ML20070R063

Testimony of DC Richardson on ASLB Question 1.4 Re Risk to Public Presented by Chain of Events Including Pressurized Thermal Shock

ML20070R063
Person / Time
Site:	Indian Point
Issue date:	01/24/1983
From:	Richardson D CONSOLIDATED EDISON CO. OF NEW YORK, INC., POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20070R032	List: ML20070R027 ML20070R043 ML20070R063 ML20070R091
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR ISSUANCES-SP, NUDOCS 8301270268
Download: ML20070R063 (13)

Text

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U iYc UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION JAll26 /d0:49

' ATOMIC 1 SAFETY.AND LICENSING ~ BOARD r

t :

Before Administrative Judges:

-James P.

Gleason, Chairman Frederick ~J.

Shon Dr. Oscar H.

Paris i

)

In.the Matter of

)

)

CONSOLIDATED EDISOU-COMPANY OF

)

Docket Nos.

NEW YORK, INC..

)

50-247 SP

-(Indian Point, Unit No. 2)

)

50-286 SP

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POWER AUTHORITY OF THE STATE OF

)

NEW YORM

)

January 24, 1983 (Indian Point, Unit No. 3)

)

)

LICENSEES' TESTIMONY OF DENNIS C. RICHARDSON AND DENNIS C.

BLEY ON BOARD' QUESTION 1.4 ATTORNEYS FILING THIS DOCUMENT:

Brent L.

Brandenburg

' Charles Morgan, Jr CONSOLIDATED EDISON COMPANY Paul F.

Colarulli OF NEW YORK, INC.

Joseph J.

Levin, Jr.

4~Irving Place MORGAN ASSOCIATES, CHARTERED New York, New York 10003 1899 L Street, N.W.

(212) 460-4600 Washington, D.C.

20036 (202) 466-7000

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8301270268 830124 PDR ADDCK 05000247 T

PDR 1

pm I.

. Introduction My name is Dennis C. Bley, Ph.D.

I-am a consultant at P2.ckard', Lowe and Garrick, Inc., in reliability, risk, and decision analysis for electrical generating plants.

I was a l principal-investigator on the Indian Point Probabilistic Safety Study.

'A~ statement of my professional qualifications is attached.

My name is Dennis C.

Richardson.

I am the Risk Assess-

-mont Technology Manager in the Nuclear Safety Department of the Nuclear Technology Division of Westinghouse Electric Corporation.

I was a principal invest! 1ator on the Indian Point Probabilistic Safety Study.

A statement of my pro-fessional qualifications is attached.

'his testimony addresses Board Question 1.4:

T Uhat risk to public health and safety is

. presented by the_ Indian Point plants

.through a' chain of events including

- pressurized thermal shock (PTS) to the reactor pressure vessels?

II.

Response

Pressurized thermal shock is a concern for the integ-rity of irradiated vessels during cooldown transients in which the vessel is repressurized.

The neutron irradiation of.the vessel increases the tendency of the vessel material to transform from a ductile to a brittle regime during these

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Leooldown events.- Thus, vessel integrity could be lost y

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'through crack extension through the vessel wall.

' TheiIndian Point"Probabilistic Safety Study.(IPPSS)

Jevaluatsd the1 frequency of a reactor vessel rupture large.

enough.to exceed.the capability.of the emergency core cooling systems...That evaluation used the same~ methodology-

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and assumptions-as the Reactor Safety Study (RSS -- Ruf. 1)

. and yields'a mean frequency of 3 x 10-7 per reactor year for allitypesief vessel failure,~which included those failures

^

induced byltransients (pressurized thermal shock (PTS) chain-

-of-events) and spurious events. ~ Subsequent to the issuance of the'IPPSS, a: specific casessment of PTS transients and 5the associated-fracture mechanics was performed by the

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Nuclear' Regulatory Commission (Commission) and Westinghouse

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Owners Group for Westinghouse pressurized water reactors.

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1The quantitative results.of this analysis are bounded by the IPPSS evaluation.

LThe-IPPSS and RSS result is based'on data and judgments

- presented-in a report by the Advisory Committee on Reactor-Safeguards (ACRS -- Ref. 2).

Because the. experience with I

nuclear pressureLvessels was limited,.the ACRS report was

-based.on more extensive but applicable ~ data for fossil-

-fueled boiler drums used.by niectric utilities for the gen-1eration of electricity.- These data are generally applicable-

.since the areas of. concern which drive the analysis results 2.

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- are the heatup/cooldown transients and pressure / temperature operating.regionr.,. which would be similar to those exper-ienced by nuclear vessels.

No catastrophic failures have been observed during operation in the United. States in more than.700,000 vessel years.

Failures during pre-operational cold hydrostatic testing were not included.

Therefore, to the. extent.that fossil plants are susceptible to preseurized thermal shock' conditions, the IPPSS included such vessel

-failures.

Furthermore, work published-since the IPPSS supports our judgment that pressurized thermal shock is not a major contributor to risk at Indian Point (Refs. 3 and 4).

Vessel rupture could cause core melt, but would not, by itself, cause loss of containment integrity.

Additional unlikely failures such as loss of coatainment heat removal must occur before a significant release of radioactivity is possible.

Based on the generic report on PTS by the Westinghouse Owners Group (Ref. 3), the Commission's generic position on PTS (Ref. 4), and nil-ductility temperature calculations specific to Indian Point Units 2 and 3, the probability of extension without arrest of the cracks which are assumed to exist in the reactor vessels at these units is conservatively approximated to be at the most 10-6 per reactor year-for current vessel fluence levels.

This frequency would be roughly an order of magnitude higher after 20 years of operation.

Further work to support rule-i l

E

, making for PTS is underway, and plant specific analysis should result in significantly-lower frequencies than the Labove generic values.for Indian Point Units 2 and 3.

The above frequencies'should be further reduced when extended to cover PTS-induced vessel ruptures large enough to exceed the capability of-the emergency core cooling systems.

The above Efrequency of 10-6 for'small cracks propagated through'the

-vessel wall should be reduced by at least an order of magni-tude for disruptive vessel failures based on the Marshall report (Ref. 5).

This is consistent wi*h the 3 x 10-7 fre-quency stated above so that the IPPSS treatment of all causes of~ disruptive vessel failure does indeed cover the PTS concern.

Current NRC policy is to key plant specific PTS analysis on a nil-ductility screening temperature limits.

The current screening criteria indicate that the Indian Point units will not approach this' limit until after the

. year 2000.

l Based on the above, pressurized thermal shock is not a risk concern for the Indian Point plants.

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References-

~

1.

Nuclear Reguletory Commission, Reactor Safety Study:

~An Assessment-of Accident Risks in U.S.

Co:amercial Nuclear Power Plants," WASH-1400 (NUREG-75/014),

October 1975.

12.

The Advisory Committee on Reac' tor Safeguards,'"Repott on the Integrity of Reactor Vessels for Light-Water Power Reactors," WASE-1285, UC-78, January 1974.

- 3.

" Summary of - Evaluations Releted to Reactor Vessel Integrity," performed for the Westinghouse owners.

Group, Westinghouse Electric Corporation, Nuclear Technology Division,.May 1982

- 4.

SECY-82-465) Policy Issue on " Pressurized Thermal Shock," William J.

Dircks, Noverher 23, 1982.

5.

United Kingdom Atomic Energy Authority, "An Assessment of the Integrity of PWR Pressure Vessels," October 1976 and as revised in 1982.

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Dennis C. Richardss- - Risk Assassant Technology Manager Fenn Stata University, S.S. !+rospace Engineerirs 1963 M.S. Control Engineertng 1965 Sdn Otago State University. M.S. Mathematics 1970 University of Pittsburgh, MBA 19E0.

Hr. Richardson has tsany years of professional and canagement egerience in the nuclear field. He joined the Fressurir.ed Water 74 actor Divisica of Westinghouse in 1972 where he managed the Reac.cr Protection Analysis Group for performing nuclear plant sarety analysis and, most recencly, has canaged the Risk Assess =en: Technology Organization.

Prior to this, Mr. Richardson was with Gulf General Atocic where he wrked on design of contrcl er,d safety systecs for the gas-cooled nuclear plant. M Westinghouse, he has participated in and dincted a number of risk assessment and safety analysis stuaies for a wide variety of applications.

He was a frincipal innstigator in both the Zion Sta-tion and Indian Pofnt Station Reactor Safety Studies.- He directed tne PRA studies for the Westingnouse Owners Group that addmssed the

- Post-TMI NUREG mquirecents on emergency procedures and operator display regut me,ents. Mr. Ricnardson was technical and program manager for the British (NMC) Referurre Water Reactor Safety Stuay. He has aise led the deveicpment of econocic and financial risk assass=ent techniques for the use in new reactor mcdel design concepts.

Mr. Richardson is a mecber of the IEEE and ANS anc has served on the working groups for t o standards corzittees. He is reviewing.he sec-tions for the PRA ranual dincted by NRC te be finished in 1981. He fs author or co-author of more than 15 reports and papers dealing with risk assessment and various aspects of nuclear plant design t

W W W W 9r sir

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NAME.

DEMN:S C. ELEY EOUCATION t.0., nuclear Reac::r Engineering, Massachusa::s inszi:uze of Tecnnelogy,1379.

C:urses in. nuclear engineering and ccm: uter science, C:rnell Universi y, 197E-L97A.

U.S. Navy Nucl ear Power School, - 1968.

University of Cincinna:f, 5.5.E.E.,1967.

Courses.in Ma:nematics and Physics, Centre College of Xantucky, 1951-1965.

-PRCF_35!0NAL EXMRIEHCE

-General Su= mary A consul tant at ? f ckard, Lowe & Garrick, Inc.,1379-present.

Technical analysis of ocwer plant availacility and risk.

C:s:-benefit analysis Of cover clant system changes.

Preparation of technical recorts, excert tastimony, and proposals.

Supervision cf the tecnnical quality of PLG reports and 'direc-icn of some PLG projects.

nstrue:Or at availacility, risk, and decision analysis c:urses effered by PLG.

Oy sta r ~ Creek Probabilistic Ri;k Assessment (OPSA).

Assistad in the cc=cle:icn and review of this c:=pleta risk assessment of an ocerating 5WR ;erf:r=ed for l Jersey Central Pcwer & Lign:.

Work Order Scheculing System (WOSS).

Assisted in develocing the San Onofre 2 and 3 plant-model for a computer

.mased work order p-icriticing, scheduling, and rec:rd keeping system for Southern California Edision Company.

Steam Turoine Diagnostics Cost-Benef t: Analysis.

Develoced and applied a procedure for evaluating diagnostic alternatives for E?RI.

Reliabili:y Analysis Of Diablo Canyon Auxiliary Feecwatar System for Pacific Gas & Electric.

Midland Pl an:

Auxiliary Feed 4atar System Reliaoflity Analysis for C:nsumers Pcwer.

- Technical Review of-the " Office of Emergency Services Recommended Emergency Planning Ione Considera:icns..." for Sou nern Calif:enia l

. Ecison.

Prioritization of NRC Action Plan for NSAC.

Devel:: ment of a me:nedelogy and participation in an AIF workshop to apply it' fer

- E?RI/NSAC.

Iien 2nd Indian Point Precabilistic Safety Studies.

Methods development, systams analysis, and plant modeling.

0:ner ? ras--LaSall e, 3r:wns Fer y, Midland, Pilgrim 1, and Ocones.

On USS Enterprise, Reactor Training Assistant, 5 months,1971.

Responsible for technical training of approximitaly a00 nuclear trained officers and men prior to annual safeguards examination.

Preculsion Plant Station Officer, 9 months, 1970-1971.

Responsible for maintenance and operation of one propulsion plant (:we reactors, eight steam generators, and asscciated eculpment) during ;cwer range testing of new i

l reactors and during deployment.

Approximately 50 enlistad perscnnel were assigned to the pl ant.

Shift Pr:pulsien Plan: Watch Officer,15 mentns, 1369-1970.

Sucervised a crew cf abou: 20 navy enlisted opera: Ors anc many shipyard werkers on 5-hcur snift rotation :encucting maintenance l-mm e e

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t ELEY - 2 anc-esting in one procuision plan during refueling-overhaul.

Shi;bearc cualifica:icns:

Prcpul sion Cuty Officar, responsible - for all prcpulsion ecuisnent during absence of Re.?::ce Officar and Engineer'Officar.

Engineering Officer of tne Waten, opert:fonal wat:n in Cantral Centrol, ressonsible for all propulsion and engineering equi;=en; and wa::h stancers. ~ Propulsion Plan: Wa ch Officer, operational watcn in one prceul sion plant, ' directed and= responsible fer all coerations in :ne

I nt.

K At Cincinnati 3eli, Pl ant staff assistant, 4 mcnths,1967. Werked in central cffica and transmissien group suoplying technical assistance to the line organfration.

Cooperative.:rainee, 3 years,1964-1967, work-study program with alternate three month periods at ne University of Cincinnati.

Chronoloeical Su==ary 1979-Present Consul tant, Pickard, Lowe and Garrick, Inc.

1974-1979 Massachusetts Institute of Technolo;y.

Research assistant for Oeoar: men cf Energy LWR Assessment Project.

Teacning assistant in engineering cf nuclear reactors.

Summer 1976 Northeast Utilities.

Engineer:

economy studies, plant startup, analysis of physics tests.

1967-1974 U.S. Naval Reserve, active duty.

Instrue:cr of naval science, Cornell University, 1971-1974; Reactor Department of USS Enterprise, deployment and refueling-overhaul, 1969-1971; Nuclear ?cvtr training program and Officer Candidata School, 1967-1969.

'1964-1967

' Cincinnati Bell.

Plant staff assistant and werk-stucy program trainee.

MEMSERSHIPS, LICENSIS, AND MONORS 3-The Sectety for Risk Assessment.

Instituta of Electrical and Electronics Engineers.

American Nuclear Society.

American Association for the Acvancemen: cf Science.

The New York Academy of Sciences.

U.S. Naval Reserve, Cc=mander.

-Regis:ared Nuclear Engineer, 5:sta cf California.

m 5 LEY - 2 Si;=a.Xi-(national science honors socie:y),1975.

Sherman R. Xnaco Fellcwsnip (Nortneas: Utilities), 1975-1975.

Sl oan Research.Traineesnip,137 *-1975.

Ita Xappa Nu (national electrical engineering honors socie y),1957.

RE? ORTS AND PUBLICATIONS

'"Seabroen Procabilistic Safety Assessment," Puolic Service Company of e

. New Hampsnire,.:s be puolisned in -1383.

Pf ckard, Lowe and Garrick, Inc., "Midlanc Procabfi f stic Risk Assessment,"

Consumers-?cwer Company, to de published in 1982.

.0conee Probabilistic Risk Assessment," a joint effort of the Nuclear Safety Analysis Center, Duke ?cwer, and otner participating utilities, :o be published in 1382.

Tennesse: Valley Authority and'Pickard, Lowe and Garrick, ~nc., "Br:wns Ferry Procamilistic Risk Assessment," to be cuolisned in 1952.

Apostolakis,_ G., M. Xa:arians, and D. C. Bley, "A Me:nodology for Assessing the-Risk from Caole Fires," accepted for puolication in Nuclear Sa fe ty, 1932.

Kaplan, S., H. F. Perl a, and D. C. Bl ey, "A Methodology for Seis=ic

. Safety ' Analysis,of Nuclear Power Plants," ;rocosed presentation at : e International Maeting on Thermal Nuclear Reactor Safety,

' Chicago, Illinois', August 29-September 2, '.982.

31cy,.D. C., S. Kaolan, and 3. J. Garrick, " Assembling and Decomposing PRA Results:

A Matrix FormaTism," preocsec presentation at the International Meeting on Ther=al Nuclear Reactor Safety, Chicago, Illinois, August 29-imptemoer 2,1982.

Garrick, 5. J., S. Kaplan, and D. C Bley, "Recent Advances in Probabilistic Risk Assessment," prepared for tr.e MIL Nuclear Pcwer Reactor Safety Course, Camoridge, Massacnusetts, July 19, 1982.

Fleming, K. N., S. Xaplan, and 3. J. Garrick, "Seabrook ?r:babilistic Safety Assess:ent Management Plan,"PLG-0239, June 1982.

Garrick, S.

J., " Lessons Learned From First Generation Nuclear Flan:

Probaciitstic Risk Assessments," to be presented at the Workshoo on Low-?robacility/High-Consequence Risk Analysis, Arling:cn, Virginia, June 15-17, 1952.

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- 3 LEY- - 4 3arrick, 3. J., - 5. Xaol an, D. C. Icen, E. 5. Cl evel and, H. F. perl a,

3. C. 31ey, 3. W. Stillwell, H. ~Y. Senneider, and G. A ost:laxis, "?:wer

?l an: Availacility E71neering:

Me nods of Analysis, ?rcerain Planning, and Acclica:icas," EH.I NP-2155, FLG-0155, May 1932.

~

Bl ey, D. C., and R. J. Mulvihill, "C:=nents en Evaluation Of Avafiabili v

.I.wrovemen: 0;;iens for Moss Landing Units 5 and 7," ?LG-0225,

~

Mar:n 1952..

5:111well, 0.. W., G. Acos:sl akis, D. C. Bl ey, ?. H. Raabe, R. J. Mulvihill, S. K4:lan, and 5. J. Garrick, "III Availabilit~y Handbeck," ?LG-0213, January 1982.

Bl ey, 0.- C., L. G. H. Sarmanian, and D. W. Stillwell, "Reliaoility Analysis of Safety Injection System Modification,. San Onofre Nuclear Generating Statica - Unit.1,".?LG-0205, Oc :cer 1951.

" Zion Pr:bacilistic Safety Study," Coatonwealth Edison C:mpany,

' September 1931.

Buttemer, 0. R., " Analysis of Postulated Accidents During Lew ?cwer Testing at the San Cncfre Nuclear Generating Stati:n--Uni 2," ?LC--019 9, Septa =ber 1981.

El ey, D. C., 0. W. Stillwell, and R. R. Fray, "Reliacility Analysis cf Diablo Canyon Auxiliary Feedwater System," presented at the Tenth Biennial Tecical Conferener on Reacter Operating Experience, Cleveland, Chic, Augus: 17-19,:1981.

Garrick, S. J., and O. C. Eley, " Lessons Learned fr:m Cur ent PRAs,"

cresented to ne ACRS Subc:=mittee en Reliacility and Prceabilistic Risk Assessment, Los Angeles, California, July 23, 1981.

Xapl an, S., G. Acest:lakis, 3. J. Garrick, D. C. 31ey, and K. Woodard,

" Method =1cgy for Probacilisti: Risk Assessment of Nuclear ?:wer Plants,"

draft version rf a book in preparation, ?LG-0209, June 1951.

?erla, H. F., "Proj ect Pl an:

Pr:babilistic Risk Assessment, Midland Nuclear Pcwer Flant," PLG-01EO, May 1981.

El ey, D. C., C. L. Cata, 0. W. Stillwell, and B. J. Garrick, "Midl and Plant Auxiliary Feecwater System Reif ability Analysis Synopsis,"

?LC-0155, March 1981.

Pickard, Lowe anc Garri:k, Inc., " A Me-hedclogy :: Quantify Uncertainty of Cos: of Electricity for Alternate Casigns of (0:=bustion) Turcine C =sined Cycle F1 ants," ?LG-0152, Mar:h 1981.

e

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3 LEY - 5 Garrick, 3.. J ', S. Ahmed, and O. C. 31ey, "A Me:nodolcEy for Evaluating ne Costs anc 3enefi:1 of Power Plan: Otagnostic Tecnnicues," succi :ed

for cresenta icn a ne Ninth Turocmacninery Symcosium, Hous

ca, Texas, Decemoer 9-11, 1980, Pickard, ;Lowe and Garricx, Inc., "$sminar:

Procacilistic Risk Assessmen:

cf Nuclear ?cwer Plants," PLG-0154, Novemcer 1980.

Pickard, Lowe and Garrick, Inc., "Pr0 ject Pl an:

Precabilistic Rick Assess =ent, Erewns Ferry Nuclear Plan: Uni: 1," PLG-0149, Oc:ccer 1980.

Garrick, 3. J., S. Kapian, D. C. Iden, E. 3. Cl eveland, H. F. Perl a, D. C. Bley, and O. W. Stillwell, "Pcwer Plant Availac411ty Engineering,

'Methocs of Analysis - Program Planning - Applications," 2 Vols.,

PLG-0148. Octccer 1980.

Bl ey, D. C., C. L. Cate, D. W. Stillwell, and 3. J. Garrick, "Midl and

?lant Auxiliary Feecwater System Reliability Analysis," PLG-0147,

-Cc ccer 1980.

131ey, O C., D. M. Wheeler, C. L. Cate, D. W. Stillwell, and

3. J. Garrick, " Reliability Analysis of 7taols Canyon Aux'liary Feecwater.

System," PLG-0140, Septemoer 1980.

Garrick, 3. J., et al, " Project Plan: Oconee Probacilistic Risk As sessment," PLG-0128,.. August 1980.

Garrick, 3. J., D. M. Wheel er, I. 3. Cl evel and, D. C. Bl ey,

L. M. Reichers, and C. 3. Morrison, "Ocerating Experience of LarGe U.S. Steam Turbine-Genera:crs; Volume 1 - Data, 'Yolume 2 - Utility

. Di rec: cry," PLG-0134, June 1980.

Garrick, 3. J., E. Kaplan, anc D. C. Bl ey, " Seminar:

2 ewer Plan:

Precabilistic Risk Assessment and Reliaofif ty," PLG-0127, May 1950.

Garrick, 3. J., and S. Xaclan, "0yster Creek Precacilistic Safety Anal-ysis (OPSA)," presented at the ANS-ENS Topical Meeting on Thermal Reac:ce l

Safety, Xnexville, Tennes.see, April 8-11, 1900.

Garrick, 3. J.,. S. Xaclan,

. E. Apostolakis, D. C. 31ey, anc T. E. Petter, " Seminar: Prcbabilistic Risk Assessmen; as Applied to Nuclear Pcwer Plants," PLG-0124, March 1990.

Garrick, 3. J., S. Ahmed, and C. C. Bl ey, "A Methodelegy for Ivaluating the Cos:s.and Benefits of Pcwer Plan: Dia9acs:ic Techniques," PLG-0115.

J anuary 1930.

ar I

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lN m h v n.$ 7.,..

3LIY - 5 Kasi an, S., 3. J. Garriek', and O'. C. El ey, "Ne tes on Ri sx, Pr:cac tiiw' and vecisten," PLG-0113, lievemcer 1979 l

31 ey, D. C., C. ' L. Ca ta, D. C. Iden, 3. J. Ga rri ck, anc J. M. Mucson,

- Sei smic Safe.y Margins Research Program (Phase I), Prcjec: yI: - Sys aas

^n aly si s,

Pla-0110, Sectammer 1979.

. Ca:e,.C. L., and 3. J. Garricx, "W-501 C:moustion ursi.9e Startine Reliacilicy Analysis,' PLG-0103, Juns 1979.

~

Pickard, Lewe and Garrick, !nc., " Plan Availacility Program Saecifica-tion, dan Onofre Nuclear Generating Station," Maren 1979.

.Pickard, Lewe and Garrick, Inc., " Work Order Scheduline Systam, Cesign Specification," March 1979. -

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