ML20125D589

Severe Accident Mitigation Design Alternative Evaluation

ML20125D589
Person / Time
Site:	05200003
Issue date:	12/15/1992
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
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ML20125D581	List: ET-NRC-92-3784, AP600 Conformance Assessment W/Alwr Utility Requirements Document,Vol Iii,Rev 3 ML20125D576 ML20125D589 ML20125D593
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NUDOCS 9212150311
Download: ML20125D589 (17)

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l ATTACHMENT 2 ET-NRC-92-3784 NSRA-APSI,92-0265 AP600 SEVERE ACCIDENT MITIGATION DESIGN ALTERNATIVE EVALUATION s

DECEMBER 15,1992

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1. INTRODUCTION AND GENERAL DESCRIPTION OF PLANT g tii[

m n APPENDIX 1B SEVERE ACCIDENT MITIGATION DESIGN ALTERNATIVES 18.

1.0 INTRODUCTION

in SECY-91-229; the NRC staf f recommends that serch accident midgation design alternatives be This report provides an evalranon of Sescre addresseo for certified designs i t single n.lemaking Accident hiitigation Design Alternauses (SAMDA) for process that would address loth tne 10 CFR 50.34 (f) the Wesunghouse APNO design. This esaluation is and NEPA considerations in the 10 CFR Part 52 design __

perfonned to evaluate whether or not the safety benefit certification rulemaking. SECY-91-229 further of the SAMDA outweighs the costs of meorporating the recommends that appheants for design cernfication S AMDA in the plant, and is conducted in accordance assess S Ah1DAs and the applicable decision rationalc as with applicable regulatory requirements as identified to why they will or will not benefit the safety of dieir below, designs. The Commission approved the staff recommendations in a memorandum date October 25, The Nauonal Environmental Policy Act (NEPA), 1991 (RefetCHCC 8)-

Section 102ACHiin requires, in p:ut, tnat 18.2.0

SUMMARY

.all apt icies of the Federal Government shall (C) include in every recommendation or report on An evaluation of candidate modificadons to the proposals for legislation and other major Federal APNX) design was conducted to evaluate the potential 4 actions significandy af fectmg die quality of the for such moi.hfications to provide significant and human environment, a detailed statement by the practical 'mprovements in the radiological risk profide of responsible official on (iii) attematives to the the APNM) design.

proposed action. The process used for identifying and seleenng candidate design attematises included a review ot 10 CFR 52.47(a)00 requires an applicant for design SAMDAs evaluated for other plant designs. Several certification to demonstrate SAMDA designs evahiated previously for other plants -

were excluded from the present evaluation because they

. compliance with any technically relevant portions have already been incorporated or otherwise atldressed of the Three Mile Island requirements set forth in in de APNk) design. These include:

10 CFR 50.34(0 .

- liydrogen ignition system A relevant requirement of 10 CFR 50.34(0 contamed in + Reactor emity Gooding system snbparagraph fl)O) requires the performance of = Reactor coolant pump seal coolmg

. Reactor coohtnt system depressurinttion a planthite specific probabilistic risk assessment, a Reactor sessel exterior cooling.

the aim of which is to seek such imprmements in the reliabihty of core and containment heat removal Additional design attematives were identified based systems as are tigmticant and practical and do not upon the results of the AP600 probabilistic risk impact excessively on the plant . assessment (Reference 1). Fourteen candidate design attematives were selected for further evaluation.

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1. INTRODUCTION AND GENERAL DESCRIPTION OF PLANT An esaluadou of each of these attematives was 6. Active high pressure safety injectmn system perfonned using a toundmg medaniology such diat the potenual benefit of each alternative is conservatively 7. Steam generator shell side passive heat removal maximized. As part of diis process, is was assumed that system cach SAh1DA perfonns beyond expectations and completely eliminates the severe accident sequenc:s that X. Steam :'encrator safety valve flow directed to in-the design attemative addresses. In addition, the capital containment refuelmg water stotage t;mk (IRWST) cost esumates for each ahernauve were intentionally biased on the low side to maximite the risk reduction 9 Increase steam pencrator secondary side pressure benefit. This approach maximucs the potendal benefits capacity associated with each alternative.

The results show that despite the significant 10. Secondary containment filtered ventilation conservatism employed in the evalcanon, none of the SAh1DAs evaluated provide risk reductions which are i1. IRWST discharge valve dnersificauon cost beneficial. The results also show diat esen a conceptual " ideal S AN1DA", one w hich reduces the total 12. Ex-sessel core catcher plant radiological risk to zero, would not be cost effectne. This is due pnmarily to die already low risk 13. High pressure containment design profile of the AiW W) design. which is approximately two onters of magnitude below existmg plants. 14. Diverse actuation system (DAS) improved reliability.

18.3.0 SELECTION OF SAMDAS A description of each design alternative evaluated Cambdate design alternatives were selected based for APM) is presented in Section 7.0 upon design n!ternatives evaluated for other plant designs Several design alternatises addressed in previous (References 2, 3, and 4) as well as suggestions from SAh1DA evaluadans for other plants were excluded AIVO design personnel. Addinonal candidate design from further evaluauon because the ahernatives are altematives were selected based upon an assessment of already incorporated into the AP6fo design. These the APo00 probabilistic risk assessment results, design features include:

Fourteen design alternatises were finally selected foi further evaluation. These fonrteen SAhlDAs are:

• Hydrogen ignibon system

• Reactor cavity thxxitng system

!. Chemical volume and control system (CVCS)

• Reactor coolant pump seal cooling uppraded to mitigate small LOCAs

• Reactor coolant system depressuritation

+ Reactor vessel exterior cooling.

2. Fihered containment sent 18.4.0 METHODOLOGY

3. Normal residual heat removal system (RilR) located inside containment The severe accident mingation design alternatives analysis employs a bounding metimdology such that the

4. Self-actuating containment isolation vahes benetit is conseivansely maximited and the capital cost is conservatisely minimized for each SAh1DA. The risk

5. Passhe containment spray IB-2 W- Westinohouse a

1. INTRODUCTION AND GEtJERAL DESCRIPTION OF PLANT reduction, capital cost estunates, and cost benetit estabhshed. For this evaluation. the nsk reduction is analysn meth(xts are docussed in this section. comerted to a capit;d benelit w hich can then be threctly coinpared with the capital costs.

18.4.1 Risk Reduction The benefit of each design rdternative is die reductmn of risk in tenns of whole Nxly inan rem per Risk for the purpne of this evaluation o the year received by the total population within a 50-mile probabihty of core damage for each accident ininator, radius of the APNK) phmt site. Consistent with previous multiphed by the consequences of the accident, SAh1DA evaluations 'md MRC regulatory an;dysis expiewed in tenus of man-rem per year. The total risk guidehnes, a value of Sl, XK) per offsite man-rsm n the sum of the inks from all the accidents. averted is used to convert man-rem per year to dollars The reducuon of rnk for emh SAh1DA is the per year. This value is intended to be the surrogate for ddlerence in nsk between the APNX) design and an all of fsite consequences includmg property damage and APNK) design with the design alternative incorporated. is referred to as the annual lesch /ed benefit.

It n assumed that each S Ah1DA works perfc(tly and The risk reduction reported as dothus per year is completely chminates the accident sequences that the then converted to a maximum capital benent which can design alternative addresses. This appnuch then be compared to the capital costs. The maximum consenatisely maximi /es the benefits associated with capital benefit is equal to the annual leseh/ed benefit each design alternative, and is not intended to imply that (dollars per year) divided by the annual leseli/ed fixed such a perfect design is p>ssible. The S AN1DA benefits charge rate.

are the reduction of risk in terms of whole body man. The annual leveli/cd fixed charge rate is determined rem per > car received by the total population withm a from a number of financial factors. These factors are 50 mile sadius of the APNK) plant sue, pisen in Table 1B A-1 and are taken from the EPRI Each design ahemative is evaluated based on how Tectuncal Assessment Guide (Peference 6). The u af fech each of .he release categories in the APou) equations used to detennine the annual leveh/ed fixed probabthstic nsk assessment. charge rate are f rom the Nnclear Energy Cost Data Base (Reference 7). For a nuclear plant economic life of N) 18.4.2 Capital Cost Estimates years and a tax hfe of 15 years, the annual levehm fixed charge rate is 15A percent in current U.S. dolktrs The capital cost estimates for each S Ah1DA are (with inflationL intentionally biased on the low side to inautmze the rnk reducuon benefit. All reasonably anticipated one. time 18.5.0 PRA RELEASE CATEGORIES capnal costs are accounted for in the estimates. Actual plant costs are expected to be higher since the cost To assess each design alternatis e's reduction of risk, esumates do not melu& the cost of testmg and the potential for each alternative to reduce the frequency m;untenance or the ermeenny cost to design the of occurrence or the consequence of each release

! alternauve to ht mto the APNXL The cost estimates are category is assessed. The steps involved in creating the l based on 1992 U.S. dollars. APN)O release categones are brictly docussed in this i section.

18.4.3 Cost Benefit Analysis The AP600 Lael I plant event trees identify the sequences that lead to core damage. Sequences that l

l In order to compare the risk reduction, which is have sunitar charactensuch are grouped together mto I

reported in man-rem per year, to the capital costs, which accident subclasses for the containment system analysis.

are reported in dollars, a conunon set of uma must be N WOSlinch0US0 u l

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). INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

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The characteristics considered in the bmning of the pLmt . Prevention of sessel failure and/or core melt arrest event sequences into the accident classes are as follows:

• Coohng of ex-sessel debris.

• The mitiating event type - such as loss of coolant acchient or anticipited transient without set:un. The endpoints of the containment event trees paths leadmg to core d;unage are grouped into appropriate source term categories based on similar fission product releases. Different

. The prunary system pressure at the tune of core endpomts for the AP600 plant are delined, depending on damage (htyh or low) the type of containment failure (bypass, isalation failure, or late overpressure due to core-concrete interaction). If

. Tuning of core d.unage (early or late) the containment thies not fail, the availability of the passise containment coohng system water has a strong

• Containment integrity at the time of core d'unage influence on the containment pressure, and therefore is Ontact or impairedi used to determine the release category. The source term for a representative sequence in each important accident

. Availability of safety systems after core damage class is calculated with the Malular Accident Analysis Drogram Version 4.0 (M AAP 4.0) cale.

. Disposinon of water in the contaimnent at the time The release categories for the AP600 are defined as of core damage follows:

• Containment pressure and temperature at the Ome of . OK -- release associated with the leakage from a core damage. containment with ptssive containment cooling water available, Containment event trees for each of the significant accident subclasses are developed and discussed in the . OKP - release associated with tl.e leakage from a AIW) piobabilistic risk assessment (Reference 1). The containment with passive containment coming water contatnment event tree analysis considers both the not asailable, contauunent and associated auxiliary systems, in panicular, the following items are considered: . CC -- release associated with the leakage from a containment that is pressurized with noticondensible

. Contaimnent isolation system gases pencrated by core-concrete interaction,

• Passive containmem cooling system . Cl -- release associated with the leakage from a containment that is bypassed or has not been

. In-containment refuehng water storage tank injection isolated (impaired).

. EUvessel debns conhng. The following sections present a brief description of the accident sequences from the probabilistic risk The functions accomplished by these systems are: assessment which represent each AP600 release i category.

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• Maintenance of containment integrity and/or the l reduction of containment pressure 18-4 W Westirloh0Use

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1. INTRODUCTION AtJD GENERAL DESCRIPTION OF PLANT

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f 18.5.1 Release Category OK for accident m:magement or use of alternative methals of wetting the containmem shell.

The r epresentative sequence for une OK release Because of the induence of water in the category has an imtiating event which is a Linch containment there is essentially no difference in fission diameter loss of coolant accident wah a fadure of the in. pn> duct release if the debris remains in the vessel or is containment refuehng water storage tank check vahes released to die conttinment. The final release frac 6ons, and nonnal RilR injection. Core d;uuage begins 2.0 at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> alter corc (kunage, are presented in Table hours into the accident. The in-contunment refueliny 11(5-1. The OKP release category frequency is 5.6 x water storage tank is not draineti into the containment 10 per year, cauty to proside external coohny to the reactor vessel, so the core debris is not mamtained in the vessel. The 18.5,3 Release Category CC sessel imls at 11.X hours, and the molten core drams into the contanunent at low pressure. The debris is quenched The representative sequence for release category CC and cooled in the reactor cavity, so there is no is imtiated by a 4-inch diameter loss of coohmt accident significant ex-tessel release. The passise containment with a failure of the in-containment refueling ,/ater cooling system and hydro"en igniters are available, and storage tank check valves, nonnal lulR injection, and containment pressure remain- below design pressure. the passive containment cooling system water now.

The nnal 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 Three out of the four core makeup tanks and dama.ee, are presented in hble,1 B.51. The OK reletse accumulatars are available, The in-conttinment category irequency is 2.5 x 10 per year, refuehng water storage tank is not drhined into the contamment cavity to provide external cooling to the 18.5.2 Release Category OKP reactor vessel, so the core debris is not maintained in the sessel. The core tkunape begins at 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. The The representatne sequence for release category sessel tails at 11.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, and the molten core drains l

! OKP is mined by a 4-inch diameter loss of coolant into the cavity at low pre %ure. The cavity dries out accident with failures of the m-containment refuchng because the water from the available core makeup tanks wa?er storage Umk check valves, rormal RilR injection, and accumulators is trapped as steam or m water holdup and passive containmem cooheg sy, tem coohng water, volumes. Passive containment coohng system Four out oi font core makeup us and accumulators condensation does not keep up with the rate of boiloff are available. The m-containment refueling water f rom the debris bed. Core-concrete interaction creates l f orac' tank is not drained into the containment cavity to noncondensible gases that pressuri/c the conttimnent.

pros ide external coohng to the reactor sessel, so the core At 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after core damage, the pressure in the debus is not mamtained m the sessel. Core d.unage containment is essentially equal to design pressure. The occurs at 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and vessel failure occurs m 15.8 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 hours. The debris is quenched and coolable in the presented in Table 1 B,.5-1. The CC release category reactor casity because all of the water holdup volumes frequency is 7:i x 10, per year.

are full and the condetnation from the passis e containment cooling system shell returns water to the 18,5.4 Release Category Cl '

contanonent sump. The containment pressure is elevated over the long tenn, but it equilibrates at a pressure well The representttive sequence for rekase category Cl l beks the ultunate capacity of the shell, so contanunent is initiated by a loss of feedwater to the steam generators integnty is inaintained. No cretht is taken in the analysis and the failure of the passive residual heat removal and automatic depressuntation systems. The containment W Westifl8h0USe

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1. INTRODUCTION AND GENERAL DESCRIPTION OF PLANT Revision: 0 E

does not isolate. The containment isolation failure is 18.7.0 SAMDA DESCRIPTION AND Effective: 12/15/92 modekd as the ladore of one purge ime.

BENEFIT The core ternperature exceeds 2500"K at 4.2 hourt The operator dumps in-containment refueling water This sechon describes each S Ah1DA and the benelit storage tank wmcr into the cavity on a high-high core ded due m the nM0cmion, in the evaluation of cut temperature. The water sutiounds and cools the the risk reduction benefit, each SAh1DA is assumed to reactor vessel, preventing sessel failure. The reactor operate perfectly with 1009 efficiency, without failure coolant system hot icy mptures due to the high of supporting systems. A perfect SAh1DA reduces the temperature and pressure m the reactor coolan system.

Irepency of accident sequences which it addresses to The remainder of the core melts and falls into the lower zero. This is censervative as it maxituires the benefit of head. Fission product released Nto the contairhnent can each design alternative. The SAh1DA will reduce the be directly transported to the enviromnent. The final M by lowering the frequency, attenuating the release, j release fractions, at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> alter core damage, are or both. The benefit will be described in tenus of the presented in Table 1B;5-1. The CI release category base accident sequences and dose unich are affected by the frequency is 2.0 x 10 per year. llowever, because the S AN1DAs, as well as the overall risk reduction, treguency of excessive leakage, which exceeds the techmcal specification leakage, from the other release 18.7.1 Upgrade the CVCS for Small categories is lumped into the Cl release category, the overa" release category frequency is taken to be 3.0 x LOCAs 10" per year, The chemical, solume, and control system (CVCS)

I' '""endy capam of maintaining the reaaor =lant 1B.6.0 TOTAL POPULATION DOSE system (RCS) inventory to a level m which the core remains covered in the event of a very small (< 3/4" To assess the potential benetits associated with a (h:uneter break) loss of coolant accident (LOCAs)I This design ahemative, esumates are made of the twal offsite S ANID A involves upgrading the pumping capacity, and populahon dose resulting from each of the release i ne sizes of the CVCS system in order to be able to use

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categories 0.c., source terms) idenhned m Section 5.0. the system to keep the core covered during small (< 4" The h1ELCOR Accident Consequence Code System diameter breaks) LOCA accidents, as well (N1 ACCS). Version 1.3 (Reference 5) is utilized for this A perfect, upgraded CVCS system is assumed to analysis. The cale input is identical to the APW) prevent core damage in all the very small and small probabihsnc risk assessment, however the consequence LOCAs in each release category. The CVCS is assumed evaluated is the ef fectise whole body equivalent dose (50 to have perfect support systems (power supply, year committed), resulting from exposure dunng the component cooling) and to work in all situations 2nitial 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 damre, to regardless of the common cause failures or oler the total populauon within a 50-inile radhn of the plant. ss stems. This resuh3 in a total averted risk of Table 1B.6-1 present.s the estimated incan and 5$80 x IM man-rem per year.

methan doses in person sieverts (1 person-scivert equals lik) man-rems) for each release category. Table 1B.6-2 1B.7.2 Filtered Vent shows the 50-mile population dose nsk for each release cmegory as well as the wtal risk of 3.42 x 10' man-rein This SAh1DA consists of placing a filtered per year for the AiWM) plant.

containment vent and all associated piping and penetratioas into the AfwH) containment design. A Gltered sent added to the containment would prevent 1B-6 W Westingl100Se

1, IfJTRODUCTIOff AtJD GEtiERAL DESCRIPTIOt1 OF PLAtJT h

omte ment f ailure Irom slow prewurt/ation events by 1 B.7.3 Locate Normal RHR inside deprewon/ing the cont uninent through a litter into the Containment environment. Fibered sentmp af fech the source tenns from release caterones OKP and rT. This S AMDA consists of placing the entire normal The sequences in release category OKP hase no residual heat remosal (RIIR) system and piping inside water conhng of the contairunent shell and pressurve die the contauunent pressure boundary, locating the normal contamment due to decay heat ste;uning from debris in RllR inside the containment would pres ent containment the cauty or the R JS. Scquences in telease category bypass due to interf acing systern LOCAs (ISLOCA) of CC dry out the exoessel debris bed, and pressurue the the RifR system. In past probabilntic risk assessments contaimnent Irom non-condernible gas generation due to of current peneration nuclear power plants, the ISLOCA co,e-concrete interaction (CCl) m the reactor cavity. is the leadmg contnbutor of plant rnk because of large flowever, neither release category contains sequences m offsite consequences. A failure of the valves which which the contaimnent fads. Release category OKP isolate the low pressure RilR system from the high caws pressuri/c the containment, but the decay heatmg pressure RCS causes the RilR system to userpressurite and the heat removal f rom the dry PCS reach and fail, releasing RCS coolant nuoide the containment equilibrmin well before the pressure cueeds ultimate where it cannot be recovered for recirculation cooling of capacuy. Release category CC cases pressun/c the the core. The result is core damage and the direct containment slowly and are not pretheted to f ail the release of fhsion products outside the containment.

contanunent before four days af ter the imtianon of die in the APru), the RilR system is designed with a accident, prouding ample time for ad lioc accident lugher design pressure than the RIIR systems m current snanagement procedures to tenninate the CCI and pressuri/cd water reactors, and an adJitional isolation prevent contamment failure. In (=nh the OKP and CC vahe is prouded in the design, in the probabilntic risk telease ca:epones, the source tenn to the environment is awewment, no ISLOCAs contribute signincantly to the not much mme than the source tenn trom the OK core damaye lrequency of thc APh00 (reference l, Table release cateyory in w hich the containment remains below 7 1 ). Therclore, relocating the normal RilR system of the design prewmc over the long term. the Ap(m inside contaimnent w til provide virtually no Filtered senting of these sequences can be assumed risk reduction benent and will not be investigated finther to t case Itm of the noble pas liwlon products and in terms of cost.

approdmately 1.0 x 10 ' of the aerosol fiwion pnxtucts tassuming a decontanunation factor of 1000 for the 1B.7,4 SeIf- Actuating Containment partiu..atest The source tenns of the OKP and CC isolation Valves release catepones in wiuth the containment rem,uns intact are sigmticantly sm:dler than die expceted source Thn S AMDA consisb of improved containment tenn from hitered sentilatmn. Therefore, the fdtered isolanon provsions on all nonnally open contaimnent sent provides no benefit, and in fact provides a hability penetrations. The category of "nonnally open" is hmited to the AP600 design by increasing the residual nsk- to nonnally open pathways to the environment dunng Jhn design ahenutne o not analy/cd funher in tenns mwer and shutdown conditioin, excludmg closed of mst.

tems inside arid outside the watainment such as raal RiiR and component conhny. The design ahemative would be to add a self-actuatmg valve or enhance the existing unide containment isolation valve to proude for self actuation in the event that cont.imment Conditiorn slidlCattse of a sCsCre accid 0nt.

3 Westinghouse

1, INTRODUCTION AND GENERAL DESCRIPTION OF PLANT Revision: 0 "y Effective: 12/15/92 To evaluate the benefit of this SAh1DA, the frequency from a pLmt with passise s#ety systems to a pluit with of all contaimnent isoladon failures are subtracted from passive plus active systems and is not consistent with the Cl release category and are added to the OK release design objectives.

category and the nsk is requantified. This does not include mduced containment failures which occur at the 18.7.7 Steam Generator Shell-Side Heat time of die accident such as in cases of vessel rupture or Removal System anticipated transients without scram ( ATWS). The benefit results in an averted nsk of 1.13 x 10 ' man-rem This S Ah1DA consists of providing a passive safety per year. grade heat removal system to the secondary side of the steam generators. The system would provide closed 18.7.5 Passive Containment Sprays loop cooling of the secondary using natund circulation and stored water cooling, thus preventing a loss of This S AMO A im olves adding a passive safety grade primary heat sink in the esent of a loss of startup spray system :uid all a.ssociated piping and support feedwater and passive RHR heat exchanger. A perfect systems to the APo00 containment. A perfect secontLtry heat removal system would eliminate containment spray with perfect support systems is transients from each of the release categories. In order assumed to provide linion product scrubbing and release to evaluate 'he benefit of this S AMDA, the frequencies reduction in tiv event of a failure of containment of all the transient sequences is subtracted from the isohition. Further, sprays ensure water coverage of any overall frequency of each of the release categories and core debris in the containment, preventing core-co? crete the risk is requantified. The total risk averted is mteracuon. To evaluate the tv nefit of conminment 6.7 x 1(f man-rem per year.

sprays, the OKP and CC release category frequencies are added to the frequency of the OK release category, and 18.7.8 Direct Steam Generator Relief Flow a dose reduction of 100 is . runed to be applied to the to the IRWST Cl release category. This natts H a total averted risk of 3.39 x 10 ' man-rem per year. This SAMDA consists of providing all the piping and valves required for redirecting the flow from the 18.7.6 Active High Pressure Safety steam generatoi safety and rehef valves to the in-Injection System containment refueling water storage tank (IRWST). An ahernate, lower cost option of this S AMDA consists of This S AMDA consists of adding a safety grade redirecting only the first stage safety valve to the actis e high pressure safety injecuon (HPSD pump and all IRWST. This system would prevent or reduce lhsion associated piping and support systems to the APtui product release from bypassing the cont:dament in tne design. A perfect high pressure safety injection system event of a steam generator tube rupture (SGTR) event.

is assumed to prevent core melt for a!! transients and in order to evaluate the benefit from this S AMDA (both small medium and hirge LOCAs in each releae opuons), the frequencies of :dl the SGTR sequences are l

ccegory. Only excessive LOCA and ATWS are subtracted from the Cl release category frequency and

awumed to lead to core damage. Therefore, the added to the OK release category frequency, and the risk

, frequency of each release category is reduced by the is requantified. The total risk averted is (i.7 x ILF man-l trequencies of all the LOCAs and transients sequences tem per year.

! in she categories, and the risk is requantified. The l aserted risk is I.X6 x 10' man-rem per year. This S AMDA wouhl completely change the design approach W Westingt10tjse l

1. INTRODUCTION AND GENERAL DESCRIPTION OF PLANT u!" 'ib i

m. ..

1B.7.9 increased Steam Generator 1B.7.11 Diversify the IRWST Discharge Pressure Capability Valves This S Ah1DA consists of increasing the design This SAMDA consists of redesigning die in-pressure of the steam generator secomtiry side and contrunment refuchng water storage tank (IRWST) safety vahe set point to the deprec that a steam disch:upe valve cantipurahon from four check valves to generator tube mpture will not cause the secondary two check valves and two air-operated valves. Tais system salcty vahe to open. The design pressure would hge w'll reduce the frequency of core melt by have to he increase sufnciently such that the combined chmin ene the commor. cause failure of the IRWST heat capacity of the secondary system inventory and the injection. 'Io estunate the benent from this S AMDA, PRllR system could reduce the RCS temperature below the frequencies of all the release categories is reduced by T mfor the secondary design pressure. Although specific the contribution of IRWST mjection failure sequences, analysis would base to be performed, it h estimated that and the risk is requantined. The total risk aserted is the design pressure would hase to be increased several 8.33 x 10 5man-rem per year.

hundred psi. mike the system described in secuon 6.S.

this design would aho present the release of fission 1B.7.12 Ex-Vessel Core Catcher pnulucis which by passes the containment via the SGiM.

Therefore, the nsk reduction is also the same as that This SAMDA consists of designing a structure m quantified in secuan 0.X. The total risk averted is the containment cavity or using a special concrete or 6.7 x IU" man rem per year, coating which will inhibit core-concrete interaction (CCD, eu n il the debrL bed dr.cs out. A perfect core 1B.7.10 Secondary Containment Filtered catcher design wotdd prevent CCI enhrely, and the Ventilation benefit from the core catcher would be estimated by assummg that all of the sequences in the CC release This S AMDA conshts of prodding the middle and category would all result in OK releases. Therefore the lower annulus (helow the 135' 3" elevation) of the frequency of the CC release category is added to the OK secondary cor. crete containment with a passive aanulus release frequency and die risk is requantified. This 6her sptem to for liltration of elevated releases. The S AMDA results in sirtually no reduction in risk s'.nce passive filter system is operated by drawmg a partial the risk from die CI release category, which dominates sacuum on the middle annulus through charcoal and the plant risk is not reduced in any way by the ex-vessel llEPA litters. The partial vacuum is drawn by means of core catcher. Therefore, this S AMDA is not considered an eductor with motive now from compressed gas t;mLs. further.

The secondary contamment w onhl then reduce particulate fission pnxtuct release from the pathways from which 18.7.13 High Pressure Containment Design the malonty of the primary containment leakage is l predicted to occur. In order to evaluate the benent from This S AMDA design consists of using the massive such a system. the of fsite doses from the containment high pressure containment design in v hi,^h the design leakage release catepones, OK, OKP and CC, and the pressure of the containment is approximately 300 psi (20 excesshe leakage frequency comnbution to the Cl bar) for the APro) containment. The masdve release category are assumed to be zero, and the nsk is containment design has a passive containment cooling l requantined. The total risk averted is 1.14 x 10 ' man- feature inuch like the AP900 containment. Tt e high

! rem per yc m design pressure h considered only for prevent, i of I

containment failures due to severe accident phenomena n

1 W_ westinEhouse l

I 1, INTRODUCTION AND GENERAL DESCRIPTION OF PLANT such as ste.un explosions mid hydrogen detonation. A The remaming design alternatives are evaluated to perfect high pressure coutmument design would reduce detennine their cost bene 5t. The results of the the probabihty of containment failures, but would have retnamine severe accident in tigation design alternative:,

no reduct.on of the frequency or magnitude of the evaluation are summarited in lante lila 1. The first release troin an unnotated containment, The AiWH column identities the design alternatise for which e probabilistic risk assessment concluded that the APNO reduction in nsk was calculated. The second column is n not susceptible to containment failure due to sesere the total man-rem reductmn per year for the design acetdent phenomena. Smcc the AIWU probabihsuc risk alternatn e. The thi. I column is the capital benefit assessment predicts no os erpressure containment failures, calculated based on the reducuon in ns This v:due _

the lugh pressure contamment design, at best, provides represents the maximum amount of capital that could be a rnk reduction of urtually /cro, and thcretore w di not spent in order for the design alternative to be cost j be considered Nrther. beneticial. The next calumn is the estimated minimum capital costs for the alternatne. The fmal column 1B.7.14 increase ReliaHlity of Diverse terresents the net capital A nent. The net benent is Actuation System calculated by subiracting the capital cost from the capital benefit A negatise benetit is idenutico by the use of Thn S AMDA design consists of unproung the parenthcses.

reliabihty of the diverse actuation system (DAS) which Fise of the design alternatises evaluated in other actuates engineered safety teatures and allows the SAMDA analpes are meluded in the current AiWO operator to monitor the plant status. A perfectly rehable design. As the ? rtm plant core damage frequency is y DAS systern wouki reduce the frequency of the release approdmately two orders of magnitude lower than for categories by the cumulative f requencies of all sequences edsitug plants, the benents of design alternauves are in which DAS Imlure leads to core damage. In order to sery small. Four et the S AMDAs analyzed prosided no evaluate the benetit trom the DAS system upgrade, the benetit at all and the others analy/ed provide neghgible frequency of the DAS lailure are subtracted from the benetits.

release category f requencies and the nsk is requannfied. Assuming an additional design alternatise was The total rak averted a 7.lX x 10' man-rem per year, deseloped which provides a 100 percent reduction in userall plant risk, representing an averted risk of _

3 A2 x 10" man-tem per year, the capital benent only 18.8.0 RESULTS amounts to s22.20.

As thwussed in Secuon 7.0, tour desien alternatnes liccause of the small innial risk associated with the APNO, none of the severe accident mitigatior, design considered for the APo00 proside no benefit for alte ,uives are cost beneficial.

reducing tesidual of fsee risk. These attematises are:

. nnerea sent 18.

9.0 REFERENCES

. Locate the nonnal residual heat remosal system 1. "Simphtied Passise Adsanced Light Water Reactor inside comainment Plant Program . APNo Probabdistic Risk AssessmentJ Westinghouse Electne Corporation

. Euessel core catcher and ENEL DE-ACO3-90SFIM95, June 26,1992.

. Ihph pressure containment de ign.

IB-10 W-Westinch00se e

1. ItJTRODUCTION AND GENERAL DESCRIPTION OF PLANT Hi !iill F ii n

2. " Supplement to the Final Environmental Statement -

Limerick Generating Station, Units I and 2,"

Docket Nos. '3-352/353, August 1989

3. " Supplement to the Final Environmental Statement -

Comanche Peak Steam Electric Station, Units 1 and 2," Di=:Let Nos. 50-445/446, October 1989

4. "Syo a 80+ Design Alternatises Report," Docket No.52-002, April 1992.

5. "MELCOR Accident Consequence Code System (M ACCS) Users Guide," NUREG/CR-4691, SANDX6-1562, Volume 1,1990.

o. " Technical Aswssinent Guide," EPRI P-65S7-L, Volume 1 Revision 6, September 19X9

7. Nuclear Energy Cost Data Ilase, DOE /NE4x)95, U.S. Department of Energy, September 19SX.

N. "SECY-Yl-224 - Severe Accident Mitigation Design Alternatives for Certified Standard De>igns,"

USNRC Memorandum from Samuel ,1. Chilk to James M. Taylor, dated October 25,1991.

[ W85thgl100S8

1. INTRCDUCTION AND GENERAL SESCRIPTION OF PLANT TABLE 1B A-1

SUMMARY

OF ANNUAL LEVELIZED FIXED CHARGE PNiu ASSUMPTIONS

'13 pe of Securit) Value Ihscount Rate (liefore tax) II.5'A/yr Intlation rate 5.0'slyr Federal anti Staic Income Tax Rate 3 S O's i

l Insestinent Tax Credit 0.04 l

l'roperty Taxes and insurance 2.04 Tax Recovery Period 15 3 cars Cony , thok late 60 years Total I.estliied I'ixed Charge Rate 15.4 G 1P-12 W Westinghouse

1. INTRODUCTION AND GENERAL DESCRIPTION OF PLANT Table i11.5-1

SUMMARY

OF FISSION PRODUCT RELEASE FRACTIONS 24 HOURS AFTER CORE DAMAGE I

l OK OKl' CC CI Xe,Kr 4.2 x 10 ' l .0 x 10 ' 6.4 x 10 3.4 x 1(r' 7

Csl 5.6 x 10 ' 2.0 x 10 ' 7.9 x 10 3.7 x 10 2 TcO; 0.0 0.0 0.0 0.0 Sr0 3.2 x 10 " X.0 x 10 ' 4.9 x 10 " 6.7 x 10 '

M oO, 5.6 x 10' 9.6 x 10 ' 6.5 x 10' l .4 x 10

Csolt 5.X x 10 ' 2.0 x 10 ' 9.0 x 10' 3.7 x 10 2 llaO 2.9 x 10 ' 6.5 x 10 ' 4.2 x 10 4.8 x 10

La;O, 2.0 x 10 " 5.5 x 10 ' 3.1 x 10 ' 2.0 x 10' CcO 2 5.9 x 10 " 1.6 x ir ' l .1 x 10 7

2.8 x 10 '

Sb 1.0 x 10

• 4.X x 10 ' l .1 x 10'* l .1 x 10 '

Te2 0.0 0.0 0.0 0.0 UO; 0.0 0.0 0.0 i 0.0 Ficqueno 2.5 x 10 ' 5.6 x 10 " 7.6 x 10" 3.0 x 10 "

8-'3 W westingnouse

1. INTRODUCTION AND GENERAL DESCRIPTION OF PLANT R ._ , . , _

TABLE Al'Mm INI'INI ATED l'Ol'UI,a lion DONE ESTIh1ATES (EDEWBODY DOTES IN PERSON-SIEVERTS)

Dose (l'erson-Sieserts)

Rescase Category Distarne (31jtes) hiean Sledian OK 50 n.93 x 10 4.96 x 10 2 Cl 50 1.14 x 10" 7.51 x 10 2 2

CC 50 9.01 x 10 6.33 x 10 OKP 50 1.34 x 10~' l.02 x 10 '

l Notes: 1. Doses are tused on the 50 year conunitted 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 d.unage.

2. One person-sievert equals f(O rnan-rem.

18-14 W- WestinEhouse

1, INVRODUCTION AND GENERAL DESCRIPTION OF PLANT sis .

\

l TABLE 18.6-2 AlWM) liase Itisk (Whole llody l'opulation I)me to a 50 Mile Radius)

Release Frequency Mean Consequence Risk Category (yr~ ' ) ( rnan-rein) (inan-rem-yr ')

OK 2.5 x 10 ' 6.93 1.73 x 10' 7

i OKP 5.6 x 10 " 13.4 7.50 x 10 7 CC 7.6 x 10 "' 9.01 6.85 x 10' Cl 3.0 x 10 ' l14(o) 3.42 x IU '

Total Risk 3.42 x 10 '

I l l l

)

l l

I i

1 i

l l

l l

l l

i l

l W Westin7hou3e

, s

1. INTRODUCTION AND GENERAL DESCRIPTION OF PLANT

• Revision: 0 "g Effective: 12/15/92 TABLE IB.X-1

. APMM) S AMDA Ill3Of fl S k

Rhk Capital Capital Net Capital De<,ign Alternathe  !! eduction lienefit Cmt Itenefit (man rem per yr) ($) ($) ($)

t--

Upgrade CVCS for Small L(X'A 5.X0 x 10' <l 1,460JN K) ( 1,460,(K K))

Seif Actuating Cont:unment bolation Vah es 1.13 x 10' 7 60,(WX) (60JMK))

Passhe Containment Spray 3.39 x 10 ' 22 3,5(M),(KK) (3,5(M)JXX))

Acthe liigh Pressure Safety injection

_ System 1.X6 x 10 ' 12 20,(x10,0(M) (20fM M)JMW))

SG Shell Side lleat Removal 6.70 x 1(r4 4 1,180,000 (1,180f100)

SG Rehet Flow to IRWST 6.70 x 10" 4 560Jxx) (560fXt))

increased SG Pressure Capabihty 6.70 x 10 4 4 2,720/100 (2,720,0( K O Secondary Containment Ventilation with Filtration 1.14 x 10' 7 2JN K),(M R) (2,(H M)JXX))

S.33 x 10' <l 300fXK) (300/KX))

])hersity IRWST Vahes blore Rehable DAS/ DIS 7.18 x 10' 5 390JXM) (3903)o0)

Y, 1B-16 Yj WestinghDUSS

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __