ML20198G246

Summary of ACRS Subcommittee on DHR Sys 860318 Meeting in Washington,Dc Re Review of NRR Resolution Effort for USI A-45, Shutdown DHR Requirements. List of Attendees Provided.Viewgraphs Encl

ML20198G246
Person / Time
Issue date:	03/19/1986
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
REF-GTECI-A-45, REF-GTECI-DC, TASK-A-45, TASK-OR ACRS-2407, NUDOCS 8605290323
Download: ML20198G246 (44)

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' Aze dsL2Kb DATE ISSUED: 3/19/86 nhIu. ACRS DHRS SUBCOMNITTEE MEETING MINUTES MARCH 18, 1986 WASHINGTON, DC PURPOSE: The purpose of the meeting was to continue review of the NRR resolution effort for USI A-45 " Shutdown Decay Heat Removal Require-ments."

ATTENDEES:

ACRS NRC D. Ward, Chairman A. Marchese  !

J. Ebersole W. Minners C. Michelson A. Buslik I. Catton 1 P. Davis UCLA L. Cave SNL D. Ericson Northern States Power S. Hatch J. Niles G. Sanders M. Bohn MEETING HIGHLIGHTS, AGREEMENTS AND REQUESTS

1. During opening remarks, Mr. Michelson asked if the DHRS Subcomit-tee would explore the pros and cons of a dedicate,d shutdown system.

Mr. Ward indicated that he did not see a need at this time for independent Subcommittee action on this item and preferred to wait and see what NRR proposes for resolction of USI A-45. Dr. Catton asked if NRR could tell the Subcommittee how it will factor the AE0D Report's conclusions into the A-45 effort. [ Note: A Report on Cold-Shutdown DHR Problems at US PWRs was scheduled to be given 8605290323 060319 , __.

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DHRS Meeting Minutes March 18, 1986 at the meeting, but was postponed due to the unavailability of the Presentor.] Mr. Minners (NRR) indicated that today's presentations would constitute a status report and that no resolution for A-45 has yet been proposed. Mr. Ebersole decried the lack of a clear focus for dedicated DHR on the standard plants.

2. A. Marchese (NRR) overviewed the day's presentations. He noted that NRR is in the process of developing a resolution position (s) for A-45. In response to Mr. Ebersole, A. Marchese indicated that one possible outcome of A-45 could be a new GDC that specifies a dedicated DHR for future plants. As a result of further dis-cussions, Mr. Ward reminded the Subcommittee that A-45 is address-ing current plant DHR systems.

In response to Dr. Catton, Mr. Minners said the AE0D concerns for DHRS during cold shutdown will be addressed under Generic Issue 99:

"RHR Suction Valve Interlocks." Mr. Ward indicated surprise that A-45 is not addressing cold shutdown-related DHR problems. Mr.

Minners said A-45 will look at the entire spectrum of DHR, but GI 99 will focus on current RHR systems. The difference is that the SNL studies themselves do not focus on cold shutdown DHR. An evaluation of the cold shutdown issue will be performed in the Staff's regulatory analysis for A-45 resolution. Dr. Catton said examination of human factors problems and addressing "little (or inexpensive) fixes will buy you alot in terms of DHR reliability.

D. Ericson (SNL) reviewed the status of the Plant Analysis Reports (Figures 1 and 2). SNL's goal is to have all Analysis Reports complete by May 30, 1986. A summary Report on the Analysis Reports is also scheduled for May 30 - being written in tandem with the individual Reports themselves. UCLA is preparing some related analyses in support of A-45 (Figure 2A). These studies are scheduled to be complete by the end of September 1986.

1 DHRS Meeting Minutes ' March 18,1986

3. S. Hatch (SNL) reviewed the results of the Cooper Case Study. He initially indicated the changes made to the Cooper analysis based on the results of the Quad Cities review (Figure 3). In response to Subcommittee questions regarding the restricted scope of the SNL studies, Mr. Hatch said other ongoing studies (PRAs RSSMAP, etc.)

are being examined for applicability to the DHR issue.

The initial insights (concerns) for Cooper were noted (Figure 4).

Mr. Ward said the fact that the qualitative analysis was in some cases confirmed by the quantitative analysis is a significant finding. He said this approach gives a good yardstick as to how well the SNL qualitative review correlates with the quantitative review.

Key results from the Cooper Study include:

• Twenty-four internal event sequences are dominant prior to recovery; 20 internal event sequences are dominant after recovery; seismic events, external floods, and fires dominate the special emergencies. Figure 5 lists the most significant internal vulnerabilities. In response to Mr. Michelson, NRR said the Event V accident is not addressed in A-45 due to  !

resource constraints and a judgment based on results of PRAs l that SB LOCAs and transients are the dominant risk contribu-tors.

1 In response to Mr. Davis, NRR said the AC and DC power up- i grades required per USI A-44 will be factored into the A-45 Study results.

Six plant modifications were proposed as a result of the internal event study. These are: (1) addition of a third diesel; (2) add a dedicated battery for diesel 1; (3) add a

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< I DHRS Meeting Minutes March 18, 1986 third 125 VDC station battery; (4) add a reactor building closed cooling water (RBCCW) manual bypass line; (5) add a RBCCW isolation valve; and, (6) provide auto-isolation for a RBSW M0V.

For the external events, the fire and flood analysis results are shown on Figures 6 and 7. For the seismic analysis, a set of vulnerabilities and modifications were proposed (Figure 8).

A set of five Alternative fixes for Cooper were defined (Figures 9 and 10). In response to Mr. Michelson, Dr. Ericson said the choices made in grouping the Alternatives were done so as to facilitate the cost / benefit analysis and optimize their coverage of vulnerabilities. In response to questions from Mr. Ebersole, Mr. Marchese said the UPPS system will be considered in the A-45 Regulatory Analysis. Mr. Davis raised a concern that Alternative 5 (add-on DHRS) may complicate recovery from an ATWS.

The value/ impact (V/I) analysis was detailed. The summary of the value/ impact analysis (Figure 11) shows that three of the Alternatives (2-4) are cost-effective fixes. If no interdic-tion or decontamination (I&D) for offsite releases are assumed, then all fixes become cost-beneficial (Figure 12).

In sumary, key points Mr. Hatch noted included: (1) station blackout sequences are the most important; (2) Cooper has good physical redundancy in DHR systems; and (3) the effect of containment venting is minimal. With regard to Item (3), Mr.

Hatch said their study of the venting issue was by no means an exhaustive one. [

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DHRS Meeting Minutes March 18, 1986

4. The Turkey Point plant analysis Report details were given by G.

Sanders (SNL). D. Ericson noted that the base case assumed no bleed and feed (B&F) capability which is the reverse of the normal assumption. (Note: this was based on information obtained from the Utility that neither B&F procedures, nor associated operator training, existed at this plant.) However, $NL did do a sensitivi-ty study which evaluated B&F capability.

Mr. Sanders' presentation followed the same format as above for the Cooper analysis. Key points of the Turkey Point analysis include:

• Initial insights for Turkey Point are given in Figure 13. The calculation of core melt contributors shows a large contribu-tion (8.8 x 10-4) for internal sequences. The dominant internal failure is a transient with loss of AFW. Comon-mode failure of the AFW pumps is the significant vulnerability.

Figure 14 lists the proposed modifications for internal event vulnerabilities.

For the external event vulnerabilities of concern for this -

plant (seismic, fire, wind, and flood), Figures 15-18 provide the results and proposed fixes. In response to a question from Mr. Ward regarding the plant's seismic vulnerabilities, SNL noted that the water storage tanks are non-conservatively designed vis-a-vis seismic margin.

• Three Alternatives (fixes) are proposed for this plant (Figure 19). In response to Mr. Michelson, Mr. Sanders said the Utility indicated that they are quite receptive to installing the modifications proposed to address internal events. The CM probabilities for the various Alternatives are given in Figure

20. These values assumed no B&F capability. With B&F, the change (decrease) in CM probability is 6.6 x 10-4 resulting in an overall CM probability of 2.2 x 10-4 .

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4 DHRS Meeting Minutes March 18, 1986 l l

• Results of the V/I analysis (Figure 21) show a cost / benefit for alternatives 1 and 2. Again, assuming no I&D all fixes become cost beneficial, Dr. Catton noted that if B&F capabil-ity is assumed, the results of the V/I analysis could change [

significantly. Mr. Ebersole noted that there is large uncer-tainty in the B&F capability for these plants (i.e., how likely one is to successfully B&F).

In summary, Mr. Sanders noted that: internal event sequences dominate core melt, auxiliary feedwater failures dominate internal sequences, emergency electric power is a key issue, and test or maintenance outages have a measurable impact. In response to Mr. Ward, Mr. Sanders said that the dominant CM sequences would remain unchanged assuming B&F capability, but the proposed modifications (fixes) may well change with B&F.

SNL ran some sensitivity studies assuming various OHR capabilities.

(With/without B&F capability, blocked / unblocked PORVs, sq ondary blowdown,etc.) The results (Figures 22-23) will be factored into the overall A-45 resolution analysis. The analysis indicates that whether the PORVs are blocked or not makes little difference on the overall CM probability. In response to Mr. Ward, SNL said the capability for secondary blowdown doesn't significantly reduce the CM probability either because of the relatively few accident -

sequences for which it is useful.

D. Ericson provided some comments on the use of B&F vis-a-vis the Point Beach and Turkey Point studies. The results (Figure 23A) indicate that the decision to B&F vis-s-vis its impact on CM probability seems to be plant specific.

5. Dr. M. Bohn (SNL) discussed the methodology used by Sandia for the seismic analyses used in the Plant Analysis Reports. He noted that

__ . . ~ . _ _

DHRS Meeting Minutes March 18,1986 the seismic analysis is performed at a level consistent with the internal event analysis effort.

i The methodology or specific steps used in the seismic analysis was outlined and detailed (Figure 24). Key points of the discussion included:

• The steps in the seismic hazard characterization were detailed

. (Figure 25). SNL does take account of the effects of local site amplification due to such phenomena as soft soil columns under the plant. For Zion (the worse case study to date) the exceedance probebility due to this phenomena alone increased by a factor of > 6. This approach was applied to Cooper since its site soil characteristics are similar to Zion. There was discussion of protection against beyond-SSE events. Mr. Ward indicated that one should be able to analyze beyond-SSE events if such events pose a unique threat (e.g. liquefication).

• In discussing the failure mode determinations, Mr. Bohn noted, in response to Mr. Michelson's questions, that relay chatter

, is not considered in the seismic analysis.

Component fragilities are determined on a generic and site-specific basis as appropriate. Figure 26 shows the equipment that required site-specific fragility analyses for the plants analyzed to date.

• The basic elements of the seismic calculational procedure were outlined. An example of the results obtained is shown in Figure 27.

Mr. A. Buslik (NRR) provided some comments on his review of the SNL seismic analysis. He indicated that the SNL CM frequency from the

l l

DHRS Meeting Minutes March 18, 1986 l seismic contributor was a median, not a mean, value. He feels this contributor should be taken as a mean value. Another concern he noted is that the uncertainties in the study are rather large and that it is not clear these uncertainties have been accounted for in l the analysis.

6. L. Cave (UCLA) discussed his work on developing value/ impact methodology to help guide the Staff in its V/I analyses for USI A-45. Figure 28 shows the scope of the work. He noted that analyses they have performed shows that the $1000/MR value used by NRR in the V/I calculations appears to be a reasonable estimate given the (now outdated?) assumptions used in the CRAC-2 code (Figure 29).

Mr. Cave showed that if on-site costs are included in the V/I analyses, fairly large sums can be justifiably spent to reduce CM probabilities (Figure 30).

Evaluation of the cost impact of a nuclear moratorium imposed in the event of a severe accident was shown. The results show costs ranging from $2.2 to $200 billion, depending on the time of shut-down (0 to 10 years after accident). Mr. Cave said that the possibility of a moratorium should be carefully considered, given the recent political decisions in Europe (and to a lesser extent in the U.S.) regarding the availability of the nuclear option.

~

Mr. Ward said the results seem to indicate that a large portion of the costs result from public reaction (fear) than from actual hardware fixes.

7. Mr. Marchese discussed the remaining schedule milestones for USI A-45. A draft regulatory position on A-45 should be available in

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. J DHRS Meeting Minutes March 18, 1986 August 1986. CRGR review would be in November 1986 with the resolution Package issued for public comment in March 1987.

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8. J. Niles (NSP) discussed the-NUMARC initiative,on decay heat .

removal. NUMARC believes DHR is a safety issue requiring industry attention. Thorough technical review and coment by the Industry on A-45 is also essential. In response to Mr. Michelson, Mr. Niles indicated that NUMARC has not organized a definitive agenda for addressing the DHR issue. NUMARC continued by stating that resolu-tion of this issue should be cost beneficial, provide obvious safety benefits without introducing down-side risk, and help improve the reliability of existing DHR systems. Close coordina-tion between NUMARC and NRC is essential to closure.

Mr. Ward noted that there are many imponderables in the ll/I analy-ses including the reaction of the public. He indicated that the line between technical issues and "public" issues becomes blurred  ;

when considering the need for fixes. .

NUMARC provided preliminary coments on the A-45 effort to date. l The remarks were of a generally negative nature (Figure 31).

Mr. Ward noted that the Subcommittee welcomes broad-scope response from the Industry on this issue. Mr. Niles in response to Mr.

Ward, said the basic Industry position for this issue would prob-ably be that modest DHR system improvements will .be sufficient., l Y Mr. Ward asked that the Industry keep the Subcomittee informed of their position as the resolution effort unfolds.

9. In response to Mr. Michelson's question, the Subcomittet requested that NRR provide the resciution Package before CRGR review, some- l time in August 1986.

i

)

CHRS Meeting Minutes March 18, 1986 Mr. Ward, referring to an earlier discussion, asked if the concern of the reliability of the RHR system will be evaluated (PRA-type study) for A-45. Mr. Minners indicated that this will not be done.

Mr. Ward said some type of systematic analysis of this issue should be conducted. NRR agreed.

10. Mr. Ward solicited the Subcommittee's comments on the day's presen-tations.

P. Davis - The Cooper analysis - Referring to the NUREG-1150 work, he noted that the CM probability for Cooper is av100 times higher than for Peach Bottom. He wondered why. Dr. Ericson said there has been a more complete study (more in depth analyses) on Peach Bottom. Mr. Davis also had some detailed comments specific to the Cooper analysis. Overall he was impressed with the quality of the Cooper Report.

I. Catton - The Cooper analysis should evaluate a UPPS-type system for the plant. He suggested NRR evaluate modest fixes. The Turkey Point report should have considered B&F. Seismic issues don't concern him for plants east of the Rocky Mountains. NRR should focus on RHR system reliability and evaluate work done by others (Levy, et. al.) on dedicated DHR systems.

J. Ebersole - NRR should focus on evaluation of dedicated DHR to avoid the detailed analyses we are now doing.

I C. Michelson - NRR should do a detailed careful analysis of a dedicated DHR system. NRC has yet to do this. NRR said their i regulatory analysis for A-45 will include this item.

The meeting was adjourned at 5:20 p.m.

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s DHRS Meeting Minutes March 18, 1986

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. t!0TE: Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, NW, Washington, DU, or can be purchased from ACE-Federal Reporters;, 444 North Capital Street, Washington, DC20001,(202)347'f700.

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CASE STUDIES STATUS REPORT i

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COMMENTS RECEIVED

' FINAL EDITORIAL WORK PENDING QUAD CITIES DRAFT SUBMITTED COMMENTS RECEIVED FINAL EDITORIAL WORK - APRIL 30 COOPER DRAFT SUBMITTED FINAL EDITORIAL WORK - MAY 30 i

DRAFT COMPLETE, NOT SUBMITTED TURKEY POINT -

FINAL EDITORIAL WORK PENDING

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REPORT ON CASE STUDIES - DRAFT SCHEDULED MAY 30

SUMMARY

REPORT ON UCLA STUDIES - INITIAL DRAFT SCHEDULED JUNE 30 FINAL DRAFT SCHEDULED SEPT. 30 VALUE-IMPACT MEASURES NON-QUANTIFIABLE VALUE-IMPACTS

' OTHER AVERTIBLE COST ISSUES COSTS OF NUCLEAR MORATORIA 4

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• INCREASED CONTROL CIRCUlT FAILURE ADDED SUPPORT SYSTEM INITIATING EVENTS

• REEVALUATED OTHER DATA l

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NO INTECRATED RE50TE SHUTDOM4 PANEL OUTSIDE CONTROL ROGA PLANT APPEARS TO BE VULNERABLE TO UPSTREAM DAM FAILURES BOTH SAFETY DIVISIONS ROUTED TMOUGH ONE CABLE 8PREADING ROGWI COteENSATE STORAGE TANKS ARE NOT SEISWIC CATEGORY I I

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l MOST SIGNFICANT VULNERABIUTIES - INTERNAL

1. LOCAL FAULTS OF TWO DIESELS

! 2. FAILURE OF DIESEL FIELD FLASHING i

3. FAILURE OF A 125 VDC POWER TRAIN

! 4. COOLING WATER MOV FAILURES l

6. COOLING WATER FLOW DIVERSIONS l

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SEISMIC-RELATED VULNERABILITIES RBCCW HEAT EXCHANGERS I

125 AND 250 VDC SWITCHGEAR RCIC MINI-FLOW VALVE CONDENSATE STORAGE TANKS 480 VAC TRANSFORMERS SEISMIC MODIFICATIONS WELD HEAT EXCHANGER CONNECTION ADD SUPPORTS / ANCHORS s

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DEFINmON OF DHR ALTERNATIVES

l. ALTERNATIVE 1 THIRD DIESEL GENERATOR l

PLUS ALTERNATIVE 4 ALTERNATIVE 2 DEDICATED BATTERY TO DG 1 l PLUS ALTERNATIVE 4 l ALTERNATIVE 3 i THIRD 125 VDC BATTERY PLUS ALTERNATIVE 4 l ,

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RBCCW ISOLATION VALVE MODIFY RBSW VALVE HPCI/RBSW FIRE BARRIER l SEISMIC MODIFICATIONS MODIFY FLOOD EOPS l

l ALTERNATIVE 5 j ADD-ON DHR SYSTEM

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SUMMARY

OF VALUE-IMPACT ANALYSIS DELTA NET BEN $ NET BEN PER TOTAL PER CORE OFFSITE P-REM ALT MELT ($XE-6) P-REM ($XE-6) 8463 -14.2 4994 1 3.8E-4 -22.6 1397 +4.9 0 2 3.0E-4 -1.9 1471 +4.4 0 .

3 2.8E-4 -1.9

-1.2 1191 +4.8 0 4 2.7E-4

-67.5 1.9E4 -57.8 1.9E4 5 4.8E-4 e

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SUMMARY

(NO INTERDICTION) '

NET BEN $ NET BEN DELTA PER TOTAL PER CORE OFFSITE P-REM ALT hELT ($XE-6) P-REM ($XE-6) l 237 +44.4 156 1 3.8E-4 +35.9 39 +51.4 0 2 3.0E-4 +44.7 41 +47.8 0 3 2.8E-4 +41.6

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+7.2 532 +17.0 453 I 5 4.8E-4 l l d

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INITIAL TURKEY POINT INSIGHTS e TWO DIESEL GENERATORS FOR TWO UNITS

• THREE TURBINE DRIVEN AFW PUMPS FOR TWO UNITS

• SAFETY-RELATED COMPONENTS LOCATED OUTDOORS e NO BLEED AND FEED PROCEDURES CURRENTLY IN PLACE BOTH SAFETY DIVISIONS ROUTED THROUGH ONE CABLE SPREADING ROOM i e REDUNDANT COMPONENTS ARE NOT SPATIALLY SEPARATED 9

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INTERNAL EVENT MODIFICATIONS 4

1. TIE A MOTOR DRIVEN STARTUP FEEDWATER PUMP TO THE UNIT 3 CST. ALSO ALLOW REDUDANT DC POWER SUPPLIES FOR AFW CONTROL VALVES.

2. ALLOW AUTOMATIC REDUNDANCY OF STATION BATTERIES FOR I STARTUP OF EACH DIESEL GENERATOR.

3. INSTALL A REDUNDANT SERVICE WATER VALVE TO ISOLATE NON-ESSENTIAL SYSTEMS.

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l h-1 a SEISMIC CORE MELT PROBABILITIES LEVEL (SSE) S 2

Ty T TOTAL 3

.5-1 4.2E-8 6.1E-7 5.7E-7 1.2E-6 l 1-2 1.7E-7 4.9E-6 5.1E-6 1.0E-5 2-3 3.9E-7 3.6E-6 5.8E-7 4.6E-6 3-4 2.4E-7 4.2E-7 1.9E-8 6.8E-7

>4 1.5E-9 5.4E-8 5.1E-8 1.1E-7 1.7E-5

, SEISMIC RELATED VULNERABILITIES / MODIFICATIONS

1. REFUELING WATER STORAGE TANKS - STRENGTHEN SUPPORTS

2. CONDENSATE STORAGE TANKS - STRENGTHEN SUPPORTS

3. ConeoNENT CootiNG WATER NEAT ExcHANGERS - Support eEoESTAtS q%

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SUPPRESSION SYSTEM 3.0E-6/RX-YR l

WATERPROOF INSTRUMENTATION

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) 19 FT - DIESEL FUEL OIL TRANSFER PUMPS FAIL FOLLOWED BY BATTERY DEPLETION 2.4E-6 20 FT - DIESEL GENERATORS AND ALL SAFETY PUMPS EXCEPT AFW (EVENTUALLY DC DEPLETED) 1.2E-5 P gg ,fgf

? rgayCzy 20.5 FT - ALL PUMPS INCLUDING AFW 4.?F-5 7?m E TOTAL 5.6E-5 MODIFICATION RAISE EXISTING TWO FOOT WALL TO FOUR FEET AND ADD STOP GATES FOUR FEET HIGH 1.9E-5 N'

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EXTREME WIND ANALYSIS VULNERABILITY 400 FOOT CHIMNEY FALLS ON THE DIESEL FUEL OIL TANK AND BATTERIES DEPLETE 1.8E-7 l CHIMNEY FALLS ON DGS AND CST 1.1E-5

! CHIMNEY FALLS ON 480 V SWITCHGEAR AND I- BATTERIES DEPLETE 1.3E-5 TOTAL 2.4E-5 MODIFICATION

USE STARTUP FEEDWATER PUMPS POWERED BY BLACK -

START DIESELS 7.2E-7 4

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1 TURKEY POINT ALTERNATIVE DEFINITION ALTERNATIVE 1: ALIGN MOTOR DRIVEN FEEDWATER PUMP TO CST PROVIDE REDUNDANT DC POWER TO AUXILIARY l

FEEDWATER VALVES i PROVIDE AUTOMATIC REDUNDANCY OF DC POWER TO DGS INSTALL A REDUNDANT SWS ISOLATION VALVE ALTERNATIVE 2: ALL ABOVE PLUS:

PROVIDE CABLE SPREADING ROOM FIRE SUPPRESSION i

EXTEND FLOOD WALL HEIGHT STRENGTHEN RWST AND CST SUPPORTS

{ STRENGTHEN CCW HEAT EXCHANGER SUPPORTS l

ALTERNATIVE 3: ADD-ON SDHR SYSTEM s - - - -

l TURKEY POINT RESULTS FOR ALTERNATIVES 1

[ ALTERNATIVE BASE CASE 1 2 1 INTERNAL 8.8E-4 1.7E-4 1.7E-4 4.9E-5 1 SEISMIC 1.7E-5 1.7E-5 3.4E-6 9.5E-7 INTERNAL FLOOD NEGLIGIBLE - - -

FIRE 7.5E-5 7.5E-5 3.0E-6 3.9E-6 WIND 2.4E-5 7.3E-7 7.3E-7 2.4E-6

, EXTERNAL FLOOD 5.6E-5 5.6E-5 1.9E-5 5.5E-6 f LIGHTNING 2.SE-S 1.0E-7 1.0E-7 2.5E-7

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1.05E-3 3.2E-4 2.0E-4 6.2E E CHANGE IN

,- CORE MELT 7.3E-4 8.5E-4 9.9E-4 NOTE: ALL VALUES ARE PER REACTOR-YEAR OF OPERATION f.

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SUMMARY

OF VALUE-!MPACT ANALYSIS DELTA NET BEN. NET BEN.

CORE OFFSITE 1 PER TOTAL 1 PER ALT. MELT (5 X 10-6) P-REM (1 X 10-6) P-REM 1 7. x-, 1o.1 1es so.2 e l

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STEPS _IN A SEISMIC PRA

1. DETERMINE THE LOCAL EARTH 00AKE HAZARD.

2. IDENTIFY ACCIDENT SCENARIOS FOR THE PLANT WHICH LEAD TO RADI0 ACTIVE RELEASE. (INITIATING EVENTS AND EVENT TREES).

3. DETERMINE FAILURE MODES FOR THE PLANT SAFETY AND SUPPORT SYSTEMS.

L (FAULT TREES)

4. DETERMINE FRAGILITIES (PROBABILISTIC FAILURE CRITERIA) FOR THE i IMPORTANT STRUCTURES AND COMPONENTS.

5. DETERMINE THE RESPONSES (ACCELERATIONS OR FORCES) 0F ALL STRUCTURES AND COPPONENTS (FOR EACH EARTH 00hKE LEVEL).

.6. COMPUTE THE PROBABILITY OF CORE DAMAGE AND RADI0 ACTIVE RELEASE USING THE INFORMATION FROM STEPS 1 THROUGH 5.

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7. ESTIMATE UNCERTAINTY IN COMPUTED RESULTS.

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J SEISMIC HA7ARD CHARACTFRIZATION (A) IF A SEISMIC HAZARD CURVE EXISTS FOR THE PLANT (E.G., FROM THE SEP PROGRAM OR FROM AN EXISTING PRA 0F A NEARBY PLANT)

IT WOULD BE USED.

(B) IF NO SUCH CURVE EXISTS, A HAZARD CURVE SCALED TO AN EXCEEDANCE PROBABILITY OF 2.5E-4/ YEAR AT THE SSE WOULD BE USED. THE SLOPE OF THE CURVE FOR HIGHER PEAK GROUND ACCELERATION (PGA) VALUES WOULD BE ESTIMATED FROM OTHER HAZARD CURVES FOR THE SAME BROAD SEISMOLOGICAL PROVINCE (I.E., CENTRAL STABLE REGION, EASTERN C0ASTAL REGION, SOUTHERN STATES, ETC.) USING NUREG/CR-37S6.

(c) ALL PLANTS WEST OF THE ROCKY MOUNTAINS REQUIRE SITE SPECIFIC HAZARD CURVES DUE TO THE HIGH LEVELS OF SEISMIC ACTIVITY, BUT MOST EXISTING PLANTS HAVE HAZARD CURVES ALREADY AVAILABLE.

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e t SITE SPECIFIC FRAGILITIES DEVEL0i)ED POINT RFACH TURKEY POINT QUAD CITIES COOPER

'RWST

- RWST CST CST CSTk CST 125V BATTERY RACK RBCCW HTX 4160 BUSES CCW HTX 250V BATTERY RACK 125V SWITCHGEAR 480 BUSES DG ACCUMULATORS SBGTS EXHAUST STACK 250V SWITCHGEAR DC BUSES STACK. RCIC LMOV DC BATTERY CHARGER HPIS LMOVs .

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BATTERY RACKS

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Table 5-7. Polat Beach Selenic Response Locattene Response Location / Elevation (ft.l* Spectral Frequency (Mgl Multiple of PGA**

1 Yard 0 EPA 1 2 Yard 8 5 2.3 3 Yard 8 5-10 2 4 Pump House 7 ZPA 1 5 Pump Mouse 7 5 2.3 6 PAD + -19 IPA 1 7 pas -19 7 1.5 8 PA8 -5 2PA 1 9 PAR O EPA 1 10 pas 8 5 2.1 11 PAR 26 2PA 1.0 12 pas 47 2PA 1.3 13 PA8 47 2-5 3.4 14 CRM+ 0 2PA 1 15 CRn 8 5-10 2 16 CRs 8 7 2 17 CRs 26 EPA 1.2 IS CRB 26 5 2.3 19 CRn 26 5-10 2.0 20 CRB 43 2PA 1.3 21 CRn 43 5-10 2.25 22 RR 6 ZPA 1.5 23 RB 21 2PA 1.75 24 RR 46 2PA 2.0 25 R8 66 ZPA 2.25 26 Rn 66 5-10 5.0

'Above mean lake level as on plant drawings.

4 Rey - PAB is Primary Auxiliary Building CRB is Control Room Building RB is t'he Reactor Building

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EARLY COMMENTS ON A-45 WORK THUS FAR:

s REALISTIC, BEST-ESTIMATE PRA APPROACH TO SYSTEM RELIABILITY ANALYSIS IS NEEDED.

e CONSERVATIVE MODELING ASSUMPTIONS DISTORT RESULTS FOR DECISION MAKERS.

o GROUPING "lMPROVEMENT MODS" CAN MASK HIGH COST, LOW VALUE MOD ITEMS, WHEN COMBINED WITH LOW COST, HIGH VALUE MOD ITEMS.

e ALTERNATE COST-BENEFIT METHODOLOGIES (i.E.; ALTERNATES TO 50.109) DO NOT CONTRIBUTE TO ISSUE RESOLUTION.

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e COMMUNICATIONS AND COOPERATION BETWEEN NRC AND NUMARC CAN PROVIDE A POSITIVE PATH ON THESE MATTERS.

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