Text

Summary of 890816 Meeting W/Numarc,Epri & Other Industry Representatives Re Applicability of 10CFR50.59 & NRC Interpretation of Phrases That Define Which Changes,Tests & Experiments Involve Unreviewed Safety Question
	ML19327A691
Person / Time
Issue date:	09/06/1989
From:	Fischer D; Office of Nuclear Reactor Regulation
To:	Calvo J; Office of Nuclear Reactor Regulation
References
	NUDOCS 8910040373
	Download: ML19327A691 (62)
	v • d • e

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1 MEMORANDUM FOR:

Jose A. Calvo, Chief I

Technical Specifications Branch Division of Operational Events Assessment, NRR I

l I

FROM David C.

Fischer, Chief Special Projects Section

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Technical Specifications Branch j

SUBJECT MINUTES OF WORKSHOP ON 10 CFR 50.59 on August 16, 1989, the NRC staff met with NUMARC, EPRI, and other industry representatives to participate in a workshop on 10 CFR 50.59.

The purpose of the workshop was to discuss the applicability of 10 CFR 50.59 and the NRC staff's interpretation i

of the phrases in 10 CFR 50.59 that define which changes, tests, and experiments involve an unreviewed safety question.

During the workshop, industry described NSAC-125, " Guidelines for

'l 10 CFR 50.59 Safety Evaluations," an industry developed guidance document that discusses the process of performing safety reviews j

for facility changes, tests or experiments at nuclear power plants.

Industry representatives also discussed the use of screening I

criteria to limit the 10 CFR 50.59 safety evaluation documentation i

process to activities consistent with the intent of 10 CFR'50.59.

j i

In addition, a representative from Westinghouse gave participants 1

a Nuclear Steam Supply System vendor's perspective of 10 CFR 50.59 and NSAC-125 and industry led a

discussion of the

terms,

)

" consequences" and " margin of safety" as used in these documents.

j Original Signed by David C. fischer David C.

Fischer, Chief Special Projects Section Technical Specifications Branch

Enclosures:

List of Attendees NSAC-125 i

Presentation slides J

DISTRIBUTION:

Please see attached (50.59M. DOC) l 7WD OTSB:DOEA:NRR OTS :DOEA:NRR NVGilles DCFischer 09/(.4 /89 09/6 /89 I

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DISTRIBUTION:

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.w/ enclosures:

j PDR i

Central Files w/o enclosures TEMurley GMHolahan OTSB Members JHSnietek CHBerlinger 0T83 R/F 1

FJMiraglia CJHaughney DOEA R/F l

DMCrutchfield JLieberman Regional Adminiatrators SAVarga BDLiaw NRC Participants CERossi JWRoe GCLainas Eh7ordan 1

JFCongol CIGrimes i

FGillespie WGKennedy

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BKGrimes MGMalsch i

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is MEETING ATTENDEES M&ME AFFILIATION J. N. Sniezek NRR J. A. Calvo OTSS/DOEA/NRR D.

C. Fischer OTSB/DOEA/NRR l

N. V. Gilles OTSB/DOEA/NRR B. Clayton OEDO/NRR J. Norris PDII-2/DRP/NRR D. Brinkman PDI-1/DRP/NRR R. Hernan PDI-4/DRP/NRR D. Hickman PDV/DRSP/NRR A. Bournia PDIV/DRSP/NRR L. Kinter PDII-1/DRSP/NRR D. Pickett PDIV/DRSP/NRR P. J.-Noonan PDIV/DRSP/NRR E. B. Tomlinson PDIV/DRSP/NRR D. L. Nigginton PDIV/DRSP/NRR W. A. Paulson PDIV/DRSP/NRR G. Klingler ILPB/PMAS/NRR E. Butcher ILPB/PMAS/NRR R. Pulsifer PDIII-1/DRSP/NRR R. Martin PDI-2/DRP/NRR E. Leeds PDI-3/DRP/NRR D. Hood PDII-3/DRP/NRR i

J.

B. Hopkins PDII-3/DRP/NRR P. W. O'Connor PDIV/DRSP/NRR l

T. V. Wambach PDIII-3/DRSP/NRR l

K. N..Tabbour PDII-3/DRP/NRR L

G. - C. Lainas ADR2/NRR l

C. Poslusny PDIV/DRSP/NRR t

B. Siegel PDIII-2/DRSP/NRR i

T. Alexion PDIII-3/DRSP/NRR T. M. Ross PDIII-2/DRSP/NRR L. Earr DRIS/NRR S. A. Reynolds PDV/DRSP/NRR M. Fleishman DRA/RES t

J. Wing PRAB/DREP/NRR J. Stone PDI-2/DRP/NRR i

t T. Reed-PDII-3/DRP/NRR S. Hoffman PDII-2/DRP/NRR W. Hodges SRXB/ DEST /NRR W. Troskoski OE C. J. Haughney DOEA/NRR D. Langford PDI-1/DRP/NRR F. Burrows SELB/ DEST /NRR T. Colburn PDIII-3/DRSP/NRR G. Dick PDIV/DRSP/NRR E. Trottier PDI-3/DRP/NRR G.

Vissing PDI-4/DRP/NRR T.

O. Martin Region III J. Durr Region I

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MEETING ATTENDRES s

R&tg AFFILIATI0H J. H. Sniesek WRR J. A. Calvo OTSB/D0EA/NRR D. C. Fischer OTSB/DOEA/NRR N. V. Gilles OTSB/DOEA/NRR B. Clayton OED0/NRR J. Norris PDII-2/DRP/NRR i

D. Brinkman PDI-1/DRP/NRR

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R. Hernan PDI-4/DRP/NRR l

D. Hickman

' PDV/DRSP/NRR l

A. Bournia PDIV/DRSP/NRR L. Kinter PDII-1/DRSP/NRR I

D. Pickett PDIV/DRSP/NRR

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P. J. Noonan PDIV/DRSP/NRR

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E. B. Tomlinson PDIV/DRSP/NRR D. L. Wigginton PDIV/DRSP/NRR W. A. Paulson PDIV/DRSP/NRR j

G.

Klingler ILPB/PMAS/NRR i

E. Butcher ILPB/PMAS/NRR I

R. Pulsifer PDIII-1/DRSP/NRR R. Martin PDI-2/DRP/NRR E. Leeds PDI-3/DRP/NRR

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D. HoM PDII-3/DRP/NRR J. B. HoPkins PDII-3/DRP/NRR i

P. W. O'Connor PDIV/DRSP/NRR j

T. V. Wambach PDIII-3/DRSP/NRR K. N. Jabbour PDII-3/DRP/NRR G.

C. Lainas ADR2/NRR j

C. Poslusny PDIV/DRSP/NRR B. Siegel PDIII-2/DRSP/NRR

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T. Alexion PDIII-3/DRSP/NRR T. M. Ross PDIII-2/DRSP/NRR l

L. Zerr DRIS/NRR l

S. A. Reynolds PDV/DRSP/NRR M. Fleishman DRA/RES l

J. Wing PRAB/DREP/NRR J. Stone PDI-2/DRP/NRR l

T. Reed l'

PDII-3/DRP/NRR t

8. Hoffman PDII-2/DRP/NRR l

W. Hodges SRXB/ DEST /NRR W. Troskoski OE C. J. Haughney DOEA/NRR l

D. Langford PDI-1/DRP/NRR B

F. Burrows SELB/ DEST /NRR T. Colburn PDIII-3/DRSP/NRR G. Dick PDIV/DRSP/NRR i

E. Trottier PDI-3/DRP/NRR l

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BACKGROUND d

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NSAC-125 L

GUIDELINES FOR 10CFR50.59 L

SAFETY EVALUATIONS l

1 NRC Seminar i

August 16,1989 I

W. B. Reuland i

f NSAC

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Chernobyl story

. Committee respect for the concerns e

facing the regions j

Originated as an educational,not enforcenient tool i

l Relationship to safety e

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Use of Guidelines l

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. The Guldelines are designed to i

accomplish three things:

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1. Help the nuclear plant determine when 50.59 applies

2. Help improve the the review process and documentation 1

3. Help narrow the threshold definitions for unreviewed safety questions

. Guidelines are not likely to settle all differences of opinion

. The "S" word

. The need to test for reasonableness l

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.lu NSAC-105

. Basis for getting involved in 50.59 r

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. Discusses the safety and control of design changes

. Used by many utilities, NRC and INPO l

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1987 Draft i

. Used process parameters as i

L consequences l

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. Allowed for small increases over bounding SAR calculated values i

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. Margin of safety based on safety limits to failure o

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. Probabilities based on accident i

categories

. Approximately 40 industry comment i

letters i

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. Consequences confined to radiological l

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. Margin based on an acceptance limit identified as that value reviewed and approved by NRC l

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. Increased input from NSSS Vendors i

36 Industry Comment Letters 1

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4 NRC Mtgs L

. First meeting with Jim Taylor and others led to agreement that NRC would give guidance to an industry initiative on 50.59 L

. Second meeting with Bill Russel provided further guidance and continuity following NRC reorganization i

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. Met with Sam Bryan committee June l

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1988 and discussed May 88 letter l

. Met with Ernie Rossi-discussed draft of l

Dec 1988 and May 1989 NRC letter o

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. May 10,1989 Letter to Tom Tipton l

. June Letter to Ernie Rossi

. Workshop participation and conclusions i

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TREATMENT OF UNREVIEWED SAFETY CUESTIONS IN PART 50.59 REVIEWS 9

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PRESENTED BY i

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DAVID FISCHER I;-

TECHNICAL SPECIFICATIONS BRANCH p

DIVISION OF OPERATIONAL EVENTS ASSESSMENT i

0FFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY C0FMISSION 1

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10 CFR 50.59 GUIDELINES WORKSHOP i'

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d SCFE E APPlICARILITY E 10 CFR 50.59 10 CFR 50.71 EQUIES THAT TE SAFETY ANALYSIS EPORT E (FDATED TE UPDATED FSAR SHG1D E CGESISTDIT WITH TE LICENSING BASIS (NOT ALL ASPECTS E TE LICENSING BASIS AE TO E IN00W0 RATED INTD TE FSAR) j LICENSING EASIS D001FENTS LICENSEE-CEERATED ISC-GEERATED 1

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RESPONSES TO NRC GENERIC LETTERS

ORDERS AND BULLETINS ENVIRONMENTAL REPORT

REGULATIONS OTHER (E.G., SECURITY PLAN, ANTITRUST

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ENVIRONENTAL ASSESSMENTS 4

FINAL ENVIRONMENTAL l

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i CHANGES THAT INVOLVE AN !CTUAL Ell)CTIGE IN MARGIN E SAFETY ftJST ET.

PRIOR NRC APPR(NAL SHOULD BE EXPLICITLY DEFIED /AERESSED IN TE TECWICAL SPECIFICATIGtS' i

BASES (IF NOT IN BASES TEN CDISULT LI&NSING BASIS) l IT MAY E SlFFICIENT TO DETEIMIE TE DIECTIGt0F TE CHANGE IN PRRGIN BASE DECISION ON PHYSICAL PARAPETERS OR CGOITIGIS WICH CAN E (BSERVED OR CALCULATED TE EY TO DEFINING TE MARGIN E SAFETY IS IDENTIFYING TE ACCEPTANCE LIMIT Ag casomoeAs l

EXAMPLES OF EASES THAT EXPLICITLY ADDRESS MARGINT OF SAFETY FUEL DESIGN LIMITS REACTOR COOLANT SYSTEM DESIGN PRESSURE PROVIDED CHANGES ARE MADE CONSISTENT WITH LICENSING EASIS-(METHODS AND SPECIFIC ACCEPTANCE CONDITIONS, CRITERIA, LIMITS, ETC.)

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PLANS AND SCHEDULES l-GOAL:

ESTABLISH A PROCESS THAT ENHANCES SAFETY REVIEWS 4

NUMARC/NSAC ISSUE GUIDANCE DOCUMENT-(NSAC-125):

JUNE 1989 TRAIL PERIOD (APPR0X.-6 MONTHS) TO GAIN EXPERIENCE WITH j

GUIDANCE DOCUMENT:

JULY 1989 TO l

JANUARY 1990-j_

STAFF PLANS TO FORMALLY ENDORSE / ISSUE GUIDANCE:

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WHAT ARE " CHANGES TO PROCEDURES AS DESCRIBED IN THE SAR"?-

WHAT ARE " TESTS OR EXPERIMENTS NOT DESCRIBED IN THE SAR"?

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7 10CFR50.59(A)(1) STATES THAT, 1 *.

"THE HOLDER.0F A LICENSE AUTHORIZING OPERATION OF A PRODUCTION L

OR. UTILIZATION FACILITY MAY (I) MAKE CHANGES IN THE FACILITY AS p

~ DESCRIBED IN THE SAFETY: ANALYSIS' REPORT, (II). MAKE CHANGES IN' l

lJ THE PROCEDURES AS DESCRIBED IN THE SAFETY ANALYSIS REPORT AND

-(III) CONDUCT TESTS OR EXPERIMENTS NOT DESCRIBED IN THE SAFETY ANALYSIS REPORT,' WITHOUT PRIOR COPEISSION APPROVAL, UNLESS.'THE PROPOSED CHANGE, TEST OR EXPERIMENT INVOLVES A CHANGE IN THE TECHNICAL SPECIFICATIONS INCORPORATED IN THE LICENSE OR AN

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UPDATED FSAR SUBMITTED BY THE LICENSE TO THE NRC.AS

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10CFR50.71(E) REQUIRES THAT THE FSAR BE REVISED TO INCLUDE j

- THE EFFECTS OF CHANGES, SAFETY EVALUATIONS, AND. ANALYSES l

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WHAT IS A " CHANGE TO THE FACILITY AS DESCRIBEQ'IN THE SAR"?

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e 10CFR50.69 IS CONCERNED WITH CHANGES WHICH' AFFECT THE DESIGN, FUNCTION, OR METHOD OF PERFORMING THE FUNCTION OF

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A STRUCTURE, SYSTEM, OR COMPONENT (SSC'S) DESCRIBED IN THE SAR BY TEXT OR DRAWING.

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CHANGES TO SSC'S NOT EXPLICITLY DESCRIBED IN THE SAR ARE

-INCLUDED'IF THE CHANGE HAS THE POTENTIAL FOR AFFECTING THE FUNCTION OF SSC'S WHICH ARE EXPLICITLY DESCRIBED IN THE SAR.

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TEMPORARY ALTERATIONS ARE INCLUDED.

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e PREVIOUSLY EVALUATED TESTS DO NOT REQUIRE SAFETY EVALUATIONS EVERY TIME A TEST IS PREFORMED.

EXAMPLES L

INCLUDE PREOPERATIONAL TESTS, STARTUP TESTS, AND PERIODIC L

SURVEILLANCE TESTS.

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LIMITING THOSE ITEMS TO BE EVALUATED ON THE BASIS OF CLASSIFICATION SUCH AS IMPORTANT TO SAFETY OR SAFETY I

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(3.10) e NON-SAFETY RELATED SYSTEMS ARE NOT EXCLUDED FROM THE SCOPE OF 10CFR50.59 (2).

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(2) o CHANGES TO NON-SAFETY RELATED EQUIPMENT NOT DESCRIBED IN l '

THE SAR CAN INDIRECTLY AFFECT THE ABILITY OF EQUIPMENT i

IMPORTANT TO SAFETY TO PERFORM ITS INTENDED FUNCTION.

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PROTECTION, HIGH ENERGY LINE BREAK, AND MASONRY BLOCK WALLS ARE SOME OF THE AREAS IN WHICH r!!LNGES TO NON-SAFETY

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EXAMPLES INCLUDE PLANT MODIFICATIONS, TEMPORARY ALTERATIONS, " MINOR MODIFICATIONS", " ENGINEERED WORK REQUESTS," AND FUEL i

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1NESE MECHANISMS OFTEN HAVE BEEN DEVELOPED BY DIFFERENT UTILITY. DEPARTMENTS AT DIFFERENT POINTS IN TIME AND MAY CONTAIN CONFLICTING GUIDANCE REGARDING SAFETY EVALUATIONS.

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e EACH MECHANISM THAT HAS THE POTENTIAL FOR CHANGING THE FACILITY AS DESCRIBED IN THE SAR SHOULD BE CHECKED TO ENSURE THE FOLLOWING:

c.

I 1)

THAT IT INCLUDES CONSIDERATION OF THE NEED TO PERFORM A 50.59 SAFETY EVALUATION.

L 2)

THAT IT PROVIDE GUIDANCE FOR PERFORMING SAFETY i

EVALUATIONS THAT MEETS INDUSTRY AND NRC EXPECTATIONS.

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FOR OLDER VINTAGE PLANTS, RELATIVELY FEW PROCEDURES ARE DESCRIBED IN THE SAR.

HOWEVER, IN RESPONSE TO CHANGING EXPECTATIONS, THE.NLMBER OF PROCEDURES IH USE AT THE PLANT HAS GREATLY INCREASED.

L e

THE COMPLETE LIST OF PROCEDURES IN USE AT THE PLANT CAN BE j

L PRE-REVIEWED TO IDENTIFY CERTAIN PROCEDURES WHICH WILL NOT 1

L REQUIRE A SAFETY EVALUATION FOR EVERY CHANGE.

EXAMPLES i

- INCLUDE CERTAIN CHEMISTRY PROCEDURES, CERTAIN ADMINISTRATIVE PROCEDURES, ETC.

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l REQUIREMENT STATING WHICH PROCEDURES REQUIRE ON-SITE

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REVIEW.

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NSAC 105 PROVIDES ADDITIONAL GUIDANCE.

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10CFR50.59 GUIDANCE PRINCIPLES i

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INTEGRATED APPROACH To AssunE COMPATABILITY / CONSISTENCY WITH REGULATIONS Section Analysis Notification Reporting Record Applying for l

or Keeping Amendment Evaluation i

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CRITERIA, AND LIMITS ARE j

i DEFINED (INCLUDING MODELS, TESTS,-UNCERTANITES,--PENALTIES, l

METHODOLOGY, ETC.)

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ANALYSIS PERFORMED CONSISTENT WITH #1 & #2 TO MEET #3 i

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F t ;.. v, f1 10CFR50.59 GUIDANCE PRINCIPLES

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'FOR CHANGES T0 QUANTITATIVE ACCEPTANCE L

CRITERIA-AND LIMITS, THE NRC HAS PREVIOUSLY L

GRANTED EITHER GENERIC OR CASE-SPECIFIC APPROVAL.

THUS, PROVIDED THAT'THE FOUR CONDITIONS ARE MET, THERE IS N0 UNREVIEWED l

SAFETY QUESTION INVOLVED.

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s 10CFR50.59 GUIDANCE PRINCIPLES i

ACCEPTANCE LIMIT CATEGORIES 2.

FOR CNANGES WHICH1ARE COMPARED T0 QUALITATIVE l

ACCEPTANCE CRITERIA.0R LIMITS (E.G. WELL l

BELOW 10CFR100 GUIDELINES)'IT MAY BE NECESSARY.TO REQUEST PRIOR NRC REVIEW AND APPROVAL FOR ANY CHANGE.IN A NON-CONSERVATIVE i

DIRECTION UNLESS THE NRC HAS GRANTED PRIOR APPROVAL.0N EITHER A CASE-SPECIFIC OR GENERIC BASIS.

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10CFR50.59 GUIDANCE-1 i

PRINCIPLES 1

~~

' ACCEPTANCE LIMIT CATEGORIES l

SOME UTILITIES HAVE ESTABLISHED CRITERIA ACCEPTABLE TO THE NRC-T0 DEFINE WHAT CHANGES WOULD NOT INVOLVE AN UNREVIEWED SAFETY QUESTION.

IT SHOULD BE NOTED,THAT PRIOR ACCEPTANCE ON.

ANOTHER DOCKET IS NOT: GENERIC UNLESS THERE IS EXPLICIT NRC STATEMENTS TO THAT EFFECT.

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15.4.8.1 RUPTURE'0F A CONTROL ROD DRIVE MECHANISM HOUSING (rod CLUSTER CONTROL ASSEMBLY EJECTION) i THE MECHANICAL FAILURE OF A CONTROL ROD MECHANISM i

PRESSURE HOUSING WOULD RESULT IN THE EJECTION OF i

A ROD CLUSTER CONTROL. ASSEMBLY.

FOR ASSEMBLIES INITIALLY INSERTED, THE CONSEQUENCES WOULD BE A J

RAPID REACTIVITY INSERTION, TOGETHER WITH AN ADVERSE CORE POWER DISTRIBUTION, POSSIBLY LEADING TO LOCALIZED FUEL ROD DAMAGE.

ALTHOUGH MECHANICAL PROVISIONS MAKE THIS ACCIDENT EXTREMELY UNLIKELY, THE APPLICANT HAS ANALYZED

.THE CONSEQUENCES OF SUCH AN EVENT.

METHODS USED'IN THE ANALYSIS ARE REPORTED IN WCAP-7588, REVISION 2, "AN EVALUATION OF THE ROD EJECTION ACCIDENT IN WESTIN'GHOUSE. REACTORS USING SPATIAL KINETICS METHODS," WHICH HAS BEEN REVIEWED AND ACCEPTED BY THE. STAFF IN A LETTER DATED AUGUST 28, 1973.

THIS REPORT DEMONSTRATED t

l THAT THE MODEL USED IN THE ACCIDENT ANALYSIS IS CONSERVATIVE RELATIVE TO A THREE-DIMENSIONAL KINETICS CALCULATION.

THE APPLICANT'S CRITERIA FOR GROSS DAMAGE OF FUEL ARE A MAXIMUM CLAD TEMPERATURE 27000F AND AN i

1 ENERGY DEPOSTION OF 200 On 225 CALORIES PER GRAM IN THE HOTTEST PELLET.

THESE CRITERIA ARE MORE CONSERVATIVE

THAN THOSE PROPOSED IN RG 1.77,

" ASSUMPTIONS USED FOR EVALUATING A CONTROL ROD

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EJECTION ACCIDENT FOR PRESSURIZED WATER REACTOR."

THEREFORE, THEY ARE ACCEPTABLE.

l L

RG 1.77 HAS AN ACCEPTANCE CRITERION OF 280 CALORIES PER GRAM ENERGY DEPOSITION AND NO CRITERION FOR CLAD TEMPERATURE OTHER THAT THAT IMPLICIT IN REQUIREMENTS FOR FUEL AND PRESSURE VESSEL DAMAGE.

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L' FOUR CASES WERE ANALYZED:

BEGINNING-OF-CYCLE AT' 102% AND ZERO POWERAND END-0F-CYCLE AT 102% AND L

'ZERO POWER.

THE HIGHEST CLAD TEMPERATURE, 24220F, AND THE HIGHEST FUEL ENTHALPY,179 CALORIES PER GRAM, WERE REACHED IN THE

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END-OF-CYCLE ZERO-POWER AND BEGINNING-0F-CYCLE FULL-POWER CASES, RESPECTIVELY.

THE ANALYSIS ALSO'SHOWS THAT LESS THAN 10% OF THE' FUEL' EXPERIENCES DN8 AND LESS THAN 10% OF THE HOT PELLET MELTS.

ANALYSES HAVE BEEN PERFORMED TO SHOW THAT THE PRESSURE SURGE PRODUCED BY THE ROD

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EJECTION IS MILD AND WILL NOT APPROACH THE RCS EMERGENCY LIMITS.

FURTHER ANALYSES' HAVE SHOWN i

THAT A CASCADE EFFECT '(THE EJECTION OF A FURTHER ROD BECAUSE OF-THE EJECTION OF THE FIRST ONE) IS NOT CREDIBLE.

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THE STAFF CONCLUDES THAT THE ANALYSIS OF THE ROD EJECTION ACCIDENT IS ACCEPTABLE AND MEETS GDC 28.

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1 LL' 10CFR50.59 GUIDANCE l::

PRINCIPLES I

DNBR i

i CRITERION EXAMPLE L

e AN EXAMPLE IS PROVIDED BY THE DNBR CRITERION CHOSEN AS THE POINT BELOW WHICH CONFIDENCE IN e

L CLADDING. INTEGRITY IS DECREASED L(ASSUMED TO BE 1

L 1.3 IN THIS EXAMPLE).

IF IN THE.SAR THE LICENSEE L

HAD CALCULATED.A DNBR OF l'.9,. AND THE NRC IN ITS 1

- SER HAD' CONCLUDED THAT THIS VALUE WAS ACCEPTABLE i

L "BECAUSE IT IS ABOVE 1.30," THE ACCEP,TANCE' LIMIT-q L

IS 1.3.

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CHANGES IN MINIMUM DNBR SUCH AS FROM 2.0 TO 1.9 L

.0R.1.35: TO 1.3.D0 NOT REPRESENT A REDUCTION IN L

THE MARGIN OF SAFETY.

ALL OF THE ABOVE' CHANGES REFLECT AN EQUAL CONFIDENCE IN THE INTEGRITY:0F THE FUEL CLADDING, IN THAT THE NEW VALUE IS ABOVE h

' THE ACCEPTANCE LIMIT OF 1.3.

J L

HOWEVER,.IF THE SER SIMPLY STATES THAT A DNBR LIMIT OF 1.9 IS ACCEPTABLE WITHOUT PROVIDING ANOTHER LIMIT.AS THE ACCEPTANCE LIMIT, 1.9 IS THE ACCEPTANCE LIMIT.

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10CFR50.59 GUIDANCE PRINCIPLES 9

10CFR100 GUIDELINES I

FOR OTHER ACCIDENT ANALYSES, ACCEPTANCE CONDITIONS ARE BASED ON 10CFR100 DOSE GUIDELINES.

TWO CASES SHOULD BE CONSIDERED:

IF THE ACCEPTANCE LIMIT IS BASED ON AN UNSPECIFIED HARGIN TO THE PART 100 GUIDELINES AND N0 SPECIFIC ACCEPTANCE CONDITIONS HAVE BEEN PRE-APPROVED, THE ACCEPTANCE LIMIT MAY BE DEFINED AS THE DOSE THAT WAS REPORTED IN THE SAR.

IF THE ACCEPTANCE LIMIT IS BASED UPON HEETING SPECIFIC ACCEPTANCE CONDITIONS PRE-APPROVED BY THE NRC, AN INCREASE IN CALCULATED OFF-SITE DOSES RESULTING FROM A CHANGE, TEST OR EXPERIMENT DOES NOT REPRESENT AN INCREASE IN RADIOLOGICAL CONSEQUENCES AS LONG AS THE ACCEF

,NCE LIMIT AND CORRESPONDING ACCEPTANCE CONDITIONS FOR THE ACCIDENT CONTINUE TO BE HET.

017?u: PAL 8/11/89 6

.I, 10CFR50.59 GUIDANCE PRINCIPLES 8

EKAMPLES A CHANGE THAT AFFECTS THE RADIOLOGICAL CONSEQUENCES OF A HAIN STEAMLINE BREAK WITH COINCIDENT IODINE SPIKE FOR THE PLANT.

THE NEW DOSE REMAINS WITHIN THE SRP 15.1.5 APPENDIX A CRITERION FOR THE PLANT (WNERE THE NRC HAS DEFIIED "SMALL FRACTION-0F 10CFR100 LIMIT" AS 30 REM THYROID AND 2.5 REM WHOLE BODY), THEN THE ACCEPTANCE LIMIT IS STILL NET.

WHERE A CHANGE IN CONSEQUENCES IS SO SHALL OR THE UNCERTAINTIES IN DETERMINING WHETHER A CHANGE IN CONSEQUENCES HAS OCCURRED ARE SUCH THAT IT CANNOT BE REASONABLY CONCLUDED THAT THE CONSEQUENCES HAVE ACTUALLY CHANGED (I.E. THERE IS NO CLEAR TREND TOWARDS-INCREASING THE CONSEQUENCES) 33333333333333333333333333333333333333333333333333 THERE IS NO INCREASE IN CONSEQUENCES THAT WOULD INVOLVE AN UNREVIEWED SAFETY QUESTION EEE=EE=E=EsazREEszEEEEEEEEEEEazEEEEEE==bEEEEEEEEEE I

0172W PAL 8/11/89 7

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i EXCERPT FROM SAFETY EVALUATION REPORT i

TORNAD0 HISSILE PROTECTION FOR i

ISOLATION VALVE CUBICLE AUXILIARY _SYSIDiS_ BRANCH l

l I

I.

JNTRODUCTION i

NUCLEAR POWER PLANTS MUST BE DESIGNED TO WITHSTAND THE EFFECTS OF TORNADO AND HIGH WIND GENERATED HISSILES S0 l

AS NOT TO IMPACT THE HEALTH AND SAFETY 0F THE PUBLIC IN i

ACCORDANCE WITH THE REQUIREMENTS OF GENERAL DESIGN j

i CRITERIA 2 AND 4.

THE CURRENT LICENSING CRITERIA l

L GOVERNING TORNAD0 HISSILE PROTECTION ARE CONTAINED IN STANDARD REVIEW PLAN (SRP SECTION 3.5.1.4 AND 3.5.2).

THESE CRITERIA GENERALLY SPECIFY THAT SAFETY-RELATED SYSTEMS BE PROVIDED POSITIVE TORNAD0 HISSILE PROTECTION (BARRIERS) FROM THE MAXIMUM CREDIBLE TORNAD0 THREAT.

HOWEVER, SRP SECTION 3.5.1.4 INCLUDES GUIDANCE ON USE 0F PROBABILISTIC RISK ASSESSMENT (PRA) METHODOLOGY IN LIEU 0F THE DETERMINISTIC APPR0ACH FOR ASSESSING t

TORNADO MISSILE PROTECTION.

THE ACCEPTANCE CRITERION u

IN THIS REGARD IS SIMILAR TO THAT IDENTIFIED IN SRP i

l SECTION 2.2.3 WHICH DEALS WITH IDENTIFICATION OF DESIGN i

BASIS EVENTS USING PROBABILISTIC NETHODS.

THE TORNADO MISSILE ACCEPTANCE CRITERION IS AS F0LLOWS:

l "THE PROBABILITY 0F SIGNIFICANT DAMAGE TO STRUCTURES, SYSTEMS AND COMPONENTS REQUIRED TO PREVENT A RELEASE OF i

RADI0 ACTIVITY IN EXCESS OF 10 CFR PART 100 FOLLOWING A SHALL MISSILESTRIKE, ASSUMING,LOSSOF0FFSITEPOWEg,PER BE LESS THAN OR EQUAL TO A MEDIAN VALUE OF 10-YEAR OR A MEAN VALUE OF 10-6 PER YEAR."...

...THE APPLICANTS ELECTED TO DEMONSTRATE COMPLIANCE WITH THE TORNADO MISSILE PROTECTION CRITERION FOR THE IVCs BY PRA METHODOLOGY RATHER THAN PROVIDE POSITIVE PROTECTION FOR THE ROOF OPENING....

L MEN?'?

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l II. EVALUATION AS PREVIOUSLY STATED, THE (

NAME

- ) PLANT 'IS e

L DESIGNED WITH F0UR SEPARATE IVCs, EACH 0F WHICH IS L

MISSILE PROTECTED FROM ALL SIDES BY HEAVY CONCRETE WALLS EACH CUBICLE IS COMPLETELY PROTECTED EXCEPT FOR THE ROOF WHICH IS OPEN.

THE HEIGHT OF THE IVC WALLS IS i

55 FEET AB0VE PLANT GRADE.

J THE APPLICANTS' PRA CONSIDERED ALL 0F THE SRP SECTION 3.5.1.4, NOVEMBER 24, 1975 MISSILE SPECTRUM A'AS l

P0TENTIAL MISSILES IGCLUDING THE UTILITY POLE AND THE L

AUTOM0 BILE.

REVISION 2 0F THE SRP HOWEVER,. ALLOWS THE L

EXCLUSION 0F THE UTILITY POLE AND THE CAR AT ELEVATIONS UP TO 30 FEET.AB0VE ALL GRADE LEVELS WITHIN 1/2 MILE OF THE FACILITY STRUCTUNES UNDER REVIEW.

AS THE HEIGHT OF l

THE IVC WALL IS 55 FEET AB0VE PLANT GRADE THE MISSILES WHICH WE CONSIDER TO APPLY FROM MISSILE SPECTRUM A ARE THE WOOD PLANK, THE STEEL R0D AND THE STEEL PIPES.

OUR

' EXAMINATION 0F ELEVATED AREAS WITHIN 1/2 NILE OF THE FACILITY. STRUCTURES DISCLOSED ONLY THE DIKE AREA AR0UND THE ULTIMATE HEAT SINK WHICH COULD BE CONSIDERED AS A POSSIBLE LAUNCH P0 INT FOR THE AUTOMOBILE OR THE UTILITY POLE.

THE APPLICANTS HAVE ASSURED US THAT THERE WILL BE N0 UTILITY POLE STORAGE ALONG THE DIKE AREA.

ADDITIONALLY, THE ONLY VEHICULAR TRAFFIC ALONG THE DIKE WOULD BE TRANSIENT IN NATURE IN ORDER TO CONDUCT INSPECTION, AND THIS TRAFFIC WILL BE CONTROLLED.

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IN ORDER FOR A MISSILE TO STRIKE ANY 0F THE COMP 0NENTS IN A GIVEN IVC, IT MUST APPROACH THE ROOF OPENING AT A STEEP ANGLE WITHIN A GIVEN SOLID ANGLE.

THE R00F i

OPENING 0F EACH IVC IS APPR0XIMATELY 745 SQUARE FEET l,

THUS PRESENTING A RELATIVELY SMALL TARGET.

I ADDITIONALLY, THE SAFETY-RELATED TARGET AREAS WITHIN i

THE IVCs ARE MUCH SMALLER THAN THE IVC OPEN ROOF l

l AREAS.

THE FACT THAT THERE ARE FOUR SEPARATE CUBICLES SUBSTANTIALLY DECREASES THE PROBABILITY 0F SINGLE MISSILE BEING CAPABLE OF DAMAGING MORE THAN THE COMPONENTS IN ONE CUBICLE.

MULTIPLE MISSILES HOWEVER, l

l COULD ENTER SEPARATE CUBICLES.

WE CONSIDER THIS A LOW i

PROBABILITY EVENT, AS DISCUSSED FURTHER BELOW....

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0169W: PAL 8/10/89

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.. 00R CONSULTANT'S EVALUATION 0F THE APPLICANTS' PRA CONSIDERED THE VALIDITY AND CONSERVATISM 0F THE APPROACH, ASSUMPTIONS, AND DATA USED IN THE APPLICANTS' ANALYSIS TO ESTABLISH THE PROBABILITY 0F TORNADO AND HURRICANE-B0RNE 4

NISSILE DAMAGE TO THE IVC EQUIPMENT.

ALS0 INCLUDED IN THE EVALUATION IS AN ASSESSMENT 0F THE CORRECTNESS 0F THE RESULTS OBTAINED IN THE STUDY.

- WE HAVE REVIEWED OUR CONSULTANT'S TER AND HIS SUPPLEMENT l

THERET0 CONTAINED IN LETTER DATED (

DATE

)

l WHICH RESOLVED THE OPEN ITEMS IDENTIFIED IN THE TER.

WE CONCUR WITH THE FINDINGS AND RESULTING ESTIMATE OF THE PROBABI j

3.X10gITY0FDAMAGET0ESSENTIALEQUIPMENTINIVC0F WE FURTHER AGREE THAT THIS VALUE IS CORRECT TO l

WITHIN AT LEAST ONE ORDER 0F MAGNITUDE UNCERTAINTITY L

THEREFORE, ADDITIONAL POSITIVE TORNADO MISSILE PROTECTION NEED NOT BE PROVIDED FOR THE IVCs SINCE THE PROBABILITY OF EXCEEDING 10 CFR 100 DOSE CRITERIA DUE TO TORNADO MISSILES IS LESS THAN THE 10-7 PER YEAR ACCEPTANCE CRITERION.

2 BASED ON THE AB0VE, WE CONCLUDE THAT THE APPLICANTS HAVE SATISFACTORILY DEMONSTRATED COMPLIANCE WITH GENERAL DESIGN CRITERIA 2 AND 4 WITH RESPECT TO TORNADO MISSILE PROTECTION i

FOR THE IVCs.

THE DESIGN OF THE IVCs IS THEREFORE ACCEPTABLE WITHOUT THE ADDITION 0F FURTHER PROTECTION FOR THE ROOF AREA.

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CONSEQUENCE AND MARGIN OF SAFETY I

i

" CONSEQUENCES" AND " MARGIN OF SAFETY" INTERPRET THE e

RESULTS OF THE SAR ACCIDENT ANALYSES IN TERMS OF

~

PUBLIC HEALTH AND SAFETY.

i I

THE ACCIDENT ANALYSES ARE THOSE OF CHAPTER 6 AND 15 e

(OR EQUIVALENT CHAPTERS) AND OTHER EVENTS WITH WHICH THE PLANT IS DESIGNED TO COPE AND ARE DESCRIBED IN THE SAR.

l i

I 2-

.. _ _ _.. _. ~ _.,

~

6

[

ACCEPTANCE CRITERI A l

THE SAR, BASED ON LOGIC SIMILAR TO ANSI STANDARDS, e

PROVIDES ACCEPTANCE CRITERIA AND FREQUENCY RELATIONSHIP l

FOR CONDITIONS FOR DESIGN.

THE UNDERLYING OBJECTIVE IS THAT:

e 1-IF THE PLANT IS OPERATED WITHIN THE ASSUMED

[

INITIAL CONDITIONS, AND l

2-IF IMPLEMENTATION SAFEGUARDS OPERATE AS ASSUMED IN SAR, THEN l

i l

THE POSTULATED ACCIDENTS OR CONDITIONS WILL BE CONTAINED I

WITHIN THE APPROPRIATE ACCEPTANCE CRITERIA.

ACCEPTANCE CRITERIA OR LIMITS DEFINE THE LEVEL OF e

DEGRADATION OF (OR CHALLENGE TO) IMPLEMENTED FISSION PRODUCT BARRIERS ALLOWED BY REGULATION.

i

!, l{

i

}

-t i

1-

+

DOSE RELEASES AND MARGINS TO DOSE RELEASES l

FOR ACCIDENTS RESULTING IN DOSE RELEASES, DOSES e

l REPRESENT CONSEQUENCES. CONSEQUENCES (DOSES) ARE l

REGULATED WITHIN ACCEPTABLE LIM!TS.

l FOR ACCIDENTS NOT CAUSING DOSE RELEASES,THEMARGINS TO e

j FAILURE OF FISSION PRODUCT BARRIERS REPRESENT THE MARGIN t

l

-OF SAFETY.

I I

e MARGIN OF SAFETY IN THE BASIS OF ANY TECHNICAL j

SPECIFICATION MAY BE EXPLICITLY IDENTIFIED IN THE T.S.

BASES, OR MAY BE IMPLICIT AS IDENTIFIED IN SAR OR SER.

l ii l

4-l I

I" 1.

-c CONSEQUENCES OF ACCIDENTS OR MALFUNCTIONS OF EQUIPMENT IMPORTANT TO e

SAFETY.

e CONSEQUENCES = DOSE

^

e LIMITS IN SAR OR SER e

DOSE LIMITS se REGULATORY

--10 CFR 100

--10 CFR 20 ee PLANT SPECIFIC (SAR AND/OR SER) ee GENERIC: SER + SRP

^

=

_ ~_

_ q g.

1 L

NO INCREASE IN CONSEQUENCES l

I i

i e

INDETERMINANT CHANGE i

e WITHIN ACCEPTANCE LIMIT 4

i

i i

~i

1a t-L INDETERMINANT CHANGE l

l l

WHERE A CHANGE IN CONSEQUENCES IS SO SMALL OR THE UNCERTAINTIES IN DETERMINING WHETHER A l

CHANGE IN CONSEQUENCES HAS OCCURRED ARE SUCH I

THAT IT CANNOT BE CONCLUDED REASONABLY THAT I

THE CONSEQUENCES HAVE ACTUALLY CHANGED (1.E.,

THERE IS NO CLEAR TREND TOWARD INCREASINGTHE CONSEQUENCES), THE CHANGE NEED NOT BE CONSIDERED AN INCREASE IN CONSEQUENCES.

i 7~

i i

s.

1 ACCEPTANCE LIMIT IS l

THAT VALUE PROPOSED I

BY LINCENSEE IN SAR AS CLARIFIED BY THE SER.

l i

~

.a

. ~ ~

i

[

EXAMPLE 1:

SER STATES: " CALCULATED DOSE IS ACCEPTABLE, THEN l

THE CALCULATED DOSE IS THE LIMIT. (PLANT SPECIFIC LIMIT) i 4

l EXAMPLE 2:

l l

SER STATES: " CALCULATED DOSE IS ACCEPTABLE BECAUSE IT IS BELOW SRP VALUE OF 25 mrem",THEN 25 mrems IS THE-l i

l LIMIT. (PLANT SPECIFIC BOUNDING LIMIT) l l

l s

l l

l i

l

l

l RELATIONSHIP BETWEEN i

ACCIDENT ANALYSIS AND TECH. SPECS.

,t I

e T.S. INITI AL CONDITIONS (CRITERION 2, PROPOSED POLICY l

STATEMENT TECH. SPEC. IMPROVEMENT):

l "THOSE PROCESS VARI ABLES THAT HAVE INITI AL VALUES ASSUMED I

IN DBAs AND TAs, WHICH ARE MONITORED AND CONTROLLED DURING l

l POWER OPERATION SUCH THAT PROCESS VALUES REMAIN WITHIN THE ANALYSIS BOUND".

e T.S SAFETY SYSTEM PERFORMANCE (CRITERION 3, P.P.S. ON T.S.

i IMPROVEMENT):

l l

" STRUCTURES, SYSTEMS, OR COMPONENT... WHICH FUNCTION OR l

ACTUATE TO MITIGATE A DBA OR TA THAT EITHER ASSUMES THE l

FAILURE OF, OR PRESENTS A CHALLENGE TO THE INTEGRITY OF A l

FISSION PRODUCT BARRIER'. PERFORMANCE TO BE INCLUDED IN i

TECH. SPECS,1NCLUDES: CONSIDERATION 0F ALL APPLICABLE EVENTS, WHETHER EXPLICITLY OR IMPLICITLY PRESENTED...TO LIMIT CONSEQUENCES TO WITHIN THE APPROPRI ATE ACCEPTANCE I

1 CRITERI A.

l !

-. - -. N

=

m a

l i

MARGIN OF SAFETY

~

i e THE OBJECTIVE OF CRITERI A 2 & 3 IS TO ASSURE THAT ACCIDENT ANALYSES

.i ACCEPTANCE CRITERI A WILL NOT BE EXCEEDED.

e ACCEPTANCE CRITERIA FOR SOME f

i PARAMETERS ARE ENTERED AS SAFETY LIMITS i

f IN TECHNICAL SPECIFICATIONS.

e ACCEPTANCE CRITERIA FOR THE OTHER PARAMETERS, WHETHER OR NOT IN TECH.

i SPECS., ARE TO BE FOUND IN SAR AND SER.

1 e ACCEPTANCE CRITERIA DEFINE THE MARGIN OF l

i SAFETY.

l h

I

{

1 l

! i

=.

e

.:c j

Example of Margin of Safety Using Containment Pressure Transient E

o i

.e l

2 A J L l

Failure Point of the g

Containment Boundary

[

a Exact Value Unknown Margin of Safety i

(Approximately 130 psig) g 0

i l

' I Acceptance Lirmt 50 psig

\\

Higher than Maximum in

{

SAR Documented 35 psh a

SAR analysis N Previously l

Peak Containment Curve from SAR Reviewed and Approved Pressure Analysis 1

i Region of Oparation and Analyzed Transients ESF Setpoint 5 psig h

l Operating Range 1

I I

l Normal Operating

r 3 r l

<IN m

t Pressure v

l Tirne 4

F%ure3-2

{

l l

1 4

)

I

~

1 4

l e IMPACT OF CHANGES IS EVALUATED IN TERMS OF THE l

RESULTS OF THE SAR ANALYSES.

i e IF A CHANGE IN INITI AL CONDITIONS, MODEL, OR SAFETY i

I SYSTEM PERFORMANCE DOES NOT INCREASE THE RESULTS l

ABOVE THE ACCEPTANCE LIMIT,THE CHANGE IS NOT A USQ.

i i

l e HOWEVER, IF NRC ACCEPTANCE SPECIFIED CONDITIONS ON A COMPUTER CODE, METHOD, INDUSTRY PRACTICE, OR l

PENALTY, CHANGES TO THESE CONDITIONS MAY INVOLVE A USQ REGARDLESS OF RESULTS.

l e TREATMENT IS ANALOGOUS TO APP. K LOCA CRITERI A l

WHERE THE DETERMINATION OF WHETHER NRC REVIEW IS f

REQUIRED IS BASED ON ANALYSIS RESULTS:

j l

i 0-49 F PCT INCREASE, NO NRC REVIEW f

50 F OR MORE, NRC REVIEW i

t

~

13

~

l

,__ _..~-

m m--.

mm

- m

-

ML19327A691

Text

Enclosures:

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