ML20248L701
| ML20248L701 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 03/02/1998 |
| From: | Collins T NRC (Affiliation Not Assigned) |
| To: | Quay T NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9803240087 | |
| Download: ML20248L701 (15) | |
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fa,5-. m 1998 tnJ3 tU MEMORANDUM TO: Theodorek. QNa , re or StandarqigtggPgestprNgtate Division of Reactor Program Management NO FROM: Timothy E. Collins, Chief / original signed by/
Reactor Systems Branch Division of Systems Safety and Analysis
SUBJECT:
SRXB INPUT TO AP600 FSER, CHAPTER 20, GENERIC ISSUES Attached is a supplemented FSER Chapter 20, Generic issues, prepared by SRXB. All issues ;
for which SRXB has primary responsibility are resolved based on responses from Westinghouse '
to staff RAls. This SE supersedes the memo from Timothy Collins to Theodore R. Quay, dated January 9,1998.
If you have additional questions, please contact David Diec at 415-2834
Attachment:
As stated cc: G. Holahan/S Newberry (w/o attachment)
T. Collins (w/o attachment)
R. Caruso DISTRIBUTION:
.PDR D. Diec SRXB RJF A.Cubbage L. Lois - T. Attard G. Themas G. Hsii S. Sun l SRXB: SRXhSA SRXB:DS D RCARUSO TECOLLINS WB $#88 DOCUMENT NAME: G:\W6FSER1.C20 l l i
i 0Ff3 140001
/
9803240087 980302 3 PDR ADOCK 0520 C_______________ _ _ _ . _ .
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. Issue A-31: Residual Heat Removal Shutdown Requirements As discussed in NUREG-0933, Issue A-31 addressed the ability to transfer heat from the reactor to the environment after shutdown, which is an important safety function. It was resolved in 1978 with the issuance of SRP Section 5.4.7, " Residual Heat Removal (RHR) System."
l l The safe shutdown of a nuclear power plant following an accident not related to a LOCA has l typically been interpreted as achieving " hot standby" condition. The NRC has placed considerable emphasis on the hot-standby condition of a power plant in the event of an accident or other abnormal occurrence and, similariy, on long-term cooling, which is typically achieved by the residual heat removal (RHR) system. The RHR system starts to operate when the reactor l
coolant pressure and temperature are substantially lower than the hot-standby-condition values.
Even though it may generally be considered safe to maintain a reactor in hot-standby condition -
for a long time, experience shows that certain events have occurred that required eventual cooldown or long-term cooling until the RCS is cold enough for personnel to inspect the problem l
and repair it.
I
! In SSAR Section 1.9.4.2.2, Westinghouse stated that the AP600 design includes passive safety-I related decay heat removal systems that establish and maintain the plant in a safeshutdown condition following design-basis events and it is not necessary that these passive system achieve cold shutdown as defined in RG 1.139.
l The passive core cooling system is design to maintain plant safe-shutdown conditions indefinitely. Cold shutdown condition is necessary only to gain access to the reactor coolant j system for inspection, maintenance, or repair. For the AP600 design, cold shutdown conditions
! can be achieved using highly reliable, but non-safety-related systems, which have similar !
. redundancy as current generation safety-related systems and are supplied with ac power from either onsite or offsite sources. Passive core cooling capability is discussed in Section 6.3 of the SSAR.
l Westinghouse states that the passive residual heat removal system can achieve hot standby conditions immediately and can reduce the reactor coo! ant temperature to 215.6 'C (420 'F) within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The reactor pressure is controlled and cars be reduced to 1.72 MPa (250 psig).
The passive RHR system also provides a closed cooling system to maintain long-term cooling.
Therefore, the AP600 complies with GDC 34 by using a more reliable and simplified system for both hot standby and long-term coolHg modes, and it is not necessary that these passive systems achieve cold shutdown as defined by RG 1.139.
GDC 34 requires a residual heat removal system to be provided with suitable redundancy in components and features to assure that, with or without onsite or offsite power, it can accomplish its safety functions so that the specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. No definition is specified as the safe-shutdown condition for which the RHR system should accomplish. EPRI in the Utility Requirements Document for Passive ALWRs proposed that the safe-shutdown condition be defined as 215.6 "C (420 OF) for the passive ALWR designs. The staff has concluded that cold shutdown is not the only safe stable shutdown condition which can maintain the fuel and reactor pressure boundary within acceptable limits. In SECY-94-084, Section C, " Safe Shutdown Requirements," the staff recommended, and the Commission approved, that the EPRI's proposed 215.6 'C (420 OF) criteria or below, rather than the cold shutdown condition required by l RG 1.139, be accepted as a safe stable condition, which the passive RHR system must be capable of achieving and maintaining following non-LOCA events. This acceptance is predicated
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on an acceptable pat * ,e safety system performance and an acceptable resolution of the issue of regulatory treatment of non-safety systems (RTNSS). The SECY paper also states that the passive safety system capabilities can be demonstrated by appropriate evaluations during detailed design analyses, including:
(1) A safety analysis to demonstrate that the passive systems can bring the plant to a safe stable condition and maintain this condition, that no transients will result in the specified acceptable fuel design limits and pressure boundary design limit being violated, that no high-energy piping failure being initiated from this condition wi!l result in violation of 10 CFR 50.46 criteria; and (2) A Probabilistic reliability analysis, inc!uding events initiated from the safe-shutdown conditions, to ensure conformance with the safety goal guidelines. The PRA would also
' determine th.e reliability / availability missions of risk significant systems and components as a past of the effort for regulatory treatment of non-safety systems.
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Issue A-9: Anticipated Transient Without Scram As discussed in NUREG-0933, Issue A-9, addressed the issue of ensuring that the reactor can attain safe shutdown after incurring an anticipated transient with a failure of the reactor trip system (RTS). An anticipated transient without scram (ATWS) is an expected operational i occurrence (such as loss of feedwater, loss of condenser vacuum, or loss of offsite power (LOOP) to the reactor) that is accompanied by a failure of the RTS to shut down the reactor.
The acceptance criterion for the resolution of Issue A-9 are as follow:
o Compliance with the nitiption requi rment of 10 CFR 50.62(c)(1) that plant equipment must automatically initiate 6 nergs.. " faec . vater (EFW) and turbine trip under conditions indicative of an ATWS. This eq?pment must function reliably and must be diverse and independent from the RTS.
- Compliance with the preventioniequirement of 10 CFR 50-62 (c)(2) that the plant must have '
a scram system that is diverse and independent from the existing RTS.
In SSAR Section 1.9.4.2.2, Westinghouse stated that the AP600 design complies with the requirements in 10 CFR 50.62, " Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants," and a discussion of the j design features to address the probability of an ATWS is included in Sections 1.9.5 and 7.7 of the l SSAR.
Westinghouse indicated that the AP600 design complies with the requirements of 10 CFR 50.62 with I a diverse actuation system that includes the AMSAC (ATWS mitigation system actuation circuitry) protection features mandated by 10 CFR 50.62 by tripping the turbine and diversely actuating
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1 3 q selected engineered safeguards functions; lk;;ar, the AF000 de;;ga deee net es;ernetb lli inttiste-a-t P//,i but b;2:j be;sdee esterne;Z ;r,;;%;;ea ef the PRllR eee;bg eje:e,T.. The l AP000 dee:ga deee ae;, theiefere eeinp;i wi;h 10 Of'R 50.02.
There are other AP600 design features aimed at minimizing the probability of ATWS occurrence and j mitigating the consequences, as discussed in Sechon 1.9.5 of the SSAR. For the AP600 design with
! passive core coohng systems, the staff requires that an ATWS analysis be performed to demonstrate that its ATW8 response is consistent with that considered by the staffin its formulation of the 10 CFR 50.62 design requwoments for cunent plant designs. The applicant has provided (response to RAI 440.26) the analysis of a complete loss of normal feedwater without reactor trip, using the LOFTRAN code.'. The ;;e" hee re-;c:td a-jds-se: ;afein. "a eene..ab; r,e Ar ,00 AT/!O enet,;h, ,,h2h 7,e +;-p"-sa; he; eg.eed te pee,ide. Thh h an egen he a. The ues of the rR:lR H
e,;b.T M lbu ef ee;ein;;b AP/J b ;%; ;a epea ;;eT..
Therefere, b;ue A 0 b net ree;lad fer:he A"000 dee'gn T$iiIstidisiTessWiiiib(siW10FNischilli@iit%P9007~ "siBiWWiHNIHRIMs 009t156:62 req 4iromimes:17tisWetelisidin6sionWeilin isausV ~L WsERMptedKUM6eliteffi thereforetcorioludwthis:isshe resolved Issue A-17: Systems Interactions in Nuclear Power Plants j i
As discussed in NUREG-0933, Issue A-17 addressed concems regarding adverse systems ;
l Interactions (Asis) in nuclear power plants. Depending on how they propagate, Asis can be classified as functionally coupled, spatially coupled, and induced-human-intervention coupled.
As discussed... NUREG-1229, " Regulatory Analysis for Resolution of USI A-17," dated August 1989, and GL 89-18, " Resolution of Unresolved Safety issue A-17, Systems interactions in l Nuclear Power Plants," dated September 6,1989, issue A-17 concems ASls caused by water intrusion, intemal flooding, seismic events, and pipe ruptures.
A nuclear power plant comprises numerous structures, systems, and components (SSCS) that are designed, analyzed, and constructed using many different engineering disciplines. The degree of functional and physical integration of these SSCs into any single power plant may vary considerably. Concems have been raised about the adequacy of this functional and physical integration and coordination process. The issue A-17 program was initiated to integrate the areas of systetas interactions and consider viable attemJives for regulatory reqt.' ements to ensure that the Asis have been or will be minimized in operating plants and new plants. knen the framework of the program, the staff requested, as stated in NUREG-0933, that plant designers consider the operating experience discussed in GL 89-18 and use the Probabilistic risk assessment (PRA) required for future plants to identify the vulnerability and reduce Asis.
This issue identified the need t: Mvestigate the potential that unrecognized subtle dependencies, or systems interactions, amcq Atructures, systems, and components (SSCS) in a plant could lead to safety significant eventt in NUREG-1 174, intersystem dependencies are categorized based on the way they propagate into functionally-coupled, spatially-coupled, and induced human-intervention coupled systems interactions. The occurrence of an actual adverse systems interaction (ASI) or the existence of a potential ASI, as well as the potential overall safety impact, I
is a function of an individual plant's design and operational features. For AP600 with new or differently configured passive and active systems, a systematic search for ASIS is necessary.
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- Westinghouse submitted WCAP-14477, Revision 0,"The AP600 Adverse System interaction Evaluation Report," dated February 1996 for staff review and approve. The purpose of the report was to identify possible adverse interactions among safety-related systems and between safety-related and non-safety-related systems, and to evaluate the potential consequences of such interactions. The staff reviewed WCAP-14477 report and provided Westinghouse with comments and questions. Westinghouse subsequently addressed the staff's questions and comments and issued a revision to the WCAP-14477 report. The staff reviews this issue as part of the regulatory treatment of non-safety systems (RTNSS) and has documented its review in Chapter 22 of the FSER.
' The staff concludes that Westinghouse has adequately assessed possible adverse systems interactions and their potential consequences in WCAP-1447, revision- 1. In addition, the staff has conducted confirmatory testing involving potential systems interactions, and has performed analyses of selected accident scenarios in which nonsafety and/or safety systems could interact.
Both the confirmatory tests and analyses showed that potential systems interactions did not have significant adverse effects on overall safety performance. Additionally, no additional unanticipated adverse systems interactions were observed. Ols 20.2-5 and 20.2-6 are closed.
This issue is considered closed.
Issue A-26: Reactor Vessel Pressure Transient Protection Since 1972, there have been many reported pressure transients which have exceeded the pressure-temperature limits specified in technical specifications (TS) for PWRS. The majority of these events occurred at relatively low reactor vessel temperatures at which the material has less toughness and is more susceptible to failure through brittle fracture. This is issue A-26 in NUREG-0933 which was resolved with the issuance of SRP Section 5.2.2, " Overpressure Protection." Applicants for cps and operating licenses were requested to design an overpressure -
protection system for light-water reactors (LWRS) following the guidance provided in SRP Section 5.2.2.
In its May 28,1993, letter, Westinghouse stated that the AP600 design conforms to the criteria in Branch Technical Position (BTP) RSB 5-2, "Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures," of SRP Section 5.2.2.
The pressurizer it sized to secommodate most pressure transients and over pressure protection for the RCS is provided by either the pressurizer safety valves or the normal residual heat removal relief valves, as described in Section 5.2.2 of the SSAR.
The staff concludes that AP600 design satisfies the BTP RSB 5-2 requirements and therefore, considers this issue closed.
Issue B-22: LWR Fuel Westinghouse identified in Table 1.9-2, of its May 28,1993, letter that it considered Issue B-22 relevant to the AP600 design; however, this issue is not required for the AP600 design to meet 52.47(a)(1)(ii) or (iv).
As discussed in NUREG-0933, issue B-22 addressed the staff concems that individual reactor fuel rods sometimes failed during normal operations and many fuel rods are expected to fail during severe core accidents. Failure of fuel rods results in radioactive releases within a plant
i 5
and is a potential source of release to the public. The resolution of this issue was to ensure that these fuel failures did not result in unacceptable releases to the public. Several problems were
' identified in the staff effort to improve the predictability of fuel perfomiance and these were addressed in the revision to SRP Section 4.2, " Fuel System Design," in 1981. The staff concluded that the then existing requirements on fuel were adequate to ensure continued low fuel defect rates and additional requirements would not significantly decrease the number of fuel defects. This issue was then dropped from further consideration.
Westinghouse stated the AP600 reactor core design complies with SRP Section 4.2 and the discussion on the fuel system design is in Section 4.2 of the SSAR.
The staff has completed its review of the VANTAGE-SH fuel for the AP600 design. The details of fuel design and acceptance criteria are discussed in Section 4.2 of the final safety analysis
- report. The staff concludes that Westinghouse has satisfactorily resolved all questions raised during the staff review of the issue, and therefore, the staff considers this issue resolved. The Open item 4.2.8-1 is closed.
Issue C-4: Statistical Methods for ECCS Analysis As discussed in NUREG-0933, Issue C-4 addressed the statistical methods used for perfor-mance evaluation of the ECCS during a LOCA. In accordance with the requirements of 10 CFR 50.46 as amended on September 16,1988, the NRC requires that the LOCA analyses for license applications use either the 10 CFR Part 50 (Appendix K) evaluation models or the statistical (realistic) models, including the uncertainty of calculation in the adverse direction. The realistic models must be supported by applicable experimental data. Uncertainties in the realistic models and input must be identified and assessed so that uncertainty in the calculated results can be estimated.
In SSAR Section 1.9.4.2.2, Westinghouse stated the AP600 methodology applied for LOCA l analysis is discussed in SSAR Chapter 15.
Appendix K of 10 CFR Part 50 specifies the requirements for LWR ECCS analysis, which call for specific conservatism to be applied to certain models and correlations used in the analysis to account for data uncerisinties at the time Appendix K was written. USI C-4 addresses NRC development of a statistical assessment of the uncertain level of the peak cladding temperature limit. In 1988,10 CFR 50.46, " Acceptance Criteria for ECCS for Light Water Nuclear Power Reactors," was revised to allow the realistic ECCS evaluation model, in addition to the evaluation model conforming to the Appendix K requirements. This BE evaluation model will use analytical technique realistically describing the behavior of the reactor system during a LOCA, with comparisons to applicable experimental data. The realistic evaluation model must identify and account for uncertainties in the analysis method and inputs so that when the calculated ECCS cooling performance is compared to the acceptance criteria, there is a high level of probability l that the criteria would not be exceeded.
As described in SSAR Chapter 15, computer codes WCOBRA/ TRAC and NOTRUMP, respectively, are used for the large- and small-break LOCA analyses. WCOBRA/ TRAC is a realistic code, and the uncertainties will be included in the analysis. NOTRUMP is a code using the Appendix K requirements.
Issue C-4 is closed.
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- Issue C-5: Decay Heat Update.
- As discussed in NUREG-0933, Issue C-5, addressed the specific decay heat models for the I LOCA analysis models. In accordance with the requirements of 10 CFR 50.46 as amended on
' September 16,1988, the LOCA analyses for license applications should use either the 10 CFR Part 50 (Appendix K) models, or the realistic models supported by applicable experimental data and including uncertainty of calculation in the adverse direction. When Appendix K models are used, the decay heat generation function should be based on ANS 5.0, " Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors," plus a 20-percent uncertainty factor. . When_ realistic models are used, the decay heat function in ANS 5.1, " Decay Heat Power ;
in Light Water Reactors,"is acceptable for licensing applications.
In SSAR Section 1.9.4.2.2, Westinghouse stated that the large-break LOCA analyses for the AP600 design, discussed in Section 15.6.5 of the SSAR, used the decay heat model identified in the 1979 ANSI 5.1 standard.
l This issue involved following the work of research groups in determining best estimate j decay heat data and associated uncertainties for use in LOCA calculations, q Appendix K of 10 CFR Part 50 requires the use of 1971 ANS Standard, " Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors," times 1.2 be used forthe heat generation rates from the radioactive decay of fission products in the ECCS calculation.
]
The staff has determined that the 1979 ANSI 5.1 is technically acceptable and has allowed this .
use in the realistic evaluation model. For the AP600 application, the 1971 ANS decay heat model and the 1979 ANSI decay heat model are used in NOTRUMP and WCOBRA/ TRAC, respectively, for small- and large-break LOCAS. The staff has completed'end documented its review of WCOBRA/ TRAC and NOTRUMP in Chapter 15 of the AP600 FSER. The staff considers issue C-5 closed.
3
- Issue C-6: LOCA Heat Sources As discussed in NUREG-0933, Issue C-6 addressed the issue identified in NUREG-0471 which
' involved staff evaluations of vendors' data and approaches for determining LOCA heat sources and the need for developing staff positions. The contributors to LOCA heat sources, along with their associated uncertainties and the manner in which they are combined, have an impact on LOCA calculations, the staff informed the Commission in SECY-83-472, " Emergency Core Cooling System Analysis Methods," November 17,1983, that statistical combination of LOCA heat sources would be allowed to justify the relaxation of non-required conservatism in emergency core cooling system (ECCS) evaluation models.
In SSAR Section 1.9.4.2.2, Westinghouse stated that the discussion of LOCA heat sources for the AP600 design is included in Section 15.6.5 of the SSAR.
The staff has completed and documented its review of WCOBRA/ TRAC and NOTRUMP in l Chapter 15 of the AP600 FSER. The staff considers issue C-6 closed. j l
- Issue 22: Inadvertent Boron Dilution Events l
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l As discussed in NUREG-0933, issue 22 addressed the possibility of core criticality during cold
, shutdown conditions from inadvertent boron dilution events. Although this issue was resolved with no new requirements, the acceptance criterion is that plants shall minimize the co-
- sequences of such events by meeting SRP Section 15.4.6, " Chemical and Volume Control !
System Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant '
(PWR)." Specifically, the plant shall respond in such a way that the criteria regarding fuel damage and system pressure are met, and the dilution transient is terminated before the l
shutdown margin is eliminated. lf operator action is required to terminate the transient, redundant alarms must be in place and the following minimum time intervals must be available between when an alarm announces an unplanned dilution and when shutdown margin 's lost: 1
- c e during refueling (Mode 6)- 30 minutes
- e during all other operating modes - 15 minutes l l
1 l Section 15.4.6 of the SSAR provides a safety analysis which demonstrates that redundant
{
l' alarms are available to enable operators to detect and terminate an inadvertent boron dilution i event within the above required time intervals, before shutdown margin is lost, i l In addition to the events in this issue, the staff has identified the following two boron dilution ;
scenarios where a deborated water slug may accumulate in the RCS and a restart of the RCPs '
l will cause this slug to pass through the core resulting in criticality or a power excursion:
l e The first scenario occurs during a plant startup when the reactor is deborated as part of startup procedures. A loss of offsite power will result in tripping the RCPs and charging pump. The subsequent startup of the diesel generator will restart the charging pump and '
l cause the accumulation of deborated water in the reactor lower plenum. The RCP restart with recovery of offsite power will cause this deborated water to pass through the core.
e The second scenario is related to transients or accidents, such as a small break LOCA
,. with heat removal by reflux condensation natural circulation that may result in an
! accumulation of deborated water in the RCS loop. This water will pass through the core with an inadvertent restart of the RCPS.
l The staff has completed and documented its review of inadvertent boron dilution issue in
! Section 15.2.6.5.4 of the staff FSER. The staff considei. Issue-22 closed. The J oen item i
20.3-2 is closed.
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ofths;URS: JSiniii%4:RC8imessurnY evaluegelittilnedeT~ . E' " itis y.
' ~
g 7.. j lncludes ^
1.fspeciscWMthstlproventsWet:006SMiiEisibeis
L _ . . . provide:Weessoni3iieiinwhat preanasiRBesip l pressp6 tenki:arietterretructums:eise ur69iintedWwatreosphhie:ws. ~
- pressute - . seecs ts.n...i.nAis.g..t'.T, . ..bu,3 7*)ntertnoing sistemio(subsyseenistierbeiiiloot(ifil.e.3.My,.y.;.t~"
MnO'M,. ,. th, c.; cur erJ8,.gy ;.je/...A.,ctoic c,.;.on.
/a .M 4, - g. >, -
y . c 4 lis.i.'.demtioiG.T..
hii,ws...y.l.lmt, ys ga. , .
c.s d 3.,,.,.,3.,,.;.;9,,w..
. ,- . . .. y , .
Isolatiordild1TtWstaffeisiustlohVthese systems followil (sh, Normal.,Res,id.,.W ua eai.j, ern._~.,S,yste ovel mi Thsp6rtiossRths;RNSlfdirHlbs RCSWtlis contaihifhentliblitlisiiiilGiiiRClVi@"iintsidi
( containment:MaesignidKtli6 RCSl operating ihissureldhdMportions:M ?M#is ischarge line deithetM.
civ a.nd ftC85
~
,4 upst.y,_sm_#.thGThe pressure.a
. - calshaft mechani smal;ofCMar.e
- de.,signe.the.ltNSLifM.-
pus ~Nd~ l.M,m ttiniURii3
! presi6relitsoo:pistgZeiwonlfportion Subsn6tionV3:2 of WCAP;14425^ discuss 6sittie*difficultiesfoidesigninithiRfWpun of thiRN8:havine~iUR8:foweslthihilhKMC8)isis full;RC8loperatiridWsesbrQA fundanssotal'pfobismLisfitst?si$gls:6f;sselfiij@sh] sit 6sisdd[
! RCSprissuni::Will likelp.tiaviiIsbnormallifast:siiisar'sfMseal facesTduring"ri6hnefplahl operatiotundWssarpr6ssais5This inemissdinese:arnormarsisiht:opeisne co6delihiii:550id 4.a _< . .
..~yyg
. - ,.n :,; s r :.a g.,
c y _
y.r .v . g x o. . .
yygg v.w,ven n l
pressure %Use.,. of hig,.,.w .,:n,nh pressure._seses will als. ,.go require more frequent , maintenance during _~.. vnnorma a :~.:.vana..cl operation [ThsreforeDtlijmpracticalLio?deaisifia"asilliestilould maintalriNftCWiressuti boundar# inh:no:leika s sianolis:bperatisatisfacsersy::str=NeonomihmirAep00 RNS"pumphechn61 cal siehl lidsilined.16,minimizsNamounh)OpekeDil[ exposed liNil SC8 pressure $X6;tNElstudfdriths; Onli8ssis Nuclear Poes/ Station decay;tselitism hal purpp;;sigilmin&disid6pessure%f.450*psisl found thit tsiotating isiialiwould ^ maintain 11ti stmetumOntegrflylst'siressbrsiiriexces*s^of 2500^pilTahdWmechanicatseilidENlwithsta6(i l isrssssel6fi1200 tom 2501si)lth6ut leaking $Ths,AP600ftNS; pump. mechanical:biisili~sihi!!si
, v,- DHR'pu ps,.l'b. u.t~i..t.,s'de. sign,<'pr.v,ss..u.re. ce is..-*twice as higheThe AP600;RN8 pump t.ai als'/.oth...e Da.4 7*.is'B.e..s. s..j,e,,.
leakage ffrom s.3-r..; hing:.,m. t'he,,MW@-
E -s -
s - -
<-ie
.a a sc -*N< - < - -* .v.? -' ' ' -
has aLdisaster t u's ;that
, limits.the the pumpM.'s--within' i capab'il'itesvs)f the normal .niskebpTsystsm iricaisWeatastiophicimechanical:seil failWelt.eaksie:cahbii i
controlled With the seal'leskoff lins'robted 16a floor drain that is%Qted toWauxilisr9 buildifid sumpd experise_ofThiiLisinorel Astrrial-condition faiorablelsliability3ih'an'a r sealLspeSially designed Ibfifull;
10 80tiisillhIsiWXUX6f)SSAR'dlibisishithiliTAP800'deslisliiiiiiidhiitWiERN8"dijiiitliHij aimed stemphoing th(meshood afiin intersystem LOPM.. 0pituliheieNong$mAlgueW Meik ido8%gelpeppef.aledMNelied ][, i e anearmdv. sh e mism m.: .~~
kuctieriflhitK9tildecontainmentishichleidesignedW ~~ -
proteollehjiyhe;RCataridjii40educithsilik4.ne sidsyv eiso m a m. dii6h m es mis m u ssset s d -
tegig9 Meet
~~~
NioVs;insidigibi:containmenfiire ireertocked tiprovent '
ackluefk.' Cg~
.f .
gggg heidifeontainsiiiiallefwehe,' .
~~~ ]
- Iom assuens holdup tenksN>coventTE:a N: -
_Jaisiingusemigente theti ylh(ihreis checit valvedishdMiso6dr<perage(Weald ( talMnths96W pressure MRN8J Ali3EthiiRN8TdisW16clGassVTnitsisshiiti6h'chihse(tfirinidioWiiWiiniUNs W~eiiiEKRNs pumpLou_cti_on operatorto>. a $mdition'oflibe9h_d.'aih. ' ".pr. essbre RC8 pressure that could ala,rm .
hi. p_rovided.<in.~6_5_.
eventuaikancoed m ..
m e_lN*con the niWdesigr.c pressure!iWhistsisperated pressuieLisolation valves'iniidinive iemote posillunfiridicatis)hiiid the:'mainloontrol roomEin addition, these pressure isoleSon vehesitiikpedfled IW88ARTibis 3.916 to bersubhet t6 technical KM6M>eo biwlthiri limitsYithl_ specinostion:LCoW63vhich RCSPN eak tssung haccordance Wthisurvoulence stepirement The:inaltdoncludss that thelRN8;desighl meets llbshquirementsM8ECL900j6?
q (bEChshil65[di6dR61GhiiiControl SyWisiTWlakeUd systshii!
sG6iWil6K'exe:60hs:ssAnisiovidsirrastaniird'dilaoites5KWeiiiraibliRTt0Hiim6547shiil .
operationsWthe chemicalland. volume control system l(CVS)3TherpdrificatiosfD6iiifpathMMid CV8 hii shigh@reissure 61osed-loop'deisigri which is'hntirely withiritheisontainmerQThi potentist 6sntdbstors 16'siilSt.OCA are ths" portions 6f the'CVSlcicated outsid('iheMii5F6iirit]
~
- ~
i.elthe istdown:lirisKthil liquid radwasts systeniland;ths?piskaup?s sishi$
Thi CValisaldisp'paiWp@persWl6termittentlMrdakeWIRRCS:Isikage3Wpui%psisisd and htop automanedviwhatthe pmss6rtser ieven maches ew bosom and sie topw viinormai Isses. barst, voipectisetMThshakeup pumpitaks sucekm:aboChhher:ess:ticataneid tsntMths domineressedwaim'atoracitanK(owsT)tsne inket inedsisi:cva;'burmesseWiliidiristum s' tream pootar@he#woup: nne nom en makeup pump discherse tese:nes hwa demon iusa gisatif eiidr3qest to the RC8'dssign' pres'
~
s ureiHowevedthiipusip sucildnllirii pipinfidd asiociated corhponentshave i design pressureRt 50)slOiidth: thaCUBS leis,thiiihths.RCs operating pressufe3
~
SUbsecti6h~3 3Yof ths.WCAP?14425'c6niihdslthititisnbiWactidiblil6"dsiidsthiil6M pressure porbons:of the: makeup s06 Hon pipingLto highier design"piessure.ilt is n6fpracti6 ibis 16 haVe a high"dssigh pressure for large tanks such"as the~ boric acid tank [which'ireisnted to::t.he l aimospheri[als 'well as thel piping directly c6nnected tolthess[ atmospheric tanki'Opjsjttye first
~
isolation italve&The suction lines each contain 'a check valve that separates the suction piping fron{a largialm6 spheric'tanklThese ch'eck valves"are. designed to'open'6n loddiffsiential
._____ _________ ________- -_-____ _ l
I -
11 Niliit
$sisiIire'75ifulhisiiUiihightedde56yll6lIsilGWsGid65*dibi*diidsilsiuliefliifsiiiWii@
the,louMWinestse portionsMthe' piping'froniloverppsesure illi$iefeuentlef ' ;dieck thi:dischsuggitneorthermafaxpension'in'6sseiefslossefaudn100W ~ tedves g gg g g.gg g g l
imhtchlg'deelensanharidle
'are radioactive nuidssindltneir[losiishnenitoredWaeno instrumentatiotG
- 3. ;..-;9./ W,-,.,.~v..i /. ' q >. sw 9, e ..? g, .u.,;.
'.~.u
'.-/,.</..#y,...
. ~
BhltWiEE. pump,dschargeside:cocurs$flieniil(Ehlgh@nesisis ~ li6ei l Milif( MMisillven0ilfL ._.Z !
inldeM$
i p' umps MFME!M(5nefoontalefinent
~~ .thejnaluissi tertninele~
terninalitisi[tCLOCANeddson7thipuriscessei)66dinlettsidosos velvam iippuld- "ts infosed elete(looselion'vneves. 'Wii " l i unlii'sateisistus;acidati6rfistonalMh6ssimultiple;Tssfetp<%946 18LOCA in thslinskesp?suok)6116eEAs :speclRed in Tabis W1h688410 l Inlofstop3falves7 arid inelurificatioh'retum fini;;Wselve"shditseisk valve 7spe,. fileif i
testingEThusi^stopNahles~are~provided;iiirittip6eition.irnNoellon liiith6lconllet teonis$1nedditi6ii!
l lise;CIVs~alsoL haVs thei'capatWilty for leikasstihd7h6d shi[psovidedeilth]iihlsu l the makesp:sstion'in the ;s6ntr61l room:st;sil tirsis&Thslsinff flhdsllist protsctio positiorilsdi , .
~
thelintentRSECy-90-016.lSLOCNp6sitl6N l (6KCMddsh1LidjuldAidweste;5psterni niii"cValinil5E016s:2666s6tsfaihihigh:EskiGis'EGiin6 isis':16spTsiildssihiiiladiinG Imenedletelydownstream:bf thisilconnehtion his high pressurehnul0;btageleidosn*erificeTwhidh reduces pressure'in thiletdown'aine from1he RCS.loperatin0 Pressure W belouitiijdpelgrL 1 pressure of thbSpreAiure"pohlonM the lesican liseRAround $is'letdsen grRepM a bypass ^ i i line containihg"a l6cked/clossdmanusfisolitionTvalyEMiliepenidIchliliiiiihditigiiriwhorl ttAi '
( RCS is depressurik6dlt6 provide' sufficient leidowdfisiir^iwheinfequired5ThiiftetdowrillniOiithiri l
l opulppsdfsith'twoisafetpielsted/ normallp closed 7fallTclosed CIVs"wheM irpenetroisis i l
containhsiint'to;thiliqUid.hidWisti (WLS);dsgasifier packaislssd EigsdThel letdown *ll6Ed6ET6 toland' including the loutts6Afd CIV hass ' design'pressursL.st:2486?psigfoownstroanfsi ths ;
60thcard ClVAths,WLS11etdowrt)ine has?a desidripf11sssGiaW50'peisliiijdjherisforeld66hT66l l insettrieCsysS;critsrii; i
i SG6iis5N65~51K6f;WCAPs14425'56hti6dithit it lihiWiiieIcil6stdiWdeisid67thi16EIjifiisisGii portions lof thisJetd6Wn'ilhelto 4 higher: design lpreamsThi;WL8;EHTi:are'large atmosphed6 tanksf'a6(are thereforein6t secuceble i for higher?desigripressure%Not ts;9ifletdown lins]
ivhich ishiGfsd to thideg~asifisi package 16i the EHTsj and'theIdsgasifisipsclinge7ishlcli >
discharges ~directly to the WLSLEHTshThelCVS litdownTspsisiiheithelfollswing1feistDNiiiil6 l Meet the;lSCOCA..criteriai$(1) the piessdreldr6placross ths CVSl letdown orificefpr6te6 tithe WLS from oVerpressurization during letdown ^ operation's1by: red 0cingithil press 0fiinittie~ WLSR2)
~
in casee ariinadvertent vslve'clo'surs~1n the~ WLS' ddri6g letdown 7a'ritiefNaise/which di6chargeALdirectly;ts the EHTfis~provided thst would protest the WLSlfrois syistpasssdflisti6h!
~
(3) duel to; the leidowri orificb7a break irithsWLS ~duiing1 letdown from thi CVShosid'isssitilri afi RCSilealithauslwithin the? capability.of the normal makeup systsim{(4)?if aril8LOCNshould
'oscuriit would'be feiminated by automatic isolation of the two'hurificatiohl6op is01stiodisisisis snd 'tWo' letdown ~ isolation valves on los pressurizer lesal'or k safeguards)ctuation"sidnaffind (5).thejetdowr[line;CIVs havitheLcapabilitj for leak _ testing"ahd havsyalve position indicstfordfi
12 tiiir6Efilsiit'Eii5HisOisilmsfs6a (6fWisWLSTdegasifkiiE^6elsisE*idntainFFhlih iliiiiiiiiiiGliR5 that WouldauamWoor$ot room operators:WietWioMapreneweilisijs: spread 6ii Dili pressureiNedditionTas:discusied'proviouslyf tissiltheipurificad check. vet #sileju6iect to;isaktestinstThe:staltandsjiiiiiAV8Jiedowri pipiniWilbs _
go.016)$t.DCA positior$
(d)TfimeToseWiblisg;sistsiid!
Di C_~"?~~~M_ rsFilsistP8sy~6616edsciiiNisiliiiiilhiiriliiiiiiiiiiws.uiginiliiiiiiviii'aclai6.a
~ .
thi@ig c . .. .
~81RWCMM428 f ewmospoonvowL nowwessw. essaior weter stonige:tenk (EWSTfdhd:ltsidrainsgs" snit:liniel ~ ~ance, eductor.dupply8umiFisil!
h6d'deminefellisd Wstef@~suppti llrisiHThsse' portions hdW~desigiiipressure RCSloperednipisuhi Theappucinisontends;thet;ltWpracucaigdcIsigriWiii[
pressure" portion of the ess ters higher design presswo becages,enify,ar4Wetemaapheds
~
pressuro and c6nrisd to;ttisiosipressure domineranted water syilsm1DNBMD6elphing1Hi EWST t6 higtfpreisiiiroTt6 ineeOSLOC6;critade would hijuiroNOW8W;beldesignedfothiili pressure 3ienichis h6t practicabis]
' Ths:P881ii~sihHistsdlisthsjtC8TthetiU@Kthil66al:Wampisji61ntilijithWRC8;hiflidif preesistag~ isha:ilauid spaceitund ths: barn makeup tenisasseligthsis:inimpany connec8en icontain(a now rest'r leting armes thst umits'ths 8cs W6isihiHtCSDMWs5Gf a s' ampts truel@andfalsd reduceltNiipresssrs in ths sampaninsesiduitng seihilind op;erationsEEach' (6s:comnion headdiMThe:Eommonheidersampling lins ~alscF6ontainislnormapy then penettsteiW4ontainmord l closed isolation iiiiliti:tili6:normelli blessd'CIVolwhicKireTales PlVisnd Wlu tis lsolated sn" safeguards.si6nal V sponW6ans8ng':
operstisnsThbisirsplinstins&csonnectitoVaample coolor?ihdluselsamplebildes31sdaddi tion?onolofMind lih6s: connected 16 thslow pressure pordorict,P88:centeisiftirili?ithock valvssl h6d ths'otheCob6tains'bhe;6 heck ~vilsW" arid ons~normany closed isolation valigiSths: sis 61 trist these:veves IssisdTthsisak would not:bverpressurtmeM6Wpressure possonsWhis system?tinhwould fl6w1directly to MEW 8T4tn theinlikely event lof(gross failure #9si:hli6 posasuiRchacOsties!thimaximuni;flodrate fWiniW RC8lwould lik wNhin fiEToepaldEpKths li ormal makelijilGsteinRIFWeter level Iri thi EW8T lihi6nitoredfiiiiWu'hlghWilitierm"Irith%
tintiwould julert tips l'6perator ts;afpotsntial Isak into;ttioWesili nuis RSSisawiipung)l6sMI6e
~
operator would.then beLable to lsolate the leak bp closing the CIVs!!The:CIVsLhave. remote
' position indlestion]n ths" control room snd sto" subject to thei:CIVJeakage teiCfAliii/M;isilisis fr6m theselCIVsthr6ughLthel1/4 finch lpiries would bolsniall3ThereforeMR88:deMhis6ts the intent.of the:1SLOCA criteria; (e) Solid.Raddastisysishii Thisolid hiidssibiFijitshi(WSSRWhish'providersi6ragsfscilitissf6F6oth%iifshdWiys51.id l
Mstes"pri6t (dand'subsequsht to probessind and packagind71s~conrictedl6*ths Sg$pfissbis l CVSidemineraliz6rs 16fabilitate transfer lof spent rssin from the;CVS;demineralizeis tithsispent
~
tosin stors'ge tanks:(SRSTs)JThe spent resin header connects ideachVthelthres CVS domineralizers with an individual,'normalVclosed isolation valvefand then"pesetrstes bostainment with tn nsrmally closed locked-closed CIVs to ths SRSTs outsideW#ms&Uiliaise is placed; downstream"of theiobtboard CIV to. isolate the:downstreaivipiping'16 fabilitats:CIV, leak L___--________
13 isi4Kg'Miiirsiuir:16sW#lplso ~doiiv~nitissinviiiirinhnselliatiigiiriniilisisT16wgunedaiiirdialin ismeMaisewawwace;epwaune prennuniwesenenzaa4 Wop smairessoaisiinit is I
! . ifyiicessorisidesignMWS$1(Whigherdesign^gegesse l6e'lilem hie (JuitiscrmoomponenteLouch ashmentashire, belli
$ll@willene nshtM pon%1 J
~ ~ ~
7.tsE.W88 " ' Ti._i.f.filis,sWno_sinitlyTsolelud._ Aic$iiiiil ~~ "siniiaE_al, controlle#h. ave position in .
mome ' ~ "
l accordenosiistiiNInissvise joshneissmutasAR ~~~
l sisA61d_enntelisssntud noimanycirentiefhimotor oossaantstospreuiestseglutistesin the,. . ihneediesedientagon vehnwA , .. .
~kimb:IsolesonWalvesliiin:hu:ciosas#Teelets ..
l the Was~ensthe::Rc8stieddition?downstroeni:of therinboard:c Vlatharesentranifitilhs!
L t6ersXiisi8erysiis;wtueKdischsige(to the wl.s contsinmenfoumpluside coiselesnwd
- TherefoniE#iii[WSSl:spsrytJoshlines;:si'wln6t required siilis:designedRa;higherdesWA l
pissidria (f)#. _hofelise.
em _d.MaterJ a_ns,.fe.n.nd~s.
r ran .-_torage- ,sy,,,.,n., stem:
l Thradrhi6siilliid TstiflffiHifdFrintif6esoi~s9Eten((DWsy~r656Wu(WiiitiFltNiih~ihii
- d6miriernepedwistsEtisatment systeinIhnd provideia resentorbtdomineralliidWiniiEsiiFisik19 l thibondsinosishi46fiisetanniand fordistresution thmuihout tisilent.jTlWdeelgriliiisNu6c6pnal' l detail (RWiKDWs:areirovidedhaubsection's:2iorsensaMtATtWassisinermarinaiminer
trenefer siempiW4:s~ uction;f%i vierdeminerenced weser storagelesiK(DWsfyihiroupp15iiifst througli'shetetync~6x9 den ^reductioii^ unit to thidemineroRzed isoler.distitpullenheedergr6hi ibis' header #derriliaralized Wataf isisuppelat isisrlous systemeM esiplans&Csis OWS espsi$
l lin(penetristes:c6hial6fsent164 supply headsE inside' coned which esissisiliilew DWs i IntwfaceJaeh tiis RSS'nnd ths:CVS dominera8airsdTh(DWS providei'deminenidebdinisterl6 the:P88:td flush thsEssjines:priottol.RCSfsampilng,TehdK!heiDV8jseminefeHseritiEipluice l tosinitoittleMS8;
, R8 b ew s w-pressure y . with e U. lo- RCS..opera -g pre.ssuten _
- t e only pos.
'and the,,CVS" ~ _sible..s s(em ,sdemineralizersre.ssu The R$8_
Ho.w DWS evert In.~l6 _.
h,a overp-- ~
.,~.._. - ds. m'
~ ~ .
ig_n. -
ion p~athways, e;the.connec.t --
con onli .
fr l 6c5rithisiiiNiiih6ftipis[fellUroslsnd misilignmordi$f isb!stionialves*ahdihocit valveshths
~
l high-presourWistimsfff6tist volvihas been added.t6 #iWDWs header inside#containmordM l pro'clurjd;#is;nssibi!itfof}veipiessurizing the DW8RiriadditiorQMW7%X!hs _
l li#$]A?concerftfo6lSLOCA$DW8 MHald most likely result hthe;; rupture;of the D sG@6iidshd c55616si6Hi Thi/ stiff b65610dsi'thif;thWAP600 dssigh I666sististultlithistaff 56 sit l6Hidisssssed]4
~
SEC%90-018]egarding ISLOCA.1Th6reforef; issue;105 is resolved!
i l
l l
I
14 156Ui32Wfdllill5iiE5siljhil Sleed Xi'alidualisWesurteoreessTibiGNM22'21Hsistigstehi:eyssiir6rmswpp1HiiiiiiWW6H16 insatm.~me.v.nimias ~ :onwniwasmenos.uswes adequaipiiffifkiiirgsnoy procedures, operatshrainingfind)wailablefissitmonitoringisyitisns36E t...wmiawed.winnisiein.4 n+bi.idooannofoaminowiino.icarmamestairm (il4.YlosijeUlmedselenE16liiidanalysisof thii[ toss offoodwelet,qHie ,5Wii tiltTouisd opwaters woetesment tFinlestefood-and.tilied operationstand m@et om)assin insananentedenassineasquetoVsistipwatorm'soWneedWiiitissiifisaisliiiMA_lbast OMWh0lWilnellonillWe failure (dia0 nose)hdjehe30rrectivi actiohi[651.inilleti '
M460leedbiouldissulWM*ois coolino.
Ifis~stil0ies;diiiisiilihipiiriiusliiiRWiianghouiii*iiniiGidiinsibHHiiklis'iitelisilitiiDisilTn~d Insediamorgenot guide 8nes Wi+t.WAP600 flesponse:WL6EE:htHistl Sin @didhii concluded thist thi#id~ add blood w6mgenepE%Firei':acceptatsei&ThwefodiNiiiGiDi iissolisdjdiMM600)isighlOpinJteri20M 6Kclosedl-Issue 1.D.5(3): Control Room Design - On-Line Reactor Surveillance Systems As discussed in NUREG-0933, issue I.D.5(3) addressed the benefit to plant safety and operations of continuous on-line automated surveillance systems. Systems that automatically monitor reactor performance can benefit plant operations and safety by providing continuous diagnostic information to the control room operators, to predict anomalous plant behavior.
Various methods of on-line reactor surveillance have been used, including neutron noise-monitoring in BWRs to detect vibrations in intemal components, and pressure noise surveillance at TMI-2 to monitor primary loop degasification. On-line surveillance data have been used to assess loose thermal shields.
Continuous on-line surveillance of the NSSS involves the following areas for which acceptance criteria are separately defined:
e vibration monitoring of resciorintemals e RCPB leakage detection e loose-parts monitoring The acceptance criteria for the resolution of Issue I.D.5(3) for monitoring vibrations in intemal components are in ANSI /ASME OM-5-1981," Inservice Monitoring of Core Support Barrel Axial Preload in Pressurized Water Reactors." This standard makes recommendations on the use of ex-core neutron detector signals for monitoring core barrel axial preload loss. This standard also documents a program containing baseline, surveillance, and diagnostic phases and makes recommendations for data acquisition frequency and analysis.
The acceptance criteria for leak monitoring are in RG 1.45 that documents acceptable methods for channel separation, leakage detection, detection sensitivity and response time, signal calibration, and seismic qualification of RCPB leakage detection systems. It defines the regulatory position for an acceptable design of these systems.
The acceptance criteria for loose-parts monitoring are in RG 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors." This RG gives guidelines on
15 such 'system characteristics as sensitivity, channel separation, data acquisition, and seismic and environmental conditions for operability, it also identifies alert levels, data acquisition modes, safety analysis reports, and TS pertaining to a LPMS.
! ' AP600 design includes the reactor coolant pressure boundary leakage detection system as
- required by 10 CFR 50, Appendix A, General Design Criterion 30, and the conformance to the l
staff regulatory positions, as indicated in RG 1.133, for the design of the loose-parts monitoring system. The detailed system' design discussions are in SSAR Chapters 5 & 7.
i The staff has reviewed the functional requirements of the metalimpact monitoring system (MIMS), which monitors the reactor coolant system for the presence of loose metallic parts according to the regulatory position requirements as indicated in RG 1.133, rev 1, May 1981, and concludes that the MIMS functional design requirements satisfied RG 1.133, with the exceptions of system surveillance and reporting requirements, which are most appropriately addressed by the Combined Operating Licensees for plant specific design.
Issue I.D.5(3) is resolved.
Issue ll.D.3: Coolant System Valves - Valve Position Indication l As discussed in NUREG-0933, Issue ll.D.3 addresses the requirements in NUREG-0737 for positive indication in the control room of RCS relief or safety valve position. The acceptance criterion for the resolution of this issue is that the plant design shall include safety and relief valve i indication derived from a reliable valve-position detection device or a reliable indication of flow in .
the discharge pipe in accordance with the requirements in NUREG-0737. This indication shall have the following design features:
e Unambiguous safety and relief valve indication shall be provided to the control room operator.
e Valve position should be indicated within the control room and should be alarmed.
e Valve position indication may be either safety or control grade; if it is control grade, it must be powered from a reliable (e.g., battery-backed) instrument bus (see RG 1.97).
Valve positio,1 inoication sh ,.Jid be seismically qualified consistent with the component or syste i to which it is attached.
e Valve position indication shall be qualified for the appropriate operating environment which includes the expected normal containment environment and an OBE.
e Valve position indication shall be human-factors engineered.
As discussed in the staff AP600 DSER, confirmatory item 20.4-1 requires that Westinghouse update SSAR Table 3.1 1 -1 to include remote positive indication for the pressurizer safety valve, j normal RHR relief valve, and steam generator safety valves. The staff has reviewed the latest SSAR revision 9, Table 3.1 1 -1, which indicates that positive indications have been included for these valves. - Item 20.4-1 is considered closed. TM i action item II.D.3 is resolved.
I' Issue ll.E.2.2: Research on Small Break LOCAs and Anomalous Transients I
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4 16 As' discussed in NUREG-0933, Issue ll.E.2.2, addressed the NRC research programs focused on small break LOCAs (SBLOCAS) and reactor transients. The programs included experimental research in the loss of flow tests (LOFT), semiscale LOFT, Babcock and Wilcox integral systems test facilities, systems engineering, and material effects programs, as well as analytical methods development and assessments in the code-development program.
The programs called for in this issue were completed by the NRC and showed that ECCSs will provide adequate core cooling for SBLOCAs and anomalous transients consistent with the single-failure criteria of Appendix K to 10 CFR Part 50. The application of the experimental data from the research programs to validate the conservatism of the licensing codes used in the SBLOCAs are addressed in issue ll.K.3(30) in this section.
Westinghouse did not address this issue in its May 28,1993, letter. It concluded, in. ~
x Table 1.9-2 of that letter, that this issue was not relevant to the AP600 design because this issue was resolved with no new requirements.
Because the AP600 design is the first passive advanced LWR design to be reviewed by NRC, the staff is considering how the research for the non-passive LWRs apply to this design. The distinguishing feature of the AP600 is a dependence on safety systems whose operation is driven by natural forces, such as gravity and stored mechanical energy.
While passive systems may be conceptually simpler than conventional active systems, they may be potentially more susceptible to system interactions that can upset the balance of forces upon which the passive systems depend on for their operation. It should be noted that these " passive" systems still rely one some active operation to place them in operation.
For a design with passive safety systems and without a prototype plant that will be tested over an appropriate range of normal, transient, and accident conditions, the following requirements, the following is required by 10 CFR 52.47(b)(2)(1)(a):
The performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, experience, or a combination thereof.
e Interdependent effects among the safety features of the design have been founti acceptable by analysis, appropriate test programs, experience, or a combination thereof.
e Sufficient data exist on the safety features to the design to assess the analytical tools used for safety and analyses over a sufficient range of normal operating conditions, transient conditions, and specified accident sequences, including equilibrium core conditions.
Westinghouse has developed test programs for the AP600 design to investigate the passive reactor and containment safety systems, including component phenomenological (separate-effects) test, and integral-systems tests. The staff has completed and documented its review of the AP600 testing programs in Chapter 21 of the AP600 FSER issue ll.E.2.2 is considered closed.
Issue ll.E.5.1: Design Evaluation
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r 4 17-As discussed in NUREG-0933, Issue ll.E.5.1, addressed the requirement for B&W licensees to propose recommendations on hardware and procedural changes relative to the need for methods for damping primary system sensitivity to perturbations in the once-through SG. In 10 CFR 50.34(r)(20(xvi), it is stated that a design criterion should be established for the allowable number of actuation cycles of the ECCS and RPG consistent with the expected occurrence rate of severe l overcooling events considering anticipated transients and accidents.
Westinghouse identified in Section 1.9.3 of the SSAR that it conaldered issue ll.E.5.1 relevant to
, the AP600 design and stated that although this issue applies only to B&W designs, the AP600 l
design uses the passive core cooling system to provide emergency reactor coolant inventory.
control and emergency decay heat removal. Component design criteria has been established for the number of actuation cycles for the passive core cooling system. The identified actuation l cycles include inadvertent actuation, as well as the system response to expected plant trip p occurrences, including overcooling events. Operation of the ADS is not expected for either design basis or best estimate overcooling events. Section 3.9.1 of the SSAR has additional information.
The s't aff reviewed Table 1.9-2, which provided status of TMl and USI/GSI related items discussions, including item il.E.5.1 in Section 1.9 of the SSAR. The staff considers Open item 20.4-15 closed.
Issue ll.F.2: Identification of and Recovery From Conditions Leading to inadequate Core Cooling As discussed in DSER Open item 20.4-17,10CFR 50.34(f), Additional TMI-related Require-ments, requires that instruments be provided in the control room, which have unambiguous indication of inadequate core cooling (ICC), such as primary coolant saturation meters in PWRS, and a suitable combination of signals from indicators of coolant level in the reactor vessel and in-
! core thermocouple in PWRs and BWRS. NUREG-0737, TMl action plan item II.F.2, discusses l the ICC phenomena and the need to have a reactor water level indication system that provides '
indication of reactor coolant void fraction when the reactor coolant pumps (RCPS) are operating, .j and reactor vessel water level when the RCPs are tripped. '
Prior to the TMI accident, an accepted operational practice of PWRs was to operate the RCPS, if
[ they were available, during a LOCA to provide continued core cooling. During the TMI LOCA event with the stuck open PORVS, the reactor coolant continued to leak through the open valves, ,
- the pressurizer level indicated high, and subsequent-lCC occurred because the reactor coolant l was highly voided. Nevertheless, core cooling was maintained with the continued operation of the RCPS. Subsequently, the RCPs were tripped and because of high void content in the coolant, the water level dropped below the .op of the core causing fuel damage. As a result of the TMl lessons leamed, the reactor vessel water level indication system was added, specifically for PWRS, to ensure operator action to trip the RCPs following a LOCA, rather than later in the i
LOCA sequence to prevent ICC event NUREG/CR-5374, Summayy of Inadequate Core Cooling Instrumentation for U.S. Nuclear Power Plants discusses acceptable approaches to instrumentations used to address ICC.
In response to staff RAI #440.162, Westinghouse explained that the AP600 design concept is 4 different from current operating plants in that the AP600 design automatically trips the RCPs and initiates safeguard injections through the passive safety systems such as CMT, ADS, PRHR and IRWST to maintain core cooling in the event of a SBLOCA. It does not rely on a reactor vessel ,
level indication system as do existing reactors, where reactor vessel level indication is important
18 for operator actions to trip the RCPS, to monitor coolant mass in the vessel and to manually depressurize the RCS in the event of ICC. There is no need in the AP600 for the operator to trip the RCPS, to inject water into the core or to manually depressurize the plant during a SBLOCA.
The instruments typically used in current PWRs include subcooling margin monitoring I capability, core-exit thermocouple, and reactor vessel level indication system, which together would provide the operator with the ability to monitor the coolant conditions and to appropriately take actions to ensure core cooling during the approach to and to recover from the inadequate core cooling conditions. The AP600 design includes subcooling margin monitoring capability, core-exit thermocouple and the hot-leg level indication system.- The AP600 hot-leg level indication system is different from the reactor vessel level indication systems currently used .
In Westinghouse plants.
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The AP600 hot-leg level indication is a safety-related level indication system, which consists of separate pressure taps that connect to the bottom of the hot leg, and to the top of the hot leg bend leading to the steam generator and has the ability to provide indication of reactor water vessel level for a range spanning from the bottom of the hot leg to approximately the elevation of the vessel mating surface, in addition, during the operation of the ADS to depressurize the plant, the reactor vessel water
. level will vary greatly and will not provide a reliable indication of ICC. The AP600 hot-leg water level indication is not used to direct operator actions even when the water level may potentially drop below the hot leg level. Therefore, the water level is not an important indication for -
mitigation of ICC in the AP600 design. The hot-leg level indication system is used, however, as a verification of reactor water inventory to terminate the recovery action in the ERGS for the ICC event.
Because the AP600 design automatically trips the RCPs during a SBLOCA event and because the operators are not prone to be mislead by forced two phase flow, the core exit temperature is an important and sufficient indication of an approach to ICC condition. The temperature reading provided by core-exit thermocouple has been appropriately included in the ERGS for plant recovery.
The staff has reviewed the Westinghouse response and has determined that for e SBLOCA event a safeguard signal would automatically trip the RCPS, passive safety systems such as the CMT would automatically inject water into the core, the ADS would automatica!!y initiate te depressurize the plant, the reactor coolant would automatically be cooled by the PRH_R, and subsequent injection from the IRWST would occur. The staff has also determined that for AP600 design, he core-exit thermocouple and the subcooling margin monitoring together would provide unambiguous indication of an approach to ICC and the safety-related hot leg level indication is only used to terminate the recovery action in the ERGS for the ICC event. Therefore, the requirements for ICC, as discuss in 10CFR50.34(f), have been satisfied and the issue is resolved.
Issuell.K.1(3): Review Operating Procedures for Recognizing, Preventing, and Mitigating Void Formation in Transients and Accidents As discussed in NUREG-0933, issue ll.K.1(3) requested licensees to have operating procedures L for recognizing, preventing, and mitigating void formation in the RCS during transients and accidents to avoid loss of the core-cooling capability during natural circulation.
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19 The staff has reviewed the resolution of issue I.C.1 and its related ERGS AES-0.2, " Natural Circulation Cooldown," and has concluded the guidelines direct the operators to cooldown and depressurize the plant using natural circulation conditions by dumping steam and subsequent RNS operation. These steps are specified to preclude any possible upper head voids formation and also direct the operators to verify that a steam void does not exit in the vessel. The staff concludes that the ERGS provide directions to plant operators to recognize and to preclude voids formation M the vessel and therefore, the staff considers issue ll.K.1(3) closed.
Issuell.K.160): Review Operating Procedures and training to Ensure that Operators are Instructed Not to Rely on Level Alone in Evaluating Plant Conditions As discussed in NUREG-0933, issue ll.K.1(4D) asked licensees to provide operating procedures to ensure that operators shall not rely on levelindication alone in evaluating plant conditions. As stated in NUREG-0933, the staff determined that this issue was covered by issues 1.A.3.1, l.C.1, and ll.F.2, and is resolved.
Issue I.A.3.1, " Revise Scope and Criteria for Licensing Examinations," was implemented by NRC by a rule changc to 10 CFR Part 55, " Operators Licenses," to require simulator as part of the reactor operator licensing examinations. The staff willimpose the requirements of 10 CFR 55.45 on simulators on the COL applicant referencing the AP600 design; therefore, Westinghouse and the staff does not have to address issue I.A.3.1 for compliance with 10 CFR 52.47(a)(1)(iv).
Westinghouse did not address thie, %v in its May 28,1993, letter. It concluded, in Table 1.9-2 of that letter, that thi< n a yes not relevant to the AP600 design because this issue is not a design certification isQ b -; the responsibility of the COL applicant. However, in response to the staff request for h e .ialinformation (RAI), Westinghouse stated that the design portion of this item is addmNd in the proposed resolution to issues I.C.1 and ll.F.2.
The staff has completed its review of Issues I.C.1 and ll F.2 and has concluded that AP600 ERGS do not instruct the operators to rely on levelindication alone in evaluating plant conditions.
The status of core cooling is determined by indications of core exit thermocouple temperature, RCS subcooling, and RCS hot leg temperature in addition with RCS level. The staff considers these issues resolved and therefore, Issue li.K.1(4d) is closed.
IssuellA.1(17): Trip Pressurizer Level Bistable so that Pressurizer Low Pressure Will Initia' Safety injection As discussed in the staff DSER Open Item 20.4-22, TMI action plan item II.K.1(17) addresses the requirement for Westinghouse plants to trip the pressurizer level bistable so that the pressurizer low pressure, rather than the pressurizer low pressure and pessurizer low level coincidence, would initiate safety injection.
AP600 design does not depend on pressuns,r low pressure and pressurizer low level coincidence to initiate safety injection in the event of LOCAS. Safety injection in AP600 design is automatic. The following safeguard signals would initiate safety injection: Low-1 pressurizer pressure or Hi-l containment pressure or Low compensated steam line pressure or !
Low-3 cold leg temperature. In addition, the AP600 design also gives the, operator manual safety injection capability. The staff concludes that any single safeguard s gnals mentioned above would initiate safety injection. Therefore, this issue is resolved.
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20 t issue ll.K.1(24): . Perform LOCA Analyses for a Range of Small-Break Sizes and a Range of Time Lapses Between Reactor Trip and RCP Trip Issue ll.K 1 (24), of NUREG-Og33 required PWR licensees to perform a LOCA analysis for a range of small-break sizes and a range of time lapses between reactor trip and RCP trip. The staff determined in NUREG-0933 that this issue for PWRs was covered by issue I.C.1, "Short-Term Analysis and Procedures Revision." -
l Westinghouse has provided for staff review of AP600 Emergency Response Guidelines (ERGS),
which addresses issue I.C.I. The staff has reviewed the responses to issue I.C.1 and specifically the emergency response guideline AE-0, "AP600 Reactor Trip or Safety injection" for small break LOCA that addresses item II.K1(24) and has concluded that the AP600 design automatically .
trips the RCPs during a LOCA event. The guideline directs the operators to verify that all reactor coolant pumps have been tripped, and if not, the operators are directed to manually trip the reactor coolant pumps. Based on the plant design features and the appropriate operator's actions using ERGS, the staff considers item ll.K.1(24) resolved. Open item 20.4-23 is closed.
1' Issue ll.K1(25): Develop Operator Action Guidelines.
As discussed in NUREG-0933, Issue ll.K.1(25) required PWR licensees to develop operator
- action guidelines based on the analyses performed in response to issue ll.K.1(24), which is
! discussed above. The staff determined in NUREG-0933 that this issue was covered by issue I.C.I.
. Westinghouse did not address this issue in its May 28,1993, letter. It concluded, in Table 1.9L2 of that letter, that this issue was not relevant to the AP600 design because .a issue had been superseded by one or more other issues. Although this issue was covered by issue l 1.C.1, as stated above, Westinghouse also did not address this later issue because it considered Issue I.C.1 the sole responsibility of the COL applicant.
The final procedures would be the responsibility of the COL spplicant; however, the range of LOCA analyses for a range of time lapses and the specific information to go into the procedures would be the responsibility of the designer, or Westinghouse in the case of the AP600 design.
Westinghouse addresses accidents for the AP600 design in SSAR Chapter 15. The staff requests that Westinghouse address operator action guidelines, or EPGS, of I.C.1 and the role of the COL applicant in issue ll.K.1(25' This is Open item 20.4-24.
The staff has completed its review of Issue I.C.1 and has concluded that Issue I.C.1 is closed, therefore, Issue ll.K.1(25) or Open item 20.4-24 is also closed.
Issue ll.K1(27): Provide Analyses and Develop Guidelines and Procedures forinadequate l
Core Cooling As discussed in the staff DSER, the AP600 design should describe analyses of ICC conditions and develop guidelines and procedures to mitigate an ICC event, and that this issue is l dependent on the resolution of Action items ll.F.2 and I.C.1, Identification of and Recovery from Conditions Leading to ICC and Guidance for Evaluation and Development of Procedures for Transients and Accidents, respectively.
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- e-l 21 Westinghouse has submitted AP600 Emergency Response Guideline (ERG) for staff review, and 1 - also responded to staff DSER Open item 20.4-17 to address Action items I.C.1 and ll.F.2, i
respectively. The staff has reviewed Westinghouse response to Action item II.F.2 and a detailed discussion of this item is documented in its respective section. In the AP600 ERG, L Westinghouse provides high-level guidance to deal with inadequate core cooling conditions.' The staff has reviewed AFR-C.1, AP600 Retponse to inadequate Core Cooling procedure and '
( analysis bases, which describes how passive safety-related systems would automatically trip the
, RCS pumps, initiate and depressurize the RCS to inject water into the core upon receiving a .
l safeguard signal, in this procedure, the operators are instructed to monitor plant conditions using core exit temperature and indicated hot leg level, which is designed to provide indication of an approach to ICC and to recover from an ICC condition. The operators are also instructed to manually initiate injection when automatic passive safety injections fail. Passive safety-related J system actuation indications of CMT, ADS, PRHR, and IRWST are integrated into the
- j. procedures, which provide operators with directions to ensure that adequate core cooling will be maintained. Therefore, the staff concludes that Westinghouse has appropriately provided analyses and procedures to mitigate ICC conditions. Issue ll.K 1(27) or Open item 20.4-25 is closed.
l Issue ll.K3(6): Instrumentation to Verify Natural Circulation
! As discussed in NUREG-0933, Issue ll.K.3(6), addressed requiring licensees to provide instrumentation to verify natural circulation during transient conditions. The staff determined in NUREG-0933 that this issue was covered by issues I.C.1, ll.F.2, and ll.F.3.
Westinghouse has provided the staff with pertinent information about the AP600 design, which adoresses TMI action items I.C.1, II.F.2 and ll.F.3. The staff has reached a conclusion that those issues relevant to the resolution of the TMl action item II.K3(6) have been resolved. The detailed discussion of the related issues are addressed in their respective TMl item discussions.
Therefore, this issue is resolved.
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Issue ll.K.3(8): Further Staff Consideration of Need for Diverse Decay Heat Removal l Method independent of Steam Generator j As discussed in NUREG-0933, issue ll.K.3(18) addressed further staff consideration of the need for diverse decay heat removal methods which were independent of the steam generators. The !
staff determined in NUREG-0933 that this issue was covered by issues ll.C.1, " Interim Reliability Evaluation Program," and ll.E.3.3, " Coordinated Study of Shut-down Heat Removal Requirements." In NUREG-0933, the staff also stated that issue ll.E.3.3 was addressed in issue i A-45, " Shutdown Decay Heat removal Requirements."
Westinghouse has provided the AP600 shutdown evaluation report for staff review. The report
. describes multiple decay heat removal capabilities independent of the steam generator. The detailed discussion of the multiple decay heat capabilities is included in Chapter 19.3 of the l FSAR. The staff, therefore, concludes that Issue ll.K.3(8) or Open item 20.4-27 is closed. ;
4 I Issue ll.K3(30): Revised SBLOCA Methods to Show Compliance with 10CFR Part 50, )
Appendix K As discussed in NUREG-0933, Issue ll.K.3(30) required licensees to revise and t,ubmit analytical methods for small-break LOCA analyses for compliance with Appendix K to 10 CFR Part 50 for l
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l 22 NRC review and approval. The re /ision was to account for comparisons with experimental data, l including data from LOFT test and semiscale test facilities. Altematively, licensees were to provide additionaljustification for the acceptability of their SBLOCA models with LOFT and semiscale test data. Clarifications were issued in NUREG-0737. The staff has reviewed NOTRUMP code and has documented its discussions in Chapters 15 and 21 of the FSER. The staff, therefore, considers this issue closed.
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BUGL STATUS BUGL TITLE DSER RESOLUTION FSER RESOLUTION !
l BL-80-12, Decay heat This bulletin dealt with The ' staff evaluates this issue removal reducing the likelihood of in FSER Section 6.3. This BL losing DHR capability. W is resolved.
stated that this issue is
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1-discussed in SSAR 7.4.1 1
! BL-80-18, maintenance of W stated that the design N/A. Closed.
l adequate mini flow through does not have CCP as part of
- CCP following secondary side Si and that it is not applicable high energylins rupture. to AP600. i BL-86-01, mini flow logic AP60n does not have valves N/A. Closed.
L problem that could disable l'1:ti flowlines. Issue l RHR pumps resolved.
l BL-89-03, potential loss of The staffindicated that SSAR Section 9.1 discusses -
required shutdown margin movement and placement of fuel storage and handling, ,
during refueling fuel during refueling is within including the refueling - {
the scope of AP600 core equipment used to safely 1 design. This issue is a COL move and store fuels. j action item. Additionally, IRWST provides i large quantities of borated water that maintains the required shutdown margin.
This BL is closed and the COL action item is still valid regarding plant specific guidelines.
GL-80-01, report on ECCS The staff requested that W The safety evaluation of this address NUREG-0630, issue is in FSER Chapter 15.
" Cladding, Swelling and The GL-80-01 is closed.
Rupture Models for LOCA Analysis." 1 GL-80-014, LWR primary W needed to indicate where W stated that SSAR 1.9.4.1.2 !
cociant system pressure- in Section 1.9 of the SSAR and USI-B-63, discuss this - !
isolation valves that this issue is discussed issue. Tfis~sWillissii~tiii: !
l thlilii^OslnWIT1Wi!!!s50s j
. 05 SL iThlOsisaji ;
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r-I l 24 GL-80-019, resolution of W stated that the fission gas The staff position has not y enhanced fission gas release release Models are changed. No additional l concem' accounted for in WCAP- action is needed. The GL 10851-P-A and WCAP- 019 is closed.
l 11873-A, " improved fuel .
performance Models and safety evaluations." This issue is resolved.
GL-81-021, natural circulation The staff stated that W W has submitted AP600 cooldown should address the ERG for ERG-GW-GJR-100, Rev 3 this event. dated 5/97 for staff review.
The staff has reviewed this submdtal and its related natural circulation ERG and has determined that guidelines are sufficiently
! given to the operator to cool down the plant using natural circulation means. This GL is closed.
GL-83-11, licensee W should address tha This issue is COL qualifications for performing qualifications for performing responsibility. This GLis safety analyses in supporting safety analysis for AP600 closed.
, licensing actions design.
GL-84-21, long-term, low- Core peaking factor may be The safety evaluation of this power operation in PWRs greater than assumed in issue is in FSER Chapter 15.
safety analysis for extended This GL is closed.
l low-power operation following a retum to full power operations.
GL-85-16, high boron This GL is resolved because The staff position has not concentration AP600 design does not have changed. No action is B'-' and boron concentration required. This GL is closed.
from CMT is much lower than BIT (22,000 ppm)
GL-87-12, loss of RHR with This GL addressed potential W has submitted WCAP-RCS partially filled for loss of RHR during 14837, Rev 2 (11/97) that midloop operation discusses shutdown risk concems, including potential ,
loss of RNS. The staff has resolved this issue and its SE l
is discussed in FSER Chapter 19. This GLis closed.
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.o 25 GL-88-17, loss of DHR This GL addressed potential W has submitted WCAP-forloss of RHR during 14837, Rev 2 (11/97) that midloop operation discusses shutdown risk concems, including potential loss of RNS. The staff has resolved this issue and its SE is discussed in FSER Section 19.3. This GL is closed.
GL-91-07, RCP seal failures GSI-23 discussed RCP pump The staff has resolved GSI-seal failures. W addressed 23 issue because W design this issue in SSAR Sections does not have RCP pump 5.1.3.3 and 1.9.4.2.3 seals, and the GL is not applicable to AP600 canned pump design. This GL is closed.
GL-93-04, rod control system W should revise the WCAP- WCAP has been revised failure 13559 to include this item of (8/96) to include reference of discussion for AP600 design this item of discussion in SSAR 3.9.4. The staff SE is discussed in FSER Chapter
- 4. This GL is closed.
GL-83-22, safety evaluation THIS ISSUE WAS NOT The staff has reviewed W ;
of ERGS INCLUDED IN DSER AP600 ERG-GW.GJR-100, '
Rev 3 dated 5/97 and has l documented its evaluation in I FSER Section 18.9.3. This GL is closed.
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.c l-26 BL-96-01, rod control - THIS ISSUE WAS NOT The BL was issued because problem INCLUDED IN DSER ofincomplete control rod l insertion (lRI) evaluation at the South Texas and Wolf Creek plants. It has been deterfraned that the IRI was l caused by thimble tube distortion resulting from excessive load. Since this is a fuel design problem, and W l has not committed to any fuel l'
manufacturers, the staff has l' concluded that W does not '
l have to address this issue, unless it has committed to l certain fuel designs l discussed in the BL. This issue should be appropriately addressed by the COL l applicant. This is a new COL action item. The BL is closed.
i GL-86-16, ECCS evabation The staff requested that W W discussed this issue in discuss the ECCS evaluation SSAR Sections 6.3.5 and Models for AP600 design 15.0.11. The staff has evaluated Westinghouse ECCS Models and has discussed this issue in FSER Chapter 15.' This GL is closed.
GL-85-05, Inadvertent Boron W discussed this issue in l
Dilution SSAR Section 15.4.6. The l staff has evaluated and has i discussed this :ssue in F"ER Chapter 15. This GL is closed.
I-GL-96-04 This issue is not part of SRXB responsibility. It is most appropriately addressed by SPLB.
GL-86-07, NUREG-1190 THIS ISSUE WAS NOT This issue is not part of l regarding the San Onofre INCLUDED IN DSER SRXB responsibility and it is Unit 1 loss of power and most appropriately addressed water hammer by HHFB. ]
l l 27 Note: All TMI Items (11.B.1, ll G.1,11. K1(28), ll.K.2(16), ll.K.3(2), ll.K.3(5) and ll.K.3(25) indicated in the 12/10/97 note from J. Sobrosky have been resolved in the DSER. The staff does not see any changes in position regarding its evaluation of these items. _ Als.o, issue 23 has been resolved and is reflected in DSER. No change in the staff position is anticipated, i
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