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| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, INDIVIDUAL PLANT EXAM (OF EXTERNAL EVENTS) | | document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, INDIVIDUAL PLANT EXAM (OF EXTERNAL EVENTS) | ||
| page count = 84 | | page count = 84 | ||
| project = TAC:M74379 | |||
| stage = Request | |||
}} | }} | ||
Latest revision as of 05:21, 23 September 2022
| ML20116C351 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 10/26/1992 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-88-20, TAC-M74379, NUDOCS 9211030334 | |
| Download: ML20116C351 (84) | |
Text
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E4 r Va er Power Stai.on
' thippingport. PA 15077.OfYa4 141M 393 SM.$
October 26, 1992 JonN o siusta Vee Presatent . 6dacy.et Group O. S. !Juclear Regulatory Commission Attn: Document Control Desk Wanhington, DC 20555 l
Subject:
Heaver Valley Power Station, Unit No. 2 Docket. No. 50-412, Licenne No. NPF-73 i
Generic Letter 88-20 (TAC No. M74379)
References:
- 1. liRC Letter tc Duquerne Light Company (DLC),
Generic Letter 80-20 Individual Plant
] Examination (IPE) For Severe Accident Vulnerabilities - Request For Additional Information (TAC No, M74379), dated July 15, 1992
- 2. DLC Letter to the NRC, Generic Letter 88-20, dated August 17, 1992
- 3. DLC Letter to the NRC, Generit: Letter 88-20, dated September 11, 1992 Please find attached the second submittal of Duquesne Light Company's responses to the NRC's P2 quest for Additional Information (RAI), Reference 1. Our plan to provide two submittals in response to *he Rn1 is stated in Reference 2. The responses from our first submittal (Reference 3) are included herein for convenience.
Should you nave any questions regarding this submittal, please
, contact Ed Coholich at (412) 393-6224.
Sincerely,
'J (. D,/
. S bw ber
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Attachment cc: Mr. L. W. Rossbach, Sr. Resident Inspector Mr. T. T. Martin, URC Region I Administrator Mr. A. W. De Agazio, Project Manager Mr. R. R. Janati, Pennsylvania Department of Environmental Resources Mr. M. L. Bowling (VEPCO) i d
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ADDITIONAL INFORMA110N POR BEAVER VALLEY [fNIT 2 INDIVIDllAL P1WT EXAMINATION Ouestion 1. a) Describe briefly the peer reviev performed on the Individual Plant Examinatici (IPE) to help assure the - analytic techniques used in the back-end analysis vere correctly applied. Identity specific areas reviewed, experti af the reviewers, and characterize the peer review findit, .d ar/ i significant comments.
h) As an example of the internal review performed,-provJda a i copy or sommary of peer reviev somments and resolutions M i appropriate) for aspects of the Probabilistic Risii l Assessment involving the " Emergency Svitchgear Ventilation" i t ma system analysis through event tree quantification, plant improvements and conclusions.- .
r Response 1. a) The analytical techniques used in the back-end analysis of the Beaver Valley Unit 2 1PE vere developed by PLG, Inc., ;
and applied using results from previous NRC and industry analyses. Particularly. heavy emphasis was placed on the '
Surry analysis in NUREG-1150, since Beaver Valley and surry-are similar plants. The. back-end analysis vas reviewed :
within Duquesne Light by the Radiological Engineering group and Nuclear Engineering group to assute that the patameters- ,
used in the input appropriately described the Beaver Valley plant. The analysis was also reviewed .vith'Sandic,.thei principle coritributors - to' the back-end analysis- for the -
NUREG-1150 analysis of -Surry. Their comments vere to .
Include provisions within the model to reduce the heactor .
Coolant System pressure prior to vessel break by way of-stuck open PORVs and Reactor- Coolant- Pump seal leaks. Ve had previously only modeled induced steam generator tube-ruptures and induced hot leg failures. as pressure reducing -
mechanisme, b) The intetnsi review focused mainly on system design and operation, and on Emergency' Operating Procedures. . Thete
- vere no specific comments ~ on the Emergency Svitchgear ~
Ventilation System in the formal internal-reviev, however, the system mode), and the success criteria, vere discussed- ,
with plant personnel during the development of the system analysis. Proposed plant - improvements were discussed with-
- the Unit 2 Operations Manager and General Manager. ;
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Question 2. Describe how containment loading vas. assessed -for each of the l Containment Event Tree (CET) end-states. Discuss the development I of plant-specific probability distribution functions of failure ,
likelihood for the range of failure pressures. l Response 2. As discussed in Section 4.4, "The Beaver Valley Unit 2 containment is very similar in design- to the Surry Unit-1 ,
containment", which was analyzed extensively in UUPEG-ll50. Both containments vere designed and constructed by-the Stone & Vebster Engineering Corporation (SVEC), a member of the BV-2 back-end analysis team. Based on a detailed review of the .BV-2 containment design and a comparison to the Surry Unit I design, it was concluded "vith a high degrer- of confidence that the ,
failure distributions'for Beaver Valley Unit 2 and-Surry Unit I '
containments vould be similar, -and that use the Surry distribution would be somewhat conservative for-the Beaver. Valley ;
Unit 2 containment". Based on this conclusion, the HUREG/CR-4551 v Carry Unit I distributions for containment failure pressure and conditional probability that the failure vould be large vere utilfred without modification in the BV-2 study.
As shown in Figure 4.5-1, the CET has 12,463- end states.
Therefore, it is assumed that this request is directed at the ;
broad categories of end state discriminators as related to CET Top Events C1, AP, C2, CE which address early containment t failures, and. Top Events C3 and C4 which address late containment- ,
failures.
No pressure loading considerations are addressed for Top Event C1' which addresses containment failure prior to vessel breach.- In the BV-2 IpE, this top event addresses only whether the containment is isolated.- I t vas assumed, as~1t was in NUREG/CR-4551, that containment threats (blowdovn or hydrogen burns) prior to-vessel breach could be ignored.
CET Top Event AP addresses containment failures due~to in-vessel >
steam explosions. Containment loading was net-evaluated for:this ;
top event. For the failure ' branch of its top event, it was assumed that the containment would fail. Containment failures, ,
resulting from_In-vessel steam explosions, were assumed to be large.
I CET Top Event C2 addresses ~ the containment loadingint vessel breach. Because of the similarity between the Surry: plant- ~
analyzed in NUREG/CR-4551 and BV-2, the containment. pressure rise distributions developed for the former vere; adapted to BV-2, with
- minor adjustments to account. for slight- differences 'in'the <
containment volume and pover ratings. These looJs distributions vere compared to the fa 'ure of various CET paths.
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i 140 specific containment loada vete calculated for Top Event Cf.
vhich addresses containment failure within four (4) hours of f vessel breach, Cie to hydrogen burns within that time period, [
including those that occurred at vessel breach in the absence of -
IIPMB. MAAP analysis performed for 3V-2 indicated that the amount .
of hydrogen generated in-vessel for most BV-2 sequences was i typically of the order of 700 J.bm (equivalent to the oxidation of i approximately 40% of the core Zircaloy). MAAP also indicated l' that the quantity of hydrogen generated ex-vessel in this time period was relatively small. Therefore, the primary source of i hydrogen in this four (4) hour time frame is that vhich is 4 -
produced in-vessel. Furthermore, the concern regatding signific_.it h;drogen burns during this time period applies only 1 to scenarion in which the steam concentrations in the containment atmosphere are lov (i.e., when containment sprays are .in ;
operation).
?vr scenarios in which the containment sprays are operating, it is likely that hydrogen burns vill occur at low concentrations if ,
hydrogen is "slovly" released into the containment. Only when the hydrogen is suddenly released into the containment (e.g., due to an induced failure of the hot leg or et vessel breach) vill l l
the hydrogen concentrations achieve significant values. -When vessel breach is accompanied by HPME, the containment. loads- 1 discussed for Top Event C2 include _the contribution of hydregen' ;
burns. However, for " pour" type vessel breaches at high- .
pressure, there could be a sudden release of-hydrogen into the
- reactor cavity and then into the containment. For those scenarios'in vnich there was a sudden release of hydrogen into a non-steam inerted containment atmosphere, it was assumed-that-if -;
the global concentration exceeded 12%, a burn vould occur which . 'l vould,-in turn, fail the containment. The intermediate logic implicit in this assumption is as follows:
- 1. A deflagration at a 12%-hydrogen concentration is not likely : '
to fall the' DV-2 containment (based on peak containment pressures determined using the adiabatic burn assumption). . ,
- 2. Although MAAp simulations showed . hat the containment was '
well mixed when sprays were, in _operatiun,-it1va.= assumed that local concentrations could be 20% higher than_the global concentration.
3.- Although the BV-2 containment _ configuration is. not necessarily amenable .to. a Deflagration to Detonation-l Transition (DDT), it was assumed- that a DDT vould occur if '
!- local concentrations exceeded a value of 15%. (minimum alue l- reported in Reference 4-8). '
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- 4. It was assumed that DDT vould result in a large containment [
failure. .
Figure 4.2 (based on the in-vessel hydrogen generation i distributions reported in Volume 2 of NUREG/CR-4551) vas used to '
determine the probability that the amount of hydrogen generated in-vessel vould exceed a level necessary to produce a global concentration of 12%. This probability vas estimated to be 1.38. L Top Event C3 addresses late hydrogen burns. If sprays are in i operation, the only late burns of significance are those !
resulting from sudden releases of hydrogen generated in-vessel-into the containment. These releases vere addressed in Top ,
Event CE. At the time that the HAAP analysis was performed for l BV-2, the MAAp program indicated that for scenarios in which ,
there was uncooled debris in the cavity hydrogen vould recombine in the reactor cavity, or burn as it exited the reactor cavity as a hydrogen-laden jet. In the obsence of containment heat i removal, the deposition of the energy associated with these burns, along with decay heat and noncondensible gases genersted from the decomposition of concrete, containment overpressurization vould eventually occur. . While the timing of--
such failure is certainly influenced by the rate of containment pressure and temperature rise.- there is considerable. uncertainty as to the failure pressure, especially when there are potentially-multiple failure modes, some of which are sensit4ve to temperature. Industry practice is to assume that the-time of -
containmentifailure corresponds to the median failure r assure.
In feet, however, there is1 a finite-probability of-containment '
failure at any pressure which exceeds the test pressure.- llence, there is significant uncertainty in the time of failure .even if the containment loading vas known precisely.
Top Event C4 addresses .slov, long-term overpressurization of-the containment. In the absence of containment heat removal,;the L
E containment atmosphere is likely to continuously pressurize until-t some mode of. containment failure occurs. HAAP analysis performed ,
for BV-2 provides containment' pressure and temperature _historias, p The same considerations regarding .the timing of containmant fai;ure that vere discusr-i above for : Top: Event C3~, apply to Top Event C4 as well.
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t Questie 3. Describe hov phenomenological uncertainties vere accounted for ;
durirg the quantification of Containment Event Trees. '
i Response 3. The phenomena of greatest interest in the Beaver Valley Unit 2 :
(BV2) Backend Analysis were the following:
- Induced failures of the hot leg, surge line, and steam generator tubes
- Containment loads at vessel breach resulting from the effects !
of high pressure melt ejection
- In-vessel hydrogen generation l Induced failures of the Reactor . Coolant System (RCS) boundary occur when RCS pressure is maintained at -high-levels and the-components of interest are heated by natural circulation and fission product deposition to temperatures' at which. their strength is significantly diminished. At these conditions, it is .
possible that fallute of these RCS components could occur before vessel breach occuts, thereby claverting a high pressure-vessel breach sequence to one that vill occur at low pressure.. As noted in Section 4.6.2.1, base .and sensitivity MAAP cases (using Version 14) for a BV2 fast station blackcut event vere performed ;
to provide temperature histeries for the RCS components of interest. For each of these cases a determination ofLthe time to failure for each of the-RCS components of interest, including the reactor vessel, was performed. Except. for the reactor vessel,-
the time to componen' failure was estimated using harson-Hiller .
thermal creep rupture data. Probability distributions were '
generated for each of the component failure times and these distributions were combined using the'STADIC program to determine the following mean ptobabilities- of induced component failure '
after core damage occurs
- Induced SGTR occurs first <10-3
- Hot leg piping or safe end' failure occurs first = 0.9 Vessel melt-through occurs first = 0.1-l The Surry analysis for NUREG-1150 cited mean values of 0.018 for induced SGTRs and '0.72 -for hot leg. failure prior to vessel ,
breach, with the RCS at the system setpoint; pressure'during core:
degradation. Because of concerns regarding the " coarseness" of!
the vay that HAAP models-the.RCS, the NUREG-1150-mean values vere used in. the BV2 -Backend quantification.- A- " double delta" distribution vas .used to represent these mean 1 values ine the generation af uncertainty distributions for the frequencies of- -
major release category groups.
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At the time of the BV2 Backend study, it vr.s judged that the '
uncertainties in the containment loads at vessel breach, due to high pressure melt ejection, vere best represented by the NUREG '
1150 distributions generated for Surry by the expert opinion process. Plant specific HAAP analyses performed for BV2, using recommended best estimate input parameters, indicated pressure rises which vere-less than the median values of the corresponding NUREG-1150 distributions. Hence, because of the similarity-between the Beaver Valley and Surry plants, the NUREG-1150 data.
for Surry was applied to DV2 vith first order scaling of the pressure rise due to differences in containment free volume and -r power level. The containment load distributions were combined ,
with the containment strength distribution in the STADIC program to determine the mean probability of containment failure fot various sets of plant damage state parameters-(e.g., RCS pressure '
prior to_ vessel breach, status of containment sprays, etc.).
Renent experiments (e.g., see Reference 3-1, below) confirm that debris vill be trapped in the lower compartment of the containment minimizing the effects of DCH. Thus, it is believed _
that the approach used in the BV2 Backend Analysis,-to assess'the prob 1bility of containmer.t failure at vessel breach, is somewhat i conJervativo. :
The first uncertainty associated with hydrogen involves the antity of hydrogen a anerated in-vessel. The BV2 analycis_is based on an eggregate of the NUREG-1150 probability distributions ,
for this parameter. HAAP analyses performed for BV2 fast station blackout sequences using the - "no blockage option" corresponds to '
approximately 46% core circaloy. oxidation. As shown in Figt re 4.2-1 of -the IPE submittal, this fraction of core oxidation corresponds to_approximately the 70th_ percentile of the assumed in-vessel hydrogen _ distribution. Hence,_the assumed distribution '
was deemed appropriate for "V2 severe accidents.
Reference 3-1. .
al. " Experiments to Investigate the Effects of Allen, H. D., et Flight Path on Direct Containment lleating (DCH) in the surtsey Test __ Facility," prepared for the~ U. S. Nuclear Regulatory Commission by Sandia- National Laboratories,- NUREG/CR-5728 (SAND 91-1105), October 1991.
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. f Question 4. Section 4.1.4, " Equipment Survivability" (Page 4.1-6) of the IPE :
states that, " survivability of equipment for DV-2-is such that ;
equipment failures under severe accident. conditions vould not create instances of Unusually Poor Containment Performance (UPCP),
given a severe accident." ,
a) State the definition of UPCP, and- discuss the basis for_this ,
definition, b) Vas the conditional and absolute probability of UPCP for internal events only estimated? If so, please provide the- -
estimates.
Response 4. The statem?nt made 'n Section 4.1.4 vas more limited in scope'than its reference in Question 4 vould indicate. The intent of that .
statement (in Section -4.1.4) was tn- confirm that no equipment failures resulting from severe accident conditions- inside ;
containment had been identified that vould, of themselves, " create' ;
instances of unusually poor containment performance". That_being said, i t caa also be stated that the overall containment performance for Beaver Valley Unit 2 is. judged to be adequate.-
The adequacy of the ' Beaver Valley Unit.2 Containment performance is_ based on the observation that the nean conditional probability '
of a large, early containment f ailut e' and containment bypass for' ;
Beaver Valley Unit 2, gietn core damage, is less than 5%. .This compares to about 13% foi Surry, including bypass, E and about; 4% 1 for Zion, including bypassi as- analyzed in NUkEG-1150-final summary report dated November _1990. It'must also-be noted, that-the Surry and Zion 'NUREG-1150.. results: include all early containment failures, large and: -small,- while1 'the Deaver Valley-Unit 2 results are for-large failuresLonly. However, it'must also be noted further that most of- the Surry and Zion early_ failures are large'enough to yield' source terms comparable to VASH-1400 l a 10% iodine- release), while the early, small:
, PVR4- (i.e.,
containment failures _for Beaver Valley _ Unit 2'are dominated byj _
small isolation- failures yielding _ iodine release fractions, an.
order of magnitude lover. On this basis, it is judged that the containment- performance of -Beaver Valley Unit 12, given-core damage, is comparable to that of : Surry. and: Zion as analyzed'in p NUREG-1150'and, therefore,--- that ' Beaver--Valley Unit. 2 doesinot-L exhibit unusually poor containment-performance.
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. t Question 5. a) Provida a concise discussion of _ how the IPE process treated equipment survivability during a severe accident scenario, b) Vas any essential equipment identified which vould fall as a result of severe environmental effcets? Hov is it determined.
which equipment (qualified for Design Basis Accident [DBA) environments), vill be usable and' assumed to operate in severe i accidents? Ilov was credit for such equipment taken-in the PRA7
/
c) Section 4.1.4.1 of the BV-2 IPE (Page 4.1-6) states that the. l containment response reported in Reference 4-7 for the Clon ,
Plant can be taken as representative of thei for BV-2. '
Discuss the applicability of the Zion analysis to BV-2. <
d) Explain how tSe information in Table 4.1-3 was used in the BV- ,
2 IPE process.
Response 5. a) Equipment survivability was treated in the IPS process as a=
means of determining the availability of equipment after core damage has occurred; i.e., beyond- the scope of the Level 1 ;
~
analysis. The Level 1 analyst uses= equipment availability and failure rate data as a means of determining the likelihood of core damage, and this infc:mation is fed into the Level 2 analysis via the' plant damage states. However, equipment'that j vas functioning properly at this endpoint of the Level 1 analysis (or equipment that may hnve been considered.to be restorable in the Level 1 analysis after some time interval) may, in fact, not be available due to conditions created within the' containment (or other relevant. environment)_that. ;
result from the core degradation itself, and about which-the--
Level 1 analyst's information is silent. This is the issue of equipment survivability that is dealt-vith in Section 4.1.4.
The following observations can be made regarding.the treatment of equipment survivability in the Bea"cr Valley Unit 2 IPE:
- The - only _ relevant environment is that within the Containment. Interfacing system LOCAs de not take '
credit for any equipment potentially- affected-by_the l' loss of coolant outside Containment.
l .
' - Any-equipment associated with ECCS is ignored beyond the onset of core damage; .in -the Beaver' Valley Unit 2 IPE all core degradation events are assumed ~to -leadLto -
vessel failure and no -recovery actions are currently. .
- included in. - the analysis. Therefore, questions >
regarding ECCS equipment -survivability beyond the onset of core damage are irrelevant.
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- Thett are no containment fan coolers which operate under ,
accident conditions in the Beaver Valley Unit _2 containment. Containment heat removal is accomplished ;
by Recirculation Spray Coolers. The recirculation f sprays are initiated by a timer once the CIB (containmrut pressure) signal is reached. The CID .
signal first ac'.uates the quench sprays and then the recirculation- sprays. There- is no reliance on ,
containment sump level indication for success of.
containment recirculation sprays.
- Auxiliary Feedvater is not eredited in. any analysis beyond the onset of core damage; therefore, questions !
related to_ Auxiliary Feedvater control survivability ,
beyond the onset of core-damage are irrelevant._ :
Uith-these observations in mind, the remaining parts of Odestion 5 l can be addressed, b) An analysis of equipment survivability for severe accident
- conditions inside Containment was performed for large, dry containment PVRs (as well as for other plant types) by the_ ,
IDCOR Program _and reported in Reference 4-7 of the Beaver Valley Unit 2 IPE. This analysis indicated that for large, I dry containment PVRs, cables inside Containment could be-vulnerable to temperatures associated with the most limiting L (highest atmosphere temperature) severe accidents.= :Therefore, for any-of the accident monitoring and-_ control items mentioned on-Table 4.1-3 that involve cable -runs inside Containment, survivability was not-- determined; i.e., -there vas no prima __
facie evidence that survival was likely, and no detailed- '
analysis was conducted.- _ HovcVer, because of the observations '
listed in Part a) of this question, - the lossoof these items would not adversely affect the conclusions of. the Besver-Valley Unit 2 IPE. l For_ the- remaining three. (3) . parameters _ in- Table 4.1-3.
(containment; pressure, -coatainment-_ area- -radiation,. and ,
containment.qatmosphere hydrogen sampling),-_ failure _ due.to l temperature is- vieved: as unlikely. Of these only. 'the.
containment pressure' indication vas- of concern, based on the findings of-the IDCOR Report. However .:this plant featurei ts l depended upon only during the phase of the accident:vhere_the pressure is within design; and, therefore, its survivability (and that of the other tvo [2] items) is viewed as'likely. ~ +
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I c) A comparison of the limiting containment pressure and ,
temperature transients (for Fast Station Blackout or THBL'), !
Figures 4.2-2 and 4.2-3 of the Beaver Valley Unit 2 IPE and .
Figures. 3-9 through 3-13, and Figures _3 46 and 3-47 of !
l Reference 4.1-3 for the IDCOR analysis of Zion with and
' vithout_a global hydrogen burn, confirms that-the limiting event for the IDCOR equipment survivability for Zion is as ,
severe than that for Beaver Valley Unit 2. In neither case i was a glubal hydrogen burn calculated to occur, but in the ;
Zion analysis, a global burn vas. forced soon after vessel failure which produced a 1160'F _ spike in the containment
' atmosphere temperature. This spike was then included in the equipment survivability assessment.
d) The response to this part of the question can be found in the response to Part b). l T
Ouestion 6. Describe briefly the plant-specific insights obtained.from.the BV-2 back-end analysis, and discuss how the BV-2 back-end insights vere or vill be used to enhance plant safety.
Response 6. As noted in Section 1.7, the DV-2 containment configuration is not conducive to flooding of the reactor cavity, either before or after vessel breach-(except fot vessel injection following. vessel breach).. The OSS provides only limited,flov-to the cavity while' it is operating, and the RSS spray coverage pattern is such that none of its ' flow reaches the cavity. The cavity does not :
communicate with the sump, and it - cannot be' flooded (vithout an external source of vater) due to_ spillover_from_the_ remainder of-the containment. ~The cost-benefit: ' aspectsof design changes to provide water to the cavity vill be examined'during the accident management phase of the'BV-2 IPE.
The BV-2 CDF contains a relatively high: percentage (approximately ,
27%) of SB0 Events. Thus, a relatively Sigh. fraction of vessel breaches occur at RCS pressures,- at ~v hlen the effects of forced- k ejection of debris from_the vessel must be considered. As.noted in-Section.l.5, "for sequences _ involving-Station blackout and no ,
steam generator cooling, current. procedures _(ECA 0.0) preclude RCS- ,
deprcssurization v.a the PORVs, as vould otherwise be directed for otner sequences per FR-Col." As_also noted, consideration vill'be
-given to extending existing. -procedural. provisions for RCS depressurization to co~er. ' Station blackout sequences _vhere appropriate.
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i Core damage scenarios involving SGTRr, and a stuck open secondary side relief valve, have the potential for significant off-site releases. If lov pJessure injection is available, depressurization could exteni the time to core damage, thereby ,
providing a much larger time vindow for recovery actions, and ;
significantly reduce the source term if core damage cannot be prevented. Existing procedures Are being revirved and updated to Perform the more explicitly instruct the operators to depressurization.for sequences in which all high head safety injection is also failed. Procedures and training are also being.
reviewed to ensure that a stuck-open main steam safety / relief valve vould be locally gagged, thereby isolating the faulted-steam.
generator.
Question 7. Discuss the considerations given to in-vessel steam evnic' ion as a contributor to early containment failure probability.
Response 7. In-vessel steam explosions vere addressed in Containment Event ,
Tree Top Event 12 - In-vessel Steam Explosion Fails Containment -
(AP). As noted in Section 4.6.3 of the IPE Summary Report, the failure fractions for this Top Event vere taken directly from Volume 3 of NUREG/CR-4551 (0.008 for lov RCS pressure melts and 0.0008 for high pressure melts). All in-vessel steam explosion caused failures were considered as large, early 'ontainment failures. The contribution of in-vessel steam explosions to the frequency of large, early . containment failures can be determined by examining the importance of CET- split fractions APL and APH in-the split fraction importance table for Release Category Group I (sae Table 7-1 attached to this response). The table is the basis foi Table 4.8-3 of the IPE submittal, which is an abbreviated versb n of the attached table. The sum (0.0465) of the importance of split fractions APL -and APH represent the fractional contribution of in-vessel steam explosions -to the frequency of ,
l Release Category Group I. This represents approximately 5% of-the-Release Category Group I frequency. In terms of abs frequency, in-vessel steam explosions account for 3.7 y-10~plute . peg reactor- year, or- approximately 0.2% of the total CDF of 1.9-x 10 per reactor year.
l.
9 o >= _ _ -- - -
4 .m.-- , ,,+ R ,, , , . _ , . , _ , _ _ . . , , . _ , , . . _ , , __
, TABLE 71. Split Fraction importanco for Larga, Early Containmsnt Failures and GypncOs MODEL Names BV2LYL2
$plit Fraction IFportance f 0f Cr0Up 3 LECFBT
' brted by leportance Group Freq;ency e 8.0195E 06 16:30:10 26 AUO 1992 I
Page 1
. . . . . . $ F None. . . I ppor t anc e. . . . . Ac h i evecent . . R educ t i on. . . Der i vat i ve. . $ f Va l ue. . . . . . . F r equenc y. . . . . .
- 1. SSF 1.0000E+00 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E+00 8.0195E 06
- 2. CP1 1.0000E +00 1.0000E*00 0. 0000+
- 00 0.0000E+00 1.0000E+00 8.0195E*06
- 3. ICF 9.3516E 01 1.0000E +00 0.0000m+00 0.0000E+00 1.0000E+00 7.4V95E 06 l 4 WRF 9.1380E 01 1.0000E + 00 0.0000E+00 0.0000E+00 1.0000E + 00 7.3282E 06 l
- . NMF 9.0300E 01 1.0000E+00 0.0000E*00 0.0000E*00 1.0000E+00 7.?415E 06 l
- 6. IBF 7.8958E-01 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E400 6.3320E 06 1 7 FEF 7.3647E 01 1.0000E +00 0.0000E+00 0.0000E MG 1.0000E+00 5.9061E 06
- 8. CCF 7.2410E 01 1.0000E +00 0.0000E + 00 0.0000E+00 1.0000E + 00 5.8069E 06 j
- 9. HHF 6.7296E 01 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E+00 5.3968E 06
- 10. SP2 5.9243E 01 1.0000E +00 0.0000E+00 0.0000E+00 1.0000E400 4.7510E 06 ,
- 11. Mr3 E.6083E 01 1.0487E +00 4.4027P01 4.8790E 06 9.2000E 01 4.4976E 06 j
- 12. WAF 5.2172E 01 1.0000E+00 0.0000E+00 0.0000E+00 1. 0000E + 00 4.1839E 06 4 1L FAF 4.9575E 01 1.0000E+00 0.0000E*00- 0.0000E+00 1.0000E+00 3.9757E 06 1
- 14. LCF 4.9391E 01 1.0000E + 00 0.0000E+00 0.0000E+ 00 1.L000E+00 3.9509E 06 l
- 15. tAF 4.8792E 01 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E+00 3.9128E 06
- 16. $EF 4.7874E 01 1.0000E*00 0.0000E*00 0.0000E+00 1.0000E+00 3.8392E 06
- 17. W8F 4.768?E 01 1.0000E+00 0.0000E*00 0.0000E*00 1.0000E+00 3.8244E 06
- 18. LHF 4.5477E 01 1.0000E +00 0.0000t+00 0.0000E+00 1.0000E+00 3.6470E 06
- 99. QSF 4.5419E 01 1.0000E+00 0.0000E+00 0.0000E 400 1.0000E
- 00 3.6424E 06
- 20. $MF 4.538?E 01 1.0000E+00 0.0000E+00 0.0000E*00 1.0000E+00 3.6400E 06
- 21. FBF 4.1475E 01 1.0000E+00 0.0C00E*00 0.0000E+00 1.0000E+00 3.3261E 06
- 22. RRF 4.0851E 01 1.0000E*00 0.0000E+00 0.0000E+00 1.0000E+00 3.2760E 06
- 23. EBF 4.0696E 01 1.0000E + 00 0.0000C+00 0.0000E+00 1.0090E+00 3.2636E 06 24 VLF 3,8451E T,1 1,0000E + 00 0.0000E+00 0.0000E+00 1.0000E
- 00 3.0836E-06
- 25. C2a 3.5658E 01 2.545?t+00 6.4342E 01 1.5251E 05 1.8750E 01 2.8596E 06
- 26. L2s 3.5658E 01 1.2488E+00 6.4342E 01 4.8550E 06 5.8900E 01 2.8596E 06
- 27. RCF 3.1569E 01 1. 0000L
- 00 0.0000E+00 0.0000E+00 1.0000E+00 2.5316E 06 28 RDF 3.1298E 01 1.0000E+00 0.0000E + 00 0.0000F+00 1.0000E+00 2.5100E 06
- 29. ME2 2.9400E 01 1.1087E+00 7.0600E 01 3.2298E 06 7.3000E 01 2.3577E J6
- 30. RPR 2.8011E 01 1
- 057E+00 8.3430E-01 4.5822E 06 2.9000E 01 2.2464E 06
- 31. RSF 2.4861E 01 1.000nE+00 0.0000E+00 0.0000E+00 1.0000E+00 1.9937E 06
- 32. 03F 2.2324E 01 1.0000E400 0.0000E+00 0.0000E+00 1.0000E+00 1.7903E 06
- 33. A0F 2.1992E-01 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E+00 1.7f26E-06 34 C2J C 0425E 01 1.8909E*00 7.9575E-01 8.7827E-06 1.8650E 01 1.6380E 06
- 35. L2J 2 ?425E 01 1.1443E+00 7.9575F-01 2.7952E-06 5.8600E 01 1.6380E 06
- 36. SAF 1.9609E 01 1.0000E+00 0.0000E*00 0.000CE+00 1.0000E+00 1.5726E 06
- 37. SBF 1.9520E 01 1.0000E+00 0.0000E*00 0.0000E+00 1.0000E+00 1.5654E 06
- 38. HCF 1.9118E 01 1.0000E+00 0.000CI+00 0.0000E+00 1 ;000E+00 1.5332E 06
- 39. elPF 1.8441E-01 1.0000E+00 0.0000E+00 0.0000E+00 1.On00E 00 1.4789E 06
- 40. AFF 1.7834E 01 1.000 W 00 0.0000E+00 0.0000E
- 00 1.0000E+00 1.4302E O',
- 41. C22 1.7787E 01 2.2227E + 00 8.2213E 01 1.1232E 05 1.2700E-01 1.4264E 06
- 42. L22 1./787E 01 1.1534E+00 8.2213E 01 2.6563E 06 5.3700E 01 1.4264E 06
- 43. IRF 1.7669E 01 1.0000E+00 0.0000i+00 0.0000E*00 1.0000E+00 1.4170E-06
- 44. IWF 1.7597E 01 1.0000E+00 - 0.0000E+00 0.0000E+00 1.0000E+00 1.4112E 06
- 45. CIF 1.7454E 01 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E+00 1.3997E 06
- 46. OFF 1.7436E 01 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E+00 1.3983E-06
! 47. 00F 1.643$E 01 1.0000E+00 0.0000t+00 0.0000E*00 1.0000E+00 1.3180E 06 l 48. RPQ 1.6037E-01 1.6226E400 8.7248E 01 6.0157E 06 1.7000E-01 1.2861E 06
- 49. OSF 1.5418E-01 1.0000E + 00 0.0000E+00 0.0000E+00 1.0000E+00 1.2365E 06
- 50. RPP 1.5195E 01 9.8012LbO1 1.0229E+00 3.42PTE 07 5.3500E 01 1.2185E 06
- 51. W94 1,4736E 01 2.7702E+00 8.5264E 01 1.5378E-05 7.6850E 02 1.1818E 06
- 52. WC2 1.4714E 01 1.4854E+02 8.5206E 01 1.1843E 03 9.9630E 04 1.1800E-06
- 53. RPT 1.4572E 01 1.0000E+00 1.0000'E+00 0.0000E+00 5.0000E 01 1.16S6E 06
- 54. VL1 1.4211E 01 1.3850E+02 8.5823E-01 1.1038E 03 1.0300E 03 1.1396E-06 L 55. OGF 1.3956E 01 1.0000E+00 0.0000E CO 0.0000E+00 1.0000E+00 1.1192E 06 l- 56. OPF 1.3784E-01 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E+00 1.105eE 06
- 57. C12 1.3753E 01 1.0000E+00 0.0000E+00 0.00u0E+00 1.0000E+00 1.1029E-06
- 58. DOF 1.2541E 01 1.0000E+00 0.0000E+00 0.000CE+00 1.0000E+00 1.0057E 06 59, lYF 1.183DE-01 1.000CE +00 0,0000E+00 0.0000E + 00 1.0000E+C0 9.4869E 07
- 60. IBF 1.1820E 01 1.0000E+00 0.0000E+00 0.00u0E+00 1.0000E+00 9,4786E 07 i
~
l
, TABLE 7+1. Split Fraction importanco for Largh Early Containm:nt Failurcs cnd Bypas:cs MODEL kneet BV2LVt2
$plit Frution insertence for Group LEEF6Y Sorted by importancs Group Frequeney e 8.0195E 06 16:30:23 26 AUG 1992 Page 2
. . . . . 5 F h ame . . . I tpor t anc e. . . . . Ach i e vement . . R edac t i on. . . Der i va t i ve. . $ F va l ue. . . . . . . f requenc y. . . . . .
- 61. 9VP I.179BE 01 1.0000E+D0 0.0000E+00 0.0000n+00 1.0000E+00 9.4611E 07
- 62. E2A 1.1613E-01 1.1081E*00 8.8387E 01 1.7979E 06 5.1800E 01 9.3130E 07
- 63. C2n 1.1613E-01 2.7185E+00 8.8307E- 01 1.4712E 05 6.3300E 02 9.3130E 07
- 64. SE4 1.1151E-01 1.0000E + 00 0.0000E*00 0.0000E+00 1.0000E+00 8.9428E 07
- 65. RE5A 1.0747E 01 1.7691E+00 8.9253E 01 7.0299E 06 1.2260E 01 8.6186E 07
- 66. Br2 9.4989E 02 6.7834E+00 9.0501E 01 4.7142E 05 1.6159E 02 7.6176E 07
- 67. BPS 9.4989E-02 1.5412E*00 9.0501E 01 5.1022E 06 1.4030E 01 7.6176E 07
- 68. Ls2 9.3145E 02 1.0000E+00 0.0000E 00 0.0000t+00 1.0000E+00 7.4697E 07
- 69. IAF 8.2166E 02 1.0000E+00 0.0000E400 0.0000E+00 1.0000E+00 6.5893E 07
- 70. MFF 8.1625E+02 1.0000t+00 0.0000E+00 0.0000E*00 1.0000E+00 6.5459E 07
- 71. L53 7.8341E 02 9.3155E 01 1.0684E+00 1.0978E 06 5.0000E 01 6.2826E 07
- 72. RP5 7.7523E 02 9.224BE 01 0.0000E+00 0.0000E+00 1.0000E+00 6.2169E 07
- 73. Act 6.5877E 02 6.9952E+01 9.3412E 01 5.534BE 04 9.5450E 04 5.2830E-07
- 74. CS4 5.9120E 02 9.5375E 01 1.0059E+00 4.1824E 07 1.1391E-01 4.7411E 07
- 75. ISS 5.0757E 02 3.6731E+00 9.5100E 01 2.iB30E 05 1.8000E 02 4.0705E-07
- 76. TB3 4.8377E 02 2.5341E+00 9.5294E 01 1.2680E 05 2.9760E 02 3.8796E 07
- 77. AF4 4.5004E 02 1.6538E*00 9.6562E 01 5.5169E 06 4.9679E 02 3.6091E 07
- 78. A02 4.4506E 02 1.3666E+00 9.5549E 01 3.2971E 06 1.0825E 01 3.5692E 07 79 RPK 4.4122E 02 9.55D8E 01 0.0000E+00 0.0000E+00 1.0000E+00 3.5384E 07
- 80. FB7 4.3226E 02 8.1779E+00 9.5677E 01 5.7910t 05 5.9860E 03 3.4665t 07
- 81. BY2 4.2890E 02 1.0000E+00 0.0000E *00 0.0000E+00 1.0000E+00 3.4396E 07
- 82. LB2 4.2890E 02 1.0000E*00 0.0000E+00 0.0000E+00 1.0000E+00 3.4396E 07
- 83. APL 4.1645E 02 6.1590E
- 00 9.5839E 01 4.17061 05 8.0000E 03 3.3397E-07 84 CDF 4.1086E-02 1.0000E*00 0.0000E+00 0.0000E+00 1.0000E+00 3.2948E 07
- 85. EB7 4.0771E 02 2.8477E*00 9.5923E-01 1.5144E 05 2.1590E 02 3.2696E-07
- 86. RE1 4 0293E 02 8.951?E+00 9.5971E 01 6.4088E 05 5.0440E 03 3.2313E 07 87 PRF 4.0204E 02 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E+00 3.2241E 07
- 88. PR9 3.9416E 02 1.3703E+00 9.6923E 01 3.2164E-06 7.6710E 02 3.1610E 07
- 89. BP7 3.4879E 02 4.1041E+01 9.6512E 01 3.2139E 04 8.7033E 04 2.7971E 07
- 90. Rt? 3.3518E 02 1.2430E*00 9.6648E 01 2.2178E-06 1.2120E 01 2.6880E 07
- 91. OP1 3.0516E 02 3,6002E+02 9.6948E 01 2.8794E 03 8.4990E 05 2.4472E 07
- 92. PL1 2.9792E 02 1.0152E*00 9.7021E 01 3.6089E 07 6.6200E 01 2.3891E 07
- 93. R11 2.9502E 02 3.0227E+02 9.7035E 01 2.4163E 03 9.7730E 05 2.3659E-07
- 94. HH1 2.6935E 02 4.6758E+01 9.7310E 01 3.6717E 04 5.8751E 04 2.1600E
- 07
- 95. RPS 2.4401E 02 1.0000F+00 1.0000E+00 0.0000E+00 5.00000 01 1.9568E 07
- 96. BVt 2.3826E-02 4.5071E + 02 9.7617E 01 S.6066E 03 5.2979E 05 1.9107E 07 97 C07 2.2985E 02 9.7579E 01 1.0019E+00 2.0933E*07 7.2680E 02 1.B433E 07
- 98. DAF 2.2490E 02 1.0000E+00 0.0000E+Dr 0.0000E+00 1.0000E+00 1.8036E 07
- 99. RPW 2.2153E 02 9.7785E-01 0.0000E+0* 0.0000E+00 1.0000E+00 1.7766E 07 100. COB 2.1797E 02 9.9737t 01 1.000$E*00 2.4796E 08 1.4950E 01 1.7480E 07 101. CSF 2.0659E 02 1.0000E+00 0.0000E
- 00 0.0000E+00 1.0000E+00 1.6567f 07 102. RE3 1.9570E 02 1.2210E+00 9.8043E 01 1.9290E 06 8.1360E 02 1.5694E 07 103. BX1 1.9164E 02 0.0000E+00 9.8084E-01 0.0000E*00 5.6635E 06 1.5368E b7 104. br3 1.9164E -02 4.210BE+00 9.8084E 01 2.5903E-05 5.9330E 03 1.5368E-97 105. IPS 1.8508E 02 9.2967E 01 1.1808E+00 -2.0142E 06 7.2000E 01 1.4842E 07 106, BP4 1.T778E 02 2.1523E*01 9.8222E 01 1.6472E 04 8.6550t 04 1.4257E 07 107. OS1 1.5939E 02 2.5027t + 00 9.8408E 01 1.2179E 05 1.0480E 02 1.4782E 07 IDS. 011 1.34B6E 02 1.0412E+ 01 9.8769E 01 7.5579E 05 1.3060E-03 1.0815E 07 109. SA1 1.3425E+02 2.1994 E + 00 9.9081E-01 9.6920E-06 7.6010E-03 1.0766E 07 110. BP6 1.3314E-02 1.1206E*00 9.8669E 01 1.0742E 06 9.9390E 02 1.0677E
- 07 111. BK1 1.2398E 02 9.87 73E + 01 1.0012E+00 1.0821E*07 9.0492E 02 9.9427E 03 112. RE7 1.22B8E-02 1.5069E+00 9.8771E 01 4.1633E 04 2.3670E 02 9.8546E 08 113. C06 1.1994E 02 1.4725E+00 9.9057E 01 3.8649E 06 1.9560E 02 9.6182E-08 114. HUF 1.0674E 02 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E+00 8.5600E-08 115. EW6 1.0569E 0; 1.0031E+00 9.8943E 01 1.0999E-07 7.7060E 01 8.4758E 08 116 EC2 1.0569E 02 1.1208E+00 9.8943E 01 1.0534E 06 8.04585 02 8.4758E 08 117 FC2 1.0569E 02 4.0834E+DO 9.8943E 01 2.4812E 05 4160E 03 8.4758E-08 118 FB6 1.0569f -02 1.0955E+00 9.8943E 01 8.5056E 07 4.9650E 02 8.4758E 08 119. EB4 1.0563E 02 1.2650E+DO 9,8944E 01 2.2101E-06 3.8330E-02 8.4712E-08 120 VL2 1.0143E-02 1 6501E+00 9.9124E-01 5.2838E 06 1.3300E 02 8.1339E 08
, TABLE 71. Split Fraction importanco for L.argo, Early Containm:nt Failures and Bypass:s MODE; hames BV2LVL2
$plit Fraction inportance f or Group EECF8Y Sorted by Ipportance Gro@ Frecuency = 8.0195E 06 16:30:36 26 AUG 1992 I Page 3
...... SF Name... l epor t anc e. . . . . Ac h i evement . . R edse t t on. . . D er i va t i ve. . SF Value. . . . . . . F requency. . . . . .
121. aff 9.762BE 03 1.0000E+0F 0.0000E + 00 0.0000E+00 1.0000r+00 7.8292E 08 122. 0 11 9.7213E 03 8.6148t+00 9.9028E 01 6.1145E 05 1.2750E 03 7.795SE 08 123. VL3 9.5258E 03 1.6959E+00 9.9083E 01 5.6543E 06 1.3000E 02 7.6392E 08 124. DA1 8.6959E 03 3.3314E400 9.9130E 01 1.876TE 05 3.7160E 03 6.9737E 08 125. R11 8.3581E 03 1.3815E+00 9.9164E 01 3.1263E 06 2.1440E 02 6.7027E 08 126. NC1 8.3197E 03 1.508Bf*01 9.9168E 01 1.1305E 04 5.8990E 04 6.6720E 08 127. $B2 8.2634t 03 1.3408E*00 9.9183E 01 2.7987E 06 2.3400E 02 6.6268E 08 128. FBB 8.1626E 03 1.2029E+00 9.9184E 01 1.6923E 06 3.8680E 02 6.5460E 08
- 29 EC1 8.1405E 03 9.2960E + 00 9.9186E 01 6.6595E 05 9.8029E 04 6.5282E 08 130. FB8 7.9913E 03 1.0728E+00 9.9201E 01 6.4766E 07 9.8950E 02 6.4086E 08 131. AF6 7.9353E 03 4.1938E+01 9.9208E 01 3.2837E -04 1.9343E 04 6.3637E 08 132. FA2 7.8870E 03 1.2222E+00 9.9211E 01 1.5448E 06 3.4285E 02 6.3249E 08 133. 184 7.8225E 03 1.1796E +00 9.9378E 01 1.4899E 06 3.3470E 02 6.2732E 08 134 EA2 7.6983E 03 1.006JE+00 9.9230E 01 5.9123E 07 1.0442E 01 6.1736E 08 135. FAI 7.5622E 03 6.8149E+00 9.9244E 01 4.6693E 05 1.2968E 03 6.0644E 08 136. Bv4 7.4571E*03 5.6300E+01 9.9254E 01 4.4354E 04 1.3483E 04 5.9802E 08 137. EA1 7.3946E 03 1.2818E*00 9.9261E 01 2.3158 06 2.5574E 02 5,9301E 08 138. RW1 5.7550E 03 0.0000E+00 9.9424E 01 0.0000E+00 4.7860E 05 4.6152E 08 139 EEF 4.8911E+03 1.0080E+00 9.9511E 01 1.0322E 07 3.8000E 01 3.9224E 08 140. LEF 4.8911E 03 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E+00 3.9224E 08 141. HEA 4.8911E 03 1.0002E+00 9.9562E 01 3.6413E 08 9.6500E 01 3.9274E 08 142. APH 4.8695E 03 6.2271E+00 9.9581E 01 4.1952E 05 8.0000E 04 3.9051E 08 143. RC1 4.8665E 03 1.1921E+00 9.9513E 01 1.5794E 06 2.4709E 02 3.902TE 08 144 OR1 4.3360E 03 1.3040E + 01 9.9566E
- 01 9.6590E-05 3.6000E 04 3.4772E 08 145. PR7 4.1252E 03 1.0633E+00 9.9671E 01 5.3iO4E 07 4.9460E 02 3.3082E 08 146. F-R V 3.7277E 03 9.9652E 01 1.0015E+00 3.9817E 08 2.9890E 01 2.9894E 08
- 47. 003 3.4900E-03 7.0 798E + 00 9.9651E 01 4.8785E 05 5.7370E-04 2.7988E 08 149. OP3 3.4519E 03 7.0760E+00 9.9655E 01 4 6754E 05 5.6780E 2.7682E 08 149. REA 3.2324i 03 1.0205E+00 9.9677E 01 1.9060E*07 1.3600E 01 2.5922E 08 150. AF3 3.0641E 03 1.0417E+00 9.9774E 01 3.5263E 07 5.1502E 02 2.4572E-08 151. 056 2.9872E 03 3.8572E+00 9.9714E 01 2.2936E 05 1.0000E 03 2.3956E 08 152. RD1 2.6511E+03 1.1 D44E +00 9.9735E 01 8.5819E 07 2.4774E 02 2.1261E 08 153. HH2 2.5193E 03 3.9055E+00 9.9749E 01 2.3321E 05 8.6265E-04 2.02C4E 08 154. Bv1 2.4611E 03 0.0000E+00 9.9754E 01 -0.000C.+00 1.7241E 07 1.9737E 08 155. 581 2.3399E -03 8.2941E-01 1.0012E+00 1,3780E 06 7,2320E 03 1.8764E 08 156. BPA 2.2671E 03 1.4333E+01 9.9773E 01 1.0694E 04 1.7000E 04 1.8181E 08 157. D02 2.2559E 03 5.6244E+00 9.9774E 01 3.7103E 05 4.8760E 04 1.8092E 08 158. Bv5 2.2309E-03 1.0711E+00 9.9777E 01 5.8789E 07 3.0431E-02 1.7890E 08 159. SA2 2.1838E*03 1.0089E+00 9.9990E 01 7.1919E 08 1.1470E 02 1.7513E 08 160. HC3 2.0731E-03 1.1521E+00 9.9798E 01 1.2356E 06 1.3090E 02 1.6641E 08 161. R12 1.9922E 03 1.0060E+0G 0.0000E+00 0.0000E+00 1.0000E+00 1.5977E 08 162. HH4 1.7802E 03 4.0789E+00 9.9822E 01 ' 4705E 05 5.7782E D4
, 1.4276E 08 163. OR3 1.7493E 03 1.1500E+00 9.9825E 01 1.t167E 06 1.1530E 02 1.402SE 08 1 64 CCB 1.7461E-03 1.2704E+00 ? 9836E-01 2.1814E 06 6.0449E 03 1.4003E 08 165 OP2 1.7182E 03 4.3786E+00 9.h'28E 01 2.7108E 05 5.0830E 04 1.3779E 08 1 66. 0A2 1.6386E 03 1.0478E*00 9.9816E 01 3.9616E 07 3.3170E 02 1.3141E 08 167. 081 1.6266E-03 1.3269E+00 9.9B,TE 01 2.6343E 06 4.94BOE 03 1.3045E 08 168. HC2 1.5954C 03 1.1167E+00 9.984iE 01 9.4825E 07 1.3230E-02 1.2794E-08 169, 191 1.5892E 03 2.8063E+01 9.984;E 01 2.1704E*04 5.8720E 05 1.2745E 08 170. RE9 1.5818E 03 1.1363E+00 9.984f E 01 1.1059E 06 1.1470E 02 1.2685E 08 171. IR1 1.4877E 03 2.5 757E + 01 9.98?iE 01 1.9855E 04 6.0090E-05 1.1931E 08 172. R02 1.4829E-03 1.0618E+00 c'52E 01 5.0755E 07 2.3430E 02 1.1892E 08 1 73. SB6 1.4722E 03 1,0155E+00 .v859E 01 1.3566E 07 8.3210E-02 1.1806E 08 174. Vlf 1.3668E-03 1.0000E*00 0.0000E+00 0.0000E+00 1.0000E+00 1.0961E 08
( 175. SE 2 1.3510E 03 1.2535E+00 .9865E 01 2.0434E 06 5.3020E 03 1.0834E 08 l 176. rR1 1.3394t 03 3.6514E+00 Y.9867E - 01 2.1274E-05 5.0210E 04 1.0741E 08 177. BP8 1.3006E 03 1.0111E+00 9.9870E 01 9.9536E 08 16479E 01 1.0430E 08 178. OS2 1.2926E 03 1.0723E+00 9.9873E 01 5.9005E 07 1.7220E 02 1.0366E-08 179. IR2 1.1804E-03 4.4 797E + 00 9.9882E-01 2.7915E 05 3.3910E 04 9.4660E 09 180 OS4 1.1696E-03 1.0000E+00 0.0000E+00- 0.0000E+00 1.0000E+00 9.3798E- 09 l
l l
14 -
l i
- ~ - , - ,,- . . , . .
TABLE 71. SpHt Fraction importanca for Largh Early Containm2nt Failurcs and Bypass:s MDOEL kster BV2LYL2 Split f raction leportance f or Group LECfBf Sorted by importance Group f requency = 8.0195E 06 i 16:31:06 26 AUG 1992 Pope 4
. . . . . . $F hue. . . Imor t ance. . . . . Achi evenent . . R edac t i on. . . Der i vat i ve. . $F Value. . . . . . . F regamcy. . . . . .
181. RE4 1.1429E 03 1.0071E+00 9.98S(1-01 6.6322E 08 1.3820E*01 9.i656E 09 182. SB3 1.122BE 03 1.0B63E*00 9.9934E 01 6.9744E 07 7.6060E 03 9.0046E 09 183. AF2 9.7147E 04 2.9590E+00 9.9905E 01 1.5718E 05 4.8585E 04 7.7907E 09 1B4 RE6A 9.6348E 04 1.0049E +00 9.9904E 01 4.6714E 08 1.6540E 01 7.7266E 09 185. Rf2 9.1242E 04 1.5134E+00 9.9909E 01 4.1242E 06 1.7740E 03 7.3171E 09 186. lW2 9.0358E 04 3. 6404t
- 00 9.9910E 01 2.1182E 05 1.4210E 04 7.2462L D9 187 RWF 9.025BE 04 1.0000E*00 0.0000E+00 0.0000E+00 1.0000E+00 7.2382E 09 1B8. MF1 8.6533E 04 1.2825E+00 9.9913t 01 2.2726E 06 3.0536E 03 6.9395E 09 189 FR6 8.4700E D4 9.6934E 01 1.0016E+00 2.5906E407 5.0960E 02 6.7925E 09 190. Ri3 8.3793E 04 0.0000E+00 9.9916E 01 0.0000E+00 3.5780E 06 6.7197E 09 191. RIF 8.3793E 04 1.0000E+00 0.0000E*00 0.0000E+00 1.0000E*00 6.7197E 09 i 192. $ES 8.28492 04 1.1605E*00 9.9917i 01 1.2941E 06 5.1340E 03 6.6440E + 09 193. DB2 8.2332E 04 1.1386E+00 9.9923E 01 1.1180E 06 5.5180E 03 6.6026E 09 194 PR8 7.7929E 04 1.0182E+00 9.9953E 01 1.4972E 07 2.5070E-02 6,2495E 09 195. Af1 7.5750E 04 0.0000F400 9.9925t 01 0.0000E+00 1.0720E 05 6.0748E 09 196. SB4 7.4990E + 04 8.875sE 01 1.0013E+00 9.1202E 07 1.1210E 02 6.013BE 09 197. 181 7.4687E 04 1.4517E *00 9.9935E 01 3.6274E 06 1.4460E 03 5.9895E 09 198. PA1 7.3445E 04 1.0448E*00 9.9927E 01 3.6515E 07 1.6130E 02 5.B899E 09 199. tR4 6.6156E 04 9.5462E 01 1.0004t+00 3.6729E 07 9.1130E 03 5.3053E 09 200. t16 6.5774t 04 1.0547E
- 00 9.9934E-01 4.4400E+07 1.1880E 02 5.2747E 09 201. HR2 6.3433E 04 1.095 7E +00 9.9937E 01 7.7286E 07 6.5820E 03 5.0870E 09 202. D01 6.098EE 04 8.3007E+00 9.9939E-01 5.8553E 05 8.3530E-05 4.8909E 09 7"3. '19 5.4355E 04 9.9738E 01 1. 0000E
- 00 2.1247E 08 1.1810E 02 4.3589E 09 404. SA4 5.2846E 04 9.9615E 01 1.0000E+00 3.1223E 08 1.1660E 02 4.2180E 09 205. LH2 4.9623E 04 1.0422E+00 9.9950E 01 3.4241E 07 1.1622E 02 3.9795E 09 206. PA2 4.B825E-04 1.0005E+00 9.9951E 01 8.0302E 09 4.8760E 01 3.9155E 09 207. CS2 4.8130E 04 8.9224E 01 1.0004 E + 00 8.6748E 07 3.8348E 03 3.8598E*09 208. HMF 4.1034E 04 1.0000E+00 0.0000E + 00 0.0000E+00 1.0000E*00 3.2907E 09 209 PRJ 3.6480E 04 9.?931E 01 1.0003E+00 7.9407E 09 3.0340E 01 2.9255E 09 210. AF5 3.5564E 04 . 5125E+00 9.9965E 01 4.1131E 06 6.8178E 04 2.8520E 09 211. RDA 3.4825E 04 1.0041E+00 9.9965E 01 3.5837E 05 7.7930E-02 2.7928E 09 212. Rx1 3.4825E 04 1.1805E+00 9.9965E 01 i.4503E 06 1.925 7F
- 03 2.7928E 09 213. FC1 3.4687E 04 0.0000E+00 9.9963E-01 0.0000E+00 2.4771E 05 2.7817E 09 214. FB4 3. 4 687F- N, 1.0178E+00 2.9965E 01- 1.4557E 07 1.9070E 02 2.7817E 09 -
215. S93 3.4665E 04 1.5740E+00 9.9966E 01 4.6056E 06 5.8975E 04 2.7800E 09 216. LH1 3.3198E 04 1.4810E+00 9.9967E 01 3.8604E 06 6.8966E*04 2.6623E*09 217. HMS 3,0278E 04 1.4054E+00 9.9970E 01 3.2536E 06 7.3718E 04 2.4281E 09 218. MU2 3.024BE 04 1.0148E+00 9.9970E-01 1.2129E 07 2.0000E-02 2.4257E 09 219. BK2 2.6812E-04 7.2190E 01 1.0001E+00 2.2313E 06 5.0316E 04 2.3106E-09 220. FB3 2.4799E 04 1.1901E+00 9.9975E 01 1.5263E 06 1.3030E 03 1.9887E 09 221. CS3 2.4156E-04 8.9129E 01 1.0009E+00 8.7930E 07 8.5727E 03 1.9372E 09 222. WA2 2.3946E 04 1.0182E+b0 9.9976E 01 1.4806E 07 1.2970E
- 02 1.9203E 09 223. DB3 2.3171E 04 1.0144E+00 9.9978E 01 1.1692E 07 1.5200E 02 1.8582E 09 224 Bv3 2.3102E 04 0.0000E*00 9.9977E 01 0.0000E+00 1.2627E 06 1.8527E 09 225. OR2 2.2900E 04 1.1276t+00 9,7977E-01 1.0248E 06 1.7920E 03 1.8365E 09 226. HM1 2.2B48E 04 1.3977E + 00 9.9977E 01 3.1916E 06 5.7410E 04 1.8323E 09 227. ASF 2.2653E 04 1.0000E+00 0.0000E+00 0.0000E+00 1,0000E400 1.8166E 09 228. EM2 2.2525E 04 1.0040E+00 9.9977E 01 3.3764E 08 5.3500E-02 1.8064E 09 229 052 2.2525E 04 1.0363E+00 9.9977E 01 2.9322E 07 6.1605E 03 1.0064E- 09 230. WB3 2.0532E 04 1.0166E+00 9.9979E 41 1.3474E 07 1.2220E 02 1.6466E 09 231. DD6 2.cs87E 04 8.1042E 01 1.0003E+00- +1.5224E 06 1.3560E-03 1.610?E 09 232. HLF 1.7963E 04 1.0000E+00 0.0000E + 00 0.0000E+00 1.0000E+00 1.4405E 09 233. PRA 1.7682E-04 9.0528E 01 1.0002E+00 -7.6117E 07 2.0010E 03 1.4180E 09 234 MSF 1.7061E-04 1.0000F^C0 0.0000E+00 0.0000E+00 1.0000E+00 1.3682E-09 235. RES 1.6108E 04 1.007Ei+00 9.9954E 01 6.379?E-08 2.0250E 02 1.2918E 07 236. ORS 1.5710E 04 1.4181E*00 9.9984E 01 L '542E 06 3. 7560E-04 1.2598E 09 237 AFA 1.5682E 94 1.0369E+00 9.99B4E 0 2.9694E 07 4.1872E 03 1.2576E 09 238. Cl2 1.4977E 'v4 1.0087E +00 9.9985E 01 7.0816E-08 1.6960E 02 1.2010E 09 239. DA3 1.4745E 04 1.0345E+00 9.9985E 01 2.7751E 07 4.2610E-03 1,1825E 09 240. PA3 1.4410E+04 1.0002E+00 9.9986E-01 2.8137E 09 4.0780E 01 - 1.1556E 09 1
- ' - - + - + - g
TABLE 71. Split Fraction importanca for Larg3, Early Centainm:nt Failurcs and Bypasses ,
, MODEL hame 8V2tVL2 '
Split fraction legertence for Group LEEFBf Sorted tiy Inportance Grc.up Fr equerry e 8.0195E 06 16:31128 26 AUG 1992 g Fege 5 i
. . . . . . if h ane. . . I nser t enc e. . . . . Ac h ievenent . . R edac t i on. . . Der i va t i ve. 5 F Value. . . . . . . F r equenc y. . . . . . l 241. C11 1.4177E 04 1. 0?T3E +00 9.9966E 01 2.w04E 07 5.1670E 05 1.1369E 09 242. AW1 1.3909E 04 1.2989E+00 9.V986E 01 2.3981E 06 4.6512E 04 1.1154E 09 243. l'R H 1.3640t 04 1.0004E + 00 9.9989E 01 4.0466E 09 2.1240E 01 1.0918E 09 244. 5A7 1./096s 04 1.00?7E+00 9.9991E 01 6.2205E 08 1 1740E 02 9.7019E 10 245. $8J 1.209BD 04 1.0033E+00 9.9?B8E 01 2.7791E 08 3.4910D 02 9.7019E 10 246. C2T 9.5981E 05 1.0959E*00 9.9990E 01 7.6972E 07 1.0000E 03 7.6972E 10 247. L2T 9.5981E 05 1.0000E+00 0.0000E+00 0.0000E+00 1.0000E+00 7.6972E 10 248. Ali 7.9922E 05 1.0079E + 00 9.999?E 01 6.4093E 08 1.0000E 02 6.4093E+10 249. Of1 6.9267E 05 9.9520E 01 1.0000E+00 3.8578E 08 9.2100E 03 5.5549E 10 250, SA5 6.3032E 05 1.0008E+00 9.9999E 01 6 7637E 09 1.3910E 02 5.0548E 10 251. IA1 $.1323E 03 3.6870E 01 1.0002E+00 5.0645E 06 3.4241E S4 4.1158E 10 252. SBE 4.9633E 05 1.0006E*00 9.9975E 01 5.2345E 09 7.4390D02 3.9803E 10 253. FB5 3.8598D 05 1.0012t+00 9.9296E 01 9.6309E 09 3.2140E 02 3.0954D10 254 EB3 3.5644E 05 1.0014E*00 9.9??6E 01 1.1285E 08 2.5330E 02 2.6sB4E 10 255, 182 2.5281E 05 1.04 78E* 00 9.9997E -01 3.834BE 07 5.287t'E 04 2.0?74E 10 256. T12 2.1019E 05 1.0000D 00 0.0000E+00 0.0000E+00 1.0000000 1.6856D 10 257. L3C 1.8478E 05 1.0000E+00 0.0000E+00 0.0000E + 00 1.0000D 00 1.4818E 10 .
258. C3C 1.8478E 05 1.0000D00 9.99980 01 3.8996E 10 3.8000E 01 1.4818E 10 259. H3C 1.8470E 05 1.0000E+00 0.0000E 00 0.0000E+00 1.0000E + 00 1.4818E-10 260. MH6 1.8215E 05 0.0000E*UD 9.9900E 01 0.0000E+00 4.7642E 07 1.4607E 10 261. 007 1.7420E 05 9.5973E 01 1.0001E+00 3.2:40E 07 1.6470E 03 1.3970E 10 2v2. $BA 1.7203E 05 9.9995E 01 1.0000[400 4.0083E 10 3.5820E 02 1.5796E 10 263. $BC 1.5854E+05 9.9753D 01 1.0000D 00 2.0003E 08 1.3550E 02 1.2714E 10 <
264. ECG 1.5537D 05 1.03716 00 9.9999E 01 2.9778E 07 2.8030E404 1.2460E 10 +
265. Rf4 1.5175E+05 0.0000E
- 00 9,9999E 01 0.0000E*00 4.3000E 06 1.2170E 10 266. 6Li 0.0000E + 00 9.9372E 01 1,0002E+00 5.1975E 06 3.0970E*02 0.0000E + 00 267 CC1 0.0000E+00 0.0000E+00 1.0000E+00 0.0000E+00 2.8563E-05 0.0000E+00 268. AF7 0.0000E + 00 9.9963E 01 1.0000E+00 2.9978E 09 2.2947D 04 0.0000E+00 269 M51 0.0000E+00 9.9381E 01 1.0000D 00 4.9671E 08 7.1010D 04 0.0000E+00 270. AFC 0.0000D 00 9.4972D 01 1.0000E+00 2 kT49E 09 4.8675E 04 0.0000E+00 271 PRI 0.0000E+00 9.9987E 01 1.00n0E+00 1.1944E 09 1.020^O 01 0.0000E
- 00 272. CC7 0.0000E+00 9.9994E-11 1.0000E*00 4.9264E 10 2.5823D 04 0.0000E+00 273. CC4 0.0000E
- 00 0.0000E+00 1.0000E+00 0.0000E+00 4.0554E 05 0.00COE+00 2 74 . CC2 0.0000E+00 0.0000E
- 00 1.0000f*00 0.0000E +00 - 3.122BE 05 0.0000E + 00 275, Of2 0.0000E+00 9.9983E 01 1.0000tt00 1.39h?E 09 3.313C? 04 0.0000E+00 276, rRK 0.0000E+00 9.9979E 01 1.0000E+00 -1.7016bL 2.0570E 03 0.0000E+00 277. LS1 0.0000E+00 0.0000E+00 -1.0000E+00 0.0000E*00 0.0000E400 0.0000E*00 270. TT3 0.0000E*00 9.9996E 01 1.0000E+00 2.9063E 10 1.6640E 02 0.0000E*00 279 111 0.0000E+00 9.8300E 02 1.0000E+00 7 *.3150 06 5.0560D 05 0.0000E+00 280. 871 0.0000E+00 0.0000E+00 1.0000E + 00 0.5000L+00 0.0000E+00 0.0000E + 00 281. 050 0.0000E*00 0.0000E-00 1.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 282. 115 0.0000E+00 0.0000E+00 1.0000E+00 0.0000E+00 0.0000E400 0.0000E+00 283. ME1 0.0000E+00 0.0000E+00 1.0000D00 - 0.0000E+00 0.0000E+00 0.0000E+0a 284, $P1 0.0000E+00 0.0000E+00 1.0000E+00 0.0000E+00 0.0000E+00 0.000nE+00 205. AFD 0.b000900 0.0000E
- 00 1.0000E+00 1.0000E+00 1.24B?E 05 0.0000E+00 286. MSO 0.0000D00 0.0000E+00 1.0000E+00 0-0000E+00 0.0000D 00 0.0000E+00 287, C11 0.0000E+00 0.0000D00 1,0000E+00 0000E+00 0.0000D00 LOOOOD00 288. 000 0.0000E+00 9.76A7E 01 1.000'D 00 1.8755E 07 2.2960E 03 0.0030E*00 289. C21 0.0000D00 9.9509E 01 1.000,7+ 00 3.9380E 08 2.0000E 04 0.0000D 00 290 IA2 0.0000E*00 8.633eD 01 1.00o1;+00 +1.0964E 06 5.8650D 04 b.0000E+0f 291 001 0.0000D 00 9.9960E 01 1.0000E+00 3.2138E 09 1.1950E 03 0.0000E+00
?92. Plt 0.00009 00 0.0000E*00 1.0000E*00 0.0000E+00 0.0000D 00 0.00000 00 293. PRO 0.0000EiD0 0.0000E+00 1.0000F+00 0.0000E*00 (. 0000E+00 C,0000E+00 294. PR2 0.0000E+00 9.9298E 01 1.0000D 00 5.6298E 08 5.2h0E 04 0.0000E+00 295. Cs6 0,0000E + 00 9.9987E 01 1.GJ00D 00 1.0699E 09 3.3740E-02 0.0000E+00 l
296. P12 0.0000E+00 8.4886E 01 1.0040E+00 -1.2438E 06 2.5(70E+0? 0.0000E+00 l 297 003 0.0000E+00 9.9998F 01 1.0000E+00 1.559BE 10 5.0000t 02 0 0000000 1 292. CSS 0.0000E+00 9.9994E 01 1.0000E*00 4.SS21E 10 4.7728t 04 0.0000E+00 299 FR5 0.0000E400 9.9905E 01 1.0000D 00 -7.7924E Of 2.59308 02 0.0000E+00 300. IC2 0.00009 00 9.9490E 01 1.00000 00 4.0889D 08 3.2777E-04 0.0000E+00 l
- %~
~ - , _ , - - . _ - -- - - _ . -
I
, TABLE 7 'i. Uplit Fraction importanco for Large, Early Containmsnt Failurcs and Bypasns
, Motett tiene: av?tvt2 to'It f rettion Inportance for Group LfCFBY 5etted oy inportance Grcup Frequeney . 8.01951<06 16:31:49 26 AUG 1992 Page 6 6
. . . . . . $ f W ame. . . letor t enc e. . . . . Ac h ie vNent . . Rutac t i on. . . Der i va t i ve. . $ F Unlue. . . . . . . F r equenc y. . . . . .
301. 101 0.0000t + 00 9.402tt 01 1.0000t+00 4.7919f 07 1.8347t+.4 0.0000t+00 307. PR3 0.0000t+00 9.9321t 01 1.0000t+00 5.4$0DE 05 5.1040t 04 0.0000t+00 .
303. HH7 0.0000t+00 9.79/9E 01 1.0000t+00 1.6228t 07 1.3939t 03 0.0000t*00 3 04. IP1 0.0000E+00 0.0000t*00 1.0000t+00 0.0000t+00 0.0000t + 00 0.0000C+ 00 305. $81 0.0000f*00 9.974Bt 01 1.0000t+00 2.0446t-Ob 1.*250E 02 0.0i'^nt+00 306. C03 0.0000 *00 9.9857t 01 1.0000t+00 1.1576t 08 6.4980t 03 0.0000E+00 307 00" 0.0000f400 9.95??t 01 1.0000t+00 3.8356E 00 9.1230t 04 0.0000t+00 308. 00 5 0.0000t + 00 0.0000f*00 1.0000t+00 0.0000t+00 0.0000t+00 0.0000t*00 309. i12 0.0000t+00 0.0000t+00 1.0000t+00 0.00008' + 00 0.0000t+00 0.0000t+00 310. 000 0.0000E*00 9.9994t 01 1.0000t+00 4.9267E 10 P.7463t 04 0.0000t+v0 311. CCJ 0.0000t+00 9.9455E 01 1.0000E+00 4.3740E 08 3.4410E 04 0.0000t+00 312. 009 0.0000t+00 9.9944f-01 1.0000L*00 -(.5324E 09 1.6030c 03 0.0000t+00 313. Pl1 C.0000t*LO 7.4574t 01 1.0000t 00 2.0394t 06 1.8120E 04 0.0000t+00 314 I$1 0.0000t+00 0.0000E+00 1.0000t+00 0.0000t400 0.0000t+00 0.0000t + 00 315. Ris 0.0000t+00 0.0000t+00 1.000vt+00 0.0000t+00 0.0000t+00 0.0000E+00 316. . *15 0.0000t+00 8.81901 01 1.0001t+00 9.47561 07 5.1$10E 04 0.0000t+00 317. OD3 0.0000t+00 9.9786t 01 1.0000E*00 1.7213E 08 1.2900E 03 0.0000t+00 318. 015 0.0000t+00 0.0000E00 1.0000t+00 0.0000t+00 0.0000t+00 0.0000E+00 319. E51 0.0000t+00 9.56??t 01 1.0000t+00 3.5115t+07 $.4721E+05 0.0000t+00 l
l
, 1 t
i
-i i
Question 8. a) Provide a discussion Of the ignition sources and limits used j in the hydrogen combustion analyses. Vere sensitivity sit' dies !
performed to evaluate the impact on the IPF results, due to the uncertainties of the ignition limits used? ,
b) Provide the information requested in NUREG-1335 (Section 2.2.2.1), i.e., accurate but- simple representations of the ;
containment choving the lustrument tunnel, reactor cavity !
compartment, loop compartment (s), annular compartment (s) and I uppet compartment, with specific identification of potential i reactor release points and vent paths indicated. Estimates of j compartment free volumes and vent path flow areas should also 6 be provided. Please address specifically how this information is used in the assessment of hydrogen pocketing and detonation, t
c) Discuss the plant-specific effects on containment integrity and equipment survivability due to local detonations. The discuss'on should cover likelihoods of local detonation and-potentials for missile generation as a result of local detonations.
d) In Page 4.6-17 on Top Event 20 - _ Late Burn of Combustibic Gases, the IPE states that, "I f . the containment is not lnerted..., hydrogen burns are assumed to be assured in this time period; however, -these burns are not expected to challenge the containment." Please discuss briefly the ;
reasons for not expecting the hydrogen burns to challenge the containment.
Response 8. a) No sensitivity studies . relative .to ignition limits vere performed for the BV2 Backend Analysis. Because Beaver _ Valley and Surry plants are essentially " sister" plants, the BV2 Backend Analysis relied heavily on the insights obtained from -
the analyses performed _for Surry for NUREG-1150. ~In addition, plant-specific MAAP analyses performed for BV2 indicated that-burns vould either be precluded by 7 team inertion or vould-occur when the _ hydrogen concentrations achieved -global ;'
flammability levels, as determined by the MAAP algorithm for most servore accident scenarios. Thus, either burns'did not
- occur or occurred at ;reIatively lov hydrogen concentration.
An exception to this observation is when hydrogen is suddenly released into a non-inerted containment. This exception was discussed in our response to RAI Item 2, and-vill be discussed-
- again later-In'this response. The burn pressures-calculated by MAAP vere significantly less than those which would cause a significant probability of' containment failure. Hand
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B calculations for adiabatic burns -Mailagrations), up te the limits of detonation limits, indier,ted that the pressure rises associated with these deflagrstions vete not likely to result in containment fallute. Although the loads associated with detonations (transitions from deflagra' ions) might be sufficient to cause containment failure, these loads are difficult to calculate and containment strength criteria for these types of loads were not available, As noted in 0;r response to RAI Item 2, the BV2 1pE adopted a conservative treatment for detonations (,esulting from transition.s from-deflagrations) and consequent containment failure. It shou 4d be noted that f4UREG/CR-4551 did not address detonations for Surry.
As noted on Page 4.2-3 of the IPE submittal, in Reference B '.,
the analysis of hydrogen combustion for the surry plant for
!JUREG-1150 assumed that if electrical power vere available during the-period of hydrogen generation, "the sprays vill keep the steam concentration lov, and sparks from electrical equipment vill cause ignition near the lovet deflagrable limit", preventing significant concentrations of hydrogen.
Based on the extent of m3xing promoted by spray c,peration and the relatively lov ignition energy levels required -for ignition, this argument was assumed to be valid-for BV2 as well, except- for the sudden release of hydrogen into the containment (e.g., vessel blowdovn at high pressure after i severe core degradation). ]
The initiation of combustion requires that the temperature of the reactant gases be raised above a " threshold _ temperature",
thereby initiating reaction - (see Reference 8-2). This' J temperatureincreasecanbecausedbya-f}ame, spark,-are,-ho*,
gas, hot particle (such as core debris) , compression, shock waves, adiabatic heating,. and the addition _of pyrophoric or hypergolic materials. The energy required to initiate combustion decreases as the hydrogen gas temperature increases, until the self-ignition temperature is reached.
According to Reference 8-3, stated flammability lit L ts -have little meaning at mixture- temperatures. in -excess of approxim*tely 1200"F. At thess temperatures, the mixture becomes near-hypergolic (i.e., spontaneously reacting)..
essentially independent of steam concentrations.
1 It should be noted that the loads associated vith high pressure-melt ejection _ include the contribution-of hydrogen combustion.
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i Minimum energies required to ignite various hydrogen-air sixtures are shown in Figure BA-1 (taken from Reference 8-4).
As noted. in Reference 8-6, near-stoichiometric mixtures (approximately 30% hydrogen) can be ignited vith spark energy levels as lov as 0.02 milli-joules. llovever the ignition energies required to ignito hydrogen mixtures increase substantially as the mixture concentrations approach the flammability limits. As noted in Reference 8-6, the er.ergy required for ignition increases rapidly as the1 : initial pressure is lovered. Conversely, the minimum spark energy decreases with increasing initial mixture tenperatures.
Reference 8-6 also notes that there is a degree of uncertainty associated with the magnitude of. the available capacitive discharge electrical energy, which is actually expended in heating the gas near the electrodes. Measured ignition energy' requirements for mixtures near the flammability limits appear to be very sensitive to the size of the vessel used in the experiment. One reported series of experiments indicated that the required spark energy to ignite a 4.5% hydrogen mixture wasintherangeof10to100jou}es,dependingonwhgtherthe-tests vere conducted in a 240 ft. vessel or a 10 ft.- vessel. .
The peak precsures associated with detonations are vell above .
the quasi-static pressures associated with deflagrations. J llovever, the energies required for detonation are many orders of magnitude above those required for deflagration. As noted' in Reference 8-7, -detonation initiation- vithin a range of hydrogen concentration from 18 to 59- volume percent (the approximate range of hydrogen detonsbility) requires an energetic ignition source, severe confinement, and/or a sufficiently large volume of gas mixture. Referpnce 8-7 concluded that, "the energy levels required to directly initiate detonation are orders of magnitude greater than those necessary to initiate burning at the same hydrogenL concentration", and that "a de facto transition to detonation-is highly unlikely in reactor containment bu.ildings ,
particularly when there are high steam concentrations or hydrogen concentrations belov -about 10 volume percent".
Minimum ignition energies of- 4100 joules have been reported (Reference 8-7) for hydrogen-air mixtures. According to-Reference 8-7, this energy level is several orders of magnitude higher than -vould be- produced from an electrical spark caused by contact arcing or by electrostatic discharge and approximately eight orders of magnitude higher than the-minimum ignition energy required to initiate deflagration.
i l
j 1
Figure 8A-2 (taken from 9eference 8-8) identifies the energy l levels of various potential ignition sources. Based on the !
ignition energy requirements shovn in Figute 8A-2. a match l' burning for only a fraction of a millisecond vould generate sufficient energy to initiate a deflagration in a hydrogen-air mixture near the lover limit of flammability. (
If electrical povet is not available, the containment sprays
- vill not operate, and the containment is likely to be inerted ;
by high concentra',lons of steam. Vhen steam inertion prevents combustion, the recovery of electrical power and containment sprays becomes a concern since operation of the sprays vill condense the steam and drive the gas mixture towards the '
flammability range. The Surry analyses performed for NUREG-1150 as;sumed that hydrogen vould be ignited and burned as soon l as the gas mixture entered the . flammability range, !
guaranteeing that the burn vould occur at lov hydrogen concentration. Operation of the centaintnent spray vould guarantee substantial mixing.
i The recovery of At power during or after coro degradation was '
not addressed in the IPE submittal. Because of potential- :
deleterious effects (such as containment deinerting), the :
strategy for recovery of mitigating systems such as containment sprays, must be carefully examined end fully ,
evaluated in the context'of an accident management-program.
As noted earlier, for scenarios in which the containment ,
sprays are operating, it is likely that. hydrogen burns vill l occur at lor concentrations when hydrogen is "slovly" released I into the containment.- Only when the hydrogen is suddenly ,
released into the containment (e.g., due to an induced failure :
of the hot. leg. or at vessel breach), will- the hydrogen ,+
enneentrations achieve significant. values. When vessel breach is accompanied ' !!PHE, the containment loads discussed for Top Event C2 y. lude the contribution of hydrogenLburns.
However, for " pour" type vessel- breaches at high pressure, ,
there could be a sudden release of-hydrogen into the reactor- '
cavity and then into the containment. -For those scenarios.in which there vas a sudden release of hydrogen into a non-steam ;
inerted containment atmosphere,- i _t vas assumed that:1_f the _
- global concentration' exceeded 12%, a burn vould occur which would, in turn, fall the containment. The logic implicit in- +
this assumption is as follovs: ,
- 1. Containment- fallure, due--to a deflagration- at a 12%
to fall the BV-2:
hydrogen concentration,-is not- likely containment (based . on peak containment pressures: ,
i
- determined using the adiabatic burn. assumption). ,
L t.
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- 2. Although HAAP simulations shoved that the containment was vell mixed vhen sprays were in operation, it vas assumed that local concentrations could be -20% higher than the '
global concentration, t
- 3. Although the BV-2 containment configuration is not- }
necessarily e menable to_ a Deflagration to Detonation ;
Transition (DDT), it was assumed that a DDT would occur if_ _;
local concentrations exceeded- a value of 15% (minimum ,
value reported in Reference 8-9). j
- 4. It was assumed that DDT vould result in a large-containment failuie. ;
Figure 4.2-1 of the BV2 IPE submittal (based on the in-vessel !
hydrogen generation distributions reported in Volume 2:of <
I NUREG/CR-4551) was used to _ determine the probabili ty ' that- the-amount of hydrogen generated in-vessel vould exceed a level ,
necessary to produce a global concentration of 12%. This ;
probability was estimated to be 0.38, and was used as the !
split fraction value for . Top Events _ C2 and CE vhen vesse) blovdown occurred at high pressure in the absence of HPHE.
b) Attached Figures 8-1 through 8-7 provide dimensions of the Beaver Valley Unit 2 containment building._ A simplified schematic of the BV2 containment. including compartment locations, compartment junctions, and potentiel break locations, is given in Figure 8-8. As shown in Table 4.1-l'of the IPE submittal . the 6 total free volume of the. BV2 containment is 1.72 x 10 eubic feet. HAAP divides the-containment free volume into four compartments. The currant HAAP Parameter File for BV2 cv... ins the-following volumes for these compartments:
6 Upper compartment - 1.02 x 10 cubic feet 5 I Lover (loops) compartment- - 4.30_x 10 cubic feet 5
Annular /dcad end compartment - 2.55 x 10 cubic feet j Reactor cavity / instrument tunnel - 7892_ cubic feet During the development of the _HAAP parameter--file, it vas noted that a lot of junction areas- existed between .
compartments relative to the -potential hydrogen' release points. This was confirmed _=during the containment _valkdown.
It was also noted in the HAAP analyses that-the hydrogen was cell mixed, especially when the containment' sprays _ vere fune;1oning. Therefore, no significant h/drogen pocketing.is expected for Beaver Valley.-
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1, c) As noted in the discussion of Top Event 18 in Section 4.6, detonations vere assumed to fail the containment, leading to a large fission product release directly to the environment. No credit vas taken for mitigating equipment following a-detonation.
d) Top Event 20 addresses containment failure due to late :
hydrogen burns. As noted earlier, -i' the containment is not l
. inerted, the only burns of significance are those resulting from sudden teleases of hydrogen generated into the containment; hovever, no- such releases are expected in this &
time frame. Sudden releases at vessel breach vere addressed in Top Events C2 and CE. Although MAAP analysis indicated '
that for scenarios in which there- vas uncooled debris in the '
cavity, hydrogen vould recombine in the reactor cavity or burn as it exited the teactor cavity as a hydrogen-laden jet. .Ir the absence of containment heat removal the deposition of tLe energy associated with these burns, along with decay heat,-and '!
noncondensible gases generated from the decomposition of f concrete, containment overpressurization vould eventually .
occur (split fraction C3A).
If the containment is inerted,_ operation of containment sprays in th time period could drive the containment atmosphere to flammable mixtures. However, as noted earlier, recovery of ,
sprays after severe core damage vas not addressed in this
- i phase of the study, l
h h
- 23'-
5
+
, . - .,-.n,.-,m ~m- y _ . . . _ , . . - , ...,,_.4g , ~.,,.-,%, ..,,,-e,. ..,,,,,-_,e-m,-,.,
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i f
1 REFF.RF.NCES 8-1 Breeding, R. J., et al, " Evaluation of Severe Accident Riskst i Surry Unit 1", NUREG/CR-4551 (SAND 89-1309), Volume 3. Revision 1,
'rarts 1 and 2, October 1990. 3 8-2 Stull, D. R., " Fundamentals of Fire and Explosion", AIChE Monograph Series, No. 10, Vol. 73, 1977. ;
8-3 Villiams, D. C., et al, " Containment Loads Due te_ Direct Containment Heating and Associated Hydrogen Behavior Analysis and Calculationt Vith the CONTAIN Code", NUREG/CR-4896, May 1987.
8-4 Camp, A., et al, " Light Vater Reactor Hydrogen Manual", NUREG/CR-2726 8-5 Hertzberg, Martin, " Flammability Limits and Pressure Development ;
in lig -Air Hixtures", PRC Report No. 4305, l'resen t ed at the Vorkshop on the Impact of Hydrogen on Vater Reactor Safety. ;
Albuquerque, New Mexico, January 25-28, 1981.
8-6 Liu, D., et al, "Some Results of VNRE Experiments in flydrogen Combustion", NUREG/CR-2017, September: 1981.
t 8-7 IDCOR, " Hydrogen Combustion in Reactor Containment Buildings".
Technical Report 12.3, September 1983.
8-8 Fauske & Associates, " Technical Support for the Hydrogen Control Requirement for the EPRI Advanced Light Vater Requirements Document", DOE /ID-10290, U.S. Department of Energy, t 8-9 -Sherman, H. P., et al, " FLAME Facility...The Effect of Obstacles and Transverse Venting on Flame- Acceleration and Transition to Detonation for Hydrogen-Air 'Hixtures at Large Scale", NUREG/CR-5275, April 1989.
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n 6 Ques tion- 9. NUREG-1335 recognizes the importance of'considering uncertainties in' the accident -progression- and'=CET. quantification.- EPRI recommends that sensitivity . studies be performed by HAAP users, which could provide qualitative insight -into - understanding uncertainties. Please specify what ' specific revision (s) of-the-MAAP-3.0B Code vere used for the BV-2 PRA. -Address the Gabor Kenton & Associates -report prepared for EPRI (" Recommended-Sensitivity Analyses for an Individutl Plant Examination Using MAAP-3.0B"). In particular. with respect to Appendix A of-the report, indicate for each of the 78 indicated-parameters: a) If the. recommended value(s) were 'ised, b) If value(s) other than the recommended value(s) were used,:and' the basis for the choice; or c) If the sensitivity study indicated was not performed,' provide the reasons for omitting the recommended analyses.
-Response
- 9. Calculations were performed vith MAAP-3.0B, -Revisions 14 and 16.
The analysis for Surry described in- NUREG-1150 vas used as the-basis for the FRA quantification and no sensitivity studies were performed by DLC. Question 10. Discuss briefly the 'quantification results for each containment isolation failure mode-(including common-mode failure). . Response 10. The containment isolation' failure modes tvhich vere considered at-
-Beaver Valley Unit 2 consist of the following:
- Small containment bypass; i.e., an SGTR
- Large containment bypass; i.e., large' interfacing systems'LOCA (V w' Sequence)-
- Containment not. isolated-or failed prior to' core damage;J1eak area less than the equivalentoof 3 inches in diameter.
- Containment not isolated or failed prior to core damage; leak area greater :than the equivalent of 3 inenes in diameter.
Due to the subatmospheric design of the Beaver ValleyJunit 2-- containment building, the -last failure . mode described above was not included in the PRA models, since preexisting' failures'of this. size vould be obvious to the operator inasmuch as he-vould be unable to meintain subatmospheric pressure. Small containment-
. - . . - ... . ~ . . _. . - . . . . _ - . . - ~ .
4.- bypasses are due to SGTR Initiating Events in the Level 1 Event _= -. Trees.- These-Initiating Events- account .for.17.0% of the Release Category Group II frequency. 'Large containment bypassesm are:due ' to interfacing system LOCAs'in the Level 1 EventLTrees, or induced SGTRs in the Level 2 Event' Trees. These two types'of large containment bypasses : account for 9.1%_ of the1 Release Category Group I frequency. As discussed above, any preexisting containmentLisolation failures vere considered to be small in nature and therefore were binned directly into Release Category: Group- II. -Containment: isolation failures that vere explicitly modeled are:
- Major containment vents and drains; e.g., nump pump discharge .
' liner !
- RCS connections; e.g., RCP seal vater return line
- Connections to containment atmosphere; e.g.,- containment.
vacuum line The following attached tables are reports generated-from the PRA-containment isolation model, a brief description for~each of these-reports is discussed in the following pages. Figure 10-1.shovs-the fault tree that was used to quantify the containment isolation' model. 3 4 m k n r L ,
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+
Table 10-1. This table lists .the; containment isolation-split-fractions that were u;ed in the-PRA. Included in this= table are split f t actf orf
~ descriptions, point estimate _( PC) and. Monte _ Carlo (MC/LH) mean~
split-fraction values,-and the basic event success / failure states for house events in the : fault tree (Figure 10-1). .It should:-be .;
-noted that split. fraction CI7 only consists- of a single basic --
event (operator-action ZHECI3) and hence was not quantified using the fault trees. w g,E-. - ~ h-1 1
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TABLE 101. CONTAINMENT ISOLATION SPLIT FRACTION REPORT. Page 1 of 2 MODEL Name BV2 Spli! F *r'8an Renart inr Ton Event Cl 15:31:14 22 SEP 1992 Page 1 Split Frection Cl1 - CONTAINMENT ISCLAliON - ALL SUPPORT PE Mean = 4,9710E-03 Date : 09 AUG 1991 14:16 HC/LH Pean = 5.16'0E-03 Date : 28 AUG 1991 00:40 Sasic Event Inpacts for Split Fraction : C11 Basic Esent State Description XXACPU S LOSS OF EMERGENCY AC PURPLE XXACOR S LOSS OF EMERGEPCY AC ORANGE XXSAFF S SSPS TRAIN A UNAVAILABLE XXNOSB F k0 LOSS OF ALL AC POWER XXSBFF S SSPS TRAIN B UNAVAILABLE Split Fraction Cl2 CONTAINMENF ISOLATION - LOSS OF AC PURPLE PE Mean = 1.6790E-02 Date : D9 AUG 1991 14:16 NC/LH Mean a 1.6960E 02 Date : 28 AUG 1991 00:40 Basic Event Impacts for split Fraction : C12 Basic Event State Description XXACOR S LOSS OF EMERCENCY AC ORANGE XXACPU F LOSS OF EMERGENCY AC PURPLE XXSAFF S SSPS TRA;W A UNAVAILABLE XXSBFF S SSPS TRAIN B UNAVAILABLF XXNOSB F NO LOSS OF ALL AC POWER Split Fraction Cl3 CONTAINMENT ISOLAi!ON - LOSS OF AC ORANGE PE Mean = 1.1050E-02 Date : 09 AUG 1991 14:1o QC/LH Mean = 1.1?40E-02 Date : 28 AUG 1991 00:40 Basic Event in1 pacts for Split Fraction : Cl3 Basic Event State Description XXACOR F LOSS UF EMERGENCY AC ORANGE XXACPU S LOSS OF EMERGENCY AC PURPLE XX5AFF S SSPS TRAIN A UNAVAILABLE XXSBFF $ $$PS TRAIN 8 UNAVAILABLE XXUOSO F NO LOSS OF ALL AC POWER Split Fraction Ul4 CONFAINMENT ISOLATION LOSS OF SSPS TRAIN A PE Mean = 5.1090E-02 Cate 4 09 AUG 1991 14:16 ' MC/LH Mean = 5.1390E 02 Date : 23 AUG 1991 00:40 BeJ c Event impsets fur Split Fraction : Cl4 Basic Event State Description XXACOR S LOSS OF EMERGENCY AC ORANGE XXACPU S LOSS OF EMERGENCY AC PURPLE XXSAFF F SSPS TRAIN A UNAVAILARLE XXS8FF S SSPS TRAIN B UNAVAILABLE XXNOSB F NO LOSS OF ALL AC POWER
TABLE 101. CONTAINMENT ISOL ATION SPLIT FRACTION REPORT P g) 2 of 2 ' , MCOEL Name: BV2 Sollt Fraction Report for Top Event Cl 15:31:23 22 SEP 1992 Page 2 Split Fraction CI5 - CONTAINMENT ISOLATION LOSS OF S$PS TRAIN B l PE Mean = 6.2710E-02. Date : 09 AUG 1991 14:16 ' MC/LH Mean a 6.2720E-02 Date ? 28 AUG 1991 00:40 Basle Event Ippacts for Split Fraction : CIS Basic Eve *t State rlption ) XXACOR 3 . + EMERGENCY LC ORANGE XXACPU S -- if EMERGENCY AC PURPLE I XXSAFF S SSPS ' n;:: - 'NAVAILABLE XXSBFF F $$PS TRAIN B UNAVAILARLE XXNOSB F NO LOSS OF ALL AC POWFR l j Split Fraction C16 - CONTAINMENT ISOLATION - LOSS OF AC ORANGE & PURPLE PE Mean a 1.1830E.02 Date : 09 AUG 1991 14:16 MC/LH Mean = 1.1880E 02 Date : 28 AUG 1991 00:!.0 Basic Event Impacts for Split Fraction : C16 Easic Event State Description
................ ..... ................................ j XXACOR F LOSS OF EM2RCENCY AC ORANGE I XXACPU F LOSS OF EMERGENCY AC PURPLE l XXSAFF S S$PS TRAIN A UNAVAILABLE l XXSBFF S S$PS TRAIN B UNAVAILABLE XXNOSB S NO LOSS OF ALL AC POWER Split Fraction C17 . CONTAINMENT ISOLAll0N - LOSS OF $$PS 1%ns ! s'=F PE Mean = 1.0000E-01 Date 4 09 AUG 1991 14:16 MC/LH Mean = 1.0270E-01 Date : 28 Aub 199". 00:40 Equation: ZHECl3 Split Traction CIF - GUARANTEED FAILURE PE Mean = 1.0000E+00 Date 09 AUG 1991 14:16 MC/LH Mean = 1.0000E+00 Date : 28 AUG 199? 00:40 l 1
( Constant Value: 1.0 l l
, . . . . , . - . . , - . . . . -, . ~_ . - _ . - . _
4 r '% Table'10-2. This report consists of- the' containment -isolation common cause
; failure modes, which vere' developed by using-the Multiple Greek-Letter (HGL) methodology.-. Incorporated -into- this table are the common cause group identifiers,.the basic events that are affected >
in_the gr ap, the order of the common cause_ failure mode.modeled, the failure mode, and then database- variables that were used to
- quantify the MGL equations.
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7
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.TASLE 10 2. CONTAINMENT ISOLATION COMMON CAUSE REPORT P:g)1 cf 3 '
MODEL Name: BV2 CCF Model Report for Top Event c t 15:30:07 22 SEP 1992 Pag? 1 i -Group 10 : MVC Basic Events Description MVFC2CMSMOV381 2CHS*MOV381 FAILS TO CLOSE
'MVF C2CHSMOV378 2ChS*MOV378 FA!LS TO CLOSE Algebraic Method: M0L Order = 1 out of 2 Failure Mode 10 : CLOSE Total Failure Rate a 2TVMOD deta a 2BVM00 Group ID : AVC Baalc Events Description AVFC2CHSLCV460A 2CHS*LCV460A FAILS TO CLOSE i AVFC2CHSA0V200A 2CHS*A0V200A FAILS To CLOSE Av? !CHSA0V200C 2CHS*A0V200C FAILS TO CLOSE AVFL.: HSA0V2008 2CHS*A0V2005 FAILS TO Cl,0SE AVFC2CHSLCV4600 2(.HS*LCV4600 FAILS TO CLOSE AVFC2CHSA0V204 2CHS*ADV204 Falls TO CLOSE
< Algebraic Method: MGL Order = 3 out of 6
' Falture Mode ID CLOSE Total failure Wate = ZTVA00 Beta = 2SVA00 Gamma = ZGVAOD Delta = ZDVA00 Group 10 : VSC Basic Events Description VSFC2CVS$0V1518 2CVS*SOV1518 FAILS TO CLOSE VSFC2CVS$0V152B 2CVS*SOV152B FAILS TO CLOSE Algebraic Method. MGL 06 der = 1 out of 2 Falture Mode 10 CLOSE Total Falture Rate a ZTVt00 l
Beta a 2BVSOD i r i o " l }
TABLE 10 3. CONTAINMENT ISOLATION COMMON CAUSE REPORT Page a of 3 MODEL Name: BV2 CCF Model Report f or top Event Cl 15:30:10 22 SEP 1992 Page 2 ) Group ID : AV1 Basic Events Description AvFC2DASA0V1008 20AS*A0V1003 FAILS TO CLOSE AVFC2DASA0V100A 2DAS*A0V100A FAILS TO CLOSE Algebraic Method: MGL Order = 1 out of 2 Failure Mode 10 : CLOSE Total Failtre Rate a Z1VA00 Bete a 7BVt00 Group it. AV2 Basic Events Description AVFC20GSA0V10SA 2DGS*A0V108A FAILC TO Cl0SE AVFC2DGSADV108B 2DGS*A0V1088 FAILS TO CLOSE Algebraic Method: MGL
=
Order = 1 out of 2 Falture Mode 10
- CLOSE Total failure Rate = ZTVA00 Otta = 'BVA00 Group 10 ' AV3 Basic Events Description AVFC2VRSA0v109A1 2VRS*A0v109A1 FAILS TO CLOSE L
+
AVFC2VRSA0v10SA2 2VRS*AOV109A2 FAILS TO CLOSE Algebreic Meth>d: MOL Order = 1 out of 2 8alture Mode ID CLOSE Tot 6l Failure Rate = ZTVACD Beta = 2BVA00 Group 10 VS1 Basic Events Description VSFC2CVSSOV153A 2CVS*SOV153A FAILS TO C15+E VSFC2C\bSOV1538 2Cvs*SOV1536 FAILS TO CLOSE l2-i
TABLE 10 2. CONTAINMENT ISOLATION COMMON CAUSE REPORT Page 3 cf 3 MODEL Nafnes BV2 CCF Model Report for Top Event Cl 15:30:30 22 SEP 1592 Page 3 Algebreic Methodt MGL Order = 1 out of 2 Fellure Mode a0 t CLOSE Total f ailure Rate
- 2TVSOD Beta = 2BVS00 Group ID : VS2 Basic Events Descripti m VSFC2CVSSOV151A 2CVS*Sov151A FAILS TO CLOSE VSFC2CVS$0V152A 2CVs* N V152A FAILS TO CLOSE AlgdrerMethod: MGL Order = 1 out of 2 Fafture Mode ID : U.0SE Total Failure Rete = 2TVSCC Beta = 2BVS00
, . . . - . -. - - . . , - . . , . . - , . . - . - - ~ . . . - - - _ . - . - -. . .
, i =
T 4 b f table 10A3. This table provides'the cause table- for each of-the: containment isolation split, fractions that vere quantified-by using:the fault = tree on Figure 10-1 The cause tables consist of the-quantified ~ j minimal cutsets forJea:hfparticular: split: fraction. .These cutrets are ranked in descendini _ order according1to - their quantified-values. Additionally, this table -shows -the % Importance,- orlthe ~ percentage that each cutset +ontributes .to the: Monte' Carlo mean- ; split fraction- alue, and the %. Cumulative, which is .the l cumulative summation of the % Importance.-.The cause table reports vere generated by using-a 99.9% _ cumulative cutoff for.cach.of the , up.;' 'ractions. The alignment of the system.when--the cutset was-quantified is also provided. These are all shown as-heing normal,. , since no maintenance or tests are performed on an unisolated component during plant operation. It _should be noted that singleton cutsets, whose basic- event identifiers' are separated- by-a comma and enclosed in brackets [ ],-are-common cause'failuresiof components. Independent failures- of common i.3use components are shown as a sing 1c basic event enclosed in brackets. r m Y e
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TAILE 10 3. CONTAINMENT ISOLATION CAUSE TABLENORT ' Page 1 cf 7 MODEL Names BV2 ~ Cause Table for Top Event C1 and $plit Fraction C11 PE value of C11 = 4.9710E-03 Date : 09 AUG 1991 14:16 MC/LH Vetue of C11 a 5.1670E+03 Date : 28 AUG 199100:40 15:23:55 22 SEP 1992 Page 1 No. . . Cut set s . . . . . . . . . . . Val ue . . . . . % loportance X Cunut at ive Al i grwent . . . 1 [VSFC2CVS50V151A,V 7.851E 04 15.1945 15.1945 NORMAL SFC2CVSSovl52A) 2 (VSFC2CVSSOV151B,V 7.851E 04 15.1945 30.3500 NORMAL SFC2CVSSOV1528] 3 (VSFC2CVSSOV153A,v 7.851E-04 15.1945 45.5335 NORMAL
$8C2CVS$0V1538) 4 OPRCl2 4.764E 04 9.2201 54.8036 NORMAL 5 (AVFC2PRSA0V109A1, 4.063E 04 7.8634 62.6669 NORMAL AVFC2VRSA0V109A21 6 [AVFC20ASA0V100B,A 4.063E-04 7.8634 70.5303 NORMAL VFC20ASA0V100A1 7 (AVFC2DGSA0V108A,A 4.063E 04 7.3634 78.3937 NORMAL VFC2DGSA0V1088) 8 [MV8C2CHSMOV331,MV 3.957E-04 7.6582 86.0519 NORMAL (C2CHSMOV378) 9 [VSFC2CVSSOV151A) 1.37M-04 2.6631 88.7149 NORMAL e
(VSFC?CVSSOV152A1 10 [VSFC2CVS$0V153A) 1.376E-04 2.6631 e 91.3780 NORMAL (VSFC2CVSSJV15381 11 (VSFC2CVSSOV1518] 1.376E-04 2.6631 94.0410 NORMAL FVSFC'CVSSOV1528) t 12 CVFP2CVS93
- 6.308E-05 1.2208 95.2619 NORMAL
-VSFC2CVSSOV102 13 (MVFC2CHE"0V381)
- 4.540E 05 .8787 96,1405 NORMAL
[MVFC2CHSMOV378) 14 (AVFC2DMA0V1008] 4.049E-05 .7836 96.9241 NORMAL (AVFC2DASA0V100A] 15 (AVFC2Vk:A0V109A1] 4.049E 05 .7836 97.7078- NORMAL [AVFC2VRSA0V109A21 16 (AVFC20GSA0V108A) 4.049E-05 .7836
=
98.4914 NORMAL [AVFC2DGSA0V1088) 17 IMVFC2CHSMOV381)
- 3.497E 05 .6768 99.1682 NORMAL CVFR2CHS173 18 AVFC2RCSA0V101
- 3.256E-05 .6302 - 99.7983 NORMAL CVFR2RCS6d TABLE 10 3. CONTAINMENT ISOLATION CAUSE TABLE REPORY Page 2 cf 7
, MODil Nanes 8V2 Cause Table for Top Event Cl ard split Fraction Cl2 PE Value of Cl2 = 1.6790E-02 bate r 09 Auc 1991 14t16 Mts t N Value of C12 e 1.6960E 02 Date 28 AUG 1991 00:40 L 15:24:47 22 $[P 1992 Page 1 N o. . . Cut t e t e . . . . . . . . . . . Va l ue . . . . . % Inportence % Coma st i ve Al l grver.t . . .
1 _(MVFC2CntMOV378) 6.091E 03 35.9139 35.9139 NORMAL 2 CVfR2CHS173 5.7',7E 03 34.003', 69.9175 NORMAL 3 (VSFC2CVS50V151A,V 7.665E 04 4.5195 74.4369 NORMAL
$FC2Cvf50V152A1 4- (V5f C2CV$sovl51B,V 7.665E 04 4.5195 78.9564 NORMAL l SFC2Cvss0V152B1 5 (vsFC2CVS50V153A,V 7.665E '4 4.5195 83.4758 NORMAL EFC2CVSSOVl538) t 6 OPRCl2 5.765E 04 3.3992 86.8750 NORMAL 7 (AVFC2VR1'A0V109A1, 4.011E-04 2.3650 89.2I.00 NORMAL l
AVfC2VRSA0V109A2] 8 IAVFC2005A?v108A,A 4.011E 04 2.3650 91.6050 NORMAL VFC2DGSA0V10881 9 (AVFC20ASA0V1000,A 4.011E 04- 2.3650 93.9699 NORMAL VFC20A5A0V100A1 f 10 (MVfC2CH5Mov381,MV 3.936t 04 2.3208 96.2907 NDkMAL FC2CHSM0V378) 11 (VSFC2CVS50V151B] 1.3190-04 .7777 97.0o84 NORMAL f (VSFC2CVSSOV1528] 12 (VSfC2CVSs0V151A1 1.319E 04 .7777 97.8461 NORMAL e (VSfC2CVSSOV152A] l 13 (VSFC2CVS$0V153A) 1.319E 04 7777 96.6238 NORMAL [V$fC2CVS$0V1538) 14 CVFP2CVs93
- 6.407E 05 .3778 99.0016 NORMAL VSFC2CVSSOV102 i 15 (AVFC2VRSA0V109A1) 4.036E n5 .2380 99.2396 NORMAL (AVFC2VRSA')V109A2] -
I 16 .2380 99.4775 NORMAL
.IAVFC2DA$A0V1008] 4.036E 05 l-l IAVfC2DASA0v100A) l I 17 (A .2380 99.7155 N0aMAL
. vfC20CSA0V'98A) 4.036E-05
[AVFC2DGSA0V10881
- lj(; -
4 TABLE 10-3. CONTAINMENT ISOLATION CAUSE TA E REPORT Page 3 cf 7 8;0 DEL Names BV2 Cauce table for Top Event C1 and Split Fraction Cl3 PE value of CI3 = 1.1050t 02 Date : 09 AUG 1991 14:16 MC/LH Valw of Cl3 = 1.1240E-02 Date : 28 AUG 1991 00:40 15:25:39 22 SEP 1992 Page 1
$.o... Cutsets........... Value..... % Inportav e % Cumulative Alignment...
1 [MVFC2CHSMov381) 6.303E 03 56.0765 56.0765 NORMAL 2 (VSFC2CVSSOV151 A,V 7.577E-04 6.7411 62.8176 NORMAL SFC2CVSSOV152A] 3 [VSFC2CVSSOV1518,V 7.577E-04 6.7411 69.5587 NORMAL SFC2CVSSOV152B] 4 (V5FC2CVS$0v153A,V 7.577E 04 6.7411 76.2998 NORMAL SFC2CVSSOV15381 5 OPRCl2 4.605E 04 4.0970 80.3968 NORMAL 6 IMvFCP';HSMOV381,MY 4.068E 04 3,6192 84.0160 NORMAL FC2CH?MOV378) 7 [AVFC2VRSA0v109A1, 3.967E 04 3.5294 87.5454 NORMAL AVFC2VRSA0V109A2] 8 (AVFC2DASA0V1008, A 3.967E 04 3.5294 91.0747 NORMAL VFC2DASA0V100A] 9 [AVFC2DGSA0V108A, A 3.967E-04 3.5294 94.6041 NORMAL VFC200SA0V1088) 10 (VSFC2CVSSOV151A1 1.275E + 04 1.1343 95.7384 NORMAL [VSFC2CVSSovl52A) 11 [VSFC2CVSSOV153A) 1.275E 04 1.1343 96.8728 NORMAL e (VSFC2CVSSOV1530) 12 [VSFC2CVSSOV1518) 1.275E-04 1.1343 98.0071 NORMAL e (VSFC2CVSSov152B) 13 CVFR2CVS93
- 6.286E-05 .5593 98.5664 NORMAL VSFC2CVSSOV102 14 IAVFC2VRSA0V109A1) 3.884E-05 .3456 98.9119 NORMAL
[AVFC2VRSA0v109A2] 15 .3456 99.2575
,tAvFC20GSA0v108t] 3.884E-05 NORMAL (AVFC2DGSA0V1088) 16 [AVFC20ASA0V1008] 3.884E 05 .3456 99.6030 NORMAL
[AVFC20ASA0v100A]
~'
AVFC2RCSA0V101
- 3.305E 05 .2940 99.8971 NORMAL CVFR2RCS68 l
TA!1E 10 3. CONV.INMENT ISOLATION CAUSE TABLE REPORT - Pag) 4 of 7 MODEL Nane Sv2 Cause Table for Top Event CI are '3ptit Frec*iui Cl4 F'E Value of Cl4 = 5.1090E 02 Date : 09 AUG 1991 14:10 MC/LH Value os Cl4 = 5.1390E 02 Date : 28 AUG 1991 00:40 15:26:04 22 SEP 1992 Page 1 N o . . . Ci t s e t s . . . . . . . . . . Va l ue . . . . . 1 Inportaree 1 Cmulative Alignment... 1 (VSFC2CVS$0V152A) 1.029E 02 20.0234 20.0234 NORMAL 2 (VSFC2CVSSOV1528) 1.029E-02 20.0234 40,0467 NORMAL 3 -(v5FC2CVSSOV1538) 1.029E 02 20.0234 60.0701 NORMAL
- - 4 (AVFC2VRSA0V109A2) $.280E 03 10.2744 70.3444 NORMAL 5 (AVFC2DASA0V100B] 5.28JE 03 10.2744 80.618E NORMAL 6 (AVFC20CSA0V1088) 5.280E 03 10.2744 90.8932 NORMAL 7 [VSFC2CVSSOV153A,V 7.640E 04 1.4867 92.3798 NORMAL SFC2CVSSOV153B]
8 (YhfC2CVSSOV1518,V 7.640E-04 1.4867 93.8665 NORMAL SFC2CVSSOV1528) 9 (VSFC2CVSSOV151A,V 7.640E 04 1.4867 05.3532 NORMAL SFC2CVSSOV152A) 10 OPRCl2 5.322E 04 1.0356 96.3888 NORMAL 11 (MVFC2CHSMOV381,MV 3.942E 04 .7671 97.1559 NORMAL FC2CHSMOV378) 12 (AVFC2VRSACV109A1, 3.87BE-04 .7546 97.9105 NORMAL AVFC2VRSA0V109A21 13 IAVFC20ASA0V1008, A 3.878E 04 .7546 98.6651 NORMAL VFC20ASA0V100A) 14 (AVFC2DGSA0V108A,A 3.878E 04 .7546 99.4197 NORMAL VFC2DCSA0V1088) 1$ (AVFC2CHSLCV4608,A 6.680E 05 .1300 99.5497 NORMAL VFC2CHSA0V204) 16 CVFR2CVS93
- 6.227E 05 .1212 99.6709 NORMAL VSFC2CVSSOV102 i
17 [MVFC2CHSMOV381)
- 4.444E 05 .0865 99.7574 NORMAL (MVFC2CHSMOV378) 18 (AVFC2CHSLCV4608) 3. 887E-05 0756 99.8330 NORMAL
* [AvFC2CHSA0v204)
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(ACLE 10-3. CONTAINMENT ISOLATION CAUSE TABLE REPORT Page 5 cf 7
. MODEL Names BV2 Cause Irbte for Top Event Ci and Split Fraction C15 PE Value of CIS = 6.2710E 02 Date : 09 AUG 1991 14:16 MC/LH Value of C15 s 6.2720E 02 Date : 28 AUG 1991 00:40 15:26:35 22 $[P 1992 Page 1 Ne . Cutsets... .... .. ValUe..... % tmpvrtance % Cumulative Alignment...
1 (VSFC2CVS$0V151A1 1.002E-02 15.9758 15.9758 NORMAL
-l 2 (v5FC2CVSSov153A) 1.002E-02 15.9758 31.9515 NORMAL 3 (v5FC2CVSSOV151B) 1.002E 02 15.9758 47.9273 :nRMAL 4 cvFR2CVS93 5.782C-03 9.2188 57.1460 NORMAL 5 CVFR2RCS68 5.782E-03 9.2188 66.3648 NORMAL 6 (AVFC2VRSA0v109A1) 5.425E 03 8.6496 75.0143 NORMAL 7 (AVFC2DGSA0V108A) 5.42SE-03 8.6496 83.6639 NORMAL 8 (AVFC20ASA0V100A) 5./-25E 03 8.6496 97.3135 NORMAL 9 (VSFC2CVSSOV1518,V 7.481E 04 1.1928 93.5062 NORMAL SFC2CVSSOV1528) 10 (VSFC2CVS$0V151A,V 7.481E-04 1.1928 94.6990 NORMAL SFC2CVSSOV152A) 11 (VSFC2CVS$0V153A.V 7.481E-04 1.1928 95.8917 NORMAL SFC2CVS50V15301 12 CPRC12 4.985E-04 7948 96.6865 NORMAL 13 (MVFC2CHSMOV381,tiV 3.988E 04 .6358 97.3224 NORMAL FC2CH$MOV378) 14 (AvFC2VRSA0V109A1, 3.977F 04 .6341 97.9565 NORMAL AVFC2VRSA0V109A2) 15 (AVFC2DASA0V1008, A 3.977E-04 .6341 98.5906 NORMAL VFC2DASA0V100A) 16 (AvFC2DGSA0V108A, A 3.977E-04 .6341 99.2246 NORMAL VFC20GSA0V1088) 17 (AVFC2CHSLCV460A,A 6.817E-05 .1087 99.3333 NORMAL VFC2CHSA0V200A) 18 (AVFC2CHSLCv460A, A 6.817E 05 .1087 99.4420 NORMAL VFC2CHSA0V200C]
19 (AVFC2CHSLCV460A,A 6.817E-05 .1087 99.550: NORMAL VFC2CHSA0V2008) 20 (MVFC2CHSMOV381)
- 4.606E-05 .0734 99.6242 NORMAL (MVFC2CHSMOV378) 21 (AVFC% dSLCV460A) 4.076E 05 .0650 99.6891 NORMAL (A'/'C2CHSA0V2000) 22 (A' FC2CHSLCV460A) 4.076E-05 .0600 99.7541 NORMAL (A IFC2CH5A0V200C1 23 (AVFC2CHSLCV460A) 4.076E-05 .0650 99.8191 NORMAL n
TASLE 10-3. CONTAINMENT COLATION CAUSE TA LE REPORT P:ge 6 of 7 MODEL Names BV2 Cause Table for Top Event Cl and split Fraction Cl5 PE Value of CI5 = 6.2710E 02 Date : 09 AUG 1991 14:16 MC/LH Value of CIS = 6.2720E-02 Date : 28 AUG 1991 00:40 15:26:42 22 $EP 1992 Pege 2 No. . . Cut s e t s . . . . . . . . . . . Va l ue . . . . . % 1mportance % Cunuletive Aligrynent. .. (AVFC2CMSA0V200A1 24 (MVFC2CHSMOV3811
- 3.564E 05 .0568 99.8759 NORMAL CVFR2CHS173 25 [AVFC2CHSLCV460A,A 1.001E-05 .0160 99.8919 NORMAL VFC2CHSA0V200A,AVF C2CHSA0V200C,AVFC2 CHSA0V2008,AVFC2CH SLCV4600,AVFC2CHSA OV2041 26 [AVFC2CHSLCV460A,A 4.688E-06 .0075 99.8994 NORMAL VFC2CHSA0V200C,AVF C2CMSA0V20081 i
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TASLE 10 3. CONTAINMENT ISOLATION CAUSE TABLE REPORT Page 7 ef 7 MODEL Nane BV2 Cause Tebte for Top Event Cl and Split Fraction C16 PE Value of C16 = 1.1830E*02 Date : 09 AUG 1991 14:16 MC/LH value of C16 = 1.1880E 02 Date : 28 AUG 1991 00:40 li ??:30 22 SEP 1992 Page 1 Wo. . . Cut s e t s . . . . . . . . . . . Va l ue. . . . % IrTortance % Ctnulative Alignment... 1 CPRC11- 7.316E 03 61.5825 61.5825 NORMAL 2 IVSFC2CVSSOV151A,v 7.644E-04 6.4343 co.0168 NOMAL SFC2CVS$0V152A] 3 (VSFC2CVS$0V153A,V 7.644E 04 6.4343 74.4512 NORMAL SFC2CVSSOV1538) 4 (VSFC2CVSSOV1518,V 7.644E-04 6.4343 80.8855 NORMAL SFC2CVS$0V1528) $ OPRCl2 5.036E 04 4.2391 85.1246 NORMAL 6 (AVFC2VRSA0V109A1, 3.851E-04 3.2416 88.3662 NORMAL AVFC2VRSA0v109A2] 7 (AVFC20ASA0V1008,A 3,851E-04 3.2416 91.6077 NORMAL VFC2DASA0V100A1 8 (AVFC2DGSA0v108A,A 3.851E-04 3.2416 94.8493 h0RMAL VFC2DGSA0V10881 9 1.0968 95.9461 NORMAL jVSFC2CVSSOV153A) 1.303E-04 (VSFC2CVSSOv1538) 10 (VSFC2CVSSov1518) 1.303E 04 1.0968 97.0429 NORMAL [VSFC2CVSSOV1528) 11 (VSFC2CVSSOV151A1 1.303E-04 1.0968 98.1397 NORMAL (VSFC2CVSSOV152A] 12 CVFR2CVS93
- 6.213E 05 .5230 98.6627 NORMAL VSFC2CVSSOV102 13 (AVFC2VRSA0v109A1] 4,003E-05 .3370 98.9997 NORMAL (AVFC2VRSA0VI(1A2]
14 (AVFC20GSA0v108A] 4.003E-05 .3370 99.3366 NORMAL
-[AVFC2DGSA0V1088) 15 (AVFC20ASA0V1000) 4.003E 05 .3370 99.6736 NORMAL (AVFC2DASA0V100A)
e i Table 10-4. This table lists'the small, early containment failure or bypass: importance, vbich -is the_ percentage - that each containment-isolation split' fraction contributes to the_ Release Category Group _- II frequency.
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n ...-. _ _ TABLE 10-4. CONTAINMENT ISOLATION SPLIT FRACTION IMPORTANCE TO SECFBY Split Fraction Split Fraction SECFBY Importance SECFBY Percent Description importance Cl1 Containment isolation - 5.45E-03 - 0.55% All Support Available "~ Cl2 Containment isolation - 1.40E-03 0.14 % Loss of AC Purple Power Cl3 Containment isolation - 2.51 E-03 0.25 % Loss of AC Orange Power Cl4 Containment isolation - 7.86E-05 <0.01 % Loss of SSPS rain A CIS Containment Isolation - 9.27E-05 <0.01 % Loss of SSPS Train B Cl6 Containment isolation - 6.94E-03 0.69% Loss of AC Orange & Purple Power Cl7 Containment Isolation - 0 0.00 % Loss of Both SSPS Trains and OS is Failed CIF Contaanment Isolatim - 7.32E 01 73.21 % l Guaranteed Failure i
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e 1 l Quession 11. The Table on Pages 2.4 1 and 2, identifying valk-throughs, does not explicitly identify any specific system valkdowns by analysts 10 account for the impact of plant modifications prior _to valk-l: throughs, or modifications conducted during- the time .rame of the IPE. In addition, in the list _ af information sources (Table
?.4-1). there is no mention of Engineering documents used to_
control plant modifications. Vhat is the " FREEZE" date used for the plant configuration analyzed in the IPE? Since there is usually a lag time between docu~+nts that. request plant modifications ana revision .o document;-
-hat were use.1 to i base.the models on, were any -modifications incorporated in the-plant tha' vere being done just before the-freeze date-that vere -
not incorporated in the model? Response 11. Host of the information sources listed in Table 2.4-l'are outputs of the desigi. +ange process, which Engineering user to. control plant modifications. The " FREEZE" date that was used for the-PRA/IPE plant configuration was December 31, 1988. Per DLC procedures, Valve Operating Number Diagrams (VONDs) must be updated and _ approved prior to any plant modifications being operationally accepted, and these measures . ensured that these documents. reflected the as-built conditions of the plant at- the time of the " FREEZE" date. -Since these VONDs vere used as the. basis for-the PRA rystem -fault -tree -models, all-plantLdesign modifications performed and_ turned over to the Station prior to - January 1, 1989 vere _ incorporated _ into ' tlut PRA. Iluman Actions used in the PRA models were based on the procedure whosel revision vas in-effect at the " FREEZE" date. Question 12. Duquesne Light Company (D'.C) has _ stated that the PRA for BV-2-vas originally-performed by Packard, Love and Garrick, Inc. (PLG) and Stone & Vebster Engineering Corporation (S&V), and thai DLC personnel incorporated pleat-specific data and requantified the-model. Ilovever, Table 5.3-1 shows minimal involvement of the DLC organization in-reviewing the quantification _ Since expertise in the . methods is- important to ensure : bat the= techniques are correctly applied,- plense discuss DLC personnel
~
participation in the update of the BV-2;Hodel and the completion-of the Beaver Valley Unit 1 (BV-1) PRA.
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Response 12. DLC personnel vere intimately involved in the development of BV-2 system models, including their quantification hence this DLC involvement satisfied most of the quantification review process and on1.y_winimal additional DLC reviev vas necessary. The update of the BV-2 Model (i.e., development of a plant. specific database, and requantification of system fault. tree and event tree models) j was performed entirely by DLC personnel. DLC personnel took the lead in the BV-1 PRA and performed the majoritv of vork for all aspects of the PRA. ; Question 13. Section 5.4 resolution of comments indicates that the review comment / resolutions vere dovimented in a*cordance with the PLG- t 0223, " Quality Assurance Program". Does conformance with'th ; program comply with the DLC in-house requirements ivr , doewnen ta t ion? l Vill comment / resolution for BV-1 use PLG's program or DLC's? Response 13. The PLG Quality Assurance Program requirements for documentation meet or exceed the DLC program reqeirements. BV-1 Comment resolutions vill be documented per the DLC program.
?
Question 14. Table 3.1.1-2 ideniffies Instrument Air as being captured ander Initiating Event "TLHFV". However, there is no discussion in Section 3.1.1 (Initiating Events) which i nd '.c a t es that the ; frequency of this event was added to the "TLMFV". Plesse_ identify 1 the frequency of Loss of Instrument Air (LOIA), and the source, i.e., whether the frequency was obtained from generic or plant-specific data. Response 14. Loss of Instrument Air (LOIA) vas not explicitly added to the Total Loss of Hain 'Feedvater (TLHFV) Initiating. Event frequency because the f requency of LOIA is a small percentage of TLMFV. The frequency of LOIA frog generic sources reported in IPE Reference 3.3.1-3 is 2.0 x 10' per year. This frequency corresponds to total Losses of Instrument Aic. It is seen _to be very small compared to the Total Loss of Halia Feedvater frequency of 0.12 per year. Partial losses of air to' individual components (e.g., HFV valves) are already accounted - for _in the frequencies-assigned to i such initiating events, i.e., such events 'are included in.the derivatir n of initiating event- frequencies for part2a1 and total - losses of Hain Feedvater. . question 15. Discuss the impact ,s LOIA :on front line and support syste.as - i designed to mit4 gate.the effects of failutas sustained during.or , after'a trip, and the rationale used in combining-the event with TLHFV as opposed to treating it as a unique Initiating Event. l l _C. ..u.__- a___... _ -___~__-..-__,___,_.-..,-.u,,.-.._...., . . ,,
1 l Response 15. The u pect at LOIA on systems is described in IPE Section 3.2.2, dependency tables. Both the intersystem dependencies associated , _vith instrument air and the. containment instrument air systems are identified. The first page of the Support-to Frontline System Depencency Table, Table 3.2.3-2, vas inadvertently omitted f rom : This first page of the table is attached as the submittal. Page 59 of this submittal. : i The loss of containment ins t ruinen t air vould cause the-air-operated containment isoletion valves inside containment to fall to the " fall-safe" position. As indicated in Table 3.1.1-2, this l vould cause the CCP isolation valves for the RCP thermal barrier ' cooling to fall closed, but RCP seal injectior. and RCP motor
- Therefore, loss of=this system cooling would remain available.
vould not cause a plant trip. The immediate impact on the plant of a loss of Instrument Air would be closure of the feedvater control valves, the condenser l steam dump valves, the isolation of letdown, and the loss of ; control for normal pressurizer spray. The HSIVs vill eventualJy-close when their accumulators are exhausted; i.e., after approximately thirty (20) minutes. The most significant impact of a loss of instrument Air or systems designed to mitigate a plant trip is the loss of Hain Feedvater and the Main Condenser. The TLHFV Initiating Event includes the same impact on main feedvater and the condenser as the LOIA. Treating LOIA as a unique Initiating Event vould add a minor
.ontribution to the total core damage frequency. This conclusion is based on the fact that the Auxiliary Feedvater flow valves enul the pressurizer. PORVs are not dependent -on compressed all at .
Beaver Valley Unit 2, as they are at some other plants.-
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p Question 16. Discuss the technical basis, or provide a reference for " assuming" that "very small LOCAs" (less than 1/2 in. equivalent diameter) are within the makeup capacity of the normal charging system and, therefore, these events could be " conservatively" included with "small LOCA" Initiating Events (Page 3.1-7 in Section 3.1.1). Response 16. The reference to conservatively include very small LOCAs within the small LOCA Initiating Event category refers to only those events which lead to an immedfate plant trip. The accident sequente model for small LOCAs is -conservative for such cases, because it assumes that recirculation f;om the containment sump. will_eventuclly be tequired. By contrast, in the analysis _for Surry in NUREG/CR-4550, for very small LOCAs, credit was taken for cooling down and going on closed loop Rl!R as an alternative success path to that of recirculation from the sump. _The omission of this succcss path makes the current analysis conservative for very small LOCAs. The FRA Team at Duquesne-Light Company is-unavare of any small or very small LOCA events at a U.S. nuclear plant, which resulted in the need for recirculation from the containment sump. Int;ead, all such LOCAs which aave occurred-to date, have been successfully mitigated by RCS depressurization and successful closed loop EllR cooling. For very small LOCAs which do not lead to an immediate plant trip, the leak rate must be within the capacity of normal charging; otherwise. the net _ loss of inventory would lead to a plant trip. Normal charging is designed for mitigation of RCS breaks up to 3/8" in diameter. For very small LOCAs, less than this size, the operators vould initiate a controlled, manual plant shu'down;- which is not-considered a plant initiator. Question 17. Discuss the impact of LOCAs or Steam Line Brorks on mitigating systems as Initiating Events. Response 17. Table 3.3.7-2 of the IPE submittal lists the *mpacted plant systems for all Initiatirg ' Event categories, including LOCAs and Steam Line Breaks. The PRA did not take credit for any equipment' that was not environmentally qualified for the conditions present Residual Heat Removal equipment. following these initiatots, e.g., i In addition, Beaver Valley Unit. 2 UFSAR -Section 3.6 addresses l protection against dynamic effects ~ associated with~the postulated- ' rupture of_ piping due to Steam Line Breaks (UFSAR Section 15.1.5) or LOCAs (UFSAR Section 15.6.5). r I L
i Ouestion 18. Unlike the intormation provided for component data, there is no l discussion or identification of plant-specific. data used in the ;
"updat!ng process" for Initiating Events. l a) Provide a listing of the frequency of Initiating-Events (e.g.,
Turbine Trip, Reactor Trip, Loss of Offsite Pover/ Main . P.V./ Instrument Air) that vere obtained from plant operating experience, as opposed to those arrived at through system 4 analysis. b) Include a discussion of the updating process. for the t~ Initiating Events and a discussion of the frequency of those events vhort total frequency is made up of multiple events (e.g., TLh.4). Section 1.1 states that in 1991- DLC developed a plant-specific database and used it to requantify the Unit 2 PRA rodel. However, Section 3.3.2.1 indicates that the plant-specific data presented and discussed in Section 3.3.2 was collected between 11/87 and ; 12/88. c) lias the data presented been captured through 1988 or 19917 d) Is the PRA model. quantified using plant-specific data different from what is presented in-the IPE? e) If the PRA model has been' quantified using plant-specific data through 1988, please provide a discussion of any plans to update'the database and the PRA model, and any component failures or Initiating Events' occurring _ since 1988, which-vould impact the IPE results.
. Response 18. Table 18-1 shows the Beaver Valley Unit 2 Initiating Events that
- vere obtained using plant operating experience. Additional information,. including a oiscussion on =the _ loss of bus, 2A '
(Initiating Event LB2A) is addressed in Section 3.3.2.4. - Ta bl e~ 3.3.1-4 lists the Initiating Event distributions- for all of the > initiators that were used in the IPE. The methodology used to develop .the- plant specific 7aitiating ' Evant-frequencies that vere based on plant operating experience-is sinilar'to the two-stage Bayesian approach used for component failure. rates, as described in Section 3.3.1.3 of the-_IPE: submittal, y i l
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~ 1ncluded information that was collected between 11/87 and 12/88. .
A speciflc plan for updating the plant specific database and i system models has not been determined, llovever, ve intend to l periodically perform these updates, as needed, based upon the. ; Any component failures or ; significance of plant changes. Initiating Events occurring after 1988 -are not expected to have a -, major impact on the IPE results. I I i 1 e h i 4 1 i J 62'- I i-
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_m.,__ . _ _ . . . _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . . . . _ _ _ . _ . . _ _ _ - . _ 4 Question 19. Generic Letter 88-20 and !JUREG-1335 request that the IPE submittal provide a list of all generic plant data for equipa:ent _ and Initiating Events, including origin and method of analysis. Since Section 3.3.1 indicates that for a majority of components the generic component failure rates vere taken f rom "Databa.% for i PRA of Light Vater liuclear Power Plants", PLG-0500, 1989 and I since this document is not in the public domain please provide a listing of the generic component failure rates used for the BV-2 IPE (or the PLG database used in the analysis). This list.should include those genreic values used as a. basis for updated values. l Response 19 Table 19-1 lists all. of the PLG- generic component failure rate distributions used in the IPE. This data vas used either . . . directly, when plant-specific data was not available, or as the basis for the plant-spec;fic updated values shovn in Table 3.3.2-2. h T p F 6 l
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1 l 1 l TABLE 19-1. GENERIC DATA DISTRIBUTION REPORT FOR COMPONENTE PAGE 1 NAMI. cf DI FI RI N T IO9 M[fr LTH 41LE MEDINI 9b7H 41LE E Tidd t ' 12LV IC FATTERY - IAI Tkt Cl NTUI ou DUVM 4.NE-04 7 . 2 0 E- M 3.3GE-04 1.10E-03 n n_h7 h 1MV rr kr.,Ti t k y - rt>H m E OF W TfDT M kD d Oi U ATION 7.530-07 6. N E-T 3. 7 4 '? 1.64E-06 Z1[cHR bh1TEk / CHA6L k - i AI Luff Ullm OLLkJsTIOr' 1. f%E 45 9.502-07 6.25E-06 S.300-05 i ZTm Ik U. ~ - I AI LUb E fxkIlt. Of LkJs7 j oN 4. ME-07 6.9 4-tm 3.400-07 1.13E-06 1 I ZTcH C c1PNIT f* EN LR 14 % VA? /WD Af G r' - TAIL To c!.03E ON 1.61L-n3 2.6Pr@4 1.070-03 3.400-03 ElcFlO CIR-MIT 14 LJyLA ( 4 % V/r AND A!WL ) - IAILUFE To OiLN O t. 4 9EW4 6.40E-05 3.($E-04 1.400 03 f.Trb !1 CJRUUIT 14FM Lk f 4 4WA'? /J:D fJ.GL)-T kANOTER GI E N FtJkIlh 0.ME-07 5.4*F-00 3.79E-07 2.2M 06 ETUU2 CI RCV! T bVk ( AC Uh 1.C,11. 3 07) - f /d LUFE TO CLOSE UJ DEM ?.21E-04 6.4tE-06 fi . 9 E- 0 5 6.510-04 2Trb29 ClkCUlT 3MRIM OR II:, L7 , 4 PV ) - FAI LUF1: 70 01 E N W DrJ4 8.34-04 e . 3 FL -M 3.2FE-04 2.40E 03 1 71CFIT C1 hCul'1 M LAElk (AC OR IC ,LT.46 WJ - ThANM LA OI EN ICR 2.HL-07 2.990-08 1.2eE-07 9.690-07' l ITCIO r h T 104 Thir OkFM Ek - EAI L TD C I EnATc DN 0DVWD 1.7 70- % 4.14E-04 1.33E503 3.72r-03 I7CHLk CCNTROL kWH VLNil!AT!W CHILLER - FAILUkE TURIN i OLEkA 9.4tE-05 s.210-05 7.0PE-05 1.94-04 ZicHLa crWT60L ho* H VE NTILATim Chill Ek - FAILUkl Tv ETART UN B.07f-03 6.2ff-04 -4.720-03 2. ME-02 7.7; A H< Mb cmI ALEJ0R FAILUkE teh1N3 CluATtou 9.91E-c' 9.94E-CE 4.98E-05 * .400-04 7TcMIF AI k ct N H la Ok (AltUpr TO OTAkT W ULMfG 3,29E-@ 2 CI E-04 1.63L-03 1.12E-02
!.TcFJW E T HX1 CONT 60L hod - I AIL TO INJELT ON DEt4/J:D 3..;L-05 ? , ' Of -M 1.02E-05 9.12L-Op 2TD% ( NEUt%i ! C PMM k - fA1LUkl TU DllkATE ON CU%F 4
- 2 E-M ?.31L-04 1.Off-03 3.320-03
;TEAOf I NEUMAT.t c IMi lk - TAl l VM TU TkJJi?ll .4 70 lAILf! 00SITI . . tM- 0 4 ' . 5 7 E46 1.04E-04 7.62E-04
*TDADT I Nf Uf %T I C ($Pi i. k - TRAhM Lk OFEN Uk @P T WIN 3 Of f kATI
- 671-07 1. %E -09 1.100-07 ti . 0 0 E-07 Z T1 + DD OA @kAFT D/NFLk - ild LURE To Ol EN m L1'UJ L .69E-04 5.33E-M 1.44E-04 6.27E-04 ETu n f,AN DhAl T NWFP - 1 HMf LR cio;LD 1.041-06 2.43E-09 7.PDE-09 2.19E-06 )
41 Dl h i rikt DAHi t P - 1HAEMT/J!T ACTUATH41 4.20E-Oe 1.E70-09 1.30E-D9 1.19E-07 l 2T N31 LMLSE1 GEM LATuk - TAILUhL lCkI!M IIEET H L. OF OFLkATION 1.760 02 1.070-03 8.240-03 5.36E-02 l ZTIG DIFFEL lENLhA1(* - FAI LUb t '10 KUN AFTER fikST HOUR 2.51L-03 2.97E-04 1.49E-03 7. 4 4E-f 3 l l
;-1 D L , IFITrF1. M NERATM - l AItUPE TO START ON DE'%9 D 2 .14 E- M . 500-03 1.3bE-02 6.44r.-0 ,
LTDRYf AIK IMEk - COMihESJLD AIR JYE7EM - FAI L DUR. OiER. 1 <_,3 E c 2.31E-06 P . 0fi E- D A 4.10E-07 l
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2T FLI T VLUl l LAT 1 W FIL1ER 1.070-06 3.04E-6h 4.16E-07 3.0$E-C6 , I 2TIL20 VINTILATION LO"YhE - Fii*0 3F D 1 PE-07 3204049 4.1(E-00 3.0LE-01 1 2TEL3F IULL OIL FILTik - FLU tlE D 1 07F-06' 3. J4 E- 09 4 .1 tb O 7 3.05E-06 l ' -l ZT N R 1 Akil FANS - F A!1 %E DURDA3 el E!+i' ION 7. HF- 06 1.5tt-06 6.23E-06 1.5BE-05 l ITibir IJ M C FAN 5 - TAILURE TO START ON DD%"D 2.93E 43 3.270-04 1.HE-03 B.35E-0S ZTIN;h VENTILAilnN IAN - rAI Lt'RE DUnim OFERATION 7. DE-06 1. ME-06 6.23E-0? 1.58L-05 2Tfh2F VFNTILATION FAN - FAILURE TO STAh! ON DEMJdD 4.04L 24 4.9fE-05 0.03E-04 1.24E-03 ! 2TTUlk PUOE - LAIL UFEN DUhlNa OlfkAT?OF 9.200-07 2 . f0 f. ~ 05 3.160-07 2.83E-06 l T Mhfi - HE AT D r Wi lf R - RUiTURt:/LXCEMIN E 1 EAEAGE DMkIE OIEkA 1. 4E-06 2.21E-07 1.32E-06 L 18E-06 l 2TINVR INVLhiTk - FAILURE DDAIN3 OPEfJJ10 1.e3E.05 -1.ME-06 1.13E '.370-05 Z Tm3R MA ok GENEhAT W - FA! LORE CURIN5 OPERATION 3. ! 4 E --% 9.60E-07 1.100-05 1.20E 04 TT H NK - !WJ4 ALLY et LRATED MMoR I*kI Vi_N PP P - T AILUhE DUf ING Of 3.36E-05 2.03E-06 !,59E-D$ 3,B30-05 I ITPMO$ NOFftAtt y OLEMTED MDTON Ok!'.D PtHrJ-FAIL TO 7 AAT CH 2 3! . 1.47E 1.4tE-03 7.3BE-03 I i ZT TM2 OTANELY M2TOk DRIVEN KHF . - FAill:nE * *? AIN T Ol E AAT!t_N 3. 4 2 E 45 2. ME-M 1.71E-05 9.320-Q5 1 ITfMr STANDBY M7 : 't ISIVE: t uMts -f AILURE To STAAT ON DEFAND 3.24E-23 2.G1E-04 1.63E-03 1.120-02
- 'T F 0 t u FIFE ,GEFJafh THAN THM E INCH, IEh FIIE CLCTION R.ME-10 1. % E-12 1. KE-10 2.02v09 2TFr2D 12M ,1ES7 THAN TREE INN, i f R FIi t M PTION - S . U E-09 1. GEE-11 1.KF 2.02E-Oe-ZTP31R FOKER TIfLY 1.71E-C5 1.03E 46 7.600-06 4. 9;E-05 )
ZT r? ?h TMPINE M1VEN AM, (TEDWA!Ek FW - I AILML IENING OPE 1, 0 3 E- M 6,90-05 4.210-04 3.01E-03 ZTf ?E TUkb!NL Dk!VFM AUX.rtELA Tik OUMP - IAILUhE TO STAh? ON 3,31E+02 6.CLE-03 2.45E-' B.25E-02 ITRLID_ -kEIAY
- FAILMRE TO OiERATE ON tat 4 D- 2. 41 E-N 2. 39E-05 1.1DE-04 7.47L-04 E T RLI F. h E l>W - FAIllhE ICk1Na e E l kAT D " 4.20L-01 2. N-% 1.9eE-07 1. 51E4 06
-7TSC10 FERVICE hATrk STRAINEL - FA!!M DL KING of ERAT!W 6,22E-h3 P.0 & O7 3.o0E-06 1.5EE-OS 2 T "2 P - STI,AINLR OTFEk %M LENVICE WATEh - FAILU'E U!RIN3 OPER
- 22E-M E.0SE-D' 3.90E-06 1. M E- 05
*T3C3F EEkVICE KtMk TRAVELING TcREEN - TAI!.URE D%IN3 CE E AA!I 6 . .J E - M 9.0eE-07 s.60E-06 1.HE-05 i -_ _ _ _ _ _ _ _ . _ _ _ _ __ _ 2 ._ _ _ . _ _ _ . - _ _ _
. . . - . . - - ~ _ . . - - - -
l
- l l
TABLE 19*1. JENERIC DATA DISTRIBUTION REPORT FOR CCt4PONENTS TAGE 2 NA% Of DISTR 1!OT10N M U.N LTH 11LE HEDI/W 9biH 41LE I
. J P. T C(61 AlhMi tJT };UtiDifG r. RAY Not?!u li AAIN) [ L'JG 7.000 00 2,70L*09 3.020-00 2.000-07
!1GTCD F L1Cok Tk!P bkUJ Ek, SHUNT 1h1P "o!L -1 AIL To of f kATE 1.40E-04 L270-05 1.000+04 2.94E 04 ITMWI'D PISTAD!E l AI LUPE 10 Of f P1+TE ON LtFMD 3.e n-o? s.nE-:a 2.stE-07 9.ict-07 773WPI D!mM E .UUM0d OT EijsTItW .,21E-06 2. ME-09 4.v1L-07 4.61E-D6 ZTTrMT MUIDR-OI LR>sTED TJ SOUNNE"i I WJ TD! + N!!4 OlLN DURIfG OF 2.M E-01 1.70E+0e -1.20E-07 6.11t-07 ZTFWPD FRfWURT SWITCII - FAIL TO CNEMATE ON CUWlD 2.690-04 1.150-OL 1.09E-04 9.37E-04 ITTVID STORAGE TANM -- kt10T@E (L'A1HG OlaATHW 2.6CE-09 7.t9L-10 1.04f>00 7.(3E-04 ZiTkfR r!OW TRNQl.It'IER - FAIL t'URING OlihAThrJ 6,200-06 6.04T-07 4. 3 9E-M 1.40E-05:
ETTRLH LEVEL TAAWH11TER - TAILURE t%11M Off hM10N 1.57E-05 3.96E-06 1.2tE-Oh 3.340-0$ ET1 H F UfSSUht TRAtaHITid - TAI!ME ICMin DIEMTION ' 00E-06 9.90E-07 4.700-06 1.96E-05 ETW@ ELACTOh Tk1 P i+EAG N UNDLRVOL1 AGE Colt -fall TO OIEN O 2.EE-0) 6,430-94 2.0AE-n3 E.77E-03 ITVAOD AIR LtERATED VALVE / ( Af LUKE TO DilJ1d L ON DOWJD 1.52E*03 2.37L-04 1.0$E 03 3.32E-03 IT/Aot Alh tmATLD VALVE rAILukE TO Tk/WSIEF TO TAILED 195IT1 2.66L-08 7.57E-06 1.04E-04 .L 62E+04 FTVA01 A.IR 00l%ATED YALVL / ThAH!if1R OitN/SWT IUh1NG OTERATIO 2,6'E-07 1.60E-06 1.10E-07 8. O t> E- 07 LTVCDD CHECN VALVE / IAILUhE TO OfEkhit ON HFAND-Of f LEIC $,090-04 1. 0 0E- 0 *, 1.37E-04 1.440-03 E1 VCCC Sh!!O CHEC /ALVE lAILS TD CLOSE ON LDtAND ft . 3 5E- 0 4 1.65E-05 2.24E+04 2. Yi '-03 nod FWING CK2 VALVE TAILJ TO OlEN ON DeM,h D 1.0 M -D4 3.60E-06 4. 91 E-O b 5.10L+04 ZTVCOL CriECE V/1VE (OTEEk THMl MON - GPon LIAEAGE DUh!H1 OF S. NE-07 9 '. - 0 8 3.17E-07 1.260-06 ZT74 CHLCV VALVE {OTHER THAN STOP) TH k CIOSEL/LLUGOLD 1.64f-00 '.w3L-09 7.60E+09 2.19E-06 ZTVCOX CHECK VALVE LEAVAGE kATE @ EATER 't, TAN 1 C s GPi ( H R HCY 'k ) . 7 '130 -O rc i.71E 09 2. PIE-06 2 81E-07 - ITVCSD CH!.CK VALVL MTCP ) FAILURE TO OLLhATE ON DEMAND 9.1 X+04 7.07E-05 4.14E-04 2.61E-03 ITVCSL CHFCK VALVE MTOP) - GEOM LEAEAGE DUM N3 O! EEATIM b. MF-07 9. 2 3E -O tt 3.170-07 1. ? t,EW ITVWP CHLCK VALVF (STON - THANSFLR CLDSLD/ PLUu-JED 1.04E-GC . 4 X-0 9 7.P0t-09 2.19E-08 Z7 VELD EllCTho-HYDhAULIC VAIVE lEXCflT TSV,TCV) fAILUFE TO DI E 1.52E-01 2 37E-04 1.06t%03 3.3?E-03
. ZTVEIT EttCT40-HYDRAULIC VALVE (LXCElT TSV,TC"i TFA'43f Ek OPEN/ 2.67E v7 .1.50E-08 1.100-01 0.06E-07
- ETVE?1 TUT +D1WE_ STot/ CONT ROL VALVE ThADSH R. CLO?ED DUh!NG DE F RA /.9PE-05 9.230-07 1,13E-05 9.2EE @$
ITVE22 '1URMHL UTOi/CONTh0L VALYL TMWSFER DI LH DO M NJ Di r AATI 1.280-OS 3.54E-01 4.05E-06 3.55E.05 ETVE2 D TU6 HINE 0100/CiWTROL VALVE FA!!a.'ht 10 ORRATE ON DEMNiD 1.2$0-04 .92E-05 9.37E-05 2.63E+04-f7VHOT FWJUAL VALVE - TRANSf LR OIEN/EUUT DUh!NG OLEhATION 4.?OE-08 1.57E-09 1.30E-OB 1.19E 07 ZTVM& CHFCK VALVE OR MOV FISK RUITU M 4.53E-09 ] 09E-09 l'.tSE-06 1.65E-01 l ITVMOD MOTok OI ELATED VALVE + TAILURE TO Of f'N/CLCeE ON DEPAND 4.300-03 .7.49E-04 0.94E-03 1.0!E-02 71 VMOE MOV IAILURE TD CLOSE ON DEf WJD WHILE SHOWING CLo3ED 1.07E-04 2.10E-05 7.4?E-05 3.07E-04
- T'.NT tt.o V. I TRANStER CLOJ'ED cn TR7W3FER OFEN TURING OIERAT % .' 7 0 4 E 9.65E-09 5.%E-00 2.DE-07 l
- ZTVRIO SAFE 1Y VALVE - FA!LUkt TO OTEN ON DD%NL 3.2BE+04 1. 21E-05 . I d 9E 04 - 1. 06E-03 -!
ZTVh1S EldETY VALVF FAILUht TO Rf SEAT ON DEMMD 2.6?E-03 8.840-0$ 1.150 03 8,21E-03 1 IfVR1W SAFETY YALVE a FAILULE TO RESf.AT A!V h MATER E. 1. r1 E 2. P OE-Q 3 1.2CE*01~ 2.50E-01.
- 0TVR20 RELIE.T VALVE (EX0EPT PORV, SAFETY) FAILUEE TO OLEN ON DE 2.42F-05 1.ZE*07- a 721-06 6.92E-05 LTVR2T AEL10F VAM TOTHER THAM POhV OR *ATETY)' - FkfMATUk' OF 6.06E-06 9. 7 6 E -4.010-06 1.44E 0$
2TVRM PA FORY FAILUPE To hESEAT ON DDi4ND 2.50E-02 5.85E-03 1.67E 5,2$0 -ETVh3G fWk PORV lo1LUkr TO DIEN ON DDWW 4.2'E-03 9. ME-04 L 20E-03 8.90E-03 l ETVSOD DOLENMD VALVE (DIhECT AC'lNW FAILURE TO OPERATE ON DL- 2.43E-03 7.640-05 9,79E-Q4- 6.94E 03 ITVSOT - FDLENDID VALVE / Th'WSFER OiEN -Oh SHUT !NRING OPERATION 1. 2 7 E- Ot. b.19E-CB 4.91E-07 4.07E-06
*TVTCP BUTTERFLY TEMERATURE CONTROL VALVE + FLT TO Of ERATE ON - -1.!!T-03 2.370-04 1. 08 fw03 . 3. 30E-03 37VTCf DUTTFRFLY TEMIERATUPE CONTRCL. VALVE - FAIL TO TR FA TO F 2 660-04 7.57E-Of 1.04E-04-l7,620,04-ITvtCT WTTEKrLY' TEMf EAATURE CONTROL LLVE-TRU OFEN/ SHUT DVR1 4. 20E-OF -' 1'. 57 E- 09 1. OE-00 1.19E47 >
KTXkik- TRANS Fuk'tEh ' (GST,UAT, RAT ) - FAI LUF O IAlh1N3 Di f RAT UN 1. LEE-06 2.63E-07 '1.10E-O' 3.160'*06
' TTXh;R TRW TOhMc R (STN. SERVICE,4.16KV TO 400V) - FAILURE DURI 6.67E-07 1,050-G7 4.OE-07 1 370, OTXR3R TRAN3I OR4;R i lHJT WENT ) i460V TO ll'il FAILURE rn INN 1. %E-06 7.940-09 7.00E-07 4.670-06 t
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e Question 20. In verifying that the subwittal contained a listing cf Initiating i' Ev;nt frequencie-, it was noted: i " That the system Initiating Event f requencies in Table 3.1.1-3-vere ; different from the values provided in_ Table 3.3.5-2. A constant value is displayed for all parameters of the- i distribution for Initiating Events: VAX, VBX, and VXB. Explain these apparent discrepancies, and provide a discussion _ , regarding eny possible impacts on the. results presented in the , IpE, due to these discrepancies. Response 20. Table 3.3.5-2 used Beaver Valley Unit 2 plant-specific data for obtaining the Initiating Event frequencies derived from system models, while Table 3.1.1-3 used PLG- generic component failure rates. The values from lable 3.3.5-2 are correct and vere the ones used to quantify the Event Trees. Table 3.1.1-3 should have reflect these values. been updated *s When developing the Service Vater System modele frr Riskman, it was necessary to break che system up into several different smaller Top Events in order to generate the cutsets. The values , for the VAX, VBX, and VXB initiators were deH /ed frem equations that' used mean values from the initiating- uncertainty distributions for ther 3 smaller Top Events, and are, therefore, only point values based on the means of the smaller Top _ Initiating ! Event frequencies. This, however, has no effect _on the core damage frequency, since Riskman only. uses the mean values of the- _i Initiating Event distributions _to quantify the Event Trees. , Question 21. The Internal Flooding Analysis indicated that mitigating features , such as redundancy and separation vere considered. Ilovever , actual operating experience has demonstrated that. separate rooms do not necessarily provide protection because .of drain _ systems that are plugged or allov backflov, unsealed doors, or maintenance actions or situations. Discuss how consideration was given:to these conditions in the flooding analysis, and hev_they impacted. < the choice or quantification of Initiating Events. , , Response 21. -The envite, quantification and impact of interna 11 flood In!.tiating Events is based on evaluation of actual flood sources at each location, as well as potential propagation into the--location'. In i considering proparation, the number' and size of drains was ; considered,Jvhether-there are seals on the door _vas considered, and whether-the door opens out of the room or i.ito the room was - e,-. -, .n - - , ...,~,--..,--m-,,--,.,,--n~ .,,w,vs w,~ - , + - - - , , , . -w, ,c om .u-c . w ..n,e w n .r-- - - ~~vr+ . , ,
P i consi6; rod. In addition, backflov through drains was revieved. < These considerations were checked in the field during valk-throughs. In general, if there vere several drains in a room, it was assumed that mos vere functional (unplugged)' and a large , flood source vas requited to impact equipment. _Haintenance actions vere considered with regard to flood Initiating Events and are included in the Initiating Event database. Maintenance was , not explicitly considered vith tegard to open Joors or plugged drains. Most doots are fire doors which require frequent , inspections or fire vatches when open. Door seals vere qualitatively considered with regard to the :nos t likely . propagation path foi smallet floods. !!ovever, door Jeals alone vete not the basis for screening out propagation to an adjoining location. As an example, the Cable Vault Flood Initiating Event WVFL) is based on a flood that occurs in Room CV-2 where there is_o11y one (1) floor dral... The flood. collects in this toom, falling one ' train of HCCs, and eventually fails the door that opens into Room CV-1. The. flood is assumed to fail redundant HCCs in CV-1. A potential flood event in CV-1 is assumed _ to be envelopeo by the initiator in CV-2 becau,e there are several drains in CV-1, there is less flood potential in CV-1, the area is much larger in CV-1, and the more likely propagation path is into rtairvells and to.-the-pipe tunnel. There are doors ' hat open into_ CV-1 from the _ Emergency Switchgear Room, however, the drains and floor areas in the Svitchgear- rooms vere judged to be sufficient to handle leakage into the rooms. Question 22. Sections 1.4 (Summary of Mr.jor Findingu), 3.3,8_(Interior Flooding Analysis), and 4.8 (Back-End Results) do not characterize the , impact of internal flooding events, either as important or not significant. However, Figure 4.8-1 shovs that Cnntrol Building _ Flood (CBFL) events contribute approximately 6.6% of'the "small 1 early containment failures or bypasses", which is- the third largest contributing inttiator.- Provide a discussion of_ the flooding analysis addressing whether , the process yields 'non-conservative, realistic _or conservative estimates, and DLC's assessment of the IPE conclusions in_ light of this, especially with regard'to CBFL. Response 22. Section 1.4 is a brief executive' summar3 of the major _ findings which did not include a summary of internal flood contributions. However, Section .l . 4 does' describe the dominant core damage requences (Leve) 1) and early large release sequences (Level 2), 1- which indicate that internal ' floods 'a,e not. dominant-sequences. !- Section 3.4 summarizes the Level il results, and as shown in this. section, internal floods provide. a minor contribution to core . damage frequency.
- 68 O
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+
1 i Section 3.3.8 provides a qualitative summary of the internal flood analysis, the tesulting initiators identified from the study, and- i insights ftcm the study. The final results from including the l initiators in the overall accident sequence model are included in . , 1 Sections 3.4 and 4.0. Sec(lon 4.8 indicates that internal floods contribute, but do not dominate releases. CBFL's contribution to small early containment - failure is based on a sn ,*
- vater flood in the Fan Room next to the Hain Control ) tom, and 1re water flood in the Cable Tunnel. ;
Both floods are assumed t pagate to Elevation 707 of the 1 Control Bui141ng v1.ich houces process racks and other electrical auipment. This vill likely cause a plant trip _and coulc i spuriously operate equipment. A detailed analysis of the impact ' was not performed. It was assumed that solid state protection vould fail after the plant trip and, if the_ operators fall to initiate safety inj ec t ion , they ;vould_ also- fail- to--isolate - containment (small early releace in Level 2), resulting in core ; da.aage. There are potential conservatisms in that the service vrter flood would most likely propagate through double doors _in the Fan Room to the outside. However, it was conservatively- -> assumed tnat the flooJ vould push the stairvell door open_to , Credit was given to= operator detection and Elevation 707. isolation for the service vater flood. This-vas not the case for j the fire water flood. Another potential conservatism could be the : human errar rate used fer responding to this flood. . As mentioned above, a detailed analysis of_ the impact of the flood was not-performed, and a detailed analysis of the timing.of operator response was not performed. The conditional probability of operator failure to initiate safety injection and iso - containmentusedforthisInitiatingEvent-vas1.04x10"} ate . =the This process is considered to be conservctive since the-quaranteed- ! failure of operators to isclate containment after failure of SSpS results in a large percentage of its end-states to be assigned to the small containment bypass-release group.. Therefore, CBFL shovs minor contribution to core damage frequency, but indicated some importance to the Release Category Group II frequency.- I Question 23. It is noted. that in the discussion of-Top Events D0, DP IE,-IB, IV and IY, the time -that power is specified to be available is-dependent on "Hov Long The Batteries Last", and is identified as either'3.5 or 8 hours. However,'the sy" tem's description for DC Electric Power (Section 3.2.1 .9) states the assumption that following a loss of AC power DC power is evaluated for a mission time of just 2 hours. The EV-2 FSAR Chapter 8 also indicates that the life of the batteries under design loads:is.2 hours. I
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,- ._~v,, .,w..
I Discuss the technical basis, or provide a reference for the assumption of battery life longer than 2 hours, as relctes to the Top Events above. Response 23. Tne emergency 12 J DC ba'teries are designed to last 2 hours based on the licensing design basis-discharge. !!ovever after evaluating the actual plant operating design, Stone 6 Vebster engineering personnel have estimated an extended battery availability to cope with station blackouts. On the basir of this, it is expected that Battery th.mber 3 (which provides power to the vital instrument bus that supplies power for centrol of the auxiliary feedvater pump) vill last for at least 8 hours, and Battery 11 umbers: 1 and 2 (which provide power to the buses reouired _to recover the emergency diesel generators) vill 'ast for 3.5 hourc. Question 24. In Section 3.1.3.1 (General Transient /Small LOCA Tree) under the description provided for Top Event CI-(Containment Isolation), a c discussion is provided which relates to the Seal h0CA HoJe1. flovever, the discussion and Section 3.3.3.-(Iluman railure Data), which is referted to therein an containing the Seal LOCA Model,_du not explicitly describe the Hodd used for the IPE submit tal. L Provide a discussion of 'he Seal LOCA Hodel, as used_ in the !W4 submittal including the variouc leak . rates, timing of seal failure, and the probability of their occurrence with and without i the seal teturn line isolated. i In addition, discuss the impact on Core Damage Frequency (CDP), if ! the assumption is incurtoct that the low press.ure seal leakoff I pipe vill vithstead high pressure on failure of the number one I seal, Response 24. The General Transient /small LOCA Event Trees are described in. Section 3.1.3.1. Section 3.3.3 provides a. summary of the electr.ic , power recovery approach. The electric power recovery tesults are also summarized in Tavle 3.L 3-ll. lL The seal L^CA model is described in Reference 3.3.3-5, Appt.idix B,.. Section B.2. The- specific seal LOCA leak rates, used as a function of _ time af ter loss of seal cooling, are provided in Table B.2-1,-copy attached. The mooel for the pump seal leak rates was based on the four-loop RCP seal LOC". study of-Reference B.2-4 for Vestinghouse RCPs with-the old style 0-rings that existed in the Beaver Valley'RCPs at 'he time of the s bdy, and scaled by the nuber of loops at Deaser Valley to reflect the leak rates per pump. The flov -rates listed in GPH. define.the effective flow: area, assuming an RCS pressure of 2250 psig. The timeLto core uncovery for_ a given leak rate, which- varies vith time, was computed accounting for the- decrecse in PCS pressure as the-accident progtesses and -includes the effects of: the operator action to depressurize the Steam Generators. Refetence B.2 4.is . as follows: i
- 70 ,
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..-._.--.g-.--.~.-..- .. ~ .~. - .- _ .- - - - - - - - .. _ -. _ .,
~!:
I t f NUREG-11560, Report 'Peactor Coolant Seal LOCa, "Results of l Expert Opinions Elicitation on Internal Event Front-End Issues . Expert Panel", NUREC/CR-5116, Volume 1 for NUREG-1150: Sandia 88-0642, April 1988. ; i A 480 gpm leak rate per RCP is expected, given the assun.ption that - the--lov presrure seal leakoff piping ruptures _after failure of the i
#1 RCP seal. On the basis of the analyses performed in the SEALOC code (Reference B.2-5) with the turbine-driven auxiliary feedvater system available, this leak rate vould result in core damage r (1,200*F) approximately 7 hours after the ini;iation of the !
station blackout if operators took action to depressurize the steam generators in 2 hours or less. If the operators depressurize after 2 hours, or completely fail to depressurize at- ' all, core damage vould occur approximately 2.9 hours after the station blackout. Using these new core damage times in the-electric power recovery m results in a total core damage frequency of 1.95 x 10~gdel or approximately a 3 % increase. l Reference-B.2-5 is as follovs: ; R. K. Deremer, " Reactor- ' Maneke, J. A., D. R. Buttemer, and Coolant Purp Seal LOCA Analysis during Station Blackout Events at Seabrook Station,"- prepared for New flampshire Yankee, _; Pickard, Love and Garrick, Inc., PLG-0724, January 1990. e 5 e
)
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i Tabin D 21 Seal LOCA Flow Rates (GPM) pN Pump with and without Primary Depressurleation
- Time after Station Disekout (hours)
' Cumulallye Probability 01.0 1.0 1.5 1.5 2.5 . 2.5 3.5 4.5 5.5 5.5+ Probability (gpm) (gpm) (gpm) (gpm) (gpm) Jgpm) 0.2712 .2712 21 21 21 21 21 21 0.0151 .2863 21 2' 21 61 61 61 0.0161 .3024 21 21 61 61 61 61 0.0181 .3205 21 61 61 61 61 61 ! 0.0120 .3325 21 61 108 108 108 108 0.0059 .3384 21 61 100 108 120 175 0.1120 .4504 21 61 250 250 250 250 0.0130 4640 21 123 250 250 250 250 0.5302 .9942 21 250 250 250 250 250 1 0.0016 .0958 21 308 308 308 308 308 0 0042 1.0000 21 400 480 480 480 480 i-E l I NSWIN0043 042591 B.2-18 PLG. . E__-.______ . _ _ _ . _ . . _ _ _ _ _ , . _ _ . _ . . _ _ _ . _ _ _ . . . _ . _ _ . _ . . ., i
e l i Question 25. Section 3.4.3 of the submittal provides information on the j importance of the five (5) systems that perform Decay lleat Removal (DilR) functions, and indicates that no particular vulnerabilities , have been found. Ilovevet. the values provided in Table 3.i. 3-1 as ! the " percentage vi CDF in which event. is failed", shov a non-- l negligible contribution for some Top Events due to loss of support- l (e.g., MFF 9.7% and AFF 20.2%). A value for lillF (High llead Safety J Injection Pumps, Suppott Unavailable) is not provi Ndi however, , Table 3.4.2-1 shovs the percentage of CDF vith this split fraction ! as 62%. t Generic Letter 88-20 and Appendix 5 therein, indicate that support systems are important to the Di!R Function and suggests- hat they be consid 7ed in the search for DilR related vulnerabi ities. l Thetefore, please dircuss the impact of support systems on these five ($) systems, differentiating between the contribution from j Loss of Pover (LOSP and BVX), and other supports such as Se' vice Vater, Ptimary Component Cooling _ Vater and Instrument /Contal unent Instrument. -f Response 25. The contribution to total cote damage f requency for the five Decay lleet Removal Systems of Section 3.4.3, due to loss of electrical E and non-electrical support systems, is presented in Table 25-1. As can be seen by Table 25-1, the majority of spilt froction failures come from contributions related to the loss of electrical power, CIA /CIB algnals, and- Initiating Events _ unrelated to electric pover. The largest contribution to-core damage frequency ' due to loss of non-electric power support _ systems is 16.4%, which is due to- the loss of both service water. headers to-.the- y llHSI/ Charging Pumps Lube -011 Coolers. The next two large4t ' contributions to CDF for non-electric support ~ systems are 2.6% and_ 2.5%, which are RHR failures due to the loss of Primary Component Cooling Vater to the RilR pump seals and heat exchangers, and AFV-failures associated with the loss of Solid State Protection System , signals, respectively. The contribution _to core-damage frequency - for Hain Feedvater failures stemming from:the loss of non-electric power support systems is less than'0.05%. i k P 6 4
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TABLE 25-1. BREAKDOWN OF CDF DUE TO GUARANTEED FAILURES OF DECAY HEAT REMOVAL SYSTEMS Split Fraction ! Core Damage Frequency (Percentage of Split Fraction) [Fercertage of Total CDF] l Decay CN Damage Guaranteed Failures Guaranteed Failures Guaranteed Failures Guarantee 3 Fattures Hest Split Frequency Related to Loss of From Non-Electric Due to Loss of Related to CIA /CIS Removal Fraction [Pe-:entage Electric Powr Power Related Non-Electric Power Sqnals System of Total CDF] Initiating Events Support Systems HHSI HHF 1.14E-04 8.35E-05 ' 3.00E-05 (73.6%) N/A (26.4%) N/A [62.0%] [45.6%] ' [16.4%) RHR P.RF 7.63E-05 2.45E-05 3.11 E-08 4.85E-05 4.69E-05 (322Y; (< 0.1%) (6.4%) (61.4%) 41.6%] [13.4%) [< 0.05%) [2.6%] [25.6%] AFW AFF 3.70E-05 3.21 E-05 3.48E-07 4.63E-06 (86.6%) (0.9%) (12.5%) N/A i [202%) [17.5%] [02%]' [2 5%]
$ MPN. MFF 1.78E-05 122E-05 3.51E-06 4.82E-08 1.99E-06 (68.8%) (19.7%) (0.3%) (112%)
[9.7%) [6.7%] [1.9%] [< 0.05%] (1.1%) Operator CDF 1.36E-05 1.36E-05 Cooldewn (100%) N/A N/A N/A [7.4%] [7.4%).
._.-f_.._ _ _E__.. _ __ - _
, _ ... _ _____ _ __ - _ _ _ ... _ _ _ . - .. _ ._ _ _ _._-_m._.-__ _
i f v I-- - Question 26. Table 3.4.3-1 shows the percentage of CDF in which the Event AFF is failed as 20.2% (3.84E-5) identifying it as due to Large Flood ; in Safeguards Area. However, Figure 3.3,0.2 (comparative contributions to-core damage from floods) shovs that only 16.6% of ; the CDP fsom all floods (7.32E-6 x 0.166 1.22E-6) is doe to ! safeguards floods. Provide a discussion of this apparent ; discrepancy and other values in the- table -which may likeviva -* iLpact the results of f,e IFE. Response 26. Split fraction AFF vas mislabeled in . Table 3.4 3-1. Its r descriptiou should read "AFF - Guaranteed Failures of Auxiliary Feedvater", since it consists of all AFV failures, not-just those due to Safeguard Area floods. Only one of the many vays of _ in -the north and south-failing AfV consists of a large flood . Safeguard Areas for a period greater than 20 minutes (i.e., Initiating Event SGFL2).- As shown in Table 25-1, this fallute_ ; mode contributes less than 1% to the total split fraction failure ; frequency, with the other 99.1% due to the loss of support systems. All the possible vays of guaranteeing the fallute AFV are shown in the AFF spilt fraction rule below: I AFF - (STEAM 4 AFP22)
- AFP23A
- AFP23B ,
E b wheret i STEAM = TT F
- MS-F + INIT.SLBC + INIT-SLBD
- MS.F AFF22 - SA.F
- SB-F
- OS=F + INIT.SGFL1 + INIT.SGFL2 +
INIT 5LBC + INSTRUH ' AFP23A = SA.F
- OS=P + A0=F + INIT-DVX + D0 F + OG F
- VA F ,
+ INIT SGFL1 + INIT-SGFL2 + BV F -
1 AFF23B = SB F
- 05 F + BF F-+ INIT.BVX + DP=F + OG F
- VB.F
+ INIT-SGFL2 + BV-F and, INSTRUM = IR F
- IV.F
- IB F
- IY-F + BV F' Therefore, the percentage of core damage freque ey shown in Table 3.4.3-1 in which split fraction AFF is prest t is_ correct at 20.2%, however, only 0.2% of 'this is attributed to' Initiating -:
Event SGFL2. Question 27. As-indicated in the paragraph on -Feed and Bleed Cooling,-the BV-2 design minimizes the f requency of sequences involving failure of l AFV and-Bleed and Feed Cooling, relative to other PVRs -eviously studied", b- Tuse of credit taken for realigning th; electric motor-drive IV pumps. It vould' appear that this capability is ; of significu benefit to DV-2.- ( t
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Discuss the benefit derived from this capability in terms of CDF l' vith and-vithout this capability. In concert with this, please provide the benefit derived from the capability to feed and bleed upon loss of all secondary cooling -(i.e., MF and AF) in terms of CDF vith and without this capability. ! Response 27. The probabilistic failura of the operators to realign th ! pumps only accounts foi a core damage frequency of 6.51 , or x 10~g less total than cote 0.1 damage % of the frequency total core damage vould frequency. be increased . Iloveverf', to 2.33 x 10~ t heor ' a factor of 1.23, if this action is assumed to be a ;uaranteed ; failure (i.e., no credit taken for the action). Likewise, the l probabilistic failure of tL) operatory to initiate feed and bleed cooling only accounts for 1.76 x 10~ or less-than 0.1 % of .the :
~
total core damage frequency. The- total core damage frequegcy, however, vould increase by a factor of 1.17, or to 2.22 x 10' .i f no credit is taken for this action. . Question 28. provide a list of the types of Initiating Events identified as "other" in Figure 3.4.0-2, and the breakdown of their contributions to CDF. Response 28. Sea attached Table 28-1 for the breakdown on all of the Beaver , Valley Unit 2 Initiating Event contributions to the total core . damage frequency. h t A 4 l l F l' , t I L a
+y , g_w ,. *' -
t 't '*'d".= fT*'P{'t
TABLE 281. Deaver Valley Unit 2 Initiating Evant Contributions ; Initiator Percemr :) of initiator Percentage of ; Initiator Core Melt Total Core Initiator Core Melt Total Core Frequency Melt Frequency Frequency MellFrequency , LOSP 2.86E 05 14.84 SGFL2 3.81E 07 0.20 BVX 2.35E 05 12.17 LCVA 3.45E-07 0.18 SLOCl 2.15E 05 11.18 VSX 3.44E-07 0.18 ! SLOCN 2.06E 05 10.69 TLMFW 3.37E-07 0.17 AOX 1.48E-05 7.67 SLBD 3.34E 07 0.17 BPX 9.31E 06 4.83 VPFL 2.90E 07 0.15 IRX 7.24E-06 3.76 MSV 2.84 E-07_. 0.15 IWX 7.23E 06 3.75 LCV 2.84E 07 0.15 SGTR 7.21 E-06 3.74 ELOCA 2.65E 07 0.14 IMSIV 4.81 E-06 2.50 AOXA 2.29E 07 0.12 TT 4.56E 06 2.37 LPRF 2.23E 07 0.12 - LB2A 4.25E-06 2.21 LB2AA 1.90E 07 0.10 CBFL 4 05E 06 2.10 DPXA 1.85E-07 0.10 WBX 3.55E 06 1.84 DOXA 1.82E-07 0.09 DPX 2.88E 06 1.49 IBXA 1.70E 07 0.09 RT 2.52E 06 1.31 LOSPA 1.49E 07 0.08 MLOCA 2.12E 06 1.10 BPXA 1.37E-07 0.07 SGTRA 2.01E 06 1.05 SLBC 1.35E 07 0.07 _PLMFWA 1.90E-06 0.99 CX1 1.22E 07 0 06 PLMFW 1.56E 06 0.81 CPEXC _1.21E 07 0.06 ISI 1.53E 06 0.79 LPRFA 1.01 E-07 0.05 WXB 1.29E 06 0.67 lYXA 9.08E 08 0.05 . _ AMSIV 1.26E 06 0.66 IRXA 8.15E 08 0.04 DOX 1.25E-06 0.65 IWXA 8.10E 08 0.04 WAX - 1.20E-06 0.62 lYX - 5.03E-08 -- 0.03 ISFL 1.13E-06 0. *,8 IBX 4.97E 08 0.03 TTA 8.89E 07 C.46 TBFL 4.72E 08 0.02 LLOCA 8.52E 07 0.44 CVFL 4.32E-08 0.02 SGFL1 8.37E-07 0.43 WXBA 1.09E 08 0.01 EXFWA 8.26E 07 - 0.43 WAXA 1.09E-08 0.01 EXFW 6.78E-07 0.35 CX1A 8.13E-09 0.00 ABFL1 5.36E-07 0.28 ABFL2 1.75E 09 0.00 SLB1 5.10E 07 0.26 WBXA 3.35E 10 0.00 TLMFWA 4.10E 07 0.21 77 - "d.p- - *- p.g.w,e-%- .p e m . gm. e -4.yn.--. q,nu ,,y -p r-ps-, e- mey -' gr - y.viv,,wa-g.m..w.e-e--. - . .,. g eer g - . w
i i i Question 29. The nubmittal identified: core damage as having occurred when loss i
>f core heat removal progressed beyond the point of core uncovery.
and core exit temperatures excted 1200*F. j llow many sequences were screened out because of' this doub1v criteria? Discuss the impact on the resultant CDF obtained using ; this criteria. Please address the tollowing: \
- The basis for the temperature chosen (1200*F).
- Do all sequences with the core uncovered go to core damage,-or was there recovery prior to reaching 1200'F7
- Vould the CDF be significantly different without the 1200*F core exit temperature criterion?
Response 29. There vere no sequences ccreened out by application of the 1200'T criterion. The use of this criterion only marginally impacts the timo available for recovery actions, if the 1200*F criterion vas replaced by just core uncovery, some sequences vould increase in frequency by a slight amount to reflect the incremental effects of recovery between the time of ~ core uncovery and the time of core- , exit temperatures reaching 1200*F. For typical sequences in which the 1200'F criterion vas applied, there vould only be about
- fifteen (15) minutes' after core uncovery until core exit thermocouple readings in the Control Room woul reach 1200*F.
This estimate is based on HAAP ' analyses performm' for Seabrook Station as documented in the following reference : Fleming, K.N., et al, "Rish Management Actions to Assure
. Containment -Effectiveness- at Seabrook Station", PLG-0550, '
prepared for New llampshire- Yankee Division of Public Service
-i Conipany of New Hampshire, July 1987, Table 6-3.
~
Vhile the use of the 1200'F criterion does not appreciably impact the core damage frequencyL in comparison with core uncovery, this value uas relected for consistency with the Functional Restoration-
- Guideljnes that form part of the Emergency Operating Procedures.
At Beaver Valley, Functional Res;9 ration Guideline- FR-C.1, '
- Inadequate ccre Cooling, is entered w en core; exit thermocouple temperatures exceed-1200*F. . See,._for example, Part 12 of Figure 3.1.1-2, the Event Sequence Diagram - f or Beaver _ Valley - Uni t 2. : The actions in~FR-C.1 vere not- credited in reducing the frequency of.
core damage. , Because_of the small incremental time .between-core uncovery and- ; 1200*F core. exit temperatures, ~ tne use. ot this criterion has no < significant impact on the' estimation of-the CDF.
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1 I Question 30. The BV-2 submittal has identified loss of Emergency Switchgent Room IIVAC as a significant contributor to CDF, due to the ] relatively rapid rise in room temperatures that vill exceed the qualification tempetature of equipment in the room. Ilovever, experiences of other plants have indicated that temperature rise determined by test on loss of IIVAC is not as rapid as determined by calculation. The possible prediction by calculation of temperature rise significantly mote rapidly than might be experienced and could cause a distortion in the identification of contributors to CDF and subsequent misapplication of resources. Is DLC giving consideration to verification of the rate of temperature rise determined for the Emergency Svitchgear Room on loss of IIVAC, to establish if the contribution from this event is appropriate? Response 30. DLC is currently evaluating performing a test in order to verify the heat-up rate in the Emergency Svitchgear Rooms folloving a loss of all llVAC. Question 31. Section 6.1 Indicates that the two (2) risk factors of merit that have been considered are CDF and early release frequency. In t.ddition, Section 6.3.1 states that in order to determine vulnerabilities the major accident " CATEGORIES" vete evaluated along vith top ranking sequences, a) Provide the definition of vulnerability, and describe the process used in conjunction with the above to huntify the vulnerabilities as requested by NUREG-1335. b) Discuss the findings related to identifying potential vulnerabilities vith respect to containment failure or by-pass, and assessing any associated plant modifications, c) Discuss the anticipated benefit (decrease in CDF or irnpact on release category), the rationale by which the listed option was chosen from the potential options, and the respective timing, if implem@ ' ion for those "under review". d) Discuss the consideras on given to independent fallute of the Service Vater Headers (VA and VB involved in 13.7%'CDF, and in top ranking sequences involving small LOCAs which contribute 21% to CDP), and the common check valve in the suction of.the
!!HSI pumps (VL-1, involved im .approximately 15% CDF, and also in top ranked sequances involving loss of vital bus and small LOCA) as vulnerabilities. ,
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Response 31. a) The evaluation of contributors to core damage f requer cy and early release frequency is performed in a top-dovn systeLatic' ) manner, vorking from the-general to the specific._ First the j results ara broken down to examine general classes of accident sequences. Several different approaches are followed to. > dsfL+ accident sequente classes by a common characteristic. These characteristics include initiating event, plant damage . f state, split fraction, and combinations of these. . Once accident sequences have been classified in this manner, the i importance of the group can be evaluated- in terms of pctcentage contribution to CDP or percentage contribution to-early release frequency. Next _ the results are examined in scenarios as defined by the Initiating Event, split fractions-of failed Event Tree Top Events, and. the end states of the-Level 1 and Level 2 Event Trees. Finally, the causes of each event in the important scenarios are delineated-to-identify the fundamental contributors to risk. These fundamental contributors to risk are defined as vulnerabilities. i b) The Beaver Vt. ley Ui.it 1 2 Containment Building appears to-be l more vulnerable to early overpressurization failures, which contribute to more than 90% of the Release Category Group I frequency, than it does to large bypass failures, vhose . failures contribute less than 10% to the group frequency. As shown in the sensitivity . studies -of Section 4.8.4, severe accident. management procedures-reg.irding_in-vessel recovery of core damagw events and RCS depressurization vould '
.significantly reduce the containment carly overpressurization failure frequency. Therefore, no design modifica' ion ~to the-Containment Pullding or Containment Spray Systems are currently planned.
c) The potential. enhancements to address the identified. l
. vulnerabilities are listed in Table 6.3-2 of. the Summary .
Report. This Table lists the percen'. age contribution to cote. damage ~ frequency of each vulnerability. . The proposed enhancements vill reduce the CDP to the extent that they vill-eliminate the vulnerability. C t benefit analyses of the potential enhancements vill be performed _to' find th , s t,
- effective means of reducing- the vulnerability and detern, ng L a schedule for implementation.
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.l d) Independent failures of the Service Vater header flow paths '
(Top Events VA and VB) are dominated by the conditional probability that the RSS Heat Exchanger supply HOV fails to i open, or the CCP/CCS isolation HOV falls to close, on one Train, given that the other Train's HOV has already failed during CIB conditions. These HOV common cause failure-on- ! demand modes account for greater than 81% of the Top Event failure frequency, with the remainder due to independent j failures of the HOVs. Existing Emergency Doerating Procedures direct operators to verify that these H0k have been properly positioned following a CIB signal, and to manually align them ; if they have not. Since these_ valves are located inside the :l Service Vater System Valve Pit, operators should not have any i' problems accessing the area.following C1B accident conditions. Hence, taking credit for operator actions to manually align , these H0Vs vould reduce the split fraction failure frequency 7 and, consequently, the Top Event importa7ce to core damage ! frequency. The failure of the RVST check valve 2055-27, common to the suction of the HHSI pumps, curt 1ntly contributes , approximately 2.1% to the total core damage frequency. This vulnerability could be resolved, as addressed in Section , 6.4.5, by realigning flov from the LHS1 pumps to the HHSI pump ; suction piping, thus bypassing check valve 20SS-27, if it fails to open. Question 32. Discuss briefly the IPE results (including the contributions to CDF) of any analysis related to a small break LOCA due to a stuck-open safety valve event if the PORVs are blocked off to stop any leakage. The discussion should address the percentage of time the PORVs are blocked off -due to- leakage and failures of operator actions to open the PORV block valve during accident conditions. ' Response 32. The PRA/IPE rest 0.ts reflect the normal Station alignment of the
- PORV block valves at the time that the RCS pressure relief system models were being developed; i.e., two (2) block valves open 100%
of the time and_one (1) block valve :losed 100% of the time, prior to any Initiating Events occurring. The PRA RCS pressure relief models include-failures of the safety rel!ef valves.(SRVs) to
- reclose in the event that they open due to P0_RVs falling to open.
These stuck-open safety _ valves result' in failures to the RCS pressure relief models and, consequently, are - modeled as small'
. break LOCAs in the- existing PRA analysis. Operator actions to open the ."0RV block valves are modeled in Top Events OB (Bleed and '
Feed Cooling) and OD (Depressurization of_RCS for RHR Entry). The human actions and their mean failure rates assgelated with these e (4.34 x 10' ), ZHE0B2 (3.65 x top 10 2) vents are ZHE0D1 (1.19 x 10~as follonj).ZHE0B1 Failures of these human' actions only contribute a total of 0.39% to the core damage frequency.
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Based on the above, this is not considered to be a credible failure modes however, Table 32-1 is provided assuming that it is. Table 32-1 shove a brief analysis of the increase in core damage frequency related to a small break LOCA due to a stuck-open SRV, assuming all three PORV block valves are always closed. This analysis was performed by using the methodology for the third sensitivity case discussed in Section 4.8.4 of the IPE submittal and the following equation l I NewCDP-OldCDP+{()(OldCDF) PRI x (New SFV - Old SFV)) B As seen in the table, core damage frequency would increase by 46.6%, assuming that the PORVs are always isolated.- It should be noted, however, that this analysis does not include any increase in core dan. e frequency due to ATVS events as a result of having only SRVs available to limit the RCS pressure surge during these transients. l f a t Y I" s
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g ' q'; 5 TABLE 321. CORE DAMAGE FREQUENCY DUt t STUCK-OPEl: :.9V LOCAS CMEl Croup Frequency = 1831CE-04 (Ol' Jore Damage Frequency) Stuck-bren SRV (NO SI) = 8.6094E-03 (New SFV for PRO thre :gh PR9) Stuck-open SRV (S'S) 2 3.0363E-01 (New SFV fcr PRA thrcagh PRW)
~ . GF Name Denvative Old SF Value New SFV Old SFV Dervivative *
[G (Delta SFV) Delta SFV PRO 0.0000E+00 0.0000E+00 8.6094E-03 ' O.0000E+00 PR1 1.9243E-03 5.0210E 04 8.1073E-03 1_5601E-03
*' PR2 9.5979F 07 5 2240E-04 8.0870E-03 7.7618E-09 PR3 1.2168E-00 5.1040E-04 8.0990E-03 9.8549E-09
[ $R4 PR5
-4.9595E-01 2.1608E 07 9.1130E 03 2.5930E-02
-5.0360E-04
-1.7321E-02 2.4976E-11 3.7877E-09 PR6 4.0825E-07 -4.2351E-02 -1.7290E-08 5 0960E-}2 [
PR7 f.9471 E-05 4.9460E-G2 , ,
-4.0851 E-02 3.2464E-06 ,.
PR8 4.6981 E-05 5070E 02 412 E-02 -7.7334E 07 , PR9 1.3512E-Ot .6710E-02 ,
,-} ,"f 01E-02 -9.2018E-06 PRA 2.7320E-04 2.00105-03 1( - 3E-01 8.2405E 05 O.0000E+00 1.0000E+00 -o.9637 E-01 0.0000E+ 00
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2.6781 E-00 2.1240E-01 9.1230E-02 2.4432E-07 Pki 2.1459E-06 1.0200E 01 2.0163E-01 4.3268E-07 - PRJ , 1.33I?E-06 3.0340E-01 2.3000E-04 3.029E-10 d I.c -3.3007E-08 2.0570E-03 3.0157E-01 -9.9540E-09 PRT -3.3957E-09 2.0020E-01 1.0343E-01 -3.5122E-10 PRU -3.10@E-09 1.0270E-01 ,. 2.0093E-01 -6.2469E-10 PRV -3.2707E-08 2.9890E-01 4.7300E-03 -1.5/ 70E-10 PRW 0.0000E+00 0.0000E+00 _ 3.0363E-01 0.0000E+00 Sum of (Derv.
- Delta SF'y = 8.5432E-05 New CDF = Old CCF + Sum of (Derv.
- Delta SFV) = 2.6850E 04 Percent increase in CDF = 46.64 %
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