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{{#Wiki_filter:.. C E(9 K E_10'- ) =,c.? m. ' T O O o 9-I u l May 28,1982 ( MEMORANDUM FOR: D. Okrent, Chairman ACRS Grand Gulf Subcommittee M. Plesset, Chairman j ACRS Fluid Dynamics Subcommittee I i
SUBJECT:
STAFF MEETING WITH J. HUMPHREY, et al., RE MARK III CONTAINMENT I attended a staff meeting regarding the concerns of Mr. Humphrey on May 27,1982. Mr. Humphrey is a former General Electric Engineer who worked on " Stride" containment systems. Mr. Humphrey wrote to MP4L expressing his concern that certain issues relating to the Mark III containment were not being properly addressed, and some of these concerns may apply to Grand Gulf. Mr. Humphrey met with officials of Mississippi Power and Light Company to discuss his concerns. The staff held a meeting at Bethesda, May 27, 1982 with MP&L, Mr. Humphrey, the General Electric Company and Bechtel to discuss the significance of Mr. Humphrey's concerns. Attached for your information is a copy of the presentation by MP&L listing each of Mr. Humphrey's concerns and MP&L's response to each. Mississippi Power and Light's position is: 1. They have accurately listed all of Mr. Humphrey's concerns. 2. They have thoroughly reviewed the concerns. 3. They have adequately responded to each of the concerns. 4. All the concerns have been addressed and accounted for in the Grand Gulf design. 5. No further action is needed. Mr. Humphrey's position is: Q "d FILE: $C p ~ ' ny of his concerns relate to the area between the NSSS vendor X-Ib{ onst 1 ,.1t.yifnd the.vtfl.ity responsibility and require furtheW[ ,= -.m 2... _. e o"* > . AC,RS, .g;gigman* gy8ysS85g w w o t.......- g ~> 7%2% m
i ( - a,* 9 O O r p A 4-The utility (MP&Ll was requested to have a third position review the concerns. MP&L has contracted with TERRA Corporation and expects a report from TERRA within two weeks. A transcript was taken of the meeting. If you so desire, I will forward a copy of the transcript to you. bA
Attachment:
As stated /1" I cc: w/a ACRS Members R. Fraley M. Libarkin T. McCreless J. McKinley G. Quittschreiber cet w/o ACRS Technical Staff ACRS Fellows l l l 4 ~h~ hh D 'T_' orricE) { SURNAMEf DATEk n.m~ ~ - _
.g w-f g,g RECEIVED AovlSORY Cat,WiUEE ON j 7-yk REACTOR StJEGUARDS, U.S.N.RA JAN 1i1982 g hfS,9,1011,12 l 2 3f gg 6 iii I BWR FUTURE _Q1 RECT 40NS. PRESENTATION BY D. R. WILKINS TO ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D.C. i JANUARY 7, 1982 /
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( MAJOR TEST PROGRAF 6 SUPPORTING BW/6-MARK 11I l e SMALL, INTERMEDIATE AND LARGE-SCALE MARK I1l CONTAINENT TESTING e FULL-SCALE FUEL BUNDLE HEAT TRANSFER TESTING l e FULL-SCALE FLCH-INDUCED VIBRATION TESTING OF REACTOR INTERNALS e ACTLERATED FULL-SCALE PIPE TESTING FOR STRESS CORROS10N PREVENT 10N e EXPANDED FULL-SCALE ECCS TESTING l e Bm/6 MECHANICAL ODMPONENT VERIFICAT!ON TESTS - FLOW CONTROL VALVE - SAFETY / RELIEF VALVES - FAST SCRAM CRD ) - INCLINE TRANSFER TUBE - FEEDWATER N0ZZLE/SPARGER - ETC. l DRW 1/7/82 l 1 i .?__, k* ha&&.... W. ~~ c g I
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1 i BMl/6 SYSTEM IMPROVEMENTS i WATER DELIVERY IMPROVEMENTS l e REACTOR CORE ISOLATION COOLING SYSTEM LOGIC l e AUTO-DEPRESSURIZAT10N LOGIC FOR NON-BREAK EVENTS e AUTO-RESTART OF HIGH PRESSURE CORE SPRAY SCRAM SYSTEM IMPROVEMENTS w SCRAM DISCHARGE VOLUME e ATWS MODIFICATION _ del %._Y.. HEAT REMOVAL IMPROVEMENT i e (DblAlfNENT OVERPRESSURE RELIEF l DRW 1/7/82 ) l t 1 i l 1 w. b,. w.: I
ESTIMATE OF CORE DAMAGE PROBABILITY PER REACTOR YEAR 10 O WASH 1400 m BWR/6 WITH ItiPROVEMENTS A - WATER DELIVERY B - ATWS MODIFICATION C - CONTAINMENT OVER PRESSURE RELIEF B C 10-6 _ A ( 10-7 * ( 10-8 l g .: REAK
- EAK BREAK m-FEEDWATER OFFSITE HEAT POWER REMOVAL
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l FUNCT10N OF CONTAINENT i e FUNCT10N PROTECT PUBLIC FROM EXCESSIVE DOSE IN SEVERE ACCIDENTS e ACCOMPLISHED TWO WAYS - CONTAlfNENT BARRIERS . PRIMARY AND SECONDARY 4 . INTEGRITY FOR ALL DESIGN BASIS EVENTS AND MQSI EVENTS BEYOND DESIGN BASIS ( - FlLTERED CONTAIitENT VENTING . ADDIT 10NAL PROTECT 10N FOR EVENTS BEYOND THE DESIGN BASIS i DRW 1/7/82 1 ~ - d { 1
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S F DxTU O E W T R A D L L P O F O D P x L N O O O I P S F S D E R = P P F U D S = S m ? YAW S l T = DT i EN m A DE P AV RE G G N ET 5' - DN I B E B l-RD U OI R FC - V W{i C nh C S GA N L I D O N BN O BA P U a SN RS N CT O S l E I '*~. RI S 3 ES E TN AA R WR PP T U S TN EM YN L RI L AA E MT W I N Y RO R PC D I 'i ) f .q: ^ Td(3,4: I
T N W I O R P D ER UL I AF EL BABO R P G T N NS EI EE RB MR = UB NU 3 [ SR SU x I S c AS EC .6R 4J e. TE T G RS NR I 0I P OP L 8S L C I 2P LO n D B A0 EL A RP CE P nh lL v w. U NI A TS EY C CE R UR EE E RU TS FT R G EA UI 1 SN RM SS 0 3 I w ,A2 LI I SP 6 S g 5 I X E 1 P VY I O R 0 T r_ I I R P4 TI N AL P s KP RA EB A RA M s P AC TNEM YN L RI L AA E MT W I N Y RO R PC D fj'
( l l Importance of Fission Product Scrubbing in Mark lli Pressure Suppression Containment EVALUATION BASIS WASH 1400
- CORE MELT SOURCE TERMS 10,000 DF = 1
- CONTAINMENT FAILURE AT 4 HOURS
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Oo FATALITY THRESHOLD (320 REM) O O m 100 - w a O {DF = 1,000 10CFR100 LIMIT (25 REM) S DF EALISTIC EVALUATION 0 1 2 3 4 5 6 7 8 9 MILES DOWNWIND FROM SITE DRW w 5, g,y p p .i-
( o D SUMARY BIR/6 MARK III e PRODUCT OF 25 YEARS OF EVOLUTION AND SIMPLIFICATION e BACKED BY 400 R-YRS. FIELD EXPERIENCE e MOST TESTED GE-BWR EVER e RETAINS PROVEN FEATURES OF EARLIER BWR'S - SIMPLICITY - MANY WATER DELIVERY SYSTEMS - DEPRESSURIZATION CAPABILITY - NATURAL CIRCULATION - PRESSURE SUPPRESSION 9 - ETC. DRW 1/7/82 %%e.. + ____--_-__A
t BM U6 - MARK Ii1 CONT. e INCORPORATES LATEST SAFETY TECHNOLOGY - WATER DEllVERY IMPROVEMENTS - ATWS MODIFICATION - OPERATOR INTERFACE - CONTAIINENT OVERPRESSURE RELIEF - SUPPRESSION POOL SCRUBBING - ETC. e BACKED BY STRIDE AND GESSAR - INTEGRATED NUCLEAR ISLAND DESIGN - SUPPORTS ONE-STEP LICENSING e FIRST UNIT IN OPERATION t - KUD SHENG - TAlWAN l - 61 MONTH CONSTRUCTION I BWR/6 - MARK 11I --- ENGINEERED l FOR THE FUTURE I DRW 1/7/82
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7 ( ( EXCERPT FROM MIN 9TES OF 261ST ACRS MEETING JANUARY 7-9, 1982 III. Boiling Water Reactor Standard Plant Design (0 pen to Public) [ Note: R. Savio was the Designated Federal Employee for this portion of themeeting.] A. Design Features _f BWR/6-Mark III D. R. Wilkins di scu ssed with the Committee the current status and future course of the General Electric BWR design (see Appendix X). He listed f ou r ma jo r principles in GE's approach to BWR design: con-tinuous simpli fication of the design; standardization of an entire l " nuclear island"; evolu ti ona ry rather than revolutions ry change; and thorough "t e s t bef ore use" of new features. A series of slides were then shown wnich described advantages of direct cycle power production which included safety and economic advantages through simplicity, lower pressure, inherent reactivity control, and direct communication between water sources and the reactor vessel, the evolutionary design of com-mercial BWRs, and descriptions of several representative plants Dresden 1, KBR, Oyster Creek, and Dresden 2. The Pressure Suppression i Containment concept was described and its advantages in seismic, spacial and constructibility considerations explained. D. W. Moeller requested an explanation of an entry on the slide entitled, Increasing Regulatory Emphasis on Nuclear Island, which pointed out 1973 regula-tory empnasis on occupational exposure. D. R. Wilkins explained tnat General Electric had deliberately laid out its STRIDE or Reactor Island design to respond to regulatory requirements as promulgated with particular attention to shielding, lay down space for equipment, and planned maintenance activities which will greatly reduce plant person-nel occupational exposures. In answer to a question by Acting Chairman
- Ray, D.
R. Wilkins indicated that the STRIDE design permi ts better electrical separation for routing cables from power sources to equip-ment in the plant to meet tne increasing regulatory requirements for divisional separation. The definition of the nuclear island design package being of fered by General Electric was discussed. D. R. Wilkins explained that General Electric was not offering a turnkey design but did indicate tnat the architect / engineer would work for General Electric and GE woulc provide technical direction to the architect / engineer and assume responsibility for system integration of the entire nuclear island. D. R. Wilkins explained, in addition, that the owner has only to furnish cooling towers, the water supply for the towers, switch yards, the turbine generator and turbine building, and the service building. Wi t h i n the service building GE provi des the offgas system. D. R. Wilkins discussed the principal features of the current BWR/6-Ma rk III. J. Ebersole questioned whether GE had developed evapora-tive cooling for the suppression pool which becomes hot in long-term GE BvR/(v nW5 .u3pp
a - c o MlWUTES OF THE 261ST ACRS MEETING JANUARY 7-9, 1982 transients. D. R. Wilkins indicated that that was a potential feature under study. D. R. Wilkins explained, in answer to a question by W. Kerr, that the containment verification testing program is directed at quantifying the containment loads in the suppression pool du ring loss of coolant accident conditions and during safety relief valve blowdowns to provide a design basis for the structural design. In answer to an inquiry by Acting Chairman Ray, D. R. Wilkins indicated that this reactor system is designed to nandle the ATWS events within the limits of the suppression pool and witnin the limits of the ATWS events. J. Ebersole expressed concern at GE's ability to depressurize the BWR/6 plant under all circumstances. He pointed out that the rapid depressurization D. R. Wilkins just described is done by opening relief valves by energizing solenoid val ves. He su dested that GE should have handled the reliability of that depressurization process by getting rid of the solenoid valves; these were used on reactor designs about 2D years ago. D. R. Wilkins commented on environmental qualification of the solenoid valves. D. R. Wilkins explained that if reactor vessel water level is main-tained, decay heat removal in the BUR /6 is passive. Strong natural circulation internal to the reactor vessel and steam release to either the main condenser or to the suppression pool heat sinks combine to provide this passive decay heat. When the subject of reactor water level instrumentation was brought up, J. Ebersole concluded from the GE presentation that the loss of coolant accident was itself being allowed to destroy the redundant characteristics of water level instrumentation. M. W. Hodges, NRC Staff, indicated that his impression was that the BWR/6 had three sets of taps around the vessel and three trains of level instrumentation such that if a failure of one train caused the inoperability of a second train during an accicent, one train of plots would still be available. The discussion that ensued was concluded by a request from Acting Chairman Ray that GE reply in v:riting concerning the redundancy of core level instrumentation. D. R. Wilkins described symptom-oriented emergency guidelines for the BWR which would help to minimize the chance of operator error. D. A. Ward questioned whether GE saw any potential for automating the logic in the guidelines with a process computer. D. R. Wilkins presented I a detailed description of simple effective operator interf ace displays but indicated that GE has no plans for automating emergency guidelines. J. Ebersole i nq ui red concerning improvements in the scram discharge I volume design such as redundant drain vent valves and diverse redundant water level indicators. D. R. Wilkins volunteered to provide a written explanation of its desi gn of control rod scram discharge volumes. Committee Members questioned G. Sherwood about inservice inspection of stub tube welds. W. Kerr was concerned that GE welds met the ASME code 6 ?. y n In a
e ~ MIMUTES OF THE:2615T ACRS MEETING JANUARY 7-9, 1982 d which includes inservice inspection. G. Sherwood indicated that the welds do meet the ASME code which requires inservice inspection but some stub tubes are not normally inspected because it is very difficult to do. J. Ebersole endorsed the concept of rapid pressure release embraced by GE of dumping steam to the suppression pool. D. R..Wilkins discussed ' accident mitigation. He indicated that the. ' function of containment at a nuclear plant in the case of the BWR pressure suppression containments is accomplished in two ways. The first way.is through containment barriers, primary and secondary-barriers which are designed to maintain their integrity' for all design basis events and have' sufficient margin to maintain their integrity for most events beyond the -design basis. The second way concerns filtered containment venting or scrubbing of potential releases f rom the containment. Containment venting and scrubbing of potential releases are an ' inherent safety features of the Mark III pressure suppression containment. D. Okrent expressed concern that. GE was not. bringing before the Committee potential pathways or scenarios in which the fission product scrubbing would be ineffective and the decontamination factor of a thousand that D. Wilkins mentioned would be inoperative. D. R. Wilkins explained that GE is focusing its attention on proba-bilistic risk analyses to show that the probability of the existence of bypass pathways is acceptably small and can be neglected. D. Okrent questioned whether GE could strengthen its containment in order to be able to handle a larger range of hydrogen events. D. R. Wilkins noted that strengthening the containment from,a risk assessment or consequence assessment point of view would be hard to defend from a benefit cost point of view. He stressed the defense in-depth concept that GE uses. J. Ebersole expressed concern that a badly degraded core would segregate the internal dry well in the BWR/6 containment from the outboard side of the drywell and limit mixing such that at the end of an accident, if you were producing hydrogen or if you failed to cool the core, you would have a relatively concentrated source of hydrogen in the drywell compared to a mixed case. D. Wilkins explained that a mixer system did exist which circulates air from the wetwell region back into the drywell as part of the existing hydrogen control. i G. Sherwood of GE described the nuclear island design and suggested that GE is dealing with problems with which the ACRS is concerned such as systems interfaces, standardization and optimization, and in-depth system interaction studies. 7 / Q*
EXCERPT FR0ti MINUTES OF 261ST ACRS MEETING JANUARY 7-9, 1982 D. Okrent questioned wnether GE had looked at design features which would be useful in helping to reduce the likelihood of successful serious sabotage by an insider. D. R. Wilkins did not respond in detail Decause of privileged information that is involved. D. Okrent suggested that a closed meeting be scheduled in the future to discuss this subject. G. Sherwood described the nuclear island licensing ,/ program, tne GESSAR program status, outstanding regulatory issues, and f.l other generic licensing issues with respect to the BWR/6 Mark III (see s Appendix XI). \\ l It was explained that station blackout capability is extended by the containment overpressure relief function. Nuclear island failure modes S and effects analyses will identify any needed corrections with regard ( to systems interactions. When G. Sherwood mentioned that GE would be ready to report on a full risk assessment in the near future, D. Okrent expressed the hope that GE would identify all of tne areas where j judgment was used in selecting either the parameter or the methods for det ermi ni ng inputs to the analysis including specifying the ranSe of (i inputs that are possiole. D. R. Wilkins indicated that GE planned to ~ include an uncertainty analysis or error budget in its risk assessment l which is aimed at addressing tnis issue. l J. Ebersole was concerned about providing reactor operators with Class IE type separation and quality levels for operator indications, input on recorde rs, and indi cators and enunciators. He questioned whether GE hos taken steps in its STRIDE design to qualify and upgrade the information flow to the operator to IE caliber. D. R. Wilkins indicated that GE is addressing tne issue of reliability of information to the operator but did not plan to upgrade the equipment to Class 1E. I
we-au wu \\ LICENSING 0F GE'S BWR 6 f4K III NUCLEAR ISLAND l I l ] ADVISORY C0f711TTEE ON REACTOR SAFEGUARDS JANUARY 7, 1982 l l l l 7DhscEfo$n!o$ l GLEi1N SHERWOOD NOR SMUARDS, U.S.N.R C. DAN WILKINS JOE QUIRK JAN 1 i1982 y, ['8 9JO4141:2g3 4[ ,,s NUCLEAR POWER SYSTEt1S DIVISION GENERAL ELECTRIC C0i1PANY L ACR$OFFICECOPYx 30 NoLRemove from AC1STice
I AGENDA I i INTRODUCTION.............................. GLENN SHERWOOD Objectives of BWR/6 FK Ill Licensing 8 'l BWR FUTURE DIRECTIONS........................ DAN WILKINS BWR Evolution Features of BWR 6 11K III Results of BWR Evolutions Plant Protection Features i LICENSING OF BWR 6 MK 111................. GLENN SHERWOOD Nuclear Island Licensing GESSAR 1 and 11 Licensing Objectives I Summary j i I l 1 1 1 I J
i GE'S NUCLEAR ISLAND DESIGN o INTEGRATED DESIGI4 o COVERS ALL RADIOLOGICAL SIGNIFICAf4T SYSETf1S AND STRUCTURES o SIPPLIFIES INTERFACES o i1AXIt'IZES STAf4DARDIZAT10N AND OPTIf11ZATION o ALLOWS TIMELY IN-DEPTH SYSTEf!S INTERACTION EVALUATIONS o STRONG ENGINEERING SUPPORT Design - One Organization Complete Design Record Detailed Plant Design Specification a ADVAf4CED DESIGN FEATURES Solid State NSPS Improved ECCS Performance Flultiple Barrier Containment 8 x 8 Fuel Bundle Compacted Control Room
NUCLEAR ISLAND = _ - - = _ +;-m ;<. Z.74 ) Nuclear. steam supply 8 J (1) Reactor bldg. NUCLEAR ISLAND Auxiliary nuclear system Scope < (2) Fuel bldg. (3) Diesel gen. bldg. (4) Auxiliary bldg. 5 7 (5) Radwaste bldg. (6) Controlbldg. w 3 6 Balance of plant (7) Turbine bldg. _I4 (8) Service water bldg. (9) Switchyard 1 I2
l NUCLEAR ISLAND LICENSING PROGRAM JAN-SEPT 1972 PHASE I: AEC/ACRS ENDORSEMENT OF STANDARD PLANT CONCEPT APRIL 1973-DEC 1975 PHASE II: NRC/ACRS CONCURRENCE - PRELIMINARY DESIGN OF BWR/6 f1 ARK III NUCLEAR ISLAND MARCH 1980-DEC 1982 PHASE III: NRC/ACRS APPROVAL - FINAL DESIGN OF BWR/6 MARK JII NUCLEAR ISLAND 1975 - 1983 PHASE IV: POWER WORTHINESS CERTIFICATE
1 1 GESSAR PROGRAM STATUS PRELIMINARY DESIGN APPROVAL (GESSAR I) APPROVED 12/75 - 12/78 EXTENSION 12/78 - 12/80 FINAL DESIGN APPROVAL (GESSAR II) TENDERED 3/80 NRC ACCEPTANCE LETTER 12/81 TO BE DOCKETED 1/82
+ 1 REGULATORY ISSUES o REGULATORY GUIDES Preliminary Design Approval on Regulatory l Guides Through 1,76 (March 1974) l All Current Regulatory Guides Assessed in 4 GESSAR II Final Design Approval Will Include Regulatory Guides Through 1982 a STANDARD REVIEW PLANS (SRP's) Assessment of Nuclear Island Against SRP's Evaluation Completed (Pre-1981 Version) l Ho Design Changes Required a ' SEVERE ACCIDENT /Tr.I ISSUES HRC to Complete Review in 1982 Proposed Key Activities and Schedule Program Management Meeting November 1981 BWR Technology Update Meetings Feb-April 1982 GESSAR Severe Accident Appendix Submittal May 1982 ACRS Review August 1982 Commissioners Approval September 1982 i o UNRESOLVED SAFETY ISSUES 25% - Not Applicable 45% - Not:Open Issue (SER, Requirements specified) 30% - On Going 1.e., ATWS, Station Blackout, Control System Failure I
'e OTHER GENERIC LICENSING ISSUES o STATION BLACK 0UT Containment Overpressure Relief Extends Capability a 3YSTEf1S INTERACTI0i1 Nuclear Island FMEA Will Identify Any Needed Corrections a SABOTAGE PREVENTION Inherent in Nuclear Island Design 9 l
\\ NUCLEAR ISLAND SAFETY REVIEWS l 1 ' ~ o TOTAL PLANT FAILURE MODE AND EFFECTS ANALYSIS Covers 76 Safety Related Systems Preliminary Analysis 80% Complete No Significant Safety Problems Identified To be Complete Fourth Quarter 1982 o TOTAL PLANT PROBABILISTIC RISK ASSESSMENT Evaluates Core Melt Probability and Societal Risk Preliminary PRA Completed in 1981 Meets Consensus Safety Goal Plant Risk Substantial!/ Below WASH 1400 To be Completed First Quarter 1982 4
( SUhMARY a GE DESIGN AND LICENSING EMPHASIS...BWR 6 NUCLEAR ISLAND o NUCLEAR ISLAND RESPONSIVE TO CURRENT REGULATORY REQUIREl1ENTS o NUCLEAR ISLAND LICENSING PROGRAfi 1 PDA Approval in 1975 FDA Submitted to NRC, Docketed in January 1982 Requesting FDA SER in September 1982 Proposing Degraded Core Rule in tiid 1982 o BWR 6 fiK III NUCLEAR ISLAND DESIGil TO BE GE OFFERING IN THE 1980's o NUCLEAR ~lSLAND GE CONTRIBUTION TO PRE-APPROVED PLANT LICENSING REFORM}}