ML20024A992

Interim Rept of TMI-2 Occurrence,Technical Review Committee

ML20024A992
Person / Time
Site:	Crane
Issue date:	06/08/1979
From:	BABCOCK & WILCOX CO.
To:
References
TASK-01, TASK-02, TASK-06, TASK-07, TASK-1, TASK-2, TASK-6, TASK-7, TASK-GB GPU-0005, GPU-5, NUDOCS 8307010270
Download: ML20024A992 (61)

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i INTERIM REPORT

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.. JUNE 8,1979 i

3 Pif. Exh..indiv Charles Shapiro CSR Q.) 9(80 Do'le Reporting Inc.

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TECHNICAL REVIEW CO MITTEE CHARTER i

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L' Ef fective insnediately, a Technical Review Cossaittee is established l

i with the following. charters (1) Review technical aspects of the THI-2 occurrence b.c.

(commerciai business or legal issues will be handled I

separately).

g (2) Develop recommendations sad general scope of engineering t.

programs to improve plant safety end reliability with L

3 emphasis on the NSS and interact

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(3) Assess impact of the THI-2 occurrence and resulting 1

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. changes in regulatory positions on NFGD technical programs, procedures, standards, etc. Consider I

relationships.among engineering, service, training.

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licensing, human engineering and R&D programs.

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TECHNICAL REVIEW COMMITIZE MEMBERS D.'

Cannon /J. E.' Burgo J. 3. Taylor, N. S. Elliott C. F. Welch

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1. N. Kubik E. A. Vomack J. R. Hamilton R.

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til Pindings and Recoacaendatione Section 1 Sequence of Events-Section 2 g

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f Programs for Evaluation of Equipment Section 3 Condition g

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Program for System Behavior Analysis Section 4 i,,

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. Prograse for Alternate Systema and Section 5 r

i Controls Configurations E

g Potential Changes in NRC Philosophy.

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Potential Changes in Organization' Section 7

,L and Interface

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Annotated'Se'quence of. Event's.. Plant Appendix 1 j'

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Computer Alarm Printout Descrip-i.

tion, Postulated HP1 Sequence o.f Operation,and Postulated j

4 Sequence o'f Op,eration of.

l Pressurizer Dto Cate Valve 8

Description of Rasetinater.

APPendia 2 i

Purther Discussion' of Prograna/

'Appendia 3' g;..

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Compilation of Technical Suggestions P

Made by. 3&R Personnel Since the F

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's FINDINGS AND RECOMMENDATIONS i

e Recomroendaticn Finding This TRC report lists several areas requiring further I

1.

The sequence of events as y,ublished by BBR,-CPU, HRC 1s substantially in agreement with our version on

. analysis. Wg'should continue to support the EPRI l'

basic facts and probable causa of each event. EPRT effort to its completion. If there are questionable events or interpretations following the EPRI study, is preparing a comprehensive sequence which probably B&W should further analyze to our own sittsfaction.

will be accepted as the " official" version.

This will probably have to be done at our own expense.

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.9 Extend studies of off-nurnal sequences to identify 2.

Events of the THI-2 sequence challenged both equ{p-improvements 1n components (valves, etc.) and system

]. Jj ment and operators.. Equipment malfunction and configuration which will reduce challenges and improve operator response led to deteriorating plant

' conditions that might have been avo'Jded.

system reliability. Hechanistic safety methodology I

(German) (KTA 3501) is a recommended approach.

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(See Section 4) g.

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,... ? T-9 Determine alterations to B&W system that would alleviate

'af d formation in the core, collapse of the pressuriser or eliminate the concerns raised by the above factrra.

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steam bubble, the collection of stdam and noncondensible Items to be considered include:

gasesinthereactorvesseldoneahdupperpartsofthe J-less were found to be real factors requiring attention 1.

Elimination of the FORV in the course of the THI-2 incident.

2.

1.arger pressuriser - relocated surge line 3

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Heans of venting reactor head and top of J-legs 4.

Ileans of increasing thermal capacity of secondary i

system heat sink inventory Y

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FINDINGS AND REC 0!tfEIIDAT10HS_

, Recoammenda ton Finding -

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Operators were " confused" by control room'information.

The more important suggestions the cosalttec can make here ares They had no " mental image" of the HSS system charac-teristics..Their incorrec,t analysis of system condi-1.

Provide additional primary information to reduce tions led to poor action decisions which caly operator need for implied analytical decisions.

aggravated the situation.

'(a) Primary t vs. t les sat hot

. (b) Positive flow indications as opposed to 1,

g valve position indications.

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2.. Provide enhanced system know-how to operators.

3.

Increase operator training requirements.

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(a) Small breaks

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-(b) Equipment malfunctions W

(c) Combinations of equipment malfunctions and operator error (d) Compulsory retraining and requalification.

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Improved control room design through hurr,an i

engineering studies to better couple the operator with the information available and E

important control stations.

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F1ND1NCS AND RECOH!!ENDATIONS, l

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Recommendation Finding.

Studies of the ef fects of the various environmental 5.

Munerous components in the TM1-2 system have been parameters on the current condition and future life subjected to unexpected conditions of high radiation, of these components should be begun as soon as pos-3 4

. high tasaperature, temperature transients, building alble. This report specifies in a general way those

.l spray including Na0N and possibly boric acid..

programa ye believe to be most important. Some can be funded;by S&W but others are sore properly funded by EPRI, p0E, and/or CPU. Some of the work can be done within S&W but other work is best done by other

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R&W NSS systems have been' combined with too many We suggest the consideration of the following recom-I:

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mendations:

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balancekof-plant (30P) designs for which we have B&W should standardize those parts of the 10P 1.

.little et no responsibility.

which are crucial to the operation of the 205

g planta.

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On -sisting and backlog plants wherein no wholesale changes to the 80F are possible.

B&W should insist on a design review of the SOP to assure ourselves that satisfactory system operation can be obtained.

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B&W should establish agreements with customers

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allowing approval rights over future changes l

in the 30P.

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.__.___m FINDINGS ILEC&cfENDATIONS si Recommendation Finding I

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Evaluation of plant transisat experience and improved The Power. Systems and Controls unit har b u n g.

established as a focal paint for evaluation understanding resulting from advances in the engineer-of plant operatinnat experience. Transients ing state of art can contribute to accident prevention -

which occur frequently, and severe transients, and mitigation if they are effectively and promptly should be evaluated assinst engineering design coansunicated to the operstors' of licensed reactor and operational bases for the operating plants I

to determine whether sy' tem, control, and oper-i plants._

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'stor performance during these transients is in

, accord with design and to provide early warning i

of potential problem areas. $1milarly, findings resulting*from improved knowledge of plant

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behavior during two engineering state of the art advances,must be promptly funneled to the

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operating plants to assure that control and f,,)

operation reflects the best possible understand-r.

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ing of both normal and off-normal conditions.

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closed loop system for accomplishing this should

.j be established as discussed in Section 4.

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[. i Index to Section 2 4, f h

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e ~~ 1 2.0 Introduction

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2.1 Events Requiring Confirmation or Further Investigation

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h 2.2 Chronological Sequence of Events t'

2.3.1 Reactor Cociant System Isometric 2.3.2 P t.I D

~0-50 minutes 2.3.3A Reactimeter and Str3.p Chart Data 0-100 rinutes 2.3.33 Reactimeter and'St' rip Chart Data 2.3.3C Reactimeter and Strip Chart Data

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2.3.3D, Reactimeter and Strip Chart' Data- ' 24)--300 minutes 2.3.35 Reactimeter an*d Strip Chart Data

- 300-400 minutes 3

.2.3.3F Reactimeter'and Strip Chart Dats

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2.3.3G Redctimeter and Strip Chart Data

- 500-600 minutes,

2.3.3H Reactimeter and Strip Chart Data.

- 600-700 minutes j+

l 2.3 3I Reactimeter and Strip Chart Data

- 700-1}00 minutes

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2.3.3J. Reactimeter and Strip Chart 9.ch 800-900 minutes.

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3 2.3.3K Reactincter and Strip Chart Dat.a

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Appendix 1 - Annotated Sequence of Events' (Computer Data) 1W Postulated HP: Sequence of Operation during TMI-2 4-M LOF'.f Incident.

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. Postulated Sequence)of 1pening/ Closing Pressurifer

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E!!O Gate Valve during T AI-2 LOFW Incident i.

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Appendix 2 - Transient Sequence of Events - Reactimeter and Strip fi Chart Data I

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SEQUENCE OF EVENTS e

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2.0 INTRODUCTION

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This report has been prepared to describe the operation of H

che Rabcock and Wilcox NSS and associate auxiliaries during D (i the first 1000 minutes of the March 28, 1979 loss-of-feedwater iJ E incident at TMI-2.

e The report is not intended to answer all of the questions n

related to the incident or t'o provide detailed comments on y

i-plant design, equipment performance or plant operator i

actions.

It is intenJcd to document the sequence of events as accura'tely and completely as possible using reactimeter,

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evaluation.

The report will be updated in its final iss se with any new

's data or analysis.

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k 2.1 EVENTS REQUIRING CONFIRMATION OR FURTHER INVESTIGATION fo r the TMI-2 ti During the preparation of the Sequence of Events h

incident, it. became evident f rom available data that the more 6((

significant events are readily explainable, however the data does indicate that further analysis could lead to a better understanding of all phases of the incident and possible design enhancements.

Identified below is a partial listing of items 7'

C needing confirmation or further investigation.

3 Water volumesdded to RC System by MU/HPI pumps.

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a Reactor building pressure and temperature just prior to M

postulated hydrogen burn; i.e., pressure went up, tempera-i Q

2.

r ture went_down_.,,,_

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Actual. cause of loss of feedwater.

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• 4 Plant computer performance.

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Letdown flow and cakeup tank level.

3 RC pressure, steam Transients indicated on source range BF 6 '.

generator pressure and }cyc1 instrumentation as a result of the postulated hydrogen, burn in the reactor building, IA f

and IB h,igh temperature, etc.

Restructure data (time and ~ amplitude) t'o' assurc that scaling k

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has not caused subtfc changes in events to be overlooked.

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Changes.in RC temperature and. pressure from 630 minutes to 3

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1000 minutes.

z Potential errors *in temperature compensated delta P measure-3-

• 9.- *ments as a result of out-of-range tenperature measurements.

Availability of safeguards data (HPI - LPI - BS, etc.).-

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Is MU-V12 isolated by ES?

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• 12. ' Functions which should be recorded in the futufe.

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Reactimeter data reduction program.

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PORY block valve open/close;' frequency.

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Were atmospheiic dump valves operried in auto or manualT T

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h "Further investigst. ions could Icad to design enhancements.

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What caused operator to believe stea::: generator "B" was leaking?

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Did operator take manual control of turbine by-pass and

!.l h steam generator "B" at 75 to SS minutes?

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Design quench capacity of RC drain tank.

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When did fuel damage occur? How long?

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', y) 2.2 CHRONOLOGICAL SEQUENCE OF EVENTS

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The following sequence of events is presented to explain the '

plant's response and conditions reached as a result of the Loss-of-Feedwater transient (LOFW) experienced by TM1-2 on M

March 23, 1979.

For this chronology, reference time was 12 identified as the time of thc turbine trip (04:00:36), this g

time is defined as time equal to zero.

h Plant status immediately before transient:

Three Mile Island Unit Two was at 97% power with the Integrated

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Control System in full automatic.

Rod groups one through five

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were fully withdrawn, rod groups six and seven were 95% with-i drawn and rod group eight was 271 withdrawn.

Reactor Coolant

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M System total flow was approximately 107.5% of design flow and i

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the Reactor Coolant System pressure was 2155 psig.

Reactor Coolant Makeup Pump B (MU-PIB) was in service supplying makeup k

g Coolant System letdown flow was approx *..nate 'y 70 gpm.

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and reactor coolant pump seal injectior flow.

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Reactor Coolant System boron concentrat, ion.as approximately E

1030 parts per million.

The Pressurizer Spray Valve (RC-V1)

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M and the pressuri:er heaters were in manual control while-y

t-tJ spraying the pressuri:er to equali:e boron concentrations,

between the pressuri:fer and the ' remainder of the Reactor Coolant j;-

g System.

The pressuri:er safety valves discharge header 'thermo-3 couples were indicating 210 F to 230 F due to leakage through t

.one of.the.Pressuri:er Safety Valves (RC-RIA and KC-RIB).

y Steam Cencrator l arameters were as sown 1,n the following table: { l i Table of Stes::r Generator Parameters l . Steam Cencrator A Steam Gener*ator'B Loop Feedwater 5.7459 MPPH" 2.7003,MPPH* '561 57.4% s Operating Level g** 'Startup Level 158.8 inches 165.4 inches. Steam Pressure 910 psig 839.6 psi's Feedwater Temperature 462.7 F 462.7F .) Steam Temperature 595 F 594 F

• MPPI'. - Million Pounds per Hour L

g j,- 5 Steam Generator Feedwater Pumps (FW-PIA and FW-PIB), Condensate b Booster Pumps (CO-P2A, CO-P2B and CO-P2C), and Condens~te Pumps @C. a (CO-PLA and CO-P1B) were in service. I i A loss of feedw'ater flow transient started when Condensate Pump E. (CO-PLA) tripped resulting in both Main Feedwater Pumps (FW-PIA ~ and'FW-PIB) tripping. The main feedwater pumps trip caused the .b I 2-5 E l, } y - -,.., - - e e o

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r4 e la /. ti. main turbine to trip. These actions took place in less than f yI i .t-- two seconds. The reactor, in the process of running back, h tripped on high RC pressure (approximately 2355 psig) about 8-10 seconds after the turbine tripped. 7 [t< w f As was determined later, a small break loss-of-coolant accident I was initiated when the Electromatic Relief Valve (RC-RV2) did y not shut properly after opening to reduce reactor coolant system g [y I The high reactor coolant system pressure was a direct pressure. ](( result of the loss of feedwater flow transient which culminated in a reactor trip due,to high reactor coolant system pressure. As the RCS pressure' spiked high, the ERV opened at the setpoint of 2255 psig, the pressure continued increasing at'a reduced el rate until the reactor tripped when RCS high pressure setpoint U rl was. reached (2355 psig). Now the RCS pressure and pressuri:er k 7 level began a rapid decrease toward 1600 psig and 155 inches. As' the pressure decreased. the ERY should have closed at about i 2205 psig. The operator started a second makeup pump (A) and isolated letdown to minimi:e the loss of pressurizer inventory. [ With two makeup pumps on and RC pressure sti'll decreasing, the ) pressuri:er level b gan increasing twoard the upper level limit, t therefore at about five minutes, the operator stopped the (C) 9 makeup pump. This delayed the level from going off scale high L .A until about six minutes into the transient. The. pressurizer E level remained off scale high until about 10 minutes. After ? coming on scale, the pressuri:er level rem > ned above 350 inches h;" for the next two hours. k During this same period, the emergency feedwater pumps started Mg but did not feed the steam generators until the operator opened ei 1 the building isolation valves at eight minutes into the transient.

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The steam generators were dr'y and OTSG pressure had.dr.opped to approximately 775 psig. When feed was again established, the Qc steam generator quickly repressuri:cd to approximately 1000 psig [ in 'about two minutes. At this point, cooling of the RCS was t reestablished as evidenced by a decrease in the hot and cold leg g. temperatures. 10 The pressuriter ERV and code safety valves discharge is piped to { the RCS drain tank located in the basement of the reactor h-building.. The RCS drain tank is designed to quench the steam relieved from either relief valve. The drain tank pressure and M. temperature were normal before the turbine trip. As soon as g - the ERV opened, the drain tank pressure began a slow rise for about three minutes; during the next minute there was a sharp gy rise in pressure from approximately 30 to 115 psig. The pres-gg sure then remained constant for about three minutes before g rising quickly to approximately 150 psig. The pressure then r actually decreased until about 12 minutes, at which time the j*k At this pressure becan a rapid rise to approximately 191 psig. point, the tank goes solid and the rupture disks blow. 2-6 F c~ . g..; -

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~.. 3 ' t ( '. w ! i: The RC drain tank pressure indication provides information that i ec' P 0* ? the ERY is open. The tank is' designed to quench a normal I transient relief without going solid or rupturing the rupture f disk. The tank is not designed for a 15 minute relief or steam P 4 or steam / water mixture to it. As a result of the rupture disk $a blowing,wthe reactor building pressure and temperature began an 4 immediate rise. g 0 3 Engineered safeguards actuation of liigh Pressure Injection-(HPI) bU occurred at approximately two minutes as RCS pressure reached 1 1640 psig. Reactor Coolant Makeup Pump B (MU-PIB) tripped h automatically as a result of the engineered safeguards actuation d,,, 9f HPI. The ES design is such that makeup pumps A ana'C are 0 used for llPI and if running, makeup pump B is tripped prio'r to E Qg actuation of makeup pumps A and C. { During the period of 20 to 74 minutes, the RCS was stabilited f h near a saturation. temperature-pressure reistionship at 542 F g and 1050 psig. The RCS loop flow rates had decreased from about 69 million pounds per hour to approximately 47 MPPli and con-li' tinued to decrease. The containment building temperature and f. } pressure had increased from 0 psig and 120 F to 1.5 psig and r. 170 F, as a. result of releasing the contents of the RC drain is tank to the reactor containment building atmosphere. The 0; 3g Electromatic Relief Valve (RC-RV2) was open. [- w. Steam generator Icycis had been restored to low 1cyci setpoint t } 30 inches, but steam generator "B" was having le, vel control 9 problems. The icvels were maintained by. manual aperation. 1 7 Steam gencedtor p,ressure was being held essentially constant Q and below.setpoint by the turbine byp.sss system. [e u b The soyrce range nucicar instrumentation began an increasing trend during the entire period. Since BKST water was being. 5 I added to the core, this increasing trend would reficct less i dense water in the duwnc.omer/ core region; therefore, inc'reas ed I neutron Icakage to the detectors.,instead of neutron multiplica-gion.in'the core. At.73 minutes, the RC flow began to oscillate, therefore the RC [ s pumps in loop ".B" were stopped to preclude possibic damage to 3e .the RC pumps from operation near saturation temperature-pressure i conditions. Once the "B" loop RC pumps were stopped, the k The "B" e . primary to secondary heat transfer rate decreased. 2 steam generator pressure then decreased due to the lack of ng primary to secondary heat. transfer?and being filled with feed- [! I water at the same time. g1 e Due to this heat transfer rate differwnce i etween the two steam s l generators, a large pressure differential existed between them. i l This could have been interpreted as a primary. co secondary leak } i which caused the operators to isolate the "B" steam generator. z .i s. t 5r, 4 2-7 h 3 B 'G t - - y .~ # ..~:._.... . q.. b T V. -.

.m %.,9 [ { iM1 ~ Tdi 1, '" 401 e The source range nuclear instrumentation indication segan a rapid increase and then decrease to its former level during y the trip and coastdown of the "B" loop RC pumps. This is A-interpreted as a change in neutron leakage instead of neutron }f (i flux increase in the core. As the pumps coasted down, the fluid in the downcomer/ core region experienced a marked decrease in density as the mass flowrate decreases followed by a quenching E 8'kp g3 effect. During the period of 74 to 94 minutes, the RCS/ core began a slow h heatup due to decreased heat transfer caused by loss of two RC h l-pumps. The RCS mass flow continues to decrease at about the [- same rate that started early in the transient (approximately 15 minutes). The Source Range (SR) and Intermediate Range (IR) @7 Nuclear Instrumentations continue to increase during this period. u Again, this is interpreted as mass decrease in the downcomer/ E core region allowing increased leakage to the detectors. At about 94 minutes, steam generator "B" was fed for about three minutes and "A" steam generator was steamed dry. 'The RCS began } l a cooling trend that continued until about.100 minutes. The SR and IR detectors rate of increase slowed as well as the signal ,t seemed to exhibit less noise. At 100 minutes, the RC pumps in Loop "A" were stopped to preclude possible damage to the RC pumps from operation near saturation [ temperature pressure conditions. Once these pumps were tripped, n( the RCS/ core had no forced circulation cooling. The operator d4 had taken control of steam generator 1evel and was raising the .1 "A" steam generator level to 50% on the operate range. This level is setpoint to induce natural circulation cooling in a hf subcooled systen. This Icyc1 was obtained at approximately 4 125 minutes.. This feeding of the steam generators continued i to cool the RCS until suf ficient blockage of the hot' legs M j. occurred with superheated steam. At about 125 minutes, essen-GE tially all cooling by the steam generators was lost as evidenced R hy the hot leg and cold Icg temperatures diverging. The steam 3i .,g# generators continued to provide little cool.ing through 150 minutes. As is noticed at about 112 and 123 min.utes, the. hot Icg temperatures began an increase that took then off scale high l r.. }p ( 620 F) at 131 and 150 minutes, respectively. The SR and IR again exhibited the same behavior as they had done 4 .at 73 minutes when the "B" loop pumps were stapped. This time, 3 fc. the spike was more pronounced as well'as the. rapid decrease in u b the signal. Then the signal began a rise to its maximum ind,i-y cated value at approximately 125 minutes. A significant change g 'L occurred in the signal's character when the "A" loop pumps were g stopped. This characteristic of the signal continued for at i least the next two hours. 'the postulation is that the signal ,y ~is an indication of voiding in the downcomer/ core ~ region.. The g P + 3 8 i. .7---- . g - ~~ 7 7.__ ...'.h n 3:L::. p l'

O

[M w rL H E D.. The SR and IR indication indicates that with no forced circu-p lation cooling, the signal is very quite probably due to voiding 9 d In the time interval 104-175 minutes, the indi-O cation is that a large density change took place. This indi-p 5. in the core. 2 cation remained until about 175 minutes when a RC pump was e g restarted for a short time; the indication at this point P decreased sharply which acts like a quenching ef fect or a f[ slug of water being pumped through the vessel downcomer. } iM ~ 142 minutes, the EMO block valve was closed, therefore a isolating the ERV and stopping the RCS leakage through the ERY. lf At ry The reactor building pressure and temperature began to decrease 'J once this valve was closed. k 3 The plant status at 150 minutes. b forced RCS flow due to all RC pumps being stopped. F-The RCS has no The "B" steam generator is isolated due to suspected primary to [. The hot leg temperatures are off-scale high secondary leak. The ERV was closed 'and the sy: tem has started to g ( 620 F). The RCS pressure was appr sxima cly 800 psig and repressuri.:e. 475-588 F. RC cold Icg temperature was This i RC makeup pump "A" was running with IIPI valves throthled. llPI condition has existed essentially since two minutes into the tra.n.s ient. e. Once the hot' legs were. filled with superheated steam, the only 6 heat transfer to the generators would be the condensing of the Since "B" generator is isolated and F steam in the generators.transfer is minimal, it.is concluded that 6 condensing steam heat cssentially no RC flo'w or cooling via the gtcan generatorsfor a brief period when RCp t exists after 150 minutes except is run at 174 minutes. At this point, the only mode of cooling for the core is by high pressure injection with flow through the [ Since the t pressuri:er relief valyc via the reactor vessql.142 minutes, there crists 3 pressufi:cr relief valve was. closed atremovat until around 200 minutes when RC pre pf. little decay heat A-sure is lower and ES starts both makeup pumps 1A and IC. 174 minutes when the RC pump-4 +

• brief cooling does take place at The Source Range 2B is stiarted and run for some 19 minutes.

y; NI indicates a quenching effect at this time but again rises 201 minutes where E rapidly until the llPI pumps are started at again they indicate a quenching effect. _ _ {Y 155 minutes, the operator began increasing "B" steam At about inches on the Startup Range to more generator level from 94 Steam generator than 65% on the Operate Range by 190 minutes. transfer occurs { pressure rises sharply, then decreases as heat when the generator is fed. W W 2-9 i e ~ '

• ~~

g 4 ~ 4 Pm' W'- -- ~. - w - g -c' c ~;

~ ~ s ~ y ENR Ru _ i' if~ [? w^ In preparation for trying to start an RC pump, the operator opens the main steam isolation valves on "B" steam generator. The b"- operator tried to start RC pumps 1A, 2A and 13 but could not get e t'j them to start. At 174 minutes, he did get RC pump 2B to start. Flow was indicated in the "B" loop for about one minutes. The s steam pressure rose after the liSIV's were closed, and the "B" f[{ The SR and IR NI's generator was isolated for second time. L 1 spiked down in response to the increased fluid density in the downcomer/ core region. p; p At about 133 minutes, steam generator pressure control was switched [ D from the turbine bypass. system steaming to the condenser to the

f. -

r atmospheric dump valves, due to loss of condenser vacuum. [ At about 215 minutes, the operator was steaming and feeding the i 7 generator. This 1esulted in decreasing the steam generator pressure further. The steam generator.s are essentially passive il4 now until the RC pump 1A is started at 932 minutes. ~ ~'fter The p H ssuri:er ERY was closed at 142 minutes and core A ~ At cooling is significantly reduced, RCS pressure began rising. about 174 minutes, RC pump 2B is started, this causes a slug of subcooled water to enter the core, as SR and IR indicate. Imme-distely the subcooled water is heated to saturation and then h superhe.ated. As this takes place, we see the pressuri:er level and RC pressure l'ncrease rapidly until the EC pressure stabilizes 3 at about 2000 psig. The water continues to 2xpand, driving PZR [1 level offscal,e high and causing RCS pressure to ncrease..The operator opens the PZR EMO at 192 minutes, allowing.both PIR [,t level and RC pressure to decrease and increased core cooling to g E/S actuates at 203 minutes and continues HPI cooling of g b e gir,. i The the core until 217 minutes when makeup pump 1C is stopped. SR and IR detectors respond to HPI actuation as they indicate an. hi increase in flow and density in the downcomer/ core region.

Also, at 217 minutes, the SR detectors see a large increase in down-comer water temperature when makeup pump 1C is stopped.

t The operat.or closes the P R EMO valve at 210 minutes and again f RC pressure and PIR level began incri:ssing. He again opens the t Q P:R E!!O at 231 minutes when P R level goes offscale high. The. ?" PIR E!!O remains open until af ter 300 minutes. The PIR level lg remains greater than 365' inches during this time period. [- Reactor building pressure rose to 4 psig at 235 minutes and S/S r actuation isolated the reactor building. The operator starts RC pump 1A at 248 minutes to verify previous starting and running amp readings. The pump was stopped after i F about one minute after the operator again observed a no-flow indication and running currents of less than 100 amps. The plant continues with these same conditions until after 300 g n J. I minutes. g g. -10 ~ w- ~"- .....~...%- 9 i

~ l I ' ;; i (; u-l: M u P i,- The reactor coolant hot leg temperatures continued to read 2 5 ~ n Q* off-scale high (greater than 620F) throughout the period. hC The reactor coolant cold leg temperatures were 221F in Loop 1 A and 243F in Loop B. Both were slowly decreasing. The U' k p level in steam generator A was 47% on the operate range and h steady. Steam generator B was isolated with a level of 67% k + on the operate range. The Power Operated Emergency Main I Steam Dump Valve (MS-V3A) was open and Loop A steam pressure [_ E was only 45 psi. Steam generator 8 pressure was 325 psi. (2 The Electromatic Relief Block Valve (RC-V2) was cycled to F._ h attempt to obtain a normal operating level in the pressure j-and establish pressure control. These efforts were not successful. Pressuri er level remained at 400 inches throughout the period. Reactor coolant pressure [ stayed between 1230 and 1350 psi until the decision was made to maintain continuous High Pressure Injection and increase to collapse the 4 Reactor Coolant System pressure to attempt This attempt h superheated steam / gas space in the systen. began.at about 313 min. and lasted approximately two hours, j 333. min. the B loop cold leg temperature dropped about 20F. Atin 4 min. the B loop cold leg temperature began incr(asing and Inte'r reached a temperature pisteau some 40F obove the 10 min. During this interval about St flow was indicated 323 min. valve. in the B loop. No explanation for these indications curr,ently cxists. 5 Following the closure of the Electronatic Relief Block Valve b (RC-V2), the Reactor Coolant System pressure. increased to about 3100 psi. This took about 40 cl n. (2100 psi at 360 min.). N.3 i For the next 100 minutes the Reactor Coolant System pressure (h . Manual control oscillat'ed regularly between fl00 and.1900* psi. (RC-V2) maintained y of the Electromatic Relief Block Valve the Electromatic Relief i At about 460 min. these oscillations. Block Valve (RC-V2) was opened and Reactor Coolant System pr' essure p was reduced to about 560 psi at 525. min. At 370 min. the level in the.A OTSG was gradually increased f-from 50% on the operate range to 99% at 433 min'. The pressure in the~A OTSG had rised slowly to about 30 psi but,' during the level increase, it' slowly decreased to 15 psi as the result of quenching action by the feedwater. Both Loop A and B cold leg temperatures slowly decreased [i (about 10F/20 min.) from 370 to 525 min. It is postulated o losses and the that this is due to the combination of ambient h I effect of " cold" high pressure inject an w.'ter. 8 [ At 325 min. the Loop A cold leg temperature suddenly dropp { l be open.(RC pressure approximately 560 psi) and a small volume h 12F. g f 2-11 e l j n, .s. ~ ~ ~ ' "~ ..-.[.f-.a.: J.a I 'h : a

r ? m, { ix ? ? I.d t v' C of cold water entering the reactor coolant system :hrough the

• [

GC core flood line (adjacent to the A loop leg temperature instru-ment being recorded) could quench some steam in the vessel L-annulus drawing flow out of the cold. leg. This, in turn, would [!fi draw cooler water up the pump suction line past the temperature C sensor. p L ((' 51 l The Loop A cold Icg temperature begins increasing from 163F g at 545 min. At 598 min. it has reached 198F. Here it increases }? to 219F. in the next 4 min. before starting to decrease again at } 602 min. During this interval (as "A" Teota is increasing) i about 6.2% flow is indicated in Loop A. The.sechanism for this is not yet understood. F [~ At approximately 589 min. the reactor building pressure spikes kE to about 28 psig. A small spike also appears on much of the system pressure and level instrumentation at this time. Source p Range Nucicar Instrumentation appears to have responded to this, t [ Reactor coolant pump. motor inlet air temperatures also responded by alatming high (IA and IB). There is a corresponding rapid l increase in the building temperature. The currently accepted f 7 k reason for this is detonation of the hydrogen gas which was present in the building. ? IJ h Plant Status at [00 min. 6 All RC pumps ar'c stopped, the systen has superheated steam / gas q (: in both hot 1 cgs with an RC pressure of 550 psig. '4akeup pump Q <Y 13 is running under throttled conditions. The p2R IcVel and A }{[ and B hot leg temperatures are all offscale high. The steam generators are not removing heat since the. atmospheric dump. . valve on "A" steam generator was closed at about 520 min. and "B" steam generated has been isolated since about 175 min. The R f( cold leg temperatures are continuing to cool due to the combina-E tion of ambient losses and "colti" high pressure injection water. The A and 8 cold leg temperature's are 218 and 142 F. The "A" steam generator pressure due to addit;on of cold feedwater and g lack of heat transfer has fallen to about ambient (15 psig) t Vf l whereas the '.'B" generator being isolated has maintained a - g s1 wly decreasing pressure which now is 260 psig. The steam o generator icvels have-' remained at about 100% and 62%.in the A-h! 9 g-and B generators for at least the last 3 hours. At'about._782. min.-the auxiliary boiler was returned to service E c and was supplying gland sealing steam to the main turbine. al f, Condenser va'cuuni was reestablished at* about 793-min. With a p. vacuum drawn on the condenser, the steam generators are again x ?' availabic to remov.e heat from the RCS. It appears that suffi-QU cient circulation on the primary side began.about 770 min., si* steam generator "A" pressure began to recover and a slight cooling of the cold le'g began.* After condenser vacuum was j' reestablished, it appears.that the."A" steam generator was 4 either fed or the turbine bypass valve opened at about 800-b- 320 min. as steam generator pressure begins to decrease. 'The j f c. e 2-12 ins m . p p....... s 1 s r:

~ .-] D v Hp tnt e L u The $D pressure had reached ambient by approximately 932 min.

.,7 )

q "3" steam generator was re-filled beginning at 696 min., u i T the secondary side pressure decreased as a result of the fill p]c_- -', ; and little heat transfer. The pressure stabili:es at about 712 min. and remains at about 160 psig until the RC pump is { started. e: The P2R block valve had been opened at -462 min, to depressurize L; the system with the intention of actuating the core flood system. l' c1 gy 550-650 psig range until -300 The RCS pressure continued in the min. when the decision was made to repressuri:e the RCS by fill- ((c ing with HPI water., The PZR block valve was closed and a makeup pucp was started around 800 min. to repressuri:e the g'"'~- system. h The RCS hot leg "A" tenperature returned on scale (<620'F) at 625 min. for the first time since going offscale high at 131 [. min. into the transient. It returned high briefl hen de-e: creased onscale and remained between 520 and 595*y, t F until the E }} 4 RC pump was started for approximately 10 sec. at 932 min. leg "B" temperature returned on scale. (<620*F) 1; The RCS hot min Once the 1A ~ M briefly at 932 min. and then again at 95e leg temperatures d apped offscale RCP was rynning, both hot low (<520'F) and the plant began an immediate cooldown onco forced circulation cooling was again established. 6 With the return of hot leg "A" te'mperature on scale (<620*F) at 625 min., there is evidently a quenching effect occurring }. in the top of the hot leg / steam generator that causes the T i 'This causes superheated steam to be condensed to saturation. a burp effect to occur when this hot water flows through the l1 "A" cold leg into the vessel. The "A" cold leg temperature - began a sharp increase of about 195 F at 672 min., the f16w 'J =. approximately 703 then slows,before another burp occurs at This second increase in cold leg temperature is sjower f i min. to increase by 165 F. the,revious one taking 2S min. I. ths;t p is taking place l1 ~~ ' The' P2R level indicates that quenching effect665 min. and S00 min. due to the decrease in icvel occuring at { Betwee'n and after these times, HPI water is added to the RCS M causing an increase in P2R level. j

f..

At about 500 miii., makeup pump IC is restarted and "A" steam h generator is again steamed to condenser, this causes a cooling g? cold leg temperature beginning effect as indicated *by the "4"F. The cold leg temperature a decrease from (95 F to 420-420 F until the makeup pump 1C was stopped at b remained at 883 min. The,RCS has now been replenished by the HPI system. 880 min'., thE~RCC is again'in a saturated /subcooled Atab[ut c state except for a bubble in the reactor vessel head and the 5 The RCS is now repressurized to approximately During this time period 835-930 min., the "A" hot [, "S" hot leg. I. 1eg and cold leg temperatures are gradually approaching e { 2350 psig. H other. It 2' g. [ l - - +,, g 15.- j h,. }: )

• e-m.,_,-

,;, t w f i U: f3 F I [ j; '^"* in the "A" loop during this period. At 932 min. (15 hrs., 32 rr.i n. ) RCP-1A is started and ran for approximately 10 sec. When the "A" pump starts reverse flow occurs through the "B" I p steam generator, once the flow falls into the "B" hot leg, the e p k superheated steam bubble is quenched and RC pressure falls. L; The system pressure rises as the bubble in the head expands I, due to system heatup. g& There was no PZR outsurge during the RCS pressure reduction I. due to the PZR being solid and the PIR block valve being closed. mi During the period enat.the pump was off (932-950 min.), RC flow [j O indicates an incretase in "A" loop flow, again it is speculated ~ that natural circulation is beginning.' c At 950 min. into the LOFW traniient, the RCP-1A was again started and remained in operation.' The Plant Status at l'000 min. K.8 h RCP-1A is running in'"A"~1 bop; - - ~ ~ ~ ~ - - PIR leve*1 is 400 inches Steam' Generator A is steaming to the main condenser. } Steam Gen'erator B'is isolated. RCS Prgssure is&1165 psig. ~ Makeup Pump 1B is operating supplying RC' pump sea. injection flow. p RCS Loop "A" cold leg temperatIare" is s275'F. b RCS Loop "B" cold. leg temperature is s275 F.. ,RCS 1.oop A 6 B hot '1cg temperatures of fsc' ale low (<520 F). E ~ 4 3 3, t } s s =, 0 ....-s I D l j r- .l. l s.-

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T' !.I C 9C L A* u 1 d g ~_

• t PROGRAM 3.1.1 CORE DAMACE ASSESSMENT l

b r. understanding of the extent of core damage

l Purpose -

To gain an sustained during the period of tine the core was un- [ k covered, an'd develop data to *se able to predict the L future perfornance of cores 1. similar accidents. b The objectives of this program are severalfold as f-l. Objectives - follows: ~~ To gain an understanding of the performance of the [ 1. core during the incident r I To develop a qualitative assessment of the perform-

• [

2. ance of the fuel and other caterials during the , incident - 3. To develop'the data necessary to benchmark analyses used to predict the behavior of a core, fuel..struc-tural, neutron poison and other materials during the gg course of an incident of a similar nature. I I. In. general, B&W has all of the necessary resource capa-T Resources Required - bility to carry out,the exanination and ana,1 ses of the 7 i Equipnent would have to be developed damaged TM1-2 core. P.. and tested t.: inspect the core in place; to retrieve fuel I pellets and debris and to map the core in detail as it is g disassembled and removeds :*one of these tools is beyond t The foundation

• the capability of B&W to design
• and test.

~ 1 laid in the past' on inspection tools, television. spent fuel examination PIE equipment and steam generator,t6cl. [ ing is sufficiedtly broad to enable B&y to design, build .'f g. and test needed equipment. No signific' ant capital facilities will be required with h g 'the possible exception of unique measuring devi.cas for 3 p. 3 hot cell.use,and possibly the planned upgrade of the hot A. ( cell. I The program would take about three years to, complete and on a cost basis. It is not expected h cost about $8,000,000 that B&Wwould be awarded the entire contract as the results 3&W needs to recognize this and f S are of national interest. [ E move rapidly to plan strategy and prepara general work scope. t. ~ -- t 3 3 A t v t sc t t j 'N ~- ^ ;- ( M ', 7 ~ ".u.2....*. I

?. *_s% ( D ' LC PROGRAM 3.1.2 CORE DESIGN ASSESSMENT <. = [f r Purpose - The data gathered in the " Damage Assessment" program can p be used to gain an understanding of the ability of the U core materials to withstand severe loss-of-coolant y accidents. C Objectives - The objectives of this program include: }1 J f ~ 1. A prediction of the extent of 'IMI-2 damage using data ~ gathered'from the Damage Assessment prograr .2. Predict the extent of damage which may result from yG several postulated incidents which invohe partial core uncovering. [ 4 3. Asses,s the suitability of fuel assembly structural ( l materials including the use of Zircaloy spacer grid. O [; 4 Coc. pare damage which may occur in R&W's cores with Wg that which may result from-other reactor manufacturers I' designs (This item for internal funding only.) g!} g k The availability of the data frca the " Damage Assessment" Discussion - should make it possible, in conjunction with 'DtI-2 thermal u 1 and hydraulic analyses results, to evaluate the effect f; of assumed cora uncovering incidents on the fuel assemblies. ,V This would permit an assessment of the time available to complemen't action to recover from us.elanned event sequences. El y In addition, the evaluation of core.aateri le, for suitability TJ J will gain by the availability of data. hon.e change in the [ -specification'of materials or selection of asterials may a result in a core design which can better withstand a loss- '( p of-coolant incident, t-Resources - The offort is analytical in na,ture requiring thermal, hydrau-lie, mechanical and metallurgy analyses. Some computar code i 3 {}. development will most likely be required. B&W has the-types of people required for this task. It is expected that ) the program would cost about $1,000,000. Work would be iif started innaediately with planning and be completed in 1982. j 'g ,r-i 1 $a q [ l g g i,- I i 3-2 r l 'W' I ~ ~ ,[ -- .. 2.:d. J.....I.: h. I 4 Y-

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PROGRAM 3.2.1 - MATERIALS BEHAVIOR f I!- V 5 Purpose - To investigate the effects of the incident on various ) materials of component construction (other than pressure g. boundary materials). Components and/or materials poten-g tially affected are: (1) instrumentation, (2) lubricants (3) electric motors. (4) pilot operated relief valve, and u (3) control rod drive mechanisms. j -- c Objectives - The objectives of this program include: 4: (1) A determination of the environmental can'itions that [. d each component / material was subjected to during the f [- incident. i 3 (2). An assessment of failure modes for all faculty com-f- 3 u ponents. (3) An identification of contaminants that could have led to failure or degraded performance. 1 (4) A quantitative indica' tion of,the degradation in L' performance, as measured against original.*specifi-cations, of the component / material and relation of ( current performance to life expectancy. Q p g ' Reccca:endations for improvements in components / p g (5) materials to assure. increased reliability and life.* i The various prujects in this program require vastly h Resources - t. different manpower and facility resources which are hi ,q" Apecified in the individual pr Ject s'sscriptions' l.a g* Appendix 3. In general, R&*J has the technical competence B and facilities to perform the majority of the work. The l largest effort. required would be in the instrumentation {,. Each of ip , area where up ta'eight ma'n-years may be needed. ( the other subjects can be dealt with in something like one' y_ i man-year. The largest project could be accomplished in ($ I approximately two years. l l l a e.m. 4 I a g b 3-3 E t r I ,,, g __,

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. 3 (! rc5r = 11 0 ' m i., 't X PROGRA:1 J.2.2 Sil'LLATION OF SCISOR CONDITIO'IS-r:

f. '

Purpose - Evahate the ability of current instru:nencation to function !b dud g an incident.. b v g Objective - Detersine what effect the TMI-2 incident conditions would [ have on current equipment. Recocnend instrumentation ' t f equipment installation improvements. [ Resource - .Obtain a proposal from independent radiation test labora-tory where activity would cover 6 to 9 months. BCC0/CPCO will require several months for site data collection to be [ 1 used by the test laboratory..- d i g g, y x o = { U ! W HI WO i

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t ~~ ?0 JDI w .iG PROCPxt 3.3.1 - PRESSURE BOUNDART !iATERIA1.$ ?, Classification Area: l'aterials Behavior 0 a h,. t W E

Purpose:

To determine whether pres.ure Soundary 5 materials have been significan ly i degraded by the environment imposed on g,~ them during the incident. k Determine the temper'ature leveis and Objectives: time at temperature' suffered by the [ reactor vessel, primary piping, and k. steam generators. r (i g Assess the impact of these temperature excursions on the acceptability of is these materials for future plant opera- {l tion. t Resources: The resources required will be heavily v dependent on the maximum ta=peratures f-experienced by the materials. If found n to be below 1200*F little additional. g- ~ ' nvestigation vill be required. If (- i above 1350*F prognosis la bad fgr = licensability. If between 1200 F and 1350*F possibilities of saving the components exist but major sampling and testing of component. saterials will f; - be needed. Depending upon scope this

• y t

program could extend from $80,000 to h. $500,000. 5 .n. I Pi. i..1 s. 3 { 3 r Y. l t". e5 ,7;. 1 ^ s. p F5 t F. { ...n_m....."_,. _*r-

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m PROGRAM 4.1 SYSTD1 BEP.AVIOR ANALYSIS tu ! e. g Purpose - To develop an improved safety analysis concept which will perwide the capability for a mechanistic and systematic analysis of sequences of events with multiple independent r The analysis of this class of events is expected l causas. to re,sult in improved nuclear plant safety systems and g operator training programs which will improve the response to such events and also. reduce their probability of occur-I rence. - - ~ ~. N.** r. R ( g Objectives - The objectives of this program are to: f l. To develcp methods for the mechanistic and systematic safety evaluation of sequences of events with multiple (;,'. independent causes. This analysis should include 3 consideration of I

• Consequential ' failure of previous events h7
• Single random failures 4,[

. Maintenance unavailability l .Coseen mode failures The method should avoid' undue reliance on reliability ~ L.[ or failure rate data. The method should also include [ I realistic simulation of plant and operator response L to a sequence of events. i, b. I 2. To apply the improved safety analysis concept to develop

/

improved safety systems and operator, training programs. which will provide greater protection against seguencee j [. I .of events with multiple independent causes. E-Discussion -

• Tlioupson and B4ckerlye stud'ied a' n'u her of" reactor accidents f~-

( and concluded that most accidents have involved sequences of e. T.hree or more independent events which interact to produce }' i (1:. acre severe consequences than any of the events alone. They 1 I defined three broad ca'tegories of basic causes. These ares F i rt-

1. ' Design or equipeent salfunction.

b-

2.. Human er or and C-3.

Instrumentation problems includeing error misinterpre-p 3 tation, or lack of instrumentation. [. h 'The NRC evaluation concludes there were mult'iple independent E I causes for the TMI-2 incident.- .)~

• The Technolony of Nuclear' Reactor Safety, Volume.1 bI t

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= -=--..n y fhD ^ [ 5 9 - l Discussion - The safety evaluation of sequences of cultiple independent (Cont'd) events is complicated by the r.any possible permutations and h'S' co=binations of events; particualrly shen dur.an error is considered in a mechanistic analysis. 0.-E ne existing 3RC standard review plan safety analysis frame-p work and criteria do not provide the needed level of safety y for sequences of events with multiple' independent causes. [ Prior to T:!I-2, it was considered that analysis of design-( basis acci'ents in accordance with the single failure t d & L criteria would provide a substantial safety nargin under note likely situations. This conclusion must now be considered. L (, The Rasmussen report

• used a risk-based event. tree / fault tree

,N, p analysis method. Probability analysis and failure rate data ( C - were used to calculate quantitative risk values. However, the present data base does not permit consensus acceptance F DC of risk, values. 6 X B&W has used an alternative nethod in Vest Germany for-- t evaluation protection systems and engineered safety features. 3 This method appears adaptable to provide a mechanistic event tree safety analysis. This method does not place undue T reliance on reliability or f ailure rate data, nor does it [ i include arbitrary assusptions regarding the safety margins afforded by a design basisi accident analysis.' This system includes systematic consideration of sequences of events O involving an initiating event, consequential failures of-Q( previous events, single randon f ailures, maintenance unavail-EBp g ability, and commqn mode, failures. ,sa t.. g [T. Resources Required - B&W has all the neces'sary resources required to develop the proposed method for U.S. application and to carry out the 5 analysis, and has done such analyses.for pressurized, water * { e. r'eactors in West Cernanys k I The analysis methodology required-devel*opment and adaptation g F-E$ for use in the United States. This includes: h 1. Development of event consequence acceptance limits for g f, U.S. application. g {. \\; ~ 2. Combination and number of additional faults to be coe-f ~ sidered. er 3. Assumptions to be employed regarding human factors and -[ 4 man-machine interface. Development of the methodology would cost approximately $150 p g '. to 4::00K and require 4-6 months to complete. gh The analysis itself would cost approximately $2M per plant I design and would require about 1 year to complete. { f.

• • Reactor Safety Study WASE-1400, o'et.1975 B

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Y i m h Purpose - Deterrine instrumentation requirements during and fo11oving 'q postulated accidents, and i= prove the capability of monitor-N Q ing the 5555 and SOP state during and following unusual p w incidents. ?: h E Objectives - The objectives of this program include: [ Q l. Characterize postulated accident conditions to define 0 the scope of conditions f,or which nooitoring instru-mentation is required. [~ 9 2. Determine variables chich must be ceasured during incidents and' deter =ine requirements for recording. ? C.~ display, and analysis.of,infornation. 8 3.

• Determine i=provements which can be made to existing h,

{'. instrumentation to better meet these requirements. y. g. Evaluate the usefulness and feasibility of new instru-f;' I.. g eentation ta me.et these requirements. WPCD, RDD,'and 3CCO would contribute tb this work. The Resources - program would also benefit by EPRI participation since 3 It has an industry wide applicstion. Estimated costs .t ~ for such a program vary, videly, depending on equipment S h' development / testing costs. Engineering time is estimated k. to be on the order of 9 =an-yeirs. p i 3.' 0 g 1 [. i. T Y f r I g 1. e= b l t F I ~ [ .. ~ t I I 5-1 O t I I l ? -=- v- ..e. .M.. e a b fh

~ ~ %.) , (. 101 [ f u.Di e PROGRAM 5.1.2 NORMAL OPEP.ATIONS I':STRUMENTATION AND COMTROI.S N. $g~" ' I .I j u9." 2C 1 g a h v.... To evaluate near-term improvements tha. cou1Ld be made in g [' U r. Purpose - the instrumentation and controls systens as a result of T11-2 to determine if there are alternate configurations which are more desirable from a total syn. tem viewpoint. w e Objectives - The objectives of this program include: Review of near-tern improvements that could be made in h _- 1. instrumentation and controls with regard to any adverse effects during plant conditions other than those made QE ij 3-prominent by T:3-2. .t L 2.. Determine alternate long range solutions which may be rg 3) y more desirable. b Work is currently being performed to investigate improvements k Discussion - to the instrumentation and control' systems in several areas. This program is proposed to review those improvements from a total system and long range viewpoint. Examples of specific Q tJ aress to review include: r-1. PORV (reliability, positive position indication. j. inferred position from existing instrumentation, E automatic closure of block valve on low pressure) A

ir 2.

Iticreased range for temperature measurement Q C. g. 3. Primary systen saturation margin instrumentation 4. Reactor trip on loss of feedwater and turbine trip f 5. Increased,use of incore,thermocouples - ~ .g$- 6. Containment isolation ;hilosophy 7. Instrumentation to monitor the state of readiness f-of Engineered Salety Features Systems. Q.*k'. [ 'N?CD would be the major resource for this program with Resources'- RDD and,ECCo support. It is estimated that the program r The would require 3 man-years and 1 year to complete. . work..shoul.d. start as 'soon as possible and be internally ~ p funded. H . 535 H: 5-z " ^ ' *- . -- - - -, 7 ;no. -C- ",n u -- ...b.'.. ;-j.'j&.^.: '. e

F.- m Ii % ~ f,c61 Q" h PROCRAM 5.2.1 RUID Rou UNDER ACCIDENT CONDITIONS d If]i E-E( [s, c iD accident conditions and to consider design changes where p[,,M 'To review the possibility of flow systecs clogging under Purpose - necessary to assure continued operation of essential systems. O h Objectives - The objectives of this program include: t. 1:2 1. Evaluate the probability of debris collecting in the [: vafious flow systens under accident conditions. 2. Rev'iew the ability of pumps, filters, and flow control E - Y devices to continue to function under accident conditions. g g k'here there is an identified need, consider design [ 3. k improvements to assure continued operation with debris b in the system. + i Q Discussion - During the TMI-3 incident, letdown ficw control devices - 1 became partially clogged, but were able to give sufficient m letdown to balance the seal injection needed for the RC 4 This program is proposed to review all flow systems !~ . pumps.

• log under accident con-I 3

with respect to their tendency t. c g ditions and evaluate design imp.ovecie: es where appropriate. 1 E Exanples of possible design improvemeats are: o t,. 5 1.' Addition of estergency filter bypass lines. ?. b 2. Addition of paralle'l flow control devices with larger 1 h j openings. 3 g + 3. Addition of r' emote means of backflushing. ~ i S' It is 9, .. qPCD yould.be the. major resource f r this program. Resources - estimated that the program would require two man-years and, hs. p 9 months to complete. The work should start within six mont g,, 1 m I n. i6 N . _ +. t I 6._ i n ls. $e' E' y. .S 5 1_ g E, P Y -z - l - -..v.--- =._ _, _ :., <3 . :. G.$.l$ ' U. *.. w,a P k-f* ---_______1_E_________________________

( .n a p \\ . C. l ,( '8 'D i _...10 i $,q f PROGRA.M 5.3.1 FEEDb'ATER SYSTUt3 h.Lr. Purpose -- Investigate methods to increase the reliability of the main feedwater system and to provide an auxiliary f feedwater systen with the necessary reliability. W Objectives - The objectives of this program include: 1. Identify the causes of loss of main feedwater events ,~ which have occurred at B&W plants and develop design requirements which will substantially improve the reliability of the main feedwater system. 2. Review tho' design (components and control) of the auxiliary feedwater systen and develop, design re-quirements to increase the reliability. ) Resources - NPCD design capability would be the primary resources i however, since the program relates to the integration i of B&W equipment with equipnant supplied by others, the i program would benefit by participation of a BOF designer. l It is esti.nated that the program would require approxi-3 l nately 4 manyears and on the order of one ye4r to A t complete. The program is appropriate for contract funding. [. J i. d.

• J k

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.j ,n = c \\b n ~CN L> i~6 f. PROCRAM 5.3.2 CONTROI. ROOM OPERATIONS 1 3,. kJ Purpose - To improve control rooms and operator training so that they C bring forth from nuclear power plant operators appro-E priate and timely action and inhibit inappropriate { action during contingency situations. M3 j Objectives - There are several objectives to this program. 1. Establish the basis for safe operation in basic a* [] terms such as heat balances, coolant inventory, and Q structural limits. Is.I I 2. Broaden understanding of what various contin-gencies will do to the plant by looking beyond the j incredible accident to the many contingenc4es that 3 are more likely to occur. 3. Establish what in' formation the operator needs, [ 3eterdine optimum display methods and how the operator 'j2 should be trained to respond to such optimum displays.' l:. De, sign a control room anii operator training upgrade ' " ~; package for existing and near-term plants based on the principles of human factors engineering. g; m, e I

• 5. ' Design a control room and. training program that 1.

clearl'y supeiior to that of other NSS vendors for W longer term and' future sale plants. This control' M-e. room design should be based on the principle of human t. {. factors engtheering. minimizing arbitrary limitations placed on the design. by traditional scope of supply and vendors. '4 g,

• 6.

Establish B&'J'as the*1eader in human centared, well enginedreB control rooms and operator training. NPCD should assume full' responsibility for control roca ' Resources - design with support given by the R&DD and Bailey. It I will take scaathing like a six mar 4 team working for at j E. least three years to achieve these objectives in t'erms of' reference designs for the short tera and long ters. h Additional effort will be required on each contract to p implement the changes. (~;, b Items 5 & 6 are strongly advocated by.-R. N. Kubik.. However the. connittee. F S did not endorse them. f g I p n h 1 - ~. - -..- g .....a. : b '... ' ~ L - ]C L

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-r .C Dh e! l n.e U Li PROGRAM 5.4.1 CORE COOLING (NATURAL AND FORCED) wy 0 is t c f5 '*.':~~ ~ E. l I Purpose - Investigate core cuoling systems con ascent with P equipment malfunctions and inappropriate <erator actions (i E and to improve the abt?.ity tol establish and monitor natural circulation under postulated accident conditions. (t t t-g. t.. Objectives - The objectives of. this program include: - o. t' Evaluation of possible' design improvenests to increase 1. the probability - of maihtainics solid water in the' y primary loop. W [.. 2. Evaluation of means to re-establish solid water con- .ditions if dissolved gases are released, or if sedaz locks chould,foria in the primary loop. p 3. Determine instrumentat'.'on stquired to monitor ths i primary loop and its ability to erstablish natural circulation and also montter natural circulation after .y it is established. W 4 Study decay heat removal under all:yssible' operating 1 conditions. E \\ . W i. i. NPCD design capability would be the major resource for 3 g,- Resources = this prograa with about 20% of the work performed by the It is estimated that the program would ( 9 R&D Division. Most require 6 man years and about 2 years to complete. ... q.k of the funding would be internal; however, some parts of Q y. tge work any be appropriate for outside funding. f. l l g Q ?.5 J c i ,g; .(6 i g E L' s

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ilt f t f w y-p 6.0 Potential Changes in N?.C Philosophy _ ,] k The challenge The changes in liRC philosophy due to TMI-2 will be najor. It b~i facing industry will be to somehow contain and lead these chantes. this time that virtually all areas of the ::SS will be af fected, p appears at i The NRC is working on a list of 10 areas which will be subject to major b, change: L fI 'J 1.. Reactor Operator Training and Licensing k e 2. Reactor Transient and Accident Analysis Instrumenta-3. Licensing Requirements for Safety and Process Equipment. f tion and Controls [ } Offsite and Onsite Emergency Preparations and Procedures 4. 5. Reactor Siting -. . L... Licensee Technical Qualification I' NRR (Of fice and IIuclear Reactor 7egulation) accident-response role. i 7. capability and nanagement. ? 8. Feedback of reactor operating experience ,I.

9. ' hT.PA Considerations

10. Licensing Reguirements for Post-accid'ent 11onitoring, Diagnosis f,J id and Control

) NSC has. reorganized the Of,fice of.*uclear Rest tor. Regulation, and a special f' group has been set aside under Drs Roger Matts Ja to prepare a " lessons They will be defining regulatory.hanges which they believe Learned" report. appropriate as a result.of Tf1-2. [

• (

The ACRS has also prepared several interim reports containinC recomendations p for dealing with the generic implications of TMI-2. They plan to i'ssue a Their recommendations address , 3 final report toward the end of the sumer. Scme represent concerns which 3 many of the same areas as the !.TC list above.are very specific, such as some are extremely broad such as operator training and qualification and g Their Q questioning the future acceptability of the single failure criterion. recommendations as of Itay 16', 1979 are included as Enclosure 1. Plant systems and equipment which heretofore have not received auch attention s. will get close scrutiny because of the vital role played at IG-2 by many Examples are the purification and letdown system. { I the pressurizer vent system, the steam genarator level ceasurezent instruments. { "non-safety" components. t etc. { R&U has now underway in the Licensing Section a review of all aspects of its licensing activities to assess ways in which the requirements and the licensing

t This effort is aimed at i

process are likely to change as a.rrsuit of IsI-2. ,r. 6-1 ff e 1 l n, - ~ .,.g.-. z-- n ..j_b :.-_...;.n.. '... * .I l l' ~ hi

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,o i - -l g 21 r" M y L Q, ] ' a a N. ( 6.0 Potential Chantes in NRC Philosophy (Cont'd) .M-assessing the impact on backlus contracts such as the Midland. Bellefonte, y*- y Erie. VPPSS. PCE, and DECO projects. An interim report is due wt as a [ f p1 result of this review on July 15. 1979. In general. it is expected that the regulatory clinate will become much {b u y ocre penetrating, such more independent (i.e. relying more on their own analyses) and more costly to deal with. It would seen that the promise of regulatory. stability and the benefits of the standard design have 5 [ disappeared at least for the next five years. it is tco early to state specifically what the NRC will [, f At this time. change, but as stated above. industry must dev. slop an ef fective channel to communicate and influence the staff. This will be dif ficult because of =E E outside criticism that the staf f has been influenced too much already f rom C,, I: the outside. w a 6.1 Potential Changes'in Plant Operations _ \\ Significant changes in the training.' qual'ification. Licensing.* and super-i 1 vision of operators for commercial nuclear, power plants appears likely. ? Immediate changes are not evident; however, increased difficulty of license 1 examination and review of the station requalification program should occur immediately. Other changes are possible. which include:

1.. Establishing a licensed supervisory position ag ' esc'h station to be manned j

by a highly trained engineer whose primary responsibility'is core safety. ,g rs 3 The position would be manned around the clock. This would provfde back y}- shitt engineering. entineer with primary responsi-The NRC may establish an on-shif t safety'nd responsibilities would be [i 2. bility for safety of the core. Duties a W L similar to those of a plant supervisor and the individual would have Ir authority to shut down any of the plants and control the plant's safety g The NRC would have to embark on a very extensive recruiting, p# systems. compaign along,with training and qualification programs to implement this. n It'would be expected that an,incrcsse in the attention pa'id to qualifi- ? 3. cation programs utilized by the utilities in preparing individuals for 7 Past utility practice in control positions will come from the NRC. Q general did not have qualification programs for promotion to seat level-p The eligibility is for the most part covered by the union con-jobs. p Institution.cf qualification prograne would help improve the } tract. technical level of prospective reactor operators. More attention to operational performance will occur. At present, 4. actual operations are the concern of the Operator Licensing Branch j( This is a group of nine to ten individuals assisted by This group must expand to provide adequate personnel to { ] of the NRC. consultants. support direct involvement in plant operations. -dperator Licensing is y This [ experimenting with decentralization to the regional I&g orifices. l t J - ' ~ ~ c y...%. ..pl~...,,..... ' f 1

T A 4 . _i l e ..4' .w ,1 -I [ %' t ' .g, ! ru 6.1 Potential Changes in Plant Operations ,.::~~"'~ o 4 (Cont'd) { will probably continue and may nake it possible to attract sufficient qualified personnel to make close supervisiois of plant opetJtions s 6 possible. U 2 E i 5. In this extreme. a change in supervision of operation that might be-I implemented would be the adapting of the U.S. hvy system of opera- [' tional readiness examination of each power plant and staff once each year by a qualified board of examiners. Annual examination would consist of k [. 4. Oral examination of nost of the plant personnel including plant C superintendent, plant' engineer and staff, auxiliary operators, f technicians, and " licensed operators." p v. ,b. 'Jeitten examination on the gaineral principles of operation. f [ c. Operation examination of shifts to determine performance levels ^ under normal and abnormal conditions. g d. Inspection of the plant's material condition'. E t g Based on tho examination. the plant would be allowed to continue to. operate I W or be shut down for repair or training. Passing a re-examination would be {,' required to t'ast.ne operation. Instituting such a' full scale examination g pgogram would have major ef fects on the plants and uti,11 ties. The most y 3 significant would be { a 1. Each plant control room would'have to have 4. simulator except if there } were two plants having duplicaGs. Oconee would have to have three (3) simulatoru; Arkansas, two (2); Midland, two (2), etc. The simulators L become immaediately required when the examining staff 1%, quires the shif t [ personnel to respond to casualty condition. The power grid nor the system can stand excessive reactor trips, loss of feedvater, nor large power I[. swings to demonstrate proficiency of the operators and supervisory personnel. 4 4 W 2. Personnel retention woul'd become a pedlem"similar to that of the U.S. l Navy. The contributing factors would be high stress for error free L l performance and the potential of long working hours. l 3.

lilitant operator unions could, develop with. c'tendant demands, for very high pay and significant reduction in work week he ars.

l:. 4. Very significant difference in pay and working con (itions between f t l fossil and nuclear would further compound utilities' personnel problems. } This is a situation similar to the' airline.pliots.,' ~ t. l e G f f ,7. .,p-3., ; ~ .g_ ,..... ~. V.. l ((

T ,3 ,{ f e wn - - glNb rds w ENCLOSURE 1 ! NC g ADDITIONAL RECC:OtE::DATIONS RELATING TO } [3..-~ TMI-2 ACCIDCTT 03 MAY 16,1979 r ACRS LETIERS b c IM A. Interim Report No. 3 dated May 16 1979 l R:connendation 1 - Examine operator qualifications, training and licensing, and requalification training and testing. i L ' R:cocoendation 2 - Establish formal procedures for the use of LER informations [J[ 3t. (a) In training supervisory and maintenance personnel f (b) In licensing and requalification of plant operacing hg-personnel (c) In anticipating. safety problems [j Rscotmnendation 3'- Consider formal review of operating procedures for severe I J transients by inter-disciplinary tean, and develop more j. standardized formats for such procedures. g f. R: commendation 4 - Re-examine comprehensively the adequacy of design, testing 58 2 i and maintenance of offsite and onsite AC and DC power supplies with emphasis ons

• -{

(a) Tailure nodes & effects analyses b m (b) More systematic testing of power system reliability K (c) Improved quality assurance and status nonitoring of y power supply systema Y R:commendatio'n 5 - Make a detailed evaluation of current capability to with-aj {@J g stand station bla,ckout, includings

• (a).Exanina' tion 'of naturar circulation capability under p

f such circumstances ? (b) Continuing avaliability of components needed for long-term cooling under such circumstances E (c) Po'tential for improvement in capability to sur9ive E; ', extended blackout Et .Recossmendation 6 - Examine a wide range of anomalous' transients and degraded l C: accidents which might 1.ead to water hasaner, with emphasis ( ons

• I G

L-(a) Controlling or preventing such.condftions ( (b) Research to provide a'better basis for control-or - t j prevention of such conditions h, e E i 6-4 i g-g t w., 3 .n. : =; - ~ =- =;; ~- ..?...a., X.'..u.z JI 'rg, 1

~' ~ r ~ [ us. i ~ O d' (~ 9 yM ?hId C 1 1 ? fi ' N'. [ ENCLOSURE 1 6 ADDITIONAL RECC:".*C' DATIO!!S RELATING TO se H TMI-2 ACCIDCIT ON MAY 16, 1979 { g ACu trrrzu 0b 1 A. Interin Report No. 3 dated May 16. 1919 F b Recomendation 1 - Examine operator qualifications, training and licensing, F and requalification training and testing. II Reconcetidation 2 - Establish formal procedures for the use of LER informations tj ) l (a) In training supervisory and maintenance personnel g-(b) In licensing and requalification of plant operating personnel { (c) In anticipating. safety problema i Recommendation 3 - Consider formal review of operating procedures for severe f transients by inter-disciplinary team, and develop more standardized formats for such procedures. Q f M 2 3' Recoassendation 4 - Ra-examine comprehensively the adequacy of design, testing f - {g acd maintenance of offsite and onsite AC and DC powet supplies with emphasis ont - h yailure nodes & effects analyses (a).More systematic testing of power synten reliability T (b) (c) Improved quality assurance and status monitoring of y power supply systema ) IIrl' Recoammendation 5,Make a detailed evaluation of current capability to with-j stand station blgekout, includingt 'li

• (a).Examina' tion of staturar' circulation capability under j'

2 such circumstances (b) Continuing avaliability of ce,ponen s needed for long-5 m tern cooling under such circumsta..ns M b (c) Po'tential for improvement in capability to surfive ni extended blackout g dd . Recommendation 6 - Examine a wide range of anomalous' transients and degra e accidents which night 1_ead to water hasseer, with emphasis h g [e.l ons Controlling or preventing.such. conditions L: (a) (b) Research to provide a better basis for control-or ( ~ 1 prevention of such conditions a i g y 4 g l b4 g t \\ k -.-y ~g .p-- .=~ l --g.- 7 .4..a :.. i. u :- 5

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%4 fr vi J :di ~ bi di +3C1 Recoczmendation 7 - Plan and define NRC role in emergencies, including con-sideration of: L 3.T [ Assurance that formal emergency plans, procedures (a) t; and organizations are in place il (b) Designation of energency technical advisory teams f (names and alternates) ? (c) Compilation of an inventory of equipnent and materials { D needed in unusual conditions or situations 0 Reconssendation 8 - Review and revised within three months: I Licensees' bases for obtaining offsite advice and g (a) assistance.in emergencies f rou within and outside t i Licensees' current bases for notifying and providing ( company b (b) information to offsite. authorities in emergencies Y' Recommendation 9 - Exanine the lessons learned at III-2. including consider-I ation.of the followings --.. -..__... (a) Behavior, failure modes, survivability and other t l aspects of TMI-2 components and systems as part f of the long-tern recovery process (b) Determine if. design chantes are necessary to g. facilitate' decontaminatic and recovery of major, et t. nuclear power plant systems g 3f. ' Recommandation 10 -Expedite resolution of unresolved safety issues by the N following means: 25 E Q Suitable studies on a. timely. basis,by licenseep to h (a) auscent NRC staff efforts { pg (b) Use of consultant and contractor support by NRC Staff t. y [ Recon =sendatiod 11 - Augment expeditiously }.he NRC staf f capability to deal with prob,lems in reactor and fuel cycle chemistry in the g g following areas Behavior of PWR g BWR coolauts and other materials h-(a) I under radiation condittens ti Ceneration, handling & disposal of radiolytic (or ).. (b) other) H2 at nuclear facilities (c) yerformance of chemical additives in containment' K., g Processing and disposal techniques for high and low y sprays (d) 1 level radioactive westes (e) Chemical operations in the other psets of nuclear j I fuel cycle 5 (f) Chemical treatment operations involved in recovery. I' decontamination of decommissioning of nuclear' }. ? facilities '~ .L 6-5 e c -- *w .c ..y 7----

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