NSD-NRC-97-4937, Forwards Draft AP600 Ssar App 1B.Response to RAI Re Severe Accident Mitigation Design Alternatives Also Encl

Forwards Draft AP600 Ssar App 1B.Response to RAI Re Severe Accident Mitigation Design Alternatives Also Encl

ML20133H273
Person / Time
Site:	05200003
Issue date:	01/10/1997
From:	Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NSD-NRC-97-4937, NUDOCS 9701170116
Download: ML20133H273 (60)

Text

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g Westinghouse Energy Systems Box 355 Pittsburgh Pennsylvama 15230 0355 Electric Corporation NSD-NRC-97-4937 DCP/NRC0703 Docket No.: STN-52-003 January 10,1997 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D. C.,20555 ATTINflON: T. R. QUAY SUBJECf: AP600 DRAFT SSAR APPENDIX IB (SAMDA) AND RESPONSE TO REQUESTS FOR ADDITIONINFORMATION

Dear Mr. Quay:

Enclosure 1 of this letter provides a draft copy of AP600 SSAR Appendix IB. Appendix IB will be included in Revision 11 to the SSAR, which is scheduled for February 28,1997. No changes are expected betwtxn the draft copy enclosed with this letter and the Appendix IB that will be included in SSAR resision 11. If there are any changes, they will Le clearly identified to the staff.

Enclosure 2 provides %estinghouse responses to NRC requests for additional infom1ation pertaining to Severe Accident Mitigation Design Alternatives (SAMDA). Specifically, responses are provided for RAls 100.14 through 100.31. The responses close, from a Westinghouse perspective, the addressed questions. The NRC should review these responses and inform Westinghouse of the status to be designated in the "hTC Status" column of the 0113.

The common theme in both the AP600 SSAR Appendix 1B and the SAMDA RAI responses is that there are no altemate severe accident mitigation design features for AP600 for which the safety benefit outweighs the costs of incorporating the design feature. This conclusion is expected considering that one of the objectives of the Utility Requirements Document (URD) is to address severe accidents. The evolution of the AP600 design has .

I considered severe accidents via the results of the level 1 and level 2 Probabilistic Risk Assessment (PRA).

Consequently, severe accident mitigation design features not included in other plants, as outline in AP600 SSAR \

Appendix 1B, are included in the AP600 plant design. Since AP600 meets the URD requirements, and has l incorporated the PRA into the design process, the conclusion that no additional SAMDA would provide j significant risk benefit should be expected. 1 i (L-

.170003 l mu 9701170116 970110 PDR ADOCK 05200003i A PDR (

I NSD-NRC-97-4937 January 10,1997 IX'P/NRC0703 Per the request of the NRC stafT, the enclosed infonnation is being provided to the NRC on an expedited schedule with the understanding that the NRC consultants who are to review the enclosed material are only available to resiew it in the first quarter of 1997. As agreed to during a telecon with Mr. Dino Scaletti, NRC, this infomiation was te provided by January 10,1997. Westinghouse expects the NRC will review the enclosures in a timely manner, in accordance with NRC consultants near-tenn schedule conunitments.

Please contact Cynthia L llaag on (412) 374-4277 if you have any questions concerning this transmittal.

% sd Brian A. McIntyre, Manag Advanced Plant Safety and Licensing

/jml Enclosures cc: D. Scaletti, NRC WNiosures)  ;

J. Sebrosky, NRC (enclosure 1) l J. Kudrick, NRC (w/o enclosures) '

N. J. Liparuto, Westinghouse (w/o enclosures) l l

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Enclosure I to Westinghouse 12tter NSD NRC-97-4937 January 10,1997 l

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1. Introduction end G:nert! Descriptio2 of Platt l

APPENDIX IB SEVERE ACCIDENT MITIGATION DESIGN ALTERNATIVES I 1B.1 Introduction l His report provides an evaluation of Severe Accident Mitigation Design Alternatives l (SAMDA) for the Westinghouse AP600 design. This evaluation is performed to evaluate l whether or not the safety benefit of the SAMDA outweighs the costs of incorporating the SAMDA in the plant, and is conducted in accordance with applicable regulatory requirements as identified below.

The National Environmental Policy Act (NEPA), Section 102.(C)(iii) requires, in part, that l

...all agencies of the Federal Government shall . (C) include in every recommendation or repon on proposals for legislation and other major Federal actions significantly affecting the quality of the human environment, a detailed statement by the responsible official on . . (iii) alternatives to the proposed action. ,

l 10 CFR 52.47(a)(ii) requires an applicant for design certification to demonstrate  ;

. . compliance with any technically relevant portions of the Three Mile Island requirements set forth in 10 CFR 50.34(f) .

A relevant requirement of 10 CFR 50.34(f) contained in subparagraph (1)(i) requires the performance of

. . a plant / site specific probabilistic risk assessment, the aim of which is to seek such improvements in the reliability of core and containment heat removal systems as are significant and practical and do not impact excessively on the plant . .

In SECY-91-229, the NRC staff recommends that severe accident mitigation design altematives be addressed for certified designs in a single rulemaking process that would address both the 10 CFR 50.34 (f) and NEPA considerations in the 10 CFR Pan 52 design cenification rulemaking. SECY-91-229 further recommends that applicants for design cenification assess SAMDAs and the applicable decision rationale as to why they will or will not benefit the safety of their designs. The Commission approved the staff recommendations ,

in a memorandum dated October 25,1991 (Reference 8). l 1B.2 Summary An evaluation of candidate modifications to the AP600 design was conducted to evaluate the potential for such modifications to provide significant and practical improvements in the radiological risk profile of the AP600 design. l Revision: 11

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1.12trodiction cnd G:ner:I Description of Platt ,

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The process used for identifying and selecting candidate design alternatives included a review

, of SAMDAs evaluated for other plant designs. Several SAMDA designs evaluated previously l for other plants were excluded from the present evaluation because they have already been incorporated or otherwise addressed in the AP600 design. These include-

• Hydrogen ignition system

• Reactor cavity flooding system

• Reactor coolant pump seal cooling

• Reactor coolant system depressurization

• Reactor vessel exterior cooling.

Additional design alternatives were identified based upon the results of the AP600 probabilistic risk assessment (Reference 1). Founeen Fifteen candidate design alternatives were selected for further evaluation.

An evaluation of each of these alternatives was performed using a bounding methodology such that the potential benefit of each allemative is conservatively maximized. As part of this process, is it was assumed that each SAMDA performs beyond expectations and completely eliminates the severe accident sequences that the design alternative addresses. In addition, the capital cost estimates for each attemative were intentionally biased on the low side to  !

maximize the risk reduction benefit. This approach maximizes the potential benefits associated with each attemative.

The results show that despite the significant conservatism employed in the evaluation, none i I

of the SAMDAs evaluated provide risk reductions which are cost beneficial. The results also show that even a conceptual " ideal SAMDA", one which reduces the total plant radiological risk to zero, would not be cost effective. This is due primarily to the already low risk profile of the AP600 design, which is approximately two orders of magnitude below existing plants.

1B.3 Selection of SAMDAS Candidate design altematives were selected based upon design alternatives evaluated for other plant designs (References 2,3, and 4) as well as suggestions from AP600 design personnel.

Additional candidate design alternatives were selected based upon an assessment of the AP600 probabilistic risk assessment results. Fourteen Fifteen design alternatives were finally selected for further evaluation. These4euneen fifteen SAMDAs are:

Chemical volume and centrol system (CVGS) upgraded to mitigate small LOCAs

• Filtered containment vent

• Normal residual heat removal system (RHR RNS) located inside containment

• Self actuating containment isolation valves

• Passive containment spray

• Active high pressure safety injection system

. Steam generator shell side passive heat removal system

. Steam generator safety valve flow directed to in-containment refueling water storage tank (IRWST) l Revision: 11 DRAFT,1997 IB-2 W65tingh0086

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1. IItroduction cnd Gener:I Description of Plant Increase steam generator secondary side pressure capacity

• Secondary containment filtered ventilation IRWST-discharge-valve 4tversification Diverse IRWST injection valves Diverse containment recirculation valves e l'x-vessel core catcher High pressure containment design Diverse actuation system (DAS) improved reliability.

A description of each design alternative evaluated for AP600 is presented in Subsection IB.7.

Several design alternatives addressed in previous SAMDA evaluations for other plants were excluded from further evaluation because the alternatives are already incorporated or otherwise addressed into the AP600 design. These design features include:

a llydrogen ignition system

• Reactor cavity flooding system Reactor coolant pump seal cooling

• Reactor coolant system depressurization

• Reactor vessel exterior cooling.

1B.4 Methodology The severe accident mitigation design alternatives analysis employs a bounding methodology such that the benefit is conservatively maximized and the capital cost is conservatively minimized for each SAMDA. The risk reduction, capital cost estimates, and cost benefit analysis methods are discussed in this subsection. ,

IB.4.1 Risk Reduction Risk for the purpose of this evaluation is the probability of core damage for each accident initiator, multiplied by the consequences of the accident (population dose), expressed in terms of man-rem per year. The total risk is the sum of the risks from all the accidents.

The reduction of risk for each SAMDA is the difference in risk between the AP600 design ,

and an AP600 design with the design alternative incorporated.

It is assumed that each SAMDA works perfectly and completely eliminates the accident sequences that the design alternative addresses. This approach conservatively maximizes the benefits associated with each design alternative, and is not intended to imply that such a perfect design is possible. The SAMDA benefits are the reduction of risk in terms of whole ,

body man-rem per year received by the total population within a 50-mile radius of the AP600 plant site.

Each design alternative is evaluated based on how it affects each of the release categories in the AP600 probabilistic risk assessment.

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1. Iltroduction cnd General Description of Plant La IB.4.2 Capital Cost Estimates The capital cost estimates for each SAMDA are intentionally biased on the low side to maximize the risk reduction benefit. All reasonably anticipated one-time capital costs are accounted for in the estimates. Actual plant costs are expected to be higher since the cost estimates do not include the cost of testing and maintenance or the engineering cost to design the alternative to fit into the AP600. 'Ihe cost estimates are based on 19921996 U.S. dollars.

IB.4.3 Cost Benefit Analysis 1

In order to compare the risk reduction, which is reponed in man-rem per year, to the capital costs, which are reported in dollars, a common set of units must be established. For this evaluation, the risk reduction is convened to a capital benefit which can then be directly compand with the capital costs.

The benefit of each design alternative is the reduction of risk in terms of whole body man-rem per year received by the total population within a 50-mile radius of the AP600 plant site.

Consistent with previous S AMDA evaluations and NRC regulatory analysis guidelines, a value of $1,000 per offsite man-rem avened is used to convert man-rem per year to dollars per year.

This value is intended to be the surrogate for all offsite consequences including property damage and is referred to as the annual levelized benefit.

The risk reduction reponed as dollars per year is then converted to a maximum capital benefit which can then be compared to the capital costs. The maximum capital benefit is equal to the annual levelized benefit (dollars per year) divided by the annual levelized fixed charge i rate.

The annual levelized fixed charge rate is determined from a number of financial factors. l These factors are given in Table IB.4-1 and are taken from the EPRI Technical Assessment Guide (Reference 6). The equttions used to determine the annual levelized fixed charge rate are from the Nuclear Energy Cost Data Base (Reference 7). For a nuclear plant economic life of 60 30 years and a tax life of 15 years, the annual levelized fixed charge rate is 4-SA 15.7  ;

percent in current U.S. dollars (with inflation). '

1B.5 PRA Release Categories To assess each design attemative's reduction of risk, the potential for each attemative to reduce the frequency of occurrence or the consequence of each mlease category is assessed.

The steps involved in creating the AP600 release categories are briefly discussed in this subsection.

The AP600 Level 1 plant event trees identify the sequences that lead to core damage.

Sequences that have similar characteristics are grouped together into accident subclasses for the containment system analysis. The characteristics considered in the binning of the plant event sequences into the accident classes are as follows:

1 id Revision: 11 DRAFT,1997 1 B.4 Westinghouse i

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1. Introduction ccd G:neral Description of PI:nt The initiating event type -- such as loss of coolant accident (LOCA) or anticipated transient without scram (ATWS), leading to core damage The primary system pressure at the time of initial core damage uncovery (high or low)

Timing of core damage (early or late)

Containment integrity at the time of core damage (intact, not isolated or bypassed or impaired)

• ----Availability-of safety-systems-after-core damage Disposition of water in the containment at the time of core damage Gontainment-pressure-and-temperature-et-the-time 4<ere d:=cge.

Containment event trees for each of the significant accident subclasses are developed and i discussed in the AP600 probabilistic risk assessment (Reference 1). Consideration of severe accident phenomena that may challenge containment integrity forms the basis for the nodes on the containment event tree. Operator actions or system top events are generally considered with respect to preventing or mitigating severe phenomena. 'Ihe containment event tree considers that the following phenomena represent the severe accident issues relevant to the l AP600 containment integrity:The-containment event-tree-analysis-considers--both-the l containment-and-associated-auxiliary-systems--In-particularr -the-fc!!cv.ing items--are l consideredt In-vessel fuel-coolant interactions

• In-vessel hydrogen generation Creep rupture failure of steam generator tubes High-pressure melt ejection Melt attack on the containment pressure boundary

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Containment overpressurization from decay heat Reactor vessel integrity i Ex-vessel fuel-coolar.t interactions

• l Core-concrete interaction'and hydrogen generation I

• Hydrogen deflagration and detonation Elevated temperatures of the containment shell (diffusion flame heating)

Elevated gas temperatures (equipment survivability)

Containment-isolation-system

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Passive-centainment-occling system In-containment-refueling water-storage tank injec4 ion Ex veze! debriwooling ThMunctions accomplished-by-these systemwret Mainte=nce of centainment4ntegrity-and/or-the-reduc 4 ion-of-containment-pressure

--Preventica of vezel-failure and/cr core melt a rest Gooling of ex veze! debris:

The end-state of each path on the containment event tree describes the effectiveness of the containment to mitigate offsite doses for that accident sequence. The radiological consequences of the core-melt accident are largely determined by three major considerations:

Revision: 11 3 W85tiligt10USS IB-5 DRAFT,1997

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1. I troduction cnd General Description of Plant '

De mode of the postulated containment failure (bypass, isolation failure, gross failure, or intact containment)

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ne time of postulated containment failure relative to the time of major fission-product release from the core or cote debris

. Fission-product removal mechanisms in the containment AP600 PRA does not credit active containment fission-product removal mechanisms such'as i containment sprays or fan coolers.c Therefore, natural deposition processes, gravitational settling, thermophoresis,:andidiffusiophoresis,are relied on to'semb aerosolsLfrom'.'the containment atmospherei ne natural processes are time-dependent, thus the~ mode ~of containment failure, timing of the containment failure, and magnitude of the offsite release are directly related and treated together for the AP600 containment event tree' via~ielease categories. The source term for each release category is calculated with the Modular Accident Analysis Program Version .4.0 (MAAP 4.0) code. The-endpoints-of-the<+ntainment+ vent trees-paths-are-grouped-into-oppropriate-source-term-cetegories-bred on similar-fission produc44eleases-Different-endpointsfor-the-AP600-plant-aredermedrdependingen4he4ype of-centaimnent4ailure-(bypassr-iselation-failure;-or-late-overpressure dce to cere-cenerete interac4 ion}--If-the-containment-does-not-failr4he-availability-ef-the-passive-containment cooling-systemwater-hase+trong-influence +n4he<+ntaimnenttressurerend4herefore is und to-determine-the4elease-eategoryr-The-source-term 4er-a-representative-sequence-in-each important seeident-elaswrealculated-with4he-Modular-Aceident-Analysis-Pfogram-Version 44(MAAP-4.0)-cede:

The release categories for the AP600 are defined as follows:

• IC - intact containment; CFE -- containment failure early, occurring in the time frame between the onset of core damage and the end of core relocation; CFI - containment failure intermediate, occurring in the time frame between the end of core relocation and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage; CFL - containment failure late, occurring later than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the onset of core damage; CI - containment isolation failure, with the failure occurring before the onset of core damage; BP - containment bypass, with the bypass occurring before the onset of core damage.

OE release-assoeiated-with4he-4eakage4 rom-a-sontaintnen: ' ith paz.ve contaimnent eooling-water-available; OEP relese azeeiated-with4he-leakage-frome<entainment+ithpassivecentainment eooling-water not available; CC release-associated-with-the-leakage 4 rom-e-containment-that4s-pressurized-with noneondensible-gasegenerated-by-c+re<enerete-interaetion; C! =!c=e anceinted ; ith4he-leakage 4 rom-a-eentainment that is-byp==d cr-has-not been :=! ted-(impaired).The4ello"/ing r,ubsec4 ions-prean: a brief-description-of-the acciden: =quences4 rem :he probabilistic risk === men * "-hich+epre=nts each AP600 relea= eetegory-Revision: 11 DRAFf,1997 13-6 WestlDgh0US8

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1. I:troduction cnd G:neral Description of Plant l

The following subsections present a brief description of the accident aquences-form-4he probabilistio-rislaissessment-whic4Hepresents-each-AP600 release categories category.

1B.5.1 Release Category IC - Intact Containment If the coritaisiment integrity is maintained throughout the accident, then the telease of radiation  !

from the containment is due to nominal leakage and is expected to be within the design basis of the containment. This is the "no failure" containment failure mode and is termed intact l containment. The main location for fission-product leakage from the containment is l penetration leakage into the auxiliary building where significant deposition of aerosol fission products may occur.

The final release fractions, at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage, are presented in Table 1B.5-1. He IC release category frequency is 1.5 x 104 per year.

IB.S.2 Release Category CFE - Early Containment Failure Early containment failure is defined as failure that occurs in the time frame between the onset of core damage and the end of core relocation. During the core melt and relocation process, j several dynamic phenomena can be postulated to result in rapid pressurization of the l containment to the point of failure. The combustion of hydrogen generated in-vessel, steam  ;

explosions, and reactor vessel failure from high pressure are major phenomena postulated to i have the potential to fail the containment. If the containment fails during or soon after the time when the fuel is overheating and starting to melt, the potential for attenuation of the fission-product release diminishes because of short fission-product residence time in the containment. He fission products rel:ased to the containment prior to the containment failure i

are discharged at high pressure to the environment as the containment blows down.

l Subsequent release of fission products can then pass directly to the environment. Containment l

failures postulated within the time of core relocation are binned into release category CFE.

The final release fractions, at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage, are presented in Table IB.5-1. De CFE release category frequency is 6.6 x 104 per year.

IB.5.3 Release Category CFI- Intermediate Containment Failure Intermediate containment failure is defined as failure that occurs in the time frame between the end of core relocation and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage. After the end of the in-vessel I fission-product release, the airbome aerosol fission products in the containment have several hours for deposition to attenuate the source term. The global combustion of hydrogen generated in-vessel from a random ignition prior to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> can be postulated to fail the containment. The fission products in the' containment atmosphere are discharged at high pressure to the environment as the containment blows down. Containment failures postulated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the onset of core damage are binned into release category CFI.

Revision: 11 3 Westinghouse 13 7 DRAFT,1997

1. Irtroductio2 cnd General Description of Plart ,

The final release fractions, at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage, are presented in Table 1B.5-1.l'Ihe CFI release category frequency is 1.3 x 10'" per year.

1B.5.4 Release Category CFL - Late Containment Failure

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Late containment' failure is'defiried as containment failure' postulated to occur laterithan 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the onset of core damage. Since the PRA assumes the; dynamic phenomena, such as hydrogen combustion, to occur before 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, this failure mode occurs'only,from the lossfoi containment heat re'moval via failure of the passive containment cooling system:

De fission products that are airborne at the time of containment failure will be discharged

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at high pressure to the environment, as the containment blows down.' Subsequent release of fission products can then pass directly to the environment. Accident sequences with failure of containment heat removal are binned in release category CFL.

He final release fractions, at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />'after core damage, are presented in Table 1B.5-1. The CFL release category frequency is 1.5 x 10~" per year.

1B.5.5 Release Category CI Containment Isolation Failure A containment isolation failure Nears because of the postulated failure of the ~ system or valves that close the penetrations between the containment and the environment. Containment isolation failure occurs before the onset of core damage. For such a failure, fission-product releases from the reactor coolant system can leak directly from the containment to the environment with diminished potential for attenuation. Most isolation failures occur at a penetration that connects the containment with the auxiliary building. He auxiliary building may provide additional attenuation of aerosol fission-product releases. However, this decontamination is not credited in the containment isolation failure cases. Accident sequences in which the containment does not isolate prior to core damage are binned into release category CI.

The final release fractions,'at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage, are presented in Table'IB.5-1. The CI release category frequency is 3.6 x 104 per year.

1B.5.6 Release Category BP - Containment Bypass Accident sequences in which fission products are released directly from the reactor coolant system to the environment via the secondary system or other interfacing system bypass the containment. The containment failure occurs before the onset of core damage and is a result of the initiating event or adye' rse conditions occurring at core uncovery. The fission-product release to the environment begins approximately at the onset of fuel damage, and there is no attenuation of the magnitude of the source term from natural deposition processes beyond that which occurs in the reactor coolant system, in the secondary system, or in' the interfacing system. Accident sequences that bypass the containment are binned into release category BP.

The final release fractions, at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage, are presented in Table 1B.5-1. The BP release category frequency is 1.1 x 104 per year.

Revision: 11 DRAFT,1997 iB-8 W85tingt100S8

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1. Introduction rnd G:nerel Description of Phnt 4.o. ._,. i. o.._ _.

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The4epresentative4equencefer4heOK-release-eategory4taran4nitiating even:-whieh4s-a4 inch-diameter !cu cf-coolant-accident-with-a4ailure-ofahe-in-cen*A=n: refue!!ng-water storage 4ank+ heck-valves andeormaLRHR-injec4 ion-Gore <lamage begin: 2.Jhoure4nto the accident-The4n-centainment-refueling-water +torage4ank-iscot<lrained4nto4he+ontainment cavity

• provide-enternalceoling-to4he-reaetor-vessel-so4he-core-debri: is not maintained in4he veze!. The veze! fails-at444 hours andahe-molten-coredrains-into4he+ontainment r

at4ow-pressure--Dealebris is quenched and cooled 4n-the-reactor-savity, ;c dere is ne significant+x-veze! release- -The-passive-containment-cooling-systentend41ydrogen-igniters are-availablerend-containment-pressure 4emains-below-design-prezure.

The4inal-releaseirac4 ions, a: 2-1 hours-efter-core-damager areyesented4n-Table 4M4Ahe OK--welease-category-frequency 4s-24-x40iper-year, gg. o.._i __ _ c,.t ,.-, ^~ "."

The4epresentative-sequence 4or4elease-eategory OKF49-initiated-by-a4mch<liameter4oss of-coolant-accident-with-failures-of-the-in-containment-refueling-water-storage-4ank-check valvesrnormaLRHR-injection rand-passive-contaimnent-cooling-system 4eoling-water-Four out-of4our-core <nakeup-tanks-and-accumulator &ere-available --The4n-centainment4efueling water-storage 4ank4s-not-<irained-into-the-containment-cavhy-to-provide-extemal+ooling-to the4eac4er-+ esse!, sc Se cere-debris 4s-not-maintained 4n4he vezel Core da=ge accurs-at 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-and-vessel 4ailure+ceurs a: 152 houn. The41ebris i; quenche44nd4eolable4n4he reactor-cevity-because-ell-cf de water-holdup-volumes-arefull-and4he-condensationfron>4he passive-centainment-cooling-system-shell-returns-water-4o-the-containment-+ ump-The centainment prezure-is-elevated-over-lhe !cng termrbut-et-equilibrates-at-a-pressurc we!I below4he-ultimate-capacity-of4he-shell,-eo-containment-integrity 4s-maintained. Nceredit4s taken-in-4he-analysw-for-acc-ident-management-er-use-of-alternative methods-of-wetting 4he centainment shelb Because-of-the-influence-of-water-in-the-containmentrthere is ezentially-no-difference-in fission-produet4elea= if de debris 4emains-in4he veze! cr is4elea,ed 10 de centainment-The-Gnal relea:e fractions, a: 21 hour2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />sefter-core 4a.._r, _. yresented in Table !B.5 ! Se OKP-release-eategory-frequency i: 5.6 x 10%er yer, Lo..e.2 o. . ._ . __ c._... . _, , _.,.

. , , . , c,.~c The4epresentative sequence for-release-cetegory-GG-is4nitiated by a ' inch dia=::: !cz of eoolant-eceedent-with-a-failur; c: Me in-c+ntainment-refueling-water-storage-tank-check valvesroormaLRHR4njec4 ion-end4he-passive-contamment-cco!!ng systernwater41ew-Three out-of-the-four-core-makeup-tank; and ac-eumulators-are-available--E.e in containment refueling wa::: . c=ge :ank-is-not-<! rained inic the contamment-eavity :c provide external ecoling :c Se reac:cr veze!, ;c 6e ecre debris-is-not-emintained in Se veze!. He core damage begin: at 2.0 houn. He veze! fails c: !!.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, and Se me!:en-core <irains-into the-eavity a: !cw prezu=. He cevity-drie cut becauz 6e water from de availab!c cere Revision: 11

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1. I:troduction cnd G:neral Description of Pla t ,,

makeup-tank: =d-aceumulators-is-4 rapped-as-steam er in vater he! dup vc!un=. Passive containment <eoling ystem conde=ation4oewo: keep up v.itlHhe-rate +f boilof0from4he debris-ted-Core-cenerete-interaction-create noncendensible-gases-that prezurize-4he l centainmen At 24-hours +fter-cere4amagerthe-prezure in the-containment i: essentially '

eq=! :c desiga-pressure-he-final <elease-fractionsr t-24 a hours after-core-damager-are presented-in-Table !B.5 !. He CC re! =e category-frequency is 74-*-10i*-per-yeae 13.5.4 Ric- Catege:y CI l l

The representative sequence-for4elease<etegory C! is initiated-by-e40ss-of-feedwater-to-the i steam-generator: =d --the-4ailure-of-the paz.ve residual--heat-remova! =d au:cmatie depressurizationeystemsr-Thecontairtmentdoe.rnotisolate--Thecontainment4solatiordailure is-modeled-as-the-failure-of-ene-purge 4ine---Re-core-temperature-exceeds 4500 K a: 1.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s-The-operator-dumps 4n-centainment4efueling-water 4torageaank-water 4nto4he+avity on-a-high-high-core-exit-temperaturer-The-water-surrounds-and-ceois-the-reactor- vessel-preventing-vessel-failure. The reac4or-coolant-system he: leg ruptures-due-to-4he-high l temperature-and-pressure 4n4he-reac4er-coolant-system---The-remainder-ofahemere-meltsend '

falls 4nte-the4ower-head.-- Fizion producas-released-into-the-centainment := be direc41y transported 4o4he-environmentr-Thefmal-release 4ractionsr et-244toursefter-c+redamagerare i presented-in-Table 4B.5 !. The C4-release-eategory-base-frequ=cy is 2.0 x-104 per yearr l lioweverr-because the frequency of excessive--leakager -witic4t-exceeds-tle-technical '

specification 4eakager fronHhe+ther4eleasecategoriesis4 umped +ntethe G44elease-category, the-overall4elease-eategory-frequency is taken-tc be 3.0 : !^m* earr l IB.6 Total Population Dose l

To assess the potential benefits associated wit h a design alternative, estimates are made of the l

total offsite population dose resulting from each of the release categories (i.e., source terms) l identified in Subsection 1B.5. The MELCOR Accident Consequence Code System (MACCS),  !

Version 1.5.11.1 M (Reference 5) is utilized for this analysis. The code input is identical to )

the AP600 probabilistic risk assessment, however the consequence evaluated is the effective whole body equivalent dose (50 year committed), resulting from exposure during the initial  !

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the onset of core damage, to the total population within a 50-mile radius I of the plant.

Table IB.6-1 presents the estimated mean 50-mile radius population whole body dose in person-rem (man-rem)randmediandoses-inferson sieverts44-person sievert +quals-1004 nan-rems) for each release categoryr-Teble-1B444 hows and the 50-mile population dose risk (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) for each release category, as well. The as4he total risk is of 7.3 x 10-3 3.42 x 10-8

-man-rem per year for the AP600 plant.

IB.7 SAMDA Description and Benefit This subsection describes each SAMDA and the benefit expected due to the modification. In the evaluation of the risk reduction benefit, each SAMDA is assumed to operate perfectly with 100 percent efficiency, without failure of supporting systems. A perfect SAMDA reduces the Revision: 11 DRAFT,1997 1B-10 Westingt10use

y  ;. mm_,.

1. I troduction and Gsneral Description of Plan

• frequency of accident sequences which it addresses to zero. His is conservative as it maximizes the benefit of each design attemative. The SAMDA will reduce the risk by lowering the frequency, attenuating the release, or both. The benefit will be described in terms of the accident sequences and dose which are affected by the SAMDAs, as well as the

~

overall risk reduction. N6te thatlfor'thipsrposes of these~ evaluations lin reases to'releiss categoiy IC are not factored into theirisk benefit calculations.ine~IC dosetislsufficiently small (3E+2) that changes to the IC total frequency do not result in an appreciable ~chinge;to i overall,results. This,is;also~alconservative representation since,thisjnaximizes3heirisk l reduction.

1B.7.1 Upgrade the CVGS for Small LOCAs The chemical, volume, and control system (CVGS) is currently capable of maintaining the I reactor coolant system (RCS) inventory to a level in which the core remains covered in the l

event of a very small (.<_3/8" -3/47 diameter break) loss of coolant accident (LOCAs). His SAMDAfinvolves' providing7in cantainment refueling water storags;tankT(IRWST)i/

containment recirculation l connections td the CVS and adding a.secondllide~frors thsCVS makeup pumps to'the RCS in" order to.be able to use the system to_keepithicore, covered during small and intermediate LOCAs, This-SAMDA-involverr-upgradig 60 pumping ,

I capacityrand-line-sizes-of-the<VCS system-inorderde4+able4e =e 6e system 404eep4he cere-covered 4uring-smallG4Z-diameterheaks) LOCA accidents, as =lb i

A perfect, upgraded CVGS system is assumed to prevent core damage in the RCS leak, passive RHR heat exchanger tube ruptures, small LOCA, and intermediate LOCA-andmedium i LOGA of all-the =y small-and-small-LOGAs-in each release category. The CVGS is l assumed to have perfect support systems (power supply, component cooling) and to work in all situations regardless of the common cause failures of other systems. This results in a total 4

averted risk of 5.5 x 10d5.80 x 10 -man-rem per year.

1B.7.2 Filtered Vent This SAMDA consists of placing a filtered containment vent and all associated piping and )

penetrations into the AP600 containment design. He filtered vent"could'be lis5dWWnt the  ;

ebntainment to prevent catastrophic ~ overpressure failure, and also provides filterinicapability I for' source term release. With respect to the AP600 PRA,'the possible scenario in"which the l filtered vent could result in risk ~ reduction would be late contalmnent overpressse failures I (release category CFL). ' Other conta'mment overpressure failures occur due to dynamic. severe accident phenomena, such as hydrogen burn, steam explosion, etc. The late' containment failures for AP600 are failures of the passive containment cooling systens(PCS)." ' Analyses have indicated that for scenarios with PCS failure, air cooling may limit the ~ containment pressure to less than the ultimate' pressure. However, for the purposes of the Level 2 PRA,

~

failure of PCS is assumed to result in containment failure based on an' adiabatic heatup! To conservatively consider the risk reduction of a filtered vent, the use of,a filtered vent to preclude a late containment failure will be evaluated. A decontamination factor (DF) of 1000 will conservatively be assumed for each PRA Level 1 accident classification,~even~though it is realized that the dose due to noble gases will not be impacted by the filtered vent since Revision: 11 T Westinghouse iB-11 DRAFT,1997 i

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1. hiroduh and General Descrl=^ =_ of Plant -

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equal t6 the; decontamination factor assumedfsince the PRA Leiel;11accident 'classifWatiori

frequencies do not change.ine; total averted risk for a filtered vontjijhus 1000tmani

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4 IB.7.3 Locate Normal RHR Residual" Heat Rimi6ial Inside Containment -

4 4

i Ris SAMDA consists of placing the entire normal residual heat removal (RHR) system .

(RNS) and piping inside the containment pressure toundary. Locating the RNS = ..' RHR i

inside the containment would prevent containment bypass due to interfacing system LOCAs (ISLOCA) of the residual heat rem 6 val (RHR) system. In past probabilistic risk assessments

] of current generation nuclear power plants, the ISLOCA is ti>e leading contributor of plant risk l

because oflarge offsite consequences. A failure of the valves which isolate the low pressure l

RHR system from the high pressure RCS causes the RHR system to overpressurize and fail, releasing RCS coolant outside the containment where it cannot be recovered for recirculation I cooling of the core. De result is core damage and the direct release of fission products l outside the containment. I e

In the AP600, the RNS RHR gn: . is designed with a higher design pressure than the RHR systems in current pressurized water reactors, and an additional isolation valve is provided in l the design. In the probabilistic risk assessment, no ISLOCAs contribute significantly to the '

core damage frequency of the AP600 (Refe-rence 1, Chapter 33 Tt!: 7 !). Derefore, Revision: 11 l DRAFT,1997 1B.12

4

$ '[

• yy 4

1. I troductica tad G neral Descriptioi of Plant f"-9 i 4 Ie 4

e i relocating the RNS'nc==! RHR :y::= of the AP600 inside containment will provide j virtually no risk reduction benefit and will not be investigated further in terms of cost.

i 1B.7.4 Self Actuating Containment Isolation Valves i

i This SAMDA consists of improved containment isolation provisions on all normally open containment penetrations. De category of "normally open" is limited to normally open pathways to the environment during power and shutdown conditions, excluding closed systems inside and outside the containment such as RNS normal-RHR and component cooling. The design alternative would be to add a self-actuating valve or enhance the existing inside containment isolation valve to provide for self actuation in the event that containment conditions are indicative of a severe accident. To evaluate the benefit of this SAMDA, this design change is assu;med;toieliminate the CI release; category. de f=qu= y of !!

centamment-isolation fai!=: = =b:=cted frc= 6: CI =!== =e;;crj =d = cdded to+he OK =!:=: =::i;; y =d 6: risk 4weqe="5ed. His does not include induced containment failures which occur at the time of the accident such as in cases of vessel rupture or anticipated transients without scram (ATWS). De benefit results in an averted risk of 7.4 x 10 4

M3-x-40 4man-rem per year.

1B.7.5 Passive Containment Sprays This SAMDA involves adding a passive safety-related grade-spray system and all associated piping and support systems to the AP600 containment. A passive containment spray system could result in risk. benefits in the following ways:

scrubbing of fission products, primarily for CI failures, assuming appropriate tiining, containment spray could be used as an alternate'means for flooding the reacter vcssci (ingessel rctcation) and for debris quenching simuld vessel failure occur, containment spray could also beused to control containment pressure for~ cases in which PCS has failed.

In order to envelop these potential risk benefits, the risk reduction evaluation ~will assume that containment sprays are perfectly effective for each of these benefits, with'the exception 'of fission product scrubbing: for- containment ' bypass. Dus the. risk; reduction can be conservatively'. estimated by assuming' all release categories .except:BP are, eliminated.

~

Herefore, passive cor'ainment sprayfresults in a total averted risk of 6.9 x 10'8 man-rem per year. , A ped::: :=tir==: :p=y wi 5 pede:: =ppc : :yst= is :=:=d c p=vid: 5=i=

p=de:: ::=bbin; :-d =!:=: redue:i= ' 6: ::=: cf a f:1!:= cf :=tir= : i=!::!=

Fur $::, sp=ys === ::::::c :=; cf =y ec= &bi i,6 :=::!r= :, p=v=:in;;;c=

eonerete-mieracti= Tc :v:!=:: de 5:=5: cf-c=tir= : :p=y:,6: OK =d CC =!==

::;;c y '=q:=:le: = d&d :: Se f=q==y of 6: OK =!== ::e;crj, =d a dc

d ::i= cf 1T i ===d :: be :pp'ird :c de CI =!== =::;crj. '"!: ==!= i, : :ct!

ve.ed -id cf 3.39 10i .=. == p:: y=.

Revision: 11 1B-13 DRAFT,1997

-m 1

1. Introduction cnd G:neral Description of Plant -

~

Additionally, a nonsafety-related containment spray system is evaluated. 'Ihe nonsafety; related containment spray system utilizes the fire protection pumps as a ~ motive force for spray,less remotely operated valves but approximately the same amount of piping. It will be conservatively assumed that the risk benefits are the same,'and that the risk nduction remains the same for the nonsafety-related containment spray system. Thus, the nonsafety!related i

containment spray system results in a total averted risk of 6.9 x 10 man-rem per year.

1B.7.6 Active Illgh Pressure Safety Injection System This SAMDA consists of adding a safety-related grade active high pressure safety injection (liPSI) pump and all associated piping and support systems to the AP600 design. A perfect i high pressure safety injection system is assumed to prevent core melt for all events but transients =d-small-medium-and-large-LOGAs-in-each-release-eategory-Only-excessive LOCA and ADVS-ere-essumed-to-lea 440-eere-damage. Therefore, to estimate the risk reduction, only the contributions to each' release category of Level 1 accident classes'3C (vessel rupture) and 3A (ATWS) need be considered. The ~ averted risk is 6.1 x 10-8 man-rem per year, the-frequency of :=h re!== =tegory-+s-reduced-by-the-freque=ie: cf allahe LOGAs-andaransients4equ=ce: in-the<ategoriesand4he-risk-is-requantified---The-averted risk-is4-86-x40i-man-rem per yer. This SAMDA would completely change the design approach from a plant with passive safety systems to a plant with passive plus active safety-related systems and is not consistent with design objectives.

IB.7.7 Steam Generator Shell-Side IIcat Removal System This SAMDA consists of providing a passive safety-related grade heat removal system to the secondary side of the steam generators. The system would provide closed loop cooling of the secondary using natural circulation and stored water cooling, thus preventing a loss of primary heat sink in the event of a loss of startup feedwater and passive RHR heat exchanger. A perfect secondary heat removal system would eliminate transients from each of the release categories. In order to evaluate the benefit of this SAMDA, the frequencies of all the transient sequences is subtracted from the overall frequency of each of the release categories d

and the risk is requantified recalculated. The total risk averted is 5.3 x 10 6.7 10d man-rem per year.

IB.7.8 Direct Steam Generator Relief Flow to the IRWST This SAMDA consists of providing all the piping and valves required for redirecting the flow from the steam generator safety and relief valves to the in-containment refueling water storage tank (IRWST). An alternate, lower cost option of this SAMDA consists of redirecting only the first stage safety valve to the IRWST. This system would prevent or reduce fission product release from bypassing the containment in the event of a steam generator tube rupture (SGTR) event. In order to evaluate the benefit from this SAMDA (both options), this design change is assumed to eliminate the BP release category. Se freq==4es-of-el! 6e SGTR

qu=ce: =: =btracted frc:n Se CI re!== =tegory freq==y =d added to Se OK re!==

4 eategory-freque=y, =d de rid inequantifie& The total risk averted is 4.2 x 10 6.7 x 10d man-rem per year.

Revision: 11 DRAFT,1997 IB-14 W85thgh00S8

1. I: trod:ction cnd General Description of Plant 1B.7.9 Increased Steam Generator Pressure Capability This SAMDA consists of increasing the design pressure of the steam generator secondary side and safety valve set point to the degree that a steam generator tube rupture will not cause the secondary system safety valve to open. The design pressure would have to be increased sufficiently such that the combined heat capacity of the secondary system inventory and the PRHR system could reduce the RCS temperature below To for the secondary design pressure.

Although specific analysis would have to be performed, it is estimated that the design pressure would have to be increased several hundred psi. Like the system described in Subsection IB.7.8, this design would also prevent the release of fission products which bypasses the containment via the SGTR. Therefore, the risk reduction is also the same as that quantified in Subsection IB.7.8. The total risk averted is 4.2 x 104 6-7-*-10d man-rem per year.

1 B.7.10 Secondary Containment Filtered Ventilation This SAMDA consists of providing the middle and lower annulus (below the 135' 3" elevation) of the secondary concrete containment with a passive annulus filter system to for filtration of elevated releases. The passive filter system is operated by drawing a partial vacuum on the middle annulus through charcoal and HEPA filters. He partial vacuum is drawn by means of an eductor with motive flow from compressed gas tanks. The secondary containment would then reduce particulate fission product release from any failed contaimnent penetrations (containment isolation failure). the-pathway *4 rom-whic4Hhe-majority-of-the primary-containment 4eakage-ispredicted4e+ccw, In order to evaluate the benefit from such a system, this design change is assumed to eliminate the CI release category. the+ffsite-doses from4he<+ntainment4eakage+elease-eategories r OKrOKPendCGrand4heexcessiveleakege frequency-contribution-to-the CI rele=e category-are-assumed 4o-be-zero r-and-4he-risk-is 4

requantiGed: The total risk averted is 7.4 x 10 4.14 x 10 4man-rem per year.

1B.7.11 Diverse IRWST Injection Valves Di=!fy Se !RWST D!=h=ge W!v=

This SAMDA consists of changing the in-containment refueling water storage tank (IRWST) injection valve designs so that two of the four lines use diverse valves. Each of the four lines is currently isolated by a squib valve ~ in series with a check valve. In order to provide diversity, the valves in two of the lines will be provided by a different vendor. This SAMDA consists-of-redesigning-the-in-containment-refueling-water-storage-tankdIR"/ST) diseharge valve-conGguration4 rom-four-check valves 4o4wo-eheck-valves-and4wsur-operated-valves, This change will reduce the frequency of core melt by eliminating the common cause failure of the IRWST injection. To estimate the benefit from this SAMDA, all core damage  ;

sequences resulting from a failure of IRWST injection are assumed to be averted. Core damage sequences resulting from a failure of IRWST injection correspond to PRA Level 1 ,

accident classification 3BE; thus, release category 3BE is eliminated. tie 4requencies-of-all '

the re!ere categories is reduced by 4he-centribution-of-IRWST-injec4 ion 4ailure-sequences, 4

and4he-risk is requantified-De total risk averted is 5.3 x 10 84h-10ir,ian-rem per year.

Revision: 11 W85thgh0USS DRAFT,1997 IB 15

1. I troductio2 and G:ncral Description of Plant ,,

i' 1B.7.12 Diverse Containment Recirculation Valves This SAMDA consists of changing' the containment recirculation valve design ~s so that two out of the four lines use diverse valves. Each of the four lines currently contains a~ squib valve; two of the lines contain check valves and the other two contain motor-operated valves.

In order to provide diversity, the squib valves in two lines will be.made diverse by supplying i

them from a ,different vendor._ This chenge will reduce the frequency of core melt by '

eliminating 1the common cause failure of the containment recin:ulationETo estimate 1the benefit from this SAMDAs' all ~ core damage sequences resultirig froni a failure of cbntainment

~

reckulation are assumed to be averted. Core damage sequences resulting from failure of containment recirculation correspond to PRA Level 1 accident classification 3BL; thus, release cateFory 3BL is eliminated. The total risk averted is 1.5 x'104 man-rem per year.

1B.7.1343 Ex Vessel Core Catcher This SAMDA consists of designing a structure in the containment cavity or using a special concrete or coating which will inhibit core-concrete interaction (CCI), even if the debris bed -

dries out. A perfect core catcher would prevent CCI for all cases. 'However, the AP600 incorporates a wet cavity design in which ex-vessel cooling is used to maintain the core debris

)

in the vessel thus preventing ex-vessel phenomena, such as CCI. Consequently, containment l

failure due to CCI is not considered in detail for the AP600 Level 2 PRA. For cases in which i reactor vessel flooding is failed, it is assumed that containment failure occurs due to ex vessel steam explosion or CCI. This containment failure is assumed to be an early containment failure, CFE, (due to ex-vessel steam explosion) even though CCI and basemat meltthrough would be a late containment failure. To conservatively estimate the risk reduction of an ex.

vessel core catcher, this design change is assumed to eliminate the CFE release category. He total risk averted is 6.1 x 10 man-rem per year. Aferfect+ornatcher-desiga-wou!d prev =t GGamtirelyrand-the4enefit4 rom 4he4 ore-catcher-would-be-estimated by assuming det a!!

cf de =qu=ee: in-the CC relea= category ><cu!d all-result-in OK relea=:. Herefere, Se frequeney-of-the CC re!=:e category i: edded to de OK relee= frequeney =d de rid i:

'; requantified. ":i: SAMDA results in "irtually ne reduction 4n-risk 4ince $e rid fic:n Se CI release <ategoryr w hieh-dominate: de p!=t-rid is net reduced in =y evey by Me = ve:2!

eore-eatcher. Herefere rthirrSAMDA is not eensidered4arther2 18.7.1443 Illgh Pressure Containment Design This SAMDA design consists of using the massive high pressure containment design in which the design pressure of the containment is approximately 300 psi (20 bar) for the AP600 containment. The massive containment design has a passive containment cooling feature much like the AP600 containment. He high design pressure is considered only for prevention of containment failures due to severe accident phenomena such as steam explosions and hydrogen detonation. A perfect high pressure containment design would reduce the probability of containment failures, but would have no reduction of the frequency or magnitude of the release from an unisolated containment (containment isolation failure or

! containment bypass). To estimate the risk reduction of a high pressure containment design, this design is assumed to eliminate the CFE, CFI and CFL release categories. The total risk Revision: 11 DRAFT,1997 IB-16 W85tiflgh0US8

l l

1. Introduction and General Description of Pisct avehed is 6.Ti10tmari;reiiipeiysir2 he-AP600probabilistio-risk aze==n: eencluded that :he AP600 i: net suzeptibisoeontainment4ailure-due te = vere acciden: phenomemw Sincelhe AP600 probabilistie-risk azenrnen: predicts-no-overpressure-containmen: failures, the4tigh-pressure <entainment<lesign, at bes:, provides a risk-reduc 4+on+f-virtually-eerer end therefore-will-not4e eensidered4urtherv IB.7.1544 Increase Reliability of Diverse Actuation System His SAMDA design censists of improving the reliability of the diverse actuation system (DAS) which actuates engineered safety features and allows the operator to monitor the plant status. A perfectly reliable DAS system would reduce the frequency of the release categories by the cumulative frequencies of all sequences in which DAS failure leads to core damage.

In order to evaluate the benefit from the DAS system upgrade, a Level 17 sensitivity ainalysis assumin~g~ perfect reliability of;DAS;was" completed.s the4requency of-the-DAs-failure-are subtrac4edfrom4herelease<4tegory4requencies and4herisk is requantined. psisigthsresults of the~ Level 1 DAS sensitivity; the total risk averted is determined to be 21)i 10-* -7718-*4&-

• man-rem per year.

IB.8 Results Auliseussed in Subsec4 ion-1B.7, fcur design-altematives-considered for the AP604 provide no-benefit 4er-reducing-residual offsite-risk-These a!:ernatives-are+

a----Filtered-vent Locate-theormal-residual heat remaval-system-insidtsentainment )

Ex vezel-core-eatcher

--High-pressure-containment desi;;n ne remainmg-design alternatives from.Section IB.7 are evaluated to determine their cost benefit. The results of the remaining severe accident mitigation design alternatives evaluation are summarized in Table IB.8-1. The first column identifies the design alternatise for which a reduction in risk was calculated. The second column is the total man-rem reduction per year for the design alternative. The third column is the capital benefit calculated based on the reduction in risk. This value represents the maximum amount of capital that could be spent in order for the design alternative to be cost beneficial. The next column is the estimated minimum capital costs for the attemative. The final column represents the net capital benefit.

The net benefit is calculated by subtracting the capital cost from the capital benefit. A negative benefit is identified by the use of parentheses.

Several Five of the design alternatives evaluated in other SAMDA analyses are included in the current AP600 design. These design features include:

• RCS depressurization system

. Passive residual heat removal system located inside containment

• Cavity flooding system

• Passive containment cooling system Revision: 11

[ Westingh00Se 1B-17 DRAFT,1997 o

1. Introduction cnd General Description of Plant

' I j

l 1

Hydr 6g~efignitdfiifa'IsicMr9:c6ntaininent

• Diverse ~ actuation ~systeiij Canned motor RCPs I

~

Interfacing systemLwith high'desigri'pressuni l 1

As the AP600 plant core damage frequency is approximately two orders of magnitude lower than for existing plants, the benefits of additional l design alternatives are very small. 7hs fifteen SAMDAs. analyzed provided little or no benefit tol th(AP.6Rdesign; Four cf the l

SAMDAs-analyzed-provided-no-benefit-at-al! =d de cie.~,-analyzed-provide-eegligible benefitsr--

Assuming an additional design alternative was developed which provides a 100 percent reduction in overall plant risk, representing an averted risk of 7.3 [10-8 3.42 10iman-rem per year, the capital benefit only amounts to $46.50 $2240, Because of the small initial risk associated with the AP600, none of the severe accident mitigation design altematives are cost beneficial.

l IB.9 References i i

1. "AP600 Probabilistic Risk Assessment," Westinghouse Electric Corporation and ENEL, l Revision 8, September 1996: June-26,-1992:

2. " Supplement to the Final Environmental Statement - Limerick Generating Station, Units I and 2," Docket Nos. 50-352/353, August 1989.

3. " Supplement to the Final Environmental Statement - Comanche Peak Steam Electric Station, Units I and 2," Docket Nos. 50-445/446, October 1989.

4. " System 80+ Design Alternatives Report," Docket No.52-002, April 1992.

5. Chanin,- D.I., Sprung, J.L., Ritchie, L.T., and Jow, H-N, MELCOR Abcident Consequence Code Systen (MACCS) User's Guide, NURGE,CR-4691, SAND 86-1562, Vol.il, Sandia i National Laboratories, U.S. Nuclear Regulatory Commission,1990. "MELCOR Accident Consequence Code Syste:n-(MACGS) Users Guide," NUREO/CR 4691, SANDS 61562, Volume 1,19 J.

6. " Technical Assessment Guide," EPRI P-6587-L, Volume 1, Revision 6, September 1989.

7. Nuclear Energy Cost Data Base, DOE /NE-0095, U.S. Department of Energy, September 1988.

8. "SECY 91-229 - Severe Accident Mitigation Design Alternatives for Certified Standard Designs," USNRC Memorandum from Samuel J. Chilk to James M. Taylor, dated October 25,1991.

Revision: 11 DRAFT,1997 1B-18 W85tkigh00$8

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1. Introduction cnd G;neral Description of Plant l

9. Chanin, D.I., Rollstin, J., Roster, J., and Miller, L., MACCS Version'1.5.11.1: A Maintenance Release of the Code, NUREG/CR-6059, SAND 92 2146, Sandia National Laboratories, U.S.

Nuclear Regulatory Commission, October 1993.

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[ W85tiligt10US6 I B-19 DRAFT,1997 9

1 l

1. Introdrction cnd General Description of Plant ,,

Table 1B.4-1

SUMMARY

OF ANNUAL LEVELIZED FIXED CIIARGE RATE ASSUMPTIO?4S Financial Factors Value l Discount Rate (before tax) 10.3% (Before Tax),9.13% (After Tax)  ;

RS%/yr  ;

Inflation rate 4.1%/yr N

Fr&ral and State Income Tax Rate 40.1%

E0% )

Investment Tax Credit 0.0%

Property Taxes and Insurance 20%

Tax Recovery Period 15 years Component Book Life 30 years

(%

Total Levelized Fixed Charge Rate 15.7 %

464 4 l

Revision: 11 DRAFT,1997 1B-20 Westinghouse i

i :' ..

1. Introduction cnd GInertl Description of Pl*nt Tehle-1Br54 SU'"'ARY OF F!ES!OF PRODUC RE! E^.SE FRAS!ONS u unemeir-rrn,~nne g g l

OE OKB CG C4 Xe Kr 4,2-*40i 44*401 64-*401 34-*401 Gsl 64-*40i 2-0-n-401 70-*-101 ar7-*-40i TeOr 04 04 04 00 S4 W 401 W-40i 44-*-401 677-*461 moo, W 40i 9.4+401 64-*401 44*40I GsOH W46' W404 00 *-401 3-7-*4 01 Ba0 W40* 64-*-401 4:2-*-40i 44-*40i La3 0,- 2# *40i Sr5-*401 3-1-*40i 24*401 Geo, 54-*401 44*40d 44-*-401 W401 Sb W401 44-*401 44+404 44-*401 Tea 04 04 04 0.0 UOr 0:0 04 04 04 Frequeney W-401 64+-l&* 74-*-10 d 14-*401 l

l l

Revision: 11

[ W85tiligl100S6 iB-21 DRAFT,1997

C lC E gQW Table IB.5-1 Il W

s- r

l 2 ENVIRONMENTAL RELEASE FRACTIONS AT 24 HOURS AFTER CORE DAMAGE

$O 4

PER RELEASE CATEGORY Release Emirammental Release Fractions at 24 Heers After Core Denmage Xe,Kr Cal TeO, SrO moo, Ch0H BeO I A,O CeO, Sb Te, UO.

IC 9.2E-4 4.7E-6 0.0E0 2.967 4.7E-6 4.5E-6 3.lE4 9.lE 7 1.0E4 8.2E4 0.0E0 0.0E0 BP 8.6&I 3.8&3 0.0E0 2.4&5 6.9E-4 3.2E-3 2.4FA 3.956 9.866 6.lE-2 0.0E0 0.0E0 Cl 8.4&1 3.4E-2 0.0B0 2.153 4.1&2 3.5&2 123-2 5.153 6.5E-3 6.1B-2 0.0E0 0.0E0 CFE 7.051 8.362 0.0E0 9.6E-4 2.7E-2 8.1&2 9.953 6.0E-4 8.854 7.252 0.0E0 0.0E0 CFI 6.2&l 3.4&3 0.0E0 5.8E 4 7.8&3 2.98-3 5.6&3 1.lE-3 1.5&3 1.052 0.0E0 0.000

- CFL I.lE-3 1.2B-5 0.0E0 5.9E 7 1 lE-5 1.165 6.lE4 1.6E-6 1.956 1.7E-5 0.0E0 0.G30 tc r

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1. Introduction cnd GIncrcl Description of Plant Td!: !E.6 !

l APf^^ EH!"^."D POPULAT!ON DOSE-ESTIMATES l (Effective-Whc!: Ecdy Equivalent-De= in Peran-S* evens) l t

De= (P:=c, Sknc) I l R&;x C!:ry D!ar=  !

l (Mge.) Mean Median j l

l OK 50 6 & -104 4-96-n-404 l

C4 50 4-14-*-40*8 7r51-*-40'8 l GG 4 50 9h40 6h40-a OMP 50 4J4-n404 442-*-404 i

Noter-4r-Doseswe4msed-on4he-50t ear <emmitted dcz fee-epowrWuring4he4nitial-24-hours fe!!cwing<ere I damage: I l \

2-One-person-s+ evert +quals-400-man 4em l

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I Revision: 11 l T Westinghouse IB-23 DRAFT,1997 l

o

L iifil"

1. I: trod:ction and General Description of Plant .

T:1!: !E.5 2

.'"!~2 S :: "!;h ("!h ! SOff " ;;!:t!;.. 0000 0 : 50 .'.'!!: ".0 '! 0)

Release F: quency Mz Cc=qu::ce %k Category (yed) (m= rem) (. . ... . . ... . jr'.)

OK 24-n-101 643 4r7h40' OKP 56n-401 414 7r50-n-10 d GG 7 4 *-40

• 944 6M-n-10*

C4 10 *-101 444000 142-*-404 Tetal-Ek 144-*40 4 l

Table 1B.6-1 l

1 AP600 BASE RISK OVhole Body Population Dose to a 50 Mile Radius) l I

Release Release Category Mean Population Dose' Risk Category Frequency (man-rem) (man-rem yr)

(yr)

IC 1.5 x 10 3.12 x 10** 4.7i x 105 CFE 6.6 x l a' 9.25 x 10*' 6.13 x 108 i CFI 1.3 x 10'" 3.35 x 10*5 4.39 x 104 CFL 1.5 x 10-" 1.05 x 10*8 4 1.59 xJ0 C1 3.6 x 10* 2.05 x 10*' 7.40 x 10d BP 1.1 x 104 3.72 x 10** 4.17 x 10d Total Risk 7.34 x 148 Note:

1. Doses are based on the 50 year committed dose for exposure during the initial 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following core damage.

Revision: 11 DRAFT,1997 IB 24 Westingh0US8 1

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1. Introduttlen and G:neral Description of Pla:t l

i Table IB.8-1 AP600 SAMDA RESULTS Risk Capital Capital Net Capital Design Alternative Reduction Benefit Cost Benefit (manrem/yr) ($) ($) ($)

Upgrade CVS for Small LOCA 640+40i 44 1 460,000 (!,450,C")

5.5 x 10d 4 1,500,000 (1,500,000)

Contanunent Filtered Vent 1.0 x 10 6 5,000,000 (5,000,000)

Self-Actu:. ting Containment Isolation 143-*401 7 60,4)00 (60,000)

Valves 7,4 x 10d 5 33,000 (33,000)

Safety Grade 3-39-*40d 22 3r500400 (17500,000)

Passive Containment Spray 6.9 x 10~8 44 3,900,000 (3,900,000)

Non-Safety Grade Containment Spray 6.9 x 10-' 44 415,000 (415,000)

Active High Pressure Safety injection 4-86-n401 44 20,000,000 (20,000,000)

System 6.1 x 10~' 39 SG Shell Side licat Removal 6-70-n40d 4d80400 (4d80400) 5.3 x 104 3 1,300,000 (1,300,000)

SG Relief Flow to IRWST 6.-70+401 560,000 (560,4)00) d 4.2 x 10 3 620,000 (620,000)

Increased SG Pressure Capability 6-70+-101 3r720.000 (2,720,000) 4.2 x 104 3 8,200,000 (8,200,000)

Secondary Containment Ventilation with 444-*401 7 2400;000 (3;004,000)

Filtration 7.4x19 5 2,200,000 (2,200,000)

Divesty4RWST-Velves 8434401 <-4 300400 (300,L")

Diverse IRWST Injection Valves 5.3 x 10-8 34 160,000 (160,000)

Diverse Containment Recire Valves 1.5xl@ <1 150,000 (150,000)

Ex-Vessel Core Catcher 6.1 x 10r' 39 1,660,000 (1,660,000) liigh Pressure Containment Design 6.1 x 10-8 39 50,000,000 (50,000,000)

More Reliable DAS/ DIS 748-n40" 5 390,4)00 (3 J C")

2.2 x 104 2 470,000 (470,000) i Revision: 11 (M W65tingt10US0 1B-25 DRAFT,1997

4 f&.-+-AlA.d'IAe.:wi E-$N-.g-Qwa-4t +A- 1, e-A'-4s4-A4 a+4n+%fsinkt.&._. Os %y-Ee+ 4ma+4 44 -+

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e Enclosure 2 to Westinghouse Letter NSD NRC-97-4937 January 10,1997 l

l NN k

NRC REQUEST FOR ADDITIONAL INFORMATION l

Hs Question: 100.14 l Describe in detail the process used to identify and evaluate candidate SAMDAs pertinent to the AP600 design. For those SAMDAs which were included in the assessments but which were not described or discussed in Appendix IB of the AP600 SSAR, entitled " Severe Accident Mitigation Design Alternatives," Revision I, dated January 13,1994, provide a description of these candidate SAMDAs, their estimated costs, and the reasons why they were not in:luded in the discussions of Appendix IB.

l

Response

The process used to identify and evaluate candidate SAMDAs pertinent to the AP600 design included a review of SAMDAs evaluated for other plant designs, including the following:

Limerick (Reference 100.14-1)

Comanche Peak (Reference 100.14-2)

. System 80+ (Reference 100.14-3)

In addition, the results of the Rev. O AP600 PRA were reviewed to assess possible design alternatives. Of the candidate SAMDAs identined from this initial review, the ones which were not included in the SSAR were those which were already included in the AP600 design. These AP600 design features include:

hydrogen ignition system )

reactor cavity Gooding system reactor coolant pump seal cooling (AP600 has canned rotor pumps) reactor coolant system depressurization external reactor vessel cooling.

All other SAMDAs are discussed and evaluated in Appendix IB of the SSAR.

References:

100.14-1 " Supplement to the Final Environmental Statement - Limerick Generating Station. Units I and 2,"

Docket Number 50-352/353, August 1989.

100.14 2 " Supplement to the Final Environmental Statement - Comanche Peak Steam Electric Stations Units I and 2," Docket Numbers 50-445/446, August 1989.

100.14-3 " System 80+ Design Alternative Recort," Docket Number 52-002, April 1992.

SSAR Revision: None.

W-Westinghouse

. 1 NRC REQUEST FOR ADDITIONAL INFORMATION 7

55 3

Ouestion: 100.15 The $20,000,000 projected cost for the addition of an active high pressure safety injection system (HPSI) app:ars to be excessive. Provide justification for this estimate.

Response-I The estimate for an active high pressure safety injection system (HPSI) was a designer's estimate; no detailed system analysis or cost calculations were performed. Part of the consideration of such a system was the costs associated with the requirements for making such a system safety-related and seismically qualified. Based on the risk reduction, a lower capital cost estimate would not effect the conclusion regarding risk vs cost.

SSAR Revision: None.

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Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

.- .n=

=

=l Question: 100.16 Provide more complete technical and cost information on the SAMDAs thus far evaluated for the AP600. The additional information should include design descriptions and definition (design feature descriptions, performance requirements, system schematics, etc.) and further details on the estimated costs for each SAMDA.

Response

The detail requested in this RAI is beyond the scope of evaluating a system which will not be implemented in the AP600 design. Detailed system descriptions, performance requirements, and system schematics were not developed as part of the SAMDA evaluation; rather the AP600 designers were provided a description of the design alternative, and an estimate of the design revisions and cost was completed for use in the SAMDA evaluation. As can be seen in SSAR Appendix IB, due to the low core damage frequency for the AP600, no design alternative is shown to be cost effective and thus no further effort to define system details is warranted.

SSAR Revision: None.

i W wesegnouse

NRC REQUEST FOR ADDITIONAL INFORMATION

= .-

? $

Question: 100.17 The design alternatives evaluated for the AP600 are stated to have been selected based in part upon design alternatives evaluated for other plant designs. References for Limerick, Comanche Peak, and CE System 80+ are cited as the other designs used for this purpose. However, no mention was made of whether plant improvements considered as part of the NRC Containment Performance Improvement (CPI) program were also included (see NUREG/CR-5367. -5575, -5630, and -5662). Please justify that the set of design alternatives considered for the AP600 include all relevant design improvements considered in these earlier evaluations.

Response

A review of NUREGs/CR-5567, -5575. -5630, and -5662 indicate that the design alternatives considered as part of the Containment Performance Improvement (CPI) program included design changes to enable:

RCS depressurization Hydrogen control Reactor cavity flooding Containment venting Corrective actions for ISLOCA Scrubbing for containment bypasses.

These design alternatives have been considered for AP600 RCS depressurization is accommodated via the automatic depressurization system. Hydrogen control is accomplished in the AP600 large dry containment with hydrogen igniters. Flooding of the reactor cavity is included in the AP600 design. Containment venting via a filtered vent is considered in the SAMDA cost benefit evaluation in the revised Appendix IB of the AP600 SSAR (revision 11).

Addressing ISLOCA via locating normal residual heat removal system inside containment is considered in the SAMDA cost benefit evaluation in the revised Appendix IB of the AP600 SSAR. Finally, scrubbing releases from containment bypasses is considered an accident management strategy versus a design alternative; such a strategy is included in WCAP-13913, Revision 1. December 1996, Framework for AP600 Severe Accident Management Guidance.

SSAR Revision: None.

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION g:  :

= m Question: 100.18 Where available, provide a comparison of the AP600 cost estimates to those for similar design alternatives considered in previous analyses, including the Comanche Peak Limerick, and Watts Bar SAMDA analyses, the NUREG-1150 studies, and pertinent SAMDA evaluations for the GE ABWR and CE System 80+ designs.

Response

Comparisons of SAMDAs for other dissimilar plant designs is not necessary for evaluation of the cost benent of AP600 design alternatives. The cost estimates provided for AP600 design alternatives have been intentionally biased on the low side to maximize the risk reduction benent. As can be seen in SSAR Appendix IB, due to the low core damage frequency for AP600, no design alternative is shown to be cost effective, even with the cost estimates minimized.

SSAR Revision: None.

i l

100.18-1 W West lnghouse i

NRC REQUEST FOR ADDITIONAL INFORMATION

= -= '*

Question: 100.19 Identify and discuss all risk-significant changes made to the design of the AP600 over the past few years which were i

based on the results of the PRA and/or consideration of SAMDA issues. In addition, specifically identify and discuss risk significant design changes and improvements made since the 1994 SAMDA submittal (Appendix IB of SSAR).

These discussions should note the risk reduction achieved by these changes, as well as their estimated costs.

1 Response: i No design changes have occurred since 1994 as a result of SAMDA issues.

The changes incorporated into the design since 1994 as a result of PRA are discussed below. These changes were based upon insights from the PRA analyses. The risk reduction, as measured by the reduction in the dose from a severe accident, was not a factor for these changes.

1. The squib valves in the IRWST injection lines are to be diverse from the squib valves in the IRWST l recirculation lines. This change improves the ability to flood the reactor cavity should the IRWST injection fail l

due to a common cause failure of the squib valves. The additional cost to make the valves diverse is estimated I at $160,000 per plant.

2. Two service water system (SWS) air-operated valves were changed to motor-operated valves. This was done ,

to improve the reliability of the SWS after the PRA analyses indicated a failure of the air-operated valves could I cause a failure of the system. The cost difference between the different kinds of valves is estimated to be

$10,000 per plant.

3. The locations of the IRWST vents and stairways for access to lower areas of containment were changed. The hydrogen diffusion flame analysis showed a potential for creep of the containment shell at the previous vent and stairway locations. To eliminate this potential, the locations were changed. There is no significant cost differential for this change.

SSAR Revision: None.

T westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

= m Ouestion: 100.20 Several of the SAMDAs identified in Appendix IB of the AP600 SSAR, dated January 13.1994, were discussed in qualitative terms and estimates of costs and risk reduction were not provided. Provide estimates of risk reduction and costs associated with each of these alternatives.

l

Response

l In the January 13,1994 SAMDA discussion, the following design alternatives were discussed qualitatively with no risk reduction or cost estimates:

• filtered vent a locating normal residual heat removal inside containment

• ex-vessel core catcher a high pressure containment design.

In Revision i1 of Appendix IB of the AP600 SSAR (February 28,1997), the filtered vent, ex-vessel core catcher i and high pressure containment design are all quantitatively evaluated with risk reduction estimates and capital cost l estimates. However, the design alternative for locating normal residual heat removal inside containment continues to be discussed qualitatively since quantitative calculations for cost estimates are not warranted due to virtually no l risk reduction benefit.

l SSAR Revision: None. 1

. l T

W-Westingtiouse

}

NRC REQUEST FOR ADDITIONAL INFORMATION

.sg  :

Question: 100.21 The SAMDAs discussed in Appendix IB were evaluated in terms of only four release categories: OK, Cl. CC, and OKP. In Revision 1 of the PRA the release categories were redefined and expanded to nine categories. Update the AP600 SAMDA evaluation relative to the expanded set of release categories.

Response

In Revision iI of Appendix IB of the AP600 SSAR (February 28.1997), the risk reduction evaluation is updated to reDect the release categories presented in AP600 PRA revision 8.

1 1

SSAR Revision: None.

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1 100.21-1 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION w

Ei. =

= T.

Cuestion: 100.22 NUREG/CR-5474, entitled " Assessment of Candidate Accident Management Strategies," presented several strategies for preventing core damage and for mitigating the effects of core damage. The SAMDAs identified in Appendix IB of the AP600 SSAR d!d not address the area of t.ccident management improvements. Discuss the basis f r excluding accident management strategies from the SAMDAs considered for the AP600.

Response

SAMDA are severe accident mitigation design alternatives. Mitigation of a severe accident involves the application of an accident management strategy. Thus, the S AMDAs evaluated consider accident management strategies via the calculation of risk reduction, and the discussion of how the design alternative would mitigate the severe accident.

Detailed accident management strategies were not developed for any design alternatives, or existing design features, and the final detailed AP600 accident management guidance (strategies) have not been developed.

The AP600 design already includes design alternatives which enhance accident management capability, such as:

• hydrogen ignition system

• reactor cavity flooding system a reactor coolant system depressurization.

I Accident management is not part of the AP60 certified design.

SSAR Revision: None.

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l 100.22-1 W-Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION EE =

F 5 Question: 100.23 The SAMDA evaluation in Appendix IB of the AP600 SSAR presented a brief evaluation of the alternative of increasing the reliability of the diverse actuation system (DAS). To what extent is this option of increasing the reliability of the DAS equivalent to making it safety grade? What consideration has been given to making the portion of the diversified actuation system that trips the reactor safety grade, and what would be the improvement in the reliability / availability of the reactor trip portion of the DAS if it were safety grade? What would be the cost of this limited scope upgrade?

Response

Increasing the reliability of the DAS is not related to, nor is it equivalent to, making the system safety-related. The DAS will be sufficiently reliable to meet the design goals, and it will incorporate industry design and quality standards including validation and verification of the system.

As noted in SSAR Appendix IB, if it were to be assumed that a DAS improvement "provided a 100 percent reduction in the overall plant risk, representing an averted risk of 7.3 x 10-3 man-rem per year, the capital benefit only amounts to $46.50." An improvement of the DAS cannot provide a 100 percent reduction in the overall plant risk, so the maximum capital benefit of a revision to the DAS would be less than $46.50.

If the reactor trip portion of the system were to be made safety-related, it would involve designing and building a safety-related sys:em for the reactor trip function. This would have to be completely separate from the rest of the nonsafety-related DAS. That is, the reactor trip function of the current DAS design is not a separate set of wires and chips. It is one function of many performed by the DAS components. If the reactor trip function were to be safety-related, it would require the design, construction, documentation, and verification of an entirely separate system for that function. This new safety-related diverse reactor trip system would have to include the additional documentation requirements of a safety-related component, as well as the additional redundancy requirements required of such components. Many, if not all, of the components of the safety-related function, would have to be custom designed and constructed instead of using readily available materials as will occur with the current design of the DAS.

The development and construction of this separate, diverse reactor trip function for the DAS is estimated to cost more than the DAS with the current design. This cost is significantly more than the maximum capital benefit of $46.50.

The additional cost to develop the safety-related reactor trip function would result in a very small improvement in the system reliability (due to the added redundancy for the trip function), and the maximum capital benefit could not be realized.

SSAR Revision: None.

[ WeStiligt10tJS8

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NRC REQUEST FOR ADDITIONAL INFORMATION l ms \

E. . ,,. l I

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Question: 100.24 l l

Based on information in Table 24-1 of the PRA, a significant number of penetrations would be screened out because l

they are 2-inches in diameter or less and may become plugged due to the particulate source term. However, in most l core melt sequences in the AP600 design the core debris would be covered by an overlying water pool, resulting in l less airborne fission products and a smaller likelihood of plugging. Without this screening, how much does the l probability for failure to isolate containment increase? How would the risk reduction (person-rem per yr) estimates

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for each SAMDA change if these leak paths were included in the evaluation?

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Response

Once all containment penetrations were identitled, screening criteria were used to eliminate containment penetrations which may not be important pathways for releases outside containment. One of the screening criterion, as noted above, is:  !

lines penetrating containment which are 2" in diameter or smaller (small lines tend to become plugged due to particulate source term) l This screening criterion would eliminate all containment penetrations of 2" diameter or less. However, there are also '

other criteria for screening containment penetrations. These include:

penetration is connected to a closed system inside containment whose integrity is not compromised during an accident, '

line penetrating containment Ms isolation valves and is part of a " closed system" outside containment, and l is capable of withstanding severe accident conditions, I penetration has at least one blind flange, penetration is administratively controlled and normally closed during power operation either by locked closed valves or power removed from valves, penetration has valves other than the containment isolation valves inside containment that are normally closed or automatically closed.

For those AP600 containment penetrations less than 2" diameter, all but four of these penetrations would have also been screened by one of the other criteria. For instance, the 2' demin water line is administratively locked closed, and thus may be screened via this criterion. The four penetrations which were screened solely on size are:

• 1" PSS line - Containment Air Sample Return

= 3/8" PSS line - RCS/PXS/CVS samples out

• 1" PCS line - Containment pressure instrument lines

. 1" WLS line - RCDT gas Thus, the lines screened out by size alone are all I" in diameter or less. Consequently, quantitatively speaking, consideration of these additional four containment penetrations in the containment isolation failure probability would increase, but less than if all lines 2" or less were included.

100.24-1 W-Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

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For consideration of risk reduction, consider that for release category IC (intact containment), a whole body population dose to a 50 mile radius is 300 man-rem. This dose is based on 6ssion product releases from a leaking containment, with the containment leak rate corresponding to the Tech Spec limit of 0.12% containment volume / day.

This limit was modeled in MAAP with a 10E-6 m' leak in containment. This corresponds to approximately a 1/16 inch diameter hole.

To conservatively estimate the impact of this on risk reduction, the 300 man. rem predicted by the intact containment cases are ratioed by the respective areas for the 1" bypass and nominal leakage (5.07E-4/2.95E-6 = 172); thus it is assumed that a 1" leak will result in a mean popuhtion dose of 5.2E+4 man-rem. Furthermore, assume that all intact containment cases now fall into this new category of 1" containment isolation failures; thus the 1" CI failure release category has a frequency of 1.5E-7 per year. Tl e total risk is thus 5.2E+4 x 1.5E-7 = 7.8E-3 man-rem / year.

Previously, the total risk was cah ulated to be 7.lE-3; with these conservative assumptions, the revised risk is estimated to increase to 1.5E-2. T he capital bene it for a " super design alternative" which results in 100 percent reduction in overall plant risk thus .ncreases fron $46.50 to $96.00.

The core damage frequency for AP600 is of such small magnitude to render consideration of these four additional 1" diameter containment penetrations insignificant to design alternative evaluations.

SSAR Revision: None.

100.24-2 W Westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION E:f SE Question: 100.25 What is the projected reliability of the software used for actuating important systems for core cooling and reactor )

shutdown? What design criteria and software quality assurance processes, including additional testing, are used for l the software to ensure it meets the reliability goal? I Response: l The software for actuating safety systems will conform to the requirements of ANSI /IEEE-ANS 7-4.3.2 (Reference 100.25-1). The quality assurance process for the software is as set forth in the Westinghouse Energy Systems ISO- I 9001 Quality Assurance program.

The PRA instrumentation and control models include common cause failure of software for actuation logic groups i in the protection and safety monitoring system (PMS), common cause failure of software for various output logic l groups in the PMS, and common cause failure of software postulated to be common to the PMS and the plant control l system (PLS). There is no fixed reliability goal for software. The values included in the PRA models for the l postulated common cause failure of the PMS software is 1.1E-05 per demand. The value for the common cause j failure of the software common to the PMS and PLS is 1.2E-M per demand. Both of these values are of sufficient magnitude to ensure that these events appear in the dominant core damage cutsets and risk rankings.

References:

100.25-1 ANSI /IEEE-ANS 7-4.3.2, " Standard Requirements for Digital Computers in Safety Systems of Nuclear 1 Power Generating Systems" l l

SSAR Revision: None. j l

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NRC REQUEST FOR ADDITIONAL INFORMATION 59 "=3

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1 Question: 100.26 l l

From the PRA it is not clear if the shutdown PRA considered situations when the containment is open. Describe I how the shutdown PRA addressed the containment being open, and discuss how consistent this model is with expected plant conditions at shutdown, especially during refueling.

Response

The AP600 Technical Specifications require that containment integrity be maintained during plant operation in Modes I through 4. Containment closure capability is required by the Technical Specifications when the plant is in Modes 5 and 6. The modeling in the Shutdown PRA addressed containme-; status in a manner consistent with expected plant conditions at shutdown, including refueling modes. Further details concerning this topic have been previously provided in the response to RAI 720.306.

SSAR/PRA Revision: None.

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l NRC REQUEST FOR ADDITIONAL INF04MATION g .. .;

= =g Question: 100.27 The AP600 PRA includes importance measures for risk increase and for risk decrease. However, we cannot find Fussell-Vesely measures of importance, or an equivalent mefuure, that provides a ranking af the importance of events to the core damage frequency (CDF). Although we can infer important event failures that contribute significantly to the CDF from the dominant sequences and their cut sets, we cannot tell the rank-ordered importance of events to CDF. Please provide a rank order of events by contribution to the CDF using the Fussell-Vesely measure or an equivalent measure.

l Response:

Chapter 59 of AP600 PRA (revision 8) includes risk increase and risk decrease importance measures for initiating events, for operator actions, and for hardware failures. Both of these importance measures provide rank-ordered indication of the importance of events to CDF. The risk decrease ranking provides information equivalent to that provided by Fussel-Vessely, in that it indicates the relative impact that could be obtained if the basic event in l question were guaranteed not to fail.

Table 100.27-1 lists the Fussel Vessely importance values for basic events in the CDF cutsets for the internal initiating events at power analysis.

SSAR/PRA Revision: None.

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100m W-Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION y= =q 5

Table 100.27-1 Fussel-Vessely Importance Values RISK IMPORTANCE CALCULATION Number of Basic Events = 536 Number of Cutsets = 14103 BEV = Basic Event Probability FV = Fussel Vessely Importance BASIC EVENT ID CUTSETS BEV PROB. FV 1 ACACV028GO 171 1.750E-03 5.590E-03 2 ACACV029GO 171 1.750E-03 5.590E-03 3 ACAOR001EB 6 7.200E-07 2.277E-06 4 ACAOR001SP 123 7.270E-04 2.325E-03 5 ACATK001AF 8 2.400E-06 7.614E-06 6 ACBCV028GO 164 1.750E-03 4.633E-03 7 ACBCV029GO 164 1.750E-03 4.633E-03 8 ACBOR001EB 3 7.200E-07 1.886E-06 9 ACBOR001SP 116 7.270E-04 1.926E-03 10 ACBTK001AF 5 2.400E-06 6.312E-06 11 ACX-CV-GO 168 5.100E-05 3.192E-02 12 ACX-TK-AF 12 1.200E-07 7.491E-05 13 AD2McD01 3 5.640E-02 1.646E-06 14 AD2 MOD 02 3 5.640E-02 1.646E-06 15 AD3 MOD 03 3 5.640E-02 1.646E-06 16 AD3 MOD 04 3 5.640E-02 1.646E-06 17 AD4 MOD 07 12 5.800E-04 4.728E-07 18 AD4 MOD 08 12 5.800E-04 4.728E-07 19 AD4 MOD 09 12 5.800E-04 4.728E-07 20 AD4 MOD 10 12 5.800E-04 4.728E-07 21 ADF-MAN 01 57 5.000E-01 6.348E-03 22 ADN-MAN 01 507 3.020E-03 1.087E-02 23 ADN-MAN 01C 4 5.000E-01 1.133E-02 24 ADX-EV-SA 1985 3.000E-05 3.332E-02 25 ADX-MV-GO 77 1.100E-03 4.400E-04 26 ALL-IND-FAIL 85 1.000E-06 5.211E-05 27 ATW-MAN 01 32 3.300E-02 5.969E-04 28 ATW-MAN 01C 53 5.170E-01 3.265E-02 29 ATW-MANO3 167 5.200E-02 4.997E-02 30 ATW-MAN 04 43 5 200E-02

. 4.627E-03 31 ATW-MAN 04C 50 5.260E-01 4.431E-02 32 ATW-MANOS 5 5.200E-03 4.208E-03 33 ATW-MAN 06 1 5.200E-03 6.680E-08 34 ATW-MAN 06C 1 5.000E-01 4.109E-03 35 BSIZE 315 5.000E-01 1.269E-01 36 BSIZE-LARGE 304 5.000E-01 1.269E-01 37 CANAV014LA 1 8.760E-03 1.998E-06 38 CANCV015GC 4 2.450E-02 2.092E-06 39 CANTP011RI 121 5.230E-03 8.784E-04 100.27-2 3 Westinghouse

l NRC REQUEST FOR ADDITIONAL INFORMATION

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BASIC EVENT ID CUTSETS BEV PROB. FV 40 CASMOD01 3 2.410E-03 3.693E-06 41 CASMOD02 11 2.310E-02 1.453E-05 42 CASMOD03 3 2.310E-02 4.582E-07 43 CAX-CM-ER 9 1.200E-04 1.041E-05 44 CCAMOD03 1 6.140E-04 1.135E-06 45 CCBMOD01 7 4.800E-02 2.461E-05 46 CCX-AV-LA 136 6.100E-05 3.947E-02 47 CCX-BC-SA 9 8.400E-06 1.533E-05 48 CCX-BL-ER 2 1.200E-05 2.618E-05 49 CCX-BY-PN 366 4.700E-05 7.314E-04 50 CCX-BY-PN1 45 5.700E-05 9.909E-06 51 CCX-EAI 1 1.270E-05 2.625E-07 52 CCX-EP-SA 34 8.620E-06 9.330E-05 53 CCX-EP-SAM 273 8.620E-06 1.391E-02 54 CCX-IN-LOGIC-SW 18 1.100E-05 7.032E-03 55 CCX-INPUT-LOGIC 70 1.030E-04 6.617E-02 56 CCX-IV-XR 130 2.400E-05 2.869E-05 57 CCX-IV-XR1 18 2.400E-05 3.541E-06 58 CCX-PL2 MOD 5 3 6.980E-05 4.628E-06 59 CCX-PL303 15 9.690E-05 2.423E-04 60 CCX-PL3EHO 2 4.030E-06 8.815E-06 61 CCX-PL3 MOD 1 24 1.410E-04 3.666E-04 62 CCX-PL3 MOD 1-SW 2 1.100E-05 2.401E-05 63 CCX-PL3 MODS 13 6.980E-05 7.474E-06 64 CCX-PL3 MOD 5-SW 1 1.100E-05 2.270E-07 65 CCX-PL403 26 9.690E-05 7.360E-05 66 CCX-PL4EHO 20 4.030E-06 2.941E-06 67 CCX-PL4 MOD 1 27 1.410E-04 1.071E-04 68 CCX-PL4 MOD 1-SW 23 1.100E-05 8.283E-06 69 CCX-PL903 6 9.690E-05 8.257E-06 70 CCX-PL9 MOD 1 12 1.410E-04 1.297E-05 71 CCX-PLA03 2 9.690E-05 3.340E-06 72 CCX-PLAMOD1 3 1.410E-04 5.474E-06 73 CCX-PLB03 6 9.690E-05 5.337E-06 74 CCX-PLPMOD1 10 1.410E-04 8.748E-06 75 CCX-PLD03 1 9.690E-05 2.294E-06 76 CCX-PLDMOD1 1 1.410E-04 3.340E-06 77 CCX-PLMMOD4 26 4.980E-05 3.782E-05 78 CCX-PLKMOD4-SW 23 1.100E-05 8.283E-06 79 CCX-PLMOD3 21 1.030E-04 1.968E-05 80 CCX-PLMOD3-SW 1 1.100E-05 2.270E-07

. 81 CCX-PLSMOD6 32 2.530E-04 5.477E-05 82 CCX-PLSMOD6-SW 1 1.100E-05 2.270E-07 83 CCX-FM-ER 2 1.400E-05 3.058E-05 84 CCX-PMA030 78 9.690E-05 8.503E-05 85 CCX-FMAEHO 20 4.030E-06 2.941E-06 86 CCX-FMAMOD1 86 1.410E-04 1.245E-04 87 CCX-FMAMOD2 4 3.040E-04 1.453E-07 88 CCX-FMAMOD4 26 4.980E-05 3.782E-05 89 CCX-PMB030 89 9.690E-05 1.832E-05 90 CCX-PMBMOD1 110 1.410E-04 2.769E-05 W Westinghouse

NRC REQUEST FOR ADDITIONAL. INFORMATION at:_ a:::

? N BASIC EVENT ID CUTSETS BEV PROB. FV 91 CCX-FMBMODE 4 3.040E-04 1.453E-07 92 CCX-FhC030 2 9.690E-05 4.410E-06 93 CCX '/MCMOD1 2 1.410E-04 6.414E-06 94 CCX-FMCMOD2 4 3.040E-04 1.453E-07 95 CCX-FMCMOD4 1 4.980E-05 1.915E-06 96 CCX-FMD030 48 9.690E-05 1.003E-04 97 CCX-PMDEHO 20 4.030E-06 2.941E-06 98 CCX-PMDMOD1 50 1.410E-04 1.460E-04 99 CCX-PMDMOD2 4 3.040E-04 1.453E-07 100 CCX-PMDMOD4 32 4.980E-05 4.784E-05 101 CCX-FMS-HARDWARE 116 7.890E-05 2.853E-02 102 CCX-FMXMOD1-SW 321 1.100E-05 1.778E-02 l 103 CCX-PMXMOD2-SW 18 1.100E-05 7.032E-03 l 104 CCX-FMXMOD4-SW 67 1.100E-05 2.671E-04 105 CCX-SFTW 191 1.200E-06 1.485E-02 i 106 CCX-TRNSM 315 4.7F'E-04 1.248E-03 I 107 CCX-TT-UF 115 1.170E-04 1.428E-04 108 CCX-VS-FA 15 3.840E-05 1.461E-04 I 109 CCX-XMTR 284 4.780E-04 2.532E-02 i 110 CCX-XMTR1 1 4.780E-04 4.452E-06 111 CCX-XMTR195 103 4.780E-04 2.495E-02 l 112 CDNTF01BRI 36 5.230E-03 2.306E-04 l 113 CIAEP014SA 1 1.710E-04 3.902E-08 114 CIB-MAN 00 54 1,840E-03 8.251E-03 115 CIB-MAN 01 51 1.340E-03 1.554E-03 116 CIX-AV-LA 1 7.700E-04 7.153E-06 117 CMA-CV 10 2.000E-06 1.016E-05 118 CMA-PLUG 97 7.270E-04 4.408E-03 119 CMAAV014LA 10 1.590E-03 1.286E-05 120 CMAAV015LA 10 1.590E-03 1.286E-05 121 CMAOR001EB 10 7.200E-07 3.656E-06 122 CMATK002AF 10 2.400E-06 1.220E-05 123 CMB-CV 2 2.000E-06 9.022E-07 124 CMB-PLUG 23 7.270E-04 3.395E-04 125 CMBAV014LA 2 1.590E-03 1.141E-06 126 CMBAV015LA 2 1.590E-03 1.141E-06 127 CMBOR001EB 2 7.200E-07 3.252E-07 128 CMBTK002AF 2 2.400E-06 1.083E-06 129 CMX-AV-LA 43 9.600E-05 4.545E-04 130 CMX-CV-GO 89 5.100E-05 3.273E-02 131 CMX-TK-AF 7 1.200E-07 7.630E-05 132 CMX-VS-FA 116 3.840E-05 2.540E-02 133 CONDVACUUM 11 1.000E-03 5.664E-05 134 CV3EPCPASA 2 1.710E-04 1.045E-05 135 CVBPM01BTM 32 2.190E-02 3.116E-04 136 CVMOD01 29 2.210E-04 5.829E-04 137 cvMOD02 8 1.410E-03 1.135E-04 138 CVMOD03 16 1.120E-02 1.445E-04 139 CVMOD04 47 7.370E-04 1.989E-03 140 CVMODOS 26 2.880E-02 5.968E-04 100.27-4 3 Westinghouse

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1 NRC REQUEST FOR ADDITIONAL INFORMATION  ;

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l BASIC EVENT ID CUTSETS BEV PROB. FV l 141 CVMOD07 26 2.710E-02 5.635E-04 142 CVN-MAN 00 5 3.100E-03 5.187E-03 143 CVN-MAN 02 2 1.580E-03 1.466E-07 144 CVN-MANO3 2 1.070E-03 8.111E-06 1 145 CVNMV090GC 2 8.760E-02 1.460E-06 146 CVNMV091GC 8.760E-02 l 2 1.460E-06 147 CVX-PM-ER 3 3.700E-05 8.194E-05 148 DAS 360 1.000E-02 1.924E-02 149 DUMP-MAN 01 4 1.320E-03 3.369E-05 150 ECOMOD01 1175 5.080E-03 1.465E-03 151 EC1BS001LF  ! 4.800E-06 8.176E-08 152 EC1BS001TM 8 44 2.700E-03 1.646E-03 153 EC1BS011TM 2 81: 2.700E-03 6.238E-04 154 EC1BS012TM 33b 2.700E-03 9.217E-04 155 EC1BS013TM 216 2.700E-03 6.626E-04 156 EC1BS111TM 201 2.700E-03 7.675E-05 157 EC1BS112TM 39 2.700E-03 4.243E-04 158 EC1BS121TM 201 2.700E-03 5.853E-05 i 159 EC15S122TM 6 2.700E-03 3.775E-06 l 160 EC1BS131TM 39 2.700E-03 4.243E-04 161 EC1CB100VO 92 4.200E-03 1.453E-05 162 EC1 MOD 11 16 4.800E-05 3.707E-06 163 EC1 MOD 12 14 4.800E-05 6.824E-06 164 EC1 MOD 13 32 4.800E-05 4.880E-06 165 EC1REDG1GA 3 4.360E-03 8.218E-07 166 EC2BS002LF G 4.800E-06 2.207E-07 167 EC2BS002TM 488 2.700E-03 6.185E-04 168 EC2BS021TM 56 2.700E-03 1.976E C4 169 EC2BS022TM 234 2.700E-03 2.896E-04 170 EC2BS023TM 211 2.700E-03 3.000E-04 171 EC2BS211TM 35 2.700E-03 9.903E-05 172 EC2BS212TM 13 2.700E-03 9.840E-05 173 EC2BS221TM 208 2.700E-03 1.377E-04 174 EC2BS222TM 9 2.700E-03 4.298E-06 175 EC2BS231TM 13 2.700E-03 9.840E-05 176 EC2CB200VO 103 4.200E-03 4.174E-05 177 EC2 MOD 21 1 4.800E-05 7.922E-08 178 EC2 MOD 22 10 4.800E-05 1.472E-06 179 EC2 MOD 221 1 1.680E-05 4.854E-07 180 EC2 MOD 23 30 4.800E-05 1.321E-06 181 EC3BS003TM 1 2.700E-03 1.608E-06 182 EC4BS004TM 2 2.700E-03 1.648E-06 183 EC4BSO41TM 2 2.700E-03 1.648E-06 184 EC4BS411TM 2 2.700E-03 1.648E-06 185 ECX-CB-GC 71 7.300E-04 6.625E-05 186 ECX-CB-GO 60 4.200E-04 3.784E-05 187 ED1BSDS1LF 5 4.800E-06 8.176E-08 188 ED1BSDSITM 170 3.000E-04 8.527E-05 189 ED1 MOD 01 85 5.040E-04 3.560E-05 190 ED1 MOD 03 190 2.700E-03 3.616E-04 191 ED1 MOD 06 66 3.480E-04 2.740E-06 T westingbuse

NRC REQUEST FOR ADDITIONAL INFORMATION n= ==.

E :ij BASIC EVENT ID CUTSETS BEV PROB. FV 192 ED1 MOD 07 107 3.050E-04 2.240E-05 193 ED1 MOD 11 81 3.170E-04 5.380E-05 194 ED1 MOD 113 81 3.170E-04 5.380E-05 195 ED1 MOD 13 115 3.170E-04 2.334E-05 196 ED2BSDS1LF 5 4.800E-06 8.176E-08 197 ED2BSDS1TM 97 3.000E-04 3.245E-05 198 ED2 MOD 03 69 2.700E-03 2.347E-05 199 ED2 MOD 11 101 3.170E-04 3.428E-05 200 ED3BSDS1TM 33 3.000E-04 2.426E-05 201 ED3 MOD 01 83 5.040E-04 6.026E-05 202 ED3 MOD 03 44 2.700E-03 2.538E-05 203 ED3 MOD 04 89 2.190E-02 6.932E-05 204 ED3 MOD 07 400 3.050E-04 6.886E-03 205 ED4BSDSITM 157 3.000E-04 2.834E-05 206 ED4 MOD 02 1 1.920E-04 8.218E-08 207 ED4 MOD 03 30 2.700E-03 5.627E-07 208 ED4 MOD 11 165 3.170E-04 3.267E-05 209 ED4 MOD 112 164 3.170E-04 3.237E-05 210 FWBMOD11A 1 3.340E-04 2.069E-06 211 FWDMOD11B 1 3.340E-04 2.069E-06 212 FWMOD010 10 1.410E-02 1.514E-05 213 FWMOD013A 32 1.410E-02 3.268E-05 214 FWMOD013B 57 1.410E-02 3.823E-05 215 FWMOD028 160 1.410E-02 7.617E-04 216 FWMOD03A 36 1.700E-02 3.993E-05 217 FWMOD03B 64 1.700E-02 4.710E-05 218 FWMOD067A 36 1.410E-02 5.969E-04 219 FWMOD067B 10 1.410E-02 1.152E-05 220 FWNCV029CO 8 2.190E-04 2.548E-06 221 FWX-MV2-GO 15 5.500E-04 3.547E-05 222 FWX-FM2-FS 15 5.400E-04 3.460E-05 223 HPM-MAN 01 1 5.020E-04 2.578E-07 224 IDABSDD1LF 20 4.800E-06 3.505E-06 225 IDABSDDITM 67 3.000E-04 2.991E-04  !

226 IDABSDK1LF 20 4.800E-06 3.505E-06 1 227 IDABSDK1TM 55 3.000E-04 2.975E-04 1 228 IDABSDS1LF 20 4.800E-06 3.505E-06 229 IDABSDS1TM 73 3.000E-04 2.994E-04 230 IDAFD003RQ 23 1.200E-05 9.030E-06 231 IDAFD004RQ 23 1.200E-05 9.030E-06 232 IDAMOD04 119 3.170E-04 1.967E-05 233 IDAMOD05 74 5.160E-04 3.666E-06 234 IDAMOD06 9 4.320E-05 7.467E-08 ]

235 IDAMOD07 26 2.190E-02 4.027E-06 236 IDAMOD08 44 3.170E-04 1.061E-06 237 IDBBSDD1LF 5 4.800E-06 1.383E-05 238 IDBBSDDITM 158 3.000E-04 1.000E-03 i 239 IDBBSDK1TM 12 3.000E-04 1.eieE-06 I 240 IDBBSDS1LF 5 4.800E-06 1.383E-05 241 IDBBSDS1TM 164 1.000E-04 1.001E-03 100.27-6 T Westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION b- . . . l BASIC EVFNT ID CUTSETS BEV PROB. FV 242 IDBFD013RQ 26 1.200E-05 3.765E-05 l 243 IDBMOD24 18 3.170E-04 1.674E-05 l 244 IDBMOD25 6 5.160E-04 5.321E-07 1 245 IDBMOD27 26 2.190E-02 4.027E-06 246 IDCBSDDlLF 3 4.800E-06 2.252E-08 247 IDCBSDDITM 203 3.000E-04 1.133E-04 248 IDCBSDS1LF 3 4.800E-06 2.252E-08 249 IDCBSDS1TM 209 3.000E-04 1.136E-04 250 IDCFD007RQ 27 1.200E-05 3.129E-06 I 251 IDCMOD28 18 3.170E-04 1.674E-05 l 252 IDCMOD29 6 5.160E-04 5.321E-07 '

253 IDCMOD31 26 2.190E-02 4.027E-06 254 IDDBSDD1LF 25 4.800E-06 1.734E-05 255 IDDBSDD1TM 169 3.000E-04 1.289E-03 256 IDDBSDK1LF 20 4.800E-06 3.505E-06 257 IDDBSDK1TM 81 3.000E-04 3.175E-04 258 IDDBSDS1LF 25 4.800E-06 1.734E-05 259 IDDBSDS1TM 175 3.000E-04 1.289E-03 ,

260 IDDFD019RQ 33 1.200E-05 4.470E-05  !

261 IDDFD020RQ 23 1.200E-05 9.030E-06 l 262 IDDMOD32 119 3.170E-04 1.967E-05 263 IDDMOD33 74 5.160E-04 3.666E-06 264 IDDMOD34 9 4.320E-05 7.467E-08 265 IDDMOD35 26 2.190E-02 4.027E-06 l 266 IDDMOD38 44 3.170E-04 1.061E-06 1 267 IEV-ATW-S 91 2.050E-02 2.256E-03 268 IEV-ATW-T 13 1.170E+00 4.210E-03 ,

269 IEV-ATWS 230 4.810E-01 5.309E-02 1 270 IEV-CMTLB 1404 8.940E-05 2.093E-02 271 IEV-ISLOC 1 5.000E-11 2.956E-04 272 IEV-LCAS 174 3.480E-02 1.024E-03 273 IEV-LCCW 244 1.440E-01 7.252E-04 274 IEV-LCCND 316 1.120E-01 6.112E-03 275 IEV-LLOCA 642 1.050E-04 2.967E-01 276 IEV-LMFW 253 3.350E-01 1.790E-03 277 IEV-LMFW1 150 1.920E-01 1.040E-03 278 IEV-LOSP 694 1.200E-01 5.956E-03 279 IEV-LRCS 34 1.800E-02 7.508E-05 280 IEV-MLOCA 1713 1.620E-04 3.683E-02 281 IEV-NLOCA 3383 ' 700E-04

. 1.863E-01 282 IEV-POWEX 391 4.500E-03 1.084E-02 283 IEV-PRSTR 330 2.500E-04 3.298E-03 284 IEV-RCSLK 802 1.200E-02 1.338E-02 285 IEV-RV-RP 1 1.000E-08 5.912E-02 286 IEV-SGTR 507 5.200E-03 3.597E-02 287 IEV-SI-LB 287 1.040E-04 2.258E-01 288 IEV-SLB-D 26 5.960E-04 5.595E-05 289 IEV-SLB-U 117 3.720E-04 7.275E-04 290 IEV-SLB-V 197 1.210E-03 2.347E-03 291 IEV-SLOCA 1657 1.010E-04 2.396E-02 292 IEV-TRANS 446 1.400E+00 6.736E-03

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NRC REQUEST FOR ADDITIONAL INFORMATION 1

I BASIC EVENT ID CUTSETS BEV PROB. FV 1

293 IRBEP117BSA 43 1.710E-04 8.858E-07 294 IRBEP123ASA 16 1.710E-04 1.157E-05 295 IRBEP123BSA 16 1.710E-04 1.157E-05 296 IRDEP118BSA 43 1.710E-04 8.858E-07 297 IRWMOD01 103 1.200E-02 1.239E-04 298 IRWMOD03 243 1.200E-02 1.960E-04 299 IRWMODOS 56 1.460E-03 4.241E-03 300 IRWMOD06 54 1.460E-03 4.239E-03 301 IRWMOD07 40 1.460E-03 1.507E-06 i

l 302 IRWMOD08 42 1.460E-03 2.861E-06 303 IRWMOD09 33 1.460E-03 1.229E-05 304 IRNMOD10 161 1.460E-03 7.289E-05 305 IRWMOD11 141 1.460E-03 2.123E-05 306 IRWMOD12 289 1.460E-03 9.370E-05 307 IWA-PLUG 229 2.400E-04 1.479E-01 308 IWACV122AO 62 1.750E-03 5.080E-03 l' 309 IWACV124AO 60 1.750E-03 5.079E-03 310 IWARS118BFA 112 8.760E-04 1.171E-05 311 IWARS123BFA 17 8.760E-04 7.133E-07 l 312 IWB-PLUG 226 2.400E-04 6.706E-05 '

313 IWBCV122AO 45 1.750E-03 1.839E-06 314 IWBCV124AO 45 1.750E-03 3.448E-06 315 INBRS118AFA 20 8.760E-04 6.410E-06 316 IWBRS123AFA 31 8.760E-04 2.538E-03 317 IWCRS120BFA 233 8.760E-04 5.313E-05 318 IWCRS125BFA 17 8.760E-04 7.133E-07 319 IWDRS120AFA 122 8.760E-04 4.080E-05 320 INDRS125AFA 31 8.760E-04 2.538E-03 4

321 INNTK001AF 45 2.400E-06 1.275E-05 322 IWX-CV-AO 1958 3.000E-05 5.179E-02 323 IWX-CV1-AO 1 5.400E-07 3.322E-04 324 IWX-EV-SA 1819 2.600E-05 4.481E-02 325 IWX-EV1-SA 1 1.000E-05 6.148E-03 326 IWX-EV3-SA 23 1.000E-05 7.525E-06 327 IWX-EV4-SA 969 2.600E-05 2.206E-01 328 IWX-FL-GP 1258 1.200E-05 1.319E-02 329 INX-XMTR 454 4.780E-04 4.125E-02 330 LPM-MAN 01 132 1.340E-03 1.044E-03 331 LPM-MAN 02 312 3.300E-03 8.427E-03 332 MDAS 751 1.000E-02 1.278E-02 333 MSAEPSD1SA 3 1.710E-04 6.958E-06 334 MSAEPSD2SA 3 1.710E-04 6.958E-06 335 MSAEPSD3SA 3 1.710E-04 6.958E-06 336 MSAEPSD4SA 3 1.710E-04 6.958E-06 337 MSAEPSD5SA 3 1.710E-04 6.958E-06 338 MSAEPSD6SA 3 1.710E-04 6.958E-06 339 MSAEPSD75A 3 1.710E-04 6.958E-06 340 MSAEPSD8SA 3 1.710E-04 6.958E-06 341 MSHTP001RI 30 5.230E-03 7.494E-03 342 MSHTP002RI 30 5.230E-03 7.494E-03 100.27-8 3 Westingh00Se

NRC REQUEST FOR ADDITIONAL INFORMATION F= =M I 35 BASIC EVENT ID CUTSETS BEV PROB. FV 343 MSMODV001 2 2.710E-02 3.453E-07 344 MSMODV003 2 2.710E-02 3.453E-07 345 MSMODV005 2 2.710E-02 3.453E-07 346 MSMODV007 2 2.710E-02 3.453E-07 347 MSX-AV-FA 15 1.500E-03 1. 36E-04 348 OTH-BL 11 1.900E-01 1.440E-04 349 OTH-MGSET 37 1.750E-03 4.599E-03 350 OTH-PO 1 1.200E-04 1.117E-06 l 351 OTH-PRES 52 2.000E-03 4.927E-04 1 352 OTH-PRESU 139 3.270E-01 3.331E-02 353 OTH-PRSOV 392 1.000E-02 1.082E-02 354 OTH-R05 694 7.000E-01 5.956E-03 355 OTH-SDMAN 44 7.700E-04 2.038E-03 1 356 OTH-SGTR 512 1.000E-02 1.468E-02 357 OTH-SGTR1 30 6.700E-03 5.413E-04 358 OTH-SLSOV 167 1.100E-02 1.076E-03 359 OTH-SLSOV1 282 2.100E-02 1.029E-02 360 OTH-SLSOV2 45 1.000E-02 7.906E-04 361 OTH-SLSOV3 109 5.400E-03 5.222E-04 l 362 PCNHR001ML 45 2.400E-06 1.275E-05 363 PL20301ASA 2 1.160E-03 1.155E-05 364 PL20301BSA 2 1.160E-03 1.155E-05 l 365 PL2 MOD 11 4 2.090E-03 2.464E-05 366 PL2 MOD 52 2 8.740E-04 5.297E-07 367 PL30301ASA 5 1.160E-03 2.428E-05 368 PL30301BSA 4 1.160E-03 2.180E-05 369 PL30302ASA 2 1.160E-03 6.952E-06 370 PL30302BSA 1 1.160E-03 4.469E-06 371 PL3 MOD 11 11 2.090E-03 5.677E-05 372 PL3 MOD 12 4 2.090E-03 1.846E-05 373 PL40301ASA 135 1.160E-03 1.019E-05 374 PL40301BSA 112 1.160E-03 9.169E-06 375 PL40302ASA 66 1.160E-03 3.270E-06 376 PL40302BSA 43 1.160E-03 2.252E-06 377 PL4EH0A1SA 32 8.000E-05 5.181E-07 378 PL4EH0A2SA 11 8.000E-05 1.563E-07 379 PL4 MOD 11 172 2.090E-03 2.025E-05 380 PL4 MOD 12 102 2.090E-03 7.809E-06 381 PL4XS00ASA 37 8.000E-05 6.017E-07 382 PL5 MOD 11 2 2.090E-03 2.755E-07 383 PL70302ASA 2 1.160E-03 7.035E-07 384 PL70302BSA 2 1.160E-03 7.035E-07 385 PL7 MOD 12 2 2.090E-03 1.266E-06 386 PL90302ASA 2 1.160E-03 7.035E-07 387 PL90302BSA 2 1.160E-03 7.035E-07 388 PL9 MOD 12 3 2.090E-03 1.905E-06 389 PLAMOD12 1 2.090E-03 6.385E-07 390 PLMMOD41 100 6.350E-04 3.733E-05 391 PLMMr042 23 6.350E-04 4.663E-07 392 PLSMOU61 1 3.460E-03 2.471E-07 393 PLSMOD62 4 3.460E-03 2.293E-06 T Westinghouse

e NRC REQUEST FOR ADDITIONAL. INFORMATION I f BASIC EVENT ID CUTSETS BEV PROB. FV 394 PMA0301ASA 135 1.160E-03 1.016E-05 395 PMA0301BSA 112 1.160E-03 9.147E-06 396 PMA0302ASA 56 1.160E-03 3.051E-06 397 PMA0302BSA 33 1.160E-03 2.033E-06 398 PMAEH0AlSA 32 8.000E-05 5.181E-07 399 PMAEHOA2SA 11 8.000E-05 1.563E-07 400 PMAMOD11 175 2.090E-03 2.024E-05 401 PMAMOD12 86 2.090E-03 7.361E-06 402 PMAMOD31 73 5.020E-03 2.740E-04 403 PMAMOD41 86 6.350E-04 4.221E-06 404 PMAMOD42 18 6.350E-04 3.694E-07 405 PMAXS00ASA 37 8.000E-05 6.017E-07 406 PMBMOD11 9 2.090E-03 6.811E-07 407 PMBMOD32 73 5.020E-03 2.740E-04 408 PMCMOD33 60 5.020E-03 2.697E-04 409 PMD0301ASA 135 1.160E-03 1.016E-05 410 PMD0301BSA 112 1.160E-03 9.147E-06 411 PMD0302ASA 56- 1.160E-03 3.051E-06 412 PMD0302BSA 33 1.160E-03 2.033E-06 413 PMDEH0A1SA 32 8.000E-05 5.181E-07 414 PMDEHOA2SA 11 8.000E-05 1.563E-07 415 PMDMOD11 186 2.090E-03 2.120E-05 416 PMDMOD12 8G 2.090E-03 7.361E-06 417 PMDMOD34 60 5.020E-03 2.697E-04 j 418 PMDMOD41 86 6.350E-04 4.221E-06 419 PMDMOD42 18 6.350E-04 3.694E-07 420 PMDXS00ASA 37 8.000E-05 6.017E-07 421 PMS-RTSWITCH 1 3.000E-05 1.271E-07 422 PRAAV108LA 21 1.090E-03 5.341E-06 423 PRAAV108TM 14 5.000E-04 2.007E-06 424 PRAMOD10 11 2.110E-03 1.072E-04 425 PRAMOD9 42 1.410E-02 8.055E-04 426 PRBAV108LA 21 1.090E-03 5.341E-06 427 PRBAV108TM 14 5.000E-04 2.007E-06 1 428 PRBMOD10 11 2.110E-03 1.072E-04 l 429 PRCEP101SA 2 1.710E-04 7.768E-06 l 430 PRCEP108SA 2 1.710E-04 7.768E-06 431 PRDEP108SA 2 1.710E-04 7.768E-06 432 PRI-MAN 01 2 C.960E-04 2.254E-05 433 PXX-AV-LA 1220 9.600E-05 1.120E-03 434 RC1CB051GO 107 4.200E-03 4.112E-04 435 RC1CB052GO 107 4.200E-03 4.112E-04 436 .RC1CB053GO 107 4.200E-03 4.112E-04 437 RC1CB054GO 107 4.200E-03 4.112E-04 )

438 RC1CB061GO 107 4.200E-03 4.112E-04 l 439 FC1CB062GO 107 4.200E-03 4.112E-04 440 RC1CB063GO 107 4.200E-03 4.112E-04 441 RC1CB064GO 107 4.200E-03 4.112E-04 442 RCX-RB-PA 161 8.100E-06 5.373E-03 i 443 REA-PLUG 166 2.40CE-04 6.006E-04 100.27-10 3 Westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION i gs y i

$ BASIC EVENT ID CUTSETS BEV PROB. FV '

444 REACV119CO 163 1.750E-03 8.747E-05 445 REAMOV117TM 18 5.000E-04 3.560E-06

.' 446 REB-PLUG 207 2.400E-04 7.895E-04 447 REBCV119GO 296 1.750E-03 1.126E-04 d

448 REBMOV117TM 96 5.000E-04 6.517E-06 449 REC-MANDAS 288 1.160E-02 1.297E-02 450 REC-MANDASC 429 5.060E-01 3.161E-02 451 REG-MAN 00 321 2.040E-01 9.911E-04 j 452 REN-MAN 04 393 1.000E-02 4.044E-02

453 REX-FL-GP 1258 1.200E-05 1.319E-02 1 454 RHN-MAN 01 124 2.900E-03 2.417E-03

, 455 FHN-MAN 01C 11 5.000E-01 9.260E-04 456 RHN-MAN 06 246 3.750E-03 5.629E-05 457 RN11 MOD 3 372 1.410E-02 1.199E-02 458 RN22 MOD 4 372 1.410E-02 1.199E-02 159 RN23 MOD 5 372 1.410E-02 1.199E-02 460 RNAEP01ASA 54 1.710E-04 4.998E-06 461 RNAEP01BSA 57 1.7109-04 4.929E-06  !

462 .RNAEP022SA 27 1. ~ 10 3.- 0 4 1.298E-04 463 RNAMOD06 418 3.C03-02 1.133E-03 t 464 RNAh0D09 130 5.010E-02 2.145E-03 465 RNBEP011SA 1.710E-04 1.298E-04 27 466 RNBMOD07 444 3.400E-02 1.236E-03

. 467 RNBMOD10 130 5.070E-02 2.145E-03 3 468 R14DEP023SA 27 1.710E-04 1.298E-04

! '469 RNNCV013GO 101 1.750E-03 1.450E-03 1 470 RNX-CV-GO 26 5.100E-05 3.876E-05 471 RNX-KV-GO 60 6.100E-04 4.894E-04 472 RNX-KV1-GO 180 4.900E-03 4.110E-03 .,

473 RNX-PM-ER 23 1.600E-05 1.205E-05 l 474 RNX-FM-FS 65 7.700E-04 6.202E-04

, 475 ROD-CTRL-SYS 24 6.600E-04 5.152E-05 e

476 RPTMOD01 52 8.760E-04 8.138E-05 477 RPTMOD02 52 8.760E-04 8.138E-05 a 478 RPTMOD03 52 8.760E-04 8.138E-05 i 479 RPTMOD04 52 8.760E-04 8.138E-05 4

480 RPTMOD05 52 8.760E-04 8.138E-05

, 481 RPTMOD06 52 8.760E-04 3.138E-05 j 482 RPTMOD07 52 8.760E-04 8.138E-05 483 RPTMOD08 52 8.760E-04 8.138E-05 4R4 RPX-CB-GO 201 4.200E-04 2.136E-02 485 SFBEP028SA 6 1.710E-04 1.830E-06 486 SFNMV067GC 2 1.100E-02 2.093E-06 487 SG1TF51ARI 6 5.230E-03 1.964E-06 488 SG2TF50ARI 8 5.230E-03 4."31E-05 489 SGAAV040LA 2 1.090E-03 1. E-06 490 SGAOR--DAS-SP 14 7.220E-03 1.2.0E-04 491 SGATL--DAS-UF 11 5.230E-03 8.714E-05 492 SGBAV040LA 51 1.090E-03 1. 2 4~. f- 0 3 493 SGBAV074LA 11 8.760E-03 8.136E-05 494 SGBA' DSLA 11 8.760E-03 8.136E-05 100.27-11

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NRC REQUEST FOR ADDITIONAL INFORMATION BASIC EVENT ID CUTSETS BEV PROB. FV 495 SGBOR--DAS-SP 14 7.220E-03 1.220E-04 496 SGBTL--DAS-UF 11 5.230E-03 8.714E-05 497 SGX-MV-GO 14 5.500E-04 3.310E-Oc 498 SW4:.tOD03 2 6.340E-04 1.525E-0e 499 SWAMvD09T 16 2.520E-04 2.613E-05 500 SWB-001TM 7 3.800E-02 1.953E-05 501 SWBMOD02 3 2.440E-02 7.207E-06 502 SWBMOD11P 1 1.410E-02 1.638E-07 503 SWN-MANO3 20 4.000E-02 4.492E-04 504 TCBMOD01B 1 2.520E-02 1.608E-06 505 VF1AV004 1 8.760E-03 8.572E-08 506 VFIAV010 1 8.760E-03 8.572E-08 507 VFOAV003 1 8.760E-03 8.572E-08 508 VFOAV009 1 8.760E-03 8.572E-08 509 VFSFRAC 2 1.200E-01 1.714E-07 510 VWAMOD01 6 2.520E-04 2.693E-05 511 VWAMOD02 12 G.120E-04 8.366E-05 512 VWAMOD03 6 2.520E-04 2.693E-05 513 VWBMOD04 74 1.830E-02 7.961E-04 514 VWBMODOS 92 2.190E-02 9.794E-04 515 VWBMOD06 16 % '80E-03 1.960E-04 516 VWN-MAN 01 16

• 60E-03 1.955E-04 517 VW/-RF-ER 2 i.200E-05 2.618E-05 518 WLIAV004LA 1 8.760E-03 7.153E-07 519 WLIAV055LA 1 8.760E-03 7.153E-07 520 WLOAV006LA 1 8.760E-03 7.153E-07 521 WLOAV057LA 1 8.760E-03 7.153E-07 522 ZANMOD01 45 8.400E-05 3.570E-06 523 ZANTR-2AHF 5 2.880E-05 4.280E-08 524 ZANTR-2BHF 2 2.880E-05 1.230E-08 525 ZO1DG001TM 561 4.600E 02 4.689E-04 526 ZOlMOD01 298 2.020E-02 1.509E-04 527 ZOlMOD04 23 1.250E-03 1.242E-06 528 ZO2DG002TM 512 4.600E-02 6.808E-04 529 ZO2 MOD 01 284 2.020E-02 2.711E-04 530 ZO2 MOD 03 2 1.000E-04 2.974E-07 531 ZO2 MOD 04 30 1.250E-03 8.848E-06 532 ZOX-BL-ES 9 6.000E-05 3.119E-07 i 533 ZOX-DG-DR 60 4.400E-04 3.966E-05 1 534 ZOX-DG-DS 42 2.800E-04 *

.535E-05 l e 535 ZOX-PD-ER 24 1.300E-04 4.414E-06  ;

536 ZOX-PD-ES 118 2.000E-03 1.932E-04 l CALCULATED CDF = 1.69E-07 1

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im.27-12 W westinghouse

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NRC REQUEST FOR ADDITIONAL INFORMATION  ;

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= -- 1 E ig Question: 100.28 )

I Please provide Chapters 55 (Sei 'n Evaluations) and 57 (Fire Evaluations) of the AP600 PRA. j 1

Responso:

Chapter 57, Internal Fire Analysis, has been provided with AP600 PRA Revision 8 (September 1996).

Chapter 55, Seismic Margins Analysis, was previously provided with the response to RAI 720.158. It is being revised to address other RAls. The updated version will be provided in early 1997.

SSAR/PRA Revision: None. I l

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J NRC REQUEST FOR ADDITIONAL INFORMATION ii!E M EFi

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Question: 100.29 )

Identify the most important structures, systems, and components (SSCs) relied upon to prevent and mitigate core damage during and following seismic events. Discuss the basis for establishing the importance of these SSCs.

Response: l l

The requested information can be found is the response to RAI 720.158. The seismic margin analysis is being l revised and will be included as Chapter 55 of the AP600 PRA in early 1997.

SSAR/PRA Revision: None.

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l 4 NRC REQUEST FOR ADDITIONAL INFORMATION mE =E 7 4 l

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Question: 100.30 l

l Provide an assessment of major contributors to risk from external events, and design alternatives considered and/or implemented by Westinghouse to reduce risk from each of these contributors.

Response

l Chapters 57 and 55 of the PRA discuss the internal fire analysis and seismic margins analysis, respectively. Chapter 58 of the PRA discusses other external events and the probability of an accident leading to severe consequences due to an external event. ,

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The AP600 SSAR (Chapter 2) also discusses the ability of the plant to withstand events such as high wind, seismic  !

events and external floods. I No design alternatives were considered or implemented on AP600 to reduce risk from external events as c result of a SAMDA evaluation.  !

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SSAR/PRA Revision: None.

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100.30-1 W-Westinghouse

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' I o l NRC REQUEST FOR ADDIT ONAL INFORMATION EE M q l Ouestion: 100.31 i

The AP600 design suggests several SAMDAs not yet considered that might prove cost effective. Evaluate and i discuss the following possible candidate design alternatives.

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1 Increased regulatory oversight of the most risk-significant non-safety SSCs 1 Improving the instrumentation and controls (quality of components, quality / maturity index of software)

Use of fan coolers (FCs) to remove fission products, and possibly upgrading the FCs and support systems to improve reliability  !

Addition of a non-safety grade in-containment spray system Increasing the thickness of the reactor cavity concrete to reduce the likelihood of containment failure by 1 cavity melt-through.

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Response

l Increased regulatory oversight of the most risk-signi6 cant nonsafety SSCs There is a program in place entitled Regulatory Treatment of Nonsafety-related Systems (RTNSS) which Westinghouse is implementing on AP600 with the NRC staff. This program evaluates the AP600 on a deterministic and probabilistic basis to determine if any further regulatory oversights is needed.

The cost associated with increased regulatory oversight is believed to be significant due to the associated administrative burdens ultimately placed on the nonsafety-related SSCs. As shown by the risk reduction of the previous design alternatives, due to the low core damage frequency of the AP600 design, severe accident mitigation design alternatives have very low risk reductions. Coupled with the significant cost, this option is not a viable SAMDA.

Improving the instrumentation and controls (ouality of components, quality / maturity index of software)

The AP600 instrumentation and controls is believed to be of sufficient quality as is evident by the low failure probability of the I&C components. As shown by the risk reduction of previous design alternatives, due to the low core damage frequency of the AP600 design, severe accident mitigation design alternatives have very low risk reductions. Coupled with the increased costs, this option is not a vaiable SAMDA.

Use of fan coolers (FCs) to remove fission products, and possibly upgrading the I Cs and support systems to improve reliability Containment fan coolers are included in the AP600 design. Finalization of the AP600 severe accident management guidance should include the use of the fan coolers in accident management strategies to remove containment energy, control fission products, and control hydrogen. These points are included in WCAP-13913. Revision 1. December 1996, " Framework for AP600 Severe Accident Management Guidance."

As seen with the evaluation of other design alternatives, upgrading the fan coolers design will result in an insignificant risk reduction (versus cost) due to the low core damage frequency of the AP600 design.

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• o NRC REQUEST FOR ADDITIONAL INFORMATION l ik

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i Addition of a non-$afety grade in-containment spray system ,

i Evaluation of a nonsafety-related in-containment spray system is included in the revised Appendix IB of the AP600 l 3 s SSAR (revision 11. February 28 1997). As shown by the risk reduction presented in SSAR Appendix IB, due to the low core damage frequency of the AP600 design, a severe accident mitigation design alternative such as a nonsafety-related spray system has very low risk reductions. Coupled with the large capital cost, this option is not ,

a viable SAMDA.

)

Increasing the thickness of the reactor cavity concrete to reduce the likelihood of containment failure by cavity melt-through Sinc: the reactor cavity flooding system provides a means to preclude CCI by maintaining the core debris in the I teactor vessel, there is little risk reduction to be gained for further design changes to address CCI. Additionally, ,

changes to the reactor cavity would result in a large cost. As with other alternatives, this is not a viable SAMDA. I l

SSAR/PRA Revision: None. l l

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100.31-2 W-Westinghouse