GNRO-2009/00036, Report of 10CFR50.59 Evaluations and Commitment Changes - April 1, 2008 Through June 30, 2009

Report of 10CFR50.59 Evaluations and Commitment Changes - April 1, 2008 Through June 30, 2009

ML092020181
Person / Time
Site:	Grand Gulf
Issue date:	07/16/2009
From:	Perino C Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GNRO-2009/00036
Download: ML092020181 (56)

Text

GNRO-2009/00036 July 16, 2009 2009 U.S. Nuclear Nuclear Reaulato Regulatory Commission Attn: Document Control Control Desk Desk Washington, DC 20555 UBJECT

SUBJECT:

eport of Report of 10CFR50.59 10CFR50.59 Evaluations and Commitment Evaluations and Commitment Changes Changes-ril 1, April 1, 2008 2008 through June 30, throu.gh June 30, 2009 2009 rend gulf Grand Nuclear Station, Gulf Nuclear Unit No, Station, Unit NO.11 Docket No. No. 50-416 50-416 License No License No. NPF-29-29

Dear Sir or Madam:

Pursuant to Pursuant 10CFR50.59(d)(2) Entergy to 10CFR50.59(d)(2) Entergy Operations, Inc. hereby Operations, Inc. hereby submits submits a a summary summary of 50.59 evaluations of 50.59 evaluations for for the period of the period of April April 1, 1, 2008 through June 2008 through 3 June 30, 2009. Also attached 2009. Also attached is the summary is the summary of commitment changes of commitment for the changes for the same period same period in accord in accordance with NE195-07 Guidelines.

If you If you have have anyny questions questions or or require additional information, require additional information, please please contact tennis contact Dennis Coultererat,60 437-65 at 601-437-6595.

This This letter does not letter does not contain contain any any commitments commitments..

rely Sincerely, LP/DMC :dmc CLP/DMC:dmc Attac eats 1. Table of Contents Attachments: tints

2. 10CFR50.59 50.59 Evaluations Evaluations and and Commitment Commitment Change Change Evaluations Evaluations cc: (See Next xt Page)

G090036 0090036

00 GNRO-2009/00036 Page 2 cc: ident Inspector NRC Senior Resident Inspector on Grand Gulf Nuclear Station Port Gibson, MS 39150 u.S. Nuclear Regulatory Commission ATTN: Mr. Elmo E. Collins, Jr. (w/2)

Regional Administrator, Region IV 612 East Lamar Blvd, Suite 400 Arlington, TX 76011-4005

u. S. Nuclear Regulatory Commission Commission ATT ATTN: DORL ((w/2)

Mr. Carl F. lyon, NRR/ADRO/ DORl ATTN: ADDRESSEE ONLY ATTN: U. S. Postal Delivery Address Only Mail Stop OWFN/8 B1 Washington, DC 20555-0001 G090036

Attachment 1 Table Table of Contents Grand Gulf Nuclear Station I

10CFR50.59 Evaluation and Commitment Change Evaluation Report Period for the Period April 1, 2008 through June 30, 2009 Acronyms I Alarm Response Instruction LOP Loss of Power TM American Society for Testing and Materials MAPLHGR Maximum Average Planar Linear Heat Generation Rate CCE Commitment Change Evaluation PR Minimum Critical Power Ratio CMWT Core Megawatts Thermal MNCR Material Non-Conformance Report CR Condition Report MOV Motor Operated Valve DCP Design Change Package MS Mechanical Standard EP Emergency Procedure MSIV-LCS Main Steam Isolation Valve Leakage Control System EPI Equipment Performance Instruction NPE Nuclear Plant Engineering E Electric Power Research stitute NSSS Nuclear Steam Supply System ER Engineering Request PDMS Plant Data Management System ES ectrical Standard PPM Parts per Million ESF Engineered Safety Feature PRA Probabilistic Risk Assessment GE General Electric PSW Plant Service Water GG Grand Gulf RCIC Reactor Core Isolation Cooling GGN Grand Gulf Nuclear RFO Refueling Outage G Gallons per Minute RHR Residual Heat Removal 101 Integrated Operating Instruction RPV Reactor Pressure Vessel lSI In Service Inspection SCN Standard Change Notice 1ST In Service Testing SERI System Energy Resources, Inc.

LBDC License Basis Document Change SGTS Standby Gas Treatment System LDC License Document Change SOER Significant Operating Experience Report R Linear Heat Generation Rate SSW Standby Service Water LLRT Local Leak Rate Test TRM ITS Technical Requirements Manual I Technical Specifications LOCA Loss of Coolant Accident UHS Ultimate Heat Sink Pa Pagee 1 of 3 G090036

Attachment 1 Table of Contents Grand Gulf Nuclear Station 10CFR50.59 Evaluation and Commitment Change Evaluation Report 09 for the Period April 1, 2008 through June 30, 2009 Safety Evaluations Evaluation Number Initiating Document Summary SE 2008-0001-ROO Engineering Change Containment Fire Water Isolation Valve to be Gagged Open (EC) 8927 SE 2008-0002-ROO EC 8474 Automatic Depressurization System Air Receiver and Accumulator Relief Re-seat Pressures lowered.

SE 2008-0003-ROO EC 9133 Leading Edge Flow Monitor Software Modified.

SE 2008-0004-ROO Cycle 17 C Cycle 17 Core Operating Limits Report Page 2 of 3 G090036

Attachment 1 Table Table of Contents Grand GulfI Nuclear lear Station Station R 0.5 Evaluation and Commitment 10CFR50.59 Change Evaluation mmitment Change Evaluation Report riod Apr for the Period April 1, 2008 through June 30, 2009 2009 Commitment Change Evaluations Commitment Source Document Summary Number CCE 2008-0003 Entergy Response to One-time extension to the 6-cycle Static Frequency Interval for the RHR Jockey Pump B Generic Letter 96-05 Discharge Block Valve. This also corresponds to a one-time14 month extension to the maximum interval between static tests.

CCE 2009-0001 GNRO 94-00059 Deleted commitment to alert plantlant personnel i procedures personnel in procedures concernin concerning the hazards of loaain emergency core cooling system working in containment and the potential for clogging suction strainers. Original commitment was based on NRCB 93-002 supplement 1.

GGNS subsequently installed large capacity passive strainers per NRCB 96-003.

Additionally, GGNS follows fleet FME procedure. Commitment no longer necessary.

CCE 2009-0008 Entergy Response to One-time extension to the 6-cycle Static Frequency Interval for the Drywell Chillers Generic Letter 96-05 Return Header Dryw n Valve.

olation Page 3 of 3 G090036 G090036

Attachment 2 luatio 10CFR50.59 Evaluations and I

Commitment Change Evaluations 0036 G090036

afety Evaluation GGNS 50.59 Safety Evaluation Number Number SE 2008-0001-ROO G090036

e 1

I. OVERVIEW / SIGNATURES Facility: GGN Evaluation # 2008-0001-ROO Proposed Change / Document: EC 8927 Description of Change: EC 8927 gags valve 1P11 F515 in the open position. This valve is a pneumatically operated butterfly valve in the condensate and refueling water storage and transfer system. The valve is one of four designed to close in the event there is a demand for fire water in containment. Closing these valves is

'intended to ensure that there is adequate fire water in containment to mitigate the fire. The loads downstream of 1~11 F515 served by the condensate and refueling water storage and transfer system have been reviewed.

The only constant demand is the flow to the seal steam generator. One condensate transfer pump pumps 600 gpm with another 600 gpm pump as backup. The 45 gpm maximum demand of the seal steam generator will have little effect on the quantity of water supplied to the fire protection system in containment. The maximum demand for fire water in containment is 635 gpm. The additional 45 gpm for the seal steam generator will result in a maximum demand of 680 gpm in the event of a fire. This is beyond the capacity of one condensate transfer pump and will require the second pump which will automatically start on low discharge pressure. In the event that one condensate transfer pump is inoperable and the need for additional fire water in containment arises, the Fire Protection system can be manually aligned to provide fire water to the containment. All fire protection in containment is manual. With a demand of 680 gpm for two hours a total of 81 ,600 gallon of water will be required. The condensate storage tank has a capacity of 300,000 gallons of which 130,000 gallon is available to fight fires. Sufficient water will remain available for other uses including ding filth fighting fires because the plant fire protection water system (backup fire water source) is the supply which is credited creel ite to meet the req requirements for a Fire Suppression Water System in containment.

Is the validity of this Evaluation dependent on any other change? DYes ~ No No If "Yes," list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action is completed.

Based on the results of this 50.59 Evaluation, does the proposed change DYes ~ No require prior NRC approval?

Preparer: K. M. Black I *: IEntergy/Engineeringl Name (print) I Signature I Company I Department I Date Reviewer: Andy Foxl *'

Name (print) I Signature I Company I Department I Date IEntergy/Engineeringl 8(13/08 OSRC: Charles Quick for Dub Barfield IEnter lEn ineerin Chairman's Name (print) I Signature I Date 2008-027 August 22. 2008 OSRC Meeting #

tore may be obtained via electronic 1 Signatures electronic processes processes (e .g ., PCRS, ER ER processes),

processes), manualmanual methods methods (e.g., irk signature),

(e .g ., ink signatu e-mail, or telecommunication. If using or telecommunication.

e.. (u:..+".... i C ";); j sing an an e-mail e-mail or t'\"",f"", J or telecommunication, telecommunication, attach

I A S attach itIit to to this this i form.

EN -101-ATT-9 Rev. 4

10 CFR 50.59 EVALUATION FORM Sheet 2 of 4 II. 50.59 EVALUATION Does the proposed Change being evaluated represent a change to a method of evaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only Question 8. If "No," answer DYes all questions below. o No Does the proposed Change:

1. Result in more than a minimal increase in the frequen rent of an accident frequency of occurrence 0 Yes previously evaluated in the UFSAR? 0 No BASIS: Valve 1P11 F515 is a pneumatically operated butterfly valve in the condensate and refueling water storage and transfer system. The valve is one of four designed to close in the event there is a demand for fire water in containment. Closing these valves is intended to ensure that there is adequate fire water in containment to mitigate the fire. The condensate and refueling water storage and transfer system is not associated with the initiation of an accident previously evaluated in the UFSAR. Gagging this valve open will not increase the frequency of occurrence of an accident previously evaluated in the UFSAR.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a 0 Yes structure, system, or component important to safety previously evaluated in the UFSAR? ~ No BASIS: Valve 1P11 F515 is a normally open valve that ensures a water source for the seal steam generator and ensures seal water for several pumps in the Radwaste System, G18, and several other intermittent demands. Gagging the valve open will ensure these demands are met and that there is an uninterrupted source of water to the seal steam generator. Gagging valve 1P11 F515 open will not result in an increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the UFSAR.

3. Result in more than a minimal increase in the consequences of an accident previously 0 Yes evaluated in the UFSAR? ~ No BASIS: Valve 1P11F515;s a pneumatically operated butterfly valve in the condensate and refueling water storage and transfer system. The valve is one of four designed to close in the event there is a demand for fire water in containment. Closing these valves is intended to ensure that there is adequate high quality fire water in containment to mitigate the fire. The loads downstream of 1P11 F515 served by the condensate and refueling water storage and transfer system have been reviewed. The only constant demand is the flow to the seal steam generator. One condensate transfer pump pumps 600 gpm with another 600 gpm pump as backup. The 45 gpm maximum demand of the seal steam generator will have little effect on the quantity of water supplied to the fire protection system in containment. The plant fire protection water system (backup fire water source) is the supply which is credited to meet the requirements for a Fire Suppression Water System in containment. Therefore, gagging valve 1P11 F515 open will not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.

EN-Li-101-ATT-9.1, Rev. 4

4. Result in more than a minimal increase in in the consequences of the consequences of aa malfunction malfunction of of a a structure, structure, Yes ystern, or system, or component component important important to to safety safety previously previously evaluated evaluated iin the UFSAR? No No BASIS: Valve 1P11 F515 is a pneumatically operated butterfly valve in the condensate and refueling water storage and transfer system. The valve is one of four designed to close in the event there is a demand for fire water in containment. Closing these valves is intended to ensure that there is adequate fire water in containment to mitigate the fire. The loads downstream of 1P11 F515 served by the condensate and refueling water storage and transfer system have been reviewed. The only constant demand is the flow to the seal steam generator. One condensate transfer pump pumps 600 gpm with another nother 600 gpm pump as backup. The 45 gpm maximum demand of the the seal steam generator will have little seal steam effe onn the little effect the quantity of water supplied to the fire protection system ystem in in contain containment. The plant fire protection otection water ater system system (backup fire water source) is the supply which hich isis credited credited to meet the requirements eats for for a Fireire Suppression Suppression Water System in containment. Therefore, gagging gagging valve IP1 F515 open will not result valve 1P11 ,ult in in mmore than than a a minimal increase in the' consequences of a structure, system, or component important to safety >afety ppreviously ated in evaluated in the the U UFSAR.

5.

5. Create a possibi possibility for an accident of a different different type any previously than anv type than previously evaluated evaluated in the Yes UFSAR? R? No BASIS: Val Valve 1P11 F515 is urnatically operated i a pneumatically operated butterfly butterfly valve valve inin the the condensate condensate and and refueling refueling waterwater rage and transfer system.

storage ystem. The The valve valve is is one of four one of four designed designed to close in to close in the event there the event there isis aa deman demand for fire fire water ontainmen Gagging water in containment. gagging this this valve valve open open willwill result result inin the elimination of the elimination of a a single single point poin vulnerability vulnerability that iat could could rresult in in aa plant transient.. Therefore, plant transient Therefore, gagging gagging valve valve I1P11p11 F515 F515 open open willwill not not create create a possibility accident of a different possibility for an accident different typetype than than any previously evaluated any previously evaluated in in the the UFSAR UFSAR..

6. reate a possibility Create bility for for aa malfunction malfunction of of aa structure, structure, system, system, or component important or component important to to safety safety 0 Yes with with a differ nt result a different result than than any any previously evaluated in previously evaluated in the UFSAR?

the UFSAR? [83. No BASIS: Valve 1P11F515 is a pneumatically urnatically operated operated butterfly butterfly valve valve inin the the condensate condensate and and refueling refueling water storage storag.e and and transfer systemsystem.. The The valve valve is is one four designed of four one of designed to to close close in in the the event event there there isis aa demand demand for fire water water in containment. inment. Closing Closing these these valves valves is is intended intended to to ensure ensure thatthat there there is is adequate adequate fire water in fire water tainme to mitigate containment itigate the the fire. Gagging this fire. Gagging this valve valve open open will result in will result in the the elimination elimination of a single of a single point vulnerability that vulnerability that could uld result result in a plant in a plant transient.

transient. From From a fire protection a fire protection standpoint, standpoint, the loss of the loss of the the P11p11 system system is is bounde bounded by y aa loss loss of of offsite offsite power power.. Therefore, gagging valve Therefore, gagging valve 1P11 1P11 F515 F515 open will not open will not create create a ossibility for possibility for aa malfunction malfunction of tructure, system, of a structure, system, or component important or component important to safety with to safety with a different result a different res than any previously viously evaluated evaluated in in tthe UFSAR.

7. design basis Result in a design basis limit limit for fission product for aa fission barrier as product barrier as des ibed in described the UFSAR in the UhSAR hero being 0 Yes exceeded or altered?altered? [83 No BASIS: Gagging BASIS: Gagging valve 1 P11 F515 valve 1P11 F515 open open ensures adequate water ensures adequate water isis available available to to both both the the seal seal steam ste generator and generator and the fire protection the fire protection system system in containment . The in containment. The P11 system and P11 system this valve and this valve havehave no no impact impact on any fission product barriers and are not on any fission product barriers and are not required required to to support support any any systems systems that that do do. . Therefore, Therefore, this cannot this cannot result in result in a a design design basis basis limit limit forfor a fission product a fission barrier as product barrier described in as described in the the UFSAR UFSAR being being exceeded exceeded or red .

altered.

8. Result in a departure from a method of a method of evaluation describe in the UFSAR evaluation described FSAR used used in establishing in establishing Yes the design design basesba safety analyses?

or in the safety analyses? No BASIS: No BASIS: No analytical lytical methods methods were changed with were changed with respect to to fire protection during fire protection during the preparation of the preparation of this this EC. Therefore, refor there ere can can be be nono departure departure from from aa method method of of evaluation evaluation described described in in the UFSAR used the UFSAR used in 1 establishing tthe design establishing design bases bases or or in in the the safety safety analyses.

analyses .

If any of the aboveabove questions questions is is checked checked "Yes," obtain NRC approval approval priorprior to implementing the to implementing the change change by initiating a change nge to to the the Operating operating License license in in accordance accordance with with NMMNMM Procedure Procedure EN-LI-103.

EN-LI-101-AT

10 CFR 50.59 EVALUATION FORM Sheet 4 of 4 EN-LI-101-ATT-9.1,, Rev.

EN-LI-101-ATT-9.1 Rev . 44

GGNS 50.59 Safety Evaluation Number SE 20 8 000 SE 2008-0002-ROO G090036

10 CFR 50.59 EVAlUATION FORM Sheet 1 of 4 I. OVERVIEW I SIGNATURES 1 Facility: Grand Gulf Nuclear Station Evaluation t# I Rev. #: 200a-QO02*ROO Proposed Change I Document: EC 8474 Description of Change: This evaluation IS for lowering the reseat pressure of the ADS air receiver and accumulator relieff valves valves from 171 pSlg to 152 pSlg This affects the air receiver capacity which affects the time durat,on before re aan operator ,s required to go and connect make up air to the ADS system The current time iIS 4 days ThisThis evaluation evaluation will will lower lower the the titime to 3 5 days aluat are depen Is the validity of this Evaluation dependent on any other change? o Ves ~ No No If "Yes," list the reqUired changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other Identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action Is completed.

Based on the results of this 50.59 Evaluation, does the proposed change require prior NRC approval?

o Ves t8J No No Preparer:

Preparer: Robei idler I1 See Robert W Fuller See lAS lAS 1I EOI EOII1 CAE-Mech/Civil/

DE-Mach/Clvlll 7-9-08 Name igna Name (print) J Signature I Company I Department I Date ate Reviewer: Dave Wilson I See lAS I EOII DE-MechlClvIl/8-8-08 Name (print) I Signature I Company I Department I Date OSRe: Michael A Krupa I See lAS I EOII Nuclear Safety Assurance I 08-26-08 Chairman's Name (pnnt) I Signature I Date 2008-028 2008=028 OSR OSRe Meeting #

1 Signatures may be obta.ned vIa electronic processes (e 9 PCRS. ER processes), manual methods (8 9 Ink signature).

I I e...mall. or telecommunlCabon. If uSIng an e...mall or telecommunlcabon. attach tt to thiS form EN-LI-101-ATT-91, Rev 4

10 CFR 50.59 EVALUATION FORM Sheet 2 of 4 of 4

u. 50.59 EVALUATION Does the proposed Change ge being being evaluated evaluated represent represent a a change c to a method of evaluation

.Qt!bY? If "Yes," Questions tions I1 - - 7 are not 7 are applicable ; answer not applicable; answer only o Question 8. If "No." answer 0 Yes all questions below. t8J No Does Does the proposed Change the proposed Change

1. Result In more than a a mammal increase In minimal Increase the frequency in the frequency of occurrence of an accldent 0 Yes previously evaluated In the UFSAR? t8J No BASIS BASIS ThisThis change change willwill not not affect affect the Inlt,al conditions, the initial prior to conditions. prior an accident, to an requi accident, required for the proper operation n of of ADS ADS or any other or any other system system or or component component The The change change willwill not not affect the function affect the functton ofof any any system s

or component. The or component.. The ADS opening and ADS opening operation is closing operatIon and closIng IS not not affected affected andand wIll will continue continue to to function function as a

previously described previously described In in the FSAR . The the FSAR. normal and The normal and low-iow low-tow set set fu funcllon of the SRVs are unaffected by the change The change lowering of The lowering of the relief valve the rehef valve reseat resent pressure pressure to 152 pSlg to 152 p will not affect the rehef valve's ability, ho however. It Will decrease the maximum number of allowable days to connect makeup air to the ADS air receivers ThiS IS WIthin the design basIs of the affected equipment and therefore wlIl not Increase the probability of occurrence of any accident previously evaluated an the FSAR 2 Result in mora than a minimal .ncrease In the likelihood of occurrence of a malfunction of a 0 Yes structure, system. or component Important to safety previously evaluated In the UFSAR? 18) No BASIS Lowering the reseat pressure of the ADS air receiver and accumulators rehef valves will not affect the operatJon of the ADS rehet valves The lower reseat pressure of the relief valve wrll result In a change to the allowable time to connect makeup air to the ADS air accumulator and receivers Calculatton MC-Q1821-96014, Rev 1 determined the decrease In pressure of the contained air for ADS operation and will result ,n a larger mass loss from the system whenn the ADS air system rehef valves are operated due to the lower reseat pressure This Will not affect the Initial al operation opera of ADS. however, It does affect Its long term operatton RevIsion 0 of Calculation MC-Q1 821-96014 Indicated ndi that there IS an air mass loss In the ADS air system due to to the the temperature indu temperature Induced pressure Increase pressure increa dunng a LOCA The ADS air system rehef valves WIll continue to to actuate actuate andand rreclose to preserve the the system rote system Integnty Lowenng the reseat pressure of the rehef valves WIll cause a larger air mass loss than previously calculated The lowering of the reseat pressure will not affect the Integrity of the ADS receivers or accumulators as the ADS atr system rehef valves WlU stilt function as deSigned and their setpolnts and rehevlng capacity rema.ns unchanged The larger air mass loss Will reduce the tong term air capacity without additional makeup from 4 days to 3 5 days ThiS reduced time requiredrequired toto supply mak, supply makeup from an outside source IS stili Significantly greater than the already established ied time limit for time hmlt for providi provld,ng ADS air makeup per ONEP 05-1-o2-V...9 of 1 6 days The plant design Includes the the capability capablhty to to suppJy suppll. pneumatic makeup at any time dunng the event and thiS reduction .n tlme does not adversely affect the capability to prOVide makeup air Per AECM 83/0672, questIon 2 response a leakage rate of 1 SCFH FH per per MSRV MSRV ,s is a a conservatIvely conservatwely assumed leakage rate ,ntended to ,nctude leakage from sources other her than the MSRVs than the MSRVs themselves themselves Per Per GE purchase specification 21A9538, Rev 4, actual maximum deSign ign leakage from IndIVidual leakage from i MSRV actuators,s 0 1 SCFH Per calculation MC-Q1B21-96014, Rev 1, the maximum imurn allowable all leakage from the ADS pneumatic system (other than the MSRVs) IS 3 21 SCFH and therefore thee Iota total leakage IS 3 71 seth (3 21 + 5 MSRVs ... 0 1 SCFH per valve) rather than an assumed 1 SCFH per valve Therefore, the change does not ,ncrease the th probability of occurrence of a malfunctIon of equipment imports Important to safety previously evaluated ated in the UFSAR

.n the UFSAR 3

3 Result In more than a mlnrmallncrease In the consequences of an aCCident previously 0 Yes evaluated ,n the UFSAR? ~ No EN-LI..101-ATI.. 91, Rev 4

10 CFR 50.59 EVALUATION FORM Sheet 3 of 4 BASIS Lowenng the rehef valve reseat pressure will not affect the operation of the ADS rehef valves The lower resea reseal pressure wilt allow more mass to leave the ADS aar accumulators and receivers The ADS elief valves relief valves will win continue continue to to operate operate as as designe designed Lowerang ring the the re reseat pressure of the reltef valve for ADS ton will operalton result in will result in larger larger air air mass mass loss loss from from tthe system m when when thethe ADS ADS air air system system relief rehef valve valves are led du operated due to the temperature Induced pressure Increase. rease . This This will not affect Will not affect the the initial inlt,al operation of ADS; however. It does affect Its long term operation Presently. itIt is IS assumed assumed tthat the ADS air capacity IS capable of maintaining the ADS valves open for long term core cooling for 4 days As a result of the lower reseat pressure the contained air is only capable of maIntaining the valves open for 3 5 days without makeup air, rather than 4 days. The change Will not compromise any safety related system or prevent a safe reactor shutdown The design change will not affect any potentlaJ radIological release po,nt or degrade any barners which are relied upon to mitigate the consequences of previously evaluated accidents. Calculation MC-Q1 B21 ..06006 determined a one tIme exposure to an Indlvidualln.tlaUy establishing an alternatenate pneumatic pneumatic source source to ADS was 1 13 rem This exposure remains valid as the dose the indiViduall willwill receive receive toto connect connect m makeup air to the ADS air receivers Therefore, this change will not Increase the consequences of an an accident aCCident previously evaluated tn the FSAR 4 Result In more than a mln,mal nimaii increase Increase iIn the consequences of a malfunction of a structure, 0 Yes system, or component Important to safety safety previously previously evaluated In the UFSAR? t8I No BAStS The lowering of the rehef valve reseat reseat set set po, point Will not affect the operation of the ADS valves valves CalculatIon MC-Q1 821-96014 determined cned thethe pres pressure decrease for the contained air for ADS operation Will result In larger air mass loss from mass Joss from thethe svste system when the ADS air system rehef valves are operated due to lower the rehef valve reseat reseat pressure pressure ThiS Will not affect the Initial operation of ADS, however, It does affect ,ts long term operatIon The The previous reVision of the calculatIon fndlcated that there IS an air mass loss tn the ADS atr system due to the temperature Induced pressure Increase dUring a LOCA The ADS air system reUef valves Will conttnue to actuate and reclose to preserve the system Integrity Lowering the relief valve reseat pressure wltl result In a larger air mass loss than previously calculated The Increase In the air temperaturel pressure Wilt not affect the Integrity of the ADS receivers or accumulators as their setpolnts and relieving capacity remains unchanged The larger air mass foss wdl reduce the long term air capac,ty w,thout add,t,onal makeup from 4 days to 3 5 days ThiS reduced time requlTed to supply makeup from an outside source Is stilt Significantly greater than the already established time limit for prOViding ADS air makeup per ONEP 05-1-Q2-V-9 of 1 6 days The pjant deSign Includes the capability to supply pneumatic makeup at any time dUring the event and thiS reduction In time does not adversely affect the capability to prOVide makeup air Therefore, the change does not Increase the consequences of a malfunct malfunction of a structure, system, or component rmportant to safety previously vio evaluated In the FSAR 5 Create a POSSibility for an acc\dent ent of a different type than any previously evaluated aluated In the 0 Yes UFSAR? ~ No BASIS The lowering of the rehef valve reseat pressure ressure will will not adversely affect the operation of the ADS rehef valves The In.t,al temperature of the air Within thin tthe components IS not affected as the contained air mperat temperature would equalize With the the drywell drywell temperature temperature over a penod of time The drywell air nature ,ncrease temperature increase due to a lOCA will result In Will result in an in an ,ncrease In the contained air temperature The increase to Increase In the air temperature/pressure sure willwIn notnot affect the integ affect the lntegrtty of the ADS receIvers or accumu'ators as as the the ADS air system rehef valves Will ill still st.1I function function as designed lowering the rehef valve reseat setpolnt as deSigned Will not Increase affect the ability of the relief valve overpressure rehef valve overpressure protection The change wilt not adversely affect any other system or component andd will not result Wilt not result inin any any fa fa.lure that would compromIse reactor safety Therefore, the change Will not create the pOSSibility for an acc.dent of a different type than any previously evaluated In the FSAR EN-LI-1 01-ATT-9 1 Rev 4 J

10 CFR 50.59 EVALUATION FORM Sheet 4 of 4 6 Create a possibility for a malfunction of a structure. system, or component tmportant to safety 0 Yes With a different result than any previously evaluated In the UFSAR? t8J No pecification M912 BASIS SpecIfication M912 0 0 lowers lowers the the reseat reseat pressure pressure of of the ADS arr the ADS air receiver receiver and and accumulator accu relief valves. Although Although the the ADS ADS air system relief alf system valves will rehef valves will stay stay open due to longer due open longer to the lower reseat pressure than previously calculated than previously calculated itIt will win notnot affect affect the integrity of the integrity of the the ADS receivers or ADS receivers or accumulators accumulators as as the the ADS tem rehef arr system relief valves vatves will still function Wlif stilI function as as designed.

designed . Therefore Therefore the the change will not change will create the not create possibility the posSlblitty alfunction of for a malfunction of equipment equIpment importantimportant to to safety safety ofof a different type a different than an type than any previously evaluated tn the FSAR R 7

7 Result Result In in aa design design basisbas.s limit hmlt forfor a a fission product barner fission product barrier as as described described In in the the UFSAR UFSAR being being 0 Yes seeded or exceeded or alt altered? [83 No BASIS.

BASIS. This This does does not not affect affect a a fission product barner fission product barrier The The maximum maximum allowable allowable timetime to to connect connect makeup makeup air air to ADS ADS system system will wIll bebe reduced reduced from from 4 days to 4 days to 3355 days days The The larger air mass larger air mass loss loss Wilt will reduce reduce the long term the long ter air capacity capacity without wtthout additional additional makeupmakeup from from 4 4 days days toto 335 5 days days ThiS This reduced reduced timetime required reqUired to to supply supp makeup makeup from from an an outside outside sourcesource IS is stili still significantly greater than sIgnificantly greater than thethe already already established established time time limit lima for for providing ADS prOViding ADS air makeup per air makeup per +DNEF o5-1-02-V-9 ONEP 05-1-Q2-V. of 11 6

. 9 of 5 days days The The connection connection of the makeup of the makeup air assure air assures that that there there is adequate air IS adequate air supply supply to the ADS to the ADS valves valves during during a LOCA The a LOeA accident is The accident IS a small hne a small line break break accident aCCident which which causes causes slow slow depressurization depressurization of of the the reactor vessel and reactor vessel and high high temperature temperature In in the drywell the drywall The The high temperature In high temperature in the drywell heats the drywell heats up the ADS up the ADS air receivers and air receivers and accumulators accumulators and causes the and causes the relief valves reUef valves to to lift 11ft.. The reseating of The reseating of the the relief relief valve valve will delayed due be delayed Will be due to the lower to the lower reseat reseat pressure.

pressure, ThiS Thf; In turn causes larger turn causes larger Sir air mass mass loss from the loss from the ADS ADS airair receivers receivers andand accumulators accumulators ThiS This loss loss of of air air mass mass Will will cause the cause the need need for ADS makeup for ADS makeup supply supply air be connected to be air to connected soonersooner The The maximum maximum allowable allowable timetime period period for connection is for connection IS 3 35 days 5 days 8 Result Result In in a a departure departure from from a a method method of of evaluation evaluation described descnbed In in the the UFSAR IJFSAR used used Inin establishing establishing 0 Yes the the design deSign bases bases or in the or In safety analyses?

the safety analyses' t8J No BASIS The BASIS The method method used used to to analyze analyze the the time t.me toto depressurlze depressurize the the receivers receivers andand accumulators accumulators was was thethe same same methmethod previously established and previously established and there there was was no no change change to to the the met methodology If any of the above questions Is checked "Yes," obtain NRC approval prior to implementing the change n e by Initiating a change to the Operating License in accordance with NMM Procedure EN...LI*103.

EN-LI-101-ATT-91. Rev 4

GGNS 50.59 Safety Evaluation Number SE 2008-0003-ROO G090036

10 CFR 50.59 EVALUATION FORM Sheet 1 of 4 1

l. OVERVIEW I SIGNATURES Facility: Grand Gulf Evaluation" I Rev. fl.: 2008..0003*ROO Proposed Change I Document: EC 9133 Description of Change: The GGNS Feedwater Leading Edge Flowmeter (LEFM) system has two meters (one for each feedwater line) Each meter has 8 paths (16 transducers) that are grouped via software rnto two essentially Independent measurement systems (planes) Plane 1 consists of paths 1.2.3 & 4 and plane 2 conststs of paths 5, 6. 7 & 8 If one path in a plane IS ~ost. the plane lS taken out of service and the system IS In Alert or Ma1ntenance mode AdditIonal path failures In th,s plane wil' have no effect on system operation. An additional path failure In the other plane win result In the system gOing to Fait mode. As documented In CR 2008-3854, Qath 1 and path 8 of LEFM meter 1 have degrading transducers that caused the paths to fad.

Since they are in different planes. the system was in Fad mode and reactor power was restricted to 3833 MWt Qer TRM 6.3 12 Temp mod EC 9203 was Issued to restore the path operabIlity but It rs possIble that the downpower scheduled for 8/29/08 Will aga.n cause these paths to fad because of their degrading condition Cameron (formerly Caldon) IS proposing that the software configuration setting (INn files be modified to allow the LEFM system to operate in Normal mode With path 1 and path 8 failed. DeSign Engineering has reVIewed and approved the proposed modification. The paths WIll be regrouped via the modified software settings Into virtual plane 1 (paths 2. 3, 4 & 5) and virtual piane 2 (paths 4. 5, 6 & 7) where the data for the faded paths has been substituted WIth data from orthogonal functional paths The Situation soecific uncertainty analysis done by Camergn IS not valid beyond RF16 Is the validity of this Evaluation dependent on any other change? DYes l8J No If "Yes," list the required changes/submittals. The change. covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request).

Establish an appropriate notiflcatlon mechanism to ensure this action is completed.

Based on the results of this 50.59 Evaluation, does the proposed change o Yes ~ No require prior NRC approval?

Preparer: Timothy M Bryant J see lAS I EOII DeSign Englneenng-I&C J 8-26-08 Name (prrnt) I Signature I Company I Department I Date Reviewer: Nathan Jones I see lAS I EOII System Engineering I 8...26-08 Name (pnnt) I SIgnature I Company I Department I Date OSRC: Michael A Kru a I see lAS /8-26-08 Chairman's Name (prrnt) I Signature 1 Date 2008-028 OSRC Meeting #

1 Sagnatures may be obtained via electronic processes (e 9 , PCRS, ER processes), manual methods (e 9 Ink signature).

t e-mail, or telecommun,cabon If uSing an e-mail or telecommunication, attach It to thiS fann EN-LJ-101-ATT-91, Rev 4

10 CFR 50.59 EVALUATION FORM Sheet 2 of 4 II. 50.59 EVALUATION Does the proposed Change being evaluated represent a change to a method of evaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only Question 8. If "No." answer ~ Yes all questions beCow. 0 No Does the proposed Change Result In more than aa minimal mrnlmal increase increase inIn the the frequency frequency of of occurrence occurrence of of an an accident accident 0 Yes previousry evaluated en n the the UF UFSAR?AR? 0 No BASIS N/A 2

2 Result In more than a minimal minimal increase Increase inIn the the likelihood likelihood of occurrence of of occurrence of a a malfunctio malfunction of a 0 Yes structure, system. or component Important to safety previously evaluated In the UFS UFSAR? 0 No BASIS: N/A 3

3 Result in more Result in more then than a minimal increa a minimal Increase In the consequences of an accident previously 0 Yes lusted in evaluated In the UFSAR? 0 No BAStS* N/A 4

4 Result In more than a mInimal Increase In n the consequences of the consequences a malfunction of a malfunction of of a structure, a structure, 0 Yes system, or component Important to safety previously usly evalu evaluated In the UFSAR? 0 No BASIS- NJA 5 reate a Create a oossibi poss,bil,ty for an accfdent of a different type than any previously evaluated In the 0 Yes UFSAR UFSAR? 0 No BASIS NJA 6 Create a possibility for a malfunctIOn of a structure, system, or component rmpo Important to safety 0 Yes With a different result than any previously evaluated in the UFSAR? 0 No BASIS- NJA 7 Result in a deSign basis Is limit r,mit for a fIssion product barner rrier as as descri described in the UFSAR befng 0 Yes exceeded or altered? D No BASIS. NJA 3

8 Result in a Result in a departure from a departure from a method method ofof evaluation de evaluatIon described In the UFSAR used .n establishing 0 Yes bases or the deSign bases in the or In safety analyses?

the safety analyses t8J No BASIS:

The LEFM system (ultrasonic flowmeter I UFM) IS required to be operable above 25% thennal power per TRM 6.3.12. If the LEFM system can not be restored wltlun 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, reactor power must be restncted to 3833 Mwt . TRM basts sectIon 6.3 12 paragraph 1 states; "The u}trasonlc tlowmeters (UFMs) operate with significantly lower uncertaInties in feedwater flow and temperature than the feedwater venturis and RTDs, respectIvely I v The The OPERABILITY OP of both UFMs ensures that core power IS known WIth suffiCIent accuracy that plant safety and deSIgn bases are supported at operatIon up to core power levels of3898 Mwt. u TRi\tl basts section 6.3.12 paragraph 4 states; "The UFM system performs continuous Internal self-morntoring, therefore Its operabIlity status IS always available The UFM system IS consIdered OPERABLE as long as this self*monltonng functIon detennines that these Internal dIagnostIcs are In the proper ranges. u EN-LI-101-ATT-9 1. Rev 4

10 CFR 50.59 EVALUAnON FORM Sheet 3 of 4 A brief dIscussion of the LEFM system is contaIned In the NRC SER for amendment 156 to the faclhty operatIng license (GNRI 2002/0121) winch approved the uprate from 3833 Mwt to 3898 Mwt. A more detailed discussion is contained in the NRC SER for Caldon engtneenng report ER-157P supplement to topical report ER-80P (GNRI 2001/0147). The ER-157P SER transmittal states; "'On the basis of our review, the statT finds the subject report to be acceptable for referencing In power uprate license applications for Waterford Steam Electric StatIon, Unit 3 (Waterford 3), Grand Gulf Nuclear Station (GGNS), and RIver Bend StatIon (RBS), to the extent specIfied, and under the Hmltations delineated in the report, and tn the enclosed safety evaluatIon (SE). The SE defines the basis tor the NRC acceptance of the report.~' A discussion of the LEFM system is also contained in the SER for ER-80P (GEXI 2001-0183).

The proposed change will not adversely affect the desIgn functIon of the LEFM system which IS to measure feedwater flow with an accuracy of +/-O.29% full scale mass flow and feedwater temperature with an accuracy of +/-O 67°F as assumed In GGNS core power uncertaInty calculation XC-Ql1 11-00001. Based on Cameron engineering report ER 699, the assumed uncertainty values are still bounding while in Nonnal (Check plus) or Alert (Check) modes of operation wIth the new virtual plane setup. Therefore the changes to the method ofperfonning and controlling the design function are also not adverse. The selfmonitoring features of the LEFM system are not affected. However, it should be noted that, because each vIrtual plane uses both rematnlng outsIde paths (paths 4 and 5), a failure of either of these paths will lead to the meter entering the "fail" mode (sInce a faIlure of eIther of these paths win lead to a path failure lD both virtual planes). The method of evaluatlon must be addressed via 10CFR50.59 evaluation sInce It was updated to account for the addItIonal uncertainties. The situatlon specific uncertainty analysts done by Cameron Cameron will not be vahd alid after after RFt6.

The analysis was based on a review of8 months ofLEFM data to determIne the variabtlity of the vIrtual meter factor against the current 8 path meter factor. The Cameron revIew of 8 months of LEFM data demonstrated clearly that the sWirl and radial velocity components are neglIgible and constant withIn the margtn already accounted in Grand Gulfs DeSIgn Basis uncertainty analysis.

As a result, substItutIng data from orthogonal paths to create new virtual planes introduces httle additional error. The Increased uncertainty aSSOcIated with the VIrtual 8 path meter factor is In the order of+/- 0.. 05%. When thIS uncertainty IS combined as the root sum square wIth the the actual ainty of uncertainty of the the 88 path path CheckPlus system (+/-

CheckPIus system (+/- 0.26%)

0.26%) the result remains +/- 0.. 26%.. SInce the Grand Gulf LIcensing Basis uncertainty is +/- 0.29%,  %, the the shght s Increase in the Loop A meter factor uncertainty can be readily accommodated ated within the current Licensing BasIS.

To provIde for the pOSSIbIlity of an additIonal fatled path, meter factors and uncertalntles for each of two virtual planes are also establIshed. They are determined from the perfonnance of each virtual plane as against the two plane meter factor over the 8 month data penod. The analySIS of each VIrtual plane demonstrates that the uncertaInty of Its meter factor is equal to or less than the Matntenance Mode uncertaInty in the DeSIgn Basis Uncertainty Report, that IS, the uncertatnty used in Grand Gulfs licenSIng Basis. It IS therefore acceptable, in the event of a faIlure of any one of paths 2, 3, 6, or 7, to use the remaining fully functIonal Virtual plane to determIne the one feedwater mass feedwater mass flow for Loop A over the remaInder of the cycle, SInce the flow measurement uncertainty will be WIthIn the current Grand Gulf LicenSIng Basts. The mass flow uncertaInty WIll uncertainty increasee to to +/-0.28%

+/-O.28% forfor aa path path in still bounded alert, stIll in alert, by the bounded by t licensing base uncertainty of +/-O.29%

EN-LI-101-ATT-91, Rev 4

10 CFR 50.59 EVALUATION FORM Sheet 4 of" The uncertainty of+/-O.29% represents +/-O.29% full scale mass flow, I.e , not percent power or percent flow readIng. The temperature uncertainty of+/-O.67°F remwns unchanged..

NEI 96...07 rev. I (Guidelines for IOCFR50.59 ImplementatIon), specIfies that methodology elements can be changed without NRC approvaltfthe results are "essentially the same". Based on the above discussion, the virtual plane setup proposed by Cameron is acceptable and IS "essentIally the same" as the standard LEFM setup outlined in Caldon topical report ER 157P. The Cameron engineering report ER699 and the supporting analysis were prepared under their 10CFR50 appendlx B QA program. The changes to the software INI files will be controlled via 01-8-17*46 If any of the above questions Is checked "Yes,t' obtain NRC approval prior to implementing the change by initiating a change to the Operating License iin accordance with NMM Procedure EN*LJ-103.

EN...LI-101-ATI*91, Rev 4

9£00608 00t:l-17000-800~ 3S JaqwnN UO!le n leJ\3 AlaJes 6S*0S SN~~

1 I. OVERVIEW I SIGNATURES Facility: GGNS lua Evaluation 2008-0004 Rev. #: 00 Proposed Change I Document: Cycle 17 COLR (LBDCR 2008-029)

Description of Change: The core reload is a recurring activity for each fuel cycle. At the 'end of each fuel cycle, depleted fuef assemblies are discharged from the core and replaced by fresh reload assemblies. The remaining bundles resident in the core are shuffled to new locations and fresh fuel is loaded in accordance with the next cycle's core design and reference loading pattern. This evaluation addresses the Cycle 17 reload changes and operation of the Cycle 17 reload core.

Is the validity of this Evaluation dependent on any other change? El Yes DYes 2

[g] No No If "Yes," list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action is completed.

Based on the results of this 50.59 Evaluation, does the proposed change Mange El Yes DYes [g] No require prior NRC approval?

Preparer: Scott Stanchfield I I ESI I BWR Fuels I Name (print) I . pan 1 epartment I Date Reviewer: Guy B. Spikes I J:b,y 6. KC::c;z I ESI I Nuclear Analysis I ld 111 JfJ8 OSRe: I Chairman's Name (print) I Signature I Date 034-2008 OSRC Meeting #

lectronic processes 1 Signatures may be obtained via electronic processes (e .g . PCRS, ER processes), manual methods (e.g., ink signature),

(e.g.,

lecommunication, aft e-mail, or telecommunication. If using an e-mail or telecommunication, attach it to this form.

EN-Li- 10 i -H- T-i-9 EN-LI-101-ATT-9.1, Rev. 4

II. 50.59 EVALUATION Does the he proposed proposed Change Change being evaluated represent being evaluated represent aa change change to to aa method of evaluationluation ONLY?  ? If If "Yes,"

"Yes," Questions Questions 11 -- 77 are are not not applicable applicable;; answer only Question 8. If "No," answer Y answer R Yes DYes Lions below.

all questions beta ~ No No Does the Does propose Change:

the proposed anger I.

1. Result in Result in more more than than aa minimal minimal increase increase in in the the frequency frequency ofof occurrence occurrence of of an acrid an accident DYes iously evaluated previously evaluated ini the UFSAR? ~ No BASIS:

ycle 17 The Cycle 17 core core loading loading and and cycle cycle operation will not result in more than a a minimal minim increase in the

'frequency of occurrence of an accident of an accident previo previously evaluated in the FSAR. The precursors precursors to these events are independent of the core design and the frequency classifications reported in in FEAR FSAR Chapter Chapter 15 are unaffected by the core parameters. The following considerations support this conclusion. ion.

Mechanical The ATRIUM-10 and GE14 mechanical designs have been reviewed for use at Grand Gulf. No unusual failure modes or increased failure frequency have been identified for this fuel design. This is the sixth cycle of operation for ATRIUM-10 fuel and this fuel design has accumulated operational experience at GGNS and other plants with no significant problems. Although the GE14 1 fuel design is being employed for the first time at GGNS, there is a significant amount of operational rational experience experience within the fleet with this fuel design. The bundles will operateAerate within within the the power history assum power history assumptions in the fuel mechanical analyses and will11 result in exposures result in exposures within within the analyzed burnup the analyzed limits of burnup limits of the the mechanical designs, including those bundles bundles that that will be irradiated will be irradiated forfor aa fourth fourth cycle.

cycle . Althoug Although an

. increased channel bow condition can result in increased friction between the control blade and i and its corresponding fuel assemblies, control rod settle and insertion testing (EPI 04-1-03-C11-7) will continue to be performed during Cycle 17 to ensure that the increased axial friction loads on the channel and fuel assembly load chain remain below acceptable limits.

Nuclear The neutronic characteristics of the Cycle 17 core design have been considered in the safety analysis. Adequate shutdown margin has been predicted by analysis and will be confirmed during startup tests. In addition, the hold-down capability of the standby liquid control system and the subcriticality of Cycle 17 fuel in the spent fuel storage racks have been confirmed. Therefore, the probability of inadvertent criticality has not been increased by the introduction of the Cycle 17 reload fuel.

Thermal-Hydraulic Cycle 17 is mixed core of ATRIUM-10 and GE14. Analyses have been performed to show thermal-hydraulic compatibility with co-resident fuel types. Analyses have been performed to demonstrate that Cycle 17 meets all Enhanced-1 A stability performance criteria without changes to the E1A hardware or power-flow map region boundaries. Therefore, the probability of thermal-hydraulic instabilities has not increased.

Analyzed Events The probability of the occurrence of anticipated operational events is not dependent on the core configuration. No changes to the plant design are required for the Cycle 17 core. The Cycle 17 core loading will not affect the precursors to any of the Chapter 15 events. The probability of an analyzed event therefore has not increased.

As described in FSAR Section 15A.6.5.3, the Control Rod Drop Accident (CRDA) results from a LN-U-M-Al 1- 9 EN-LI-101-ATI-9.1,Rev.4

failure of the control rod-to-drive mechanism coupling after the control rod becomes stuck in its fully inserted position. Although an increased channel bow condition can result in increased friction between the control blade and its corresponding fuel assemblies, analyses have shown that there would not be sufficient friction to result in a mechanical failure of the coupling. Additionally, the control rod drive mechanism would not produce enough force to result in a mechanical failure of the coupling even if the channel bow was so severe that the assemblies would preclude blade movement. As such, channel bow is not considered a precursor to the CRDA, and any increased bow associated with the high exposure ATRIUM-10 bundles would not increase the probability of this event.

On these bases, the th probability robability of occurrence of accidents previously identified in the FSAR is not increased for increased for the the Cycle Cvcle 17 17 core core with increased channel bow.

2. Result in more than a minimal imal increase increase in in the the likelihood likelihood of of occurrence occurrence of of a malfunction of a DYes structure, system, or component ponent important important to to safety previously evaluated safety previously evaluated in in the UFSAR? [R1 No BASIS: No plant modifications are required to accommodate the Cycle 17 core design. The only additional loads placed on plant equipment uipment would would bbe due to increased friction between the control blades and excessively ively bowed bowed ATRIUM-10 ATRIUM-10 bundlesbundles.. This This probability probability has been reduced has been reduced byby p previous re-eling campaigns channeling campaigns for for fuel bundles most fuel bundles susceptible to most susceptible abnormal bow.

to abnormal bow . Based Based on previous experience with bowed fuel fuel atat GGNS GGNS and and other other BWR-6's, BWR-6's, increased control blade increased control blade friction friction can result in can result in increased increased control rod settle times but settle times but is is not expected to not expected significantly impact to significantly scram times.

impact scram times . Technical Technical Specification Specification time testing scram time testing andand control control rodrod settle settle and and insertion testing (EPI 04-1-03-C11-7) will continue to be performed d during during Cycle Cycle 17.17. T These actions would identify any potential scram time or other impacts and such that appropriate corrective actions appropriate corrective actions are are to taken. These actions will ensure that the increased control blade blade friction loads are friction loads are not not sufficient to cause sufficient to anv failures cause any failures associated associated with the control with the control bl blades or the control blade drive drive system, system, the the fuel fuel assembly assembly loadload cha chain, or the vessel internals.

A reactor internal structural assessment has been performed for the reload fuel. This assessment included normal, upset, emergency and faulted condition loads on the reactor internals and concluded that the reactor internals remain qualified for the reload fuel.

nse five vessel A conservative vessel overpre overpressurization analysis has been performed, which shows that the e vessel vessel pressure limit limit iis not exceeded.

The precursors to any malfunction of equipment important to safety fetv are not affected by this the Cycle 17 reload core. Therefore, there is not more than a minimal imal increase increase in in the the likelihoo likelihood of an occurrence of a malfunction of a sse important to safety preViously Viously evaluated in the FSAR.

3. Result Result inin more more than than aa minimal minimal increase increase in the consequences of an accident previously DYes evaluated evaluated in in the the UFSAR? [R1 No BASIS BASIS:: As As reported reported in in Attachment Attachment 1, 1, the the acceptance criteria reported in FSAR Section 15.0.3.1 and the Technical Technical Spec Specifications are satisfied for each event classification. Core operating limits have been evelODed to developed to enensure that moderate frequency events do not violate the MCPR safety limit or fuel cladding strain limits. The consequences of infrequent events have been shown to meet the appropriate acceptance criteria riteria while the individual acceptance criteria for the limiting faults have been demonstrated to be be satisfied satisfied.. As As such, such, the the consequences consequences of of in infrequent events and limiting faults described in the FSAR are are unchanged unchanged for for the the Cycle Cycle 17 17 reload reload core core.. The The following following considerations considerations support support these these conclusions conclusions..

Moderate Frequency Events The Cycle 17 core operating limits have been developed with NRC-approved methodologies such that the MCPR safety limit and the fuel cladding strain limit will not be violated by any analyzed EN-LI-101-ATT-9.1, Rev. 4

moderate frequency transient initiated from any statepoint available ilable, to to GGNS.

GGNS. As As such, such. no no fuel foe failures are expected to result from any moderate frequency event. These analyses I considered GGNS-specific operational modes such as MEaD, SLO, FHOOS, and EOC-RPT PT inoperable.

ino These core operating limits consist of MCPR, MAPLHGR and LHGR curves thatt are are functions funs of flow, power, and exposure. These limits consider conservative channel bow assumptions for for the expected TRIU increased bow associated with the highly exposed ATRIUM-10 Aerating limits will fuel. These core operating rated into the core monitoring be incorporated ring system.

I Infrequent Ev Events The consequences of the limiting infrequent events have been evaluated and shown to meet their respective acceptance criteria. These events include the pressure regulator failure downscale, misplaced (i.e., misoriented and mislocated) bundle and single loop operation pump seizure

'Iaccidents. Radiological analyses using the alternative source term (AST) have been performed to ensure that these events will not result in an increase in offsite or control room doses or doses greater than their respective acceptance criteria. These evaluations include conservative channel bow assumptions that bound the current measured bow data and the expected increased bow associated with the highly exposed ATRIUM-10 fuel.

Limiting Faults The limiting faults at GGNS include the fuel handling accident, the control rod drop accident, and the design basis LOCA. The radiological analyses for these events have been developed as part of the GGNS AST effort and bound the Cycle 17 core parameters. For the LOCA, MAPLHGR operating limits and single-loop mul.tipliers have been developed for the Cycle 17 core configuration such that the requirements of 10CFR50.46 are satisfied. The containment response for the Cycle 17 core was found to be bounded by previous cycles as is the hydrogen analysis. The seismic/LOCA response of the Cycle 17 core has been confirmed to be acceptable.

Therefore, the proposed change does not result in in more more than than a minimal increase a minimal increase in the loafed in consequences of an accident previously evaluated in the FAA the FSAR.

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, DYes system, or component important to safety previously evaluated in the UFSAR? [8J No BASIS: The Cycle 17 GE14 reload fuel design has been shown to be compatible with the co-resident ATRIUM-10 fuel inserted in previous cycles. The malfunctions of key plant components are analyzed as part of the reload process with the results reported in various sections of the FSAR. The consequences of these malfunctions have been shown to remain unchanged for Cycle 17 operation.

Therefore, Cycle 17 operation will not result in more than a minimal increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the FSAR.

5. Create a possibility for an accident of a different type than any previously evaluated in the DYes UFSAR? [8J No BASIS: The Cycle 17 GE14 reload fuel has been shown to be compatible with the ATRIUM-10 fuel that was inserted in previous cycles. The details of this design have been specifically considered in the safety analysis and the core monitoring system. No plant modifications are required to accommodate the new core design or Cycle 17 operation. The GGNS Cycle 17 fuel has been approved for GGNS reactor chemistry conditions.

The GGNS operational parameters (water chemistry re uirements, sspectral-shift requirements, ift cor core designs, and EN-LI-101-ATT-9.1, Rev. 4

MEOD rod-lines) rod-lines) have have bee been reviewed and are not expected to result result in unusual crud in unusual crud buildUD li buildup like observed on that observed on the high the high-power GE111 bundles bundles atat R River Bend. Inspection of a high-power, once-burnt representative fuel bundle during ing GGNS GUNS RF1 RF10 0 has has confirmed confirmed thatthat the hi the high-power GGNS Cycle 10 fuel fuel bundl usual crud bundles have no unusual crud buildup buildup..

Therefore, Cycle 17 operation will not create a possibility possibility for an accident for an accident of of a a different different type type than than an any previously evaluated in the FSAR.

6. re, system, Create a possibility for a malfunction of a structure, system, or or component i component important to safety Yes DYes with a different result than any previously evaluated in n the the UFSAR?

UFSAR? ~ No No BASIS: The Cycle 17 GE14 reload fuel design has been shown to be mechanically, neutronically, and thermal-hydraulically compatible with the co-resident ATRIUM-10 fuel. A mixed core analysis has shown that the reload fuel will not introduce any adverse flow distribution effects. No plant modifications are required to accommodate the new core design and no additional loads will be imposed on any existing equipment. The GE14 bundles provide sufficient clearance for proper control blade operation and allow sufficient bypass flow in the bypass region to provide adequate cooling for control blades and in-core detectors. There are no special operational considerations associated with the Cycle 17 core other than those associated with the increased bow condition. Control rod settle and insertion testing (EPI 04-1 C11-7) will continue to be performed during Cycle 17 to ensure that the increased control blade friction is not sufficient to cause any failures associated with the control blades or the control blade drive system, the fuel assembly I~ad chain, or the vessel internals.

Therefore, Cycle 17 operation will not create the possibility possibility for for aa malfunction malfunction of an sse important to safety with a different result than previously evaluated!ted in in tthe FSAR.

J. barrier as Result in a design basis limit for a fission product barrier described in as described in the the UFSAR UFSAR beingbeing El DYesYes exceeded exceeded or or altered altered? ~ NoNo BASIS: Mechanical analyses have been performed to ensure that all fuel in the Cycle 17 core meet the mechanical design limits for steady-state operation as well as transient conditions including fatigue damage, creep collapse, corrosion, fuel rod internal pressure, rod bow, internal pressure, etc. Additionally, no Cycle 17 fuel will exceed the applicable burn-up limits.

Core operating limits have been developed using NRC approved codes in order to ensure that the Cycle 17 fuel will not exceed the MCPR safety limits for steady-state operation and anticipated operation occurrences. Similarly, operating limits have been developed to ensure that the Cycle 17 fuel will not exceed the 1% cladding strain limit or experience core-wide fuel melt during steady-state operation or AOO's. Although some vessel blowdown to the suppression pool may be experienced during some AOO's, which would increase the suppression pool temperature, the bulk containment pressure increase is negligible and would not exceed the design limit.

As described in Attachment 1, a bounding pressurization event with a failure of the direct scram has been analyzed for Cycle 17 to ensure compliance with ASME code requirements. This analysis indicates that the vessel pressure safety limit is not exceeded for Cycle 17.

A design basis limit for the peak fuel enthalpy of 280 cal/gm has been n established established for the control for the control rod drop accident (CRDA) to preclude significant fuel cladding failure such ch that that core geometry and cooling may be impacted. An evaluation has been performed to shown shown that that the generic GNF CRDA analysis is applicable to GGNS. This This generic generic analysis analysis sshows that a CRDA will not exceed the 280 I

cal/gm peak enthalpy limit. Since this accident is a localized event and the peak enthalpy does not exceed 280 cal/gm, there is no impact on on the the vessel vessel or containmen pressures. As such their or containment respective limits are not exceeded.

EN-LI-101-ATT-9.1, Rev. 4

FIR UATION FORM 10CFR50.46 provides limits associated with the ECCS performance analysis (LOCA analysis). Two such limits are Peak Clad Temperature (PCT) and local clad oxidation. Although these limits are not subject to 10CFR50.59, they are discussed in this evaluation for completeness. Grand Gulf specific analyses have been performed for ATRIUM-10 and GE14 fuel in accordance with 10CFR50.46.

These analyses, which are applicable to Cycle 17, show that the PCT and local oxidation are well below the limits set forth in 10CFR50.46. These analyses also show that the core-wide metal water reaction, which is used to evaluate compliance with the containment design limit, is less than the 10CFR50.46 limit. The remainder of the existing containment analysis associated with LOCA events is applicable to Cycle 17 as described in Attachment 1. As such, the containment pressure design limit will not be exceeded in Cycle 17.

IIAn ATWS evaluation has also been performed for Cycle 17. As described in Attachment 1, the aping vessel pressure remains below the ASME emergency vessel pressure limit of 1500 psig resulting and temperature response the temperature and the response used in the used in the existing exi ATWS containment analysis is applicable to Cycle 17. Thus, Cycle 17. Thus, the the containment containment pressure design limit will not be exceeded for the ATWS event..

pressure design Additional evaluations have been performed for Cycle 17 including Appendix radix R (Fire Protection),

R (Fire Protect hydrogen analyses, and SSO as described in Attachment 1. These evaluations I show t!that the existing evaluations are applicable to Cycle 17 and that their respective limitslimits are not exceeded are not exceeded.

Therefore Therefore, Cycle 17 operation will nott result result in in aa design basis limit design basis for aa fission limit for product barrier fission product barn as described in described the FSAR in the FSAR being being exceeded or altered.

8. Result in a departure from a method of evaluation described in the UFSAR used in establishing DYes the design bases or in the safety analyses? [8j No No BASIS: The reload analyses currently described in the UFSAR are based on AREVA's NRC approved methodology. With the transition to GNF for Cycle 17, GGNS will be implementing GNF's NRC approved methodology, which is covered by GESTAR. The NRC has reviewed and approved the GESTAR methodologies for use in ,neutronic, thermal-hydraulic, transient, and LOCA licensing calculations. As such, replacing the current AREVA methodology with GESTAR does not result in a departure from a method of evaluation described in the UFSAR. Additionally, GESTAR is currently included in the list of NRC approved methods given in Technical Specification 5.6.5. As such, there is no Technical Specification change required to implement GNF's reload methodology.

For Cycle 17, a new Appendix R safe shutdown analysis was performed using the SAFERIGESTR-LOCA methodology. This methodology is different from the SAFE/CHASTE method currently described in the FSAR (Appendix 9C). SAFER has been approved by the NRC for long-term reactor vessel inventory and peak fuel cladding temperature analysis for BWR LOCA's as welt as other off-normal reactor transients. SAFER has also been approved for use in the long-term loss of feedwater (LOFW) event evaluation. The LOFW event is similar to the Appendix R event. SAFER has been used for Appendix R analysis for several BWR plant power uprate projects, as well as for stand-alone Appendix R analyses, which have been accepted and approved by the NRC. As such, replacing the current Appendix R SAFE/CHASTE method with the new SAFERIGESTR-LOCA method does not result in a departure from a method of evaluation described in the UFSAR.

A new ATWS analysis has also been performed for Cycle 17. The GGNS UFSAR does not explicitly describe the methods used in the ATWS analyses. However, the new ATWS evaluation analyzes potentially limiting events for both reactor vessel integrity and containment response. The vessel integrity evaluation uses GNF's ODYN code which has been approved by the NRC for ATWS evaluations. The containment integrity evaluation is performed using GNF's STEMP code. STEMP has been used in the generic BWR ATWS analysis as well as for several plant-specific BWR ATWS analyses for power uprate projects, which have been accepted and approved by the NRC. As such, EN"LI"101"ATT"9.1 ,Rev. 4

pproved methods the Cycle 17 ATWS analysis uses NRC approved methods and does not result in a departure from a method of evaluation.

Other current analyses described in the UFSAR such as the dose analyses, containment analysis, hydrogen analysis, station blackout, have been shown to be applicable to Cycle 17. As such, no new methods were used. The GGNS EP basis calculation has been updated to reflect changes to fuel-related parameters due to the introduction of the GE14 fuel. This revision was performed using the same methods applied in the current EP basis calculation.

GNF developed a critical power correlation called GEXL97 to monitor MCPR for ATRIUM-10 fuel.

GEXL97 was developed in a similar manner to the correlations for GNF fuel designs with the exception that GEXL97 is based on data generated by AREVA's SPC critical power correlation instead of using actual test data. This method introduces additional correlation uncertainty that has been accounted for in the MCPRSL calculation as required by GNF methodology. The GEXl97 correlation has been NRC approved in Amendment 179 and will be incorporated into the list of NRC approved methodology provided in Technical Specification 5.6.5 by LBDCR 2008-032 Therefore, Cycle 17 operation will not result in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses.

If any of the above questions is checked "Yes," obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN*LI*103.

U 'l -/-\ I I -~ . 1, Mt EN-tJ-101"A.If"~l.1, V.

~ev. 4

Attachment 1: GG GGNS Cycle 17 17 Reload Reload Analysis Summary The infonnation in this attachment is a summary of that provided in EC 11027.

1.0 1.0 Cycle Cycle 17 17 Core Core D Description The GGNS Cycle 17 core will be a mixed core of ATRIUM-10 and of ATRlUM-I0 and GE14 fuel assemblies.

Specifically, the Cycle 17 reload licensing calculations are based on a core containing 248 fresh

<;1E14 bundles and 232 once-burnt, 232 twice-burnt, and 88 thrice-burnt ATRIUM-IO assemblies.

The The ATRIUM-I0 fuel type was introduced at GGNS in Cycle 12 and this is the first cycle for GE14.

The Cycle 17 core has been designed for 533 Effective Full Power Days (EFPD), which is based on 544 operating days with a 98% load factor and a core thermal power of 3898 MWt. A breakdown by bun bundle type and bundle average enrichment is provided below:

Cycle 17 Core Composition Cycle Number of Average Bundle Inserted Bundles Enrichment (w/o) 13 4 3.86 14 84 4.043 15 92 3.589 15 140 4.044 16 88 4.039 16 144 4.046 17 24 3.957 17 128 3.908 17 96 3.951 Consistent with previous reloads, Cycle 17 is a conventional scatter load with the lowest reactivity bundles placed in the peripheral region of the core. This loading pattern is designed to maximize cycle energy and minimize power peaking with an aggressive spectral shift operating plan.

During Cycle cle 14 14 operation, control rod operation, control rod behavior indicative of behavior indicative of abnonnal abnormal (shadow corrosion) channel (shadow corrosion) charm bow

)served in was observed in some some control control cells cells loaded loaded with with high high exposure ATRIUM-10 bundles exposure ATRIUM-I0 bundles (CR-GGN-2005-(

03287). The The increased increased bbow condition has also been observed at GGNS and other BWR-6's using GE's vance, (thick-thin) advanced (thick-thin) chachannels. During RF14 and RF15, several ATRIUM-10 channels were examined and measured d to quantify bow to quantify bow and and b bulge. The channel inspections results show that the abnonnal channel bow bow is Sistent with that applied in consistent with that applied in the the safety safety analyses. Several ATRIUM-I0 bundles have been previously ~hanneled to mitigate abnormal channel previously rechanneled to mitigate abnonnaI bow. Zircaloy-4 channels have been used for the channel bow.

past past two reloads since,e they are less s they are less susceptible to abnormal channel bow. Additionally, the Cycle 17 core deli process exp design process explicitly sidered abnormal channel bow resulting considered abnormal channel bow resulting in a core design that reduces the probability of control of blade blade friction friction.. Consistent with Drevio Consistent with previous GGNS and industry experience, increased channel channel bow is not expected to to significantly impact scram significantly impact scram times times or or adversely adversely affect the ability affect the ability to insert

,ontrol blades control blades.

EN-LI-101-ATT-9.1 Rev. 4 t

EVALUATIO 2.0 2.0 F el Mechanical Fuel Mechanical D Design The GE14 fuel design will be the reload reload fuel fuel type type for for GGNS GGNS Cycle 17. Thi Cycle 17. This is is the the first cycle for first cycle for GE GEI4I fuel while the ATRIUM-IO -10 fuel fuel type type was was introduced introduced inin Cycle 12. Although Cycle 12. Although GE14 has nott been been used used atat GGNS, there is a significant amount of operating experience associated with the GE 14 design across the e fleet. Each fuel assembly is a lOx I Ox 10 array 10 array with part-length rods and debris filter lower tie plates.

Consistent with GEl 1 bundlesfiles previously previously used used at FNS the GEl4 bundles will use barrier cladding at GGNS, adding toto ml mitigate PCI related I failures fuel failures. The GE 14 fuel meets the el requirements of Technic fuel requirements of Technical Specification 4.2.1.

2.1 Design Analyses Mechanical design analyses have have been performed with NRC-approved d methodology to evaluate tresses, transient mechanical criteria including cladding steady-state strain and stresses, transient strain strain and and stresses, fatigue damage, creep collapse, corrosion, hydrogen absorption, fuel rodd. intern internal pressure, etc. All parameters were found to meet their respective design limits for the Cycle 17 core.

2.2 Seismic/LOCA Response faun of the GE14 and ATRIUM-IO was evaluated considering the effects of a The seismic perfonnance combined earthquake/LOCA. The maximum stresses on all fuel assembly components (including the interactive channel) were determined to be acceptable. The mechanical design limits of the ATRIUM-IO were shown to remain applicable to the re-channeled bundles. A cycle-independent GEI4 fuel lift assessment was performed for various conditions including nonnal and faulted. In all cases, there remains positive margin to the minimum GE14 bundle uplift forces indicating positive engagement. Evaluations have also been perfonned to assess the impact of the increased channel bow condition on ATRIUM-10 bundles. These evaluations indicate that any increase in bundle uplift force due to axial friction loads is not sufficient to result in an uplift condition, even during seismic events. Additionally, the strength margins for the ATRIUM-IO fuel channel, vessel upper tie plate post, and fuel assembly load chain are acceptable considering the expected increased control blade friction loads. Control rod settle and insertion testing (EPI 04-1-03-CII-7) will continue to be performed during Cycle 17 to ensure that the increased control blade friction is not sufficient to cause axial friction loads greater than that considered in the analysis.

A reactor internal structural assessment was performed for the reload fuel. The rector internals assessed included the shroud, core plate, top guide, orificed fuel support, control rod guide tube, and control rod drive housing among others. This assessment considered all applicable nonnal, upset, emergency, and faulted condition loads for the reload fuel such as seismic, SRV, LOCA, core flow loads and others. The assessment concluded that the reactor internals remain qualified for the reload fuel.

2.3 Top ofActive Fuel Consistent with previous cycles, a generic fuel stack of 150 inches will continue to be used to represent the "top of active fuel" as currently reported in various plant documents such as Figure B 3.3.1.1-1 in the Technical Specification Bases. This stack height is based on a ISO-inch pellet stack height of previous fuel designs. The GEI4 reload eload bundles bundles have a ISO-inch pellet stack height while the ATRIUM-10 bundles have a pellet stack of 149.45 inches for the full-length rods. This small variation is in the upper axial reflector region of the bundle (i.e., the stack height of the enriched region is the same).

The current Level 1 setpoint is conservative and need not be changed. This setpoint and the pellet stack height of the GE 14 and ATRIUM-10 fuel designs has been consideredd Iin the ECCS LOCA analyses. In EN~lj~101~ATT"9.1, Rev. 4

addition, addition, accor according to EP-2A, the operators are directed to reduce dace water water level level during during an an ATW ATWS to -167 lies (i.

inches e. the TAF for fuel designs with ISO-inch pellet stacks) in (i.e., in order order to to decrease decrease core core power power by reducing thethe static head on the core. This action will remain appropriate ate forfor the the ATRILTM-10 ATRillM-I0 bundles tJUIlal(~S since the static head on the power-generating region of these assemblies the enriched uranium) will be essentially the same as the static head on the power-generating region e 150 inch fuel stack height of GE 14 and previous fuel designs.

3.0 Nuclear Design 3.1 Shutdown Margin II The minimum shutdown margin during Cycle 17 is predicted to be at least 1.7% 'Vo Ak/k Llk/k based based on on actual actual Cycle 16 operation. Therefore, the design objective of 1% reported in FSAR Section Section 4.3 .2 .4.1 is satisfied 4.3.2.4.1 is SatlSI1(:~a and there is more than 95% probability/95% confidence fidence that that the the T/

TIS S 33.1.1

.1 .1 required 00.38% .38 0//o Akik value value willwill be satisfied considering the shutdown margin prediction e uncertainties uncertainties.. A A shutdown shutdown margin (SDM (SDM) measurement will be performed during Cycle 17 restart for confirmation of )f required required margin margin..

3.2 SLC Capability The Standby Liquid Control system is designed to inject a quantity of boron that iat produces produces aa concentration concentration of no less than 660 ppm in the reactor core accordingg to to the the Bases Bases for Technical ical Specification 3.1.7.

Specification 3.1 .7.

Calculations for the Cycle 17 core predict a shutdown margin margin of of 11.3%

. ilk/k based based on on aa conservative) conservatively short Cycle 16 operation. The shutdown margin SLC shutdown The SLC margin reauireme requirement of 1% 1% ~k/k for for CA GGNS NS Cycle 17 is tis therefore satisfied.

3.3 Criticality Technical Specification 4.3.1.1 requires that the fuel storage racks cks (upper (upper containment containment pool pool andand spent spent fuelfuel pool) be maintained with a keff ~ 0.95 if fully flooded flooded wi nborated water with unborated water including including an an allowance allowance for for uncertainties as described in Section 9.1.2 of the FSAR. Confin -natory analyses Confirmatory analyses have have beenbeen performed performed to demonstrate that the bundle reactivities and rack predictions ns on Boraflex performance on Boraflex perfonnance through through the the end end oof Cycle 17 are bounded by the assumptions in the current criticality ticality safe safety analysis analysis for for the the Region Region II storage storage locations. This ensures that the k,ff rage locations keff for Region I storage locations will be less than 0.95 less than 0.95 as as required required by by Technical Specification 4.3.1.1 (and nd repeated repeated in in FSAR FSAR Section 9.1). Further, her, the the separation separation requirements requirements associated with the "6 of 16 blocked" clod" Region II storage locations ions ensures ensures that that thethe k ff for for Region Region II locations will also be less than 0.95. The impact of increased channel bow results in itIally results in no no change or a slight reduction of reactivity, which would remain bounded by the current analyses, analyses.

3.4 Void Reactivity Coefficient Section 4.3.1.1.2 of the UFSAR states that the void coefficient "shall shall be be negative negative over over the entire entire opoperating tinge. GNF has perfonned analyses on the GE14 fuel design range." sign for for numerous numerous operating conditions and operating conditions a oderator in concluded that boiling of the moderator in the active channel channel flow flow areaarea results results in in aa negative react -.

reactivity feedback for all expected modes des of power operation" 3.5 Channel Management Channel management is discussed in FSAR Section 4.2.3.3.10 and is designed to minimizee channel channel distortion-related control rod interference in the core. GNF has developed a channel hannel management management, program for BWR's called the cell friction model (CFM). In previous reload core designs, designs, Entergy Enter has considered high neutron flux gradient ent than channel bow for fuel located located on on thethe core periphery and core periphery has taken and has taken EN-LI-101-ATT-9.1, E' I I I 1 1 . Rev.

Rev . 44

VALUATION FOR f 16 the position that all edge bundles, whether adjacent to a control blade or not, will be re-oriented between cycles. The risk of excessive channel distortion due to flux gradient effects in peripheral core locations is included in the CFM model. Therefore, Therefore, excessive excessive channel channel bow bow due due to to high neutron flux gradient is addressed in Cycle 17 in n the CF analysis for bladed core locations. In addition, all Cycle 17 reinsert the CFM fuel assemblies that were located on the edge edge of the cycle of the Cycle 16 have been core have 16 core been verified to have been rotated if located in non-bladed locations, or are directly considered in the CFM if located in bladed locations.

4.0 ThermalI Hydrau Hydraulic Analysis 4.1 COlnpatibility The Cycle 17 core will contain 248 GE14 fuel fuel bundle bundles and 552 ATRIUM-l 0 bundles. GNF performed a mixed-core thermal-hydraulic analysis considering multiple core configurations ranging from a full core of ATRIUM-10 to a full core of GE 14. This evaluation tiara demonstrated that the flow distribution between the co-resident fuels does not result in any degradation of of the the thermal-hydraulic thermal-hydraulic performance or operation of the mixed core during the transition process.

4.2 Stability GGNS implemented the BWROG Long-Term Stability Solution Enhanced Option I-A (ElA) during Cycle

11. Consistent with the approved EtA methodology, GNF performed additional analyses to confirm the continued applicability of the EIA stability solution for the new Cycle 17 reload fuel design, and that operation of the Cycle 17 core design will continue to meet the required stability performance criteria with the existing E1A hardware and region boundaries. The current stability analysis is applicable to the increased channel bow associated with the high exposure ATRIUM-lO fuel and is not affected by the ATRIUM-IO channel replacement (re-channeling) performed during previous refueling outages. tage 5.0 Safety Limits 5.1 MCPR Safety Limit The Cycle 17 MCPR safety limits for two-loop and single-loop operation have been developed using GNF's NRC-approved methodology that accounts for the the Cycle 17 mixed core. These limits are least 99.9% of the rods established to ensure that at least acts in the core in the core are are expected expected to to avoi avoid boiling transition and are developed through a statistical convolution of of all all the uncertainties associated the uncertainties associated withwi the calculation of thermal margin.

The Cycle 17 MCPR safety limit analysis supports the existing MCPR safety limits of 1.08 for two-loop operation and 1.10 for single loop operation. As such, no changes to the existing MCPR safety limits in Technical Specification 2.1.1.2 are required for Cycle 17. The Cycle 17 safety limit analyses explicitly accounts for abnormal channel bow.

5.2 Pressure Safety Limit The worst-case pressurization event, MSN closure with failure of direct scram on valve position, was evaluated for GGNS Cycle 17 for compliance to ASME code requirements. The minimum number of SIRVs permitted by Technical Specification 3.4.4 (7 S/RVs in safety mode and an additional 6 SIRVs in relief mode) was assumed in the analysis with the worst-case selection of available S/RVs. This evaluation reported a peak vessel pressure of 1308 psig and a peak steam dome pressure of 1276 psig which is well below the 1325 steam dome pressure safety limit reported in Technical Specification 2.1.2.

A '1' ft .4 EN-LI-101-ATT-9.1.

all - Ll - 1 V I -tA I I I, Rev.

F-,UV . 4

n 6.0 Analyzed Events The events analyzed for each reload can be broken down into three categories based on their frequency of occurrence: moderate frequency events, infrequent incidents, and limiting faults. As outlined in FSAR Section 15.0.3.1, each event classification has its own set of acceptance criteria which are most stringent for those events with the highest frequency. Other events, such as the control rod drop accident and fuel handling accident, are considered limiting faults but have specific acceptance criteria reported in the Standard Review Plan (SRP) and the applicable Regulatory Guides. These events were analyzed considering the available GGNS operational modes of maximum extended operating domain, single loop operation, feedwater heaters out of service, and EOC-RPT inoperable. The events analyzed in support of the Cycle 17 reload are reported in Table 1 along with the applicable acceptance criteria.

6.1 Moderate Frequency Events In order to ensure moderate frequency events do not result in fuel failures, acceptance limits have been developed in order to preclude transition boiling or excessive cladding strain. Compliance with the MCPR safety limit ensures that rod failures are not expected to occur. As discussed in FSAR Section 4.2.1.2.15, the cladding plastic strain must remain below 1 percent ensuring no fuel rod failures due to excessive clad strain. To ensure that the MCPR safety limit and fuel cladding strain limit are satisfied during moderate frequency MCPR and LHGR operating limits have been developed based on the results of the transient analyses. These operating limits are reported as functions of core flow, power, and exposure in the COLR. To account for the higher SLO MCPR safety limit, the SLO MCPR operating limits must be increased by the difference between the SLO MCPR safety limit and the TLO MCPR safety limit. For MCPR performance, channel bow is explicitly considered in the MCPR safety limit and c'alculation of the MCPR operating limits in accordance with GNF's approved methodology.

As required by Technical Specification 5.6.5, these events were analyzed with the NRC-approved methodologies listed in Technical Specification 5.6.5. As listed in Table 1, the moderate freque"ncy events used to develop the GGNS Cycle 17 core operating limits are:

• Feedwater Controller Failure (Upscale) No Bypass,

• Load Reject No Bypass,

• ass, Turbine Trip No Bypass,

• Control Rod Withdrawal Error,

• Recirculation Flow Control Failure, and

• ater Heating.

Loss of Feedwater Heati GNF has reviewed all other moderate frequency events in GGNS FSAR Chapter 15 and confIrmed that listed above.

they are bounded by the events listed above.

Table 1: Cycle 17 Reload Event Analyses Event FSAR Acceptance Criteria Section Feedwater Controller Failure, 15.1.2 No fuel rod failures Maximum Demand, No Bypass Generator Load Rejection, No 15.2.2 No fuel rod failures Bypass urbine Trip, Turbine Tri No Bypass 15.2.3 No fuel rod failures of Feedwater Loss of Fee water Heating Heatine 1 15.1.1 15.1 .1 No fuel rod failures Control Rod Withdrawal Error 15.4.1, No fuel rod failures LN-U- 1 U I -

EN-Li-101-ATT-9.1, Rev. 4

15.4.2 Recirculation Flow Control 15.4.5 No fuel rod failures Failure with Increasing Flow Pressure Regulator Failure 15.2.1 Offsite doses :::;100/0 10CFRSO.67 Down Scale (closed) CR doses :S 10CFR50.67 Mislocated Bundle 15.4.7 Offsite doses ~10% IOCFR50.67 CR doses:s IOCFRSO.67 Misoriented Bundle 15.4.7 Offsite doses ~10% IOCFR50.67 CR doses S 10CFR50.67 Single Loop Operation Pump 15.3.3 Offsite doses $10% IOCFR50.67 Seizure Accident CR doses S 10CFR50.67 Fuel Handling Accident 15.7.4, Offsite doses $25% IOCFR50.67 15.7.6 CR doses :s 10CFR50.67 Control Rod Drop Accident 15.4.9 Offsite doses $25% 10CFR50.67 CR doses S 10CFR50.67 Deposited Enthalpy S 280 caVg Loss of Coolant Accident 15.6.5 Offsite doses :s 10CFR50.67 (LOCA) CR doses :Sl OCFR50.67 Peak Clad Temp. :s 2200 of Maximum Clad Oxidation :s 17%

Maximum H2 Generation S 1%

Coolable Geometry Maintained ASME Overpressurization 5.2.2.2 Pressure Safety Limit protected 6.2 Infrequent Incidents As listed in Table 1, the infrequent events analyzed in support of Cycle 17 are the pressure regulator failure downscale, misplaced (mislocated, misoriented) bundle accidents, and the single loop operation pump seizure accident. GNF applied the more conservative acceptance criterion for a moderate frequency event to the single loop pump seizure accident. The analysis detennined that the MCPR remains greater than the SLO MCPR safety limit for this event. The consequences of the remaining events have been trol room quantified in terms of fuel failures and the offsite and control roo doses have been evaluated with GGNS alternative source term models and methodologies. The conseqt consequences of these events were confirmed to result in offsite doses less than 10% of the IOCFR50.67 value (25 Rem value (25 TEDE) and control room doses Rem TEDE) remain less than the value in IOCFR50.67 (5 Rem TEDE As such, the consequences of the infrequent events remain bounded by the acceptance criteria reported in the FSAR.

6.3 Limiting Faults As listed in Table 1, the limiting faults analyzed in support of Cycle 17 are the fuel handling accident, the control rod drop accident, and the loss of coolant accident. The analyses of these events have been shown to be applicable to the revised core loading and to the abnormal channel bow condition.

Fuel Handling Accident The current Fuel Handling Accident calculation applied the alternative source term, to which GGNS is now licensed, and confirmed the 10CFR50.67 acceptance criteria would be satisfied fled for for the Cycle 17 EN-LI-101-ATT-9.1, Rev. 4

EVAL core. As such, there is no increase in consequences for this event. The existing light-load movement curves in 07-S-05-300 have been confirmed to be applicable for the Cycle 17 core.

Control Rod Drop Accident The control rod drop accident dose analysis has been revised for the alternative source term effort. The source term releases in this calculation have been confirmed to be applicable to the GE14 and ATRIUM-10 fuel types and the GGNS Cycle 17 core. As such, there is no increase in consequences for this event.

In addition, the peak deposited fuel enthalpy was determined to be less than 280 cal/gm.

Loss Of Coolant Accident An Appendix K LOCA evaluation was performed considering the GE14 and ATRIUM-IO fuel types using GNF's NRC-approved SAFERJGESTR-LOCA methodology. The bounding LOCA calculations resulted in a peak licensing basis peT of 1630°F for GEI4 and 1880°F for ATRIUM-10, which are well below the 10CFR50.46 acceptance criterion of 2200°F.

The GE14 and: ATRIUM-l 0 LOCA analyses also confirmed the 10CFR50.46 requirements that the core-,

wide metal-water reaction is less than I % and that the total local oxidation remains well below 17%.

Therefore, no changes in geometry are expected in the event of a LOCA. On the basis of this analysis, the the Cycle Cycle 1717 core core was found to satisfy the requirements of 10CFR50.46. The LOCA analysis is ble to the increase applicable to the increased bow condition associated with the high exposure ATRIUM-IO fuel and to the re-channeledled ATRfUM- bundles.

ATRIUM-I0 The LOCA dose analysis was previously revised for the alternative source term effort. The core source terms in this calculation have been confirmed to be applicable to the GE14 and ATRIUM-IO fuel types and the GGNS Cycle 17 core. As such, there is no increa is no increase in consequences for this event.

6.4 Other Analyses The following analyses have been shown to be applicable to the increased d cchannel bow condition considered for the high exposure ATRIUM-IO fuel.

Anticipated Transient Without Scram (ATWS)

The limiting ATWS events were analyzed for vessel integrity and containment response considering the GE14 reload fuel. The limiting vessel pressurization event was detennined to be the Pressure Regulator Failure Open (PRFO). The Loss of Offsite Power (LOOP) is analyzed as potentially limiting for containment response due to the conservative assumption of a reduced number of residual heat removal trains available. The ATWS peak pressure analysis resulted in a vessel pressure of 1299 psig, which is well below the ASME emergency pressure acceptance limit (1500 psig). The containment containment analysis resulted in a peak suppression pool temperature of I70<>P and containment pressure of 8.2 psig, which are well below the 1850P and 15 psig containment design limits. Therefore, GGNS can load and operate with GEI4 fuel and continue to meet the ATWS acceptance criteria.

Appendix REvaluation The Appendix R safe shutdown scenario in FSAR Appendix 9C was re-analyzed by GNF for a full core of GE14. This analysis demonstrates that, consistent with IOCFR50 Appendix R, cladding failure would not occur in the event of a major fire. Acceptable performance (peak clad temperature response) of the ATRIUM-IO fuel has also been demonstrated by AREVA considering the FSAR safe shutdown scenario.

EN-Li-EN-ll-101-ATT-9.1, Rev. 4

Station Blackout The core parameters assumed in the GGNS station blackout ut re response reported in FSAR Appendix 8A remain applicable to the Cycle 17 mixed xed care core, including the reactor coolant and condensate inventory assumptions.

The Cycle 17 core design has been reviewed for any changes to parameters that may impact the GGNS Emergency Operating Procedures. This review concluded that two parameters, the mass of clad and channels (Mclad) and mass of fuel (MfueI), increased slightly due to the GE14 bundles loaded for Cycle

17. However, the impact on the existing EOP's is extremely small such that changes in curves are not discemable and no EOP revision is required. The minimum covered active fuel length to maintain peT <

I 800°F without injection (FaIf-I8) increased due to the GE14 bundles. This change resulted in an ~2 inch increase in the Minimum Zero Injection Reactor Water Level (MZIRWL). This change has been implemented in revisions to the affected plant procedures.

Hydrogen Analysis The reload fuel can impact the design basis and degraded core hydrogen analyses with the changes in the cladding surface area or volumes. The current design basis hydrogen analysis assumes a full core of ATRIUM-I 0 fuel with an additional 20% margin. The Cycle 17 core is composed of31%

GEI4 and 69% ATRIUM-IO fuel. GE14 has slightly more zirconium available for metal-water reaction than ATRIUM-IO. A comparative cladding analysis performed by GEH has shown that the post-accident hydrogen production for a full GE14 core is <0.2% more than a full ATRIUM-I0 core.

Since the Cycle 17 core composition is only 31% GE14 fuel, and the design basis hydrogen analysis assumes a full core of ATRIUM-IO plus 20% margin, the current analysis remains bounding for Cycle 17. Similarly, an evaluation of the degraded core hydrogen analysis concluded that both a full core of ATRIUM-I 0 and a full core of GEl4 are bOWlded by the current analysis. As such, the grade core hydrogen analyses are applicable to Cycle 17 mixed core.

design basis and degraded Containment Analysis The GGNS drywell and containment analyses are dependent on the amount of stored energy in the core as well as the extent of zirc-water reaction. The Appendix K LOCA analysis determined the core-wide metal-water reaction to be <0.1 % compared to the 5% of the active Zircaloy that ABD-4 Section 5.4.2 reports was applied in the long-term containment analysis. No changes have been made to impact the initial coolant enthalpy since the power-flow map is Wlchanged for Cycle 17.

The stored energy of the GEI4 reload fuel and ATRIUM-I0 fuel designs have been evaluated and shown to be bounded by the stored energy of the GE 8x8 design on which the containment analysis is based. As such, the existing GGNS drywell and containment analysis remains applicable to GGNS Cycle 17.

Vessel Pressure/Temperature (PIT) Curves The core design can impact the neutron fluence assumptions made in the calculation of the P/T limit curves. Consistent with previous cycles, the Cycle 17 core is a conventional low leakage design with the lowest reactivity bundles placed on the peripheral region of the core. As such, the Cycle 17 core will not result in increased neutron fluence rates on the vessel walls or impact the assumptions in the prr limit curve calculations.

calculation The PIT limit curves are exposure dependent. Projections through EOC 17 plus 30 days (i.e., long

Cycle 17 operation) indicate that the vessel service life will exceed the range of applicability of the Pff limit curve currently in use (16 EFPY to 20 EFPY) approximately 80 days after the beginning of Cycle 17. As such, the TRM will need to be updated during Cycle 17 to reflect the change in the applicable Pff limit curves. LeTS # A-35826 has been initiated to insure the correct Pff curve is implemented within 60 days of Cycle 17 startup.

Core Monitoring System The 3D Monicore (3DM) system version 6.58 has been installed at Grand Gulf for use in cycle 17

~nd beyond. 3DM replaces the Powerplex system previously used for monitoring AREVA fuel.

3DM is built around the PANACEA code which is a coupled nuclear thermal-hydraulic diffusion theory model providing 3 dimensional simulation of BWR performance. The basic accuracy and approval of PANACEA has been confirmed in NEDE-32695P-A. The current version of 3DM and PANACEA is in service at numerous BWR plants in the US and overseas.

The 3DM code package was tested by GNF at the factory and at Grand Gulf with a Site Acceptance Test (SAT). The results of both GNF and Entergy testing ensure adequate capability exists to monitor the cycle 17 Grand Gulf core consisting of GE-14 new fuel and AREVA Atrium-lO fuel (mixed core).

Live data input, TIP transfer and processing, and actual case data were reviewed and analyzed in the Site Acceptance Test (SAT) including operation of the a special module which performs AREVA fuel preconditioning calculations on the AREVA fuel only. The results showed acceptable agreement with actual plant data and the Powerplex system which was in use at the time in cycle 16.

The testing was performed in accordance with Entergy station procedures to ensure verifiable, reliable results were obtained.

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-Entergy NUCLEAR MANAGEMENT (NON)-QUALITY RELATED EN*LI*l10 I REV. 1 MANUAL INFORMATIONAL USE PAGE 10F5 CMSID:

P-32889 P-32889 Plant Licensing Plant LicenSIng Trac Tracking Number: ceE 2008-0003 Source CNRO-97/00004, dated 3/17/97, IS the Entergy 180 Day Entergy 180 Day Response Response to to NRC NRC Gener Generic Document/Date: odic Verification Letter 96-05, "Periodic Verification of Design-Basis Capability of Design-Basis Capability of of Safety Safety-Related Motor-Operated VaJves" CommItment: Deletion 0 ReviSion [83 (one-time extension)

Haslithe original commitment been Implemented? 181 YES YES "

D NO, Notify Plant NO, Notify Plant sin LicenSing Original Commitment

Description:

Commitment +ant P-32889 P-32889 is IS addressed addressed inin th the original GL 96-05 response per CNRO-96/00016, dated 11/15196, and restated verbatim In the 180-day response per CNRO-97/00004. ThiS commitment tracks the tmplementation of general program gUidance and methodology at each EOI site In site specifiC programs. The 180-day response provides a summary of the "EOI Safety-Related Motor Operated Valve Periodic Verification Program" and Includes the following:

"The following matrIx provides the resutts of the comb,ned risk ranking and setup ratio used to determine a maximum bounding test frequency. tn addition, the maximum Interval between static tests is not to exceed 10 years.

STATIC IC TEST TEST FREOUENCY FREQUENCY MAT MATRIX H U3_

H (3) 3 -Cycles 3 Cycles 3 Cyctes 3 C cles g..gycles 2 Cycles 2C 2 cles Cycles M (2)

M 2 6 Cycles 6 gycles 4 C cfes 4 Cycles 4 Cycles 4 Cycles 3 Cycles 3 Cycles L 1 L (1) 6 Cycles 6 Cycles 6 Cycles 6 Cycles 6 Cycles 6 Cycles 4 Cycles 4 Cycles t

RISK T I/ RATIO RISK RATIO ~ -+ H-H (1)

H-H (1) H (2)

H (2) M(3)

M(3) IL (4)

L (4)

Revised Commitment

Description:

Make a one-cycle one-time extensIOn to the 6-cycle Static Test Frequency interval for 1E12F290B (RHR (RHR Jockey Pump B discharge block valve), that wdl also correspond to a one-time 14 month extension to to the maximum interval between static tests.

Summary of Justification for Change or Deletion:

1E12F290B was scheduled for static testing on 717/08. Working this valve requires unavatlability time to be taken on the AHA 8 system. To reduce plant risk and to preserve limited remaining unavailability for RHR B system, management requested that we defer the late date for performing the diagnostic test on valve 1E12F290B until 8/30/2009. To approve thiS deferral requires a one-time one-cycle test-interval-extension for thIS valve to be made. This.8 acceptable because th,s valve has a hlgh*high safety margin (1542%) and the risk ranking for this valve is low (reference Engineering Report NO. GGNS-94-0025, Rev. 0). The valve has a PM task that monitors the condition of the assembly every 36 months and no abnormalities have been identified. No failures of this MOV has been observed over the past 10 years and the quarterly stroke time history for the valve does not show any abnormalities.

NUCLEAR NUCLEAR (NON)-QUAD (NON)-QVALITY RELATED RELATED EN-LI-110 EN-LI-II0

, REV.

REV. 11

- Ent nteW et8Y MANAGEMENT MANAGEMENT MANUAL MANUAL INFORM'TIONAL INFORMATIONALUse USE PAGE PAGE 2 of 5 20F5 r ~ _ _ 1 P~~~dB~_D_~_~_I_G~._H_e_rr_~~~~~~.~*.~~~)~~D~:

Print Name/Signature Management Approval: WIlliam H. Parman I fk "'&/1?<bIl b4117*,--t Date: 7-10 "01 Pnnt Name/Signature Plant LicenSing Management Mike Larson I ~

Concurrence:

Prtnt Name/Signature

-Entergy NUCLEAR MANAGEMENT MANAGEMENT (NON)-QVAUTY RELATEP tU r

EN-LI-IIO I REV. 1 MANUAL MANUAL INFORMATIONAL INFORMATIONAL USE PAGE PAGE 30F5 3 OF 5 PART I 1.1 Is the existing commitment located in the Updated Final Safety Analysis Report, Emergency Plan.

Quality Assurance Program, Fire Protection Program. or Security Plan?

o YES STOP. Do not proceed with this evaluation. Instead use appropriate codified process (e g., 10 CFR 50.71 (e), 10 CFR 50.54,10 CFR SO.59. 10 CFR 50.55) to evaluate commitment.

(:gJ NO Go to Part II.

PART III PART 2.1 Could the change negatively impact the ability of a System, Structure, or Component (SSe) to perform its safety function or negatively impact the ability of plant personnel to ensure the sse is capable of performing its intended safety function?

o YES Go to Question 2.2.

(:gJ NO Continue with Part III. Briefly describe rationale:

With the high-high safety margin and low PRA ranking for thiS valve, a one-cycle delay In static testing for thiS valve will have no SSC personnet effect, nor will it negatively affect the ability of SSCs to perform their safety functions.

2.2 Perform a safety evaluation using the following 10 CFR 50.92 criteria to determine if a significant hazards consideration exists:

• Does the revised commitment involve a significant increase in the probability or consequences of an aCCident previously evaluated?

DYes 0 No Describe basis below:

- Does the revised commitment create the poSSibility of a new or different kind of accident from uated?

any previously evaluated?

DYES 0 NO Desc Describe basis below:

Does the revised commitment involve a significant reduction in a margIn of safety?

DYES 0 NO Describe basis below:

" any of the above questIOns are answered Yes, STOP Do not proceed With the reVISIOn, OR dISCUSS change with NRC and 00ta1l'l necessary approvals poor to mplementatton of the proposed change If all three questtons are answered NO. go to Part III. (Attach additional sheets as necessary )

.Entergy

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NU CLEAR NUCLEAR MANAGEMENT MANAGEMENT (NON)-QUALITY (NON)-QUALITY RELA TED RELATED EN-LI-110 EN-LI-II0 I REV.. 1I REV MANUAL MANUAL INFORmAinONAL ORmAin INroRMA ONALU DONAL SE USE PAGE PAGE 4 OF 5 40F5 PART III 3.1 ponse co Was the original commitment (e.g. t response to NUV, etc .) to NOV, etc.) to restore restore an obligation (e.g., rule.

regulation, order or license condition)?

o YES Go to question 3.2.

t8l NO Go to Part IV 3.2 Is the proposed revised commitment date necessary and justified?

o YES Briefly describe rationale (attach additional sheets as necessary) and notify NRC of revised commitment date prior to the onginal commitment date.

DNO STOP. Do not proceed With the revision,i OR apply for appropriate regulatory relief.

PART IV 4.1 Was the original commitment: (1) explicitly credited as the basis for a safety decision in an NRC SER, (2) made In response to an NRC Bulletin or Generic Letter, or (3) made in response to a request for information under 10 CFR 50.54(f) or 10 CFR 2.204?

IZI YES Go to Question 4.2.

ONO Gota PartV.

4.2 nal comm Has the original commitment been implemented?

IZJ YES YES STOP. You STOP, You have have completed this evaluation. Revise the commitment and notify NRC of revised commitment in summ.ary report.

ONO Go to Question 5.1

Ent&V I I

NUCLEAR NUCLEAR (NON)-QUALITY (NON)-QuALITY RELATED RELATED EN-LI-110 EN-LI*IIO REV.

REV. 11

.Entetgy MANAGEMENT MANAGEMENT MANUAL INFORMATIONAL INFORMATIONALUSE PAGE 550F5OF 5 MANUAL USE PAGE PARTTV "T

5.1 Was the original commitment made to minimize It recurrence of a condition adverse to quality (e.g.,

a long-term corrective action stated in in ann LER)

LER)?

F]

DYESYES Go Go to to Quesbon Question 5.2.

5.2.

ONO STOP. You have completed this evaluation. Revise the commitment. No NRC notification required.

5.2 Is the revised commitment necessary to minimize recurrence of the condition adverse to quality?

o YES Revise the commitment and notify NRC of revised commitment in next annuaVRFO interval summary report.

o NO Revise commitment. No NRC notification is required.

REFERENCES List below the documents (e.g., procedures, NRC submittals, etc.) affected by this change.

Doc. NumberllD Description CNRQ-97/00004, Entergy 180 Day Response to NRC Generic Letter 96-05 dated 3/17/97

GGNS Commitment Change Evaluation Number CCE 2009-0001 G090036

NUCLEAR NUCLEAR (NO?+QUMATYRELATED (NON)-QUALITV RKIATM EN-LI-11,0 EN-LI-IIO REV. I REV.t Entergy Ljqt"eqy MANAGEMENT MANAGEMENT MANUAL IN FORMA"ONAL USE USE MANUAL INFORMATIONAL PAGE 21oF31 PAGE 210F 31 ATTACHMENT 9.4 COMMITMENT CHANGE EVALUATION FORM NOTE Forward completed form to Plant Licensing.

eMS 10: _P_-_2_5_05_7 ......_ sing Tracking Plant Licensing T Number: ceE 09-0001 Source GNRO-94J00059 J 4-19-1994 Document/Date:

Commitment Deletion t83 Revision oEl Has the original commitment been implemented? t83 YES o NO, Notify Plant Plant Licensing Original Commitment

Description:

The procedures governing the control of work on plant equipment and housekeeping will be revised to alert plant personnel to the potential hazards which exist when working inside containment with material that could potentially clog the ECCS suction strainers.

Revised Commitment

Description:

N/A Summary ary o stification fo of Justification for Change or Deletion:

Commitment No. P-25057, based on information contained in GNRO-94J00059, in response to NRC bulletin (NRCB) 93-02, Supplement 1, was implemented in GGNS Procedure 01-S-07-9 (Revision 19) issued January 11 1995. The actions to be performed for NRCB 93-02 were interim actions due to the J

lack of a final resolution by the BWR owners group. Several years later NRCB 96-03 was issued based on the events in NRCB 93-02.

With the issuance of NRCB 96-03 and RegUlatory Guide 1.82 Revision 2, "Water Sources for Long-Term Recirculation Cooling Following a Loss of Coolant Accidenf, as well as the final resolution by the BWR owners group, BWR owners had to implement one of three options to be in compliance.

Grand Gulf Nuclear Station elected to implement "Installation of a large capacity passive strainer design" in the Suppression Pool. The installation of the large passive strainer was completed in May 1998 to satisfy actions reqUired by NRCB 96-03. Commitment A-32649 was closed on May 28, 2002; this was the commitment that implemented installation of the large passive strainer.

F NUCLEAR NUCLEAR (NON)-QUALITY RELATED (NON)-QUALITY RELATED EN-LI-110 EN-LI-JIO REV. 1I REV.

-W, tfff9y

-::::::- En ergy MANAGEMENT MANAGEMENT MANUAL MANUAL INFORMATIONAL USE INFORMATIONAL USE PAGE PAGE 220F31 22oF3l ATTACHMENT 9.4 COMMITMENT CHANGE EVALUATION FORM The intent of GGNS Commitment No. P-25057 is also addressed in Entergy fleet Procedu~e EN-MA-118, Jl "Foreign Material Exclusion EN-MA-118 states, "The BWR 6 Suppression Pool is an FME high risk area". The definition of a high risk area in the FME procedure is: "Intrusive work or work adjacent to open components involving safety systems that provide a direct path to the reactor vessel or spent fuel pool."

Additionally, the Suppression pool has been cleaned of foreign material and accumulated debris during outages in 1998,2002,2004 and 2007. Periodic cleaning of the Suppression Pool is an integral part of an ongoing maintenance action intended to eliminate or minimize accumulation of debris in the pool. Based on initial actions taken pursuant to the Commitment No. P-25057 and the final design configuration achieved by installation of the passive strainer, the objectives of this GGNS Commitment have been met.

It is thus concluded that Commitment No. P-25057 is no longer warranted due to the current Suppression Pool design configuration and the closure of commitment A-32649.

Prepared By: Jason Keir I Date:

Pr" Management Approval:

---

~~ ............~....................-- Date:

Plant Licensi Licensing Management Concurrence:

Concurrence:

NUCLEAR NUCLEAR (NON)-QU (NON)-QUALITY W RELATED TED EN-LI- 0 EN-LI-lI0 REV. I1 REV.

Entergyy MANAGEMENT MANAGEMENT MANUAL INFORMATIONAL USE INFORMATIONAL PAGE 230F31 ATTACHMENT 9.4 COMMITMENT ~r1jlooll * * "*'Itr:" EVALUATION FORM {TYPICAL}

Sheet 3 of5 PART I 1.1 Is the existing commitment located in the Updated Final Safety Analysis Report, Emergency Plan, Quality Assurance Program, Fire Protection Program, or Security Plan?

o YES STOP. Do not proceed with this evaluation. Instead use appropriate codified process (e.g., 10 CFR 50.71 (e), 10 CFR 50.54,10 CFR 50.59,10 CFR 50.55) to 0

evaluate commitment.

~ NO NO Go to Part II.

Go PART II 2.1 Could the change negatively impact the ability of a System, Structure, or Component (SSC) to perform its safety function or negatively impact the ability of plant personnel to ensure the sse is capable of performing its intended safety function?

o YES Go to Question 2.2.

~ NO Continue with Part III. Briefly describe rationale:

No this point is answered in Safety evaluation 97-0016-ROO and ER-1997-0089-000.

2.2 Perform a safety eva.luation using the following 10 CFR 50.92 criteria to determine if a significant nsideration exists hazards consideration exists::

- Does the revised ised commitment commitment involve a significant increase in the probability or consequences of an an ac accident previously previously evaluated?

evaluated?

DYes 0 No Cue Describe basis below:

- Does the revised commitment create the possibility of a new or different kind of accident from any previously evaluated?

DYES 0 NO Describe basis below:

Does the revised commitmentt involve involve aa signific significant reduction in a margin of safety?

DYES 0 NO Describe Describe basis basis below below:

IfIf any any of of the the above above questions questions are are answered Yes, STOP.

answered Yes, STOP. DoDo not proceed with not proceed with the the revision revision, OR discuss change with NRC and obtain necessary approvals prior necessary approvals prior to to implementation implementation of of the proposed change.

the proposed change. If If all all three three questions questions are are answered answered NO, go to Part Itt (Attach additional additional sheets as necessa sheets as necessary.)

f& NUCLEAR (NON)-QUALITY RELATED EN~LI- 1 EN-LI-IIO REV. 1 Entergy MANAGEMENT MANAGEMENT MANUAL MANUAL INIFORMA"ONAL INFORMATIONAL USE USE PAGE 24 PAGE OF 31 24oF31 ATTACHMENT 9.4 COMMITMENT CHANGE EVALUATION FORM {TYPICAL}

Sheet 4 of 5 PART III 3.1 Was the original commitment (e.g., response to NOV, etc.) to restore an obligation (e.g., rule, regulation, order or license condition)?

o YES Go to question 3.2.

r8J NO Go to Part IV.

3.2 Is the proposed revised commitment date necessary and justified?

o YES Briefly describe rationale (attach additional sheets as necessary) and notify NRC of revised commitment date prior to the original commitment date.

DNO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relief.

PART IV 4.1 Was the original commitment: (1) explicitly credited as the basis for a safety decision in an NRC SER, (2) made in response to an NRC Bulletin or Generic Letter, or (3) made in response to a request for information under 10 CFR 50.54(f) or 10 CFR 2.204?

YES Go to Question 4.2.

DNO Go to PartY.

4.2 Has the original commitment been implemented?

r8J YES STOP, You have completed this evaluation. Revise the commitment and notify NRC of revised commitment in summary report.

DNO Go to Question 5.1.

NUCLEAR NUCLEAR (NON)-QUALffY RKLAW (NON)-QUALITY RELATED.D EN-LI-110 EN-LI-110 REV. I REV.l "tergy Entergy MANAGEMENT MANAGEMENT INFOP-MIATIONAL USE MANUAL MANUAL INFORMATIONAL USE PAGE 25 OF 31 PAGE 250F31 ATIACHMENT9.4 COMMITMENT CHANGE EVALUATION FORM {TYPICAL}

Sheet 5 of 5 PART V 5.1 Was the original commitment made to minimize recurrence of a condition adverse to quality (e.g.,

a long-term corrective action stated in an LER)?

DYESYES Go to Question 5.2.

DNO STOP. You have completed this evaluation. Revise the commitment. No NRC notification required.

5.2 Is Is the revised commitment the revised commitment nice necessary to minimize recurrence of the condition adverse to quality?

o YES Revise the commitment and notify NRC of revised commitment in next annuallRFO interval summary report.

o NO Revise commitment.. No NRC notification is required.

REFERENCES List below the documents (e.g., procedures, NRC submittals, etc.) affected by this change.

Doc. Number/IDr/ Description 09 01-8-07-009 This procedure 1 This procedure will will be be cancelled cancelled Commitment This document will be closed P-25057

9£00608 8000-600~ 3~~

JaqwnN uOllenleA3 a6uell~ luaWI!WWO~ SN~~

Atk, NUCLEAR NUCLEAR (NoNyQuALITY RELATED (NON)-QUALITY ftLATED EN-LI-110 EN-LI-IIO REV. 1-REV.~'

Entergy

-=7-MANAGEMENTNFORMATIONAL MANAGEMENT MANUAL MANUAL INFORMATIONAL USIE USE PAGE PAGE I OF 5 10F5 "eMS 10:

Source P-32889 Plant Licensing Tracking Number:

CNRO-97/00004, dated 3/17/97, is the Entergy 180 Day Response to NRC Generic ceE 09-0008 Document/Date: Letter 96-05, "Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves" Commitment: Deletion 0 Revision to

[8J (one-time extension)

Has the original commitment been implemented? [8J YES o NO, Notify Plant Licensing Original Commitment

Description:

Commitment P-32889 is addressed in the original rigin GL 96-05 response per CNRO-96/00016, dated 11/15/96, and restated verbatim in the 180-day response response per per CNRO-97/00004. This commitment tracks the implementation of general program guidance and nd methodology methodology at at eeach EOI site in site specific programs. The 180-day response provides a summary summary of of the"EOI Safety-Related Motor Operated Valve Periodic Verification Program" and includes the followi fonowing:

"The following matrix provides the results of the combined risk ranking and setup ratio used to determine a maximum bounding test frequency. In addition, the maximum interval between static tests is not to exceed 10 years.

STATIC TEST FREQUENCY MATRIX

--H (3) 3 3C Cycles eles 3 3 Cycles C cles 2 2 Cycles C etas 2 Cycles M (2) 6 Cycles 6 C etes 4 Cycles 4 C eles 4 Cycles 4 C etes 3 Cycles L (1) 6 Cycles 6 C etas 6 Cycles 6 C ctes 6 Cycles 6 C eles 4 les RISK RATIO H-H (1)

H-H (1) H (2)

H (2) M (3)

M (3) L (4)

Revised Commitment

Description:

Make a one-cycle one-time extension to the 6-cycle ycie (9 (9 years) static T years) Static Test Frequency interval for 1P72F123 (Drywell Chillers Return Header Drywell well Isolation Isolation Valve), that will also correspond to a one-time 6 month extension to the maximum interval al (10

( years) between static tests. The valve was last tested on 121211999 and to meet the commitment, the valve must be tested by 12/212009. The valve will be tested during RF17 (May 2010) so GGNS is requesting a 6 month extension beyond the 10 year commitment.

Summary of Just of Justification for Change or Deletion:

1P72F123 was was scheduled scheduled for for static diagnostic testing during CYCLE 17. To diagnostically test 1 P72FI test 1P72F123 ssary to stroke the valve CLOSED and OPEN. The normal configuration of the valve it is necessary is iin the ode is OPEN position position to support full to support full power power opera operation. Stroking the valve closed will result in a loss of d of drywell chilled water and this water and is an this is an unacceptable unacceptable level of risk at full power operation and potentially is a pre-cursor to a plant plant transient transient event.

event. To To reduce reduce plant plant risk, the site (GGNS) is requestingrequesting a defers of the late date for a deferral performingng the the diagnostic diagnostic testtest on on valve 1, valve 1P72F123 (which is until 6/0212010 (which after Refu is after Refuel 17) to odate diagnostic accommodate diagnostic testing testing thethe vaIv valve in Refuel 17. To approve ve this this deferral deferral requi reqUires a one-time one-cycle cycle test-interval-extension test-interval-extension for for this valve to be made. This is acceptable this valve  :-e-ptable, because tai valve has a high because this safety margin (113.2%)

safety margin (113.20/0) and and thethe ririsk ranking for this is valve valve is is Ilow (reference reference Engineerir Engineering Report NO.

-0025. Rev.

GGNS-94-o025, Rev. 0) 0).. The The valve has a valve has a PM PM task that monitors the that monitors condition of the condition of the a the assembly every 36 months and no abnormalities have been identified. ntifi. No failures failures of this MOV has been observed over the past 10 years and the the cold cold shutdown surveillance stroke shutdown surveillance e time time historv history for for the th~} va valve does not show any abnormalities. "

NUCLEAR NUCLEAR (NUNYQUALITY (NON)-QUALITY RELATEDRELATED EN-LI-110 EN-LI-IIO REV, REV. II E ,pr;jy MANACEMENT MANAGEMENT MANUAL MANUAL INFORMATIONAL INFORMATIONAL ASE USE PACE PAGE 2 OF 5 20F5 Prepared By: Daniel G. Herro Date:

Print Name/Signature Wi /1 Management Approval: Matt Rohrer / IVl~ Date:

Print Name/Signature Plant Licensing Management Christina Perino I Concurrence:

---

..........--.....................-....-_....._- Date:

NUCLEAR NUCLEAR (NoN)-QuALffy RELATED (NON)-QUALITY ftLATED EN-LI-110 EN-LI-IIO REV.

REV.lI MANAGEMENT MANAGEMENT MANUAL MANUAL INFORMATIONAL USIE INFORMATIONAL USE PAGE PAGE 3 OF 5 30F5 PART I 1.1 Is the existing commitment located in the Updated Final Safety Analysis Report, Emergency Plan, Quality Assurance Program, Fire Protection Program, or Security Plan?

o YES STOP. Do not proceed with this evaluation. Instead use appropriate codified process (e.g., 10 CFR 50.71(e), 10 CFR 50.54,10 CFR 50.59,10 CFR 50.55) to evaluate commitment.

~ NO Go to Part II.

PART II

• 2.1 Could the change negatively impact the ability of a System, Structure, or Component (SSC) to perform its safety function or negatively impact the ability of plant personnel to ensure the sse is capable of performing its intended safety function?

~ YES Go to Question 2.2.

Continue with Part III. Briefly describe rationale:

2.2 Perform a safety evaluation using the following 10 CFR 50.92 criteria to determine if a significant hazardsrds consideration consideration exists:

- Does Does the used com the revised commitment involve a significant increase in the probability bilitv or conseauences of or consequences of an accident previously evaluated?

o Yes ~ No Describe basis below:

Per Per Engi Engineering Report GGNS-94-Q025 the 1P72F123 is normally OPEN during full power operation to ide flow to the drywell coolers. The Drywell Chilled Water system is not required for safe shutdown provide or GGNS Level 1111 event tree success. The e, con containment isolation that this valve supports is backed up by MOVs 1P72F122, 1P72F125 and 1P72F126. W126. These These th steel on three valves were last tested o the following dates and have the following safety margin:

1P72F122 - tested on 10/11/05 and has 75.2% safety margin 1 P72F1 25 -tested 1P72F125 - tested on 4/2 on 4/23/01 and has 82.1% safety margin 1019/08 and 1P72F126 - tested on 10/9/08 a had 27.6% safety margin The 1P72F123 The 1 P72F1 valve has a PM task that monitors the condition of the assembly every 36 months and no abnormalities have beenbeen identified. No failures of this MOV have been observed over the past 10 years and the cold sh tdown survei shutdown surveillance stroke time history for the valve does not show any ny abnormalities.

Therefore Therefore based o the discussion above, a one time extension of the committed testing frequency based on would would not cause a not cause revious y a significant increase in the probability or consequence of an accident previously evaluated evaluated.

A-4,N-2 (NON),--QUALrff (NON)-QUALITY RELATED NUCLEAR NUCLEAR RELATED EN-LI-110 EN-LI-IIO REV REV.l . 1.

'MANAGEMENT MANAGEMENT MANUAL INFORMATIONAL USE INFORMATIONAL MANUAL USE PAGE PAGE 440F5 oF 5

- Does the revised vised commitment ment create create the possibil of a new or different kind of accident from the possibility any previously sly evaluated evaluated?

n YES I5<J NO Describe basis below: below:

Allowing the 1P72F123 valve valve to to be be ddiagnostically tested at a one time frequency of 10 years and six months will not create the possibility ibility of of aa new or different kind of accident from any previously eval evaluated sincesince the the diagnostic diagnostic testin testing, performed on a 9 year frequency, primarily monitors for valve and actuatorr degradation and test the the actuator capability to produce enough fhrust to close the valve againstt design design basis accident pressures. sores. Failure Failure of of this valve to stroke close will not create any new ident scenario accident scenario since since the the Drywell filled Chilled Water system is not required for safe shutdown or GGNS Levell/II event tree success.

Does the revised commitment involve a significant reduction in a margin of safety?

DYES [8j NO Describe basis below:

With the high safety safety margin margin and and low low PRA ranking for this valve, a one-cycle delay in static testing for testing for this valve will11 notnot involve involve a a significant significan reduction in a margin of safety nor will it negatively affect iffectthe the ability of SSCs to perform their it safety safety functions.

If any any ofof the the above above questions questions are are answered Yes, STOP answered Yes, STOP.. Do Do not proceed with not proceed with the revision, OR the revision, discuss change OR discuss change with NRC and with NRC and obtain obtai necessary necessary approvals prior to approvals prior to implementation implementation of of the the proposed proposed change change.. If all If three questions all three are answered questions are answered NO, go to NO, go Part III.

to Part Ill. (Attach (Attach additional additional sheets sheets as necessary .)

as necessary.)

PART III 3.1 Was the original original commitment commitment (e.g., (e.g., response response to NOV, etc.)

to NOV, to restore etc.) to an obligation restore an obligation (e.g., rule, regula regulation,on, order order or or license license condition) condition)?

o El YYES Go to questionestion 3.2.

[8j NO Go to Part IV.

3.2 Is the proposed sed revised rev commitment date necessary and justified?

D YES Briefly describe Briefly describe rationale rationale (attach (attach additional additional sheets sheets as necessary) and as necessary) and notify notify NRC NRC ofof iced commitment revised commitment date date prior t the original commitment date.

prior to DNO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relief.

PART IV 4.1 Was the original commitment mitme (1) explicitly credited as the basis for a safety decision in an NRC SER, (2) made jade in in response response to to an NRC Bulletin or Generic Letter, or (3) made in response to a request for information ation u under 10 CFR 50.54(f) or 10 CFR 2.204?

~YES tea Go to to Question 4.2.

DNO Go to Part V.

F fML

-0 , NUCLEAR NUCLEAR (NON)-QU AMY WIA (NON)-QUALITY TED RELATED EN-LI-1,10 EN-LI-IIO REV.

REV.l1 Entergy E- /'I, tergy MANAGEMENT MANAGEMENT MANUAL MANUAL INFORMAMNAL INFORMATIONAL USE USE PAGE OF 5 PAGE 550F5 4.2 Has the original commitment been implemented?

rgJ YES YES STOP, You have completed this evaluation. Revise the commitment and notify NRC Of revised comm of revised commitment in summary report.

DNO Go to Question 5.1.

PART V 5.1 Was the original commitment made to minimize recurrence of a condition adverse to quality (e.g.,

a long-term corrective action stated in an LER)?

El DYESYES Go to Question 5.2.

DNO STOP. You have completed this evaluation. Revise the commitment. No NRC notifica notification required.

6.2 Is the revised commitment necessary to minimize recurrence of the condition adverse to quality?

o YES Revise the commitment and notify NRC of revised commitment in next annuallRFO interval summary report.

o NO Revise commitment. No NRC notification is required.

REFERENCES List below the documents (e.g., procedures, NRC submittals, etc.) affected by this change.

Doc. Number/ID Description CNRO-97/00004, Entergy 180 Day Response to NRC Generic Letter 96-05 dated 3/17/97