ML20205S494

Summary of 860429-30 Meeting W/B&W Owners Group in Bethesda, MD Re B&W Owners Group stop-trip & Transient Response Improvement Program.List of Meeting Attendees & Viewgraphs Encl

ML20205S494
Person / Time
Issue date:	05/19/1986
From:	Paulson W Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8606120534
Download: ML20205S494 (53)

Text

{{#Wiki_filter:- _ _-_ _ May 19, 1986 . LICENSEE: Babcock & Wilcox Owners Group (B&WOG)

SUBJECT:

Summary of April 29 and 30, 1986 Meetings with the B&W Owners Group .On April 29 and 30, 1986, the NRC staff met with representatives of the B&WOG in Bethesda, Maryland. The purpose of the meeting was to present information to the NRC regarding the B&WOG stop-trip and transient response improvement program. The topics discussed were (1) sensitivity studies; (2) integrated control system (ICS)/non-nuclear instrumentation (NNI) evaluation; and (3) risk assessment review. A list of attendees for each session is enclosed -(Enclosure 1). A copy of the handouts provided by the owners group is also enclosed (Enclosure 2). The B&WOG sensitivity study of B&W plant transient responses was discussed in the morning session on April 29, 1986. The objectives of the program were discussed by the owners group. MPR is performing the sensitivity study under contract to the B&WOG. MPR summarized the status of the study including a discussion of the modeling to be used. The staff comments included: o The B&WOG needs to justify the adequacy of the calculational models for the purpose that they are being used. In this regard, the staff would like to see more detail on the steam generator modeling. o The need for objectivity and independence of the MPR review of sensitivity was emphasized. e o The staff would like to be kept informed of those recommendations from the owners group program that the utilities will adopt, and if a recommendation is not adopted, the basis for this decision. The staff also requested that schedules be provided for the implementation of recommendations. o The need for timely coordination and integration by the owners group of the results from each of the tasks was emphasized. The integrated control system (ICS)/non-nuclear instrumentation (NNI) program was discussed in the afternoon session on April 29, 1986. Following the Rancho Seco December 26, 1986 event, the B&WOG I&C Committee developed two recommendations for immediate action: 1. Each utility should evaluate the loss of ICS/NNI power and make appropriate changes to assure itself that the plant will go to a known safe state without any operator action; and 2. Each utility should evaluate the restoration of power to ICS/NNI and make appropriate changes to assure itself that the plant will remain in a known safe state upon restoration. In addition, the I&C Comittee developed three short-term actions which are to be completed by December 31, 1986: , '// NW B606120534 860519 pR TOPRP EPfVBW C PDR r

be-on . 1. Each utility shall develop and implement a recommended preventive maintenance program for ICS/NNI; 2. Each utility shall examine the wiring of the power supply monitors in its ICS/NNI cabinets and make a modification if necessary to wire that monitor directly to the output bus after the auctioneering diodes; and 3. Plan and initiate an exhaustive evaluation of the ICS. The owners group conducted a test of the ICS at Davis-Besse that showed that upon restoration of power, the position of the controllers is reproducible at the hand stations; however, the controllers will not come back to the positions that existed prior to loss of power. The owners group concludes that the 820 ICS system (CR-3; Davis-Besse; Rancho Seco and ANO-1) is reproducible upon restoration of power. In addition, bench tests on some analog memory modules at Rancho Seco showed that some of the modules were defective. The staff requested that the owners group provide the report on the results of the Davis-Besse-test and the Rancho-Seco bench tests. In the longer term, the owner's group will evaluate the benefit and determine the timing on replacing the ICS/NNI. The staff indicated that criteria need to be developed to determine whether or not to make a change. On April 30, 1986, the risk assessment review was discussed. The B&WOG indicated that the objective of the risk assessment is to provide a perspective on the complex event sequences (B&W reactor transients which required safety system to mitigate the transient) relative to other possible sequences leading to core melt. The owners group plans to compare the dominant accident sequences from the PRA's for Oconee and Crystal River-3 with the complex event sequences to determine if the events dominating core melt are the same in both cases. The owners group will then generalize results from the Oconee/ Crystal River-3 PRAs to the other B&W plants as discussed in enclosure 2. The staff requested that the owners group provide the transient Assessment Program report on transient events and the compilation of configuration differences between Oconee and Crystal River-3 and the other B&W plants. nw we tn ' Walter A. Paulson, Project Manager PWR Project Directorate #6 Division of PWR Licensing-B

Enclosures:

As stated P -6 WPaulson;jak JS 5/f/86 5/p/8

~ MEETING

SUMMARY

DISTRIBUTION Licensee: Babcock & Wilcox Owners Group (B&WOG)

• Copies also sent to those people on service (cc) list for subject plant (s).

Docket File NRC POR L PDR PBD-6 Rdg JStolz WPaulson OELD EJordan BGrimes .ACRS-10 NRC Participants WPaulson JTBeard RJones GSchwenk DCrutchfield CThomas ODParr JWermiel LMarsh PShemanski RBorgen HGarg KWichman RFerguson DBasdekas HBailey RKendall BAgrawal MRubin

ENCLOSURE I ATTENDANCE LIST APRIL 29, 1986 MEETING - B&W OWNERS GROUP AND NRC STAFF B&WOG SENSITIVITY STUDY NAME AFFILIATION WT~Faulson NRC/NRR/PBD#6 J.T. Beard NRC/NRR/0 RAS R.C. Jones NRC/NRR/TS G.A. Schwenk NRC/NRR/RSB-B D. Crutchfield NRC/NRR/PWR-B G.R. Braulke GPU Nulcear C. Thomas NRC/NRR/RSB-B 0.D. Parr NRC/NRR/PEICSB J.S. Wermiel NRC/NRR/PEICSB R.W. Ganthner BAW J.S. Carlton B&W L.B. Marsh NRC/NRR/RSB-B P. Shemanski NRC/NRR/PWR-B/PEICSB R. Borgen NRC/NRR/INEL H. Garg NRC/NRR/PWR-B K. Wichman NRC/NRR/PWR-B T. Dragone MPR J.H. Taylor PAW R.L. Ferguson NRC/PWR-B/PEICSB D.L. Basdekas NRC/RES/DET H. Bailey NRC/IE/DEPER/EAB R.B. Borsum B&W-Bethesda S. Connor Doc-Search Associates i N. Ehrich MPR H. Estrada, Jr. MPR S.T. Rose Duke Power Company G.R. Braulke GPUN R. Kendall NRC/DPWRL-B/PEICSB B. Agrawal NPC/RES/DAE/CSRB i T. Myers BWOG-Toledo Edison i h ll

ATTENDANCE LIST APRIL 29, 1986 MEETING - B&WOG AND NRC STAFF ICS/NNI NAME AFFILIATION WI~Paulson NRR/PWR-B L.B. Marsh NRR/RSB/PWR-B J.T. Beard NRC/NRR/0 RAS R. Kendall NRC/DPWRL-B/PEICSB G.R. Braulke GPU Nulcear C. Thomas NRC/RSB-B S.T. Pose Duke Power Company T. Myers Toledo Edison R.W. Ganthner B&W J.D. Carlton B&W R. Oakley Arkansas Power & Light R. Borgen NRC/INEL P. Shemanski NRC/NRR/PWR-B/PEICSB H. Estrada MPR H. Garg NRC/NRR/PWR-B/PEICSB B. Agrawal NRC/RES/DAE/CSRB 0.0. Parr NRC/NRR/PEICSB J.H. Taylor B&W E. Swanson B&W W.E. Wilson B&W R.S. Enzinna B&W R.W. Dorman B&W C.B. Doyel FPC R.C. Jones NRC/NRR/PWR-B

.. ~.... 4 I ATTENDANCE LIST APRIL 30, 1986 MEETING NRC STAFF i AND B&W Ok'NERS GROUP } R F AT$ESSMENT REVIEW i. NAME AFFILIATI'ON W. Paulson NRC/PWR-B R.C. Jones NRC/PWR-B 3 i D.L. Basdekas NRC/RES/DET B. Agrawal NRC/RES/DAE 1 I L.B. Marsh NRC/NRR/RSP-B N. Rubin NRC/NPR/FOR-B B. Yourgblood 90volchewer National Lat-E. Swanson B&W 1 R.S. Enzinna B&W T. Daniels Duke Power Company G. Hudson Duke Power Company R.W. Ganthner B&W G.R. Brau ne GPUM i M. Averett FPC J.D. Carlton B&W i H. Bailey NRC/IE/DEPER g i a i i i i l I-i 'I y -e -, = .~m y. y, ,,,..p.7.... ~.... ,p-. -of-y- ,,-e-m, _-w-- --g ---,_--,._,y_.--_=,, .7

z r;.,-~ e B&WOG SENSITIVITY STUDY MEETING AGENDA APRIL 29,1986 l. INTRODUCTION GEORGE BRAULKE (GPUN) 11. OVERVIEW OF SENSITIVITY STUART ROSE (DPC) STUDY OBJECTIVES ~ lli. MPR STUDY - DETAILED HERB ESTRADA (MPR) REVIEW OF PLAN, SCOPE, & REPRELIMINARY RESULTS IV. NRC RESPONSE, COMMENTS NRC ON B&WOG PLAN V.

SUMMARY

STUART ROSE (DPC) i l \\ o

Stop-Trip Program Process ~ ~ I. Information Gathering II. Integration Ill. Implementation I R TAP DATA i i l EXISTIIIG l 88WOG l PR(UECTS l " \\. I ~ LIST ~ ~ l letc 0F ^ PRaxCr CCIICDINS DPROYDeli DEFILE PRIORITIZE BU ET RECD 9 0 BA-CONCERflS. SCHEDULE. TI M M UTILITT AREAS CONCERNS AND -e N TS pygg IN PERFORM TD STEERIIIS I DFLDerT- ~* jf i PROJECTS M ITTEE ADON PLAIIT 3 IlfTERVIEWS l UI Tis l ggtgp l 8 INDUSTRY 1 PROJECTS HetPOWDIT l' lc REVIEW l OF BWIlliotg REPORT SDISIT!YITY l_ STATUS OF l UPLDEITATION l l-I 1 m ont L 604I3 1' ACHIEVDUIT l l' 2 l l

MEETING PURPOSE 1. TO DEFINE THE B&WOG SENSITIVITY STUDY, ITS SCOPE AND OBJECTIVES 2. TO DESCRIBE THE MPR STUDY PLAN IN DETAll 3. TO RECEIVE NRC COMMENTS, QUESTIONS, AND h l AREAS OF CONCERN RELATIVE TO B&WOG PLANS FOR CONDUCTING SENSITIVITY STUDY i am - + _ -, - - -, _ ,_m

---

p

---

.ym,

,,w

SENSITIVITY STUDY OBJECTIVES 1. TO QUANTIFY THE PERCEPTION OF " SENSITIVITY" BY DEFINING MEASURABLE, QUANTITATIVE INDICES OF i THERMODYNAMIC BEHAVIOR DURING NORMAL { OPERATIONS, ANTICIPATED OCCURRENCES, AND i DESIGN BASIS ACCIDENTS. i 2. TO REAFFIRM REACTOR SAFETY MARGIN i REQUIREMENTS. 3. TO EVALUATE THE RELATIVE DIFFERENCES IN THESE INDICES AND MARGINS AMONG PWR DESIGNS. 4. TO IDENTIFY AREAS OF POTENTIAL IMPROVEMENT IN ^ THE B&W DESIGN AND OPERATION BASED UPON AN ANALYSIS OF THE OBSERVED DIFFERENCES AND THElR SIGNIFICANCE. 5. TO IDENTIFY SPECIFIC AREAS WHICH WOULD BE APPROPRIATE FOR CONSIDERATION IN RISK OR OTHER STUDIES.

THREE ELEMENTS OF SENSITIVITY COMPARISON 0 COMPARIS0N OF PWR RESPONSES DURiNG NORMAL OPERATIONS 0 COMPARISON OF PWR RESPONSES DURING OFF-NORMAL TRANSIENTS ANTICIPATED OCCURRENCES DBAs O COMPARIS0N OF HARDWARE DESIGNS MECHANICAL PROTECTION CONTROL OPERATOR INTERFACE l l

SCOPE OF COMPARISON WITH SIMPLIFIED PLANT MODELS 0 PRELIMINARY PLANT MODELS B & W UNIT PRE-75 CE UNIT POST-75 JL UNIT 0 FOLLOW-UP ANALYSIS P OST-75 CE P RE-75 JL 5 ,__o-

( REFERENCES A REPORT TO THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS BY A TASK FORCE OF THE ACRS STAFF AND

FELLOWS,

" REVIEW AND EVALUATION OF THE BABC0CK & WILCOX NUCLEAR STEAM SUPPLY SYSTEM (WITH EMPHASIS ON THE OPERATION AND CONTROL OF THE ONCE-THROUGH STEAM GENERATOR)." S. P. KAIRA, "MODELING IRANSIENTS IN PWR STEAM GENERATOR UNITS," NUCLEAR SAFETY, VOL. 25, NO. 1, JANUARY-FEBRUARY 1984.

NRC, NUREG-0667,

" TRANSIENTS RESPONSE OF BABCOCK & WILCOX-DESIGNED REACTORS," MAY 1980. NRC, NUREG/CR-4385, " EFFECTS OF CONTROL SYSTEM FAILURES ON TRANSIENTS, ACCIDENTS, AND CORE-MELT FREQUENCIES AT A WESTINGHOUSE PWR," NOVEMBER 1985. l NRC, NUREG/CR-4385, " EFFECTS OF CONTROL SYSTEM l FAILURES ON IRANSIENTS, ACCIDENTS, AND CORE-MELT FREQUENCIES AT A BABC0CK & WILC0X PRESSURIZED WATER REACTOR," DECEMBER 1985. NRC, NUREG/CR-3958, " EFFECTS OF CONTROL SYSTEM FAILURES ON IRANSI'ENTS, ACCIDENTS, AND CORE-MELT FREQUENCIES AT A COMBUSTION ENGINEERING PRESSURIZED WATER REACTOR, MARCH 1986.

REFERENCES (CONT'D) NRC, NUREG/CR-4471, "LOS ALAMOS PWR DECAY-HEAT-REMOVAL STUDIES

SUMMARY

RESULTS AND CONCLUSIONS," MARCH 1986. S. LEVY AND J. E. HENCH, "A STUDY OF SIMULATION AND SAFETY MARGINS IN LIGHT WATER REACTORS," PREPARED FOR THE PRESIDENT'S COMMISSION ON THE THREE MILE ISLAND ACCIDENT, AUGUST 25, 1979. FSARS LERS INCIDENT REPORTS AND RELATED CORRESPONDENCE IN THE PUBLIC DOCKET IE INFORMATION NOTICES IE BULLETINS VENDOR TECH MANUALS PLANT DESIGN DATA TRADE PRESS

~ I MODEL VERIFICATION 0 FIRST PRINCIPLES BASIS 0 COMPARISONS WITH ACTUAL FEEDWATER UPSETS 0 COMPARISONS WITH ACTUAL STEAM FLOW UPSETS 0 COMPARIS.0NS, IF APPROPRI ATE, WITH MORE DETAILED MODEL RESULTS w -G---

y e MODELING APPRUACH 4 OVERALL PLANT PKz. p AtiAwnc. sre$u/a>we.imeA.mc-mau HoHoGewooS h Lugni:Gt o Mims e oorsve4e = w / he.v SSV Aov Tsv k ,Msi v o q o sv l TCV m o m Q <.g d FWV ey LeoPs SCT P NOT LEfr EWEBCy

• 1 DOVW

'A O D 1 GesuP rourrtCS, 30 J scr

• DePet.egt ceEFPct% #

, VA%%g % u

• mmret coweicisur

. cot.D ttGr R9820/ 1 Du4MY, 97gg R*.MTMMERC PLUQ, 1 LUMP FUEL NEMY

• h BLE SToQAc g Floto SIMCrLe kobe RJ

= coc= ae-xm.. -r

.MODELING APPROACH (CONT'D) UTSG Ws w 4 , m @elRs . megyg EQuiusesu4 % gu VAWIC,htstBE WirTcuesu va m e_ w o ucuib i Mn HMS STtME . Tnev-e=".4@ Gttset Fw'0 m% gg Wl'M M M STE %,g 1 - c.m iu9tweur mw.c -p MOMeNW WCtr m; i cr 7 9 .DVuwgicxu.v mu=9 3bwucagee__ D cu.3 Co99thpT BEAT e l bp!Drreau kkTe

• i W49 p%3ngt, gusea,y BALAuca neoupD WAcd e5 WO 14 */Utsg Bou0tt w

Wbm8 M6 s ---{ncu; -m

• • • D

-InwuceMet. uD9EE ugvowose - a m -touwcoget Louw R. VELoct rsus L l T.4/ \\Tc 1 1 L,1 4 4 D W M A M IC. W N M %.wes muuur wea'r m* To rum soubut St.oid i (

1 MODELING APPRO ACH- (CONT'D) OTSG l ' 1 LutAO bypnAlc., yg l WW ILM.AUCE N:= A200% SH N-rf O d

• DTE op CHung 69 VAR $!LM5s e

ERAJ narceAT:vs HenT rt2AWSFsg,, esigt s W " Q" BOG = m e.- t weom v. -m,% P4kK.*1D S d GER "T,o4NestarD* Sacries f?-Edreou L N l ummmmusMN ws w e.eus nem. EntAuce tusm pseiexrioc;, Nw .h a huJucoMap_ hw bO/A4% ese oe. % w ved w Mouwrug ).hiMT Abbrnet) MS COMS 'b4W2.t M 4 fu Beltw4 -REAb oF FLuib iu rust SEcreou OC N@ LE* 'TOSE *Rou BLN J i - HEAD op. Ft08b 10 DA'?E.1 ,,,,,,,,_,,y 8 %M ND MQuig i I LOkAP %t Ltu G 9 Et".rt N % VEL.coTits %N L.oS S -l LUMP 'Dcxouc Mte Posmaa Ms C eD Fics DM) A - 1. Lou p A DAJAMic Tc EMEEGY ME /wbBouuG stcTrou 1 l L

n 9 UNIT c THIP DATA SG 22 4500 of FEED,TLOW 10 22 (X,J/lR )- = ~~ I i-4030 t J l 3600 3150 1-2700 ~ 2250 1860 1350 \\ 3 900 t-45g 0 0 18 36 54 6 72 90 108 126 144 162 180 TIME l

a ..-.-.-.n . s, e e CD ^ CJI gg

e. -.e M

g2.e _w s=a ec. M ---e lll=3 EWlm Car"Ji 3 smJ cC: i CD 8lh e w ---e EN3 C4 _CJB CD CM Cs4 w C4 W CJB C.O E M I N C

• N3 A

=--e 83.e j m W GB GB N W y# H = s.2 M \\ CElB E CCD M M "*t- ' Q* < - \\ '1 N cs3 w \\e \\ I i _W C53 cu G3 G Lt"> . CEB tr:p C5h trJe C53 Lt"3 G3 tr"3 C53 M3 CCD W Lt":p 'C' c4 ---e cr% CC:n w n G 8:F% CF% 8lllF% ClT% ClF% alllF% ClCI CClB CCB CCll> - -e

UNIT 2IRIPDATASG22 SG 2h LEYEL'(IN.) i-- 1,' ~ -12 .' h.' -24 ~ -36 -48 -60 -72 8' -84 -96 (s s -108 \\D -120 0 10 20 30 40 50 60 70 80 90 100 TIME e

S- ?- REACTOR TRIP WITH RFR PUMP INITIALLY AT.4800 RPM vs n-a- .C U C OC*> l~ W D W 1 \\ t yC t O* i \\ Actual Data C \\ y Model Sin atlation g I wOe i g e s C \\ C \\ W \\ l l H N N If1 0 - N T s ~ ~ - ~ ~ ~ ~ - - = 's f , = = %f l g-i I k I ) 0 P.O 40 60 80 100 120 TIME seconds l l I

-. ~ o O1 cw 8~ o N m REACTOR TRIP WITH RFR PUMP INITIALLY AT.4800 RPM o 8-o e 0 e g-

c..

CA 30JO g o-o u3 - -- Actual Data O Model Simulation o* A / t o i_-- o-i s. o ,I t N s__ N--- l i , ~... ----- l l o i i i i i i 0 20 40 60 80 100 120 TIME seconds f

~ COMPARISONS OF RESPONSE DURING NORMAL OPERATIONS "0 PEN LOOP RESPONSE" TO DISTURBANCES IN: 0 REACTIVITY a STEAM FLOW 0 FEED FLOW l

o O .ov.. W STEAM GENERATOR MODELS l 10% REACTOR POWER STEP CHANGE DOWN -r 100* POWER ~.o N. c Babcock & Wilcox o C ~ ~ ~ Combustion Engineering C --e ~~'~,s,- ~s ~ s ' w. a Thot ,= C 1e_ T a superheat n_ -a iac ,0 ,o <r, e o T oN-p cold o o ,o o-Tsat #"** ec T, i '0 '. 0 0 i0 00 i0.00 60.00 30.00 i 00.00 120.00 - '. v c

r s._ s

%e

o o i

Oo_,

oo O_ (.', go O_ 'o C - N ~ O o_ W Babcock & Wilcox Cn ~'* L Combustion Engineering g o 0-(MS..- oo .o.. -r o O STEAM GENERATOR MODELS o_ m 10% FEEDWATER FLOW STEP CHANGE DOWN 100% POWER o O 4'.00 6'0.00 5'0.00 i00.00 i20.00 O 00 20.00 Mc - =_. c i. 0 v ~

1 Q u m-O O o o o .o_ v. 00 N STEAM GENERATOR MODELS o 10x FEEDWATER FLOW STEP CHANGE DOWN O O 100* POWER o O 'o N o o o o o .o_ D_ a C 2 2 ^_ i -,. c 2 Zo O o o ~_ ~~ >v. W- ' 's, L: i ' __J ' ~s, ~~s CD o N s,N a (.0 c o o o_ N_ e t oo o Babcock & Wilcox o o o 'o _ ~~~ Combustion Engineering e I oo ao o' m o_ .m e oo O 2 I T' T.00 2'.00 IG.00 6'0.00 5'0.00 100.00 120.00 0 TIME SEC li _, Q--

5 oo -~ .oo N.- oo .o in oo .oo co x O_ ,s- / o O / / e s vc Cn o - Q_ ~ s' O U"J o G_ o / o a n-s m - Babcock & Wilcox --- Combustion Engineering o o ,s d e' o. o s' / o o STEAM GENERATOR MODELS C l 4-10% STEAM FLOW STEP CHANGE DOWN oc o ....s 1004 POWER =c .= C T.00 2'0 00 i0.00 6'0.00 S'0.00 1'00.00 i20 00 7 i t i._ .c r r vp s 4

~ ~ D. o g O o O O-g- co g o O o o o'. o a s' tn m' ,s s' s e o s' o o s' o .o s' O. co. N s' d ,s'/ o = C ,a o A. 29 E _s bl c >v. N-og Y Babcock & Wilcox Co O (no O Combustion Engineering ~ ' " " O O" 7-cn o O o 5 9 o e, e 9 STEAM GENERATOR MODELS o". C-104 STEAM FLOW STEP CHANGE DOWN ,m 100% POWER oo .o. 5 o y ~. 0 00 20.00 40 00 6'O.00 S'O.00 1'00 00 i '20.00 7 i. M. C ^ 7 p< 3-J

COMPARISONS OF ANTICIPATED UPSETS AND DBAs O REACTIVITY R0D WITHDRAWAL FSARs FSARs R0D EJECTION O COOLANT FLOW UPSETS ALL PUMPS FSARs ONE PUMP MODEL 0 STEAM FLOW TURBINE TRIPS --

FSARs, MODEL SAFETY, DUMP AND BYP ASS VALVE FAILURES MODEL STEAMLINE BREAK FSARs l

l - m

COMPARIS0NS OF ANTICIPATED UPSETS AND DBAs (CONT'D) O FEED FLOW OVERFEED MODEL LOSS OF ONE PUMP MODEL LOSS OF ALL PUMPS

FSARs, MODEL EFW OPERATION, MODEL SENSITIVITY TO OVERFEEDING 0

DECAY HEAT REMOVAL VIA OTHER BLEED & FEED ANALYSES 0 LOSSES OF COOLANT FSARs 0 EXCESSIVE COOLANT INJECTION FSARs 0 STEAM GENERATOR TUBE RUPTURES

FSARs, OTHER ANALYSES l

l

COMPARISONS OF SYSTEM SPECIFICS 0 MECHANICAL COMPONENTS 0 PROTECTION SYSTEMS ENGINEERED SAFETY FEATURES REACTOR PROTECTION SYSTEM COMPARISON OF TRIP FREQUENCIES INPUT SIGNAL COUNTS TESTAB ILITY COMP ARIS0NS 0 CONTROL FUNCTIONS /0PERATOR INTERFACES REACTOR POWER AND TEMPERATURE REACTOR PRESSURE FEEDWATER FLOW / STEAM GENERATOR INVENTORY STEAM FLOW / TURBINE GENERATOR POWER l l MODEL ANALYSES AD HOC ANALYSES .----_.-,,.--_y-_.

B&WOG ICS EVALUATION MEETING AGENDA APRIL 29,1986 1. INTRODUCTION GEORGE BRAULKE (GPUN) 11. ICS EVALUATION CHRIS DOYEL DETAILED REVIEW (FPC) OF PLAN, SCOPE l & RELATED ACTIVITIES Ill. NRC RESPONSE, NRC COMMENT ON B&WOG PLAN IV.

SUMMARY

CHRIS DOYEL ( (FPC) i

l MEETING PURPOSE 1. TO DEFINE THE B&WOG ICS/NNI EVALUATION, ITS SCOPE, OBJECTIVES, AND ITS RELATIONSHIP TO OTHER ICS/NNI ACTIVITIES. 2. TO DESCRIBE THE ICS EVALUATION PLAN IN DETAll AND HOW THE RESULTS WILL INTERFACE WITH OTHER ACTIVITIES. 3. TO RECEIVE NRC COMMENTS, QUESTIONS, AND AREAS OF CONCERN RELATIVE TO B&WOG PLANS FOR CONDUCTING ICS/NNI EVALUATION. I i

B&WOG l&C COMMITTEE RECOMMENDATIONS 1. IMMEDIATE ACTIONS 2. SHORT--TERM ACTIONS (BY 12/31/86) 3. LONG-TERM ACTIONS )

l IMMEDIATE ACTIONS 1. EACH UTILITY SHOULD EVALUATE THE LOSS OF ICS/NNI POWER AND MAKE APPROPRIATE CHANGES TO ASSURE ITSELF THAT THE PLANT WILL GO TO A KNOWN SAFE STATE WITHOUT ANYOPERATOR ACTION REQUIRED. 2. EACH UTILITY SHOULD EVALUATE THE RESTORATION OF POWER TO ICS/NNI AND MAKE APPROPRIATE CHANGES TO ASSURE ITSELF THAT THE PLANT WILL REMAIN IN A KNOWNSAFESTATE UPON RESTORATION. 4

l SHORT-TERM ACTIONS (BY 12/31/86) 1. EACH UTILITY SHALL DEVELOP AND IMPLEMENT A RECOMMENDED PREVENTIVE MAINTENANCE PROGRAM FORICS/NNI. 2. EACH UTILITY SHALL EXAMINE THE WIRING OF THE POWER SUPPLY MONITORS IN ITS ICS/NNI CABINETS AND MAKE A MODIFICATION IF NECESSARY TO WIRE THAT MONITOR DIRECTLY TO THE OUTPUT BUS AFTER THE ACTIONEERING DIODES. 3. PLAN AND INITIATE AN EXHAUSTIVE EVALUATION OF THE ICS. I

_. ~.. 1 ICC OWNERS GROUP STUDIES INDUSTRY RECOMMENDATIONS RECOMMENDATIONS 4 TROM PREVIOUS FOR DESIGN 1f INTOFJtATION FMEA's REQUIREMENTS s GATHERING NURIG.IE BULL

• KNOWN (EXISTING)

INDEPENDENT REVIEW OF DESIGN RIQUIREMTS SENSITIVITY INTCRMATION (TODAY) STUDY TAP INTERIUM pA A REPORT ORIGINAL REVIEW DESIGN REQUIRDfENTS 'I REVIEW OPERATING COMPARISON: O PHILOSOPHY ^ b LIST OT ICS/NNI P OBOS AND K EXISTING W s DESIGN P.ECOMMENDED PATH REQUIREMTS. TO S0WTION LIST or AS-BUILT N ICS/NNI ICS/NNI 's T MODITICATIONS CONTIGURATION PLANT OPERATING (TODAY) INTERVIEWS A PROCEDURES AND IEC OWNERS PM PROGRAMS i GROUP COMMITTEE OPERATOR SUPPORT AND IEC COMMITTEE LATEST SYSTEM DRAWINGS U COMPARISON EVALUATE ALL AS-BUILT m CHANGES FOR TO'ALL GENERIC L PLANTS IMPLICATIONS

LONG-TERM ACTIONS FOR UPGRADE AND/OR REPLACEMENT 1. COMPLETE AND DOCUMENT THE ICS EVALUATION. 2. USE THE EQUIPMENT OBSOLESENCE PROGRAM TO EVALUATE THE BENEFIT AND DETERMINE THE TIMING ON REPLACING THE ICS/NNI. 3. DEVELOP A GENERICSPECIFICATION FOR REPLACING THE ICS/NNI. I

I L ill L ,k ~3- ,l R 3l s !i l l il h[. ll fltl 9 ~ i I 11 il 1 ! 11 1 I I O !!l lill as Ii th a

B&W OWNERS GROUP TRIP REDUCTION AND TRANSIENT i RESPONSE IMPROVEMENT PROGRAM RISK ASSESSMENT REVIEW i ( l l APRIL 30,1986 l l l L..

B&WOG RISK ASSESSMENT MEETING AGENDA APRIL 30,1986 i INTRODUCTION GEORGE BRAULKE (GPUN) RISK ASSESSMENT MARK AVERETT DETAILED REVIEW (FPC) i OF PLAN, SCOPE, & RELATED ACTIVITIES 1 NRC RESPONSE, NRC COMMENTS ON B&WOG PLAN

SUMMARY

MARK AVERETT (FPC) l

Stop-Trip Program Process e

1. Information Gathering II. Integration 111. Implementation l

TAP -A I I I EIISTIE I OmER B3WDG PRa!ECTS I l LIST l -o "TEGR^ 1stC OF g 9cET ECON O EA-DEFIE PRIORITIZE COEDtitS Drit0VD0lT + C0KERits

SCHEDULE, TI M M UTILITY AND

+ N TS pgggg IN AltEAS CONCERMS 4

• l 17FLD0lT-

~# ^ PERFORM TO STEERIE ATION PROJECTS N ITTEE INTERVIEWS l EIISTiE gg gg I & IMXJSTRT PPOJECTS IlOEPDODIT REVIEW ~l IOllITDR, REPORTk 0F I_ STAiti,OF SDtSITIVITY '~ l DFLDOITATICII j 1 -l I E,0RT GOALS ~ i ADf1EVD0li i I i l

.s s.- 0 0 MEETING PURPOSE i 1. TO DEFINE THE B&WOG RISK EVALUATION, ITS SCOPE AND OBJECTIVES. 2. TO RECEIVE NRC COMMENTS, QUESTIONS, AND AREAS OF CONCERN. I

9 4 D RISK ASSESSMENT OBJECTIVE O PROVIDE A PERSPECTIVE ON THE COMPLEX EVENT (CATEGORY "C") SEQUENCES RELATIVE TO OTHER POSSIBLE SEQUENCES LEADING TO CORE MELT l l l l l l l l l

TOGLS O TWO B&W PLANT PRA'S - OCONEE - CRYSTAL RIVER O THE TWO PRA'S SELECTED ARE: COMPLETED UP-TO-DATE: THOROUGH ANALYSIS l - SIMILAR EVENT AND FAULT TREE STYLES O CONFIGURATIONS REPRESENT A GOOD CROSS-SECTION OF MAJOR FEATURES IMPORTANT TO COMPLEX EVENTS - WITH AND WITHOUT MSIVs - WITH AND WITHOUT EFIC - S.G. LEVEL CONTROL BY O FEEDPUMP SPEED O CONTROL VALVES - 721 AND 820 ICS MODELS

RISK ASSESSMENT APPROACH l. ASSESS THE IMPORTANCE OF THE 10 OPERATING EXPERIENCE COMPLEX EVENT SEQUENCES TO CORE MELT. II. COMPARE INITIATING EVENTS FROM OPERATING EXPERIENCETO PRA INITIATING EVENTS. lit. EVALUATE PRA DOMINANT ACCIDENT SEQUENCES, SYSTEMS AND COMPONENTS FOR CONTRIBUTION TO CORE MELT. IV. GENERALIZE THE RESULTS AND DRAW CONCLUSIONS FOR THE B&W PLANTS. t l 1

l. OPERATING EXPERIENCE SEQUENCE ASSESSMENT 1. ASSEMBLE INFORMATION ON THE 10 CATEGORY "C" EVENTS: 4 O SYSTEMS / COMPONENTS / FAILURE MODES O OPERATOR ACTIONS O SEQUENCE OF EVENTS 2. MATCH OPERATING PLANT SEQUENCES WlhH THE PRA SEQUENCES. 3. IDENTIFY SEQUENCE DIFFERENCES: O OPERATOR ACTIONS O EQUIPMENT SUCCESS / MALFUNCTIONS 4. EVALUATE AND RECONCILE DIFFERENCES. 5. DETERMINE WHAT ADDITIONAL FAILURES (OPERATOR OR EQUIPMENT) MUST TAKE PLACE TO REACH CORE MELT. l

II. INITIATING EVENT ASSESSMENT 1. 209 REACTOR TRIPS TO BE " BINNED" INTO INITIATING EVENT GROUPS. 2. CALCULATE AND COMPARE INITIATING EVENT FREQUENCES TO: O PRA INPUT FOR OCONEE/CR-3 O GENERIC SOURCES FOR ALL PWRS 3. EVALUATE AND RECONCILE DIFFERENCES l

lil. DOMINANT ACCIDENT SEQUENCES FROM OCONEE AND CRYSTAL RIVER-3 PRAs

1. IDENTIFY AND RANK O

DOMINANT SEQUENCES O DOMINANTSYSTEMS/ COMPONENTS 2. COMPARE DOMINANT SEQUENCES TO COMPLEX OPERATING EXPERIENCE EVENTS O DETERMINE IF EVENTS DOMINATING CORE MELT ARE THE SAME OR DIFFERENT FROM COMPLEX OPERATING EVENT SEQUENCES

IV. GENERALIZE RESULTS FROM OCONEE / CRYSTAL RIVER-3 PRAs FOR B&W PLANTS: CONTRAST SYSTEMS AND COMPONENTS IMPORTANT TO COMPLEX EVENTS WITN THOSE FOR CORE MELT SEQUENCES NOTE CONFIGURATION DIFFERENCES FROM OCONEE AND CR-3 DETERMINE IF THE CONFIGURATION DIFFERENCE WOULD SIGNIFICANTLY: O INCREASE INITIATING EVENT FREQUENCY USED IN PRA O DECREASE THE ABILITY TO MITIGATE THE j EVENT l i. i O CHANGE THE RISK IMPORTANCE OF COMPLEX EVENT SEQUENCES

OUESTIONS TO BE ANSWERED ARE INITIATING EVENTS AND COMPLEX SEQUENCES ADEQUATELY PORTRAYED BY THE PRAs? ARE COMPLEX EVENT SEQUENCES SIGNIFICANT TO CORE MELT RISK 7 CAN RISK METHODS BE USED TO AID PRIORITIZATION OF TRIP REDUCTION AND TRANSIENT RESPONSE i IMPROVEMENT PROGRAM CONCERNS? l ~

4 'g 9 b SCHEDULE EXPECTED RESULTS FOR OCONEE/CR-3 MID JULY ) GENERALIZED RESULTS FOR ALL B&W OCTOBER t - - - -}}