ML20206B218
| ML20206B218 | |
| Person / Time | |
|---|---|
| Issue date: | 08/01/1988 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| REF-GTECI-099, REF-GTECI-A-45, REF-GTECI-DC, REF-GTECI-NI, TASK-A-45, TASK-OR ACRS-2593, NUDOCS 8811150377 | |
| Download: ML20206B218 (37) | |
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&RS- a593 s j DATE ISSUED: 8/1/88 qkSk6 Advisory Committee on Reactor Safeguards Decay Heat Removal Systems Subcomittee Meeting Minutes Resolution of USI A-45/ Status of Resolution of Generic Issue 99 July 27, 1988 Washington, D.C.
PURPOSE: The purpose of the meeting was to: (1) continue review of the NRC Staff's resolution position for USI A-45; and (2) discuss the status of resolution for Generic Issue 99 - loss of DHR in PWR's at shutdown conditicns.
ATTENDEES:_
Principal meeting attendees included:
ACRS NRC D. Ward, Chairman W. Minners, RES W. Kerr, Member R. Woods, RES C. Wylie, Member K. Kniel, RES i I. C6tton, Consultant W. Hodges, NRR P. Davis, Consultant '
P. Boehnert, Staff ERC International -
EPRI/NUMARC D. Ericson
- G. Vine h J. Haugh p W. Parkinson R. Newton -
G. Neils D. Paddelford '.
! MEETING HIGHLIGHTS, AGREEMENTS, AND REQUESTS !
i 1. Mr. Ward indicated that most of the information to be heard on the f A-45 resolution position has been discussed in prior meetings of l the Subcommittee. He noted that NRC has requested ACRS coment on }'
- the A-45 retolution proposal. The discussions on GI-99 are for our g
information at this tine; definitive resolution of this issue is not yet in hand. ;
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L DHRS Meeting Minutes- July 27, 1988
- 2. R. Woods (RES) discussed the proposed resolution of USI A-45. He noted that A-45 was made a USI in late-1980. Key questions to be L
answered were: (1) "Do current regulations provide sufficient
' assurance that risk from DHR failures is acceptably law? and, (2) t Are improvements to DHR function in operating plants cost-beneficial?"
1 The key conclusions of the Staff's resolution effort are:
- DHR failures are significant contributors to core damage frequency from SB-LOCAs, trarsients !
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- Vulnerabilities, corrective actions, and cost / benefit ratios f for fixes are plant specific !
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- The dedicated DHR system is not cost beneficial f
- Designs of DHRS's evaluated were not compared to "current" f OBA-be, sed requirements [
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I Dr. Kerr asked if question (1) above has been addressed by the staff. Mr. Woods indicated that his presentation below will answer i l
I the question to the extent of: is the core melt risk acceptably [
low.
h fiRC assumed a core damage frequency (CDF) goal of 10-5/RY for A-45 {
resolution-consistent with the A-44 and A-49 resolution approaches.
i Dr. Kerr again questioned the term "sufficient assurance" as used above. Mr. Minners indicated that no obvious deviations from N DBA-requirements were seen, but RES did not attempt to determine if :
the plants met the staff's regulations. Dr. Kerr said that work on fiUREG-1150 and the A-45 analyses indicate that the current
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DHRS Meeting Minutes . July 27 1988 f
regulations are not sufficient to assure adequate plant safety.
Further discussion'resulted in Mr. Woods indicating that current ,
regulations do not assure adequate safety and more needs to be done in the area of DHR reliability. Mr.' Ward added that, given that no change to the determistic regulations will address this problem, the staff has concluded that a PRA-type analysis is needed to resolve this issue.
The scope and content of the A-45 regulatory analyses were dis-
! cussed (Figures 1 and 2). A total of six alternative fixes were i considered. NRC has now decided to recommend Alternative 2 - a limited scope PRA to be conducted via the IPE and later on the IPE external event analysis (IPEE).
Technical findings from the A-45 studies were discussed (Figure 3).
The key findings were that the relative importance of vulnerabilities is plant-specific, as are the effectiveness of corrective actior.s. In response to Mr. Ward, Mr. Woods indicted that a change in regulations is not ruled out, given discovery of some IPE-found vulnerability when all plants are evaluated.
Mr. Woods detailed the results of the value-impact analyses for the six alternatives. Three V/I ana*yses were performed (Figure 4 A-C).
Dr. Kerr asked if any fixes are justified if a plant meets the safety goal. Mr. Minners indicated that this is acceptable pursuant to the backfit rule. Mr. Minners said he believes all plants meet the health objectives of the safety goal, and any A-45 fixes take the plant beyond the CDF safety goal (10~4/RY).
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DHRS Meeting Hinutes July 27, 1988 Because many risk contributors are plant-specific, the only way to identify PHR vulnerat.ilities is throt.gh plant-specific examina-tions. The evfects of corrective. accions are plant-specific, and the CDF at many plants is above the staff-selected goal. Given all this, NRC concludes that A-45 should be subsumed into the IPE/IPEE.
NRC reviewed the NUMARC/EPRl/WOG reanalysis of the Point Beach
' plant analysis report conduced in support of the A-45 resolution effort. NRC reviewed the Industry reanalysis for two reasons: (1) ,
to answer the possible claim that Alternative 1 (no action necessary)isjustifiableiftheNUMARCPRAiscorrect,andif Point Beach is a "bounding" plant (then Alternative I would be acceptable); and (2) to provide an example of NRC review of an IPE, and to highlight differences between NRC and industry. Mr. Woods
! showed the results of the NRC review which shows that the difference between NRC and the Industry is a factor of ~ 3.5 (Figure 5). Dr. Kerr said that the results show that the PRAs can
- not be held to an accuracy of less than a factor of ten. Further
! discussion led Mr. Minners to say that the "right" answer is i
believed to be somewhere between the two numbers and one should
- focus on the reasons for the differences seen. Mr. Davis said we i
don't know the "right" answer (for the C0F) and will probably never know, as it's a moving target.
Mr. Woods discussed the rationale for use of a 10-5/RY CDF. Key points noted were:
- The Staff is currently considering proposing more general use of the overall 10'4/RY CDF
- The Staff believes DHR failure related CDF is 1/3 to 1/2 of overall CDF and therefore, one needs a total DHR failure 5
related CDF of + 3x10
DHRS Meeting Minutes July 27, 1988 , ;
To achieve that, one needs a e,Jantifiable DHR failure related !
to CDF of 1x10-5 (to account for operator errors, acts of comission, etc.)
- 10-5 is intended only for A-45. NRC is not implying that the IPE should use this !
i In response to Mr. Ward, Mr. Minners said the above rationale is generally consistent with the requirements of the staff's safety goal implementation plan. Mr. Davis said he believes the DHR failure related CDF is probably less than the 1/3 to 1/2 the CDF -
value noted above. !
Dr. Kerr questioned why IPE's should not use the 10-5 CDF goal
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i vis-a-vis the A-45 analyses. Mr. Minners said the only direction the Comission has given the staff in this area is to require the IPE's. Mr. hard indicated that the IPE program requires additional direction.
. In response to Mr. Davis, Mr. Minners said the real difference between the NRC and NUMARC Point Beach PRA's is the credit given to ope *atoraction(s), i x t f
) 3. G. Neils (NSP) overviewed the NUMARC Working Group's presentation i
l on the industry r.tsponse to the A-45 resolution position. Mr. ,
l Neils said Industry saw the A-45 study as addressing the need for a ,
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dedicated DHR system. He said the Sandia Plant Case Studies were a :
) "modified" IPE and he agrees with NRC that DHR weaknesses are plant I specific, not generic. Other NRC Programs have addressed almost every other DHR vulnerability. Dr. Kerr questioned this, citing ,
Appendix R as a case in roint. Mr. Neils said that in his opinion, j
- the Appendix R fixes will rc^Jee the fire CDF risk to an acceptable value. NUMARC believes that the value of the IPE is the potential 1
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DHRS Meeting Minutes' July 27, 1988 for plant-unique identification of low cost /high value improve-ments.
o Mr. J. Haugh discussed the objective of the reanalysis of the Point Beach Case Study. The central goal is to identify and quantify the conservatisms in the N'RC Case Study and provide a quantitative bases for NRC/NUMARC discussions on A-45 resolution.
Results of technical exchanges between EPRI/WOG and NRC/SNL, reduced original case study estimate of core melt frequency (CMF) due to OHR at Point Beach from 3.0E-4 to 9.3E-5 per reactor-year.
The NSAC-113 estimate is 1.0E-5 per reactor-year. These total CMF estimates include contributions from internal and external ("spe-
)' cial emergency") events. Figure 6 details the principal contribu-tors to the CMFs in the NUMARC study. The main reasons for the
, differences between the NRC/NUMARC CMF's/CDF's result from: choice l of input data, modelling assumptions, and allowance for recovery actions.
Details of the internal events portion of the NUMARC reanalysis l (NSAC-113) was described by W. Parkinson. In particular, he j focused on four specific sequences where significant differences l
wereseenbetweenNRCandNUMARC(Figures 7-14). The focus of the r
discussion centered on the credit assumed given for proper operator l
l actions and plant personnel recovery actions. Mr. Parkinson argued that one of the above sequences (Event Q - struck open relief valve
! - Fiqures 9-10) should be deleted from consideration as it is not a i
j viable event.
4 j J. Haugh discussed the seismic and fire risk contributors for the NSAC-113 reanalysis. For the seismic analysis, NUMARC said the f
j significant contributor to the differences between the NRC/NUMARC l
analyses is the hczard curve used by Sandia (Figure 15).
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-0 DHRS Meeting Minutes July 27, 1988 The fire risk assessment was detailed. Mr. Haugh said this area was where the biggest difference was seen between the two studies (2.2x10-5 for NRC Study - 6.3x10-8 NUMARC Study). In response to Dr. Kerr, Mr. Haugh indicated that the fire risn assessment proce-dure used by NUMARC (Figure 16) is the procedura they recommend for general use, industry-wide.
Discussing the assessment procedure, NUMARC indicated that the central difference between the two analyses is the numerical value assigned to success for fire suppression (automatic and manual).
Dr. Kerr asked what is the relevance of the above information vis-a-vis subsuming A-45 into the IPE. Mr. Vine said the issues of severe accidents and treatment of the IPEs for external events remain outstanding. He said the above external events analyses show that the CDF for external events is a non problem: 1.e. the CDF is acceptably low. Further discussion indicated that NUMARC believes the outcome of the NRC/NUMARC discussion will be applied -
to resclution of the two concerns discussed above. Mr. Ward added that, in his opinion, NUMARC is maintaining that the NRC analyses is overly conservative and thus adds further weight to not requiring backfitting of a dedicated DHRS on all plants.
R. Newton discussed the proposed modifications to the Point Beach plant resulting from issues considered both "old" (e.g. TH! re-quirements) and "new" (A A-45 requirements). Figure 17 lists these proposed modifications to address tha above. concerns includ-ing (it is hoped) the severe accident issue. In response to Mr.
Ward, Mr. Newton said they are not availing themselves c,f the ISAP methodology.
4 D. Ericson (ERC International) discussed the NRC staff's revised estimates based on NRC/NUMARC discussions following the NSAC-113
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DHRS Meeting Minutes July 27, 1988 reanalysis. Mr. Ericson n ad the following general points regard-ing the differences between the two studies: (1) NRC conducted a conserystive generic analysis where the NUMARC analysis was "best.-
estimate for a specific plant (Point Beach); (2) NRC made some as'.umptions where NUMARC obtained detailed data; and (3) one must ce cautious in generalizing the NSAC-113 results beyond the Point i Beach - specific analyses.
Mr. Ericson proceeded to detail the differences between the speci-fic accident sequences discussed in the two reports. For the seismic and fire (Figures 18-19), Mr. Ericson noted that there is discussion on-going that is attempting to resolve the disagreement on the seismic hazard curve to be used. Dr. Beckner (NRC-RES) said NRC is attempting to reach agreement with the industry on just what l ha:ardcurve(s)shouldbeusedtoaddressalloutstandingseismic issues. Figure 20 gives a summary of the differences for the :
t internal and external initiators.
Mr. Davis said he thought that the initial Sandia analyses were supposed to be best estimate, not conservative. Mr. Minners said he believes the Sandia studies gave conservative estimates of accident sequences, given the limits on funding a"ailable and the r need to simplify the analyses. Dr. Kerr expressed concern that extensive backfits may be required based on flawed analyses because
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'i of extensive conservatism. Mr. Minners replied that even if a BE analyses is done, the issue of uncertainty still remains. ;
Dr. Kerr also comented that this is the first reanalysis he has !
been where all the risks ended up being decreased, f L
! Mr. k'ard asked NRC what the Subcomittee should conclude from the ;
j above comparison. The staff said the comparison shows the value of (
1 the IPE, process as plants are making modifications (as noted above) !
.. IRS Meeting Minutes July 27, 1989 independent of NRC urging, which is a healthy sign. Mr. tiinners said the process shows that vulnerabilities have been identified and are being fixed. Mr. Neils said the process shows that a generic fix is not appropriate here.' It also shows that the IPE should be a good tool to catch plant-unique vulnerabilities. Mr.
Newton said he believes the vulnerabilities seen at Point Beach are unique to the age of the plant. He expects newer plants will find a different, probably smaller, set of vulnerabilities.
In response
- Dr. Kerr, Mr. Newton said he believes advanced plant designs address the DHR issue in different ways and a dedicated DHR system is not necessary.
Mr. Ward said are really needs to have a goal (bottom line) of some sort to benchmark to, and he is concerned that the IPE program has no such goal to measure against. Further discusi, ion noted that the safety goal policy statement does not provide definitive guidance
- /is-a-vis regulatory decisions on such issues as A-45.
Mr. Vine provided some concluding comments on the NUMARC prospec-tive for the A-45 studies. Key points noted were:
- The dedicated DHR system fails all cost benefit measures by a wide margin
- Many of the reasons for much lower risk at Point Beach generally apply to other case study plants
- With possible exception of seismic, all other "external risk" contributors selected for analysis by A-45 case studies have been shown to be insignificant
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DHRS Meeting Minutes July 27, 1#88
' Best-estimate analysis is essential for credible useful results. Any additional margin (if needed) should be added at the end of the analysis. This is an mportant lesson for the IPE process
- U.S. nuclear power plant operating experience is the best source of credible data for best-estimate analysis, and the best foundation for "defining the problem" (e.g. USI A-44)
Dr. Catton questioned the use of plant data experience for external i events such as flooding and seismic where very long time intervals are seen between limiting (i.e. extreme) events. ,
! Mr. Ward asked NRC how the industry is to deal with the "A-45
- issue" for future plant designs. Mr. Minners said the staff has no l position here yet, but this is being addressed by T. King under the advanced reactors programs.
- 5. W. Hodges (NRR) discussed the status of the resolution effort for
- GI Loss of DHR in PWRs at shutdown conditions. NRR has i assumed the lead from RES on resolution of GI-99. This was done since the focus of the issue has shifted to resolution of the l concernt, arising from the loss-of-DHR event at Diablo Canyon Unit l 2. Specifically, NRR is addressing the issues associated with loss
! of DHR at mid-loop operation conditions. RES had sponsored a PRA of loss of DHR while shutdown; this study was conducted at BNL.
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Key points noted by NRR were:
' NRR is seriously concerned that PWR operation during decay heat removal system cooling is a significant contributor to the likelihood of a release due to a core damage accident.
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4 DHRS Meeting Minutes July 27, 1988 .
- Loss of DHR at mid-loop operation events continue to initiate at an unacceptably high rate. Two were reported in May.1988.
- Numerous publications and meetings have not led to a solu-tion (s).
- Phenomena associated with pressurization during core boiling have been identified which potentially could lead to severe core damage in a shorter time than previously believed. Other "new" phenomena affect the reactor coolant system (RCS), decay heat removal (DHR) system, instrumentation and other equipment.
- Licensee responses to the Generic Letter on issues related to the Diablo event have not been satisfactory. Some licensees' responses were unsatisfactory in every one of 12 categories evaluated. There is a serious lack of understanding and inadequate preparation for operation was identified. Some licensees are not taking corrective action of any kind. This situation has improved considerably in the past couple of l months as word of the staff's concern has spread. Individual licensees have shown excellent insight into selected areas l
such as RCS draining, containment closure, instrumentation, l
DHR system operation, t.tc. However, such information is not effectively shared.
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- Events continue to initiate which have the potential to become ,
serious. Mitigation planning to prevent core damage is often poor. Planning to prevent a release should core damage occur l
l 1s often nonexistent. Analyses is often nonexistent. Plants are operated in unanalyzed areas where implications are not l
) understood. In response to Mr. Ward, Mr. Hodges said in some cases, no analyses were performed to determine what recovery I methods / procedures were available, given the situation where l
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DHRS Meeting Minutes July 27, 1988 many systems / pieces of equipment are out of service. One-third of the loss of DHR events occurred during mid-loop operation.
- NRR has a three-pronged approach to resolve this issue: (1) initiate a set of""expeditious actions" to assure immediate reduction of the release likelihood of core damage occurs; (2) develop longer term "programmed enhancements;" and (3) modify
%1av) "upd 4tious actions" as programmed enhancements are put in ek w r:gure 21 lists the actions for (1) and (2) above. Key La the above requirements is that procedures must be in place to require fast closure of the containment, given a loss of DHR.
Mr. Ward raised questions regarding the containment closure re-quirement. NRR said they are attempting to strike a balance between lowering the likelihood of a major release versus over burdensome limits on plant operations. Mr. Ward said he didn't understand why a core melt at shutdown is less threatening than at operation. He doesn't see the rationale for requiring "half-closure" of containment vis-a-vis the releases expected during a core melt. Mr. Mazetis (RES) noted that the staff attempted to reduce the overall plant risk (core melt + release), in response to Mr. Ward, NRR said they have not evaluated the containment failure probability for this event. Mr. Ward said he doesn't see the trade off here as the consequences of the ultimate event (core nelt) would be as severe either at shutdown or at power.
Further discussion indicated that the containment hatch is usually open only af ter considerable time has elapsed af ter shutdown.
However, technical specifications do not prohibit mid-loop opera-J tion with an open containment shortly after shutdown (i.e. at high decay heat levels).
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i The longer term programed enhancements ould require independent . ,
hard-wired (not tygon tubing) level instrumentation and temperature monitoring during mid-loop operation. In addition, procedures and ;
analyses will be required to assure plant operations can promptly address any loss of DHR. Technical specifications will be proposed
! to limit key DHR support equipment that could be out of service at any one time.
HRR is proposing to issue a Generic Letter transmitting the above
' requirements to licensees. It is expected that the Letter will be issued in the mid-August time free. In response to Dr. Kerr, Mr..
Hodges said no new regulatiora opnu to k weded to addrets this concern. Messrs. Kerr and Ward indicated that a regulation may be more appropriate here in order to assure unambigucas instructicns to the licensees as to what is required, especially for future l
j plant designs.
Mr. Ward asked if the core melt risk comes mainly from loss of DHR .
at mid-loop operation. Mr. Mazetis indicated that the BNL study showed that loss of core cooling at mid-loop operation was 80% of the core melt risk. Mr. Mazetis said the PRA also showed that even a "leaky" containment (hatch pushed against seal) provided signifi-cant risk reduction ("a factor of 3). In response to Mr. Ward and Dr. Kerr, Mr. Hodges said NRR believes this is a serious problem and requires expeditious action; thus, it is not a candidate for the IPE program.
Mr. Mazetis (RES) said the NRR generic letter on this issue will resolve GT-99 as far as RFS is concerned; GI-99 will not address the issue of deletion of the auto closure interlock between the high and low pressure portions of the OHR system. The slight increase in risk is accomodated by other compensating actions.
Mr. Hodges said he will provide the Subcomittee a copy of the proposed Generic Letter around mid-August.
f DHRS Meeting Minutes July 27, 1936
- 7. Mr. Ward asked the Subcommittee for their views on the resolution approach to GI-99 and A-45. ,
GI-99 P. Davis - No strong feelings; Staff actions appear appropriate.
I. Catton - Approach appears sensible, but he'd like to see the Generic Letter.
C. Wylie - Proposed fixes dre not a big burden and seems to be a sensible approach. One needs to be prudent ,
during shutdown operations.
W. Kerr - !s uneasy about the staff approach. He is not !
completely sure he understands what the true risk is I here. Believes GI-99 resolution probably should be part ;
of the A-45 resolution approach.
D. Ward - Also uneasy; on one hand CDF risk as given by ,
NRR is significant, but, on the other, he is not sure the risk dictates such a stronc response by the staff. ,
t The Subccnnittee agreed that NRR should discuss this issue with the ACRS at its September meeting. Mr. Ward also asked for RES to discuss the results of the BNL risk analysis at the September [
Meeting as well.
US! A-45 W. Kerr - Supports resolution via the IPE. Puzzled as to j how one will separate the seismic risk contributor for the external analysis from the overall CDF though, j l
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C. Wylie - Agrees with Dr. Kerr. Suggest reference be made to ACRS letters on integration of the severe acci- j dent issue into the IPE program, h I. Catton - ACRS Letters on IPE already address the resolution of A-45. He is concerned that the issue has :
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- been "lost" in the IPE rubrick and one should "bite the bullet" and squarely address the problem (s) seen with .
I DHR. Mr. Ward said that the dedicated Dt!R system was ,
given consideration and the system was, in the end, found wanting. Dr. Catton expressed skepticism with the t i results of the value/ impact analysis used to justify l incorporation of A-45 into the IPEs. ;
i P. Davis - Endorse resciution via the IPE's. Concerned i how external events will be handled as most PRA's show this to be a significant CDF risk contributor.
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D. Ward - Endorse A-45 via the IPE but noted that IPE i
process is at present an "empty box." The A-45 case j studies provided a discipline for the selection process 1 that is lacking for the IPE's. He is concerned that a j
pattern is developing for resolution of significant issues (A-45,severeaccidents)inwhichtheStaff l! approach is to develop a process of some sort and then assume something good will come out of the process. He ,
l 1s also concerned that no gem ric lessons or traditional regulatory insight emerged from the A-45 studies. He
' said the risk analyses can't "da it all." INPO has examined the DHR issue and developed generic correnon sense actions to improve the situation, l
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r DHRS Meeting Minutes July 27, 19'88 The Subcommittee decided to bring the A-45 discussion to the ACRS at its August meeting. Presentations were requested from NRC and NUMARC representatives.
- 8. The meeting was adjourned at 3:45 p.m.
NOTE: Additional meeting details can be obtained from a transcript ,
of this meeting available in the NRC Public Document Room, 1717 H Street, N.W., Washington, D.C., or can be purchased from Heritage Reporting Corporation, 1220 L Street, N.W.,
Suite 600, Washington, D.C. 20005, (202) 628-4888.
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O USl A-45 SCOPE i
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SIX CASE STUDIES, DHR FAILURE RELATED PRA'S (SUMiiARIZED lit ENCLOSURE B)
LIMITED TO SYSTEMS NEEDED TO RESPOND TO TRANSIENTS AND SMALL-BREAK LOCAs EVALUATED SUCH SYSTEMS' VULNERABILITY TO FIRE, FLOOD, SEISMIC, INSIDER SABOTAGE 4
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CONTENT OF REGULATORY ANALYSIS APRIL, 1988 DRAFT REVISION DESCRIBES SIX ALTERNATIVES AND PROPOSES RESOLUTION WITil ALT. # 2 (PLANT-SPECIFIC ANALYSIS)
ALTERNATIVE 1 - NO ACTION COULD BE ACCEPTABLE 1E NRC ANALYSIS RESULTS ARE CVERLY CONSERVATIVE (EPRl/WOG: POINT BEACl!)
/LTERNATIVE 2 - LIMITED SCOPE PRAs SEVERE ACCIDENT PROGRAM IPE/IPEE ALTERNATIVE 3 - SPECIFIED SYSTEMS MODIFICATIONS USIs AND Gls ALTERNATIVE 4 - DEPRESSURIZATION AND COOLING PWR - FEED AND BLEED BWR - CONTAINMENT VENTING ALTERNATIVE 5 - DEDICATED HOT SHUTDOWN CAPABILITY ALTERNATIVE 6 - DEDICATED COLD SilVTDOWN CAPABILITY 3
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TECHillCAL FINDINGS FREQUENCY OF CORE DAMAGE DUE TO DilR FUNCTION FAILURE (P(CM)DHR 1 AVERAGES 2 TO 3 x 10-4 PER R-YR (INCLUDE INTERT1AL AND EXTERfiAL CAUSES)
SUPPORT SYSTEM FAILURES (E.G., EMERGEf'CY POWER, SERVICE WATER, COMP 0iiEf1T C00LliiG) C0fiTRIBUTE SIGNIFICANTLY TO P(CM)DHR
- REDUf1DAliCY CONCERNS AND C0f4SIDERABLE SHARING 0F SYSTEi1S, PARTICULARLY AT SUPPORT SYSTEM LEVEL FOR SOME PLANTS CONCERNS WITil OVERALL GENERAL ARRAllGEMENT OF EQUIPMENT FROM A SAFETY VIEWP0lf1T, E.G., LACK OF INDEPENDENCE, SEPARATION -
& PHYSICAL PROTECTIOil 0F REDUNDAlli SAFL60ARD TRAltlS i
- FIRE, FLOOD, SEISMIC, SA30 TACE RISK CONCERNS RELATIVE IMPORTAf4CE DE VULNERABILITIES IS PLANT-SPECIFIC EEELc_TIVENESS QE CORRECTIVE ACTIONS jl{ REDUCING P(CWp;;g 11 ELMJI SPECIFIC
/$ h
i REG. ANALYSIS DIFFERENT APPROACHES TO VALUE-IMPACT ANALYSIS VALUE-lMPACT ANALYSIS PERFORMED 3 WAYS:
A. VALUE TERM: AVERTED DOSE TO POPULATION IMPACT TERM: COST OF IMPLEMENTAT10fl B. VALUE TERM: SAME AS METHOD "A" EXCEPT AVERTED ONSITE DOSE INCLUDED IMPACT TERM: COST OF IMPLEMENTAT10tl LESS AVERTED ONSITE COSTS C. SAME AS METHODS A & B PLUS THE SAVINGS FROM SPECIAL CONSIDERATIONS (E.G., SAB0TAGE, MORATORIUM, RESOLUTION OF OTHER GENERIC ISSUES, UNCUAtlTIFIABLES)
RESULTS:
METHOD A - ALTERNATIVES 2, 3 a 4 MAY BE COST-EFFECTIVE
- METHOD B - ALTERNATIVES 2, 3 a 4 MAY DE MORE COST-EFFECTIVE l
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EPRl/WOG STUDY RESULTS (NSAC - 113)
CORE DAMAGE FREQUEf1CY PER YEAR SOURCE OF RISK NRC EPRl/WOG (RF)* REVISED NRC (RF)*
IllTERNAL 1,4E-4 2.5E-5 (6) 2:6E-6 (54)
SEISMIC G,1E-5 7.4E-6 (8) 4.1E-5 (1,5)
FIRE 3.2E-5 6.3E-8 (500) 2.2E-5 (1,5)
INTERNAL FLOOD 7.75-5 1,0E-8 (7700) 9.8E-7 (79)
EXTERNAL FLOCD 1,9E-3 1,0E-6 (2) - -
WIND 4.0E-6 1,0E-8 (400) 1,7E-7 (24)
LIGHTNiliG 5,8E-8 1,0E-8 (6) - -
TOTAL 3,1E-4 1.0E-5 (31) 9E-5(J.f) 99)
'8 EDUCT 10N EACTOR COMPARED To "NRC" 9
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PRINCIPAL CONTRIBUTORS TO REVISED CASE STUDY CMF .
THE PRINCIPAL INTERNAL EVENTS CONTRIBUTING TO THE CHF ESTIMATES INCLUDE:
LONG TERM STATION BLACK 0UT (LTSB SE0VENCES)
SMALL BREAK LOCA WITH FAILURE TO IMPLEMENT SUMP RECIRCULATION (S2MH1'H2' SEQUENCE).
LOSS OF 0FFSITE POWER TRANSIENTS (TIMLE SE0VENCE)
REACTOR / TURBINE TRIP TRANSIENTS INVOLVING A STUCK OPEN RELIEF VALVE (T30H1'H2' SEQUENCE)
THE PRINCIPAL EXTERNAL EVENTS CONTRIBUTING TO THE CHF ESTIMATES INCLUDE:
SElSMIC SE0VENCES FIRE SEQUENCES (AFW PUMP ROOM AND 4160V SWITCHGEAR ROOM)
THE CONTRIBUTION DUE TO SEISMit AND FIRE SEQUENCES IS LARGER THAN THE CONTRIBUTION DUE TO ALL INTERNAL EVENTS
- _ - _ _T,
- SMALL-BREAK LOCA AND FAILURE T0 IMPLEMENT SUMP RE S2MH1'H2': SBLOCA + MFW FAILURE
- HPRS 8 LPRS FAILURES CASE STUDY: 4.7E-5 REVISED CASE STUDY: 7.0E-6 NSAC-113: 5.8E-7 PR!NCIPAL DIFFERENCES: .
- 1. SBLOCA FRE0VENCY CASE STUDY: 2.0E-2 BASED ON LEAKS <2-IN. DIA: DERIVED FRO ANO-1 IREP (MURLEY MEMO: ISOLABLE LOCAS '
DOMINATED)
NSAC-113: 3.0E-3 BASED ON LEAKS <2.0-IN DIA DERIVED FROM CREDITS OCONEE PRA AND INDUSTRY EXPERIENCE.
ISOLATION PRIOR 10 REClRCULATION AT -20 HRS SUMP RECIRCULATION NOT REQUIRED FOR ANY EXPERIENCED S0 FAR l i
l STATUS: REVISED CASE STUDY ACCEPTS 3.0E-3 SBLOCA l
FREQUENCY t
- 2. OPERATORS Fall TO IMPLEMENT SUMP RECIRCULATION .
CASE STUDY: lE-3 REVISED CASE STUDY: lE-3 ,
NSAC-ll3: lE-4 BASED ON POST TH1 ERGS /EOPs AND ST REQUIREMENTS AS WELL AS LONG TIME TO DE RWST STATUS: NO AGREEMENT
SMALL-BREAK LOCA AND FAILURE TO IMPLEMENT SUMP RECIRC (CONTINUED)
- 3. RECOVERY FROM R'. CIRCULATION FAULTS CASE STUDY: CREDITED LOCAL VALVE OPERATION TO ESTABLISH RECIRC FLOW PATH IN SOME INSTANCES REVISED CASE STUDY: SAME AS AB0VE ilSAC-ll3: FULLY CREDITS POINT BEACH PROCEDURE ECA-1.1: INCLUDES LOCAL VALVE OPERATION AND RWST REFILL (BUT NOT APPLIED WHEN OPERATOR FAILS TO IMPLEMENT REClRCULATION.)
STATUS: NO AGREEMENT EPR!/WOG CONCLUSIONS THE CASE STUDY TREATMENT OF OPERATOR FAILURE TO SWlic. HOVER TO RECIRCULATION IS INCOMPLETE. A COMPLETE APPLICATION OF NUREG/CR-1278 WOULD YlELD A LOWER PROBABILITY OF FAILURE. THE NSAC-113 METHODOLOGY IS BASED ON SIMULATOR DATA NSAC-113 FULLY CREDITS THE POINT BEACH PROCEDURES THAT ADDRESS RECOVERY FROM EQUIPMENT FAILURES DURING SNITCH 0VER
EVENT 0: STUCK-0 PEN RELIEF VALVE T30H1'H2': REACTOR / TURBINE TRIP (MFW AVAILABLE) +
SRVs Fall TO CLOSE + HPRS & LPRS FAILURE CASE STUDY: 2,5 E-5 REVISED CASE STUDY: 3.6E-6 NSAC-ll3: NOT APPLICABLE I
PRINCIPLE DIFFERENCES:
CASE STUDY: SRVs ASSUMED TO HAVE A 7% CHANCE OF OPENING IMMEDIATELY FOLLOWING TRIP:
PORVs ASSUMED TO BE BLOCKED WEli C2 0 5A;O CONTINUOUSLY REVISED CASE STUDY: SRVs ASSUMED TO HAVE A 1% CHANCE OF OPENING IMMEDIATELY F0Lt.0 WING TRIP 4
BASED ON SNL RSSMAP STUDY FOR A B&W PWR NSAC-113: EVENT 0 SEQUENCES 00 NOT EXIST FOR REACTOR OR TURBINE TRIPS (T3) AT PT, BEACH: NEITHER PORVs NOR SRVs WILL BE CHALLENGED BASED ON WESTINGHOUSE PWR OPERATING EXPERIENCE EVENT 0 CONSERVATIVELY INCLUDED THE OPENING OF BOTH PORVs IN ALL LOSS OF 0FFSITE POWER (TI) AND LOSS OF MAIN FEEDWATER (T2) SEQUENCES STATUS: NO AGREEMENT
EVENT 0: STUCK-OPEN RELIEF VALVE (CONTINUED)
EPRl/WOG CONCLUSIONS:
THE REVISED CASE STUDY IS NOT CONSISTENT WITH THERMAL HYDRAULICS AND OBSERVED W PWR EXPERIENCE, AND IS INCONSISTENT WITH CURRENT PRA PPACTICE (1.E.,
BOTH INDUSTRY AND NUREG-llSO STUblES)
,/
- flb0
LONG TERM STATION BLACK 0UT CASE STUDY: 3.6E-5 REVISED CASE STUDY: 9.9E-6 :
I NSAC-ll3: 5.flE-7 l l
PRINCIPAL DIFFERENCES:
l CASE STUDY: -
USED IREP DATA FOR DIESEL GENERATOR FAILURE AND 8
- HOUR MISSION TIME USED LOSS OF 0FFSITE POWER FREQUENCY BASED ON i NUREG-1032 GENERIC DATA a
RECOVERY INCLUDED OFFSITE POWER AND DIEEE.L l j GENERATOR REPAIR ;
) l NSAC-ll3:
- l USED NSAC-108 DATA FOR DIESEL GENERATOR FAILURE !
(WITH 2-3 HOUR EFFECTIVE MISSION TIME)
USED LOSS OF 0FFSITE POWER FREQUENCY BASED ON !
PLANT SPECIFIC DATA !
! - RECOVERY INCLUDED OFFSITE POWER, REFILLING CST !
! AND BALANCING SERVICE WATER LOADS TO DIESELS [
i !
- L
! STATUS:
- REVISED CASE STUDY ACCEPT:i NSAC-113 LOSS OF
) -
0FFSITE POWER AND DIESEL GENERATOR DATA !
NO AGREEMENT ON RECOVERY ACTIONS l l l .
?
LONG TERM STATION BLACK 0UT (CONTINUED)
EPRl/WOG CONCLUSIONS:
i
. r.ST REFILL RECOVERY ACTION IS REASONABLE: HOWEVER, PLANNED POINT BEACH MODIFICATION WILL PROVIDE ADDITIONAL 200,000 bALLONS OF CAPACITY, ELIMINATING THIS CONCERN ;
l SERVICE WATER RECOVERY IS NOT SIGNIFICANT TO THE FINAL CHF '
ESTIMATE l
l RECENT GAS TURBINE GENERATOR RELIABILITY DATA FROM POINT '
BEACH INDICATES LTSB FREQUENCY SHOULD BE REDUCED BY A FACTOR OF 2 TO 3 ,
I i
i
(
LOSS OF 0FFSITE POWER -
TIMLE: LOSP + MFW FAILURE + AFW FAILURE + F8B FAILURE (PREDOMINANTLY SHORT TERM STATION BLACK 0UT)
ESTIMATES: CASE STUDY B.7E-6 REVISED CASE STUDY: 4.9E-6 NSAC-ll3: 7.7E-7 PRINCIPLE DIFFERENCES:
- 1. USE OF NEW STATION BATTERIES CASE STUDY: ANALYSIS CONDUCTED PRIOR TO INSTAL".ATION OF NEW BATTERIES NSAC-ll3: INCLUDED OPERATOR ACTION TO USE NEW BATTERIES TO START DIESELS AFTER INDEPENDENT OR COMMON i: CAUSE FAILURE OF STATION BATTERIES STATUS: REVISED CASE STUDY HAS NOT YET CREDITED NEW j BATTERIES IN THIS SEQUENCE (NEW BATTERIES l
CREDITED IN REVISED CASE STUDY SEISMIC i EVALVATION)
- 2. LOSS Oc 0FFSITE POWER FREQUENCY CASE STUDY: 8. tie-2 BASED ON NRC GENERIC ESTIMATE (NUREG-1032)
NSAC-ll3: 6.2E-2 BASED ON PT. BEACH SPECIFIC DATA STATUS: REVISED CASE STUDY ACCEPTS THE NSAC-ll3 DATA
/m.Til
- e 1
2.
t LOSS OF OFFSITE POWER (CONTINUED) :
l
- 3. DIESEL GENERATOR FAILURE PROBABillT'i i CASE STUDY: 3.8E-2 BASED ON IREP GENERIC DATA ,
~ :
> NSAC-ll3: 2.2E-2 BASED 0N NSAC-108 l STATUS: REVISED CASE STUDY HAS NOT YET ACCEPTED I l
DIESEL DATA IN Thid SEDUENCE (BUT CREDITED I 1 LONG TERM STATION BLACK 0UT) '
i EPRl/WOG CONCLUSIONS: '
1 l
- INCORPORATING NSAC-113 DATA AND RECOVERIl CREDITED IN OTHER REVISED CASE STUDY SEQUl REDUCES THE REVISED CASE STUDY CMF TO 8.6E-f
, L l l 4 :
l ;
I I 1 .
- 1 1 _
\
SEISMIC SEQUENCES CASE STUDY: 6.lE-5 REVISED CASE STUDY: 4 lE-5 NSAC-113: 7.4E-6 PRINCIPAL DIFFERENCES:
- 1. HAZARD CURVE CASE STUDY: -
GENERATED A SCISMIC HAZARD CURVE BASED ON ZION (SSMRP) AND CALCULATED LOCAL S0ll COLUMN EFFECT.
NSAC-il3: -
SEISMIC HAZARD CURVE WAS REDUCED FROM CASE STUDY VALUES BY FACTORS OF TWO FOR <3*SSE AND FIVE FOR
>3*SSE BASED ON EPRI VS LLNL EVALUATIONS OF BRAIDWOOD STATUS: NO AGREEMENT
- HAZARD CURVE BEING ADDRESSED SEPARATELY--SEISMIC MARGINS RESEARCH e
lm wi .
t FIRE RISK ASSESSMENT PROCEDURE
[ ,_ _ _, l IIRE . GENERIC FREQUENCY POSTULATE TRANSIENT I INITIATORS I
APPROPRIATE TO BLDG. COMBUSTIgLE fzgg l
l 4 FREcutNCY l l SORT BY APPLICABLE f!RE TYPES
- APPLICABLE TO EACH f!RE ZONE IN BLDG.
- =
' I APPLY WEIGHTING FACTORS APPROPRIATE I TO EACH f!RE TYPE IN EACH FIRE ZONE l
_ -_ . . ._. _ _. _ _ _ 4 -. _ _ - - _ _ - -
FIRE CALCULATE FIRE ENERGETICS APPROPRIATE l SEVERITY
' To EACH f!RE TYPE IN EACH f!RE ZONE l I APPLY CONDITIONAL PROBABILITIES ron l l
ENERGETICS APPROPRIATE TO EACH FIRE l TYPE IN EACH f!RE ZONE
_-..___._---g----__----. --\
FIRE EVALUATE SUCCESS Or AUTO. SUPPRESSION l '
, SUPPRESSION
' I
{ j l EVALUATE SUCCESS or MAN. S U P P R E S S I O N -<-
_ _ _. _ _ _ - - ; - _ _- _ .- - - - - - -- l RECOVERY EVALUATE RECOVERY l g _ _ _ _ _ _ _ _ 4 -. _--;
EVALUATE CHf
{_ _ _ _. __ _ -- _ - - - W i
I l
i f ___
j):=Jg'. //l -
PROPOSED MODIFICATIONS ,
.. l
- INSTALL SWITCHGEAR TO BY-PASS 4160 SWITCHGEAR ROOM i f
- MODIFY 13.8 KV BUS SECTION
- INSTALL A THIRD SAFETY RELATED EMERGENCY DIESEL GENERATOR (EDG) WITH ASSOCIATED SWITCHGEAR, I t
BATTERY, COOLING, AND BUILDING.
l l
- MODIFY THE 4160 VOLT SYSTEM TO FACILITATE THE l 2
NEW EDG -
l l !
!
- INSTALL THIRD LOW VOLTAGE STATION AUXILIARY i TRANSFORMER ;
- INSTALL A SEISMIC MAKE-UP WATER STORAGE TANK ,
l
! (MUWST) l I
i
- INSTALL AN UNDERGROUND CONNECTING DUCTBANK f i
! BETWEEN THE UNIT 2 FACADE AND NEW EQUIPMENT f l
- LOCATION OF THE NEW EQUIPMENT BUILDING WOULD BE 1 IN THE AREA CURRENTLY OCCUPIED BY WAREHOUSE 2
a COMPARISON OF SEQUENCES (CONTINUED) .
t l P(CM) PER RX-YEAR SE_QUENCE _ ORIGINAL EPRI/WOG R_EVISED
\
l SEISMIC 6.1 E-05 7.4 E-06 4.1 E-05 l
l RATIONALE:
AGREED WITH CREDIT FOR ADDED BATTERIES 1
DISAGREED WITH EPRl/WOG RECOVERY POST-CUAKE D!SAGREED ON SURVIVABILITY OF RWST i
e ik, i
s
4, I
~
COMPARISON OF SEQUENCES (CONTINUED) i l P(CM) PER RX-YEAR SEED R E N_C_E DRIGINAL, EPRl/WQG REVISED
! FIRE 3.2 E-05 6.3 E-08 2.2 E-05 l
i i
RATIONALE- i
, AGREED WITH CREDIT FOR SECOND HALON SYSTEM IN AFW ROOM 1 !
l DISAGREED WITH LOWER HALON FAILURE RATE ,
l 1
[
wj l3i
- i
_ _ _ _ - _ _ ._ __ . _ _ - _ - - - _ . _ _ _ - . _ _ . . . . - . _ . = _
o ,
i 2 j
SUMMARY
P(CM) PER RX-YEAR
$_EQUENCE j)RIGINAL EPRI/WO_G REVISED _
INTERNAL 1.3 E-01 2.5 E-06 2.5 E-05 EVENTS EXTERNAL 1.7 E-04 7.5 E-06 6.4 E-05 EVENTS TOTAL 3.0 E-04 1.0 E-05 8.9 E-05 3
N' ' .
b' t
N.
. . s .
ci, S. EXPEDITIOUS ACTIONS .
- 1. CONTAINMENT CLOSURE
- 2. RCS TEMPERATURE
- 3. RCS LEVEL
- 4. DO NOT PERTURB RCS
- 5. BACKUP EQUIPMENT
- 6. HOT AND COLD LEG CLOSURE
- 7. DISCUSS DIABLO CANYON IMPLICATIONS WITH OPERATIONS STAFF C. PROGRAMMED ENHANCEMENTS
- 1. INSTRUMENTATION
- 2. PROCEDURES
- 3. EQUIPMENT g 4. ANALYSES sk 5. TECHNICAL SPECIFICATIONS
( -
s i
._ _ . . . . _ _ . - _ _ _ . . . _ - _- _. _. _ _ _ - -