ML101390318

Calculation 80096650, Revision 0, Leakage Reduction Program Calculation, Attachment 6 to LR-N10-0163

ML101390318
Person / Time
Site:	Hope Creek
Issue date:	11/15/2009
From:	Gita Patel Public Service Electric & Gas Co
To:	Office of Nuclear Reactor Regulation
References
LR-N10-0163 80096650, H-1-ZZ-MDC-1880, Rev 3
Download: ML101390318 (20)

Text

Attachment 6 LR-NIO-0163 HCGS 10CFR50.59 Evaluation No. HC 2008-215: "H-1-ZZ-MDC-1880, Revision 3"

50.59 REVIEW COVERSILEET FORM LS-AA-104-1001 Revision 2 Page 1 of 4 Station/Unit(s): Hope Creek Generatinp, Station Activity/Document Number: 80096650

Title:

Leakage Reduction Program Calculation, Revision 0 NOTE. For 50.59 Evaluations, information on this form will provide the basisfor preparing the biennial summary report submitted to the. NRC in accordance with the requirements of 10 CFR 50.59(dX2).

Description of Activity:

(Provide a brieft concise description of what týie proposed activity involves.)

The following activities are performed:

Increased the allowable Engineered Safety Feature (ESF) leak rate to 2.85 gpm Updated the Main Steam Isolation Valve (MSIV) lealcage release model Increased the primary containment isolation valve (PCIV) maximumn isolation time to 120 seconds, and Revised the offsite and control room doses, the doses for the vital area missions, and the doses for areas requiring continuous occupancies to reflect the preceding activities.

HopeCreek Calculation H-1-ZZ-MDC-1 880, Revision 3, evaluates the post-LOCA offsite and control room radiological impact of the following three changes:

1. Primary containment isolation valves (PCIVs) are proposed to remain open for 120 seconds during a LOCA. This change introduces a potential radioactive release path to the environment through the open drywell and suppression chamber purge exhaust valves.

2.- Increase in the allowable ESF leak rate from 1.0 gpm to 2.85 gpm.

3. Update of the MSIV leakage release model to the current regulatory accepted model. The model changes include:

a) Credited the elemental iodine removal by the containment wetted surface area b) Revised the aerosol gravitational deposition in the MSIV lines beyond the outboard MSIVS to account for the finer aerosol particles by creditinga smaller aerosol removal rate than that in the current analysis, c) Modeled less elemental iodine removal in the main steam lines than that in the current analysis, and d) Redistributed the remaining MSIV leakage of 100 scfh in one intact main steam, line instead of two steam lines in the current analysis, The above changes result in decreases in the offsite radiological consequences, and an increase in the control room (CR) radiological consequences. The increase in the CR radiological consequence is both less thali the regulatory allowable dose limit and can be defined as a minimal increase per the guidance in the 10 CFR 50.59 resource manual.

In addition to Calculation i-I--ZZ-MDC-1 880, Revision 3, Technical Evaluation DCR # 80096650-02 10, Rev 0 was originated, to determine tlhe desiga functional impact on systems & components located downstream of the outboard PCIVs, which are expected to remain open for 120 seconds during a LOCA and exposed to peak LOCA pressure and temperature. The evaluation also assesses the impact of the increased maximum isolation time on the Primary Containment Integrated Leak Rate Test (PCILRT) Program, post-LOCA EQ temperature and doses; and structural integrity of system exposed to the post-LOCA peak pressure and temperature higher than the design condition.

Hope Creek Calculation H-1-ZZ-MDC-1923, Revision 2, evaluates the post-LOCA doses to area requiring continuous occupancy at the Technical Support Center (TSC), Guard House (GH), and Operational Support Center (OSC) due to changes in the post-LOCA release. The resulting TEDE doses are less than those calculated in the current analysis.

Hope Creek Calculation H-1-ZZ-MDC-1927, Revision I, evaluates the post-LOCA mission doses to various vital areas due to changes in the post-LOCA release. The resulting TEDE dose rates are less than.those calculated in the current analysis.

These changes hereafter are collectively called "proposed activity."

Reason for Activity:

(Discuss why the proposed activity is being performed.)

50.59 REVIEW COVERSHEIET FORM LS-AA-104-1001 Revision 2 Page 2 of 4 Station/Unit(s): Hope Creek Generating Station Activity/Document Number: 80096650

Title:

Leakage Reduction Program Calculation, Revision 0 The maximum PCIV isolation thiewas increased because compliance with the instantaneous closure times .of5 seconds listed in Hope Creek Technical Requirements Manual (TRM) Table 3.6.3-1 consistently became a difficult task for large bore PCIVs, mainly the drywell and suppression chamber purge supply and exhaust valves. The adverse impact of current instantaneous closure times is as follows.

• Large momentum associated with instantaneous closure time causes valve.seat damage.

• Valve seat damage adversely impacts containment leak rate characteristics, long-term reliability, and PCILRT.

• Long-term valve seat damage results in an expensive valve replacementjob.

The benefits of extended closure times are as follows:

• Permitted by the NRC in Regulatory Guide. 1.183, Section 1.3.2, which allows -the licensees to increase the PCIV maximum isolation time to 30 seconds without reanalyzing the design basis LOCA.

• Improves the long-term reliability of the PCIVs for the entire plant design life, including life extension of the plant Maintains the leak-tight containment pressure boundary and thereby improves the PCILRT results

• Provides operational flexibility..

, Eliminates potential for expensive valve rpja~cem~ent costs Eliminates potential for. a forced outage associated with the PCIV repairs Per RG 1.183, Section 1.3.2, for the selected timing characteristics of the Alternative Source Term (AST) methodology, e.g.,

change in the closure timing of a containment isolation valve, re-analysis of radiological calculations may not be necessary if the modified elapsed time remains a fraction (e.g., 25%) of the time between accident initiation and the onset ofthe gap release phase, which is 2 minutes or 120 seconds (RG 1.183, Table 4). This means that the NRC Staff allows the licensees to increase the PCIV maximum isolation time to 30 second (0.25 x 120 seconds = 30 seconds) without reanalyzing the design basis LOCA.

For longer time delays, the regulatory guidance requires that the affected design basis analyses are to be re-calculated, all affected assumptions and inputs should be updated and all selected characteristics of the AST and the TEDE criteria should.be addressed.

The PCIVs listed in the HC TRM Table 3.6.3-1 have been relocated from the HC Technical Specification by the HC operating license amendment No. 171. The PCIV maximum isolation times are proposed to increase up to 120 seconds due to the associated benefits listed above. The HC TRM maintains the maximum isolation times for the PCIVs, therefore the change to these times does not require an operating license amendment and NRC approval. The proposed changes can be adopted under the provisions of 10CFR50.59 guidance.

The review of the proposed change was performed using the applicable P&IDs to determine whether any of these open valves establishes a direct release path to the environment that bypasses the reactor building. The review indicates that the drywell purge exhaust (Penetration # 23, Isolation Valves GS-V024, V025, & V026) and suppression chamber purge exhaust (Penetration# 219, Isolation Valves GS-V027 & V028) could establish a direct release path to the environment during a LOCA.

These purge exhaust isolation valves are proposed to remain open longer than 30 seconds; therefore, the evaluation of the radiological consequences became necessary for the 120 seconds closure time in H-1-ZZ-MDC-1880, Rev 3 to determine if the increases'in the total dose consequences are less than minimal dose margins and if the total doses are less than the regulatory allowable limits to adopt the above changes for the current 6perating license under the provisions of 10CFR50.59 guidance.

A technical evaluation was performed to determine the design functional impact of systems & components located downstream of the outboard PCIVs, which are expected to remain open for 120 seconds during a LOCA and exposed to peak LOCA pressure and temperature. The evaluation also assesses the impact of the increased maximum isolation time on the PCILRT. The gas filled systems and components are exposed to the post-LOCA peak temperature and pressure, whiile the PCIVs are remaining open for 120 seconds during a LOCA. There is a potential impact on the systems and components design functional requirements due to the additional exposure to the post-LOCA conditions beyond the system design condition while the PCIVs remain opened.

50.59 REVIEW COVERSHEET FORM LS-AA-104-001, Revision 2 Page 3 of 4 Station/Unit(s): Hope Creek Generating Station Activity/Document Number: 80096650

Title:

Leakage Reduction Program Calculation. Revision 0 The allowable ESF leak rate was increased from 1.0 gpm to 2.85_gpm because plant maintenance has determined that meeting the 1.0 gpmn leakage limit is difficult, a higher leak rate is believed to be bounding for future plant maintenance surveillances,,

and because the resultant total doses would not substantially increase above the current reported UFSAR doses.

The MSIV leakage release model was updated.to be consistent with the current regulatory MSIV leakage model accepted and implemented to reflect the most recentNRC guidance as promulgated through NRC reviews and acceptance of the MSrV leakage model for the Peach Bottom Atomic Power Station. The adoption of the lately developed MSIV leakage model is appropriate to address the NRC concern about the lightly packed aerosols behavior in the mrain. steam lines beyond the outboard MSIVs along with the reduction of tie MSIV leakage based on the drywell pressure and temperature. The MSIV leakage model in the current analysis is extremely conservative, which unnecessarily expended the CR dose margin without having any prudent benefits. The newly adopted MSIV leakage is still conservative, complies with the NRC defense-in-depth philosophy, and is beneficial to the CR dose margin.

The radiological evaluation in H-1-ZZ-MDC-1880, Revision 3, determines that the total increase in dose consequences is minimal and that the total dose consequences are within the regulatory allowable liniits. The technical evaluation concludes that the integrity of systems d0Awstream of the outboard PCIVs are maintained without any adverse impact on their design functions time during either totally eliminates or substantially reduces the large bore valve Seat damage resulting in a leak maximum and the increased isolation tight pressur.e pbotmdary the PCILRT and following a LOCA.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases., or safety analyses described in the UFSAR.)

Post-LOCA exclusion area boundary (EAB), low population zone (LPZ) and control room (CR) doses are dependent on the activity released to the environment via different release paths. The inclusion of an additional bypass release path through open PCIVs, the increase in the allowable"ESF leak rate to 2.85 gpin, and the updated MSIV leakage. release model, collectively reduced the offsite radiological consequences, and increased the control room radiological consequence. The total CR dose consequence and increase in the total CR dose consequence are both less than the regulatory allowable dose limit and can be defined as a minimal increase per the guidance in the 10 CFR 50.59 resource manual.

The increased maximum isolation time provides the operational flexibility and reduces the refueling outage critical time, and costs by either eliminating or minimizing valve seat damage and the need for repair or replacement of the large bore PCIVs having a virtually instantaneous closure time of 5 seconds. The containment pressure boundary can be tightly controlled during a LOCA to reduce the resulting dose consequences.

The increased ESF leak rate also provides operational flexibility by minimizing the likelihood of failed leak rate surveillance.

The structural integrity and. design function of the systems downstream of the outboard PCIVs are not adversely impacted by their exposure to the post-LOCA peak pressure and temperature while these PCIVs remain open for 120 seconds during a LOCA.

The proposed change neither modifies the plant equipment design functions nor impacts the equipment reliabilities. It requires the revisions of valve testing procedures and HC TRM Table 3.6.3-1 for the increased closure time.

The post-LOCA dose rates to various vital access areas have been reduced with con-esponding increases in occupancy times to perform the vital functions. The reduction in the post-LOCA vital access area dose rates and increases in occupancy times are not considered adverse because they are beneficial for the performance of the post-accident vital functions. Therefore these changes are screened out and are not subject to a 10 CFR 50.59 evaluation, even though the changes call for the LOCA safety analysis to be updated.

The Hope Creek UFSAR Change Notice No. HCN 08-028 identifies appropriate UFSAR changes.

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 2 Page 4 of 4 Station/Unit(s); Hope Creek Generating Station Activity/Documeni Number: 80096650

Title:

Leakage Reduction Program Calculation, Revision 0 Summary of Conclusion for the Activity's 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

The reanalysis of the radiological consequences of a LOCA to include the addoidonal bypass #elease path.thr6ugh open PCIVs, which remain open for. 120 seconds durhig a LOCA, to increase in the allowable ESF leak rate to 2.85 gpm, to update the MSIV leakage release model, and to revise the affected doses to various vital areas-combined to result in decreases in the offsite radiological consequences, an increase in the control room radiological consequence, and an increase in allowable occupancy times to perform variousvital functions. The total CR dose and increase in the CR dose consequence are both less than the regulatory allowable dose limit and can be defined as a minimal increase per the guidance in the 10 CFR 50.59 resource manual.

The reanalysis does not:

0 Adversely affect UFSAR described SSC design functions 0 Adversely affect how UFSAR described SSC design functions are performed or controlled 4 Result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses e Involve a test or experiment not described in the UFSAR 0 Increase the frequency of occurrence of accidents

• Increase the likelihood of occurrence of malfunctions

• Increase the consequences of a malfunction 0 Increase the possibility of an accident of a different type than is already analyzed in the UFSAR

" Create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR

• Result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered Attachments:

Attach all 50.59 Review forms completed, as appropriate.

(NOTE: if both a Screening and Evaluation are completed, no Screening No. is required.)

Forms Attached: (Check all that apply.)

[]Applicability, ,Review 50.59 Screening 50.59 Screening No. N/A Rev. N/A

[ 50.59 Evaluation 50.59 Evaluation No. HC 08-215 Rev. 0

50.59 APPLICABILITY REVIEW FORM LS:-AA-104-1002 page I Pf 1 Aptivit/Documnent Number. 0965 Address the questionsbelowfor all aspzctsof the Actlvity. Ifthe answer is yes for ny:portion ofthe Activity, apply Vie identified process(s) to that portion ofbhe Acivit,. Notle UAtit is',not unisual to have more than one przocess-apply to a given Activity.

See Section 4 of the *Resburc Manual (WM) f'or aOdjtfona! guidphnce.

1. )Doesthe proposed Actvity nvolve a change- '.' .. . " . ..

1. Technical Specilications or Operatinig License NOY E] YES See Secton 4.2.1 ZICR09) of the:RM,

2. Conditions of Licen.,e.

Quality -Apuranco program (IOCFRSO.54(a))? ' NO. El YS Security Plan (10CFR5O.54(p))? ONO 0]YES See Section 4:.2.2 of theRM Emergency P1aA (JiOCRk0.54.(q))7

3. Codes and-Sn490s ISTPrograPlan (IOC-FRSO.55af))? *..NO "S ee-Section 4.2.1.3 of theR*M 151 Rrogr=Pazl VIa (lCM.O,5.5e(g)? j ES

4. ECS- Acceptance Criteria (l0CF 0A6)? " NO El YES See Section 42.loftheRM

z. pecnc~xeipons(lCFRO.2)2fENO ElYBS 'SeeSecticou 4.2.1.5 of the RM.

6. adiation Proteetion Program (OF2) O YSSeScin4216o h ~

io Pogam(aplicable roeain ies NOA Y.ES See SiactioAi 4.21.V 6fthe ~

8. prorax0. contrOlle.*by the Operting Liceense or the Technical' = NO S See section 4.2.17 of the P,,

Specifications (such as the ODCM). . . O S.e. n.

9. Environmental~rtection Program " NO F YES- See S6e i6h42..7 of the RM 0 lOther. pro o leby etegulaons O YES See ection 4.2 of the R.

,1L Doesthi proposed Acity involve maintenance Whic4 restores S$Cs to their original.oondlitiaoffri inovp a*emPo~rary a~teratioznsuporting'- '

n~an~aana tatWill bel dn ffe~ct dttring.ai-poweropeain for 90 days or.LSNOE YESý See.Section 42.2 of the RM

___less? _______________________

111. Does the proposed Activity inv.olve a chanige to the: ______

1. U.FS AR (ipcdingq documen.ts ijdofagated by rfernce~thatis epciadedi-om therea . iretnenttperform f a 5.50. eiewbyNEI96-07

• NO l.YES Se Section 4.2-3 of the PM:

orN*-Y99_03? "_......... ... ........ . .............. ....

2. Managerialor administrative procedures goveming the*,onducdof Y See Section 41 of t R..

faeiity, operations (subject to'te cartr6l of IOCFR*S1, ApPaendix 13) W NO El YS See Section 4.2.4 oftheRM

{3. Procedures forpeforming mamntesnaneaotvities (subject to IDCFRSO,, I jjO, V S 'CSi,,"4 ofte Appedix B)?,q.4ý4o k R4egulatory commitnit notmovered byan6ober Treulition based. chnge process 99.!4)2 e YRS Ye -t See S.ec-.on 4 2 3/4 A. i." f e. .

.:TVDpethe proppsed Act ivty inplva cbpTtS Spn ih Fnepnun IF4 fuict PSoaeIntlain conrlb 0.CFR72.4S) y Fuel., Oý.O [],YES. See Section 4,24 pfthe.j Check one ofthe folowivng:

[fall f spectkof the Activity are conrtioled'byone or:moreeoflhhe above processes, then a50.59 Screeing:is requiredand the* AciMity.:may'be implemrented in accordauce with. its governing procedure,

• If .oofoionofhqA ity irnt.r8 rner offe 6be oesseten proess a559 r ng loedby for the poAifo*n ot boVered by any of the abov~processes. Theremaintng portion ofthe .ctivitysho.ld be implemented in accordance with-its governing procedure.

150.59 Scren 50.59 Eva[Uator: Gonal. T. Metel .Sg P*ate; _110/=2009.

.(C~irme OJ.1) {(Print name) /v4Si'giat'ure)

50.59.SCRE NING FORM. LS-AA-104.-10o3 Pageý of`4 50.59 Screening No. HC 2008-215 Rev. No. 0 Activity(Doeument Number: 80096650'

• ~.

.$*50.59 Seening Questicene (Chec'k 'czirect response and provide separatc writter respo*te providing t¢eha.sis.for the .a fet

-to each qu stion)(See Section 5 of the ~souceMa.ual (M) for dWditional guidance).

Do.e the proposedActivity involve a.hange to 'an3 SSC .thatadversely affects., 'I.*JAR 'BYES: NO described design: function? (See Sectipon.2,,.I of ffie:RM)

Theproposed activity of an increase in the allowable Engineered Safeý reawe(SF)eak rte!

and In update of eai Steam IsolaonValve (MSI:V)Tealagerlease.re odel do not invplve a change to amSSC. The proposed activity of ati.additiorial release patthiasoeiateh with an ncr~ease in the primary containment isolation valve (PCIV) maximum isolation time'to 120 iszconds'during a LOCAý,:nd.the design functional impacts.on the systems, structures and components.($SCs),

downstream ofthe:open PCIVs, are-evaluated in Reference iL2 for the SSC exposures to the post-LOCA containment peak pressure and temperature. The design pressures and temperature I"f all systems downstream of the open PCIVS are less than the post-LOCA containment pressUre and temperature, except for the primiary containment .instrument gas system (CGS) (Ref 112. Table 5), whichl ias: adesignitemperaturetmat is less:. thant the ps-.LOCA cortainment'peaktempiemtare.

The.structural integrity ofPCIOGS: is further evaluated for the. system. exposire to a higher post-LOCA temperature imposing additional thermal expansion stress.. TIl evaluation in. Reference 11.2, indicated that the total stres .ofvarious.piping segments including the additionalstress; resulting. from the post-LOCA temperature exposure is less than the allowable stress. Therefore; it is concluded that the structural integrity of the P'IGS will be maintained to perform intended..

normal and, abnormal -system,Jucitions.

The SSC design, finetions described in the UFSAR'are not adversely impacted by-tei proposed activity ofthb increased ESF leakage, updatedMSThV leakage model, and opened .PCiVsupto1,20 seconds. duringa LOCA used in. establishing the curent design basis or ased in the existing safety analysis. Although the open, PCIVs establish the additional post-LOCA release'path which contributes to offsite and control room doses thatremain within' th6regulatory allowable limits (Ref L.1, Sections 81..I&".2) any.increase~in the:dose exposure: is considered adverse'and requires fuzther'evaluation in the attached 50.59 Evaluation Formr.The.vital area tission dose.

rates are red*ued and consequ ently theoccupancy times-have -lengthened (Refs. 11.8 & 11.9). The reduction in the post-LOCA vital; accessmareacdose rates and increases in.ocupailcy times are not considered adverse because, they are beneficial forfthe performance of the post-accident vital functions. Therefore these cbanges are screened out and are notsubject to a 10 CFR.50.59 evaluation, even thoughlthe changes call for the LOCA safety analysis to be updated. These changes are pmurly academic nature to demonstrate compliance with the NUREG-0737, Section, l.B.2 shielding adequacy for performance dofthe potential.vital fnctions. This infornation:is historical in nature, never used forany post-accident action plan and does:not adversely impact the design ffifirtions of SSC.

In sunmary,.the prop?sed'activity, alters the radioiogical design basis inthe safety analysis, wh needs to be evaluated:in the aitached 50.,59 Evaluation. Fomn,

2. Does 'the proposed Activity involve a change to a procedute that adversely affects how UFSAR X .

4e*cribed SSC design fimt-io are plerformed or controlled? (See Section 5.2..2 of theR The proposed ficre.*e inithe aiowaIble E8 leak rate fom. £0 to 2.85 gpm prondes the operational. flexibilityin t¢ analyzed condition,4The ESF leaka*e is 'postulatedtto determine the.

vaWi4 siting criteria for the liope .Creek site in compliance tAihthe regulatofy reqtiirement (Ref 111), TbedncnaedaE.SF 3 eak rate heitheir introduces a conltrol meahanism nor alters Ve designt functiong related to the existing system, configuration, and therefore does not adversely impact the.

nwner in Whi*h t*e SSC design fntons are perfo-rmed or controlled.

The proposed update of the MSV leakge release model qdoes notrequirea procedure hehge.

50.59 SCREENING FORM LS-AA-304-1003 Revision 1.

Page .2 of 4 5059 Screening.No. HC 2008-215 Rev. No.......0........

Activity/.Document:Number: .80096650 The new limit of 120 seconds: closure time forfthe PCIVs results in.the*revisions of M &C testig:

procedures, 11C. TRM Table 3.,63-1 (Ref Iff. , Section 1310), and UFSAR Table,6,2-1 6 to ielude the new clostreitime, As discussed in The response to Question I.,te increased 0osurtit (me ha' no impact on SSC design functions'. Te increased closure time iieither introduces a new cvntrol mechanism nor alters the design functions ofthe existing.systoni configtn'atinn, and tbereofotedes not adversely htipact the manner in which the SSC design functions are performed gr controIled.

In sumnnry, the revision :of affected procedured do not adversely impact the. UFSAR described SSC design function performance and control.

3. Does the proposed, Activity involve an adverse change to an, element of a LUFSAR described El O]s/NO evaluation methodologyj, or use ofan alternative evaluation melodoalogy, thati*s usedin

.establishing the.design bases or used. in. the safety analyses? (See Section 52.2.3 of the RM)

The proposed activity is analyzed in Reference 11.1 using the AST methodology and. TEDE.dose criteria in accordance. with Reg. Guide 1.,183 (Ref 11.3) and ARCON96 .atmospheric dispersion.

methodology in RG .194 (Ret 11.5). The NRC Staff.approved these.source term and atmospherto dispersion methodologles, and the TEDE dose criteria as HCGS licensing bases by issuarice of operating license amendment: 134.(Ref IL4), The use ofAST metbodology, TEDE dose criteria, and ARCON96 atmospheric dispersion methodology to eva *.emte radiological. impact ofthe proposed activity is not an, adverse change to an element. of aXUFSAR described evaluation.

methodology, or use of an. alternative evaluatiohnmethodology, that is used in establishing the design bases or used in the safty.analyses.

The proposed activity of an increase in the allowable ESF leak rate is consistent.with the guidance of RG 1.183, Sections A5. tthrough A5.6..

Theproposed update offhe MSIV leakage releasemodel is also :consistentmwth the guidance of:

RG 1.183, Sections AU. through. A6.5 and is.consistent with.the rmost recent.NRCguidance as.

promulgatedthroug-IiNRC reviews and acceptance of the MSIV leakage models for'the Peach Bottom Atomic Power Station.

The proposed activity of an* additionailrelease.path associated with an increase. in the PCIV maxmum, isolation time to 1.20.seconds is consistent w'ithi the guidance ofRG 1.183, Section 1.3.2.

In summaiy,, the proposed activity is not an adverse change to an. element ofa UFSAR described evaluation methodology, or use of an altemative evaluation methodology,. that is used; in establishing the design bpses or used fn thesr.ety analyses

4. Does the propoSed Activity involve:a test or experimentnot described ia the UFSAR, where an. [IYES O NO S.Sc is utIlize~d Or controlled in a tnanner that is Qutaid the ref£erence bounds .offthe design* for that SSCor is inconsistent wIth analIyses or descriptions i the UFSAR? (See Section 5.2.2.4 of the The proposed activity of an increase in.the allowable FSF Jeak ate, an update of the MSIV lteakage release model, and the additional release path associated wvth an increase in the PCIV Inaximum iso46atitn4tiime. ejthr lyeotve a testpQn an eXperiment that is not desribed in,Vie UFSAR.

The testing ofthe PCIVs will be perfornet 6in the sanme mndner as behfore with the sazge applicable reg-ulatory compliances witlh a newly established closure time with0uth*virg any adverse effect on the plant safety and public health & safety.

In sunmmary, the proposed activity does ott itvolve any test or expelimuet not des4ribed in.the

50.59 SCREEN G FORM L.S-.AA-104-1003 Page 9df4 50.59Screening No. HC 2008-215,. Rev."No..0 Activity/Document.Number: 80096650; UFSAR. The-.esting of the PC~lVs willlbe performed in consistent with applicable tesong.

procedures$with a new closure, time limit.

5, Does the prpposýed A_4ivityreqqrea RM)

Licese?" (See $ection 5.2.2.5 :ofthe.

change in the Technical Specifications or Operating [' YES f:NO The ESF lak: rate is pot reflected in the Technical Speefic tons or' the Operating ticense.

The MSIV teak ie i* zpodeled onsi:tnt with t;e Tecl*rcal. Spccification 3 6.12,c, "P.rimary containmeni Leakage imng Condition For Opcrationi" The PCL1Vs listed in.the Hope Creek Generating Station Technical Requitremenfs Manual:(HC Table. 3.6.3-.1(RefN1A,6")have been relocated.fi'om the.HC Techncal Specification by the HC operating license amendment No. 171 (Ref. 10A.7) and their maximum isolation timeg ate maintained in the HC*TRM Since the PClV isolationtlimes are controlled: andmaintained by ihe, HC TRM outside.the HC Technical Specifications, the change toisolation~times does.not constitute a change intbe:TechnicalS.pecifications or Operating License.

Ir surnmaiy,. the change -does not constitute a change inthe Technical Spec'ifications or Operating, License.

11. List the documents (e.g&, UFSAR, Technical Specifications, other licensiog~basis, :technical, commitments, etc.) reviewed, including sections numbers. whee relevant informationiwas found (if ot.identified inthe. response to each question).

1. Hope Creek Calculation No. H-.IzZ-MDCr.880, RSev 3,.Post-LOCA.EAB., LPZ, and CR Doses.

2. Hope Creek Technlcal Evaluation DCR # 80096650-0210, Revision 0, Technical Evaluatiolto Determine tie post-LOCA Design Functional Impact on Systems & Components-Located Downstream of Outboard Containment Isolation Valves whicl are Expected to Remain Open for 120 seconds at the Hope Creek Generating Station:(HCGS)

3. NRC Regulatory Guide !183, "Alternafive Radiological Source Terms for Evaluating Design Basis Accidents. at Nuclea.r Power Reactors",.July.2000

4. NRIC.SafetyEvaluation.Repot Hop;e Creek GenemtingStation.- Tssdance of Amendment No..134 for Increase in Allowable MSTV Leakage Rate and Elimination of MSIV Sealing :System

5. U.S. NRC Regiilatory Guide 1,194, June 2003,0'"Aimospherýc Relattve Concentrations For.ControI Room Radiological Habitability Assessments At Nucleari1ower Plants."

6. Hope Cre&k Generating Station Tec*hnical Requirements Manual, (HC TRN, Revision J1*, Table 3.6-3A- Primary Containnment lsofation.Valves 7., Hope Creek Operating Liceoisq Amrý.,dno6.* I17, ,R:, Relocate Compo0nent Lists For Primary Containment'lsolation Vabes Front Ticltica Specifications (TACNo.: D3600),

8. Hope :Creek Calculation No. H-.I-Z;-R-MD 923, Rev 2, Vital Area Mission Doses

9. Hope Creek Calculatlion.No,4. li-ZZ ,0C-I1927j Rev 1. Areas Requiring Continuous Occupancy tO. CFR 50.67,Accident Source Term IJIL Select the appropriate conditions:

F ifUaH questions are anstered:NO, then complete the 50.59 Screening and impl*eent the Aqtivity per the applicabte governing procedure.

] )'question .:, 2,) , or 4 is answered YES and question . is auswered NO, then a 50.59 Evalnation shall be perform~ed,."

[]If questions 1, 2, 3, and 4 aret answered NO. and question 5 is answered YES, thena License Amendment is required

50.59 SCREENMG FORM LS-AA-104*1003 Revisjon I Page. 4of64 50.59 Screening No, HC 2008-215 Rev. No. 0 Aetivity!Document Number: 80096650 prior to implementation. of the Activity.

l1faosoon 5 is answ-Pro yB's or aoy pprlion of an Aqdyity leiaLcs,4en etisquedpor&

impIernei10tionA ofOat por1ion .oftd.ie Acyvity. Jz addition, if.qusfion 11,2,., o,,4 i.s answ e'YES for thde remainilig portibns o*the Actvit then a 50.59: EbaluationS I he:pezfork-ed for the remaining p~rtjQns 6f the

.Acfvity,.

IV, $cranidng Signoffs:

50.59 Screener: Gopa J. Patel (Si ature) Date,- lfl5/2009) print i~ame) 50.59'Reiiewer. __ .,_.__ ~ignt~

(.Pxntnarnae) 1gpatýrý)

50.59 EVALIUATION IFORM LS-AA-104-104.

Revision 2 Page I of 7 50.59 -EvaluationNo.,: HC 2O08215: Rev. No" 0 Acfivity/Dfocument Number: 80096650 I Completethe 50.59 Evaluation:

NOTES& Providea separate written response~providing the basis.for the answer to each.question below. The Resource Manual (RM) shlould be'used to detemineihe fcontent ofeach response:(see Section 6.2 for additi6nal guidance).

Ilf ie Screeniiug indicated that. oly a change in method.of evaluation exists,bondy Questionl8'is required t6:be answered. If the Sctreening indicated that no.change. in method of evaluation exists, Question 8 does need not be answered... *

-Des the proposed acivity result in more than a.miniimal increase in.the frequency of occurrence of an. accident previously evaluated intlhe .FSSAR?. (See.secon 6 . fthe6 .R")6 []2S 16N0:

The re-analySis of the.:radiological Coniseqinnces due toan ineroase in th. allowable.Engineered Safety Feature (ES ) leak rate, an ,.updtate ofihe Main. Steam: Isolation Vnv! ( I~o:SiV) 1a.release6,'mdl. 6d te addiftinal_reease path. a ocated ith an, increase imthe. primary containment isolatio valve.(,CTY) nmitmi n imolatia. time. does not introdace the possibilbofat hange int e -frequency of an accldentbdcau.ethese clianges arae7nt ifiators of any acci dent and no liew failure mode a:re0intro duced.

The degign basis LOCA is categoried as.a'Limitirwg :auW,.in heH.*GS TJEAiRS cetno1.6.5.1. Ref. flA) The LOCA is an ev,*e wtareJplab, wlch is n xpected to but isOi..tula4od becapse its .oeec ,ould inctie the potential for the release. ofsignificant amounts oftadioactJve material (Re14 M).. . Siq6et*e.affectedaMA is postulated.to evaluateits' dose cofihsequence, ihb Aieqehoy ofoccu-end!. of the DMA is de m eýd.ba4sd on the P!oCbab"listic kisk Assessment (RA) and notadendent on tbO$Fe ~ .A-~rate, the MIV i eOedaeree o~ilr

] the z diditoa rCejse pat4 nrdcd yteoe In.sumimary, e propdsd actity des no iihppct the:!f.quedn of o du.eof ain accideOn lprevio0tly,eValuat* in the

2. Does tbe propo.ýO actiV!ty resWl,.nm rmtrea6= a melnkral hinr.ase in Ole l1kelihd0dof tconturece o~fa zuallinedon ofA. SSC important to safety previously evaluated in the.UFSAR? (See Section 6.2.2. of the RM), D YES 0m teMre-analysis ofradiologieal consequences does not.introdilce thepossibility ofa change in thelikelihood ofa malfunction because !be re-analysis is not an initiator of any newmalftnctionsandl no-new failure modes-are introduced.

The re-analysis of tfe.radiologicaleonsequences -does,notintroduce the possibility :of a change in the likelihood of a malfunction because thedesign parameter values used in the analyses-are consistent with the perfbrmance ofthe credited S.SC: s. The sety redatedfb nction of the 2Firation Recirculation and. Ventilation stYnFRVS) m ýxaust aTd Co.trol Room:

1~nergency.Filfragon. (CR.3F) System is:to mitigate the post-apc de*et dses, The 1?RVs and CREF are techT*caly credited in hie revised analyse '"witl e sae filtratitn efficiencies and flow rates-as. those in the previous-revision of the analyses.,

Thereforelthe proposed change does not impose any addifioonal: chall.enges to tbeir intended inucon.and required performance, and dondt increase.floe, likelihood ot.occurrence of any mafwfiiions .

In summary, the proposed-activity does not impact.the likelihood of oc.dlrrene ofa malf4uckion ofbn ssc imnportent to safety previously evaluated in the UFSA".

3. Does the proposed aztivity result in more *t*n a mninial increase Wthe consequences o=an, aiecident p.eviously evaluate&d.i the UFS -.2 (See Seon . ...

4'on ...2, . nheR) b.. R.., . [

[4JYES 1

• The6 following tables indlcate at.tlhe c1hages inbhe revisedxradiological consequences of the; design basis.LOCA at the vfi0oi receptor loeati0n e lees trani*ini.miat and the catlcaated total doses are4less fh allowable regulatory limits (Ref.

1l.1, S.e!to5.',4). 'Therefo..t the prbposedactivity does not result inaore than.a minimal increase in the consequences of an attident pTeviod*ly evalnard.in thfe'-FSAR.

I---_ .___I.......... ....

50.59 EVALUATION FORM.

Revision 2 Page 2 of 7 S.59 E0valuation No.: HCC200815 .. Rev. No.: 0 Acfviýy/Document Number: 80096650 Curet Proposed Reultoy Irpoe Mimu Design Basis Aeidetit Toar: ToWta Dws. Dose Dose. Dose Dose Dose Limit Increase Increase Limit (iem) (rem) (rei.) (rem) (rem): (rem)

TEDE9 TEDE: TDE TEDE TEDE TEDE A .B C. D=B-mA E0.1c- . F Loss of Coolant Accident " I-ZZ-MDO- . .. .. Re. 3 (LOCA) 1880., Rev2 "J._

_ __o__"'"__"

Contr1f1 Rom.. 4.1.6 4.17. 0.01 0.084 5 Exclusion Area Bouinary 3.10 1.4 " 25 -1.67 .2.19 25.

Low Population Zone; . 0.696- 0.548 25 1r05 U : 243 25 B F~rom 1-1-1-ZZ-iMDC-18O, Rev 3.(~ 113I)

C From., 10 cQW59.670 (R *ef A FFrogiRG .183, Table 6(Ref>1L2) 00e inýlsi ofa diinlbps ees a~througioe P(lVs,, the icesei h alwblEFlal ae o28 gpm, ndtflie update of tb MSWI Iebage:reteas*nmodl combine to resaut in decreess.te offsite irdiol0gkal conseqtences, and an iricr6ame in the control room radiological conseuences. ThZe post-LOCA proposed o0ntrol iom dose Inc.rease (in Colum D)is .less than the -minimal dobices euaoylimit (0 lhin E,),.a~nd the pOst-LOCA total ptoposed cohntl roo dose.(i Cohl1uxl ) is.legs tan tie4lowable regaory lirit.(it Coin F.). Therefor, thl proposed atvity does not reswlt in more thau.:aninimal intbh conseaueices of the LOCA gs previously eoaliated ýn tde

!increas.e The vital area mission doses.in calcuIration*.-.-ZZ-MD>I7 927, i .e I (Ret f.41) 6rian* tei adequacy of the plant shieldingt*o providethe'required ptoteotion.t *a operator perforqming a post-acc dett vital fpnction ad occupancy based on the calculated. mamum JoqCation specific dose.rates, The post-LOCA dose rates adtresulting occupancies in various ,ital raf5 are reported in Sa WOV PtEýJao eotft~oeCekCntn resr oe irt Rf 10,TbeBI he vital area minsson dose rais I, the revtsed nalyTs~sare subst*nti*ly euced; Wh.ich l ailoved for increased 6..upancy times. for the:Njtal area'.Consdering that the aed"ced radiatio exposures are ess~pfiaf!y the same andthaitey are calculated in a conservative manner, the. resglting vital area mission doses and. oocpaies ar.eacceptable ioat havingany advrse Simrniliy the dpes or the areaS equing continuous occupancy (e.g., TSC, OSO, Ctad!ouse) Mn calcation f t.-ZZ-MDC- 1923, Rev 2 (Ref 11.2) becarme less than the previoIusly calclated doses. The post-LOCA 0oses f-or the Veas req~irom ooupac ~ in21 PPtiuri efornc120 ITe Pale 2e PA2 &-4r.g CP64, ýsjhl~iq that tqerduced dose are essentially the sane had ihat they are clc*i, ed in a .conseiwveinaner the revised dpses tie acceptabje without baving ty adiverse impact on the current.plant licensing bases.

Insummarythiere are orly minimal increases in:the tofal Consequences ofth atcident pd total dose consequences remain witli the regulatoD allowable limits.,

4. Does theproposed activity esmult ij more than atninimial increase in the aodtetences of a malfunction ofan SSC important to safety previously evaluated in the UFSAi? (See Sectiod 6.2.4 offt RM)

• YES [ NO The re-analysis ofradio] ogical :onsequenmees does not introdgce tlhepossbility of a change in the cosequerces of a malfunction bepase the re-ana!ysig is not at initiator ofdany new malfunctions and no new Tilure modes are introduced.

As discussed in Response to Question 2 above, the2-proposed change does not introdue anY kind Qf malftnedIon of the safety related systemncr&di*ed in 0he revissed dose gonsequenoe nith cdi a* additonaltsafety fonction nor invOlve oyphysicalchangge totime SSC &0ctio. Therefore, the proposed change does not'create ate possibiitY for a

5059 )EVALLUATION FORM LS-AA-104-1-004 Revision 2 Page 3 of 7 50.59 Evaluation No., HC 2009;:1I1 '8! ev.No.: 0

-Activity/Doeu meat Number. ' 80096650 malfianotion of an..SSC* important to safely previously evaruate~d in V1gAR and. flier~by it 4'oes not result in arty telated increase in the.onsequannye.la tn summary,. iere is tn increase in the consequences cf a mall-irnction .

5, Does The proposed activity create a possibility foran,accident af. diffie *itype.than any previousl eP'atuated inthe UFSAR? (See:Section 6.23 ofthe RM) O]YES,2NO The re-analysis of radio logical consequences, does not Intiroduce the possibility of.a new accident because the'reAnalysis is notan initiator-ofany accident and no new failure modes are introduced;.

As discussed. in Response to Question I above; the analyzed design basis LOCA Is.a *hypothetical condition postulated&

because its consequences would include thepotential forthe release of signiflcant amounts of radioactiveimaterial. The proposed change of an increase: in the alfowable ESF leak rate, an update ofthe MSIV leakage release mo del and the additionalrelease path associated, with an increase in.t1he PlQV maximum, isolatioa time are not related to: any mechanism or process that creates anacdi dent. They simply represent ite activity releasepaths conitributing theý dose consequences after the accidentbhas already occurred. Therefore, the propose!d activity does notcreate a possibility. for6 acidentfadiffrent type.than: any previously evaluated in the.UTSAR.

In' summary, there is no, increase, in the possibility of an accident of-a different type than is afreadyanalyzed I'SAR.'

in the dy aazeinte. ". PL

6. Does the proposed activitycreate a possibility fore mainalfuntio of an SSCJimportnt to safety with a different result than any previoosly. evaluated. in UFSAR? (See Secion 621.6 of the ~M ElYES O"NO Tbe proposed i.crease in the allowable"ESF lee.-rate, Aniupdato of the MSIV leakage rpleage mo6del, and. the additional release path asocjiated With. inceae in tee PCIV maxmum fiolation time neither imPact nor modfy the SS f in "

Consequently, the proposed a-tiiy either invotveý any physidlchangeto any SSC noroifiT theirexistirg impotant to safety functions.:Theei-bre, the proposed activy does not'retelt hepos~ibity for;a maluntion-ofe SSC important to safety wit a diffefent reWitan any pre iously evaltated'ii UFSAK.

In summpary the proposed activity does not cre.te : possibility for amalfin.tioh ofan SSC 'importt to safety with

• dtifferent resut than any:prevxy evab.ated in UFS.AR.

7. Dle6 the proposed actvitr* ettlt In a desig basis Imit, for a alssion ptoductbarrier as described 1 SA bein exceeded or altered? (See. Section, 62.7 ofthe.R1.)'. YES ZNO There are threa(3) fission product barriers.namely tlie.fuel cladding, reactor coolknt system pressureboundary, and drywell (primary containment): pressure boundary. Two but of three fission productbarriers - fuel,claddlni and reactor.ooolant system pressure boundary -arepostnIated to rupture during a typical large breakLOCA. The ractions offel failure (core

.lnve*tory)'used in,tbhereanalysis are the satne a$.thosemo deled in the previomUSalalysls. The:PC-!ls listed in the .HopeCreek Generating StationTechnical Requh'ements Manual. (H-C TRM) TaVlCe 3.6.3-1. kef ,1f.9)'haye been relocated ffom the 1iC Technical Specification.Tabie 36.34-1 by the :He: operating license amendment.No. 171 (Ret 1348).and their maximum' isolation times are maintained in the HC TRM, The containme Viegity is relaxed for 120 seconds W~en the drywell and suppression chamber pi*g exhaust.PCIVs are postulated'to remain open resulting in an additional bypass release path during a LOCA, which was analyzed. The instantaneous-dlosure OfPCIVs becane necessary to maintain, the containment integrity asa fission product barrier inthe previrs.tlC licensing basis basei on the TID (Trch'cn1 3 nformation Doconent)-l48*

source term, which postulates the instantaneou8s release: of core inventory inltier'coxtainmeat. To Mncorporate the NRC.

defense-in-depth philosophyto mitigate the dose consequences, t.e 'containment integrity was maintained by instantaneously closing those PCIVa, wlichb establish a direct fe*le pat!littothe pn*yironment.

The NRC Staff approved the Alternative .Sourc* Term-(AST), T MPE dose crteria and A ON96'atmOsphrcdispersion.

metiod6ology (Reot IL 0) as the HCOGSlicensing bases by issvance of opeuiing iese amements 134: and: 46 (Refs; fl.5 and!1L8)AIn the AST'(ýef. 02, Table. ), o41y 5% of the core iodin&and noble gas activity: are expected to release into iae

5..59 EVALUATION FORM LS-AA-,04-1o004 P4.e 4of 1

.. .y .. .

SO.59 .Evalukdlon No.: 'HC20138-21.5 . "*.:. ... iRev.. .os 0i Activity/Document Nii tber:. 80096650 ontainment 120 seconds (2 miniates) after oxvset ofýa LQCA. Thetefbore, lihe priiýky containmcit tegrity as a ission product barrier is not*require %orthe first 2 minutes after onpet of, LOCA f!l fnProdu~ts J.ause t Pcre ar telea§ealin tle containlient during this time period which Would need cqotainment confinementa The oyily acivity available¢fotrelease through the open containment barrier (PCIVs) is the, teact6r coolant speci. c aetivi whic ijs n~gligibjy sman]. in comparison to tOe core gap activityi. The dose co*sequences from the open containment barrier via open PCI Sfor 120.seconds before the onset of gapvactivity re ease :are nayzed and added to o0ther pos-LOCA dose contrb'utions. The ttal icreaset in dose

. consequencs are less than applicable minimal dose margins as shown ii Response to Question 3.. 'Thetfore, the.relaxation of containment Btri product barrier while therl is no tision produci reelead inithe coftncnt isthe as a fisi tesinically and. egally acceptable or RG l1g3, Section 1.3 . The re lysis-of other post4LOCA4 relse paths credited

• containment.as a.fission product battier during and following the fission. prodoct release inMthe c~ntahment as described in theHoe ree IP$o ndetoid býith Tcicand Specifiain 4(eat~ .. D11.6)chreortinete aisdon pi0duit arrier :limits deseribd 'he.TU'SAR Min are neither fexcedednor zltered:vesly.

insurmaary; theproposed actroty neither excee4 lor-alte the d gesigi baislimit fr fsion-prdttbarrer a.fdescrbed in the VJSAPR.

bPO.the proposed4ti Ityr "dpart f t thod o eValuatiol cdjei i &UFSAR d ieabsng'the design bases or in the safety analyses? (See Section 6.2.8 of the 1 Y'* 'E NO YES The.revised analysis uses-fte AST methodology and TEDE dose criteria in accordan.e ite. Rag. Guide 1..18 (Ref. "" 11.2) n6d' ARCON96 atnosphenic dispersion methodology:i RG 1.94 . (Ref. ff. 10). The NRC Staffapproved these methodlog ies as 1-ICGS licensin..g-bases by issuance of operating license.amendments l.34. and 146 (Refs. It5 and IL1)., The aerosol deposition on the main steam lines surface areas used.in the existingand.reavise anlyses consistent.with the NRC approved guidance in.AEB98-03 (Ref* 11.13). The aerosil deposition inlhe 1prvious'Revision 2'of the LOCA analysis was deve oped in a very consety4tive pnor for the industy's first deposition model "

The prevrious depositio modelbadfthefoll*wing ponsrva isms in the analysis:

1- The M$S leacage`Wajs assunmed to be constant in both the MSIVfailtdsteearn line (150 scfli) and. infa& steamlines *(5.0 2,, ;Fo .r 40):percentile aerosol settling velocity in piping b6tween the RPV nozzle and Turbine Stop Valve (TS").

3. Eklbmental iodine temoval by telwetted surface area.was not credited.

Te.previous deposifio' model was inconsistent with.flte latest regulatory developmentfor aerosol depositionin.the following imanner";

1, The MSI faild lin. boundry was 4nof clearlydefined. The NRC lates't understanding and. dafidion of the MýSi

• faWiled-line is thatit is the main steam line between.the RPVand inboard MSIV and tha aeroeso and,elemental iodirge depostiou ard mixing i. this failed. line ghould not be,credited. The MSIV failed line modeled in tlee previous analysis is

• 'incbnsisi.tent~vith: the: 1inste2fRC deliniton of~tle MSIV :fail~ed lin~e in that.the previous analysis credited *lie aerosl .and.

elementaleremoval and mixing in this segnenrtofMSIV failed.line withoutconsidering the volume ofthe.pipe segment be6,tween the inboard and outboard MSIVs. In the revised LOCA analysis,the MSIV failed line boundaryis ldearly Sdefined without aaosoland elemental iodineremoval and withoat mixing In th.MSIV failed line between the RPV nozzle and inboard MSIVS.

2. In thepravious analysis, forty (40). percentile aerosol settling velocity was constantly usedin!both piping segments-RPVito outboard M$JV and outboard MSIV to TSV;: which did notaccountfor the ziewly dek~eoped iNRC .encept of lesser deposition oftthe fighter aerosol particle 'by gravitational deposition in piping beyond the outboard MSIV. The. use of40 percentile .aerdsol settlingvelocity iSconservative in compiarson To the.NRCrecommended. aerosol settling -

velocity of 50 perceatile for thelheavder aerosol particlesin Ile steam.line betweentSh RPV nozzle and outboard:MSIV but its use for theentire release path. from the :RPVnozzie to TSV"is non-eonservative-becauseit does not account for the lesser deposition.of the lighter aerosol particles by gravitational deposition. in the main steam line beyond the o4tboard MSIV.

The combined effect ofbothconservatisms.ad inconistencies: in the: deposition modelswas suclh that the aerosol deposition allowed the licensees:.

model. was-very conservative in the previous analysis bebauseprw.ously the *RC unconditionally

50.59 EVALUATION FORM LS-AA-304-1004 Revisibo 2 Page. 50 7 50.59 Evaluation No.; ReC 2009-215s ... No, 0 Activity/Document Number.-80096650 1.Te r of IliaVISIV leekage based onthe 0 .t-LCAdrywsejll p PuAio sresi etem.erOat jur ind,

2. The 6lementtal iodine deposition o the iWetteadrywell, surface af.as9, Subsequent t6othe HCGS AST approval the NRC Sta rdefined ' the c ept acceptable MSIV leaae nodel of'aobased oi the research end experience gained withthe ASM' The revised LOCA analysis uses exatlty the aýme MSTV lekge model that has been accepted by the NRC Staff in the foblowing AST licne artendmentg:

1. The redtdn n the MSVe pst7L CA&drywel pres9 re.enmpehiture in ahnost -dlB AST license amendments including, but not lirited to, Dresden 2 nu3 :(Ref. 11.14), Quad Cites *nd 2 (Rf [1.14), Peach

']

Bottom 2 and. 3 (Ref i1.15), Liiner, 1pk and 2 (Ref, 11,.16), and Clinton (Re-f 117)D in the AST 1icente amendment.s credited 2-, 'Theadsor~ption fthe lemental idine by the drywell wetted:sutfnce iýa: was for Dresden 2 .and' '(Ref. iL14)' Quadl Cities. I and 2 (Ref 1J,14X and Peacit Bottom 2.and * (Ref. .I.I$).. .

3, 'Ten*n-wervatiwsreapssiated with the liglter *erosol prticle rbo;vaby the* *ta1eposonwaaairy new condern addressed by the NRC. Thref&or, itis adressed only in the ltest AS? lieean6.endrmens :for Peal Bottom 2 and 3 (Refp tits. I5), which already received AST !iend *tnendnent.

To make the revised analysis current and address the !atst reguJ tory concern aJout the non-conser,.tismasociated , th the

• depostion oflighter aerospl paortcles, Trevisedanalisis s aJ~/zed" oxactly in the samo Manner as in the NRC approved '

AST? icense a"nenines*.0.asfollows*

i. The:MS1Vleak~age is postulate: to release to the enVirf0nent:through two main steam l inesistead of three lines in.the prevtous~anaoysis.

2 Eac1h main steam release path: is-divide.into. tw0 well- )6ed Voulinis to be consistent with A0B: 99ý0,

2. A *-Viry well defined MSIV *Ied line boui~d..y Without creditingany deposition.ata niixing in t steam line betWeetn thb. PV"and Inboard M4SIY4-

4. The (50) percentileaerosol depositior ve~city in The MSIV TV line Jd twten itboard .4 outboar MSIsand intactfst.am line between. te R)V nozzle and: oUtbo'd MSV is credited tojus that t heavier aerosolIaI pfrticlee subje*cted lto a larger removal rate by gvitation and no aerosol, deposition tredited !n the ýS1V failed line-beteexi the RPV nozzle and. inboard MSIV.

45,The thirty (30) perceile velocity in the MSW failed: ad intact seam Vines beyond thbe ulbotrd MSIVs tO

%Veposiicm justify the teductijo ip Th deposWion. ofthe lighter aerosol particle by gravitationpi depo sition, 6, The elemental iodine removal by te dye wettd surfhe area is credited th -in anadysis.

7*. The leser elemental iodine.rtemnval e iey of 50% iý -d fo 0-24 hr in,te main steam lnes. 0 No ren.val of 1eiemenbla iodine is assumned after 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />si,.*  !

8. No aerosol gravitation-deposition of the lighter aerosol.particles in the maini steam lines is credited after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Per PE&G Procedure LS-AA-104-1000 (Ref IL19), Section 3.4, rather than maling a minor change to an existing method of evalUation, a liqens~e may adopt a completely new methodology without prior NRC approval provided the new method is

approved by the NRCfor theintonded application. As discussed.in Section:6.2.8, a new method is "approved by he NRC-for

'the intended applicafion?' if it is; approved forThe type of analysis being conducted, the licensee satisfies applicable terms and conditions:for it use and the mrethodlis approvedin an:NRC SERor ot1qensis accepted:by the NRCas part of a planxs licensingbasis: e.g., the iaethod'is described in the plant's UFSAR or the NRC has accepted licensee comm:inents made in docketed-licensing correspondence such as responses to NRC GenericLetters or Bulletins. Sections 3.4and 6.2.8 note tat the "conserviv.e" and "e*sentially the same" cfiteria do not apply when evaluating thfeuse ofa new methodolgy approved by the NRCfor the intended application. For example, thleuse of anew NRC-approved methodology may providenon-conservative results; however, that. s acceptale as loig as the methodology peviously eas been approvedty the MC for the. intended application.

The P'oeedure LSAAlQ4-:00O (Ref. [.19), Sectiins 3,4a41d 6,2,8 deseribe two peanS In which* one may depart from a method of evaluation described n the UFISAt The second Means i's 4a4ing fum,a mnethod described in te Pb Rto another method unless that method bas beer approved. by NRC brmtheinteonde*4.pliciOn.. Taesecond eans is cosidered here approppintely a4t fully discussed: in tl)eabove' sectio.. .

50.59 EVALUATION FORM .  :

LS-AA-104-o004

' evisio'n 2 Page 6 of7 50.59 EValonition, No.: .14C 2008-4I5 Re.N6.6 Q ActivityDocument NuMber- 80096650 The revised analysils uses , he same fegulat~y' gunce descPrieýd in the. UfS A"R. ue In tecrataayi.T rvsd anaysis capt*ues the excesive conseryatisnii theaeroscol deposition mode1.and~updates te mode to the latest l requirements. The resulting dose marginis used.to increase the FSt leakageand cntainment iso]ation time There.ore, the ir&analysis doesnotr6uitin a:d0parture 0fom a method ofevaluation that ba.s ben appr6ved by.NRC for the inended application without imposing any site-speoific terms and.conditionD on other licensees. The' change is not a departure from a method of evaluation because it is appropriate for the intended application, it complies with regulatory requireneits itis.

adopted in a:conservative manner* and itlias been approved by the NRQ *iQnother AST license am.thents; In summary, the proposed activity does not result i*na departure from amethod' ofeValuation described in the IUSAR used in establishing the design~bases or in the safey tanalyses, IL Identify references used to perifrm the.evaluation (ifotaprovided in 1he tesponse-to each question).

1. H CiZZ-MDC-1.880, Rev *, 'Post-LOCA EAB, LPZ, and CR Doses' 2., NRC Regulatory Guide. .183, "AlItemsaive. i10giel Source Tes fr Evaluating D.es- Ba*sisAccident at,.

Nuclear Power Reactors", XUty 000:

1 *' 0 CER.50.67, I'Accident Source Term."':

4, iCC -QLIS.AR Section 15.0...2,6AFrequtency Classifkatian

5. C.... f y . ....o G e "13i to...Ameod ent N".

... -4' n, .rase AlIoWableMS7V Leakage Rate and Elimination of MSIV Sealing ystem

6. tqOGS. Technical Specficatio. 6.*,44f, PrimyContaiuient Leakag ARt Tg*est g. Prgran.

7. JHope C-reek Tejzhnical Specifi-cation LimitinCnitofrOerio(L ) 3/4.4,5 SecfcAciiy

8. Ipe lCreek:Operating-License. Auendnent.o. ~Ii,~RE. R ocate C obponct Lits-Fi. Prhary Containment-Isolaotion.VAIvs PhomrTbchnjiqSpeeificatitns (TAOlNo. MD3600) 9, Hope Creek*Generating Sta4ou Techical Requiients Manual(0 T.M). evisibn I,.Table6.3-1, Pm6,ina Conttainment Isolation Val~es

10. tS. NRC kegulator (uijde 1.0.94, Juie2Q03, Amo6ph Relative Concentrtions f.r Contro Roomk .adi oogical Habitability Assessments At l.Uclear PowerPlants."

1i., Hope:Creek Calculation No. I{-rRZZ-MDC-l927* Rev 1,,VitalAroa MissionDoses *

12. Hope cre.k CalcICuationNo. H-)-ZZ:-MDQ-1923, Rev 2, Afres Rquirig i ontintous Ogupancy0 13.: NRC RepbrtAEi-9$-,03, -'-A- essnn ofR. iloia C1`qecsPrte er ioPatAplcto sn Revised NMU G-1465) Source Term

14. Dresden Nuclear-Power Statdo; Units2 ad 3,-atdQud CttesNucwer Stater Stton, Units ,and,2-ssabee:,of A*.end*erentsKRE:.Adoptioa OfAhternalive Souce T160 MetJiocio (TAO NOs ,*B6V53.1,tBT653Q

'6532, MB6538, MC8275, MC8276; MC82V-7, MC8278), Septeniber -11, 200* *(DAMS. Acicession N1%. ML062070292)

IS. Peach Bo.ttoM Atomic P-wer Statidn,V tbits 2 and 3 ISsuapn.of-Anendni*eti.s ,: Applcation.of-A.temative Source Term ,Mthodology (TAC NOS; NM680Q6:and M: 607), September5, 2006 (ADAMS Accession No. ML082320257)

16. LimeridkGenerating Station, Units I and2 -IsgtgnceofýAmendmnents Re-:.ApplicatonfAlterative Souce Term Mehdlgy (TAO Wo,qt MC2295 adMC229), Auput3 206 (ADAS A c~sion 0o L O022 1,)14

17. Clintpn Power Station, Unit I.- issuance o04an Aiendmen..t'R Appiicakion of-Alternative .Suree Tetur MeWhodbogy (TACNo. UMB'365), September 1%.2005 (ADAMS AccpssionN . ML05270,61) 18.. NRC lettey to P5E Nu.a dated Ap~i~. 15.2003. "Hope Creek eneatinvg StPto *-Issu~cee o n eAm 146 Re:

jdjent Contwinot. ýequirem.ents During Fuel I4andfinmand.Rem0ov f Charcoal Filters. (TAO N,.MB5548J."

'19* "PSE&G Procedure LS-AA-104000, Rev3., S059 Resorce Manual.

.20. GE NED-33076P Rev% :.Safety Analysis Report for Aope Creek ConstaPressre Power Ipate Mll. based upo.n the re$uits:of this EVahataio-: (Selec eone offite f.llbwing)

• .itplqmerit thp Activity per plant procedures without obtaining a License Amendment.

fl]

. Request and receive a License Amendre*t prior to inpl rentatior,

50.59 EVALUATION. FORMI LS-AA-10442004

.RevisioO~

50.59 Evaluation No:: wJC2008-15 N04 Activity/Docmn Nurn r: 80096650 IV. Sigo-offs:

50,5 ~vav~t~:

Gp~J.Patel - Date: COO15/2009 prirtI Namec) 50.59 Reviewer:

(Printed: Name) .

(Si tV. .

.E~J~'E-w~.FAA 4A, _b atd: i~T .1 O ii"1'aie

.PORC... tting-Numbr LR-N10-0163 Additional Proposed Changes to the HCGS' Technical Specifications (Facility Operating License NPF-57)

Technical Specification Pagqe 5.3.1 5-4

DESIGN FEATURES

.5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 764 fuel assemblies. ond shall be limited to th*o ao

. . bl!io.... hl.h Aao.. b*or

......... f uoe i BWRB Each assembly shall consist of a matrix of Zircalloy or ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material and water rods. Limited su~bstitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with NRC staff approved codes and methods and have been shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions A maximum of twelve GEI4i Isotope Test Assemblies may be placed in non-limiting core regions, beginning with Reload 16 Cycle 17 core reload, with the purpose of obtaining surveillance data to verify that the GE14i cobalt Isotope Test Assemblies perform satisfactorily in service (prior to evaluating a future license amendment for use of these design features on a production basis). Each GE14i assembly contains a small number of Zircaloy-2 clad isotope rods containing Cobalt-59. Cobalt-59 targets will transition into Cobalt-60 isotope targets during cycle irradiation of the assemblies.

Details of the GE14i assemblies are contained in GE-Hitachi report NEDC-33529P, "Safety Analysis Report to Support Introduction of GEl4i Isotope Test Assemblies (ITAs) in Hope Creek Generating Station," Revision 0, dated December 2009, CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 cruciform shaped control rod assemblies.

The control material shall be boron carbide, powder (B C) and/or hafnium metal.

The absorber material has a nominal absorber length of 143 inches.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of:

1. 1250 psig on the suction side of the recirculation pump.

2. 1500 psig from the recirculation pump discharge to the jet pumps.

c. For a temperature of 575'F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately'21,970 cubic feet at a nominal steam dome saturation temperature of 547'F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.

HOPE CREEK 5-4 Amendment No. *-3