ML13163A266

from Bhalchandra Vaidya to Samson Lee: G20120172, 2.206 Petition

ML13163A266
Person / Time
Site:	FitzPatrick
Issue date:	05/18/2012
From:	Bhalchandra Vaidya Office of Nuclear Reactor Regulation
To:	Michelle Albert, Collins T, Cook W, Dennig R, Eul R, Robert Fretz, Jennerich M, Samson Lee, Mauri Lemoncelli, Sammy Mccarver, David Mcintyre, John Monninger, Kimyata Morgan-Butler, Richards K, Andrea Russell, Carrie Safford, Catherine Scott, Thadani M, Ulses A, Bhalchandra Vaidya Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation, NRC/OE, NRC/OGC, NRC Region 1
References
2.206, FOIA/PA-2013-0010, G10120172, GL-89-016
Download: ML13163A266 (29)

Text

Dorlen . Larec Doerflein, Lawrence From: Vaidya, Bhalchandra Sent: Friday, May 18, 2012 3:26 PM To: Lee, Samson; Vaidya, Bhalchandra; Bickett, Brice; Doerflein, LawrenceR'-rnf-ri-h Matthbw; Dennig, Robert; Ulses, Anthony; MorganButler, Kimyata; Fretz, Robert;] b)(7)(C)

Ryan; Richards, Karen; Safford, Carrie; Monninger, John; McIntyre, David; Collins, Timothy; Scott, Catherine; Albert, Michelle; Cook, William; Thadani, Mohan; Russell, Andrea; Cc: McCarver, Sammy; Lemoncelli, Mauri Wilson, George

Subject:

G10120172, 2.206 Petition; PRB Meeting of May 17, 2012 to make Initial Recommendation to accept or reject Attachments: Dec 6,1991 Document.pdf; Aug 14.1992 Document.pdf

Folks, Subsequent to our discussion in the PRB meeting on May 17, 2012, 1 have had more conversations with Mohan Thadani, who was the PM for the GL 89-16. The following points would help to clarify the claims of the Petitioners with respect to the NRC efforts during the GL 89-16 process as well as Post-Fukushima Events:

(1) Contrary to our discussion. GL 89-16 addresses all contents of "Vent Products.' such as Steam, Hydrogen, Nitrogen, etc. This is mainly because the BWROG criteria that were used to evaluate the licensees' responses, included Hydrogen concems (In addition to Sept 28, 1992 NRC Approval, I am providing two more documents from ADAMS Archive for your information). GL 89-16 was part of the NRC's program to enhance the Containment Performance in response to "Beyond Design Basis Accidents, namely SBO TW Sequence," in addition to the Design Basis Accidents. Therefore, the Petitioners' claim that the I NRC's evaluation of the licensee's GL 89-16 responses being 'improper," "in-adequate," "faulty," etc., is not correct. Whether, the modification to Vent system were installed, or not, FitzPatrick was found to meet the performance requirements per I BWROG Criteria. Ni (2) After Fukushima Accident, the NRC through the near term task force (NTTF), and Japan Lessons Learned Directorate (JLD),

has issued the Hardened Vent Order to achieve the "Reliable" Vent System to address the performance issues on the Vent System, all constituents of the Vent Product. The Hardened Vent Order has a prescribed time line for completion, which will require a 10 CFR 50.90 process that allows for public Participation. FitzPatrick is no different from the other 24 odd BWR, MARK I plants. Therefore, the Petitioners' claim that Fukushima event makes the FitzPatrick Vent System unreliable, undependable is true only for "Fukushima Type Event," and is no different than other 24 odd BWR, MARK I plants. The Commission has established the process to resolve the issue for all BWR, MARK I Plants-for Fukushima type of events, including FitzPatrick.

With the above discussion, I believe that the Table in the Final PRB Notes of our meeting yesterday, May 17, 2012, should indicath "reject" or "no" for all items. ,

If you believe that we need another meeting to discuss this more, let me know and Andrea and I will make arrangement for the meeting.

Thanks, Bhalchandra K. Vaidya Licensing Project Manager NRC/NRRIDORL.LPL 1-1 (301)-415-3308 (0) bhalchandra.vaidVarc.nrc.gov

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.i r. . 199 JP.9,1-W65 U.S. Nuclear Regulatory Commn'ssion Atnn.: Document Control Desk Mail Station P 1-137

.. Washington, D.C. 20555

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Hardened Wetwell Vent Capability for the James A. FitzPatrick Nuclear Power Plant

Reference:

1. NRC Generic Letter 89-16, 'Installation of a Hardened Wetwell Vent,' dated September 1, 1989.

2. NRC letter. S. A. Varga to J. C. Brons, 'Installation of a Hardened Vent Capability at the James A. FitzPatrick Nuclear Power Plant,'

dated January 24, 1991.

Dear Sir:

Generic Letter 89.16 (Reference 1) requested that utilities with BWR Mark I containments volunteer to install a hardened wetwell vent system. The Authority's response to this Generic Letter requested that the decision whether to modify the existing FitzPatrick plant hardened wetwell vent be deferred until after completing the RtzPatrick individual Plant Examination (IPE) which was then under development. The NRC concurred and requested that the Authority submit its final position regarding the Boiling Water Reactor Owners Group (BWROG) hardened vent design criteria within 60 days of

• completing the FitzPatrick IPE (Reference 2). In addition, the NRC requested that the Authority use the results of the IPE to re-examine venting procedures and training of operators.

This le"ter contains the Authority's response to Reference 2. Attached is an evaluation of venting during severe accidents at the FitzPatrick plant. it includes the Authority's final position regarding the BWROG hardened vent criteria and a re-examination of venting procedures and training. In addition, it describes insights the Authority gairneld from pedorming the IPE and the status of investigations into accident management strategies associated with severe accidents.

The Authority concludes that the current design of the FitzPatrick hartdened vewetwl0 vent meets Generic Letter 89-16 by providing a reliable venting capability with significant scrubbing of fission products during specific severe accident conditions. The evaluation also concludes that several procedural changes associated with the operation el the vent

/I.o

I equipment may be beneficial as accident management strategies. These venting strategies include:

Venting until the containment pressure iS reduced to near etmospheric

'1 pressure; and,

• Initialing the vent early, i.e., venting under cortain circumstances prior to the containment pressure reaching the currently established vent pressure.

Implementation of these procedural enhancements would require changes to the currently approved emergency procedure guidelines and emergency operating procedures. The Authority will bring these issues to the attention to the BWROG for generic consideration. When these changes are approved by the BWROG tKnn.N;),. thc Authority will provido the NRC with an implementation schedule.

Ifyou have any questions, please contact Mr. J. A. Gray, Jr.

Very truly yours.

Ralph E. Boedlo Executive Vice Presidont Nuclear Generation cC! U.S. Nuclear Regulatory Commission 475 Allendale Road King o0Prussia. PA 19406 Office of the Resident Inspector U.S. Nuclear Regulatory Commisston P0. Box 136 Lycoming. NY 13093 Mr. Brian C. McCabe Project Directorate 1I-Division of Reactor Projects I1/11 U.S. Nuclear Regulatory Commission Mail Stop 14 B2 Washington, D.C. 20555

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New York Power Authority Attachmont to JPN 91-0G5 Hardened Wetwell Vent Capability for the James A. Fi.zPatrick Nuclear Pow,,er PlAqt 1*. Background '"' '"

Generic Letter 89-16 (ReferenCe 1) requested that utilits with GWR Mark I containments volunteer to install a hardened wetwoll vent system. The Authority response (Reference 3) stated that any decision whether at not to modfy RtzPatrick's existing hardwned wotwoll vent design should be deferred until after completing the FitzPatrick Individual Plant Examination (IPE), then boing developed in accordance with NRC Generic Letter 88,20. The. AVthotity prov.'i'd a significant amount of additional information (References 6,8, 9. and 1i) to supPoIrt.Ihislpoz;,-ur The NRC concurred and requested that the Authority submit within 60 days 0f the complolion of the FitzPatrick IPE, its final position on the Boiling Waler Reactor Owners Qroup (BWROG) hardenod vent criteria (Reterence tO). In addition, the NRC requested that the Authority use the results of the IPE to reexamine the venting procoduros and training of operators. Qn Scptqmbcr 13, 1991. the New York Power Authority provided the James A. Fi*t Patrick Nuclear Power Plant Individual Plant Examination to the NRC (Reterence 1t).

This attachment provides the Authority's detailed response to Rolerence O10. Soctior II describes several insights concerning post-accident venting. These insights wcfo gained trom pedorrning the IPE and from performing other severe accident studios and evaluations. The Authorily's detailed evaluation of the FitzPatrick vent design to the BWROG hardened wetwell vent design criteria is provided in Section Ill. Section IV discusses issues associated with venting procedures and training. The status of investigations into cortain accident management strategies associated wrth venting and station blackout events, is contained in *oc, V As detailed in References 6 and 9, the FitzPatrick plant already has a harTdened vent. The vent piping originates at the primary contlinmont suppression chamber air space and terminalos al the inlet to the standby gas treatment system (SGTS), which is located in a separate structure adjacent to the reactor building. All ot the vernt piping is rated for 150 psig internal pressure.

Because the FitzPatrick plant already has a hardened wotwoll vent, the remaining vont-roeted issues are mostly related to the need, it any. for modifying the existing venting procedures or adding new vent equipment to meet the BWROG vant.design criteria, The nood for. and tho nature of such additional venting procedures and equipment has been determined from insights gained from the FitzPatrick plant IPE and other severe accident analyses.

11. Severe Accident (TW and $5OJ Venti.ng n slhts TSeqence Definition In TW accident sequences, a plant transient causes tho reactor to trip. The reactor is automatically shutdown with reactor water level maintained above Iho reactor core, providing core cooling by using oerror more plant systems (e.g., condensate, HPCI, RCIC, LPCl, core spray. etc.). However, normal moans of decay heat removal (e.g., steam dump through the turbine bypass valves, AHR shutdowrf cooling, RHR Suppression pool coolinq, and AN-i LPCI or

NewYork Power Authority..

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Altachmenl toJPN.9],-064

.page2of 18.

containment spray injection through an RHR heat exchanger) are all unavailable. Instead, decay heat Is transferred to the suppression pool by steam flow through one or more Safety/relief valves (SRV). This heats up the suppression pool, and when pool temperature, in conjunction with reactor pressure, approaches the heal capacity temperature limit (HCTL) curve, the operator manually depressurizes the reactor by opening additional SRVs. The suppression pool absorbs tlhis additional heat and continues to condense the steam gencraled from decay heat.

When the water temperature approaches 212 0 F. the suppression pool can no longer condense all of the steam being added through the SRVs. Steam then evolves from the surtace of the suppression pool into the containment atmosphere. Eventually this begins to pressurize the containment. Atter approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, the containment pressure approaches 44 psig.

which is the FitzPatrick primary containment pressure limit (PCPL). The emergency operatino procedures (EOPs) then direct tVe operators to vent the containment to maintair PrS.,ur.-L the PCPL.

Watwell venting at this time prevents the contairment from overprossurizing, while providing a stable, long term decay heat removal pathway (the steam generated olf the suppression pool is directed to the envrronment). The release to the environment from TW venting would be slightly contaminated steam mixed with the inventory of nitrogen gas which was initially in the containment. When one of the normal decay heat removal pathways is restored, the TWV accident is over. The vent is then closed with the plant remaining in a stable cold shutlown condition.

I11the containment is not vented, then pressure continues to rise. VVhen the containment pressure rises to within 50 psi ol the containment instrument nitrogen pressure (or at approximately 70 psig), the SRVs close due to insutficient dp across their air actuators and the reactor repressurizeos Itthe high pressure coolant injection (HPCI) system is not available. [hen all coolant injetion to the reactor ceases. Reactor water level then drops duo to uoi-Ott ;'*,

core damage tollows. The containment continues to pressurize from gases generated from metal.waler reactions within the reactor and, following reactor vessel failure, from core-concrete reactions within the drywall. When the pressure within the containment exceeds its failure pressure, structural failure occurs. With HPCI available, core cooling is maintained as the reactor repressurizes. However. this does not prevent containment from ovorpressurizing from steam generated off the suppression pool. HPCI is then assumed to faitlwith the olher ECCS systems when the containment fails as described below, and core damage follows.

The sudden depressurizatron at the containment upon its lailure has several eftects: Some of the water in the suppression pool flashes 1o steam, and with reduced containment pressure, the SRVs may reopen. dopressurizing the reactor. Suppression pool flashing reduces the net positive suction head (NPSH) available to the emergency core cooling system (ECCS) pumps taking suction from the suppression pool and may cause them to fail from cavitation. It these pumps do not fail from reduced NPSH. then the IPE.assumos that all ECCS pumps (RHR. Core Spray. and HPCI) fail from environmental effects (temperature/humidity) due to steam ri,:r.

from a rupture or leak of the suppression pool. Since one or more of these pumps were providing cooling water to the reactor, core damage would follow.

New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Attachment to JPN-91-065.

page 3 of 18 SBO Sequence Definition In a long term station blackout (SBO) all sources o1 AC electrical Powoer are unavailable. The reactor Is aulomalically Shutdown and the steam driven HPCI and RCIC systems maintain water level above the reactor core. As with the TW sequences, decay heat is transferred to the suppression pool by steam flow through the SRVs and the containment eventually pressurizes.

Since the HPCt and IRCIC turoines are dependent upon DC control power. depIction of the station batteries alotr about eight hours loads to their failure. Once HPCI and RCIC tail, the water inventory In the reactor boils away and core damage begins The onset of core damage occurs approximately 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> into the SBO. At this time; the containment pressure is significantly below the PCPL vent setpoint pressure of 44 psig.

A. iPE Insiqlhts

• TW Scauences The FitzPatrick IPE was used to estimatc the rsk roeucrton associaltd w-th successful venting of the wetwell airspace for the TW accident sequences. This is shown by comparing the total core damage frequency (CDF) with and without venting Based on theF.lzPatrick IPE. we find:

CDF, without venting - 2.72 E-5/yr.

CDF, with venting - t.92 E-6/yr.

The total core damage frequency is reduced by a factor of 14 due to vntirng Ouriog- TW sequences. This factor of 14 is conservativj because it assumes that all ovorpressurlations of the containment from 1W sequinces lead to a loss of core cooling. Novertrefess, ventin, during TW sequences is an ;mportant miiyjating action For FitzPatrick, the dominant TV sequences are those that result from a nonrecoverabl0 loss of either of the two 4160 VAC emergency busses (10500 or 101)00) Loss of either of these busses results in having three of the four RHR pumps tunctionally unav3xatMe Ifo decay heat removal. The RHR and 9HR service water (RHRS.144 pumps are organized and powered as follows:

Division A Division B RHR pump A RHR pump B (10500) t05m0)

RHR pump C i WHI pump D (t060l I (,,i0 RHRSWV pump A&C I ri4FpSW Lxr'n P*.L, i 10500W) * !', 0.)OI

New.Yorlt Power; Authority .

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Attachment to JPN-9i-065

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To provid decay heaI removal, at leatst ono RHR and one RHFSv pump Irom the same frvision must function. Given a loss of a single emergency buS, both RHRSW pumps hfom one division as well as one RHR pump from tho other division lose power, leaving only one RHR pump and two RHRSW pumps available for decay heal removal. Random failure of that one remaining RHR pump loads to a loss of decay heat removal, i.e. a TW accident sequence.

Since loss of an emergency bus is the dominant TW event initiator, in all likelihood, the vent would have to be manually operated. This is Decause valves in the vent path are powered or controlled from both emergency busses Loss of either 4160V AC emergency bus require that at least some of the valve manipulations would be performed manually. In addition, it offsite power is also lost, then the wetwell vent containment isolation valves. 27AOV. 117 and IS, wOuld have to be manuJally oponoo due to loss of the instrument air system. (Thes" 1

valves are dosignod to close upol loss of instrument air.) All of tho valvesin the vent .'.'

are located within the reactor building and accessible depending on radiological conditions

.wthin the building. Since the PCPL pressure is reached only after 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> into the TW event, the IPE determined that sufficient time is available to predict that venting would be toccsssary and carry out all manul actions necessary to establish the vent path prior to the containmnt reaching the PCPL B. Other, Non. PE Ins9*hts -rWSouencs The Authority looked at venting at containment pressures below the EOP specitied vent pressure to see if this affects the risks associated with TW. Early venting nmght cause RHR aird core spray pumps to cavitate eorn reduced NPSH due to flashing within the suppression pool Flashing couldn't occur untli the pool temperature is above 212°F.

Analyses indicate that it would take approximately ten hours for the supworsriý,n pont approach 212cF Verting prior to 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> inoe a TW sequcnce does rot result in PC,,,

llashinq and should not cause an NPSH concern Alter 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> into a TW event. the core spr ay and RHR pumps might tail from cavitation upon vonting However. the. IPE assumes that the RHR and core spray pumps will fail upon venting anyways, with core cooling only available through alternatve means such as the condensate. condonsate booster, condensate transfer, RHRSW, and tire pumps. Low pressure venting could only accelerate this effect The potential earlier toss Of the core spray or RHR Systems duo to early venting would not alter IPE calculated core damaqe frequencies.

Ifthe HPCI Or RCIC systems wore being used for core Cooling. they would be unilfcccte by earty venting These Systems initially takeotheir sucltion from the condensate storage tanks and their suction auto. sovap to the Suppression pool is roqurrod to be inhibi*ed by 1

procodure Early. ieO,.pressure venting does not increase IPE calculatt W.iSk ,

i *.....

New' orkPowerAuthority, JAMES

• FITZPATRICK NUCLEAR POWER PLANT AttachmentlooJPN'91-0G65-page 5 of 18 C, IPE Insights - Long leirm SOO Sequences The FiVP rick IPE did not lake c~redit lfor using the haidened.wo Itwll voi' during SO0 sequences. 8, s -- , on the EOPs, th Operatorts would not vent until the containment pressure approached the PCPL. Studies performed as part of the IPE project show that SOO cote damage following battery depletion occurs 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> after the SOO began. This is significantly less than the 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> necessary to roach the PCPL venting pressure. At the lime of reactor vessel breach, a containment pressure spike may occur. Then containment pressurization would continue at an accelerated pace from the gasses evolved from core-concrete Interactions and wtll eventual!y reach tire PCPL pressure. However, once core damage occurs the operators lose access to the reactor building, and cannort manually open the vent path.

The IPE confirmed the previous analyses which supported tie FiltzPa*rick 10 CFR 50.63 SOO coping duration. As noted above, core damage does not occur untit 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> into an SOO, whereas under 10 CFR 50 63. the FrtzPatrsck plant is required to be able to withsltnd' an SOO of tour hours duration.

D. Other, Non-lPE Insqhts_*Long Te'm SO0 Sequences Ifnew equipment such as a small engine driven source of DC power is irStallted and .venting occurs during a postulated SOO, the SOO is essentially converted into a "T'IV sequence by providing long term core cooling with the HPCi or RCIC systems and decay heal removal to the atmosphere with the vent. (See Section V.) In this scenario, venting is idenlical to that described fot the TWVwith the exception that all valves would hive to be Manually operated.

In this situation core damage frequency from an SOa would te reduced.

As noted earlier, core damage during an SOO occurs long before the containment reachIs the PCPL vent pressure. The Aulhority investigated the possibility of opening the tvtwetl venl early during an $s0 - bolote core damage occurs. In this scenario, the containment is vented signiticantly below the PCPL pressure. The HPCI and RCIC pumps, which draw their water from the condensate storage tanks, woulo be un.ffecled by venting. Early venting does not prevent nor postpone cote damage, since core damage follows directly from station batteries depletion. However, once core damage occurs, the previously opened wetwell vent could scrub the evolvcd gasses in the Suppression pool and reduce the amount Oftissioni products riloased from the containment. Without the vent. IhQ:

containment would eventually fail. releasing the fiSsion products without the benbirt of scrubbing. Therefore, initiating the vent at a lower containment prTss$re Ouiirg an SOO' would reduce the severity of the release to the environment.

E. Potential Vent Modifications The Authority has identitioo several possi$le nardoaio modifications to tne FitzPatriCk plant which may hIelp rThitt3le Cure damage dur4ng OB TIVir, Oai squer*cOS Thiso I

New York Power Authorty. , ,

JAMES A. FITZPATRICK NUEAR ty.RER PLANT..

AttachnenttoJPN-9g-065 , ,

page 6of 18 J

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'moditicatioos,are unreloied to wctwal vent operation an arc'di.lcuss- S incldent management stragegics in SectionV..

The potontial to identify cost be'neOicial rnoditrialions to the vent itself is small, vent modifications c-3nsidored include providing dedicated motive power (AC power andrdor compressed air) for the normally closed valves in the vent pathway. and providing a hard piped bypass around the %TS titer trains Ther are several reasons why these vent modifications are not cost beneficial.

(1) The mean.;ore damage frequency at FitzPatrick duo to SBO sequences is already relatively low at 1.75 E-6/yr. (Reference 1?,FittPatrick IPE, Table 1.4.5.1),

Consequently, the economic value of reducing this frequency is atso low. This is demonstrated by comparison to an earlier NRC analysis of venting at the FizPatrick plant (Reterence 5) This NRC analysis assumed 25 more years 0?Fit.iltPil.

operation and calculated 1637 person-renms averted as a result 01reducing the FitzPatrick core damge frequency by 4.5 E-5/yr. Using a simple ratio of Core damage trcquencic-s, the person-rcms averled by eliminating SBO sequences at the FizPairick F'" plant would b "b:

1 75 E-6 t1637) 63 7 person Forns averlea 45 E.5 At $1.0l per perssot, rem aveted. onrly aboul SG.000v OuId be justified tor the eost of modifying tric Fit-Patrick plant to reduce SOO oftfste hcalth arid property risks to

• zero This ustifiable expenUtture is less than 10% of the V.b0.000 the Authority estimatld to provide a hard pipe bypass around the FitzPatrick Standby Gas Treatment System, (SGTS) tilter trains JRelerenco 3). Although the Cost ol providing 'edicdrtcd ;',w.r supplies to the vent valves has not been determined, it wil most likely also exce-*

564.(J Ho'wever. neither Of the modi*ications reduce the CDF, their ben.elits only' reduce. but not elihrna3te the olfsit dose. Theotoloe, their benotfi*ts are considerably less than S84,000 (2) Based upon expert opinion, the two dornimarit containment failure modes during long term S30 sequences are diroct lincr attack from the Core debris and rapid containment overplersSuritation at tio tIme of reactor vessel breach. (ele0tenc i,1 FitzPatrick IPE, Section 4.7). Should containnunt Strul.ural intogrity be iostl i oithli of th*ose mechanisms, wetwell venting would be ineffective (i a., fission products' '

would escape from the containment thtouh the breach caused by the failuro rather Mtlarn oe scrubbed by the wetwell vent).

(31 Severat accident management strategies based upon insights gained'from lfe" F,t.,Pa'rici IPE are under review (See Section V.) Should any of thC ,g.:..

prove t0 tDetechnlically feasible and cost benelicial, it wilt be considered fur iiprjrpenet,'rr~r* It irnlplomri)rodi l1tIpquorIry of SBO Sequences loading to a core

New York Power Authiorlty..

JAMES:A. FITZPATRICK NUCLEAR POWER PLANT Atftcim~bent toJPN.9f.055 , I page 7 of. t8.

moll would be roduced. These strategies would also reduce the cost benefit of modifying the vent to. cope with SBc conditions.

I Based on the above. haCware moditications to the FitzPatlick hardened vent to cope with SBO sequences are not justified.

i111, Evaluation of the FitzPatrick Plant to the BWROG Design Criteria In Reference 10, the NRC compared the Ft,,Patrick plants existing vent capability I. the BWOG hardoned wetwelt ven esign crilteria The NRC Concluded that tho existing FitzPatrick vent design salisties several o1 these criteria and may have acceptable deviations trom several oth*rs. The tollowing inlotr, ilion Provides the Authority's final position rogmding each of the BWROG criteria.

BWROG Criterion Lal The, vtl shall be sized such that under conditions of (1) constant heat input at a rate equal to one percent of rated thermal power (unless lower limit justified try analysis), and (2) containment pressure equal tc the primary containment pressure limit (PCPLI. the exhaust flow through the vent is sufficient to prc%+ni the containment pressure from increasing.

NYPA Response to Criterion The Authority compared the FitzPatfick v otwel( vent configuration to the NRC approved design for Boston Edison's Pflg-im Plant. The Pilgrim plant is rated at 1998 i h and has a imilrng vent diameter of 8 inches (approximately 50 in2). The Fi.zPatrick plant is rated at 2436 M'fh arid has a limiting vent area through a set of par allel 6 and 12 inch valves

{27MOV.t21 and 120) comprising a total cross section area of over 140 !r;,. "mc!irn1;i1r; FitzPatrick vent flow area Is over 2 3,4 times the size of Pilgrim's. and the limiting flow ar'ea Oer MWth is over 2 t.,4 times greater than that of the Pilgrim design.: However, the FitzPatrick 1B inch watwell vent containment isolation valves. 27A0V-1 t7 and 1 are blocked from opening beyond 50W(Rat. Technical Specification 3.7.D) NeverthesS. the wetwoll vent should be capable o0 passing steam ntows corresponding to greater than 24.36 MVth (one percent 01 rated thermal power). The Authority will confirm the minimum heat removal capability of the hardened wetwelt vent by formal calculation Since the NRC considers the PIlgrim design to be acceptable, the FitzPatrick design, being proportionally larger, should also be acceptable. In addition, the NRC concluded in Reference 10 that the suppression chamber vent path was acceptable for satisfying Criterion (a) when using the main suppression chamber vent containment isolation valves (27AOV- 177 and 118).

Tilts criterion is fully met

i i, ... . . ..

New YoM.PowerAuthrlty':

.JAMEi A. FITZPAT.RICK NUCLEAR POWERPLANT:

Attachment oJPN-91.065..

pageo8 of Ia 1BWROG Curltrion b)

. The haidcnod vent shall be capable ot operating up IQ the PCPL. l1shilt not compromise the existing containment design basis.

BWROG Criterion (ft The hmrd vent path shalt be capable of withstanding, without toss of functional capability.

expected ven*ling conditions associated with the tVJ Sequence NYPA Respore to Crirteria ib) and iM With the exception of the Slandby GaS Tre.,atnmnt System (SO I S) toorn the vent path trom the primary containment to the enironment has 3 oesign pressure riing of 150psig.

which is greater than the design pressure riling 0o the containment and signiticantly groeaer than the PCPL pressure 0144 psig. The PCPL pre'-suie O 44 psig 1s jut thai t1(

56 psig deosign pressure of the cont'inme-nt due to imitations on the closing Capability of the containment vent isolation valves.

The Autlhoriy analyzed the SGIOS and determined that ithi inlet transition pieces are capable at withstanding an internal pressure ot approximately 1,-2 psig. The SGTS filter train enclosures are rated for an internal pressure of a few pstg. At the pressures expected to be encountered durin; TW venting, the transition pioceS may rupture and; possibly. the SGTS enclosure as .r.elI The vent Cftluent woulo then enter the SOTS room at'j, or1ce the room is slightly pressurized, would be reliev*d through a set of double doors that open to the envronment The lunctional requirement of the wctwell vent (ie., to remove decay heat from the containment and provide a scrubbed vent path) would not be compromised.

In Reference t0. the NRC concluded that Ihis vent flow path, including damage to the SGIS. could be an acceptable deviation frOm this criterion Based on the IPE. the Authority concludes that loss of the SGTS is an acceptacie consequence of 5svere accidont ventinq Neverlhetess. the Authority ,s investigating the pOssibilily of Opening the SGTS charcoal filter access do*rs and the SGTS room door to the envirnrnrent pno, It) veting r ficts.in No,,Id make the SGI S inte)iale. therchy hI ehiminarinq the roquilr'J sccordlary Containment ditterential pressure and, coisequently.

making the secondary containmenl inapieatie as well But this action may reduce the darnage to the S3TS during severe accioent venting. Tmis procedural action wilt be investigatedl urthter It the Authorily determines that these actions will prevent SOTS damage during venting and allow the S(TS to be returned to service tollovinq venting.

then it vil bie considered for inclusion in the venting procedure The rtaaric; vent drsiqn is an aCceptable dewiatruri lroni Criteria bI) aind ty)

New York PowerAuthorlty JAMES A. FITZPATRICK NUCLEAR POWER PLANT Attachment to JPN.9 .l065 page 9 of 18 BWROG CriterIon (Cl The hardened vent shalt bWdesigned 1o operate during conditions associatle wilh the TWV sequence. The need lot station blackout venting will be addressed during the IPE.

NYPA Response to Criterion jc.

Venting During the TW Sequence:

The capa)srtty of the FitZPatrick hardened vent to operate during poStulated T'V sequences is discussed in the responses to crileria ta). (b) and (fIl The FittPatrick haderled vent should be capable 0t removing at lenst one percent of thn rated thermal pov.-et and o0 wtlhslandinc the anticipated pressures lot tri Section of tire vent palh within the reactor building Opealior, ul thI valves in the vent flovw path has been cemnstr"lrted since the sjmG path iS uSed 10 depressur'te tire suppression Chamlrel lollowing p*rmary Conta,nment integrated leakage rate tests (ILRT). During both ILRT and TW venting operation, the containment is initially at high pressure (44 psig lot TW and 45 psig fto IL.RTI The flaw path for both modesof operations is virtually identical, initially tr.rough a'2 inch bypass line arounrI ,he 18 inch containment isolalion valves, 27AOV. 117 ard It18, and later through the larger valves The only signiicant diffterence is that during TV' ventinJ, the 2 Inchr bypass line maynot De sufficient to dcpressuri/o the containment and the transition to the larger valves v.ould OCC-ir sooner IN'hS procedjre is5Successlully employed during an ILRT and Should also ;vork for veritinci duting a TV! o.'ent The FitIPitrick hardened vent satisfies lo'A 'ate, pressuie otlention arilopierabilty requatireents during TIW sequences Venting During the Station Blackout (S10) Sequence:

The B3'. sequences requirt* Vssenlliaiiiy tic $iimc tIILlt reliruval I'lte arid &irr'.sure rctenlto,i c.tpabilities Is dr the I"Vsequel ices Tho valve operatcis lor ve*ilhiNg ufrciq SBU must toC done mnuatly clue to Itr lack of motlve poA.Otr.to tM' valves (AC and compresscd air) All of these valves are located in the reactor building find Could be accessed during SEC prior It onsel of core damage. The effectiveness, or need for venting durinq an SAO. ,it oplsed to the Capability to Vent. Is OIScussed InISections It aili V this Cr,1etion 1i fully MCIf 1

New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Attachment to JPN-9t 065, page 100l t8 BWROGCnawrin (d)

The hardened vent s-,all includ* a means to prevent inadvertent actuation.

NYPA Response to Criterion (d)

The NRC'S rev.wt ot this criterion (Reference 10) correctly stated thal -10 prevent rnadvertenl actuation of the vent, the plant relies on operator trai*rlng and adherence to the EOPs." it is highly unlikety thait an Operator would devatoe from thu EOPs The actions neCCssary to initialo ihe vent are signihcantly more complicatea than merely turning a c;ontrol switch Venting the containment requires an exact sequence of mnanipui3tion$s involving six ValveS within the containmrnt vent and purge system. These valves are operated from the primary con'tainment panel in the relay room. not from the control room And in the momt likely TW setquences floss of emergency bus tO150 or 10100). venting wivuld reuire manual operation of many of those ,,Vves in the react0r buldinrg In addaition. dr-'fdvll pressure and possibly high containment radiation instrumentation would be gitnor at;r'iq continuous conlainment rSOlatiOl signals to the vent isolation valvets Trese signralt., voulo have to be bypassed as directed by the ventinq proCedure F AOF Leaving tie conutol room to operate valve controls in the relay roomn, to 'ypas conrtainment ISCi*aion signals. or to manually reposetion valves durirng a severe accident

,ruUld occur oflly urlder the specific direcliun of the Shift Suptrvisor or the Enrerqcncy Director Iradvertent actuation ourting an SBO is even less likely than Outing a TW event. in .50 sequences, every valve manipulation would have to be porlormnO manuaJly. Theelcore. il is highly unlikely that inadvertent vent actuation would lake place during an SBO.. For.

mor0 detailed information. see the FitzPatrick IPE. Appendo% E The Autholriy has deterrnirned thia wradvrltOrt actuation ol.thC* FtPailhCk iirctniiO*, vrirll is higtriy unlikely. Eist~rrg procedure ada trairing provide air adequolle ncalis to prevent inadvertent ventl actuationi Ihis Crt:erion is ftully rne.

BWR_*G cierion le:

The vent pait up to ano including the secono coritaninncnit isolation barrier sthall be designod COn,1i,%rrt Vth the design* bOzsS, Q1trioCplant R.4 PA Respons.. to Criterion ie.,

The irai'JCne * *.Vt oath uSeS the C,.rrenriy .riSisilea suppression Chamcr purge eChaus!

corrlat rlpro pernetral,on The corrta,,rinrnr! iS.-tljf) pro,3s,nns for tijs MnrlrnotOn m LCI

, , ll=l

New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLAN*.T.,

Attachment to JPN.91.065 page 11 of 1' the desiqn Dasis f the FitzPat,,&k plant. The NRC also concluded in Rpference 10 that'Inc existing FitzPatrick'cdesign satisfies Criterion (c).

This Crilerion is lutty mci.

BWP OG Criterion (ta Radiation monitoring shall be proviced to alert conirot room operators o0 radioactive releases during venting.

NYP.AResponse to Criterion ,.)

The currently installed post accident sampting system (PASS) has the capability to sample the wetwo(t atmosp' ere lo tte presence of radionuclides such as noble gases and iodines This system is designed for the environmental conditions associated with accidents and would be available during TWevents The Fit. Pati ck vent procedurp calls for its use stating. "t1 time peimits, the Radiological and Environmental SCr v~IL..s Departmen'ts shoula analy.e the Containmont atmosphere prior to venting to aid in formulaton of prolecitve action rocommen'ations."'

As discussed previously, TW accoent sequences evolve slowIly. providing ample time 10 draw and analyze a PASS sample of the wotwell atmosphere prior to venting. By sampling trom the wotwetl atmosphere, the PASS system wilt analyze the aclual gasses that will be released from the wlotwell vent. Therefore. accurtle venting source terms.can be determined. Once venting has begun the operators will know what is being released and additional samples can be drawn as necessary to deterrmine it the radiological content of the vent eftluent has changed In addition. the PASS System is desUgnod to sample the dcywvvll atmosphr*e Wan suppression pool water for radioactive li*lutds, gasses. and disolved gasses. The drywell PASS sample. in addition to Other accident quahlied containment radiation detection instrumentation. can be used to assess the radrological consequences, vi v~iilii;!.

the drywell atmosphrce has not boon scrubbeod through the suppression pool, those radtological assessments would give cOnServativo indications of whajI would be released IOfrn the wet wvell vent There are conditions under w-hich operators aro directed to vent the drywol, directly (See Section IV ) Under these conditiunt. the Cry-.%ell PASS sample toqether ilth Continfment radiatlion monitocing instrumenial-on would provide drect indication of the radioactivity asoCo cd writh dryv.ell ventinq The Authority cornsioers ttie inst,.*lle raaijitun duteo0 r arid Sampling systems A..oqut,'0 for TW vehiiiTiJ Sairce operation of trio PASS requires AC power, it ,ould niot be available undief S3O concittions

New York Power Authority JAMES A. F ITZPATRICK).NUCLEAR POWER PLANT Attachment to JPN.91l-W5 page 12 of 18 Assuming 6nt AC injopendent dlecicaled v.erting radiation rnonitoring systlem were to bý

installe,.i w.ould have little accident mitigaticn/rnanagernent Ilun iorl. Only 6uIring tt;gso rare, long term SBO. soquencos leading to core cdagM9e whore crlty, tow pressuro wet'wel(

venting is employed (contrary to current emergency procedures) and the contaur j, .n:

does not tail from other SBO phenomna, could signiticant radiation be detected in the vent path.

Even if a dedicated vent monitor detected radioactive materials passing through the vern.

this would not be the basis to' any operator action. The operators are instructedl to vent regardless of radiation releases. In addition, the vent cannr, t rv- manually isolated during the release because of the high radiation levels expected within the reactor building.

Therefore. a dedicated vent radiation monitor would not serve any useful operational purpose.

In the most likcly. long evolving TW and S80 acojnt sequences. emergency response decisions and recommendations to County and State agencies for Sheltering or evacuation of the public, would be based on conserative iadiotogi( .31estimates prior to venting. The actual dose rates and radiological assessments made during ventIng would not have any effect on the previous decision and recommnendations' -DrPx,.,ic protective actions. Radiation monitoring needs associat*ed with offsite emergency responses are met by currently installed sampling systems, instrumentation and olfsite monitoring capabilities.

A AC independent dedicated vent f adiation monitoring system serves no useful purpose in the decisions to open or Close the vent nor in decisions related to oftsile responses. The existing radiation monitoring capability at the FLt Patrick plant ts sufficient lor venting during both TW and SBO events.

The Fi*,PatriCk vent desqn is an acceptable deviation from criterion (q).

BWROG Criterion (Nh The hardened vent design shall ensure that no ignition sources are prtseirt In 1t1e p,; ;. .'..

N Response

.PA to Criteion Hydlogern and other combustible gases would not 00 generated during venting under TVI conditions because core damage would not have occur-ed Thereore. the piesence of ignition sout ces aie unimporlant during TW Combustible gases such as hydrogen or carbon mono-ide are generated only during core damage events such ai SBO Not only ae $SBO sequcnces.uNtikely. but by their very nature they mingmie concerns about equipment betinq energied (,qni*ion sources) in the vent path including the SGTS room Once AC powe is Ietored. accodent mitigation activities ,fOuiJ reduce or e01110la1e COrnl)'usible g9,rs t15

.rrier rtoi

New York Power Authority.

JAMES A. FiTZPPTRICK NUCLEAR POWER PLANT Attachment to JPN.91-065 page 13 of 18 In letters previously providod the.NRC on the hardened vent (References and 9), the.

Authority notedt that:

Combustion wtthih the vent piping and SGTS room is unlikely due to high steam concentrations.

Even it combustion were to occur, Itis unlhkely tO Cause structural failure of the SGTS room, particularly failure of the 2 fool thick reinforced concrete common wall between the SGTS room and the reactor building, and Even it the SGTS room failed structurally. it would be unlikely that this failure would propagate Such that the primary containment would also fail.

The alternate action would be to install a hard pipe bypass around the SGTS at an estimated cost of S60,000 (Reference 3) Since combustion in the egisting vent path is not risk significant, the Authority does not plan to modify the FityPalrick vent- design to reduce ignition sources The it,?Palrick vent design is an acceptable deviation from criterion th)

IV. Examination of Ventlns Procedure, and Training In Section 2 0 of the Enclosure to Reference 10, the NRC concluded that.

(1) operators are untiained regarding venting consequences and do not expect a rupture in the SBGT porton of the ventfiiq path, ay:

lej Operalo'S ate not tamiliar with other methods expec. -J to be employed to stretch out the time to reach containment tailure pressure and otner decay heat removal pathways.

(3) present simulator scenarios ivo'ving loss of lecay heat removal

!,equences do not result in containment venting: ano 141 procoeuraj guidance is not provido: to dctermin vtihen to secure venting once it has be-en stanoe In adlditon, the p'ocedures do rnot clearly irndicate the conditions which would require ubc of the diyo(ell. suppression chamber., or both. vent paths Also. F.

AOP.35 contains human factors weaknesses v,hrch could prove detrimental to operator uSe 01the procedure W*'h tespecl t*issue, I. the imorc general ISSue *"uperaftiairarnq ii tire v3riOus severO accident ris'iqnls atro unhctomCna icveiled througjh lile IPE process, Flcomjurratirions lot ope*atui t*raor i.rth iSL)t to the resulis ol the IPE have tx-eii i*Oentiti anid are Leng

saw New York Power Author(Ity JAMES A. FITZPATRICK NUCLEAR POWER PLANT Attachment to JPN*9t-065

. . . . p..* .4o116,.. * ,,

integrantld into the. operator training program. Included wIt this-trainring willL .erqining in sev'Ir.o during spyeto.accidcgs.

accident phenomena such as the conseqpences of venting Item (2) is more appropriately handled under the accident management proqqram which in beineg cOordinated by utllies through NUMARC and the NRC. In any veortl, both TW and SOD sequences evolve slowly and, even without employfng other melhods ot heat removal. rcquite approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> to reach the PCPL pressure. This provides ample time lot operator action. The IPE dd not consider means of extcngdnq the timo to rcacti the PCPL or Conlainment failure pressure or alternate methods otrheal removal necessary. Siccuss'ul ope.ation, o( tho FitPatrick vrnt resolves any concern about reaching containment Iailurc pressurc, climinatling the ne6d to proviing measures to stletch out the tcmo.lo.caCh 11iis prcssurc.

With regald to pre(set t simulalof scenaros in ttrem (3), 1he Fit/.Patrlct si,'ulator CanDnot simulat0 TWV sequences which irquirc venttng. The FitzPatrick simulator is designeo to meet the ie(.uiitemeritf ot ANSI.'ANS.3 5.1985. °Nucleai Power Plant Simulators 10r Use in Operator Tiatrtjig' w? ch has very specific iequi:cmcnts lot the capatbiliies ol the Srmri-latui and for val)daolio Cof,irrulalo n)odels Opeialors ate not simulalor trainecd wih unviilnt.td simulatllor models bOcause modeling inaccuiaces may mislead them. The simulator sip" "ession pool tem'peraluro model has not been validaled for beyond design basis events tabovc ap. axim....:

1,0"LF). Such as wo-ld t1 qeCnrte* by long lern TW sequences. This lihniatioti( 01 little siqvificance for sCernanos which evollve slowly. es*ocially where enier.gercy proced ires aie available arid operators well trainr*. Io addition. thie actual valve manipulations to etffcct -venting

'cannot be perfomedf ir. the simulato( becausc the vent valve control switches ar not l*oca*tcejd i tfie cont(rl roor'r ;ind. therofoic. are not pr)vj- Inl the siniulatot O,.,'itos ae fujlly cl,*,,srto,.on trained on lfioe porti*os or the F..OPs ,hich rpre.ent beyond dcSqn Ltasis evetSts With respect to lieni (4). F AQP *5 was siqr,*iicanily revtrod. p(oviuinq more djeailed rstlrhichovis and impioving huinart factors conis4eations. Subsequent to the Nih's viS,I to the Fit P,1,iicnI pla1t (Relerericeo) aru th**r ferviw o1 Itte procedure. The revised tormuat ot F AUi 35 sho'uld alleviatlo he htunrar flactors weaiknostes noted by the NRC F-AOP.35 MrWS ur oralors tovn*tl ootly a5 ni.cess-ry to redi airi*

in imitain P)lrriary :

Containmem pressure below the Primary Containment Piessue limit A piusse coControl bandr Should be LroSon Such that f1fS.t1 .iaitologtcal consequences ,yre expected to ' iin-imh'.c1 "

This rNiic,-Owit ,imi t1 ievised hurnr'or to preovye moie 51pocitic *di*xodce -?1SC b.) i 1110. 1t1hi i 'n 01 Ihte plre'....u i' ir .ti liar (N I t,( v.'vr .c, rIinirtjliri P C(ulI1 hivr; 1o .cconor fini tter.rii l)('5SUUri/,i4I)ii ',iln , I, u)n irmaniy poY.lij*l.ili ,Kr:Ldeitr iti ' ,llld i I r ve.y i.itCrsnrrue N!ui1 and WOUId IrIoi ue vent opr*f ;itilon I'rU' The Authorit1y Is trive*stgilrrq.g .ilether this pfo..e'dute shOutld be mnOdilioc to direct tre operator to Contirnue ven*t(i' until the corttainnient has been depressuh.,od to near ar*rrtsphefhc: pressure instead of using a pressure control batnd there arc only n iinim, adverse radiofologcal cortsequertcreti Ittic 1W reqn.ernce by alio,%,rrq ihc vent to remtanrl open until nicar atmospheric cofn*rirs iti fe.oa.t*h- in the currilrtnir t*ri ) Of Other Core danrle ovents. this opproach It1 hf j ;i ,Ar it ki.f pr*i)fNirrl t r*.*efi.fit Tr*P ~iait'r['I ve'tiiiK(j at lOW COiit.iriflilit p.rit!ýSUIe iS rtirCti'...l1 i ii I -- i ,'.n. '-i V

' ' " '. I . .. I "  !' . !

New York Power Authority JAMES A. FITZPATRfCK NUCLEAR POWER PLANT Attachment toJPN.91,065 page 15 of 18 FAQP.35 is very specific on when the venting.of the suppression charpber or dryweall hallbe conducted: This procedure states. *Aprimary.Containment vent path taking its suction on the Torus airspace is preferred due to.an expected decrease in radioactive release rate due t.6 PO61 scrubbing of fission products.* Itthe suppression pool wat~r leve' is abovp 28.5 tfet [ust Wf ow the height of the vent penetration[, then the procedure requires that venting be conducf'd directly from the oryvaell. regardless of the radiological consequences. There are no conditions under which simultaneous venting frnm both the suppressicn chamber and the drywall N,'ould be required. Furthermore, the flow limiting valves in the vent path are common to bDth the suppression chamber and crywell vent paths. Opening both paths simultaneously does not increase the limiting flow area and should not significantly increase the overzail vent floW rate.

V. Accident Management Strategies The Authority is currently examining several accident management strategies. Neither the feasibility nor the cost effectiveness of these strategies ha, -, been determined. .Hoinever, preliminary analyses indicate that should those strategies br- implemented, the risk significance of SOd sequences could be lowered by a factor of about 4 to 6.

One accident management strategy is to provide a small engine driven 9vncratol to supply GOO VAC power to one or both battery chargers or to supply 125 V\DC power directly to one or both DC systems. The Fit2PatrIck IPE verified that SO sequences fead to coTr damage due 0o inability to maintain core cooling once OC power is exhausted. Use of an addotional electric genor ator could extend the availability of DC power and prevent core damage.

Another fPE insight is to install a crossfie between tho fire protection system and the emergency service water system jESW). This will allow one or both of the diesel dinven fire pumps to supply cooling water to one or more of the jacket water cooling heat exchangers in the lout emergency diesel generators jEDGs). Based on the FitlePatnck tPE, loss of ESW was a mator contributor to EDG urnavailability. When at least one EOG becomes functional. core cooling would be possible through use 0f the core spray 4,ystem or the LPCI moce of RHR. thereby lerminaling the SBO.

A single operating diesel generator can power an RHR service water pump anr alt necessary auxiliary equipment in addition to the RHR and core spray pumps to provide simultaneous core cooling and cr ntainmenl heat removal. With an EDG providing AC power, there is no need ;o initiate the vent, and it the vent is already open, there is AC power to close it. Three rmdes of ECCS system operation can provide the heat removal path from the reactor COie to thfe ultimate heal sink (Lake Ontaro) using a single EDOG. shuldown cooling, suppression pool cooling, arid contanmin1 spray throughi an PRR heat oexhangoe in shuldou6.n cooling moov ar, aIlltRt pinup Cilculales reactur Coolant through atr RfIR heat excnangei and thncy back to the reaclor vessel In suppression poo co0ing mode an flHR pump cliculateS wafer from the suppression pool through an RHR heat exchanger and then back to thi sup*iess~On pool In containment spray mode. an RHR pump circulates water from the sukppiesioir pool through an RHR heat exchanger and then through the containment spray headers. The t,.at(e then collet-5 on the floor of the drywtll and spills back into the suppression poCotlhrouqh thc o;,incomers Trey,r. iast two modes reauire using the care spray system to

New York Power Authority JAMES A: FITZPATRICK NUCLEAR POWER PLANT Attachment to JPN.9 1-065

., *. pag,-.t ,.oTB. .

provide core cooling in all three of these modes, an RHRSW pump providcs cooling waler from Lako Onjanio to the tube side of the RHR heat exchanger and relurns the heat ed water back to the lake.

Another IPE insight is the use of the existing fire protection system crossfio to the RHRSW to provido CoOling water flow to an RHR heat exchanger instead of using the RHRSW pumps. Tho crosslie was originally installed to provide cooling water to the reactor COre during SBO conditions. This now use of the crosslie would mitigate certoin TW sequences where an RHR pump is available, but the corresponding RHRSW pumps are not. This strategy can be implemented through simple procedure changes since all necessary crosstie hardware have been previously installed and only new valve manipulations are required.

The Authority plans to bring several other venting strategies to the attention of the BWROG for generic consideration prior to any decision to implement on a planl specific basis. TeO of these strategies are early (low pressure) ventlnq during SO and venting the containment to near atmospheric pressure instead of maintaining containment pressure just below the PCPL. These strategies were rnenlioned previously and are described in greater delail below.

The first strategy involves manual opening of the vent valves prior to the containment pressure reaching the PCPL during an SBO if loss of core cooling is imminent. Due to the long tine between loss of core cooling arid core damage during SEO sequences, there should be sufficient Time To manually open The vent valves. This would not significantly increase risks, and could be uselul in SBO sequences it the containment does not fail early from other SBO phooomena (e g, direct drywell liner attack). In fact, venting in SBO sequences. may else reduce the probability of early containment failure from postulated conta:nment pressure spikes at the time of teacti vessel failure. Rogardless of when the vent is opened, the suppression pool would be available to scrub radionuclides from the vent effluent. Therefore. the release to

. the environment would not be adversely affected by initiating the vent at a lowec containment

. pressure during SBO. 11this strategy were not implemented, then adherence to the cu'rrent EOPs and postulated radiaological conditions would prohibit any vent operation during an SBO.

1Tho socond stralugy is to modify vent operation such that instead of mainl.tining a containment pressure control Dond, the vent would remain open until the containment -s fully depressurized.

This mode ot vent operation would apply to TW sequences where core cooling is provided throughout the vent operation. Itearly venting during SBO is ;mplernented as described Rbove, then orice core darrage has occuired the vent could nol be reclosod duo to fadiological conditions within the reactor building. Nevertheless, lt)eta are sjignilicant boenfi*s tO maintaining fthvent open during both TW and SBO. Maintaining the venT open will:

Reoajce the rumt)er of op*eraTor actions and vent valve manipulations. Theie would be no concern tlmt the vent could nrt be reclosed, and it reclosed, that it could not be reopened 05 would be nrý.cLIa* To0 maintatin a pressure control band. This improves the recahihliy 0f the Verit

. fulqrre Tn'j.t

' ii nbr:i tca.kýCf Ifuon, the writ air nrurit

. . NeOy York PoWer Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT

',,- *i... Attachment to;Jf, N-9.1-O65 .'

• i , .- ... , .

page i7 o7 18 Minimize the driving forde (pressure 'di ference between ihe cohtainment and Mte atmosphere) for fission product release through other openings in the containment.

• Cool the containment and reactor vessel surfaces to retain radionuclides. And, S,'crub all vent effluent through the suppression pool to lap the maximum amount ot

'%dionuclides.

VI. Conclusion The existing FitzPatrick hardened wetwell vent is adequate to meet the accident conditions associated with TW and 580 conditions. It meets many of the BWROG design criteria and represents an acceptable deviation from the remainder. The hardware modifications needed to fully meet the BWROG design criteria are not necessary to ensure that the vent. performs its decay heat removal and scrubbing functions and would not produce significant public benefits.

Procedures and operator training with respect to vent operation are adequate and the Authority is fully confident that plant operators will use the existing hardened wetwell vent when plant conditions dictate.

The hardware modifications associated with the accident management .tralegics described in Section V are unrelated to the hardened wotwell vent and would be preventive in nature, These strategies are currently being evaluated, with the fire protection to ESW crosslie modification expected to be installed in the near future. Accident management procedural changes which have the potential to improve the operation of the vent will be brought to the attention to the BWROG for generic consideration. When these operational changes are approved by the BWROG and NRC, the Authority will provide the NRC with an implementation schedule.

VII. References I. NRC Generic Letter 89.16, "Installation of a Hardened Werwell Vent,' clalti Saptemuor 1, 1989.

2. NRC memo. B.W. Shoron to kC. Thadani, *Reduction in Risk from the Addition ol Hardened Vents in BWR Mark I Reactors," dated October 19, 1989,

3. NYPA letter, J.C. Brons to the NRC, (JPN-89-070) providing the Authority's initial response to Generic Letter 89-16. dated October 27, 1989,

4. NRC memo, W. Minners to A.C. Thadani, 'Draft Environmental Assessment and Plant-Specific Regulatory Analysis for Instalialion of Hardened Vents in BWRs With Mark I Containments." dated January 8, 1990.

Ir,*,v~ J- .. .. . . I -II

New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Attachment to JPN-91-065 page 18ot 18

5. NRC letter, T.E. Murley to J.C. Brons, -Staff's Backlil Analysis for James A. FitzPatrick Nuclear Power Plant Regarding Installation of a Hardened Wetwell Vent,* dated June 15, 1990.

6, NYPA letter, J.C, Brons to the NRC, (, iN-90-M55) ptoviding the Authority's comments on the NRC's Hardened Vent Backfit Ai talysis, dated July 25, 1990.

7. NRC letter, D.E. LaBarge to J.C. Brons, 'Topics for Discussion During Planned Site Visit to Address the Hardened Wetwell Vent at the FitzPatrick Nuclear Power Plant,' dated August 17, 1990.

8. NRC staff visit to the FitzPatrick plant to gain further insight into tho desiqn of the FitzPatrick Wetwell Vent Path, hold on August 22, 1990.

9. NYPA letter, J.C. Brons to the NRC, (JPN-90.061) providing two supplementary Authorty's analyses to the NRC staff, "Radiological Benefits of an Elevated Release" and "Combustion MAlysis," dated September 7,1990.

10. NRC letter, S.A. Varga to J C. Brons, 'Installation of a Hardened Vent Capability at the James A. FitzPatrick Nuclear Power Plant," dated January 24, 1991.

11. NYPA loller, J.C. Brons to the NRC. (JPN.91-048) providing the compl*etd Level I and Level If IPE for the James A. FitzPatrick Nucicar Power Plant, dated September 13,1991.

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• ~U.S. Nucl'ear Requlatory Commission ATTN: Document Control Desk

.Maji St~itiun P1-137 Waqhnqtn, C. 26555 ,

SUBJECT:

Jkmes A. FruPatrick Nuclear Power Pl...."

Docket No. 50-333 HordtancAweltUVmnt -fRviji Referenwe NRC letter, R. A. Plasse to R, E. Beedle, 'equest for Additional Information -Nardened Wetwell Vent Capability for the James A.

FUTuPatrick Nuclear Power Plant (TAC No. M823641.' dated July 2, 1992.

Deal Sir:

In a May 19. 1992 conference call with the NRC stall, the Authority agreed to provide additional information regarding the FitzPatrirr -. etwell vent design. The referenced letter details the specific information requested by the NRC in order to resolve the outstanding concerns of Generic Letter 89-16 for the FitzPatrick plant.

This letter provides the Authorty's response to these three concerns.

I) Perfurm the calculation to confirm the minimum heat removal capability of the hardened vent and provide the results to the NRC stall.

Thr. calculation has been performed. The calculation determined that one percoret decay heat 124.36 MW) produces 25.183 lbmisec of steam at 44 pslg Wth:f PCPL piessure) by evaporation from the suppression pool, at a rate of 269.964 lt'Isec 'This is the volurnetric flow rate of the vent necessary to prevent one percent doecy heat from causing pressure to cuortinue to increase within the containment. Assuming a vent effluent of puro nitrogen (the initial gas contained in the wetwelli. a vent mass flow rate of 44.21 Itff*sec is required. Although the acttual %trit ftfflutrnit would be a moikh e of riltioger dtrd other gases. prirrlarily ICeBar the highrer ditrnst y of nit r Ufei li;,ls Iit )n.*eivahive results in this calculatrio).

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The venting procedure direcis venting through parallel 6 and -12. inch. lines.

This parallel path, as well as each individual path. were evaluated for thei,r ability to pass.the required 1low tale. Venting through the 6 inch line alone is capable of passing 17 Ibm/sec. whichis insufficient for reducing the containment pressure.

The 12 inch line alone is capable of passing 71 Ibm/sec. which is significantly greater thin the 44.2 bin/sec required. The combined flow through both linesis 78 lbm/sec. The calculation concludes that veeting through the combined flow path.or only.through tihe 12 inch line is sufficient to remove one.pefcent .deciy heat. Thenflore, the FitzPatrick design hully meets the BWROG hardened wetwell vont heat removal design criteria.

Because the vent path can pass signiticantly more flow than required (approsinrately 60% miargin for venting through the 12 inch line aloine) ample vent ciipa(ity is avijilithle to Support the pli;nned 4% power uprate. This will be formally

2) 1ho vent path at FtiPatrfik for the wetwell may tlclude up to 7 different valves. -As:,tiiurcv; ol virim operatiin is vital to tile surcsess Mf the lrzirdoned vent opuraituti. Provide confirmation that (tie valves iised in tihe vent path am- cptoblitir of opeatilon ,up) to the PCPL (44 The upwrability uf .he wetwell vent icontaininent isolation valves (27AC-V-1I 7 anid 1181 were factored into the retlemination of the primary containment piessuire Imit (PCPLI. These large diameter hiaterfly valves required modification of the shallt tu disk pin to ensure their ability to open and close against a differential piressure of 44 psig (the PCPL).

The smnall diamneter cr.nluilaiieit isslaioni bypass valves I27MOV-1 17 and 1231 have dtsiut'n pressures in excess.of the PCPL. 27MOV- 117 is designed to opeaote against a diflerential pressure of 56 psig. the primary contuinment design pressure. 27MOV 123 was origirnally designed for a PWR primary coolant sysimc apphlcation with a design differential pressure of 1621 psig. These two valves are opierrer to rJm.pressuizre the containment following conduct ol the pornmay corrtrirnerrt intiegated leakage rate test IPCILRT) ftom an initial pressure of.45 psig. Thirefure. thrilr ability to open frOmri a piessurir in excess of the PCPL is peCodir; l;ly delmOrIStr alted, thu pr;i;illu 6 and 12 inch valv.5 liuiitlwi driwnstretin in tle vent path 127MOV. 121 amid 120) were desigiied to ovr.malt ;ringainst a differeintial pressure Ul 10^ W.G. 1tie mindatory opening sequence for thi,. vent valves requires-tlese valves to be opened before the conrlallomrtrii isrnilni.oni valve% to ensure that the valves which can handle the PCPL pressure are operred la8S.

The: urily other valve in the vrlt 1jt1h i.%one of the slandby gas treatment system iSGTS) irnlt isolation valves (01 125MOV. 14A. B). Siince the vent pro oini emrliiius one itain of fhe SGTS to fm sit service prior to venting. ti s valve di*i li*i

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Wvillaiiieady be0 Upell p)ilot l0 iopeling ally ofi the1 witir vaipes Ill the Vent ptrti zind will not bt' exSpo ed to sigriti fr; an diffto nlia I pi tssr i Therre fore tflre Fit zPa trik iiisgt ileeS. (he DW.ROG 1air da'ic ( wetwu I vulit crieria or1 b~eing capable of oppunnig~ up to tire PCPL.

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in the contaniunt al mosphrer, "Thon, ivwiding by II)r vent flow rate Icc/5sec. and considering the sciubbing etfecls of tho suppresson pool, a release lato ICisecI can be 0elermined, Once held teading'. are obtained. the release rate can be refined by back calculatvon. 1ifrs accidenrt uniontoring eq*iprmentr mees the minent of the E3WROG omaritea to provide radiatuin rmonitrririrj of the vent affluem.

It yDou have any rjue*,tlins. please contact Mr. J. A. Gray. Jr.

Veiy tfuly yours, Ralph E. BSUL.dle E xui:w.tjve Vi(.(! Pfesidemt Ni~arf G uneIwrtofIr CC: Regjiornal AdmtiO'Itr at or U.S. Nuclear eqUioatofy Commrns'ui.r

'475 Atlendale Road King of Prossia. PA 19406 Office of the Residenm Inspector uVS. Nucl*rlat RPgilat ory Cfmnmnisssion P.O. Box 136 Lycmitrir)Jg. NY 13093 Mr. 811m) C. MtCrtru Prtojec t Dorr.cltate I I DivisioIr of Rei mtetr PlIoLfis" 111t UVS. Ncle:rw Rrr*ulat ry Conl 'i*in, iii M:ad Stop 14 82 Washiligt*in. DC 20555