Inc. Entergy Waterloo Road Port Gibson, MS 39150 · 5.S3 reviewed, approved and Issued drawings and assignment of drawing coordinator to coordinate design drawings and resolve problems

" EntergyEntergy Operations, Inc.Waterloo RoadPO. Box 756Port Gibson, MS 39150Tel 601 437 6299

Charles A. BottemillerManagerPlant Licensing

GNRO-2004/00029

April 15, 2004

U.S. Nuclear Regulatory CommissionAttention: Document Control DeskWashington, DC 20555

Subject: Report of 1OCFR50.59 Safety Evaluations and CommitmentChanges - November 01, 2002 through March 31, 2004Grand Gulf Nuclear StationDocket No. 50-416License No. NPF-29

Ladies and Gentlemen:

Pursuant to 1OCFR50.59(d)(2), Entergy Operations, Inc. hereby submits thesummary of 10CFR50.59 evaluations for the November 01, 2002 through March31, 2004 period. Also attached is the summary of commitment changes for thesame period made in accordance with NEI 95-07 Guidelines.

If you have any questions or require additional information, please contact ChuckHolifield at 601-437-6439.

This letter contains no commitments.

Yours truly,

Pel

CAB/CDH;cdh

attachments:

cc:

1. Table of Contents2. 1 OCFR50.59 Evaluations and Commitment Change

Evaluations(See Next Page)

I eq 17

April 15, 2004GNRO-2004/00029Page 2 of 3

cc: HoegLevanwayReynoldsSmithThomas

T. L. (GGNS Senior Resident)D.E (Wise Carter)N. S.L. J. (Wise Carter)H. L.

(wla)(wla)(wla)(wla)(w/o)

U.S. Nuclear Regulatory CommissionATTN: Mr. Bruce S. Malleft (w/2)Regional Administrator, Region IV611 Ryan Plaza Drive, Suite 400Arlington, TX 76011-4005

U.S. Nuclear Regulatory CommissionATTN: Mr. Bhalchandra Vaidya, NRRIDLPM (w12)ATTN: FOR ADDRESSEE ONLYATTN: U.S. Postal Delivery Address OnlyMail Stop OWFN/7D-1Washington, D. C. 20555-0001

.

I Roland Wood - page3of3.pdf 0 Page_ 1 ]

April 15, 2004GNRO-2004/00029Page 3 or 3

bcc:OUTLOOK MAIL: DISTRIBUTION IS ALL ELECTRONIC

James D. E. (ANO-NSAPL)Miller 0. J. R. A. (GG-TRNG)Bottemiller C. A. (GG-NSAPL)Mitchell T. G. (ANO-NSA)Williams G. A. (GG-VP)Edwards J. B. (GG-GMPO)GGN CENTRAL FILE (175)GGN PLANT LICENSINGSen G. (W3-NSAPL)King R. J. (RR-NSA)Burford F. G. (ECH-NSL)Lorfing D. N. (RB-NSAPL)Krupa M. A. (GG-NSA)Wiles D. P. (GG-ENG)

File (LRSDOCS Directory - GNRI or GNRO)OTHER:

(

ATTACHMENT ITABLE OF CONTENTS

GRAND GULF NUCLEAR STATION1 OCFR50.59 SUMMARY REPORT FOR THE PERIOD

STARTING NOVEMBER 01, 2002 AND ENDING MARCH 31, 2004

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Page 1 of 4

MEANING OF ACRONYMSARI Alam, Response Instruction LOP Loss of PowerASTM Amer-can Society for Testing and Materials MAPLHGR Maximum Average Planar Linear Heat Generation RateCCE Commitment Change Evaluation MCPR Minimum Critical Power RatioCMWT Core Megawatts Thermal MNCR Material Nonconformance ReportCR Cond tion Report MOV Motor Operated ValveDCP Design Change Package MS Mechanical StandardEP Emergency Procedure MSIV-LCS Main Steam Isclatlon Valve Leakage Control SystemEPI Equipment Performance Instnrction NPE Nuclear Plant EngineeringEPRI Electric Power Research Institute NSSS Nuclear Steam Supply SystemER Engineering Request PDMS Plant Data Management SystemES Electrical Standard PPM Parts Per MillionESF Engineered Safety Feature PRA Probabhlistic Risk AssessmentGE General Electric PSW Plant Service WaterGG Grand Gulf RCIC Reactor Core Isolation CoolingGGN Grand Gulf Nuclear RFO Refueling OutageGPM Gallons Per Minute RHR Residual Heat Removal101 Integrated Operating Instruction RPV Reactor Pressure VesselISI In Service Inspection SCN Standard Change NoticeIST In Service Testing SERI System Energy Resources, Inc.LBDC License Basis Document Change SGTS Standby Gas Treatment SystemLDC License Doctinent Change SOER Significant Operating Experience ReportLHGR Lnear Heat Generation Rate SSW Standby Service WaterLLRT Local Leak Rate Test TRM Technical Requirements ManualLOCA Loss of Coolant Accident UHS Ultimate Heat Sink

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GRAND GULF NUCLEAR STATIONiOCFR50.59 SUMMARY REPORT FOR THE PERIOD


Page 2 of 4

SAFETY EVALUATIONS

Evaluation No. Initiating Document SummarySE 2002-0007-R02 LBDR 2002-104 Core Operating Limits Report Cycle 13 Rev. 2SE 2002-0008-ROl LBDC 2002-116, Revision 1 TRM 7.6.3.3, Item g.1, Inservice Inspection and Testing Program Surveillance

3.6.1.3.6SE 2003-0001 -ROO ERGG-2003-0152-000 Recirc flow control valve min position change analysis

Revision 0SE 2003-0002-ROO Calculation XC-Q1 N11-94004 Update of Offsite and Control Room doses from a MSL break

Revision 0SE 2004-0001-ROO LDC 2004-009 Core Operating Limits Report Cycle 14 Rev. 0SE 2004-0001-RO1 LDC 2004-009 Core Operating Limits Report Cycle 14 Rev. 1SE 2004-0002-ROO Temp Alt 04-005 (Division 1) Manual operation of SSW cooling tower fans In Modes 4 and 5 to assist in

Revision 0 maintaining basin temperature within admin limits during cold weatherTemp Alt 04-006 (Division 2)Revision 0

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ATTACHMENT 1TABLE OF CONTENTS

GRAND GULF NUCLEAR STATION10CFR50.59 SUMMARY REPORT FOR THE PERIOD


COMMITMENT CHANGE EVALUATIONS

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Page 3 of 4

Commitment No. Source Document SummaryCCE 2002-0007 10CFR, Part 55.45 Annual transient and steady state simulator testing requirement was deleted by

a revision to 1 OCFR55.45.CCE 2002-0008 AECM-90/0004 Procedural weaknesses Identified by response to SALP report for inclusion of

appropriate Improvements into the 10CFR50.59 process. Completely removethis commitment.

CCE 2002-0009 AECM-821490 Valve P53F006 seat leakage will te tested at a frequency not to exceed 60months versus every 18 months.

CCE 2003-0001 AECM-38/C024, Aft. 1 Area Delete the drawing control procedure to provide for ready acceptance of NPE5.S3 reviewed, approved and Issued drawings and assignment of drawing

coordinator to coordinate design drawings and resolve problems. Thisprocess Is no longer applicable.

CCE 2003-0002 AECM-8910003.88-26-01.111 NMM NF-104 has been revised to include detailed instructions for completing2.B DOE/NRC form 741.

CCE 2003-0003 AECM-8910003.88-26- Delete the commitment to reassign responsibility of SNM program to the01.111.3.a&b GGNS General Manager

CCE 2003-0004 GNRO 93100029 Paragraph 3, Delete the commitment to track and trend Rosemount transmitters that areSentence 4 susceptible to fill-oil loss.

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GRAND GULF NUCLEAR STATIONIOCFR50.59 SUMMARY REPORT FOR THE PERIOD


COMMITMENT CHANGE EVALUATIONS

Page 4 of 4

CCE 2003-0005 GNRO 94/103.94-13-01.11I.S2 Miscellaneous material storage procedure was revised to identify the personnelwho are required to maintain copies of the inventory sheets and item locationmaps for items stored in the pools.

CCE 2003-0006 VOIDEDCCE 2003-0007 AECM 85/0201 Response 2, Commitment specifying that GGNS Admin procedure Determination of

Paragraph 3 SafetylQuality Classifications' will be used for guidance Is no longer valid sincethe Master Equipment List and its successors (CDB, EDB) have satisfied that

_ requirement.CCE 2003-0008 GNRO-2000100011 99-1903 Since Information tags are now caution tags, 'Protective Tagging" procedure

Paragraph 3 will be revised to more clearly describe how caution tags will be used In placeof Information tags.

CCE 2003-0009 GNRI 99100047 Check valves E38FO02AIB and E38FO03ANB will no longer be disassembledand inspected on a sample basis. Instead, they will be full stroke opened andclosed on a cold shutdown frequency per Ma-1 988, Part 10.

CCE 2003-0010 GNRO 02100054 Delete the commitment to perform a parametric study at the uprated conditionsto quantify the Impact of Thermal Power Optimization on GGNS wear ratesand update the CHECWORKS model as necessary.

CCE 2003-0011 SIL 156 Delete the commitment to verify neutron monitoring instrument tubes areproperly seated after maintenance activities are performed in the vicinity of thetubes.

ATTACHMENT 2

10CFR50.59 Evaluationsand

Commitment Change Evaluations

GGNS 50.59 Safety Evaluation Number

SE 2002-0007-R02

EN-S NUCLEAR OUALrTY RELATED LI-i 01 Revision 2

MEneN MENT ADMINISTRATiVE

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ATTACHMENT 9.1 50.59 REVIEW FORM Page I of 12

1. OVERVIEW / SIGNATURES

Facility: GGNS

de4 t/,Td V 03/JeopOA - I /*Sl/ 0 03

Document Reviewed: Core Operating Limits Report (COLR) Change/Rev.: LOC 2002-104

System Designator(s)/Descriptlon: N/A

Description of Proposed ChangeThis evaluation addresses the Cycle 13 reload changes and operation of the Cycle 13 reload core asgiven In the Core Operating Limits Report (COLR). Revision 1 address the explicit incorporation ofthe increased GE11 channel bow described In CR-GGN-2002-01810 Into the Cycle 13 core designand operation. Revision 2 removes the lImitatIons on equipment out of service allowancespreviously imposed for Cycle 13 exposures >11 GWd1MTU.

If the proposed activity, In Its entirety, involves any one of the criteria below, check the appropriatebox, provide a justlificatlon/basis in the Description above, and forward to a Reviewer. No further50.59 Review is required. If none of the criteria Is applicable, continue with the 50.59 Review.

El The proposed activity is editorial/typographical as defined in Section 5.2.2.1.

Dl The proposed activity represents an OFSAR-only" change as allowed in Section 5.2.2.2 .(Insert item # from Section 5.2.2.2).

El The proposed activity is controlled by another regulation per Section 5.2.2.3.

If further 50.59 Review Is required, check the applicable review(s): (Only the sections Indicated mube Included In the Review.)

El SCREENING Sections l, II, and liI required

D 50.59 EVALUATION EXEMPTION Sections I, Ii, 1II, and IV required

El 50.59 EVALUATION (#: 2002-0007-R. 2 ) Sections I, I, IlIl, and V required

Preparer: Guy B. Spikes/ / A /EOI/Nucl. Eng. - SAI II2//Name (print) / S natur6 Company I Department I Date

G.E. Broadbent/ vIEwerNuc. Eng. - S Li Vj3CName (print) I Signature I Company / Department i Date

OSRC .tI)-c Eer/1J. 1 rt& 1/2Chairman's Signature / Date (N/A for Screenings and 50.59 Evaluation Exemptions)

List of Assisting/Contributing Personnel:Name: Scope of Assistance:J. A. Elam (Echelon Core Design) Core Deslgn and neutrionic inputD. L. Smith (Echelon Fuel Fabrication) Fuel mechanical inputJ. P. Head (Echelon Fuel Fabrication) Core stability and hydraulic compatibility InputG. W. Smith EP Input

II. SCREENING

A. Licensina Basis Document Review

1. Does the proposed activity impact the facility or a procedure as described In any of thefollowing Licensing Basis Documents? (Check "WNA" for those documents that are not applicableto the facility.)

Operating License YES NO WA CHANGE # and/or SECTIONS TO BE REVISED

Operating LUcense 0 =TS 0 0NRC Orders D 0 3

If "YES", obtain NRC approval prior to Implementing the change. (See Section 5.1.13 for exceptions.)

LBDs controlled under 50.59 YES NO WA CHANGE # andlor SECTIONS TO BE REVISED

FSAR El LDC 2002-106

TS Bases 0 0Technical Requirements Manual 0 El

Core Operating Limits Report 0 0 - LDC 2002-104

Offslte Dose Calculations Manual 0 0NRC Safety Evaluation Reports' 0 0If "YES", perform an Exemption Review per Section IV OR perform a 60.59 Evaluation per Section V.

LBDs controlled under other regulations YES NO WA CHANGE # and/or SECTIONS TO BE REVISED

Quality Assurance Program Manual 2 0 0

Emergency Plan2 0 0Security Plan2 3 0 0Fire Protection Program' 0 E0 0(includes the Fire Hazards Analysis)

If "YES", evaluatelprocess any changes In accordance with the appropriate regulation.

2. Does the proposed activity involve a test or experiment not described In theFSAR?if "yes," perform an Exemption Review per Section IV OR perform a 50.59Evaluation per Section V.

00Z

YesNo

3. Does the proposed activity potentially Impact equipment, procedures, or facilitiesutilized for storing spent fuel at an Independent Spent Fuel Storage Installation?(Check "N/A" If dry fuel storage Is not applicable to the facilIty.)If "yes," perform a 72.48 Review In accordance with NMM Procedure LI-112.(See Sections 1.5 and 5.3.1.5 of the EOI 1OCFR5O.59 Review Program Guidelines.)

000

YesNoNIA

' If 'YES," see Section 5.1 5.2 if 'YES." notify the responsible department and ensure a 50.54 Evaluation Is performed.3 The Security Plan is classified as safeguards and can only be reviewed by personnel with the appropriate security clearance. The

Preparer should notify the security department of potential changes to the Security Plan.it " Eb, evaluate the change in accordance with the rtquirenients of the iacillty s Operaung Llcense Condition.

EN-S NUCLEAR OuAuTY RELATED LI-101 Revision 2EnteW MANAGEMENT ADMINISTRATIVE

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ATTACHMENT 9.1 50.59 REVIEW FORM Page 3 of

B. Basis

(Provide a clear, concise basis for the answers given in the applicable sections above. Adequate basis mustbe provided within the Screening such that a third-party reviewer can reach the same conclusions. Simplystating that the change does not affect TS or the FSAR is not an acceptable basis.)

The current MCPR Safety Limit has been shown to be applicable to the Cycle 13 core. As such,Tech Spec 2.1.1.2 does not need to be revised. There are no other Tech Spec, Bases, or TRMchanges required for Cycle 13 operation. There are no NRC orders applicable to the Cycle 13reload campaign.

The Cycle 13 core will contain fuel types currently described in the FSAR, however, the corecharacteristics and response will be somewhat different than currently described in the SAR. Assuch, Cycle 13 analyses have been performed for the new core and the FSAR will need to beupdated appropriately as will the COLR.

The Cycle 13 core design and operation will not affect the OAPM, Emergency Plan, Security Plan, orFire Protection Program.

C. References

[Discuss the methodology for performing the LBD search. State the location of relevant licensing documentinformation and explain the scope of the review such as electronic search criteria used (e.g., key words) orthe general extent of manual searches per Section 5.3.6.4 of LI-1 01.]

LBDs/Documents Reviewed: Keywords:

OLM, FSAR, TS, Bases, TRM

D. Is the validity of this Review dependent on any otherchange? (See Section 5.3.4 of the EOI 10CFR50.59 ProgramReview Guidelines)

D] YesI0 No

If "Yes," list the required changes.

EN-S NUCLEAR OUALrY RELATED LI-101 Revision 2

Ent gMANAGEMENT ADMINISTRATIVEINFORMATION USE

ATTACHMENT 9.1 50.59 REVIEW FORM Page 4 of 12

Ill. ENVIRONMENTAL SCREENING

If any of the following questions Is answered "yes," an Environmental Review must be performed Inaccordance with NMM Procedure EV-115, "Environmental Evaluations," and attached to this 60.59Review.

Will the proposed Change being evaluated:

Yes No

0 El Involve a land disturbance of previously disturbed land areas in excess of one acre (i.e.,grading activities, construction of buildings, excavations, reforestation, creation or removal ofponds)?

O 0 Involve a land disturbance of undisturbed land areas (i.e., grading activities, construction,excavations, reforestation, creating, or removing ponds)?

o El Involve dredging activities in a lake, river, pond, or stream?

o 0 Increase the amount of thermal heat being discharged to the river or lake?

O 0 Increase the concentration or quantity of chemicals being discharged to the river, lake, orair?

o 0 Discharge any chemicals new or different from that previously discharged?

o E Change the design or operation of the intake or discharge structures?

El 0 Modify the design or operation of the cooling tower that will change water or air flowcharacteristics?

0 0 I Modify the design or operation of the plant that will change the path of an existing waterdischarge or that will result in a new water discharge?

o 0 Modify existing stationary fuel burning equipment (i.e., diesel fuel oil, butane, gasoline,propane, and kerosene)?

o 0 Involve the installation of stationary fuel burning equipment or use of portable fuel burningequipment (i.e., diesel fuel oil, butane, gasoline, propane, and kerosene)?'

0 E0 Involve the installation or use of equipment that will result in an air emission discharge?

o El Involve the installation or modification of a stationary or mobile tank?

O 0 Involve the use or storage of oils or chemicals?

0 E0 Involve burial or placement of any solid wastes in the site area that may effect runoff,surface water, or groundwater?

'See NMM Procedure EV-117, 'Air Emissions Management Program,' foi guidanue in wi-weriny this questiuo.

EN-S NUCLEAR OUALfTY RELATED LI-101 Revision 2

EntMANAGEMENT ADMINISTRATIVEINFORMATION USE


V. 50.59 EVALUATION

A. Executive Summary (Serves as input to NRC summary report. Limit to one page or less. Send an electroniccopy to the site licensing department after PSRC approval, if available.)

Brief description of change, test, or experiment:

This safety evaluation assesses the reload-related changes associated with Cycle 13operation as presented in the Core Operating Limits Report (COLR) located in theOperating License Manual (OLM). Cycle 13 has been designed for 492 Effective FullPower Days with a core consisting of 240 fresh ATRIUM-10 assemblies, 204 once-burntATRIUM-1 assemblies. 228 twice burnt GE assemblies. and 128 thrice-burnt GE11assemblies. There are no TS, TS Bases, or TRM changes required to operate with thisnew core, however, the FSAR does require updates. The Cycle 13 core has beendesigned and analyzed for a 1.7% power unrate at a rated thermal power of 3898 MWt insupport of the Cycle 13 implementation of the Appendix K uprate. As such, the reloadanalyses are applicable to both the current power level of 3833 MWt as well as theuprated power level of 3898 MWt. The Appendix K uprate is being reviewed by the NRCand is not addressed in this evaluation. Individual design changes on GGNS systems arenot addressed in this evaluation. Individual design changes on GGNS systems areassessed in the safety evaluation associated with the specific change package and arenot addressed in this evaluation. Attachment 1 provides a detailed description of theCycle 13 reload analysis and the issues considered in this evaluation. Revision 1provides additional evaluations associated with the explicit incorporation of the increasedGE11 channel bow described in CR-GGN-2002-01810 Into the Cycle 13 core design andoperation. Revision 2 removes the limitations on equipment out of service allowancespreviously imposed for Cycle 13 exposures >11 GWdIMTU.

Reason for proposed Change:

Cycle 13 operation will require new core operating limits and the Core Operating LimitsReport has been revised to include these new limits. These limits include flow-, power-,and exposure-dependent LHGR, MAPLHGR, and MCPR limits.

50.59 Evaluation summary and conclusions

The Cycle 13 core configuration and operation has been evaluated with respect tomechanical. neutronic. thermal-hVdraulic, dose, thermal performance, and methodsconsiderations for GGNS. This evaluation concludes that the reload-related changesassociated with Cycle 13 operation will not constitute an unreviewed safety question.

EN-S NUCLEAR QUALUT RELATED LI-1 01 Revision 2

'Etee MANAGEMENT ADMINISTRATIVEMANUAL INFORMATION USE


B. License Amendment DeterminationDoes the proposed Change being evaluated represent a change to a method of 0 Yesevaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only 0 NoQuestion 8. If "No," answer all questions below.Does the proposed Change:1. Result In more than a minimal Increase in the frequency of occurrence of an accident 0 Yes

previously evaluated in the FSAR? 0 NoBASIS:The Cycle 13 core loading and cycle operation will not result in more than a minimalincrease in the frequency of occurrence of an accident previously evaluated In theFSAR. The precursors to these events are independent of the core design and thefrequency classifications reported In SAR Chapter 15 are unaffected by the coreparameters. The following considerations support this conclusion.

MechanicalThe ATRIUM-10 mechanical design has been reviewed for use at Grand Gulf. Nounusual failure modes or increased failure frequency have been identified for this fueldesign. This is the second reload at GGNS with ATRIUM-10 fuel and this fuel designhas accumulated significant problem-free operational experience at other plants.The Cycle 13 bundles will operate within the power history assumptions In the fuelmechanical analyses and will result in exposures within the analyzed burnup limits ofthe ATRIUM-10 and the GE11 mechanical designs. All design criteria for the GE11bundles have been shown to meet their respective limits Including those that will beIrradiated for a fourth cycle.

NuclearThe neutronic characteristics of the Cycle 13 GE11 and ATRIUM-10 mixed coredesign have been considered in the safety analysis. Adequate shutdown margin hasbeen 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 ofCycle 13 fuel in the spent fuel storage racks have been confirmed. Therefore, theprobability of inadvertent criticality has not been increased by the introduction of theCycle 13 reload fuel.

Thermal-HydraulicFRA-ANP's modeling of the GE fuel and the thermal-hydraulic compatibility of theATRIUM-10 and GE11 fuel have been reviewed and found acceptable. In fact, toaccurately model the GE1 1 bundle hydraulics, a Cycle 11 GE1 1 bundle was shippedto FRA-ANP for hydraulic testing in their hydraulic test facility. Analyses have beenperformed to demonstrate that Cycle 13 meets all Enhanced-1A stabilityperformance criteria without changes to the ElA hardware or power-flow map regionboundaries. Therefore, the probability of thermal-hydraulic instabilities has notincreased.

Analyzed EventsThe probability of the occurrence of anticipated operational events is not dependenton the core configuration. No changes to the plant design are required for the Cycle

Am_ I EN-S NUCLEAR UALuTY RELATED LI-101 Revision 2

I EnteMANAGEMENT ADMINISTRATIVE

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13 core. The Cycle 13 core loading will not affect the precursors to any of theChapter 15 events. The probability of an analyzed event therefore has notincreased.

As described in UFSAR Section 15A.6.5.3, the Control Rod Drop Accident (CRDA)results from a failure of the control rod-to-drive mechanism coupling after the controlrod becomes stuck in its fully Inserted position. Although the increased channel bowcondition may result in increased friction between the control blade and itscorresponding fuel assemblies, there would not be sufficient friction to result in amechanical failure of the coupling. Additionally, the control rod drive mechanismwould not produce enough force to result In a mechanical failure of the coupling evenif the channel bow was so severe that the assemblies would preclude blademovement. As such, channel bow is not considered a precursor to the CRDA, andthe increased bow associated with the GE11 bundles would not increase theprobability of this event.

On these bases, the probability of occurrence of accidents previously identified In theSAR is not increased for the Cycle 13 core with increased channel bow.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction 0 Yesof a structure, system, or component important to safety previously evaluated in the 0 NoFSAR?BASIS:

The Cycle 13 reload fuel Is a design that has been shown to be mechanically,neutronically, and thermal-hydraulically compatible with the co-resident GE11 fuel.No plant modifications are required to accommodate the new core design and theonly additional loads placed on plant equipment would be potential friction betweenthe control blades and excessively bowed GE11 bundles. Based on previousexperience with bowed fuel at GGNS and Clinton Power Station, this increasedfriction is not expected to impact scram times. Technical Specification scram timetesting requirements during Cycle 13 would identify any potential scram time Impactsand the appropriate actions would be taken in accordance with TechnicalSpecifications. Additionally, this Increased friction would not be sufficient enough toprovide any failures associated with the control blades or the control blade drivesystem.

A conservative vessel overpressurization analysis has been performed, which showsthat the vessel pressure limit is not exceeded. The precursors to any malfunction ofequipment important to safety are not affected by this change.

Therefore, there is not more than a minimal increase in the likelihood of anoccurrence of a malfunction of a SSC important to safety previously evaluated in theFSAR.

EN-S NUCLEAR OUALITY RELATED LI-101 Revision 2

NA ENT ADMINISTRATIVE;JMANUAL. INFORMATION USE

ATTACHMENT 9.1 1 50.59 REVIEW FORM Page 8 of 12

3. Result In more than a minimal increase in the consequences of an accident previously E Yesevaluated in the FSAR? 3 NoBASIS:As reported in Attachment 1, the acceptance criteria reported in SAR Section15.0.3.1 and the Technical Specifications are satisfied for each event classification.Core operating limits have been developed to ensure that moderate frequencyevents do not violate the MCPR safety limit or fuel cladding strain limits. Theconsequences of infrequent events have been shown to meet the appropriateacceptance criteria while the individual acceptance criteria for the limiting faults havebeen demonstrated to be satisfied. The following considerations support theseconclusions.

Moderate Frequency EventsThe Cycle 13 core operating limits have been developed with NRC-approvedmethodologies such that the MCPR safety limit and the fuel cladding strain limit willnot be violated by any analyzed moderate frequency transient Initiated from anystatepoint available to GGNS. As such, no fuel failures are expected to result fromany moderate frequency event. These analyses considered GGNS-specificoperational modes such as MEOD, SLO, FWHOS, and EOC-RPT inoperable. Thesecore operating limits consist of MCPR, MAPLHGR and LHGR curves that arefunctions of flow, power, and exposure. These limits consider conservative channelbow assumptions that bound the current and expected increased bow associated withthe GE1 1 fuel.

These core operating limits will be incorporated into the core monitoring system,however, as with previous cycles, the GE1 1 operating limits illustrated in the COLRwill differ somewhat from those limits in the core monitoring system. GE hasgenerated pellet-based exposure-dependent LHGR and lattice-based MAPLHGRlimits for their GE11 fuel bundles; however, for competitive reasons, GE hasdesignated the limits for only the most-limiting or least-limiting lattices as non-proprietary. The COLR, which is submitted to the NRC for information only, willtherefore report these non-proprietary limits for reference purposes only, and refer tothe appropriate GE proprietary document for the actual limit. It is recognized thatmost lattices will operate at higher limits than these most-limiting COLR and TRMreference curves, however, this is acceptable since the Technical Specifications3.2.1 and 3.2.3 require that all APLHGRs and LHGRs, respectively be less than orequal to the limits specified in the COLR, which refers to the proprietary GEdocument for the actual limit.

Infrequent EventsThe consequences of the limiting infrequent events have been evaluated and shownto meet their respective acceptance criteria. These events include the pressureregulator failure downscale, misplaced (i.e., misoriented and mislocated) bundle, andsingle loop operation pump seizure accidents. Radiological analyses using thealternative source term (AST), which has been recently licensed for GGNS, havebeen performed to ensure that these events will not result in offsite or control roomdoses greater than their respective acceptance criteria. These evaluations includeconservative channel bow assumptions that bound the current and expectedincreased bow associated with the GE1 1 fuel.

EN-S NUCLEAR OUALM RELATED LI-101 Revision 2MANAGEMENTADMINISTRATIVEz:EntergANGEENMANUAL

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ATTACHMENT 9.1 _ 50.59 REVIEW FORM Pg

Limiting FaultsThe limiting faults at GGNS include the fuel handling accident, the control rod dropaccident, and the design basis LOCA. The radiological analyses for these eventshave recently been developed as part of the GGNS AST effort and bound the Cycle13 core parameters. For the LOCA, MAPLHGR operating limits and single-loopmultipliers have been developed for the Cycle 13 core configuration such that therequirements of 10CFR50.46 are satisfied. The containment response for the Cycle13 core was found to be bounded by previous cycles as is the hydrogen analysis.The seismic/LOCA response of the Cycle 13 core has been confirmed to beacceptable. The Cycle 13 core design results in minor changes to two EPparameters (Mclad and Mfuel), however, the existing EP's remain applicable to Cycle13.

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

4. Result in more than a minimal increase in the consequences of a malfunction of a 0 Yesstructure, system, or component important to safety previously evaluated in the 0 NoFSAR?BASIS:

The Cycle 13 reload fuel is a design that has been shown to be compatible with theco-resident fuel types. The malfunctions of key plant components are analyzed aspart of the reload process with the results reported in various sections of the SAR.The consequences of these malfunctions have been shown to meet their respectiveacceptance criteria.

Therefore, Cycle 13 operation will not result in more than a minimal increase in theconsequences of a malfunction of a structure, system, or component important tosafety previously evaluated In the FSAR.

5. Create a possibility for an accident of a different type than any previously evaluated in 0 Yesthe FSAR? E0 NoBASIS:

The Cycle 13 reload fuel is similar to and compatible with the fuel that was insertedin previous cycles. The details of this design have been specifically considered inthe safety analysis and the core monitoring system. No plant modifications arerequired to accommodate the new core design or Cycle 13 operation. The GGNSCycle 13 fuel types have been approved for the Cycle 13 reactor chemistryconditions.

The GGNS operational parameters (water chemistry requirements, spectral-shiftcore designs, and MEOD rod-lines) have been reviewed and are not expected toresult in unusual crud buildup like that observed on the high-power GE1 1 bundles at

n EN-S NUCLEAR OuALury RELATED LI-1 01 Revision 2

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ATTACHMENT 9.1 50.59 REVIEW FORM I Page 10 of

River Bend. Inspection of a high-power, once-burnt representative fuel bundleduring GGNS RF10 has confirmed that the high-power GGNS Cycle 10 fuel bundleshave no unusual crud buildup.

Therefore, Cycle 13 operation will not create a possibility for an accident of adifferent type than any previously evaluated in the FSAR.

6. Create a possibility for a malfunction of a structure, system, or component important to 0 Yessafety with a different result than any previously evaluated in the FSAR? El No

BASIS

The Cycle 13 reload fuel is a design that has been shown to be mechanically,neutronically, and thermal-hydraulically compatible with the co-resident GE1 1 fuel.The ATRIUM-10 fuel will not introduce any adverse flow distribution effects such aspreferential flow through the ATRIUM-10 bundles that may negatively impact the GE11 bundles. No plant modifications are required to accommodate the new coredesign and no additional loads will be imposed on any existing equipment. TheATRIUM-1 0 bundles provide sufficient clearance for proper control blade operationand allow sufficient bypass flow in the bypass region to provide adequate cooling forcontrol blades and in-core detectors. There are no special operationalconsiderations associated with the Cycle 13 core other than those associated withthe increased bow for the GE1 1 bundles. The higher friction expected between thecontrol blades and GE1 1 bundles experiencing increased bow is not sufficient tocause a failure of the fuel bundle, control blade, or control rod drive coupling.

Therefore, Cycle 13 operation will not create the possibility for a malfunction of anSSC important to safety with a different result than previously evaluated in the FSAR.

7. Result in a design basis limit for a fission product barrier as described in the FSAR 0 Yesbeing exceeded or altered? E NoBASIS:

Mechanical analyses have been performed to ensure that all fuel in the Cycle 13core meet the mechanical design limits for steady-state operation as well as transientconditions including fatigue damage, creep collapse, corrosion, fuel rod internalpressure, rod bow, internal pressure, etc. Additionally, no Cycle 13 fuel will exceedthe applicable burn-up or residence time limits.

Core operating limits have been developed using NRC approved codes in order toensure that the Cycle 13 fuel will not exceed the MCPR safety limits for steady-stateoperation and anticipated operation occurrences. Similarly, operating limits havebeen developed to ensure that the Cycle 13 fuel will not exceed the 1% claddingstrain limit or experience core-wide fuel melt during steady-state operation or AOO's.The results of these analyses show that the vessel pressure does not exceed thevessel pressure safety limit. Although some vessel blowdown to the suppressionpool may be experienced during some AOO's, which would increase the suppression

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pool temperature, the bulk containment pressure increase is negligible and would notexceed the design limit.

As described in Attachment 1, a bounding pressurization event with a failure of thedirect scram has been analyzed for Cycle 13 to ensure compliance with ASME coderequirements. This analysis indicates that the vessel pressure safety limit is notexceeded for Cycle 13.

A design basis limit for the peak fuel enthalpy of 280 ca/gm has been established forthe control rod drop accident (CRDA) to preclude significant fuel cladding failuresuch that core geometry and cooling may be impacted. The CRDA has beenevaluated for Cycle 13. This evaluation considers all potential withdrawal sequencesand concludes that a CRDA will not exceed the 280 cal/gm peak enthalpy limit.Since this accident is a localized event and the peak enthalpy does not exceed 280cal/gm, there is no impact on the vessel or containment pressures. As such theirrespective limits are not exceeded.

1 OCFR50.46 provides limits associated with the ECCS performance analysis (LOCAanalysis). Two such limits are Peak Clad Temperature (PCT) and local cladoxidation. Although these limits are not subject to 1 OCFR50.59, they are discussedIn this evaluation for completeness. Grand Gulf specific analyses have beenperformed for ATRIUM-1 0 and GE1 1 fuel in accordance with 1 OCFR50.46. Theseanalyses, which are applicable to Cycle 13, show that the PCT and local oxidationare well below the limits set forth in 1 OCFR50.46. These analyses also show that thecore-wide metal water reaction, which Is used to evaluate compliance with thecontainment design limit, is less than the 1OCFR50.46 limit. The remainder of theexisting containment analysis associated with LOCA events is applicable to Cycle 13as described in Attachment 1. As such, the containment pressure design limit willnot be exceeded for Cycle 13.

An ATWS evaluation has also been performed for Cycle 13. As described inAttachment 1, the resulting vessel pressure remains below the ASME emergencyvessel pressure limit of 1500 psig and the temperature response used in the existingATWS containment analysis is applicable to Cycle 13. Thus, the containmentpressure design limit will not be exceeded for the ATWS event.

Additional evaluations have been performed for Cycle 13 including Appendix R (FireProtection), hydrogen analyses, and SBO as described in Attachment 1. Theseevaluations show that the existing evaluations are applicable to Cycle 13 and thattheir respective limits are not exceeded.

Therefore, Cycle 13 operation will not result in a design basis limit for a fissionproduct barrier as described in the FSAR being exceeded or altered.

8. Result in a departure from a method of evaluation described in the FSAR used in 0 Yesestablishing the design bases or in the safety analyses? E0 NoBASIS:

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The reload analyses performed by the fuel vendor utilized NRC approved methodsas listed in Technical Specification 5.6.5 and described throughout the FSAR. Thesemethods were consistent with those used for Cycle 12. As described in Attachment1, the uncertainty applied in the Safety Limit calculation associated with each of theequipment out of service combinations was calculated in accordance withFramatome's NRC approved methodology. All remaining GGNS evaluationscurrently described in the FSAR have been shown to be applicable to Cycle 13. Assuch, no new methods were used. Finally, the GGNS EP calculation has beenupdated to consider the minor changes to two fuel parameters. This revision did notIncorporate any new or different methods.

Therefore, Cycle 13 operation will not result in a departure from a method ofevaluation described in the FSAR used in establishing the design bases or in thesafety analyses.


SE 2002-0008-ROI

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1. OVERVIEW I SIGNATURES

Facility: Grand Gulf Nuclear Station

Document Reviewed: LDC 2002-116 ChangelRev. I

System Designator(s)/Description: B21 - MSIV's

Description of Proposed ChangeTRM 7.6.3.3, Item g.1., Inservice Inspection and Testing Programs Surveillance 3.6.1.3.6, MainSteam Isolation Valve (MSIV) requires MSIV full close stroke within 12 hours after reaching 600psig prior to entry to Mode 1. This evaluation will remove that requirement for the MSIV being fullclosed stroke at 600 psig and will permit MSIV IST closed stroke testing during "cold" shutdown.The requirement will be based upon the NUREG-1482 definition of "cold" shutdown testing, whichallows cold shutdown frequency testing including valves tested either while decreasing power toreactor cold shutdown for a outage or while increasing power to seady state power operaton aftera planned or forced shutdown. A method for fast closing the MSIVs while shutting down iscurrently addressed in Integrated Operating Instruction 03-1-01-3 step 5.29.3 and will be added to06-OP-1B21-V-0001. This procedure step would now be credited for MSIV fast closure testing tofulfill [ST cold shutdown testing requirements. The MSIVs, which are cold shutdown frequencytesting valves, will now be treated like other cold shutdown valves. During a refuel outage, allcold shutdown valves are to be stroked. But, during forced outages and non-refuel outages, coldshutdown valves are stroked per a scheduled sequence until startup, at which time testing iscurtailed - not all cold shutdown valves (including the MSIVs) are required to be stroked prior tostartup after a forced outage.If the proposed activity, in its entirety, involves any one of the criteria below, check the appropriatebox, provide a justificationibasis in the Description above, and forward to a Reviewer. No further50.59 Review Is required. If none of the criteria is applicable, continue with the 50.59 Review.o The proposed activity is editoria~ltypographical as defined in Section 5.2.2.1.

o The proposed activity represents an FSAR-only' change as allowed in Section 5.2.2.2(insert Hem # from Section 5.22.2).

If further 50.59 Review Is required, check the applicable review(s): (Only the sections indicated mustbe Included In the Review.)

D | SCREENING Sections l, II, l1l, and IV required

I | 50.59 EVALUATION EXEMPTION I Sections I, ii, liI, IV, and V required

E 1 50.59 EVALUATION (#: 2002-0008-ROI) Sections 1, 11, III, IV, and VI required

Preparer: Gary D. Young / x.4x I EO1/Engineering Programs I'lName (print) / Signaturb I C6m~e71 Department / Date

Reviewer. Robert W. Fuller I ,a -' tl EOI / Design Engineering/ I )-2/-a 35 Cg2pe .int) / Signature / Company / Department / Date

OSRC: 4 A / s , e/ Th e. / .O

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Chairman's Name (print) / Signature / Date[Required only for Programmatic Exclusion Screenings (see Section 5.8) and 50.59 Evaluations.)

List of AssistinglContributing Personnel:Name: Scope of Assistance:

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II. SCREENING

A. Licensing Basis Document Review

1. Does the proposed activity Impact the facility or a procedure as described In any of thefollowing Licensing Basis Documents?

Operating License YES NO CHANGE # and/or SECTIONS IMPACTED

Operating License El 1H1

TS E El

NRC Orders 0 E

If 'YES", obtain NRC approval prior to Implementing the change by Initiating an LBD change In accordance with NMM Lt-113(Reference 2.2.13). (See Section 5.1.13 for exceptions.)

LBDs controlled under 80.59 YES NO CHANGE # (if applicable) and/or SECTIONS IMPACTED

FSAR 1H 0 LDC 2002-116, FSAR Section 16B.1-212

TO Dasez E H

Technical Requirements Manual 1H 0 LDC 2002-116, TRM 7.6.3.3

Core Operating Limits Report 0 H

NRC Safety Evaluation Reports' 13 H

If 'YES", perform an Exemption Review per Section V OR perform a 50.59 Evaluation per Section VI AND Initiate an LBDchange In accordance with NMM LI-113 (Reference 2.2.13).

1LBDs controlled under other regulations YES NO CHANGE # (if applicable) and/or SECTIONS IMPACTED

Quality Assurance Program ManuallO H

Emergency Plan2 0 H

Fire Protection Prograrm1 0 H(includes the Fira Hazards Analysis)

Offsite Dose Calculations Manual 1 HII

if 'YES', evaluate any changes in accordance with the appropriate regulation AND Initiate an LBD change In accordance withNMM LI-113 (Reference 2.2.13).

2. Does the proposed activity Involve a test or experiment not described In theFSAR?If "yes," perform an Exemption Review per Section V OR perform a 50.59Evaluation per Section VI.

3. Does the proposed activity potentially Impact equipment, procedures, orfacilities utilized for storing spent fuel at an Independent Spent Fuel StorageInstallation?(Chock "N/A" if dry fuel storage is not applicable to the facility.)If "yes," perform a 72.48 Review In accordance with NMM Procedure Li-112.(See Sections 1.5 and 5.3.1.5 of the EOI IOCFR50.59 Review Program Guidelines.)

D YesP9 No

D Yesa NoEl N/A

B. Basis

'If 'YES,' see Section 5.1.4.2 If YES,' notify the responsible department and ensure a 50.54 Evaluation is performed. Attach Ihe 50.54 Evaluation.' If YES.' evaluate the change in accordance with the requiremintq of the facility's Operating LUcnaC Condition.

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Provide a clear, concise basis rut tire answers given In the applicable sections above. Explain why theproposed activity does or does not impact the Operating Ucense/Technical Specfications and/or the FSARand why the proposed activity does or does not involve a new test or experiment not previously described Inthe FSAR. Adequate basis must be provided within the Screening such that a third-party reviewer can reachthe same conclusions. Simply stating that the change does not affect TS or the FSAR is not an acceptablebasis. See EOI 50.59 Guidelines Section 5.6.6 for guidance.)

The purpose of the evaluation is to provide the rationale for enhancing the test requirements forMSIV fast closure. The enhancement treats the MSIVs like other IST cold shutdown frequencyvalves by allowing MSIV fast closure testing (if required per IST frequency requirements) anytimeduring cold shutdown (including while decreasing power to reactor cold shutdown or whileincreasing power to steady state power operation) as defined in NUREG-1482. Currently MSIVfast closure testing is done during startup at 600 psig in 06-OP-1B21-V-0001. It will now beacceptable to perform MSIV fast closing IST testing when decreasing power to reactor coldshutdown, during reactor cold shutdown, or while increasing power to steady state poweroperation. It is preferred to fast closure test the MSIVs at low steam flow because the MSIVs wouldbe 'wet and shot' but any pressure and temperature in accordance with IST test frequencyrequirements is acceptable.

During a refuel outage, all cold shutdown valves (including the MSIVs) are to be stroked. But,during forced and non-refuel outages, cold shutdown valves are stroked per a scheduled sequenceuntil startup, at which time testing is curtailed - therefore, not all cold shutdown valves (including theMSIVs) are required to be stroked prior to startup (per NUREG-1482).

Operating License:

The Grand Gulf Nuclear Station (GGNS) operating license does not affect any MSIV fast closuretesting. The Technical Specifications and the Environmental Protection Plan are not impacted bythis project. Therefore, the proposed activity does not impact the GGNS operating license.

Technical Specifications:

The MSIV fast closure testing is covered by Technical Specification Surveillance Requirement3.6.1.3.6. However, none of the Technical Specification requirements for MSIV fast closure testingare affected. In addition, the evaluation will not create a system configuration or operatingcondition such that a Technical Specifications LCO or surveillance requirement is no longeradequate. Likewise, the evaluation will not bypass or invalidate automatic actuation featuresrequired to be operable by the Technical Specifications or exceed any limits specified in theOperating License and Technical Specifications. Therefore, no Technical Specifications change isrequired for the issuance of this evaluation.

UFSAR:

The UFSAR Is affected by this evaluation. UFSAR Section 16B.1 page 16B.1-212 (which is theTRM in the UFSAR) will be changed by this evaluation. MSIV fast closure testing is discussed hereand the pressure at which the MSIVs are tested. UFSAR Section 16B.1 section 7.6.3.3 g.1. will bedeleted allowing MSIV fast closure testing during per IST cold shutdown frequency requirements.This new information will allow MSIV fast closure testing (if required per IST cold shutdownfrequency requirements) during cold shutdown as defined in NUREG-1482, including testing whiledecreasing power to reactor cold shutdown, during reactor cold shutdown, or while increasingpower to seady state power operaton, both after a planned or forced shutdown.

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NRC Ordors:

The NRC Orders issued at Grand Gulf are not affected by this evaluation because this evaluationdeals with MSIV fast closure testing and this evaluation is not to be used for security reasons whichis what Grand Gulfs current NRC Orders deal with.

Technical Specification Oases:

There are no Technical Specifications or Bases impacted by this activity. The TechnicalSpecification for Containment isolation is 3.6.1.3 and the surveillance requirement under thisTechnical Specification is 3.6.1.3.6 for fast closure testing will remain the same. This is anevaluation for assuring that MSIV fast closure testing can be accomplished during cold shut asdefined in NUREG-1482. including while decreasing power to cold shutdown, during reactorshutdown, or while increasing power to steady state power operation.

Technical Requirements Manual (TRM):

There is impact to the Technical Requirements Manual affected by this activity. The affected TRMsection is 7.6.3.3. This section will be changed to delete the requirement of 7.6.3.3.g.1 thatspecifies at reactor pressure of 600 psig during reactor startup. The method of MSIV fast closuretesting during shutdown is already performed in Integrated Operating Instruction 03-1-01-3, step5.29.3. The procedure for MSIV fast closure testing, 06-OP-1 B21-V-0001 will reflect MSIV fastclosure testing during cold shutdown as defined in NUREG-1482, including while dereasing powerto cold shutdown, during reactor shutdown, or while increasing power to steady state poweroperation.

Core Operating Limits Report:

This activity does not impact the COLR (GGNS Core Operating Limits Report). This evaluationdiscusses MSIV fast closure IST cold shutdown frequency testing during cold shutdown as definedin NUREG-1482. It does not have any impact on the COLR and does not affect any licensingactivities.

Offsite Dose Calculations Manual:

This activity does not impact any equipment required to monitor offsite dose. Therefore, nochanges to the ODCM is required.

NRC Safety Evaluation Reports.

There is no impact to any SERs by providing an evaluation for MSIV fast closure IST coldshutdown frequency testing being performed during cold shutdown as defined in NUREG-1482.

Quality Assurance Program Manual:

This evaluation complies with all requirements of the Entergy Quality Assurance Program Manual,as applicable. This activity does not change any commitments contained in the QAPM. Therefore,this activity does not require a change to the QAPM.

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Emergency Plan:

This evaluation does not impact the interaction of GGNS personnel and offsite agencies inresponse to an emergency.

Security Plan:

This evaluation does not impact the Security Plan since it does not require the breaching of securitybarriers. Therefore, no change to the Security Plan is required.

Fire Protection Program (includes the Fire Hazards Analysis):

This evaluation does not impact the Fire Protection Program by providing MSIV fast closure testingduring controlled reactor vessel shutdown.

C. References

Discuss the methodology for performing the LBD search. State the location of relevant licensing documentinformation and explain the scope of the review such as electronic search criteria used (e.g., key words) or thegeneral extent of manual searches per Section 5.3.6.4 of LI-1 01, NOTE: Ensure that electronic andmanual searches are performed using controlled copies of documents. If you have any questions,contact your site Licensing department

LBDs/Documents reviewed via keyword search:

UFSAR 16.1 (page 16B.1-212), TechnicalSpecification 3.6.1.3 (SR 3.6.1.3.6),

TRM 7.6.3.3 g 1

LBDs/Documents reviewed manually:

NIA

Keywords:

MSIV, Main Steam Isolation Valve, 600psig

D. Is the validity of this Review dependent on any otherchange? (See Section 5.3.4 of the EOI 10CFR50.59 ProgramReview Guidefinec.)

o YesB No


NIA

References

1.GIN 2001/00986

2. Integrated Operating Instruction (101) 03-1-01-3

3.06-OP-1 B21 -V-0001, MSIV Operability Test

4.TRM 7.6.3.3 section g paragraph I

5.Technical Specification Surveillance 3.6.1.3.66.NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plant

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111. ENVIRONMENTAL SCREENING

If any of the following questions is answered "yes," an Environmental Review must be performedIn accordance with NMM Procedure EV-115, "Environmental Evaluations," and attached to this50.59 Review. Consider both routine and non-routine (emergency) discharges when answeringthese questions.


Yes No

1. 0 10 Involve a land disturbance of previously disturbed land areas in excess of one acre (i.e.,grading activities, construction of buildings, excavations, reforestation, creation or removalof ponds)?

2. 03 0 Involve a land disturbance of undisturbed land areas (i.e., grading activities, construction,excavations, reforestation, creating, or removing ponds)?

3. 03 0 Involve dredging activities in a lake, river, pond, or stream?

4. 0 I0 Increase the amount of thermal heat being discharged to the river or lake?

5. 03 0 Increase the concentration or quantity of chemicals being discharged to the river, lake, orair?

6. 03 0 Discharge any chemicals new or different from that previously discharged?

7. 03 0 Change the design or operation of the intake or discharge structures?

8. 0 0 Modify the design or operation of the cooling tower that will change water or air flowcharacteristics?

9. 0 0 Modify the design or operation of the plant that will change the path of an existing waterdischarge or that will result in a new water discharge?

10. 0 11 Modify existing stationary fuel burning equipment (i.e., diesel fuel oil, butane, gasoline,propane, and kerosene)?'

11. 0 0E Involve the installation of stationary fuel burning equipment or use of portable fuel burningequipment (i.e., diesel fuel oil, butane, gasoline, propane, and kerosene)?'

12. 0 E 0 Involve the installation or use of equipment that will result in an air emission discharge?

13. 0 01 Involve the installation or modification of a stationary or mobile tank?

14. 0 0 Involve the use or storage of oils or chemicals that could be directly released into theenvironment?

15. 0 0l Involve burial or placemernt of any solid wastes in the site area that may affect runoff,surface water, or groundwater?

See NMM Procedure EV-117, 'Air Emissions Management Program,' for guidance in answering this question.

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IV. SECURITY PLAN SCREENING

If any of the following questions is answered "yes," a Security Plan review must be performed bythe Security Department to determine actual Impact to the Plan and the need for a change to thePlan.

A. Could the proposed activity being evaluated:

Yes No

1. 0 E31 Add, delete, modify, or otherwise affect Security department responsibilities (e.g., includingfire brigade, fire watch, and confined space rescue operations)?

2. 0 VJ Result in a breach to any security barrier(s) (e.g., HVAC ductwork, fences, doors, walls,ceilings, floors, penetrations, and ballistic barriers)?

3. 0 0a Causc materiels or equipment to be placed or Installed within the Security Isolation Zone?

4. 0 08I Affect security lighting by adding or deleting lights, structures, buildings, or temporaryfacilities?

5. 0 0 Modify or otherwise affect the intrusion detection systems (e.g., E-fields, microwave, fiberoptics)?

6. 0 0l Modify or otherwise affect the operation or field of view of the security cameras?

7. 0 0i Modify or otherwise affect (block, move, or alter) installed access control equipment,intrusion detection equipment, or other security equipment?

8. 0 0 Modify or otherwise affect primary or secondary power supplies to access controlequipment, intrusion detection equipment, other security equipment, or to the CentralAlarm Station or the Secondary Alarm Station?

9. a 0 Modify or otherwise affect the facility's security-related signage or land vehicle barriers,including access roadways?

10. 0 0 Modify or otherwise affect the facility's telephone or security radio systems?

The Security Department answers the following questions if one of the questions was answered44yes".

B. Is the Security Plan actually impacted by the 0 Yesproposed activity? la No

C. Is a change to the Security Plan required? 0 Yes Change # (optional)

09 No

Name of Security Plan reviewer (print) I Signature I Date

VI. 50.59 EVALUATION

A. Executive Summary (Serves as input to NRC summary report. Limit to one page or less. Send anelectronic copy to the site licensing department after OSRC approval, if available.)


Currently TRM 7.0.3.3 g.1. requires MSIV fast closure testing to be performed during startup at 600psig reactor pressure. This testing is done for Technical Specification 3.6.1.3 under SurveillanceRequirement 3.6.1.3.6 which is MSIV fast closure testing. This evaluation deletes the TRMrequirement. Procedure 06-OP-1 B21 -V-0001 will need to be revised to allow MSIV fast closuretesting during cold shutdown as defined in NUREG-1 482, including while decreasing power to cold |shutdown, during reactor shutdown, or while increasing power to steady state power operation.There is a minor procedure change for 101 03-1-01-3. The tables will be removed from the 101 03-1-01-3 to the 06-OP-1B21-V-0001 because it currently fast closes the MSIV and records therequired information. This will eliminate the need for MSIV fast closure testing during startup afteran outage, unless required for post-maintainence retest per the 01-S-07-28 ASME Section XlRepair/Replacement Program checklist.


During controlled reactor shutdown MSIV fast closure testing is performed by 101 03-1-01-03. ISTMSIV fast closure testing is currently performed and credited during startup. This may not be theoptimum time to test the MSIVs. During controlled shutdown it would be desirable to fast closuretest the MSIVs during shutdown around 60 psig reactor pressurc. Fast closure testing during stiutdown reduces the possibility of preconditioning the MSIVs prior to fast closure testing. Currently,the valves may have maintenance performed on them during an outage, thus giving theimpression of preconditioning of the MSIVs. Fast Closure testing of the MSIVs during controlledshutdown of the reactor would eliminate this situation and would be an IST enhancement becauseno perceived MSIV preconditioning occurs.Additionally, MSIV post-maintenance retest is currently performed during restart at approximately600 psig. It may be decireable to perform those retests during reactor cold shutdown. Thisevaluation will provide flexibility to perform testing and retesting at the optimum time to supportOperation's activities and outage schedules.


For the MSIV fast closure testing, the requirements are going to be deleted in TRMSection 7.6.3.3 9.1. This testing currently performed by Integrated OperatingInstruction 03-1-01-3 but not utilized for IST testing will be moved to 06-OP-1 B21-V-0001. This will allow MSIV fast closure testing during cold shutdown as defined inNUREG-1482, including while decreasing power to cold shutdown, during reactorshutdown, or while increasing power to steady state power operation. The procedurefor this testing is 06-OP-1 B21-V-0001. Procedure 06-OP-1 B21-V-0001 needs to berevised to include MSIV fast closure testing during cold shutdown as defined inNUREG-1 482, including while decreasing power to cold shutdown, during reactorshutdown, or while increasing power to steady state power operation.

Stroking of the MSIVs will be performed per the limitationsfrequirements of GIN2001/00986. If the valve is "wet" (i.e., the steam line is filled with steam) the valve canbe repeatedly stroked with no minimum wait time between valve cycles and the valvewill not experience valve galling or damage due to stroking. Although experience has

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shown that 'cold' stroke of the MSIVs are consistent with the "hot" stroke of theMSIVs, the 06-OP-1 B21-V-0001 will be revised with enhanced stroke timeacceptance criteria.

If fast closure testing of the MSIVs is performed at 60 psig reactor pressure, thesaturation temperature is greater than 300*F. The MSIVs are still 'hot' even with thetest pressure reduced from 600 psig to 60 psig. The MSIV fast closure time will beunaffected by reducing the test pressure to 60 psig. The MSIVs fast closure stroketime can be performed at any pressure and temperature. Limitations about strokingthe MSIVs scold" and 'dry" has been communicated to operations via GIN2001100986.

The requirement to MSIV fast closure test at 600 psig was not found in any of theMSIV technical specifications, vendor manual or design documents for the MSIVs. Itappears that there was no basis for this reactor pressure other than the MSIVs wasdesired to be tested 'wet' and 'hot". Therefore 600 psig was chosen as the testpressure apparently arbitrarily chosen based on steam conditions.

Changing the TRM to allow MSIV fast closure testing is an IST enhancement. Iteliminates preconditioning of the MSIVs prior to as-found testing. This TRM changedoes not affect MSIV design functions and continues to provide safety pressureboundary, seismic, and tornado protection requirements established In the design andlicensing basis for the MSIVs. All essential plant systems and equipment will functionas assumed in the Accident Analysis. Therefore, this change will have no effect onany consequences of the accidents evaluated previously in the UFSAR, will notchange offsite dose to the public, will not affect any fission product barriers, and doesnot alter any assumptions previously made in evaluating the radiologicalconsequences of an accident described in the UFSAR. As a result, this change willnot increase the consequences of an accident or create an accident of a different typethan any evaluated previously in the UFSAR.

The MSIV are designed to meet the current licensing and design requirements. Thisevaluation does not change the MSIV system actuation, flow parameters, or thepressure boundary requirements and they will function as assumed in the AccidentAnalysis. Therefore, this change will not increase the consequences of a malfunctionof equipment important-to-safety or create the possibility of a malfunction ofequipment important-to-safety of a different type than any evaluated previously in theUSAR. Also, this change will not reduce the margin of safety as defined in the basisfor any Tcchnical Specification.

Based on the results of this safety-evaluation, the affects associated with thisevaluation are inconsequential and, therefore, do not constitute an Unreviewed SafetyQuestion (USQ).

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This change is an enhancement to MSIV fast closure testing allowing testing during cold shutdownas defined in NUREG-1482, including while decreasing power to cold shutdown, during reactorshutdown, or while increasing power to steady state power operation. It does not affect plantinfluents or effluents. Therefore, this evaluation does not represent a change to the EnvironmentalProtection Plan or a change that will affect the environment There is no potential for anUnreviewed Environmental Question, therefore, there is no need to perform an environmentalevaluation.

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B. License Amendment Determination

Does the proposed Change being evaluated represent a change to a method of 3 Yesevaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only IS NoQuestion 8. If "No," answer all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an 0 Yesaccident previously evaluated in the FSAR? i No

BASIS:This evaluation does not make a physical change to the Plant. It changes the TRM requirementsfor MSIV fast closure testing. This enhancement will allow MSIV fast closure testing during coldshutdown as defined in NUREG-1482. including while detrr.asing power to cold shutdown, duringreactor shutdown, or while increasing power to steady state power operation. This is a change totesting conditions and does not change frequency of occurrence of an accident. A change to theTRM is made to allow fast closure testing of the MSIVs during cold shutdown as defined inNUREG-1482, including while decreasing power to Cold shutdown, during reactor shutdown, orwhile increasing power to steady state power operation. The MSIV isolation mode of operation isnot changed or affected by this evaluation.

Chapters 3, 6 and 15 of the USAR were reviewed to determine the impact of this evaluation onany accidents previously analyzed. Several accidents and transients analyzed in Chapter 15were determined to be unaffected this testing enhancement.

The change in TRM 7.6.3.3 allowing testing during cold shutdown as defined in NUREG-1482,including while decreasing power to cold shutdown, during reactor shutdown, or while increasingpower to steady state power operation has no impact on the probability of occurrence of allowingreactor vessel inventory to leak into the environment. Based upon the above discussion, it isconcluded that the proposed evaluation does not result in an increase in the frequency ofoccurrence of an accident previously evaluated in the USAR.

2. Result in more than a minimal increase in the likelihood of occurrence of a 0 Yesmalfunction of a structure, system, or component important to safety previously A Noevaluated in the FSAR?BASIS:The TRM change and evaluation meets the original licensing and design requirements for theassociated MSIVs. The change does not affect required MSIV design or system operation, meetscontainment isolation requirements, does not invalidate existing seismic, pipe stress, or tornadodesign criteria, moderate energy line crack flooding and spray analysis, high energy line break jetimpingement and pipe whip evaluations, pose a fire hazard, and dues not impact RPV Internals,core analysis or thermal hydraulic design criteria and analysis, or affect any UFSAR Chapter 15Accident Analysis. Based upon the above discussion, it is concluded that this change does notincrease the likelihood of occurrence of a malfunction of equipment important to safety previouslyevaluated in the USAR.

The change in the TRM allows MSIV fast closure testing to occur either during cold shutdown asdefined in NUREG-1482, including while decreasing power to cold shutdown, during reactorshutdown, or while increasing power to steady state power operation.

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3. Result in more than a minimal inr.repse in the consequences of an accident 13 Yespreviously evaluated in the FSAR? Ed No

BASIS:This evaluation allows MSIV fast closure testing to occur during cold shutdown as defined inNUREG-1482, including while decreasing power to cold shutdown, during reactor shutdown, orwhile increasing power to steady state power operation. The TRM 6.7.3.3 will be revised toreflect these changes, Procedures to collect the MSIV fast closure data exist and will be utilizedas needed to meet the test conditions. All affected piping, fittings, and valve pressure boundariesare qualified to the appropriate fluid transients and operational conditions and meet the designand licensing requirements for pressure boundary integrity and containment isolation provisions.

This TRM change and evaluation affects equipment located within the primary containment and theauxiliary building, and does not penetrate any structural wall or barriers. Therefore there is no affect to theboundary Integrity of any fire area. Theretbre this evaluation will not compromise the function nor integrityof structures, systems or components important to safety and has no effect on the Fire Hazards AnalysisReport.

A review of USAR Chapter 15 Accident Analysis was performed. The proposed TRM change wasevaluated against the existing safety analyses to determine if any of the analyses are impacted.The criteria used in this evaluation were that the change should not impact the ability of MSIVs toprovide containment isolation within the Technical Specification Surveillance requirement, shouldnot create an event of a type not previously analyzed, and previous component analyses shouldnot be negatively impacted. The proposed TRM and FSAR change satisfies the evaluationcriteria, and the TRM and FSAR change is within the bounds of the existing safety analyses.

The other potentially impacted accidents are a FSAR Chapter 6 analysis. These accidents areconsidered limiting faults. For the case of the recirculation line break inside containment (i.e.,drywell), the reactor vessel depressurizes very rapidly and the MSIVs would go closed within therequired Technical Specification Surveillance requirement (SR 3.6.1.3.6).

MSIVs reactor/containment isolation remains unchanged. All essential plant systems andequipment will function as assumed in the Accident Analysis. There is no increase in offsite dosedue to any accident previously evaluated. Therefore the proposed activity does not increase theconsequences of an accident evaluated previously in the UFSAR.

4. Result in more than a minimal increase in the consequences of a malfunction of 3 Yesa structUre, system, or component important to safety previously evaluated in the E NoFSAR?BASIS:This evaluation does not make a physical change to the Pladit. MSIV fast closure testing Iscurrently performed and credited during startup. This may not be the optimum time to test theMSIVs. During controlled shutdown it may be desirable to fast closure test the MSIVs duringshutdown around 60 psig reactor pressure. Fast closure testing during shut down eliminates thepossibility of preconditioning the MSIVs prior to fast closure as-found testing. Currently, thevalves may be worked on during an outage, thus giving the impression of preconditioning of theMSIVs. Fast Closure testing of the MSIVs during cold shutdown as defined in NUREG-1482,including while decreasing power to cold shutdown, during reactor shutdown, or while increasingpower to steady state power operation.

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As an IST enhancement, the TRM is changed to allow testing during restart or shutdown.Integrating Operating Instructions 03-1-01-3 is a procedure that currently tests MSIV fast closureduring Shutdown. Procedure 06-OP-1 B21 -V-0001 would be revised with the MSIV fast closuretesting instructions from 101 03-1-01-3 when shutting down.

This TRM change and evaluation meets the current design and licensing basis such that allaffected systems, structures, and components, including the RPV and its internals that areImportant to safety meet all required operational modes and will function as assumed in theAccident Analysis. Fast Closure Containment isolation of the MSIVs does not compromisecontainment integrity because the inboard and outboard MSIV containment isolation valves willstill meet the Technical Specification Surveillance Requirement 3.6.1.3.6 for closure time.Evaluation of the various piping and hydraulic transients were performed to ensure that noadverse impact to the MSIV lines were created. Any breaks and/or malfunctions of safety relatedor important to safety equipment and the mitigating actions for these failures or malfunctionsremain the same. As such there is no change in the radiological consequences at the siteboundary. Therefore this evaluation will not increase the consequences of a malfunction ofequipment important to safety evaluated previously in the UFSAR.

5. Create a possibility for an accident of a different type than any previously 3 Yesevaluated in the FSAR? Ed No

BASIS:This evaluation meets and does not invalidate the current design and licensing basis for thefollowing:

* Containment isolation provisions.* USAR Chapter 15 - Accident Analysis.* Fluid transient and system operational conditions.

The TRM is being changed to allow MSIV fast closure testing during either restart or shutdown.There are procedures in place that currently collect the MSIV fast closure times. The MSIVisolation time is not be changed and the enhancement to the TRM provides a test condition that isconservative with respect to the existing test condition of reactor vessel pressure at 600 psig.There are no other events postulated as a result of this evaluation which could create thepossibility of an accident of a different type than any evaluated previously in the USAR.

Therefore this evaluation as previously described will not create the possibility of an accident of adifferent type than evaluated previously in the UFSAR.

6. Create a possibility for a malfunction of a structure, system, or component 0 Yesimportant to safety with a different result than any previously evaluated in the E NoFSAR?BASISThis evaluation meets the current design and licensing basis such that all affected systems,structures, and components including the RPV and its internals that are important to safety meetall required plant operational modes and events. This includes tornado, seismic, and pipe stresscriteria, moderate energy line crack flooding and spray analysis, high energy pipe break jetimpingement and pipe whip target evaluations, fire hazards, and UFSAR Chapter 15 AccidentAnalysis.

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MSIV fast closure testing is currently performed and credited during startup. This condition is notconservative with respect to the proposal to test the MSIVs during cold shutdown as defined inNUREG-1 482, including while decreasing power to cold shutdown, during reactor shutdown, orwhile increasing power to steady state power operation. During controlled shutdown it may bedesirable to fast closure test the MSIVs during shutdown around 60 psig reactor pressure. Fastclosure testing during shut down eliminates the possibility of preconditioning the MSIVs prior tofast closure IST as-found testing. Currently, the valves may be worked on during an outage, thuspossibly giving the impression of preconditioning of the MSIVs. Fast Closure testing of the MSIVsduring cold shutdown as defined in NUREG-1482, including while decreasing power to coldshutdown, during reactor shutdown, or while increasing power to steady state power operationwould eliminate this situation.

There are no other postulated events which could create the possibility of a malfunction ofequipment important to safety of a different type than any evaluated previously in the UFSAR.Therefore this evaluation as previously described will not create the possibility of a malfunction ofequipment important to safety of a different type than any evaluated previously in the UFSAR.

7. Result in a design basis limit for a fission product barrier as described in the 3 YesFSAR being exceeded or altered? 0 No

BASIS:The MSIV capability of containment isolation is not changed or affected by this evaluation. TheMSIV fast stroke time for containment isolation will perform as required in accordance with thedesign and licensing basis including USAR Chapter 15 Accident Analysis. Therefore, thisevaluation causes no change to plant equipment or the design basis that is addressed by theOperating License. A review of the Facility Operating License Conditions and the TechnicalSpecification show that these documents are not affected by this evaluation. TechnicalSpecifications, Section 3.6, Containment Systems, Subsection 3.6.1.3, Primary ContainmentIsolation Valve (PCIVS) address containment isolation. No change to these sections of theTechnical Specification is required. The Technical Specification Surveillance Requirement3.6.1.3.6 will maintain the current isolation valve stroke times.

8. Result in a departure from a method of evaluation described in the FSAR used in Cl Yesestablishing the design bases or in the safety analyses? El No

BASIS:This evaluation does not make a physical change to the Plant. Containment isolation is achievedby closing the MSIVs. The TRM section 7.6.3.3 g.1 will be deleted. A change to the 06-OP-1 B21-V-0001 is made to incorporate Integrated Operating Instruction 03-1-01-3 for MSIV fastclosure. The MSIV mode of operation is not changed or affected by this evaluation.

As an IST enhancement, the 06-OP-1 B21-V-0001 is changed to allow testing during coldshutdown as defined in NUREG-1482, including while decreasing power to cold shutdown, during |reactor shutdown, or while increasing power to steady state power operation. IntegratingOperating Instructions 03-1-01-3 is a procedure that currently tests MSIV fast closure duringShutdown. Procedure 06-OP-i B21-V-0001 would be revised with the MSIV fast closure testinginstructions from 101 03-1-01-3 at reactor pressure of low steam flow during controlled shutdown.The MSIV fast closure testing would be allowed during cold shutdown as defined in NUREG-1482, including while decreasing power to cold shutdown, during reactor shutdown, or whileIncreasing power to steady state power operation.


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I. OVERVIEW / SIGNATURES e9 SiG h4FJW a 4v/6


Document Reviewed: ER-GG-2003-0152-000 ChangelRev. 0

System Designator(s)/Description: B33

Description of Proposed ChanqeER-GG-2003-0152 performs the analysis to increase the recirc flow control valve (FCV) min position (MINED) from 7% to 12% to alleviate FCV sticking problems after recirc pump upshift.

If the proposed activity, in its entirety, involves any one of the criteria below, check the appropriatebox, provide a justification/basis In the Description above, and forward to a Reviewer. No further50.59 Review Is required. If none of the criteria is applicable, continue with the 50.59 Review.

o The proposed activity is editorialtypographical as defined in Section 5.2.2.1.

o The proposed activity represents an 'FSAR-only" change as allowed in Section 5.2.2.2--(Insert item # from Section 5.2.2.2).

If further 50.59 Review is required, check the applicable review(s): (Only the sections indicated mustbe Included In the Review.)

0 SCREENING Sections I, II, III, and IV required

3 50.59 EVALUATION EXEMPTION Sections I, II, III, IV, and V required

0E 50.59 EVALUATION (#: 2 e q -eld) ll 6 . Sections I, II, Ill, IV, and VI required

Preparer: G.E Broadbent /EOI/Engineerinal/Name (print) /Inature / Company/ Department I Date

Reviewer: S.C. Stanchfield/ / I EOI / Engineering / Yz/,&-7Name (print) Tsi any I Department I Date ' {

Shore- 1OSRC:Chairman's Name (print) I Signature I Date %i % U

(Required only for Programmatic Exclusion Screenings (see Section 5.8) and 50.59 Evaluations.]

List of Assisting/Contributing Personnel:Name: Scope of Assistance:

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II. SCREENING


1. Does the proposed activity impact the facility or a procedure as described In any of thefollowing Licensing Basis Documents?


Operating License 10 E

TSO I D

NRC Orders 00 E

If "YES", obtain NRC approval prior to Implementing the change by Initiating an LED change In accordance with NMM U-113(Reference 2.2.13). (See Section 5.1.13 for exceptions.)


FSAR El 0 LDC 2003-051

TS Brues 0 I

Technical Requirements Manual 0 _0

Core Operating Umits Report 0 01

NRC Safety Evaluation Reports' 0

If YES, perform an Exemption Review per Section V OR perform a 50.59 Evaluation per Section VI AND Initiate an LEDchange In accordance with NMM U-113 (Reference 2.7.13)

LBDs controlled under other regulations YES NO CHANGE # (it applicable) and/or SECTIONS IMPACTED

Ouality Assurance Program Manual 2 0 0

Emergency Plan E0 __

Fire Protection Program3 S 00(includes the Fire Hazards Analycic)

Offsite Dose Calculations Manual 0 00

It "YES", evaluate any changes In accordance with the appropriate regulation AND Initiate an LBO change In accordance withNMM Li-113 (Reference 2.2.13).

2. Does the proposed activity involve a test or experiment not described in theFSAR?If "yes," perform an Exemption Review per Section V OR perform a 50.59Evaluation per Section VI.

o YesEl No

3. Does the proposed activity potentially impact equipment, procedures, orfacilities utilized for storing spent fuel at an Independent Spent Fuel StorageInstallation?(Chcck "N/A" If dry fuel storage Is not applicable to the facility.)If "yes," perform a 72.48 Review in accordance with NMM Procedure LI-112.(See Sections 1.5 and 5.3.1.5 of the EOI 1 0CFR50.59 Review Program Guidelines.)

E3

E3El

YesNoN/A

If YES.' see Section 5.1.4.1 I YES.' nctify the responsible department and ensure a 50.54 Evaluation is performed. Attach the 50.54 Evaluation.It 'YES.' evaluate the change in accordance with thA renqttiramonte of the facility's Operoting Licenre Cu idItton.

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B. Basis

Provide a clear, concise basis for the answers given in the applicable sections above. Explain why theproposed activity does or does not impact the Operating License/Technical Specifications and/or the FSARand why the proposed activity does or does not involve a new test or experiment not previously described inthe FSAR. Adequate basis must be provided within the Screening such that a third-party reviewer can reachthe same conclusions. Simply stating that the change does not affect TS or the FSAR is not an acceptablebasis. SAA FOI 50.59 Guidolines Soction 6.6.6 for guidance.)

The FCV MIN ED position is not governed bv the Operatina License. Tech Specs. Tech SpecBases, TRM or NRC Orders. It is described in the FSAR and is credited in the Idle Loop Startupanalysis in FSAR Section 15.4.4. A revised reload analysis has confirmed the operating limits inthe current Core Operating Limits Report are unaffected by this change in MIN ED.

C. References

Discuss the methodology for performing the LBD search. State the location of relevant licensing documentinformation and explain the scope of the review such as electronic search criteria used (e.g., key words) or thegeneral extent of manual searches per Section 5.3.6.4 of LLI-11. NOTE: Ensure that electronic andmanual searches are performed using controlled copIes of documents. If you have any questions,contact your site Licensing department.

LBDs/DocumAnts reviewed via keyword coarch: Keywords:

FSAR. Operaling License Manual arecirc', 'flow control valve", "loop startup",.min ed"


D. Is the validity of this Review dependent on any otherchange? (See Section 5.3.4 of the EOI 10CFR50.59 ProgramReview Guidelines.)


o1 YesEl No

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If any of the following questions Is answered 'yes," an Environmental Review must be performedin accordance with NMM Procedure EV-115, "Environmental Evaluations," and attached to this50.59 Review. Consider both routine and non-routine (emergency) discharges when answeringthese questions.


Yes No

1. 0 El Involve a land disturbance of previously disturbed land areas in excess of one acre (i.e.,grading activities, construction of buildings, excavations, reforestation, creation or removalof ponds)?

2. 0 E0 Involve a land disturbance of undisturbed land areas (i.e., grading activities, construction,excavations, reforestation, creating, or removing ponds)?

3. 0 E0 Involve dredging activities in a lake, river, pond, or stream?

4. 0 E Increase the amount of thermal heat being discharged to the river or lake?


6. 0 [0 Discharge any chemicals new or different from that previously discharged?

7. El 0 Change the design or operation of the intake or discharge structures?


9. 0 0 Modify the design or operation of the plant that will change the path of an existing waterdischarge or that will result in a new water discharge?

10. 0 0l Modify existing stationary fuel burning equipment (i.e., diesel fuel oil, butane, gasoline,propane, and kerosene)?'

11. 0 Wl Involve the installation of stationary fuel burning equipment or use of portable fuel burningequipment (i.e., diesel fuel oil, butane, gasoline, propane, and kerosene)?'

12. 0 0 Involve the installation or use of equipment that will result in an air emission discharge?

13. El 0 Involve the installation or modification of a stationary or mobile tank?


15. E 0 Involve burial or placement of any solid wastes in the site area that may affect runoff,surface water, or groundwater?

'See NMM Procedure EV.117. Air Emissicns Management Program.' for guidance in answqring this quQstion.

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If any of the following questions is answered "yes," a Security Plan review must be performed bythe Security Department to determine actual impact to the Plan and the need for a change to thePlan.


Yes No

1. 0 0 Add, delete, modify, or otherwise affect Security department responsibilities (e.g., includingfire brigade, fire watch, and confined space rescue operations)?

2. 0 El Result In a breach to any security barrier(s) (e.g., HVAC ductwork, fences, doors, walls,ceilings, floors, penetrations, and ballistic barriers)?

3. 0 m Cause materials or equipment to bo placed or inotolled within the Security Isolation Zone?

4. 0 El Affect security lighting by adding or deleting lights, structures, buildings, or temporaryfacilities?


6. 0 0D Modify or otherwise affect the operation or field of view of the security cameras?

7. 0 E0 Modify or otherwise affect (block, move, or alter) installed access control equipment,intrusion detection equipment, or other security equipment?

8. 0 0l Modify or otherwise affect primary or secondary power supplies to access controlequipment, intrusion detection equipment, other security equipment, or to the CentralAlarm Station or the Secondary Alarm Station?

9. aJ 0 Modify or otherwise affect the facility's security-related signage or land vehicle barriers,including access roadways?


The Security Department answers the following questions if one of the questions was answered"yes".

B. Is the Security Plan actually impacted by the 0 Yesproposed activity? 0 No


0 No

Name of Security Plan reviewer (print) I Signature / Date

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V. 50.59 EVALUATION EXEMPTION

Enter this section only if a "yes" box was checked in Section IL.A, above.

A. Check the applicable boxes below. If any of the boxes are checked, a 50.59 Evaluation Is notrequired. If none of the boxes are checked, perform a 50.59 Evaluation in accordance withSection VI. Provide supporting documentation or references as appropriate.

0 The propoaed activity meets all of tle foluwii 1criteria regarding design function per Section5.6.1.1:

The proposed activity does not adversely affect the design function of an SSC as describedin the FSAR; AND

The proposed activity does not adversely affect a method of performing or controlling adesign function of an SSC as described in the FSAR; AND

The proposed activity does not adversely affect a method of evaluation that demonstratesintended design function(s) of an SSC described in the FSAR will be accomplished.

0 An approved, valid 50.59 Roviow(c) covcring associated aspects of the proposed activityalready exists per Section 5.6.1.2. Reference 50.59 Evaluation # (if applicable)or attach documentation. Verify the previous 50.59 Review remains valid.

o The NRC has approved the proposed activity or portions thereof per Section 5.6.1.3.Reference:

o The proposed activity is controlled by another regulation per Section 5.6.1.4.

B. Basis

Provide a clear, concise basis for determining the proposed activity may be exempted such that a third-partyreviewer can reach the same conclusions. See Section 5.6.6 of the EOI 1 OCFR50.59 Review ProgramGuidelines for guidance.

-

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ER-GG 2003-0152 000 increcoes the recirc system flow control valve MIN ED position from 7% to12%.

Reason for proposed Change:As described in CR-GGN-2003-00352, plant startups have been adversely affected by the recircflow control valves sticking after recirc pump upshift. An increase in the MIN ED position will allowthe recirc pumps to be upshifted with the valves in a more open position which will alleviate thisproblem.


This 50.59 Evaluation concludes that the oniv analysis impacted bv the chan e to MIN ED is theidli lnnn ntrtrin Thik anailvsi- has been revised with approved method, I remains a non-limiting event. No fission product barriers are impacted since no fuel failure is calculated to occurfor this event. A detailed evaluation in the reviewed document concludes that this change will notadversely impact the recirculation system. cause an unplanned scram, or violate fuel requirements.Consequently. this change does not increase the frequency of occurrence of accidents or SSCmalfunctions previously evaluated in the FSAR. Since there is no fuel failure, there are noincreases in the consequences of accidents or SSC malfunctions previously evaluated in the FSAR.

- -

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D. License Amendment Determination

Does the proposed Change being evaluated represent a change to a method of 3 Yesevaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only E NoQuestion 8. If "No," answer all questions below.


1. Result in more than a minimal increase in the frequency of occurrence of an 01 Yesaccident previously evaluated in the FSAR? El No

BASIS:The only FSAR event that is impacted by an increase in MIN ED is the idle loop startup (ILS),described in FSAR Section 15.4.4. This event involves the startup of a second cold recirc loopwith an accompanying operator error. The increased FCV position when the idle loop is startedresults in a faster influx of cold water into the reactor and consequently a larger reactivity-inducedpower swing. The ILS is classified as a moderate frequency event, which may occur at afrequency of once per calendar year to once every 20 years per FSAR Section 15.0.3.1.The increase in MIN ED would not impact the number of times an idle loop would need to bestarted nor the potential for an operator error in starting this loop. Consequently, this MIN EDchange will not result in an increase in the frequency of occurrence of an accident previouslyevaluated in the FSAR.

2. Result in more than a minimal increase in the likelihood of occurrence of a 3 Yesmalfunction of a structure, system, or component important to safety previously 0l Noevaluated in the FSAR?BASIS:ER-GG-2003-0152 evaluated the impact of this change on the recirculation system, cause anunplanned scram, or violate fuel requirements. For the recirc system, the normal and structuralloads on the jet pumps were found to be unaffected since the maximum flow through the jetpumps is not increased. The recirc pump motor was found to have only a small increase in BrakeHorsepower and winding heating which would have no appreciable affect on motor life orreliability. The neutron flux transient generated by the pump upshift was calculated to not exceed80% power, thereby maintaining significant margin to the flux scram setpoint at 118% power. Thewidth of the neutron flux spike generated by the pump upshift is predicted to be sufficiently smallthat a thermal power scram would also not occur. The ILS event has been re-analyzed andconfirmed to meet the applicable fuel thermal requirements.On these bases, this ER evaluation concludes that this change will not adversely impact therecirculation system, cause an unplanned scram, or violate fuel requirements. Therefore, thisMIN ED change will not increase the likelihood of occurrence of a malfunction of a SSC importantto safety previously evaluated in the FSAR.

S. Result in more than a minimal increase in the consequaenres of an accidrent 0 Yespreviously evaluated in the FSAR? E0 No

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BASIS:As described in the Basis to Question 1, the only FSAR event that is impacted by an Increase inMIN ED is the ILS. Per FSAR Section 15.4.4, the ILS was found to be a non-limiting event,whose results met the MCPR safety limit and LHGR overpower criteria.

As described in detail in ER-GG-2003-0152, the ILS event has been re-analyzed with theincreased MIN ED. This evaluation concluded that the MCPR safety limit and LHGR overpowercriteria are still not violated in an ILS. Therefore, fuel failures are still not predicted to occur as aresult of this change in the event of an ILS.

With no fuel failures, there is no radiological release associated with the ILS. Consequently, thisMIN ED change will not result in an increase in the consequences of an accident previouslyevaluated in the FSAR.

4. Resilt in morn than a minimal incrn;sA in the consequences of a malfunction of 0 Yesa structure, system, or component important to safety previously evaluated in the E NoFSAR?

BASIS:The FSAR analyses postulate a number of different malfunctions of the recirc system that may beimpacted by the flow control valves. Specifically, these failures include fast opening, fast closing,and line break.

The increase in MIN ED does not impact the rate at which the FCV can open or close such thatthe flow-dependent reload analyses are not impacted. In the LOCA dose analysis, the sourceterm release is governed by the non-mechanistic NRC guidance in Reg Guide 1.183, which Isindependent of FCV position at the onset of the event. In the ECCS LOCA performance analysis(FSAR Section 6.3.3), the break area is governed by the flow area at the jet pump nozzles and isnot impacted by the position of the FCV.

Consequently, this MIN ED change will not result in an increase in the consequences of amalfunction of a SSC important to safety previously evaluated in the FSAR.

5. Create a possibility for an accident of a different type than any previously 0 Yesevaluated in the FSAR? E0 No

BASIS:

The evaluation in ER-GG-2003-0152 concludes that this change will not adversely impact therecirculation system, cause an unplanned scram, or violate fuel requirements. Therefore, thisMIN ED chango will not croato tho possibility for an accident of a different type than previouslyevaluated in the FSAR.

6. Create a possIbIlIty for a malfunction of a structure, system, or component 0 Yesimportant to safety with a different result than any previously evaluated in the El NoFSAR?

BASIS

No new SSCs are required and no new operational modes or failure modes of the FCV areintroduced with this MIN ED change. The impacts on the recirc system have been evaluated inER-GG-2003-0152 arid concluded lu be acceptable. Therefore, this MIN ED change wIll notcreate the possibility for a malfunction of a SSC important to safety with a different result than anypreviously analyzed in the FSAR.

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7. Result in a design basis limit for a fission product barrier as described in the D YesFSAR being exceeded or altered? 0M No

BASIS:The fission product barriers potentially impacted by this change are the fuel cladding and reactorcoolant system. ER-GG-2003-0152 concludes that this change will not adversely impact therecirculation system or violate fuel requirements. Therefore, this MIN ED change will not result ina design basis limit for a fission product barrier as described in the FSAR being exceeded oraltered.

8. Result in a departure from a method of evaluation described in the FSAR used in 03 Yesestablishing the design bases or in the safety analyses? El No

BASIS:SAR 15.4.4.3.1 describes the analytical model used to evaluate the ILS event during Cycle 1. GEapplied their point kinetics code called REDY to confirm that this event was non-limiting for theoperating limits. Per SAR Table 1 5D.4-1, this same methodology was applied when GE re-analyzed this event for the GGNS Maximum Extended Operating Domain (MEOD).The current GGNS fuel vendor, Framatome, has performed the ILS evaluation supporting thisincrease in MIN ED. This evaluation applied Framatome's one-dimensional transient code, calledCOTRANSA2. This methodology has been used for other events at GGNS, is part of the GGNSCOLR methods in Technical Specification 5.6.5, and has been NRC-approved for analysis of theILS transient (letter: S Richards (NRC) to J.F. Mallay (Siemens), dated May 31, 2000).This re-analysis is conservative in that the results are more severe (i.e., larger ACPR) since theincreased FCV position during the pump startup results in a faster influx of cold water into thereactor and consequently a larger reactivity-induced power swing.Since the elements of the method described in the FSAR were updated in a manner to generatemore conservative results and the new methodology is NRC-approved for this application, thisMIN ED change does not represent a departure from a method of evaluation described in theFSAR.


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I. OVERVIEW I SIGNATURES 0 '

Facility: Grand Gulf Nuclear Stationt

Document Reviewed: Calculalion XC-01 N1 -94004

System Designator(slDescrIption: Ni1

Description of Proposed Chanae

_ ze210,3

D 5r SR M 8/I/Zo 4023°3

PAT~E". 83/21/Zco3

Change/Rev. -&.

This calculation undates the otfsite and control room doses from a main steam line break outsidecontainment.

If the proposed activity, In Its entirety, Involves any one of the criteria below, check the appropriatebox, provide a Justificatlon/basis In the Description above, and forward to a Reviewer. No further50.59 Review Is required. If none of the criteria Is applicable, continue with the 50.59 Review.

O The proposed activity is editorial/typographical as defined in Sectlon 5.2.2.1.

O The proposed activity represents an MFSAR-only change as allowed in Section 5.2.2.2(Insert item # from Section 5.2.2.2).

If further 50.59 Review ls required, check the applicable review(s): (Only the sections Indicated mustbe Included In the Review.)

O SCREENING Sections l, II, il, and IV required

O 50.59 EVALUATION EXEMPTION Sections l, ll, iII, IV, and V required

- 50.59 EVALUATION (#: 9 64-9 -04040-2, 7 ) , Sections 1, 11, III, IV, and VI required

G.E. Broadbent/ (i E n -IA orePreparer:

Reviewer:

OSRC:

-

Name (print) / Signature / Company/ Department / Date

Tc C4 Cgo ber-j-' Chairman's Name (print) I Signature I Date

[Required only for Programmatic Exclusion Screenings (see Section 5.8) and 50.59 Evaluations.]

List of Assisting/Contributing Personnel: NLCName: ScoDe of Assistance: i-- D- _

1 EN-S NUCLEAR OUALITYRELATED Li-10i Revision 3Enter* y MANAGEMENT APUINSTRATWE

n~e~z MANUAL. _INFORMATION USE

ATTACHMENT 9.1 50.59 REVIEW FORM Pae 2 of 11

11. SCREENING

A. LIctnaing Basis Document Revieyr

1. Does the proposed activity impact the facility or a procedure as described in any of thefollowing Licensing Basis Documents?

Operating License YES NO CHANGES and/or SECTIONS IMPACTED

Operating Ucensa a 0TS 0 151

NRC Orders 0 1m1

If 'YES, obtain NRC approval prior to Implementing the change by Initiating an LSD change In accordance with NMM LI-U13(Refernce 2.2.13) (S Section S.1.13 or exceptions.)

LEDs controlled under 50.59 YES NO CHANGES (If applicable) end/or SECTIONS IMPACTED

FSAR _~l 0 LDC 2003-059

TS Bases m l

Technlcal Requirerntsn Manuel D mCore Operating ULimts Report a 1

NRC Salety EvaluatIon Reporbt a mn "YES', pertorrn an Exemption Review per Section V QB perform a 50.59 Evaluation per Section VI M2 Initiate an LBDchange In accordance with NMM L-113 (Reference 2.2.13),

LEDc controlled under other regulations YES NO CHANGE U (If applicable) and/or SECTIONS IMPACTED

Quality Assurance Program Manualf 0 1 O

Emergency Planer? 3 C

Fire Protection Prograrm 0 l(includes the Fire Hazards Analysis)

Offsite Dose Calculations Manual 0 1ID

It 'YES", evaluate any changes In accordance with the appropriate regulation PND Initiate an LSD change In accordance withNMM LI-113 (Reference 2.2Y13)

2. Does the proposed activity Involve a test or experiment not described In theFSAR?If "yes," perform an Exemption Review per Section V OFl perform a 50.59Evaluation per Section VI.

3. Does the proposed activity potentially Impact equipment, procedures, orfacililtles utilized for storing spent fuel at an Independent Spent Fuel StorageInstallation?(Check NWA" If dry fuel storage is not appilcable to the facility.)It 'yes," perform a 72.48 Review In accordance with NMM Procedure L1112.(See Sections 1.5 and 53.1.5 of the EOI 10CFRSO.59 Review Program Guidelines.)

° Yesl No

0E3l

YesNoWA

Iit YES.' see Section 5.1.4.2 if 'YES.' notify the responsible departnent and ensure a 50.54 EvaluatIon Is performed. Attach the 50.54 Evaluation.3 If YES.' evaluate the change In accordance with the requirements of the faciltys Operating Ucense Condition.

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ATTACHMENT 9.1 5 50.59 REVIEW FORM Page 3 of 11

B. Basis

Provide a clear, concise basis for the answers given In the applicable sections above. Explain why theproposed activity does or does not Impact the Operating Ucense/Technical Specifications and/or the FSARand why the proposed activity does or does not Involve a new test or experiment not previously described Inthe FSAR. Adequate basis must be provided within the Screening such that a third-party reviewer can reachthe same conclusions. Simply stating that the change does not affect TS or the FSAR Is not an acceptablebasis. See EOi 50.59 Guidelines Section 5.6.6 for guidance.)

This calculation updates the offsihe and control room doses from a main steam line break ousidecontainment. The requirements in the Tech Soecs. TRM. and Operating License are inpute to theanalysis and are not affected by the calculation. NRC Security Orders are also not Impacted. TheTS Bases do not describe the radioloaical consequences of this event. The COLR Is unaffectedsince this event Is bounded by the recirc line break which is addressed by the current limits in theCOLR. This event is described In SAR Sedion 15.6.4 to demonstrate that the applicable reaulatoaccentance criteria are satisfied. This event Is not affected bv the OA manual. Emeroencv Dlan.Fire Protection prooram. or ODCM.

C. References

Discuss the methodology for performing the LED search. State the location of relevant licensing documentinformation and explain the scope of the review such as electronic search criteria used (e.g.. key words) or thegeneral extent of manual searches per Section 5.3.6.4 of 1-1I 01. NOTE: Ensure that electronic andmanual searches are performed using controlled copies of documents. tf you have any questions,contact your site Licensing department.

LBDs/Documents reviewed via keyword search:

FSAR. Operating License Manual

Keywords:

"steamline break, 'radiological dose'

LBDs/Documents reviewed manually.

0. Is the validity of this Review dependent on any otherchange? (See Section 5.3.4 of the E0I 10CFR50.69 ProgramReview Guidelines.)

If Yes:, list the required changes.

o1 YesE No

EN-S NUCLEAR QuALrTY RELATED LI-101 RevisIon 3& En t ADMINSTRATIVElg,.e,~ MANAGEMENTADITAIVgy ", 'MANUAL INUSEMAION*US

ATTACHMENT 9.1 50.59 REVIEW FORM Page 4 'of 1

ill. ENVIRONMENTAL SCREENING

If any of the following questIons Is answered "yes," an Environmental Review must be performedIn accordance with NMM Procedure EV-115, "Environmental Evaluations," and attached to this50.59 Review. Consider both routine and non-routine (emergency) discharges when answeringthese questions.


Yes No

1. 3 0E Involve a land disturbance of previously disturbed land areas in excess of one acre (i.e.,grading activities, construction of buildings, excavations, reforestation, creation or removalof ponds)?


3. 03 0 Involve dredging activities In a lake, river, pond, or stream?

4. 03 0 Increase the amount of thermal heat being discharged to the river or lake?


6. 0 - -- 0 Discharge any chemicals new or different from that previously discharged?


B. 03 0 Modify the design or operation of the cooling tower that will change water or air flowcharacteristics?

9. 03 0 Modify the design or operation of the plant that will change the path of an existing waterdischarge or that will result In a new water discharge?

10. 03 0 Modify existing stationary fuel burning equipment (i.e., diesel fuel oil, butane, gasoline,propane, and kerosene)?

11. 03 0 Involve the Installation of stationary fuel burning equipment or use of portable fuel burningequipment (i.e., diesel fuel oil, butane, gasoline, propane, and kerosene)?'

12. 0 0 Involve the installation or use of equipment that will result in an air emission discharge?


14. 0 0i Involve the use or storage of oils or chemicals that could be directly released into theenvironment?

15. 03 0 Involve burial or placement of any solid wastes In the site area that may affect runoff,surface water, or groundwater?

See NMM Procedure EV-117, 'Air Errissions Managemnent Program,' or guidance In answering his question.


if any of the following questions Is answered "yes," a Security Plan review must be performed bythe Security Department to determine actual Impact to the Plan and the need for a change to thePlan.


Yes N-o

1. 0 El Add, delete, modify, or otherwise affect Security department responsibilities (e.g., Includingfire brigade, fire watch, and confined space rescue operations)?

2. 0 El Result In a breach to any security barrier(s) (e.g., HVAC ductwork, fences, doors, walls,ceilings, floors, penetrations, and ballistic barriers)?

3. 0 El Cause materials or equipment to be placed or Installed within the Security Isolation Zone?


5. 0 81 Modify or otherwise affect the Intrusion detection systems (e.g., E-fields, microwave, fiberoptics)?

6. 0 El Modify or otherwise affect the operation or field of view of the security cameras?

7. 0 El Modify or otherwise affect (block, move, or alter) installed access control equipment,intrusion detection equipment, or other security equipment?

8. 0 M1 Modify or otherwise affect primary or secondary power supplies to access controlequipment, Intrusion detection equipment, other security equipment, or to the CentralAlarm Station or the Secondary Alarm Station?

9. 0 El Modify or otherwise affect the facility's security-related signage or land vehicle barriers,Including access roadways?

10. 0 E Modify or otherwise affect the facilitys telephone or security radio systems?

The Security Department answers the following questions If one of the questions was answered"yes".

B. Is the Security Plan setually impacted by the 0 Yesproposed activity? 0 No


0 No

Name of Security Plan reviewer (print) I Signature / Date

. EN-S NUCLEAR QUAUTY RELATED LI-101 Revision 3

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INFORMATION USE



Enter this section only if a "yes" box was checked In Section II.A, above.

A. Check the applicable boxes below. If any of the boxes are checked, a 50.59 Evaluation Is notrequired. It none of the boxes are checked, perform a 50.59 Evaluation In accordance withSection VI. Provide supporting documentation or references as appropriate.

D The proposed activity meets all of the following criteria regarding design function per Section

The proposed activity does not adversely affect the design function of an SSC as describedin the FSAR; AND

The proposed activity does not adversely affect a method of performing or controlling adesign function of an SSC as described In the FSAR; AND

The proposed activity does not adversely affect a method of evaluation that demonstratesintended design function(s) of an SSC described in the FSAR will be accomplished.

O An approved, valid 50.59 Review(s) covering associated aspects of the proposed activityalready exists per Section 5.6.1.2. Reference 50.59 Evaluation # (if applicable)or attach documentation. Verify the previous 50.59 Review remains valid.


O The proposed activity is controlled by another regulation per Section 5.6.1.4.

B. BasisProvide a clear, concise basis for determining the proposed activity may be exempted such that a third-partyreviewer can reach the same conclusions. See Section 5.6.6 of the EOI I 0CFR50.59 Review ProgramGuidelines for guidance.


A. Executive Summary (Serves as Input to NRC summary report. Limit to one page or less. Send anelectronic copy to the site licensing department after OSRC approval. If available.)


Calculation XC-QIN1I-94004, Rev. 0 updates the offsite and control room radiological dosesassociated with a main steam line break outside containment. This 50.59 evaluation addresses thechanges made to the inputs of the analysis and a new dose methodology. The input chancesinclude higher reactor coolant source term concentrations and uodated dose conversion factors anddiffusion factors (Y/Qs), which have been-approved by the NRC in association with the alternativesource term licensing effort This calculation was performed with a new dose methodology calledRAPTOR. which has been determined to be essentially the same as Previous NRC-approvedmethods via benchmark evaluations.

Reason for proposed Change:A calculation revision is necessary to correct the deficiencies identified in CR-GG-2003-01876,and to include revised atmosoheric diffusion factors as tracked under LO-GLO-2001-00034. CA-12. As described in CR-GG-2003-01876. the original Bechtel calculation was found to be deficientIn that it did not (1) address the control room dose although GDC-19 reported explicit re ulatoryacceptance criteria for the control room doses for accidents. nor (2) consider this worst caseiodine spiking scenario (4 uClia DE 1-131) by only evaluating the equilibrium Iodine case (0.2 uCVaDE 1-131).


This 50.59 Evaluation concludes that, based on a comparison to the original Bechtel calculation,this revised calculation does not increase the conseguences of an accident previously evaluated Inthe FSAR. The computer code and Inputs applied In the analysis were found to not be a departurefrom a method of evaluation described in the FSAR used in establishing the design bases or in thesafety analyses. Since this calculation does not Impact any SSC or plant procedure. this changedoes not Impact fission product barriers, or Increase the freguencv of occurrence of accidents orSSC malfunctions previously evaluated in the FSAR. or create the possibility for an accident ormalfunction of a different type than Previously evaluated in the FSAR.

EN-S NUCLEAR QuALrT RELATED i-101 Revision 3Enteg MANAGEMENT ADMINISTRATIVE

j INFOLRMATON USE



Does the proposed Change being evaluated represent a change to a method ofevaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer onlyQuestion S. If "No," answer all questions below.

00

YesNo


1. Result In more than a minimal Increase In the frequency of occurrence of anaccident previously evaluated In the FSAR?

o YesIxi No

BASIS:This calculation updates the offsite and control room doses associated with a main steamlinebreak outside containment and does not impact any SSC or plant procedure. Since thiscalculation only Impacts the plant analysis, it will not impact the frequency of occurrence of anaccident previously evaluated In the FSAR.

2. Result in more than a minimal Increase In the likelihood of occurrence of amalfunction of a structure, system, or component Important to safety previouslyevaluated in the FSAR?

o Yes0 No

BASIS:This calculation updates the plant dose analysis. No changes are being made to any SSC orplant procedure. Therefore, this calculation will not impact the likelihood of occurrence of amalfunction of a structure, system, or component important to safety previously evaluated in theFSAR.

3. Result In more than a minimal increase In the consequences of an accidentpreviously evaluated In the FSAR?

o Yes0 No

BASIS:The radiological consequences of a main steamline break outside containment are reported InSAR Section 15.6.4 as developed In Bechtel Calculation 5.3.27. As described In CR-GG-2003-01876, this Bechtel calculation was found to be deficient in that it did not (1) address the controlroom dose although G DC-1 9 reported explicit regulatory acceptance criteria for the control roomdoses for accidents, nor (2) consider this worst case Iodine spiking scenario (4 uCVg DE 1-13 1) byonly evaluating the equilibrium iodine case (0.2 uCVg DE 1-131). To assess whether this changerepresents a significant increase in consequences, the new dose results are compared to theoriginal Bechtel results In the following table.

Equ1librium Iodine Case (0.2 uCV DE 1-131)Exclusion Area B oundary Control Room

Thyroid Whole TEDE Thyroid Whole TEDE_ _ _ _Body Body

Bechtel Calculation 582 55E2 Not Not5.3.27 5. _ _ 5.5E 2 Calc'd. Calc'd.

Updated Calculation 2.43 3.8E-2 1.1 9E-1 4.55 2.99E-3 1.53E-1

QuAuTY RELATED L-101 Revision 3ADmNISThIRAVE

IbfORMAMiON USE

IATTACHMENT 9.1 50.59 REVIEW FORM Page 9 _ 1

As shown above, the only available case for comparison Is the equilibrium iodine case at the EABlocation. The new calculation reports EAB thyroid and whole body doses of 2.43 and 3.8E-2Rem, respectively which are less than the original Bechtel results of 5.82 and 5.5E-2 Rem,respectively. The only available direct comparison therefore shows a reduction In consequences.The analysis Input changes that caused this dose reduction are addressed in Question #8 sincethey relate to the methodology.

As described In CR-GG-2003-01876, the original Bechtel analysis did not calculate a control roomdose. This oversight on Bechtel's part does not mean the control room dose was zero, only It wasnot quantified. Based on the offsite comparison performed above, the control room would also beexpected to see a similar trend, with a reduction in the calculated dose from the Bechtelcalculation.

CR-GG-2003-01876 also reports that Bechtel did not evaluate another more-limiting case, Inwhich iodine spiking is assumed to put the coolant iodine concentration at the maximum allowedTechnical Specification limit of 4.0 jiCVg Dose Equivalent 1-1 31. As shown In the following table,the new analysis demonstrates that this case meets the NRC acceptance criteria. The resultsfrom this more-limiting case with higher doses will be added to the FSAR representing a separatecase from the equilibrium case currently reported in the FSAR. Based on the offsite comparisonperformed above, these results would be expected to be lower than the doses calculated byBechtel, if Bechtel had generated them.

Iodine Spiking Case (4.0 uCIsg DE 1-131)| Exclusion Area Boundary Control Room

Thyroid Whole TEDE Thyroid Whole TEDEBodyBodyv___

Bechtel Calculation Not Not Not Not5.3.27 Calc'd. Calc'd. Calc'd. Calc'd.Updated CaCculation 477 7.21E-1 2.32 89.2 5.88E-2 3.01

10 CFR 50.67 _ __Acceptance Limit 2 5.0

In summary, for the cases In which a comparison can be made, the new analysis reports lowerdoses than the previous Bechtel analysis. This trend would extend to the new dose point (controlroom) and the worse scenario (iodine spiking) such that the new doses would be less than theassociated Bechtel results, if Bechtel had generated them. On these bases, the results of thenew MSLB radiological analysis do not represent an increase in consequences of an accidentpreviously evaluated in the FSAR.

4. Result in more than a minimal Increase In the consequences of a malfunction of 01 Yesa structure, system, or component Important to safety previously evaluated In the Is] NoFSAR?BASIS:This calculation updates the radiological analysis of the main steam line break outsidecontainment. Consequently, this change will not affect the consequences of a malfunction of aSSC important to safety previously evaluated In the FSAR.

5. Create a possibility for an accident of a different type than any previously 0 Yesevaluated in the FSAR? Et No

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ATTACHMENT 9.1 60.59 REVIEW FORM Page 10 Of 11

BASIS:

This calculation updates the plant dose analysis. No changes are being made to any SSC orplant procedures. Therefore, this calculation will not create the possibility for an accident of adifferent type than previously evaluated in the FSAR.

6. Create a possibility for a malfunction of a structure, system, or componentImportant to safety with a different result than any previously evaluated In theFSAR?

0lIm

YesNo

BASISThis calculation updates the plant dose analysis. No changes are being made to any SSC orplant procedures. Therefore, this calculation will not create the possibility for a maltunction of aSSC important to safety with a different result than any previously analyzed In the FSAR.

7. Result in a design basis limnt for a fission product barrier as described in theFSAR being exceeded or altered?

3 YesEI No

BASIS:The fission product barriers potentially Impacted by this change are the fuel cladding and reactorcoolant system. This calculation updates the plant dose analysis and makes no changes to theRCS, MSIV closure timing, or plant procedures. On these bases, this calculation will not result Ina design basis limit for a fission product barrier as described In the FSAR being exceeded oraltered.

EN-S NUCLEAR OUALITY RELATED LI-101 Revision 3Enter ANAGEMENT ADMINISTRATIVEMANUAL

.I INFORMATION USE


B. Result in a departure from a method of evaluation descnbed in the FSAR used in 3 Yesestablishing the design bases or in the safety analyses? I No

BASIS:MethodoloavThe dose analysis methodology Is not described in the FSAR In any detail. The computer codeoriginally applied by Bechtel In the analysis of this event was called @ PUFF. The GGNScomputer code applied for the updated calculation is RAPTOR. As documented in EngineeringReport GG-SA-2003-0002, RAPTOR has been benchmarked to the NRC's RADTRAD code andto the TRANSACT computer code used for the GGNS alternative source term (AST) analyses.Since RAPTOR has been shown to provide results that are essentially the same as anothermethodology previously accepted by the NRC, the use of the RAPTOR computer code Is not adeparture from a method of evaluation described in the FSAR used In establishing the designbases or In the safety analyses.

Elements of the MethodologyAs discussed In the response to Question 3, the results of the dose analysis decreased with thenew analysis. Per the EOD 50.59 guidelines, this change could be construed as 'nonconservative' such that margin to the acceptance criteria Is gained via the re-analysis. However,the dose reduction Is due to changes to the code inputs as a result of the AST analyses, whichwere NRC-approved In GNRI-2001/00032.The three significant differences that resulted in the reduced EAB dose results are the (i) FGR-11basis for reactor coolant Iodine source terms, (ii) FGR-11 dose conversion factors, and (ii)updated diffusion factors. Each of these inputs were specifically submitted by GGNS andreviewed and approved by the NRC in association with the AST licensing efforts. The newcalculation merely updates the analysis based on the NRC-approved inputs.On these bases, although the results change in the non-conservative direction, this new analysisdoes not represent a departure from the method of evaluation described in the FSAR used in thesafety analyses since It applies NRC-approved changes to the elements of the analysis.


SE 2004-0001-ROO

I. OVERVIEW I SIGNATURES

Facility: GGNS

Document Reviewed: Core Operating Limits Report (COLR)

System Designator(s)IDescriptlon: J1i1

Description of Proposed Channe

ChangelRev. LDC2004-009

This evaluation addresses the Cycle 14 reload changes and operation of the Cycle 14 reload core as givenin the Core Operating Limits Report (COLR).

If the proposed activity, in Its entirety, Involves any one of the criteria below, check the appropriatebox, provide a Justlficationlbasis in the Description above, and forward to a Reviewer. No further50.59 Review Is required. If none of the criteria Is applicable, continue with the 50.59 Review.

o The proposed activity is editorial/typographical as defined in Section 5.2.2.1.

o The proposed activity represents an 'FSAR-only" change as allowed in Section 5.2.2.2(insert Item I trom Section 5.2.22).

If further 50.59 Review Is required, check the applicable review(s): (Only the sections Indicated mustbe included In the Review.)

o SCREENING Sections I, II, l1l, and IV required

o3 50.59 EVALUATION EXEMPTION *9 , os/ Sections I, II, 1II, IV, and V required

l 50.59 EVALUATION (#: 5 Sections l, II, II, IV, and VI required

_ _5E-zoo41 - 0oo/-2OO .

Preparer: Guy B. Spikes/ X)2 d. LG. /EOI/Nucl. Eng. - SA 2 - q - vcfName (print) / Signature I Co pany / Department I Date

Reviewer A,' <52214eiA-5 ___ <-^ , - //Mlov.77Name (print) / Signa 7r -VepaKEnt gate -

A-k, Kt 75 bet Io q

.

.;,- -/

OSRC: 67;z0czfm-4~Chairman's Name (print) I Signa1re I Date .

[Required only for Programmatic Exclusion Screenings (see Section 5.8) and 50.59 Evaluations.)

List of AccictinglContributing Pcrsonnol:Name: Scope of Assistance:

J. A. Elam (Central Ennineerina BWR Fuels) Core desian and neutronic InPut

.41

D. L. Smith (Central Engineering DWR Fuels)J. P. Head (Central Engineering BWR Fuels)G. W. Smith (GGNS-PSA)

Fuel mechanjical InputCore stability and hydraulic compatibility inputEOP Input

EN-S NUCLEAR QUAUy RELATED LI-01 Revision 3

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II. SCREENING




Operating License 0 H

TS 0 H

NRC Orders 0 H

If 'YES", obtain NRC approval prior to implementing the change by Initiating an LBD change In accordance with NMM U-113(Reference 2.2.13). (See Section 5.1.13 for exceptions.)


FSAR H 0 LDC 2004-010

TS Bases 0 El

Technical Requirements Manual H E LDC 2004-012

Core Operating Limits Report H 0 LDC 2004-009

NRC Safety Evaluation Reports' 1 0

If 'YES", perform an Exemption Review per Section V OR perform a 50.59 Evaluation per Section VI AND Initiate an LBDchange in accordance with NMM U-113 (Reference 2.2.13).

LBDs controlled under other regulations YES - -NO CHANGE # (if applicable) and/or SECTIONS IMPACTED

Qualty Assurance Program Manual' O H

Emergency Plan2 0 H

Fire Protection Program' 0(includes the Fire Hazards Analysig)

Offste Dose Calculations Manual E X

If 'YES', evaluate any changes In accordance with the appropriate regulation AND Initiate an LSD change In accordance withNMM Li-113 (Reference 2.2.13).


3. Does the proposed activity potentially Impact equipment, procedures, orfacilities utilized for storing spent fuel at an Independent Spent Fuel StorageInstallation?(Check IN/A" If dry fuel storage i3 not applicable to the facility.)If "yes," perform a 72.48 Review In accordance with NMM Procedure LI-112.(See Sections 1.5 and 5.3.1.5 of the EOI 10CFR50.59 Review Program Guidelines.)

O YesE No

o YesO NoEZ NIA

'If "YES.' see Section 5.1.4.2 If YES,' notify the responsible department and ensure a 50.54 Evaluation is performed. Attach the 50.54 Evaluation.3If YES.' evaluate the change in accordance with the requirements of the facility's Operating License Condition.

A" EN-S NUCLEAR QUALary RELATED L4l01 Revision 3MANAGEMENT ADMINISlIAmV_:`En)terg MAUA

INFORMATION USE


B. Basis

Provide a clear, concise basis for the answers given in the applicable sections above. Explain why theproposed activity does or does not impact the Operating LicensefTechnical Specifications and/or the FSARand why the proposed activity does or does not involve a new test or experiment not previously described Inthe FSAR. Adequate basis must be provided within the Screening such that a third-party reviewer can reachthe same conclusions. Simply stating that the change does not affect TS or the FSAR is not an acceptablebasis. See E0I 50.59 Guidelines Section 5.6.6 for guidanc .)The current MCPR Safety Limit has been shown to be applicable to the Cycle 14 core. As such,Tech Spec 2.1.1.2 does not need to be revised.

Tech Spec 3.4.11 includes exposure-dependent PIT limit curves. The cumulative reactor vesselservice life (EFPY) is expected to increase beyond the applicability of the P/T limit curve currentlyin use (L1 6 EFPY) approximately 5 months after BOC14. The TRM will need to be updated toreflect the change in applicable PJT limit curve in TS 3.4.1 1.

There are no other Tech Spec, Bases, or TRM changes required for Cycle 14 startup and initialoperation. There are no NRC orders applicable to the Cycle 14 reload campaign.

The Cycle 14 core will contain FANP 9x9-5 reinsert fuel bundles last used in Cycle 11 and the corecharacteristics and response will also be somewhat different than currently described in the FSAR.As such, Cycle 14 analyses have been performed for the new core and the FSAR will need to beupdated appropriately as will the COLR.

The Cycle .14 core design and operation will not affect the OCDM, QAPM, Emergency Plan,Security Plan, Fire Protection Program, or any NRC SERs.

C. References

Discuss the methodology for performing the LBD search. State the location of relevant licensing documentinformation and explain the scope of the review such as electronic search criteria used (e.g., key words) or thegeneral extent of manual searches per Section 5.3.6.4 of LI-101. NOTE: Ensure that electronic andmanual searches are performed using controlled copies of documents. If you have any questions,contact your site Licensing department.

LBDs/Documents reviewed via keyword search: Keywords:

OLM. FSAR. TS Bases. TRM Fuel, reload, channel. COLR. P/T

LBDslDocuments reviewed manually:

D. Is the validity of this Review dependent on any other El Yeschange? (See Section 5.3.4 of the EOI OCFR50.59 Program 0 NoReview Guidelines.)


An acceptable final core loading. For a final core loading not exactly as provided in FAB03-1155, an evaluation of the as loaded core would need to be performed to ensure that theCycle 14 reload analyses would continue to be acceptable.

Revise the TRM prior to BOC14+5 months to reflect change In applicable TS 3.4.11 PIT limitcurve (LCTS 35421)

1 EN-S NUCLEAR QuAuTf RELATED LI-101 Revision 3En',, , jMANAGEMENT AvMINISTRATIVE

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111. ENVIRONMENTAL SCREENING

If any of the following questions is answered "yes," an Environmental Review must be performedIn accordance with NMM Procedure EV-115, "Environmental Evaluations," and attached to this50.59 Review. Consider both routine and non-routine (emergency) discharges when answeringthese questions.


Yes No

1. 0 E Involve a land disturbance of previously disturbed land areas in excess of one acre (i.e.,grading activities, construction of buildings, excavations, reforestation, creation or removalof ponds)?

2. 0 El Involve a land disturbance of undisturbed land areas (i.e., grading activities, construction,excavations, reforestation, creating, or removing ponds)?

3. 0 0E Involve dredging activities in a lake, river, pond, or stream?


5. 0 0l Increase the concentration or quantity of chemicals being discharged to the river, lake, orair?


7. El Change the design or operation of the intake or discharge structures?


9. 0 El Modify the design or operation of the plant that will change the path of an existing waterdischarge or that will result in a new water discharge?

10. 0 0l Modify existing stationary fuel burning equipment (i.e., diesel fuel oil, butane, gasoline,propane, and kerosene)?'.

11. 0 0 Involve the installation of stationary fuel burning equipment or use of portable fuel burningequipment (i.e., diesel fuel oil, butane, gasoline, propane, and kerosene)?'

12. 0 E 0 Involve the installation or use of equipment tI-at will iesult In an air emission discharge?


14. 0 E 0 Involve the use or storage of oils or chemicals that could be directly released into theenvironment?

15. 0 El Involve burial or placement of any solid wastes in the site area that may affect runoff,surface water, or groundwater?

'See NMM Procedure EV-117, 'Air Emissions Management Program,' for guidance In answerina this question.

A__ EN-S NUCLEAR QUALrTY RELATEo LI-101 Revision 3Entef yMANAGEMENT ADMINIsrAIIVE

E n MANUALINFORMAT1ON USE



If any of the following questions is answered "yes," a Security Plan review must be performed bythe Security Department to determine actual Impact to the Plan and the need for a change to thePlan.

A. Could the proposed activity being evaluated;

Yes No

1. 0 E Add, delete, modify, or otherwise affect Security department responsibilities (e.g., includingfire brigade, fire watch, and confined space rescue operations)?

2. 0 E Result In a breach to any security barrier(s) (e.g., HVAC ductwork, fences, doors, walls,ceilings, floors, penetrations, and ballistic barriers)?

3. 0 0 Cause materials or equipment to be placed or installed within the Security Isolation Zone?



6. 0 0 Modify or otherwise affect the operation or field of view of the security cameras?

7. 0 0 Modify or otherwise affect (block, move, or alter) installed access control equipment,intrusion detection equipment, or other security equipment?

8. 0 l0 Modify or otherwise affect primary or secondary power supplies to access controlequipment, intrusion detection equipment, other security equipment, or to the CentralAlarm Station or the Secondary Alarm Station?

9. 0 El Modify or otherwise affect the facility's security-related signage or land vehicle barriers,including access roadways?


The Security Department answers the following questions if one of the questions was answeredYoeI".

B. Is the Security Plan actually impacted by the 0 Yes .

proposed activity? 03 No


0 No


AftEN- NUCLEAR O UALITYRELATED 1401 RevIson 3EnteMANAGEMENT ADMINISTATrVT

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A. Executive Summary (Serves as input to NRC summary report. Limit to one page or less. Send anelectronic copy to the site licensing department after OSRC approval, If available.)

Brief description of change, test or experiment:

This safety evaluation assesses the reload-related changes associated with Cycle 14operation as presented in the Core Operating Limits Report (COLR) located in theOperating License Manual (OLM). Cycle 14 has been designed for 538 Effective FullPower Days with a core consisting of 244 fresh ATRIUM-10 assemblies, 239 once-burntATRIUM-10 assemblies, 204 twice-burnt ATRIUM-10 assemblies, 28 thrice burnt GE11assemblies, and 85 previously loaded twice-bumt FANP 9x9-5 assemblies. Eight of thefresh ATRIUM-10 assemblies are supplied with lead fuel channels manufactured by KobeSteel Ltd. There are no TS or TS Bases changes required to operate with this new core,however, the FSAR does require updates. A TRM change will also be needed later duringCycle 14 to change the applicable TS 3.4.11 P/T limit curve. The Cycle 14 core has beendesigned and analyzed for a rated thermal power of 3898 MWt. Attachment 1 provides adetailed description of the Cycle 14 reload analysis and the issues considered in thisevaluation. Increased (abnormal) channel bow described in CR-GGN-2002-01810 hasbeen explicitly considered in the Cycle 14 reload analysis for the GE11 and potentiallysusceptible high exposure ATRIUM-10 fuel types using the advanced (thick-thin) channeldesign.




The Cycle 14 core configuration and operation has been evaluated with respect tomechanical, neutronic, thermal-hydraulic, dose, thermal performance, and methodsconsiderations for GGNS. This evaluation concludes that the reload-related changesassociated with Cycle 14 operation does not require NRC review.

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Does the proposed Change being evaluated represent a change to a method of a Yesevaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only E NoQuestion 8. If "No," answer all questions below.


1. Result in more than a minimal increase in the frequency of occurrence of an accident 0 Yespreviously evaluated in the FSAR? 0 No

BASIS:

The Cycle 14 core loading and cycle operation will not result in more than a minimalincrease in the frequency of occurrence of an accident previously evaluated in theFSAR. The precursors to these events are independent of the core design and thefrequency classifications reported in FSAR Chapter 15 are unaffected by the coreparameters. The following considerations support this conclusion.

MechanicalThe ATRIUM-10 mechanical design has been reviewed for use at Grand Gulf. Nounusual failure modes or increased failure frequency have been identified for this fueldesign. This is the third reload at GGNS with ATRIUM-10 fuel and this fuel design hasaccumulated significant problem-free operational experience at other plants. TheFANP 9x9-5 fuel type was introduced at GGNS in Cycle 5 and was last used in Cycle11. The mechanical design of this fuel has been reviewed for use at GGNS in themixed (ATRIUM-10, GE11, 9x9-5) Cycle 14 core. The Cycle 14 bundles will operatewithin the power history assumptions in the fuel mechanical analyses and will result inexposures within the analyzed burnup limits of the ATRIUM-10, GE11, and 9x9-5mechanical designs. All design criteria for the GE11 bundles have been shown tomeet their respective limits including those that will be irradiated for a fourth cycle.

NuclearThe neutronic characteristics of the Cycle 14 9x9-5, GE11 and ATRIUM-10 mixedcore design have been considered in the safety analysis. Adequate shutdown marginhas been predicted by analysis and will be confirmed during startup tests. Inaddition, the hold-down capability of the standby liquid control system and thesubcriticallty of Cycle 14 fuel In the spent fuel storage racks have been confirmed.Therefore, the probability of inadvertent criticality has not been increased by theintroduction of the Cycle 14 reload fuel.

Thermal-HydraulicFANP's modeling of the GE fuel and the thermal-hydraulic compatibility oG theATRIUM-10, GE11, and 9x9-5 fuel have been reviewed and round acceptable. Toaccurately model the GE1l1 bundle hydraulics, a Cycle 11 GE1 I bundle was shippedto FANP for hydraulic testing in their hydraulic test facility. Analyses have beenperformed to demonstrate that Cycle 14 meets all Enhanced-1A stabilityperformance criteria without changes to the ElA hardware or power-flow map regionboundaries. Therefore, the probability of thermal-hydraulic instabilities has notincreased.

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Analyzed EventsThe probability of the occurrence of anticipated operational events is not dependenton the core configuration. No changes to the plant design are required for the Cycle14 core. The Cycle 14 core loading will not affect the precursors to any of theChapter 15 events. The probability of an analyzed event therefore has notincreased.

As described in FSAR Section 15A.6.5.3, the Control Rod Drop Accident (CRDA)results from a failure of the control rod-to-drive mechanism coupling after the controlrod becomes stuck in its fully inserted position. Although an increased channel bowcondition could result in increased friction between the control blade and itscorresponding fuel assemblies, analyses have shown that there would not besufficient friction to result in a mechanical failure of the coupling. Additionally, thecontrol rod drive mechanism would not produce enough force to result In amechanical failure of the coupling even if the channel bow was so severe that theassemblies would preclude blade movement. As such, channel bow is notconsidered a precursor to the CRDA, and any increased bow associated with the highexposure ATRIUM-10 bundles (the GE11 bundles are loaded in unrodded locations)would not increase the probability of this event.

On these bases, the probability of occurrence of accidents previously identified in theFSAR is not increased for the Cycle 14 core with increased channel bow.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction D Yesof a structure, system, or component Important to safety previously evaluated in the El NoFSAR?BASIS:

The Cycle 14 ATRIUM-10 reload fuel and 9x9-5 reinsert fuel are designs that havebeen shown to be mechanically, neutronically, and thermal-hydraulically compatiblewith the co-resident GE11 fuel. No plant modifications are required to accommodatethe new core design and the only additional loads placed on plant equipment wouldbe potential friction between the control blades and excessively bowed ATRIUM-10bundles (the GE11 bundles are loaded in unrodded locations). Based on previousexperience with bowed fuel at GGNS and Clinton Power Station, this Increasedfriction is not expected to impact scram times. Technical Specification scram timetesting and appropriate channel bow surveillances will be performed during Cycle 14.These actions would identify any potential scram time or other impacts and theappropriate actions would be taken. Additionally, this increased friction would not be-sufficient to provide any failures associated with the control blades or the controlblade drive system.

A conservative vessel overpressurization analysis has been performed, which showsthat the vessel pressure limit is not excccded. The precursors to any malfunction ofequipment important to safety are not affected by this change.


3. Result in more than a minimal increase in the consequences of an accident previously 0 Yesevaluated in the FSAR? E NoBASIS:

As reported in Attachment 1, the acceptance criteria reported in FSAR Section15.0.3.1 and the Technical Specifications are satisfied for each event classification.Core operating limits have been developed to ensure that moderate frequencyevents do not violate the MCPR safety limit or fuel cladding strain limits. Theconsequences of infrequent events have been shown to meet the appropriateacceptance criteria while the individual acceptance criteria for the limiting faults havebeen demonstrated to be satisfied. The following considerations support theseconclusions.Moderate Frequency EventsThe Cycle 14 core operating limits have been developed with NRC-approvedmethodologies such that the MCPR safety limit and the fuel cladding strain limit willnot be violated by any analyzed moderate frequency transient initiated from anystatepoint available to GGNS. As such, no fuel failures are expected to result fromany moderate frequency event. These analyses considered GGNS-specificoperational modes such as MEOD, SLO, FHOOS, and EOC-RPT inoperable. Thesecore operating limits consist of MCPR, MAPLHGR and LHGR curves that arefunctions of flow, power, and exposure. These limits consider conservative channelbow assumptions for the GE11 and highly exposed ATRIUM-10 fuel that bound theincreased bow previously observed In the GE11 fuel.

These core operating limits will be incorporated into the core monitoring system,however, as with previous cycles, the GE11 operating limits illustrated in the COLRwill differ somewhat from those limits in the core monitoring system. GE hasgenerated pellet-based exposure-dependent LHGR and lattice-based exposure-dependent MAPLHGR limits fortheir GE11 fuel bundles; however, for competitivereasons, GE has designated the limits for only the most-limiting or least-limitinglattices as non-proprietary. The COLR, which is submitted to the NRC for informationonly, will therefore report the non-proprietary limits for the most limiting enrichedlattice at each exposure for reference purposes only, and refer to the appropriate GEproprietary document for the actual limits. The actual LHGR and MAPLHGR limits formost lattices will be higher than the COLR reference curves and it is recognized thatmost lattices could operate at higher values than the reference limits. However, thisis acceptable since Technical Specifications 3.2.1 and 3.2.3 require that all APLHGRsand LHGRs, respectively be less than or equal to the lImits spectfied In the COLR,and the COLR refers to the proprietary GE document for the actual limits.

Infrequent EventsThe consequences of the limiting infrequent events have been evaluated and shownto meet their respective acceptance criteria. These events include the pressureregulator failure downscale. misplaced (i.e., misoriented and mislocated) bundle, andsingle loop operation pump seizure accidents. Radiological analyses using thealternative source term (AST) have been performed to ensure that these events willnot result in offsite or control room doses greater than their respective acceptancecriteria. These evaluations consider conservative channel bow assumptions for theGE11 and highly exposed ATRIUM-10 fuel that bound the increased bow previouslyobserved in the GE1 1 fuel.

Limiting FaultsThe limiting faults at GGNS include the fuel handling accident, the control rod dropaccident, and the design basis LOCA. The radiological analyses for these eventshave been developed as part of the GGNS AST effort and bound the Cycle 14 coreparameters. For the LOCA, MAPLHGR operating limits and single-loop multipliershave been developed for the Cycle 14 core configuration such that the requirementsof 10CFR50.46 are satisfied. The containment response for the Cycle 14 core wasfound to be bounded by previous cycles as is the hydrogen analysis. Theseismic/LOCA response of the Cycle 14 core has been confirmed to be acceptable.The Cycle 14 core design results in minor changes to two EP parameters (Mclad andMfuel); however, the existing EP's remain applicable to Cycle 14.


4. Result in more than a minimal increase In the consequences of a malfunction of a l Yesstructure, system, or component important to safety previously evaluated In the 0 NoFSAR?BASIS:

The Cycle 14 ATRIUM-10 reload fuel and 9x9-5 reinsert fuel are designs that havebeen shown to be compatible with the co-resident GE1 I fuel type. The malfunctions ofkey plant components are analyzed as part of the reload process with the resultsreported in various sections of the FSAR. The consequences of these malfunctionshave been shown to meet their respective acceptance criteria.

Therefore, Cycle 14 operation will not result in more than a minimal increase in theconsequences of a malfunction of a structure, system, or component important tosafety previously evaluated in the FSAR.

5. Create a possibility for an accident of a different type than any previously evaluated In 0 Yesthe FSAR? E NoBASIS:

The Cycle 14 ATRIUM-10 reload fuel and FANP 9x9-5 reinsert fuel are similar to andcompatible with the GE11 fuel that was inserted in previous cycles. The details ofthis design have been specifically considered in the safety analysis and the coremonitoring system. No plant modifications are required to accommodate the newcore design or Cycle 14 operation. The GGNS Cycle 14 fuel types have beenapproved for the Cycle 14 reactor chemistry conditions.

The GGNS operational parameters (water chemistry requirements, spectral-shiftcore designs, and MEOD rod-lines) have been reviewed and are not expected toresult in unusual crud buildup like that observed on the high-power GE1l1 bundles atRiver Bend. Inspection of a high-power, once-burnt representative fuel bundleduring GGNS RF10 has confirmed that the high-power GGNS Cycle 10 fuel bundleshave no unusual crud buildup.

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6. Create a possibility for a malfunction of a structure, system, or component important to 0 Yessafety with a different result than any previously evaluated in the FSAR? E NoBASIS

The Cycle 14 ATRIUM-10 reload fuel and 9x9-5 reinsert fuel are designs that havebeen shown to be mechanically, neutronically, and thermal-hydraulically compatiblewith the co-resident GE1I fuel. The ATRIUM-10 and 9x9-5 fuel will not introduceany adverse flow distribution effects such as preferential flow through the ATRIUM-10 bundles that may negatively impact the GE 11 bundles. No plant modificationsare required to accommodate the new core design and no additional loads will beimposed on any existing equipment. The GE11, ATRIUM-10 and 9x9-5 bundlesprovide sufficient clearance for proper control blade operation and allow sufficientbypass flow in the bypass region to provide adequate cooling for control blades andin-core detectors. There are no special operational considerations associated withthe Cycle 14 core other than those associated with the increased bow condition.The higher friction expected between the control blades and anyATRIUM-10bundles experiencing increased bow would not be sufficient to cause a failure of thefuel bundle, control blade, or control rod drive coupling. Appropriate channel bowsurveillances will be performed during Cycle 14 to monitor this condition. The GE11bundles are loaded in unrodded locations on the core periphery and increased bowin these bundles will not affect control rod movement.


7. Result In a design basis limit for a fission product barrier as described in the FSAR 0 Yesbeing exceeded or altered? 0 No

BASIS:

Mechanical analyses have been performed to ensure that all fuel in the Cycle 14core meet the mechanical design limits for steady-state operation as well as transientconditions including fatigue damage, creep collapse, corrosion, fuel rod internalpressure, rod bow, internal pressure, etc. Additionally, no Cycle 14 fuel will exceedthe applicable burn-up or residence time limits.

Core operating limits have been developed using NRC approved codes in order toensure that the Cycle 14 fuel will not exceed the MCPR safety limits for steady-stateoperation and anticipated operation occurrences. Similarly, operating limits havebeen developed to ensure that the Cycle 14 fuel will not exceed the 1% claddingstrain 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 duringsome AOO's, which would increase the suppression pool temperature, the bulkcontainment pressure increase is negligible and would not exceed the design limit.

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ATTACHMENT 9.1 50.59 REVIEW FORM Page 12 Of 13


A design basis limit for the peak fuel enthalpy of 280 calIgm has been established forthe control rod drop accident (CRDA) to preclude significant fuel cladding failure suchthat core geometry and cooling may be impacted. The CRDA has been evaluated forCycle 14. This evaluation considers all potential withdrawal sequences andconcludes that a CRDA will not exceed the 280 cal/gm peak enthalpy limit. Since thisaccident is a localized event and the peak enthalpy does not exceed 280 cal/gm,there is no impact on the vessel or containment pressures. As such their respectivelimits are not exceeded.

IOCFR50.46 provides limits associated with the ECCS performance analysis (LOCAanalysis). Two such limits are Peak Clad Temperature (PCT) and local cladoxidation. Although these limits ere not subject to 1 OCFR50.59, they are discussed inthis evaluation for completeness. Grand Gulf specific analyses have been performedfor ATRIUM-1 0, 9x9-5, and GE11 fuel in accordance with 10CFR50.46. Theseanalyses, which are applicable to Cycle 14, show that the PCT and local oxidation arewell below the limits set forth in IOCFR50.46. These analyses also show that thecore-wide metal water reaction, which is used to evaluate compliance with thecontainment design limit, is less than the 1OCFR50.46 limit. The remainder of theexisting containment analysis associated with LOCA events is applicable to Cycle 14as described in Attachment 1. As such, the containment pressure design limit will notbe exceeded in Cycle 14.



Therefore, Cycle 14 operation will not result In a design basis limit for a fissionproduct barrier as described in the FSAR being exceeded or altered.

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8. Result In a departure from a method of evaluation described in the FSAR used in D Yesestablishing the design bases or in the safety analyses? E No

BASIS:

The reload analyses performed by the fuel vendor utilized NRC approved methods aslisted in Technical Specification 5.6.5 and described throughout the FSAR. Since theMCPR operating limits developed for the ATRIUM-10 fuel are conservatively appliedto the GE11 and 9x9-5 fuel types, the references listed In Technical Specification5.6.5 need not include the supporting NRC approved methodologies for these non-limiting fuel types. As described in Attachment 1, uncertainty applied in the SafetyLimit calculation associated with each of the equipment out of service combinationswas calculated in accordance with Framatome-ANP's NRC approved methodology.All remaining GGNS evaluations currently described in the FSAR have been shown tobe applicable to Cycle 14. As such, no new methods were used. Finally, the GGNSEP calculation has been updated to consider the minor changes to two fuelparameters. This revision did not incorporate any new or different methods.



SE 2004-0001-ROI

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Facility: GGNS

Document Reviewed: Core Operating Limits Report (COLR) Change/Rev. LDC 2004-009

System Designator(s)/Descriptlon: JI I

Description of Proposed ChangeThis evaluation addresses the Cycle 14 reload changes and operation of the Cycle 14 reload core as givenin the Core Operating Limits Report (COLR). Revision I evaluates the impact on the Cycle 14 core designand operation of the identification of fuel failures during RF13 and changes to the reference core loading toreflect the replacement of one failed once-burnt ATRIUM-10 fuel bundle.

If the proposed activity, In Its entirety, Involves any one of the criteria below, check the appropriatebox, provide a justification/basis in the Description above, and forward to a Reviewer. No further50.59 Review Is required. If none of the criteria Is applicable, continue with the 50.59 Review.

O The proposed activity is editorialtypographical ar, defined In Section 6.2.2.1.

o3 The proposed activity represents an 'FSAR-only" change as allowed in Section 5.2.2.2(Insert item # from Section 5.2.2.2).

If further 50.59 Review Is required, check the applicable review(s): (Only the sections Indicated mustbe Included In the Review.)

i] | SCREENING I Sections I, II, III, and IV required

o 50.59 EVALUATION EXEMPTION Sections I, II, Ill, IV, and V required

0 50.59 EVALUATION (#: SE-2004-0001-R.1 ) Sections 1, 11, III, IV, and VI required

Preparer: Guy B. SOikes/ ):&, 4. sL4 /EOI/Nucl. Eng. - SAI 3 - 16 -O4Name (print) I Signature / C pany / Department ( Date

Reviewer: 5X!/4/ 6.ezzz//-Name (print) / Signatu TDe / Deftmenn V Date

Ws,3 1- -o _(0q/ r / a.

OSRC:Chairman's Name (print) / SignApre / Date[Rcquircd only for Programmatic Excluaion screenings (see Section 5.8) and 50.59 Evaluations.

List of Assisting/Contributing Personnel:Name:

J. A. Elam (Central Engineering BWR Fuels)D. L. Smith (Central Engineering BWR Fuels)J. P. HRad (Central Engineering DWR Fuels)G. W. Smith (GGNS-PSAI

Scope of Assistance:Core design and neutronic inputFuel mechanical inputCore stability and hydraulic compatibility InputFOP Innut

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II. SCREENING



Operating License YES NO CHANGE # andlor SECTIONS IMPACTED+__,

Operating License 0 __

TS 0 __

NRC Orders 0 _

If 'YES', obtain NRC approval prior to implementing the change by Initiating an LBD change In accordance with NMM L-113(Reference 2.2.13). (See Section 5.1.13 for exceptions.)

LBDs controlled under 50.69 YES NO CHANGE # (if applicable) andlor SECTIONS IMPACTED

FSAR E 0 LDC 2004-010

TS Bases 0 0 1

Technical Requirements Manual I 0 LDC 2004-012

Core Operating Limits Report E 0 LDC 2004-009

NRC Safety Evaluation Reports' 03 6

If 'YES", perform an Exemption Review per Section V OR perform a 50.59 Evaluation per Section VI ADl Initiate an LBDchange In accordance with NMM L1.113 (Reference 2.2.13).

LSDs controlled under other regulations YES NO CHANGE # (if applicable) andlor SECTIONS IMPACTED

Quality Assurance Program Manual2 0' IEmergency Plan2 O 0

Fire Protection Program' 0 B(induces the Fire Ha7ards Analysis)

Offste Dose Calculations Manual 0 B

If 'YES", evaluate any changes In accordance with the appropriate regulation AND Initiate an LBD change In accordance withNMM U-113 (Reference 2.2.13).

2. Does the proposed activity Involve a test or experiment not described in theFSAR?If "yes," perform an Exemption Review per Section V OR perform a 50.59Evaluation per Section VI.

3. Does the proposed activity potentially Impact equipment, procedures, orfacilities utilized for storing spent fuel at an Independent Spent Fuel StorageInstallation?(Check "NIA" if dry fuel atorage is not applicable to the facility.)If "yes," perform a 72.48 Review In accordance with NMM Procedure LI-112.(See Sections 1.5 and 5.3.1.5 of the E01 10CFR50.59 Review Program Guidelines.)

O Yesl No

o YesO No0 NIA

If YES.' see Section 5.1.4.2 If 'YES.' notify the responsible department and ensure a 50.54 Evaluation is performed. Attach the 50.54 Evaluation.3 If YES.' evaluate the change in accordance with the requirements of the facility's Operating License Condition.

B. Basis

Provide a clear, concise basis for the answers given in the applicable sections above. Explain why theproposed activity does or does not impact the Operating LicenselTechnical Specifications and/or the FSARand why the proposed activity does or does not Involve a new test or experiment not previously described inthe FSAR. Adequate basis must be provided within the Screening such that a third-party reviewer can reachthe same conclusions. Simply stating that the change does not affect TS or the FSAR is not an acceptablebasis. See EOI 50.59 Guidelines Section 5.6.6 for guidance.)

The current MCPR Safety Limit has been shown to be applicable to the Cycle 14 core. As such,Tech Spec 2.1.1.2 does not need to be revised.

Tech Spec 3.4.11 includes exposure-dependent P/T limit curves. The cumulative reactor vesselservice life (EFPY) is expected to increase beyond the applicability of the PIT limit curve currentlyin use (L16 EFPY) approximately 5 months after BOC14. The TRM will need to be updated toreflect the change in applicable P/T limit curve in TS 3.1.1 1.

There are no other Tech Spec, Bases, or TRM changes required for Cycle 14 startup and initialoperation. There are no NRC orders applicable to the Cycle 14 reload campaign.

The Cycle 14 core will contain FANP 9x9-5 reinsert fuel bundles last used in Cycle 11 and the corecharacteristics and response will also be somewhat different than currently described in the FSAR.As such, Cycle 14 analyses have been performed for the new core and the FSAR will need to beupdated appropriately as will the COLR.

The Cycle 14 core design and operation will not affect the ODCM, QAPM, Emergency Plan,Security Plan, Fire Protection Program, or any NRC SERs.

C. References

Discuss the methodology for performing the LBD search. State the location of relevant licensing documentInformation and explain the scope of the review such as electronic search criteria used (e.g., key words) or thegeneral extent of manual searches per Section 5.3.6.4 of L-101. NOTE: Ensure that electronic andmanual searches are performed using controlled copies of documents. If you have any questions,contact your site Uconcing dopartmont.


OLM. FSAR, TS Bases. TRM Fuel, reload, channel. COLR. P/T


D. Is the validity of this Review dependent on any other RI Yeschange? (See Section 5.3.4 of the EOI 10CFR5O.59 Program 0 NoReview Guidelines.)


Revise the TRM prior to BOC14+5 months to reflect change In applicable TS 3.4.11 P/T limitcurve (LCTS 35421)

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If any of the following questions Is answered "yes," an Environmental Review must be performedIn accordance with NMM Procedure EV-115, "Environmental Evaluations," and attached to this50.59 Review. Consider both routine and non-routine (emergency) discharges when answeringthese questions.


Yes No

1. 0 0 Involve a land disturbance of previously disturbed land areas in excess of one acre (i.e.,grading activities, construction of buildings, excavations, reforestation, creation or removalof ponds)?


3. 0 0 Involve dredging activities in a lake, river, pond, or stream?






9. El 0 Modify the design or operation of the plant that will change the path of an existing waterdischarge or that will result in a new water discharge?

10. 0 0 Modify existing stationary fuel burning equipment (i.e., diesel fuel oil, butane, gasoline,propane, and kerosene)?.

11. 0 0 Involve the installation of stationary fuel burning equipment or use of portable fuel burningequipment (i.e., diesel fuel oil, butane, gasoline, propane, and kerosene)?'

12. El 0 Involve the installation or use or equipment that will result In an air emission discharge?



15. 0 E0 involve burial or placement of any solid wastes in the site area that may affect runoff,surface water, or groundwater?

'See NMM Procedure EV-117, 'Air Emissions Management Program. for guidance in answering this question.

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If any of the following questions Is answered 'yes," a Security Plan review must be performed bythe Security Department to determine actual Impact to the Plan and the need for a change to thePlan.


Yes No

1. 0 El Add, delete, modify, or otherwise affect Security department responsibilities (e.g., includingfire brigade, fire watch, and confined space rescue operations)?

2. 0 El Result in a breach to any security barrier(s) (e.g., HVAC ductwork, fences, doors, walls,ceilings, floors, penetrations, and ballistic barriers)?

3. 0 El Couso matorials or equipment to be placed or installed within the Security Isolation Zone?

4. 3 i1 Affect security lighting by adding or deleting lights, structures, buildings, or temporaryfacilities?


6. 0 E Modify or otherwise affect the operation or field of view of the security cameras?

7. 0 0 Modify or otherwise affect (block, move, or alter) installed access control equipment,intrusion detection equipment, or other security equipment?

8. 0 M Modify or otherwise affect primary or secondary power supplies to access controlequipment, Intrusion detection equipment, other security equipment, or to the CentralAlarm Station or the Secondary Alarm Station?

9. 0 21 Modify or otherwise affect the facility's security-related signage or land vehicle barriers,including access roadways?


The Security Department answers the following questions If one of the questions was answered"yes".

B. Is the Security Plan actually Impacted by the 0 Yesproposed activity? 0 No


0 No




Brief description of change, test, or experiment:This safety evaluation assesses the reload-related changes associated with Cyclc 14operation as presented in the Core Operating Limits Report (COLR) located in theOperating License Manual (OLM). Cycle 14 has been designed for 538 Effective FullPower Days with a core consisting of 244 fresh ATRIUM-10 assemblies, 239 once-burntATRIUM-10 assemblies, 204 twice-burnt ATRIUM-10 assemblies, 28 thrice burnt GE11assemblies, and 85 previously loaded twice-burnt FANP 9x9-5 assemblies. Eight of thefresh ATRIUM-10 assemblies are supplied with lead fuel channels manufactured by KobeSteel Ltd. There are no TS or TS Bases changes required to operate with this new core,however, the FSAR does require updates. A TRM change will also be needed later duringCycle 14 to change the applicable TS 3.4.11 PIT limit curve. The Cycle 14 core has beendesigned and analyzed for a rated thermal power of 3898 MWt. Attachment 1 provides adetailed description of the Cycle 14 reload analysis and the issues considered in thisevaluation. Increased (abnormal) channel bow described in CR-GGN-2002-01810 hasbeen explicitly considered in the Cycle 14 reload analysis for the GE11 and potentiallysusceptible high exposure ATRIUM-10 fuel types using the advanced (thick-thin) channeldesign. Revision 1 incorporates the results of evaluations associated with a revision to thereference core loading pattern. The original core design considered replacement of oneassumed failed once-burnt ATRIUM-10 fuel bundle. The revised core loading reflects theidentification of that failed bundle and its replacement with another non-failed once-burntATRIUM-10 bundle. The revised reference core loading is provided in FAB04-072 and theCycle 14 core has been verified to be correctly loaded per 17-S-02-108-RF13-031504.




The Cycle 14 core configuration and operation has been evaluated with respect tomechanical, neutronic, thermal-hydraulic, dose, thermal performance, and methodsconsiderations for GGNS. This evaluation concludes that the reload-related changesassociated with Cycle 14 operation does not require NRC review.

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Does the proposed Change being evaluated represent a change to a method of 0 Yesevaluation ONLY? If "Yes," Questions 1 -7 are not applicable; answer only El NoQuestion 8. If "No," answer all questions below.


1. Result in more than a minimal increase in the frequency of occurrence of an accident 0 Yespreviously evaluated in the FSAR? E No

BASIS:

The Cycle 14 core loading and cycle operation will not result in more than a minimalincrease in the frequency of occurrence of an accident previously evaluated in theFSAR. The precursors to these events are independent of the core design and thefrequency classifications reported in FSAR Chapter 15 are unaffected by the coreparameters. The following considerations support this conclusion.

MechanicalThe ATRIUM-10 mechanical design has been reviewed for use at Grand Gulf. Nounusual failure modes or increased failure frequency have been identified for this fueldesign. This is the third reload at GGNS with ATRIUM-10 fuel and this fuel design hasaccumulated significant problem-free operational experience at other plants. TheFANP 9x9-5 fuel type was introduced at GGNS in Cycle 5 and was last used in Cycle11. The mechanical design of this fuel has been reviewed for use at GGNS in themixed (ATRIUM-10, GE1 1, 9x9-5) Cycle 14 core. The Cycle 14 bundles will operatewithin the power history assumptions in the fuel mechanical analyses and will result inexposures within the analyzed bumup limits of the ATRIUM-10, GE11, and 9x9-5mechanical designs. All design criteria for the GE11 bundles have been shown tomeet their respective limits including those that will be irradiated for a fourth cycle.

NuclearThe neutronic characteristics of the Cycle 14 9x9-5, GE11 and ATRIUM-10 mixedcore design have been considered in the safety analysis. Adequate shutdown marginhas been predicted by analysis and will be confirmed during startup tests. Inaddition, the hold-down capability of the standby liquid control system and thesubcriticality of Cycle 14 fuel in the spent fuel storage racks have been confirmed.Therefore, the probability of inadvertent criticality has not been increased by theintroduction of the Cycle 14 reload fuel.

Thermal-HydraulicFANP's modeling of the GE fuel and the thermal-hydraulic compatibility of theATRIUM-10, GE11, and 9x9-5 fuel have been reviewed and found acceptable. Toaccurately model the GE11 bundle hydraulics, a Cycle 11 GE1l1 bundle was shippedto FANP for hydraulic testing in their hydraulic test facility. Analyses have beenperformed to demonstrate that Cycle 14 meets all Enhanced-1A stabilityperformance criteria without changes to the ElA hardware or power-flow map regionboundaries. Therefore, the probability of thermal-hydraulic instabilities has notincreased.

EN-S NUCLEAR QUALrrY RELATED L-401 Revision3MANAGEMENT AUMINIKATIVE

~Entergy MANUALINFORMATION USE


Analyzed EventsThe probability of the occurrence of anticipated operational events is not dependenton the core configuration. No changes to the plant design are required for the Cycle14 core. The Cycle 14 core loading will not affect the precursors to any of theChapter 15 events. The probability of an analyzed event therefore has notincreased.

As described in FSAR Section 15A.6.5.3, the Control Rod Drop Accident (CRDA)results from a failure of the control rod-to-drive mechanism coupling after the controlrod becomes stuck in its fully inserted position. Although an increased channel bowcondition could result in increased friction between the control blade and itscorresponding fuel assemblies, analyses have shown that there would not besufficient friction to result in a mechanical failure of the coupling. Additionally, thecontrol rod drive mechanism would not produce enough force to result In amechanical failure of the coupling even if the channel bow was so severe that theassemblies would preclude blade movement. As such, channel bow is notconsidered a precursor to the CRDA, and any increased bow associated with the highexposure ATRIUM-10 bundles (the GE11 bundles are loaded in unrodded locations)would not increase the probability of this event.

On these bases, the probability of occurrence of accidents previously identified in theFSAR is not increased for the Cycle 14 core with increased channel bow.

2. Result In more than a minimal increase in the likelihood of occurrence of a malfunction 0 Yesof a structure, system, or component important to safety previously evaluated In the E NoFSAR?BASIS:

The Cycle 14 ATRIUM-10 reload fuel and 9x9-5 reinsert fuel are designs that havebeen shown to be mechanically, neutronically, and thermal-hydraulically compatiblewith the co-resident GE11 fuel. No plant modifications are required to accommodatethe new core design and the only additional loads placed on plant equipment wouldbe potential friction between the control blades and excessively bowed ATRIUM-10bundles (the GE11 bundles are loaded in unrodded locations). Based on previousexperience with bowed fuel at GGNS and Clinton Power Station, this increasedfriction is not expected to impact scram times. Technical Specification scram timetesting and appropriate channel bow surveillances will be performed during Cycle 14.These actions would identiTy any potential scram time or other Impacts and theappropriate actions would be taken. Additionally, this increased friction would not besufficient to provide any failures associated with the control blades or the controlblade drive system.

A conservative vessel overpressurization analysis has been performed, which showsthat the vessel pressure limit is not cxcccded. The precursors to any malfunction ofequipment important to safety are not affected by this change.


EN-S NUCLEAR QUALTY RELATED LI-101 Revision 3MANAGEMENT ADMINISTRATIVE

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3. Result in more than a minimal increase in the consequences of an accident previously 0 Yesevaluated in the FSAR? El No

BASIS:

As reported in Attachment 1, the acceptance criteria reported in FSAR Section15.0.3.1 and the Technical Specifications are satisfied for each event classification.Core operating limits have been developed to ensure that moderate frequencyevents do not violate the MCPR safety limit or fuel cladding strain limits. Theconsequences of infrequent events have been shown to meet the appropriateacceptance criteria while the individual acceptance criteria for the limiting faults havebeen demonstrated to be satisfied. The following considerations support theseconclusions.

Moderate Frequency EventsThe Cycle 14 core operating limits have been developed with NRC-approvedmethodologies such that the MCPR safety limit and the fuel cladding strain limit willnot be violated by any analyzed moderate frequency transient initiated from anystatepoint available to GGNS. As such, no fuel failures are expected to result fromany moderate frequency event. These analyses considered GGNS-specificoperational modes such as MEOD, SLO, FHOOS, and EOC-RPT inoperable. Thesecore operating limits consist of MCPR, MAPLHGR and LHGR curves that arefunctions of flow, power, and exposure. These limits consider conservative channelbow assumptions for the GE11 and highly exposed ATRIUM-10 fuel that bound theincreased bow previously observed in the GE11 fuel.

These core operating limits will be incorporated into the core monitoring system,however, as with previous cycles, the GE11 operating limits illustrated in the COLRwill differ somewhat from those limits in the core monitoring system. GE hasgenerated pellet-based exposure-dependent LHGR and lattice-based exposure-dependent MAPLHGR limits for their GE11 fuel bundles; however, for competitivereasons, GE has designated the limits for only the most-limiting or least-limitinglattices as non-proprietary. The COLR, which is submitted to the NRC for informationonly, will therefore report the non-proprietary limits for the most limiting enrichedlattice at each exposure for reference purposes only, and refer to the appropriate GEproprietary document for the actual limits. The actual LHGR and MAPLHGR limits formost lattices will be higher than the COLR reference curves and It Is recognized thatmost lattices could operate at higher values than the reference limits. However, thisis acceptable since Technical Specifications 3.2.1 and 3.2.3 require that all APLHGRsand LHGRs, respectively be less than or equal to the limits specified in the COLR,and the COLR refers to the proprietary GE document for the actual limits.

Infrequent EventsThe consequences of the limiting infrequent events have been evaluated and shownto meet their respective acceptance criteria. These events include the pressureregulator failure downscale. misplaced (i.e., misoriented and mislocated) bundle, andsingle loop operation pump seizure accidents. Radiological analyses using thealternative source term (AST) have been performed to ensure that these events willnot result in offsite or control room doses greater than their respective acceptancecriteria. These evaluations consider conservative channel bow assumptions for theGE11 and highly exposed ATRIUM-10 fuel that bound the increased bow previouslyobserved in the GE II fuel.

EN-S NUCLEAR QuALTY RELATED Li-101 Revision 3En MANAGEMENT ADMINISTRATIVE

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Limiting FaultsThe limiting faults at GGNS include the fuel handling accident, the control rod dropaccident, and the design basis LOCA. The radiological analyses for these eventshave been developed as part of the GGNS AST effort and bound the Cycle 14 coreparameters. For the LOCA, MAPLHGR operating limits and single-loop multipliershave been developed for the Cycle 14 core configuration such that the requirementsof 10CFR50.46 are satisfied. The containment response for the Cycle 14 core wasfound to be bounded by previous cycles as is the hydrogen analysis. Theseismic/LOCA response of the Cycle 14 core has been confirmed to be acceptable.The Cycle 14 core design results in minor changes to two EP parameters (Mclad andMfuel); however, the existing EP's remain applicable to Cycle 14.


4. Result in more than a minimal increase In the consequences of a malfunction of a 0 Yesstructure, system, or component Important to safety previously evaluated in the 0 NoFSAR?BASIS:

The Cycle 14 ATRIUM-10 reload fuel and 9x9-5 reinsert fuel are designs that havebeen shown to be compatible with the co-resident GE1l1 fuel type. The malfunctions ofkey plant components are analyzed as part of the reload process with the resultsreported in various sections of the FSAR. The consequences of these malfunctionshave been shown to meet their respective acceptance criteria.

Therefore, Cycle 14 operation will not result in more than a minimal increase in theconsequences of a malfunction of a structure, system, or component important tosafety previously evaluated in the FSAR.

5. Create a possibility for an accident of a different type than any previously evaluated in 11 Yesthe FSAR? 0 NoBASIS:

The Cycle 14 ATRIUM-10 reload fuel and FANP 9x9-5 reinsert fuel are similar to andcompatible with the GE11 fuel that was inserted in previous cycles. The details ofthis design have been specifically considered in the safety analysis and the coremonitoring system. No plant modifications are required to accommodate the newcore design or Cycle 14 operation. The GGNS Cycle 14 fuel types have beenapproved for the Cycle 14 reactor chemistry conditions.

The GGNS operational parameters (water chemistry requirements, spectral-shiftcore designs, and MEOD rod-lines) have been reviewed and are' not expected toresult in unusual crud buildup like that observed on the high-power GE11 bundles atRiver Bend. Inspection of high-power, once- and twice-burnt representative fuelbundles during GGNS RF13 has confirmed that the high-power GGNS Cycle 13 fuelbundles have no unusual crud buildup. As described in Attachment 1, the fuel

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IATTACHMENT 9.1 1 50.59 REVIEW FORM Page I11 I jt3

failures during Cycle 13 are not indicative of generic or core-wide fuel performanceissues.


6. Create a possibility for a malfunction of a otructuro, system, or component important to l Ye3safety with a different result than any previously evaluated in the FSAR? 0 No

BASIS

The Cycle 14 ATRIUM-10 reload fuel and 9x9-5 reinsert fuel are designs that havebeen shown to be mechanically, neutronically, and thermal-hydraulically compatiblewith the co-resident GE11 fuel. The ATRIUM-10 and 9x9-5 fuel will not introduceany adverse flow distribution effects such as preferential flow through the ATRIUM-10 bundles that may negatively impact the GE 11 bundles. No plant modificationsare required to accommodate the new core design and no additional loads will beimposed on any existing equipment. The GE1 1, ATRIUM-1 0 and 9x9-5 bundlesprovide sufficient clearance for proper control blade operation and allow sufficientbypass flow in the bypass region to provide adequate cooling for control blades andin-core detectors. There are no special operational considerations associated withthe Cycle 14 core other than those associated with the increased bow condition.The higher friction expected between the control blades and any ATRIUM-10bundles experiencing increased bow would not be sufficient to cause a failure of thefuel bundle, control blade, or control rod drive coupling. Appropriate channel bowsurveillances will be performed during Cycle 14 to monitor this condition. The GE11bundles are loaded in unrodded locations on the core periphery and increased bowin these bundles will not affect control rod movement.


7. Result In a design basis limit for a fission product barrier as described in the FSAR 11 Yesbeing exceeded or altered? E NoBASIS:

Mechanical analyses have been performed to ensure that all fuel in the Cycle 14core meet the mechanical design limits for steady-state operation as well as transientconditions including fatigue damage, creep collapse, corrosion, fuel rod internalpressure, rod bow, internal pressure, etc. Additionally, no Cycle 14 fuel will exceedthe applicable bum-up or residence time limits.

Core operating limits have been developed using NRC approved codes in order toensure that the Cycle 14 fuel will not exceed the MCPR safety limits for steady-stateoperation and anticipated operation occurrences. Similarly, operating limits havebeen developed to ensure that the Cycle 14 fuel will not exceed the 1% claddingstrain 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 duringsome AOO's, which would increase the suppression pool temperature, the bulk

EN-S NUCLEAR QUAuTY RELATED LU-101 Revision 3

INFORMATION USE


containment pressure increase is negligible and would not exceed the design limit.


A design basis limit for the peak fuel enthalpy of 280 cal/gm has been established forthe control rod drop accident (CRDA) to preclude significant fuel cladding failure suchthat core geometry and cooling may be impacted. The CRDA has been evaluated forCycle 14. This evaluation considers all potential withdrawal sequences andconcludes that a CRDA will not exceed the 280 cal/gm peak enthalpy limit. Since thisaccident is a localized event and the peak enthalpy does not exceed 280 calgm,there is no impact on the vessel or containment pressures. As such their respectivelimits are not exceeded.

10CFRSO.46 provides limits associated with the ECCS performance analysis (LOCAanalysis). Two such limits are Peak Clad Temperature (PCT) and local cladoxidation. Although these limits are not subject to 1 OCFR50.59, they are discussed inthis evaluation for completeness. Grand Gulf specific analyses have been performedforATRIUM-10, 9x9-5, and GE11 fuel in accordance with 1OCFR50.46. Theseanalyses, which are applicable to Cycle 14, show that the PCT and local oxidation arewell below the limits set forth in 1 OCFR50.46. These analyses also show that thecore-wide metal water reaction, which is used to evaluate compliance with thecontainment design limit, is less than the 10CFR50.46 limit. -The remainder of theexisting containment analysis associated with LOCA events is applicable to Cycle 14as described in Attachment 1. As such, the containment pressure design limit will notbe exceeded in Cycle 14.



Therefore, Cycle 14 operation will not result in a design basis limit for a fissionproduct barrier as described in the FSAR being exceeded or altered.

EN-S NUCLEAR CUALMY RELATED LI-101 Revision 3MANAGEMENT ADMINISTAWET

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ATTACHMENT9.1 50.59REVIEWFORM Page 13 o

8. Result in a departure from a method of evaluation described in the FSAR used in 0 Yesestablishing the design bases or in the safety analyses? Ed No

BASIS:

The reload analyses performed by the fuel vendor utilized NRC approved methods aslisted in Technical Specification 5.6.5 and described throughout the FSAR. Since theMCPR operating limits developed for the ATRIUM-10 fuel are conservatively appliedto the GE1 1 and 9x9-5 fuel types, the references listed in Technical Specification5.6.5 need not include the supporting NRC approved methodologies for these non-limiting fuel types. As described in Attachment 1, uncertainty applied in the SafetyLimit calculation associated with each of the equipment out of service combinationswas calculated in accordance with Framatome-ANP's NRC approved methodology.All remaining GGNS evaluations currently described in the FSAR have been shown tobe applicable to Cycle 14. As such, no new methods were used. Finally, tle GGNSEP calculation has been updated to consider the minor changes to two fuelparameters. This revision did not incorporate any new or different methods.



SE 2004-0002-ROO

. a

I .



Document Reviewed: Temp Alt 04-005 (Division 1) Change/Rev.: 0TemD Alt 04-006 (Division 21 ChangelRev.: 0

System Designator(s)!Description: P41. Standby Service Water System

Description of Proposed Change

Manual operation of SSW cooling tower fans will assist In maintaining SSW basin temperature withinadministrative limits during cold weather. During Shutdown Cooling operations while In Modes 4 and 5,manual operation of the SSW system and fans will control the heat removal capability of the UHS. Thiswill be accomplished by placing a Jumper around the contacts for a SSW system initiation based uponmanual starting of the RHR pump (A or B) that Is In service for shutdown cooling. The effect will be thatthe SSW system will still manually and automatically Initiate as designed, except for the response to aRHR pump manual start. Any automatic initiation signal required in modes 4 and 5 will cause SSW tooperate as designed. There Is no automatic initiation signal that will only start the RHR pumps.Installation of the temporary alteration will require manual alignment of SSW valves and manual start of theSSW pump prior to manually starting the RHR pump. This Is In accordance with the current proceduresand places no additional burden on operations (Ref. SOI 04-1-01-E12-1, Rev. 122 and SOI 04-1-01-E12-2,Rev 101). This temporary alteration does not Impact the ability to control SSW or RHR from the remoteshutdown panels.

Temporary Alteration 04-005 and 04-006 discusses temperature limits for starting and stopping SSW fans.The fans can be cycled to maintain the SSW pump discharge temperature between 500F and 750F. Thetemporary alteration is only applicable In modes 4 and 5.

If the proposed activity, In Its entirety, Involves any one of the criteria below, check the appropriatebox, provide a Justification/basis in the Description above, and forward to a Reviewer. No further50.59 Review Is required. If none of the criteria Is applicable, continue with the 50.59 Review.

o The proposed activity Is editorial/typographical as defined in Section 5.2.2.1.

o The proposed activity represents an FSAR-only" change as allowed in Section 5.2.2.2(Insert Item # from Section 5.2.2.2).

If further 50.59 Review Is required, check the applicable review(s): (Only the sections Indicated mustbe Included In the Review.)

o SCREENING Sections 1, 11, III, and IV required

o 50.59 EVALUATION EXEMPTION Sections 1, 11, III, IV, and V required

0 50.59 EVALUATION (#- 5- 2 c,0 Z.- KO Sections I, ii, III, IV, and VI required

.. I

Preparer: I %fe+ Cf-erv- I EOI l DE - Mech /Name (print) / Signature I Company I Department / Date

Reviewer: AI,(1 1)$ 'k IA erf Fi//cn EOI/DE-MechIName (print) / Signature I Company I Department / Date

Z-27-,0

V v PC>5- ((2e7ecN GOSRC K % 2-7 1-og tt oo7- oodf

Chairman's Signature / Dae[Required only for Programmatic Exclusion Screenings (see Section 5.9) and 50.59 Evaluations.1

List of Assisting/Contrlbuting Personnel:Name:

Jeff Brown. Robert Burrell. Andrew Fox, AmandaGallagher. Gerald Lantz, Tom Matson, DennisChiplev. and Ron Roma.

Scope of Assistance:Preparation of 50.59. Additional review andcomments on 50.59. Preparation of Temn Ailts

II. SCREENING



Operating License YES NO CHANGE # and/or SECTIONS TO BE REVISED

Operating License 0 0

TS IOL 1

NRC Orders 0 I11 "YES", obtain NRC approval prior to Im plementing the change by Initiating an L ED change In accordance with NMM U -113(Reference 2.2.13). (See Section 5.1.13 for exceptions.)

LBDs controlled under 50.59 YES NO CHANGE # and/or SECTIONS TO BE REVISED

FSAR 0 0 Various sections addressed In Section 11.B. No LDC will be Issued since thetemp Bit will only be In place during the outage.

TS Bases 0 0 TS Bases addressed In Sectlon ll.B. No LDC will be Issued since the tempalt will only be In place during the outage.

Technical Requirements Manual 0 0 TRM addressed in Section 11.B. No LDC will be issued since the temp altwill only be In place during the outage.

Core Operating Limits Report 0 0

NRC Safety Evaluation Reports' 0t_If "YES'. nerform an Exematlon Review per S ection V OR perform a 50.59 Evaluation per S ection VI AND Initiate an LBD chan.s Inaccordance with NMM U.113 (Reference 2.2.13).

LBDs controlled under other regulations YES NO CHANGE # and/or S ECTIONS TO BE REVISED

Oluallty Assurance Program Manuel2 0 C 0

Emergency Plan2 0 0

Fire Protection Program' 0 0(Includes the Fire Hazards Analysis)

Offsite Dose Calculation Manual3 0 0

If "YES', evaluate any changes In accordance with the appropriate regulation AND Initiate an LBD change In accordance with NMMU-113 (Reference 2.2.13).


0 yes0 No

3. Does the proposed activity potentially Impact equipment, procedures, or facilities 0utilized for storing spent fuel at an Independent Spent Fuel Storage installation? 0(Check "NIA" If dry fuel storage Is not applicable to the facility.) 0If "yes," perform a 72.48 Review In accordance with NMM Procedure LI-112.(See Sections 1.5 and 5.3.1.5 of the EOI IOCFR50.59 Review Program Guidelines.)

YesNoN/A

'If YES. seeSection5.1.4. NoLBDchangeIsrequired.' If 'YES.' notify the responsible department and ensure a 50.54 Evaluation Is performed. Attach the 50.54 Evaluation.' If 'YES.' evaluate the change In accordance with the requirements of the facility's Operating License Condition.

B. BasisProvide a clear, concise basis for the answers given In the applicable sections above. Explain why the proposed activitydoes or does not Impact the Operating License/Technical Specifications and/or the FSAR and why the proposed activitydoes or does not Involve a new test or experiment not previously described In the FSAR. Adequate basis must be providedwithin the Screening such that a third-party reviewer can reach the same conclusions. Simply stating thsat the change doesnot affectTS or the FSAR Is not an acceptable basis. See EOI 50.59 Guidelines Section 5.6.6 for guldance.)

The Temp Alts allow for bypassing the applicable contacts in the SSW Initiation logic to prevent anSSW automatic initiation due to a manual RHR pump (A or B) start for Shutdown Cooling mode ofRHR. All other Initiation signals are unaffected and will start the respective SSW subsystems. Forshutdown cooling, fuel pool cooling assist, and any other manual RHR system start it will benecessary to manually align and start SSW prior to starting the RHR pump. This Is required bycurrent procedures (Ref. SOI 04-1-01-E12-1, Rev. 122 and SOI 04-1-01-E12-2, Rev 101) and canbe performed In the control room. Therefore, the temporary alteration will not result in anyadditional operator burdens. This method will allow manual control of the applicable cooling towerfans in order to aid in maintaining SSW temperature Within administrative limits. The temp alts donot bypass or disable any other automatic initiation signal. There is no automatic initiation signalthat will only start the RHR pump. Any automatic Initiation signal will also start other ESF equipmentand result in the initiation of SSW as designed. SSW response to LOCA, LOP, EDG, or the othercomponent start signals will remain as designed. The Temp Alts do not change the ability of theSSW system to perform its design safety functions.

Temporary Alteration 04-005 and 04-006 discusses temperature limits for starting and stoppingSSW fans. The fans can be cycled to maintain the SSW pump discharge temperature between 500Fand 75°F. The temporary alteration Is only applicable In modes 4 and 5.

OPERATING LICENSE: The operating license was reviewed and does not contain specific limits orrequirements of the SSW fans or RHR Shutdown Cooling. Therefore, no change Is required to theOperating License.

TECHNICAL SPECIFICATIONS: The TS were reviewed for applicability. The Temp Alts are onlyallowed for use with Shutdown Cooling In Modes 4 and 5. Tech Spec 3.7.1 addresses the SSW andUHS operability during Modes 1, 2, and 3. This Is not applicable since the temporary alterations areonly applicable during modes 4 and 5. No other TS LCOs other than TRM TR3.7.1 (addressedseparately below) pertain to the SSW system and fans being manually controlled for the specificpurpose of maintaining SSW basin temperature within administrative limits during shutdown cooling.Therefore the TS are not affected by this change.

NRC ORDERS: The GGNS NRC Orders required licensees to establish Interim compensatorymeasures to delineate licensee responsibility in response to the high level threat environment.Licensing was contacted because of the change to the Ultimate Heat Sink. The proposed changesto the SSW system and its control during Modes 4 and 5 do not pertain to station security.Therefore the NRC Orders are not affected by this change.

FSAR: The SSW System Is discussed In several areas of the FSAR.

Section 9.2.1 discusses the SSW system In detail. Since the proposed Temp Alts will not disable orotherwise prevent an SSW automatic Initiation from a LOCA, LOP, EDG start, or other componentstart, It still fulfills the design safety function as described in section 9.2.1. Manual control of theSSW system and fans for shutdown cooling Is in accordance with the Intent of section 9.2.1.1.2which states that the SSW system 'provides cooling to plant components, as required, duringnormal, reactor shutdown, and reactor isolation modes. The use of a temporary alteration to allowmanual control of SSW system and fans during shutdown cooling will aid in keeping SSWtemperature above administrative limits. The UHS temperature can still be controlled by the fans Inthe cooling tower; however, fan operation will only be manual during Shutdown Cooling or any othernon-automatic function. Any SSW initiation signal except for RHR pump start will result In thesystem and fans running as designed. The worst case design basis accident for the UHS(LOCAJLOP with single failure) in Section 9.2 assumes that the fans will run for Post-LOCA heatremoval. This assumption remains valid since the LOCA signals from high drywell pressure or lowreactor water level will still result in an automatic SSW system Initiation. The Temp Alts also do notchange the requirement that the SSW system respond automatically to a LOCA with no operatoraction (Section 9.2.1.5). With the Temp Alts installed, the UHS still functions as described In section9.2.5 because it will provide the designed cooling for LOCA conditions, and will provide cooling asrequired for normal shutdown operations.

Section 7.3.1.1.7 describes the SSW system Instrumentation and logic. With the Temp Altsinstalled, the system will initiate as described with the exception of response to an RHR pump beingstarted without another automatic SSW Initiation signal present. The Temp Alts require Operationsto manually align and start SSW to the applicable RHR pump to provide cooling to the pump sealcooler and pump room cooler prior to manually aligning RHR for Shutdown Cooling. Per currentprocedures (Ref. SOI 04-1-01-E12-2, Rev 101), SSW is manually aligned to RHR (valves alignedand SSW pump started) prior the RHR pumps being started. Additionally, the requirement that theSSW system fans and other essential components are prevented from being shutdown with anautomatic initiation signal present Is still satisfied except for manual start the RHR pump. Thereforein a LOCA or LOP SSW will function as designed.

Section 7.3.2.7.1 describes conformance with 10 CFR 50 General Design Criteria. GDC 20requires that 'protection systems shall be designed (1) to initiate automatically the operation ofappropriate systems including the reactivity control systems, to assure that the specified acceptablefuel design limits are not exceeded as a result of anticipated operational occurrences and (2) tosense accident conditions and to initiate the operation of systems and components Important tosafety.' This temp alt does not (1) impede the automatic operation of systems to protect fuel limitsnor does it (2) Impede the ability of the plant to sense accident conditions and respond to them.

Response to Criterion 20, Protection System Functions, states that SSW operates to serve otherESF systems and Is Initiated when they start to perform their ESF functions. This will remain truefor all components performing automatic ESF functions except for the manual start of the RHRpumps (A or B), which are affected by the temp alt. For the pump used for Shutdown Cooling, SSWwill be manually aligned before the pump Is started to ensure the pump Is protected fromoverheating as required by current procedures.

TS BASES: According to the TS Bases for the SSW system (p. B3.7-3) and TRM 3.7.1, an SSWsubsystem is considered operable when the associated pump Is operable and the associatedpiping, valves, instrumentation, and controls required to perform the safety related functions(including LOCAILOP) are operable. Operability of the UHS requires four fans (two per cell) to beoperable. Since the SSW system will automatically start on any required automatic signal besidesmanual start of the RHR pump, the system will be fully capable of performing the safety relatedfunctions for LOCA, LOP, and component starts other than RHR pump A or B. Additionally, thesystem will be manually placed in service with the fans manually controlled to maintain SSWtemperature when supplying the RHR pump for shutdown cooling, and are therefore performing thedesign function of protecting the RHR pump, although not automatically.

A| It 50.59 REVIEW1 -neg FORM ITEMP ALT 04.005 & 04-006 Pg f1"EnPae e6 rgy4

TRM: Section 3.7.1 discusses SSW operability in modes 4 and 5. Conditions A and B addressinoperability of one or more SSW fans. These requirements are not affected by TemporaryAlterations 04-005 and 04-006. Although the fans will not automatically start on a manual RHRpump (A or 8) start they will automatically start on all other initiation signals. Since the SSW basintemperatures will be maintained within administrative limits, thus within design limits, the UHS will befully capable of removing all residual heat from the reactor during manual control of the fans, and inthe event of any subsequent automatic initiation. SSW fans can be started and stopped from thecontrol room, if required to maintain basin temperature within the acceptable band, thus ensuringthat the associated RHR pump is adequately cooled.

COLR: The Core Operating Limits Report does not contain discussion of methods of controllingSSW temperature during Shutdown Cooling. Therefore there is no impact to the COLR.

NRC SERs: The NRC SER describes the NRC's response to the design of the plant as described Inthe FSAR. The intent of the SSW design as described In the FSAR Is not affected by thesetemporary alterations as described above. Therefore, there is no impact to the SERs..

OAPMIEPIFPP/ODCM: These documents do not contain Information relevant to control of the SSWsystem. Therefore they are unaffected by this change.

The proposed changes do not represent a test or experiment not previously described In the FSARbecause proper cooling will be provided to the components served by SSW as required. Automaticresponse to LOCAILOP Is maintained, and the system will be capable of performing its designsafety function. Additionally, the proposed change does not pertain to spent fuel storage.

C. References

Discuss the methodology for performing the LBD search. State the location of relevant licensingdocument Information and explain the scope of the review such as electronic search criteria used(e.g., key words) or the general extent of manual searches per Section 5.3.6.4 of LI-1 01. NOTE:Ensure that electronic and manual searches are performed using controlled copies ofdocuments. If you have any questions, contact your site Licensing department.


FSAR. TS. TS Bases. Operating License, TRM Shutdown cooling, fans. SSW

LBDs/lDocuments reviewed manually:

NRC Orders. NRC SERs. QAPM. ODCM. COLR,EP. FPP

D. Is the validity of this Review dependent on any other 0 Yeschange? (See Section 5.3.4 of the EOI 10CFR50.59 PrDgram 0 NoReview Guidelines)


N/A

III. ENVIRONMENTAL SCREENING

If any of the following questions Is answered "yes," an Environmental Review must be performedIn accordance with NMM Procedure EV-115, "Environmental Evaluations," and attached to this50.59 Review. Consider both routine and non-routine (emergency) discharges when answeringthese questions.


Yes No

1. 0 0 Involve a land disturbance of previously disturbed land areas In excess of one acre (i.e.,grading activities, construction of buildings, excavations, reforestation, creation or removalof ponds)?

2. 0 0 Involve a land disturbance of undisturbed land areas (I.e., grading activities, construction,excavations, reforestation, creating, or removing ponds)?

3. 0 0 Involve dredging activities In a lake, river, pond, or stream?


5. 0 0 increase the concentration or quantity of chemicals being discharged to the river, lake, orair?


7. 0 0 Change the design or operation of the Intake or discharge structures?


9. 0 0 Modify the design or operation of the plant that will change the path of an existing waterdischarge or that will result In a new water discharge?

10. 0 0 Modify existing stationary fuel burning equipment (i.e., diesel fuel oil, butane, gasoline,propane, and kerosene)?'

11. 0 0 Involve the Installation of stationary fuel burning equipment or use of portable fuel burningequipment (i.e., diesel fuel oil, butane, gasoline, propane, and kerosene)?1

12. 0 0 Involve the Installation or use of equipment that will result In an air emission discharge?

13. 0 0 Involve the Installation or modification of a stationary or mobile tank?

14. 0 0 Involve the use or storage of oils or chemicals that could be directly released Into theenvironment?

15. 0 0 Involve burial or placement of any solid wastes In the site area that may affect runoff,surface water, or groundwater?

I See NMM Procedure EV-117. Alr Emissions Management Program. for guidance In answering this question.

Page 8 Iof 1 4


If any of the following questions Is answered "yes," a Security Plan review must be performed bythe Security Department to determine actual Impact to the Plan and the need for a change to thePlan.


Yes No

1. 0 0 Add, delete, modify, or otherwise affect Security department responsibilities (e.g., Includingfire brigade, fire watch, and confined space rescue operations)?

2. 0 0 Result in a breach to any security barrier(s) (e.g., HVAC ductwork, fences, doors, walls,ceilings, floors, penetrations, and ballistic barriers)?

3. 0 0 Cause materials or equipment to be placed or Installed within the Security Isolation Zone?

4. 0 0 Affect security lighting by adding or deleting lights, structures, buildings, or temporaryfacilities?


6. 0 0 Modify or otherwise affect the operation or field of view of the security cameras?

7. 0 0 Modify or otherwise affect (block, move, or alter) Installed access control equipment,Intrusion detection equipment, or other security equipment?

8. 0 0 Modify or otherwise affect primary or secondary power supplies to access controlequipment, Intrusion detection equipment, other security equipment, or to the Central AlarmStation or the Secondary Alarm Station?

9. 0 0 Modify or otherwise affect the facility's security-related signage or land vehicle barriers,Including access roadways?


The Security Department answers the following questions If one of the questions was answeredyes".

B. Is the Security Plan actually Impacted by the 0 Yesproposed activity? O No

C. Is a change to the Security Plan required? O Yes Change # (optional)O No


TEMP ALT 04-005 & 04-008 Page 9 1of 141


Enter this section only If a "yes" box was checked In Section Il.A, above.

A. Check the applicable boxes below. If any of the boxes are checked, a 50.59 Evaluation Is notrequired. If none of the boxes are checked, perform a 50.59 Evaluation In accordance withSection V. Provide supporting documentation or references as appropriate.

o The proposed activity meets all of the following criteria regarding design function per Section5.6.1.1:

The proposed activity does not adversely affect the design function of an SSC as described Inthe FSAR; ANDThe proposed activity does not adversely affect a method of performing or controlling adesign function of an SSC as described In the FSAR; ANDThe proposed activity does not adversely affect a method of evaluation that demonstratesintended functions of an SSC described in the FSAR will be accomplished.

o An approved, valid 50.59 Review(s) covering associated aspects of the proposed changealready exists per Section 5.6.1.2. Reference 50.59 Evaluation # - (if applicable) or attachdocumentation. Verify the previous 50.59 Review remains valid.


o The proposed activity Is controlled by another regulation per Section 5.6.1.4.

B. BasisProvide a clear, concise basis for determining the proposed activity may be exempted such that a third-partyreviewer can reach the same conclusions. See Section 5.6.6 of the EOI 10CFR50.59 Review ProgramGuidelines for guidance.

N/A


A. Executive Summary (Serves as Input to NRC summary report Limit to one page or less. Send an electroniccopy to the site licensing department after OSRC approval, if available.)


Temp Alt 04-005 and 04-006 will allow manual control of SSW cooling tower fans during ShutdownCooling for plant Modes 4 and 5. The manual control of the SSW system and fans will control theheat removal capability when the fans are secured during cold weather which will aid in maintainingthe basin water temperature within administrative limits. The SSW fans will be cycled to maintainSSW temperature between 50'F and 75'F. SSW temperatures are indicated/recorded on severalInstruments In the control room and logged hourly. SSW fans can be controlled in the control room.This temp alt adds minimal operator responsibility.

The Temp Alt will bypass the RHR pump (A or B) signal to the associated SSW Initiation logic.Therefore, the SSW division for the affected RHR pump will not start automatically In response tothe manual start of the RHR pump. The affected division of SSW will respond as designed to allother automatic Initiation signals. There are no automatic Initiation signals that only Initiate the RHRpumps. The Temp Alt will Include Instructions that require manual alignment of SSW to the RHRpump prior to starting the pump with the Temp Alt Installed. Operation procedures currently requiremanually starting and aligning SSW prior to starting a RHR pump for shutdown cooling. Alloperation of the RHR pumps and SSW system can be performed from Ihe control room. Thereforethere Is no additional burden placed on operations by implementing the temporary alterations.There Is no impact on the operation of the remote shutdown panel (except that a RHR pump startwill not auto start SSW).


During shutdown cooling operations and cold weather, SSW temperature may approach theadministrative temperature limits for water In the SSW basin. By providing a method to runshutdown cooling with the ability to start and stop fans manually, SSW basin temperature can moreeasily be controlled within administrative limits.


Bypassing the RHR pump (A or B) signal to the SSW Initiation logic Is an acceptable method ofallowing manual control of SSW system and fans during shutdown cooling in mode 4 or 5 duringcold weather operation. The SSW system will still perform required safety related functions with theTemp Alts installed, with the exception of automatic initiation In response to RHR A or B pump start.Manual operation of SSW will be required to ensure SSW Is properly lined up to the RHR pump forshutdown cooling, as is currently done per the procedure for shutdown cooling, 04-1-01-E12-2, Allother automatic initiation signals will cause SSW to automatically Initiate as designed. The ability ofthe RHR system to automatically respond to ECCS signals Is also not affected. Relying on manualoperation to cool the RHR pumps for shutdown cooling does not pose more than a minimal increaseIn risk to the ability of the RHR system to perform Its design safety function.


Does the proposed Change being evaluated represent a change to a method of 0 Yesevaluation ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only 0 NoQuestion 8. If "No," answer all questions below.


1. Result In more than a minimal'increase in the frequency of occurrence of an accident 0 Yespreviously evaluated In the FSAR? o No

BASIS:

As mitigating systems, the SSW system and UHS are not the direct subject of any UFSAR Chapter15 accident analyses. The Temp Alt bypasses the RHR pump (A or B) manual start signal In theassociated SSW division initiation logic. All other Initiation signals remain in place and will functionas designed. The temp alts do not bypass or disable any other initiation signals from causing anautomatic initiation to occur as designed. There Is no automatic Initiation signal that will only startthe RHR pump. Any automatic initiation signal will also start other ESF equipment and result In theInitiation of SSW as designed. These temp-alts are only applicable for the manual alignment of theShutdown Cooling Mode of RHR during plant modes 4 and 5.Although the RHR pump manual start initiation signal Is bypassed, these temp alts will notadversely affect the function or operation of the SSW system or the interfacing systems and will notcompromise any safety related system, structure or component. The proposed change does notchange the applicable SSW system or UHS design (other than the RHR AIB pump automatic SSWinitiation), material, or construction standards, does not degrade overall SSW system or UHSreliability, and does not cause the SSW system or UHS to operate outside design limits. Thus, thetemporary alterations do not result in more than a minimal increase in the frequency of occurrenceof an accident previously evaluated In the FSAR.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction 0 Yesof a structure, system, or component Important to safety previously evaluated In the o NoFSAR?

BASIS:

The shutdown cooling mode of RHR Is a manually aligned safety related function that requiresproper functioning of both the RHR system and SSW system. The temp-alts provide means tomanually control the SSW fans during Shutdown Cooling Mode of RHR In plant modes 4 and 5 inorder to maintain SSW basin temperature within administrative limits. This would be achieved bybypassing the RHR pump start Initiation signal in the SSW logic. This will provide the ability tomanually start and stop the SSW fans from the control room, provided there is no other SSWinitiation signal present. Operator action to manually align SSW to the RHR system prior to startingthe RHR pumps is being relied upon to provide cooling to the RHR pumps (pump protection). If thisaction was not performed, a potential exist to damage the RHR pumps and lose associated decayheat removal capability. The likelihood of this occurrence is no greater than the likelihood ofoccurrence of failure to perform similar operator actions to manually align SSW for shutdowncooling since the current procedures for shutdown cooling requires that SSW be In service prior tostarting the RHR pumps. Thus, installation of the temp alts would not result in more than a minimalincrease in the likelihood of a malfunction of the RHR pumps or system.

The possibility for a human error allowing SSW basin temperature to increase above the designlimit of 90OF also requires evaluation. Allowing basin temperature to exceed 90'F Is unlikely. TheSSW SOI 04-1-01-P41-1 requires SSW temperature to be maintained at or below 850F duringnormal operation. RHR SOI 04-1-01-El 2-2 requires that SSW temperature be logged hourly during

EFRter 1 FORM TEMP ALT 04-005 04-006 Page 12 |of 14

shutdown cooling operation. SSW temperature exceeding 90¶F is not considered credible since theTemp Alts are intended only to be used in cold weather and with minimal heat load. Temperaturelimits have been specified for cycling the SSW fans to ensure that the administrative and designlimits are maintained.If an operator were to fail to shutdown the SSW fans prior to the temperature dropping below 50Fthen the SSW system would continue to function as currently designed. The upper limit (75F) Issufficiently low that if operators failed to manually start the fans prior to reaching 75*F there wouldbe multiple opportunities to realize this mistake prior to reaching the current administrative limit of859F. Although, given the low heat load in modes 4 and 5, and restriction on use in cold weather, itIs unlikely that the administrative limit of 85"F could be reached at all.

SSW temperature Indication and fan controls are located in the control room, therefore, thistemporary alteration does not Introduce additional operator burden.

The fans will still start upon an automatic initiation signal with the exception of a RHR pump start,thus maintaining basin temperature below the design limit of 90 degrees. Therefore, the operabilityof SSWIUHS Is not challenged by operating the fans manually In Mode 4 and 5 since there Isadequate assurance that temperature will be maintained within limits.

Therefore, Installation of the temp alts do not result more than a minimal Increase In the likelihoodof a malfunction of the SSW or RHR system (components, etc) since It will have no adverse affecton any equipment which Is Important to safety and does not cause the RHR and SSW system(including components) to operate outside design limits.

3. Result in more than a minimal Increase in the consequences of an accident previously 0 Yesevaluated In the FSAR? 0 No

BASIS:

There is no Increase In the consequences of an accident previously evaluated In the FSAR. Thetemp alts aids operation in maintaining the SSW basin temperature within administrative limits bymanually controlling the SSW fans during shutdown cooling mode of RHR In cold weather operationand In mode 4 and 5. The SSW fans would operate as designed with all automatic Initiation signalswith the exception of a manual RHR pump start. The SSW system would still be able to perform Itsdesign safety function since the system will still be operating within t design limits. Any signalgenerated by an accident (e.g. drywell pressure, reactor water level) would result In SSW Initiatingas designed. Therefore, any accident would have consequences bounded by those currentlyanalyzed within the FSAR.

With the temp alts Installed, the automatic pump protection feature for the RHR system Is bypassedfor RHR pumps A and B. Operator action to manually align SSW to the RHR system prior tostarting the RHR pumps Is being relied upon to provide cooling the RHR pumps (pump protection).However, the likelihood of an operator failing to perform this action is no greater than the likelihoodof an operator failing to perform this action under current operating conditions. If this action was notperformed, a potential exist to damage the RHR pumps and lose decay heat removal capability. Thelikelihood of this occurrence is very low since the current procedures for shutdown cooling requiresthat SSW be In service prior to starting the RHR pumps. In the unlikely situation that this occurred,it would not result In a different consequence than those already evaluated In the UFSAR. TheTechnical Specifications provide requirements for adequate decay heat removal capabilities duringmodes 4 and 5 to ensure that component failures do not result In a total loss of decay heat removalcapability.

50.59 REVIEWFORM TEMP ALT 04-005 & 04-008 Page 13 ofj 14

4. Result In more than a minimal Increase In the consequences of a malfunction of a 0 Yesstructure, system, or component important to safety previously evaluated In the 1 NoFSAR?

BASIS:

With the Temp Alt Installed, the SSW system will respond as designed to all automatic Initiationsignals with the exception of the RHR pump start. Operator action will be relied upon to manuallyalign SSW to RHR prior to starting the RHR pumps. If this action was not performed, a potentialexists to damage the affected RHR pump and lose the associated decay heat removal capability.However, the likelihood of an operator failing to perform this action Is no greater than the likelihoodof an operator failing to perform this action under current operating conditions. The currentprocedures for shutdown cooling require that SSW be in service prior to starting the RHR pumps.A pump failure is an active failure that Is already considered in station design. No new Initiatingconditions or design operating parameters are allowed by these Temp Alts.

The possibility for a human error allowing SSW basin temperature to Increase above the designlimit of 90°F also requires evaluation and Is discussed In more detail In question 2 above.

Thus, the Installation of the temp alts do not result In more than an Increase In the consequences ofa malfunction of a structure, system, or component Important to safety previously evaluated In theFSAR.

5. Create a possibility for an accident of a different type than any previously evaluated In 0 Yesthe FSAR? 0 No

BASIS:

The temp alts aids operation in maintaining the SSW basin temperature within administrative limitsby manually controlling the SSW fans during shutdown cooling mode of RHR In cold weatheroperation with the plant In mode 4 or S.The temporary alterations do not disable the essential automatic SSW responses to ECCS andEDG equipment. The SSW system will operate within current design basis requirements and thesetemp alts do not create the possibility for a different type of accident than previously evaluated In theFSAR.

With the temp alts installed, the automatic pump protection feature for the RHR system Isbypassed. Operator action to manually align SSW to the RHR system prior to starting the RHRpumps Is being relied upon to provide cooling the RHR pumps (pump protection). A potential existto damage the RHR pump if this action Is not performed, this could lead to a loss the associateddecay heat removal capability. It Is not likely to occur since current procedures for shutdowncooling requires SSW to be In service prior to starting the RHR pumps.

Also, loss of an RHR pump (loss of decay heat removal capability) is already analyzed. Thus,installation of the temp alts does not create the possibility of an accident of a different type than anypreviously evaluated In the FSAR.

TEMP ALT 04-005 & 04-006 Pagle| 14 of 14

6. Create a possibility for a malfunction of a structure, system, or component Important to 0 Yessafety with a different result than any previously evaluated In the FSAR? 0 No

BASIS:

The temp alt does not create a new failure mode. By removing the automatic SSW start on amanual RHR pump start the potential exists to operate an RHR pump without adequate cooling.However, current procedural guidance reduces the likelihood of this occurrence. Current operatingprocedures require manually starting and aligning SSW prior to manually starting the RHR pump.This temp alt removes the built In operational backup should procedural guidance not be followed.The active failure of an RHR pump Is already evaluated and is part of the reason for backup decayheat removal capability. Therefore, the possibility of a malfunction of a different result thanpreviously evaluated does not exist.

7. Result In a design basis limit for a fission product barrier as described In the FSAR 0 Yesbeing exceeded or altered? No

BASIS:

The Temp Alts do not adversely affect any fission product barrier. The Temp Alts will allow manualoperation of the SSW system to maintain SSW water temperature within administrative limits duringshutdown cooling mode of RHR while In plant mode 4 and 5. The SSW basin temperature will bemaintained within design limits as specified by Tech Specs and TRM requirements. Thus sincedesign limits will be maintained, SSW will perform as designed to protect the fission product barrier.

8. Result In a departure from a method of evaluation described In the FSAR used In 0 Yesestablishing the design bases or In the safety analyses? 0 No

BASIS:

The Temp Alts will not allow SSW to be operated outside of administrative limits, which are withinexisting design limits. No design bases are being changed as a result of the temp alt. The manualcontrol of the SSW fans during shutdown cooling will help maintain the SSW water temperaturewithin administrative limits. The Temp Alts do not make any changes to the methodology forestablishing any design bases, nor do they make a change In establishing a safety limit.

GGNS Commitment Change EvaluationNumber

CCE 2002-007

COMMITMENT CHANGE EVALUATION FORMCommitment Number: A-29293 IPlant Licensing Tracking Number: ICC- 2.007- Oo7

Source Document: i0CFR.Part 5545

Commitment: Deletion? 1 Revision? -Has the original commitment been Implemented? I L YES I C NO, Notify Plant LicensingOriginal Commitment Description:Perform transient tests and steady state tests on the simulator annually

Revised Commitment Description:

Summary of Justification for Change or Deletion:IOCFR55 45 was revised earlier this year and deleted a requirement to report on simulator test failures and theirresolution and also deleted a requirement to submit the test plan for the upcoming four yearsEntergy has implemented the 1998 version of ANSI 3 5 and issued TQ-202,Simulator Configuration Control, Rev IThis procedure implements the testing requirements of ANSI 3 5 as it applies to GGNSANO,RBS, and WF3

(Attach additional sheets if necessary)Refer to Attachment 9.4 for a flow diagram that outlines the commitment change evaluation process

Prepared By: S R Kaskie R 2 -02a

_ Print Name/Signature DateManagement

Approval: 6,&/ 6IMma &,4, okJZ/i2t

Print Name/S ign reatePlant Licensing

Management

Print Nam/cSienature Date

10 oefL -S'5S t 6-'- ,(-. 4./

r owtOA nrcoRD

V ,

V -

REF U*110

PART I1.1 Is the existing commitment located In the Updated Final Safety Analysis Report, Emergency Plan, Quality

Assurance Program, Fire Protection Program, or Security Plan?

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

3 NO Go to Part 11.PART II

2.1 Could the change negatively impact the ability of a System, Structure, or Component (SSC) to perform Itssafety function or negatively Impact the ability or plant personnel to cnsure the SSC Is capable ofperforming its Intended safety function?

C YES Go to Question 2.7.

2 2 Perform a safety evaluation using the following 10 CFR 50.92 criteria to determine If a significant hazardsconsideration exists:

Does the revised commitment involve a significant Increase in the probability or consequences of anaccident previously evaluated?

C3 YESBasis:

C3NO

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

E YES aNOBasis:

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

OYES ONOBasis:

If any or the above questions are answered Yes, STOP. Do not proceed with the revision , OR discusschange with NRC and obtain necessary approvals prior to implementation of the proposed change. If allthree questions are answered NO, go to Part HI(Attach additional sheets as necessary.)

REF LI-110

-PART III

3.1 Was the original commitment (e.g., response to NOV, etc.) to restore an Obligation (ie., rule, regulation,order or license condition)?

O E YES Go to question 3.2.

0 NO Go to Part IV.

3.2 Is the proposed revised commitment date necessary and justified?

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

5 NO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relief.

PARTIV4.1 Was the original commitment: (1) explicitly credited as the basis for a safety decision m an NRC SER, (2)

made In response to an NRC Bulletin or Generic Letter, or (3) made in response to a request forinformation under 10 CFR 50.54(Q or 10 CFR 2.204?

a YES Go to Question 4.2.

O NO Go to Part V

4.2 Has the original commitment been implemented?

O YES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment In summary report

ONO Go to Question 5.1.

PART V5.1- NA-as the ornginsl commitment made to minimize recurrence offa condition adverse to quality (e.g., a Jong-

term corrective action stated In an LER)?

D YES Go to Question 5.2.

0 NO STOP. You have completed this evaluation. Revise the commitment. No NRC notificationrequired

REF U-10

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 annualIRFO

interval summary report.

0 NO Revise commitment: no NRC notification Is required:

REFERENCESetc) affected by

REF U-11O


CCE 2002-008

COMMITMENT CHANGE EVALUATION FORMCommitment Number. P-32395 Plant Licensing Tracking Number' IC

Source Document: AECM-90/0004

Commitment: Delehon? 4 Revision? UHas the onginal commitment been implemented? I -YES I U NO, Notify Plant LicensingOriginal Commitment Description:PROCEDURAL WEAKNESSES IDENTIED BY RESPONSE TO SALP REPORT FOR INCLUSION OFAPPROPRIATE IMPROYEMENTS INTO THE I1CFR50 59 PROCESS

Revised Commitment Description:Completcly remove the commirutment

Summary of Justification for Change or Deletion:There is no regulatory requirement or basis for this comnutiient

(Attach additional shects if necessary)Refer to Attachment 9.4 for a flow diagram that outlines the commitment change evaluation process.

Prepared By. Guy H Davant A 9/27/02

. Mrint Name/Siiture DateManagement

Approval: k.6 9*27.OZ

Print Name/StWAture DatePlant Licensing

Management (Concurrence: _ ( -2'J, , . IL4. 10-14.l_

Print Name/Sianature Date

A ECC.

QAO 0M~ -A OOUMSER

REF LI-1O

PART I1.1 Is the existing commitment located in the Updated Final Safety Analysis Report, Emergency Plan, Quality


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

_ NO Go to Part 1LPART II

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

O YES Go to Question 2.2

- NO Continue with Part 1lL Briefly describe rationale:Deleting this conmmitment has no impact on the safety function of any SSC There is no regulatory requirementor basis to establish a S0 59 Rcview Program Tius commitment was madc in response to a SALP report thatstated," 10CFR5O 59 evaluations and events continue to be weak"

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

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

D YES a NOBasis:

Does the revised commitment create the possibility of a new or different land of accident from anypreviously evaluated?

D YES a NO

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

DYES 3NOBasis:

If any of the above questions are answered Yes, STOP. Do not Proceed with the revision , OR discusschange with NRC and obtain necessary approvals prior to implementation of the proposed change. If allthree questions are answered NO, go to Part 111.(Attach additional sheets as necessary.)

REF LI-1tO

PART HI

3.1 Was the original commitment (cg., response to NOV, etc.) to restore an Obligation (Le-, rule, regulation,order or license condition)?

o YES Go to question 3.2.

¢ NO Go to Part IV.


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

Rationale:

] NO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relief.

PART IV4.1 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 forinformation under 10 CFR 50.54(f) or 10 CFR 2.204?

o YES Go to Question 4.2.

- NO Go to Part V.


o YES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment in summary report.

0 NO Go to Question 5.1.

PART V5.1 Was the original commitment made to minimize recurrence of a condition adverse to quality (eg., a long-

term corrective action stated in an LER)?


4 NO STOP. You have completed this evaluation. Revise the commitment. No NRC notificationrequired

REF U-110

S.2 Is the revised commitment necessary to minimize recurrence of the condition adverse to quality?Ql YES Revise the commitment and notify NRC of revised commitment in next annuallRFO

interval summasy report.

O NO Revise commitment: no NRC notification Is required:

REFERENCESList documents (e.z, procedures, NRC submittals, etc.) affected by this chanme.

Doc. Number DescriptionAECM-90/0004 Rcsponsc to SALP Report, dated November 22, 1989

10 CFR 50 59, Changes, Tests, andExperiments Regulation

NEI 96-07, Guidelinesfor lO CFR 50 59 Implementation Industry guidance document endorsed by the NRC m RegGuide 1187

REF L.-110


CCE 2002-009

I

COMMITMENT CHANGE EVALUATION FORM C CCCommitment Number: P-23936 Plant Licensing Tracking Number: 1o2 b2 - o

Source Document: AECM-821490

Commitment: Deletion? LJ Revision? 1Has the original commitment been implemented? YES | NO, Notify Plant LicensingOriginal Commitment Description:Valve P53F006 of the ADS SYS will undergo valve seat leakage test every 18 mos.

Revised Commitment Description: teSlValve P53F006 of the ADS SYS will undergo valve seat leakage~at a frequency not to exceed once every 60 months.

Summary of Justification for Change or Deletion:The local leak rate test program was changed to a performance based testing program to be consistent with NRC approvedindustry guidance. The change to the program was approved under SE 96-0099-ROO.A performance based testing program allows for a test frequency of 60 months, changed from 18/24 month frequency, forvalves that pass two consecutive local leak rate test. P53F006 has passed two consecutive tests and is currently on a 60month frequency.

(Attach additional sheets if necessary)Refer to Attachment 9.4 for a flow diagram that outlines the commitment change evaluation process.

Prepared By: , V 0i;) ) $ )/6/ZY2

Peint Name/Signature DateManagement

Approval:

-Print Name/Signature DatePlant Licensing

Management g ^ {

Concurrence. (41 -e 6

Print Name/Slenature Date

- OA LE6D. -

|NON 0A RECORD -

LATED ER_:EX CSEt

REF: LI-11O

PART I

1.1 Is the existing commitment located In the Updated Final Safety Analysis Report, Emergency Plan, QualityAssurance Program, Fire Protection Program, or Security Plan?

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

1 NO Go to Part II.PART 11

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


ED NO Continue with Part III. Briefly describe rationale:Changing the test frequency has no impact on the safety function of any SCC. There is no regulatoryrequirement to perform a seat leakage test every 18 months. SE 96.0099-ROO documented the approval of goingto a performance based test program. The frequency for P53F006 was/is determined using the criteria set forthin a performance based test program.

2.2 Perform a safety evaluation using the following 10 CFR 50.92 criteria to determine If a significant hazardsconsideration exists:

Does the revised commitment involve a significant Increase in the probability or consequences of anaccident previously evaluated?

El YES 3 NO


1 YES 0NOBasis:A test frequency change to the original commitment does not create the possibility of a new or different kind ofaccident from any previously evaluated.

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

E YES 0NOBasis:Revising the commitment does not involve a significant reduction in a margin of safety.

If any of the above questions are answered Yes, STOP. Do not proceed with the revision, OR discusschange with NRC and obtain necessary approvals prior to implementation of the proposed change. If allthree questions are answered NO, go to Part fi1.(Attach additional sheets as necessary.)

REF: LI 10

PART III

3.1 Was the original commitment (e.g., response to NOV, etc.) to restore an Obligation (Le., rule, regulation,order or license condition)?

a YES Go to question 3.2.

0NO Go to Part IV.


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

Rationale:

o NO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relieL

PART IV4.1 Was the original commitment: (I) 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 forinformation under 10 CFR 50.54(0) or 10 CFR 2.204?

O YES Go to Question 4.2.

0NO Go to Part V.


D YES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment In summary report.

O NO Go to Question 5.1.

PART V5.1 Was the original commitment made to minimize recurrence of a condition adverse to quality (e.g., a long-



0 NO STOP. You have completed this evaluation. Revise the commitment. No NRC notificationrequired.

REF: 14 110

5.2 Is the revised commitment necessary to minimize recurrence ofthe condition adverse to quality?El YES Revise the commitment and notify NRC or revised commitment In next annuaIRFO


ONO Revise commitment: no NRC notification is required:

REFERENCESList documents (e.g., procedures, NRC submittals, etc.) affected by this change.

Doe. Number DescriptionCommitment P-23936 P53F006 (AD System valve) will undergo valve seat leakage

at a frequency not to exceed once every 60 months.

REF: LI-1IO


CCE 2003-001

COMMITMENT CHANGE EVALUATION FORMCommitment Number: 24290 PlantLicensingTracking Number: IcE c 300-o

Source Document: AECM-88/0024 Att. 1 Area 5.S3

Commitment: Deletion? 3 Revision? LHas the original commitment been implemented? I ;YES I EJ NO, Notify Plant LicensingOriginal Commitment Description:Plant Staff revised its drawing control procedure to provide for ready acceptance of NPEreviewed, approved and issued drawings and the assignment of the drawing coordinator tocoordinate design drawings and resolve drawings problems.

Revised Commitment Description:None

Summary of Justification for Change or Deletion:This process is no longer applicable, documents do not go to a pending file they aretransmitted directly to records and within a 24 hour time frame records Issues andpopulates EDMS or rejects and sends drawing/document back to NPE for correctionsand NPE has 30 days to resubmitted corrected drawingsidocuments.


Prepared By: I.

_ Print Name/Signature D__teManagement

Approval: ,L0 L ____/___1

-Prdlt Name/Signature D atePlant Licensing .

Management DConcurrence:

IPrint Name/bIgnature Date

ad.,_" , _

I.-.- .I . -.

I ,I.. t,.,,

Vr.

; t,..,'ti

I Fi>,E:o,,C:u~~

REF: LI- 10



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

3 NO Go to Part IL.PARTII

2.1 Could the change negatively Impact the ability ora System, Structure, or Component (SSC) to perform Itssafety function or negatively impact the ability of plant personnel to ensure the SSC Is capable ofperforming its Intended safety function?


M NO Continue with Part lII. Briefly describe rationale:This was an old process that is no longer applicable.


Does the revised commitment Involve a significant Increase in the probability or consequences of anaccident previously evaluated?

E YESBasis:

iQ NO


DYES a NOBasis:


EYES QNOBasis:

If any of the above questions are answered Yes, STOP. Do not proceed with the revision, OR discusschange with NRC and obtain necessary approvals prior to Implementation of the proposed change. If allthree questions are answered NO, go to Part 111.(Attach additional sheets as necessary.)

REF: 1-1l1D

- PART III

3.1 Was the original commitment (e.g., response to NOV, etc.) to restore an Obligation (i.e., rule, regulation,order or license condition)?

O YES Go to question 3.2.

ED NO Go to Part TV.



Rationale:

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

* PARTIV4.1 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 forInformation under 10 CFR 50.54(t) or ID CFR 2.204?

Q YES Go to Question 4.2.

3 NO Go to Part V.

4.2 Has the original commitment been Implemented?

El YES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment in summary report.

]NO Go to Question 5.1.





REF: L-I IO

5.2 Is the revised commitment necessary to minimize recurrence of the condition adverse to quality?C YES Revise the commitment and notify NRC of revised commitment in next annual/RFO


QNO Revise commitment: no NRC notification is required:


Doc. Number Description

. ...

REF: U-110


CCE 2003-002

COMMITMENT CHANGE EVALUATION FORMCommitment Number: P-24401 | Plant Licensing Tracking Number: Le 2003-007-.

Source Document: AECM-89/0003.88-26-01.I1 2.b

Commitment: Deletion? ILi Revision? 0Hals the original commitment been Implemented? I l YES I H NO, Notify Plant LicensingOriginal Commitment Description:

PAP 01-S-06-15 has been revised to include detailed instructions for completing DOE/NRC form 741. The personnelresponsible for SNM receipt and completion of the form have been made aware of the requirement to fill out the form asrequired by NUREGIBR-0006.

Revised Commitment Descrintion:NMM NF-104 has been revised to include instructions for completing DOE/NRC form 741. .The personnel responsible forSNM receipt and completion of the form have been made aware of the requirement to fill out the form as required byNUREGIBR-0006.

Summary of Justitication for Change or Deletion:Under NMM procedure NF-104 rev. 2 (to be issued with an effective date of 7/1/2003), overall control of the Special NuclearMaterials Program is assigned to the Manager of the Nuclear Engineering Analysis department at Echelon. PAP 01-S-06-15will be superceded by NF-104. The SNM Management and Control process will be standardized, with NEAD responsible forall EN-S sites. Thus, this commitment will be met by NF-104.


Prepared By: -P24 Le mar-L /,2 1i2 A1'( Ži ,__|_______=_ ___,__ Print tame/Sig nature Date

Management 'Approval: /Ae~C / 1/ 47d9i L- 6 . oD%

Print Name/Signature DatePlant Licensing

Management >Concurrence: t6W) C ktr, ) I 9-'A an

Print Name/Signature Date

REF: LI-O0




1 NO Go to Part I.LPART 11

2.1 Could the change negatively Impact the ability of a System, Structure, or Component (SSC) to perform Itssarety function or negatively impact the ability of plant personnel to ensure the SSC Is capable ofperforming Its Intended safety function?

D YES Go to Question 2.2.

E NO Continue with Part III. Briefly describe rationale:Administrative change with no impact on SSC's.


Does the revised commitment Involve a significant increase In the probability or consequences of anaccident previously evaluated?

E YES E JNO


a3 YES ONO


El YESBasis:

El NO

If any of the above questions are answered Yes, STOP. Do not proceed with the revision, OR discusschange with NRC and obtain necessary approvals prior to implementation of the proposed change. If allthree questions are answered NO, go to Part III.(Attach additional sheets as necessary.)

REF: U-1.10

PART III


DYES

E NO

Go to question 3.2.

Go to Part IV.

3.2 Is the proposed revised commitment date necessary and Justified?

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

Rationale:

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

PARTIV4.1 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 forinformation under 10 CFR 50.54(0 or 10 CFR 2.204?


ED NO Go to Part V.


O YES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment In summary report.


PART V


0 YES Go to Question 5.2.

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

REF. L-1 10

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 annual/RFOInterval summary report.

1 NO Revise commitment: no NRC notification Is required:

REFERENCESList documents (e.,., procedures, NRC submittals, etc.) affected by this change.

Doc. Number Description01-S-06-15 GGNS SNM procedure. Procedure will be superceded by

NF-104, rev. 2

REF: U-11I

GGNS Commitment Change Evaluation-Number

CCE 2003-003

COMMITMENT CHANGE EVALUATION FORMCommitment Number: P-24402,24403 I PlantLicensingTrackingNumber: | CC- 7o0

Source Document: AECM-89/0003.88-26-01.11.3 .a&b

Commitment: Deletion? Revision? OTias the original commitment been implemented? I 3YES I L NO, Notify Plant LicensingOriginal Commitment Description:

SERI has determined that the SNM Management and Control process could be streamlined by assigning complete control ofthe SNM program to the GGNS General Manager. This reassignment of responsibilities has been implemented consolidatingactivities under one department and location. Future SNM reports will be sent out by the GGNS General Manager removingthe NMM and the Nuclear Licensing Section from the SNM reporting process.

PAP 01 -S-06-15 has also been revised to reflect transfer of responsibility of the SNM program to the GGNS GeneralManager.

Revised Commitment Description:DELETE these commitments.

Summary of Justification for Change or Deletion:Under NMM procedure NF- 104 rev. 2 (to be issued with an effective date of 7/1/2003), overall control of the Special NuclearMaterials Program is assigned to the Manager of the Nuclear Engineering Analysis department at Echelon. PAP 01-S-06-15will be superceded by NF-I 04. The SNM Management and Control process will be standardized, with NEAD responsible forall EN-S sites. In addition, per NEI 99-04 guidelines, these are not commnitments.


Prepared By. -i4J IJ / 1G /3-3' PriAtName/Signature | _Date___ |

Management , ¢Approval: AQ J/ A j~ ) 6-/A-0

Print Na e/Signature DatePlant Licensing

Management c ( '1-° J

Print Name/Signature I Date

REF: U-IIO



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

0 NO Co to Part II.PART II

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

C YES Go to Question 2.2.

0 NO Continue with Part 111. Briefly describe rationale:Administrative change with no impact on SSC's.



D YES a NOBasis:

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

a YES D NO


D YESBasis:

ENO

If any of the above questions are answered Yes, STOP. Do not proceed with the revision, OR discusschange with NRC and obtain necessary approvals prior to implementation of the proposed change. If allthree questions are answered NO, go to Part HI.(Attach additional sheets as necessary.)

REF: LU-t1O

PART III


a] YES Go to question 3.2.

0 NO Go to Part IV.


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

El NO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relief.

PARTIV4.1 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 forinformation under 10 CFR 50.54(f) or 10 CFR 2.204?

El YES Go to Question 4.2.

[D NO Go to Part V.


Dl YES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment In summary report.

El NO Go to Question 5.1.



Z YES

DNO

Go to Question 5.2.

STOP. You have completed this evaluation. Revise the commitment. No NRC notificationrequired.

REF: L-110

5.2 Is the revised commitment necessary to minimize recurrence of the condition adverse to quality?Cl YES Revise the commitment and notify NRC of revised commitment in next annuaLlRFO


E NO Revise commitment: no NRC notification Is required:

REFERENCESList documents (e.c., procedures. NRC submittals. etc.) affected by this chance.

Doc. Number I Description01-S-06-15 GGNS SNM procedure. Procedure will be superceded by

NF-104. rev. 2

.4

.4

.4

.4

REF: U-LIO


CCE 2003-004

COMMITMENT CHANGE EVALUATIQN FORMCommitment Number: 25027 Plant LicensingTracking Number| CCE2003-0004

Source Document: GNRO-93/00029 Paragraph 3, Sentence 4

Commitment: Deletion? 3 Revision? LHas the original commitment been implemented? I g YES I D NO, Notify Plant LicensingOriginal Commitment Description:"Transmitters confirmed of having a loss of fill-oil failure will be dispositioned in accordance with our non-conformanceprocess or the applicable technical specification."

Revised Commitment Description:Delete commitment

Summary of Justification for Change or Deletion:Instruction 17-S-06-3 provides the means to track and trend Rosemount transmitters that are susceptible to fill-oil loss.GNRO-93/00029, Paragraph 3 states that trending is required to detect this condition until replacement or the appropriate psi-month threshold recommended by Rosemount is achieved. Grand Gulf has either replaced or achieved the psi-monththreshold for these transmitters (see attached). Tbis commitment has been has been implemented and is now complete.Current non-conformance programs will continue to identify component issues.

(Attach additional sheets if necessary)Refer to Attachment 9.4 for a flow diagram that outlines the commitment change evaluation process

Prepared By: Le Ec Hi t. 7

Vrint Name/Signature DateManagement Ax 91slo3

Approval: A *p ark70Print Name/Signatufe Date

Plant LicensingManagement

Concurrence: _ _ _____ _A_ _ _ _ _ _ _

Print Name/Signature I Date

/INCO*4O PRrco

MJa~ a PAGES

RaA1'w ~xC~i

REF: U-110



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

3 NO Go to Part If.PART II

2.1 Could the change negatively impact the ability of a System, Structure, or Component (SSC) to perform itssafety function or negatively Impact the ability of plant personnel to ensure the SSC is capable ofperforming Its intended safety function?


ID NO Continue with Part IlI. Briefly describe rationale:Susceptible Rosemount transmitters in service less than the psi-month threshold are vulnerable to fill-oil lossfailures. GGNS achieved the psi-month threshold for all susceptible transmitters. Therefore, removing thetrending requirements will have no negative impact (See Attached)



[J YESBasis:

[]NO


Q YES a NO


I] YESBasis:

[]NO

I

If any of the above questions are answered Yes, STOP. Do not proceed with the revision , OR discusschange with NRC and obtain necessary approvals prior to implementation of the proposed change. If allthree questions are answered NO, go to Part III.(Attach additional sheets as necessary.)

REF: LI-110

PART III

3.1 Was the original commitment (e.g., response to NOV, etc.) to restore an Obligation (iLe., rule, regulation,order or license condition)?


ONO Go to Part V.


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

Rationale:

O NO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relief.

PART IV4.1 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 Lettcr, or (3) made in response to a request forinformation under 10 CFR 50.54(1) or 1 0 CFR 2.204?


ONO GotoPartV.


3 YES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment In summary report.





o NO STOP. You have completed this evaluation. Revise the commitment. No NRC notificationrequired.

REF: U-110

5.2 Is the revised commitment necessary to minimize recurrence of the condition adverse to quality?Q YES Revise the commitment and notify NRC of revised commitment In next annuai/RFO


QNO Revise commitment: no NRC notification is required:


Doc. Number DesKcri~flonGNRO-93/00029 Response to NRC Bulletin 90-01, Supplement I

17-S-06-3 Rosemount Enhanced Monitoring Program

REF: LI.110

: *

Attachment to Commitment Change EvaluationCCE 2003-0004

BackgroundCommitment #25027 is addressed in GNRO-93/00029, Paragraph 3, sentence 4. It isimplemented within directive 17-S-06-3. The committed sentence states "Transmittersconfirmed of having a loss of fill-oil failure will be dispositioned in accordance with ournon-conformance process of the applicable technical specification." 17-S-06-3 is themeans to trend these susceptible transmitters. GNRO-93/00029, Paragraph 3 also states:"This program will remain in effect, with the transmitters listed in Tables 2, 3, and 4

being monitored on an 18-month (cycle) basis until replacement or the appropriate psi-month threshold recommended by Rosemount is achieved. The transmitters listed inTable 1 will be replaced during the next refueling outage, currently scheduled for October1993."

This attachment will evaluate if the remaining PSI-Months has been achieved, thatGGNS is no longer susceptible to this type failure, and that the commitment is no longerapplicable.

Transmitter ScopeNOTE: Reference GNRO-93/00029 for more table information.Table 1 transmittersAccumulated 53,450 PSI-Months as of May 1993130,000 PSI-Month ThresholdWorst case 76,550 PSI-Months remaining as of May 1993NOTE: All five transmitters were replaced. No further actions are required for thesetransmitters.

Transmitter Replacement WO Date1E31N086C 98003 10/6/93IE31N089A 98308 and 10/6/93

141807 4/18/95IE31N089B 98313 10/5/93IE31N089C 98314 10/1/931E31N089D 98318 10/4/93

Table 2 transmittersAccumulated 53,320 PSI-Months as of May 199360,000 PSI-Month ThresholdWorst case 6,680 PSI-Months remaining as of May 1993

Page 1 of 3

Table 3 transmittersAccumulated 53,320 PSI-Months as of May 199360,000 PSI-Month ThresholdWorst case 6,680 PSI-Months remaining as of May 1993

Table 4 transmittersLowest accumulated transmitter is 6,640 PSI-Months as of May 199360,000 PSI-Month ThresholdWorst case 53,360 PSI-Months remaining as of May 1993

On-Line DataOn-line data was referenced from the Unit's Gencration Performance Indicators toacquire the number of GGNS on-line hours since May 1993. The affected transmittersare pressurized whenever the unit is on-line. Therefore, on-line hours will be used todetermine when these transmitters were pressurized.

Year Hours On-Line Comments

1993 starting in June 3223 6846 hrs (1993 total) - 3623 hrs (hrs before June)1994 82861995 68321996 76981997 87601998 76421999 69462000 86342001 80412002 81402003 to July 17th 4566

78,768 total

Assumptions* All Rosemount transmitters requiring trending by GNRO-93/00029 are sensing

either Main Steam Line (MSL) pressure or Reactor pressure. Reactor and MSLpressure is approximately 1030 psig or above when the unit is synchronized to thegrid. To ensure the calculation is conservative 1000 psig will be used.

* Since particular off-line months were not determined the calculation will use 31days per month. This wNill ensure a conservative calculation.

Page 2 of 3

CalculationThis calculation will determine the cumulative PSI-Months since the GNRO-93/00029was issued on May 1993:

Total on-line hours from June 1993 to July 17, 2003 = 78,768 hours (see summed hoursabove)On-line months = 78,768 hrs /24 hrs / 31 months = 105.9PSI-Months = 1000 psig * 105.9 on-line months = 105,900

105,900 PSI-Months have accumulated since May 1993. This is well above theremaining 53,360 PSI-Months necessary to satisfy the trending requirements of GNRO-93/00029. Therefore, trending transmitters identified in GNRO-93/00029 is no longerrequired, and commitment #25027 is no longer applicable.

Performed by

Reviewed by

vp -. /7/zwo 3I ?>6

Page 3 of 3


CCE 2003-005



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

3 NO Go to Part 11.PART 11

2.1 Could the change negatively impact the ability of a System, Structure, or Component (SSC) to perform Itssafety function or negatively impact the ability of plant personnel to ensure the SSC Is capable ofperforming its Intended safety function?

E| YES Go to Question 2.2.

C NO Continue with Part m. Briefy describe rationale:The change allows whoever is responsible for what is stored of itemsmaintaining an inventory of what is stored in the pools.


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

E YESBasis:

ONO


E YES E NOBasis:

Does the revised commitment Involve a significant reduction In a margin of safety?

0 YES ONOBasis

If any of the above questions are answered Yes, STOP. Do not proceed with the revision, OR discusschange with NRC and obtain necessary approvals prior to implementation of the proposed change. If authree questions are answered NO, go to Part III.(Attach additional sheets as necessary.)

REF: U-110

COMMITMENT CHANGE EVALUATION FORMCommitment Number: P-29077 Plant Licensing Tracking Number: CCE2003-0005

Source Document:GNRO 94/103.94-13-Ol.HIIS2

Commitment: Deletion? U Revision? DQHas the origin al commitment been Implemented? | YES I D NO, Notify Plant LicensingOriginal Commitment Description:The miscellaneous material storage procedure was revised to require copies of the inventory sheets and item location maps tobe maintained by Health Physics personnel.

Revised Commitment Description:The miscellaneous material storage procedure was revised to identify the personnel who arc required to maintain copies ofthe inventory sheets and item location maps for items stored in the pools.

Summary of Justification for Chnnge or Deletion:Original violation was that written records of items stored in the pools were not available to inform workers of hazardspresent for items stored in pools. At the time Health Physics was assigned the responsibility of maintaining the inventory.Currently Reactor Engineering is responsible for all items in the refueling pools and has the responsibility to maintain aninventory. Following RF- 13 the Outage Group will be responsible for storage of miscellaneous items in the pools. Theissue was not properly maintaining an inventory available and not who was responsible for maintaining the inventory. Thechange meets the intent of the commitment by identifying in the procedure who is responsible for maintaining the list.(Attach additional sheets if necessary)Refer to Attachment 9.4 for a flow diagram that outlines the commitment change evaluation process

Prepared By: John D. Williams / ; 9 Q ,08-20-2003

____ _ N_ Name/Signature Date

Management f _

Approval: k. ?la0<X1A A ,~ 1/03Print Namc/Signature Date

Plant LicensingManagement a I L S

Concurrence: _ I & -.

I Print Name/Signature Date

_ RIOf atIq ) I

RI.

_A H UE N

RELATEt GCUEM

REF: U-110

PART III

3.1 Was the original commitment (e.g., response to NOV, etc.) to restore an Obligation (I.e., rule, regulation,order or license condition)?

E YES Go to question 3.2.

0 NO Go to Part [V.


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

El NO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relief.

PART IV4.1 Was the original commitment: (I) 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 forInformation under 10 CFR 50.54(f) or 10 CFR 2.204?


ED NO Go to Part V.


0 YES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment in summary report.


PART V5.1 Was the original commitment made to minimize recurrence of a condition adverse to quality (e.g, a long-


0YES

A] NO

Go to Question 5.2.


REF: Ut110

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 annuailRFO

Interval summary report.

JKNO Revise commitment: no NRC notification is required:

REFERENCESList documents (e.v.. procedures. NRC submittals. etc.) affected by this chanve.

i2o. Number Descrietion01-S-08-6 GGNS Radioactive Material Control Procedure. Assigned

Rx. Engineering Responsibility for inventory,17-S-02-300 GGNS SNM Movement and Inventory Control Procedure.

Has Attachment used for the inventory..17-S-02-301 GGNS NNM Movement and Inventory Control Procedure.

Has Attachment used for the inventory.

REF: U-t10


CCE 2003-007

---

COMMITMENT CHANGE EVALUATION FORMCommitment Number: p 346091 Plant Licensing Tracidng Number: -oOo7

Source Document:AErCM 85/0201 RE5PONSE 2, PAVAGRAPH 3

Commitment: Deletion? [0 Revision? ElHas the original commitment been Implemented? I M YES I E NO, Notify Plant LicensingOriginal Commitment Description:

THis COMITrMuJ1 SPEcI lIES THAr GGNS AOMINISTEATIViE PROCEDURE "DE1ERMItA410EN

OF SAFETY/O0ALITrY CLASSIFICATONS" WILL SE USED FOR GUIDANCE UNTIL THE MASTER

ECUIPMENT LIST DATABASE IS CEIELOPED-


Summary of Justification for Change or Deletion:

TH4vs COMMITMENT IS NO LONGER VALID. WE MA5rER EQUIPME91 LIST, AND WvS

Succ.Essovs (COSEMS) I-v4E SATISFIED 'ME REQUIREMENT rAX PRlOMPTED ElE

ISSUANCE OF TkiS coMmrMSKT.

(Attach additional sheets If necessary)Refer to Attachment 9A for a flow diagram that outlines the commitment change evaluation process.

Prepared By: E. J ,a _ & E R IZ Z003

Print Name/Signature DateIsanagement .

Approval:Cho es ();c (,(, , l -Zo

Print Name/Slntr DatePlant Licensing

Management _Concurrence: - __ ll g.


a PC RD M w

j IRFT- D M s7 1- -WNON A pE _Qi

I tNMA11 U INlr4 ,NsUMU4 Cd ftt;ES I !4

_DATE | | It \-nj

REIATEb DOCMNNWMBEW

REF: LW11I



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

i NO Go to Part 11.PART T

2.1 Could the change negatively Impact the ability of a System, Structure, or Component (SSC) to perform itssafety function or negatively impact the ability of plant personnel to ensure the SSC Is capable ofperforming Its Intended safety function?


dNO Continue with Part III. Briefly describe rationale:

THitS CoMmiMiV.T CONTAINS NO REQUiReME-TS OR CONTROLS rTMAr AVFEcr

OR MEtqTjOH4 ANY GGUaS -SC.s.

22 Perform a safety evaluation using the following 10 CFR 50.92 criteria to determine Is a significant hazardsconsideration exists:

Does the revised commitment involve a significant increase In the probability or consequences of anaccident previously evaluated?

D YESBasis:

El NO

|/A.


EYES ONOBasis:

N/A


ElYES NOBasis:

N/A

If any of the above questions are answered Yes, STOP. Do not proceed with the revision , OR discusschange with NRC and obtain necessary approvals prior to implementation of the proposed change. If allthree questions are answered NO, go to Part III.(Attach additional sheets as necessary.)

REF: U1-110

PART III

3.1 Was the original commitment (e.g., response to NOV, etc.) to restore an Obligation (L~e, rule, regulation,order or Ucense condition)?

O YES Go to question 3.2.

RNO Go to Part TV.


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

Rationale:

N/A

O NO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relief.


made In response to in NRC Bulletin or Gcneric Letter, or (3) made In response to a request forInformation under 10 CFR 50.54() or 10 CFR 2.204?

51'YES Go to Question 4.2.

C NO Go to Part V.


rYES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment In summary report.

| NO Go to Question 5.1.




a NO STOP. You have completed this evaluation. Revise the commitment. No NRC notificationrequired.

REF: LU-110

5.2 Is the revised commitment necessary to minimize recurrence of the condition adverse to quallty?a YES Revise the commitment and notify NRC of revised commitment In next annual/RFO


ENO Revise commitment: no NRC notification is required:

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

Doc. Number Description

I .

I

REF: U-t10


CCE 2003-008

COMMITMENT CHANGE EVALUATION FORMCommitment Number: |I.3 3Y77233 Plant Licensing Tracking Number: |4 ; nog

Source Document: |N ,z'F?-45O3. o*f" 33

Commitment: Deletion? n Revision?Has the original commitment been implemented? | U YES I NO, Notify Plant Licensing

Original Commitment Description: aj' 5sIG 71

CeJ }t**d /~4; -4-5 (W he Uxs-.O


4,2 -5-0-< 3 , 5, -1i 7t Cr '

Summary of Justification for Change or Deletion:

4i ,.4 - sC . ,9 rzcSsa(Attach additional sheets If necessary)Refer to Attachment 9.4 for a flow diagram that outlines the commitment change evaluation process.

Prepared By: of >_

- 7Print Name/Signature I7 [fateManagement

Approval: //34/o 3

Print Name/Signature DatePlant Licensing

MAnagement ( C.' ,) 0 2A.


OA AL D _ J

NON-CA nECOQJ

_ NrnALf rlNUJ1BXR d4 PAGES|I

RELATED OOCuSEN'rtMJMBEA -

PER LI-F1O

* PART I1.1 Is the existing commitment located In the Updated Final Safety Analysis Report, Emergency Plan, Quality


a YES STOP. Do not proceed with this evaluation. instead use appropriate codined process(e.g., 10 CFR 50.71(e), 10 CFR 50.54) to evaluate commitment.

(t NO Go to Part 11.PART .

2.1 Could the change negatively Impact the ability of a System, Structure, or Component (SSC) to perform Itssafety function or negatively Impact the ability of plant personnel to ensure the SSC Is capable ofperforming Its intended safety function?

E YES Go to Question 2.2.


Does the revised commitment Involve a significant Increase In the probability or consequences of anaccident previously evaluated?

E YES E1 NO


a YES ff'NoBasis:

I /V


E YES [jNOBasis:

If any of the above questions are answered Yes, STOP. Do not proceed with the revision, OR discusschange with NRC and obtain necessary approvals prior to Implementation of the proposed change. if allthree questions are answered NO, go to Part lil.(Attach additional sheets as necessary.)

REF:M L 10

PART III

3.1 Was the original commitment (e.g., response to NOV, etc.) to restore an Obligation (I.e., rule, regulation,order or license condition)?

E YES Go to question 3.2.

X@ NO Go to Part IV.


N S~friefly describe rationale (attach additional steets as necessary) and notify NRC of revisedcommitment date prior to the original commitment date.

O NO STOP. Do not proceed with the revision, OR apply for appropriate regulatory reilef.


made In response to an NRC Bulletin or Generic Letter, or (3) made in response to a request forinformation under 10 CFR50.54(f) or 10 CFR 2.204?


lNO Go to Part V.


.CZYES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment In summary report.


PA RT V5.1 Was the original commitment made to minimize recurrence of a condition adverse to quality (e.g, a long-



O NO STOP. You have completed this evaluation. Revise the commitment. No NRC notificationrequired.

REF. U-1 10

5.2 Is the revised commitment necessary to minimize recurrence of the condition adverse to quality?a YES Revise the commitment and notify NRC of revised commitment In next annuaVRFO


*!$N(Y Revise commitment: no NRC notification is required:

REFER ENCESList documents (e.g., procedures, NRC submittals, etc.) affcted by this change.

Doc. lNumber Mscrieflon

REF: U-11O


CCE 2003-009

COMMITMENT CHANGE EVALUATION FORMCommitment Number: 34432 Plant Licensing Tracking Number: CCE 2003-0009

Source Document: GNRI-99/00047

Commitment: Deletion? DQ Revision? aHas the original commitment been implemented? I N YES | NO, Notify Plant LicensingOriginal Commitment Description:Check valves E38FG02AIB and E38FO03A/B, have historically been disassembled and inspected on a sample basis per 06-ME-1E38-R-0001 to ensure Code compliance (ref. Valve relief request VRR-E38-01). The Closed safety position was testedon a Cold Shutdown frequency and the Open safety position was tested on a Refuel frequency.

Revised Commitment Description:Check valves E38FD02AJB and E3BFO03AJB will now be Full Stroke Opened and Closed on a Cold Shutdown Frequency(including refueling outages) per Ma-1988, Part 10 to ensure obturator travel to the Open and Closed safety positions.

Summary of Justification for Change or Deletion:Check valves E38FO02AIB and E38FO03A/B, have historically been disassembled and inspected on a sample basis per 06-ME-lE38-R-0001 to ensure Code compliance as documented in CEP-IST-1, 2 and 3 (ref. Valve Relief Request VRR-E38-01). This relief request will be deleted and the valves will now be Full Stroke Opened and Closed on a Cold ShutdoWnFrequency (including refueling outages) per Ma-1988, Part lto ensure obturaor travel to the safety postions.i1-4i7

1!4c,# bry- Gufq x t2-7410o5(Attach additional sheets If necessary)Refer to Attachment 9.4 for a flow diagram that outlines the commitment change evaluation process.

Prepared By: GaryD. Young _ i 12/11/03

Prin Namel ignat DateManagement

Approval: /j ido)6l/1t

_Pnt Name/Signature DatePlant Licensing

Management |Concurrence: '~ot.v. I


1 OA AE(A$6-._

NO'.OA AE&

_NUMBER . PAGES -OATE -

RELATED DoUMENTNUMSER.

REF. U-1 10



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

3 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 itssafety function or negatively Impact the ability ofplant personnel to ensure the SSC Is capable ofperforming Its Intended safety function?

El YES Go to Question 2.2.

I3 NO Continue with Part InI. Briefly describe rationale:


Does the revised commitment Involve a significant increase In the probability or consequences of anaccident previously evaluated?

E YESBasis:

ONO


DYES O NOBasis:


EYES ONOBasis:

If any of the above questions are answered Yes, STOP. Do not proceed with the revision, OR discusschange with NRC and obtain necessary approvals prior to implementation of the proposed change. If allthree questions are answered NO, go to Part IIL(Attach additional sheets as necessary.)

REF: U.110

PART III

3.1 Was the original commitment (e.g., response to NOV, etc.) to restore an Obligation (i.e, rule, regulation,order or license condition)?


3 NO Go to Part IV.


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

Rationale:

a NO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relief.


made In response to an NRC Bulletin or Generic Letter, or (3) made In response to a request forinformation under 10 CFR 50.54(f) or 10 CFR 2.204?


Z NO Go to Part V.


o YES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment In summary report.

a NO Go to Question 5.1.




C NO STOP. You have completed this evaluation. Revise the commitment. No N`RC notificationrequired.

REF: LI-i1

52 Is the revised commitment necessary to minimize recurrence of the condition adverse to quality?C YES Revise the commitment and notify NRC of revised commitment In next annuaIIRFO


ElNO Revise commitment: no NRC notification is required:

REFERENCESList documents (e.y.. nrocedures. NRC submittals. etc.) affected bv this chance.

Doe. Number Description

J.

REF: L-1hI


CCE 2003-0010

COMMITMENT CHANGE EVALUATION FORMCommitment Number: A-35207 Plant Licensing TrackingNumber: Cfc-l e oo -00 0

Source Document: GNRO-02/00054

Commitment: Deletion? 0 Revision? UHas the original commitment been Implemented? i L YES i 1 NO, Notify Plant LicensingOriginal Commitment Description:Prior to RF13, GGNS will perform a parametric study at the uprated conditions to quantify the impact of TPO (ThermalPower Optimization) on GGNS wear rates and update the CHECWORKS model as necessary.


Summary of Justification for Change or Deletion:See attached.

(Attach additional sheets if necessary)Refer to Attachment 9A for a flow diagram that outlines the commitment change evaluation process.

Prepared By: gr? a ' c ,z- z-

Print Mne/Signature _ Date

ManagementApproval: 0. 8'n' 1n m'% % C

Print Name/Slgnature Date

Plant LicensingManagement q-1 -Concurrence:

Print Nane/rinature Date

OAA FrRRT

.A M - -

RELATED DOC6MENNUMBER.

REF: LI-110



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

56 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 Itssafety function or negatively Impact the ability of plant personnel to ensure the SSC Is capable ofperforming Its Intended safety function?




E YES D NO


DYES EONO


DYES ONOBasis:

I.

. I

If any of the above questions are answered Yes, STOP. Do not proceed with the revision, OR discusschange with NRC and obtain necessary approvals prior to implementation of the proposed change. If allthree questions are answered NO, go to Part III.(Attach additional sheets as necessary.)

REF: LMIo0

PART mI


Q YES

go"

Go to question 3.2.

Go to Part IV.


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

[l NO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relief.


made In response to an NRC Bulletin or Generic Letter, or (3) made in response to a request forinformation under 10 CFR 50.54(f) or 10 CFR 2.204?


Go to Part V.


Cl YES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment In summary report.

El NO Go to Question 5.1.





REF: U-110

5.2 Is the revised commitment necessary to minimize recurrence of the condition adverse to quality?I] YES Revise the commitment and notify NRC of revised commitment in next annualIRFO


ONO Revise commitment: no NRC notification Is required:

REFERENCESList documents (e.g., procedures, NRC submittals. etc.) affected by this change.

Doe. Number Description I

REF: LI.110

Commitment A-35207 Deletion Justification

At the time commitment A-35207 was made, GGNS plans included implementing anextended power up-rate, in addition to the App. K up-rate, during 2003. The originalintent was to revise the CHECWORKS model to account for both the extended and theApp. K up-rates in one model revision. However, the extended up-rate project waspostponed indefinitely in early 2003.

Per the commitment description, GGNS is to perform a parametric study to quantify theimpact of the TPO on GGNS wear rates and update the CHECWORKS model, ifnecessary. The parametric study will in fact include a revision of the CHECWORKSmodel to account for the increase in power level due to the App. K up-rate. As a result ofinitial investigation into the affect of the TPO on wear rates, the inspection scope of RF12was increased to include highly susceptible components and components with lowmargins that were due for inspection within one or two outages after RF12. Theinspection results noted no significant wear rate increase in any of the components whichcould have been attributed to the TPO. Additionally, more data on other highlysusceptible components will be collected during the upcoming refuel outage (RF13-2/22/04). The RF13 results are expected to show, similar to RF12 results, that nosignificant wear rate increases will be found which could be definitively attributed to theTPO. These components along with all the other modeled program components willcontinue to be inspected per the requirements of the FAC program. The quantification ofthe impact of the TPO on all modeled systems will be addressed as a result of normalFAC program maintenance and a commitment was notlis not required to track completionof this task. However, to track completion of this task and for planning purposes, WT2003-0520 has been assigned to EP&C to revise the CHECWORKS model to the up-rated conditions in order to quantify the impact on model wear rates.


CCE 2003-0011

COMMITMENT CHANGE EVALUATION FORMCommitment Number: 25105 | Plant Licensing Tracking Number: | C6 o 003-0011

Source Document: SIL 156

Commitment: Deletion? M Revision? ElHas the original commitment been implemented? I M YES I a NO, Notify Plant LicensingOriginal Commitment Description:Verify neutron monitoring instrument tubes arc properly seated after maintenance activities are performed in the vicinity ofthe tubes except for normal fuel movement.

Revised Commitment Description:No longer performing a visual check of instrument tube seating during refueling after maintenance activities.

Summary of Justification for Change or Deletion:The instrument bosses were chamfered after initial core loading years ago where GGNS actually experienced unseatinginstrument strings during fuel movements. This design change greatly reduced the likelihood of the instrument stringinadvertently being dislodged. GGNS OE since that time has been no further instances of this event through a dozenrefueling outages with fuel movement and other cell maintenance activities. Hence, it is prudent based on OE andmodifications to remove this vendor recommended practice.


Prepared By:7) \ A* /PrprdB:Paul AlI. Differe~nt D zet , \a. 124[18 103

Pnt Nane__gnature__/ DateManagement I

Approval: Ken L Walker A41 tJe /W b y J2- A/0

Print Name/Signature /atePlant Licensing

Management Charles A. Bottemiller -Concurrence:

Print Name/Slinature Date

I NON-A 4£Ct-pi

NUM6EA d FAGCIA~ -L Yd

RELATSb Ooczum&TINu =Ll °-

ef-. N-D RE

REF. U-IO




0 NO Go to Part H.PART 1I

2.1 Could the change negatively Impact the ability of a System, Structure, or Component (SSC) to perform Itssafety function or negatively Impact the ability of plant personnel to ensure the SSC Is capable ofperforming Its intended safety function?


C NO Continue with Part HI. Briefly describe rationale:Commercial item only. If a string were to be dislodged it would fall into open space where fuel bundle wasremoved. Subsequent attempts to load the intended fuel assembly would notice the dislodged string as the stringwould hinder or prevent the ability to load the fuel assembly. The string would then be either reseated orreplaced if damaged.



l YES E NO


E YESBasis:

0JNo


D YESBasis:

ONO

__IIf any of the above questions are answered Yes, STOP. Do not proceed with the revision, OR discusschange with NRC and obtain necessary approvals prior to Implementation of the proposed change. If allthree questions are answered NO, go to Part III.(Attach additional sheets as necessary.)

RE: LU-11

PART III

3.1 Was the original commitment (e.g, response to NOV, etc.) to restore an Obligation (I.e., rule, regulation,order or license condition)?

o YES Go to question 32.

0U NO Go to Part IV.



0 NO STOP. Do not proceed with the revision, OR apply for appropriate regulatory relief.


made In response to an NRC Bulletin or Generic Letter, or (3) made In response to a request forinformation under 10 CFR 50.54(f) or 10 CFR 2.204?

H0 YES Go to Question 4.2.

O NO Goto Part V.


El YES STOP, You have completed this evaluation. Revise the commitment and notify NRC ofrevised commitment In summary report.






REF: L-It O

N

5.2 Is the revised commitment necessary to minimize recurrence of the condition adverse to quality?a YES Revise the commitment and notify NRC of revised commitment In next annualIRFO


ONO Revise commitment: no NRC notification Is required:

REFERENCESI 1.t documenlto Ig nroceur.is WIRC subthmittalse, e.) affeeted hv this chanve

Doc. Number Description03-1-01-5 Delete step 2.14 from Refueling IOI

RER: U-110