month (Ref.

II-1-1017).

On February 3,

1979,

Revision 6 of Procedure 1670.9 was issued (Ref.

4).

This change stated that only those individuals trained in specific duties could
be assigned those duties in an emergency, thereby deleting the requirement that
all individuals with specific job titles be trained in potential emergency duties.
This change essentially substituted a requirement to limit persons assigned
duties in an emergency for the training requirement.
In comparing the emergency duties of selected employees during the period
March 28 to March 30,

1979, with training records,

an investigator determinedthat several individuals who had not received Procedure 1670.9 training wereassigned to Radiological Monitoring Teams and Repair Party Teams (Refs.

2,

5).

(See Sections 2.2 and 3.2 for details.)
Interviews with individuals who were trained in accordance with Procedure
1670.9 indicated that the training provided was inadequate for them to properly
perform their assigned emergency functions (Int. 154,

156, and 187).

SeveralRadiation Monitoring Team members stated that they were unsure of whether tomake radiation surveys with the meter window open or closed.

Engineer C, who

performed the initial offsite dose projection calculations on the morning ofMarch 28, and Engineer D, who assisted him, both indicated that they did notunderstand how to properly read the containment dome monitor (HP-R-214)
as a basis for their calculations (Int.

174).

used

The actions and resulting radi-

ation exposures of other emergency workers demonstrated a general lack of knowledge of radiation protection precautions to be observed during emergencies.
There was an admitted lack of understanding of how to effectively decontaminate
skin contaminated with radioactive iodine.

During an interview, the Supervisor,

Radiation Protection and Chemistry stated, "In the four and a half years that
I've been here...

I don't think we've ever faced an iodine contamination problem...

and personally,

I don't think that I really...

to handle that.

It

had the off-the-cuff knowledge

took some time and some talking with people to start getting

to a point where I felt comfortable with..,contamination" (Int. 79).

iodine contamination and external

Users of emergency monitoring equipment demonstrated

II-l-lland expressed a lack of knowledge regarding operation of the SAM-2 (Int. 154
and 156) and doubts about which of the elements of emergency organization teams
were responsible for the functions of in-plant radiation control,
repair functions,and notifications to offsite agencies.

emergency(See Sections 2.2, 2.3,

3.2, and 3.3.1 for details relating to the situations discussed above.)
1.2.2

Radiation Protection Training
Radiation protection training required for individuals at Three Mile Island

is described in Procedure 1690,"Training Requirements" Revision 6,

dated March 22,

1978, which states that the Radiation Protection Department is responsible forthe development and implementation of the program (Ref.

18).

The program wasdeveloped to meet the needs of different plant groups.

Listed below is a summaryof training programs and the groups they served.
TABLE 11-1-1
SUMMARY OF RADIATION PROTECTION TRAINING PROGRAM
Program Title
1.

Basic IGroup Served
Temporary personnel on site less than
one day.

2.

Basic II (1 hour)Temporary personnel working outside
restricted areas.

3.

Basic III (3 hours)

Permanent personnel working outside
restricted areas.

Temporary personnel

in restricted areas for more than
one day.

11-1-12

4.

All radiation workers.

Intermediate i (3 hours)

All personnel

under radiation work permits (RWP).
5.

Maintenance personnel,

Intermediate II (8 hours)

engineers,

supervisors, others requiringradiation work permit clearance.
6.

7.

Advanced Radiation

Auxiliary operators,

control room

Protection (2 weeks)

operators, senior licensed operators

Comprehensive Radiation

Radiation/Chemistry TechniciansProtection (3.months)
8.

General Employee Training

Selected temporary personnel(No time specified)(all permanent personnel once per
year)

9.

Training for local fire, police
and, civil defense departments(no time specified but detailsof training described in Procedure 1670.9)
The investigator reviewed the Generation Division, TMI Station Resume todetermine if 50 selected individuals had received the appropriate radiation
protection training described in Procedure 1690 (Ref.
that, for the individuals reviewed,

7).

This record showed

all had received the appropriate radiation

protection training.
Training of the Radiation Protection Department was reviewed in detail.
Most radiation/chemistry technicians interviewed expressed dissatisfaction withthe extent and quality of the radiation protection training they had received
in recent years.

(Int. 42, 47, 49,

50, 58, 68,

106)

11-1-1378, 82, 89,

96, 98, 99, andSection 13.2.1.5 of the Final Safety Analysis Report (FSAR),

states thatradiation/chemistry technicians will receive annually 40 hours of "Rad Chem
Tech Training Program" and 8 hours of "general employee training/retraining"and "on-the-job training."
Amendment "A" to Procedure 1690,"Three Mile Island Comprehensive Training

for Radiation Chemistry Technicians/JRS,"

Revision 6,

dated March 22, 1978,

contains documentation sheets which, when signed by an instructor,

verify thatan individual has successfully demonstrated adequate knowledge of a specifictopic.

Amendment A states that Section III training should be completed every

2 years.

A review of Section III documentation sheets showed that only one

radiation/chemistry technician had any items signed off as completed.

However,some training had been provided to the radiation/chemistry technicians as
documented on the Generation Division TMI Station Resume (Ref.

7).

Table 11-1-2 summarizes the radiation protection and chemistry related
training as documented in the Station Resume that had been provided for 22 of
the 24 radiation/chemistry technicians.

Two technicians are not included since

they have been employed for a short time (less than 3 months).
that for Calendar Year 1978,

19 of 22 technicians did not receive 40 hours ofradiation protection and chemistry training.
of General Employee Training.
on-the-job training.

The table shows

No technicians received the 8 hours

All technicians appear to have received some

For Calendar Year 1977,

15 technicians did not receive

40 hours of radiation protection and chemistry training and 12 technicians did
not receive 8 hours of General Employee Training.

All but one technician appears

to have received on-the-job training.
Technical Specification 6.4 states, "A retraining and replacement training program for the unit staff shall be maintained under the direction of the
Director-Generation Training and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.l-1971 and Appendix A of 10 CFR Part 55."

11-1-14TABLE 11-1-2
SUMMARY OF RADIATION/CHEMISTRY TECHNICIAN TRAINING
1978
Hours of Training
Radiation
Protection

I-.
C-,,

Rad.
Rad.
Rad.
Rad.
Rad.
Rad.
Rad.
Rad.
Rad'.
Rad.
Rad.
Rad.
Rad.
Rad.
Rad.
Rad.
Rad.
Rad.
Rad.
Rad.
Rad.
Rad.

Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.
Chem.

Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.
Tech.

7
8
7
17
9
13
8
15
11
9
11
12
5
9
13
8
8
10
12
14
26
11

Chemistry
80*
6
1
1
83*
1
85*
2
2
1
2
1
81*
1
2

On JobGeneralTraininq Employee

24
24
24
24
24
24
24
24
24
24
24
24
24
24
24
24
24
24
24
24
24
24

Drills
6
3
13
5
3
5
4
6
5
6
5
3
5
3
6
3
5
4
3
5

Radiation
Protection
324**
4
238**
4
4
238**
3
5
7
3
4
4
4
240**
237**
4
3
5**

239**
5
8
3

1977
Hours of Training
On JobGeneralChemistry Training Employee Drills

80*
88*
80*
8
8
8
8
11
86*
80*
1
8
81*
9
2

8
33
20
32
20
20
32
28
30
29
32
28
28
21
22
28
33
20
20
8
39

4
8
8
4
8
8
8
8
4
4
2
8
4
4
8

6
6
7
8
9
2
8
16
7
5
7
4
4
5
8
9
6
11
11
11
5

NOTES:
1.
2.
3.
4.

5.

Hours of training have been rounded off to the highest whole number.
*Indicated a two week vendor water chemistry school.
**Indicates 6-8 week comprehensive Health Physics school provided on site by contractor.
The 24 hours of on-the-job training for 1978 was described on the "Training Program Administrative Form:as "Unit
II HP and Chem Startup."
During interviews, no technician could describe this training. Item 7 on the "Training
Program Administrative Form" titled "Instructor/Trainee's Evaluation" contained the following instructor's comment,"During the past 6 months, the techs have been getting good instructions in both HP and Chem aspects of Unit 2 Start-up."
The only training provided in 1979 prior to the accident was a 4-hr chemistry class to six technicians.

Section 5.5 of ANSI N18.1-1971 (Ref.
"5.519) states:

Retraining and Replacement Training.

A training program shall be established which maintains the
proficiency of the operating organization through periodic training
exercises,

instruction periods,

and reviews covering those items andequipment which relate to safe operation of the facility and throughspecial training sessions for replacement personnel.

Means should

be provided in the training program for appropriate evaluation of
its effectiveness."
5.5.1

Retraining

A retraining program should include:
1.

Plant startup and shutdown procedures;

2.

Normal plant operating conditions and procedures;

3.

Operational

4.

Emergency plans and security procedures;

5.

Abnormal operating procedures;

6.

Emergency shutdown systems;

7.

Changes in equipment and operating procedures;

8.

General safety, first aid, and radiation safety;

9.

Alarms and instrumentation signals; and10.

Operation of selected auxiliary systems important to overall

limitations, precautions,and set points;

plant safety."
In reviewing the retraining described by Procedure 1690 and from interviews
with respresentatives of the Radiation Protection Department, an investigator
determined Items 1, 2,

3, 5,

6, 7, 9 and 10 of Section 5.5.1 of ANSI N18.1-1977

were not included in the retraining program for radiation/chemistry technicians.
One item of particular importance was the method for informing technicians ofchanges in procedures.

According to the Radiation Protection Supervisor, anote was placed on the bulletin board stating that a procedure had been changed

11-1-16

(Int. 124).
No acknowledgement sheet was routed to ensure that each technician
read the change.
Acknowledgement sheets were not used, according to the Radiation
Protection Supervisor, because the technicians had refused to sign them.
The Supervisor Radiation Protection and Chemistry and the Radiation Protection Supervisor stated that there was no formalized retraining program forthemselves or their foremen (Int.

20 and 124).

Station records showed all had

received some retraining but the training received did not include all of theareas specified in Section 5.5.1 of ANSI N18.1-1971 (Ref.

7).

Members of the radiation protection and chemistry staff provided the
majority of radiation protection training for other members of the facility
staff.
The licensee has not maintained a retraining program for members of the
Radiation Protection Department that met or exceeded the requirements andrecommendations of Section 5.5 of ANSI N18.1-1971.
1.3

EMERGENCY DRILLSParagraph 6.1.2 of the Emergency Plan requires the licensee to conduct,

among others,

an annual Site or General Emergency drill and an annual Repair

Party Drill.

The procedure to be followed in planning, conducting,menting drills is specified in Procedure 1670.9,

and docu-"Emergency Training and Emer-

gency Exercise."
An investigator reviewed available documentation'and interviewed licensee
employees to verify that an annual Site or General Emergency Drill and a Repair
Party drill were performed during Calendar Year 1978, as required.
During Calendar Year 1978, the licensee conducted seven radiation emergency
drills having a scope equivalent to Site/General Emergencies.
drills included Repair Party activites.
drill were as follows (Ref.

One of theseThe date and general scenario of each20):

11-1-17

Date

Scenario

October 23,

1978

Waste Gas Decay*Tank Release (Unit 1)

October 25,

1978

Radioactive Liquid Release (Unit 2)

October 27,

1978

Once-Through Steam Generator(OTSG)

Tube Rupture (Unit 1)
October 30,

Loss of Water from Spent Fuel Pool

1978

with Fuel Raised (Unit 1)
November 2, 1978

Loss-of-Coolant with severe core damage,fuel melting and little core cooling,
containment leak at maximum design rate
(Unit 2)

November 6,

1978

Same as above, but Unit 1

November 8, 1978

OTSG Tube Rupture,

Unit 1.

During each of the seven drills, licensee observers were positioned atvarious locations to evaluate the response organizations'activities.

NRC repre-

sentatives were present for and witnessed the conduct of the November 8,
drill.

1978(Results of this inspection were reported in IE Inspection Report Number

50-320/78-34.)

Each of the seven drills was followed by a critique in which

the discussion highlighted areas for further evaluation and possible improvement.
An investigator reviewed action item assignment sheets that were preparedpursuant to Procedure 1670.9 after the series of seven radiation emergency drills
held in 1978.

The purpose of this review was to determine the status of theitems on March 28 (Ref.

21).

'

This review indicated that the following action

items were still open:.

11-1-18

a.

Post RM-L7 sensitivity, alert, and high-alarm values adjacent to theNo.

b.

6 recorder in the Unit 2 control room,

Ensure that newly revised site maps [in monitoring kits] replacethose outdated ones,c.

Ensure that emergency monitoring equipment has had an inventory andoperational check out prior to use (especially SAM-2),

d.

Change accountability,e.

Check out operation of Unit 2 450 MHz radio communications and black

Procedure 1670.7,

phone line circuit with offsite agencies,f.

Remove labels from Radeco air samplers used in kits,g.

Review radiation monitoring team communications with recommendations
for improvement,

h.

and

Establish routine surveillance program for GAI page system.

The following action items were identified as being closed:a.

Check page system units for proper operations in both units; ensure
merge-isolate switch is correct,b.

Ensure all operations personnel know where to report and know their
responsibilities during an emergency (especially auxiliary operators),c.

Need additional status board in control room,d.

Replace emergency kit air sampler,e.

Change Emergency Plan for new telephone numbers,

II-1-19

f.

Ensure operations personnel read calculations Procedure 1670.4,
especially liquid release calculations,g.

Engineers should immediately relieve the operators in the ECC using
communications hardware and the status board,h.

Investigate walkie-talkie availability for emergency use only,i.

Communications improvement ECC/ECS,j.

Need communications hardware for green office building (west side),k.

Check monitoring kits for high-range dosimeters,

1.

ECS dose assessment/RMS/ECS communications coordination,m.

Ensure Site and General Emerqency criteria are reviewed by all
operations personnel,

n.

Ensure replacement of AC power cord in SAM-2 Kit No.

o.

Check out SAM-2 instruments.

2,

The investigator compared the list of "open" action items with events thattook place on March 28 to determine if

their being open led to any degradationof the licensee's response.
The "open" action item relating to inventory and operational check out ofemergency equipment prior to use had a parallel occurrence on March 28.
A radiation monitoring team left the site for Goldsboro at approximately 0730.

They

did not inventory/check their emergency montioring kit before leaving.

Upon

arrival in Goldsboro at approximately 0830 they found that the SAM-2 in the
kit would not operate.

11-1-20

With regard to the "open" action item regarding a change to Procedure 1670.7,
the investigators noted that the licensee's actions on March 28 appeared toconstitute noncompliance with Technical Specifications relative to the implementation of changes to Emergency Plan Implementing Procedures.

As a result of the

Calendar Year 1978 emergency drills, licensee management determined that themethod for conducting accountability and assembly of personnel needed to'be
improved.

The task to revise the applicable Emergency Plan Implementing Procedure

1670.7 was assigned to the Radiation Protection Supervisor (Ref.

21).

On October 13,

1978 the Radiation Protection Supervisor issued a memorandum to all departments
outlining a revised assembly and accountability method (Ref.
Through interviews with site protection personnel,

22).

the investigator determinedthat the proposed method of assembly and accountability outlined in the October 13
memorandum was implemented during the response to the March 28 incident (Int.
72,

52,

32).

The investigators determined that initially some members of thesite protection force followed the accountability Procedure 1670.7, while others
followed the method described in the October 13 memorandum.

This created somedegree of confusion and delayed the prompt attainment of full accountability.
The investigator also noted that Procedure 1670.5,"Onsite Radiological

Monitoring," had not been revised to include the dispatch of an assembly area
monitor to the North Warehouse (an additional assembly area established by the
October 13 memorandum).

During an interview of the Radiation Protection Super-visor, he acknowledged that the procedure should have been revised before thememorandum was issued.

He also stated that Procedure 1670.7 had not been revisedby March 28,

172).

1979 (Int.

The investigators determined that the use of the accountability and assembly
method described in the Radiation Protection Supervisor's October 18 memorandum
is an item of noncompliance with Technical Specification 6.8.2, which requiresthat Emergency Plan Implementing Procedures be reviewed and approved by the
Plant Operations Review Committee (PORC)

prior to implementation.

The investigator also compared the action items listed as "closed" with
events that took place on March 28 to determine the adequacy of the corrective
actions.

11-1-21

Of the items listed as "closed," two items appear to have recurred.

Basedon the demonstrated performance of individuals in recognizing and declaring
the emergency,

it appears that the review of the Site and General Emergency

criteria with operations personnel was not adequate.

Interviews with the shiftoperating crew of March 28 indicated some uncertainty regarding the applicability of certain conditions to Site and General Emergency criteria still remained (Int.

145) (see Section 2.1).

The check out of SAM-2s appeared to have corrected an operational problem
However, since two of four SAM-2s were inoperable on

existing at the time.
March 28,

itis apparent that the corrective action was not adequate to prevent

recurrence of random inoperability (Ref.

23,Int. 82) (see Section 1.5).

The investigators noted that there were no explicit regulatory requirements
pertaining to the verification of the timeliness of corrective action implementation or evaluation of the adequacy of corrective action.
INSTALLED PLANT EQUIPMENT

1.4
1.4.1

Area,
1.4.1.1

Process,and Atmospheric Monitors

Operational Status

The investigators determined the operational status of area, process,andatmospheric monitors on March 28 by review of selected sections of records (Ref.
24-40),

discussions with licensee staff, and observation of monitor operation.

The Out-of-Service Logs showed that the status of the monitors on March 28
was as follows (Ref.

41,

42):

Description

Total Installed

Total in Service

Unit 2 area monitors

21

20

Unit 2 atmospheric monitors

16

15

Unit 2 liquid monitors

1212

Unit 1 area monitors

15

15

10

8

9

8

* Unit 1 atmospheric monitors
Unit 1 liquid monitors

11-1-22

The following process, area and atmospheric monitors or monitor recorders
were out-of-service on March 28 (for the reasons stated in the Out-of-ServiceLogs):a.

Unit 2 makeup tank area radiation monitor (HP-R-206)(since February 15,

b.

1979)

Unit 2 waste gas decay tank 1A gas monitor (WDG-R-1485)
inoperable (since February 16,c.

Unit 2 liquid discharge monitor readout (RM-L7)

Unit 1 radiochemicalmonitor (RM-A12)

f.

- "signal to
1979)

Unit 1 miscellaneous sump discharge liquid monitor (RM-L8)"decon liner" (since April 22,e.

- pump1979)

the RM-L7 recorder is false" (since March 13,d.

- "bad cables"

-

1977)

laboratory and nuclear sampling room air- "wiring not hooked up" (since April 18,

Unit 1 control tower air monitor (RM-Al)(since Febuary 22,

1978)- pump inoperable1979).

The record of the weekly surveillance check of RM-A12 conducted on March
17 shows the monitor was out-of-service.

Records of monthly surveillance per-formed on March 18 and calibration on March 19,
operational on March 19 (Ref.

35, 39, 43).

showed that the monitor was

The weekly surveillance check per-formed on March 24 showed the monitor out of service again (Ref.

40).

Main-

tenance Foremen E and F, in discussions, and Radiation Protection Foreman B,
in an interview,stated that there had been problems with the pump on the sampler

in the past (Int. 21).
been repaired.

However,

Maintenance Foreman F thought the pump had

He said that he recalled that, when he examined the monitor

after the calibration on March 19, the pump was operational.

He stated thatthe monitor would very frequently alarm when samples were taken and therefore
it was unplugged to stop the irritating noise from the alarm.

He stated that

he had discussed with Radiation Protection Foreman D the possibility of changing

11-1-23

the sampler alarm setpoints so that the sampler would be used, but this change
had not been made before March 28.

Because of the high dose rates in the area,

the sampler had not been checked since March 28.
If the Unit 2 makeup tank area radiation monitor had been operational during
the accident,

it

could have provided an accurate measure of increasing radiationlevels in the auxiliary building.
tion levels, but it

This monitor did respond to increasing radia-

did not respond accurately.

The nuclear sampling room air

monitor could have been used to collect air samples when reactor coolant samples
were taken in this room.

Because the control tower air monitor was inoperable,

there was no real-time monitor of Unit 1 control room air radioactivity until
the monitor was placed back in service on March 30 at 0600.
Unit 2 Technical Specifications 3.3.3.1 and 3.3.3.6; Unit 2 Environmental
Technical Specifications 2.1.1.A, 2.1.2.A, and 2.1.2.B.1; and Unit 1 Environmental Technical Specifications 2.3.2.A and 2.3.1 require that certain radiation
monitors be operable.

The monitors required by these Technical Specifications

were operable on March 28.
1.4.1.2

Calibration

The licensee is required by Unit 2 Technical Specifications 4.3.3.1 and4.3.3.6,

by Unit 2 Environmental Technical Specification 2.1.1,

and by Unit 1

Technical Specification 4.1.1 to channel check, functionally test, and calibrate certain process, area,

and atmospheric monitors at specified frequencies.

Sections 11.4 and 12.1 of the Final Safety Analysis Report (FSAR)

list othermonitors installed in Unit 2 that are not required by technical specifications.
Investigators examined records of monitor calibrations,functional tests,

and channel checks done in normal surveillance to determine if monitors were
calibrated on March 28 (Ref.

24-36, 38-40,

44).

Technical Specifications were in calibration,
identified.

All monitors required by

but certain discrepancies were

The preoperational test results of the atmospheric monitors,"Radiation System Test - Atmospheric Monitors," TP 360/lA, were incomplete.
The tests showed that ventilation exhaust flow rates in the auxiliary and fuel11-1-24

handling buildings were different than designed (Ref.

26).

Engineer M statedin discussions that the air sampler flow rates had been adjusted for isokinetic
sampling and that the test would be signed as complete after new sampler nozzles
were installed and all sampler flow rates adjusted to be the same (Ref.
The vent stack monitor (HP-R-219)

45).

sampler consisted of two sample probes

and a flow measuring probe located at different elevations in the center ofthe stack (Ref.

44, 45, 46).

The licensee considered,prior to startup of Unit

2, modifications to exhaust sampling systems in order to meet the recommendationsof Regulatory Guide 1.21,"Measuring,

Evaluation and Reporting Radioactivity

in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous
Effluents from Light Water Cooled Nuclear Power Plants," but did not make themodifications since Regulatory Guide 1.21 was not a requirement (Ref.
Preoperational

47).

test results, "Radiation Monitoring System Test - Area

Monitors" TP 360/lC, showed that although the output from the containment dome
monitor (HP-PR-214)

to the meter in the control room was calibrated, the output

was not calibrated to the readings on the chart recorder (Ref.

24).

The chart

recorder scale read from 0.1 to 104 mR/hr but the output from the meter ranged
from 0.1 to 107 mR/hr.

The output from the meter read less than full-scale on

the chart recorder (104 mR/hr) when the meter read full-scale (107 mR/hr).
Work Request 2761 for installation of a different recorder was cancelled (Ref.
48).

Maintenance Foreman G stated in discussions that an instrument technician

had begun to calibrate the monitor output to the existing recorder but the work
was not completed prior to the accident.

There was no regulatory requirementthat this monitor have a chart recorder.

Any value for this monitor taken off

the chart during the period March 28 to 30 would have been in error.

A recordof the readings from this monitor could have been useful for postaccident
evaluation.
1.4.1.3 Set Points
In examining the bases for alarm setpoints for Unit 2 monitors,

an inves-

tigator determined that the setpoints for air particulate monitors wereincorrectly based on one-half of the Unit 1 Technical Specification release11-1-25

rate limit of 0.63 pCi/sec rather than the one-half of Unit 2 Technical Specification limit of 0.3 pCi/sec.

The licensee had reduced the alarm setpoint by

25% to account for possible meter error (Ref.

49).

Even though the basis forthe setpoint was incorrect, the alarm setpoint was set less than the Unit 2
Technical Specification release rate limit.
An investigator also determined that the alarm setpoint for the reactor
coolant letdown monitor was based on operation of the monitor without an attenuating shield, but Engineer M stated that the shield was in place (Ref.

44).

If this monitor were used to determine the amount of radioactivity in the reactor
coolant,

it would result in an underestimation.

1.4.2

Radwaste Systems
Liquid Radwaste System

1.4.2.1

An investigator reviewed the Primary Auxiliary Operator's Log to determine
the status of the radwaste system prior to the reactor trip at 0400 March 28
Liquid radwaste system tanks were filled to near 60% capacity

(Ref.

50).

(Refs.

50, 51).

The Supervisor,

Radiation Protection and Chemistry statedthat Unit 2 depended on Unit 1 for radwaste processing support.

Unit 1 wasstarting up-f-rom-a-refueling outage and generating radwaste that required
processing in order to continue startup.

A Unit 1 spill of 20,000 gallons ofcontaminated water from the fuel transfer canal into the reactor building
generated more radwaste near the end of the outage than would normally bepresent (Int. 79).

The investigator concluded that there were no known liquidradwaste problems existing at the time of the accident that caused theaccident to be significantly worse than itotherwise would have been.

Table

11-1-3 shows the capacity of the Unit 2 radwaste system as recorded sometime
between 2300 March 27 and 0400 March 28 (Ref.

11-1-26

50).

TABLE 11-1-3
AUXILIARY BUILDING RADWASTE CAPACITY
Approximate
Sump/Tank
NumberName
Designation
Auxiliary Bldg
Sump
Auxiliary Bldg
Sump Tank
WDL-T-5
Miscellaneous
Waste Storage
Tank
WDL-T-2
Neutralization
Tank
WDL-T-8A
Neutralization
Tank
WDL-T-8B
ContaminatedDrain Tank
WDL-T-IIA
ContaminatedDrain Tank
WDL-T-IIB
Reactor Coolant
Bleed Holdup
Tank
WDL-T-lA
Reactor Coolant
Bleed Holdup
Tank
WDL-T-lB
Reactor Coolant
Bleed Holdup
Tank
WDL-T-IC
Evaporator
Condensate
Test Tank
WDL-T-9A
Evaporator
Condensate
Test Tank
WDL-T-9B
Concentrated
Waste Tank
WDS-T-2
Containment
Spray Room
Sump
1A
Containment
Spray Room
Sump
IB
Decay Heat
Removal Sump
1A
Decay Heat
Removal Sump
IB

Capacity(1Gal
Ft

Last Routine( 2 )Reading (ft)Percent
Capacity
Remaining

7645

6+

3.8

37%

3215

4.5+

3.4

24%

20040

10.25+

6.6

33%

8780

10.75

10.5

< 1%

8780

10.75

10.0

< 1%

2660

5.25+

4.0

23%

2660

5.25+

1.25

76%

82280

13

6.5

59%

82280

13

9.0

39%

82280

13

9.0

39%

11863

14

0.4

100%

11863

14

11.5

18%

9649

10

0

100%

359

4

2.3

43%

359

4

1.7

58%

359

4

2

50%

359

4

2.5

38%

(1)

From FSAR and plant procedure 2104-4.1(2)

Primary Auxiliary Operator's Log for March 28 prior to 0400.
11-1-27

1.4.2.2

Gaseous Radwaste System

An investigator reviewed United Engineers and Constructors,
Sheets, Testing Data of Piping Systems," MTX No.
124.3.4,

124.3.5 and 124.3.6.

Inc.,

"Data

124.3.1, 124.3.2, 124.3.3,

These data sheets indicated that the gaseous

radwaste system had been pressure tested by sections prior to final functional
testing.

"Waste Gas Disposal Functional Test," TP 231/3, had been completed

and approved by March 21,

1979.

This procedure contained documentation thatseveral deficiencies noted during the functional test were addressed by either
correcting the deficiency or changing operating procedures to compensate forthe deficiency.
There have been a number of maintenance problems with the waste gas
system since plant startup.

Examples are:Work Request 4979

Waste gas compressor, WDG-P-lB,

Work Request 4985

Waste gas compressor, WDG-P-lA, makes loud noise.

Work Request 4262

Waste gas compressor,

WDG-P-lA,

overpressurized

no seal water levelin tank, check level control pump operation.
These work requests were outstanding at the time of the accident.

Inaddition, Control Room Operator J stated that makeup tank vent valve MU-V-13
was suspected to be leaking (Int. 57).

Some of these problems may have caused

releases to be larger than they would have otherwise been.
1.4.2.3

Heating and Ventilation Systems

An investigator reviewed the following documents pertaining to theauxiliary building and fuel handling building ventilation systems:
Auxiliary Building Ventilation Functional Test, TP 173/2

11-1-28

Fuel Handling Building Ventilation Functional Test, TP 177/2
Filter Efficiency Test, (SP-2)
Burns and Roe Field Questionnaire No.

590

Burns and Roe Field Questionnaire No.

2513

GPU Start-up Problem Report 2646
Amendment 3 to Operating License No.

DPR-73,

March 24,

Attachment 2 to Operating License No.

DPR-73,February 8,

1978
1978

Technical Specifications 3.9.12 and 4.9.12.

In addition, an investigator discussed the history of these ventilation
systems with the Unit 1 Maintenance Supervisor; GPUSC Project Site Manager,
Superintendent of Maintenance; Supervisor,

Radiation Protection and Chemistry;and Unit 2 Maintenance Supervisor.
The history of these filters prior to the accident may have had asignificant impact on their performance during and after the accident.
During the testing of the auxiliary building and fuel handling building
ventilation systems on February 13-16,

1978,

leakage through bypass dampers

AH-D-4020A on the auxiliary building and AH-D-5683 and AH-D-5890 on the fuelhandling building caused the tests to be unacceptable (Ref.

52).

The bypass

dampers were temporarily sealed and acceptance retesting was performed on
March 9,1978.

After retest, the filters met or exceeded the acceptancecriteria of 99.97% removal of dioctylphthalate (DOP)

in the HEPA filters and99.95% removal of freon in the charcoal adsorbers (Ref.

11-1-2952).

No removal efficiency was performed after the bypass lines weresemipermanently sealed on April 7,
since that date.

1978.

No retesting has been performed

There were no regulatory requirements to conduct retests.

No record of any other alterations has been found at this time.
Investigators attempted to determine if

the filters could have beenpoisoned by fumes from large scale painting operations.

The GPUSC Project

Site Manager stated that any painting performed in the fuel handling andauxiliary building after acceptance testing of the filter trains was done by along-term contractor who is no longer onsite.

The GPUSC Project Site Manager

also stated that he knew of no way to establish how much painting was done,

ifDuring tours of the auxiliary building and fuel handling building by anany.

investigator, there were no indications of recent large-scale painting operations.

Several high radiation areas were not examined.

A review of testing requirements showed that there were no technicalspecification requirements for particulate and halogen collection efficiency
determinations for the auxiliary building ventilation system filters.
Technical Specification 3.9.12 requires the fuel handling building air
cleanup systems to be operable whenever irradiated fuel is in the pool.
has been no irradiated fuel in the pool.

However,

There

the Units 1 and 2 fuel pools

are not physically divided for ventilation flow purposes and the.Unit 2 fuelhandling building exhaust ventilation draws from this common air space.

Thereis irradiated fuel in the Unit 1 fuel pool.
Technical Specification 4.9.12.c requires laboratory analysis of a representative sample from the fuel handling building charcoal adsorbers after every
720 hours of operation.

Item F.2 of Attachment 2 to operating license DPR-73

defers the applicability of Technical Specification 4.9.12.c until replacementof charcoal adsorbers during the first refueling outage.
The Supervisor,

Radiation Protection and Chemistry stated that the charcoal

halogen adsorbers were ordered by Burns and Roe before they learned that they
(fuel handling building filters only) must meet Regulatory Guide 1.52,

11-1-30

Revision 1,

July 1976 guidelines for methyl iodine removal efficiency.

Testing of theordered adsorber indicated that the removal efficiency was slightly less than
the regulatory guidelines.

According to discussion Item No.

6 of "Safety

Evaluation by the Office of Nuclear Reactor Regulation supporting Amendment
No.

3 to Facility Operating License No.

DPR-73," the deferment of applicability

for Technical Specification 4.9.12.c was intended to cover minor deviations
from the specified chemical analysis requirements of Regulatory Guide 1.52,
Revision 1, July 1976.
GPU Startup Problem Report 2646, February 14,

1978, discussed the problemof leaking bypass dampers and referenced Field Question No.
1978.

2513,

March 30,

The resolution was to "Block off the following dampers on either the

upstream or downstream side with 18 gauge sheet metal making sure all joints
are airtight:lift

AH-D-5890,

In terminal box IT2782

and tag wire "45" of cable IT 27826 in terminal box IT 3119,

wire "31"of cable IT 3120C.

per this ECM (ECM
No.

AH-D-5683, and AH-D-4020A.

April 3,and tag

Wires should be tagged to indicate their removal

S-5915 dated March 24,1978)."

Step seven of Field Question

2513 indicates that ECM S-5915 was completed on April 7,

Question No.

lift

1978.

Field

2513 was approved by the Plant Operations Review Committee on

1978.

Subsequently, ECM S-5915 Revision 1 was initiated to purchaseand install new bypass dampers of a type that would not permit leakage whenclosed.

Proposed intallation was to be during the first refueling.

A "Burns and Roe Nuclear Safety Review" on ECM S-5915 was signed on March 24,
1978,

and ECM S-5915 Revision 1 was signed on June 20,

1978.

The review of

ECM S-5915 Revision 1 does not appear to address adverse effects on the adsorberthat might result from continuous use, although the system description statedthat the purpose of the bypass system was to prolong the life of the filter
trains (Ref.

53).

It was determined from discussions with the Unit 2 Maintenance Supervisor
and the Supervisor,

Radiation Protection and Chemistry that, since the completionof acceptance testing, (approximately 1 year prior to the accidest) all ventilation flow from the fuel handling and auxiliary buildings had been through thefilter banks.

According to the Superintendent of Maintenance and Shift Supervisor

F, there were no run-time records kept on these filter banks.

11-1-31

1.4.3

Communications Equipment
The March 28 availability and operability of the following normal andemergency communications equipment listed in paragraph 5.4 of the Emergency
Plan were determined through discussions with licensee personnel and review ofavailable records:a.

Radiation Emergency Alarm;b.

Public Address System;c.

Dial telephone (Metropolitan Edison Tie Line);d.

Direct current (battery) powered telephones;e.

Pennsylvania Bell telephone;f.

Dauphin County Radio Monitor;g.

Metropolitan Edison System Radio (Utility Radio);h.

Frequency modulation (FM)i.

NAWAS (National Warning System, connected to Pennsylvania Emergency

radio; andOperations Center and State Police Troop Headquarters).
An investigator determined that the only inoperative item was the Dauphin
County radio monitor in the Unit 2 control room.
frequency scanner and, prior to March 28,
frequencies.

Work Request No.

the problem (Ref.

54).

This radio monitor is ait would not receive certain emergency

C1406 was prepared on February 27,

1979 to correct

The set was still at an offsite repair facility andwas not in service on the morning ot March 28.

The absence of this item had

no effect on the licensee's ability to communicate or coordinate with the Dauphin
County Civil Defense and other agencies since it was only intended to monitor
radio communications.

The licensee used telephone to communicate with Dauphin

County Civil Defense and other officials during the emergency and had backup
communications provisions through the utility radio.
A licensee evaluation of radio communications between onsite and offsite
Radiation Monitoring Teams, the Emergency Control Station (ECS),

and control

rooms had been ongoing since shortly after a 1977 radiation emergency drill.

11-1-32

Areas under evaluation were:range,

number available,(1) the portable walkie-talkie transmit

and "dead spot" locations,(2)plant page system in all areas within the site boundary,

the audibility of the(3)

the procedure toensure merging of the Unit 1 and Unit 2 communication systems during emergencies,(4)timely reporting and repair of communication system malfunctions,lation of an override to the plant page external speaker timer,

(6)

(5)

instal-and improve-

ment of the Emergency Control Station Emergency Control Center communication
link.
For

the 9 listed available communications systems,

the following werethe methods in existence for testing the operability of the systems (Ref.
a.

Radiation Emergency Alarm - tested every Friday at noonb.

Public Address System - no surveillance testing, however,55):

maintenance was to be performed when malfunctions were reportedc.

Dial Telephone (Metropolitan Edison Tie Line) - no surveillance
testing, however,maintenance was to be performed when malfunctions

were reportedd.

Direct Current (Battery) Powered Telephones - no surveillance testing,but to be checked periodicallye.

Pennsylvania Bell Telephone - no surveillance testing, but maintenance
was to be performed when malfunctions were reportedf.

Metropolitan Edison System Radio (Lebanon Frequency) - the transmitter
was to be checked yearly for frequency and modulation per Federal
Communications Commission regulation; was to be checked weekly forproper operationg.

Dauphin County Radio Monitor - to be tested yearlyh.

Frequency Modulation (FM)Radio - to be frequency calibrated every 6months11-1-33

i.

NAWAS - to be tested weekly via NAWAS roll call.

An investigator noted that, with the exception of the Radiation Emergency
Alarm,
1.5

there are no regulatory requirements in this communications area.

RADIATION PROTECTION EQUIPMENT AND SUPPLIES
Portable Radiation Survey Instruments

1.5.1

Section 12.3.2.3 of the FSAR states, "Portable radiation survey instrumentsA
are maintained under the responsibility of radiation protection personnel.
sufficient inventory is available to allow for periodic calibration, maintenanceand repair.

These include detection instruments for measuring alpha, beta,

gamma, and neutron radiation.

Monitoring instruments and count rate meters

are located at the Access Control Point."

Table 12.4-1 of the FSAR lists several

the type of radiation detected, and their range and use."

types of instruments,

An investigator examined Forms 1763-1 maintained pursuant to Health Physics
Procedure 1763,
1977,"H.P Instrument Calibration/Status,"

Revision 1, dated December 1,

to determine the availability of portable radiation protection equipment

on site at 0400 March 28.
available for use if

Portable radiation survey instruments were considered

they were operable and had been calibrated within thefrequency specified in Health Physics Procedure 1749,
Calibration Schedule,""Portable/Lab Instrument

Revision 4, dated December 1, 1977.

The results of thisreview are summarized below.

Instrument

Description

Total
Inventory

Available for
Use on March 28

Eberline Teletector

Wide range beta-gamma;
maximum gamma 1000 R/hr

16

4

Eberline PIC-6A

Wide range beta-gamma;
maximum gamma 1000 R/hr

14

4

Eberline RO-2

Intermediate range betagamma; maximum gamma 5 R/hr

125Eberline E-520

Intermediate range betagamma; maximum gamma 2 R/hr

14

6

11-1-34

Eberline PNR-4

Intermediate range neutron;
maximum neutron 5 rem/hr

20Victoreen 808 Vamp

Intermediate range gamma;
maximum gamma 100 mR/hr

50

Eberline PAC-4S

Alpha count rate
maximum 200 Kcpm

20Eberline RM-14
and HP-210

Bet -glawa coumt rate
maximum 50 Kcpm

18

18

Radeco Grab Sampler

Air sample collection
devices14

14

Staplex Grab Sampler

Air sample collection
devices

4

2

UNICO Grab Sampler

Air sample collection
devices

1

0

MSA Model F

Air sample collection
devices

55

The above table indicates that less than one-half of the inventory of portable radiation dose rate instruments was available for use at the time of theaccident.

The Radiation Protection Supervisor (Int.

Technicians C, I, and L (Int. 42,

124) and Radiation Chemistry

58, and 82) indicated this shortage was not

unusual considering that Unit 1 had just completed a refueling outage.

Thesesame individuals indicated that the availability of portable survey instruments
had been a chronic problem.
Radiation survey instrument maintenance was the responsibility of the Instrument and Control Section.

Calibration of survey instruments was the responsibilityof the radiation/chemistry technicians.

Radiation/Chemistry Technician 0 statedthat a few days prior to the accident he and another technician attempted tocalibrate over a dozen instruments (Int.
be calibrated.

78).

Three of the instruments could

While performing an inventory of survey instruments,

the tech-nician reported finding "most of our instruments in the repair shop, and when
they weren't in the shop being worked on, they were back in the parts cage,sitting, because they didn't have parts to fix it.
instruments.

Very few of them were available.

you almost had to fight for it."That was the status of the

When you wanted an instrument

The Radiation Protection Supervisor stated

11-1-35

that they had repeatedly contacted the Instrument Section to repair survey
instruments; but, because of apparent manpower shortages,
seemed to be done (Int.

124).

the repairs neverInan effort to alleviate this problem, thelicensee was evaluating the possibility of having an outside vendor perform
radiation survey instrument repair and calibration service.
Interviews with Auxiliary Operator C and radiation chemistry/ technicians
showed that thelack of radiation survey instruments directly contributed tounnecessary personnel exposure as discussed in Section 3.2 of this report (Int.
164).
Many additional survey instruments and air samplers were necessary tosupport the in-plant and environmental monitoring after the accident.

The

Supervisor, Radiation Protection and Chemistry stated that there was a shortageof equipment because plans had not considered that an emergency would require
such frequent operations in very high exposure rate areas.

No clear regulatory

requirement or licensee commitment established minimum inventories for portableradiation survey instruments at this facility.
1.5.2.

Radiochemistry Laboratory Equipment

Section 12.3.2.7 of the FSAR states, "The Unit 1 Radiochemistry Laboratory
in conjunction with a separate shielded Counting Room is used for radioactivity
analysis.

Counting equipment such as G-M,

scintillation and proportional counters

is available."
The status of radiochemistry or health physics laboratory radiation measurement equipment was established by review of the "Laboratory Instrument Backgroundand Efficiency Log Book," discussions with members of the radiation protectiongroup,and direct observation of'equipment status by the inspector.
Unit 1

Instrument

Description

Total
Inventory

Ortec 401A

Gross beta

22

Ludlum 2000

Gross beta22

11-1-36

Available
For UseLudlum 2000

Gross alpha1

I

Packard Model 3320

Liquid scintillation
spectrometer system

1

1

Beakman Wide Beta 11

Gross alpha/beta

1

0

NMC Proportional
Counter PCC-iiT

Gross alpha/beta

10Lithium Drifted
Germanium, Ge(Li),Multi-CharTnel Analyzer

Gamma

1

1

Unit 2
Ludlum 2000

Gross beta

3

2

BeckmanLiquid scintillation

1

0

Lithium Drifted
Germanium, Ge(Li)Multi-Channel Analyzer

Gamma

1

0

Interviews with Radiation Protection Foreman A (Int. 7)

indicated that

the Ge(Li) and Beckman Liquid Scintillator had never been operable since
startup of Unit 2.
Although there appears to be more equipment than listed in the FSAR,
during the accident the use of vital (multi-channel analyzer) counting equipment was lost because of the presence of airborne radioactive material
equipment area.

in theThis directly affected operations of the facility as describedin Section 3.2.
1.5.3

Personnel Thermoluminescent Dosimetery
Section 12.3.3.1.2 of the FSAR states, "All radiation workers at thestation are issued film badges/TLDs and are required to wear such at all times
while within the controlled areas.
luminescent dosimeters (TLD)

Special or additional film badges/thermo-

are issued as may be required under unusual condi-

tions at the discretion of radiation protection personnel.
ditional devices are processed as required.

These special or ad-

The film badge/TLD of any individualis processed whenever it appears that an overexposure may have occurred."

11-1-37

The licensee maintained an onsite dosimetry program using a Harshaw,
TLD System.

The Harshaw system used a natural lithium fluoride (LiF)

dosimeter and Model 2271 reader.

Health Physics Procedure 1642,

and Calibration of the Thermoluminescent Dosimetry System,"
September 28,

Inc.,

two-chip"Operation

Revision 1, dated

1977, provided the information necessary for the radiation/chemistry

technicians to operate the dosimetry function.

Radiation/Chemistry Technician

C and the Radiation Protection Supervisor stated that no one individual had
been assigned specific responsibility for dosimetry (Int. 82).

Each radiation/

chemistry technician performed this function as a routinely assigned task.

Aradiation/chemistry technician stated that one might be expected to read TLD's
twice a year.

Radiation/chemistry technicians stated that extremity monitoring

was available by taping a TLD badge to the extremity.
The operation and management of an acceptable dosimetry system requires
specialized training and consistent attention to detail.

This concept wasaddressed in a review titled "Evaluation of the Health Physics Chemistry
Organization at Three Mile Island Nuclear Station, Units 1 and 2," performed
in September 1977 and again in a review performed by a consultant in March 1979
(Ref.

56).

Both reviews suggested the need to assign one individual the sole

responsibility for the dosimetry program.
had not been made prior to March 28,

Implementation of this suggestion

1979.

No regulatory requirements or license commitments eststablish minimum
standards for management of personnel dosimetry systems.
1.5.4

Self-Reading Dosimeters
Section 12.3.3.1.1 of the FSAR states, "Self-reading dosimeters are issued

in addition to a film badge/TLD to individuals as required.

Each individual

is instructed in the necessity of reading his self reading dosimeter at frequentintervals while in radiation areas."
An investigator examined forms 1772-1 maintained pursuant to Health Physics
Procedure 1772,
November 22,"Dosimeter Calibration and Leak Test," Revision 3, dated

1977 to determine the availability of self-reading pocket

11-1-38

dosimeters on March 28.

This review showed that a sufficient number of 0-200mR dosimeters were available for use.

Fifty 0-5 R and fifteen 0-20 R dosimeters

were tested and available from the Unit 1 health physics office.
The lack of an adequate supply of 0-5 R dosimeters directly contributed
to one individual being exposed to whole-body gamma radiation in excess of aregulatory limit.
1.5.5.

This is further discussed in Section 3.2.

Internal Dosimetry

Section 12.3.3.3 of the FSAR states, "whole-body counts will be taken on
a special basis if

the nature of exposure or suspected exposure indicates aneed for a whole-body count. This procedure acts as a confirming check on theadequacy of the station contamination control practices.

The TMI Unit 2 Radiation

Protection Procedures require that all contractor personnel provide a urine
sample, which is used as a baseline sample.
tractor people on completion of the job.

Urinalyses are also run on con-

The Radiation Protection Supervisor

has the option of ordering additional bioassay analysis and/or whole-body counts
on any personnel should the need arise."
Procedure 1628,"Program for Medical and Bioassay Examination,"

3, dated September 30,

1977,

Revision

implements the internal dosimetry program.

whole-body counting and urinalyses were performed by offsite vendors.
body counter was on site on March 28.

The
A whole-This program meets present regulatory

requirements.
1.5.6.

Respiratory Protective Devices

Section 12.3.2.5 of the FSAR states, "Respiratory protective devices may
be required in any situation in which airborne radioactive contamination exists
or is expected.

In such cases,

the airborne concentrations are monitored forconcentration and type of airborne contaminants present.
tective devices are specified accordingly.
use include:

11-1-39

The necessary pro-Respiratory devices available fora.

Half-face respirator (Filter Type)b.

Full-face respirator (Filter Type)c.

Full-face respirator (Supplied Air)d.

Self-contained breathing apparatus.

A self-contained breathing apparatus will also be used in any situation
involving oxygen deficient atmospheres.
The respiratory protective program is designed in accordance with thepractices recommended by the American National Standards Institute (ANSI
Z88.2-1969)."
An investigator examined Forms 1616-1 maintained pursuant to Health Physics
Procedure 1616,"UseApril 21,

to establish the availability of these devices.

1979,

1979,

of Respiratory Protection Devices," Revision 13,

there were 50 self-contained breathing devices on site.

were of the emergency egress type.
bottles were available.
18,

124,

dated

As of February
Six of theseIn addition, about 15 backup air supply

Based on interviews with licensee representatives (Int.

133) about 150 full-face respirators with particulate cartridges andabout 25 half-face respirators were available.

No iodine absorbing cartridges

were available on site, since they had not been approved for protection factors.
There was a breathing air compressor located in the west end of Unit 1
intermediate building.
bottles.

This equipment had been used to refill air supply

During the accident this compressor was not used since the qualityof air being compressed could not be verified as acceptable because of highairborne radioactivity in the compressor suction area.
Based on interviews with Auxiliary Operator C (Int.
Chemistry Technicians E and H (Int. 49, 50),104) and Radiation/

it was determined that an

adequate supply of readily accessible self-contained breathing devices was notavailable during the first 2 days of the accident.

11-1-40

1.5.7.

Protective Clothing And Supplies

Section 12.3.2.4 of the FSAR states "All personnel entering a contaminatedarea are required to wear specified protective clothing.

The nature of thework to be done and the level of contamination are the governing factors inthe selection of protective clothing to be worn.
able includes shoe covers,

head covers,

Additional items of specialized apparel,shields and respirators,

The protective apparel avail-gloves,and coveralls or lab coats.

e.g.,plastic or rubber suits, faceare available for operations as required.

In all cases,

radiation protection personnel evaluate the radiological conditions and specify
the required items of protective clothing to be worn."
Based on discussions with licensee representatives and vendors (Int.

155),

it appears that an adequate supply of protective clothing was available on site,
including:paper and cloth coveralls; hoods; cotton, plastic, rubber, and neo-prene gloves; plastic shoe covers,
beta glasses and surgeon caps.
of step-off-pad material,

rubbers and rubber boots; plastic suits,

In addition,itappears there was an ample supply

radiation warning signs,

labels, boundary rope, masking

tape, and plastic bags.
Paragraph 4.2.1 of the Three Mile Island Site Emergency Plan states,"Appropriate on-site assessment surveys performed by the Radiation Monitoring
Teams include general in-plant and site 0-y surveys,
gas sampling, assisting repair and rescue teams,
control points,

airborne particulate andpersonnel monitoring at accessand manning the North or South Gate monitoring points or assemblyareas."
From the preceeding statement,in conjunction with a review of theemergency equipment set aside for emergency use (see Section 1.6),

itisapparent that the "routine" complement of radiation protection instrumentation
is intended to be the primary source of instrumentation to support in-plantradiological controls and measurements during emergencies.
The licensee's reliance on the routine complement of health physics
instrumentation and respiratory protection equipment to support in-plant P-y

11-1-41

surveys,and gas sampling and for assisting repair teams

airborne particulate,degraded the response to in-plant radiological conditions resulting from theemergency.

Significantly greater numbers and types of instruments and equipment

were needed to support the in-plant efforts than were actually available.
Teletectors,

high-range survey meters,

and air packs were all in short supply.

The loss of the counting room capability (GeLi)
efforts to establish radiologicalfurther compounded the in-plant

controls and necessitated the set up of sub-stitute methods having less reliability.
An interview of the Supervisor,

Radiation Protection and Chemistry sub-stantiated the observed shortcomings in the area of equipment (Int. 25).

Planners

had not conceived of the type and extent of the maintenance and operations
activities that would have to be performed in areas of such high radiationdose levels as were encountered.
1.6

EMERGENCY EQUIPMENT
Paragraph 4.2.2 of the Emergency Plan and Procedure 1778,"Emergency

Readiness Checklist," establish an inventory of emergency supplies andequipment to be maintained at specific locations throughout the facility (Ref.
57).

This equipment consisted of radiation detection and measurement devices,protective clothing and supplies, and procedures.

The emergency kit

designations and their locations were:Emergency Kit

NumberLocation

Emergency Monitoring

4

North Search Facility(Processing Center)

Washdown Area

2

North Search Facility(Processing Center)

Ambulance

1

Machine Shop

Emergency Clothing

1

Unit 1 Storage Area outside
Access Control Point

Control Room2

One in Each Control Room

Medical Emergency
Cabinet1

Control Building

11-1-42Paragraph 4.2.2 of the Emergency Plan states, "Emergency Monitoring kits
containall instrumentation, protective clothing and supplies needed for on/
off-site field radiation monitoring."

Procedure 1778 lists the following total

inventory of radiation detection or measurement instruments,tion devices,

respiratory protec-

and dosimetry to be reserved for emergency use in the kits:

Item

NumberLocation

TLDs

20

5 in *each of 4 monitoringkits

Air Samplers(SAM-2)4

1 in each of 4 monitoringkits

Stabilized Detector (RD-19)

4

1 in each monitoringkit

High-Range Dosimeters

10

5 in each control roomkit

Dosimeter Chargers

2

1 in each control roomkit

High-Range Survey Meters2

1 in each control roomkit

Half-Face Respirators(particulate cartridge)

20

10 in each control
room kit

RM-14

1

Health Physics Lab

GM Survey Meter

1

Health Physics Lab

This procedure requires that the emergency equipment kits be inventoried
and the instruments operationally checked at least once each calendar quarter.
To determine the status of the licensee's emergency kits on the morningof March 28,

an investigator reviewed records of an emergency equipment inventory

and operational check performed on March 11, 1979,

and interviewed radiation/

chemistry technicians who were the initial users of the equipment on the morningof March 28 (Ref.

23,Int. 82).

The review of the inventory and operational

check records showed that one of the four SAM-2s in the monitoring kits at theNorth Search Facility (Processing Center) was out-of-service.

11-1-43

Radiation/ChemistryTechnician C stated that the SAM-2 was still out-of-service on the morning ofMarch 28 and not located in the north search facility.
Radiation/Chemistry
Technican C further stated that during a preoperational checks of the instruments by one of the survey team members on the morning of March 28, a second
SAM-2 was found to be inoperable (Int. 82).

Consequently, only two of four

field counting systems were operational and available during the first several
hours of the licensee's emergency response.
Paragraph 5.5.4 of the Emergency Plan states that "at least two company
vehicles can be quickly equipped...for off-site monitoring." On the morningof March 28, Radiation/Chemistry Technician C, assigned to an emergency survey
team, had to "commandeer"

a company vehicle to use for transportation toGoldsboro to perform offsite surveys,

and it was necessary for Radiation/

Chemistry Technician N to use his personal vehicle.
1.7
1.7.1

(Int. 47, 82).

ENVIRONMENTAL MONITORING PROGRAM
Sampling Equipment
1.7.1.1

Air Samplers

An investigator observed the operation of an environmental air sampler
and discussed sampler operation with the Supervisor, Radiation Protection andChemistry and Radiation Protection Consultant A.

The samplers consist of asample head, a pump, and a dry test meter, a rotameter, and a timer (forlocations see Section 1.7.2).
An investigator discussed the calibration checks of environmental air
samplers with licensee personnel and examined surveillance records to determine
the status of the samplers on March 28.
IE Report Nos.

This item was originally discussed in

50-289/78-08 and 50-320/78-16.

Unit 2 Environmental Technical

Specification 5.5 requires that written instrument calibration procedures bedeveloped and followed.

Unit 2 Environmental Technical Specification 3.2.A

outlines the detailed sampling and analysis program.
"Environmental Monitor

Unit 1 Procedure 1302-5.24,

Calibration," Revision 3 dated December 19,

annual air sampler calibration checks (Ref.

11-1-4458).

1974 requires

Review of the surveillance records and discussions with Maintenance Foreman
G, the Radiation Protection Supervisor,
and Chemistry,

and the Supervisor,

Radiation Protection

showed no Unit 2 procedure equivalent to Procedure 1302-5.24.

In discussions, Technical Analyst A stated that the Surveillance Task Index
showed that the Unit 1 calibration of the eight environmental air samplers had
not been scheduled to be done since 1974 (Ref.

59).

In discussions, Technical Analyst B stated that he thought that Procedure1302-5.24 had been withdrawn by PORC several years ago because it
technical specification requirements.

did not contain

The central procedure files contained

no record that the procedure had been withdrawn.

An investigator noted thatthe procedure was in the active operating procedure file in the Unit 1 control
room on June 15,
1979 (Ref.

1979 and was listed as active on the procedure list dated July 6,

60).

To determine if

the samplers were operating on March 28, an investigator

examined records of sample collections just prior to and during the period
March 28-31 (Ref.

61,

62,

March 1 through April 15,63,64) and the Work Request Log for the period

1979.

The records completed by the individuals

collecting the samples showed on March 22 the Falmouth substation air sampler
was blowing air out of the filter rather than pulling air through the filter
and, on.March 29 the Observation Center air sampler flow meter was erratic.
1.7.1.2

Water Samplers

An investigator discussed the operation of environmental water samplers
with the Supervisor,

Radiation Protection and Chemistry,

Consultant A, and Radiation Protection Consultant E.
matic liquid samplers,

The licensee had two auto-

one at the Columbia water works,

drinking water, and the other at York Haven Dam.
drew 10 ml of water 10 times an hour.

Radiation Protection
sampling downstream

When operating,

In discussions,these samplers

the Supervisor, Radiation

Protection andChemistry stated that individuals collecting the samples wereto check-the sample-volume each time a sample was collected as a check of theproper operation of the sampler.

The other samples routinely taken were grab

samples or grab composite samples.

11-1-45

To determine if

the samplers were operating on March 28 to 30,

aninvestigator examined records (Ref.

61,

September 27,

1979 and the Unit 1 Work Request Log (the1978 through March 30,62) of sample collections for the period

samplers were under Unit 1 maintenance) for the same period.

Investigators

discussed the samplers with Consultant E who collected the samples on March 29
and 31,

1979.

operational,

The records showed that the sampler at the York Haven dam was
but that grab samples,

rather than a composite sample,

taken at Columbia since January 31,

1979.

had been

Prior to the accident, weekly grab

samples had been taken at Columbia since the composite sampler was not operating.
Radiation Protection Consultant E stated that the composite sampler at Columbia
was not operating on March 29 or 31.
the Supervisor,

Radiation Protection Consultant A andRadiation Protection and Chemistry stated that there had been

operational difficulties with the sampler in the past and that it had been
replaced after the accident.
1.7.1.3

Environmental Thermoluminescent Dosimetry

The primary dosimeter used by the licensee at 20 locations (15and 5 controls)

indicators

is the Teledyne Isotopes environmental dosimeter (Int.

30, 74).

A representative of Teledyne stated in discussions that this dosimeter contains
CaSO 4 :Dy elements with a minimum sensitivity of 0.5 mR.
sandwiched between two 500 mg/cm
of the dosimeter.
plastic holder.

These elements arecopper sheets to flatten the energy response

The TLD elements and shields are contained ina rectangular
After a readout,

annealing at 250'C for one hour, an irradiation,

and a re-annealing step, the dosimeters are sent from the Teledyne facility at
Westwood,

NJ to Three Mile Island.

In-transit controls are sent with the dosimeters.

After arrival at the site, the dosimeters are reannealed at 250 0 C for 1-2 hours
prior to placing them in the field.
Ten of the twenty locations have RMC TLDs in addition to the Teledyne TLDs
as a quality control check (Int.
UD-200S dosimeters,

30, 160).

Each location has two RMC Modeleach containing two CaSO 4 :Tm phosphorus inside of a plastic

and metal shield (Ref.

65).

11-1-46

Functional dosimeters were in place at all twenty locations at the timeof the March 28 accident.

Dosimeter locations are shown in Table 11-1-4.

In

addition to the seven TLDs located on Three Mile Island as shown in Table 11-1-4,
the licensee also had distributed, as part of an informal program,

21 LiF TLDs

of the type used in the personnel dosimetry program (see Section 1.4.3) at various
locations immediately outside of the plant, most of which were on or near theplant security fence (Ref.

66).

Radiological Environmental Monitoring Program

1.7.2

The radiological environmental monitoring program, which includes the abovementioned sampling and monitoring equipment (air and water samplers; TLDs),
addresses the following environmental media:
Number of Locations
Required byTech Specs

MediumNumber of Locations
Routinely Monitored (Ref.

Air iodines

4 ind.*; 2 bkg.

4 ind.; 1 bkg.

Air particulates

5 ind.; 3 bkg.

5 ind.; 3 bkg.

Immersion dose
(TLD)

15 ind.; 5 bkg.

15 ind.; 5 bkg.

Milk

5 ind; 1 bkg.

5 ind.; 1 bkg.

Surface & drinking
water

5 ind.; 2 b kg.

5 ind.; 2 bkg.

180)

Other*"-Ind.
Bkg.

- indicator-(near.-field locations)- background (control or far-field) locations*"The program also includes samples of fish, aquatic plants, sediment, green
leafy vegetables, and precipitation. These sample types were not listed
above because they were either unimportant as far as the March 28 incident
was concerned or represented only an indirect pathway to man.

11-1-47

TABLE 11-1-4
THERMOLUMINESCENT DOSIMETER LOCATIONS
Licensee Designation

Location

Distances and Direction*

IS2**

North Weather Station

0.4 mi N

2S2

North Bridge

0.7 mi NNE

4S2**

Top of Dike

0.3 mi ENE

5S2**

Top of Dike

0.2 mi E

9S2South TMI

0.4 mi S

1IS1**

Mech.

0.1 mi SW

14S2

Shelley Island***

0.4 mi WNW

16SI**

North Boat Dock

0.2 mi NNW

4AI

Laurel Road

0.5 mi ENE

5AI**

Ob.

0.4 mi E

16AI

Kohr Island***

0.4 mi NNW

1OB1

Shelley Island***

1.1 mi SSW

12BI

Goldsboro Air Station

1.6 mi WSW

lCl

Middletown Substation

2.6 mi N

8CI**

Fallmouth Substation

2.3 mi SSE

7FI**

Drager Farm****9 mi SE

4CI**

Rt. 241****10 mi ENE

7GI

Columbia Water Plant

15 mi SE

9GI

York Med Ed Station

13 mi S

15GI**

West Fairview Substation*"*15 mi NW

Draft Cooling Tower

Center Bldg.*****Relative to a point midway between the two containment buildings.
**Location also has RMC TLD for quality control purposes.
***Island locations contained two Teledyne TLDs on 3/28/79.
****Location also has a dosimeter which is readout on a monthly basis.

11-1-48

Analyses of the above sample types are performed by Teledyne Isotopes ona frequency which varies from weekly (air iodine and particulates; water) tosemi-monthly (milk during the grazing season; monthly otherwise) to quarterly(TLD).

A representative of Porter-Gertz Consultants stated in discussions that,in addition to the routine analyses performed by Teledyne Isotopes,
samples (4 air, 10 TLD,

additional

1 milk, and 3 water) from the same locations have been

analyzed by the Radiation Management Corporation as part of a quality assurance
program.
Based on a selective review of this area by the investigator, itappearsthat the above-described program was in effect at the time of the March 28
incident.
1.7.3

Meteorological Monitoring Program
The Three Mile Island meteorological tower is located 2200 feet north of

the Unit 2 reactor building.

Monitoring, instrumentation on the tower consistsof redundant Teledyne/Geotech wind speed and direction indicators at the 100 ft.
level,

Rosemount temperature sensors at the 33 and 150 ft. levels.

difference is measured between the 33 and the 150 ft. level (Ref.

A temperature
67,

Int. 60).

A calibration of the instrumentation on the tower was performed on October 26,
1978, in accordance with Technical Specification 4.3.3.4.

The calibration consistedof the following (Int. 60, Ref. 68):a.

A visual inspection of the wind vane and a test of its azimuth
indication accuracy.

The Crawford Plant west stack (a coal-fired

plant upstream of TMI),
point of reference.

having an azimuth of 3360, was used as a

Alignment was checked at 00, 900,

2700.

11-1-49

180',

and

The wind vane bearings were also checked to ensure that the starting
torque was acceptable.

One sensor was noted to have a torque of 2.0

gm-cm relative to a maximum acceptable value of 1.4.

It was adjusted

to 1.4 gm-cm.
b.

Anemometer cups previously wind tunnel tested and certified (Int.
60), were given a visual inspection fordamage and none was apparently
noted.

A known torque was applied to the anemometer to ensure thatthe starting torque of the bearings corresponded to a wind speed ofless than 1 mph.

One of the anemometers was noted to have a starting

torque of 1.5 gm-cm,
gm-cm.
c.

relative to a maximum acceptable value of 0.74

It was adjusted to 0.14 gm-cm.

Temperature sensors were placed in a 32.0°F ice bath and the response
noted.

Other temperatures,from -40°F to 100 0 F, were electronicallysimulated relative to the ice bath response.
d.

No problems were noted.

The remainder of the system, up to and including the control room
recorder, was electronically calibrated.

No significant problems

were noted.
Prior to the event,the meteorological tower had been operating reliably,providing better than 98% data recovery (Int.

60).

The tower has continued tooperate equally well post-event with no significant problems noted (Int. 60).

11-1-50

2.

2.1

INITIAL EMERGENCY RESPONSE

DETECTION AND CLASSIFICATIONThe Emergency Director was responsible for classifying a situation as an

emergency in accordance with the conditions in Table 1 of the Emergency Plan
and for initiating actions according to the Emergency Plan Implementing Procedures and his own best judgment.

The Emergency Plan requires that the actions

listed in Table 1 be considered for each type of emergency,
other actions only be taken if

but that these or

they are appropriate.

The first conditions in Table 1 that appeared to have been met were thosespecified in Criterion (c)for a Site Emergency.

0415 on the morning of March 28.

These conditions were met byThis Site Emergency action level states:"Loss of primary coolant pressure coincident with high reactor building pressure
and/or high reactor building sump level."
By 0415,

reactor coolant system pressure had dropped-from 2435 psig at

the time of reactor trip (0400)

to approximately 1275 psig (Figure 11-2-1).

This pressure (1275 psig) was below the reactor coolant low pressure trip
setpoint (1940 psig) and the setpoint for emergency core cooling system initiation (1600 psig).

At 0415, a pressure rise of about 1.4 psig inside thereactor building was detected.

The duty shift supervisor was aware of thedrop in reactor coolant system pressure and increase in the reactor building
pressure.

Initially, he evaluated these conditions in relation to the Emergency

Plan and determined that they were not indicative of an emergency since theprimary coolant system pressure had stabilized and there were no increased
radiation levels either in or being released from the facility (Int.

145).

Information regarding the reactor building sump level is provided to thecontrol room via printout on the alarm printer or by phone or page from anindividual at the radwaste panel in.the auxiliary building.

Reactimeter data

for March 28 indicated that a high reactor building sump level was reached at

11-2-1

REACTOR COOLANT PRESSURE V S.
TIME ON MARCH 28, 1979,
AS DISPLAYED ON THE WIDERANGE PRESSURE RECORDER
2500 --2000

u.l

D 1500
C/)

I0)

1000

I

I

I

I

I

I

0340 0400

0430

0500

0530

0600

0630

500

TIME
Figure 11-2-1

0411.

At this time, the alarm printer was printing numerous alarms and plantparameters.

The high reactor building sump alarm printout went unnoticed

until approximately 0430 when an auxiliary operator,building near the radwaste panel,

located in the auxiliary

noted that both reactor building sump pumps

were running and that there was a high reactor building sump alarm on the radwaste panel.

Upon leaving the auxiliary building, the individual also notedan alarm (500 cpm) of a count rate meter (RM-14/HP-210)
room door.

located near the model

These conditions were reported to an operator in the control room.

The operator directed the auxiliary operator to survey the area.

This survey

was performed using an E-520 instrument and indicated radiation levels of lessthan 0.1 mR/hr.

Based on the survey results and the location (normally aradiation area),

the significance of the RM-14/HP-210 alarm was discounted.

The operator did verify, via the alarm printer in the control room, that thereactor building sump level was "high."

This was reported to the shift super-visor who became concerned about available capacity in the liquid radwaste
system and directed that the sump pumps be turned off (Int.

15).

Since the drop in primary system pressure had been stabilized by thistime and there were still no alarms on the radiation monitors in the control
room, the shift supervisor did not interpret the high reactor building sumplevel and the earlier noted conditions of decreased primary system pressure
and increased reactor building pressure to meet the conditions for a Site
Emergency (Int.

145).

The investigator determined that the lack of specifity

regarding the definition of the terms "loss of primary system pressure" and"high reactor building pressure" appeared to contribute to the failure to
declare the Site Emergency by 0430.
At 0518, the reactor building air particulate monitor (HP-R-227(P))

passedthrough the alarm setpoint (50,000 cpm)(Ref.

102).

recall having received this alarm (Int.

Two intermediate cooling water let-145).

down cooler radiation monitors (IC-R-1091

The shift supervisor did not

and IC-R-1092) alarmed at 0500 and 0518,

respectively. The detectors for these monitors were located near the reactor building
sump and were intended to monitor for leakage from .the primary to secondary side ofthe letdown cooler.
were 5,000 cpm.

The alert setpoints were 1,000 cpm and the alarm setpoints

The alarm setpoints were designed to be 10 times background11-2-3

to provide positive identification of leakage of the letdown cooler (Ref.

3).

Upon receipt of the intermediate cooling water letdown cooler radiation monitor
alarms, operators verified that the intermediate closed cooling water surge
tank level had not increased,

indicating that there was no leakage from theprimary to the secondary side of the letdown cooler.

This is a specified

action for alarm of these monitors.
Since the intermediate cooling water letdown cooler monitor and reactor
building air sample particulate monitor alarms occurred in the same building,itappears that criterion (b)was met.

Condition (b)of Table 1 of the Emergency Plan, "Local

Emergency,"

states, "More than one radiation monitor in a single

building reaches the alarm setpoint."
The shift supervisor evaluated these alarms (intermediate letdown coolingwater letdown cooler),but did not interpret them to be applicable for serving

as a basis for the classification and declaration of a Local Emergency (Int.

145).

Based on their training and the wording of the action levels it was the shift
supervisor's and operating crew's understanding that the action levels in Table 1of the Emergency Plan relative to "radiation monitors" referred only to area radiation monitors.

Since the intermediate cooling water letdown cooler monitors andreactor building air monitor were process monitors, their alarms were not thought
to meet specified actionlevels.

In addition, prior to this incident,

the inter-mediate cooling water letdown cooler radiation monitors were often in alert due
to background radiation.

The significance of their alarm on March 28 was further

discounted by the shift supervisor,

based on his assumption that the alarms re-

sulted from a "crud burst" through the letdown cooler.
Procedure 1670.1,(b)

An investigator noted that

Criterion 3.1 contained the same wording as Table 1, Criterionof the Emergency Plan and did not specify whether the criterion pertained toarea monitors,

process monitors, or both.

The investigator did note,however, that relative to a Site Emergency,

Criterion 3.6 of Procedure 1670.2,
monitors, whereas condition (f)"Site Emergency Procedure," specifies areaof Table 1 in the Emergency Plan,

11-2-4

relative toa Site Emergency,

does not.

An interview of the operating crew indicated that

their training had not distinguished or made clear the applicability of areaor process radiation monitors in relation to the action levels of the EmergencyPlan.
The investigators have determined that the alarms of the intermediatecooling water letdown cooler radiation monitors were a result of increased
radiation levels in the area of the detectors,i.e.,

the reactor buildingsump.
Review of strip chart records HP-UR-1901 indicates that alarm of theincore instrument panel area monitor (HP-R-213)
0625.

If the alarm occurred,itshould have occurred at about

apparently went unnoticed.

Review of strip chart records HP-UR-1901 indicates that alert of thecontainment dome monitor (HP-R-214)

should have occurred at about 0635. Alertof this monitor and its correlation with Site Emergency criterion (e)went unnoticed.

Criterion (e)apparentlyof Table 1 states, "Reactor building high range

gamma monitor alert alarm."
At approximately 0655, there were many indications of a sudden increase
in in-plant radiation levels.

Radiation/chemistry technicians located in thearea of the Unit 1 primary sample station heard local alarms of area radiation
monitors and noted increased readings on portable radiation survey instruments,
while Unit 2 control room personnel noted remote alarms of various area radiation monitors on the area radiation monitor panel almost simultaneously.
Radiation/chemistry technicians determined the source of these increased
levels to be the reactor coolant letdown sample lines (Int.

20, 42,145).

A Site Emergency was declared at 0655 on March 28 by the shift supervisor
in coordination with the Unit 2 Superintendent, Technical Support based on thealarm of numerous area radiation monitors in the Unit 2 control room (Int.

145).

This action level was consistent with the Emergency Plan, Table 1, criterion (f),11-2-5

for a Site Emergency and with paragraph 3.6 of Station Radiation Emergency
Procedure 1670.2,
1978.

"Site Emergency Procedure",

Revision 8,

dated February 13,

The 0655 time is based on the stated sequence that restart of the 2B

reactor coolant pump (RCP)

preceeded the sudden increase in radiation levels.

Alarm printer data shows that the 2B RCP was restarted at 0654.
For the next 30 minitues,
monitors,

radiation levels on many of the area radiation

including the containment dome monitor (HP-R-214),

Upon declaration of the Site Emergency,
radiation exposure rate was performed.

steadily increased.

a calculation of projected offsite

This calculation was completed by 0710

and indicated a whole body exposure rate of 40 R/hr in Goldsboro (Ref.
Int. 48).

Apparently,

70,

no attempt was made to calculate a projected dose byfactoring in the anticipated duration of the condition that precipitated thenumber.

The initial projection of offsite exposure rate was believed to beoverly conservative since pressure in the reactor building was only 1.4 psig,
far below the 56 psig pressure on which the calculational model was based.
At the time, there were no other data available to use in evaluating the validityof this calculation.

Consequently,

it would appear that the initial projected

exposure rate should not have been discounted so readily.

As a minimum itshouldhave been evaluated in light of the general emergency criterion (d) which states,"Offsite projected doses downwind from the site boundary are greater than 25 Rem
thyroid and/or 5 Rem whole body."(See Section 3.3.1 for a detailed discussion

concerning this calculation and evaluation.)
At 0724,

the Station Manager declared a General Emergency based upon acontainment dome monitor reading of greater than 8 R/hr.

This action was

found to be consistent with paragraph 3.1 of Station Radiation Emergency Procedure 1670.3,"General Emergency Procedure",

Revision 6, dated February 13,

1978.
The emergency classifications,

respective criteria (action levels) andthe times when the criteria were apparently met are shown in Table 11-2-1.

11-2-6

TABLE 11-2-1
EMERGENCY CONDITIONS AND TIMES MET ON MARCH 28
Emergency
Classification
Personnel
Emergencyor
Local
Emergency

Site
Emergency

GeneralEmergency

Conditionsa) Personnel injuries may involve
contamination or excessive radiation exposures.
b) More than 1 radiation monitor
in a single building reaches
their alarm setpoint.

Time Met*

0520
0625

Notes

IC-R-1091
IC-R-I092
HP-R-227(P)c) Report of an unexpected increasein the level of radiation or airborne activity in a work area.
d) Report of a radioactive spill ina work area.
e) Flooding or localized fire,that could affect a release ofradioactivity.

0655

Radiation/
Chemistry
Techniciansa) A unit vent gas monitor indidicated 100 times the instantaneous release limit specified
in the technical specifications.
b) The radiation level at the station
security fence is 125 mR/hr.
c) Loss of primary coolant pressure,coincident with high reactor
building pressure and/or highreactor building sump level.
d) Reactor building evacuation alarmfrom the source range instrumentation (Unit 1), manual initiation(Unit 2).
e) Reactor building highrange gamma monitor alert alarm.
f) The high alarm of the radiation
monitoring system in two separatebuildings due to a single event.

0730-0740

HP-R-219(P, I, and G)

1720

140 mR/hr at
GE-10
RC pressure1275 psig
RB pressure
1.4 psig
Hi RB sump

A General Emergency will be declared
when a Site Emergency has been declaredand one or more of the following conditions exists:a) Reactor building high range gamma
monitor high alarm.
b) The radiation level at the station
boundary is >125 mR/hr.
c) The liquid effluent radiation
monitor indicates.
>6.8 E-3
pCi/cc.
d) Offsite projected doses
downwind from the site boundaryare >25 rem thyroid and/or 5 rem
whole body.

0430

0635

Unnoticed

0655

Site Emergency
Declared

0720

General Emergency declared
140 mR/hr,
GE-10

1720

0710

10-40 R/hr(dose rate)*Times based on an evaluation of information obtained from interviews, strip chart
recorders, logs and reactimeter. Times reported were selected based on the relative
reliability of the various sources and reconciliation with other activities or eventsoccurring at the same approximate times.

2.2

ORGANIZATION ACTIVATION
Shortly after the reactor tripped, workers on site initially phoned
chemistry, and radiation protection supervisors at homeselected operations,

to notify them of the trip and request their assistance (Int. 20, 27,
145).

During the following 2 hours,

operational problems,120,

as onsite personnel sought solutions toother supervisors were alerted at home.

Consequently,

at 0655, when several area radiation monitors showed sharp increases, many key
TMI employees were either on site or enroute.

In addition, the remainder ofthe licensee's day shift employees were arriving for the beginning of theirshift.
On declaration of the Site Emergency at 0655, TMI's response organization
pursued actions along five parallel paths:
operating conditions,(2)(1) continued assessment of plant

assessmentof actual and projected radiologicalconditions in the plant and in the environment,(3)

notification of and commu-

nication with offsite support agencies and individuals,
TMI personnel,

(5)

(4)

accountibility ofand consideration of corrective actions to mitigate or

terminate the emergency.
Initially, the emergency response organization approximated the "planned"
organization established in the Emergency Plan and procedures which is shownin Figure 11-2-2.

Slight deviations from the "planned" organization occurred,

and there was some confusion regarding who was performing and in charge ofdirecting certain emergency actions.

Initially, the Supervisor,

Radiation

Protection and Chemistry assigned Radiation/Chemistry Technician I to assumecommand of the Emergency Control Station until relieved by a radiation protection foreman or the Radiation Protection Supervisor (Int. 79).

The emergency

responsibilities of the individual in charge of the ECS are specified in
Paragraph 4.2.7 of Procedure 1670.2 (Ref.

71).

At about 0700, Radiation Protection Foreman B arrived at the ECS andRadiation/Chemistry Technician I assumed that Radiation Protection Foreman B
had taken control of the ECS.

Radiation/Chemistry Technician I then began tofunction as a Radiation Monitoring Team member (Int. 42).

11-2-8

According to Radiation

"NORMAL" EMERGENCY ORGANIZATION

I ECC

ECS
ECS DIRECTOR'ACCOUNTABILITY'

Radiation Protection Supervisor

Site Protection
Sargent/Site Protection Officers

ECC - EMERGENCY CONTROL CENTER, UNIT 2 CONTROL ROOMECS-EMERGENCY CONTROL STATION. UNIT 1 CHEMISTRY/HEALTH PHYSICS LAB AREA
-FUNCTIONAL TITLE ADDED FOR CLARITY, LICENSEE'S PLAN LISTS NORMAL DUTY TITLES

Figure 11-2-2

NORTH
AUDITORIUM

Protection Foreman B, during the time frame of about 0700 to 0800, he primarily
performed in-plant (Radiation Monitoring Team) radiation surveys (Int.
Atapproximately 0705,

21).

Radiation Protection Foreman D arrived at the ECS

shortly after Radiation Protection Foreman B and, for a short time, assisted
Radiation Protection Foreman B in getting onsite and offsite survey teams
underway to the processing center.
1 control room (Int.

He then left the ECS and went to the Unit

18).

Radiation Protection Foreman A stated that he arrived in the ECS at about
0715 and, observing that no one appeared to be in charge,

(nt.

7).

he took charge.

At about 0735, the Radiation Protection Supervisor arrived andassumed control of the ECS (Int. 22).
According to the Emergency Plan and Emergency Plan Implementing Procedures,a single Repair Party would form at the Emergency Control Station (ECS),
functioning under the Supervisor of Maintenance or a maintenance supervisory
foreman who would also be located at the ECS.

The general activities of theRepair Party would be coordinated by the Emergency Director,
Emergency Control Center (ECC),

located in thethrough the Radiation Protection Supervisor or

his alternate, who would be in charge of the ECS.

This method of coordinationand control was designed to ensure that radiological conditions and controls
were noted and observed prior to, during, and after each emergency repair job.
Italso established a single point of command and control of Repair Party

activities, avoiding confusing,counterproductive,

duplicative, or "unauthorized"

efforts or requests.
Procedures governing the assembly,

direction, and functions of the Repair

Party Team are covered in Paragraph 4.4 of Procedure 1670.2 and Procedure
1670.8.

11-2-10

After the Site Emergency was declared at 0655, a Repair Party composed ofsix maintenance shift workers was formed at the ECS under the control of
Maintenance Foreman B (Int.

187).

A second Repair Party,of daylight instrument and control personnel,

composed primarily

was formed in the Unit 2 control

room under the direction of the Superintendent of Maintenance and the Unit 2
Supervisor of Maintenance.

The Superintendent of Maintenance and Unit 2

Supervisor of Maintenance were aware that the assigned location for either ofthem during an emergency was the ECS where they would act as the Repair Party
leader.

However,to ensure prompt availability of their expertise, and sincea Repair Party was already formed at the ECS under the direction of a maintenanceforeman,they decided to remain in the control room (Int.

was reinforced by the Station Manager.

120).

This decision

On assuming the position of Emergency

Director, he announced that the Superintendent of Maintenance would be the one
in the control room to be in charge of emergency repair functions (Ref.
120).

72,

Int.

Some time later, the Superintendent of Maintenance directed the Repair

Party at the ECS to move to the Unit 2 control room.

By 0800, all Repair

Party personnel were assembled in the Unit 2 control room,

separated from theECS.
The emergency duties of the chemistry supervisor are outlined in
Paragraph 4.2.8 of Procedure 1670.2.

Since the TMI position of chemistrysupervisor was vacant, Chemistry Foremen,

A and C, filled this position in theemergency organization (Int. 79).
The licensee operated generally with the organizational configuration
shown in Figure 11-2-3 from approximately 0730 until 0900 when the ECS wasevacuated to the Unit 2 control room.

Since the Unit 2 shift supervisor's

office, the planned alternate location of the ECS,
people in it,

already had a number ofthe Radiation Protection Supervisor established the ECS in thecontrol room proper.

He was careful to establish the location of the ECS

activities in an area where they would not interfere withplant-operations
(Int.

172).

The configuration of the emergency organization upon relocationof the Unit 2 control room is shown in Figure 11-2-4.

11-2-11

EMERGENCY ORGANIZATION IN EFFECT 3/28/79,0730 - 0900

Processing
Center andAssembly Areas

ASSEMBLY
AREA
MONITORS
Radiation/
Chemistry
Technicians

ECC--Emergency Control Center, Unit 2 Control RoomECS-Emergency Control Station, Unit 1 Chemistry/Health Physics Lab Area*- Functional Title Added

Figure 11-2-3

GATE
MONITORS
Chemistry
ForemenECSMONITOR
Radiation/
Chemistry
Technicians

EMERGENCY ORGANIZATION IN EFFECT 3/28/79,0900-1100

MONITORINGTEAM
Radiation/
Chemistry
Technicians

MONITORINGTEAMS
Radiation/
Chemistry
Technicians

I

AREA
MONITORE
Radiation/
Chemistry
Technician!

ECC-Emergency Control Center - Unit 2 Control Room
AECS-Alternate Emergency Control Station - Unit 2 Control Room*-Functional Title Added

Figure 11-2-4

Shortly after relocation of the ECS to the Unit 2 control room, theRadiation Protection Supervisor and the Supervisor,

Radiation Protection andChemistry met to discuss the direction in which the organization should proceed(Int.

79, 172).

The Supervisor, Radiation Protection and Chemistry determinedthat the Radiation Protection Supervisor had onsite and offsite monitoring
activities fully implemented and under control.

To better divide the workload,

the Supervisor, Radiation Protection and Chemistry decided to assume directcontrol of all in-plant radiation protection functions.

He determined that

this was necessary because the Emergency Plan and procedures did not containadequate provisions to deal with radiological controls under the type ofconditions that existed in the plant.

During an interview, the Supervisor,

Radiation Protection and Chemistry stated, "The procedures themselves do notaddress in any great detail in-plant radiation protection support... or coordination of maintenance and operations.. .
own mind,can also say quite honestly,in my

I never conceived of the type of maintenance and operations activities

that would have to go on in areas of such tremendous radiation dose ratelevels as we had" (Int. 79).
The ECS,
monitoring,which had limited its activities to directing onsite and offsite

began a relocation to the Unit 1 control room at about 1030,

shortly before 1100 was fully operational at its new location.

Figure 11-2-5

shows the organizational configuration from that point in time (Int.
25,47, 79,

and

20,

22,

124).

Two chemistry foremen assumed the function of gate monitors shortly
before evacuation of the island.

The normal individuals designated to perform

this function are radiation/chemistry technicians or auxiliary A operators.
Neither chemistry foremen had received emergency plan training in this duty.
At about 1130,

a washdown (decontamination) area was established at the

500 kV substation on the direction of Radiation Protection Foreman D to serve
as a monitoring and decontamination point for individuals evacuating theisland (Int.

18, 45,51).

The preplanned locations for the washdown areas

were the north washdown area, 2 miles north of the north gate on route 441 at

11-2-14

EMERGENCY ORGANIZATION IN EFFECT 3/28/79, 1100-2000

VI

VICE PRESIDENT-GENERATIONEMERGENCY
DIRECTOR

ITN

Ob
servation
Ce nter

ECC

Station Manager

ECC

SECURITY
Site Protection
Sergeant-SiteProtection
Officers

ECC

IN PLANT RADIA-TION PROTECTIONREPAIR
IP AND
A
N
IRS
PARTY
MONITORSI
L------------EMESRGNYH
CHEMISTRY
Chemistry
Foremen-HP
Foremen____________________________________________________I

ONSITEMONITORINGTEAM
Radiation/
ChemicalTechnicians

OFFSITEMONITORINGTEAMS
Radiation/
Chemistry
Technicians. Auxiliary
Operator "A"

ASSEMBLY
AREA
MONITORS
Radiation/
Chemistry
Technicians

ECC-Emergency Control Center--Functional Title Added**

Line intentionally omitted

Figure 11-2-5

WASHDOWN
AREAMONITORS
Radiation/
Chemistry
Technicians-HP
ForemenGATEMONITORS
Chemistry
Forementhe Kiwanis Field adjacent to the Swatara Creek, and the south washdown area,
3 miles south on route 441 at the Collins Substation.
the selection,

Procedures governing

operation and manning of washdown areas are specified in

Paragraph 2.3 of Procedure 1670.6.

Prior to the accident, washdown areas wereintended to be used for vehicle decontamination; but on March 28 the designatedwashdown area (500 kV substation) also served to decontaminate individuals.
The washdown area was staffed by radiation/chemistry technicians,

two of whom

(Radiation/Chemistry Technicians S and X) were relatively new employees andhad never been trained in Radiation Monitoring Team functions.
From 0705 to approximately 2000, the Station Manager (Emergency Director)
maintained ultimate control and direction of site activities in coordination
with the Vice President, Generation, who was located at the Observation Centerby about 1145.

At approximately 2000 on March 28, the nature of the relation-

ship between the Station Manager and Vice President, Generation changed fromone appropriate to the emergency to one more in line with the management
interface that would be expected during a recovery operation.

During aninterview, the Station Manager stated that, in his mind they (the site) "hadentered a recovery phase,since, for the first time, the plant appeared stable

and that there was necessary flow and cooling to the core" (Int. 71).
prepared testimony, the Station Manager stated that "By 8:00 p.m.,In

I was stillin charge of the control room but I was essentially under the direction of
Senior Management..."

(Ref.

72).

An investigator reviewed this statement in light of the radiological andplant operational conditions existing at-the time and compared them with thefollowing policy set forth in Procedure 1670.15,"Post Accident Re-entry andRecovery Plan":"...the immediate phases of responsive action include initial actions
directed toward protection of the public and immediate and planned action
directed toward termination of the incident, containment of effluents,
establishment of incident boundaries,
neling of information,

establishment of control and chan-and protection of the facility and equipment.

11-2-16

After these actions have been effectively accomplished, emergency actions
will be directed toward restoring the facility to its normal operating
condition."
By about 2000,

the operational elements of the emergency organization had

generally "stabilized" the reactor to the point where their actions weregenerally not of an emergency response nature.
The radiological elements ofthe emergency organization were, however, still responding to emergency conditions, attempting to reestablish routine radiological controls in plant (protective actions),

continue in-plant, onsite, and offsite radiological monitoring(assessment actions),

and implement corrective actions to terminate or reduce

actual or potential releases of radioactive materials.
An example of the increasing influence being exerted by groups outside thelicensee's emergency organization was evident in the taking of a reactor coolant
sample on March 29 at about 1630.

During an interview, the Supervisor, Radiation

Protection and Chemistry indicated that there was concern regarding the boron
concentration of the primary system and he was involved in a great deal of discussion concerning the necessity of taking a primary coolant sample.
source of the "order" to secure the sample,
and Chemistry stated,"There were many individuals,

kind of hard to put names or even companies.
Met Ed,and GPU.

the Supervisor,

Regarding the

Radiation Protection

and by many individuals it's

I think they spanned B&W, the NRC,

There was probably somebody from every organization." The Super-visor, Radiation Protection and Chemistry indicated that he felt the samples werenecessary,

but not critical,cal (Int.

79).

individuals,

because of the concern that the core was still criti-

The taking of this sample resulted in the overexposure of twoneither of whom were informed that the operation would be per-formed on a voluntary basis. Examples of activities that were performed without
adequate radiation protection support are described in Section 3.2.

11-2-17

The deviations and shortcomings listed above, i.e.,

separation of theRepair Party Team from its usual location under the Radiation Protection
S~upervisor; the lack of a clearly established element in the emergency organization responsible for in-plant radiation protection during emergencies;
realignment of the emergency duties of the Radiation Protection Supervisor andSupervisor,

Radiation Protection and Chemistry; and an apparent premature

shift to a recovery mode appeared to contribute to a breakdown in radiation
protection and Repair Party controls.
statements,

actions,

This breakdown was evidenced by theand radiation exposures of emergency personnel.

Section 3.2 for a detailed account of these statements,
During an emergency,

actions,(Seeand exposures.)the site protection force is initially responsible

for accountability of personnel on site.

Once accountability is complete,they are to continue to maintain security in accordance with establishedsecurity procedures (Ref.

71).

The investigators reviewed applicable pro-cedures and interviewed members of the licensee's site protection force todetermine the manner in which the security force performed these responsibilities.

(Int. 32, 52, 72)Upon declaration of the Site Emergency,

site protection officers closed

the north gate access point to the owner-controlled area, and initiated accountability (Int.

32, 52, 72).

received from the ECC.

Entry to the site was granted only on clearance

After declaration of the Site Emergency,

personnel who

were permitted entry to the site, with the exception of non-Metropolitan
Edison personnel i.e.,
in time.

NRC and contractors, were not logged in from this point

Accountability was completed by 0830,

declaration of a Site Emergency (Int.

1 hour and 35 minutes after72).

The site protection force maintained their security posts until approximately 0900 when radiation protection personnel directed evacuation fromcertain areas inside the plant itself.

At approximately 1015, the entire site

11-2-18

protection force shut down the security console, thereby defeating security
controls for vital areas within the plant, and evacuated to the north andsouth gates of the owner-controlled area.

Security checkpoints were establishedat both gates, with only the north gate used for access (Int.

72).

Withdrawal from the owner-controlled area to the north and south gate
access control points was made without taking normal security compensatory
measures to these new checkpoints.

A five-man, armed force was,standby at the Observation Center for deployment if
security operations during the emergency,
sergeant who stated,
Director.

Had he,needed.

however,on

The nature of

was described by a site protection"We are directly under the direction of the Emergency

in fact, felt that it was required at the time to maintainthe security posture we would have done so."The Supervisor of Generation

Security stated, "We've operated on the premise that safety overrides security
and that's what we were doing at this point" (Int.

72).

The investigators noted that the licensee's emergency planning provisions
did not contain contingency security procedures for implementation underemergency conditions where the normal security procedures may be inappropriate.
In regard to the planning for contingency security measures during emergencies,
the Supervisor of Generation Security stated,"The only involvement that

security had in any radiation emergency was strictly accountability.
planning it went into, contingency as you described,
knowledge,The

did not occur, to myin the last five years," (Int. 72).

Consequently,

when the normal security measures were abandoned to facili-tate the response to the emergency and ensure the safety of the site protection
force,

contingency measures were developed on the spot and modified as theneed arose.
During the period March 28 through the afternoon on March 30, personnel
and deliveries of emergency supplies and equipment were processed through theowner-controlled access point at the north gate.

11-2-19

Clearance for each entry wasobtained from the ECC,ECS,or Command Post at the Observation Center.

Searchrequirements were dispensed with during this period, and once individuals had
shown some form of identification and were cleared through the north gate they
had unrestricted access to all areas on the site, to include vital areas andareas having radiation and/or contamination at significant levels.
to the suspension of search requirements,

In addition

licensee employees were not logged

on or off site, making accountability in the event of further problems impossible.
During later periods,

on March 30, a means of accountability based on dosimeter

issuance was instituted (Int. 72).
Approximately two days after the event, on March 30, a roving patrol was
instituted outside the protected area barrier.

By March 31,

four days after

the incident, one site protection officer was stationed in the processing
center.

It was approximately nine days after the incident before the site

protection force fully reentered the processing center and began reconstituting
a return to more normal operations (Int. 72).
The investigators determined that the abandonment of the normal security
measures was appropriate under the circumstances existing at the time.

However,the failure to have contingency security measures available for implementation
during such conditions appeared to have the potential for jeopardizing thesafety of individuals for several reasons.

The failure to maintain records ofentries and exits from the site would have made it

impossible to ensure that

all individuals could be accounted for in the event of an escalation of eventson site requiring the evacuation of nonessential personnel.

The Supervisor of

Generation Security further confirmed that the security force could not have
performed accountability because,"...our personnel...

had not been respirator

qualified... [and] would not have been able to go into various places in theplant to get muster sheets,"(Int. 72).

In addition, the lack of positive

security controls for certain vital areas left these areas open to possible
sabotage or unauthorized manipulation of vital equipment.

The potential forthe occurrence of either such act could have seriously jeopardized the public

11-2-20

health and safety through the release of radioactive materials resulting fromequipment damage or operation.

This observation was of particular significancein light of the plant and core condition at the time.
The investigators noted that there were no regulatory requirements relative
to minimum security measures during emergency situations wherein the normal
methods and level of safeguards must be abandoned for reasons of safety.
2.3

NOTIFICATIONS AND COMMUNICATIONS
2.3.1

Notifications

Initial notifications that a Site Emergency had been delcared were begun
at 0702.

By 0720, the licensee had made all applicable notification calls

required by Station Radiation Emergency Procedure 1670.2,
results of these notifications were as follows (Ref.

Figure 4.

The

73-78, Int. 70):

Time

Agency/Individual

0702

PEMA

Duty Officer

0704

NRC Region I

Answering Service

0705

Vice President Generation

0709

Dauphin County Civil Defense

Duty Officer

0709

Dept. of Energy-Radiological
Assistance Program

Duty Officer

0709

Manager, Generation Operations

Persons Contacted*kManager, Generation
Engineering

0713

Radiation Management Corp.

No answer

0718

Pennsylvania State Police

Dispatcher

0720

American Nuclear Insurers

No answer*The Vice President, Generation was contacted earlier in the morning and participated in a conference call concerning the incident.
The individuals making theinitial notifications were not aware that the Vice President Generation had been
contacted and could not be reached at the telephone number listed in theprocedures (Int. 91).
11-2-21

On declaration of a General Emergency,

each of the aforelisted agencies

were recontacted to inform them of escalation of the emergencyto the generalcategory.

The results of these notifications were as follows:Time

Agency/Individual

Persons Contacted

0730

Dauphin County Civil Defense

Duty Officer

0730

Manager,

0735

PEMA

Duty Officer

0735

DOE-RAP

Duty Officer

0735

American Nuclear Insurers

No answer

0740

NRC:

0740

Pennsylvania State Police

0740

Vice President, Generation

0740

Radiation Management Corporation

Generation Operations

Manager, Generation
Engineering

Answering Service

Region I

Dispatcher

Philadelphia
Electric Company
Load Dispatcher

Communications

2.3.2

Subsequent to the respective notifications,open telephone lines between

the control room and the Pennsylvania Bureau of Radiological Health and control.
room and the NRC Region I Incident Response Center were established at 0740
and 0800,

respectively.

An interview of individuals of the Pennsylvania Bureau of Radiological
Health and discussions with the Director, Dauphin County Civil Defense andDirector, Pennsylvania Emergency Management Agency (PEMA)

indicated that thelicensee's initial notifications were in accordance with existing agreements.
The continued flow of updated information to the Pennsylvania Bureau of Radiological Health via the open telephone lines was characterized as good (Int.
46).

11-2-22

On the morning of March 30,

at 0710, Shift Supervisor C directed Control

Room Operator C to open MU-V-13 and vent the MUT.

Shift Supervisor C did not

believe this operation would result in a significant release to the environment.
He did, however,believe that a small release was possible and contacted the

ECS to suggest that a helicopter be dispatched if

it was not already (Int.

173).

Venting of the MUT had been in progress for about 50 minutes when Shift
Supervisor C became aware of a reading of 1200 mR/hr at an elevation of 130
feet above the Unit 2 auxiliary building.

About this same time, the Emergency

Director (Station Manager) and NRC Inspectors B and D arrived in the control
room.
Inspectors B and D were told by a control room operator that an uncontrolled release had occurred during transfer of gases from the MUT to the
WGDTs.

The reason for the release was described to be the lifting of relief
valves due to overpressurization of the vent header and/or WGDTs.
This information was,in turn relayed to NRC Headquarters at 0845 (Ref.

79).

The Emergency

Director instructed Shift Supervisor C to contact the ECS and have them inform
PEMA and BRH.
Interviews of BRH personnel and a discussion with the Director,

PEMA

indicated that two simultaneous but conflicting phone calls were received fromthe site concerning this release.

One call,from the Unit 2 Supervisor ofOperations, was in an "excited" tone and recommended evacuation.

The second

caller, who was not identified, presented details of the release in a calmer
tone and did not recommend protective action.
two calls did not create confusion.

The contradictory nature of the

The BRH was in contact with the facility

via the open telephone line established two days earlier.

They were provided

with current offsite dose rates, meteorological conditions and plant conditions
which indicated that radiation levels were decreasing (Int. 46).
An interview of the Unit 2 Supervisor of Operations indicates that his
recollection of his involvement in the events regarding this release arecontradictory to the recollections of several other individuals (Int. 46,
31,

173,

189,

Ref. 80).

11-2-2363,

According to the Unit 2 Supervisor of Operations,his telephone callserved "to check on the readiness condition of State Civil Defense to evacuate
people...I needed the contingency of knowing the status of State Civil Defense."
Upon further discussion, the Unit 2 Superintendent of Operations provided
additional insight into the reason why he made the call directly to PEMA
rather than having the ECS notify BRH and PEMA.
"It

was not clear to me on the morning of the 30th whether the generalemergency had ever been terminated or not.
the wrong direction,fying them and if

Rather than assume something in

I attempted to act in a conservative direction by renoti-

they were already in a general emergency with the State

Civil Defense and Bureau of Rad Health,If they were not geared up,

it would only be an information call.

then it was anticipated that this call would, infact, bring them to an operational readiness state" (Int.

157).

The investigator determined that the licensee's communications facilities
and provisions generally were adequate to support implementation of the emergency
plan with respect to initial notifications and the relay of updated information
regarding radiological aspects of the response.
it

Early in the event, however,

became apparent that additional telephones were needed to support theoperational aspects of the emergency.

(See Details I for specific information)

A potentially significant communications problem was averted early in theevent.

Except for a very few direct outside lines, all calls to the site must

be manually routed through a main switchboard.

When the switchboard is unat-

tended, no more than one call at a time may be received through the main
telephone number.

Recognizing that the normal operator would not be available

and realizing the potential difficulties that would occur, the Emergency
Director located an individual who was able to operate the switchboard andassigned him to attend it.

11-2-24

3.
3.1

ASSESSMENT, CORRECTIVE AND PROTECTIVE ACTIONSEFFLUENT MONITORING AND CORRECTIVE ACTIONS

3.1.1

Liquid and Gaseous Pathways
At 0400 on March 28, TMI Unit 2 experienced a feedwater pump trip, fol-

lowed by a turbine trip, and approximately 8 seconds later by a reactor trip
(Ref.

81).

Within 3 to 6 seconds of the turbine trip, the reactor coolant systempressure reached the electromatic relief valve (EMOV)

opening setpoint (2255 psig)

and within 13 seconds was back down to the EMOV closure setpoint (2205 psig)
(Ref.

82,

83).

The EMOV failed to reseat, however,and continued to allow reactor

coolant to be discharged to the reactor coolant drain tank (RCDT).

The pressure

and temperature of the RCDT started up almost immediately and continued at a
rapid rate for approximately 3 minutes,
mately 120 psig (Ref.

84, 85).

at which time it

At this point, it

leveled off at approxi-

appears that the RCDT relief

valve opened and started releasing steam and water into the reactor building
where water,

including condensed steam, flowed into the reactor building sump.

At 0411 the reactor building sump pumps activated and pumped as much as 8100
gallons of this low activity water to the auxiliary building sump tank (ABST).
(There is some low probability that the valve lineup was to the miscellaneous
waste holdup tank (MWHT)

but evidence presented later strongly indicates thatthe lineup was to the ABST.)
pump started at 0411.

The first pump started at 0408 and the second

Both pumps remained on until approximately 0438 when

they were manually turned off. The RCDT rupture disc failed at approximately
0415 when the maximum RCDT pressure of 192 psig was reached.

(Ref.

86,

Int.

10).
The estimated flow of 8100 gallons represents a flow rate of 140 gpm perpump.

This data was for single pump operation; with two pumps operating theflowrate was probably less than twice the flowrate for a single pump (Ref.

87).

If the discharge from the sump was routed to the auxiliary building sump tank,as suspected, the flow would be further reduced since the line size reduces'11-3-1from 4 inches to 2 inches (Ref.

88).

Although normally aligned to the MWHT,

it appears from plant records and from an observation by Auxiliary Operator Bthat the level'in the MWHT did not change on March 28.
the sump pump discharge was aligned to the ABST (Ref.

This indicates that50, 89,Int. 10).

Details

I of this report further states that the ABST level was observed to increase,apparently as a result of the RB sump pump operation.

The last recorded reading

prior to the accident indicated approximately a 700-gallon'capacity remainedin the auxiliary building sump tank (Ref.

50,

51).

This tank would have over-flowed into the auxiliary building sump causing water to backup through floor
drain piping in the auxiliary and fuel handling buildings.
The amount of radioactivity transferred from the reactor building sump
was relatively small,
failure (Ref.

90).

since the transfer was completed prior to fuel cladding
Flooding of the auxiliary building sump created a situa-tion where extremely contaminated reactor coolant could later travel into various
parts of the auxiliary'and fuel handling buildings via flooded floor drains.
Investigators evaluated the possibility of siphoning from the sump to the auxiliary building and determined this to be unlikely because the maximum elevationof the piping above the water-levelin the reactor building exceeded the heightto Which the combination of atmospheric pressure and reactor building pressure
could lift

a column of water (Ref.

92-95).

Retracing the sequence of events to approximately 0415,

investigators deter-

mined that high pressure in the reactor coolant drain tank just prior~to failureof the rupture diaphram caused a small release of radioactive gases,
detected by the radiation monitoring system (Ref.

99).

which wasThis provided the firstindication of a leak in the waste gas vent header system that would later be
the primary pathway for the release of gases.
The RCDT was connected to the vent header via two paths (Ref.
the first pathway,

96).

In

a 2-inch vent line connects the RCDT to the reactor coolant

bleed holdup tanks (RCBHT)

through valve WDL-V-1905.

This valve automaticallycloses at 10 psig to protect the RCBHTs but reopens at 6 psig to reestablish

11-3-2

communication with the RCBHTs (Ref.

97).

The RCDT pressure exceeded 10 psig

approximately 2 minutes after the reactor trip, which should have closed valve
WDL-V-1095 (Ref.

84,

85).

This valve would have reopened after the rupture

diaphragm failed (at approximately 0415) and the pressure in the RCDT decayed
to less than 6 psig (at approximately 0435).

This would have provided an open

line between the RCDT and the RCBHTs which would have remained open until reactor
building isolation at approximately 0756 (Ref.

84, 85, 98).

psig spike at 0715 and a 12 psig spike at 0745,

pressure was on the order of

Except for a 4.5

0 psig to 3 psig in the RCDT from approximately 0500 until reactor building
isolation at approximately 0756. The 12 psig spike spanned the time period during
which reactor building isolation occurred.
The second pathway is a 1-inch vent line from the RCDT to the waste gas
vent header, which is normally isolated from the vent header by valves WDL-V-126
and WDL-V-127 (Ref.

96).

Operators have stated in discussions that these valves

were open on the morning of March 28.

Pressure in the RCDT would have pres-surized the waste gas vent header and created a driving force as highas 192 psig which would have been reached just before the rupture disk failed(Ref.

84, 85).

The pressurization of the RCDT to the rupture disk failure point

coincides with a step increase in gaseous activity as measured at 0415 by theventilation system monitors.

A gradual decline of the monitor readings after the

RCDT rupture disk failure indicates that the leak rate into the auxiliary building
was dropping with the pressure in the RCDT (Ref.

84, 85, 99).

The actual pressure

on the vent header could have been as high as 150 psig, which is the reactor building
vent header relief valve (WDG-R-3)

setting (Ref.

100).

WDG-R-3 relieves tothe reactor building atmosphere so that its opening would not have resulted ina release to the environment.

Flow via either of the two pathways just des-cribed would have eventually been released via the leaking vent header.
Early in the morning on March 28 and prior to major fuel cladding failure,
the operators became concerned with the increasing pressure in the reactor bui-lding
(Ref.

101).

Reactor building pressure had increased rapidly by 1.4 psig when

the rupture diaphram on the RCDT failed at 0415 (Ref.

11-3-384, 85, 95).

The operators

apparently were not aware of the failure of the rupture diaphram on the RCOT,or they didn't realize that the leakage into the RCDT was large enough to pressurize the reactor building.

In any event, B once through steam generator (OTSG)

was isolated at approximately 0527 because the operators suspected a steam leakfrom B OTSG into the reactor building (Ref.
was not suspected at that time.

101).

A primary to secondary leak

Isolation of B OTSG may have prevented signifi-

cant releases from occurring through a primary to secondary leak that developed
sometime prior to 0656 (Ref.

102).

Prior to 0527, both OTSGs were using thepower-operated emergency mainsteam dump valves MS-V-3A and MS-V-3B (atmosphericreliefs) because of a high hotwell level (Ref.

103).

After B OTSG was isolated, A OTSG continued to discharge through the atmospheric relief valve until approximately 0656 when the restarting of the circulating water pumps would have automatically shifted control to the turbine bypass
mode of operation (Ref.

104).

Records indicate that at approximately 0842 A

OTSG was not steaming and it appears to have remained in this condition until
approximately 091,5.

Between 0915 and 1100, steaming from the atmospheric relief

on A OTSG was reinitiated.

B OTSG remained isolated on the main steam sidefrom 0527 until the main steam isolation valves were briefly opened (7 seconds)
at 0656 (Ref.

104).

It

is possible that steam from the B OTSG was directed tothe condenser when the OTSGs switched back to turbine by-pass operation.
B OTSG turbine by-pass valve was closed at 0704 (Ref.
pump discharge monitor, VA-R-748,

106).

The

The condenser vacuum

alarmed at 0656 due either to the brief openingof the MSIVs or to the switching of OTSG pressure control back to the turbine
bypass mode of operation.
from the OTSGs (Ref.

102).

This was the initial release of detectable activity
This activity was primarily noble gases and wasdischarged to the vent stack (Ref.

107,

108).

Data indicate that B OTSG was effectively isolated and the secondary system, except for B OTSG itself, remained relatively free of contamination.

Samplesfrom condenser vacuum pump discharge taken at 0524 and from the condensate pumpdischarge taken at 1500 showed no iodine activity (Ref.

11-3-4

107,

109).

Levels ofnoble gas in the range of E-3 pCi/ml on the condensate pump discharge sample
probably were due to background-at the time of counting.

An isotopic -analysis

of a sample taken from B OTSG at 1030 on March 30 showed 7.9 pCi/ml of iodine-131
and 2.1 pCi/ml of iodine-133 (Ref.

110).

These levels are so high that it seems

likely that any leakage out of B OTSG would have been easily detected at thecondenser or condensate vacuum pump discharge.

Cooling using the A OTSG was

discontinued at approximately 0842 on March 28 but was initiated again sometime
after 0915 (Ref.

111).

When the atmospheric relief valve on the main steamline from A OTSG was reopened,a man was sent to the roof of the turbine building

at approximately 1100 to survey near the release point (Int. 41).
was detected by this survey.

No radiation

Cooling A OTSG using the atmospheric relief valve

continued until approximately 1230 or possibly as late as 1315 (Ref.

112).

No

further use of the atmospheric relief valve was recorded.
NRC personnel performed radiation surveys in the visible steam plume fromthe atmospheric relief valves between 1130 and 1200.

Radiation levels as highas 20 mR/hr were attributed to the plume (Ref.

Based upon data showing

effective isolation of the B OTSG,

113).

investigators concluded that these readings

were caused by the noble gas plume (which was invisible) coming from the Unit
2 vent stack.
The misidentification of OTSG sample lines that run to the primary sample
room contributed to release of iodines through the industrial waste treatment
system.

In addition to the normal OTSG sample lines which are piped to thesecondary system lab, sample lines are also piped to the primary sample room
for the express purpose of taking OTSG samples when there is a primary to secondary leak.

Since the unit had no history of primary to secondary leaks, themisidentification of the lines had not been discovered earlier.

In an attempt

to confirm which OTSG was leaking, samples from both OTSGs were taken at approximately 0724 using the Unit 2 0TSG sample station in the Unit 1 primary sample
room (Ref.

108).

When surveyed with a GM survey meter the A OTSG sample appearedcontaminated and the B OTSG sample did not (Int. 23).

11-3-5The operations person-

nel requested another sample since they felt sure it was the B OTSG that had
the leak.

A chemistry foreman and the Supervisor, Radiation Protection andChemistry decided that the sample lines might be mislabelled (Int.

23, 79,130).

The foreman then went to the normal (secondary side) steam generator sample
station in Unit 2 which they knew to be piped correctly, and resampled the two
OTSGs.

This time the B OTSG was contaminated and A OTSG was not.

these samples,

While taking

the sample lines were flushed until radioactivity was detectedin the B OTSG sample.

This flush water went into the control and service building

sump where it was later pumped to the industrial waste treatment system (IWTS).
A diagram of the sump discharge system is shown in Figure 11-3-1.

This wasperhaps the first and most significant source of contaminated water to the IWTS
during the first three days.

It was not determined when the turbine building

sump became contaminated or what contribution the turbine building sump madeto activity levels found in the IWTS.
IWTS,

It

is known that the activity in thewhich had been declining on March 30,

increased on March 31 (Ref.

114).

This coincided with the sampling of the B OTSG at 2045 on March 30 in the secondary sample room and also with the pumping of the turbine building sump to the
IWTS (Ref.

110).

In summary,the EMOV was allowing reactor coolant to flow to the RCDT throughthe failed rupture diaphram into the reactor building sump.

The reactor building

sump pumps were turned off at approximately 0438 after pumping as much as 8100
gallons of relatively low activity water to the ABST, which would have overflowed
into the auxiliary building sump and caused the auxiliary building sump to overflow (Ref.

86,

87).

The RCDT was open to the waste gas system through two path-ways and had provided indication that the waste gas system vent header was alreadyleaking (Ref.

96, 99).

B OTSG developed a leak sometime prior to 0656 but did

not become a significant contributor to the effluent releases because it was
isolated early due to suspected (but nonexistent) steam leaks in the reactor
building (Ref.

102).

Despite these plant conditions,

there were still no significant radiologicalindications outside of the reactor building.

11-3-6SUMP PUMP DISCHARGES TO THE INDUSTRIAL WASTE TREATMENT SYSTEM.Control

Unit-Service Bldg Sump

Bldg Sump_TurbineI

Aux. Bldg

TurbineBldg.
Fuel Handling Bldg

Containment

Tendon Gallery Sump

Diesel Gen Bldg Sump

Industrial
Waste
Treatment
Plant
Mechanical
Draft CoolingTowers(MDCT)

MDCT

Radiation
DDetector
Pit

Susquehanna River

Figure 11-3-1

At approximately 0622 on March 28, the first area radiation monitor response to fuel cladding damage was detected (Ref.
area monitor, HP-R-213,

115).

The incore instrument

located on the 347-ft elevation in the reactor building

began increasing at a rate that took itfrom approximately 1 mR/hr to greaterthan 10 R/hr (off scale) in 14 minutes.

This monitor had previously dropped

from 100 mR/hr to 1 mR/hr almost simultaneously with the reactor trip becauseof the rapid decay of nitrogen-16 in the primary coolant,

indicating that themonitor could detect radiation in the primary coolant loops.

It

is thus possiblethat HP-R-213 was initially detecting the collection of fission product activity
in the hot leg loops to the OTSGs,or itinto reactor building via the RCDT.

may have been detecting gases discharged

In either case, this was the first clear

indication of fuel cladding failure as detected by the radiation monitoring
system.

Data from HP-R-213 is supported by data from HP-R-214, the containment

dome monitor.

Monitor HP-R-214 is a heavily shielded ionization chamber mounted

on the top of the reactor building elevator shaft (Ref.

44,

116).

This monitor

was increasing by 0627 at a rate comparable to HP-R-213.
Major fuel cladding damage was particularly significant from a radiologicalsafety and environmental release standpoint since millions of curies of radioactive nuclides were present in the gap between the fuel pellets and the cladding.
Very little of this activity normally escapes from the fuel cladding.
the huge inventory of gap activity released in such a short time,Withthe activity

per unit volume of liquids and gases in the reactor coolant system became sohigh that almost every leak, no matter how small in volume, became significant
from a radioactivity standpoint.

Gaseous radioactivity was detected in theauxiliary building by the atmospheric radiation monitor system within 15 minutesof the first indications of gap activity release as indicated by radiation monitors
in the reactor building.

(Ref.

99).

Since the only sources of such high radio-

activity were the reactor building and systems containing reactor coolant, oneof these sources had to be leaking.

11-3-8

The reactor building was probably not a major contributor to releases tothe fuel handling and auxiliary buildings since no detectable changes in radioactivity in these buildings occurred when the reactor building pressure went
negative at 1530 on March 28 (Ref.

95,

118).

The RCDT vented some highly radio-

active gases to the waste gas system vent header which leaked into the auxiliarybuilding and were detected by the atmospheric monitors (Ref.

99,

117).

RCDT pathway was closed by reactor building isolation at 0756 (Ref.

The

98).

Investigators have concluded that reactor coolant flowing through the makeup
and purification system was the major source of releases both initially andfor the duration of the accident.

The reactor coolant makeup and purification

system continually takes a portion of the water out of the reactor coolant systemand passes the water through demineralizers then through filters and into themakeup tank (Ref.

119).

The makeup pumps return water from the makeup tank to

the reactor coolant system (RCS).
provide the inventory control,
surizer level.

In addition,

The rate of return to the RCS is varied towith the inventory need being determined by presa small but important amount of the water is pumped

to the reactor coolant seals.

This system isolated automatically on reactor

building isolation but was quickly returned to service in order to maintainseal water to the reactor coolant pumps and control pressurizer level (Ref.
120).

The makeup tank is designed to operate with a gas space occupying about

one third of the normal tank volume (Ref.

114).

It was in this space that gases

evolving from the cooled and depressurized reactor coolant collected.

One indi-

cation that this tank was a source of gaseous leakage was a correlation of monitor
readings with the makeup tank level.

This correlation was considered reliable

since the makeup tank is not normally vented and a change in level causes a
corresponding change in pressure.
closed (as it

normally is)past its seat; however,if

If the makeup tank vent valve, MU-V-13,

wasduring this period, there was apparently leakage

the investigators were unable to definitely establish

it was closed.
Starting at approximately 0635 there appears to be a direct relationshipbetween some of the atmospheric monitor responses and the level in the makeup
tank (MU-T-l)(Ref.

118, 121).

An example of a monitor which correlated well

11-3-9

with the makeup tank level was the gas channel for HP-R-228,
exhaust duct B.

auxiliary building

The correlation is such that the monitor responded upward rapidly

as the level was increased in the makeup tank and when the level in the makeup
tank decreased the slope of the monitor response started decreasing.

Itappearsthat the vent header was pressurized on a rising tank level; however a fallingtank level did not draw gas back out of the header since there is a check valve
to prevent flow back to the tank (Ref.

The vent header pressure would

96).

then decrease because of a leak in the vent header or because the waste gas
working properly) was pumping the gas into a waste gas decay

compressor (iftank (Ref.

122).

Early on March 29, the buildup of noncondensible gases in the makeup tankcaused a reduction in letdown flow due to the back pressure from the makeup
tank (Int.

173,

189).

Manual venting of the makeup tank to reduce pressure

was initiated at 0710 on March 29 and continued on a periodic basis until sometime after March 30 (Int.

157,

173,

189).

The venting process initially con-sisted of short periods of time when MU-V-13 was opened to vent makeup tankpressure into the waste gas decay tanks via the leaking vent header (Int.
The pressure buildup,173).

however became too rapid to control with short venting

periods and on the morning of March 30, the pressure in the makeup tank caused
the liquid relief valve (MU-R-I)

to open, forcing all of the water from the

makeup tank into the reactor coolant bleed holdup tanks.

Sometime after the

opening of MU-R-l (between 0600 and 0710 on March 30) the pressure in the RCBHTs
This pressure exceeded

was observed to be greater than 30 psig (Int.

173).

the normal relief valve setpoint of 20 psig.

Because of the potential opera-

tional problems created by the high pressure and resulting low water level inthe makeup tank, two shift supervisors decided jointly to vent the makeup tankin an attempt to reduce pressure and reestablish a normal water level in the
makeup tank.

The makeup tank vent valve (MU-V-13)

0710 and remained opened for at least 2 hours (Int.
of the high pressure in the RCBHTs,

was opened at approximately
157,

173,

189).

Becausean uncontrolled release was likely alreadyin progress via the relief valves on the RCBHTs.

11-3-10

A 1200 mR/hr beta-gamma exposure rate was measured from a helicopter 130
ft above the Unit 2 reactor building on March 30 at approximately 50 minutes
after the opening of the makeup tank vent valve.

This exposure rate was due

to gas leakage from the waste gas vent header which was released to the environment via the auxiliary building ventilation exhaust.
The investigators concluded from the makeup tank level that the makeup
tank relief valve was closed at the time of the measurement.
This indicatesthat the RCBHT relief valves were closed.
vent valve,

Manually closing the makeup tankhowever, would likely have caused the makeup tank relief valve toreopen, causing the pressure in the RCBHTs to increase, resulting in the reinitiation of an uncontrolled release through the RCBHT relief valve.
It is likelythat for some period of time during this event, a release occurred over which
operators had no practical control.
Communications with the State and otheractions of a less operational nature are discussed in section 2.3.2.
Discussions with licensee personnel involved in a continuing program to
identify the sources of leaks which occurred during the accident indicate thatthe waste gas compressors were the most likely pathway for gases which leaked
from the waste gas system.

These compressors use a water seal and discharge a

mixture of liquid and gases into a separator tank (Ref.

123).

High pressuresin the vent header and/or liquid in the vent header may have caused these pumps
to lose the water seal and thus the ability to move the gas into the waste gas
decay tanks (Ref.

123).

Gas could have been discharged through a relief valve

on the separator or possibly thorough the automatic liquid overflow to the auxiliary building sump tank (ABST)

either as a gas or as a water-gas mixture.

The rupture disc on the ABST had failed prior to March 28 and any gases going
into the ABST would have escaped immediately to the auxiliary building.

Therelease pathways of the gases from the auxiliary building to the environment
are discussed in Section 3.1.2.

1173-11

The amount of liquid released into the auxiliary building and the effectof this release on airborne radioactivity levels in the plant are not accurately
known.

However,liquid leaking from the reactor coolant system into the fuelhandling and auxiliary buildings was cooled and depressurized,
entrained and dissolved gases.

thus releasing

Evaporation of water leaking onto the floor

probably contributed to airborne iodine and particulate activity inside theauxiliary and fuel handling buildings.

Several pathways for liquid releases

into fuel handling and auxiliary buildings from the makeup and purification
system are considered possible.

Relief valves for those systems containing

highly radioactive material were evaluated.
MU-R-l

Makeup tank relief valve; setpoint 80 psig; discharges to the reactor
coolant bleed holdup tanks (Ref.

MU-R-2100, 124),

Upstream of seal return cooler; setpoint 150 psig; discharges to themakeup tank (Ref.

MU-R-3100, 124),

Letdown relief valve; setpoint 130 psig; discharges to the reactor
coolant bleed holdup tanks (Ref.

MU-R-4

The following valves were considered:

Inlet to the core flood tanks,
floor drain (Ref.

100, 124),
setpoint 700 psig; discharges to the100, 124),

MU-R-5A/B Makeup and purification demineralizer relief valves; setpoint 150
psig; discharges to the floor drain system (Ref.

100, 124).

Of these valves, MU-R-5A and B were determined to be likely pathways tothe auxiliary building sump.

There were difficulties in maintaining letdown

flow on March 28 and the block orifice MU-I-FE was bypassed in an attempt toincrease letdown flow (Ref.

125).

Letdown flow restrictions could have caused

the pressure to exceed the setpoint of MU-R-5A and 5B (150 psig).

An auxiliary

operator stated that letdown flow was oscillating at times as if a relief valve

11-3-12Iwere opening and closing (Ref.

125).

MU-R-3 may have been opening and closing,-

however this valve does not discharge to the floor drains (Ref.
have contributed to the inventory in the RCBHTs,

124).

It

could

increasing the potential forthis tank to discharge liquid into the waste gas system vent header and out
thru vent header liquid drains.

A liquid release via the RCBHT vent line probably

did not occur before March 30.
On March 30, makeup tank relief valve MU-R-I lifted at approximately 0430
and stayed open until approximately 0715 (Int.

173,

189,

Ref.

121).

Duringthis period, a large amount of water was pumped into the RCBHTs (Ref.
Two of these tanks indicated a high level (full)

68,

158).

and tank pressures of greaterthan or equal to 30psig, which means water from the tanks could have flowed
into the vent header (Int.

173).

Water from the tanks could have flowed into

the sump via the vent header liquid drains, the waste gas compressor drains,
and waste gas decay tank drains (as discussed previously).
Relief valves MU-R-2 and MU-R-4 were not determined to represent significant potentials for release.

MU-R-2 discharges to the makeup tank and MU-R-4

could only be a pathway during filling of the core flood tanks (Ref.

124).

Another potentially significant pathway developed at approximately 1352
on March 28 when seal water to various radwaste system pumps was lost.
water was lost because motor control centers (MCC)
erable (Ref.

126).

Seal

2-32A and 2-42A became inop-

MCC 2-32A was made operable again at 0700 on March 29.

Loss of seal water resulted in significant leakage from pumps WDL-P-5A & B,
which take suction on the reactor coolant bleed tanks (Ref.
letdown relief valve (MU-R-3)

127).

Because ofdischarges and because of their use in the letdown

flow path, these tanks contained high levels of radioactivity.

A sample of

RCBHT-C on March 30 showed an iodine-131 concentration of 109 pCi/ml (Ref.

128).

Other radwaste pumps were of less significance since the pumps were taking suction
on tanks containing pre-accident water.

11-3-13

There was known leakage from seals on the river water pumps (RR-P-lA,
and IC) which, although not radioactive,IB,

contributed to flooding around theauxiliary building floor drains on the 281-ft elevation.

This input of cleanwater into the sump may have forced contaminated water out of the floor drains
and into the control and service building sump and the diesel generator sump.
These two sumps were eventually pumped to the industrial waste treatment system
(Ref.

129).

(See Figure 11-3-1.)

Additional leaks of radioactivity probably occurred through system leaksin the letdown and makeup systems (Ref.
discussions with licensee personnel,

130).

From maintenance records andinvestigators determined that there were

small Tiquid leaks from valves and instruments in the makeup and purification
system130).

The contribution from these sources to the activity released

to the auxiliary building sump could not be determined.

The following are examples

of some of the potential pathways taken from maintenance work requests:
Letdown sample monitor (MU-R-720)

sample pump had aleak at the pump suction union.

This had created aradioactive hot spot.
22,

1979 (Ref.

Work request 1806 dated March

130).

Makeup Pump 1C, casing drain plug leaked.
1297 dated March 20,

1979 (Ref.

Makeup valve 17, packing leaked.
dated March 22,

1979 (Ref.

Work request

130).
Work request 1825

130).

Flow Transmitter MU-7-FIT leaked downstream of MU-V-383D
and MU-V-384D.
(Ref.

Work request 1856 dated March 27,

130).

11-3-141979

There are numerous other examples similar to those above indicating leaksin the makeup system (Ref.

130).

Discussions with licensee personnel

indicatedthat such leaks ranged from a few drops per hour to perhaps a few tenths of a
gpm.

From a review of preaccident leak rate determinations required by Technical

Specifications,an investigator estimated the total preaccident leakage fromall reactor coolant system sources in the auxiliary building to be from 0.4 to
1.6 gpm.

Leakage from other sources which developed during the accident would

have added to this basic leak rate.

There are some potential liquid release

paths such as drain pots off of the waste gas vent header and automatic drains
off of the waste gas compressors and waste gas decay tanks that have not beenthoroughly evaluated.
Although liquid volume release rates appear to be small,the specific activityof the reactor coolant was so high that any leakage was significant (Ref.

131).

Gases evolved from reactor coolant released into the fuel handling and auxiliarybuildings contributed to airborne radioactivity inside these buildings.

After

March 30 this source of airborne radioactivity may have been of little significancefor noble gases but could have been a contributor to the radioactive iodine
which evolved more slowly through evaporation of liquid on the floors.
The following table lists the major nuclides for a sample of reactor coolant
taken on March 29.

NuclideHalf life

iodine-131
iodine-133
cesium-134
cesium-136
cesium-137
barium-140
strontium-89/90
gross u8 d20.8 h
2 y
13 d
30 y
12.8 d50 d/29y

Coolant Concentration
PCi/cc'
1.3 E 4
4.6 E 4
6.3 E 1
1.8 E 2
2.8 E 221.0 E 1
5.3-E 03.6 E-4*Reactor coolant sample taken at approximately 1700 on March 29.

11-3-15

3.1.2

Determination of Release Rates and Amounts
3.1.2.1

Noble Gases

This section is a summary of release rates calculated by the licensee using
TLD data and actual X/Q values for the periods that the TLDs were exposed (Ref.

85).

In order to provide a perspective of the releases for the first three daysof the accident, data for the period March 28 through April 30 is shown eventhough this investigation was generally limited to the period March 28 throughMarch 30.

These data represent approximately 99% of all of the noble gas releases.

The following noble gas release values are from Table IV-B-lof the li-

censee's "Third Interim Report on Three Mile Island Nuclear Station Unit 2 (TMI-2)
Accident, July 16,

1979" (Ref.

84, 85). The release values in this table were

calculated by the licensee by applying atmospheric dispersidn factors to environmental results.

These calculations were not verified by the investigators.

The licensee values are consistent with a preliminary assessment by the NRC
staff which estimated a release of 1.3 E 7 Ci for the period March 28 throughApril 5.

11-3-16

PERIOD OF TLD EXPOSURE (MO,. DAY,HR)

03-29-17
to
03-31-16

03-31-17
to
04-03-15

04-03-16
to
04-06-13

04-06-14
to
04-03-24

TOTAL

Isotope03-28-07
to
03-29-16

Xe-133
Xe-133m
Xe-135
Xe-135m
Kr-88

4.9E6
1.2E5
1.5E6
1.4E5
6.1E4

2.1E6
3.9E4
7.7E4
1.3E3
0

I.IE6
1.5E4
1.4E3
0
0

2.7E5
1.9E3
0
0
0

1.5E4
0
0
0
0

8.3E6
1.7E5
1.5E6
1.4E5
6.1E4

TOTAL

6.6E6

2.2E6

1.1E6

2.7E5

1.5E4

I.OE7

*The last three weeks of the month are combined into one groupsince the contribution is less than 1% of the total.
Using this table, the noble gas releases can be compared to 10 CFR 20.106
limits. A footnote to Appendix B, 10 CFR 20 states that releases to unrestricted
areas should be limited, such that:
CA
+

MPCA

CCB
MCPB

+

CCC

<

1

MPCC

where CA is the concentration of radionuclide A divided by the maximum permissible concentration of nuclide A as given in Appendix B to 10 CFR 20.
quantities (Ci)

If theof noble gases released between March 28 and April 30 were assumed

to be the total releases for a year, the quantities could be divided by the
number of seconds in a year to obtain an average annual release rate in curies
per second.

This yields:

xenon-133

8.3 E 6 curies/year
3.15 E 7 seconds/year

xenon-133m

1.7 E 5 curies/year
3.15 E 7 seconds/year

xenon-135

1.5 E 6 curies/year
3.15 E 7 seconds/year

xenon-135m

1.4 E 5 curies/year
3.15 E 7 seconds/year

4.4 E-3 Ci/sec

krypton-88

6.1 E 4 curies/year
3.15 E 7 seconds/year

1.9 E-3 Ci/sec

11-3-17-

2.6 E-1 Ci/sec=

5.4 E-3 Ci/sec-

4.8 E-2 Ci/sec

Applying the dispersion factor (X/Q)
Specifications,

given in the Environmental Technical(6.7 E-6 sec) one obtains the followng:=M_xenon-133:

2.6 E-1

Ci

x 6.7 E-6 sec= 1.7 E-6 Ci

sec

FS

xenon-133m:

5.4 E-3

Cisec

x 6.7 E-6 sec
mu= 3.6 E-8 Ci

xenon-135:

4.8 E-2

Ci

x 6.7 E-6 sec= 3.2 E-7 Ci

secm
xenon-135m:

4.4 E-3

Cisec

x 6.7 E-6 sec
m9= 2.9 E-8 Ci
W9krypton-88:

1.9 E-3

Cisec

x 6.7 E-6 m-Ssec= 1.3 E-8 Cim-

Ci/m 3 equals Pci/cc, therefore the above concentrations can besubstituted into the summation equation as follows:

I Ci

=

1.7 E-6
3 E-7

=

11

i MPCi+

3.6 E-83 E-7+

3.2 E-71 E-7+

2.9 E-83 E-8+

1.3 E-8
2 E-8This means that noble gas releases when averaged over a year are at least
11 times the limit in 10 CFR 20.
3.1.2.2

Iodine

The amount of radioactive iodine released was calculated based on continuous samples collected from the Unit 2 station vent.
from monitor HP-R-219.

The samples were taken.

A sample for the period 0400 on March 28 through 1900

on March 28 was lost prior to analysis. The amount of radioactivity released
during this period was estimated by a licensee contractor based on samples fromauxiliary building and fuel handling building ventilation exhausts.

11-3-18

Data shown below was extracted from TABLE IV-C-l of the "Second Interim
Report on the Three Mile Island Nuclear Station Unit-2 (TMI-2) Accident, "June 15,
1979, published by the licensee (Ref.

84).

TABLE IV-C-1
Smoothed Iodine Release Rate Data
Used in Dose Assessments
131

I Release RatepCi/sec
4.21
22.7
2.7
9.7
2.3
7.0
0.43
3.7
6.9
12.7
0.46
1.3
2.2
4.1
6.6
8.6
14.0
6.0
11.0
3.0
5.5
7.5
2.0
5.5
1.5
2.5
1.0
3.8
1.5
0.80
0.50
0.46
0.37
0.38
0.56
0.48

Start Date
(Mo. Date Hr.)
79032804
79032819
79033022
79040106
79040303
79040319
79040519
79040615
79040706
79040803
79040909
79041016
7904111979041323
79041410
79041505
79041508
79041518
79041616
79041624
79041716
79041804
79041808
79041914
79042022
79042213
79042304
79042312
79042316
79042406
79042516
79042600
79042700
79042808
79042900
79043000
1

Iodine release rates are roughly constant over the period from one start timeto the next.

11-3-19

From this table, investigators estimated that the amount of iodine-131
released through April 30 was approximately 14 curies.

This number differswith results of an analysis by a licensee contractor of samples taken on April
12 from the charcoal adsorber in the fuel handling building and auxiliary building
ventilation systems.

This analysis determined 26.84 Ci of iodine-131 were released

from March 28 through April 12. The detailed data for the charcoal adsorber
analysis are in a licensee contracted report by Nuclear Consulting Services,Inc.,"Analysis of the Adsorbers and Adsorbents from Three Mile Island Unit 2,"

May 25,

1979. The method used for this analysis was not reviewed by the investi-

gators.
If the release rate in curies per second is calculated for the 26.84 curies
above, the release rate can be multiplied by the X/Q in the Environmental Technical
Specifications to show that concentrations of iodine-131 (only) at the siteboundary would have been less than 10 CFR 20 Appendix B limits if

averaged overa year as allowed by 10 CFR 20.106.
The radiation monitoring stripcharts indicate that significant noble gas
releases started at approximately 0700 on March 28 but significant iodine releases
did not start until a few hours later (Ref.

99,

117).

Unit 2 gas monitors rapidly

went offscale, with the Unit 2 particulate and iodine monitors going offscale
within 1-2 hours after the gas monitors.

A review of data from the Unit 1 fuelhandling building and auxiliary building ventilation system monitors which weremonitoring leakage into Unit 1 from the Unit 2 fuel handling building shows
the iodine channels increased and fell rather sharply back to near previous
levels.

This indicates that the monitors were responding to noble gases rather

than iodine since iodine would have accumulated on the charcoal sample cartridgeand caused a continuous increase in the reading (Ref.

69).

Neither particulate monitors nor iodine monitors on the effluent release
paths provided any quantitative information because of the high noble gas background.

Even though HP-R-219 particulate and iodine channels came back on scale

11-3-20

during the period under investigation,itis not known what fraction of thereading should be attributed to iodines and what faction to noble gas.

For

one nine hour period on March 30, the iodine channel reading on HP-R-219 decreased instead of increased.
Since the iodine detector was measuring a buildup of iodine on a charcoal
cartridge, the rate of increase (cpm/min) must be determined in order to establish a concentration.

A declining monitor response would probably indicatethat noble gases adsorbed on the filter were being removed at a rate exceeding
the iodine buildup or the background radiation was declining at a rate greaterthan iodine buildup rate.

Either situation means the monitor reading should

not be used.
It

is probable that most of the releases were via the Unit 2 auxiliarybuilding and fuel handling building ventilation systems with possibly some shortreleases from the relief valve vent header directly to the vent stack (Ref.

122).

There was some leakage into Unit 1 via the fuel handling buildings (Int. 79).
Gaseous radioactivity in Unit 1 increased significantly whenever the Unit 2
ventilation systems were turned off (Ref.

69).

The Unit 2 and Unit 1 fuel handlingbuildings are not physically separated in the spent fuel pool areas.

Shift

Foreman B stated that the Unit 2 ventilation system supply fans tripped andremained off because of high radiation levels,

but the exhaust fans operated

continuously except for a few brief periods when the ventilation systems wereturned off in an attempt to reduce the release rates.
Securing the fuel handling building and auxiliary building ventilation
systems early on March 28 and again on March 29 caused exposure rates to increasesignificantly in the Unit 2 auxiliary building, thus hampering emergency activities.

Perhaps more important was the fact that control room airborne radioactivity

levels started increasing when the ventilation systems were shutdown.
Difficultyin maintaining control room isolation due to heavy personnel traffic into andout of the control room in conjunction with poor meteorological conditions allowed

11-3-21

gases escaping from the plant to be picked up in the ventilation system supply
tunnel.
room.

This was the likely cause for gaseous activity increases in the control
Because of the need to ensure habitaility of the control room and tokeep dose rates as low as possible in the auxiliary building to facilitiate
emergency activities,
3.1.2.3

the ventilation systems were subsequently kept in operation.

Liquid Release

The only significant release of liquid occurred through the industrial
waste treatment system (IWTS).

The IWTS does not normally process radioactive

fluids, and there are no radiation monitors in this system.

The effluent, after

dilution, is sampled continuously by plant discharge monitor RM-L7 just prior
to being discharged into the Susquehanna River.
An investigator determined from discussions with Chemistry Foreman A that
contaminated water probably was initially introduced into the industrial waste
treatment system from the control and service building sump.
the discharges from the secondary sample system sink.

This sump receives

The sample line fromthe contaminated B OTSG was flushed into the secondary sample sink for approximately 45 minutes commencing at 1000 March 28.

Other sumps were known to containradioactive water after March 30 as a result of the accident; but only the turbine
building sump should have been contaminated because of the primary-to-secondary
leak.

It

could have been a source of contamination to the IWTS during the first

three days.

How soon contaminated water was pumped to the IWTS could not be

determined; but levels of iodine-131 as high as E-4 pCi/cc were detected in
the IWTS as early as 1000 on March 29.
At 0400 on March 28, a discharge from the IWTS was in progress.

This dis-charge continued until 0900, when it was stopped because of the problems in
Unit 2 (Ref.
and 0900.

138).

Approximately 29,100 gallons were released between 0400

There is no evidence to suggest that this release was contaminated.

A small increase in the reading on RM-L7 during this period was considered by

11-3-22

the investigator to be due to increasing background radiation from Unit 2 gaseous
releases. This is supported by the fact that the reading on RM-L7 did not change
after the liquid discharge was terminated.
On March 29, a release was started at 1315 but terminated at 1410 becauseof high levels of xenon reported in a sample of the IWTS water taken at 1225
(Ref.

138, 139).

Van),

was later determined by the laboratory personnel to be error due to theThis analysis,high noble gas background.

performed in the NRC mobile laboratory (NRC R-I

This was reported to the licensee and the release

was restarted at 1610 and continued until 1815.

The release was terminated at1815 on orders originating from the NRC executive management team at NRC Headquarters (Ref.

140).

A sample of the IWTS water taken between 1000 and 1100

on March 29 and analyzed on March 31 contained iodine-131 at a concentrationof 1.2 E-4 pCi/ml (Ref.

139).

For estimating releases,

it was assumed that

1.2 E-4 pCi/ml was the maximum value during the releases on March 29.

A smallincrease in the reading on RM-L7 was noted at approximately the time the release began at 1610 (Ref.

102).

The reading on RM-L7 remained relatively con-stant during the release,indicating that the radioactive characteristics ofthe release were not changing significantly during the release (Ref. 102).
Releases from the IWTS on March 29 totaled 29,950 gallons containing an estimated
12.66 millicuries of iodine-131.
The IWTS remained shut down from 1815 on March 29 until 0300 on March 30.
During this period of time,

there were several communications between the NRC

and various State and Federal agencies (Pennsylvania Bureau of Radiological
Health, a representative of the State of Maryland,

and Region III of the Environ-mental Protection Agency) regarding discharges from the IWTS (Ref.

141).

The

most significant communications were from the Pennsylvania Bureau of Radiological
Health (BRH)to the NRC,

requesting that the NRC stop the licensee from dis-charging liquid waste to the river.

The last such BRH request was at 1825 onlarch 29 and asked the NRC to holdup discharges to the river.
-his requst at 2205 on March 29.

The BRH rescinded

At approximately 0005 on March 30,

11-3-23

the NRCexecutive management team instructed NRC personnel onsite to inform the licensee
that releases from the IWTS could resume (Ref.

141).

to notify the NRC when releases were resumed (Ref.

The licensee was instructed

141).

At 0300 on March 30, a release was initiated from the IWTS and continued
for the remainder of the day (Ref.

During the period from 0300 to 2400

138).

three samples were taken which, when analyzed on March 31,
for iodine-131 (Ref.

139).

These results were used,sample results from the March 29 sample,

showed positive resultsin conjunction with thewhich also showed iodine-131,

mate average concentrations over the period of the release (Ref.

139).

to estiIt

isestimated that the releases from the IWTS for March 30 totaled 206,400 gallons
containing 59.91 millicuries of iodine-131.
In summary,itfrom 0400 March 28,

is estimated that releases from the IWTS for the period
until 2400 March 30,

72.56 millicuries of iodine-131.

totaled 265,450 gallons containing

Based on an annual average,

the releases were

below 10 CFR Part 20, Appendix B, limits for releases to an unrestricted area.
3.2

IN-PLANT ASSESSMENT AND PROTECTIVE ACTIONSThis section describes,

in chronological order, the in-plant radiation

protection actions taken by the licensee during the period 0400 March 28 to2400 March 30.
The section is divided into five subsections.

The first four subsections

describe in-plant radiation protection during discrete phases of the emergency
organization.

The fifth subsection is a summary of the radiation protection

findings.

11-3-24

3.2.1.

Radiation Protection and Chemistry Prior to Declaration of
the Emergency

Shortly after hearing the announcement that Unit 2 reactor had tripped,
Radiation Chemistry Technicians L and P placed the Unit 2 reactor coolant sample
system on recirculation in anticipation of.performing an isotopic analysis forradioiodine as required by Technical Specification 4.4.8 (Int.
this same time (approximately 0430),
cpm)

58, 96).

About

Auxiliary Operator B noted an alarm (500on the count rate meter (RM-14/HP-210)

located near the model room accessdoor and an indication that both reactor building sump pumps were running (Int.
10).

The auxiliary operator adjusted the RM-14 meter from the Xl to the XIO

scale, performed a survey with an exposure rate meter (E-520)
found less than 0.1 mR/hr.
his findings.

in the area, andHe informed Radiation/Chemistry Technician I of

He also informed Control Room Operator D of his findings andconcern.
About 0445,

Radiation/Chemistry Technician L was requested to collect areactor coolant sample and to analyze it to determine boron concentration (Int.
58).

This analysis,

performed about 0515,

indicated approximately 700 ppm boron.

This was a decrease of about 300 ppm from the 1026 ppm determined at 0330, prior
to the reactor trip (Ref.

90).

The technician stated that he did not understand

why the boron concentration had decreased, since he had been informed that waterfrom the borated water storage tank (BWST)
system.

had been added to the reactor coolant

He informed Shift Foreman C of the boron-analysis results.

Shift Foreman

C stated that he also did not understand the decrease and requested a second
sample be analyzed.
During this same period-of time, Radiation/Chemistry Technicians B and I
sampled the main condenser vacuum pump discharge as required by Technical Specification Table 2.3-2, Note (3)

(Int.

42,

106).

This sample was analyzed at

0548 and did not appear to indicate a primary-to-secondary

11-3-25leak.

At 0545 the Supervisor,
control room.
(ES)

Radiation Protection and Chemistry arr.ived in theHis major concern at that time was that an engineered safeguards

function had occurred resulting in the injection of sodium hydroxide into

the primary coolant system.

He did not address this concern at that time, since

he was immediately approached by the Unit Superintendent,

Technical Support

and advised that there were indications of a leak inside the reactor building
and that a sample of the reactor building atmosphere was needed (Int.

20).

Indication of a leak included both reactor building sump pumps starting
automatically,followed by a reactor building sump high water level; an increase

in reactor coolant drain tank pressure, burst of the rupture disk followed bya rise in reactor building pressure to 1.4 psig; increase of the intermediateclosed cooling water coolant letdown monitor (IC-R-1092)

to five times its alarm

setpoint, followed by an alarm on the particulate channel of the reactor building
air monitor (HP-R-227(P))

(Ref.

142).

In addition, at 0521 an operator had

requested from the computer several parameters which indicated the electromatic
relief discharge pipe temperature was at least 64'F hotter than discharge pipe
from the code safety valves (Ref.

143).

These indications were consistent witha release from the pressurizer.

Interviews with the Supervisor,

Radiation Pro-

tection and Chemistry did not indicate that he was informed of these indications.
The Supervisor,

Radiation Protection and Chemistry and Radiation/ Chemistry

Technician I immediately proceeded to the 305-ft elevation of the auxiliarybuilding to take a gas sample and remove the charcoal cartridge from monitor
HP-R-227.

As the technician loosened the wing nuts that secured the charcoal

cartridge holder, water began to spray with enough force to propel the cartridgefrom the holder.

The technician caught the cartridge and forced it

the holder and secured the wing nut.

back into

The water spray contaminated his hand.

He was later decontaminated (Int. 20, 42).

The Supervisor, Radiation Protection

and Chemistry returned to the control room and informed Unit 2 Superintendent,
Technical Support of the unsuccessful attempt to obtain the air sample.
Unit 2 Superintendent, Technical Support informed the Supervisor,

Radiation

Protection and Chemistry of concern over the 700 ppm boron concentration.
requested that the analysis be verified.

11-3-26

TheHe

Returning to the Unit 1 laboratory area, the Supervisor,

Radiation Pro-

tection and Chemistry discussed the boron analysis results with Radiation/Chemistry
Technician L and learned that a confirmatory sample was being taken, by Radiation/
Chemistry Technician P (Int.
the Supervisor,96).

At about this same time (approximately 0620),

Radiation Protection and Chemistry was requested to make prepar-

ations for an entry into the reactor building.

He asked Radiation/Chemistry

Technician I to accompany the operator assigned to enter the reactor building.
Radiation/Chemistry Technician I suggested that, since he and the other radiation/
chemistry technicians were fatigued, the entry be delayed until the oncoming
shift arrived in about 30 minutes.

The Supervisor,

Radiation Protection andChemistry agreed with this suggestion, but asked that he and.Radiation/Chemistry
Technician B prepare to make the entry in the event a delay was impossible.
The two technicians left for Unit 2 to make the preparations (Int.
At this time (approximately 0630),

the Supervisor,

106,

156).

Radiation Protection

and Chemistry was informed that two independent boron samples had been completed(one indicated 402 ppm, the other 405 ppm) and that a sodium analysis indicated0.153 ppm (Int.

20,

94).

This concerned him,since he believed it

indicatedthat deboration of the primary coolant system was taking place (Int.
requested the control room be informed of the sample results.

20).

He

Radiation/Chemistry

Technician P had measured the gross degassed beta-gamma activity of the reactor
coolant and found that it

had increased from 0.38 pCi/ml at 0330 to 4.0 pCi/ml

at 0602.

This information, with the boron results, was phoned to the control

room (Int.

96).

A reactor coolant sample collected about 0643 indicated a gross gamma activity of 140.73 pCi/ml (Ref.

144).

The Unit 1 nuclear sample room area radiation

monitor RM-G3 alarmed (2.5 mR/hr) during collection of this sample.

Radiation/

Chemistry Foreman B surveyed the area and found 200 mR/hr on contact with thesample lines (Int.

77).

These exposure rates were not recorded.

coolant recirculation flow was turned off.
RM-A12,The primary

The nuclear sampling room air monitor,

was inoperable during this period of time (Ref.

42).

No other measure-ments of the concentration of radioactive materials in air to which these technicians were exposed were made.

11-3-27

As the Supervisor, Radiation Protection and Chemistry was about to leave
the Unit 1 laboratory area for the Unit 2 control room, he heard an area radiation
monitor alarm (Int.
(RM-G4),20).

This alarm, the Unit 1 hot machine shop area monitor

was set at 2.5 mR/hr (Ref.

145).

A review of Unit 1 strip chart (RMR-2)

showed that this monitor had been reading a steady 0.85 mR/hr until 0645 (Ref.
69).

From 0645 to 0650, the reading increased from 0.85 mR/hr to 400 mR/hr.

Radiation/chemistry technicians were returning from Unit 2 via the model room
door and walking towards the catwalk that crosses over the railroad tracks inthe fuel handling area, when they heard the hot machine shop radiation monitor(RM-G4)alarm (Int. 42).

Their first action was to order a security guard sta-tioned near the hot machine shop door to leave the area.

Then,

as one cut thelock off the door, the other returned with a survey instrument (RO-2).
Supervisor,

The

Radiation Protection and Chemistry arrived on the scene and observed

as the technicians surveyed the area.

The Survey indicated general area gamma

exposure rates of 500 mR/hr and up to 1 R/hr several feet from the unshielded
sample lines located in the hot machine shop.

The technicians determined thesource of radiation to be the sample lines.
Unit 2 FSAR Section 12.1.2.11,"Sample Station Shielding," states, "Allsample and recirculation lines are shielded by placing them in shielded cubicles
or pipe chases which are located at sufficient distances from normally occupied
areas."

An inspector observed that the reactor coolant sample lines were notshielded or placed in a pipe chase as they run from the Unit 2 fuel handlingbuilding through the normally occupied access pathways of Unit 1 fuel handlingbuilding through the hot machine shop into the nuclear sampling room.
Based on the survey, the Supervisor,

Radiation Protection and Chemistryconcluded Unit 2 had suffered severe fuel degradation (Int. 20).

He contacted

the Unit 2 Superintendent, Technical Support and relayed his findings.

The

Unit 2 Superintendent, Technical Support indicated to him that they were alreadydoing everything they could to get cooling water and boron to the reactor coolingsystem (Int. 20).

He was informed that a Site Emergency was being declared.

11-3-28

3.2.2.

Initial In-Plant Assessment and Protective Actions

Day shift radiation/chemistry technicians were arriving in the laboratory
area at this time.

The Supervisor,

Radiation Protection and Chemistry directed

Radiation/Chemistry Technician 0 to proceed to Unit 2 to conduct surveys andwarn personnel

(Int.

78).

He directed Radiation/Chemistry Technician U to callin the Radiation Protection Supervisor and a radiation protection foreman.

Heplaced Radiation/Chemistry Technician I in charge of the Emergency Control Station
until relieved by a foreman (Int.

6).

He told the technician to start assemblingother technicians into onsite and offsite survey teams.
At this point, the Supervisor, Radiation Protection and Chemistry ran to
the Unit 2 control room to assume his Emergency Plan responsibilities.
passed the Unit 1 condensate water treatment area,
gency alarm (Int.
Upon arrival

As he

he heard the radiation emer-

20).
in the control room, the Supervisor,

Radiation Protection

and Chemistry proceeded to the radiation monitoring system portion of the control
panel.

He noticed that the condenser off gas monitor (VA-R-748)

and concluded that a primary to secondary leak had developed.

was in alarm,
He noted that

other radiation monitors were increasing and that the containment dome monitor(HP-R-214) wa•

in the alert and rising (Int.

20).

A review of multipoint recorder strip chart HP-UR-1901 indicates that thefirst area radiation monitor to increase was the incore instrument panel area
monitor (HP-R-213)
115).

located on the 347-ft elevation of the reactor building (Ref.

At 0400, following the reactor trip, this monitor's reading decreased

from 100 mR/hr to 1 mR/hr due to the rapid decay of the nitrogen-16 inventory
in the reactor coolant system.

At 0622,

this monitor increased rapidly from 1

mR/hr to greater than the range of the instrument (10 R/hr) in 14 minutes.
This may be indicative of clad failure and corresponding relase of gap activity
into the coolant system.

The containment dome monitor (HP-R-214)

11-3-29

responded ina similar manner.
to 150 mR/hr.

The containment dome monitor increased rapidly from 1 mR/hr

By 0652, the containment dome monitor had increased to 350 mR/hr.

The first indications of increasing radiation levels recorded by the area radiation monitoring system in the Unit 2 auxiliary building were noted at 0638 by
the makeup tank area monitor (HP-R-206).

It

increased from 0.7 mR/hr at 0636

By 0700, this monitor was reading 1 R/hr.

This

monitor had been tagged out of service before the accident because it

under-

to 100 mR/hr in 16 minutes.
responded to radiation.

A review of multipoint recorder strip charts HP-UR-1907 and HP-UR-2900
indicated that at 0640 most of the auxiliary building and fuel handling building
ventilation systems'
(Ref.

99,

monitors were indicating a rapid rise in airborne activity

117).

In the period 0645 to 0730,

Radiation Protection Foreman B, several radiation/

chemistry technicians and auxiliary operators were in the.auxiliary building(Int. 21).

These radiation/chemistry technicians were performing beta gamma

general area surveys and collecting air samples.

Radiation/Chemistry Technician

J was at the Unit 2 health physics control point area counting air samples (Int.
51).
A review of pre-accident radiation surveys indicates that radiation levels
were generally less than 1 mR/hr in all areas of the auxiliary building except
in valve rooms and cubicles containing demineralizers,(Ref.

filters, and waste tanks

179).
The results of the majority of these radiation surveys performed on March 28

after 0400 were not recorded.

In some cases survey results were recorded but

these records were not retained.
Table 11-3-1 summarizes survey data from about 0700 to approximately 0730.
The data are based on information obtained during interviews with several radiation/chemistry technicians and and. Radiation Protection Foreman B (Int.
42, 51,78).

11-3-30

21,

TABLE 11-3-1•SUMMARY' OF SURVEY DATATYPE OFRADIATION

LOCATION

INSTRUMENT

281 water on floor

Teletector

305 valve alley

RO-2

305

door to makeup
purification
valve room

RO-2

I

200 mr/hr

305

door to makeup
purification valve
room (re-survey)

RO-2

y

5 R/hr305

door to makeup
tank room

RO-2

'y

1 R/hr305

door to makeup
tank room(re-survey)

E-520

'y

2 R/hr305

hydrogen line justinside makeup tankroom

Teletectory10 R/hr305

door to makeup
tank room(re-survey)

RO- 2

y

5 R/hr at
at 10 ft305

Waste-gas-decay
tank roomRO-2

y10-20 mR/hr

305

radwaste operating

RO- 2

N

50 mR/hry

RESULTS
0.2 mr/hr*
100 mr/hr

panel
305

HP-R-227

RO-2

N

10 mR/hr

305

near containment
personnel access
hatch

Instrument unknownN

1-2 R/hr

328

near HP-R-219
a rapid series ofmeasurements taken
just prior to 0655

E-520

y

10 mR/hr
50 mR/hr
75 mR/hr
175 mR/hr*This survey was performed by Radiation Protection Foreman B who walked throughthe water on the floor without wearing protective clothing (Int. 21).
He surveyed
himself immediately on leaving the auxiliary building and did not find himself tobe contaminated.
Radiation Chemistry Technician I also toured this area shortly
thereafter and reported water on the floor but no increase in radiation levels(Int. 42). He had worn a particulate respirator during his tour. He also was notcontaminated.

11-3-31

Air samples were collected on the 305-ft elevation near the containment
air monitor (HP-R-227)byand on the 328-ft elevation near the stack monitor (HP-R-219)
These samples were given

Radiation/Chemistry Technician U at about 0650.

to Radiation/Chemistry Technician J for analysis.
not maintained.

A record of the results was

Radiation/Chemistry Technicians 0 and U surveyed themselves

after leaving the auxiliary building.
to levels of up to 30 mrad/hr.

Both found they were uniformly contaminatedNo record of this survey was maintained.

they were in the process of decontamination,

While

the background radiation levelsin the area of the Unit 2 health physics control point began to increase,causing

the area to be evacuated to the hall between the control and auxiliary buildings.
A step-off-pad was established in the hall (Int. 51,

78).

Air samples taken about 0700 near the entrance to the auxiliary building
were remembered by Radiation/Chemistry Technician L to be in the range of E-8
pCi/cc of beta gamma particulate radioactivity (Int.

58).

At about this time, the auxiliary building and fuel handling building ventilation was secured.

Radiation/Chemistry Technician U, at the health physics

control point area, noted the ventilation had been secured and called the ECS
to remind them to have someone close the model room door (Int.

78).

When Unit

2 fuel handling building ventilation is off, Unit 1 fuel handling building ventilation will draw air from Unit 2 if

the model room door is open.

This appears

to have occurred in this case since Unit 1 fuel handling building ventilation
monitor (RM-A4)
at 0726 (Ref.

reached the gross particulate alarm set point of 1 E-8 PCi/cc
146).

Ventilation of the Unit 2 auxiliary and fuel handling buildings

was started at 0900 according to the notes of Control Room Operator J (Ref.
147).

With ventilation secured,

the airborne activity increased throughout

the auxiliary and fuel handling buildings.
By 0730,

a General Emergency had been declared, word had been passed via

the paging system to evacuate the auxiliary buildings and the ECS had been established at the Unit 1 chemistry and health physics control points.

11-3-32

Health Physics Procedure 1670.8 "Emergency Re-entry for Repair or Rescue,"
Revision 1, dated January 16,

1978 and Procedures 1670.5 and 1670.2 provide

discussion and guidance for in-plant radiological assessment and protective
actions in a radiation emergency (Ref.

148).

These procedures do not specifi-

cally establish what methods the licensee will use in controlling a sustained
in-plant adverse radiological environment.

The Supervisor,

Radiation Protection

and Chemistry stated that, in retrospect, the Emergency Plan Implementing Procedures do not adequately address an in-plant radiation emergency (Int. 25).
Section 12.3.2.10 of the FSAR states, "a radiation work permit (RWP)

isrequired for all personnel working in radiation areas or posted RWP areas except
in emergency situations where action is necessary for the protection of personnel
or safety of the Unit.

In such a case, the entry should be made by qualified

personnel carrying radiation monitors.

In the event such an entry is made,entry shall be documented."
The first Emergency Repair Party Team (ERT)

action performed by the teamlocated at the ECS was to close the model room door.

The Repair Party Leader,

Mainteneance Foreman B, and two assistants were told by Radiation/ Chemistry
Technician I that the exposure rate was 400 mR/hr in the work area (Int.
The Repair Party wore full face particulate respirators,
and their street. clothes.

187).

appropriate dosimetry,

They were not accompanied by a Repair Party Team

monitor and could not locate a radiation survey instrument so they took a countrate meter (RM-14/HP-210)

with them.

This instrument is normally used for measuring contamination.
ment deflected full scale prior to reaching the work location.
closing the door and exiting in about 10 minutes.
were found to be contaminated.
decontaminated by showering.

The instru-

The team continued,

On return to the ECS all

The individuals removed their clothing and were
Their entry was not under a RWP,

mented, no air activity measurements were made,no continuously indicating dose

rate instrument was used in the high radiation area,

11-3-33.

it was not docu-

and no record of. the extent

or distribution of personnel contamination was maintained.

The investigators

review of emergency training provided to these individuals, as discussed in
Section 1.2.1,

indicates the training they had received was not consistent withthat required by Procedure 1670.9.

The performance of this Repair Party tends

to substantiate the investigators evaluation of training.
In the time period from 0730 until 0807 (the time the model room door was
closed) two independent surveys of the auxiliary building were performed (Ref.

149).

One survey performed by Radiation Protection Foreman B was intended to determine
the exposure rate on the roof near the stack (Int. 21).
with an exposure rate meter (E-520,

closed-window)

The measurements taken

read 4 mR/hr.

descended from the roof through the auxiliary building.

The foreman

He did not wear any

protective clothing.

He attempted to return to ECS through the model room door,

but found it

He did not perform-any radiation measurement in the auxi-

closed.

liary buidling.

On exiting at the newly established step-off-pad he was foundto be uniformly contaminated to 20 mR/hr as measured with an exposure rate meter(E-520/HP-177)

closed window probe.

Unit 1 for decontamination.

He removed his clothes and returned to

No record of this entry or his skin contamination

was made.
The other independent survey was performed by Radiation/Chemistry Technician I (Int. 6).

On his own volition, he decided to make what he believed

to be the first re-entrysurvey into the auxiliary building after evacuation.
The protective measures he took included full protective clothing,high- andlow- range pocket dosimeters,

and a high-self-contained breathing apparatus,

range exposure rate instrument (Teletector).

He found the gamma radiation levels

had increased to 1.5 to 2 R/hr on the 281-ft elevation.

The general area exposure

rates in the 305-ft elevation hallways were 1.5 to 2 R/hr, with 4 R/hr on thedoor to the makeup tank room and 1 to 2 R/hr near the reactor building air monitor(HP-R-227).
average,

He found the exposure rates the 328-ft elevation to be,less than 100 mR/hr.

on theOn exiting the area and removing the protective

clothing, he surveyed himself and found that he was not contaminated.
several notes of exposure rates but these notes were not maintained.

11-3-34

He madeBased on the above examples,the initial entries into the auxiliary building

to assess radiological conditions were not adequately planned.

Preparations

did not include consideration of appropriate protective clothing for entires
nor did they include recognition of the need for surveys that could be used toplan future entires into the area.

Air samples were not analyzed to identify

isotopic constituents of the activity.
3.2.3

Implementation of Emergency Plan
The Radiation Protection Supervisor arrived on site about 0730 (Int.

22).

He proceeded to the ECS and assumed control as called for in the Emergency Plan.
His first efforts were to complete establishment of communications and assignmentof survey teams.

He repeatedly attempted to contact the Supervisor,

Radiation

Protection and Chemistry to determine the extent and nature of the emergency.
He did not make this contact until the ECS was evacuated to the Unit 2 control
room at 0910 (Ref.

150).

In the period from 0745 to 0900, several samples weretaken and analyzed to evaluate the extent of fuel degradation and primary system
integrity.
The Radiation Protection Supervisor stated that the licensee did not havea "Chemistry Supervisor" and that this position was filled during the Emergency
by Chemistry Foreman C (Int.

22).

Interviews with chemistry foremen indicatedthat two of them, Chemistry Foremen A and C, assumed responsibility for fillingthe "Chemistry Supervisor" position described in the Emergency Plan (Int.
132).

A third chemistry foreman,

Chemistry Foreman B, had arrived about 0700

and gone to the Unit 2 control room (Int. 23).
possible primary to secondary leak fromcontrol room operator.

130,

There,

he became aware of a

discussions with the shift foreman or

He proceeded to the Unit 2 secondary laboratory to posi-tion valves so that potentially contaminated samples from the steam generators(OTSG)

could be collected in the Unit 1 primary sample room.

Collection of

OTSG samples was consistent with the requirements of Unit 2 Emergency Procedure
2202-2.6, "OTSG Rupture," Revision 4, dated October 6,

11-3-35

1978 (Ref.

151).

Chemistry

Foreman B proceeded to the Unit 1 primary sample room and collected a liter
sample from each steam generator.
meter (RM-14/HP-210)

He surveyed each bottle with a count rate

and found that A OTSG sample caused the survey meter topeg full scale (50,000 cpm).

He immediately informed the control room that

the A OTSG showed high activity.

Because of increasing air activity in thechemistry laboratory area, he did not perform any further analysis on thosesamples.

However,sample records indicate that the B OTSG sample taken at 0724

was gamma scanned at 0745 and did not have indications of a primary-to-secondaryleak.
The control room, when informed that A OTSG had a leak, told Chemistry
Foreman B that they did not understand,with a leak in the B OTSG.

since their indications were consistent

Chemistry Foreman A had a preconceived notion thatthe sample points in the primary chemistry laboratory may have been labeled
incorrectly.

A confirmatory sample at the normal sample point in the Unit 2

secondary laboratory was taken at about 1000 by Chemistry Foremen A and B (Int.
23,

130).
The one liter sample from the B OTSG caused the survey meter (RM-14/HP-210)

to rapidly peg full scale.

The two chemistry foreman concluded that the samplelines in the primary sample room were in fact improperly labeled and that the
B OTSG had developed the primary to secondary leak.
to Shift Foreman C.
physics area (ECS)This information was provided

The samples were taken to the Unit 1 chemistry and healthto be analyzed.

The samples were not analyzed because of

the high background radiation.
Chemistry Foreman A arrived on site at about 0800 (Int.
to the ECS,
time.

130).

He proceededmustered in, and reviewed what chemistry data was available at thatHe then went to the Unit 2 control room to discuss plant conditions with

the Supervisor,

Radiation Protection and Chemistry.

Returning to Unit 1 ECS,he decided that a sample of the reactor coolant was necessary.

He instructed

two radiation/chemistry technicians to collect and analyze a reactor coolant

11-3-36

sample.

Radiation/Chemistry Technician K, when told to collect the primarycoolant sample,

hesitated because of the information he had become aware of

regarding the exposure rates (2 R/hr) measured on the primary sample lines (Int.
24).

He was also concerned that collection of the sample would result in highairborne activity and radiation levels in the ECS since it
sample room.

adjoins the nuclear

The spread of airborne activity from the nuclear sample room toother parts of Unit 1 control building,observed in the past (Int.

including the control room, had been

11).

The Radiation Protection Supervisor (the ECS Director) was nc. informedthat the chemistry foreman had directed a reactor coolant sample be taken (Int.
22)

A review of the Emergency Organization as presented in Station Radiation

Emergency Procedure 1670.2,
22,"Site Emergency Procedure," Revision 9,

dated November

1978 does not specifically describe the need for the chemistry supervisor

to coordinate his activities with the radiation protection foreman (Ref.
However,71).

figure 4 of this procedure shows the chemistry supervisor operating

under the direction of the Radiation Protection Supervisor.
Radiation/Chemistry Technicians K and I, told to collect the reactor coolant
sample,did not take special precautions such as temporary shielding,shielded containers,

use ofremote handling tools, verification of ventilation, andspecial plans to dispose of liquid waste.

Neither considered themselves asvolunteers and neither discussed their planned actions with the Radition Protection Supervisor.

Both dressed in a complete set of protective clothing,including self-contained breathing device operated in the pressure demand mode,
and a plastic rain suit. Both wore 0-200 mR, and 0-5 R pocket dosimeters and
TLD badges.

Neither wore extremity monitoring.

No provisions for air samples

were made (Int. 24, 42).
When reactor coolant sample recirculation flow was initiated at 0845 thehot machine shop area radiation monitor (RM-G4)
800 R/hr in a period of about 7 minutes.

11-3-37

increased from 200 mR/hr to

The nuclear sample room area radiation

monitor (RM-G3)

increased from 300 mR/hr to 25 R/hr in about 7 minutes.

radiochemistry laboratory monitor (RM-G2)
after RM-G4 and RM-G3,

increased,from 0.3 mR/hr to 8 mR/hr (Ref.

The

starting a few minutes
69).

One radiation/chemistry technician entered the nuclear sample room with a
Teletector and measured an exposure rate of 200 R/hr at the sample hood.

Thesecond radiation/chemistry technician entered the room, opened the sample isolation valve for a few seconds and allowed about 50 ml to flow into the sample
bottle.

He estimated his total time in the sample room to be 10 seconds.

review of the Unit 1 effluent monitors, fuel handling building (RM-A4)
liary building (RM-A5),
monitors (Ref.

152).

since February 27,A

and auxi-

indicated a rapid increase on all channels of these
The Unit 1 control room air monitor had been out of service

1979.

The technicians exited the nuclear sample room, leaving

the sample in the hood (Int.

24, 42).

The Radiation Protection Supervisor at the ECS noted increasing radiationand levels of airborne activity in the ECS at this time (Int. 22)
and foot monitors and count rate survey meters (RM-14/HP-210)
ECS alarmed.

.

The handlocated in theLevels of radiation were recorded as 2 to 3 mR/hr at 0850 to 50

mR/hr at about 0900 in the ECS area (Ref.

153,

Int. 22).

The Radiation Protection

Superv-isor-decided,

based on increasing radiation and airborne activity levels(9.36 E-11 pCi/cc),

to evacuate the ECS to Unit 2 control room.

He was informedof the increasing levels on RM-G3 and RM-G4 but was not aware of the reactor
coolant sample.
The reactor coolant sample recirculation flow was secured at 0900 (Ref.

150).

This resulted in a decrease in radiation levels measured by monitor RM-G4 from
700 R/hr to 4 R/hr in about 10 minutes.

The RM-A4 and RM-A5 monitors responded

in a similar fashion, returning to their initial levels in about 30 minutes.
This appears to indicate a gaseous release resulted from this sample.

11-3-38

The radiation/chemistry technician measured the 50 ml sample with a Teletector and found it
area,

read 200 R/hr at 6 inches (Int. 42,78).

On exiting the

one radiation/chemistry technician checked his pocket dosimeter and noted

a dose of 200 mR had been received in the 10-second period.

This technician

was not surprised that airborne activity was increasing at the ECS.

Both techni-

cians removed their protective clothing, performed a rough check for personnel
contamination and helped evacuate the ECS to Unit 2 control room.
At 0917,
the ECS was established in the Unit 2 control room.

At 0925 and 1004, the ECC

Log indicates a request to preform an air sample in the Unit 1 control room
(Ref. 153).
No air sample results were recorded; but at 1015 the log indicates
Unit 1 control room ventilation was put in air recirculation mode.
In the absenceof air sample data, the control room area monitor RM-Gl strip chart was reviewed
and did not indicate an increase in radiation levels from 0700 to 1200 (Ref.

154).
As the ECS was being evacuated,

Radiation/Chemistry Technician U and Radi-

ation Protection Foreman C stayed behind to complete the reactor coolant sample
(Int.

43,

78).

The foreman stationed himself outside the ECS door to control

access.
Radiation/Chemistry Technician U dressed in protective clothing with afull-face particulate respirator, wearing a high-and low-range pocket dosimeter
and TLD,prepared to enter the sample room.

rates were in the 200 R/hr range.

He had been informed that the doseNo extremity monitoring was worn and no special

precaution such as sheilds or handling tools were used.

No air samples weretaken in the nuclear sample room (Int. 78).
The radiation/chemistry technician entered the sample room, removed 1 mlof sample and diluted it

in a waiting volumetric held by another radiation/chemsitry technician in the primary sample laboratory.
This sample was analyzed
on the gamma multichannel analyzer.
He reentered the nuclear sample room andwithdrew 5 ml of coolant for a boron analysis.

11-3-39

He checked the exposure rate

with a Teletector and remembered it

to read about 2 R/hr.

Radiation/Chemistry

Technician U recalls a pocket dosimeter exposure of 400 mR for the sum of both
entries.

He remembered the sample boron concentration was 248 ppm (Int. 78).

The activity of this sample indicated the following (Ref.
50 pCi/ml
181 pCi/ml
144 pCi/ml
449 pCi/ml
247 pCi/ml
604 pCi/ml
29 pCi/ml

It

155):

iodine-132
iodine-131
iodine-133
xenon-133
xenon-135
rubidium-88
krypton-85.

is important to note that this reactor coolant sample was initiatedand taken without the cognizance of the Radiation Protection Supervisor.
sample was not recirculated for the normal 30 to 45 minutes,

Theand the sample

line was not flushed for 15 or 20 seconds as is normally done to remove thestagnant volume of water in the sample leg.

Radiation/Chemistry Technician K

stated that the sample was not representative of water in the reactor coolant
system because the recirculation and flush were inadequate (Int.

24)

This con-clusion has since been supported by evaluation of radiation data from monitors
RM-G2,

RM-G3,

and RM-G4 (Ref.

69,

154).

These monitor data indicate the total

activity of the reactor coolant sample was probably several times higher than
the reactor coolant sample collected on March 29

at 1615.

Failure to take special precaution resulted in unnecessary exposure to
this individual.

11-3-40

Another important result of this sampling effort was its effect on theemergency response capability of the facility staff.

This highly radioactive

sample caused an increase in airborne activity and radiation levels that resulted
in evaucation of the ECS area.

Italso resulted in the loss of analytical equip-

ment that included:a.

Both Ortec 401A beta counting systemsb.

Packard Model 3320 Liquid Scintillator, andc.

Multi-channel Analyzer

Loss of this equipment left the licensee without isotopic identification
capability until the arrival of the NRC Mobile Laboratory at 1930 on March 28
(Int.

73).

Loss of this capability resulted in the inability to confirm thepresence of iodine-131 in air and water samples inplant and to verify,
timely manner,in a

the results of environmental samples.

At about 0800 to 0830, Maintenance Foreman G was directed by operations
supervision to enter the 305-ft elevation ofthe auxiliary building to bypass
a fire defeat switch that would allow starting of the ventilation fans.

This

Repair Party entry was coordinated through the ECC rather than the ECS (Int.
143).
The foreman was briefed by the Supervisor, Radiation Protection and Chemistry
as to the possible radiation hazards.

He proceeded to the Unit 2 health physics

control point, which was manned by radiation/chemistry technicians.

He dressedin full protective clothing and wore a self-contained breathing apparatus,and pocket dosimeter.

TLD,

He entered the area carrying a Teletector survey instrument.

Just inside the entrance doors he recalled an exposure rate of 150 mR/hr.

The

fire detection system panel is located just about 20 to 25 ft beyond these doors.
He recalled an exposure rate of 20 mR/hr at the panel.
minutes.

On exit of the area,

contamination.

The job took about 5

he was surveyed and found to be free of personal

He returned to the control room and debriefed with the Supervisor,

Radiation Protection and Chemistry.

11-3-41

While the primary coolant sample was being collected, Radiation/Chemistry
Technician T, at the request of Radiation Protection Foreman B, donned a full-faceparticulate respirator and proceeded to the 328-ft elevation of the auxiliarybuilding to change the station vent stack monitor (HP-R-219)
charcoal samples (Int. 75).

particulate andHe wore high- and low-range pocket dosimeters anda TLD badge and carried a Telectector.

The entry took place about 0830 to 0900.

He changed the charcoal cartridge; however,the particulate filter was missing

and was not replaced with a new particulate filter since the filter housing
was missing.

The technician recalled exposure rates of 2 R/hr in the area ofmonitor HP-R-219.

He stated that it

took 45 seconds to change the samples andabout 2 minutes total time in the building.
450 mR,

by TLD it was 395 mrem.

His dose by pocket dosimeter wasThis indicates exposure rates higher than 2

R/hr were probably present in the auxiliary building at the time of this entry.
Results of the charcoal cartridge analysis have not been identified.
The ECS personnel,

including the Radiation Protection Supervisor, arrived

in the Unit 2 control room and established the alternate ECS off to one sideof the operating area.
room (Int. 7, 22).

This resulted in about 40 to 75 people in the control

The ECS group had brought what equipment they could from

Unit 1. This included portable survey instruments,
radios.

Ludlum scalers, and portable

The Supervisor, Radiation Protectionand Chemistry briefed the Radiation

Protection Supervisor.
At 0800, the control room ventilation system was placed in a recirculation
mode (Ref 147).

This means the ambient control room air is recirculated throughparticulate filters and charcoal adsorbers.

Fresh air is supplied to the control

room in this mode through particulate filters and charcoal adsorbers to makeup for air leakage from the room.

The control room air monitor (HP-R-220)

this air supply before filters and adsorbers,

samplesand thus is not representativeof actual control room atmosphere in the recirculation mode.

In this condition,

the actual air activity in the control room should be less than the concentration
measured by HP-R-220 as long as the control room doors are infrequently opened.
Entry into the control room by contaminated personnel and the leaving of doors
open result in introduction of radioactivity into the control room.

11-3-42The monitor HP-R-220 particulate channel alarmed at 0948 (Ref.

99).

A

review of strip chart HP-UR-1907 indicates the monitor HP-R-220 particulate
channel increased from 40 cpm at 0930 to 30,000 cpm at 1040 (Ref. 99).
gas channel increased from 30 cpm to 200 cpm in the same time period.

However,the iodine channel

The alert

increased from 1 cpm at 0945 to 3000 cpm at 1130.

alarm for the iodine channel is set by procedure at 100,000 cpm (Ref.
Using the calibration data from the license procedure,

The

49).

the table below is anindication of incoming concentrations of airborne activity for the period described above (Ref.

49).
Unit 2 Control Room Incoming Air Activity(0930-1130,

Gross particulate
Gaseous
Iodine 131

calculated from HP-R-220)
2.2 E-8 pCi/cc
5.6 E-5 pCi/cc
7.3 E-12 pCi/cc

At 1010, the gas channel of monitor HP-R-220 reached the high alarm setpoint
(Ref.

99).

room.

The Radiation Protection Supervisor considered his main duty the control

A decision was made to move the ECS personnel to the Unit 1 control

of onsite and offsite survey teams and that this would be performed from Unit
1 control room (Int.

22).

Some radiation/chemistry technicians remained in the Unit 2 control roomin support of the Supervisor,

Radiation Protection and Chemistry.

At 1017,

all personnel in the Unit 2 control room donned respiratory protective devices
equipped with particulate filters (Ref.

156).

A sample taken in the control

room at 1030 indicated 1 E-7 pCi/ml gross beta-gamma particulate activity.

Noisotopic identification was possible because of high background radiation inradiochemistry laboratories.

Several additional air samples were taken in thecontrol room during the next 6 hours that particulate filter masks were worn.
These samples indicated initial gross particulate activities of up to 1 E-5
pCi/cc.

Half-life determinations were made that indicated short-lived isotopes(Int. 24).

11-3-43

The radiation/chemistry technicians and the Supervisor, Radiation Protection and Chemistry discussed these findings and reviewed the monitor HP-R-220
indications (Int.

24).

Records of the survey results were not maintained.

The charcoal cartridge for monitor HP-R-220 was put in service on March 13,
1979 and removed April 4,

1979.

Sample results indicate an average iodine-131

activity 1.14 E-12 pCi/cc.
The strip chart for HP-UR-1901 was reviewed and indicated that the control
room gamma radiation levels,

as measured by monitor HP-R-201 were generally

less than 0.5 mR/hr for the first three days, except for a 1-hour period from0030 to 0130 on March 29 when the maximum reading was 1 mR/hr (Ref.
TLD badge located in the control room from February 28 to May 31,
a total dose of 25 mrem gamma and no beta (Ref.

115).

The

1979 indicated157).

The licensee's decision to wear respiratory protection in the control roomfor 6 hours was conservative from a radiation protection point of view and consistent with station Health Physics Procedure 1616,"Use of Respiratory Protective

Devices," Revision 9, dated March 20,

158).

1978 (Ref.

The Supervisor, Radiation Protection and Chemistry stated that he pursued
the following general course of action in implementing his Emergency Plan responsibilities (Int.
a.

133):

Establish communications,

get the monitoring teams into action, andmaintain a feed-back of survey data,b.

Support entries into the auxiliary building, andc.

Provide input to the operations staff based on his understanding of
the radiation monitoring system.

By 1100, the ECS was reestablished in the Unit 1 control room and a flowof information had commenced (Ref.

153).

From the time of his arrival in the

Unit 2 control room, the Supervisor, Radiation Protection and Chemistry was

11-3-44

communicating directly with the Emergency Director,
shift supervisors,

supervisor of operations,and control room operators on matters relating to interpreta-tion of radiation monitoring systems and data.
The Supervisor,

Radiation Protection and Chemistry stated his plan forcontrolling the in-plant radiation hazard was to attempt to personally brief
each individual as to the radiological conditions expected in the area, review
ways to minimize exposure, and provide general precautions or recommendations.
During the period of the accident up to 2400 March 30,

the Supervisor,

Radiation Protection and Chemistry stated that no entries were made into areas
for the protection of vital equipment that might have warranted an emergency
dose.

The planned dose commitment per entry was not expected to exceed 1.5

rem.
Because of increasing air activity, the radiation/chemistry techniciansat the step-off-pad access to the auxiliary building evacuated the area andrelocated to the Unit 2 control room about 1100 (Int.
rate meter with (RM-14/HP-210),

51).

They set up a countLudlum beta scaler, and a couple of E-520s andRO-2s monitors on a table in the control room.
Evacuation of the technicians left no positive control over entrance tothe auxiliary building.
About this same time all nonessential personnel were evacuated from thesite.

Radiation Protection Foreman A and several technicians proceded to thenorth auditorium and surveyed the individuals assembled there for contamination
prior to release (Int.

7, 51).

Other individuals,

including Chemistry Foreman

B surveyed vehicles at the gates as they left (Int. 23).

At about 1100, word

was passed to have all vehicles and personnel proceed to the 500 kV station
for survey and release.

This assembly area was established at the directionof Radiation Protection Foreman D and was used to monitor and decontaminate

11-3-45personnel (Int. 18).
This represents a departure from the Emergency Plan since
this location differs from the two "washdown areas" designated in implementing
Procedure 1670.6.
3.2.4

In-plant Radiation Protection
This section presents a review of radiation protection,

during the period 1100 March 28 to 2400 March 30.

including examples,

Information presented isbased on interviews with each member of the radiation protection staff.
During this period, the RWP procedure was not used (Ref.
each entry into the auxiliary building was made.

159).

No log of

No temporary pocket dosimeter

dose accum.<7ation system was put into effect for entries into high radiation
areas such as the auxiliary building (Int. 133).
Records of dose rate surveys
in the auxiliary building were made at times, but these records were not maintained (Int.

133).

Surveys of personnel contamination were made,one instance were records maintained (Int. 23).

but in only

High radiation areas were notcontrolled in accordance with Technical Specification 6.12.
Station Radiation Emergency Procedure 1670.2,"Site Emergency Procedure",

Revision 9, dated November 22, 1978 defines Emergency Repair Party duties asfollows:"4.4.2.3

Specific duties of the ERP will depend upon the nature of theemergency.

However,the goal, of the ERP is to correct any

malfunction and return the plant to normal operating condition
as soon as possible."

Entries into the auxiliary building to operate valves,

pumps,

breakers;

to perform. inspections and maintenance; and to troubleshoot or repair are considered to have been Repair Party activities.

In the absence of documentation,

several of the entries will be discussed; however,

the investigators may not

be aware of all entries into the auxiliary building.

11-3-46

3.2.4.1

About 1100 March 28 Shift Foreman A entered the 305-ft elevationof the auxiliary building to close electrical breakers that permit operationof the decay heat suction valves (Int.

107).

The entry was made alone.

tective clothing and a pocket dosimeter were not worn.

Pro-

This individual carrieda survey meter, which he remembered indicated exposure rates in his work areaof 2 R/hr.

He stated that his entry took about 5 minutes,

building he was found to be contaminated.

and on leaving the

Review of the individual's TLD results

indicate that a dose of 550 mrem gamma and 660 mrem beta may have resulted fromthis entry (Ref.

160).

No records of the survey of his personnel contamination

were maintained.
3.2.4.2

Between 1100-1600 March 28 the Supervisor,

Radiation Protection

and Chemistry told Radiation/Chemistry Technician U to escort two individuals
into the auxiliary building to check valves and close the containment building
spray pump breakers (Int. 78).
workers,

The two individuals were not shift maintenanceand neither had received emergency Repair Party Team Training as requiredby Procedure 1670.9.
They dressed in full protective clothing, which included cloth coveralls,
paper coveralls,

plastic and rubber boots, gloves,wore self-contained breathing devices,logged.

They

TLDs, and 0-200 mR pocket dosimeters.

They did not wear high range pocket dosimeters.
an RWP nor was it

hood, and rainsuit.

The entry was not made underThe radiation/chemistry technician carried a Teletector.

The three individuals entered the auxiliary building on the 305-ft elevation
where a survey of the seal injection filters was made.

They spent about 10 minutes

attempting to locate the decay heat valves of concern.

Not finding these valves,

they proceeded to the 328-ft elevation via the stairway near the elevator.
Exposure rates increased rapidly as they entered the 328-ft elevation.

Theradiation/chemistry technician terminated the entry when the exposure rate reached
100 R/hr.

The radiation/chemistry technician did not document his survey results,

but remembered the following exposure rates:

11-3-47

Elevation
(ft)

Area

Exposure Rate

Preaccident Exposure
Rate(mR/hr) Ref.179<1

Seal injection filters

281

50-75 R/hr at 1 ft

Hallways

281

2 R/hr<1

Decay heat room281

50 mR/hr<1

Top of stairwell by
elevator just before door

281

1 R/hr<1

Just inside door fromwell by elevator328

5 R/hr

<1

Near fuel handlingbuilding supply unit

328

100 R/hr<lNear auxiliary-building
supply unit

328

>1000 R/hr

<1

On exit from the area, the individuals noted that their pocket dosimeters
were off scale.

Two individuals returned to Unit 2 control room, the radia-

tion/chemistry technician went to Unit 1.
A review of TLD data indicates 560 mrem and 440 mrem gamma for the two
individuals and about 1.8 rem gamma for the radiation/chemistry technician.
About 1100, USNRC Inspector H arrived in the Unit 2 control room (Int.
40).

He was present in the control room until about 1800.

In an interview

with the investigator, he described conditions during this period.

His de-scription is summarized below:a.

The control room was very busy.

b.

The door to the control room was continuously being left open.

c.

People in the control room were wearing protective clothing.

11-3-48

d.

The auxiliary building had high airborne and radiation levels.

e.

No positive control was established over entry into the auxiliarybuilding.

f.

RWPs were not being used.

g.

An auxiliary building entry log was not being maintained.

h.

Survey results were not being documented.

i.

The Supervisor,

Radiation Protection and Chemistry was involved

with operations in evaluating valve lineups and their effect.

j.

The control room was not equipped with an operational continuous
airborne activity monitor.

k.

The control room appeared to be part of the controlled areafrom a radiological point of view.

1.

Respiratory protective devices hindered communication.

m.

Exposure rates in the control room were measured as high as 10 mR/hr,with beta-gamma survey instruments.

3.2.4.3.

Shift Supervisor C and Radiation/Chemistry Technician J entered

the auxiliary building at about 1300 to transfer radioactive water (Int.
51).

13,

Both dressed in full protective clothing and wore self-contained breathing

11-3-49

devices,

high-range dosimetry, and TLDs.

ing, their radiation survey meter (RO-2)

Upon entering the auxiliary buildpegged full scale (5 R/hr).

Both with-drew to the health physics area and waited until a Teletector was available.
When the Teletector arrived from Unit 1, the two individuals reentered the 305-ft
elevation and proceeded to the radwaste operating panel.
rates were 8-10 R/hr.
to transfer water.

The general area dose

The shift supervisor was unsuccessfulin starting pumps

On exiting the area, a survey revealed both were contaminated.

No records of this survey were maintained.

Their high range pocket dosimetersindicated about 1 R of exposure.
3.2.4.4.

In the afternoon on March 28, Radiation/Chemistry Technician H

and an electrician entered the 328-ft elevation of the auxiliary building to
make an electrical breaker lineup (Int.

49).

No RWP,

survey data were reviewed prior to the entry.

no log of entry, and no

They had difficulty locating

self-contained breathing devices and could not find rainsuit hoods.

They entered

the building and proceeded to the 328-ft level to make the breaker alignment.
The following survey data was taken with a Teletector and recalled by the technician:Elevation (ft)Exposure
Rate (R/hr)

Preaccident Exposure
Rate (mR/hr)

305

5-7 R/hr<1

to 328-ft level305

10 R/hr<1

Top of west stairs

328

20 R/hr<1

Motor control centers328

50-100 R/hr

<1

Motor control centersaverage328

20 R/hr<1

Area
Entrance to auxiliarybuilding
Bottom of west stairs

They did not finish the job because they ran out of air.
building and returned to the Unit 1 decontamination area.

They left the-The radiation/ chem-

istry technician stated that he was so contaminated that he pegged the RM-14/HP-210

11-3-50

monitor (50,000 cpm) at a distance of 3 feet.
E-520 on contact with his head.

He measured 35 mR/hr with an

Both men showered,

but the radiation/ chemistry

technician had residual contamination of 1500 cpm on his hair.

No dose rateor contamination survey records were made.
The electrician received a gamma dose of 1.33 rem, the Radiation/Chemistry
Technician a dose of 1.76 rem by TLD data.
3.2.4.5.

On the afternoon of March 29,

Radiation/Chemistry Technician Q

was instructed by a radiation protection foreman to lead a Repair Party Teamconsisting of about four auxiliary operators (Int. 142).
According to the Emergency Plan,

auxiliary operators are not designated as Repair Party Team members.
This team was to place plastic sheeting over the water on the floors in theauxiliary building on the 281-ft elevation.
any radiation survey results to review.

Prior to entry, they did not have

However, they were told that exposure

rates were 10 R/hr near the sump drain and that 6 inches of highly contaminatedwater was on the floor. The team dressed in full protective clothing, including
self-contained breathing devices operating in the demand mode.
and low-range pocket dosimeters and TLD badges.

They wore high-No extremity monitoring wasworn on their feet or ankles.
They entered the 281-ft elevation by the elevator and found exposure ratesof 2 R/hr and 10 R/hr near the sump drain. The floor had spots of standing
water, some up to 6 inches deep.
They covered most of the floor, using three
to four, 80-pound rolls of plastic.

The job took about 20 to 30 minutes.

No air samples were taken in the area during this work.

On exit from thearea and removal of protective clothing, no one was found to be contaminated.
The radiation/chemistry technician did not log his pocket dosimeter reading
for this entry. He recalls rece iving a dose of about 1 rem, and stated that
his TLD also indicated a dose of about 1 rem.

11-3-51

Review of this individual's TLD data for the period of March 1 to March
29 indicates a dose of 0.44 rem.

The individual stated that all team members

had received a dose of about 1 rem.
indicates a dose for the period

Review of one known team member's TLD data

March 29 to March 30 of 1.5 rem gamma and 2.42

rem beta.
3.2.4.6
decided,On the evening of March 28, two radiation/chemistry technicians

based on Unit 1 control room conversation,

to enter the auxiliary building

and pull the HP-R-219 charcoal and particulate samples for analysis and to collecta water sample from the floor of the 281-ft elevation of the auxiliary building(Int. 49 and 5Q).

Radiation/chemistry technicians E and W were under the impres-sion that the NRC wanted the samples.

They announced to the Radiation Protection

Supervisor that they would collect the samples.

He was in agreement.

Theyplanned the job for about 1 hour, deciding to carry a lead shield with whichto wrap the HP-R-219 cartridge.
They were unaware of previous entries into the building and did not review
any survey information.

They stated that at the time of their entry into theauxiliary building there was no control over entries, either in the form of aperson or locked door to restricted access to the building.

Prior to the entry,

they went to the Unit 2 control room but could not locate the Supervisor,
tion Protection and Chemistry or his alternate.

They entered the area dressedin full protective clothing with the exception of rainsuit hoods,
could not find.

Neither wore extremity monitoring.

Teletector and the other an E-520.

Radia-

which they

One technician carried a

They did not record their survey results;

but recalled them as follows:

Elevation (ft)Exposure
Dose Rate (R/hr)

HP Labby hand and foot monitor

305

2

<1

Hallways305

Up to 15

<1

On floor near HPR219

328

50-70

<1

Area

11-3-52

Preaccident Exposure
Rate (mR/hr)

Just inside the auxiliary building doors, the technicians split up.

The

one with the Teletector went to the 328-ft elevation to sample HP-R-219.

Theother technician's E-520 pegged full scale (2 R/hr) as he passed near HP-R-227
on the 305-ft elevation.

He did not stop, and the meter remained off scale as

he went down to the 328-ft elevation to scoop a water sample from the floor.
While on the 281-ft elevation,teries to see if

he found a Teletector.

He checked the bat-it worked but did not take an exposure rate measurement.

He

scooped a 100 ml sample and left the building.
They returned to the Unit 1 health physics area.
bottle read 40 mR/hr at contact.
100 mR/hr.

An exposure rate on the

A dose rate on one technician's head read

Both were contaminated to about the same extent on their heads andstated they were able to decontaminate themselves.

No survey records were made.

Failure to leave the area when the individual's dose rate instrument pegged
full scale represents on unsafe radiation protection practice, which can resultin unnecessary exposure.
Radiation/Chemistry Technician E estimated his total time to change thecharcoal cartridge (the particulate filter was still missing) was between 3.5
and 5.5 minutes (Int.

50).

He stated his TLD indicated a dose of 870 mrem.

The other technician stated his TLD read 720 mrem.
At 1930 March 28, the NRC Mobile Laboratory arrived at the Observation
Center (Int.

72).

Two in-plant air samples were analyzed for iodine activity

between 1930 and 2400.
detector.

These charcoal cartridges were counted with a Ge(Li)Neither indicated above the minimum detectable activity for iodine-131(2 E-8 pCi/cc).
3.2.4.7.

At about 2100 March 28, Auxiliary Operator I was directed by a

Shift Supervisor B to repressurize the core flood tanks with nitrogen (Int.
114).

He understood this action to be necessary but was not told it was for

11-3-53

the protection of vital equipment.
for this entry.

No one told him how much dose was permitted

He stated no surveys were available for review, so he listened

to radiation protection personnel discussing dose rates and concluded he might
be entering fields of 100 to 150 R/hr.

He did not specifically tell the radiation

protection personnel he was going into the building and they did not ask.

He

did not have an RWP and an entry log was not being used.
The auxiliary operator dressed himself in protective clothing but did notwear a rainsuit.

He did not have a high-range pocket dosimeter.

He donned aself-contained breathing apparatus at the step-off-pad outside the entrance tothe auxiliary building, picked up a Teletector, and entered the building.

There

was no one there to challenge his entry and no locked access to restrict entry
into the building at that time.
He entered on the 305-ft elevation,noting an exposure rate of 20 R/hr,

proceeded upstairs to the 328-ft elevation, then back through the fuel handlingbuilding to near the penetration room door where the core flood nitrogen valves
100, 30,

are located.

He recalls dose rates of 70,

valve area.

He estimates it took him between 5 and 10 minutes for the total

and finally 10 R/hr in thejob.
On leaving the building,scale.

he found his 0-200 mR pocket dosimeter was off

He returned to Unit 2 control room and, on entering, alarmed an RM-14/

HP-210 count rate meter on the table.

Assuming his clothing to be contaminatedwith gaseous activity, he decided to return to the auxiliary building and attempt
to transfer water to Unit 1.
A radiation/chemistry technician observed Auxiliary Operator I alarm theRM-14/HP-210 probe and told him he would have to go to Unit 1 and decontaminate.
This comment reaffirmed his decision to reenter the auxiliary building at thattime before decontaminating himself since he thought he would become contaminatedagain.

He told one shift foreman that his pocket dosimeter had gone off scale.

Subsequently, he planned the liquid radwaste transfer with Shift Supervisor D
but did not inform him of his previous exposure.

11-3-54

Shift Supervisor D was concerned about exposure for the transfer and discussed possible low background areas near the radwaste panel.

Radiation protection

personnel did not participate in these discussions.
The auxiliary operator searched out protective clothing but was unable tolocate a high-range dosimeter either in Unit 1 or 2.

He asked Radiation/ Chemistry

Technician K for one and discussed his planned entry into the building.
At this point he rezeroed a 0-200 mR dosimeter, dressed in the same manner,and once again entered the auxiliary building unchallenged.

This time he walked

the 305-ft elevation through the switch gear room, hoping the exposure rates
would be lowered.

He remembered 30 R/hr as he passed the makeup tank room and

10 R/hr at the radwaste panel.
room to reset a trip.

The pump would not start so he called the control

He then waited in the model room (3 R/hr) for 2 or 3

minutes before attempting to again start the pump.
energized, so he checked the breakers,
building.

The pumps still did not

found they were tripped, and left theHe estimates about 10 minutes for this entry.

On exiting the area, he discovered his pocket dosimeter was again off scale.
He undressed and returned to the Unit 2 control room.
RM-14/HP-210 probe.
gone off scale.

He again alarmed the

He informed the Shift Supervisor D that his dosimeter had

The shift supervisor told him he could not be used any more

and to return to Unit 1 and decontaminate.

The auxiliary operator was about

to shower when a radiation chemistry technician told him that, because of highairborne activity, he should wear a mask while showering.

The technician did

not survey the operator.
After decontamination,dose of 3.170 rem.

the operator's TLD was read, indicating a gamma

His Form NRC-5 equivalent indicated a total first quarter

gamma dose of 3.87 rem.

The licensee submitted a report prusuant to 10 CFR

20.405, dated May 1, 1979,

indicating that this individual had received a firstquarter whole body dose of 3.870 rem.
This entry examplifies the lack of good health physics practices.

11-3-55

3.2.4.8.

At about 2200 on March 28, Shift Foreman B requested Auxiliary

Operator C to open the auxiliary spray valve DH-V-187 located on the 328-ft
elevation of the auxiliary building in an elevated pipe run between the fuelhandling iand reactor buildings (Int. 104).

In order to operate the valve,one

must start from a platform and climb up a hand rail onto a pipe run which islocated over an open space that drops to the 281-ft elevation.
views,

itBased on inter-appears that this area was not surveyed prior to this entry.

Thefollowing quotation is provided to describe execution of this Repair Party function:"Then the foreman came up to me...
heat...

and when we go into decay

we have an aux spray line we can set up from decay heat

for the pressurizer.

So my job then was to establish the lineup

for auxiliary spray in the pressurizer.... So I went down to our
HP area and there was no Telectectors there at that time.

There

was E-520s but they were all pegged so they were useless.

So I

went over to Unit 1 HP and it was contaminated.
Teletector there.

There was no

There was nothing there I could use.

So, onthe way back, I just made up my mind that I was not going touse a Teletector.

I knew where I was going and I knew which

valve I had to open.

It

was a locked valve.

I had no intentionsof entering any rooms; so I thought it'd be --I had no idea atthat time what the radiation levels were.

As I got there again,

an HP fellow came out of the aux building.

He told me the radia-

tion levels in the hallways were 100 R/hr, and I asked him atthat time for his Teletector.
not surrender it.
shielding.

But he still needed it and could

So I just told myself; time, distance,

In other words,

bag itthrough the hallways, andonce I got up in the area of the valve....
shielded-I thought I'd be all right.

Scott Air Pacs.
time,

and

it was pretty well

Then the game was getting

It was quite a few of them being used at that

so I used the bottle that was half full.

11-3-56

There was noone there to assist us; no one there to help you.

You were pretty

well on your own ....

and I got myself dressed with what we had

there and went in.

I climbed up to the valve.

of running through the hallway, though.

I made the mistake

And I ended up hyper-

ventilating and suJffring from claustrophobia as a result.

But

I got that under control after a bit and then got up to the valve....
got it open.

My alarm is ringing.

And then I left.

Then Iran out of air, so I had to remove the mask just, oh, maybe 25
to 30 feet before the door where you get out.
there and got undressed.

And everything was contaminated.

was no clean area that you could use.
the same thing with me...
they had in there...

So I got out of
There

So I went upstairs andas soon as I opened the door the monitor

the frisker in there went off.

So I told

them I had the valve open and left."
This individual wore a TLD and 0-200 mR pocket dosimeter.
lenged access to the auxiliary building.

He had unchal-

He did not inform anyone as to thetime he was entering the building.
Upon returning to the Unit 2 control room and opening the door, the contamination on him immediately caused the RM-14/HP-210 to peg full-scale.
to Unit 1 and showered.

He proceeded

He could not estimate his contamination because theRM-14/ HP-210 there was alarming.

He returned to the Unit 2 control room andfound he was still contaminated but that it did not exceed 5000 cpm on the RM-14/
HP-210.

He then told the shift foreman that he would not be of any further

use, since he was "pretty sure he was overexposed."
and he remembers a dose of 1.25 rem.

His TLD badge was readA review of his TLD data sheet is notconsistent with his recollection of this entry and needs to be resolved by thelicensee.

The TLD data printout of his exposure for the period of March 29 toMarch 30 indicates a dose of 0.655 rem.

The licensee was informed of this finding.

11-3-57

This individual had not been designated as a Repair Party Team member andtherefore did not receive Repair Party Team training.
Radiation Training" according to Procedure 1690.

He had received "Advanced

His entry demonstrates a lackof implementation of basic radiation protection practices that could have resulted
in serious consequences.
3.2.4.9.

On the morning of March 29, a radiation chemistry technician

made a survey of the auxiliary building.

He prepared a survey map of his findings.

These records were not maintained; however,

the table below summarizes the Teletector

readings he recalled.

Area

Elevation (ft)Exposure Rate

Water on floor

281

50-60 mR/hr

Bleed tank door (closed)

281

35-40 R/hr

Bleed door (open)

281

300 R/hr

Seal injection filters

281

600 R/hr

B makeup pump room justinside door

281

100 R/hr

B spray vault

281

100 R/hrA bleed tank (inside door)

281

200-300 R/hr

Streams (4 or 5)281

1000 R/hr

General hallways

281

300-400 mR/hr

Makeup tank door

305

40-50 R/hr

Streamers (door to makeup
and purification valve room)

305

1000 R/hr

Outside valve alley305

15 R/hrDoor to makeup purification room305

50-70 R/hr

11-3-58

This radiation/chemistry technician received a dose of 1.2 rem by TLD asa result of surveys he performed on the first 2 days.
On the afternoon of March 29, additional supplies of self-contained breathing devices, face masks, and combination particulate and charcoal cartridges
began arriving on site (Int.

155).

Radiation protection support personnel werearriving with their own instrumentation at the Observation Center (Int. 54).
With few exceptions, these individuals were not used in-plant until after March 30.
A consultant's mobile laboratory arrived at the Observation Center about
1400 and began setting up. The mobile laboratory did not begin counting samples
until 0030 on March 30,

1979.

A review of all air samples counted during theperiod 0400 March 28 to 2400 March 30 by NRC and vendor mobile laboratories
indicates the highest iodine-131 concentration to have been 5 E-9 pCi/cc at
2203 on March 30 in the primary chemistry laboratory.

A total of 48 air samples

were gamma scanned (33 from both Unit 1 and 2 control rooms).

The highest con-centration in either control room was reported as 3 E-IO pCi/cc in Unit 1 at
1255 on March 30.

The remaining samples were taken in the control,

and turbine buildings.

service,

The highest level of iodine-131 activity measured was

1 E-9 pCi/cc in the Unit 2 access control area.
During the period 0800 March 28 through the afternoon of March 29,

rLDbadges were read on an as requested basis on site at the normal location.

The

TLDs were read by any radiation/chemistry technician or radiation protection
foreman that was available at the time.

In some instances,

technicians acted as Repair Party monitors (e.g.,radiation/chemistry

see paragraph 3.2.4.5) andthen read the TLDs worn by themselves and the other team members.
Procedure 1642 states that a TLD reader calibration test should be doneon or near each badge exchange o'r reading period (Ref. 161, 181, 182).
This
test was not performed prior to reading TLD badges during the incident.
Thelast recorded test was performed on February 28,
acceptable results.

11-3-59

1979.

That test indicatesSometime on the afternoon March 29, the TLD reader and support equipment
was moved from its normal location in a trailer on site to the Observation Center.
The equipment was placed into operation and some TLD badges were read.

No records

indicate that a reader calibration test was performed.
The TLD system was operated until about 0700 March 31 by Radiation/Chemistry
Technician C who had received two hours of on-the-job training on June 6, 1977,
and had not operated the equipment in about a year and a half (Int. 82).

He

did not have a copy of the procedure for either operation or documentation of
TLD results available.

He performed this job for about forty-eight continuous

hours without sleep.
His work included zeroing all the April TLD badges that had been stored
on site in the TLD trailer and reading TLD badges as they were turned in.

Heindicated that he had little or no help through this period.
Review of the licensee's TLD exposure data (Form NRC 5 equivalents)

dated

May 21 and July 8, 1979 for 200 individuals indicated several errors. Of the
200 individuals,

ten badged employees did not have entries in their dose records

for the period March 1 - 31, 1979.

Five of the 10 individuals would have received

dose during this period by virtue of their job functions.

In addition, doses

recorded for four individuals appeared inconsistent with information provided
in interviews (Int.

35, 49, 104, 114,

that changes to individuals'

Ref. 160).

The review also indicateddose records were made,but no record of the required

evaluation leading to the changes were maintained (Ref.

163, 164).

The licensee

was informed of these records discrepancies.
The licensee's method of evaluating skin dose during the period March 28
to 30 appears in error.

No dose contribution from gamma radiation was indicatedon the TLD data sheets provided to the NRC on May 21, 1979 (Ref.
The licensee dosimetrist was informed of this finding.
made to individuals'

Corrections are being

dose records as errors are identified.

11-3-60

163, 164).

Using the licensee's TLD data, an investigator estimated a total dose of
104 man rem for 270 individuals monitored during the period March 28 throughMarch 31,1979.

During the period of this investigation,

numerous individuals in the auxi-

liary building were exposed to high concentrations of noble gas activity.
several occasions,Onthe whole body exposure rate as measured by Teletector was

remembered to be 100 R/hr.
xenon-133 and xenon-135.

The majority of this radiation was probably due to

Review of data from individuals'

TLDs and background

TLD badges indicated several apparent inconsistencies.
One such inconsistency is discussed in Section 3.2.4.5 of this report.
In that example,

two individuals who were similarly exposed had different dosi-

metry results. According to TLD results one individual received 0.44 rem gamma
while the other received 1.5 rem gamma and 2.42 rem beta dose.

Since no records

were maintained of how individuals were dressed or where their TLDs were worn,
this investigation cannot accurately explain this inconsistency.
In other instances,some background TLD badges measured beta radiation

while others, exposed in similar surroundings,

measured none.

Orientation ofthese badges in relation to the source of radiation is unknown.
A possible source of measurement error applicable to all cases involves
loading of the TLD chips into the holder.

When loaded into the TLD badge, thechips can easily be positioned such that the capability to measure beta radiationislost and the response to low energy gamma radiation is diminished.

Thisoccurs when the TLD card is placed in the badges with chip G2 in front of thebeta window or when the card is placed such that both chips are shielded by
the plastic badge case.
An investigator discussed these problems with the Supervisor,

Radiation

Protection and Chemistry, who stated additional evaluations are in progress inan effort to resolve dosimetry questions.

11-3-613.2.4.10.

The Supervisor, Radiation Protection and Chemistry instructed

Radiation Protection Foreman D to take a reactor coolant sample on the afternoon
of March 29 (Int.

18).

Radiation Protection Foreman D and Chemsitry Foreman B

decided they would take the sample (Int. 23).

They were aware of the need todetermine the boron concentration of the coolant.
any other specific analysis to be performed.

They were not informed of

They discussed the assignment

for a period of less than one hour, deciding to use a three-man approach andtemporary shielding to minimize their exposure.
cruited to perform the boron analysis (Int.

Chemistry Foreman C was re-132).

The individuals dressed in protective clothing, which included several
pair of coveralls,
a rain suit.

shoe covers,

and three pair of gloves.

One individual wore

Two wore full face respirators with iodine adsorbing cartridges.

The other wore a full face mask with particulate filter.

The chemistry fore-

man's TLD and pocket dosimeters were worn inside the protective clothing.

The

high-range pocket dosimeter was not accessible to be periodically checked.

No

one wore extremity monitoring on their hands.

The radiation protection foreman

wore a high-range pocket dosimeter taped to his forearm.
during collection of the sample.

This was knocked off

Prior to taking the sample,

they built a leadshield in the primary sample lab and in the nuclear sampling room.
Prior to recirculation of the sample,

the room was surveyed with a Tele-

tector and found to have an average exposure rate of 8 R/hr.
sample room area monitor,

RM-G3,

the nuclear

indicated 1 R/hr at this time.

Just prior to initiation of recirculation flow, Chemistry Foreman B was
asked to collect a Unit 1 reactor coolant sample for boron.

He collected thissample but did not check his high-range pocket dosimeter prior to starting on
the Unit 2 Sample.
The valve lineup was made for Unit 2.

When flow was initiated at about

1545,

RM-G4 the hot machine shop area monitor increased from 850 mR/hr to 2

R/hr.

The RM-G3 monitor increased from 1 R/hr to 6 R/hr (Ref.

was allowed to recirculate for about 45 minutes.

11-3-62154).

The sample

Several times in that period

the "drag" valve had to be adjusted in order to maintain flow.

The radiation

protection foreman entered the nuclear sample room and attempted to collect asample.

When he opened the sample valve he observed an "alka seltzer-like"fluid.

He closed the valve and checked the dose ratenear the sample line andfound itread 90 R/hr.

They reviewed the valve lineup and asked the control room to cycle thevalves.
They reentered the nuclear sample room and collected about 300 ml ofsample in a one liter bottle.

The one-liter sample bottle was handled un-shielded at contact with their gloved hands.

One hundred ml was.poured into a

graduated cylinder and the exposure rate measured.

At contact, the graduatedcylinder caused the Teletector to deflect full scale (1000 R/hr).

The graduatedcylinder was set on the floor and the exposure rate measured 400 R/hr at 1 footand 10 to 15 R/hr at 3 feet.
hood.

Meanwhile,

The graduated cylinder was returned to the sample

Chemistry Foreman B had drawn an additional 40 ml of reactor

coolant into a 100-ml breaker and carried it

into the primary sample lab.

primary sample laboratory area radiation monitor (RM-G2),
to 800 mR/hr as he entered the room.

increased from 20 mR/hr

The sample was placed behind a lead shieldand pH adjusted by addition of hydrochloric acid.
Foreman C for the boron analysis.

The

Five mls were given to Chemistry

Chemistry Foreman B returned to the nuclear

sample room and poured the 100 ml sample back into the sample bottle.

The radi-

ation protection foreman took the 40-ml sample from the primary sample laboratoryand dumped it

down the nuclear sample room sink.

Radiation Protection Foreman D and Chemistry Foreman B removed their protective clothing.

Each checked their high-range pocket dosimeters.

The radiation

protection foreman's read about 0.8 R, the chemistry foreman's read about 4 R.
They left for the process center to be surveyed.
RM-14/HP-210.

On arrival,

both alarmed theSomeone surveyed them with an E-520 and found the radiation protec-

tion foreman's wirst read 200 to 250 mR/hr and the chemistry foreman's finger
read 50 mR/hr at contact.

Other areas of their bodies were also contaminated.

11-3-63

Both returned to Unit 1 to decontaminate.
the boron analysis,

Chemistry Foreman C, after completingconcluded that he was contaminated.

He immediately began

decontamination in the shower prior to the return of the radiation protection
foreman.
These three individuals attempted decontamination for the next several
hours.

They were periodically resurveyed at the 500 kV substation.

All threewent home with areas of their body having maximum fixed contamination of up to
50 mR/hr on a finger and 30 mR/hr on wrists.

Each individual supplied urine

samples the next day which were not analyzed by the licensee as of May 24, 1979.
The licensee submitted a report pursuant to 10 CFR 20.405 dated May 1,
1979 indicating that Chemistry Foreman B had received a whole body gamma doseof 4.115 rem and beta dose of 0.090 rem for the first quarter of 1979.
The NRC has evaluated the dose to the extremity and skin of each invidual
involved in this sampling effort.
as Appendix B.

This evaluation is attached to this report

One chemistry foreman received an extremity exposure in therange of 50-147 rem and a skin dose of approximately 6 to 13 rem for the firstquarter of 1979 and a skin dose of 4 to 19 rem for the second quarter of 1979.
Radiation Protection Foreman D received an extremity dose in the range of 44
to 54 rem for the first quarter of 1979.
The following reactor coolant activity results weresee (Ref.

165).

reported to the licen-

The results were developed from an undiluted sample by a venderlaboratory.
iodine-131
iodine-133
strontium-89/90
cesium-134
cesium-136
cesium-137
barium-140

1.3 E 4
6.5 E 3
5.3
6.3 El1.8 E 2
2.8 E 22.0 E 2

11-3-64

pCi/ccpCi/ccpCi/ccpCi/ccpCi/ccpCi/ccpCi/cc

In addition, the NRC reviewed the exposure to other personnel who handled
this sample.
necessary.

The licensee was informed that several exposure evaluations wereOn the night of March 29 and the early morning of March 30,

Radiation

Protection Foreman B and Radiation/Chemistry Technician L removed a 1 ml sample
for shipment to a vendor laboratory (Int. 21,

58).

Radiation/Chemistry Techni-

cians B and V diluted a 1-ml sample reading 11 R/h with a Teletector into 1
liter of demineralized water and found it

read 1 R/hr (Int.

98, 106).

One mlof this sample was diluted in I liter of demineralized water and that read 50-100
mR/hr.

One ml of that sample diluted into 100 ml of demineralized water

1 mR/hr.

read

Five samples of this diluted coolant were packaged in vials for analysis

on site (Int. 98).
During this operation no air samples were taken,
was worn,no records of surveys were maintained,

no extremity monitoring

nor were records of personnel

contamination made.
A review of the TLD data report issued July 5,
dose results for the period March 1 through 31,

1979 does not indicate any

1979 for Radiation Chemistry

Technician L who handled the sample.
3.2.4.11.

The evening of March 29, Nuclear Engineers J and E, discussedwith the Supervisor, Station Operations the importance of finding out if there
were any major leaks in the auxiliary building (Int. 33, 35).

They decided totour the area since both were familiar with radwaste operations.

They werebriefed by Radiation Protection Foreman B, dressed appropriately for the degreeof hazard, and were provided an E-520 and high-range gamma instrument.
Shortly after entering the 305-ft elevation of the building, the high-range
gamma instrument failed.

They continued on together noting that the E-520 fre-quently pegged full scale (2 R/hr).

They operated a valve on the 305-ft elevationand then toured the 281-ft elevation.
the 281-ft elevation,
panel,

they split up,

Again, the meter frequently pegged.

On

Engineer E going back up to the radwaste

Engineer J touring near the south access corridor (a previously unsurveyedarea) and the bleed tank doors.

Engineer J remembered his instrument was pegged

11-3-65

in these areas.

On exit of the building,the individual that had the E-520 in

the basement found his high-range pocket-dosimeter read about 3 Rem.

The indivi-

dual that went to the radwaste panel recalled his pocket dosimeter read lessthan 1 R.
Both returned to the control room and were debriefed by the Supervisor,
Radiation Protection and Chemistry and Supervisor, Station Operations.
TLDs were read and indicated 3.14 rem and 0.170 rem.

Their

The 3.14 rem, when added

to Engineer J's previous quarterly exposure resulted in a total first quarter
whole-body exposure of 4.175 rem.
10 CFR 20.405 dated May 1, 1979,

The licensee submitted a report pursuant toindicating that this individual

had received

a whole body gamma dose of 4.175 rem during the first quarter of 1979.
This entry demonstrated a lack of implementation of basic radiation protection training that could have resulted in serious consequences.
3.2.4.12.
On March 29,

Not all Repair Party activities were handled as described above.

a Repair Party activity to install a temporary pipe permitting

venting of the waste gas decay tank to the reactor building was installed. Thisjob was planned in advance.

The Supervisor Radiation Protection and Chemistrybriefed the Repair Party Team leader (Maintenance Foreman G).

The individuals

were properly dressed and personnel pocket dosimetry records were kept.

The

job was completed within the dose allotment and no one became contaminated (Int.
143).
On March 30 about 0500,

NRC Inspectors H and 0 met with the Superintendent,

Technical and Administrative Services and Unit 1 Superintendent Technical Support
to express concerns in regard to the radiation protection program (Int. 40).
These concerns included:

adequacy of personnel surveys,

areas, documentation of surveys,

contamination control,

toring.

11-3-66

control of high radiation
RWPs,

and effluent moni-On March 30, more control was established over in-plant radiation hazards.
Radiation/protection technicians were being used on site to establish control
points.

Temporary dose accountability forms were implemented that evening at

the site access point (Int.

106)

The whole-body counter was in operation andindividuals were being counted.
On the afternoon of March 30, the NRC became more actively involved in
the radiation protection program.

Five NRC radiation specialists were assigned

to each of the three shifts providing around the clock coverage.

In addition,

two NRC health physicists provided technical guidance.
3.2.5.

Summary of Radiation Protection Findings

3.2.5.1

Facilitiesa.

Airborne radioactive materials spread from Unit 2 into Unit 1.

b.

High radiation areas were created in Unit 1 from unshielded Unit
2 reactor coolant sample lines.

c.

The combination of unshielded sample lines and poor ventilation
resulted in the loss of:

a.
b.
c.

d.

Counting room equipment
The primary ECS
In-plant personnel decontamination facilities

Airborne radioactivity monitoring devices were not available onsite.

These instruments were needed in the control rooms andlaboratories.

These devices would have helped to provide infor-

mation necessary to make decisions on the need for respiratory
protection.

1173-673.2.5.2

Equipmenta.

Lapel air samplers were not available.

b.

Appropriate extremity monitors were not available.

c.

The number of wide range beta-gamma survey meters (0-1000 R/hr)
available was insufficient.

No beta-gamma survey meters with arange exceeding 1000 R/hr were available for persons enteringradiation fields in that range.
d.

An adequate supply of high-range pocket dosimeters were not available.

e.
3.2.5.3

Alarming personnel dosimeters were not available.
Procedures

The Emergency Plan Implementing Procedures do not adequately address sustained in-plant radiation hazards.
3.2.5.4

Training

Actions of some workers,

including members of the radiation protection

and chemistry staff, did not reflect comprehension of the following:a.

The need to know exactly when individuals entered and returned fromareas of radiation hazard.

b.

The need to measure and document the airborne radioactivity towhichworkers are exposed.
c.

The need to perform detailed surveys of personnel contamination withappropriate instrumentation.

11-3-68

d.

The consequences of handling highly radioactive material.

e.

The necessity of appropriate personnel monitoring devices.

f.

The methods used to evaluate internal and skin dose.

g.

The limitation of portable survey instruments,

including range,en-ergy response, and significance of open and closed window readings.
h.

The plant systems and components.

i.

Concepts and techniques to minimize dose.

j.

The regulations,k.

The emergency organization.

3.2.5.5

license conditions,and procedures.

Dosimetry

The TLD dose printout for the first quarter of 1979 is not accurate orcomplete.
3.2.5.6

Air Sampling

Air samples collected in other than the auxiliary and fuel handling buildings
did not indicate concentrations of iodine-131

in excess of the values listed

in 10 CFR 20 Appendix B.
A review of the records for these samples does not indicate that air samples
for particulate,

iodine, or noble gas were collected in the auxiliary building

breathing air zone from 0800 March 28 to 2400 March 30,
were made.

11-3-69

a period in which entries3.2.5.7

Respiratory Protection

Since iodine adsorbing respiratory protective devices are not approved
for use pursuant to 10 CFR 20.103(c),the licensee did not have any such devices

on site until sometime on the afternoon of March 29.
The iodine-adsorbing cartridges would have been of value in reducing theintake of iodine if

emergency egress of personnel had been necessary becauseof the development of an iodine release on site.
The quantity of self-contained breathing devices was not adequate,

causing

individuals to do without such devices at times when they should have been used.
The use of a partially filled self-contained breathing device constituted an
unnecessary risk for at least one individual who was alone and under stress ina hazardous environment,

without backup search and rescue assistance.

Whole body count data were reviewed by the investigators.

No apparent

exposures in excess of the limits specified in 10 CFR 20.103 (a) were identified.
However, the investigators are not aware of all entries to high airborne activity
areas and therefore cannot conclude that no excessive exposures have occurred.
3.2.5.8

High Radiation Area Control

From about 1100 March 28 to the afternoon of March 30,

control of accessinto the auxiliary buildings was not maintained to restrict access by unauthorized
or unprepared individuals to hazardous radiation environments.
3.2.5.9

Management Control

During the period 0400 March 28, through midnight on March 29,

licenseemanagement did not make every reasonable effort to maintain radiation exposures
to workers as low as was reasonably achievable considering the emergency conditions at the time and the equipment available for use.
actions were not taken.

11-3-70

Specifically the following

a.

Positive access control was not established to prevent entry of unprepared individuals into hazardous radiological environments.

b.

An effective method was not implemented to inure that all individuals
entering hazardous radiological environments were fully briefed as
to the hazard and the degree of urgency with which the task was tobe performed.

c.

Equipment such as high range pocket dosimeters and survey meters were
not controlled to insure that each individual entering high radiation
areas was provided with the appropriate instruments.

d.

Individuals that became contaminated were not properly surveyed anddecontaminated to insure their dose was minimized.

e.

Planning of those tasks which presented significant potential forradiation dose,such as reactor coolant sampling,

was not reviewed

by a knowledgeable management representative to insure that reasonableprecautions were to be taken.
3.3

ENVIRONMENTAL ASSESSMENT AND PROTECTIVE ACTIONS

3.3.1

Initial Offsite Dose Calculation
Environmental assessment actions began at 0655 on March 28 with the declar-

ation of a Site Emergency.
Procedure,"

Section 4.1.11 of Procedure 1670.2,"Site Emergency

requires the Emergency Director (Station Superintendent) to "Direct

evaluation of off-site monitoring team data and project offsite dose consequencesin accordance with 1670.4" (Ref.

166).

Procedure 1670.4 produces estimates ofprojected thyroid and whole body doses at down wind locations as an end product
(Ref.

167).

1173-71

Nuclear Engineer C, working in the ECC (Unit 2 control room), completed
the initial projection of environmental impact at 0710 on March 28 (Int. 48,
174). A whole-body exposure rate of 40 R/hr was calculated for location W-1l
(Goldsboro), 1.2 miles west of the plant (Int 48, 174,; Ref 168). The locationof survey point W-ll with respect to the plant and the onsite survey is shownin Figure 11-3-2. The calculation was based on the following input parameters:
TABLE 11-3-2
Offsite Dose Calculation Results (Actual)- Procedure 1670.4

Monitor reading

HP-R-214
HP-R-219

300
6 E+5

R/hr
cpm (gas channel)
cpm/min (iodine
channel)Noble gas release rate

HP-R-214
HP-R-219

1325
0.24

Ci/secCi/sec1-131 release rate

HP-R-214
HP-R-219

0.083

Ci/secCi/secAtmospheric dispersion
factor (x/Q)
Wind direction (toward)

2700

Wind Speed

Vent flow rates

4
AH-FR-5720
AH-FR-5284
AH-FR-5286
AH-FR-5659
AH-FR-5063
AH-FR-5075

0014,000mph
cfm

15,00000Projected whole body dose
rate due to noble gases

40,000mrem/hr***

Projected thyroid dose rate
due to iodine-131

8.5

mrem/hr****Not determined.
"*The value used for the atmospheric dispersion factor is not known because theinitial calculation sheet cannot be found (Ref. 168, Int. 48, 174).
The x/Q
value (normalized to a wind speed of 1 mph) required to yield an exposure rate3
of 40 R/hr from a release rate of 1325 Ci/sec is 9.4 E-4 sec/m , using themethods outlined in procedure 1670.4.
This x/Q value appears incorrect for
two reasons:(1) the companion calculation for the offsite iodine concentra3
tion used a x/Q value of 2.5 E-4 sec/m and (2) the specific value 9.4 represents an interpolation between the x/Q isopleths (referenced in procedure
1670.4) that is unlikely to have been made.
Had the same value (2.5 E-4 sec/m 3 )been used for the noble gas calculation that was used in the iodine calculation,
an exposure rate of 10 R/hr would have been calculated for Goldsboro.
***Dose rate is based on the addition of release rates determined from monitor
reading.

11-3-72

ON ISLAND (GE SERIES) SURVEY POINTSAND NEAR-FIELD TLD LOCATIONS

NORTH
GATE

0
r'-

E 4A1

0
0

4

5A1

SHELLEY
ISLAND

CENTER

W-11

12B1

wsw.

r:3

+

LAND OWNI

I-= BY MET ED

E3LANDNO
SBY MET ED

Figure 11-3-2
11-3-73

The investigators have identified several probable errors in the preceding
tabulation of input parameters employed in the offsite dose calculation,

thelargest of which was an error of three orders of magnitude (in the conservative
direction) in the reading of HP-R-214.

The containment dome monitor (HP-R-214)

consists of a Victoreen 845 Series Area Monitor having an 8-decade response_1from 10

to 107 mR/hr (Ref.

44).

The ion chamber detector is encased in a

2-inch lead shield in order to extend the range of the instrument.

The monitor

can be read out in two primary modes:(1) with the range selector switch on"all," the meter readout covers the entire eight decades or (2) with the range
selector switch on 102,103 .....

or 107, any of six 3-decade expanded scale

readouts are available (the switch position indicates the top decade of the
3-decade scale).

The monitor readings contained in Table 11-3-1 and in the text

which follows have not been corrected for the attenuation provided by the leadshielding.

Direct readings from the monitor were used in accordance with Pro-

cedure 1670.4.
Based on a review* of the chart recorder (HP-UR-1901)

for the HP-R-214

monitor, the instrument was actually indicating an exposure rate of 400 mR/hrwhen the engineer read it at approximately 0655 on March 28,

rather than 300

R/hr.**The investigator's reconstruction of this chart is shown in Figure11-3-3.

The probable cause of this error is attributed to the fact that theengineers involved with the offsite dose calculations were unaware of the "expanded scale" feature of the instrument and misread it

(Int.

174).

The mis-interpretation of this monitor reading was an indication of a weakness in thelicensee's emergency plan -training program (see Section 1.2.1).

*A discussion with Maintenance Foreman G indicated that during the period
beginning about March 26,

197g,a measurements program was undertaken to cali-brate the output signal from the HP-R-214 monitor so that the recording of
this signal (eight decade response) on HP-UR-1901 (five decade recorder) could
be interpreted after the chart was removed from the recorder.

These measure-ments were part of the routine calibration program for these monitors.
this program has not yet been completed,

Althoughthe calibration of the HP-R-214 signal

was completed (after March 28) and was used by the investigators to interpretthe chart from this monitor.
*A reading of 400 mR/hr on the expanded scale (l-l03 range) was misinterpreted
as 3 x 10s mR/hr by the individual

reading the instrument.

11-3-74

10

DOME MONITOR HP-R-214
063/28/79

Note: The response of this monitor should
be interpreted only as a relative indication of radiation level in containment.
The actual level is not shown because theattenuation provided by the detector
shield was not accounted for.

105

104

)
C:

Alarm

--

(8,000 mR/Hr)

--

L ac

E 1oa

102

Alert
mR/Hr)(•-'25

10.,111

0625

0635

0645

0655

0705
Time

Figure 11-3-3

0715

0725

0735

The investigators have determined that if the licensee had read the instrumentation in the control room properly and had used the proper atmospheric dispersion factor for Goldsboro, the following would likely have resulted:
TABLE 11-3-3
Offsite Dose Calculation Results(Reconstructed by Investigators)
Procedure 1670.4
Monitor reading

HP-R-214*
HP- R-21 9

400
500
22

Noble gas release rate

HP-R-214
HP-R-219

1.8
6 E-3

Ci/secCi/secIodine-131 release rate

HP-R-214
HP-R-219

6 E-4
9 E-6

Ci/secCi/secAtmospheric dispersion
factor**4 E-4

sec/m3

Wind direction (toward)Wind speed

2700
4Vent flow rates AH-FR-5720
AH-FR-5284
AH-FR-5286
AH-FR-5659
AH-FR-5063
AH-FR-5075

7000
040,50043,00000mR/hr
cpm (gas channel)
cpm/min (iodine
channel)mph

cfm

Projected whole body dose
rate due to noble gases

2.2

mrem/hr***

Projected thyroid dose rate
due to Iodine-131

0. 1

mrem/hr****The release rates shown for the HP-R-214 monitor assume a reactor building
pressure of 56 psig following a LOCA (Int. 48).
The actual reactor building
pressure at 0700 was about 1 psig; hence, the corresponding release rate wasmuch lower than assumed in Procedure 1670.4.
**Value is for centerline of plume (2700).
Location W-I1 (Goldsboro) is slightly
off the centerline (2650).
The x/Q estimated for this location is 6 E-5 sec/m 3 ;the corresponding whole-body and thyroid dose estimates are 0.3 and 0.02 mrem/hr,
respectively.
***Dose rate is based on the addition of release rates determined from eachmonitor reading.
11-3-76

In summary,by 0710 on March 28, the licensee had estimated that the whole-body exposure rate to an individual

residing in Goldsboro was 40 R/hr.

The

investigators concluded that if the proper input parameter had been used, thelicensee's procedure would have yielded an exposure rate of 2 mR/hr at Goldsboro.
The basis for the 40 R/hr could not be determined by the investigators.
Supervisor,

The

Radiation Protection and Chemistry later (prior to 0800) performed

a cursory check of the calculation performed by the engineer and apparently
concluded that no errors had been made (he also misread the HP-R-214 monitor).
He recalled that this calculation resulted in an exposure rate of 10 R/hr rather
than 40 R/hr. However,

because of the very low reactor building pressure at

the time (less than 1 psig),

the supervisor felt that the offsite dose calcu-

lation was totally unreliable (Int.

133).

The State Bureau of Radiological Health (BRH)

was-notified at approximately

0725 that the estimated exposure rate in Goldsboro was 10 R/hr.
parently the same estimate that was made at 0710; it
than 40 R/hr was reported (Int.

46).

in the footnote to Table 11-3-2.
cluded (Int. 46)

This was ap-

is not known why 10 rather

The basis for the 10 R/hr is discussed

Actions taken by the BRH at this point in-"....We then contacted Civil Defense [now the Pennsylvania

Emergency Management Agency (PEMA)]...

to inform them where the wind was blowing

and to be prepared to evacuate on the west side of the river...".

After being

informed by Metropolitan Edison that there were no sigrificant radiation levels
detected outside of the plant, BRH
hold tight.
3.3.2.

"....

notified the Civil Defense [PEMA] toThis was all before 8:00..."Initial Confirmatory Surveys

Although a site emergency was declared at 0654, the first environmental
radiation survey was not performed until 0748 at the screenhouse (between survey points GE-8 and GE-9 on the Island).
factors:

This delay was apparently due to two(1) a certain amount of time (estimated to be 10-15 minutes) was re-quired to perform an operability check of each emergency kit, and (2)sion existed at the time with respect to initial dispatching of teams.

11-3-77

confu-

The second point became apparent during interviews of radiation/chemistry
technicians:
Radiation/Chemistry Technician A (Int. 99)"When I got there [approximately 0700, March 28],

it wasn't too longthat we were assigned the job of the offsite team...
to go ...
the haste,

we were readywhere they needed some onsite readings right away...

so in

I radioed back and said we would take over the onsite team..."Radiation/Chemistry Technician C (Int. 82)
I arrived at work approximately 7:00 a.m.

[March 28] we walked

into the processing center and everyone was standing around ... thefirst foreman we saw, he came up to me and said go out and check theradiation emergency kits...

so I went out there and we got the SAMs

out and made sure they responded...

I guess it was approximately 8:30...

[a radiation protection foreman] asked us if we were on our way toGoldsboro and I proceeded to tell him that we had never received any
instructions where to go..-. so in about ten minutes I guess we had
finally rounded up the vehicle and proceeded to Goldsboro..."Radiation/Chemistry Technician N (Int. 47):"...about 7:30 or 25 to 8.

The [radiation protection] foreman...

was-trying to determine wind direction and other factors involved insetting up these monitoring teams.
monitoring kit...

He told me...

and go to Goldsboro by vehicle..,to get a radiation
we got to Goldsboro

about 8:45..."
Team alpha,consisting of two radiation/chemistry technicians (Int. 99),

performed the survey at the screenhouse (between survey points GE-8 and GE-9)
at 0748.

The result was less than 1 mR/hr (the detection limit for the instru-

11-3-78ment used).

This survey confirmed that the significant offsite radiation levels

calculated at 0710 (10 and 40 R/hr in Goldsboro) did not exist (Int.

127).

Other confirmatory surveys were performed at 0832 and 0842 in Goldsboro (Ref.
153,

169).

Results of both surveys were less than 1 mR/hr.

The calculated radioiodine concentration at Goldsboro, which corresponded
to the initial thyroid dose rate projection of 8.5 mrem/hr (Table 11-3-1) was
5 E-6 pCi/cc (Ref.

168).

Air samples (filter plus charcoal)

were collected at

0900 at location GE-8 (on-island; 700 ft WSW of Unit 2 vent) and at location
WSW-11 (South Goldsboro; 1.4 mi WSW of Unit 2 vent).

Results (Ref.

169) forthese samples, which were counted in the field with a SAM-2 with RD-19 detector,
were 1.2 E-8 pCi/cc for location GE-8 and 5.2 E-9 pCi/cc for location WSW-II,
both of which are very close to the minimum detectable activ-ity (MDA)
pCi/cc applicable to the sample volume and background (Int.

22).

of 5 E-9

These results

were interpreted as iodine-131 activity in accordance with Procedure 1670.6,"Off-site Radiological Monitoring."
A radioiodine sample collected at 0940 at location W-11 (Goldsboro; 1.2
mi W) was transported to the Commonwealth of Pennsylvania Bureau of Radiological
Health in Harrisburg for analysis on their Ge(Li)system (Int.

46).

The resultsof that analysis completed at 1300 on March 28 indicated that most of the radioactivity on the charcoal was due to the 81 keV gamma rays from xenon-133 andthe 250 keV gamma rays from xenon-135 (Ref.

170)

was reported as less than 1.5 E-11 pCi/cc.

The field-determined iodine-131.

concentration for this sample was 3 E-8 pCi/cc (Ref.

The iodine-131 concentration
170).

These results are

discussed further in Section 3.3.3.2 which follows.
3.3.3.

Offsite Radiation Measurements

The offsite radiation survey effort performed by the licensee during March 28
to 30 was directed by personn'el in the ECS (Int. 22,

11-3-79

28, 76,

110).

The licensee's radiation survey effort consisted of the following typesof measurements:(a)Direct radiation measurements performed with portable radiation sur-vey instruments by land-based and helicopter-based teams (Ref.
ments used (Int.

Instru-169).

22, 28) were generally the PIC-6A* (an ion chamber type instru-

ment having a range from 1 mR/hr - 1000 R/hr),the RO-2* (an ion chamber type

instrument having a range from 0-5000 mR/hr with four scale multipliers) andthe E-520* (a GM-type instrument having a range from 0-2 R/hr with five scale
multipliers).
(b)Collection of short-term air samples (particulate and iodine) for

field determination of radioiodine concentrations (primarily March 28).
samples were later (primarly after March 28) counted with a Ge(Li)in a mobile laboratory (Int.
(c)

system based

73, 93).

Collection of air, water, and milk samples and TLDs as part of theroutine Radiological
(Int.

These

Environmental Monitoring Program (REMP)

beginning on March 29

19, 28).
3.3.3.1

Direct Radiation Measurements

Beginning with the initial surveys (discussed earlier) performed early on
March 28,

a radiation survey program continued beyond the period of this investi-

gation (Ref.

169).

directed by the ECS.

As mentioned above,

the offsite and onsite surveys wereTeams consisted initially of two radiation/chemistry techni-

cians and later (primarily March 29 and 30),

as more manpower became available,

teams consisted of three individuals (Int. 22,

182).

Three teams were dispatchedinitially and up to six teams were later making surveys on March 29 and 30 (Int. 76,

82).

Manufactured by the Eberline Instrument Co.

11-3-80

Offsite teams were dispatched by the ECS to specific, predetermined points(most were within 5 miles from the site) located on a 1:24,000 scale U.S. Geological Survey Quadrangle Map (Int. 48).

A general area map is shown in Figure11-3-4. A wind speed and direction recorder is located in both control rooms;
these parameters are measured at the site meteorological tower located 2200 feetnorth of the Unit 2 reactor building (Figure 2.3-22 TMI-2 FSAR).

Plume dispersion

at any given location was determined by persons in the ECS using the methods
and atmospheric dispersion factor (x/Q)

isopleth overlays (also drawn to a

1:24,000 scale) referenced in Emergency Procedure 1670.4, "Radiological Dose Calculations" (Ref.

167).

According to this procedure, a stable, neutral,or un-

stable atmospheric dispersion isopleth is used, depending whether the winddirection range (over a 20-minute interval) is <450,>750,

between 450 and 750,

orrespectively.
In general,

teams were dispatched in the downwind direction to performsurveys at points that were inside the x/Q isopleths (i.e.,
extent of the plume).

Onsite teams,

in general,

inside the known

were dispatched in the down-

wind direction to one of the GE-series locations shown in Figure II-3-5.
Radiation levels measured by land-based onsite and offsite teams firstbegan to increase at 1020 on March 28, with an onsite team in the north parking lot reporting a radiation l evel of -3_mR/hr _(Ref.

169).

Onsite radiationlevels generally continued to increase during the next 12 to 13 hours,

peaking

at 300 mR/hr (P,y) near the north warehouse at 2130 on, March 28 and 365 mR/hr(p,y) at survey point GE-10 at 2325 on March 28.
Offsite radiation levels reported during this period, with the exceptionof 50 mR/hr measured at 1548 on PA Route-441 about 1500 feet south-of the northgate (see Figure 11-3-4 for location), were not above background until 2238
when a radiation level of 13 mR/hr was measured near Kunkel School (5.6 mi NNW)
(Ref.

169).

Several other radiation levels above background were noted in thisgeneral area prior to midnight.

However,the 13 mR/hr value was the highest

one measured until 30 mR/hr (p,y) was measured in Goldsboro at 0600 on March 29.

11-3-81

TMI-GENERAL AREA MAPNDAUPHIN COUNTYA\cP

ADAPTED FROM FIGURE 2.2-1
TMI UNIT 2 FSAR
03 MILES

1
SCALE

Figure 11-3-4

ON-ISLAND (GE SERIES) SURVEY POINTS

SAND
BEACH
ISLAND

EAST CHANNELSUSQUEHANNA RIVER

Figure 11-3-5

Several comments relative to the Kunkel School and Goldsboro measurements
referenced above need to be made at this point:(1) During the 7-hour period from 1700 to 2400 on March 28, the wind was
blowing consistently toward the northwest at a speed of 5-15 mph; during the
2-hour period from 0340 to 0540 on March 29, the wind was blowing consistently
toward the west-nbrthwest at-a speed of 4-8 mph; and during-the 3 1/2-hour-period
from 0540 to 0910 on March 29 the wind was blowing consistently toward the west
at a speed of 2-6 mph (Ref.

171).

During these periods of time (12 1/2 hours

total), the effluent plume from the plant should have been well defined.
(2)

According to an analysis made by the licensee,

period from 0700 on March 28 to 1600 on March 29,
were released (Ref.

84, 85).

through the end of Marchduring the 33-hour

6.6 E+6 Ci of noble gases

This quantity corresponds to 75% of that released

and 66% of the total released from March 28 throughApril 30.
(3)

The plume positions for the periods of time discussed in (1) above

are shown in Figures 11-3-6,-3-7,

and -3-8.

The plumes shown are for stable

atmospheric conditions (wind direction range 450)
idealistic (e.g.,and, although they are somewhat

they do not depict the turbulence that might have been intro-

duced by Hill Island - Figure II-3-6),

they served as a reasonable means of

directing the survey teams to the higher concentrations of radioactivity offsite
(Ref.

172,

Int. 22)(4)

From an examination of Figures 11-3-6 through 11-3-8 and the survey

times contained in the ECS Radiation Survey Log (Ref.
a.

169),

it

is apparent that:

The licensee failed to perform radiation surveys in the first plume
until 2238 (at Kunkel School)- more than 5 hours after the plumeappeared to have been established;

113-84b.

The licensee failed to perform any radiation surveys in the second
plume - a period of 2 hours;c.

Surveys were performed in the third plume (beginning at 0600) in Goldsboro with a resultant radiation level of 30 mR/hr (•,(y).

'), 20 mR/hr

Offsite radiation levels measured by land-based teams during the remainderof March 29 were generally less than 1 mR/hr, with the maximum noted as 3 mR/hrin Royalton at 2355 (see Figure 11-3-4 for location).

Onsite radiation levels-

measured during March 29 were generally less than 10 mR/hr, with the maximum
noted as 150 mR/hr (p,y) and 100 mR/hr (y)at location GE-9 at 0532 (see Figure11-3-5 for location).
Offsite radiation levels measured by land-based teams during March 30 weregenerally less than 1 mR/hr, with the maximum noted as 15 mR/hr (p,y) and 5 mR/hr(y)

at standard survey point S-ll (approximately 1 mile south of the plant) at
0906. Onsite radiation levels measured during March 30 were generally lessthan 10 mR/hr, with the maximum noted as 110 mR/hr (P,y) and 20 mR/hr (y)location GE-9 at 1445 (Ref. 169).

at

The numbers of radiation measurements made by the licensee's land-based
teams during March 28 through March 30 at onsite and offsite locations is asfollows (Ref.

169):
Shift

On-site(Outside)

Off-site

March 28

0700-1500
1500-2300

20
29

19
27

March 292300-0700
0700-1500
1500-2300

14
15
36

14
8
3

March 30

2300-0700
0700-1500
1500-2300

54
68
67
M

45
36
19
171

Date

Totals*Includes standard survey locations GE-I through GE-1O only
(Figure 11-3-5); estimated by the investigator to represent
about 80% of the total effort on site (outside of buildings).
11-3-85

PREDICTED PLUME LOCATION (1700-2400, MARCH 28)
1700-2400
3/28/79ATOMSHPERIC DISPERSION FACTOR (x/Q)
FOR A STABLE ATMOSPHERIC CONDITION(WIND DIRECTION VARIABILITY <450),
NORMALIZED TO A WIND SPEED OF 1 MPH.

* PREDETERMINEDRADIATION SURVEY
POINT

PLUME

G SURVEY PERFORMEDBY MET ED DURING
STATED INTERVAL

3240
HOST INN

42*

TLD LOCATION(ROUTINE PROGRAM)

40

2.5 x 10-6
2.5 x 10-8I

5 x 10-8

38
36
34)NGVIEW ACRES(D EBENEZER CHURCH RD& PA-441

3230
28

TPKE EXIT 19

26
24

22~
20 0
18 U

16~
14~
12

10
8
6
4
SHELLEY
IS./

2
0

10

8

6

4

2

0

2

4

6

CROSSWIND DISTANCE, 103FT

8

10Figure 11-3-6

00

PREDICTED PLUME LOCATION (0340-0540, MARCH 29)
PLUME
ATOMSHPERIC DISPERSION FACTOR (x/Q)
FOR A STABLE ATMOSPHERIC CONDITION(WIND DIRECTION VARIABILITY <45'),NORMALIZED TO A WIND SPEED OF 1 MPH.

0340-0540
3/29/79
2880
PREDETEFRMINEDRADIATIC)N SURVEY
POINT
2.5 x 10-6

10-6

2.5 x

I 2.5 x 10-

0-8

2.5 x 10-8*40SURVEYS PERFORMEDBY MET E-D
DURING.S3TATEDINTERVALL38

TLD LOCi ATION(ROUTINE PROGRAM)

36

1 x 10-534
32-30 lzWNW-51

28 0

.. .

26
24

WNW-41

22 0
20
0

18
16

I WNW-31

14
12

WNW-21
WNW-11

10-

GOLDSBORO-

8

12B1 rM

4

SHELLEY ISLAND

2

101B1

10

8

j

14S2

4
6
3 FT
CROSSWIND DISTANCE, 10
6

4

2

0

2

8

10

0
Figure 11-3-7

PREDICTED PLUME LOCATION (0540-0910, MARCH 29)
PLUME

054 0-0910

1
2700

ATOMSHPERIC DISP ERSION FACTOR (x/Q)
FOR ASTABLE ATMC SPHERIC CONDITION(WIND DIRECTION VARIABILITY <450),'TNF) .•PFFfl fF 1 MPW .....
NORMALIZED TO A W
OF I MPH

3129/79..-

IND SPEEDPREDETERMINEDRADIATION SURVEY
POINT

2.5 × 10-6
2.5 x 10-81xE) SURVEYS PERFORMED
ofr-, I,u L.,ir,,,-40
STATED INTERVALU TLD LOCATION
38

2.5 x 10-6
-in-5

I 2.5

110-8

(ROUTINE PROGRAM)

36
34
32

W-51

30

28
26
W-41

24

WN.-41
N

t
22

--

20

z-w
18 0

W-31

6.

. . ... . . . . !....... .1...
.. ..'' ....
W-21..--. .----.......
.. . . .. . .. . . .

14 w3

..

0

_12

10

WNW-11I

BORO.i!

:

88

-

12B1
6

:_

. . ..

.

.

4---------.EY ISLAND:------

1.OB

14

13

12

11

10

8

0

144S2

..-- ....... -A-.

6
4
4
2
0
2
6
3
CROSSWIND DISTANCE, 10 IFT

8i---N--N-A-

10

11'-A-J.

12

13

..

2

. 2

14

Figure 11-3-8

The above radiation survey effort represents only those surveys performedby the licensee and those performed under the licensee's direction. Surveys
performed by the Commonwealth of Pennsylvania,
its contractors,

the Department of Energy and

the Nuclear Regulatory Commission,and other Federal agencies

during March 28 through 30 are not included above since, with the exception ofa few Commonwealth and NRC surveys, results were not made available to the licensee (Ref.

169,

Int. 48).

Details of these surveys appear in Appendix II-C,

Tables II-C-1 through II-C-6.
Helicopter-based teams were used by the licensee during the period fromMarch 28 through 30, although the heaviest usage during the period was on March 30.
This type of survey capability is not addressed in the Emergency Plan (Int.
however, both the Supervisor,

20);

Radiation Protection and Chemistry and the Radiation

Protection Supervisor indicated that such capability was very helpful (Int.
20, 22).
The helicopter-based teams were called upon to find the noble gas effluent
plume and to make sufficient surveys to define it with reference to certain
landmarks (Int. 25).
Initially one helicopter belonging to the State Police,
and later, three helicopters chartered by the licensee were used in this effort(Int. 25).

Date
March 28
March 29
March 30

A summary of these surveys follows (Ref.
Number of
Flights

Number of Radiation
Measurements

1
1
13

24

5
295

169):

Range of results - mR/hr-(Py)<1 - 20<0.1 - 3000
<0.1 - 1200

The maximum radiation level measured during each survey was generally
directly above the plant. The 3000 mR/hr (y component was 400 mR/hr)
was measured approximately 15 ft above the Unit 2 plant at 1410; the1200 mR/hr was measured approximately 130' (600' above sea level) above
the Unit 2 reactor building at 0801.

11-3-89

3.3.3.2

Airborne Radioactivity Measurements

Including the three initial air samples discussed earlier, a total of 28
air samples were counted in the field with the SAM-2 during the period March 28
through 30 (24 on March 28 and two on both March 29 and March 30) (Ref.

169).

Eleven of these samples were also counted by the Pennsylvania BRH Laboratory(Ref.

170); the Radiation Management Corporation's Fixed and Mobile Laboratories(Ref.

178, 173); and by the Nuclear Regulatory Commission's Region I Mobile

Laboratory (Ref.

177).

Radioiodine concentrations on these 11 samples wereless than the minimum detectable activity (MDA)

for each laboratory, which ranged

from 2 E-11 to 6 E-IO pCi/cc (Int. 46, 73,

Licensee results for these93).

same samples ranged from <5 E-9 to 6.8 E-7 pCi/cc (Ref.

169).

In summary,thesamples collected and method of counting were as follows:

Date
March 28
March 29
March 30

Number Countedwith SAM-2 only

Number Countedwith SAM-2 & GeLi

Number Countedwith GeLi only

Total

15
1
1

9
1
1

3
12
14

27
14
16

Radioiodine was not detected (i.e., was below the MDA)

in any of the 40178); however,samples counted by Ge(Li) Spectrometry (Ref.

170,

and xenon-135 were detected in many of them.

These radioxenon's interferred

173,

177,

xenon-133

with the ability of the SAM-2 to quantify radioiodine according to a Metropolitan
Edison consultant who was present in the ECS at the time (Int.

19,

28):

....... I suspected that it might be what we called white out .....

[there]was just absolutely so much xenon there that the crystal was saturated
because there are small crystals in the SAM-2s .......
Another difficulty encountered with the use of the SAM-2 was the fact thatseveral radiation/chemistry technicians stated that they either had not been
trained in the use of the SAM-2 or that they were not confident of their ability
to use the instrument because of the small amount and infrequent nature of the
training that they did receive (Int.

82, 99,11-3-90156).

3.3.3.3

Radiological Environmental Monitoring Program Data

Samples from the routine Radiological Environmental Monitoring Program(REMP)

were collected on March 29 after comparing the benefit of early sample

collection and early knowledge of environmental

impact-with the benefits oflater sample collection (and being able to integrate more of the early environmental impacts) (Int. 19, 28).
Discussions on this matter were held between
the Supervisor,

Radiation Safety and Environmental

Engineering and his staff

and Radiation Protection Consultant A on March 28 (Int.

19,28, 30,160).

Samplesfrom the REMP collected on March 29 were analyzed by Teledyne Isotopes and theRadiation Management Corporation (telecon with representative of Porter-Gertz
Consultants).

The data available to the licensee prior to 2400 on March 30

are contained in Table 11-3-3 (Ref.

175).

After the March 29 collection, thelicensee conducted the REMP on an increased regime basis, which included daily
milk and water sampling, daily TLD changeout, and air sample changeout every
three days (Int.

19, 28).

Based on a telephone conversation with a representa-

tive of Porter-Gertz Consultants,
at two additional locations,Inc.,

air iodine measurements were also mademaking a total of eight locations (ten resultsappear in Table 11-3-4 because two locations have additional samples for quality
control purposes).
With regard to the TLD data shown in Table 11-3-4, two results for Kohr
Island contained in the footnote (908 and 453 mR/6 mo.),at this point.

need further elaboration

During a telephone conversation with a representative of Teledyne

Isotopes, an investigator determined that the Kohr Island TLD's were exposed
for a six month period (ref.

103); however,most of that exposure was received,during the last two days that the TLD was exposed (March 28 and 29).
ground radiation exposure for these TLDs,

using the 9G1,

The back-

7G1 and 15GI locations

as controls, was estimated to be 40 mR for the six month period.

If this 40

mR is subtracted from the two dosimeters and the remaining exposure is divided
by the exposure time after the event (29 hours from 0700 on March 28 until approximately 1200 on March 29),

average exposure rates for Kohr Island are calcu-

lated to be 30 and 14 mR/hr.

Since the radiation being measured is gamma radiation

the simplifying assumption is made that a 1 mR exposure is approximately equal

11-3-91

to a dose of 1 mrem.

Therefore,

rates of about the same value.

these exposure rates are equivalent to dose
These dose rates are in excess of the 2 mrem/hr

limit for unrestricted areas contained in 10 CFR 20.105(b).
Table 11-3-4
Radiological Environmental Monitoring Program
Data Available to the Licensee by March 30,

Type
TLD

Location
N. Bridge (1S2)
N.Wthr.

Result (mR/qtr)- 0.4 mi N

Stn (2S2)

Falmouth Subst. (7B3)

Available

37.5

3/30/79 @1230

81.0

3/30/79 @1230- 0.2 mi NNW

923.0

3/30/79 @1230- 1.6 mi SE

11.6

3/30/79 @1230-0.7 mi NNE

N. Boat Dock (16S1)

1979

Ob. Center* (5Al)- 0.4 mi E

15.6

3/30/79 @1230

Goldsboro (12BI)- 1.6 mi WSW

15.1

3/30/79 @1230

16.7

3/30/79 @1230

21.4

3/30/79 @1230

19.3

3/30/79 @1230

22.4

3/30/79 @1230

Middletown Substn. (1Cl)
Drager Farm* (7FI)- 9 mi SE

N. York Subst. (9GI)Columbia (7GI)-2.6 mi N- 13 mi S- 15 mi SE

W. Fairview* (15GI)- 15 mi NW

17.2

3/30/79 @1230

S. Beach TMI (9S2)- 0.4 mi S

23.0

3/30/79 @1230

Laurel Rd (4A1)- 0.5 mi ENE

17.0

3/30/79 @1230

Kohr Is.

- 0.4 mi NNW

8.0**

3/30/79 @1230(lOBl)

7..0**

3/30/79 @1230

6.9**

3/30/79 @1230(16AI)

S. Shelley Is.
Shelley Is.

(14S])- 1.1 mi SW- 0.4 mi WNWMonthly TLDs were also read at these locations.
Data were consistent
with the quarterly TLD, except for the Observation Center which read only
0.5 mR/mo.
**

Results for TLDs that were sent to TMI for changeout at the end of the fourth
quarter; however, TLDs at these locations were not changed out because of
weather conditions and remained in a lead container (i.e., were unexposed)
for the first quarter of 1979, Results for these locations, which were exposed for the fourth quarter of 1978 and the first quarter of 1979, (two TLDs
per location) were: 14S2-131 & 148 mR; 16AI-908 & 453 mR; and 1OBI-40.6 & 36.6
mR (all values are mR/6 month interval). According to a telecon with a representative of Teledyne Isotopes, the factor of two difference between the Kohr
Island TLDs cannot be explained.
11-3-92

Table 11-3-4 Continued
Radiological Environmental Monitoring Program
Data Available to the Licensee by March 30, 1979

Type

Location

ResultAvailable

Air iodine

Middletown Substn (ICl) - 2.6 mi N*
Drager Farm (7FI) - 9 mi SE*
N. Weather Stn (IS2) - 0.7 mi NNE
Falmouth Substn (8Cl) - 1.6 mi SE
Observation Cntr (5Al) - 0.4 mi E
W. Fairview (15Gl) - 15 mi NW
Drager Farm (7Fl) - 9 mi SE
Middletown Substn (lCl) - 2.6 mi N
Goldsboro (12Bl) - 1.6 mi WNW
N. York Substn. (9Gl) - 13 mi S

0.05 pCi/m3
<0.02
0.47
<0. 02**<0. 02<0.03<0. 04
0.03
0.30
<0.02

3/30
3/30
3/30
3/30
3/30
3/30
3/30
3/30
3/30
3/30

Water
IodineSwarata Creek (upstream)<10 pCi/l

3/30 @1230

Steelton Water Works (upstream)

<10
<10
<10
<10

3/30 @1230
3/30 @1230
3/30 @1230
3/30 @1230

York Haven Gen. Stn.
Columbia Water Plant
York (downstream)(downstream)(downstream)*Additional sampler for quality control purposes.
**Result is suspect. Air sampler did not appear to be working properly.
(Section 1.7.1.1)

11-3-93

0945
0945
1230
1230
1230
1230
1230
1230
1230
1230

Data from 21 locations on the island having LiF TLDs (the same as thoseused to assess occupational exposure - see Section 1.4.3 and 1.6.1), although
not available to the licensee during the period covered by this investigation(March 28 to 30),need to be briefly recognized at this point.

of results for these dosimeters,

which were generally changed out on a monthly

basis (near the end of the month),and April is as follows (Ref.

A comparison

with the environmental TLD results for March

166):(1) Environmental TLDs indicated that the highest offsite dose rates occurred during the first three days following the event, approaching backgroundlevels levels prior to mid-April (Ref.
(2)

84, 85).

The twenty-one onsite LiF TLDs indicated dose rates during the firstfew days in April may have been higher than dose rates during March 28-30 if
most of the April exposure occurred during the first few days of the month.
If a longer exposure period is assumed, however,these TLDs appear to be con-sistent with the results of the environmental TLDs.
Further review of this area is beyond the scope of this investigation.
3.3.3.4

Assessment/Response to Environmental Measurement Results

Based on the environmental impact information gathered by the licensee
during March 28 through March 30,

as discussed in the previous sections,itappears to the investigators that:1.

The licensee's actions taken to confirm the initial prediction of 40

R/hr in Goldsboro could have been more timely.

Even though the licensee believed

the predicted exposure rate to be unrealistic,

confirmationthat such levels

did not exist outside of the plant required 38 minutes (prediction completed
at 0710; confirmatory survey made at onsite location GE-8,
at 0748).

11-3-94

720 ft. SW of vent,2.

The principal

radiation exposure pathway for individuals residing in

the vicinity of the site was direct radiation from radioactive noble gases (primarily xenon-133) released to the atmosphere.

Radioiodines and other nuclides

released in gaseous and liquid effluents constituted a much less significant
exposure pathway.
3.

The methods employed by the licensee to assess doses received via

the principal pathway were:(a)

Use of TLDs which had been deployed (prior to -the event-) at 15 indi-cator and 5 background locations as part of the routine monitoring program.
It appears that, on occasion,
TLD locations.

the plume centerline was located between

The deployment of additional TLDs early in the event would haveincreased the probability of a given TLD being in the plume.
nation of meteorological data, however,Based on an exami-

the time periods when the wind blew

consistently into a given sector for several hours at a time were <30% of thetotal 68-hour period examined during this investigation (Ref.

171).

Duringthe major portion of the 68-hour period examined, the plume was either not welldefined or tended to meander (Int.

19, 22, 28, Ref.

171).

This latter condition

likely caused exposure rates at a given point to fluctuate considerably duringmuch of the time (Ref.

169,

Int. 22).

Both of the above conditions (a well-defined

plume in between TLD locations and an ill-defined, meandering plume) could havebeen evaluated better by the licensee had additional TLDs been deployed earlyin the response.
. .

This point is addressed in procedure 1670.6 (Ref.

176):

Place sufficient TLDs at either the continuous air monitor or at aconvenient,

representative location in the designated area to permit reading"of TLDs every four hours during the emergency .

11-3- 95

The deployment of additional TLDs apparently had not been consideredby the licensee during early stages of event (Int.

22, 25); however,even if

this had been considered, according to Metropolitan Edison Corporate Environmental Scientist B the licensee was not prepared to implement this portion of
the program (Int.

74):"I was aware that [the deployment of additional TLDs] was in the procedure;
however, I was going to change the procedure because the environmental
TLDs cannot be picked up and read on site.

They are special TLDs whichhave to be sent off to Teledyne Isotopes and Radiation Management Corporation
to be read.

So an offsite monitoring team could not collect those environ-mental TLDs and bring them here to be read."b.

Use of survey teams with portable radiation survey instruments toperform surveys at predetermined offsite locations and to track effluent plumes(primarily with the helicopter).
The data gathered by such teams were the licensee's primary means ofmonitoring the environmental

impact of effluent releases until about mid-day

on March 30 when TLD results (changed out March 29) were available.

The survey

instrument results were the only means of quickly assessing changes in environmental impact and were a necessary adjunct to the TLDs.
Results of offsite radiation surveys were evaluated by personnel in the
Emergency Control Station.

An environmental consultant was called upon to playa lead role in the evaluation.

The consultant used the following criterion

for assessing whether offsite radiation levels were increasing to the point
where action might be required (Int.
1.

19,

28):

my trigger point is that if we're going to be giving people in theenvironment somewhere between 50-100 millirem, at that point we have to
let the State know .

.

..

we had a hot line to the State and every, practi-

11-3- 96

cally every number that came in, we just relayed to the State .
was just a matter if

it

.

..

so it

looked like the collective dose was up there between

50-100 mR then it was up to me to say to the Emergency Director,
point, I think we better tell the state that it'At this

looks like the collective

dose is reaching this point and so therefore, they need to think whether
or not they want to take protective action.'that point

But we never even got to..

As noted earlier, the investigators have identified problems withthe surveys conducted by the licensee,i.e.,

the periods from 1700 to 2400 on

March 28 and 0340 to 0910 on March 29 when an inadequate assessment of radiological conditions was made.

In addition,

survey records have been discarded (Ref.

itappears that original radiation

194),

leaving the only record of sur-vey results being those radioed in and recorded by personnel in the ECS (Ref.
174).

169,

The ECS survey log, however, does not contain (with few exceptions) any

notation of the specific instrument used to perform the survey, making it

very

difficult to determine the correction factor (due to the 81 keV gamma ray fromxenon-133) to be applied to the measurements (Ref.

183).

also partially or totally absent from this survey log:
instrument was used,i.e.,

The following were(a) mode in which thewhether open or closed window; (b)orientation of

the instrument with respect to the source being measure; (c)the duration of

the measurement; and (d)4.

name of the individuals performing the survey.

Radiation dose estimates for the hypothetical maximum exposed individual

have been published previously by the licensee and by an Ad Hoc Committee consisting of NRC,

EPA and HEW representatives (Ref.

65, 84, 85).

Estimates ofchild and adult thyroid doses from the inhalation and ingestion (milk) pathways,contained in these reports was 5 mrem or less for the highest exposed individual.
With respect to radioiodine releases to the atmosphere,

results of fifty-sevenairborne radioiodine measurements based on short-term samples and ten airborne
radioiodine measurements based on long-term samples from the routine radiologicalenvironmental monitoring program (REMP)

were available to the licensee durihg

11-3-97

the period March 28-30.

Fifty of these samples (40 of 57 short-term and 10 of

10 REMP samples) were analyzed by Ge(Li)spectrometry for 1-131.

These analyses

indicated that all results, with the exception of five REMP samples, were below
the minimum detectable activity (MDA) which ranged from 2 E-14 to 6 E-I0 pCi/cc.
The lower MDA values were associated with the larger volume samples from. theREMP, while the higher MDA values were associated with the small volume (grab)
samples collected by the survey teams.

Results of the five REMP air iodine

samples above the MDA ranged from 0.03-0.47 pCi/m 3 .
air for unrestricted areas is 100

The MPC for iodine-131

in

pCi/m 3 .

Doses received by way of the principal environmental pathway (noble gas
releases to the atmosphere) were also included in each report (Ref.

65, 85).

The licensee's estimate of doses to the maximum exposed individual was 75 mremthrough April 30 for a location of 1200 meters NNE,

while that of the Ad Hoc

Committee was 83 mrem through April 7 for a location 800 meters ENE (Ref.
84, 85).

65,

Although different time frames are referenced in these two estimates,

they are fairly comparable (in time) since, based on estimates made by the licensee, greater than 99% of the noble gas releases through April 30 occurredby April 7 (Ref.

84, 85).

While the investigators did not perform a detailed, independent assessmentof individual or population doses calculated by the licensee and by the Ad Hoc
Committee,

a review of environmental radiation levels due to noble gas releases

was made (Ref.

65, 84, 85).

Special attention was paid to those periods oftime when the licensee did not, in the judgment of the investigators, perform
a sufficient number of surveys in a given area and other data suggested that
relatively high radiation levels were likely to be present.

One of these time

periods was from 1700-2400 on March 28 in the northwest and north-northwest
sectors (see Section 3.3.3.1).

The dose received by an individual residing

continuously in this area, based on the isodose contours in Appendix A of the
Ad Hoc Committee Report, was estimated to be 100-200-mrem during the period
from March 28 to April 3 (Ref.

65).

A significant fraction of this dose was

11-3-98likely to have been received during the evening of March 28.

This report (Ref.

65) also states that the dose rate measurements performed by DOE (which form
the basis for the dose contours) are likely to be conservative by at least afactor of three, so that the actual dose was probably less than 100 mrem.

11-3- 99

R E FE R E N C E S

REFERENCES
1.

Letter dated March 5,2.

Computer Printout, Emergency Training/Qualification,

3.

"Radiation Monitoring System Alarm Responses,
Station,4.
5.

1979 from Metropolitan Edison to NRC
Status as of 3-1-79

Three Mile Island Nuclear

Unit 2," Revision 2, dated November 9,

1978

Procedure 1670.9,"Emergency Training and Emergency Exercise," Revision 6,

dated February 3,

1979Training Program Administrative Forms,

1978 and First Quarter 1979 -Emergency Plan Training
6.

1978 Listing of Shift Maintenance Personnel

7.

Generation Division, TMI Station Resume

8.

1977 SAM-2 Training Form

9.

Lesson Plans - EP Training

10.

Record of Telephone Conversation Between the Group Supervisor - Technical11.

Training. and the Supervisor, Mechanical Maintenance,

dated February 1, 1979

Memorandum from Supervisor of Training dated April 5,

1979,

Subject- Pre-

paration of Lesson Plans for Radiation Emergency Drill Training Programs 1978
12.

Memorandum of Meeting Between the Radiation Protection Supervisor,
Supervisor,

Technical Training and Group Supervisor,

dated September 21,
13.
14.

Technical Training to Superintendentof Maintenance,

1978,

dated November 28,

Memorandum from Group Supervisor,

Subject: REPAIR PARTY TEAM TRAINING

Technical Training to Nuclear Engineer

1978, Subject: HPP 1670.9

Memorandum from Group Supervisor,
dated December 20,

16.

1978

Memorandum from Group Supervisor,

dated November 29,
15.

Technical Training

Technical Training to Nuclear Engineer,

1978, Subject: EMERGENCY TRAINING

Record of Telephone Conversation Between the Group Supervisor, TechnicalTraining and the Supervisor,
February 2,

Radiation Protection and Chemistry,

1979

1

dated

17.

Record of Telephone Conversation Between Group Supervisor, TechnicalTraining and the Superintendent of Maintenance,"Training Requirements,"

18.

Procedure 1690,

19.

ANSI N18.l-1971 Section 5.5

20.

Scenarios and Observer Sheets,

dated February 9,

1979

Revision 6 dated March 22,

1978

Radiation Emergency Drills Conducted

During 1978
21.

Radiation Emergency Drill Critique Individual Action Item Assignment Sheetsfor 1978 Radiation Emergency Drills

22.

Memorandum from Radiation Protection Supervisor to All Departments dated
October 13,

23.

1978, Subject: ACCOUNTABILITY DURING A RADIATION EMERGENCY

Surveillance Performance Form,
dated March 11,

Emergency Equipment and Inventory Control,

1979
TP 360/IC

24.

Radiation Monitoring System Test - Area Monitors,

25.

Radiation Monitoring System Test - (Liquid),

26.

Radiation Monitoring System Test - Atmospheric Monitors,

27.

Surveillance Performance Form and Waste Gas Decay Tank Monitor Calibration

TP-360/lB

Data Sheets for Procedure 2322-Ql on February 9,28.

TP 360/lA

1979

Surveillance Performance Form and Liquid Effluent Radiation Monitor Data
Sheets for Procedure 2322-2Q dated February 13,

1979

29.

Surveillance Performance Form and Post Accident Monitoring Instrumentation
Channel Check Data Sheets for Procedure 2301-Ml dated on February 28, 197930.

Surveillance Performance Form and Waste Gas Unit Vent Functional Test Data
Sheets for Procedure 2322-Al dated September 8,

31.

RMS Instrumentation Channel Check Data Sheets for Procedure 2301-SI dated
March 22-28,

32.

1978

1979

Radiation Monitoring System Data Sheets for Procedure 2612-R5 dated
March 20-26, 197933.

Surveillance Performance Form and Radiation Monitoring System Atmospheric
Monitor Calibration Data Sheets for Procedure 2612-R2 dated on March 4-5,
1979 and October 14,

1978

2

34.

Surveillance Performance Form and Cohtainment Monitor Functional Test Data
Sheets for Procedure 2303-M9 dated on March 1, 197935.

Surveillance Performance Form and Radiation Monitoring System Monthly Test

36.

Data Sheets for Procedure 1303-4.15 dated on March 18, 1979
Surveillance Performance Form and Radiation Monitoring System Weekly Surveillance Check Data Sheets for Procedure 1301-4.1 dated on March 17,

37.

1979

Surveillance Performance Form and Radiation Monitoring System Data Sheets for
Procedure 1302-3.1 dated on January 10, 197938.

Surveillance Performance Form and Liquid Radiation Monitor Calibration Data
Sheets for Procedure 2612-R3 dated March 4,
January 4,

39.

1979

Surveillance Performance Form and Radiation Monitoring System Data Sheets
for Procedure 1302-3.11 dated March 19,

40.

1979, October 20, 1978 and1979

Surveillance Performance Form and Weekly Surveillance Check Data Sheets
dated March 24,

1979

41.

Unit 2 Out-of-Service Log

42.

Unit 1 Out-of-Service Log

43.

Work Request 0579

44.

Operation,

Maintenance Instructions and Parts Catalog for the Three Mile

Island Nuclear Power Plant Unit 2, "Operating Instructions for Radiation
Monitoring Systems," E-07582,
45.

Engineering Change Memo No.
July 28,

1976

S-9137,

Problem Report No.

2699 and Memo dated

1978

46.

Engineering Change Memos 4722 Rl,

47.

Field Questionnaire 1597; Letter dated August 7, 1978 and Memo dated
January 15,

1975,

and Technical Memo 668 dated March 18,

48.

Work Request 2761

49.

Procedure 2105-1.12,
dated November 9,

R2 and R3
1975"Radiation Monitoring System Setpoints," Revision 6

1978

3

50.

Primary Auxiliary Operators Log dated March 29,

51.

Procedure 2104-4.1,
dated March 14,

1979"Miscellaneous Liquid Radwaste Disposal," Revision 2,

1979

52.

Filter Efficiency Test (Ventilation Systems)

53.

System Description #36 Heating and Ventilation System, Auxiliary Building

54.

Work Request C1406 dated February 25,

55.

Information Provided by Met-Ed - Communications Systems and Maintenance56.

NUS Audit 3364 and cover letter HPS-79-053

57.

Procedure 1778,. "Emergency Equipment and Inventory Control," Revision 4,
dated January 26,

58.

1979

1978

Procedure 1302-5.24, "Environmental
dated December 19,

Monitor Calibration," Revision 3,

1974

59.

Surveillance Task Index dated June 19, 1978,

60.

Unit 1 Procedure Index dated July 16,

61.

TMI Environmental

62.

Form 1660-1,

Page 16

1979

Sample Collection Sample Sheet dated March 22,

TMI - Radiological

1979

Environmental Monitoring Program - Surface

Water dated March 1, 1979
63.

Work and Station Status Sheet dated March 31,

64.

Work and Station Status Sheet dated March 29, 197965.

NUREG-0558,

1979"Population Dose and Health Impact of the Accident at the Three

Mile Island Nuclear Station," Ad Hoc Population Dose Assessment Group (NRC,
EPA and HEW),
66.

Metropolitan Edison Company,
dated July 5,

67.

May 10, 1979
NRC Form-5 Request for Area TLD Badges,

1979

Three Mile Island Unit 2,

Surveillance Procedure 2302-SAI,"Meteorological

Instrumentation Calibration," Revision 1, 3/13/78
68.

Metropolitan Edison Company, Generation Maintenance System, Manual Performance
Form,

Procedure No.

2302-SAI,

dated 10/26/78

4

69.

Unit 1 Radiation Monitor Strip Chart from Recorder RMR-2

70.

Offsite Dose Calculation Logs (Enclosures 2,

71.

Procedure 1670.2,

3 and 4 of Procedure 1670.4)"Site Emergency Procedure," Revision 9 dated November 22,

1978
72.

Station Manager Testimony

73.

Information Transcribed From Unit 2 Emergency Status Board

74.

Telephone Message Forms,

NRC Region I Answering Service,

dated March 28,

1979
75.

Bell of Pennsylvania Itemized Call Listing, Three Mile Island, dated
April 13,

76.

1979,

for Calls Made on March 28,

1979

Dauphin County Civil Defense Draft Three Mile Island, Incident Log Covering
the Period March 28,

1979 through April 2,

1979

77.

Appendix II-A, Item 169

78.

Notes of Commonwealth of Pennsylvania Nuclear Engineer

79.

Appendix II-A, Item 286

80.

Memo dated June 7,
A. F. Gibson,

1979,

Leader,C. 0. Gallina, Investigation Specialist, to

TMI Radiological

Investigation Team,

Subject:

RADIOACTIVE GASES RELEASED FROM TMI ON THE MORNING OF MARCH 30,

1979

81.

Appendix I-A, Time 082.

Appendix I-A, Time 3-6 sec

83.

Appendix I-A, Time 13 sec

84.

Second Interim Report on the Three Mile Island Nuclear Station Unit 2(TMI-2)

85.

Accident, June 15,

1979

Third Interim Report on the Three Mile Island Nuclear Station Unit 2(TMI-2)

Accident,

July 16,

86.

Appendix II-A Item 8

87.

Acceptance Test Data

1979

5

88.

Burns and Roe Drawing 2045, Revision 19

89.

Informal Record of Tank Levels on March 29

90.

Unit 2 Chemistry Log

91.

Unit 2 Chemistry Log

92.

Burns and Roe Drawing 2487

93.

Burns and Roe Drawing 2485-WDL-551,

94.

Burns and Roe Drawing 2064, Revision 16

95.

Strip Chart for Unit 2 Reactor Building Pressure

96.

Burns and Roe Drawing 2632, Revision 9

97.

System Description

98.

Appendix I-A, Time 3 hrs.

99.

Unit 2 Radiation Monitor Strip Chart from Recorder HP-UR-1907

Revision 0

56 min.

100. Valve List, Issue 6, dated 6/29/77
101. Appendix I-A, Time 87 min.
102. Radiation Monitor Strip Chart from Recorder HP-UR-3236
103. Appendix I-A, Time 60 min.
104. Appendix I-A, Time 2 hrs.

56 min.

105. Appendix I-A, Time 6 hrs.
.106. AppendixI-A, Time 3 hrs.

53 min.

107.

4 min.

Condensate Pump Discharge Sample Analysis Sheet 3/28 1500

108. Steam Generator A Sample Analysis Sheet 3/28 0724
109. Condensor Vacuum Pump Discharge Sample Analysis Sheet 3/28 0542
110. Steam Generator B Sample Analysis, 3/30,
111. Appendix I-A, Time 5 hrs.

15 min.

112. Appendix I-A,

15 min.

Time 9 hrs.

1030

113. Appendix II-A, Item 177
114. Industrial Waste Treatment System Sample Results
115. Unit 2 Radiation Monitor Strip Chart from Recorder HP-UR-1901
116. Burns and Roe Drawing 2062

6

117.

Unit 2 Radiation Monitor Strip Chart from Recorder HP-UR-2900

118.

Unit 2 Radiation Monitor Strip Chart from Recorder HP-UR-1902

119.

System Description #17,

Reactor Coolant Makeup and Purification System

120. Appendix I-A, Time 4 hours
121.

Unit 2 Strip Chart for Makeup Tank Level

122.

Burns and Roe Drawing 2028

123.

Nash Engineering Co.

Inc.,"Instructions for Operation of Size 623C Waste

Gas Compressor"'
124.

Burns and Roe Drawing 2024, Revision 25

125. Appendix I-A, 4 hrs. 26 min.
126.

Appendix I-A, 9 hrs. 51 min.

127.

Burns and Roe Drawing 2027, Revision 24

128.

Reactor Coolant Bleed Holdup Tank Samples Analysis Sheets

129. Burns and Roe Drawing 2496,

Revision 10

130.

Maintenance Work Request Log

131.

Unit 2 Reactor Coolant Analysis Sheets

132. Sample Analysis Sheets for HP-R-219
133.

Radiation Management Corporation Counting Room Log

134. Appendix II-A, Item 131
135. Appendix II-A, Item 194
136. "Analysis of the Absorber and Absorbents from Three Mile Island Unit 2"
dated May 25,

1979,

by Nuclear Consulting Services,Inc.

137. Not used
138.

Industrial Waste Treatment System Log

139.

Industrial Waste Treatment System Log Sample Result Sheets

140. Appendix II-A, Item 249
141.

NRC RI Incident Message Forms

142. Appendix lI-A, Item 25
143. Appendix I-A, Time 0521

7

144. Results for reactor coolant sample taken at 0643 on March 28,

1979145. Appendix II-A, Item 33
146. Unit 1 Shift Foreman Log
147. Control Room Operator J's Notes
148. Procedure 1670.8, "Emergency Re-entry for Repair or Rescue," Revision 0
dated January 16,

1978

149. Appendix II-A, Item 21
150.

Shift Supervisor C's Notes

151.

Procedure 2202-2.6,"OTSG Tube Rupture," Revision 4 dated October 6,

1978

152. Unit 1 Radiation Monitor Strip Chart from Recorder Unit 1 RMR-5
153.

Emergency Control Center (ECC)

Log

154. Unit 1 Radiation Monitor Strip Chart from Recorder RMR-l
155.

Results for reactor coolant sample taken at 0850 on March 28, 1979

156.

NRC Incident Messageform C5

157. TLD Printout for 1-1 to 7-5-79 Background Badges
158. Procedure 1616,
April 21,

"Use of Respiratory Protection Devices," Revision 13 dated

1979

159.

Radiation Work Permit Log

160.

1st Quarter Licensee TLD Printout

161.

Procedure 1642,"Operation and Calibration of the Thermoluminescent Dosimetry

System," Revision 1 dated September 28,
162.

1977

Teletector Technical Manual for Total

163. NRC Memorandum:

M. Slobodien (Radiation Specialist) to G. Smith (Director,

Health Physics Operations),
March 28,

"Analysis of Personnel Dosimetry at TMI since

1979: dated May 16,

1979

164.

TLD data sheets dated May 21,

1979

165.

Bettis Sample Results

166. Three Mile Island Radiation Emergency Procedure 1670.2,
Procedures,"

Revision 9 dated November 22,

8

1978

"Site Emergency

167.

Three Mile Island Radiation Emergency Procedure 1670.4,
Calculations," Revision 3 dated February 15,168. Offsite Dose Calculation Log (Enclosures 2,"Radiological Dose1979
3 and 4 of Procedure 1670.4)

169.

Emergency Control Station (ECS)

Survey Log

170.

Pennsylvania Bureau of Radiological Health Sample and Laboratory Report
Sheets for Samples Analyzed 3/28/79

171.

TMI-l Wind Speed and Direction Chart (Recorder #42B)

172.

ECS Wind Data Summary Log

173.

Radiation Management Croporation Mobile Laboratory Sample Analysis Sheets
dated 3/28 - 3/30/79

174.

Undated Memorandum from Radiation Protection Supervisor, Relative to
Radiation Survey Records Following the March 28 Event

175. NRC Region I Incident Message Form
176. Three Mile Island Health Physics Procedure 1670.6,
Monitoring," Revision 2, dated January 16,
177.

"Off-site Radiological

1978

NRC Region I Mobile Laboratory Sample Analysis Log - Analytical Results
dated 3/28 - 3/30/79

178.

Radiation Management Corporation/Philadelphia Laboratory Analytical Result
dates 3/28/79 (received via telephone by Met-Ed)

179.

Preaccident Unit 2 Auxiliary Building Surveys

180. Metropolitan Edison Company Radiological

Environmental Monitoring Report -1978 Annual Report (prepared by Teledyne Isotopes, Westwood,
181.

Procedure 1643,

NJ)"Quality Assurance Tests for TLD Dosimetry System," Revision 0

dated April 11, 1977
182. TLD Calibration Data
183. Memorandum NBS to NRC (in preparation)

9

Appendix II-A
RadiologicalSequenceof
Events

Notes:(1) All events pertain
to Unit' 2 unlessotherwise specified(2) Times are approximate

TMI-2 RADIOLOGICAL SEQUENCE OF EVENTSAS OF JULY 11, 1979ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCES

1

0400

The reactor tripped.

1. Alarm printer

2

0400-0618

Reactor coolant pressure dropped from 2435 psig just after the reactor
1. Metropolitan
trip to 1600 psig at 0402 resulting in emergency core cooling system
Edison Interim
initiation. The pressure continued to drop until it reached approximately
Report Figure 121015 psig at 0420 where it remained relatively constant until about
2. RC wide range0530 when it decreased again reaching a low point of 660 psig at 0618.
pressure recorder

3

0408

The reactor building sump pump (WDL-P-2A)

4

0410

A second reactor building sump pump (WDL-P-2B)

5

0411

There was a reactor building sump high water level alarm.

1. Alarm printer

6

0415

The reactor coolant drain tank rupture disk burst at approximately 190
psig. A pressure rise of 1.4 psig was measured in the reactor
building.

1. Second Interim
Report Figure 4E

7

0426

The steam driven emergency feedwater pump was turned off.

1. NRC Operational
Sequence ofEvents

80438

An operator manually turned off the reactor building sump pumps
after up to 8,120 gallons had been pumped to the auxiliarybuilding.

1.
2.

II-A-1started.
started.

1. Alarm printer
1. Alarm printer

Alarm printerInterview of
AuxiliaryITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

3.

NOTE:

The quantity 8120 gallons was calculated by multiplying the runtime of each sump pump by 140 gpm (the capacity of each pump operating
singularly as determined by preoperational testing). With two pumps.
operating, the total capacity should be less than 280 gpm. In
addition, since the discharge of these pumps was probably to theauxiliary building sump tank instead of to the miscellaneous waste
holdup tank, as it was during preoperational testing, the line
size is reduced from 4 in. to 2 in. This would also result in a
reduced flow rate. By procedure, the sump pumps are normallyaligned to the miscellaneous waste holdup tank. The miscellaneous
waste holdup tank level, however, did not appear to change
significantly from 3/27 to 3/30. The primary auxiliary operator
on shift at the time of the reactor trip stated that the reactor
building sump pump discharge was aligned to the miscellaneous
waste houldup tank; however, it appears likely that thedischarge was actually aligned to the auxiliary building
sump tank. Procedural guidance allows the sump pumps to bealigned to the auxiliary building sump tank if the activity ofthe sump is low, if the miscellaneous waste holdup tank is full,or if both sump filters (WDL-F-8A/B) are at high differential
pressures.

II-A-2

Operator B(Int. 10)
Primary
Auxiliary
Operator Log(OP 2104-4.1)

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

90445-0515

A radiation/chemistry technician took a reactor coolant sample. The
boron analysis result was approximately 700 ppm.
He phoned the results
to the control room. The control room asked for another sample andanalysis. A radiation level survey of the sample was not made.

1. Interview ofRadiation/
Chemistry
Technician L(Int. 58)100450-0500

The reactor intermediate closed cooling water coolant letdown
monitor (IC-R-1091) increased from 2000 cpm to 3000 cpm.

1.

NOTE:

Chart recorder
HP-UR-3264

This indicates an increase in background radiation in thereactor building.

11

0500

The Supervisor, Radiation Protection and Chemistry was phoned at homeby the Unit 2 Superintendent, Technical Support, and told to report tothe site.

1. Interview ofSupervisor,
Radiation
Protection
and Chemistry
and Unit 2
Superintendent
Technical
Support(Int. 20 and
27)

12

0500

The circulating water pumps were turned off to switch the steam
generators onto the atomospheric relief valves.
Emergency feedwater
pump suction was to the hotwell.

1. NRC Operational
Sequence ofEvents

II-A-3ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

130515-0545

The intermediate closed cooling water monitor (IC-R-1092) increased fromIt reached the alarm setpoint (5000 cpm) at
3,500 cpm to 22,000 cpm.
0518.

1. Chart recorder
HP-UR-3236
2. Procedure
2105-1.12

NOTE:
14

0518

This indicates an increase in background radiation in the reactor
building.

The reactor building air sample particulate monitor (HP-R-227 (P))
reached the alarm setpoint (50,000 cpm).

1.
2.

NOTE:

Chart recorder
HP-UR-3236
Procedure 21051.12

This alarm corresponded to 9.4 E-8 uCi/cc. The licensee established this setpoint at 25% of an equilibrium concentration of
krypton-88 and rubidium-88 expected to be present in the reactor
building due to a 1 gpm reactor coolant leak with 0.1% failed
fuel.

15

0527

A control room operator isolated the B steam generator because operators suspected that a steam leak from this generator was causing an
increase in reactor building pressure.

1.

Unit 2 Foreman
Log

16

0540

The Supervisor, Radiation Protection and Chemistry arrived on site.

1.

Blue Badge
Access List
Interview ofSupervisor,
Radiation
Protection
and Chemistry(Int. 20)

2.

II-A-4ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTION

17

0542

The condenser vacuum pump exhaust was sampled.
radioactivity above background.

The results showed no

1. Licensee sample
result18

0545

Radiation/chemistry technicians took a reactor coolant sample. The
analysis result showed a boron concentration of approximately 400
ppm. A radiation level survey of the sample was not made. The
result was phoned to the shift foreman. Another radiation/chemistry
technician analyzed a separate sample to confirm the results.

1. Interviews ofSupervisor,
Radiation
Protection and

REFERENCE

Chemistry

Technicians
L and I
(Int. 20, 58,
42)
19

0545

The Supervisor, Radiation Protection and Chemistry arrived at theUnit 2 control room and was requested to change the charcoal cartridgeand particulate filter in the reactor building air sample monitor(HP-R-227).

1. Interview ofSupervisor,
Radiation
Protection
and Chemistry(Int. 20)

20

0550

A radiation/chemistry technician attempted to change the iodine cartridgeand particulate filter in the reactor building air monitor(HP-R-227), but the charcoal cartridgeiand water blew out under pressure.
The charcoal was reinstalled. This indicated the possibility of asteam environment in the reactor building atmosphere.

1. Interviews ofSupervisor,
Radiation
Protection andChemistry andRadiation/
Chemistry
Technician I(Int. 20, 42)

II-A-5

REFERENCEITEMDATE AND TIME
3/28/79

EVENT DESCRIPTION

21

0602

A radiation/chemistry technician took a reactor coolant sample for radio- 1. Interview ofRadiation/
Normalactivity analysis. A gross beta-gamma analysis showed 4.0 uCi/ml.
Chemistry
This result
activity for this analysis should be around 0.4 uCi/ml.
Technician L
was phoned to an operator in the control room. A gamma isotopic analysis(Int. 58)
was not done on this sample. A sodium analysis showed 0.153 ppm.
NOTE:

The sodium results indicate that no appreciable amount of sodiumhydroxide was present in the reactor coolant samples for which theboron results were much lower than expected.

220615-630

A telephone conference call to discuss the conditions of the plant wasplaced between the Vice President, Generation; the Station Manager; theUnit 2 Technical Support Superintendent; and a Babcock and Wilcox
representative.

1. Interview ofUnit 2
Superintendent,
Technical Support(Int. 27)

23

0618

Leakage through the pressurizer electromatic relief valve was stopped byclosing a block valve.

1. NRC Operational
Sequence ofEventsNOTE:

This stopped the loss of coolant. This valve continued to becycled throughout the day to control pressurizer level.

24

0622

The incore instrumentation area monitor (HP-R-213) appears to havedetected the initial release of fuel cladding gap activity.

1. Chart Recorder
HP-UR-1901

25

0624

The reactor building air particulate sample monitor (HP-R-227(P))
reached the alarm setpoint (50,000 cpm) for the second time since theincident.

1. Chart recorder
HP-UR-3236
2. Procedure 21051.12

II-A-6ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONNOTE:REFERENCE

This corresponded to 9.4 E-8 uCi/cc. The monitor had been above
the alarm setpoint for approximately 15 minutes from 0518 to
0533.

26

0626

The area monitor in the reactor building on the 347 ft. elevation near
the incore panel area (HP-R-213) reached the alarm setpoint (50 mR/hr).
Also, the reactor containment dome monitor started to respond to therelease of gap activity.

1. Chart recorder
HR-UR-1901
2. Procedure 21051.12

27

0630

The Supervisor, Radiation Protection and Chemistry, was asked to arrange
for an entry into the reactor building. The entry was not made due tothe shift change and subsequent increase in radiation levels.

1. Interview ofSupervisor,
Radiation
Protection andChemistry andRadiation/
Chemistry
Technician I(Int. 20, 42)

28

0630

A radiation protection foreman arrived on site.

1. Interview ofRadiation
Protection
Foreman B(Int. 21)
2. Blue Badge
Access List

II-A-7ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

29

0630-0700

A radiation/chemistry technician surveyed the Unit 2 auxiliary building
1.
using a portable beta-gamma survey meter (RO-2).
He reported rapidly increasing levels of radiation during approximately twenty minutes it took
to perform this survey. Initially, on the 305 ft. elevation the generalarea dose rates near the valve alley were 100 mR/hr, and the door to
makeup purification valve room was 200 mR/hr. The door to the makeup
tank room read 1 R/hr. Another radiation/chemistry technician, using
a GM radiation level meter (Teletector), measured 10 R/hr just inside
the makeup tank room on the nitrogen line. The dose rate in the waste
gas decay tank room was 10 to 20 mR/hr, and at the radwaste operating
panel area, the dose rate was 50 mR/hr. A second check on the makeup
purification valve room door caused the portable beta-gamma survey
meter (RO-2) to read off-scale (5 R/hr). The radiation/chemistry technician
was advising personnel to leave the area when the site radiation emergency
signal was sounded and an announcement was made to evacuate the Units 1
and 2 auxiliary buildings. As the technicians exited the building, theportable beta-gamma survey meter (RO-2) read off-scale at a distance of ten
feet from the hallway door to the makeup tank room. This radiation/chemistry
technician was contaminated to levels of 30 mR/hr.

Interview ofRadiation/
Chemistry
Technician 0(Int. 78)

30

0635

The containment dome monitor (HP-R-214) reached alert.

Chart recorder
HP-UR-1901
Procedure2105-1.12

1.
2.

31

0643

A reactor coolant sample was collected. The analysis results showed
1. Licensee sample
140.73 uCi/ml gross gamma activity. The Unit 1 nuclear sample room, which
resultcontained Unit 2 reactor coolant sample lines, area radiation monitor
2.
Interview of(RM-G3) reached the alarm setpoint (2.5 mR/hr). The reactor coolant
Radiation/
sample recirculation was stopped at the sample room by a radiation/
Chemistry
chemistry technician.
Technician L(Int. 58)

II-A-8ITEM•DATE AND TIME
3/28/79

EVENT DESCRIPTION
NOTE:

32

0645

REFERENCE

The alarm showed rapidly increasing radiation levels from thereactor coolant sample lines.

A radiation/chemistry technician was directed by the Supervisor, Radi1. Interview ofation Protection and Chemistry to call in the Radiation Protection
Radiation/
Supervisor and a radiation protection foreman.
Both calls were completed.
Chemistry

Technician U(Int. 78)
33

0648

The Unit 1 hot machine shop area monitor (RM-G4) reached the alarm setpoint (2.5 mR/hr). Unshielded reactor coolant sample lines ran throughthis room. A survey identified 1.5 R/hr on the reactor coolant samplelines and 500 mR/hr general area dose rates in the hot machine shop.
Unit 2 control room was notified.

1. Interviews ofSupervisor,
Radiation
Protection andChemistry,
Radiation
Protection
Foreman B
and Radiation/
Chemistry
Technician I(Int. 20, 21
and 42)

340648

The station vent stack particulate monitor (HP-R-219(P))
alarm setpoint (1,300 cpm).

1. Chart recorder
HP-UR-1907

NOTE:

reached theThis corresponded to 2.5 E-9 uCi/cc or approximately a release rateof 0.3 uCi/sec, the technical specification instantaneous release
rate limit for iodine-131 and particulates.

II-A-9

REFERENCEITEMDATE AND TIME
3/28/79

EVENT DESCRIPTION

35

0650

1.
A radiation protection foreman surveyed the auxiliary building using a
He entered through the model
portable beta-gamma survey meter (RO-2).
room door. The sample line to the reactor building air sample monitor(HP-R-227) read 10 mR/hr at contact. The makeup tank room door read 1 R/hr.
The hall on 305 ft. and 281 ft. elevations had no unusual levels of radiation. The Unit 2 health physics control point, auxiliary building stairwell,
decay heat vaults, isolated areas near drains and main halls had water
2 to 3 inches deep at drains. He waded through a few inches of water on
the floor. When he exited the auxiliary building there was no contamination on his clothes or body.
NOTE:

36

0651

This indicated that the water on the auxiliary building floor may
not have been highly contaminated at this time.

The auxiliary building exhaust duct B particulate monitor (HP-R-228 (P))
reached the alarm setpoint (4,000 cpm).

NOTE:

Interview ofRadiation Protection Foreman
B (Int. 21)

1. Chart recorder
HP-UR-2900
Procedure 21052.
1.12This corresponded to 7.5 E-9 uCi/cc or a release rate of 0.3
uCi/sec.

370654

Reactor coolant pump 2B was started.

38

0654

The condenser vacuum pump discharge monitor (VA-R-748)
setpoint (2,000 cpm).

II-A-IO

1. NRC Operational
Sequence ofEventsreached the alarm

1. Chart recorder
HP-UR-3236
2. Procedure 21051.12

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTIONNOTE:This corresponded to a release rate of 0.024 uCi/sec of iodine-131
and particulates, a technical specification quarterly average limit.

39

0654

The Unit 2 Superintendent, Technical Support, directed that there be an
emergency boration.

1. Interview ofUnit 2 Superintendent ofTechnical
Support(Int. 27)

40

0655

A Site Emergency was declared by the shift supervisor based on the alarmsof process and area radiation monitors.

1. Interview ofUnit 2
Superintendent
Technical Support(Int. 27)
Unit
1 CRO Log
2.
.
3.
3. Emergency Status
Board

41

0655

The Supervisor, Radiation Protection and Chemistry put a radiation/
chemistry technician in charge of the Emergency Control Station and
The Emergency Control
directed him to form teams per the Emergency Plan.
Station was established at the 305 ft. elevation in the Unit 1 control
building at the Unit 1 health physics control point.

1. Interview of

420655-0800

1.
The Supervisor, Radiation Protection and Chemistry directed a radiation/
chemistry technician to go to the Unit 2 auxiliary building to take radiation measurements and air samples. The technician started an air sampler
near the reactor building air sample monitor (HP-R-227) and noted radiation levels were increasing. He measure 2 R/hr at the hallway door to the

II-A-II

Supervisor,
Radiation
Protection
and Chemistry(Int. 20)
Interview ofRadiation/
Chemistry
Technician U(Int. 78)ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

makeup tank room. While collecting an air sample near the station vent
stack monitor (HP-R-219) on the 328 ft. elevation, the dose rates began
rapidly increasing from 50 to 175 mR/hr. Several local radiation alarms
sounded, and the technician left the area. At this point another radiation/
chemistry technician returning from the 281' elevation reported 1 to 3 in.
of water on the floor on that elevation. The radiation/chemistry technician
who had taken the air samples was found to be contaminated.
43

0655

An engineer arrived at the control room and began offsite dose
calculations.
The containment dome monitor (HP-R-214) reading at thistime was recorded as 300 R/hr by the engineer.

NOTE:

44

0655

This monitor was later determined by the NRC to have been initiallymisread.
The actual reading appears to have been approximately
400 mR/hr.

The Unit 1 liquid discharge from the B waste evaporator storage tank
was stopped. The Unit 1 gaseous discharge authorized by permit 33-79-G
was stopped.
NOTE:

1. Interview ofEngineer C(Int. 48,
127, 174)
2. Chart recorder
HP-UR-1901
3. Procedure 21051.12

Both of these releases were started and stopped intermittently
during the next few days.
These releases were of materials thatwere accumulated prior to the Unit 2 accident and the releases did
not contribute significantly to total release from the site.

II-A-12I.

Unit 1 CRO
Log

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

45

0656

The circulating water pumps were restarted. This should have switchedsteam generators back to the turbine by-pass mode of operation. B
steam generator was unisolated for 7 seconds which was long enough to
put the condenser vacuum pump discharge monitor into alarm.

1. NRC Operational
Sequence of Events

46

0658

The Supervisor, Radiation Protection and Chemistry moved to the radia1.
tion monitoring system panel and observed that the condenser offgas
monitor (VA-R-748) was in alarm and the containment dome monitor (HP-R-214)
was in alert and increasing,

NOTE:

47

0700

Interview ofSupervisor,
Radiation Protection andChemistry(Int. 20)

The condenser offgas monitor in the alarm mode indicated a potential
primary to secondary system leak. The containment dome monitor in thealert mode indicated degradation of the fission product boundary.

The area monitor near the personnel access hatch inside the reactor
building (HP-R-211) reached the alarm setpoint (50 mr/hr).

1. Chart recorder
2.

HP-UR-1901
Procedure 21051.12

48

0700

Two Emergency Repair Party Teams were formed and standing by one at theEmergency Control Station and one outside the Unit 2 control room.

1.

49

0700

A radiation/chemistry technician set up Emergency Control Station
communications.
He was subsequently relieved by a radiation protection
foreman.

I.. Interview ofRadiation/Chemistry Technician
I (Int. 42)

II-A-13

Discussion with
Maintenance Foremen B and G

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

50

0700-0713

The reactor coolant letdown monitor (MU-R-720) gross activity increased
from 4.5 E 4 cpm to greater than 1 E 6 cpm reaching the alarm setpointof 5 E 5 cpm at 0712.

1. Chart recorder
HP-UR-3264
2. Procedure 21051.12

510700-0730

1. Interview ofBecause the air radioactivity increased to greater than 3 E-10 uCi/cc,
Radiation/Chemthe control point for entry into the auxiliary building was moved fromistry Technician
the doors to the auxiliary building to the hallway in the control building.
I (Int. 42)

52

0701

The fuel handling building exhaust particulate monitor upstream of thefilters (HP-R-221A (P)) reached the alarm setpount (8000 cpm).

1.
2.

NOTE:

53

0701

This corresponded to 1.5 E-8 uCi/cc or 0.3 uCi/sec. This monitor
was probably measuring effluent and direct radiation in the areaof the detector outside the duct.

The fuel handling building exhaust iodine monitor downstream of filters(HP-R-221B (I)) reached the alarm setpoint (200,000 cpm).

NOTE:

Chart recorder
HP-UR-1907
Procedure 21051.12

This corresponded to 5.1 E-8 uCi/cc-sec or 50% of the 0.30 uCi/sec
technical specification release rate limit. This monitor wasprobably measuring effluent and direct radiation in the area ofthe detector outside the duct.

II-A-141.

Chart recorder

2.

HP-UR-1907
Procedure 21051.12

REFERENCEITEMDATE AND TIME
3/28/79

EVENT DESCRIPTION

54

0702

1. Interview of
The shift supervisor phoned the Pennsylvania Emergency Management Agency
Unit Super(formerly the Pennsylvania State Council of Civil Defense) Duty Officer
intendent
and informed him of the Site Emergency and requested that the Pennsylvania(Int. 99)
Bureau of Radiological Health be notified.
2. Emergency Status
Board
3. Pennsylvania
Emergency Management Agency Log

55

0704

The reactor building purge exhaust duct A particulate monitor(HP-R-225 (P)) reached the alarm setpoint (10,000 cpm).

NOTE:

56

0704

1. Chart recorder
HP-UR-2900
2. Procedure 21051.12

This corresponded to 1.9 E-8 uCi/cc or 0.3 uCi/sec. This monitor
was probably measuring direct radiation in the area of thedetector outside the duct.

The answering service
An engineer phoned the NRC Region I Office.
received the call. The answering service attempted to contact the duty
officer at home.. There was no answer. The answering service paged theWhile waiting for beeper response, theduty officer on his beeper.
answering service recalled the duty officer at home and was informed that
he was enroute to the office. The beeper call was not answered (notreceived by duty officer's beeper). The answering service then called
the Deputy Director's home and was informed that he.was enroute to theoffice. The answering service paged the Deputy Director. His beeper did
not activate.

II-A-15

1. Interview of
Station Manager(Int. 41)
2. Emergency Status
Board
3. NRC telephonemessage
4. Bell of PA
itemized calllisting dated
4/13/79ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

57

0705

The. Station Manager arrived at the Unit 2 control room and assumed theposition of Emergency Director.

1. Interviews ofStation Manager
and Supervisor,
Radiation Protection and Chemistry (Int. 71,
41 and 20)

58

0705

An engineer attempted to phone the Metropolitan Edison Vice President,
He could not be reached via any of the numbers listed
Generation.
in the emergency procedures.

1. Interview ofEngineers
I and A
(Int. 70)

NOTE:

59

0706

The Vice President of Generation had been contacted earlier in themorning in Philadelphia and had participated in a conference callconcerning the incident.

The reactor building purge exhaust duct B particulate monitor(HP-R-226 (P)) reached the alarm setpoint (10,000 cpm).

1.
2.

NOTE:

This corresponded to 1.9 E-8 uCi/cc or 0.3 uCi/sec. This monitor
was probably measuring direct radiation in the area of thedetector outside the duct.

II-A-16

Chart recorder
HP-UR-2900
Procedure 21051.12

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

60

0709

The site phoned the Department of Energy Radiological Assistance Plan
Office at Brookhaven National Laboratory and informed them of the Site
Emergency.

1. Interview ofPlant Management and Unit
1 Superintendent(Int. 41)
2. Discussion withDOE
3. Bell of PA
itemized calllisting dated
4/13/79

61

0709

A senior engineer attempted to contact the Manager, Generation
Operations, Metropolitan Edison. He was not in the office. Subsequently,
the Manager of Generation Engineering was called and notified that a
Site Emergency had been declared.

1. Interview ofPlant Manager
and Unit Superintendent(Int. 41)

62

0709

An engineer called Dauphin County Civil Defense to notify theagency of the Site Emergency.

1. Discussion with
Civil Defense
Director

63

0710

The first offsite dose calculations were completed by an engineer.
1.
The initial calculations indicated a total body exposure rate of 40 R/hr
in Goldsboro (1.3 miles west of the plant). The calculations were verified
shortly thereafter by the Supervisor, Radiation Protection and Chemistrybut, together with the engineer, he concluded that the calculations
were overly conservative because the reactor building pressure wasonly about 2 psig, and hence the leak rate was much lower than assumed

II-A-17

Interview ofEngineer C(Int. 48, 127,
174);

ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

in the source term estimates (approximately 50 psig) used in the offsite
dose calculation procedure (1670.4).
Following these calculations, the licensee
dispatched survey teams in the downwind direction to the west side of theisland (where the initial survey was made at 0748) and toGoldsboro (where the initial survey was made at 0832).
64

0710

A Pennsylvania Bureau of Radiological Health nuclear engineer was calledby the Pennsylvania Emergency Managemient Agency and notified that TMI
had declared a Site Emergency.
The engineer was instructed to call thesite to obtain details.

1. State Engineers
Notes
2. Discussion with
State Engineer

65

0712

The station vent stack gas monitor (HR-P-219 (G))
setpoint (10,000 cpm).

1. Chart recorder
HP-UR-1907
2. Procedure 21051.12

NOTE:

66

0712

reached the alarmThis corresponded to 2.8 E-4 uCi/cc or approximately the quarterly
average technical specification limit for xenon-133 releases. This
monitor sampled the air going out the vent stack; however, since
the monitor was located inside the auxiliary building on the 328 ft.
elevation, the monitor reading probably reflected an increase inbackground radiation and an increase in sample activity.

The Pennsylvania Bureau of Radiological Health nuclear engineer attempted
to call the site but was not able to get through the switchboard to thecontrol room.
He left his phone number with the switchboard operator
and asked to have the control room call him.

II-A-18

1. State Engineer's
Notes

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

67

0713

An individual at the site phoned Radiation Management Corporation, thelicensee's health physics consultant firm. The phone was answered by
the Philadelphia Electric Company load dispatcher.

1. Interview ofEngineers I
and A(Int. 70)

68

0713

Reactor coolant pump 2B stopped.

1. Alarm printer

69

0714

The reactor building hydrogen purge duct iodine monitor (HPR-229(I))
reached the alarm setpoint (4 E 5 cpm).
This monitor was adjacent tothe vent monitor (HP-R-219) on the 328 ft. elevation and was probably
responding to increasing background radiation. There was no known flowfrom this system.

1. Chart recorder
HP-UR-1907
2.
Procedure 21051.12

NOTE:

70

0715

The strip chart for reactor building hydrogen purge flow showedintermittent flow of up to 15 cfm on the morning of 3/28. Discussions
with licensee operating personnel and instrument and controls personnel indicate that the flow recorder was detecting pressurechanges caused by varying flows in the auxiliary building and fuelhandling building ventilation systems and was not indicating purge
flow. For leakage to occur through this path, four butterfly valveswould have had to leak. The radiation monitor samples this exhaust
downstream of the four valves.

The shift supervisor returned the call to the Pennsylvania Bureau ofRadiological Health nuclear engineer and informed him that a transient
had occurred, the plant was shutdown, safeguards were operational, andthere was a slight pressure increase in reactor building but no offsite
releases.

II-A-19

1. Discussion with
State engineer
2.

Interview of the
Plant Manager andStation Superintendent (Int. 99)

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

71

0715

A radiation/chemistry technician measured dose rates of 1 to 2 R/hr
fifteen feet from the reactor building personnel access hatch.

1. Interview ofRadiation/Chemistry Technician
L (Int. 58)

72

0715

A rapid increase in readings of most radiation monitors on stripcharts
HP-UR-1907, HP-UR-1901 and HP-UR-2900 is attributed to the EMOV being
opened. The pathway was probably from the reactor coolant drain tankto the vent header and out through a leak in the vent header into theauxiliary building.

1. NRC Operational
Sequence of Events

NOTE:

This is a normal access route.
have been due to noble gases.

The high level exposure rate may

73

0715

74

0715-0719

The Emergency Director assigned an auxiliary operator to attend thetelephone switchboard.
A radiation/chemistry technician toured the auxiliary building andinformed emergency workers (Repair Party and Monitoring Team) in the
He phoned the control room to inform them thatbuilding to evacuate.
the auxiliary building had been evacuated.

75

0716

The auxiliary building exhaust duct B gas monitor (HR-P-228(G))
the alarm setpoint (20,000 cpm).

NOTE:

reachedThis corresponded to 6.8 E-4 uCi/cc or approximately the quarterly
average technical specification limit for xenon-133 releases. The
monitor was probably measuring effluent and direct radiation in thearea of the detector outside the duct.

II-A-20

1. Station Manager
testimony
1. Interview ofRadiation/Chem-.
istry Technician
I (Int. 42)
1. Chart recorder
HP-UR-2900
2. Procedure 21051.12

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

76

0718

The reactor building purge exhaust duct B gas monitor (HP-R-226(G))
reached the alarm setpoint (80,000 cpm).

1. Chart recorder
HP-UR-2900
2. Procedure 21051.12

NOTE:

This corresponded to 2.3 E-3 uCi/cc or approximately the technicalspecification limit for xenon-133 releases. The monitor wasprobably measuring direct radiation in the area of the detector
outside the duct.

77

0718

The Pennsylvania State Police was contacted by an individual in thecontrol room.

1. Interview of
Station Manager
and Unit Supertendent (Int. 41)

78

0719

The reactor building purge exhaust duct A gas monitor (HP-R-225(G))
reached the alarm setpoint (80,000 cpm).

1. Chart recorder
HP-UR-2900
2. Procedure 21051.12

NOTE:

79

0719

This corresponded to 2.3 E-3 uCi/cc or approximately the quarterly
average technical specification limit for xenon-133 releases. This
monitor was probably measuring direct radiation in the area of thedetector outside the duct.

The reactor building purge unit area monitor (HP-R-3236)
alarm setpoint (20 mR/hr).

II-A-21reached the

1. Chart recorder
HP-UR-1902
2. Procedure 21051.12

REFERENCEITEMDATE AND TIME
3/28/79

EVENT DESCRIPTION

80

0720

The site phoned American Nuclear Insurers.
phone call.

81

0720

The Unit 1 fuel handling building air particulate monitor (RM-A4)
the alarm setpoint (I E-8 uCi/cc).
NOTE:

82

0720

0723

reached

1. Interview of
Station Manager
and Unit Superintendent (Int.
41)
1. Unit 1 Foreman Log

This monitor was probably measuring radioactive materials spreadingfrom Unit 2.

The reactor containment dome monitor (HP-R-214)
NOTE:

83

There was no answer to theincreased to 8 R/hr.

1. Chart recorder
HP-UR-1901

This monitor is an environmentally protected ionization chamber
located on top of the elevator shaft enclosure. It has a 2 inch
thick lead shield to which the manufacturer has assigned an
attenuation factor of 100. This means that an indicated 8 R/hr
was actually 800 R/hr. Because of the shielding, the monitor does
not measure the radiation from xenon-133, which has a very low
energy gamma, 80 kev.

The fuel handling building exhaust gas monitor downstream of filters(HP-R-221B(G)) reached the alarm setpoint (30,000 cpm).

II-A-22

1. Chart recorder
HP-UR-1907
2. Procedure 21051.12

1TEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONNOTE:REFERENCE

This corresponded to 8.5 E-4 uCi/cc or approximately the quarterly
average technical specification limit for xenon-133 releases. This
monitor was probably measuring effluent and direct radiation in thearea of the detector outside the duct.

84

0723

The fuel handling building exhaust unit area monitor (HP-R-3240)
the alarm setpoint (20 mR/hr).

85

0724

A General Emergency was declared by the Station Manager based on thegreater than 8 R/hr reading on the reactor containment dome monitor(HP-R-214).
This reading corresponded to 800 R/hr when corrected forshielding around the detector.

1. Emergency Status
Board

86

0724

The Unit 2 A steam generator was sampled in the Unit 1 sample room;
the results recorded as the B steam generator included:

1. Licensee sample
result2. Interviews ofChemistry Foreman
B and Supervisor, Radiation
Protection andChemistry(Int. 39, 20)

cobalt-58 xenon-135 -reached1.5 E-6 uCi/ml
1.8 E-7 uCi/ml

The results were phoned to the control room.
NOTE: It appears the A and B sample point identifications in the
Unit 1 sample room were reversed.
87

0725

The fuel handling building exhaust gas monitor (HPR-221A(G))
of filter reached the alarm setpoint (40,000 cpm).

II-A-23

upsteam

1. Chart recorder
HP-UR-1902
2. Procedure 21051.12

1. Chart recorder
HP-UR-1907
Procedure 21052.
1.12

ITEM

DATE AND TIME
3/28/79

NOTE:

88

0728

0729

This corresponded to 1.1 E-3 uCi/cc or approximately thequarterly average technical specification limit for xenon-133
releases. This monitor was probably measuring effluent anddirect radiation in the area of the detector outside the duct.

The auxiliary building exhaust duct A gas monitor (HPR-222(G))
the alarm setpoint (20,000 cpm).

NOTE:

89

REFERENCEEVENT DESCRIPTION'

reached1. Chart recorder
HP-UR-2900
2. Procedure 21051.12

This corresponded to 5.6 E-4 uCi/cc or approximately thequarterly average limit for xenon-133 releases. This
monitor was probably measuring effluent and direct radiationin the area of the detector outside the duct.

The reactor building purge exhaust duct A iodine monitor (HP-R-225(I))
reached the alarm setpoint (200,000 cpm).

1.
2.

NOTE:

90

0730

Chart recorder
HP-UR-2900
Procedure 21051.12,

This corresponded to 5.1 E-8 uCi/cc-sec or the technicalspecification limit of 0.3 uCi/sec. This monitor wasprobably measuring direct radiation in the area of thedetector outside the duct.

The first onsite survey team, designated as alpha, and consistingof two radiation/chemistry technicians, was dispatched to the west sideof-the island (downwind direction). A radiation survey was performed at
0748 near the screen house.

TT-A-9A

I . ECC Log
2. Interview ofRaditation/Chemistry Technician
D (Int. 82)

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

91

0730

An engineer recalled the Dauphin County Civil Defense to informthem of the General Emergency.

1. Discussion with

A plant liquid effluent sample was taken from sample point at the plantliquid effluent discharge monitor (RM-L7).

1.

92

0730

Civil Defense
Director
Interview ofRadiation Protection Foreman
B and Radiation/Chemistry
Technician
C' (Int. 21 and

82)
NOTE:
93

0730

This sample may not have been analyzed.
available.

No record is

While a radiation/chemistry technician held a survey meter on the sample
line on the reactor building air sample monitor (HP-R-227) the reading went from 50 mR/hr to 1 R/hrl A radiation protection foreman phoned
the control room and requested that the sample flow be turned off.

1. Interview ofRadiation Protection Foreman
B (Int. 21)

94

0730

Offsite survey team "charlie," consisting of a radiation/chemistry
technician and an auxiliary operator, was dispatched to Goldsboro.
The initial survey was performed at this location at 0832.

1. Interview ofRadiation/Chemistry Technician
C (Int. 82)

95

0730

An engineer was assigned to watch the station vent monitor (HP-R-219)
The engineer performed
and reactor containment dome monitor (HP-R-214).
this duty until about 1100.

1. Interview ofEngineer H(Int. 97)

II-A-25ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

96

0731

The auxiliary building access corridor area monitor (HP-R-232)
the alarm setpoint (2 mR/hr).

reached1. Chart recorder
HP-UR-1902
2. Procedure 21051.12

97

0734

An onsite survey team was sent to survey the Unit 2 auxiliary building.
The results reported were:

1. Interviews ofRadiation Protection Foremen
A and D
(Int. 11 and 18)
2. ECC Log

I•

The spent fuel resin area located at the 305 ft. elevation had very
The GM radiation level survey meter (Teletector)
high dose rates.
moved toward full scale on the 1000 R/hr scale.

2.

Outside the makeup tank room the dose rates were 1 R/hr.

3.

The spent resin storage'tanks read 3 R/hr (normally the dose rates
were 0.1 mR/hr).

NOTE:

Since the resins had not been used,
was not from the resin.

this indicated the radiation
was alarming.

4.

The reactor building air sample monitor (HP-R-227)

5.

There was water on the floor of the 281 ft. elevation.

II-A-26ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

98

0735

The site called the Department of Energy Radiological Assistance Plan
Office to inform them of a General Emergency. The Department ofEnergy was subsequently requested by the Commonwealth of Pennsylvania
at 1118 to provide assistance.

1. Interview of
Station Manager
and Unit Superintendent(Int. 99)
2. Bell of PA
phone list
dated 4/13/79
3. Discussion with
DOE

99

0735

The station vent stack iodine monitor (HP-R-219(I))
setpoint (160,000 cpm).

1. Chart recorder
HP-UR-1907
2. Procedure 21051.12

NOTE:

reached the alarmThis corresponded to 2.8 E-7 uCi/cc-sec or the technical specification limit of 0.3 uCi/sec.

100

0735

A shift supervisor contacted the Pennsylvania Emergency Management Agency
and reported the escalation to a General Emergency. The supervisor
reported failed fuel and a small offsite release.

1. Pennsylvania Emergency Management
Agency Log

101

0735

The Radiation Protection Supervisor arrived on site and assumed control.of the Emergency Control Station.

1. Interview ofSupervisor, Radiation Protection(Int. 22)

II-A-27ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

102

0736

The Dauphin County Civil Defense was contacted by the Supervisor, Radiation Protection and Chemistry, to verify that the Emergency:Director,
Dauphin County Civil Defense, had been personally notified. He spokewith the Director, Dauphin County Civil Defense.

1. Interview ofSupervisor, Radiation Protection
and Chemistry(Int. 20)
2. Discussion with
Civil Defense
Director

103

0737

A Metropolitan Edison representative from Reading, Pennsylvania, phoned
NRC Region I again. He left word with the answering service that a
General Emergency had been declared.

1. NRC telephonemessage1040737

1. Interview ofYork Haven (hydroelectric plant) was called by an individual in the conStation Manager
trol room and instructed to turn on the radiation survey meter and to startand Unit Supertaking readings.
intendent (Int.
41)

105

0737

The reactor building purge exhaust duct B iodine monitor (HPR-226(I))
reached the alarm setpoint (20,000 cpm).

NOTE:

This corresponded to 5.1 E-8 uCi/cc-sec or the technical specification limit of 0.3 uCi/sec. This monitor was probably measuring
direct radiation in the area of the detector outside the duct.

II-A-28

1. Chart recorder
HP-UR-2900
2. Procedure 21051.12

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

106

0738

The NRC Region I duty officer was paged for a second time. The duty
He was approximately 5 minutes from theofficer's beeper activated.
Region I office.

1. Discussion with
NRC Duty Officer

107

0740

A representative of the Pennsylvania Bureau of Radiological Health called
the control room on a direct outside line using a phone number which had
been previously given to the Commonwealth for such events. The Unit 2
Superintendent, Technical Support, briefed Pennsylvania Bureau ofRadiological Health on the situation. The phone line was left open.

1.

Interview ofUnit TechnicalSupport Superintendent(Int. 27)

108

0740

The site phoned the NRC, Region I, to inform them that a General
Emergency had been declared. The answering service informed the callerthat they had been unable to contact designated Region I individuals, but
the regional office would open the switchboard at 0745 to receive any
messages.

1.
2.

NRC telephonemessage
Interview of
Station Manager
and Unit Super-3.

intendent(Int. 99)
Bell of PA
itemized call list
dated 4/13/79

109

0741

The Emergency Control Station established communications with the Unit 1
control room.

1. ECC Log

110

0748

The first out-of-plant radiation surveys were made on the island. The
exposure rate at the screen house (near west shore of island) was lessthan 1 mR/hr. This measurement was used to revise the initial predictionof offsite radiation levels in Goldsboro.

1. Emergency Status
Board

II-A-29ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

111

0750

The NRC Region I established telephone contact with the Unit 2 control
room.

1.

Discussion with
NRC Branch Chief

112

0755

An offsite survey team reported less than 1 mR/hr at both the north gate
and Route 441 at the Observation Center.

1.

Licensee notes

113

0756

The reactor building was isolated by high reactor building pressure
at 4 psig.

1.

Alarm printer

114

0800

Director, Region I NRC notified Acting Director, Inspection and Enforcement of the incident.

1.

Discussion with
Director, Region
I

115

0800

A radiation/chemistry technician toured the auxiliary building basement
aMd found water on the floor around drains, but concluded it was nothighly radioactive based on direct readings with a high range beta-gamma
The Teletector indicated less thansurvey instrument (Teletector).
1 mR/hr on the water.

1.

Interview ofRadiation/Chemistry Technician
I (Int. 42)

116

0800

The auxiliary building and fuel handling building exhaust fans werestarted.

1.

Licensee notes

117

0800

The auxiliary building exhaust duct B iodine monitor (HPR-228(I))
reached the alarm setpoint (100,000 cpm).

1.

Chart recorder
HP-UR-2900
Procedure 21051.12

2.

NOTE:

This corresponded to 2.6 E-8 uCi/cc-sec or 0.3 uCi/sec.

II-A-30ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

118

0800

A radiation protection foreman surveyed the Unit 2 auxiliary building
vent for gamma radiation with a GM survey meter, E-520 (closed window).
The stack read 4 mR/hr on contact with the vent pipe.

1. Interview ofRadiationProtection Foreman
B (Int. 21)

119

0800

The Region I NRC Incident Response Center was activated.

1. Discussions with
NRC Branch Chief

120

0800

The open phone line established between the Unit 2 control room and NRC
Region I was transferred to the Region I Incident Response Center. An
open line was established between Unit 1 control room and Region I Incident Response Center.

1. Discussions with
NRC Branch Chief

121

0800

A survey team was dispatched to Middletown to survey a train that had
passed the site after the incident.

NOTE:

122

0805

1.
2.

Licensee notes
Interview ofRadiation/Chemistry Technicians
X, J and F(Int. 51)

After the team was dispatched, an individual from the control room
called the train yard and said the trains could be allowed to continue without a survey.

The NRC Headquarters Incident Response Center was activated.

1. Discussions with
NRC Branch Chief

123

0807

The model room door between the auxiliary building and the fuel handlingbuilding was recorded as closed.

II-A-31

1.

ECC LogITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONNOTE:REFERENCE

When open, this door was a pathway for airborne radioactivity tomove from Unit 2 to Unit 1.

124

0810

The NRC Region I Incident Response Center was fully manned and activated.

I.

125

0812

The result of the first air sample collected on the island outside of theplant (counted in the field using a portable sodium iodine single channel
analyzer, SAM-2) waj less than minimum detectable activity of 5 E-9
uCi/cc (5,000 pCi/m ) between GE-8 and GE-9. Locations GE-8 and GE-9
are respectively 700 feet west south west and 750 feet west northwestof the Unit 2 plant vent.

1. Emergency Status
Board
2. Interview with
the Radiation
Protection
Supervisor(Int. 20)

NOTE:

Discussion with
NRC Branch Chief

The 10 CFR 20, Appendix B, concentration for iodine-131 in unrestricted areas is 1 E-10 uCi/cc.

126

0815

Recirculation of a sample line for a reactor coolant sample was begun.
Dose rate at contact with the sample lines were measured to be 200 R/hrwith a high range GM survey meter (Teletector).

1. Interview ofRadiation/Chemistry Technician.U (Int. 78)

127

0815

Two members of the emergency Repair Party Team and a foreman
entered the auxiliary building to close the model room door
through which air was flowing from Unit 2 to Unit 1.
It took about
ten minutes. The three individuals were found to be contaminated onleaving the auxiliary building.

1.

NOTE:

There appears to be a long-standing problem with balancing ofventilation systems to prevent flow between Unit 1 and Unit 2.

II-A-32

2.

Interview ofMaintenance
Foreman B(Int. 21)
ECC Log

REFERENCEITEMDATE AND TIME
3/28/79

EVENT DESCRIPTION

128

0819

The Unit 1 fuel handling building air particulate monitor (RM-A4)
NOTE:

alarmed.

1. ECC Log

This corresponds to I E-8 uCi/cc.

129

0820

The radiation protection foreman returning from the roof of the Unit 2
auxiliary building through the 305 ft. elevation became contaminated to alevel of 20 mR/hr as measured with a GM survey meter (E-520), closed
window. He concluded that the contamination appeared to be due to airborne radioactive materials since he was evenly contaminated from head
to foot.

1. Interview ofRadiation Protection Foreman
B (Int. 18)

130

0826

The letdown high temperature alarms were received.

1. Alarm printer

NOTE:
131

132

0830 - 0900

0830

This should have isolated the letdown and purification demineralizers.

The station vent stack monitor (HP-R-219) charcoal cartridge was changed
by a radiation/chemistry technician. The particulate filter and holder
were missing. The general area dose rate near station vent stack monitor(HP-R-219) was 2 R/hr. The technician was alone and wore a respirator
with a particulate filter. The change took 2 minutes and he received adose of 395 mrem.

1. Interview ofSurvey team "bravo" was dispatched to Goldsboro.

1. Interview ofRadiation/
Chemistry
Technician C(Int. 82)

II-A-33

Radiation Chemistry Technician
T (Int. 75)

REFERENCEITEMDATE AND TIME
3/28/79

EVENT DESCRIPTION

133

0830

1.
The Station Manager, the Unit 2 Technical Support Superintendent, andthe Station Operations Superintendent discussed the status of the plant
and the Emergency Plan implementation. They concluded that the Emergency
Plan was properly implemented and actions were being taken consistent withprotecting public health and safety.

134

0830

NRC Region I Mobile Laboratory was recalled from Millstone Point by the
Region I Incident Response Center and directed to Three Mile Island.

135

0831

The power operated emergency mainsteam dump valve (MS-V-34)
induce natural circulation in steam generator A.

136

0832

Survey team "charlie" reported no detectable radiation (less than 1 mR/hr) 1. ECC Log
This measurement, along with the survey
at survey point W-11 (Goldsboro).
made on the island at 0748, confirmed that the initial prediction (made at0710) of 40 R/hr in Goldsboro was in error.

137

0833

The waste disposal storage area monitor (HP-R-218)
setpoint (50 mR/hr).

II-A-34

was opened toreached the alarmInterview ofUnit 2 TechnicalSupport Superintendent(Int. 27)

1. Interviews of
NRC Inspector
N (Int. 137)
2. NRC Incident
Response Log
1. Metropolitan
Edison Second
Interim Report

1. Chart recorder
HP-UR-1902
2. Procedure 21051.12

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

138

0835

A Pennsylvania State Police helicopter landed north of the parking lot.

1. Security Log
2. Interview ofRadiation Protection Foreman
D (Int. 18)
3. Discussions with
pilot

139

0842

Neither A nor B steam generators. were steaming to the atmosphere.

1. NRC Operational
Sequence ofEvents

1400843

The Unit 1 nuclear sampling room monitor (RM-G3) increased to 10 R/hr.
This detector is located about five feet from the Unit 2 sample lines.

1. ECC Log

141

0846

The Unit 1 hot machine shop monitor (RM-G4) read 1000 R/hr.
is located about six feet under the Unit 2 sample lines.

142

0847

A 5 man NRC inspection team from Region I left King of Prussia for TMI.
The team consisted of an investigator, 3 reactor health physicists, anda reactor operations inspector.

1. NRC Inspector
D's Notes

143

0850

A 50 to 100 ml reactor coolant sample was collected in a.250 ml bottle.
A radiation/chemistry technician took 1 ml for analysis. Five mls of
this sample was added to 95 ml of demineralized water for boron analysis.
The boron concentration was reported as 248 ppm. The sample as drawn
read 200 R/hr at six inches from the sample bottle. The results of gamma
analysis were:

1. Interviews of

II-A-35

This detector 1. ECC Log

Radiation/
Chemistry
Technicians U,
K, and I(Int. 78, 24,
and 42)ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

2.
3.

NRC Region I
Incident Messageform R19
Licensee sample
results

50 uCi /ml iodine-132
81 uCi /ml iodine-131
144 uCi /ml iodine-133
449 uCi /ml xenon-133
247 uCi /ml xenon-135
604 uCi /ml rubidium-88
29 uCi /ml krypton-85
NOTE:

This was the first reactor coolant sample drawn that had extremely
high radioactivity levels although it had already been establishedfrom the radiation monitoring system readings and various healthphysics surveys that high activity existed in the reactor coolant
as early as 0630 and was a consideration in the decision to announcea Site Emergency (0655).

144

0855

A two man backup team from NRC Region I was dispatched.

145

0900

The reactor coolant sample recirculation was shut off.
shop monitor (RM-G4) started to decrease.

II-A-36

1. Interviews of
NRC Inspector A(Int. 62)
The hot machine

1. CRO J's notes

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

146

0900

A control room bypass supply fan (AH-E-4B) was started 'nd the Unit 2
control room was placed on recirculation air. This was the first time
the control room had been placed on recirculation.

1. CRO J's notes

147

0900

An air-sample taken in the southern part of Goldsboro was lcss than
5 E-9 uCi/cc (minimum detectable activity for iodine-131).

1.

ECS Log

148

0900

The industrial waste water treatment system (IWTS)
Unit 2 Site Emergency.

I.

IWTS Log

149

0910

The primary Emergency Control Station was evacuated from the Unit 1
health physics control point to the alternate Emergency Control Station,
the Unit 2 control room.

II-A-37

was shutdown because

1. Licensee notes(collected by
Shift Supervisor C)
2. Interviews ofRadiation Protection Supervisor, Radiation
Protection Supervisor B and
Radiation/
Chemistry
Technician I(Int. 20, 21,
and 42)ITEM

DATE AND TIME
3/28/79

NOTE:

150

0915

REFERENCEEVENT DESCRIPTION
The decision to evacuate the Emergency Control Station was basedon increasing air activity and general area dose rates. The GM
counter rate meter (RM-14 with a HP-210 probe) at the Emergency
Control Station had increased from 100 counts per minute to 3000
counts per minute. The increasing background radiation resulted
in a loss of counting capability, especially the lithium drifted
germanium detectors. After the loss of the lithium drifted germanium detectors, the licensee had to count samples on a lessaccurate instrument.

The steam generators were reported isolated.

NOTE:

1. NRC Operational
Sequence ofEvents

The term "isolated" may only mean that the steaming generators
were not steaming through the atmospheric reliefs due to
It may not mean that theconditions in the steam generators.
steam generators had been isolated with the purpose of reducing
radioactive releases.

151

0917

The Emergency Control Station was established at the Unit 2 control room.

1. ECC Log

152

0922

An air sample taken in Goldsboro indicated 1 E-8 uCi/cc iodine-131.

1. NRC Region I
Incident Messageform R7

NOTE:

10 CFR 20, Appendix B, concentration for iodine-131 in unrestricted
is 1 E-10 uCi/cc (100 pCi/m ). This was the same location as the0900 sample.

II-A-38ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

153

0930

The Metropolitan Edison Vice President, Generation, was directed by the
President, Metropolitan Edison, to leave Philadelphia for TMI.

1. Interview ofVice President,
Generation(Int. 91)

154

0930

American Nuclear Insurers returned the licensee's phone call.

1. Interview of
Station Manager
and Unit Superintendent(Int. 41)

155

0945

The fuel handling bridge area monitor (HP-R-215) located in the fuelhandling building, reached the alarm setpoint (20 mR/hr).

1. Chart recorder
HP-UR-1902
2. Procedure 21051.12

156

0945

The control and service building corridor area monitor (HP-R-234)
the alarm setpoint (2 mR/hr).

1. Chart recorder
HP-UR-1902
2. Procedure 21051.12

157

0948

The control room intake particulate monitor (HP-R-220 (P))
alarm setpoint (300 cpm).

NOTE:

reachedreached theThis corresponded to 5.7 E-10 uCi/cc. The 10 CFR 20, Appendix B,
concentration for strontium-90 in restricted areas is 1 E-9 uCi/cc.

II-A-39

1. Chart recorder
HP-UR-1907
2. Procedure 21051.12

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

158

0950

An air sample taken in the northern3 part of Goldsboro indicated 1.2 E-8
uCi/cc of iodine-131 (1.2 E 4 pCi/m ).

1. NRC Region I
Incident Messageform R7
2. Offsite Dose Calculation Log

NOTE:

159

0950

The 10 CFR 20, Appendix B, concentration f r.iodine-131 in unrestricted areas is 1 E-1O uCi/ml (100 pCi/m )The control room air monitor (HP-R-220)
channels were increasing.

particulate,

iodine,and gas1. Chart recorder
HP-UR-1907

160

1000

A chemistry foreman sampled secondary water from the A and B steam
The A steam generator sample
generators at the Unit 2 sample station.
indicated slight contamination based on a qualitative gross beta-gammaThemeasurement with a GM count rate meter (RM-14 with a HP-210 probe).
B steam generator caused the GM count rate meter to go off scale (50,000
The shift foreman was informed of these results.
cpm).

161

1010

The control room gas monitor (HP-R-220(260 cpm).

(G))

reached the alarm setpoint

1.

Interviews ofChemistry Foreman B andA(Int. 21, 11)1.

Chart recorder
HP-UR-1907
Procedure 21051.12

2.

NOTE:

162

1010

The 10This corresponded to a concentration of 7.3 E-6 uCi/cc.
CFR 20, Appendix B concentration for xenon-133 is 1 E-5.uCi/cc.

The NRC Region I inspection team arrived on site and went to the Unit 1
control room upon direction of Station Manager.

1.
2.

II-A-40

Security Vistors
Log
Licensee notes

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

163

1012

The Emergency Control Station moved from the Unit 2 control room to theUnit 1 control room to reduce number of people in Unit 2 control room.

1. Licensee notes

164

1017

Control room personnel donned respiratory protection.
particulate filters were worn.

1. NRC Region I
Incident Messageform C5

165

1022

An open phone line between the Unit 1 control room and the NRC Region I
was established.

1. Bell of PA listof calls dated
4/13/79

166

1030

An air sample collected in the control room indicated 1 E-7 uCi/cc ofbeta-gamma activity. Since the isotopic content had not been determined
the licensee assumed the 10 CFR 20, Appendix B, concentration was
3 E-1O uCi/cc.

1. NRC Region I
Incident Messageform R3

167

1034

A two man backup team consisting of a reactor inspector and an investigator arrived on site from NRC Region I.

1. Visitors Log

168

1053

The plant was steaming from the atmospheric reliefs.

1. NRC Operational
Sequence ofEvents169

1055

A State Police helicopter transported an air sample (taken in Goldsboro)
to Holy Spirit Hospital for pickup by Pennsylvania Bureau of Radiological
Health personnel.

1. Discussion with
Pilot

170

1100 - 1130

NRC inspectors began making radiation measurements on site and in theenvironment to check the accuracy of the licensee's results.

1. Interview of
NRC Inspectors
I and J
(Int. 146)

II-A-41

Face masks withITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

171

1100

A liquid sample from the plant effluent monitor point (RM-L7) showed
iodine-131 less than 4 E-8 uCi/ml and iodine-133 less than 1 E-8 uCi/cc.

1. Licensee sample
result

NOTE:
172

1104

The 10 CFR 20, Appendix B, concentration for iodine-131 in water
for unrestricted areas is 3 E-7 uCi/ml.

The Unit 2 ventilation system was turned off.

NOTE:

1.

NRC Region I
Incident Messageform C6

The NRC expressed concern to the licensee that this action mightlead to ground level releases.

173

1104

The gross beta-gamma airborne particulate activity in the Unit 1 shift
supervisor's office was reported as 1 E-7 uCi/cc. The licensee assumed
the 10 CFR 20, Appendix B, concentration to be 3 E-1O uCi/cc.

1. NRC Region I
Incident Messageform C6

174

1110

The island was evacuated of all non-essential personnel.

1. Security Log

175

1110

The air sample from Goldsboro was delivered to the Pennsylvania Bureau ofRadiological Health.

1. Discussion with
State Officials

1761130 - 1215

Two NRC inspectors arrived in Unit 2 control room and made preliminary
assessment of radiological controls based on discussions with the Supervisor, Radiation Protection and Chemistry, and direct observations oflicensee activities.

1. Interview ofInspector H(Int. 40)

II-A-42ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

177

1130 - 1200

An NRC inspector measured 20 mR/hr at ground level in the plume from theplant. The plume was visible because it contained steam from the atmospheric dumps.

1. NRC Region I
Incident Messageform C12
2. Interview of NRC
Inspector J(Int. 146)

178

1200

A shift foreman, an engineer and a radiation/chemistry technician entered 1.
the auxiliary building to check on three decay heat valves in preparation
for starting of decay heat and to close sump pump breakers for an auxiliarybuilding water transfer. High range pocket dosimeters were not available
for entry. A survey of the seal injection filters showed 50 and 75 R/hr
at 1 foot using a GM radiation level survey meter (Teletector).
The
general area dose rate on the 281 ft. elevation was 2 R/hr. The decay heatremoval pump vault floor was dry and the general dose rate was 50 mR/hr.
After about 10 minutes the individuals could not locate the desired decay
heat valves and left for the 328 ft. elevation. The dose rate at the door
located at the west stair entrance to the 328 ft. elevation read 1 R/hr,
5 R/hr just inside the door, 100 R/hr near the air filter system, and1000 R/hr 8-10 ft. from the equipment hatch. The individuals left the area
without closing the sump pump breakers.
These individuals received about
800 mrem each during this entry.

179

1245

Route 441 was blocked to traffic by Pennsylvania State Police.

1. NRC Region I
Incident Messageform Rll

180

1230

The A steam generator atmoshperic relief valve was closed.

1. NRC Inspector
D's notes

II-A-43

Interview ofRadiation/
Chemistry
Technician U(Int. 78)ITEM

DATE AND TIME
3/28/79

NOTE:
181

1230

REFERENCEEVENT DESCRIPTION
Steaming from the A steam generator atmospheric relief valvecommenced after 0915 (See Item 150).

An NRC inspector reported to Region I that steam dumps were closed andradiation level readings appeared to be going down (2-4 mR/hr).
NOTE:

1. NRC Region I
Incident Messageform C14

Investigators concluded based upon data showing effective isolationof the "B" OTSG that these readings were caused by the noble gas
plume (which was invisible) coming from the Unit 2 vent stack.

182

1300

1.
The Commonwealth of Pennsylvania Bureau of Radiological Health Lab
completed the lithium drifted germanium analysis of the charcoal cartridgefrom Metropolitan Edison collected from location WSW-21 on 3/28/79 at 0940.
The Pennsylvania Bureau of Radiological Health relayed the result to Metropoless than
Result:litan Edison shortly after 1300 (exact time unknown).
The Metropolitan Edison portable sodium
1.5 E-11 uCi/cc of iodine-131.
iodide single channel analyzer (SAM-2) results had been 3 E-9 uCi/cc.
Seven additional charcoal cartridges (all collected on 3/28) were sent to
Pennsylvania Bureau of Radiological Health Lab; four were less than
1.5 E-11 uCi/cc (mlinimum detectable activity), and three were not analyzed
because of higher priority samples.

183

1310 - 1314

An NRC inspector left for Harrisburg to assist the Commonwealth withsurveys. He began performing surveys at 1400.

1. Interview with

There was a reactor building pressure spike which isolated the containment. The pressure went up to 28 psig, then dropped to 1.5 psig.

1.
2.

184

1350

NOTE:

This pressure spike resulted from a hydrogen burn in thereactor building.

II-A-44Penna. Bureauof Radiological
Health Laboratory Analysis
Sheets

NRC Inspector J(Int. 146)

Alarm printerUnit 2 Foreman Log

ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

185

1359

Unit 1 control room air activity was reported as 1 E-7 uCi/cc iodine and9.4 E-8 uCi/cc particulate.

1. NRC Region I
Incident Messageform C14

186

1400

The Metropolitan Edison Vice President, Generation; the Station.Manager(Emergency Director); the Unit 2 Technical Support Superintendent; andothers left the north gate to meet with the Lt. Governor.

1. Interview ofVice President,
Generation(Int. 91)187

1400-1430

An NRC inspector (dispatched 1310-1314) made surveys in Harrisburg andfound no radiation levels above background.

1. NRC Inspector J's
notes
2. Interview of
NRC Inspectors
I and J
(Int. 146)

188

1415

The Department of Energy Aerial Monitoring System helicopter arrived andbegan tracking the plume.

1. Discussion with
DOE

189

1430

The Metropolitan Edison Vice President, Generation; the Station Manager;
and the Unit 2 Technical Support Superintendent met with the Lt. Governor

1. Interview ofVice President,
Generation(Int. 91)

190

1600

The reactor building area radiation monitors located at the fuelhandling bridge-north (HP-R-209), equipment access hatch (HP-R-212) andin-core instrument panel area (HP-R-213) were tagged out of service.

1. Out-of-Service
Log
2. Out-of-servicetag

II-A-45ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

191

1630

The Station Manager and the Unit 2 Superintendent, Technical Support
returned to the plant after briefing the Lt. Governor.

1. Interview ofVice President,
Generation(Int. 91)

192

1700

Hourly X/Q values for the period 0700-1500, 3/28/79, were provided by
telephone to the Unit 2 control room by Pickard, Lowe, and Garrick
Similar data was provided on 3/29 and(meteorological consultant).
3/30/79.

1. Interview ofMeteorological
Consultants A
and B (Int. 60)

193

1900

A representative of Porter-Gertz, the licensee's environmental monitoring
and analysis consulting firm, arrived at the Observation Center andreported to the Vice-President, Generation.

1. Interview ofRadiation Protection Consultant
A (Int. 19, 28)

194

1900

1. Interview of
Two radiation/chemistry technicians entered the auxiliary building tochange the charcoal cartridge on the station vent stack monitor (HP-R-219)
Radiation/
Chemistry
and collect a water sample from the basement floor. The charcoal
Technicians W
cartridge was changed. Auxiliary building dose rates were:and H (Int. 49)
305 ft. elevation at entrance doors, 5-7 R/hr and 10 R/hr at the steps;
328 ft. elevation, 20 R/hr at the top of the steps; 50 R/hr at HP-R-219.
Dose rates up to 100 R/hr at other locations. A water sample from the
The sample was taken to the Obser281 ft. elevation floor was collected.
vation Center for analysis. The sample results are unknown, however areading of 40 mR/hr was measured on contact with the 100 ml sample
container.

195

1900

While enroute to the TMI site from Region I, NRC inspectors made thefollowing radiation surveys:

II-A-46

1. Interview of
NRC Inspector N(Int. 137)ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION

Location

Met Ed Descriptor

Exposure Rate (Open End Window GM)

1 mile east of Exit 19
on Pennsylvania Turnpike
Olmstead Plaza

NNW-412 mR/h

NNW-31

10-15 mR/h

196

1930

The NRC Region I Mobile Lab arrived at the TMI Observation Center,and informed the licensee that it was available to analyze samples.
The licensee was also informed of the results of radiation surveys
made by the NRC at 1900.

1. Interview of
NRC Inspector
N (Int. 137)

197

2000

The overall direction and control of the emergency organization wastransferred from the Emergency Director (Station Manager), in thecontrol room, to the Metropolitan Edison Vice President ofGeneration, in the Observation Center.

1. Interview of
Station Manager(Int. 71)

198

2030

A representative of the licensee's environmental consultant firm
arrived at the Unit 1 control room and reported by-phone to theEmergency Director.

1. Interview ofRadiation Protection Consultant
A (Int. 19, 28)
2. Visitors Log

199

2100

An auxiliary operator was directed by a shift foreman
to enter the auxiliary building and increase the nitrogenpressure on the core flood tanks from 400 to 600 psig.
He could not find a high range pocket dosimeter so theentry was made with a 0-200 mr dosimeter. He entered thearea alone and without a safety man. The dose rate just

1. Interview of
Auxiliary
Operator I(Int. 114)

II-A-47ITEM

DATE AND TIME
3/28/79

REFERENCEEVENT DESCRIPTION
inside the 305 ft. elevation entrance doors was 20 R/hr. He
measured dose rates of 70 and 100 R/hr as he walked acrossthe 328 ft. elevation. The dose rate near the nitrogen valve
was 10 R/hr. The task took about 10 minutes. On exiting
the building he found his pocket dosimeter had gone off-scale.
On entering the control room he caused the GM counter
Since(RM-14/HP-210 probe) to go off-scale (50,000 cpm).
he had to decontaminate himself he decided to re-enter theauxiliary building and start a pump from the radwaste panel.
He discussed this with a shift supervisor. He did nottell the shift supervisor about his previous entry. He
re-zeroed his 0-200 mR pocket dosimeter and entered theauxiliary building and received a dose that caused his
His thermoluminescentpocket dosimeter to go off-scale.
dosimeter was read and the results indicated a dose of 3.2 rem.

200

2100

Radiation surveys made by an NRC inspector, using an end-window
GM survey meter, along PA Rte. 230, in the vicinity of the Harrisburg International Airport indicated an exposure rate of 12 mR/hr.
The inspector made 4 passes through the plume.

1. Interview of
NRC Inspector
J (Int. 146)

201

2130

Unit 2 began transferring the neutralizer tank to the Unit 1 miscellaneous waste holdup tank. This was stopped on 3/29 at 0020.

1. Unit 2 CRO Log

NOTE:The Unit 2 neutralizer tank contained preaccident water andwas clean enough to be processed and released by Unit 1.
Movement of this water from Unit 1 would provide capacity
to start getting the water in the auxiliary building
off the floor and into the tanks.

II-A-48ITEM

DATE AND TIME
3/28/79

EVENT DESCRIPTIONREFERENCE

202

2150

Four health physics technicians from the Salem Nuclear Station
arrived with monitoring equipment to assist in performing offsite surveys.

1. Visitors Log

203

2325

Radiation levels of 365 mR/hr beta-gamma and 50 mR/hr gamma were
measured at GE-10 (1000 feet northwest of the Unit 2 plant vent)on the island.

1. Observation
Center (OC) Log

NOTE:
204

2400

This was the highest beta-gamma radiation levelmeasured at ground level outside of the plant.

Releases from industrial waste treatment system are estimated at
29,100 gallons released, with no activity.

II-A-49

1. NRC calculation basedon licensee
sample results
and IWTS LogITEMDATE AND TIME
3/29/79

EVENT DESCRIPTIONREFERENCE

205

0001

A long range planning and logistical support organization composed ofsenior Metropolitan Edison and GPU management and various site personnel
was established at the Observation Center.

1. Interview with
Radiation/
Chemistry
Technician E(Int. 50)

206

0020

The transfer of liquid radwaste from the Unit 2 neutralizer tank
(WDL-T-8B) to the Unit 1 miscellaneous waste holdup tank was stopped.

1. Unit 2
Foreman Log

207

0055

The auxiliary building and fuel handling building fans were stopped.

1. Unit 1 andUni~t 2 Foreman
Logs

208

0120

The miscellaneous waste holdup tank level was 7.5 feet. The level was
6.6 feet in this tank at the start of the accident. There were no known
operations involving this tank during the period between the two readings.

1. Unit 2
CRO Log

2090210

The auxiliary and fuel handling building ventilation systems werestarted.

1. Unit 2 Foremen
and CRO Logs

210

0305

Although there was no liquid release being made, the liquid radiation
monitor alarmed due to background radiation.

1. Region I
Incident
Messageform
R42

211

0435

The makeup tank (MU-T-I)
gas decay tanks.

was vented to the vent header to the waste

II-A-50

1. Unit 2 CRO andForeman Logs

ITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTIONNOTE:REFERENCE

Operations personnel knew that the vent header leaked and that
venting to the vent header resulted in releases to the auxiliarybuilding exhaust system.

212

0530

1.
The Vice-President, Midwest Division, Radiation Management Corporation
was phoned by the Philadelphia Office of Radiation Management Corporation and requested to report to the site as soon as possible to set up
the Radiation Management Corporation lithium drifted germanium detector
for sample counting and to ensure that the Radiation Management Corporation
whole body counter (on site prior to the event because of the Unit 1
refueling outage) was functional.

213

0600-0700

Additional health physics technicians began arriving from Nuclear
Support Services and Rad Services.

NOTE:

214

0830-1130

Nuclear Support Services provided 14 technicians on MarchRad
29 and an additional 4 technicians on March 30.
Services provided two instrument technicians and 7 teletectors and approximately 27 survey meters, and pocket
dosimeters.

A nuclear engineer proposed changing makeup filters to increase
A radiation/chemistry technician was requested by an
letdown flow.
The survey instrument readengineer to survey the makeup filters.
full scale (1000 R/h) when placed through a port hole into the
The room outside the filter cubicle had generalfilter cubicle.

II-A-51

Interview with
Radiation
Protection
Consultant C(Int. 93)

1. Interview ofProtection
Foreman A(Int. 11)
2. Discussion
with Manager
Instrument
Services,
Rad Services
and letterfrom Nuclear
Support Services
1. Interview with
Radiation/
Chemistry
Technician D(Int. 89)

ITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTIONREFERENCE

area dose rates of 2-5 R/hr and the reactor coolant letdown monitor(MU-R-720) read 90 R/hr. When discussing the dose rates on the makeup filter with the operations. staff, the radiation/chemistry technician
expressed his concern about high radiation exposures which could be received during the proposed makeup filter change out. The technician was
asked to resurvey the makeup cubicle. The technician reentered
the area and escorted two repairmen to the 281 ft. elevation to repair
a leak in the river water cooling system. The resurvey of the
makeup filters again indicated greater than 1000 R/hr inside and
2 R/hr on top of the shielded cubicle. The technician advised thefour mechanical maintenance personnel not to change the makeup
filters. He also informed his foreman since the engineer who requested the work did not appear responsive to his warnings. The
radiation/chemistry technician received 1.4 rem. The filters were
not changed.
215

after 0700

A radiation/chemistry technician surveyed the auxiliary building :with
the following results:

II-A-52

1. Interview ofRadiation/
Chemistry
Technician I(Int. 42)ITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTIONREFERENCE

281 foot elevationwater on the floor, 50-60 mR/hr (the water may not have been thesource of radiation).
35-40 R/hr - hall to bleed tanks.
300 R/hr - inside the door to bleed tanks.
500-600 R/hr - by the seal injection filter.100 R/hr - inside the door by makeup pump "B".
100 R/hr - inside the south reactor building spray pump vault.
200 R/hr - inside the door of the "A" bleed tank cubicle.
1.5-2.5 R/hr - near the pit from the 305 ft. elevation at the northend of the building.
40-50 R/hr - at the makeup tank room door.
305 foot elevation
300-400' mR/hr - general halls.
1000 R/hr - localized radiation level on the door to the valve room
containing the makeup filter bypass valve (MU-V-l05).
15 R/hr - halls near the makeup filter bypass valve (MU-V-l05)
room.
50-70 R/hr - at the valve room door (which was locked).
328 foot elevation
20-30 mR/hr - in the area of the station vent stack monitor(HP-R-219).
The survey results were recorded and put in the control room.

II-A-53ITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTIONREFERENCE

216

after 0700

A radiation/chemistry technician turned the reach rod to open amakeup system valve (MU-V-l05) to bypass makeup and purification
demineralizer filters to reduce pressure upstream of the filters.
The valve was partially open prior to this action.

1. Interview ofRadiation/
Chemistry(Int. 42)NOTE:

The letdown flow monitor in the control room showed noincrease after the valve was opened.

217

0715

The auxiliary building sump tank was pumped to a neutralizer tank
(WDL-T-8B) to provide room in the sump tank for auxiliary building
sump water.

1. Unit 2 CRO andForeman Logs

218

0716

The makeup tank was bypassed and reactor coolant letdown was directed
to reactor coolant bleed holdup tank B because of high pressurein the makeup tank.

1. Unit 2 CRO
and Foreman Logs

NOTE:

The control room operators had noted that when the makeup tankwas vented, the radiation levels in the auxiliary building
increased.

219

0815

Industrial waste treatment system and the industrial waste filter
system sumps were sampled for activity prior to starting up thesystem.

1. IWTS Log

220

0835

Liberty Fire Company trucks arrived to refill bottles for selfcontained breathing devices.

1. OC Log

II-A-54ITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTIONREFERENCE

221

0845

A Unit 2 neutralizer tank (WDL-T-8A) was being transferred to Unit
1 miscellaneous waste holdup tank, moving preaccident water to Unit
1 to make room for highly radioactive water in the auxiliary buildingsump.

1. Unit 1 CRO andUnit 2 Foreman
Logs

222

0847

Workers exiting the plant were advised by security personnel at thenorth gate to go to the 500 kV station to be surveyed for contamination.

1. OC Log

223

morning

A water sample from the plant effluent monitor (RM-L7 discharge fromthe mechanical draft cooling towers) was analyzed approximately
5/4/79 by Teledyne Isotopes for Porter-Gertz.
Sample results
showed 500 + 90 pCi/l tritium (5 E-7 uCi/ml).

1. Discussions with
Radiation Protection Consultant A(Int. 19, 28)

224

0903

A radiation protection foreman and a radiation/chemistry technician
brought the thermoluminescent dosimeter readers to the Observation
Center. The background exposure rate at the normal thermoluminescentdosimeter reader location had increased to 40 mR/hr due to gaseousemissions.

1.
2.

OC LogInterview ofRadiation/
Chemistry
Technician A(Int. 99)

225

1000

Records showed the industrial waste treatment system was sampled
at 1000 with the following results: iodine-131, 1.2 E-4 uCi/ml;
iodine-133, 1.1 E-4 uCi/ml.
The sump was not being discharged.

1.

Licensee sample

The NRC reported to the Emergency Control Station that there wasno iodine in the Unit 2 control room air samples collected by thelicensee.

1. OC Log

226

1032

II-A-55

resultITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTIONREFERENCE

227

1100

The industrial waste treatment system and industrial waste filtration system were resampled. A radiation protection foreman informed
the operator that there might be a xenon problem and did not know
if it was one or both systems.

1. IWTS Log

NOTE:This was later determined to be xenon background on the
counter. The sample point was not identified on the results
sheet, but samples were normally taken at sample point 107
for the industrial waste treatment system and 104 for theindustrial waste filtration system. These two points correspond to the discharge of the respective systems prior to theeffluent entering the discharge from the mechanical draft
cooling tower. One or two early samples for the industrial
waste treatment system were dipped directly from the sump.

228

afternoon

A radiation/chemistry technician accompanied an electrician to the328 ft. elevation of the auxiliary building to operate switch gear.
They approached from the east side of the building and founddose rates in the range of 2-3 R/hr.

1. Interview with
Radiation/
Chemistry
Technician R(Int. 75)

229

1215

The floor surface of the 281 ft. elevation of the auxiliary building
was covered with plastic sheeting to reduce the release rate ofradioactive gases from water on the floor into the auxiliarybuilding atomsphere.

1. NRC Region
I Incident
Response
Messageform C46

230

afternoon

The routine environmental samples (thermoluminescent dosimeters,
air, water, milk) were collected. The decision to collect the
samples at the time was made by a Metropolitan Edison radiation
safety and environmental engineer and a licensee's environmental
consultant.

1. Interview ofRadiation
Protection
Consultant A(Int. 19, 28)

II-A-56ITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTIONREFERENCE

231

1223

The Lt. Governor arrived at the north gate and went to the Observation
Center for a briefing from Metropolitan Edison.

1. OC Log

232

1240

The turbine building, control building, and control and service
building sump pumps were turned off due to a high level in theindustrial waste treatment system.

1. Unit 2
Foreman Log

233

1315

The industrial waste treatment system was started at chemistry department instructions. The sump level was greater than 100 percent.
The sump was overflowing to a yard drain which drains to a settling
pond called the east dike drainage area. The east dike drainage
area was containing most of the water but did have a small (undetermined) amount of leakage. The eventual destination of thisleakage was the river.

1. IWTS Log
2. Discussion
with IWTS
operator

234

1328-1340

The supervisor, technical training, phoned the Department of Environmental Resources, Pennsylvania Bureau of Radiological Health concerning the industrial waste treatment system release to the river.

1. OC Log

235

1330

The industrial waste treatment system discharge was sampled. The
results showed iodine-133 concentrations to be 3.6 E-4, but noiodine-133 was recorded.

I.

NOTE:

236

1400

IWTS Log

Because of the presence of iodine-133 with no iodine-131, thisresult was investigated and found to have been in error.
Neither isotope of iodine was present in measurable amounts.

A representative of Radiation Management Corporation (a healthphysics consulting firm employed by the licensee) arrived at theObservation Center and began to set up the whole body counter andthe mobile counting laboratory.

II-A-57

1. Interview ofRadiation
Protection
Consultant C(Int. 93)ITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTION

237

1410

A helicopter measured 3 R/hr, beta-gamma,
at 15 feet above the stack.

238

1410

An industrial waste treatment system operator was informed by theshift supervisor to terminate industrial waste treatment systemrelease due to xenon in the sample.
The release had been approximately 90 gpm for 40 minutes (4950 gallons).
NOTE:

239

1455

REFERENCE
and 400 mR/hr, gamma,

I. OC Log
2. Interview ofRadiation/
Chemistry
Technician C(Int. 82)
1. IWTS Log

The xenon was later determined to be counter background andnot in the sample.

The transfer of water from Unit 2 neutralizer tank WDL-T-8A to
Unit 1 miscellaneous waste holdup tank was completed.

I.

Unit 2Foreman Log

240

1500

The Supervisor, Radiation Protection and Chemistry directed aradiation protection foreman and a chemistry foreman to take areactor coolant sample within 2 hours.
They were told theexposure rate from a 100 ml sample would be approximately 800
to 1000 R/hr.
They took approximately 30 minutes to plan the job.

1.

Interview ofRadiation
Protection
Foreman D(Int. 18)

241

153,5

The Radiation Management Corporation mobile laboratory arrived
at the Observation Center.

1.

OC Log

242

1600

The auxiliary building sump tank was pumped to a neutralizer tank
(WDL-T-8A), to provide space for water from the auxiliary building
sump.

1.

Unit 2
CRO andForeman Logs

II-A-58ITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTIONREFERENCE

243

1610

The industrial-waste tank system processing was started (at approximately 200 gpm) after receiving authority from the Supervisor,
Radiation Protection and Chemistry. The sump level was greaterthan 100 percent.

1. IWTS Log

244

1615

A 100 ml reactor coolant sample was taken by a radiation protection
The sample read greater than 1000
foreman and a chemistry foreman.
R/hr on contact, 400 R/hr at 1 foot and 10-15 R/hr at 3 feet. They
did not wear extremity dosimeters on their hands, and there were no
air samples taken. The chemistry foreman received 4.1 rem dose to
his whole body. The radiation protection foreman had nonremovable
contamination of 150 mR/hr on his forearm, the chemistry foreman
had nonremovable contamination of 25 mR/hr on his hands.

1. Interviews of

2.
3.
245

246

1616

1652

One health physics supervisor, one health physics foreman and four
health physics technicians from Oyster Creek Nuclear Station arrived
with 3 high range GM survey meters (Teletectors) and self-contained
breathing devices.

1.

The Radiation Management Corporation whole body counter was put into
operation at theQ500 kV Station.

I.

II-A-59

2.

Radiation
Protection
Foreman D and
Chemistry
Foreman B and
C and Radiation
Chemistry
Technician X(Int. 18, 23,138, 151)
Form 5
Printout
Unit 2 CRO
Log
OC LogDiscussion
with Oyster
Creek HP
Supervisor
OC LogITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTIONREFERENCE

247

1700

An industrial waste treatment system sample was taken at the dischargeof the industrial waste treatment system prior to mixing with any othersamples. A note in the log stated that samples were to be taken every
2 hours. The sample point on the industrial waste treatment systemsump pump discharge line can be used either during industrial waste
treatment system recirculation or discharge.

1. IWTS Log

248

1740

Two health physics supervisors and two health physics techniciansfrom Salem Nuclear Plant arrived with an emergency van with 5-10
survey instruments, pocket dosimeters, respirators and protective
clothing.

1. OC Log

249

1755

The NRC Executive Management Team directed the licensee to stop dumping
all water. Region I notified headquarters that stopping the dumping
would cause water to backup into the turbine building. The Region I
Director personally discussed this order with the Station Manager.

1. NRC Region I
Incident
Messageform C49
2. Discussions
with NRC Region
I Director

250

1815

The industrial waste treatment system release was stopped after
approximately 25,000 gallons were released. This release started
at 1410.

1. IWTS Log

251

1850

Four health physics supervisors and ten health physics techniciansarrived from Peach Bottom Nuclear Station. They brought approximately
6 air samplers, 36 radiation survey meters, respirators and protective
clothing.

I.
2.

One ml of the reactor coolant sample collected at 1615 read 4 R/hr
It was later diluted for boron analysis.
on contact.

1. OC Log

252

1859

II-A-60

OC LogDiscussion with
Peach Bottom
HP Supervisor

ITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTIONREFERENCE

253

after 1900

A radiation/chemistry technician and a shift supervisor took a water
sample from the auxiliary building basement floor.

1. Interview ofRadiation/
Chemistry
Technician I(Int. 42)

NOTE:

Analytical results for this sample have not been located.

254

2020-2030

An operator began to depressurize the makeup tank (MU-T-I) to thewaste gas vent header through the sample system. The attempt was
stopped.

1. Unit 2
CRO Log

255

2045

The makeup tank (MU-T-I) was vented to the waste gas header tolower the pressure to approximately 55 psig. Operators had noted
increases in radiation monitor readings and concluded there was aleak in the waste gas system and that the leak rate increased aspressure in the vent header increased.

1. OC and Unit 2
CRO Log

256

2100

The auxiliary building sump tank was pumped to a neutralizer tank
(WDL-T-8A) and the auxiliary building sump to the sump tank.

1. Unit 2
Foreman Log

257

2133

Respiratory protection was required on the island from the northbridge to the warehouse.

1. OC Log

258

2204-2213

The contaminated chemistry foreman was whole body counted and sent
home with one hand reading 25 mR/hr. The radiation protection
foreman also had fixed contamination on his forearm, however,records of the levels were not maintained.

1. OC Log

II-A-61ITEM

DATE AND TIME
3/29/79

EVENT DESCRIPTIONREFERENCE

259

2300

Two engineers entered the auxiliary building with a high range
and a low'range survey meter to look for water leaks and to checkradwaste system status. The high range survey meter failed and they
entered several areas with radiation levels above the range of theother meter (2 R/hr), including the area near the reactor coolant
bleed holdup tank room. One engineer received 3.140 rem.

1. Interviews ofEngineers E
and J(Int. 33, 35)

260

2350

The makeup tank (MU-T-I)(WDG-T-IB).

261

2400

The Radiation Management Corporation lithium drifted germanium detector
system was ready to begin counting samples.

1. Interview ofRadiation
Protection
Consultant C(Int. 93)

262

2400

Releases from industrial waste treatment system for 3/29 (only) were
estimated to be 12.65 millicuries. A total of 29,950 gallons werereleased from the industrial waste treatment system on 3/29.

1. Licensee sample
results
2. IWTS Log

NOTE:

was vented to the waste gas vent header

Since 10 CFR 20 allows releases to be averaged over a
year, 10 CFR 20 limits would not have been exceeded.

II-A-62

1. Unit 2
CRO LogITEM

DATE AND TIME
3/30/79

EVENT DESCRIPTIONREFERENCE

263

0130

A nuclear engineer at the Emergency Control Station stated that henotified a representative of the NRC in the Unit 1 control room andthe Commonwealth of Pennsylvania of a planned venting of the makeup
tank to the vent header at approximately 0200-0300.

1. Interview ofEngineers C
and D(Int. 48, 127,
174)

NOTE:

The nuclear engineer stated that he notified the Commonwealthof Pennsylvania and NRC each time he knew of a vent to thevent header.

264

0150

The makeup tank was vented to the B waste gas decay tank (WDG-T-lB).
The A waste gas decay tank was at 50 psig; B waste gas decay tank,
75 psig.

1. OC Log

265

0152

A helicopter was sent to the south end of the island to monitor while
Unit 2 vented..

I.

266

0155

The transfer of liquid from the auxiliary building sump tank to theneutralizer tank (WDL-T-8A) was stopped.

1. Unit 2 CRO
Log

NOTE:

267

0213-0215

OC LogA log entry at 1600, 3/29/79 indicated that this transfer was
initiated at or before 1600. A log entry at 2100 indicatedthat during this period water was also being pumped into thesump tank from the auxiliary building sump.
The exact times
and sequences of this operation can not be determined.

The helicopter reported no radiation readings south of the plant.
The Emergency Control Station directed the helicopter to the eastside of the island.

II-A-63

1. OC Log

REFERENCEITEMDATE AND TIME
3/30/79

EVENT DESCRIPTION

268

0215

All sump pumps in turbine building and control building areas wereshut off.

269

0215

The venting of the makeup tank to a waste gas decay tank was stopped.
This venting had been in progress since 0150.

270

0300

An industrial waste treatment system discharge began at 190 gpm.
sump level was 100%.

The

1.

IWTS Log

271

0315

The control building area sump was pumped to the turbine building sumpwith temporary pump.

1.

Unit 2Foreman Log

272

0330

An operator began venting the makeup tank to the waste gas vent header.

1.

Unit 2 CRO
Log

273

0350

The venting of the makeup tank was stopped. The pressure in the A
waste gas decay tank was 50 psig. The pressure in the B waste
gas decay tank was 80 psig. Venting had been in progress since 0330.

I.

Unit 2
CRO Log

NOTE:
274

0430

The relief valves for the waste gas decay tank were set at 120
psig. The relief valve for the makeup tank was set at 80 psig.

An industrial waste filtration system discharge was started to theriver via the discharge line from the mechanical draft cooling tower
blowdown.
NOTE:

Unit 2 CROLog

The amount-of radioactivity discharged from the industrial
waste filtration system was insignificant for the period
3/28-3/30.

II-A-64

1. Unit 2 CRO
and IWTS Logs

ITEM

DATE AND TIME
3/30/79

EVENT DESCRIPTIONREFERENCE

275

0435

The liquid pressure relief valve (MU-R-I) on the makeup tank opened
draining the makeup tank to a reactor coolant bleed holdup tank.
The makeup tank level dropped straight down to zero. An unsuccessful attempt was made to pump B reactor coolant bleed holdup tank
(RCBHT) to the makeup tank but the transfer pumps would not pumpagainst the pressure in the makeup tank. The loss of level caused theoperations personnel to switch the suction of the makeup pumps fromthe makeup tank to the borated water storage tank. Operation in thismode caused operators concern about the loss of water inventory in theborated water storage tank that might be needed for future core coolingand for reactivity control and prompted them to vent the makeup tankto establish normal makeup path. Venting of makeup tank (see 0710)
The pressure in the makeup
caused an increase in gaseous discharges.
tank was 80-84 psig. Because of the discharges to the RCBHT's via
relief valve MU-R-l on the makeup tank, the pressure in the RCBHT's
was offscale at > 30 psig which is 1 1/2 times the design pressure ofthese tanks.
It's probable that the relief valves on the RCBHT's
were opening at least periodically.

1. Unit 2
CRO Log
2. MU-T-l level

NRC inspectors met with the Superintendent, Technical and Administrative
Services, to discuss concerns with the radiation protection program.
The following areas were discussed:

1.

Interview of
NRC Inspector H(Int. 40)

2.

NRC Region
I Incident
Messageform
R87

276

0455

1.

No positive control was established over entries
Access control.
into high radiation areas such as the auxiliary building.

2.

Exposure control. The radiation work permit procedure was not
being followed and equivalent measures were not being taken.

II-A-65

3.

4.

chart recorderInterview of
StationOperations
Supervisor(Int. 157)
Interviews of
Shift Supervisors
C and E
(Int. 173, 189)

ITEM

DATE AND TIME
3/30/79

REFERENCEEVENT DESCRIPTION
3.

Effluent monitoring. The installed samplers were not being changedout in a timely manner.

277

0630

The industrial waste treatment system flow dropped from 190 to 150
gpm (0300 hrs to 0630 hrs at 190 gpm equaled 39,900 gal).

1. IWTS Log

278

0650

A representative of Pickard, Lowe, and Garrick, the licensee's environmental consultant firm, went on site.

1. OC Log

279

0700

The A reactor coolant bleed holdup tank sample was 3.22 E-2 uCi/ml
and the C reactor coolant bleed holdup tank was 109 uCi/ml.

1.

NOTE:

Licensee sample
resultThis indicated that very little reactor coolant had been put
in A reactor coolant bleed holdup tank. Since the tanks wereseparated only by a loop seal it also indicated the reactor
coolant bleed holdup tank probably did not overflow.

280

0700-0800

One ml of the reactor coolant sample taken on 3/29 was removed fromthe sample container by a radiation protection foreman and placed in
No air samples were taken and no extremity
a 6 ml glass vial.
dosimeters were worn.

1. Interview ofRadiation
Protection
Foreman B(Int. 21)

281

0710

An operator opened the makeup tank vent valve (MU-V-13) and started
venting the makeup tank to the vent header. At least two shift
supervisors and one'operator were involved in this decision. They
decided this step was necessary in order to reduce pressure in the
makeup tank so that water could be transferred from the reactor
coolant bleed holdup tank to the makeup tank and so that the feedfrom the borated water storage tank could be stopped.

1. Unit 2
Foreman Log
2.
Interviews of
the Unit Supervisor, Station
Operations andEngineer C

II-A-66ITEM

DATE AND TIME
3/30/79

REFERENCEEVENT DESCRIPTION(Int. 157, 127,
174)
282

0744

The Unit Supervisor, Operations phoned the Emergency Control
Station to have survey teams and the helicopter dispatched.

I.

OC Log283

0744-0845

An engineer at the Emergency Control Station notified the NRC
representatives in the Unit 1 shift supervisor's office and theCommonwealth of Pennsylvania via telephone that makeup tank had
been vented and that increased releases were expected.
He later
notified the NRC and Commonwealth that the vent would last longer
than originally anticipated.

1.

Interview ofEngineers C
and D(Int. 48, 174)284

0756-0801

Radiation levels of 1000 and 1200 mR/hr (beta-gamma) were measured
at 0756 and 0801 respectively, from a helicopter at an elevation of
600 feet.
This was 130 feet above the Unit 2 auxiliary building.
The helicopter decreased altitude in order to better define thesource of radiation.
As the helicopter decreased altitude, theradiation level decreased to 600 mR/hr.
The helicopter returnedto the location of the original 1000 and 1200 mR/hr readings, but
no further readings in that range were found.

1.

Unit 1
CRO andForeman Logs
Interviews of
the Unit Supervisor, Station
Operations andRadiation/
Chemistry
Technician J(Int. 157, 51)

The Unit Supervisor, Station Operations, called Civil Defense to reportthat there was a release of radioactive materials and stated that the

1. Civil Defense
Logs

285

0834

II-A-672.

ITEM

DATE AND TIME
3/30/79

EVENT DESCRIPTIONREFERENCE

site was prepared to evacuate, that the site had its own buses andAt the same time,
recommended that the State prepare to evacuate.
another call was received from the site which provided details concerning the release. The caller explained that an evacuation of thesite was not planned.

2.

3.

4.

NOTE:

286

0845

Discussions

with Civil
DefensePersonnel
Interviews ofUnit Supervisor,
StationOperations(Int. 157)
NRC Regioh
I Incident
Messageforms C56
and C57

The Unit Supervisor, Station Operations, stated that there were
two calls, one he placed and one confirmatory call back from
Civil Defense. He said that these calls contained routine emergency wording which would include the request of the status of
the Civil Defense evacuation plans, the availability of buses
to provide transportation, and the status of the plant's
evacuation planning.

An NRC inspector in control room reported to NRC headquarters that theseal return to the makeup tank was causing excessive gas pressure in
the makeup tank which was directed to the waste gas decay tanks whichwere full.
The waste gas tanks were being released to the stack.
This was resulting in an increased stack release rate. He said the
Civil Defense and Commonwealth of Pennsylvania were being notified
by the licensee.

II-A-68

1. NRC Region
I Incident
Messageform
C56

ITEM

DATE AND TIME
3/30/79

EVENT DESCRIPTIONREFERENCE

287

0900-1300

A radiation protection foreman accompanied a sample of reactor coolant which had been collected at 1615 on 3/29/79 to the airport. The
sample was flown to an outside laboratory for analysis.

I.
2.

ý3.

Unit 2
CRO Log
Interview ofRadiation
Protection
Foreman B(Int. 21)
Licenseesample results

The sample results were:iodine-131
iodine-133
cesium-134
cesium-136
cesium-137
strontium-89/90
ruthenium-106
barium-140
tellurium-132
288

0945

1.3 E+4 uCi/cc
6.5 E+3 uCi/cc
6.3 E+l uCi/cc
1.8 E+2 uCi/cc
2.8 E+2 uCi/cc
5.3 uCi/cc
2 E+2 uCi/cc
2 E+2 uCi/cc
2 E+2 uCi/cc

Air iodine results for two locations (2.6 mi north and 9 mi southeast)
sampled weekly (off 3/29) were telephoned to the Metropolitan 3 Edison
Corporate office. Results were 0.05 and less than 0.02 pCi/m
respectively.

II-A-69

1. Telecons withrepresentativesof Porter-Gertz
Consultants andTeledyne Isotopes

ITEM

DATE AND TIME
3/30/79

EVENT DESCRIPTIONREFERENCE

289

1000

The A and B waste gas decay tank pressures were 79 psig.

I.

Unit 2Foreman Log

290

1020

Water iodine results for five locations (2 upstream and 3 downstream)
were telephoned to the Metropolitan Edison Corporate office. The results
were less than the minimum detectable activity (approximately 10 pCi/l).

1.

Telecons withrepresentativesof Porter-Gertz
Consultants andTeledyne Isotopes

291

1030

The Unit 2 B steam generator was sampled with the following results
indicating a primary to secondary leak:1.

Licenseesample results

iodine-131
iodine-133
iodine-135
cesium-136

7.9 uCi/ml
2.1 uCi/ml
8.5 E-2 uCi/ml
6.8 E-3 uCi/ml

292

1045

The area radiation monitors (HP-R-231, HP-R-233, HP-R-3238) located
1. Out-of-service
at the auxiliary building sump tank filter room on the 281 ft. elevation,
Log
in the access corridor on the 305 ft. elevation and at the building exhaust
unit were tagged out of service because they did not source check.

293

1220

The transfer of Unit 2 miscellaneous waste holdup tank to Unit 1 wasstarted.

1. Unit 2
CRO Log

294

1230

The first industrial waste filtration system sample was taken at
National Pollution Discharge Elimination System point 104. This
samples point was prior to the discharge into the blowdown from themechanical draft cooling tower. No activity above background wasobserved.

1. IWTS Log

II-A-70ITEM

DATE AND TIME
3/30/79

EVENT DESCRIPTIONREFERENCE

295

1230

The Metropolitan Edison Corporate Office received results of thesamples collected on the afternoon of 3/29. The results were:

1. Interview ofRadiation
Protection
Consultant A(Int. 19, 28)-

Preliminary results for offsite thermoluminescent dosimeters werereported as less than 25 mrem/quarter.
The maximum result reported
for thermoluminescent dosimeters located on the island (location
16 S 1) was 923 mrem/quarter.
(The final result for this location
was 1044 mrem/quarter).
NOTE:

296

1440

Results from dosimeters located on Kohr and Shelley Islands
were not available at this time. The highest value subsequently received for these locations was 908 mrem/quarter
for Kohr Island (location 16AI).

-

All iodine a r samples were less than 0.03 pCi/m3 except one with
a 0.47 pCi/m at IS2 (north-northeast) on the island. The MPC
for unrestricted areas is 100 pCi/m .

-surface water samples from one upstream and four downstream location
were less than minimum detectable activity for iodine (0.5 pCi/l).
The MPC for unrestricted areas is 300 pCi/l.

A waste decay gas valve was opened to transfer B waste gas decay
tank into the reactor building.

II-A-71

I.

Unit 2
CRO LogITEM

DATE AND TIME
3/30/79

EVENT DESCRIPTIONREFERENCE

297

1500

The pressures in A and B waste gas decay tank were 80 psig.

1. Unit 2
CRO Log

298

1545

Representatives of NRC,

299

1900

The station vent stack monitor (HP-R-219) charcoal cartridge waschanged. The results were iodine-131, 1.4 E-6 uCi/cc; and iodine-133,
5.2 E-7 uCi/cc.

1. Licensee
sampleresults

300

2045

The Unit 2 B steam generator was sampled using the normal Unit 2
sample point. The results were indicative of a primary to secondaryleak:

1. Licensee
sampleresults

iodine-131
iodine-133
xenon-135
cesium- 136
cesium-137
301
302

2128
2200

NRR, arrived in the Unit 1 control room.

1. OC Log

4.2 uCi/ml
1.4 uCi/ml
2.7 E-1 uCi/ml
2.2 E-2 uCi/ml
3.8 E-2 uCi/ml

The pressurizer was vented through vent valve (RC-V-137)
reactor coolant drain tank.

to theAfter arrival on site at 2000, the Science Appl ication, Incorporated
mobile lab performed operability checks and cal ibrations and were
They did not receive
ready to accept the first samples for counting.
samples until 3/31/79.

II-A-72

1. Unit 2
Foreman Log
1.

Interview ofRadiation
Protection
Consultant B(Int. 69)ITEM

DATE AND TIME
3/30/79

EVENT DESCRIPTIONREFERENCE

303

2217

The reactor coolant system was vented through vent valve RC-V-137 to
the reactor coolant drain tank.

I.

304

2229

The miscellaneous waste holdup tank contents were being transferred
to Unit 1.

1. Unit 2
CRO Log

305

2310

The reactor coolant system was vented via vent valve RC-V-137 to thereactor coolant drain tank.

I.

306

2400

The transfer to Unit 1 from the miscellaneous waste holdup tank was
stopped.

1. Unit 2
CRO Log

307

2400

The releases from industrial waste treatment system for 3/30 (only)
are estimated at 59.91 millicuries.

1. NRC Calculation

The total releases from the industrial waste treatment system for theperiod 3/28 through 3/30 is estimated at 72.56 millicuries.
A total of 206,400 gallons were released on 3/30.
A total of 265,450 gallons were released from 3/28 at 0400 to 3/30 at 2400.
Based on an annual average,

no limits would have been exceeded.

II-A-73

Unit 2
CRO Log

Unit 2
CRO LogAPPENDIX II-B
DOSIMETRY EVALUATIONSPOTENTIAL OVEREXPOSURESUNITED STATESIREGULATORY COMMISSIONtNUCLEARREGION I
631 PARK AVENUE
KING OF PRUSSIA, PENNSYLVANIA 19406

MEMORANDUM FOR:

A. Gibson,

FROM:

M. J. Slobodien,

SUBJECT:

DOSIMETRY EVALUATIONS POTENTIAL OVEREXPOSURES

NRC:IE Investigation Team
Radiation Specialist

I have reviewed six potential overexposures identified by the InvestigationTeam.

Two individuals appear to have been exposed in excess of regulatory

requirements. Tables II-B-1 and II-B-2 summarize these overexposures giving
lower and upper bounds. The detailed analyses of these overexposures are given
in Tables Ii-B-3 and II-B-4 to this memorandum.

Four of the six individuals

identified during the investigation as being potentially overexposed were notexposed in excess of regulatory limits.

ichael . Slobodien
Radiation Specialist
Enclosures:
As Stated

TABLE II-B-1
EXPOSURES TO SUBJECT A IN IST QTR 1979
Area of Body
Whole Body

Lower Bound

Upper Bound

0.88 rem (caic.)

0.89 rem (TLD)

Reg Limit
3 rem

Extremities
Fingers

1739

18.75 rem

Hand (palm)*

173918.75 rem

Forearm (skin)*

44 rem54 rem18.75 rem

The estimate of dose of fingers, hand and palm are the same as it was not possible
to make accurate estimates of the various finger/hand configurations.
The forearmis taken as the area around the wrist.

II-B-2TABLE II-B-2
EXPOSURES TO SUBJECT B IN IST QTR 1979
Area of Body
Whole Body

Lower Bound

Upper Bound
4.1 rem (TLD)

4.0 rem (calc.)

Reg Limit

3 rem

Extremity
Fingers
Skin (Top of Head)*

144 rem

147

6

13

rem*

18.75 rem
rem*

7.75 rem

These values will increase since only 1-131 contamination was considered.
+ The licensee reported the 4.1 rem W/B dose only on May 1, 1979 pursuant to 10 CFR 20.405(b),extremity or skin doses were reported.
EXPOSURES TO SUBJECT B IN 2ND QTR 1979
Skin (Top of Head)*19.75 rem

7.75

To this value must be added the dose due to contamination (Cs-134, Cs-137, 1-131) which waspresent from the time of contamination until the first whole body count.
Details for the data on this table are on page II-B-29 and II-B-30.

II-B-3noTABLE II-B-3
Subject - A
Circumstances of Exposure
A is a health physics foreman.
On March 29, 1979 he assisted a chemistry
foreman (subject B) in taking and measuring a sample of primary coolant fromthe letdown system sampler located in the Unit 1/2 nuclear sampling room. As
a result, he handled high level radioactive sources with gloved hands, worked
in high radiation areas, and sustained significant forearm skin contamination.
The following outline summaries A's actions on March 29,

1979 to the present:

March 29, 1979
TIME
15:00

ACTIONA and chemistry foreman (B) proceeded to Unit 1/2
sampling room to prepare to take a sample ofletdown.

15:00-15:30

A and B reviewed the piping runs.

15:30-15:45

Wearing full protective clothing including a MSA
full face respirator equipped with a particulate/iodine canisters A and B entered the Nuclear
Sampling room. A radiation measurement with ateletector indicated ambient radiation levels of
6-8 R/hour.
B proceeded to the sample cooler, aligned thevalues and placed the letdown system line for Unit
2 on Recirc.
A and B exited.

16:30

45 minutes were allowed for sample recirculation.
A entered the sample room and measured radiationlevels of 90 R/hour at the "drag" valve. He
measured 17-20 R/hour ambient radiation levels.
A moved into the sample room, placed a 400 ml
beaker in the sample sink, opened the letdown
valve one full turn and took 30-100 ml.
Thesample was very gassy. A closed the valve andleft the sample in the sink. A exited to discuss
the situation with B.

II-B-4

B confirmed the valve alignment. B may haveentered the sample room to do this, or may have
It isconfirmed valve locations from a diagram.
not clear from available data.
B re-entered the sample room, opened the letdownline valve at the sink and drew a 300 milliliter
sample while holding the bottle with his left hand.
As the bottle filled A stood behind the valve control
panel east of the sample sink. B closed the letdown
valve. A and B exited.
B entered and poured 100 ml into a small polyethylene
graduated cylinder then withdrew 5 ml then took a 40
ml sample and exited with it. A hand carried thegraduated cylinder to the southeast corner of thesample room and placed it on the floor. A used ateletector to measure greater than 1000 R/hour oncontact with the beaker, 400 R/hour at one footand 40 R/hour at three feet.
A returned the graduated cylinder to the sample sink
As he left the area,and exited the sample room.
A ripped his wet suit on the valve.
In the primary chemistry lab A assisted B bypicking up a 40 ml sample of primary coolant anddumping it down the sample sink and by returning
the sample of 260-300 ml of primary coolant to theshield in the storage cabinet.
16:30-17:00

Upon removing protective clothing and frisking
with an Eberline E-520 geiger counter with an HP177 side window probe, A identified a 5-10 cm
diameter spot on the left forearm which indicated
115 mR/hour on contact.

3/30/79

At 24:00 the contaminated spot on A's forearm
indicated 40 mR/hr. This was after 6 hours ofintensive attempts at decontamination.

4/2/79At 16:00 on 3/30/79 the contaminated spot indicated
2 mR/hour.

4/6/79At 16:00 on 4/6/79 the contaminated spot indicated
At 16:07 A received an initial whole bodybackground.
1-131 as the principle
count which indicated 329 nCi.
contaminant with most material on the skin.

4/11/79

A was whole body counted indicating 70 nCi primarily as skin contamination 18 nCi in the thyroid.
II-B-5

DOSE EVALUATION
1ST QTR 1979
SUBJECT A
ActionDose Rates* (R/sec)
Whole Body
Duration3

12 sec2. 2x10

Entry to sample
sink

3 sec

4. 7x0-3

Place beaker insink, open valve
take 30-100 ml
sample5 sec

4. 7x10-

4.

Closes sample valve

2 sec

4. 7xl 0- 3
3
4.7xi0-

5.

Exits area

3 sec

6.

Reentry and placement of 1 literin sample sink1.

Initial Survey
Nuclear Sample Room

2.
3.

3

Hand/Forearm
2.2x10-3

Dose (rem)Whole Body

Fingers2.6xi0 2

2.2xi02

3

1.4xl 0-2.5x10-2

2. 3x102

9. 4x10-3

1.4xl04.7x10-3

2. 3xi03

2.8x104.7xl0-3

4.7xl0-2.5x10-2

2.5x10-2

2.5xl0-4.7x10-3

4.7xl0-4.7x10-3

3

4.7xl0-2
32

Hand/Forearm
2.6x102

1. 4x10-5x10

2. 6x102

1. 25x102

Fingers

1.4x101

2

2

1. 25x105x10

1.4x10-22

1. 4x102

4.7x10- 3

5 sec

4. 7x102-2

2.3x102. 3x102.8x10-1

2. 8x10

-2

-3

7.

Stands behind control panel whilebottle fills

60 sec4.7x10-4.7x]0-II-B-6

1

1

DOSE EVALUATION
1ST QTR 1979
SUBJECT A
Action
8.

9.

10.

11.
12.

Dose Rates* (R/sec)
Duration
Whole Body

Hand/ForearmFingers

Dose (rem)Whole Body

2.5x10 1

7x10 1

1.1x10 2

Hand/ForearmFingers

7.2x10 2

1.5

4.2

1.1x10 2

8.8x10 2

8.8x10 2

8.8x10 2

7x10-

7.2x10 2

1.5

4.2

Picks up the lOOml
graduated cylinder
filled w/primarycoolant and carries
it to southeast
corner of lab
6 sec

1.2x10 2

Measures exposure rates
standing at 6'
from the cylinder

8 sec

1.1x1O

Picks up 100 ml
cylinder returns
it to sample
sink

6 sec

1.2x102.5x10 1

Exits Nuclear
Sampling Room

3 sec

1.1x10 2

1.1x1O

1.1x1O

3.3x10 2

3.3x10 2

3.3x10 2

Picks up 40 ml
coolant in 100 ml
beaker (drawn by
subject B) anddumps into sample
sink

3 sec

4.4x10 2

2.8x10-2.8x10-1.3x10 1

8.4x10 1

8.4x10 1

-2

-2

-2

II-B-7

2

DOSE EVALUATION
1ST QTR 1979
SUBJECT A
Action13.

14.

Dose Rates* (R/sec)
Whole Body
Duration

Picks up 260 ml
primary coolant
in 11 bottle carries to cabinet. Distance to
WB=30', fingers
on contact, hand,
3"

6 sec

1. 7x10

Forearm skincontamination
LEFT FOREARM
3/29/79 @ 16:00
to 3/31/79 @ 24:00

1384 uCi-hrs

N/A

2

Dose (rem)Hand/ForearmFingers

Whole Body

Hand/ForearmFingers

0.6

1.2

1. Ox10 1

3.6

7.2

N/A

36-46 rem

N/A

0.884 rem

8.1 rem di17.1 rem
rect;
36-46 remcontamination
44-54 total,
contamination &directTOTAL #The dose rates in this calculation for contact doses are based on the method of calculating a photon flux with
E=.44MeV based on the assumed fission product inventory. This method appears in pages II-B-16 - II-B-22 of thisreport. Dose rates other than those on contact are based on teletector measurements. The skin dose calculations
are found on pages II-B-23- II-B-25 of this report.

II-B-8

BASIS FOR DOSE RATES
Action No.

Basis

I.

Dose rates measured by subject 6-8 R/hrselected 8 R/hr for W/B.

2.

Dose rates measured by subject 15-17 R/ hrafter sample circulated.

3.

Dose rates measured by subject using te letector.

4.

Dose rates measured by subject using te letector,
90 R/hr on contact with value.

5.

Dose rates measured by subject using te letector.

6.

Dose rates measured by subject using te letector.

7.

Dose rates measured by subject using te letector.

8.

W/B dose rates measured by subject using
teletector, extremity dose rates calculated forsource.

9.

Dose rates measured by subject using teletector.

10.

W/B: dose rates measured by subject using
teletector.
Extremity from calculation asshown (after Kase et. al.) pages 17-21 of thisreport.

11.

Subject measured dose rates.

12.

See No.

10.

13.

See No.

10.

14.

MIRD calculations, done in accordance with
Phamplits 10, 11 and ICRP-23 contamination values
checked w/HP-177 probe and E-520 on disc source
whole body count data, see pages 23-25 of thisreport.

II-B-9

DOSE EVALUATION SUBJECT A
2ND CALCULATIONI.

Ist QTR 1979
A.

Exposure prior to 3/29/79
Whole Body
ExtremityB.

.09 rem.09 rem

Data available for 3/29/79
1.

TLD read 0.89 for W/B.

2.

No extremity dosimetry worn.

3.

A measurement of 100 ml of primary coolant with a teletector
indicated:4.

a.

> 1000 R/hr on contact.

b.

400 R/hr @ 12" (the distance was determined using a floortile).

c.

40 R/hr @ 3'
tiles).

(the distance was determined using 3 floor

Primary coolant analysis (see data sheet),through II-B-22.

pages II-B-20

7.53xi08 photons sec-I ml-I
E = 0.44 MeVC.

Calculations
1.

Dose rate from 100 ml to deep tissue.
By assuming a point source and using the measurements made by the
subject.

k 1 inch=(400 R ) (122)2
r-157600 R/hr

k 30 inch =(400 R ) (122)2
hr-02

64 R/hr

k 30 inch =(40 R ) (36)2
hr (302)
II-B-IO-

57.6 R/hr

x 30 inch =2.

64 + 57.6
2=

Time Distance Study

II-B-1160.8

DOSE EVALUATIONIST QTR 1979
SUBJECT A
2ND CALCULATION.
ActionDose Rates* (R/sec)
Duration
Whole Body

Hand/ForearmFingers

Dose (rem)Whole Body

2.2x10 3

2.2x10 3

4.7x10 3

Hand/ForearmFingers

2.6x10 2

2.6x10 2

2.6x10 2

1.4xlO 2

1.4x10 2

1.4x10 2

12 sec2.2x10 3

Entry to sample
sink

3 sec

4.7x0-3

Place beaker insink, open valve
take 30-100 ml
sample

5 sec

4.7x102.5x102.5x102.3x101.25x101.25x10 1

4.

Closes sample valve

2 sec

4.7xi0- 3

2.5xi0-2

2.5xi0- 2

9.4x10- 3

5x10- 2

5x10- 2

5.

Exits area

3 sec

4.7x10- 3

4.7x10- 3

4.7x10- 3

1.4xlO- 2

1.4x10- 2

1.4x10-2

6.

Reentry and placement of 1 literin sample sink

5 sec

4.7x10 3

4.7x10 3

4.7x10 3

2.3x10 2

2.3x10 2

2.3x10 2

Stands behind control panel whilebottle fills

60 sec

4.7x0-3

4.7x10 3

4.7x0-3

2.8xi0 1

2.8x10-1

2.8x10-1.

Initial Survey
Nuclear Sample Room

2.
3.

7.

-3
4.7x10-3

2

II-B-122

2

DOSE EVALUATIONIST QTR 1979
SUBJECT A
2ND CALCULATION
Action
8.

9.

10.

11.
12.

Dose Rates* (R/sec)
Whole Body
Duration

Picks up the lOOmi
graduated cylinder
filled w/primarycoolant and carries
it to southeast
6 sec
corner of lab

1. 2x10

Measures exposure rates
standing at 6'
from the cylinder

8 sec

1.1x1O

Picks up 100 ml
cylinder returns
it to sample
sink

6 sec

1.2x10-2

Exits Nuclear
Sampling Room

3 sec

1.1x10-

Picks up 40 ml
coolant in 100 ml
beaker (drawn by
subject-B) anddumps into sample
sink

3 sec

4. 4x10

2

2

2

2

Hand/ForearmFingers

Dose (rem)Whole Body

1.lx10-1

1.33 @ 1"

7. 2x10-2

1.1xlO

2. 5x102. 7x102. 8x101

3

1

II-B-131.1x102

8.8x10-2

7. 2x10-

1.33 @ 1"
1.1x1O

O.3x10-

2

1

2

2

8. 3x10-3@ 1"

1. 3x10-1

Hand/Forearm

6. 6x10-8. 8x1012

8.0

8. 8x10-

2

8.0

1.5
3. 3x10Fingers2

8.4x10-1

3. 3x10-

1.6

2

DOSE EVALUATIONIST QTR 1979
SUBJECT A
2ND CALCULATION
Action13.

14.

Dose Rates* (R/sec)
Whole Body
Duration

Dose (rem)Hand/ForearmFingers

Whole Body

Hand/ForearmFingers

3.5 @ 1"

1.OXlO'

3.6

21

N/A

36-46 rem

N/A

Picks up 260 ml
primary coolant
in 11 bottle carries to cabinet. Distance to
WB=30', fingers
on contact, hand,
3"

6 sec

1.7x10"0.6

Forearm skincontamination
LEFT FOREARM
3/29/79 @ 16:00
24:00
to 3/31/79 @
11384 uCi-hrs

N/A

36-46 rem

2

.860

*The dose rates for extremities in this calculation are based on teletector
For fingers, the radiation sources were considered as line sources.
data.
Dose rates f~r a line source at 1 inch are estimated from teletector data to
at 1"
1.33 R sec
12" = 4800 R hrbe 400 R hr
1"
lOOml

II-B-14

7.2 rem direct36-46 rem contamination
44-54 total, contamination& direct

39 remBASIS FOR x VALUES SELECTED
IN 2ND CALCULATION
Action No.

Basis

1.

Subject indicated geuieral area exposure rates
6-- 8 R/hr (2.2xi0
R/sec).

2.

Subject measured exposure rates @ sample sink of 15-17
R/hr; 16 R/hr = 4.7xi0
R/sec.

3.

As in #2 above for W/B.
Extremity dose rates based on
subjects measurement of valve dose rate of 90 R/hr
2.5x10- 2 R/sec.

4.

As in #3 above.

5.

As in #2 above.

6.

As in #3 above.

7.

As in #2 above.-

8.

30" W/B based on correction from 40 R/hr @ 3' assuming
a line source. Extremity k based on 400 R/hr @ 1 footand assuming hand was at one foot.

9.

As in #8 above.

10.

30" k based on 400 R/hr @ 12".
12" and assuming deep tissue.

I].

Subject measurement of 40 R/hr @ 3' corrected to 6'.

12.

As in #10 above.

13.

As in #10 above.

14.

As in #2 above.

15.

MIRD 10, 11, ORNL 5000 assuming uniform deposition
throughout skin.

II-B-15

1" k based on 400 R/hr @ASSUMPTIONS AND CALCULATIONSI.

The exposure times are estimated from data provided by subjects A and B in
investigation interviews.

2.

The dose rate from a right circular cylinder containing the primary coolant
sample is based on the material inventory determined by sample analysis ofa 3/29/79 sample at the Westinghouse - Bettis Laboratory. The calculation
which follows is for a one liter bottle. The dose rates from the 100 ml
sample would be N 10% of that from a one liter sample. Thus, the contact
dose rate is 0.7 rad/sec (2480 rad/hr). This may be low by a factor of 3
depending on the accuracy of extrapolation to a=o in accompanying figure on
page II-B-19 of this report.
Note this is not in agreement with point
source assumption used in the "2nd calculation".
The Ist and 2nd
calculation would indicate lower and upper extremity dose estimates
respectively. The estimate of 2480-3500 R/sec is in agreement with theassumption of a the graduated cylinder being a line source at distances <12". In that case the dose rate at one inch would be approximately:{400}

R {12"}
hr F=

4800 R
hr

This is reasonably close to the value determined in the following calculation:

II-B-16

CALCULATION OF DOSE RATE FROM RIGHT CIRCULAR CYLINDER

Sources:1.

Concepts of Dosimetry K. Kase and W. Nelson, Pergamon Press, New
York, 1978 pp 118-140.
2.

Engineering Compendium of Radiation Shielding, Vol. I,
Jaeger, ed., Springer-Verlag, New York, 1968 Chapter 6 pp374-387.

II-B-17

D = (l.6x]0-8 rad gram) (0) (S) (V) (E) pen

p

MeV
where

photon flux photon cm- 2 sec-I normallized to unit activity

0S=

volume activity concentrationV=

volumeE=average photon energy

V

=

len/P

0

=

mass attenuation coefficient for tissue

SvRo 2 (F(O 1 , b 2 + F(E 2 , b 2 )4(a+Z)

where
3

S

=

photon/cm

R0=radius of sample containera,z=distance to absorbing mediumF=

the volume integral taken from reference 2

for the case in point 0 is taken from the figure which follows.

0

has
c2

units

disentegrations sec

0

v 0

photons cm 2 sec-l-Iis calculated from the curve by taking the ratio of R0 for the one liter

SvR 2 bottle used to that of the cylinder used in the example.
1.7 /5. 5=0.32.

R 1/R 0Note that the dose calculated here is a photon dose only. The beta dose in thissituation is considered to be negligible due to self absorption in solution, absorption in container walls, and protective clothing. Bremsstrahlung will be
neglected.

II-B-18

EVALUATION OF PHOTON FLUX (0) AT THE
SURFACE OF A RIGHT CIRCULAR CYLINDER
I0-I

-

1I

I

I

7-

1-

a = 0 (3x10 2 )--9aDetector

No External Shield
Ro= 5.5 Inches _h=14 Inches-

lO\
C,,

4Data Normalized to Calcula-tion at a=20 Inches_:_

a/Ro<lO

a/Ro0lO -Data Taken with GM

Counter ("C Activity in Tank).
-3 Calculation (Excluding Buildup)
Line Drawn By Eye.
0

I

I

I

I

10

20

30

40

I
50

60

70

a (inches)

For R/R = 0.32,
0

0 = 2.6 X10-2
2

II-B-19= 1.3

XIO- 2 photons
secondPHOTONS FROM PRIMARY COOLANT
N
photons sec

E (MeV)f

Sourcedps/ml

1.
2.
3.
4.
5.
6.
7.
8.
9.
10.
11.

.080
.284
.364
637
.723
.530
.875
1.24
1.29
.856
.66

.026
.054
.82
068
.016
.87
.036
.016
.019
.01
.01

12.

1-131
1-131
1-131
1-131
1-131
1-133
1-133
.A-133
1-133
1-133
1-133

.57

.23

Cs-1344.8x0 825x107
4.8x10 8
4.8xi04.8xi0 8
4.8x108
2.4xi0 8
2.4x108
84X08
2
2.4x10 8
2.4xI08
2.4x10 6

13.
14.
15.
16.
17.
18.
19.
20.
21.

.61
.79
1.03
1.17
1.36
.067
.086
.16
.273

.98
.99
.01
.019
.034
.11
.06
.36
.18

Cs-134
Cs-134
Cs-134
Cs-134
Cs-134
Cs-136
Cs-136
Cs-136
Cs-13633xI0
2.33x016
2.33x10 6
2.33xi06
2.33xi0 66.66xI06
6.66xI0 6
6.66x10 66.66x10 6

2.28xi06
2.30x1064
2.33x10 4
4.43XI04
7.93xI05
7.33x1 9

5.18x105
1.08xO 4
4.91xI1R

4.0x1O6
2.4xI0 6
1.2xlO6

3.4xI0 5
3.8x10 5
3.28xI00

22.
23.
24.
25.

.340
.818
1.05
1.25

.53
1.0
.82
.20

Cs-136
Cs-136
Cs-136
Cs-1366.66x10 6

3.5xi0 6
6.66x606
5.46xI0 6
1.33xi0

1.19x1O6
5.44x10 65.73xI0 6
1.66xIO

II-B-20

2.33x1o 6

6.66xI0
6.66xi0 6

mlrI(E)(N)

2.6xlO-9
894xI03
3.26x107
7.68x18
2.IxlO
8.7xlO6
3.86xlb 6
4.6X606
2.4x106

Ix10 6
7.4xi0 8
1.43xi07
2.08xi0 6
5.55xi0 8
l.llxlO56
5.61xl0
4.79xI06
5.93XI06
2.06xl6

2.4xi0 5

1.6xlO 55.36x1o 6

3.o5x1R

1.4xlO 6
62xi0
1.8x14i

E (MeV)f

Sourcedps/ml

26.
27.
28.

.662
.512
.622

29.
30.
31.
32.
33.
34.
35.
36.
37.
38.

1.05
1.13
1.55
.03
.163
.305
.438
.537
.053
.23

.85
.21
.11
.015
.005
.002
.06
.06
.06
.05
.34
.17
.90

Cs-137
Ru-106
Ru-106
Ru-106
Ru-106
Ru-106
Ba-140Ba-140Ba-140Ba-140Ba-140
Tc-132
Tc-132

1.03xl'7
7.4xI07
7.406
7.4x10 6
7.4x10 6
7.4x10 6
7.4x10 6
7.4x10 6
7.4x106
7.4x10 6
7.4x10 6
7.4x10 6
7.4x10II-B-21- Rh-106- Rh-106- Rh-106- Rh-106- Rh-106N
photons sec8.8x10 6
8.14x105
1.llxl5
3.7xi00
1.48xI0 5
8.14xlO5
4.44xi05
4.44x1
3.7x1005
2.52xi0 6
1266x06.66ximl(E)(N)
5.83xlO6
5
4.55xI06
7 94xI0
1.17xlO5
4.18xlO4
2.29xl0 4424105
1.35xI0 5
1.62x10 56.67x10 4.67xI0.53xi0 6

E

INiEi
IN.1

INi

=

7.53x10 8 photons sec- 1 ml-

EiN=

3.32xi0 8 MeV photon sec-

E=

0.44 Mev@ 0.44 MeV pen/p for tissues • pen/p forH 20
P/p•

0.1 cm2 from Radiological Health Handbook p.
g

D

=

k( 0
)(S )(V)(E)(p
/p)
en
v
RoVS

D= (l.6xlO 8 rad MeV-)(l.3xlO
photon*2 cm- 2 sec-=

6.89 rad/sec=

24800 rad/hr

dis- sec)(7.53xlO

133*

photon sec

ml)(lxlO ml)(.44 MeV)(.1 cm )g

The value P/p is the mass attenuation coefficient.
If the linear coefficient is used (.03 cm2 /gm) the dose
rate from a one liter sample would be approximately 7440 rad/hr which is in better agreement with Teletector
data. The principle difference between the two'coefficients is that the mass attenuation coefficient includes
compton, photoelectric and scattering interactions while the linear attenuation coefficient does not include
scatter.

II-B-22

ASSUMPTIONS AND CALCULATIONS (Continued)
3.

The skin dose (beta and gamma components) is evaluated using the techniqueof Walter Snyder et. al. in Absorbed Dose Per Unit Cumulated Activity For
Selected Radionuclides and Organs, Society for Nuclear Medicine, New York,
October, 1975.
The skin activity was estimated from the measurements made by subject A on
March 29, 1979. A measured skin dose rates with a side window geiger tube.
On June 6, 1978 a Ba-133 source set up to simulate 1-133 was assayed with aside window geiger counter similar 2 to that used by A. The source was a disc4.7 cm in diameter with a 7 mg/cm mylar cover. It had a 1-131 equivalent
activity of 0.19 pCi. The count rate was 0.2 mR/hr closed window and 0.4
mR/hr open window. For calculation purposes the open window data was used.
Subject A indicated that the skin on left forearm was contaminated in a
spot about 2 inches diameter (5.08 cm).
Therefore, the 4.7 cm Ba-133 discsimulates well the contamination geometry.
To obtain the total skin dose in the period 3/29-31/79, the integrated
activity time is required.
Figure 1 is a curve of activity in thecontaminated area as a function of time. The integral of the curve is
pCi-hrs.
Using the formulation of Snyder et. al. the dose to the skin may beobtained by assuming uniform concentration throughout the volume contaminated. The dose to the affected area is given by

D=

SftA(t)dt
fwhere s = absorbed dose per pCi-hr. for entire skin organ assuming
uniform concentrationfA(t)dt is the graphically determined pCi-hr. contaminationf=

The fraction of the entire organ which is contaminated.
This fraction is required to correct for the fact-that the
Snyder calculation averages the dose to the entire organfrom uniform contamination. Here we assume that there is alinear relationship between activity per unit area anddose.

II-B-23

From ICRP 23, Report of the Task Group on Reference Man Pergamon Press, New
York, 1975, the following skin data was taken:
1.8x10 4 cm2
42
2.5x10 4 cm2
1.6xlO cm

Surface area of typed adult male
Surface area fore subject
by height = 188 cm
by weight = 72 kg

Since subject is thin, the values of 1.65x10 4 cm2 will be used.
The contaminated area was estimated by the subject to be a circle about
5-10 cm diameter.

2

The area would be 19.6-78.5 cm .

purposes a 100 cm2 area will be used.
f

=

1xlO2

=

1.65xi0

For calculational

From this area f may be determined.

6.1xlO3

4

From Snyder et. al. (MIRD No. 11 p. 185, 1975) S = 1.6xlO- 4 rad/pCi-hr. for
1-131.
Since 1-131 was the major contaminant, we assume it was responsible
for all skin dose. This is a nonconservative assumption, since we know that
1-131, C -134, C -137 and Co-60 were also present between the onset of contaminatign and the second whole body count.
Therefore, the forearm skin dose is estimated:(l.6xlO- 4 rad)(1.38x10 3 pCi-hr)
pCi-hrDskin

3

6.1xlO=

36.2 rad = 36.2 rem

Since the area contaminated is probably less than 100 cm2 , this dose repre
sents a lower limit. If an area of 78.5 cm2 (10 cm diameter) is used, thedose will be 46.1 rem. This would represent the upper bound of skin dose.
Corroborating Information
The body worn TOD for subject A indicated a dose of .89 rem for 3/29/79. This isessentially the same value as calculated in the time/distance study. While thismay be fortuitous, it is nevertheless an indicator of the reasonableness of theII-B-24

time/distance study. Also dose rates for the 100 ml graduated cylinder whentreated as a line source at distance less than 12" and by calculation as a photon
source of a right circular cylinder indicate compatible dose rates of 2400-4000
R/hr at contact - 1". This agrees with teletector measurements. Since it forms
the basis for other whole body dose rates and since there is close agreementwith the whole body TLD, credence is lent to the calculation.
Stay times for the cases of A and B were based on interviews with subject A andre-enactments of mockups of the area by the inspector.

II-B-25

I'II-

DD~

50
CONTAMINATION vs. TIMESUBJECT A

45-

40-

z 35
30
0

25 I-

2015k

56 hrsflO1

A(t)dt 13841ACi-hrs

N

0

N.

N.

N'.
N

5k

0

10

20

30

40

50

60
3/31/79
24:00

3/29/79*16:00Figure II-B-1
II-B-26

70

80

90

TABLE II-B-4
DOSE ESIMTATE FOR SUBJECT B
Circumstances of Exposure
B is the chemistry foreman identified as subject B in Enclosure 3.
The following outline summarizies B's actions on March 29, 1979 to present.
March 29, 1979TIME

ACTION"15:00

B and A proceeded to the Unit 1/2 sampling room i'npreparation for taking a letdown sample.

15:00-15:30

B and A review pipe runs.
B enters room. Lineup and collects a Unit 1 letdown sample.

15:30

B and A entered sampling room for initial survey.
12 sec, dose rate 6-8 R/hr.

Stay time

B aligned valves at sample cooler and started letdown
sample recirculation.
Stay time ".180 sec. Dose rate W/B
and extremity 10 R/hr.
B exited area 3 sec 6-8 R/hr.
B entered area, adjusted letdown drag valve to establish
flow 120 sec @ W/B 17-20 R/hr hand 90 R/hr. B took a 300 ml
sample in a polyethelene bottle (2.1 rad/sec contact). He
then poured 100 ml into a 100 ml graduated cylinder. B also
took a 40 ml aliquot capped the bottle and left room. Saipple
handling time 60 sec contact (W/B dose ^- 3.33xi0
rad/sec). 120 sec handling time @ .28 rad/sec w/40 ml
aliquot in the primary sample lab.
B entered sample room and shut off sample cooler.
B exited.
Pocket dosimeter was > 5R.
was 50 mR/hr W/E-520, w/HP-177.

One spot on finger

B was whole body counted on 3/29/79.
contamination was noted.

At that time, surface

II-B-27

On 3/29 a WBC indicated:
1-131
1-133
Cs-134
Cs-137
Co-60

=
=
=

6875 nCi
3228 nCi
293 nCi
372 nCi
172 nCi

The position scan is strongly indicative of surface
contamination.
Approximately 50% of the initial activity
was on the hands and 20% on the surface of the head.
Initially, a small fraction (less than 5%) was internal.
By April 3, the hand contamination was negligible. Surface
contamination on the head was essentially all 1-131 and wasapproximately
600-977nCi.
One
fingerremained
contaminated.
A WBC on April 24 indicated 70 nCi on the surface of thehead. The data is plotted in figure 1.
Th2 area of the head estimated to be contaminated is 50-100
cm
An evaluation of B's dose for March 29 and dose due to skincontamination follows in Table 1.

II-B-28

TABLE 1
EVALUATION OF B's DOSE ON 3/29/79 AND SUBSEQUENTDOSE DUE TO SKIN CONTAMINATIONDOSE RATES * (R/sec)
ACTIONDOSE (rem)DURATION

WHOLE BODY

EXTREMITIES (fingers)

WHOLE BODY

B and A enter
sample room for
initial survey

12 sec2.2x10 3

2.2x10 3-

2.6x102.6x102.6x1-

B aligns valves
on sample cooler
drag values

180 sec3
2.8x10 3

2.8x10 3

5x1O-

5x1O 1

5x1O-

3.

B exits area

3 sec

2.2x10- 3

2.2x10- 3

6.6x10- 3

6.6x10- 3

6.6x10- 3

4.

Entry to dragvalves12 sec5.6x105.6x106.7x106.7x106.7x10Adjust dragvalves

120 sec

5.6x10 3

2.5x10 2

6.7x10-

3.0

6.7x10 1

Collect 300 ml
primary coolant
in 1 liter polybottle60 sec1.9x10 2

1.4

1.17

84

1.17

Draw 40 ml sample

6 sec

5.6x10-3

2x10-1

3.4x10- 2

1.2

3.4x10- 2

1.

2.

5..
6.

7.

EXTREMITY (fingers) SKIN OF HEAD

-2

2

2

II-B-292

2

TABLE 1
EVALUATION OF B's DOSE ON 3/29/79 AND SUBSEQUENTDOSE DUE TO SKIN CONTAMINATIONDOSE RATES *ACTION
8.

9.

10.

11.

DOSE (rem)DURATION

WHOLE BODY

EXTREMITIES (finqers)

WHOLE BODY

Pour 100 ml into
graduated cylinderfrom 300 ml bottle

16 sec1. 9x10-2

1.4

3x10-

Dump 100 ml
back into 1 1bottle66x10-Work with 40
ml sample inprimary lab

120 sec

1. 2x10Skin contamination on head

1ST QTR 1979
2ND QTR 1979

3

4. 6x102

2. 8x10-1

1

3. 6x10EXTREMITY (fingers) SKIN OF HEAD

22.4

3x10-1

2.7

3. 6x10-

33.6

1.44

2

1

1.44

2

1.8 - 8.8 rem

3.9 to 19.7 rem
TOTAL

4.25

147

**6 - 13 rem in

the 1st qtr
4 - 19 rem 2nd qtr
For extremities the sources are considered to be line sources*Dose rates are based on teletector data.
at less than 12" and point at distances greater than 12"**This is the sum of the whole body dose and the dose to the skin of the head.
For step 8 the doses may have been somewhat reduced if the 40 ml sample had beep taken from the 300 ml aliquot
drawn earlier.
In that case whole body dose rate would have been about 1.6x10
R/sec and the extremity dose
rate would have been 1.2 R/sec.
The net effect would be to reduce the calculated whole body dose to 4 rem andthe finger dose to 144 rem.
The value of 4.0 rem for W/B is used in Table II-B-2 for the lower bound for wholebody dose for the finger dose in Table II-B-2.

II-B-30

1.20

9'
-9

1.15
1.10

HEAD & NECK CONTAMINATION vs. TIME

1.05
1.00

SUBJECT B

0.95
0.90
0.85

9'

\'

0.80

z
I-

\'

0.75

9'

0.70

9'

0.65
0.60

z

0.55 -

LU

z

9\

-

z
0

9\

-

27\
9'

A(t)dt • (0.5•\(A) (t) -- APPROXIMATE AREA BY A TRIANGLE

-

0.50

0

0.45
0.40

t= 27d\f",

0.35 _

0.30

9'•14.8 jCi Days

t=0345 pCi-hours

D = 11--22 REM%

t = 56.5

0.25
0.20
0.15

f

65.05 Ci-hrs

D = 1.8-8.8REM---290.4 pCi-hrs

D = 7.9 -

t= 0
27d

0.10
0.05

19.7 + W/B

56.5

0

5

10

3/29

4/3

4/13

15

20

25

4/23

30

35

5/3

TIME PAST CONTAMINATION (Days)
Figure II-B-2
II-B-31

40
5/13

45

EVALUATION OF SKIN DOSE DUE TO CONTAMINATION AND DIRECT RAOIATIG9SUBJECT:

B, Chemistry Foreman

Assumptions
1.

The area contaminated is uniformly contaminated.

2.

All body skin is of uniform thickness.
This is a slightly conservative assumption if the thickness selected is that of trunk.
The 1-131 is assumed to be deposited within the skin. This is apparent
since portions of the originally identified contamination were removed
upon washing, scrubbing and chemical (KMnO 4 ) treatment.
Following sucha regimen residual contamination would be within the tissue.

3.

The mass of any skin section is in proportion to its area fraction of thetotal body.

4.

Contamination of the hair resulted in a minimum of 50% being deposited on
the skin of the scalp, however for determining the upper bound to skin dose
100% deposition will be assumed.

II-B-32

Evaluation of HP-210 Counting Efficiency for 1-131 on Skin.
1-131 (0.156 pCi) was spread over a circular section of fresh, shaved pig skin.

The area contaminated was (nt(2.5cm 2 )=19.63 cm2
An HP-210 probe coupled with an RM-14 rate meter was used to count the skin on
contact.
The count rate was 2600-3000 cpm.
The area contaminated is somewhat
larger than the area of the HP-210 pancake probe.
HP-210 Efficiency:
2800 cpm=

2.156 pCi/19.63 cm

3.52xi0 5 cpm
PCi cm-2

Data1.

HP-210/RM-14 counting efficiency for 1-131 as measured on pig skin is
3.5xi0 5 cpm/PCi/cm2 .

2.

Whole Body Counts (WBC).
a.

The WBC of 3/29/79 identified a total of 6875 pCi 1-131 as both internal and external.
No thyroid shield was used.
In addition,
1-133, Cs-134, Cs-137 and Co-60 were identified.

b.

A WBC on 3/29/79 with a thyroid shield in place indicated 7045 nCi
1-131 total body.

c.

Analysis of the positional WBC and scan data from self monitoring
indicated about 20-25% of activity on head and neck with remainder
on hands.

d.

WBC of 4/3/79 indicated only 1-131 and 10 nCi Cs-134.
The very rapid
elimination of Cs-137 and Co-60 is strongly indicative of skin contamination.

e.

WBC of 4/3/79 indicated 1-131 and 10 nCi Cs-134 only. Major location
is at head.
The failure to see Co-60 or Cs-137 and great reductionin Cs-134 is very strongly indicative of almost all external.

II-B-33

2.

Evaluation of Contamination on Head, Neck, Hands w/HP-210*a.

LEFT WRIST AND FOREARM

DATE

At

HEAD

NECK

3/29/79

Od> 5xlO4 cpm

3xlO4 cmp

> 5xlO4 cpm

4/4/795d

3xIO 3bkg

7xlO

4/11/7912d

3xIO 3bkg

2.5xi 03

MDA = l.1xlO- 2 pCi = 11 nCi/20cm2
2

20 cm

II-B-34

3

b.

3.

The WBC for 4/3/79 indicates on the positional scan approximately
90% of activity on the head. The hands had been shielded by
4/3/79. The GMSM of the neck indicated'background.
The only
other head area indicating contamination with a GMSM was the topof the head, or the temple.

Dose to Head Skin
Figure 1 plots the WBC data which indicates a major contamination on
the head.
These are the data from the 4/3 and 4/24/79 whole body counts.
The positional counts indicate that this activity was largely on the head.

CALCULATION METHOD
1.

The area under the curve in figure 1 is calculated in terms of pCi-hrs.
the first quarter then for the second quarter.
For the first quarter, the area is approximated by a rectangle (y axis
at t = 0 n 1.17 uCi. At = t = o to t = 56.5 hrs)
55.6
f A(t)dt(1.17 pCi)(55.6 hrs) = 65.05 pCi-hrs.

0

II-B-35

forFor the second quarter, the integrated contamination time is from t=55.6
hrs. to t=23 days (55.6 hrs. to 552 hrs.).
552
f A(t)dt is the area of a triangle = (½) (Activity) (At)
55.6
=

2.
3.

(0.5)(1.17 pCi)(552-55.6) = 290.4 p Ci-hrs.

The skin contamination on the head is estimatid conservatively to be 100% of
The area of contamination < 20 cm
the total.
2
4From ICRP 23 we estimate the skin area of this individual to be 1.7xlO cm .

4.

From MIPD II the dose to the skin organ from uniformly deposited 1-131 is
1.6xlOrad per pCi-hr.

5.

The skin dose is calculated as follows for the Ist quarter.
56.5 hr
D = (fA(t)dt)(fs)(l.6xlO-4
0

rad
)(Ab)
pCi-hrAcwhere fA(t)(dt is the total activity time for the period t=O to t=24:00 3/31/79.
f

is the fraction of contamination which was on the hair and skin which i-s
estimated to be on the skin only (1.0 for skin of temple).

AB is the total body skin area = 1.7xlO

4

2

cm

AC is the area of contamination = 2xlOI cm2 (temple).
D=

1st qtr-24(65.05 pCi-hrs)(1.6xlO-4 rad)(1.7x19 c )mCi-hr
(2xlO" cm-)

8.8 remThis is the maximum dose due to skin contamination on the head (temple) for
the first quarter 1979.
To this the first quarter whole body dose must be
added. That was 4.11 rem as measured by licensee TLD, 4.25 rem by calculation.
Therefore,

the total quarter dose to the head is

Dcontamination + Ddirect = 8.8 + 4.1 = 13.3 rem.

II-B-36

If tihe area ,of the skin contamination is 100, 50,
slki:m doses result:or 20 cm2 the following(17000 cm2 total body)(65 pCi-hrs.)(l.6xl0- 4 rad)=1.8 rad in the 1st qtr=

3.6 rad in 1st qtr

100 cW2 contaminatedpCi-hrs.

(17000 cm2 total body)(65 pCi-hrs)(l.6xlO- 4 rad)50 cm2 contaminatedpCi-hrs.

(17000 cm2 total body ) (65 uCi-hrs) (1.6x10-4 rad )(20 cm2 contaminated)=

8.8 rad in the 1st qtr

uCi-hrs

1.8 rem represents the lower bound for head dose, due to contamination of
the skin.
8.8 represents the upper bound due to contamination of the skin.
Subject B's chest worn TLD was read out on 3/30/79.
It indicated awhole body gamma dose of 4.1 rem. This value is within 15% of thevalue calculated herein for whole body dose and lends credence to thecalculated value.
Further supporting data is indicated by the pocket
dosimeter which read 5R at the end of the activity on March 28.

II-B-37

6.

The maximum skin dose due to skin contamination of the head for the second
quarter is calculated as follows:648
0 = (f A(t)dt)(f s(l.6xlO 4 rad)(A
56.5
pCi-hr.
bAc= (290.4 pCi-hrs.)(0.5)(l.6xlO-4 rad

pCi-hr.
Here f)(1.7xlO4 ) = 19.75 rem

2xlO1

is estimated at 0.5 as B was not positive of the fraction oflong t~rm contamination on the temple.
For a contaminated area of 100 cm2 the skin doses would be -.4 rem.
To this the whole body dose for the second quarter must be added.
In the first quarter, the skin dose will be higher than calculated
here due to contamination by 1-133, Cs-134, Cs-137 and Co-60.
The licensee should make additional evaluations of the dose due tocontamination for the first and second quarter 1979.
In addition,
to the dose due to 1-131, the dose due to 1-133, Cs-134, Cs-137 andCo-60 needs to be evaluated. The evaluation here is, therefore
non-conservative in that it intentionally neglects dose contributions
from 4 isotopes. The purpose here was to demonstrate that if, in the firstquarter, a significant skin dose could be realized from 1-131 alone, thenthe contributions of other contaminants added in would surely be a significant dose.

II-B-38APPENDIX II-C
DIRECT RADIATION MEASUREMENTS MADE BY METROPOLITAN EDISONCOMPANY ON AND OFF THREE MILE ISLANDNOTES:

1.

See Figure 11-3-2 for location of GE series survey points.

2.

A "+" in the offsite survey table indicates that the surveyý
was made in the indicated sector but at a greater distancefrom the plant.

3.

A "+" in the onsite survey table indicates that the survey
was made at a location other than the GE series survey
point.

4.

A "x/y" indicates a ýy/y measurement.
If only a single
value without a diagonal (/) is shown, it is not known
whether the measurement is ýy or y only. A dash (-) indicates that the ýy or y component was not measured.

5.

All values are in mR/hr.

6.

N-OI, -11, -21, etc. refers to 0.5-1,
respectively north of the site.

1-2, and 2-3 miles,

TABLE II-C-I
Direct Radiation Measurements Made by Metropolitan Edison Off Three Mile Island(Locations shown are closest survey points in each sector)
Date

Time

3/28/79

Prior to0700

N-l1 NNE-01

0700

NE-0l

ENE-O

E-01

<1

ESE-OlSE-O

SSE-OI

S-0l SSW

SW-ll

WSW-11

W-1l

<1

<1
<1

WNW-l1

NW-21

<1

0800

0900
1000

<1

<1

+

<1
1100

<1

2

1200
1300

1

1400

++

1500

3
<1

+

20
50

3

<1

2.5
<1
<1

+

<1

<1

NNW-21TABLE II--C-i (Contiriued)
Direct Radiation Measurements Made by Metropolitan Edison Off Three Mile Island(Locations shown are closest survey points in each sector)

Date

Time

N-11 NNE-01

3/28/79

1600

+

NE-Ol

ENE-01E-01 ESE-ol

<1

<I

<1

SE-O

SSE-01

S-01 SSW

SW-]i

WSW-11

W-l1

WNW-11

NW-21

NNW-21

<1

<1
1700

<1
<1

1800

<1
+
0.5
<.05 <1
<1
<.05

C-,

1900

<I

<.05
+

2000

2100
2200

2300

12
<1

+

+
+
+

TABLE II-C-2
Direct Radiation Measurements Made by Metropolitan Edison Off Three Mile Island(Locations shown are closest survey points in each sector)

Date

Time

N-11 NNE-01

3/29/79

0000

+

NIE-O

ENE-Ol

E-01 ESE-Ol

SE-0l

SSE-01

S-O

SSW

SW-Il

WSW-11

W-ll

WNW-1I

+

++

0200

.0300

C)

27

0500
0600

30/20

0700

1/1

6/3

2/2

+

0800
0900
1000

<0.5

NNW-21
+

3.5
3.5

0100

0400

NW-21

+

TABLE II-C-2 (Continued)
Direct Radiation Measurements Made by Metropolitan Edison Off Three Mile Island(Locations shown are closest survey points in each sector)
Date

Time

N-ll

NNE-01

NE-01

ENE-OlE-01

ESE-OT

SE-OI

SSE-01

S-01 SSW

SW-II

WSW-l1 l

W-ll WNW-ll

1100<.5/<.5

1200

+

1300

-/<.05

1400
1500

1600
1700
1800

+

1900

+

<I/-

2000
2100
2200
2300

+++

0.5

+

NW-21

NNW-21TABLE II-C-3
Direct Radiation Measurements Made by Metropolitan Edison Off Three Mile Island(Locations shown are closest survey points in each sector)
Date

Time

N-li NNE-01

NE-O

3/30/79

0000

+

+

<.1

4+

ENE-O

E-01 ESE-O

SE-O

S-0l SSW

SW-l1

WSW-I1

W-11 WNW-11

NW-21

NNW-21.15.2/<.l

.2/-

0.2/-

0.2/-

0200

2.0
0.4

0300

SSE-01

+

+

2.0/-

2.5/S.1-.15

0400

+
.02

.1/.5

.02

+

.031.020500
0600+.++ .5
+
.6

<.I

.2

.15/.05 .1/.05
+

.75
2
0700

.35

.15/.15+.15/.05TABLE II-C-3 (Continued)
Direct Radiation Measurements Made by Metropolitan Edison Off Three Mile Island(Locations shown are closest survey points in each sector)
Date

Time

N-11 NNE-01

NE-0l

ENE-OI

E-O

ESE-01SE-0l

SSE-010800

5-01 SSW SW-II
+

10/.4

0900

1.0

+
8/4.5
...

+

.2/.05WSW-I I

W-1l
1.5/.4

++
...
8/4
3.0/-

1000

1.2/.5

5.5/2.0

1100

1.0/.44.0/1.8

1200.11.02

2/1

2/.5
1300
1400
1500

21.7

1600

6/I/-

WNW-ll

NW-21

NNW-21TABLE II-C-3 (Continued)
Direct Radiation Measurements Made by Metropolitan Edison Off Three Mile Island(Locations shown are closest survey points in each sector)

Date

Time

N-11 NNE-01

NE-0l
N

ENE-01E-01 ESE-O

SE-0l

SSE-01

S-01

SSW

SW-l1

WSW-l1

W-ll WNW-ll

NW-21

NNW-21++.1/.051.2/.2

1700

6/1
.6/.l
.3/.05.06/.0l.02/.011800.04/.02

+

1900
C-)

<.01/<.0l

2000
2100

+
+

2200

+
+

2300

5/2

<.01/-TABLE II-C-4
Direct Radiation Measurements Made byMetropolitan Edison On Three Mile Island

Time
Interval

GE-I

GE-2

GE-3

GE-4

GE-5

GE-6

GE-7

GE-8

GE-9

GE-IO

Surveys at
Other
Locations
On Three
Mile Island

3/28/79
Prior
to 0700
+

0700
0800
C-

+

0900

+

7

1000

+
+

+

2

1100

1200

<1

1300

3

+

8

+

+÷

NOTE:

GE number = sample location on Three Mile Island

TABLE If-C-4 (Continued)
Direct Radiation Measurements Made byMetropolitan Edison On Three Mile Island
Surveys at
Other
Locations
On Three
Mile Island

Time
Interval GE-I

GE-2

1400

30

++

1500

70
8
12

+

1600

50

+
++

1700

GE-4

GE-5

GE-6

GE-7

GE-8

GE-9

GE-10

140

40

<1

+
+

+

26

1800
1900

GE-3

3

10

+++
+.+

2000

++

2100

+

TABLE II-C-4 (Continued)
Direct Radiation Measurements Made byMetropolitan Edison On Three Mile Island

Time
Interval GE-i
Interval GE-1
2200
2300

C)

30/20

GE- 2
GE-2

GE- 3

GE-3

GE-4

GE-4

GE-5

GE-5

-6
GE
GE-6

GE
GE-7-7

GE-8

GE-9

GE-10

1

-/10

365/50Surveys at
Other
Locations
On Three
Mile Island+

TABLE II-C-5
Direct Radiation Measurements Made byMetropolitan Edison On Three Mile Island

Time
Interval

GE-1

GE-2

GE-3

GE-4

GE-5

GE-6

GE-7

GE-8

GE-9

GE-10

30/20

1

50

25

150/100

3/2

Surveys at
Other
Locations
On Three
Mile Island

3/29/79
0000
0100
T•
0
!

0200
0300
0400

26

28

0500
0600

0.5/0.5

0700
0800

6/3.5
<0.5

0.5/-

0.5/-

5/1

11/5
11/5

<1/<I

+

TABLE II-C-5 (Continued)
Direct Radiation Measurements Made byMetropolitan Edison On Three Mile Island

Time
Interval

GE-I

0900

<0.5

1000

<0.5/<0.5 <0.5/<0.5

GE-2

GE-3

GE-4

GE-5

1/<.5

<0.5/<0.5 .5/<.5

GE-8

l/<l

GE-10

10/10
3/3
1.5/<0.5

1.5/.8
4/3

5/2

5/3

2.5/1.5

0.5/<0.5<0.5/<.5

1/0.5

3/<0.5

1/<0.5

5/3

4/2

5/3.5

+

0.5/<0.5

1300
1400

GE-9

0.8/-

0.5/<0.5

IN

3/3

GE-7

....

1100

1200

GE-6

Surveys at
Other
Locations
On Three
Mile Island

8/4

10/4

<1/<1

3/1

10/3

+

2/1

10/4

+

2/1
1500
1600

9/4

+

TABLE II-C-5 (Continued)
Direct Radiation Measurements Made byMetropolitan Edison On Three Mile Island

Time
Interval

GE-i

GE-2

GE-3

GE-4

GE-5

GE-6

GE-7

1800

GE-10

0.4/0.3

5/1.8

2/1

0.6/0.25

1.4/0.9

1.4/0.9

0.4/0.3

0.2/0.2

0.15/0.1<.01/<.01 .08/.08

13/7

0.4/0.2

0.2/0.1

0.05/0.04<.01/<.0l .16/.02.2/.08.2/.l

1.8/0.4

0.9/0.2

0.04/0.01 0.05/0.025

2000.04/.01

2100.02/.0l

2200<.01/<.01 .02/<.0l

2300

GE-9

7/3

1700

1900

GE-8

Surveys at
Other
Locations
On Three
Mile Island.05/.03.02/.02.03/.0l.03/.Ol.3/.14++TABLE II-C-6
Direct Radiation Measurements Made by
Met-Ed On the Island

Ti meInterval
3/30/79

GE-I

0000
0100

GE-2

GE-3

GE-4

.03/.01.05/.03

GE-5

GE-6

GE-7

GE-8

GE-9

GE-10

.04/.01.05/.04.2/.05.01/<Ol.1/.05

3/0.8.7/.25.3/.15.2/<01

0.1/0.03
0.4/0.1

0200

.8/.2
.7/.25/.3/-

-C,

0300

Surveys at
Other
Locations
On Three
Mile Island.3/.08

.3/.l

0400

0500

0.03/0.01 0.01/<.010600

2.5/0.5.01/.01.25/.07

1.4/0.4.3/.1.2/.06.3/.0714/3.5
4/1
5/1

.3/.1

.3/.13.06/.04.1/.03.14/.08

.71.2

2/0.5

<.01
<.01

.1/.1

+

<.01/<.01,4/.l.4/.08.65/.06.15/.05.2/.05.4/.1

2.0/0.5

3.5/1.0+

TABLE II-C-6 (Continued)
Direct Radiation Measurements Made by
Met-Ed On the Island

Time
Interval

GE-I

GE-2

GE-3

GE-4

GE-5

GE-6

GE-7

GE-8

GE-9

GE-10

0700

Surveys at
Other
Locations
On Three
Mile Island+

0800

0.3/8/4
30/9

0900

4/1

10/-

14/8
25/8
0.3/0.1

0.5/0.2

++

++
1-4

-_J

(31

1000

.3

0.2/-

1100

1.7/0.7
8/1

4/1
7/2.5
8/2.5

1200

6/1
4/.7

2/0.3

1300

6.5/2.5

0.3/-

0.3/<0.1

9/3

13/6

7/4
19/7
25/9

8/3.5
15/5
10/3
11/3

6/4
5/1.5
7/1
6/1

+

4/0.5
1.5/0.5
4/1.8

18/5
18/3.5
3/0.4

-/3.5
17/5
5/2

+

3/0.5

3.5/1

.8/.2
7/3

5.5/1.5
5/1.5
3/1.2

12/2
6.5/2.5
17/6+

TABLE II-C-6 (Continued)
Direct Radiation Measurements Made by
Met-Ed On the Island

Time
Interval
1400

GE-I

GE-2

GE-3

GE-4

GE-5

0.8/0.1
1.3/.43/1.2

0.8/0.1

4/0.3

4/0.5

GE-6

GE-7

GE-8

1600

1700

GE-10

7/1
110/205/3

6/1<0.1/-

2.5/-

40/24
50/10
20/-

8/2
30/

+
+

11/3

100/

.6/.2

+

5/-

30/-

.15/.05

80/15
20/10
14/6
60/7

0.9/0.1
3/0.6

++
++
+

10/1.2

10/1.8

8/1

5/0.5

1800

1900

.1/.03<.01/<.O1

2000.15/.03.04/.01.04/.01.01/.01.07/.O1.02/.01.01/.01

+

+
+

90/10
40/25

1500

I
I
,=•
Cr•

GE-9

Surveys at
Other
Locations
On Three
Mile Island.4/.01.02/.01.01/.O1.01/.01<.01/<.01 <.01/<.01 .3/.l

0

TABLE II-C-6 (Continued)
Direct Radiation Measurements Made by
Met-Ed On the Island

Time
Interval

GE-I

GE-2

GE-3

GE-4

GE-5

GE-6

GE-7

GE-8

2100
2200
2300

0!

2/0.1/-

0.2/-.15/0.1/-

.05/-

.25/-

GE-9

GE-10

20/-

20/-

12/12/2.5/-

8/2

Surveys at
Other
Locations
On Three
Mile Island++

APPENDIX II-D
CAUTIONS REGARDING THE USE OF INSTALLED RADIATION MONITORING SYSTEM DATACAUTIONS REGARDING THE USE OF INSTALLED RADIATION MONITORING SYSTEM DATA
When using radiation monitoring system data, it

should be remembered thatthe radiation monitoring system provides very little quantitative data.

Duringthe accident, atmospheric and process radiation monitors were measuring highbackground radiation from radioactive gases in the auxiliary building and highlevels of direct radiation from the reactor coolant system letdown piping andother sources such as ventilation ducts.

Area radiation monitors were,insome cases, measuring both gases around the monitor and direct radiation fromcertain process components.
the monitors made it
addition,it

The influence of a high gaseous activity aroundimpossible to quantify radiation being measured.

In

is not always possible to rely on the time indications on the

strip chart because they were not clearly and regularly marked and/or because
the paper did not advance properly on the recorder.

Certain of the monitor

traces were impossible to identify because the recorder was not inking properly.
Since many of the monitors were responding simultaneously, their traces overlapped and were difficult to follow.
Certain monitors can, however,assurance.

be used to establish times with reasonable

The times may be established based on the monitors'

in relation to certain operational occurrences.
HP-213,

Incore Instrumentation Area Monitor,

Closed Cooling Water Letdown Monitor.

general response

An example of these would be

and IC-R-1091,

Intermediate

Both of these monitors responded downward

almost instantaneously on reactor trip, probably due to the decay of Nitrogen
16 which has a 7 second half life.

II-D-1APPENDIX II-E

ACRONYMS,

DEFINITIONS,AND UNITS OF MEASURE

APPENDIX II-E

ACRONYMS,

1.

DEFINITIONS,AND UNITS OF MEASURE

ACRONYMS

ABST

Auxiliary building sump tank

AMS

Aerial monitoring system

ANI

American Nuclear Insurers

BRH

Bureau of Radiological Health

BWST

Borated water storage tank

CaSO4 :Dy

Calcium Sulfate: Dysprosium

CaSO 4 :Tm

Calcium Sulfate:CRO

Control room operator

DCCD

Dauphin County Civil Defense

DER

Department of Environmental Resources

DOE

Department of Energy

DOP

Dioctylphythalate

ECC

Emergency Control Center

ECS

Emergency Control Station

EMOVElectromatic relief valve

EPA

Environmental Protection Agency

ES

Engineered safeguard

FM

Frequency Modulation

Thulium

II-E-1FSAR

Final Safety Analysis Report

GAI

Plant page system

Ge(Li)

Lithium Drifted Germanium

GM

Geiger Mueller

GPU

General Public Utilities

GPUSC

General Public Utilities Service Corporation

HEPA

High efficiency particulate air (filters)

HEW

Department of Health, Education,

HP

Health physics

IWFS

Industrial waste filtration system

IWTS

Industrial waste treatment system

LiF

Lithium fluoride

LOCA

Loss of coolant accident

MDA

Minimum detectable activity

MET ED

Metropolitan Edison

MUT

Makeup tank

MWHT

Miscellaneous Waste Holdup Tank

MWST

Miscellaneous Waste Storage Tank

NAWAS

National Warning System

NPDES

National Pollution Discharge Elimination System

NRC

Nuclear Regulatory Commission

OTSG

Once-through steam generator

PEMA

Pennsylvania Emergency Management Agency

PORC

Plant Operations Review Committee

PSE&G

Public Service Electric and Gas

RAP

Radiological Assistance Plan

II-E-2and Welfare

RB

Reactor Building

RCBHT

Reactor coolant bleed holdup tank

RCDT

Reactor coolant drain tank

REMP

Radiological environmental monitoring program

RCP

Reactor coolant pump

RMC

Radiation Management Corporation

RMS

Radiation monitoring system

RWP

Radiation work permit

SAI

Science Application Incorporated

SAM-2

Stablilized assay meter

SCBA

Self-contained breathing apparatus

SOP

Step-off-pad

TLD

Thermoluminescent dosimeters

TMI

Three Mile Island

WECST

Waste evaporator condensate storage tank

WGDT

Waste gas decay tank

II-E-3

2.

DEFINITIONSAssessment actions - those actions taken during an accident toobtain and process information that is necessary to make decisions to
implement specific emergency measures,to include protective actions.

Corrective actions - those emergency measures taken to ameliorate or
terminate an emergency situation at or near the source of the problem inorder to prevent an uncontrolled release of radioactive material or to
reduce the magnitude of a release,e.g.,

shutting down equipment,

repairand damage control.

Protective actions - emergency measures taken after an uncontrolled
release of radioactive material has occurred for the purpose of preventing
or minimizing radiological exposures of persons that would be likely todevelop if the actions were not taken.

Recovery actions - those actions taken after the emergency to restore
the plant or facility as nearly as possible to its preemergency condition.

Protective Action Guides (PAGs)- projected radiological doses-or

dose commitment values to individuals in the general population thatwarrant protective action following a release of radioactive material.
Protective actions would be warranted provided the reduction in individual
dose expected to be achieved by carrying out the protective action is not

II-E-4

offset by excessive risks to individual safety in taking the protective
action.

The PAG does not include the dose that has already occurredprior to the assessment.

Action levels - radiological dose rates; specific contaminationlevels of airborne, waterborne,

or surface deposited concentrations ofradioactive materials; or specific instrument indications (including
their rates of change) that may be used as threshholds for initiatingsuch specific emergency measures as designating a particular class of
emergency,

initiating a notification procedure, or initiating a particular

protective action.

II-E-5

3.

UNITS OF MEASUREMENT
cubic centimetercc
cfm

cubic feet per minute

=

curie

Ci
cpm

=

counts per minutegpm=

gallons per minute

1

=

liter

2

mg/cm

-

milligram/square centimeterml-milliliter

mrem

-

milliremmR-milliroentgen

ppm-

parts per million

pCi-picocurie

R/hr-roentgen per houruCi-microcurie

II-E-6

APPENDIX II-F
POTENTIAL ITEMS OFNONCOMPLIANCEAPPENDIX II-F
POTENTIAL ITEMS OF NONCOMPLIANCE
Listed below are items being evaluated as potential items of noncompliancewith requirements and commitments.
These matters will be handled through thenormal enforcement channels of the Office of Inspection and Enforcement.
A.

Technical Specification 6.2, "Organization,"states in Sections 6.2.1 and6.2.2 that the unit organization and the organization of the corporate
technical support staff shall be as shown on Figure 6.2-1.
Contrary to the above,on March 28,

1979,

the unit and corporate technicalsupport staff was different from that specified in that:1.

A position titled, "Superintendent of Administration and TechnicalSupport" was added to the organization on September 18,
filled on March 1, 1979,

1978 andsuch that the "Supervisor, Radiation Pro-

tection and Chemistry," reported to this new position rather thandirectly to the "Station Superintendent/Senior Unit Superintendent;"
and2.

On March 5,

1979,

the position "Station Superintendent" was changed

to "Station Manager," which reports directly to the "Vice President,
Generation,"rather than to the "Manager,

Generation Operations;"

and3.

The position of "Chemical

Supervisor" has been vacant since theissuance of the Technical Specifications; and4.

There are two "Supervisor of Maintenance" positions, one for each
unit, rather than one; andII-F-i5.

A position titled "Superintendent of Maintenance" has been added
such that the "Supervisors of Maintenance" report to this new position
rather than directly to the "Station Superintendent (Station Manager)/
Senior Unit Superintendent."

B.

(See Details II,

Technical Specification 6.8, "Procedures,"

Section 1.1)

requires in Section 6.8.1 thatwritten procedures shall be established, implemented and maintained for
implementing the Site Emergency Plan.
Three Mile Island Nuclear Station Health Physics Procedure 1670.9,"Emer-

gency Training and Emergency Drills," Revision 4,

1978:

1.

Identifies in Section 3.1,

dated January 16,

the on-site emergency job categories andrequires that training programs for these categories will be conducted
on an annual (calendar year) basis; and2.

Describes in Sections 3.1.1 through 3.1.9, the training program forall on-site emergency job categories; and3.

Requires in Section 3.1.3.2 that Radiological Monitoring Teams becomposed of Radiation/Chemistry Technicians

and Auxiliary "A"

Operators who have received the appropriate training; and4.

Requires in Section 3.1.5.2 that Repair Party Teams be composed of
Maintenance Shift workers who have received the appropriate training.

Contrary to the above,

during calendar year 1978,

all individuals having

emergency responsibilities were not trained in that two Emergency Directors,
one Accident Assessment individual,
Members,

eight Radiological Monitoring Teamand 37 Repair Party Team Members had not received the training

II-F-2

specified in Procedure 1670.9. On March 28,

1979,during an emergency,

atleast four individuals assigned to Radiological Monitoring Teams andseven individuals assigned to Repair Party Teams performed these emergency
duties without having been trained as specified in Procedure 1670.9.
Details II,
C.

(See

Section 1.2.1)

Technical Specification 6.4, "Training," requires that a retraining andreplacement training program for the unit staff be maintained that meetsor exceeds the requirements and recommendations of Section 5.5 of ANSI
N18. 1-1971.
ANSI N18.l-1971 recommends in Section 5.5.1,"Retraining," ten topics to

be included in a retraining program.
Contrary to the above, as of March 28, 1979,

a retraining program meetingor exceeding ANSI N18.l-1971 recommendations has not been maintained formembers of the Radiation Protection and Chemistry staff in that only 2 of
the 10 topics specified were included in the program.

(See Details II,

Section 1.2.2)
D.

Technical Specification 6.8, "Procedures," requires in Section 6.8.2 thatprocedures which implement the Emergency Plan be followed, and thatchanges thereto shall be reviewed by the Plant Operations Review Committee
and approved by the Unit Superintendent prior to implementation.
Contrary to the above, an improper change to Station Health Physics
Procedure 1670.7,"Emergency Assembly, Accountability and Evaluation,"

was implemented on March 28, 1979, without the required review andapproval in that:

II-F-3

By a memorandum dated October 13,

1.

1978,

from the Radiation Protection

Supervisor to all departments an additional assembly area was designatedand the method to be used to perform personnel accountability was modified without subsequent change to Procedure 1670.7; and2.

On March 28,

1979,

in response to an emergency,

some licensee personnel

followed the approved procedure for accountability while others
followed the accountability guidance in the October 13,

1978 memorandum,

creating some confusion and delaying prompt attainment of fullaccountability.
E.

(See Details II,

Section 1.3)

Environmental Technical Specification 5.7 requires that detailed written
procedures for instrument calibration be prepared and followed.
Three Mile Island Nuclear Station Surveillance Procedure 1302-5.24,
Revision 3,

dated December 19,

and requires that it

1974, specifies the method of calibration

be performed annually.

Contrary to the above, as of March 29,

1979, eight environmental air

samplers had not been calibrated since 1974.

(See Details II,

Section

1.7.1.1)
F.

Technical Specification 6.8, "Procedures,"written procedures shall be established,requires in Section 6.8.1 that
implemented and maintained for

implementing the Emergency Plan.
Three Mile Island Nuclear Station Administrative Procedure 1004,
2,"Three Mile Island Emergency Plan," Revision 2,

1978:

II-F-4

Sectiondated February 15,

1.

Requires in Section 2.1,
Unit Superintendent,

that the "Station Superintendent/Senior

Unit Superintendent/Shift Supervisor/Unit

Superintendent-TechnicalSupport in the Control Room will,after

reviewing the emergency conditions, classify the emergency as one of
the following:a.

Personnel or Local Emergency,b.

Site Emergency,c.

General Emergency.

andHe will make this classification according to the condition of Table
1 of this Plan, and initiate actions according to the Emergency Plan
Implementing Procedures,
2.

and according to his own best judgment."

States in Table 1 of Section 2.1 that a Site Emergency exists when
there is a:a.

Loss of primary coolant pressure, coincident with high reactor
building pressure and/or high reactor building sump level(Condition c); orb.

Reactor building high range gamma monitor alert alarm (Conditione).

Contrary to the above, the shift supervisor:a.

Failed to declare a Site Emergency at 0415 hours on March 28,
1979,

at which time Condition "c" of Table 1 of Section 2.1 of

the Three Mile Island Emergency Plan had occurred; andII-F-5b.

Failed to declare a Site Emergency at 0635 on March 28, 1979,at which time Condition "e"Plan had occurred.

G.

10 CFR 20.106,of the Three Mile Island Emergency(See Details II,

Section 2.1)

"Radioactivity in effluents to unrestricted areas," requires

in Section (a) that a licensee not release to an unrestricted area radioactive material in concentrations which exceeds the limits specified inAppendix B, Table II of 10 CFR 20 when averagedover a period not greaterthan one year.

For mixtures of radionuclides,

the limiting value isdetermined as follows:
Ca+

and MPCa,

MPCb,

<1

+

MPC

MPCb

MPCa
Where Ca) C

Cc

Cband C are the concentrations of radionuclides a,b, and c,and MPCc are the maximum permissible concentrations of

nuclides a, b, and c established in Appendix B.
Contrary to the above:
The release of xenon-133,

xenon-133m, xenon-135,

during the period 0700 on March 28, 1979,

xenon-135m and krypton-88

through 2400 April 30,

1979,.

produced an average annual release rate approximately eleven times greaterthan is permitted.
H.

10 CFR 20.401,(See Details II,"Records of surveys,

requires in Section (a)Section 3.1.2.1)
radiation monitoring, and disposal,"that each licensee maintain records showing theradiation exposure for all individuals for whom personnel monitoring isrequired on a Form NRC-5 or equivilant and in Section (b)requires thateach licensee maintain records of the results of surveys required by 10
CFR 20.201(b).

II-F-6

Contrary to Section (a) of the above:
During the period March 1-31,

1979,

records of radiation exposure for atleast 5 individuals for whom personnel monitoring was provided were notmaintained on a Form NRC-5 or equivilant; and,Contrary to Section (b)a.

of the above:

Records made by offsite survey teams during March 28-30,

1979, ofresults of approximately 500 surveys performed in unrestricted areas
bordering Three Mile Island Nuclear Station were not maintained.
These results were transmitted by radio and telephone to the Emergency
Control Station where they were recorded in the Emergency Control
Station Survey Log, which was maintained; however,in this log are incomplete (e.g.,

results recorded

specific instrument used, whether

window open or closed) in nearly every instance.

(See Details II,

Section 3.3)b.

During March 28-30,

1979,

records of surveys of at least ten individuals

who were found to be contaminated with radioactivity were not maintained.
(See Details II,
c.

Section 3.2)

During March 28-30,

1979,

records of radiation surveys necessary toassess dose of at least 5 individuals who handled samples of highly
radioactive primary coolant were not maintained.

(See Details II,

Section 3.2)
I.

Technical Specification 6.12,"High Radiation Area," requires that eacharea in which the intensity of radiation is greater than 1000 mrem/hr beprovided with locked doors to prevent unauthorized entry into the area
and that, any individual entering the area be equipped with a continuously
indicating dose rate monitoring device.

II-F-7

Contrary to the above:a.

From 1100 on March 28 until the afternoon of March 30,

1979,

the

doors to the auxiliary building were not locked and access was nototherwise controlled even though the intensity of radiation in thebuilding was known to be a high radiation area with radiation levels
much greater than 1000 mrem/hr during this period; and,b.

From the evening of March 28 to the evening of March 29, 1979,atleast three entries into the auxiliary building were made by individuals
who were either not equipped with a radiation monitoring device
which continuously indicated the dose rate, or were equipped with an
instrument with insufficient range such that the instrument "pegged"or read full scale in some areas.
J.

(See Details II,

Section 3.2)

10 CFR 20.103, "Exposure of individuals to concentrations of radioactive
materials in air in restricted.areas,"requires in Section (a)(3) thatthe licensee make suitable measurements of the concentration of radioactive
materials in air for detecting and evaluating airborne radioactivity in
restricted areas for the purposes of determining compliance with theregulation in 10 CFR 20.103(a)(1).
Contrary to the above,no measurements were made of the concentrations ofairborne radioactive materials in the Unit 2 auxiliary building forperiods during which individuals were exposed from 1100 on March 28
through midnight March 30,

1979, nor in the Unit 1 nuclear sample roomand primary chemistry laboratory for periods during which individuals
were exposed from 0400 March 28 through 0800 March 30,
II,

Section 3.2)

II-F-8'1979.

(See Details

Contrary to the above,no measurements were made of the concentrations ofairborne radioactive materials in the Unit 2 auxiliary building forperiods during which individuals were exposed from 1100 on March 28
through midnight March 30, 1979,

nor in the Unit 1 nuclear sample roomand primary chemistry laboratory for periods during which individuals
were exposed from 0400 March 28 through 0800 March 30, 1979.
II,
K.

(See Details

Section 3.2)

FSAR Section 12.3.2.10 states that a Radiation Work Permit (RWP)

isrequired for all personnel working in radiation areas except in emergency
situations.

In emergency situations the entry should be made by qualified

personnel carrying radiation monitors and each entry must be documented.
Contrary to this commitment,
emergency situation,

during the period March 28-30,

1979,

an

individuals were permitted to enter high radiation

areas with radiation levels much greater than 1000 mrem/hr and withunknown levels of airborne activity, without an RWP,
were not documented.
L.

and these entries(See Details II, Section 3.2.1)

FSAR Section 12.1.2.11 states that all sample and recirculation lines areshielded by placing them in shielded cubicles or pipe chases which arelocated at sufficient distances from normally occupied area.
Contrary to this commitment,

the primary coolant sample lines are notshielded or located in pipe chases at sufficient distances as they leave
the Unit 2 fuel handling building and instead, pass through the normallyoccupied areas in the Unit 1 fuel handling building and hot machine shop.
(See Details II,
M.

10 CFR 20.201,

Section 3.2.2)"Surveys,"requires in Section (b)that each licensee

shall make or cause to be made such surveys as may be necessary to comply
with the regulations 10 CFR 20.

Also,

in Section (a),II-F-910 CFR 20.201

defines a "survey" to mean an evaluation of the radiation hazards incident
to the production,use,

release,disposal,or presence of radioactive

materials or other sources of radiation under a specific set of conditions,and that when appropriate,such evaluation includes a physical survey ofthe location of materials and equipment and measurements at levels ofradiation or concentrations of radioactive material present.
1.

10 CFR 20.101,"Exposure of individuals to radiation in restricted

areas," limits the dose to the whole body,to the extremities,

andto the skin of the whole body, to values therein.
Contrary to the above, surveys to assure compliance with 10 CFR
20.101 were not made when:a.

On March 29, 1979,

an Auxiliary Operator was permitted to enter

areas of the auxiliary building where an exposure rate of 30
R/hr was measured, without first performing a survey of hisquarterly accrued dose.

The individuals whole body dose wasmeasured on return from the auxiliary building and when added
to his previous dose resulted in a whole body dose in excess ofthe specified limit; and,b.

On March 29,(See Details II,

Section 3.2.4)

1979, a Nuclear Engineer entered an area of theauxiliary building where the radiation level was greater thanthat which could be measured by his portable survey instrument(2R/hr).

Failure to perform a survey of the exposure rate inthis area resulted in the individual receiving a whole bodydose in excess of the specified limit; and,(See Details II,

Section 3.2.4)

II-F-IO

c.

On March 29,

1979,

a chemistry foreman was permitted to repeatedly

enter high radiation areas and handle highly radioactive reactor
coolant samples without performing a survey of his accumulated
dose.

The whole body dose received by the foreman during thissample handling exceeded the specified limit; and,(See Details II, Section 3.2.4)
d.

On March 29,

1979, a chemistry foreman and a radiation protection

foreman were permitted to handle a highly radioactive reactor
coolant sample without first performing a survey of hand andforearm exposure rates.

Handling of this sample resulted in adose to the hand of the chemistry foreman and forearm of theradiation protection foreman in excess of the specified limit;
and,(See Details II,e.

Section 3.2.4)

On March 28 and March 29, several individuals received skincontamination of the head and other parts of the body sufficient
to cause exposure rates, in the range of 20-100 mR/hr whenmeasured with a hand held survey instrument and no survey ofthe dose to the skin of these individuals has been made.
(See Details II,

2.

Section 3.2.4)

10 CFR 20.105, "Permissible levels of radiation in unrestricted
areas," limits radiation levels in unrestricted areas.
Contrary to the above,from 1700-2238 on March 28,

0340-0540 on March 29,

1979,

1979,

and from

surveys to assure compliance with 10

CFR 20.105 were not made in that measurements of the levels ofradiation were not made in areas where the radioactive effluent

II-F-I1

-plume from the plant stack was predicted to have been and which
could reasonably be expected to cause radiation levels in excess ofthe specified limit.
N.

10 CFR 20.101,(See Details II,

Section 3.3)

"Exposure of individuals to radiation in restricted areas,"requires that no licensee possess,use, or transfer licensed material insuch a manner as to cause any individual in a restricted area to receive
in any period of one calendar quarter a dose in excess of three rem tothe whole body, or 18 3/4 rem to the hands and forearms,

or 7 1/2 rem tothe skin of the whole body.
Contrary to the above, during the first calendar quarter of 1979:1.

Two individuals working in restricted areas received doses to thewhole body of 4.175 rem and 3.870 rems as measured by personnel
dosimetry devices; and,2.

One individual working in a restricted area received a dose to thewhole body of 4.115 rem as measured by personnel dosimetry devices,
and a calculated dose to the hands and forearms of from 50 to 147
rem; and3.

One individual working in a restricted area received a calculated
dose to the hands and forearms of from 44 to 54 rem.
II,

0.

(See Details

Section 3.2.4.5)

10 CFR 20.202,"Personnel Monitoring,"

requires that the licensee supply

appropriate personnel monitoring equipment and require its use for eachindividual who enters a restricted area and is likely to receive a dosein excess of 25 percent of the applicable value specified in 10 CFR
20. 101.

II-F-12

Contrary to the above, on March 29,

1979 the licensee did not provide

personnel monitoring devices to measure the dose to:1.

The hands of at least two individuals received doses in excess of
the limit specified in 10 CFR 20.101 during collection and handlingradioactive samples of reactor coolant; nor to

2.

The hands or feet of the members of a repair party team who placed
plastic over the radioactive water on the floor of the auxiliarybuilding.

P.

10 CFR 20.105(b),(See Details II,

Section 3.2.4.5)"Permissible levels of radiation in unrestricted areas,"requires in Section (b)that no licensee shall possess,

licensed material in such a manner as to create,area,

radiation levels which,in the area,

if

use or transfer

in any unrestricted

an individual were continuously present

could result in his receiving a dose in excess of two mil-lirems in any one hour.
Contrary to the above,

licensee dosimetry shows that radioactivity released

from Three Mile Island Nuclear Station caused average radiation levels inunrestricted areas on Kohr and S. Shelley Islands,

which were about 30

mrem/hr and 3.1 mrem/hr for approximately a twenty-nine hour period on
March 29 and 30,

1979,

respectively.

These levels would have resulted in

an individual present receiving a dose in excess of two millirems in an
hour.

(See Details II,

Section 3.3)

II-F-13

0

ENCLOSURE I
This Enclosure contains the memoranda establishing the Office of Inspection
and Enforcement investigation into the Three Mile Island Unit 2 accident ofMarch 28, 1979.

El-1

UNITED STATESNUCLEAR REGULATORY COMMISSIONWASH IN'7TON, D. C. 20555

April 20, 1979

MEMORANDUM FOR:

Chairman Hendrie
Commissioner Gilinsky
Commi ssi oner Kennedy
Commissioner Bradford
Commissioner Ahearne

THRU:

Lee V. Gossick
Executive Director for Operations

FROM:

John G. Davis, Acting Director
Office of Inspection and Enforcement

SUBJECT:INVESTIGATION OF THREE MILE ISLAND ACCIDENT BY THE
OFFICE OF INSPECTION AND ENFORCEMENT

This is to confirm our discussion of April 13, 1979, concerning theThree Mile Island accident investigation underway by the Office ofInspection and Enforcement (IE).
The investigation performed by IE has two basic goals:
1. To establish, in a comprehensive manner, the facts concerning
the Three Mile Island accident. The parameters of this effort
are further described in Attachment A.
2.

To evaluate the performance of the licensee in association
with the Three Mile Island accident as a basis for corrective
action or enforcement action as appropriate.

The Office of Inspection and Enforcement investigation does notinclude, under our current plans, the following:
1. Any evaluation of the actions of the NRC or any of its organizational components during the course of this accident or
recovery period.
2.

Any evaluation of the actions of other agencies during thecourse of the accident, or during the recovery period of
the accident.

The Commission-

3.

-2-

April 20, 1979

Any review and evaluation of the NRC regulatory process as
it relates to the Three Mile Island accident for "lessons
learned." IE is not collecting information concerning nor
evaluating:oLegislative authority of the NRCoRules and regulations of the NRC

oSafety researchoLicensing processoInspection and enforcement process

The Office of Inspection and Enforcement will expand its investigation
In the absence of suchas directed by the Commission or the EDO.
direction, IE is proceeding with its investigation as outlined inthis memorandum.

oln G. DavisActing Director
Office of Inspection
and Enforcement
Enclosure:
IE/TMI Conceptual Outline

OFFICE OF INSPECTION AND ENFORCEMENT
THREE MILE ISLAND ACCIDENT INVESTIGATION
Conceptual Outline

1. The IE investigation of the Three Mile Island accident isdirected toward:

2.

3.

a.

Within the time period of the investigation, establishing
the facts concerning the inmnediate causes of the accident
and the actions of the plant and the licensee staff during
the course of the accident.

b.

Within the time period of the investigation, establishing
the facts concerning the actions of the licensee, the NRC,
other Federal agencies, and appropriate state agencies.

c.

Evaluating the performance of the licensee during and in
response to the Three Mile Island accident as a basis forcorrective action or enforcement action as appropriate.

The investigation consists of two parts conducted in parallel:a.

Operational - The inplant, reactor operations situation.
This will cover the time period from the closing of theauxiliary Feedwater System valves (or other earlier
immediate cause of the accident) to the restart of reactor
coolant pump 1A (about 8:00 p.m., March 28, 1979).

b.

Radiological - The inplant and environmental radiologicalconditions. This will cover the time period from the
beginning of the accident until about midnight on
March 31, 1979.

The Office of Inspection and Enforcement will develop factual
information concerning the condition of the plant and theenvironment and the performance of the licensee for the time
periods subsequent to those described in Item 2, above.
However, this information will not be included in the IE
Three Mile Island Accident Investigation.
(Note that thisdoes not include development of information concerning NRC
activities. The information will include the licensee's /reaction and response to NRC activities but-will not develop,
for example, how an NRC position or recommendation wasformulated.)

4.

The investigation will include:a.

Sequence of Events(1) Operational
Identify the sequence of events in relation to facility
operation. Determine licensee actions or lack of actions
related to operation of the facility within licensed
parameters, constraints and limits. Determine theparticipation of licensee supervision, management andengineering support. Determine causes of the event.
(2) Radiological
Identify the sequence of events in relation to licensee
activities in radiation control on site and off site.
Determine licensee actions or lack of actions related
to controlling and monitoring on site exposures,protection of workers, and control and monitoring ofoff site releases.
b.

Immediate Cause of Accident(1) Equipment
Trace the performance and maintenance history of importantequipment which malfunctioned at the beginning or during
the early phases of the incident. Define the signals orother intelligence provided to operators concerning
serviceability and availability of.e.quipment. Identify
the serviceability of equipment required Tor operation.
(2) Procedures
Determine the requirements contained in surveillance andmaintenance procedures and the appropriateness of procedures.
Determine approIdentify failures to follow procedures.
priateness of communications between plant groups on plantand equipment status. Determine whether emergency operatingprocedures were appropriate and were followed.
(3) Staff Performance
Describe the performance of the operators and other licensee
personnel during the accident. Review training of operators
particularly training concerning response to off-normal
limits.

-3-c.

Licensee Management of Accident
Describe the response of the licensee from the standpointof management and supervision of the accident.
Identify
the engineering and radiological support requested andreceived and its source.
Determine the licensee's reaction
to and analysis of the accident as it unfolded withparticular emphasis on engineering analysis of alternativemodes of accident recovery.

d.

Emergency Plan Activation
Examine the licensee's emergency plan implementation to
include preplanning and tests. Develop a detailed
chronology of the implementation of the emergency planwith particular emphasis on timeliness of notification.

5.

This investigation is being managed by the NRCoRegion I
Office.

-UNITED STATES.

'9NUCLEAR REGULATORY COMMISSION

ib~fl~~

0

WASHINGTON, D. C. 20555~.

~i~JIItItP~

5 i979MEMORANDUM FOR:

Chairman Hendrie
Commissioner Gil'insky
Commissioner Kennedy
Commissioner Bradford
Commissioner Ahearne

THRU:

Lee V. Gossick
Executive Director for Operations

FROM:

John G. Davis, Acting Director
Office of Inspection and Enforcement

SUBJECT:INVESTIGATION OF THREE MILE ISLAND ACCIDENTBY THE OFFICE OF INSPECTION AND ENFORCEMENT

Reference is made to my memorandum of above subject dated April 20, 1979.
The investigation by the Office of Inspection and Enforcement of theaccident and of the licensee's actions during the course of the accident
has been underway for approximately seven weeks.
During the conduct of
this investigation, IE has interviewed approximately llO people and has
examrined other information sources in its efforts to determine what
transpired, the proximate causes, and licensee actions before and during
the incident proper.
As a result of a current review of the progress of the investigation,
believe the following matters should be specifically called to yourattention:i.

The investigation is complex, involving extensive interplayof mechanical and control systems with human actions. This
complexity has been further compounded by the need to release
information prior to the completion of the investigation, byother concurrent investigations and by the need to reply to
specific questions prior to completion of the IE investigation.
This *m.v lead to IE being unable to meet its projected August 1,
1979 date for the investigation report. We are currently
examining our schedules to account for anticipated additional
investigatory work. We will inform you by June 25, 1979, of
any necessary rescheduling of the projected date for theinvestigation report.

I. .3-cc:

E. Kevin Cornell, EDO
Harold R. Denton, NRR
Saul Levine, RES,
Robert B. Minoguel'/SD.
William 'J.Dircks, NMSSJames-J.' Cummings, OIA
Robert G.- Ryan, SP
James R. Shea,; IPHarold D. Thornburg, IE
Norman C. Moseley,.IE
E. Morris Howard, IE
James H. Sniezek, IE
Boyce H. Grier, IE
James P. O'Reilly, IEJames G. Keppler, IE
Karl V. Seyfrit, IE
Robert H. Engelken, IE
Dudley Thompson, IE
Leonard I. Cobb, IE
James M. A;lan, IE
DISTRIBUTION:
L. V. Gossick.
T. A. Rehm
-J. G. Davis'EDO RF .
IE RF
SECY (3)
MPA

I

AD

(5069-)

NRC

EDO

6/ ..
./79~

............
/..
... ...............................

0240*ov u2.0v5~"~........

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ELD

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Too~K

ENCLOSURE 2
This Enclosure contains a listing of the interviews conducted during thecourse of this investigation.

E2-1

ENCLOSURE 2
iNTERVIEW LISTINTERVIEWNO.

POSITION TITLE

1

Station Manager & Unit 2 Superintendent

2

Shift Foreman C

3

Control Room Operator D

4

Control Room Operator C

5

Shift Supervisor E

6

Radiation/Chemistry Technician I

7

Radiation Protection Foreman A

8

Engineer H

9Control Room Operator C

10

Auxiliary Operator B

11

Radiation Protection Foreman A

12

Shift Foreman B

13

Shift Supervisor C

14

Shift Foreman C

15

Control Room Operator D

16

B&W Test Coordinator

17

Shift Supervisor E

18

Radiation Protection Foreman D

19

Radiation Protection Consultant A

20

Supervisor, Radiation Protection &ChemistryE2-2

INTERVIEWNO.

POSITION TITLE

21

Radiation Protection Foreman B

22

Radiation Protection Supervisor

23

Chemistry Foreman B

24

Radiation/Chemistry Technician K

25

Supervisor, Radiation Protection &Chemistry

26

Unit 1 Supervisor, Station Operations

27

Unit 2 Superintendent,
Technical Support

28

Radiation Protection Consultant A

29

Radwaste Foreman

30

Corporate Technical Analyst

31

US NRC Inspector D

32

Site Protection Sergeant B

33

Engineer E

34

Unit 2 Supervisor,

35

Engineer J

36

Auxiliary Operator G

37

Shift Foreman C

38

Auxiliary Operator H & Shift
Supervisor E

39

Radiation Protection Foreman B

40

US NRC Inspector H

41

Met Ed Group DiscussionEmergency PlanningE2-3

Station Operations

INTERVIEWNO.

POSITION TITLE

42

Radiation/Chemistry Technician I

43

Radiation Protection Foreman C

44

US NRC Inspector K

45

Radiation Protection Technician S

46

Pennsylvania Bureau of Radiation
Protection

47

Radiation/Chemistry Technician N

48

Engineers C and D

49

Radiation/Chemistry Technicians H and W

50

Radiation/Chemistry Technician E

51

Radiation/Chemistry Technicians F, J

52

Site Protection Sergeant C

53

B&W Site Manager B

54

Supervisor Contract Radiation
Protection

55

Region 1, US NRC,
Control Center

56

Preliminary Inquiry Interviews

57

Control Room Operator J

58

Radiation/Chemistry Technician L

59

US NRC Inspector H

60

Meterological Consultants A and B

61

Control Room Operators C and
Auxiliary Operator D

62

US NRC Inspector A

63

US NRC Inspector B

64

US NRC Inspector E

E2-4

Operations

INTERVIEWNO.

POSITION TITLE

65

US NRC Inspector G

66

Engineer F

67

B&W Operations Engineer

68

Radiation/Chemistry Technician M

69

Radiation Protection Consultant--B

70

Engineers A and I

71

Station Manager

72,

Supervisor, Generation Security &Site Protection Sergeant A

73

US NRC Inspector F

74

Corporate Environmental

75

Radiation/Chemistry Technicians R and T

76

Corporate Supervisor C

77

Unit 1 Superintendent

78

Radiation/Chemistry Technicians 0 and U

79

Supervisor, Radiation Protection &Chemistry

80

Group Interview of Training Staff

81

Control Room Operator I

82

Radiation/Chemistry Technician C

83

Unit 2 Superintendent

84

B&W Principle Engineer B

85

B&W Service Engineer

86

B&W Principle Engineer A

87

B&W Manager D

88

B&W Manager A

E2-5

Scientist, B

INTERVIEWNO.

POSITION TITLE

89

Radiation/Chemistry Technicians D and I

90

Vice President of Generation

91

Vice President Generation

92

B&W Senior Engineer

93

Radiation Protection Consultant C

94

Superintendent Administration andTechnical Support

95

Shift Supervisor A

96

Radiation/Chemistry Technician P

97Engineer H

98

Radiation/Chemistry Technician V

99

Radiation/Chemistry Technician A

100

Consultant A

101

US NRC Inspector M

102

Engineer K

103

Daupin County Civil Defense
Director and Supervisor, Radiation
Protection and Chemistry

104

Auxiliary Operator C

105

Shift Supervisor D

106

Radiation/Chemistry Technician B

107

Shift Foreman A

108

Environmental Scientist A

109

Auxiliary Operator E

110

Unit 1 Superintendent, TechnicalSupportill

Control Room Operator J

E2-6

INTERVIEWNO.

POSITION TITLE

112

Maintenance Foreman C

113

Unit Supervisor Station Operations

114

Auxiliary Operator I

115

Auxiliary Operator J

116

Control Room .Operator F

117

Control Room Operator A

118

Control Room Operator H

119

Shift Supervisor B

120

Maintenance Superintendent

121

Engineer B

122

Staff Chemist A

123

Auxiliary Operator G

124

Radiation Protection Supervisor

125

Auxiliary Operator H

126

Maintenance Foreman D

127

Engineer C

128

Auxiliary Operator D

129

Unit 2 Superintendent TechnicalSupport130

Chemistry Foreman A

131

Control Room Operator B

132

Chemistry Foreman C

133

Supervisor, Radiation Protection &Chemistry

134

Control Room Operator E

135

Unit 2 Superintendent TechnicalSupportE2-7

INTERVIEWNO.

POSITION TITLE

136

Auxiliary Operator F

137

US NRC Inspector N

138

Maintenance Foreman E

139

Control Room Operator G

140

Shift Supervisor C

141

Chemistry Foreman B

142

Radiation Chemistry Technician Q

143

Maintenance Foreman G

144

Radiation Protection Consultant A

145

Group Interview-Control

146

US NRC Inspectors I&J

147

Shift Suprvisor E

148

Unit 2 Supervisor, Station Operations

149

B&W Test Coordinator

150

Shift Supervisor D

151

Control Room Operator D

152

Shift Supervisor A

153

Control Room Operator C

154

Radiation Chemistry Technicians Q and V

155

Radiation Protection Consultant D

156

Radiation Chemistry Technicians B and G

157

Unit 2 Supervisor,

158

Corporate Supervisor A

159

Manager Generation Engineering

160

Corporate Supervisor B

E2-8

Room Staff

Station Operations

INTERVIEWNO.

POSITION TITLE

161

Corporate Environmental Scientist A

162

GPU Manager G

163

GPU Manager A

164

GPU Director Technical Functions

165

GPU Manager C

166

GPU Senior Engineer B

167

GPU Senior Engineer A

168

GPU Manager B

169

GPU Manager E

170

GPU Engineer B

171

GPU Manager D

172

Radiation Protection Supervisor

173

Shift Supervisor C and Control
Room Operator A

j174

Engineers C and D

175

Radiation Protection Foreman D

176

GPU Manager A

177

GPU Manager F

178

GPU Engineer A

179

GPU Mechanical Engineer

180

Unit 1 Superintendent Technical Support

181

Instrument Man A

182

US NRC Inspector C

183

Maintenance Foreman A

184

Control Room Operator J

185

B&W Manager C
E2-9

INTERVIEWNO.

POSITION TITLE

186

Instrument Man B

187

Maintenance Foreman B

188

Auxiliary Operator A

189

Group Interview Control Room Staff

190

Shift Foreman B

191

B&W Test Coordinator

192

Engineer B

193

Unit 2 Maintenance Supervisor

194

Repairman A
Repairman B

195

Engineer I

196

Shift Supervisor A

197

Unit 2 Superintendent
Technical Support

198

Maintenance Foreman

199

Auxiliary Operator

200

Shift Foreman B

201

Control Room Operator K

202

Control Room Operator L

203

UC and M Man A
UC and M Man B

E2-10

NRC FORM 335(7-77)

1. REPORT NUMBER (Assignedbv DDC)

U.S. NUCLEAR REGULATORY COMMISSION

BIBLIOGRAPHIC DATA SHEETNUREG-0600

4. TITLE AND SUBTITLE (Add Volume No., if appropriate)

INVESTIGATION INTO THE MARCH 28,

2. (Leave blank)

1979 THREE MILE ISLANDACCIDENT BY OFFICE OF INSPECTION AND ENFORCEMENT
Investigative Report No.

3. RECIPIENT'S ACCESSION NO.

50-320/79-10

7. AUTHOR(S)

5. DATE REPORT COMPLETED
YEAR

MONTH

July
9. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (include Zip Code)1979- DATE-REPORT ISSUED
YEAR

MONTH

Office-of Inspection and Enforcement

July

U.S. Nuclear Regulatory Commission
Washington,

1979

6. (Leave blankl

D. C. 20555
8. (Leave blank)

12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (Include Zip Code)

10. PROJECT'TASK/WORK UNIT NO.

11. CONTRACT NO.

13. TYPE OF REPORT

15. SUPPLEMENTARYPERIOD COVERED (Inclusive dates)

NOTES

14. (Leave blank)

16. ABSTRACT (200 words or less)

On March 28, 1979, the Three Mile Island Unit 2 Nuclear Power Plant
experienced the most severe accident in U.S. commercial nuclear power
plant operating history.
This report sets forth the facts concerning
the events of the accident determined as a. result of an investigationby the NRC Office of Inspection and Enforcement.
The IE investigation,
which is based on the information available at this time, is limited
to two aspects of the accident:
1.

Those related operational actions by the licensee during the period
from before the initiating event until approximately 8:00 p.m.,
March 28, when primary coolant flow was re-established by startinga reactor coolant pump, and2.

Thosesteps taken by the licensee to control the release of radioactive material to the off-site environs, and to implement his
emergency plan during the period from the initiation of the eventto midniqht, March 30.

17. KEY WORDS AND DOCUMENT ANALYSIS

17a. DESCRIPTORS

17b. IDENTIFIERS/OPEN-ENDED TERMS

18. AVAILABILITY

STATEMENT

19. SECURITY CLASS (This report)

21. NO. OF PAGES

Unclassified
Unlimited

20. SECURITY CLASS (Thispage)

22. PRICESNRC FORM 335 (7-77)UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON. 0. C. 20555
OFFICIAL BUSINESSPENALTY FOR PRIVATE USE, $300

POSTAGE AND FlEES PAID
U.S. NUCLEAR REGULATORY
CO M MISSION