ML20050A184
| ML20050A184 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Limerick |
| Issue date: | 03/19/1982 |
| From: | Thadani A Office of Nuclear Reactor Regulation |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20033C447 | List: |
| References | |
| FOIA-82-176 NUDOCS 8203310375 | |
| Download: ML20050A184 (25) | |
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Docket 30s. 50-352 50*353 MEMORANDUM FOR: Albert Schwencer, Chief Licensing Branch #2, DL l
FROM:
A. C. Thadant, Chief Reliability and Risk Assessment Branch, DST a
SUBJECT : -
REQUEST FOR ADDITIONAL INFORMATION AND CLARIFICATION ON LIMERICK PROBASILISTIC RISK ASSESSMENT (FRA) - LIMERICK GENERATING STATION Enclosed herewith is the request for additional infonnation and clarification 1
on Limerick PRA submitted by Philadelphia Electric Company.
Please transmit the enclosure to the applicant. A meeting with the applicant will be necessary in the near future to facilitate the resolution of some of our con-cerns.
I suggest a combined meeting and site visit on April 15 and 16,1982.
' h[D. c 9 11 -
A hok C. Thadani, Chief
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Reliability and Risk Assessment Branch Divisiori of Safety Technology Encle cre:
As stated
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S. Hanauer R. Houston D.'Eisenhut J. Hulman H. Ernst W. Pasedag R. Tedesco S. Acharya F. ' Co f fman I. Papazoglou, BNL H. Abelson T. Spets E. Chelliah R. Mattson H. Thompson L
Contact:
E. Chelliah, NRR R. Vollmer
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y;ctogggg REQUEST FOR ADDITIONAL INFORMATION AMD i
LIMERICX PROBABILISTIC Ri5X ASSESSMENT (rRA's !
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This is the first round of questions on the Lim y
are groupec according to the Chapters and Appendices of e ri ck. PRA.
The questions the report.
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CHAPTER 1
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PRA 1.01 The text 'c'onveys the notion 'tha't no cross-ties betwee
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i were taken into account (p.1-18).
n Unit 1 and Unit 2 i
between units (e.g., RHRSW, RHRHX)?dundancy as well as ad Are there cross-ties l--
g considering them in the analysis.
If yes, provide rationale for.not n
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The ultimate containment capability is calculated to be i PRA 1.02
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Given the fact that the containment exhibits leakege und n excess of 140 a
1-19).
ditions, what is the increased leakage rate of the c er design con-reaching 140 psig?
-e ontainment prior to primary 2nd secondary containment was modeled. Provide a description of ho 4
g PRA 1.03 that sore amount of ' containment environmentBased on t
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- p. 1-20), it is expected the reactor building; this may b I
ment is at an elevated pressure.ecome more pronounced when the contain sients, what is the probability of hydrogen combustiGiven the long-time na
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In the event that there
" ?n is containment failureJrior to coreon inside melt, what is the likelihood of hydrogen combusti l
ing? Does. hydrogen combus vate radioactive releases, tion inside the reactor building further aggra-on and if so, in what way?,
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..' l he s urces of data used for th? Limerick study 3RA 1.04 s
categories.
over the other?Please provide criteri.a for the selectionwere summarized into four sources in some instances and not others?What was the rationale for co to deternine whether or not the data base sh diWhat are the guidelines used eral data I
d be integ, rated (p. 1-23)?
o 13A 1.05 fications of the Peach Bottom Station and thIt is stated that 1
ysis, the Technical Speci-i Susinehanna Station were used (p. 1-24 e test frequencies from the this ccmbination as representative of LGS
, 1-25).
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PRA 1.06 What is the rationale for Guideline No.11 (p.1-32)?
Provide the reference for the" improved chronic-health-effects model" I
referred to (p.1-10).
PRA 1.07 " Table 1.2 Summary Of Success Criteria For The Mitigating Systems Tabulated As A Function Of Accident Initist::rs (p.1-26)."
For each t
initiator, including all 5 transients, reference the section of the 1
FSAR that describes the adequacy of the selected success criteria.
For those initiators, including all 5 transients with success criteria not described ir. the FSAR, provida the reference that justifies the adequacy of the selected success criteria.
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CHAPTER'3 PRA 3.01-For the event trees shown in Chapter !!!, provide the reference for the procao111 ties assigned, to each system success or failure and/or frequ-ency of initiators.
Provide rationale sad method used unenever dif ferent probability values are used for the same event.
Identify *<alues obtained from f ault trees ar.d provide cross reference to corresponding fault tree figure.
PRA 3.02 The Lirerick FSAR reported that the vapor suppression systent reliability and effectiveness varies as a function of the LOCA size.
However, in the Liecrick PRA. study, it does not appear that this particular aspect of the I
system has been incorporated into tha containment event trees.
If it was neglecttd, what is the justification?
If it was included, provide addi-tional details on how the system was modeled.'
I PRA 3.03 In addressing manual shutdown as an initiating event, there are situa-tions in which the reactor operator is required to shutdown the reactor in order to be in compliance with technical specifications due to the unavailability of certain safety systems.
Provide a summary of how these types of manual shutdowns were included in the event tree depicted in Fi gure 3.4.2?
PRA 3.04 Plateout and settling is assumed to " remove" radioactivity..
Can the radioactivity be released back to the environment by some 1
physical means, for instance water flash.(p. 3-125)?
PRA 3.05 Isn't the 6' sequence a drywell overpressure and not a wetwell overpres-sure as labeled in the far right column of the containment event trees (see.for example p. 3-82)? Explain the difference between 6. 6' and 6".
The definition of 6 is confusing.
In the containment event trees it reans containrent overpressure either drywell or wetw' ell.
The definition on the top of page 3-133 indicates it is a drywell f ailure.
In Section 3.2, the text states that one of the most important aspects of PRA 3.06 the event tree technique is that it ensures that all of the key accident initiators are identified.
How does the event tree technique identify key initiators, and how does it identify all key initiators?,
PRA 3.07 Further justify the statement in Secticn 3.a.3.2 that potential failures of the reactor p. essure vessel as an initiatinc event havra. very low probability of expected occurrence.
What is the ef fect on the overall consecuences by omission of this event?
PRA 3.08 According to p.1-14, Section 3.2 will discuss the9ubject of complete-Further details as to'why those events noted in tpe se.ction
. ness.
satisfy the completeness requirements, are needed. Have events like, RCP seal f ailure, loss of instrument and control air, loss of DC power, etc.
been examined in the, Limerick study?
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t In order to successfully operate the ADS, it mus* be manually initiated PRA 3.09 5
in a timely f ashion (p. 3-17).
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,I Provide the basis for the time limit on how Loon the depressurization should begin.
i Is there a time limit beyond which cepressurization is not possible?
l Is there any requirement on th'e rate of dep,ressurization?-
It is stated (p. 3-18) that the alternate methods of depressurization are given W probability for success, since they involve."cf eltive operator actions under potential stre.ssful conditions".
Witl* there be approved
, procedures delineating steps. required to implement Ehese alternate
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methods?- Are these methods included.in the cuantification of the se-quence?
Uhy coes a." controlled" manual shutdown require SRV actuation?
PRA 3.10 Isn't the plant scrammed from a power level below the bypass valve capacity?
plain the major secuence of events which are expected for a normal'.re-Ex-actor shutdown.
PRA 3.kl Explain why a value of 1.1x10-4 was used forTne unavailability of RHR/RHRSV or PCS, given a failure of the SRV's to reclose, in Figure-3.a.3.
This is the same 'as for turbine trip or manual scram event trees.
The additional problem of recovering feedwater (the initiatino event) should increase the unavailability as stated on p. 3-27 uncer' event W de-scription.
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.~rovide supporting documentation and/or calculations showing hat a
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feadwater pump can add water to the reactor vessel following a scram and a subsecuent stuck coen SEV.
Tne event trees for turbine trip and MSIV closure show the feedwater availability to be the same, indecencent of
.the condition of the SRV.
It is realized that, should the feedwater pump r.ot be able.to continue running, due te low steam :ressure, the conden-sate pure viculd take over at accroximately 600-700 psig pressure. How-ever; operator actions and additional valve operation wouig..seem to e: ce the.cbability cf successful coarati:n.
Have these ite.s been co.sicerec?
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PRA 3.13 Ine event tree for rianual shutdown has dif ferent feed > tater system un-availabilities depending upon the condition of the SRV's.
Wny does the dif ference exist in this case?
The statement at tfie top of p. 3-20 discusses overriding of the low vacu-um intericcks for the turbine bypass valves.
Have the operator actions required to bypass the MSIV low vacuum interlocks been considered in cal-culating the unavailability of the power conversion system?
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i PRA 3.14 For the MSIV transient, densing cdde was n:t 2*.' luated (;. 3-28).the report indicated ~that the RCIC steam c the RCIC steaming mode was included in the turbina trip event treeOn p.
is this decay heat removal method not consistently included in the anal t
Why y' sis? -
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.a PRA 3.15 In orcergo establish natural ventilation in tne HPCI and RCIC rooms operator action is required.
Is ~this going to be part of the. emergen,cy procedures?
'.y PRA 3.16 Further eltboration on the removal of the edergency core coal'i functionability from the event tree is.requirsd- (p3-40) ng PRA 3.17 Are there any erroneous actions expected'upon a pl and RFPT) due to actual or sensed level swell?i.e., containme ant scram condition, main taken into account in the accident sequehces?
If so, how has this been at Level 2,or Level I?
Is MSIV closure trip. point
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PRA 3.18 Pleasg explain the basis for assigning a reactor scrEm f il
- a 1x10-for
4 large, LOCA (p. 3-42) and 3x10-5 a ure of LOCAs (p. 3-45 & 3-47).
for medium and snall PRA 3.19 :ne text' indicates (p.
- robability'of lona t 3-43) that the success critaria anc calculated injection are simiia'r.erm coolant recirculation and short term coolant' l
short term oemands the same?!!hy are the success criteria for long term and s uraticn between coolant injection and coelent recirculation?What is the diff Gi ten the lonc time nature of some of the accident scenari6s W
Lh system actuation addressed in the Limerick study (fa
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l yes, were'the ceoraded environmental conditions uncer which the syst If must operate taken into consideration?
ems PRA 3.20 t! hat'are the set points for the high radiation interlock for the COR?
What would 'the containment radiation level be from this the COR actually be available?
and would y
i PRA 3.21 Some sequences on the Turbine Trips ATWS eve t t signated as negligible.
ree (p. 3-53) n The TT CgC12 U sequence is 8 ent tree which are assigned probability values such a s on the ame ev-
-' r Ty C D.
This discrepancy is present in other. sequences.
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probability as negligible?What is your criterion for assigning sequence and for' path d
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PRA 3.22 What is the probability of failure for the secondary containment '
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Given the unity probability for a number of the branches with MSIY not open, what do TW, TWE, TA, TAE, TQ, and TQE signify?
r PRA 3.23 The reprt states (p. 3-56) that with multiple relief valves failed open, the RH.s is required to operate successfully.
Is there a time Ifmit on how long multiple relief valves could stay open 4
before Exceeding the capabi.ity of the RHR system? Has this been ac-counted for in the PRA?
PRA 3.24 The T CnC2 sequence (p. 3-57) does not use COR due to "high radia-T tion associated with incipient fuel failure".
Why is there no incipient fuel failure with the T CgR sequence on that same page? A related T
s question is to give.the basis of the 907. MSIV isolation assumption for the T C CT n 2 sequence.
The T*T nR sequence (p. 3-57) states that it is " assumed" that RPT C
PRA 3.25 and FW runback are tripped from the same set of logic and sensors. Are they in fact tripped from the same logic and sensors? What flow rate
,does the FW run back to? Ha's the c'ase been investigated in which the FW runback does occur,.but the RPT ioes not? This would seem to be a core-limiting case, since vessel inventory would be rapidly decreasing.
PRA 3.26 Page 3-69 states that ARI is successful. if, and only if RPT is success '
ful. Provide detailed information on ARI.
PRA 3.27 Page 3-65 shows ARI either working or failing independent of a success or failure of RPT.
Is this consistent with the requirement on page 3-69 that ARI is successful if, and only if RPT is successful?.
c PRA 3~.28 (Top Paragraph, p. 3-86) The statement is made that the diaphragm floor is drained into a sump and the downcocer pipes. This drainage capability eliminates the possibility cf r. molten core dropping u onTlarge mass from the vessel directly into a pool of water. How does this statement apply if containment spray is used? The downcomers are approxicately one foot above the floor level so a large amount of Water can accumulate on the floor prior to the molten core dropping.
It is realized that no j
credit for containment spray has been assumed, but have negative effects, such as the above or excessive steam production, been accounted for?.-
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I Limer,ick takes a 10-2 failure rate per demand for COR (see bet, tem of p. 3-102). A Sandia Report on " Risk Assessment of PRA 3.29
- Filtered-Vented Containment Options for a Mark-1 Containment" shows a Limarick's maximum reduction in core melt probability of a factor of 10.
value icoks to be optimistic.
(See also the bett:m cf p. 3-21).
Explain why Limerick COR has been given an unavailability of 10-2 per demand.
What hYe the bases for the selec, tion,cf the probabilities on the con-tainment event tree? Address each contlinment f ailure mode in detail.
PRA 3.30 Why was-an average value of 10-3 per event used for a cgherent in-fotsa steam PRA 3.31 vessel steam explosion when more detailed values of 10-explosion during a LOCA event and 10-1 for a steam explosion during non-LOCA events were stated on p. 3-114?
Provide supporting analysis and/or calculations to show that RCIC (as PRA 3 32 stated on the bottom of p. 3-104) alone or HPCI alone is adequate for coolant inventory makeup during an ATWS conditidn and does.not result in core meltdown.
Why are transients in which the SRV's fai.1 to open transferred to only FRA 3.33.
the large LOCA event tree?
The.Erergency Operator Guidelines (Rev. 0) state in Steps 'LC-2.5 and SD-2.2.that if the SRY's are cycling, the operator is supposed to manual-PRA 3.34 ly open one SRV to reduce pressure to 150 psi below the SRV set points.
Page C-29 shows the reactor pressure to be cycling about the SRV set points indicating that this action has not been included in the analysis.
Has this operator action been included in the event tree and fault tree quantification and if not, why hasn't it?
N The containment event tree assumes an equal' probability oI containment PRA 3.35 j
failure occurring in the wetwell or drywell.
Interpretation of the in-formation in Appendix J, and the information presented at the February meeting results in our assumption that the containment fait 0re always h
starts at the midwall of the wetwell and rapidly progresses upward to the Thus, both the wetwell and drywell will be failed simultane-drywell.
h This would remove the distinction of whether or not the failure ously.s was in the drywell or wetwell and whether or not the suppression pool has i
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been lost, i.e., all failures rapidly lead to drywell failure and loss of j
'l What is the rationale for the suppression pool scrubbing is irrelevant..
using the containment event tree sequences as presented irn Figure 3.4.14?
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PRA 3.36 Where in the report is the propagation of uncertainties for the dominant "h
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in the table in page 2-122, lodine released to the environment 3RA 3.37 is typed as vapor.
Clarify whether its deposition to the ground seas considered in the subsequent CRAC analysis or not.
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PRA 3.38 Provide discussions as to how the following paracet* rs in Table 3.6.5, e
I which were part of the inputs' to.CRAC, teere 'ditermiped.
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- Time of release Duration of release 1:arnino time El'evati::n of release, and s
Enercy release u
1: bat 5:ere the values of the above rentioned parameters for the 'sequen,ces
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C Y' a"d C Y' in the same table?
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PRA 3.39 Provide the description of those failures that were identified by the study that could disable more than one ADS valve.
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PRA 4.01; Explain in detail how Figure 4.3 was generated. What exactly was used to i,
- ccupute the risk of Limerick at WASH-1400 composite site with WASH-1400 t
data and methods?
l How was, Figure 4.2 " WASH-1400 BWP. with updated methods and data" ob-I tained?
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PRA A.01 Tables A.1.2 and A.1.3 list the anticipated transients considered in the EPRI-SAI study and GE assessment.
Provide clarification as to why tran-9 sients !14-19 and 122 of Table A.1.2 do not appear in Table A.1.3.
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't In Table A.1.3, under Turbine trip with bypass, transients #36 and i37 are indicated; there are no corresponding numsers 36 and 37 in Table f
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A.1. 2.
In view of the EPRI survey and the GE assessment, discuss the major dif-
~ 1 ferences noted in Table A.1.3, e.g., loss of condenser, inadvertent open-ing o h ypass, turbine trip with bypass...etc'.
h PRA A.02 The electric load rejection with bypass valve failure in Table A.1.3 is
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listecl under turbine trip.
Shouldn't this be listed under MSIV closure so as to be consistent with the statement on p. 3-15? This would change the transient initiator frequencies used in the PRA for MSIV closure and turbine trip.
.i PRA 403 In the footnote on page A.12, a statement is made to the ef fect that due to the controlled nature of manual shutdown, there is an increased re-liability of feedwater to maintain reactor inventory.
What is the quali-I tative and quantitative basis for such a statement?
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Provide clarification for the stateme~nt:, "However, coolant injection
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functions, ATWS, and LOCA sequences are not af fected by these initiators when they are quantified (p. A-12)."
PRA A.04 On the. discussion of reported f ailures for all sizes of piping, the sum total of all the failure percentages comes to 64.4% (p. A-13);
furnish information on the remaining 35.6%.
Given the magnitude of the balances.
in failures (35.6%) versus the largest failure category (25.1%), how does ene justify the accuracy of the data if the 35.6% is not included in the data base?
PRA A.05 In addition to pipe rupture, there are other causes which could lead to LOCA, for instance, valves failed open, failure of recirclil'ation pump seals. L'ere th.?y addressed tad preferly included in the analysis as LOCA initiators?
-e-PRA A.06 The use of a 10% reduction of the probability of pipe. rupture for the probability of LOCA seems to be a rough estimate (p. A-16). Are pipe failur.e probability data given per unit length? Are there any data on primary piping ruptures?
If yes, have.they been compared to the 10%,.es-timate?
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- Table A.I.6 gives the probabilities of a LOCA for various cases. Discuss the method, analysis and criteria used in the selection of the Limerick.
values.
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- p. A-22 that "mean values of failure rates used in WASH-l*00 appear lower
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than man values reported in other sources".
Where is this sh::wn?
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PRA A.09 Provide more detail on the modeling of how a-component could fail to run for the duration of the accident.
Also, explain the basis or justifica-
'f tion on how 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> was selected (p. A-29).
PRA A.10 The f act' that values in Table A.2.4 agree does not necessarily mean that the 4 cases can always be indiscriminately used. What criteria was used c
in the.telection of cases in the LGS /PRA?
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(p. A-62, top paragraph, last sentence).
This sentence states that "com-PRA A.11 ponents involved in the roem cooling and ventilation are not included in the estimate of maintenance unavailability".
Page B-5 (bottom) states that " room cooling must be available to maintain acceptable temperatures in the HPCI compartment" for long term operation.
Is there an inconsist-ency in ignoring the cooling system? The same comment applies to RCIC (see p. B-8).
^ PRA 5.12 In the fault tree model.of the diesels (p. A-72), not all the depend-encies are shown, for instance, based on Table A.4.1 if one diesel is out
'of service, the LPCI, both core sprays, remaining diesel generstors and the containment cooling systems have to be operational. Hov are these dependencies accounted for in the fault tree model?
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PRA A.13 The average demand of 65.4/ diesel-year seem.s to be 10., ccmparad to the data given for Zion and Cook. We could not verify this number because of the lack of necessary data regarding Plant X.
Provide additional inforbation (p. A-91).
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w APPENDIX B 9ii PRA B.01 A statement is made (top paragraph, p. B-91) about the large unc6rtainty of brir.ging the reactor from hot to coio shutdown.
Is this consistent with the statement at the bottom of page 1-17 which states that the operation is of a routine nature?
PRA B.02 Explain why ADS pressure sensor is not included in Table B.5.5 1
(p. B-5.3).
PRA B.03 F4re B.9.2 depicts a generic fault tree of a MOV. How are redundant demands on a MOV modeled? If one assumes a situation in which the first demand is to close the valve and the second de=and is to open the valve, how is this modeled in the, study?
PRA B.04 Have DC failures been considered and if so, did they include operator and maintenance contributions?
PRA B.05 There are more than 600 penetrations in the containment. Has their ability to withstand pressure up to 145 psi been evaluated?
Would some of these penetrations, for instance the electrical penetrations, yield to excessive leakage under elevated temperature and pressure environments?
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PRA C.01 How was the metal water reaction of the fuel bundle zirconium channels considered?
Which core melt model and metal-water reaction model is. assumed?
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PRA C.02 The statement is made (top paragraph of p. C-16) that HPCI is allowed to stay on even af ter high exhaust pressure trip point is reached (i.e.,
the operator overrides the interlock). Does the operator have enough.
time to do this since the containment fails in less than 50 minutes?
What were your assumptions?
c PRA C.03 What it the basis for assuming that the diaphragm floor fai.ls at 2/3 of the floor penetration (70 cm) (p. C-19)? What happens to the core after floor failure?
PRA C.04 Please provide the modification that was done to INCOR which tracks the water level in the vessel and assigns 30% power to covered nodes and ce-i cay heat power to uncovered nodes (p. C-15).
'PRA C.05 _ Provide all data for input decks,of,RACAP including documentation of calculations performed in order to obtain the required inputs.
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APPEtlDIX D 1
PRA D.01 Provide the basis for the assuh.ption that 98% of the secondary contain-3 ment building air flow is filtered and 2; is not (p. D-13)?
T _ 0.02 Justify the iise of the CORRA!. value instead of the REACT value for the Tellurium release fraction.
assumption (p. 0-28)?
How sensitive are the consequences to this PRA D.03 in the tabulation of nuclide species, iodine is listed as elemental (a) and/or orga'nic.
In the fission product transport calculations, howeves, iodine is assumed to be Cs! ( Appendix D).
In the estimation of SGTS ef fectiveness, dif ferent DF values for ~ elemental and organic forms are quoted (Appendix D).
Please identify what forms of iodine were assumed in what proportions, and why; then determine decontamination factors consistently for this form (s).
(b)~
Indicate the applicability of the decontamination factors of Table.3.6.4 with respect to fission product element and physical /
chemical form.
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(c)
Section D.2.3.1 states that the SGTS was assumed to achieve cer'tain i
decontamination factors independent of the accident sequence.
The-evaluation.of filtration, systems as ESFs in NUREG-0772, in contrast, indicates susceptability of these systems to plugging as a result
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of hi.gh aerosol loading for some sequences.
Discuss.the particulate loading capability of the SGTS, and compare with the expected aerosol loading (inclu fing non-radioactive materials) f or the various accident sequences.
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What' is the basis of the statement that the 'three conditions listed on page D-8 " dictate" the degree of suppression pool s
decontamination?
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Indicate the relative importance of such variables as degree of subcooling, gas composition (non-s condensible gas.f raction), gas flow rate, and iodine concentration.
4 (e)
On page D-9 it is stated that the reason for increasing the i
saturated pool DF for Csl is the greater solubility of Csl.
i Since saturated pool DFs are limited by reduced surface interaction, as stated on the previous page, explain how a dif ference in solubility of highly soluble compounds can produce an order of magnitude change in DF.
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Since any cesium iodide reaching the suppression pool is in particulate form, explain why Csl is treated differently than-
,j other particulates.
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achieved by pH, particularly in view of the discussion of Csl in the. previous paragraph. _
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PRA D.04 The natoral deposition analysis of WASH-1400 assumed iodine
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p; of Csl discussed in the previous section, explain how the
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WASH-1400 model is applicable.
i FRA D.05 Provide the " data from the TMI accident" which indicate that Csl is_a "much larger constituent than previously believed," and discuss this data with respect to the expected partitioning of a
elemental iodine.
i PRA 0.06 Ihe first paragraph of p. 0-12 states that RB overpressurization results in ground level releases, while the last paragraph k
states that pressurization of the RB would result in release via the SGTS exhaust stack.
Please clarify.
1
' PRA D.07 The discussion of radioactive material inventory and risk as'sociated with the spent fuel pool is inaccurate in several respects-a) The spent fuel pool inventories quoted f rom WASH-1400 are m.!
necessarily applicable to LGS.
Past experience indicates that
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inventories of many discharged cores must be expected to be j
stored in the pool.
As a result, the inventory of several radiologically significant long-lived isotopes (e.g. Sr-g0) may be substantially larger than the core inventory. -The. text
, should be revised accordingly.
3 i
b)N NUREG-CR/0603 discusses risks from Classes 3 - 8 only, and therefore, provides no basis for the claim that rists (including Class 9 events) from spent fuel pool events are
' negligible. This section should be revised to provide a
.j basis for the claims made concerning accidents inv Quing j
the spent fuel pool or, if no such basis is provided, the conclusions should be revised accordingly.
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FRA E.01 What ef'::t d::s th: formati:n of Cs! have en the ;cstulated acc--*
sequencf:? How much Tellurium is oxidized during the various sequences?
Why.is there no Co-58 or Co-60 at the Limerick plant? Why are Cs-134, Cs-136 and Cs-137 inventories so much smaller than WASH-1400 (p. E-30)?
I PRA E.02 In a/Eident sequences in which the containment building fails due to
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steam overpressurization, there is the chance of 'a large amount of " fog" or vapor formation around the building due to the' expanding steam. What effect would this have on the release fractions as calculated by CORRAL?
PRA E.03
-It is not made clear in Section 3.7 or in the Appendix E as to what type of meteorological sampling scheme was ac'tually used in the Limerick site-specific consequences analysis.
Clarify whether it was the stratified meteorological sampli'69 of WA6E-1400 using 91 start times, which is the normal sampling method for the CRAC analysis, for which 8760 consecutive hourly met-data nust 'be input; or.was it a non-standard sampling scheme of in, variant ceteorology Tfor the Start Code 9 of the CRAC IIanual) generated by joint frequency distribution of meteor'ological data over any non,-specific' period f
- (continuous, or with gaps) of tice?
Use of the latter sampling sche:r.e of CRAC is known to result in acute f atality CCDF about an order of nagnitude lower compared to the acute fatality CC0F generated by the forver (the standard) sampling.
Mention of featurcs of both sampling scher.es in Item II in Table E.1.has
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led to this lack of clarity.
PRA E.04 Prov)de justification for use of the numerical values of 25 m'iles, 1.2 mph, -and 0.0 days respectively (See Table E.6) for the evacuation disdance, ef fective speed of evacuation and time lag before evacuat. ion in Limerick site-specific consequence analysis.
Values of these parameters should have been obtained f rom evaluation of*Eirecick site's plans for emergency response within the, plume exposure pathway Emer.gency Planning Zone which is approximately a circular region, centered at the reactor, of about 10-mi ra di u s.
C What is the basis for assuming that 95% of the people participate in the emergency evacuation? How sensitive are the consequences to this assump-tion (p. E-17, 18, 19)?
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PRA E.05 Provide bases for (a) the value of 0.29 used ~for the ground shielding
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factor without sheltering in' Table E.6 in contrast to 0.33 assumed 1
in WASH-1400 for situations of normal activities of people, and 10-4 m3 sec rather than the (b) the. breathing rata of 1,1 :3 J
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standard value of 2.3 x 10-* m fgec, PRA E.06 Clarify as to which j
is the spatial interval over which the shielding factors for sheltering 'in Tables E.2a., E.2b and E.3 were used in CRAC analyses and aMo the impact on the Limerick site-specific consequences.
.in sitq,tions where the emergency response would be the sheltering PRA E.07 a
mode rather than the evacuation or the no-response modes, there l
would still be a time lag before people would actually be in the sheltering mode (due to delay in notification advising people to situation of normal activities of people as as,f actors only for the shelter).
During this tine-lag the shielding sumed in WASH-1400 would be appropriate.
Further, for deriving any benefit from the J
improved shielding f actors for inhalRion (given the sheltering
.i made) it is also necessary to advise the people to open the -
windows and enhance ventilation to expel the contaminated air 1
trapped inside the buildings for exchange with the outside fresh-1 air af ter the radioacti,ve plume has lef t the area.
Unless this latter action were taken, the dose from prolonged inhalation of
.the, contaminated air trapped in the buildings would result in higher doses from plume inhalation exposure pathway (see WASH-1400, Appendix VI page 11-8 and Figure VI 11-5).
Therefore, provide a discussion of the emergency response scenario used and matching analysis of how the shielding protection f actors in Tables E.2a, E.Ib and E.3 have been factored-in in the Limerick site-specific consequence analysis.
N PRA E.08 Provide the following additional information for use in staff's confirmatory Limerick site-specific consequence calculations:
Population input for CRAC standard spatial grid, i.e."for a.
each area element generated by the 16 direction sectors of the compass and the 34 rings of outer radii as specified I
in the description of the Subgroup SPATIAL in page 49 of'the CRAC P.anaul, for the year.1970 and the year 2000.
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Estimates of evacuatior. times including the notification J
times, and travel (respoli~s'e) tihes for clearing a 10-mile plume exposure pathway Emergency Planning Zone under normal and. adverse cortfe-e consist:r.t with the er.;:ected traffic l
loading on the ey.is Q read net-works and for varicus segments of the population (in scnools, factories, hospitals, etc.).
PRA E.09 Provide a basis for the " conservative estimates of saturated pool DF", as well as a reference for the "other data evaluations."
Provig the data for these evaluations and discusa their applica-bility to the accident conditions at LGS.
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not have made any dif ference (from a mathematical convenience point of view) whether Gamma or lognormal distributions are used. A list of all i
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the prior distributions (for each input parameter) along with the important characteristics (like mean, median, or other parameters)
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should be provided.
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PRA G.01 It is stated that the use of mean values in point estimates of a fault tree will result in the mean value of the top. event, provided that the basic events are independent.
This is net t'e case, 5 wever, if the basic events of the fault tree (or a.ny ctI2r t'echnique) include iden-2 tical ccaponents tnat faii indeoendently, out are cnaracterized by the same failure rates.
Was this effect taken into consideration in esti-mating the mean values of the accidant sequences?
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PRA G.02 In orgr to compare results of the Limerick PRA to those of WASH-1400, median. values were estimated for the Limerick results based on cean values. What distributions were assumed in this process? Are the Limerick median results shown in the report?
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3 APPENDIX H'~
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PRA H.01 The Core Dispersal Model descr.ibed~in.'Ap End,ix H. involves a molten jet-exiting the vessel and attacking the concrete. THow does the erosion of concrete influence the strength of the diaphragm floor andpotentially the pedestal wai.:
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2 PRA H.02 The Core Dispersal Model, as described in Appendix H, involves the rapid cool,ing of 50% of the core materials. This appears to be inconsistent with Appendix C, which considers the steam' spike associated with vessel failure to be uncertain.
Hence, containment failure is based on a gradual pressure rise and is predicted to occur seyeral hours af ter ves-sel "fa il ure.
Explain this apparent inconsistency.
PRA H.03 _. There appears to be drains directly below the vessel through the dia-phragn floor covered only by thin steel plates.
These. steel. plates would of fer little resistance to the attack of a colten jet of ccre
. materials at vessel failure.
Failure of these plates would open up a direct path between the wetwell and the drywell.
The core materials could then mix with water increasing the potential for steam explosions and/or rapid steam generation.
Would this increase the potential for containment failure at vessel f ailure?
PRA H.04 The core dispersal model involves large dispersal forces. Rhat.is the ef fect on the integrity of the containment of such large dispershi
"- 4 forces?
PRA H.05 The above questions imply (for those accident sequences with the con-trinment intact at vessel failure) that there would be the potential for contlinment f ailure at vessel f ailure rather than due to gradual over-pressurization failure.
What is the impact of this on the appropriate-ness of the release categories and its influence on risk?
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I APPENDIX I-7-C,' w.
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I PRA I.01 Provide a description of the process used to. discover Limerick plano-specific intersystems ' dependencies and common cause failures.
For example, those compromises in redundancy due to maintenance and testing procedures, HVAC dependencies, AC pc. ear dependence upon DC control (DC control of EDG art), EDG support systems dependencies, the assumption thet equipmerit can
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perfom in the hostile environment resulting from the accident initiator, and location dependent failures.
Include a summary of.Jhe contrasts between the process used to discover inter-systems dependencies at Limerick versus the process used in WASH-1400.
Provide a description of the mathematical and graphical (ET/FT) methods used to accommodate the increased probabilities of failure due to the discovered dependencies.
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