ML061360473

Report of 10CFR50.59 Safety Evaluations and Commitment Changes - April 1, 2005 Through March 31, 2006

ML061360473
Person / Time
Site:	Grand Gulf
Issue date:	05/10/2006
From:	Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GNRO-2006/00025
Download: ML061360473 (167)

Text

son, MS 39150

-172800 GNRCK200E100025 May 1C" :2CKOY6 U.S. Nuclear Regulatory Commission Attention : Document Control Desk Washington, DC 20555 Subject : Report of 10CFR5a59 Safety Evaluations and Commitment Changes - April 1, 2005 through March 31, 2006 Grand Gulf Nuclear Station Docket No. 50-416 License No. NPF-29

Dear Sir or Madam :

Pursuant to 10CFR50 .59(d)(2) Entergy Operations, Inc. hereby submits a summary of 50.59 evaluations for the period of April 1, 2005 through March 31, 2006. Also attached is the summary of commitment changes for the same perio in accordance with NEI 95-07 Guidelines.

If you have any questions or require additional information, please contact Dennis Coulter at 601-437-6595 .

This letter does not contain any commitments .

Yours Truly, CAB/DMC/dmc Attachments : 1 . Table of Contents

2. 10CFR50 .59 Evaluations and Commi nt Change Evalua cc: (See Next Page)

GNRO-2006/00025 Page 2 cc NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150 U .S. Nuclear Regulatory Commission ATTN : Dr. Bruce S. Malleft (w/a}

Regional Administrator, Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 U.S . Nuclear Regulatory Commission ATTN : Mr. Bhalchandra Vaidya, NRR/DORL {w/a)

ATTN : ADDRESSEE ONLY ATTN : U.S. Postal Delivery Address only Mail Stop OWFN/0-7DIA Washington, D.C. 20555-0001 Mr. D. E. Levanway {Wise Carter) (w/a}

Mr. L. J . Smith use Carter) (w/a}

Mr. N. S. Reynolds (w/a}

Mr. J . N. Compton

Attachment I Table of Contents Grand Gulf Nuclear Station I OCFR5 0.5 9 Evaluation and Commitment Change Evaluation Report for the Period April 1, 2005 through March 31, 2006 Acronyms ARI Alarm Response Instruction LOP Loss of Power ASTM American Society for Testing and Materials MAPLHGR Maximum Average Planar Linear Heat Generation Rate CCE Commitment Change Evaluation MCPR Minimum Critical Power Ratio CMWT Core Megawatts Thermal MNCR Material Non-Conformance Report CR Condition Report --MOV Motor Operated Valve DCP Design Change Package Mechanical Standard EP Emergency Procedure OLALHWFA~' Main Steam Isolation Valve Leakage Control System EPI Equipment Performance Instruction NPE Nuclear Plant Engineering EPRI Electric Power Research Institute Nuclear Steam Supply System ER Engineering Request PIDMS Plant Data Management System ES Electrical Standard PPM Parts per Million ESF Engineered Safety Feature PRA Probabilistic Risk Assessment GE General Electric PSW Plant Service Water GG Grand Gulf RCIC Reactor Core Isolation Cooling GGN Grand Gulf Nuclear RFO Refueling Outage GPM Gallons per Minute RHR Residual Heat Removal 101 Integrated Operating Instruction RPV Reactor Pressure Vessel IS1 In Service Inspection Standard Change Notice IST In Service Testing System Energy Resources, Inc.

LBDC License Basis Document Change Standby Gas Treatment System LD(C License Document Change Significant Operating Experience Report LHGR Linear Heat Generation Rate Standby Service Water LLRT Local Leak Rate Test Technical Requirements Manual / Technical Specifications I LOCA Loss of Coolant Accident Ultimate Heat Sink Page 1 of 3

Attachment 1 Table of Contents Grand Gulf Nuclear Station I OCFR50.59 Evaluation and Commitment Change Evaluation Report for the Period April 1, 2005 through March 31, 2006 Safety Evaluations Evaluation Number Initiating Summary Document SE 2005-0002-ROO Calculation Revised the LOCA dose analysis 1) to apply the new RAPTOR dose methodology, 2) to XC-Q I I 11 -98017, modify the control room model by deleting the need for control room fresh air after 3 days of REV. 2 isolated operation, 3) to consider stable isotopes so that HEPA loadings can be generated, and 4) to reformat the calculation to meet procedure ENS-DC- 126 format.

SE 2005-0003-ROO LDC 2005-028, Revised FSAR SECTION 15.4.1 .1 .3 to remove reference to single rod out shutdown margin REV. 0 check following refueling . This check has is a hold over from startup testing .

SE 2005-0004-ROO ER-GG-2003-0359- Evaluation of acid flushing the tube side of an RHR heat exchanger. The acid flush is 000 facilitate the Eddy Current testing of the heat exchanger..

SE 2005-0004-RO I ER-GG-2003-0359- This first revision of SE 2005-0004 reassesses the operability of the RHR and SSW systems 000 when the inboard heat exchanger drain valves are closed.

SE 200&000&RIO LDC-2005-060 and Cycle 15 reload changes and operation of the cycle 15 cue as given in the Core Operating COLLR Limits Report (COLR).

SE 200500001110 Calculations XC- Calculations associated with offsite and control room doses associated with secondary QIP53-05011 and containment bypass leakage through the instrument air and service air piping.

XC-Q 1 M46-04004 SE 2005-0007-ROO LBDC 2004-0095 Modification of the ODCM/TRM 6 .3 .9 required actions and operability requirements aNsdicable to the discharge canal flow monitoring instrumentation .

SE 2005-0008-ROO ER 2004-0234001 Extend the DIV 11 Diesel Generator fuel oil storage tank inspection by three months.

SE 20000001-1010 ER-2005-0197-000 Change the fuel pool decay heat analytical method from the Branch Technical Position AS9--

9-2 to the OA Ridge Isotope Generation and Depletion code - ORIGEN V2 .1 .

SE 2006-0002-ROO LDC-2006-002 TRM 6 .3.8 Relaxation of turbine overspeed trilt) ATT testing LCO actions .

SE 200600ON1110 ER-GGN-2005- Removed logic for diesel generator low control air pressure trip during a LOCA 0110-00-00 Page 2 of 3

Attachment I Table of Contents Grand Gulf Nuclear Station I OCFR5 0 .5 9 Evaluation and Commitment Change Evaluation Report for the Period April 1, 2005 through March 31, 2006 Commitment Change Evaluations Commitment Source Document Summary Number CCE 2005-0002 AECM 860395 Deleted 1) Independent verification of amendments implementation checklist developed for each TS amendment . 2) Hold points and final verification will be established on the checklist prior to declaring the system operable .

CCE 2005-0003 AECM 86/0077 Revised dose related restricted locations in the spent fuel pool per analysis documented in the source document.

CCE 2005-0004 AEC14 860077 Revised the requirement that dose related restricted locations had to be filled with fuel bundles with one year of decay for cycle I discharged fuel.

CCE 2005-0005 AECM 860089 Deleted pre-NEI 99-04 guidance for justification of UFSAR and commitment change or deletion.

Page 3of3

Attachment 2 1 OCFRS 0.5 9 Evaluations and Commitment Change Evaluations

GGNS 50.59 Safety Evaluation Number SE 2005-0002-R00

60.59 REVIEW FORM Page I of 9

1. OVERVIEW I SIGNATURES Facility:

Document Reviewed : - Calculation XC-Q1 111 -98017 Change/Rev .:

System Des tor(s)/Desc  : Various Description of Proposed Activity:

This calculation revises the GGNS LOCA dose analysis to 0) apply the new RAPTOR dose methodology, (ii) revise the control room model to delete the need for control room fresh air after 3 days of isolated operation (due to the large assumed inleakage rate of 2000 cfm), (iii} consider stable isotopes so that HEPA loadings can be generated, and (iv) re-format the calculation to the ENS-DC-126 format.

Check the applicable review(s) : (Only the sections indicated must be included in the Review.)

EDITORIAL CHANGE of a Licensing Basis Document Section I SCREENING Sections I and 11 required 60.59 EVALUATION EXEMPTION Sections 1, 11, and ill required 50.59 EVALUATION (#: a-605-(') D-) I Sections 1, 11, and IV required Prepare Reviewer:

Name (print) / Signa OSRC: V.A S Chairman's Name (print) / Signature / Date (Required only for Programmatic Exclusion Screenings and 50.59 Evaluations.)

LI-101-01, Rev. 8; Effective Date: 6/23/05

11. SCREENINGS A. Licensing Basis Document Review

1. Does the proposed actin facility or a procedure as described in any of the following Licensing Basis Documents:

Operating License YES NO CHANGE # and/or SECTIONS IMPACTED Operating License 1:1 1 104 TS E3 EEO NRC Orders 1 1:1 1 N I If "YES," obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMM LI-113 . (See LI-101 for exceptions.)

LBDs controlled under 50 .59 YES NO CHANGE # (if applicable) and/or SECTIONS IMPACTED FSAR 04 Tables 15.6-9,15.6-13,15.6-14, Figure 15.6-3 TS Bases E3 QE Technical Requirements Manual 10 0 Core Operating Limits Report NRC Safety Evaluation Report and supplements for the initial FSAR 1 NRC Safety Evaluations for [E] 041 amendments to the Operating License' If "YES," perform an Exemption Review per Section III OR perform a 60 .59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMM LI-113. If obtaining NRC approval, document the LBD change in Section II .A.S . However, the change cannot be implemented until approved by the NRC. Complete Section 11.

LBDs controlled under other YES NO CHANGE # (if applicable) and/or regulations SECTIONS IMPACTED Quality Assurance Program Manual2 Emergency Plan2, 3 El 1041 1 3' 4 Fire Protection PrograM En ED (includes the Fire Hazards Analysis)

Offsite Dose Calculations Manua,3,4 D If "YES," evaluate any changes in accordance with the appropriate regulation AND initiate an LBD change in accordance with NMM LIA 13 .

If YES," see U-1 01 . No LBD change is required .

If YES," notify the responsible department and ensure a 50 .54 evaluation is performed . Attach the 50 .54 evaluation .

3 Changes to the Emergency Plan, Fire Protection Program, and Oftite Dose Calculation Manual must be approved by the OSRC in accordance with NMM OM-1 19, 4 0 YES : evaluate the change i accordance he requirements of the tacility ting License Condition or under 50.69, as appropriate .

LI-I 01 -01, Rev. 8; Effective Date: 6123105

2. Does the proposed activity involve a test or experiment not described in the FSAR? El Yes 0 No If "YES," perform a 50 .59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change AND initiate an LBD change in accordance with NMM LIA 13, if applicable. If obtaining NRC approval, document the change In Section II.A.S. However, the change cannot be Implemented until approved by the NRC . Complete Section It.

3. Basis Explain why the proposed activity does or does not impact the Operating License/ Technical Specifications and/or the FSAR . If the proposed activity involves a potential test or experiment not previously described in the FSAR also include an explanation. Discuss other LBDs if impacted. Adequate basis must be provided within the Screening such that a third-party reviewer can reach the same conclusions. Simply stating that the change does not affect TS or the FSAR is not an acceptable basis.

Tech Specs/Operating License The current GGNS Tech Sped and Operating License are inputs into the LOCA dose analy identified or proposed by this analysis.

FSAR The LOCA dose analysis is reported in SAR 15.6.5. Several changes to this section have been identified as noted in Section 11 of this 50.59 review. LDC 2005-037 makes the applicable changes .

Test or ExUriment not Described in the SAR This calculation revision only updates the methodology applied in the LOCA dose analysis . This calculation does not call for any action in the plant or changes to plant procedures.

4. References Discuss the methodology for performing LBD searches . State the location of relevant licensing document 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 . NOTE: Ensure that manual searches are performed using controlled copies of the documents. If you have any questions, contact your site Licensing department.

Electronic search method used: Keywords :

Tech Specs, Operating License, FSAR, "RAPTOR"; "LOCA Dose". "LOCA Radiological" COLR, ODCM, Emergency Plan, SER LBDs reviewed manually:

SAR 15.6.5

5. Is the validity of this Review dependent on any other change? 0 Yes 1f' No

!f "YES," list the required changeslsubmittals. The changes covered by this 50.59 Review cannot be implemented without approval of the other identified changes (e.g ., license amendment request). Establish an appropriate notification mechanism to ensure this action is completed .

LIA01-01, Rev . 8; Effective Date : 6123105

0.59 REVIEW Page 4of9 B. ENVIRONMENTAL SCREENING If any of the following questions is answered "yes," an Environmental Review must be performed in accordance with NMM Procedure EV-115 and attached to this 50.59 Review. Consider both routine and non-routine (emergency) discharges when answering these questions.

Will the proposed activity being evaluated :

Involve a land disturbance equal to or in excess of one acre (i.e., grading activities, construction of buildings, excavations, reforestation, creation or removal of ponds)?

2. E] Involve any land disturbance of undisturbed land areas (i.e., grad activities, construction, excavations, reforestation, creating, or removing ponds)?

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

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

Increase the concentration or quantity of chemicals being discharged to the river, take, or air?

Discharge any new or different chemicals that are currently not authorized for use by the state regulatory agency?

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

8. Modify the design or operation of the cooling tower that will change wafer or air flow characteristics?

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

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

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

12. El Involve the installation or use of equipment that will result in a new or additional air emission discharge?

Involve the installation or modification of a stationary or mobile tank?'

Involve the use or storage of oils or chemicals that could be directly released into the environment?

15. Q 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 for guidance in answering this quest LI-101-01, Rev. 8; Effective Date: 6/23"05

50.59 REVIEW FORM Page 5 of 9 C. SECURITY PLAN SCREENING If any of the following questions is answered "yes," a Security Plan Review must be performed by the Security Department to determine actual impact to the Plan and the need for a change to the Plan.

Could the proposed activity being evaluated :

NO

1. E Add, delete, modify, or otherwise affect Security department responsibilities (e.g.,

including fire brigade, fire watch, and confined space rescue operations)?

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

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

4. © Affect (block, move, or alter) security lighting by adding or deleting lights, structures, buildings, or temporary facilities?

a Modify or otherwise affect the intrusion detection systems (e.g., E-fields, microwave, fiber optics)?

6. ;11 Modify or otherwise affect the operation or field of view of the security cameras?

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

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

9. 1

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

lo. Modify or othe t the facility's telephone or security radio systems?

The Security Department answers the following question if one of questions C.1 through C.10 above was answered "yes."

Is a change to the Security Plan required? El Yes No Attach to this 50.59 Review or reference below documentation for accepting a "yes" answer for any of Questions C.1 through C.10, above.

Name of Security Plan reviewer (print 1 Signature / Data LI-101-01, Rev. 8; Effec e Date: 6123/05

50 .59 REVIEW FOR Page 6 of 9 D. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) SCREENING

( NOTE : This section is not applicable to Grand Gulf or Waterford 3 and may be removed from 50 .59 Rev performed for Waterford 3 proposed activities.)

If any of the following questions is answered "YES," a 72.48 Review must be performed in accordance with NMM Procedure LI-112 and attached to this 50.59 Review .

Will the proposed activity being evaluated :

1. F Any activity that directly impacts spent fuel cask storage or loading operations?

2. F1 Involve the ISFSI including the concrete pad, security fence, and lighting?

3. Involve a change to the on-site transport equipment or path from the Fuel Building to the ISFSI?

4. M Involve a change to the design or operation of the Fuel Building fuel bridge including setpoints and limit switches?

5. a Involve a change to the Fuel Building or Control Room(s) radiation monitoring?

6. 0 Involve a change to the Fuel Building pools including pool levels, cask pool gates, cooling water sources, and water chemistry?

7. El Involve a change to the Fuel Building handling equipment (e.g., bridges and cask cranes, structures, load paths, lighting, auxiliary services, etc)?

s. a Involve a change to the Fuel Building electrical power that could potentially impact cask loading or storage activities?

9. a" Involve a change to the Fuel Building ventilation that could potentially impact cask loading or storage activities?

10. r-1 Involve a change to the ISFSI security?

11 . n Involve a change to off-site radiological release project from non-ISFSI sources?

12. n Involve a change to spent fuel characteristics"?

13. Redefine/change heavy load pathways?

14. Involve fire and explosion protection near or in the on-site transport paths or near the ISFSI?

15. El ED Involve a change to the loading bay or supporting components power that could potentially impact cask loading or storage activities?

16. El New structures near the ISFSI?

17. Q Modifications to any plant systems that support dry fuel storage activities?

18. Q Involve a change to the nitrogen supply, service air, demineralized water or borated water system in the Fuel Building?

LI-101-01, Rev . 8; Effective Date: 6/23105

Ill . 50.59 EVALUATION EXEMPTION A. Check the applicable box below. If a box is checked, clearly document the basis in Section III .B, below. If none of the boxes are appropriate, perform a 60.59 Evaluation in accordance with Section IV. Provide supporting documentation or references as appropriate .

0 The proposed activity meets all of the following criteria regarding design function :

The proposed activity does not adversely affect the design function of an SSC as described in the FSAR; AND The proposed activity does not adversely affect a method of performing or controlling a design function of an SSC as described in the FSAR; AND The proposed activity does not adversely affect a method of evaluation that demonstrates intended design function(s) of an SSC described in the FSAR will be accomplished.

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

The NRC has approved the proposed activity or portions thereof.

Reference:

B. Basis Provide a clear, concise basis for determining the proposed activity may be exempted such that a third-party reviewer can reach the same conclusions, LI-101-01, Rev. 8; Effective Date: 6/23/05

Page 8 of 9 IV. 50.59 EVALUATION License Amendment Determination Does the proposed Change being evaluated represent a change to a method of evaluation 0 Yes ONLY? If "Yes," Questions I - 7 are not applicable; answer only Question 8. lf"No,"answer El No all questions below .

Does the proposed Change :

1, Result in more than a minimal increase in the frequency of occurrence of an accident E3 Yes previously evaluated in the FSAR? [1 No BASIS:

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a El Yes structure, system, or component important to safety previously evaluated in the FSAR? Ej No BASIS:

3. Result in more than a minimal se in the consequences of an accident previously El Yes evaluated in the FSAR? E] No BASIS:

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, F-1 Yes system, or component important to safety previously evaluated in the FSAR? F1 No BASIS:

5. Create a possibility for an accident of a different type than any previously evaluated in the E] Yes FSAR? El No BASIS:

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

7. Result in a design basis limit for a fission product barrier as described in the FSAR being El Yes exceeded or altered? 0 No BASIS:

8. Result in a departure from a method of evaluation described in the FSAR used in establishing Yes the design bases or in the safety analyses? No LI-1 01 -01, Rev . 8; Effective Date: 6/23/06

50.59 REVIEW FORM Page 9 of 9 BATS:

The document under review is Revision 2 to Calculation XC-Q1 111-98017, which converts the LOCA dose analysis described in SAIR 15,6.5 from the TRANSACT computer code to the newer RAPTOR computer code (SCR-2004-0735) . Also, a small change to the model was also made in the control room model. The other changes including the addition of stable isotopes and re-formatting, do not affect the results . Thus, there were two changes to the elements of the methodology: (1) a change in the model for the control room, and (ii) a change in the computer code-Control Room Model Change:

The change in the model to the control room deleted the modeling assumption that fresh air is introduced into the control room after 3 days and the control room is assumed to be in the recirc mode for the duration of the LOCA analysis. Based on a carbon dioxide buildup and oxygen depletion analysis, it was found that fresh air is no longer needed since the very large assumed inleakage rate of 2010 cfm would provide sufficient fresh air for the control room and TSC personnel . Thus, to more accurately reflect the expected plant response considering the assumed elevated inleakage, fresh air intake was not modeled .

The radiological impact of this change on the control room doses is insignificant since the source term release is very small after 3 days. Thus, since this change to the elements of analysis methods yield results that are essentially the same, it is not considered to be a departure from approved methods .

Computer Code Change:

As described in Engineering Report G-SA-2003-001, Rev . 2, the RAPTOR methodology has more capabilities than the older TRANSACT methodology including (i) tracking daughter products and stable isotopes, (ii) more isotopes, (Q more volumes and flows, and (iv) better numerical stability and has been successfully benchmarked to TRANSACT. These additional capabilities are the reason for the transition to newer methodologies .

In addition, the RAPTOR code has been rigorously benchmarked to the NRC's RADTRAD code and was found to generate results that are essentially the same as RADTRAD . The code benchmarks are documented in Engineering Reports G-SA-2005-001, -002, -003, -004, -005, -006, and -007. The new calculation therefore applies a methodology that is essentially the same as the method applied by the NRC for this application .

The GGNS LOCA dose analysis is reported in FSAR Section 15.6.5; however, the method of evaluation is not explicitly described . SAR la&5 only states "[t1he methods, assumptions, and conditions used to evaluate this accident are in accordance with those guidelines set forth in Regulatory Guide 1 .183.'

Section 121 of RG 1 .183 specifically endorses the RADTRAD code as a suitable methodology for evaluating control room doses. In addition, Appendix A to RG 1 .183 also mentions the RADTRAD methodology as acceptable for evaluating spray and aerosol removal factors. Thus, RADTRAD is deemed to be a methodology approved by the NRC for this application . In fact, since RADTRAD is a standard industry code, many utilities have prepared the AST submittals with RADTRAD and have received SERB on their proposed changes. the Thus, in using RAPTOR methodology, GGNS is applying a method that is essentially the same as NRC methodology that has been explicitly endorsed for this application_

The results of this new revision are compared to the current SAR results below. The doses at all lava have decreased slightly due to the application of the new RAPTOR methodology- These decreased doses are due to TRANSACT's very conservative core release model, which is more realistic in RADTRAD and RAPTOR. These results could be classified as "non-conservative" (as described in the 50.59 guidelines in ENS-Lt-101) in that they are lower than the previous values and yield more margin to the applicable acceptance criteria. However, since they were developed with a methodology that has been shown to be essentially the same as an NRC-approved method, these results are considered to be acceptable.

Dose Results (Rem TEDE)

Location SAR Table New Results in Calculation 00-14 XQ (911111 -48 0 17 , Rev,.2 Exclusion Area Boundary 8.78 8.41 Low Population Zone 4.60 4.46 Control Room 1 3.65 3.64 If any of the above questions is checked "YES," obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure LI-113.

LI-101-01, Rev . 8; Effective Date: 6123/06

GGNS 50.59 Safety Evaluation Number SE 2005-0003-R00

50.59 REVIEW FORM page I of 10

1. OVERVIEW/ SIGNATURES Facility: Grand Gulf Nuclear Station Document Reviewed : LDC 2005-028 Chanae/Rev. :

System Designator(s)/Description :

NIA Description of Proposed Activity :

Change FSAR Section 15.4.1 .1 .3 to remove reference to single rod out shutdown margin check following refueling. The single rod out shutdown margin check following refueling is not required per any safety analysis or ing bast and has been done historically only as a hold over from early startup testing for added conservatism . This change does not involve an unreviewed safety question .

Check the applicable review(s): (Only the sections indicated must be included In the Review.)

EDITORIAL CHANGE of a Licensing Basis Document Section I SCREENING Sections I and 11 required 50.59 EVALUATION EXEMPTION Sections 1, 11, and III required 50.59 EVALUATION {#: A Sections 4 14 and IV required Preparer: Ken L. Walker Name (print) Signature Reviewer-. P.M. Different flame (print) I Signatur OSRC: v. F. V-1 Chairman's flame (print) / Signature (Required only for Programmatic Exclusion Screenings and 50.59 Evaluations.)

LI-101-01, Rev . 8; Effective Date: 6123/05

50 .59 REVIEW FORM Page 2 of 10 If. SCREENINGS A. Licensinci Basis Document Review

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

Operating License YES -T CHANGE # and/or SECTIONS IMPACTED NO Operating License TS 13 El

-OR an M

NRC Orders 0 E If "YES," obtain NRC approval prior to implementing the change by initiating an LSD change in accordance with NMM 1-1413 . (See Lt-101 for exceptions.)

LBOs controlled under 50 .59 YES NO CHANGE # (if applicable) and/or SECTIONS IMPACTED FSAR 15.4.1 .1 .3 LDC 2005028 TS Bans El EN Technical Requirements Manual [] IZ Core Operating Limits Report NRC Safety Evaluation Report and supplements for the initial FSAR' NRC Safety Evaluations for amendments to the Operating License' If "YES," perform an Exemption Review per Section ill QR perform a 50 .59 Evaluation per Section IV QR obtain NRC approval prior to implementing the change by initiating an LBO change in accordance with NMM Lt-113 . If obtaining NRC approval, document the LSD change in Section II.A.5, However, the change cannot be implemented until approved by the NRC. Complete Section If.

LBDs controlled under other YES NO CHANGE # (if applicable) andlor regulations SECTIONS IMPACTED Quality Assurance Program ManUa,2 13 0 Emergency Plan'"

Fire Protection PrograM 3,4 E3 it (includes the Fire Hazards Analysis)

Offsite Dose Calculations ManU313,4 1 13 Vol If "YES," evaluate any changes in accordance with the a Ppro priate re g elation AND initiate an LSD change in accordance with NMM LI-113 .

'VYES,"seet-1-101 . No LBD change is required .

2 If "YES," notify the responsible department and ensure a 50.54 evaluation is performed. Attach the 50,54 evaluation .

3 Changes to the Emergency Plan, Fire Protection Program, and Oftsite Dose Calculation Manual must be approved by the OSRC in accordance with NMM OM-119.

4 U "YES," evaluate the change in accordance with the requirements of the facility's. Operating License Condition or under 50 .59, as appropriate .

LI-1 01 -01, Rev. 8; Effective Date : 6/23/05

50.59 REVIEW FORM Page 3 of 10

2. Does the proposed activity involve a test or experiment not described in the FSAR? El Yes 0 No If "YES," perform a 50.59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change AND Initiate an LBD change in accordance with NMM L1-113, if applicable. If obtaining NRC approval, document the change in Section II.A.S. However, the change cannot be implemented until approved by the NRC . Complete Section 11.

3. Basis Explain why the proposed activity does or does not impact the Operating License/Technical Specifications and/or the FSAR. If the proposed activity involves a potential test or experiment not previously described in the FSAR also include an explanation . Discuss other LBDs if impacted . Adequate basis must be provided within the Screening such that a third-party reviewer can reach the same conclusions . Simply stating that the change does not affect TS or the FSAR is not an acceptable basis.

The change does impact the FSAR as described in Section IV.

4. References Discuss the methodology for performing LBD searches . State the location of relevant licensing document 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. NOTE : Ensure that manual searches are performed using controlled copies of the documents. If you have any questions, contact your site Licensing department.

Electronic search method used: Keywords :

Autonomy (All LSD} "Shutdown margin", "SM", "refueling" within 10 words of "margin" "Single rod", "one rod" LBDs reviewed manually: None

5. Is the validity of this Review dependent on any other change? [] Yes No If "YES," list the required changes/submittals . The changes covered by this 50.59 Review cannot be implemented without approval of the other identified changes (e.g., license amendment request) . Establish an appropriate notification mechanism to ensure this action is completed .

LI-1 01 -01, Rev . 8; Effective Date: 6/23105

50.59 REVIEW F Page 4of10 B. ENVIRONMENTAL SCREENING If any of the following questions is answered "yes," an Environmental Review must be performed in accordance with NMM Procedure EV-115 and attached to this 50.59 Review. Consider both routine and non-routine (emergency) discharges when answering these questions .

Will the proposed activity being evaluated :

NO Involve a land disturbance equal to or in excess of one acre (i.e., grading activities, construction of buildings, excavations, reforestation, creation or removal of ponds)?

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

3. [] Involve dredging activities in a lake, river, pond, ditch, or stream?

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

5. [] Increase the concentration or quantity of chemicals being discharged to the river, lake, or air?

6. [~ Discharge any new or different chemicals that are currently not authorized for use by the state regulatory agency?

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

Modify the design or operation of the cooling tower that will change water or ow characteristics?

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

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

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

Involve the installation or use of equ will result in a new or additional air emission discharge?

Involve the installation or modification of a stationary or mobile tank?'

14. [] Involve the use or storage of oils or chemicals that could be directly released into the environment?

15. 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 for guidance in answering this question .

LIA01-01, Rev. 8; Effective Date: 6123105

50.59 REVIEW FORM Page 5 of 10 C. SECURITY PLAN SCREENING If any of the following questions is answered "yes," a Security Plan Review must be performed by the Security Department to determine actual impact to the Plan and the need for a change to the Plan.

Could the proposed activi being evaluated :

1. [~ Z Add, delete, modify, or otherwise affect Security department responsibilities (e.g.,

including fire brigade, fire watch, and confined space rescue operations)?

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

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

4. [] 10 Affect (block, move, or alter) security lighting by adding or deleting lights, structures, buildings, or temporary facilities?

Modify or otherwise affect the intrusion detection systems (e .g., E-fields, microwave, fiber optics)?

6. 041 Modify or otherwise affect the operation or field of view of the security cameras?

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

8. C,) Modify or otherwise affect primary or secondary power supplies to access control equipment, intrusion detection equipment, other security equipment, or to the Central Alarm Station or the Secondary Alarm Station?

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

1t} . ~] [D Modify or otherwise affect the facility's telephone or security radio systems?

The Security Department answers the following question if one of ques C.1 through C.10 above was answered "yes."

Is a change to the Security Plan required? C] Yes 0 No Attach to this 50.59 Review or reference below documentation for accepting a "yes" answer for any of Questions C.1 through C .10, above.

Name of Security Plan reviewer (print 1 Signature I Data LI-101-01, Rev . 8 ; Effective Date: 6/23105

5+0 .59 REVIEW FORM Page 6 of 10 D. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) SCREENING NOTE: This section is not applicable to Grand Gulf or Waterford 3 and may be removed from 50.59 Reviews performed for Waterford 3 proposed activities.)

If any of the following questions is answered "YES," a 72.48 Review must be performed in accordance with N MM Procedure LI-112 and attached to this 50.59 Review.

Will the proposed activity being evaluated :

NO

1. D OR activity that directly impacts spent fuel cask storage or loading operations?

2. [1 Z Involve the ISFSI including the concrete pad, security fence, and lighting?

3. El 0 involve a change to the on-site transport equipment or path from the Fuel Building to the ISFSI?

4. 0 1Ar involve a change to the design or operation of the Fuel Building fuel bridge including setpoints and limit switches?

5. 1 Involve a change to the Fuel Building or Control Room(s) radiation monitoring?

6. r Involve a change to the Fuel Building pools including pool levels, cask pool gates, cooling water sources, and water chemistry?

7. CJ Involve a change to the Fuel Building handling equipment (e.g ., bridges and cask cranes, structures, load paths, lighting, auxiliary services, etc)?

8. El 0 involve a change to the Fuel Building electrical power that could potentially impact cask loading or storage activities?

9. "

1 Involve a change to the Fuel Building ventilation that could potentially impact cask loading or storage activities?

1Q. n Involve a change to the ISFSI security?

11 . Involve a change to off-site radiological release projections from non-ISFSI sources?

12. [1 Involve a change to spent fuel characteristics?

13. Redefinelchange heavy load pathway

14. 0 Involve fire and explosion protection near or in the on-site transport paths or near the ISFSI?

15. 11 involve a change to the loading bay or supporting components power that could potentially impact cask loading or storage activities?

16. C} New structures near the ISFSI?

17. 0 Modifications to any plant systems that support dry fuel storage activities?

18. [] involve a change to the nitrogen supply, service air, demineralized water or borated water system in the Fuel Building?

1..1-101-01, Rev . 8; Effective Date: 6123105

60.59 REVIEW FORM Page 7 of 10 111 . 50.59 EVALUATION EXEMPTION A. Check the applicable box below . If a box is checked, clearly document the basis in Section III .B, below. If none of the boxes are appropriate, perform a 50.59 Evaluation in accordance with Section IV. Provide supporting documentation or references as appropriate.

Ej The proposed activity meets all of the following criteria regarding design function:

The proposed AND activity does not adversely affect the design function of an SSC as described in the FSAR; The proposed activity does not adversely affect a method of performing or controlling a design function of an SSC as described in the FSAR; AND The proposed activity does not adversely affect a method of evaluation that demonstrates intended design function(s) of an SSC described in the FSAR will be accomplished .

El An approved, valid 50.59 Review(s) cove associated aspects of the proposed activity already exists. Reference 50.59 Evaluation # - (if applicable) or attach documentation . Verify the previous 50.59 Review remains valid.

El The NRC has approved the proposed activity or portions thereof.

Reference :

B. Basis Provide a clear, concise basis for determining the proposed activity may be exempted such that a third-party reviewer can reach the same conclusions .

LI-101-01, Rev . 8; Effective Date: 6123/05

50.59 REVIEW FORM Page 8 of 10 IV. 50.59 EVALUATION License Amendment Determination Does the proposed Change being evaluated represent a change to a method of evaluation [l Yes ONLY? If "Yes," Questions 1 - 7 are not applicable; answer only Question 8. If "No," answer Z No all questions below .

Does the proposed Change :

1. Result in more than a minimal in the frequency of occurrence of an accident C( Yes previously evaluated in the FSAR? No BASIS:

Inadvertent withdrawal of a control rod during refueling resulting in criticality is considered an infrequent event in the FSAR. There is no postulated set of circumstances which results in a rod withdrawal error during REFUEL Mode. With the mode switch in the SHUTDOWN position, a control rod block also prevents the withdrawal of a control rod . The proposed change makes no physical modifications to any plant systems, interlocks, or components. It makes no change to any process used in control blade replacement activities . There is no change to refueling, fuel movement, or core loading verification processes . The SER for TS Amendment 120 addressed this issue directly, stating: -Although the shutdown margin may not have been demonstrated in Mode 5 , shutdown margin calculations would have been performed and, along with procedural compliance for any Core Alterations, would provide assurance that adequate shutdown margin is available."

Thus, there is no increase in the frequency of occurrence of an accident previously evaluated in the FSAR by removing the requirement to perform a single rod out shutdown margin (SDM) check following completion of refueling.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a [] Yes structure, system, or component mportant to safety previously evaluated in the FSAR? No BASIS :

The proposed change to remove the single rod out SDM check does not physically modify any structure, system, or component (SSC). This check is not relied upon by any analysis nor is it needed to prevent an inadvertent criticality from occurring . The check was simply considered an industry good practice at one time and was never required to prevent occurrence of any analyzed event. The FSAR describes this event as "precluded", and this check as only an "experimental" verification . No other events such Drop Accident, Mislocated Fuel Assembly, or Rod Withdrawal Error during operation are impacted.

Removing this verification does not increase the likelihood of malfunction on any SSC or the likelihood of the event itself.

3. Result in more than a minimal increase in the consequences of an accident previously F1 Yes evaluated in the FSAR? No BASIS :

The single rod out check was never intended to mitigate consequences of an inadvertent criticality during refueling . It was meant only as a loose verification that the reactor would indeed not go critical with strongest rod out once reloaded . In effect, this check depended on an analytical determination of the strongest worth control rod, so the check was no more reliable than the analysis it was attempting to check. No changes to any processes, systems, interlocks, or release barrier used to prevent or mitigate the consequences of an accident are being made by this revision.

LI-101-01, Rev . 8; Effective Date: 6123105

54.59 REVIEW FORM Page 9 of 10

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, [] Yes system, or component important to safety previously evaluated in the FSAR? No BASIS :

There are no modifications to any SSC as a result of the proposed change . The change does not make the consequences of an inadvertent criticality (in the unlikely event one were to occur) more severe . It does not impact any fuel movement procedures . No reliance has been assumed on this check in order to prevent or mitigate the consequences of malfunction of a SSC . Control Rod Drive system, fuel movement equipment, containment systems, and safety interlocks are unaffected . Thus there is no increase in the consequences of an SSC malfunction .

5. Create a possibility for an accident of a different type than any previously evaluated in the 0 Yes FSAR? No BASIS:

Inadvertent criticality during refueling has already been considered in the FSAR, and determined to be precluded by plant design . This change makes no physical changes to the plant. Other types of possible events such as multiple rod withdrawal during refueling or unrecognized multiple fuel movement errors are not created by this change . Not performing the single rod out check does not create the possibility of a new operating event . SDM is confirmed for each Core Alteration that loads a fuel bundle to core (unless doing a spiral reload) and additionally SDM must be confirmed at initial criticality per Technical Specification requirements . Thus, no new type of event or accident is created by this change .

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

C]04 Yes No BASIS:

No SSC are being modified by the proposed change . No operating procedures (other than the requirement to perform this check) are being revised . Inadvertent criticality, which this check never was intended to prevent or mitigate, has already been evaluated. Processes used to ensure proper core loading remain unchanged . There is no possible new type of SSC malfunction created by this change.

Result in a design basis limit for a fission product barrier as described in the FSAR being [] Yes exceeded or altered? No BASIS:

The proposed change does not alter any barrier . No physical changes at all are being proposed. There is no impact on fuel, vessel or containment design . No process, procedure, or analysis changes impacting barriers are being made as a result of removing the single rod out SDM check. Thus, no barrier limit is being exceeded or altered.

t_I-101-01, Rev . 8; Effective Date: &123145

0.59 REVIEW FORM P 10 of

8. Result in a departure from a method of evaluation described in the FSAR used in establishing Yes the design bases or in the safety analyses? No BASIS :

No changes to methods or analytical bases are being made. The methods used to calculate SDM are unchanged- This check has not been used to benchmark or verify any analytical methods . It only served as a rough verification of subcriticality to back up the analytical determinations made in the design and licensing of the new reload core. There are no impacts on the uncertainties used to establish the SDM limits in the Technical Specifications . Thus, there is no departure from established methods .

If any of the above questions is checked "YES," obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure Lt-113.

L1-401-01, Rev . 8; Effective Date: 6/23/05

GGNS 50.59 Safety Evaluation Number SE 2005-0004-R00

50 .59 REVIEW FORM Page I of 13 SIGNATURES Facility : Grand Gulf Nuclear Station Document Reviewed . ER-GG-2003-0359-000 Change/Rev .: Q System Designator(s)IDescription : 1E12 1 Acid Flush of Residual Heat Removal Heat Exchan Description of Proposed Change:

The ER reviewed approves an acid-flush solution and develops a process to acid-flush the tube side of an RHR heat exchanger unit (1131213001A/2A or 113121300IB/2131) . It also provides a guideline to the number of acid flushes allowed.

The purpose of the acid flush is to facilitate eddy current testing of the heat exchanger tubes. It could also improve the thermal performance of the flushed heat exchanger unit.

Check the applicable review(s): (Only the sections indicated must be included in the Review .)

EDITORIAL CHANGE of a Licensing Basis Document Section I SCREENING Sections I and 11 required 60.69 EVALUATION EXEMPTION Sections 1, it, and tit required

- ~ 50.59 EVALUATION (#: ~=I"~ij~"~,'~~I ~ } ~ Sections t, ll, and IV required Preparer : Sinyy4ong D. Lin /EB/EP&C/ 9, Name (print) / SigpAre/ CiIffipany / Department / Date Reviewer: MIKC- CA~SE~I~~~~..~--t ~~° ~,jS~'I5 ft-, 9/

Name (p=) / Signature /Comp.ar(y/ Departmenttl Date '

OSRC :

Chairman's Name (pr/ Signature / Date

[Required only for Programmatic Exclusion Screenings and 50759 Evaluations.]

LI-101-01, Rev. 7 Effective Date : 2/3105

60 .59 REVIEW FORM 13 It. SCREENINGS A. Licensing Basis Document Review

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

F- Operating License. YES NO] CHANGE # and/or SECTIONS IMPACTED 1Operating License F"

If YES", obtain NRC approval prior to implementing the change by initiating an LBD change in If ';

acc 'yo"r7dalce with NMM ENS-LI-113. (See Section 5.2[13] for exceptions.)

LBDs controlled under 60 .59 ' ~ NO CHANGE # (if applicable) and/or SECTIONS IMPACTED FSAR ED R TS Bases ED Z Technical Requirements Manual 0 n I Core Operating Limits Report 0 N NRC Safety Evaluation Report and R supplements for the initial FSAR' NRC Safety Evaluations for amendments to the Operating License' it "YES", perform an Exemption Review per Section lit OR perform a 50.59 Evaluation per Section IV gR obtain NRC approval prior to implementing the change. If obtaining NRC approval, document the LBD change in Section II.A.5;, no further 50 .59 review is required. However, the change cannot be implemented until approved by the NRC. AND initiate an LBD change in accordance with NMM1 ENS-1-1-113-LBDs controlled under other YES NO CHANGE # (if applicable) and/or SECTIONS regulations IMPACTED Quality Assurance Program ManUaj2 CD 2,3 Emergency Plan E0 IOR 4

Fire Protection Program' ED N (includes the Fire Hazards Analysis) 3,4 Offsite Dose Calculations Manual ED Z If "YES", evaluate any changes in accordance with the appropriate regulation AND initiate an LBD change in accordance with NMM ENS-1-1-113. No further 50.59 review is required.

on 5.2[51. No LBD change is required .

2 If "YES," notify the responsible department and ensure a 50.54 Evaluation is performed. Attach the 50.54 Review .

3 Changes to the Emergency Plan, Fire Protection Program, and Offsite Dose Calculation Manual must be approved by the OSRC in accordance with NMM OM-119.

4 If "YES," evaluate the change in accordance with the requirements of " facility's Operating License Condition or under 50 .59, as appropriate.

L1-101-01, Rev. 17 Effective Date: 213105

0.59 REVIEW FORM Page 3 of 13

2. Does the proposed activity involve a test or experiment not described in the FSAR?

If "yes," perform a 60.59 Evaluation per Section IV QR obtain NRC approval prior to implementing the change AND initiate an LBD change in accordance with NMM Lt-113 If obtaining NRC approval, document the change in Section II.A.5; no further 50.59 review is required . However, the change cannot be implemented until approved by the NRC .

LI-101-01, Rev . 7 Effective pate: 213105

50.59 REVIEW FORM Page 4 of 13 Explain why the proposed activity does or does not impact the Operating License/Technical Specifications and/or the FSAR and why the proposed activity does or does not involved new test or experiment not previously described in the FSAR . Discuss other LBDs if impacted. Adequate basis must be prgvided within the Screening such that a third-party reviewer can reach the same conclusions . Simply stating that the change does not affect TS or the FSAR is not an acceptable basis. '

The ER reviewed approves an acid cleaning solution for cleaning the RHR Heat Exchanger tubes, describes are acid mixinglinjecting process to Inject and recirculate the acid solution through the heat exchanger to perform the cleaning, and provides a guideline on the. allowable number of acid cleaning for the RHR Heat Exchangers. Similar acid flush processes, except that Injecting and collecting ports were readily available, using 4% citric acid solution have been performed routinely to clean the (T46) ESF Switchgear Room Coolers.

Safety evaluations have been performed for the acid flush. For some of the room coolers, more than 20 flushes have been performed. Similar acid flushes using 2.5% citric acid have also been performed before for flushing the SSW "A" and "g" piping. Safety evaluations have also been performed for those flushes. This safety evaluation reaffirms some resu#s from two previous safety evaluations, namely SE 88-OM and SE 87-0045, on similar acid flushes and focuses on the additional components present in the chemical cleaning boundary for the acid flush of the RHR Heat Exchangers.

The RHR Heat Exchangers are safety related heat exchangers cooled by the Standby Service Water. They are described in the FSAR and their heat removal capability requirements are specified therein. They are also mentioned in other licensing basis documents. However, cleaning or method of cleaning of the RHR Heat Exchangers is not mentioned in any LBD. Acid flush of some of the SSW-cooled heat exchangers, excluding the RHR Heat Exchangers, because of cross-tie of the SSWpiping with the Plant Service Water piping was committed by GGNS in the NRC GL 89-13 heat exchanger program. As expected, searches through all LSD's via AUTONOMY using*eywords "acid flush"; "acid cleaning"; and "chemical cleaning" yielded a number of hits, and the only relevant ones are NRC inspection reports regarding commitments and program establishment to acid flush those heat exchangers only, none about the RHR Heat Exchangers and none about the method of acid flush or the acid-flush chemical. Therefore, neither the proposed acid flush of the RHR Heat Exchangers nor the proposed acid-flush solujlon, let alone the allowable number of acid flushes, is described in any LSD, and implementation of this ER would not violate any LBD or require any changes to be made to any LSD.

The proposed acid mixing method would result In proper mixing of the Betz KI-2 (containing 40°A citric acid) and a Nalco penetrant (Nalco 73551 preferred), both approved by Chemistry, with SSWto form a 10% Betz Kl-2 solution containing approximately 4% citric acid and 200 ppm of the pentrant. A minute amount o¬ a defoaming agent, Betz Foamtrol CT, also approved by Chemistry, might be added at the discretion of Chemistry. Being a weak acid, the 4% citric acid solution would pose more a nuisance than a safety hazard.

Proper Personnel Protection Equipment Is to be worn during the acid cleaning work as directed by the RP.

Cautions are provided in the work instructions against spills and splashes. The floor drains will be covered by securely taping the cover to the floor, but not plugged, before the acid cleaning is started, as is done in acid flushing of the ESF Switchgear Room Coolers. This will make it easier to stop the acid cleaning process and restore the system if a postulated accident should occur. Upon completion of the acid cleaning, discharging the used 10% Betz KI-2 solution and Nalco penetrant contained in the SSW in the RHR Heat Exchangers and associated piping as well as the various drums after acid cleaning of the RHR Heat Exchangers via the SSW basin is approved by the State of Mississippi per an NPDES permit.

The system boundaries established for cleaning the RHR Heat Exchangers (1E12BOOIAIB and 1E12B002AIB) are shown on P81D M-1061CID between valves IP41FO14AIB and IP41FO68AIB as well as valves 1P41F120AIB,1P41F121AIB,1P41F166AIB,1P41F214A1B,1P41F167AIB,1P41F164A1B,1P41F158AIB,and 1P41FI65AIB. The proposed acid cleaning solution was evaluated in the ER with respect to its corrosion effect on and compatibility with all components within the chemical cleaning boundary and determined to be acceptable since the expected corrosion extent at the end of one acid cleaning of the specified duration would be well within the corrosion allowances. Other types of corrosion (crevice, IGSCC, pitting, etc.),

corrosion of weldingtbrazing metal or other corrosion mechanism possibilities were reviewed and determined not to be credible factors due to the nature of the selected chemical cleaning process, the specified chemical solution, the wetted materials within the established boundary and by following the prescribed process controls . The above-cited safety evaluations specifically discuss crevice corrosion and corrosion of welding1brazing materials as not being a concern when using a chemical cleaning process with a similar citric acid solution. Also, a non-metallic material previously not evaluated, the EPT material In the SSW Inlet and outlet butterfly Isolation valves, has been found, by reference and by testing, to be compatible with the proposed chemical solution. Another non-metallic material Is the plastic, KEL-F81, used for the valve seat in Anderson-Greenwood relief valves i P41 F100AIS, just like those for the ESF Switchgear Room Coolers, 1 P41 F127AIB, 1 P41 F138A/B,1 P41 F151 A/B, 1 P41 F194AIB, and 1 P41 F157A1B . No detailed information about the plastic Is available in the vendor manual and it does not appear that the Impact of 4% citric acid on KEL-F81 has been specifically evaluated. However, none of the ESF Switchgear Room Cooler relief valves, which LI-101-01, Rev. 7 Effective Date : 213105

9 REVIEW FORM Pace 5 of 13 have been through numerous flushing by the acid solution, or 11341Fi00A18, which have also been through two SSW piping acid flushes before, was actually found to have suffered noticeable damages, except perhaps some minor leakage, for which the reasons were unknown. Also, relief valves 1P41F100A1f3 are located about 30" more or less vertically up from the junction to the SSW main piping . As discussed in the ER, unless the valve Is leaking during the acid flushing, the turbulent eddies alone could not possibly carry the citric acid solution into the small branch line to any noticeable distance within the time frame of the acid cleaning .

Indeed, relief valves 1 P41 F100AIB are currently not leaking. Therefore, it Is believed that the upcoming acid cleaning of RHR "B" Heat Exchangers will not affect relief valve 11341 F100B. It is for this same reason that rinsing of this branch line, which will require manipulation of the relief valve, will not be performed immediately after the acid flush. In reality, since the butterfly valves 11341 FO14A1B S 1 P41 F068AB are not leaking and are more than 18' away from the heat exchanger nozzles, they are not expected to be affected by the cleaning solution either. The temporary hoses, valves, fittings, and injection pumps used are similar to those used for the ESF room coolers or brand new and compatible with the cleaning solution . No other non-metallic materials are known to exist within the system boundary identified Based solely on the corrosion rate of the limiting component, carbon steel, 11 acid cleaning could be allowed. However, two acid cleanings, each with 9-hour acid recinculating1soaking after at most 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> of acid mixingfnjecting, are approved based conservatively on the MIC Program data.

The proposed method of acid injection would, for current lack of acid injecting and collecting locations, require conversion of the RHR Pleat Exchanger outboard drain valves into injection and collection ports for the acid solution with the heat exchanger Inboard drain isolation valves closed. The affected train of .the SSWIRHR system would be declared INOP upon the valve conversion, and an LCO entered. The conversion would be performed as a part of the acid cleaning In accordance with applicable procedures such as that for welding, and the modified valves and the properly rated temporary hoses, isolation valves, and Injection Pump would form an adequate new pressure boundary for the RHR fluid. During the cleaning process while the inboard isolation drain valves are open, the RHR Heat Exchangers and SSW System would be maintained functional by closing but not tag-closing the SSW Inlet and outlet isolation valves for the RHR Heat Exchangers and posting a dedicated Operator near the inboard isolation valves to close the valves In case of actuation of the SSW System. After completion of the acid cleaning, a Maintenance Leak Test will be performed. Also the RHR Heat Exchangers monthly EPI performed by OPS and currently scheduled In the same week as, but just before the acid cleaning, will be performed only after the acid cleaning as a post-cleaning test to verify that the SSW flow rate has not been adversely affected. These are considered sufficient for OPS to clear the LCO.

The 10% Betz KI-2 solution obtained from mixing the chemicals in one of the Mixing Drum would be Injected into the heat exchangers and associated piping through the injection port, pushing the existing SSW out of the collection port into the other Mixing Drum. After completion of the acid injection within a maximum of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, a recirculating loop including only one Mixing Drum would be established to recirculate the acid solution through the loop. At the end of the recirculation phase, the outboard drain valves would be restored and the RHR Heat Exchangers declared Operable. The cleaning process may Include a period of time for soaking the tubes with the acid solution thereafter before starting the SSW Pump to rinse off the acid solution. The acid cleaning process is monitored by taking SSW samples at 3, 6, and 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> into the Recirculating Phase, and by watching closely the RHR-side pressure (1E12N026A18) and conductivity alarm (1E12L602AIB) to detect a tube leak. The total duration of the recirculating and soaking is limited to 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.

The recirculating1soaking Is also to be deemed complete when the maximum allowable copper concentration In the SSW sample exceeds 4725 ppm. The cleaning process would be promptly terminated upon SSW system automatic Initiation, detecting RHR-side pressurization and confirming an RHR Heat Exchanger tube leak, or detecting radioactivity in SS Wsamples by isotope analyses (by Chemistry Dept.). In case of a tube leak with the RHR side leak tight, only a small amount of SSW could leak into the RHR side when the SSW side pressure is higher, thus raising the reading of 1E12N026AIB. The SSW would be flushed to the Suppression Pool, thus diluting the concentration of the citric acid, and then cleaned up by means of the precoat system . One quart of 4% citric acid solution would be diluted to a undetectable concentration level of 10 ppb In the Suppression Pool. The concentration would be further reduced after cleaning by the pmcoat system before any chance of the citric acid entering the reactor coolant system. If the SSWpressure drops below the RHR-side pressure with a tube leak, the SSW side would be contaminated by the RHXside fluid.

Significant leakage flow would be detectable by observing the levels In the two Mixing Drums and isotope analysis of the SSW sample or even by the Area Radiation Monitor, The waste SSW collected and the excess 10% Betz KI-2 solution remaining in the Mixing Drums at the end of the acid cleaning will be pumped back to the SSW basin via any of the nearby SSW system valves identified in the procedure (07-1-34-T46-BOOR-2) for acid flushing of the ESF Switchgear Room Coolers, such as 11341F352,11341F337, etc. afteran isotope analysis of a sample has veered no radioactive contamination. The relatively low concentration of citric acid and removed corrosion products In the acid solution remaining in the cleaning solution Is not expected to cause any damage to the SSW basin area, based on previous experiences with the SSW piping flushes. In particular, only slight etching on the concrete has been noticed after the SSW piping flushing. The tower fan blades would not be wetted . The ceramic fill material is compatible with citric acid and no adverse effects LI-101-01, Rev. T Effective Date: 2!3105

50.59 REVIEW FORM Page 6 of 13 have been noticed before . The small total amounts of citric acid, Nalco penetrant, and removed corrosion products after being diluted by the vast basin volume will be harmless to the system and the SSW basin water will be allowed to be discharged to the river per a NPDE5 permit A permit for storing combustible material in the work area will be obtained since the total amount would exceed the normal allowance but would not be excessive. '

The acid mixingfnjecdng equipment, drums, and hoses would be set up mainly on the grating of the spacious RHR Heat Exchanger Room at El. 119' of Auxiliary Building in the low-dose area . All pumps, valves, fittings, drums, and hoses will be either new or previously verified to be free from radioactive contamination before use. The drum setup would be such that the grating loading will not be exceeded. The Acid Drums and filled Waste Drums may be set up outside the RHR Heat Exchanger Room,. but the door will be closed, except when entering and leaving during acid Injection for the purpose of draining a required amount of acid from the Acid Drum to be added to the Mixing Drum inside the room and replacement of the filled Waste Drum with an emptied Acid Drum .

4. References Discuss the methodology for performing LBO searches . State the location of relevant licensing document 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 5.5 .1(5)(d) of L1-101 . NOTE: Ensure that manual searches are performed using controlled copies of the documents. If you have any questions, contact your site Licensing department, Keyword searches were performed with all LBDs listed in AUTONOMY selected. The hits were reviewed to ensure that they are not related to the proposed acid flush of the RHR Heat Exchangers and associated piping.

LBDs/Documents reviewed via keyword search : Keywords :

All LBDs listed in AUTONOMY "acid flush" ;

"acid cleaning",

al cleaning".

LBDsIDocuments reviewed manually:

None

5. Is the validity of this Review dependent on any other change? [l Yes No If "YES", list the required changesisubmittals . The changes covered by this 50.59 Review cannot be implemented without approval of the other identified changes (e.g., license amendment request). Establish an appropriate notification mechanism to ensure this action is completed .

(List the required changes Isubmitfals.)

LI-101-01, Rev. 7 Effective Date: 213105

50.59 REVIEW FORM Page 7 of 13 B. ENVIRONMENTAL SCREENING If any of the following questions is answered "yes," an Environmental Review must be performed in accordance with NMM Procedure ENS-EV-115, "Environmental Evaluations," and attached to this 50.59 Review. Consider both routine and non-routine (emergency) discharges when answering these questions .

Will the proposed Change being evaluated :

Yes No 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 of ponds)?

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

3. [1 Involve dredging activities in a lake, river, pond, or stream?

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

5. F] Increase the concentration or quantity of chemicals being discharged to the river, lake, or air?

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

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

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

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

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

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

12. (Q Involve the installation or use of equipment that will result in a new or additional air emission discharge?

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

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

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

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

1-1-141-01, Rev . 7 Effective Date: 213105

50.58 REVIEW FORM Page 8of13 C. SECURITY PLAN SCREENING If any of the following questions is answered "yes," a Security Plan Review must be performed by the Security Department to determine actual Impact to the Plan and the need for a change to the Plan.

Could the proposed activity being evaluated :

Yes No

1. C] Add, delete modify, or otherwise affect Security department responsibilities (e.g.,

luding fire brigade, fire watch, and confined space rescue operations)?

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

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

4. [ :1 Affect (block, move, or alter) security lighting by adding or deleting lights, structures, buildings, or temporary facilities?

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

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

7. 014 Modify"or otherwise affect (block, move, or after) 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 control equipment, intrusion detection equipment, other security equipment, or to the Central Alarm Station or the Secondary Alarm Station?

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

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

Documentation for accepting any "yes" statement for these reviews will be attached to this 50.59 Review or referenced below.

LI-141-01, Rev. 7 Effective Date: 213105

50.59 REVIEW F Page 9 of INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) SCREENING (NOTE. This section is not applicable to Waterford 3 and may be removed from 50.59 Reviews performed for Waterford 3 proposed activities.)

If any of the following questions is answered "yes," an ISFSI Review must be performed in accordance with NMM Procedure ENS-t.I-112, "72.48 Review," and attached to this Review.

Will the proposed Change being evaluated :

1. [] Any activity that directly impacts spent fuel cask storage or loading operations?

Involve the Independent Spent Fuel Storage Installation (ISFSI) including the concrete pad, security fence, and lighting?

Involve a change to the on-site transport equipment or path from the Fuel Building to the ISFSI?

01 Involve a change to the design or operation of the Fuel Building fuel bridge including setpoints and limit switches?

5. 11 Involve a change to the Fuel Building or Control Room(s) radiation monitoring?

6. [] 01 Involve a change to the Fuel Building pools including pool levels, cask pool gates, cooling water sources, and water chemistry?

[] Involve a change to the Fuel Building handling equipment (e.g., bridges and cask cranes, structures, load paths, lighting, auxiliary services, etc)?

8. a Involve a change to the Fuel Building electrical power?

9. Involve a change to the Fuel Building ventilation?

10. 0 I" Involve a change to the ISFSI security?

11 . [] Involve a change to off-site radiological release projections from non-ISFSI sources?

12. o Involve a change to spent fuel characteristics?

13. [] Redefine/change heavy load pathways?

14 . © Fire and explosion protection near or in the on-site transport paths or near the ISFSI?

15. ~] involve a change to the loading bay or supporting components?

16 . C] 111 New structures near the ISFSI?

17. © Modifications to any plant systems that support dry fuel storage activities?

18. [] Involve a change to the nitrogen supply, sere ir, demineralized water or borated water system in the Fuel Building?

LI-101-01, Rev . 7 Effective Date: 213105

50 .59 REVIEW FORM Page 1 0 of 13 5159 EVALUATION License Amendment Determination Does the proposed Change being evaluated represent a change to a method of evaluation Yes ONLY? If "Yes," Questions 1 - 7 are not applicable ; answer only Question 8. If "No," answer No all questions below.

Does the proposed Change:

1 Result in more than a minimal increase in the frequency of occurrence of an accident El Yes previously evaluated in the,FSAR? Z No 4W LI-101-01, Rev. 7 Effective Date : 213105

60.59 REVIEW FORM Page 1 1 of 13 BASIS:

The corrosion impact of the proposed 4 % citric acid solution on all components, including weld material, within the chemical cleaning boundary for the acid flush of the RHR Heat Exchangers was evaluated. Crevice corrosion attack was not considered a concern since chemical environments associated with the cleaning process are not of a nature to create an aggressive environment for crevice corrosion mechanisms in the case of the SSW system, and the crevice corrosion requires before accelerated metal dissolution begins some incubation period that would not be available because the length of time crevices in the SSW system were to be exposed to the chemical cleaning process would be sufficiently short and areas that could be saturated by this chemical environment would be subsequently flushed. The critical compondrits affected by an acid flush were found to be carbon steel piping, 70-30 CuNI heat exchanger tubes, Ethylene Propylene Terpolymer (EPT) elastomer seats of SSW isolation butterfly valves (I P41F014A/B &

I P41 F068AIB), and KEL-F81 plastic seats of the Anderson Greenwood relief valves 1P41F100A1B .

The corrosion impact on carbon steel and 7530 CuNI by a 9-hour acid recirculating1soaking, following less than 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> of mixinglinjecting of the acid solution has been determined to be insignificant and well within the corrosion allowance, The APT elastorner is quite compatible with citric acid solution . No detailed information is available on the KEL-F81 plastic in the vendor manual. However, no adverse effect on these non-metallic materials has been observed from numerous 4% citric acid flushing of the ELIO Switchgear Room Coolers or a previous flush of the entire SSW system using a 2.6 % citric acid solution for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A recent 24-hour soaking of the EPT seat material from a similar butterfly valve in the proposed acid flush soluti also resulted in no significant visible changes and no noticeable change in its Shore "A" Hardness value (approximately 65). Since the safety positions of these safety-related valves are "OPEN" so as to allow SSW flow under postulated accident conditions, leak-tightness is not required for their safety function . Therefore, even if unlikely, slight degradation in the EPT material should result in minor leakage after such exposure, it would not pose a safety concern. Also, even with slight leakage through the EPT packing, there would still be sufficient torque for the valve operators to open the valves were they found to be closed unexpectedly . In reality, these butterfly valves as well as the relief valves, being leak-tight and far away from the Junctions with the main piping, are not expected to be affected by the cleaning solution during the short cleaning duration . Therefore, no components will be degraded by the acid flush as to increase the frequency of occurrence of an accident previously evaluated in the FSAR .

In the beginning of the proposed acid cleaning, the HX outboard isolation drain valves (I P41 F1 65A/B & 1 P41 F1 67A/B) of the affected train would be temporarily converted into injection and collection ports for the acid solution, and the RHR train would be declared INOP but functional . The conversion is to replace the valve stem and other internals with a quill, which will be held in place to seal off the outlet opening and allow the acid solution to go through the inlet opening. The inboard drain isolation valves (IP41FI66A/B, 1P41F214A/B, 1P41F158AlB, and I P41F164A/B) would serve the function of isolation during the conversion work ; therefore, the conversion work would not increase the frequency of occurrence of an accident previously evaluated in the FSAR. During the acid mixing and injecting as well as the acid recirculating/soaking, the SSW inlet and outlet isolation butterfly valves would be closed but not tag-closed so that they could open automatically in case of an actuation of the SSW system . The inboard isolation valves 1 P41F214AlB and 1P41F164AlB, and the temporary injection and collection ports would be open ; therefore, by posting a dedicated Operator to close the inboard drain isolation valves incase of an accident resulting in actuation of the SSW Pump, the RHR system would remain functional . The only hazard that this could present in case of an accident resulting in actuation of the SSW Pump during these periods would be a potential for SSW corn out of the collection port or even the injection port and briefly overflowing the Mixing Drums.

Caution would be placed in the work package to minimize this hazard, which would not be a plant safety concern. Thus, the inboard isolation valves would serve the function of isolation during these periods of acid cleaning work and the frequency of occurrence of an accident previously evaluated in the FSAR would not increase as a result .

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a Yes structure, system, or component important to safety previously evaluated in the FSAR? No LI-101-01, Rev. 7 Effective Date : 213105

50.59 REVIEW FORM Page 12 of 13 BASIS:

The corrosion impact on limiting components within the chemical cleaning boundary, namely carbon steel piping and 70-30 CuNI tubes, by the proposed acid flush has been determined to be well within the corrosion allowances. The safety function of the isolation valves, butterfly valves or otherwise, would not be affected by the acid flush. Therefore, no structures, systems, or components important to safety within the chemical cleaning boundary would be affected by the acid flush as to increase the likelihood of occurrence of a malfunction . The inboard drain valves are about 15' from each other; therefore, a single dedicated Operator can isolate them In case of accident within a relatively short time. This factor would not increase the likelihood of occurrence of not being able to close the drain valves.

3, Result in more than a minimal increase in the consequences of an ace! nt previously n Yes evaluated in the FSAR? ' No BASIS :

The integrity of the RHR Heat Exchangers and associated piping would not be compromised by,the acid flush since the corrosion effect would be well within the allowance. The safety function of all isolation valves would bot be impaired by the acid flush . Therefore, no systems or components within the chemical cleaning boundary would be prevented from performing their safety function during an accident previously evaluated in the FSAR as to cause any increase in the consequences of the accident. The dedicated Operator would be able to close the inboard drain isolation valves in case of an accident within a relatively short time so that they could perform their safety function,-luring the accident and would not increase the consequences of the accident.

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, Yes system, or component important to safety previously evaluated in the FSAR? No BASIS:

The acid flush would not compromise the integrity of the SSW system boundary or degrade the heat exchanger function or affect the operation of any other safety system/component within the chemical cleaning boundary required for mitigating the consequences of an accident; therefore, it would not cause any increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the FSAR.

5. Create a possibility for an accident of a different type than any previously evaluated in the El Yes FSAR? No BASIS:

The impact of the proposed acid flush on all exposed components within the chemical cleaning boundary is insignificant and well within the corrosion allowances. Leaving the inboard drain valves and the injection and collection ports open during the acid cleaning work would only add the need to post a dedicated Operator to close the inboard drain valves but would not create a possibility for an accident of a different type to occur since the inboard drain valves would serve the isolation function. Therefore, no possibility for an accident of a different type than any previously evaluated in the FSAR could be created .

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

The impact of the proposed acid flush on all exposed components within the chemical cleaning boundary is insignificant and well within the corrosion allowances. The inboard drain valves would be closed by the posted dedicated Operator in case of a postulated accident resulting in actuation of the SSW System to serve the isolation function as usual. The results of any malfunction of a structure, system, or component important to safety previously evaluated would not be made different by the acid flush. Therefore, no possibility for a malfunction with a different result than any previously evaluated in the FSAR could be created.

7. Result in a design basis limit for a fission product barrier as descritaed in the FSAR being Q Yes exceeded or altered? 0 No LI-101-01, Rev. 7 Effective Date: 213105

50.59 REVIEW FORM Page 13 of 13 BASIS:

The impact of the proposed acid flush on all exposed components within the chemical cleaning boundary is insignificant and well within the corrosion allowances. The inboard drain isolation valves would be closed within a relatively short time by the posted dedicated Operator in case of a postulated accident resulting in actuation of the SSW System, causing negligible loss in the SSW.

Therefore, the acid flush would not change any result of accidents previously analyzed in the FSAR. Hence, It could not result in a design basis limit for a fission product 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 establishing Q Yes the design bases or in the safety analyses? No BASIS:

The proposed acid flush is to chemically clean the RHR Heat Exchangers and associated piping to be better prepared for the Eddy Current Testing and possibly improve the heat exchanger thermal performance. The heat removal capability of the RHR Heat Exchangers used in safety analyses was based on the design fouling level and would not be affected by the acid flush . Therefore, the proposed acid flush would not affect any method of evaluation described in the FSAR used in establishing the design bases or In the safety analyses.

if any of the above questions is checked "YES", obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure ENS-LI-113 .

LI-101-01, Rev. 7 Effective Date: 213105

GGNS 50.59 Safety Evaluation Number SE 2005-0004-ROI

50.59 REVIEW FORM Page I of 13 OVERVIEW / SIGNATURES Facility: Grand Gulf Nuclear Station Document Reviewed : ER-GG-2003-0359-000 Change/Rev .: 0 System Designator(s)/Description : I E12 / Acid Flush of Residual Heat Removal Heat Exchangers Description of Proposed Activity :

This safety evaluation is a revision of the previously performed safety evaluation (#SE 2005-0004-Roo) to re-assess the operability of the RHR and SSW systems when the heat exchanger inboard drain valves are closed.

The ER reviewed approves an acid-flush solution and develops a process to acid-flush the tube side of an RHR heat exchanger unit (1E1213001A/2A or 1E121300113/2B). It also provides a guideline to the number of acid flushes allowed . The purpose of the acid flush is to facilitate eddy current testing of the heat exchanger tubes.

It could also improve the thermal performance of the flushed heat exchanger unit.

Work order #35963 provides instructions per the ER to perform the acid flush of RHR "B" Heat Exchangers 1E12B001B12B. It also includes instructions and cautions to cutout the RHR "B" Heat Exchanger outboard drain valve. 1P4 I F1 67B, and to weld back a like-for-like replacement valve.

Check the applicable review(s) : (Only the sections indicated must be Included In the Review.)

EDITORIAL CHANGE of a Licensing Basis Document Section I SCREENING Sections I and 11 required 50.59 EVALUATION EXEMPTION Sections 1, 11, and III required 150 .59 EVALUATION (#: SE 2005-0004-ROI Sections 1, 11, and IV required Preparer: Shyy-Jon g D. Lin / / EOH EP&C /

Name (print) / SignnaffrV CdI`np6ny / Department / Date Reviewer MIWC CAU56Y su. f 4~" ay/- A, Name (print) / Signatur6 / Compan 'ate

, ,r Chairman's Name (6rint) / Signature / Date

- (Required only for Programmatic Exclusion Screenings and

'- 50 .59 Evaluations .)

LI-1 01 -01, Rev. 8; Effective Date: 6/23/05

50.59 REVIEW FORM Page 2 of 13 II. SCREENINGS A. Licensina Basis Document Review I. Does the proposed activity impact the facility or a procedure as described in any of the following licensing Basis Documents?

Operating License YES NO CHANGE # and/or SECTIONS IMPACTED Operating License EO GO TS NRC Orders If "YES," obtain NRC approval prior to implementing the change by initiating an LSD change in accordance with NMM LI-1 13. (See Lt-10 1 for exceptions .)

LBDs controlled under 50.59 YES NO CHANGE # (if applicable) and/or SECTIONS IMPACTED FSAR Ej 0 A TS Basses Technical Requirements Manual Core Operating Limits Report 0 0211 NRC Safety Evaluation Report and 0 101 supplements for the initial FSAR' NRC Safety Evaluations for amendments to the Operating License' If "YES," perform an Exemption Review per Section ](I OR perform a 50 .59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change by initiating an LSD change in accordance with NMM LI-113 . If obtaining NRC approval, document the LSD change in Section II.A.5. However, the change cannot be im- plemented

-- ---- until

- - approved

--- - - - by-the NRC. Complete Section Ii.

LBDs controlled under other YES NO CHANGE # (if app{icable) andfor regulations SECTIONS IMPACTED Quality Assurance Program ManUO12 C] Z4 Emergency Plan" Fire Protection Prograrn 3 4 ~

(includes the Fire Hazards Analysis)

Offsite Dose Calculations MamaN If "YES," evaluate any changes in accordance with the appropriate regulation AND initiate an LBD ---~

change in accordance with NMM LI-1 13.

If "YES," see LI-10i . No LBD change is required.

If "YES," notify the responsible department and ensure a 50 .54 evaluation is performed. Attach the 54.54 evaluation.

Changes to the Emergency Plan, Fire Protection Program, and Oftsite Dose Calculation Manual must be approved by the OSRC in 3

accordance with NMM OM-119.

' If 'YES,' evaluate the change in accordance with the requirements of the facility's OnratAg License Condition or under 50 .59, as appropriate.

LI-101-01, Rev. 3; Effective Date: 6/23105

0 0 Page 3 of 13

2. Does the proposed activity involve a test or experiment not described in the FSAR? Yes If "YES," perform a 50.59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change AND initiate an LSD change in accordance with NMM Ll-113, if applicable . If obtaining NRC approval, document the change in Section li .A.5. However, the change cannot be implemented until approved by the NRC . Complete Section Ii .

3. Saris Explain why the proposed activity does or does not impact the Operating LicensefTechnicai Specifications and/or the FSAR . If the proposed activity involves a potential test or experiment not previously described in the FSAR also include an explanation. Discuss other LBDs if impacted . Adequate basis must be provided within the Screening such that a third-party reviewer can reach the same conclusions . Simply stating that the change does not affect TS or the FSAR is not an acceptable basis .

The ER reviewed approves an acid cleaning solution for cleaning the RHR Heat Exchanger tubes, describes an acid mixinglinjecting process to Inject and recirculate the acid solution through the heat exchanger to perform the cleaning, and provides a guideline on the allowable number of acid cleaning for the RHR Heat "S" Heat Exchangers Exchangers . The work order provides instructions to perform the acid flush of RHR 1E12BOOI S12B. It includes instructions and cautions to cutout the RHR "B" Cleat Exchanger outboard drain valve, 1P41F167S, and to weld back a like-for-like replacement valve.

Similar acid flush processes, except that injecting and collecting ports were readily available, using 4% citric acid solution have been performed routinely to clean the (T46) ESF Switchgear Room Coolers. Safety evaluations have been performed for the acid flush . For some of the room coolers, more than 20 flushes have been performed . Similar acid flushes using 2 .5'10 citric acid have also been performed before for flushing the SSW "A" and "B" piping. Safety evaluations have also been performed for those flushes . This safety evaluation reaffirms some results from two previous safety evaluations, namely SE 88-0006 and SE 87-0045, on similar acid flushes and focuses on the additional components present In the chemical cleaning boundary for the acid flush of the RHR Heat Exchangers.

The RHR Heat Exchangers are safety related heat exchangers cooled by the Standby Service Water. They are described in the FSAR and their heat removal capability requirements are specified therein . They are also mentioned In other licensing basis documents . However, cleaning or method of cleaning of the RHR Heat Exchangers is not mentioned In any LSD . Acid flush of some of the SSW-cooled heat exchangers, excluding the RHR Heat Exchangers, because of crosstie of the SSW piping with the Plant Service Water piping was committed by GGNS In the NRC GL 89-13 heat exchanger program. As expected, searches through all LED's via AUTONOMY using keywords "acid flush", "acid cleaning", and "chemical cleaning" yielded a number of hits, and the only relevant ones are NRC inspection reports regarding commitments and program establishment to acid flush those heat exchangers only, none about the RHR Heat Exchangers and none about the method of acid flush or the acid-flush chemical . Therefore, neither the proposed acid flush of the RHR Heat Exchangers not the proposed acid-flush solution, let alone the allowable number of acid flushes, is described in any LSD, and implementation of this ER would not violate any LSD or require any changes to be made to any LSD .

RHR & SSW Svstems Operable When Inboard Drain Valves Are Closed The proposed method of acid Injection would, for current lack of acid Injecting and collecting locations, require conversion of the RHR Heat Exchanger outboard drain valves into injection and collection ports for the acid solution with the heat exchanger inboard drain isolation valves closed . An evaluation of the operability of the RHR "S" system and SSW "$" system, which would be applicable to the "A" train as well, for the period between the time any of the heat exchanger outboard drain valves or the piping downstream of the Inboard isolation valves are first modified and the time when the inboard drain valves are opened to start acid-solution injection, considering the following results of further evaluations by Design Engineering-Mechanical and Piping/Civll :

1. All of the inboard drain valves are maintained normally closed per P&ID M1061D Rev. 38 and SOI 04-1 P41-1 Rev. 122. Additionally, the downstream outboard drain valves 1P41F1658 and IP41Fi67B do not have any operational function to support the SSW and RHR system safety functions. They are normally closed during all modes of SSW operation, as are the upstream valves.

LI-101-01, Rev . 8 ; Effective Date : 6123105

50.59 REVIEW FORM Page 4 of 13

2. Per MS-05, Rev . 5, for systems with less than 900psig pressure rating are only required to have a single drain isolation valve. The HBC drain line classification is rated for 150psig . Therefore the downstream valve is not required in each line per this standard.

3. The line class for all of the associated piping is 1'~-HBC-104. Per MS-02, Rev. 50 the design conditions for this piping is 180 josig (195 psig for SSWpiping below el. 1331) at 150F. Per US-03 Rev. 1, for 2 - and smaller HBC line classes, valves are 1500# socket welded material class CBC (1500 psig, carbon steel, ASMEIII-3) . Therefore the piping and single valve have sufficient pressure rating to maintain the boundary.

4. Calculation MC-01P41-03016 Rev. 0 determines the maximum allowed SSW system leakage to be 15 gpm given all leakage allowed by current procedures, calculations, programs, etc . Therefore even if the upstream valves leak by, as long as the total is less than 15 gpm, the SSW system will still have sufficient inventory for 30 days post-LOCH operation and will remain operable . This can be verified by opening the IP41FI65B and IP41FI67B and measuring the leakage past the closed upstream valves at standby conditions, and using a ratio to compare to the maximum allowed leakage at design conditions (Flow rate is proportional to the square root of the pressure difference) . Per drawing M1348C, the RHR heat exchanger elevation is at 104'8-314" and the valves are all located at elevation 93'6" per FSK-S-1061D-053-B through -056-B. This is an elevation difference of 11.23 feet, which corresponds to 4.86 psig of static head. Using a ratio to compare the maximum allowed 15 gom at 195 psig, the maximum allowed total leakage past the 4 boundary valves is Z36 Upm at static conditions .

piping I train from RHR Heat Exchanger 1E1280028 has been designed by utilizing criteria specified in Engineering Standard A118 far small bore piping. These drain piping have two "three directional" pipe supports in the vicinity of valves IP41FI648 and IP41FI588 for RHR Heat Exchanger OIE125001a and two "three directional" pipe supports in the vicinity of valves 1P4 1F1668 and IP41F2148 for RHR Heat Exchanger Q1E12B002B . Per M-18 design criteria, 1" schedule 80 piping all system is acceptable for unsupported piping span of 22" from "three directional" support for loading conditions, including seismic event.

6, The rubber hoses, fittings, temporary isolation valves, and the Injection Pump are all rated for at least 150 psig, and the hoses connected to the quills installed on the outboard drain valves will be taped down on the floor in the immediate vicinity of the quills and tied to handrails and other appropriate structure as they are run to the side of the room and up to the 119' elevation, Thus the weights of the whole not placed assemblies would be on the piping and would have negligible impact on the piping in case of a seismic event.

It can be concluded that, as long as the total leakage rate from the four inboard drain valves is less than 2 .36 g,pm and that the piping cut to remove an outboard drain valve is made at a location that is within 22" from the nearest "three dimensional" Support, Me SSW "B" system integrity will be maintained by the piping and the inboard drain valves while modifications are made to the downstream piping or valves.

KQ# 3063 provides notes and cautions to achieve the following:

1. Perform a leakage test for the inboard drain valves, IP4117214B and 1PAiF1668 for 1E1280028 and IP41FI64B and IP41FI58B for 1E1280018, before cutting the piping to replace the outboard drain valve IP41FI67B, and later during the acid flush work session, before starting the valve conversion to ensure that the total leakage rate from the 4 inboard drain valves is less than 2.36 gpm, Z Identify the cut locations for cutting the pipe to replace the outboard drain valve fP41F167B, ensuring that the cut locations are within 22" from the nearest "three dimensional" support, Thus, the RHR "B" system and the SSW '1B " system will remain fully capable of performing their design safety functions during the pipe cutting, welding, and valve disassembly and modification associated with the conversion of the outboard drain valves while the inboard drain valves are closed ; Therefore, they need not be declared INOP during this time, In performing some of the work, the High Energy Line Break (HELB) door #1A202 has to be blocked open. An evaluation has previously been performed for blocking open HELB doors including #1A202 in ER-GG-2005-0038-000, Rev. 0 for which a 50.59 screening was performed. The results of the ER showed it to be acceptable to block the door open as long as the annual limits of 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br /> and 3.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> for the case RHR not operating and the one RMR operating, respectively, are not exceeded. The door blocking will be performed in accordance with plant procedure Of-S-06-2 . As of today, the number of hours expended are 0.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and 0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> for not-operating case and operating case, respectively. The number of hours that might be used in this acid cleaning work will not exceed 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br />, thus within the limit established in that ER.

L1-40'l-01, Rev. 8; Effective Date: 6/23/05

50 .59 REVIEW FORM Page 5of13 Once the inboard drain valves are opened (following modifications for acid injection capability) to allow acid infection, both RHR "B"and SSW "B" will be declared INOP and LCO entered, but they will be maintained functional by posting an Operator nearby the inboard drain valves (within about 15' between the two trains) to close them in case of an SSW "B" pump initiation . The SSW inlet and outlet isolation butterfly valves, 1P41F014B and 1P41F06813, will be closed but not tagged during the acid mixing, injecting, and recirculating, so that they will open automatically upon an SSW system initiation.

Valve Convers The valve conversion would be performed as a part of the acid cleaning in accordance with applicable procedures such as that for welding, and the modified valves and the properly rated temporary hoses, isolation valves, and Injection Pump would form an adequate new pressure boundary for the SSW. After completion of the acid cleaning, a Maintenance Leak Test will be performed. Also the RHR Heat Exchangers monthly EPI performed by OPS will be performed only after the acid cleaning as a post-cleaning test to verify that the SSW flow rate has not been adversely affected . These are considered sufficient for OPS to clear the LCO.

Acid Flush The proposed acid mixing method would result In proper mixing of the Betz K¬-2 (containing 40414 citric acid) and a Nalco penetrant (Nalco 73551 preferred), both approved by Chemistry, with SSW to form a 10°14 Betz KI-2 solution containing approximately 4% citric acid and 200 ppm of the pentrant . A minute amount of a defooming agent, Betz Foamtrol CT, also approved by Chemistry, might be added at the discretion of Chemistry. Being a weak acid, the 44/4 citric acid solution would pose more a nuisance than a safety hazard .

Proper Personnel Protection Equipment Is to be worn during the acid cleaning work as directed by the RP .

Cautions are provided in the work instructions against spills and splashes . The floor drains will be covered by securely taping the cover to the floor, but not plugged, before the acid cleaning is started, as Is done In acid flushing of the ESF Switchgear Room Coolers. This will make it easier to stop the acid cleaning process and restore the system. The requirements for covering and uncovering the floor drains will be per established plant procedures and maintenance practices for similar activities . Upon completion of the acid cleaning, discharging the used 10°14 Betz KI-2 solution and Nalco penetrant contained in the SSW in the RHR Heat Exchangers and associated piping as well as the various drums after acid cleaning of the RHR Heat Exchangers via the SSW basin ¬s approved by the State of Mississippi per an NPDES permit.

The system boundaries established for cleaning the RHR Heat Exchangers (1E128001A/B and 1E128002AIB) are shown on P&ID M-1061 C/D between valves 1 P41 F014A/B and 1 P41 F068A/B as well as valves 1P41F120At8,1P41F121A/B,1P41F166A1B,1P41F214At8,1P41F167AlB,1P41F164A/B,1P41F158A/B,and 1 P41 F165AIB. The proposed acid cleaning solution was evaluated in the ER with respect to its corrosion effect on and compatibility with all components within the chemical cleaning boundary and determined to be acceptable since the expected corrosion extent at the end of one acid cleaning of the specified duration would be well within the corrosion allowances . Other types of corrosion (crevice, IGSCC, pitting, etc.), corrosion of welding/brazing metal or other corrosion mechanism possibilities were reviewed and determined not to be credible factors due to the nature of the selected chemical cleaning process, the specified chemical solution, the wetted materials within the established boundary and by following the prescribed process controls. The above-cited safety evaluations specifically discuss crevice corrosion and corrosion of weidinglbrazing materials as not being a concern when using a chemical cleaning process with a similar citric acid solution .

Also, a non-metallic material previously not evaluated, the EPT material in the SSW inlet and outlet butterfly isolation valves, has been found, by reference and by testing, to be compatible with the proposed chemical solution . Another non-metallic material is the plastic, KEL-F$1, used for the valve seat ¬n Anderson-Greenwood relief valves 1P41F100AlB, just like those for the ESF Switchgear Room Coolers, 1P41F127A/B, 1P41F138AIB,1P41F151AlB, 1P41F194A/B, and 1P41F157A/B. No detailed information about the plastic is available in the vendor manual and it does not appear that the Impact of 4814 citric acid on KEL-F$1 has been specifically evaluated. However, none of the ESF Switchgear Room Cooler relief valves, which have been through numerous flushing by the acid solution, or IP41F100AlB, which have also been through two SSW

!ping acid flushes before, was actually found to have suffered noticeable damages, except perhaps some minor leakage, for which the reasons were unknown. Also, relief valves IP41F100AIS are located about 30" more or less vertically up from the junction to the SSW main piping . As discussed In the ER, unless the valve is leaking during the acid flushing, the turbulent eddies alone could not possibly carry the citric acid solution into the small branch line to any noticeable distance within the time frame of the acid cleaning . Indeed, relief valves IP41F100A/B are currently not leaking. 'therefore, It is believed that the upcoming acid cleaning of RHR "B" Heat Exchangers will not affect relief valve 1 P41 F100B. It Is for this same reason that rinsing of this branch line, which will require manipulation of the relief valve, will not be performed immediately after the LI-101-01, Rev. 8; Effective Date; 6/23/05

50 .59 REVIEW FORM e 6of13 acid flush. In reality, since the butterfly valves 1P41F014AIS & 1P41F068A1B are not leaking and are more than 18' away from the heat exchanger nozzles, they are not expected to be affected by the cleaning solution either . The temporary hoses, valves, fittings, and Injection pumps used are similar to those used for the ESF room coolers or brand new and compatible with the cleaning solution . No other non-metallic materials are known to exist within the system boundary identified . Based solely on the corrosion rate of the limiting component, carbon steel, 11 acid cleanings could be allowed . However, two acid cleanings, each with 9-hour acid recirculating/soaking after at most 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> of acid mixing/injecting, are approved based conservatively on the MIC Program data .

The 10% Betz KI-2 solution obtained from mixing the chemicals In one of the Mixing Drums would be injected Into the heat exchangers and associated piping through the injection port, pushing the existing SSW out of the collection port into the other Mixing Drum . After completion of the acid Injection within a maximum of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />, a recirculating loop Including only one Mixing Drum would be established to recirculate the acid solution through the loop . At the end of the recirculation phase when the inboard drain valves are closed, the RHR "B'%SSIN "B" systems can be declared Operable . The cleaning process may include a period of time for soaking the tubes with the acid solution thereafter before starting the SSW Pump to rinse off the acid solution . The acid cleaning process is monitored by taking SSW samples at 3, 6, and 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> into the Recirculating Phase, and by watching closely the RHR-side pressure (1 E12N026AtB) and conductivity alarm (1E12L602A/B) to detect a tube leak. The total duration of the recirculating and soaking is limited to 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.

The recircuiatinglsoaking Is also to be deemed complete when the maximum allowable copper concentration in the SSW sample exceeds 4725 ppm. The cleaning process would be promptly terminated upon SSW system automatic initiation, detecting RHR-side pressurization and confirming an RHR Heat Exchanger tube leak, or detecting radioactivity in SSW samples by isotope analyses (by Chemistry Dept .) . In case of a tube leak with the RHR side leak tight, only a small amount of SSW could leak into the RHR side when the SSW side pressure is higher, thus raising the reading of 1E12N426A1B. The SSW would be flushed to the Suppression Pool, thus diluting the concentration of the citric acid, and then cleaned up by means of the Suppression Pool Cleanup system . One quart of 40/6 citric acid solution would be diluted to a undetectable concentration level of 10 ppb in the Suppression Pool . The concentration would be further reduced after cleaning by the Suppression Pool Cleanup system before any chance of the citric acid entering the reactor coolant system . If the SSW pressure drops below the RHR-side pressure with a tube leak, the SSW side would be contaminated by the RHR-side fluid . Significant leakage flow would be detectable by observing the levels in the two Mixing Drums and isotope analysis of the SSW sample . The waste SSW collected and the excess 10% Betz KI-2 solution remaining in the Mixing Drums at the end of the acid cleaning will be dumped or pumped back to the SSW basin via any of the nearby SSW system valves identified in the procedure (07 34-T46-BDGX-2) for acid flushing of the ESF Switchgear Room Coolers, such as 1P41F352,1P41F337, etc. after an isotope analysis of a sample has verified no radioactive contamination. The relatively low concentrations of citric acid and removed corrosion products in the acid solution remaining in the cleaning solution is not expected to cause any damage to the SSW basin area, based on previous experiences with the SSW piping flushes. In particular, only slight etching on the concrete has been noticed after the SSW piping flushing.

The tower fan blades would not be wetted. The ceramic fill material is compatible with citric acid and no adverse effects have been noticed before. The small total amounts of citric acid, Nalco penetrant, and removed corrosion products after being diluted by the vast basin volume will be harmless to the system and the SSW basin water will be allowed to be discharged to the river per a NPDES permit . A permit for storing combustible material in the work area will be obtained since the total amount would exceed the normal allowance but would not be excessive.

The acid mixinglinjecting equipment, drums, and hoses would be set up mainly on El. 999' of Auxiliary Building outside the RHR Heat Exchanger Room . All pumps, valves, fittings, drums, and hoses will be either new or previously verified to be free from radioactive contamination before use.

LI-101-01, Rev. 8; Effective Date : 8]23105

50.59 REVIEW FORM 13 References Discuss the methodology for performing LBD searches . State the location of relevant licensing document 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 . NOTE: Ensure that manual searches are performed using controlled copies of the documents. If you have any questions, contact your site Licensing department.

Keyword searches were performed with all LBDs listed In AUTONOMY selected. The hits were reviewed to verify that they are not related to the proposed acid flush of the RHR Heat Exchangers and associated piping.

Electronic search method used: Keywords:

All LBDs listed in AUTONOMY "acid flush",

"acid cleani "chemical cleaning" .

LBDs reviewed manually:

None

5. Is the validity of this Review dependent on any other change? j~ Yes No If "YES," list the required changes/submittals . The changes covered by this 50.59 Review cannot be implemented without approval of the other identified changes (e.g., license amendment request) . Establish an appropriate notification mechanism to ensure this action is completed .

LI-101-01, Rev. 8; Effec 6/23105

50.59 REVIEW FORM 8of13 B. ENVIRONMENTAL SCREENING If any of the following questions is answered "yes," an Environmental Review must be performed in accordance with NMM Procedure EV-115 and attached to this 50.59 Review. Consider both rout non-routine (emergency) discharges when answering these questions .

Will the proposed activity being evaluated:

YES Involve a land disturbance equal to or in excess of one acre (i .e., grading activities, construction of buildings, excavations, reforestation, creation or removal of ponds)?

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

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

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

o "t Increase the concentration or quantity of them ing discharged to the river, lake, or air?

6. Discharge any new or different chemicals that are currently not authorized for use by the state regulatory agency?

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

8. E] Z Modify the design or operation of the cooling tower that will change water or air flow characteristics?

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

1 t). [] Modify existing stationary fuel burning equipment (i.e., diesel fuel oil, butane, gasoline, propane, and kerosene)?'

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

12. C] Z Involve the installation or use of equipment that will result in a new or additional air emission discharge?

13. M Involve the installation or modification of a stationary or mobile tank?'

14. 11 Involve the use or storage of or chemicals that could be directly released into the environment?

15. [] 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 for guidance in answering this question .

LI-101-01, Rev . 8; Effective Date: 6123105

0 C. SECURITY PLAN SCREENING If any of the following questions is answered "yes," a Security Plan Review must be performed by the Security Department to determine actual impact to the Plan and the need for a change to the Plan.

Could the proposed activity being evaluated :

YES No

1. © Add, delete, modify, or otherwise affect Security department responsibilities (e.g .,

including fire brigade, fire watch, and confined space rescue operations)?

2. (1 Result in a breach to any security barrier(s) (e.g., HVAG ductwork, fences, doors, walls, ceilings, floors, penetrations, and ballistic barriers)?

3. [1 Cause materials or equipment to be placed or installed within the Security Isolation Zone?

4. Affect (block, move, or alter) security lighting by adding or deleting lights, structures, buildings, or temporary facilities?

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

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

7. 01 Modify or otherwise affect (block, move, or alter) installed access control equipment, intrusion detection equipment, or other security equipment?

8. C] Modify or otherwise affect primary or secondary power supplies to access control equipment, intrusion detection equipment, other security equipment, or to the Central Alarm Station or the Secondary Alarm Station?

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

1t3 Modify or otherwise affect the facility's telephone or security radio systems?

The Security Department answers the following question if one of questions C.1 through C.10 above was answered "yes ."

Is a change to the Security Plan required? [] Yes

[] No Attach to this 50.59 Review or reference below documentation for accepting a "yes" answer for any of Questions C.1 through C.10, above.

Name of Security Plan reviewer (print 1 Signature t Data L.I-101-01, Rev. 8; Effec Date: 6123105

50.59 REVIEW FORM Page 10 of 13 D. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) SCREENING Mt OTE: This section is not applicable to Grand Guff or Waterford 3 and may be removed from 50.59 Rev performed for Waterford 3 proposed activities.)

If any of the following questions is answered "YES," a 72.48 Review must be performed in accordance with NMM Procedure LI-112 and attached to this 50.59 Review.

Will the proposed activity being evaluated :

YES

1. 12 Any activity that directly impacts spent fuel cask storage or loading operations?

2. C1 13 Involve the ISFSI including the concrete pad, security fence, and lighting?

3. 0~ Involve a change to the on-site transport equipment or path from the Fuel Building to the ISFSI?

Involve a change to the design or operation of the Fuel Building fuel bridge including setpoints and limit switches?

5. V Involve a change to the Fuel Building or Control Room(s) radiation monitoring?

o fr,5 Involve a change to the Fuel Building pools including pool levels, cask pool gates, cooling water sources, and water chemistry?

7. [1 Involve a change to the Fuel Building handling equipment (e.g., bridges and cask cranes, structures, load paths, lighting, auxiliary services, etc)?

8. Ft Involve a change to the Fuel Building electrical power that could potentially impact cask loading or storage activities?

Involve a change to the Fuel Building ventilation that could potentially impact cask loading or storage activities?

10. Q OR Involve a change to the ISFSI security?

11 . M Involve a change to off-site radiological release projections from non-ISFSI sources?

12. n 0 Involve a change to spent fuel characteristics?

13. [] Redefine/change heavy load pathways?

14. 0 Involve fire and explosion protection near or in the on-site transport paths or near the ISFSI?

15 . [~ involve a change to the loading bay or supporting components power that could potentially impact cask loading or storage activities?

16. New structures near the ISFSI?

17. a 0 Modifications to any plant systems that support dry fuel storage activities?

18. E3 Z involve a change to the nitrogen supply, service air, demineralized water or borated water system in the Fuel Building?

LI-101-01, Rev . 8; Effecti Date: 6/23105

59 REVIEW FORM Page 11 of 13 IV. 50.59 EVALUATION License Amendment Date anon Does the proposed Change being evaluated represent a change to a method of evaluation Yes ONLY? If "Yes," Questions I - 7 are not applicable ; answer only Question 8. If "No," answer No all questions below .

Does the proposed Change .,

1. Result in more than a minimal increase in the frequency of occurrence of an accident El Yes previously evaluated in the FSAR? Z No BASE:

The corrosion impact of the proposed 4 % citric acid solution on all components, including weld material, within the chemical cleaning boundary for the acid flush of the RHR Heat Exchangers was evaluated . Crevice corrosion attack was not considered a concern since chemical environments associated with the cleaning process are not of a nature to create an aggressive environment for crevice corrosion mechanisms in the case of the SSW system, and the crevice corrosion requires before accelerated metal dissolution begins some incubation period that would not be available because the length of time crevices in the SSW system were to be exposed to the chemical cleaning process would be sufficiently short and areas that could be saturated by this chemical environment would be subsequently flushed . The critical components affected by an acid flush were found to be carbon steel piping, 10-30 CuNI heat exchanger tubes, Ethylene Propylene Terpolymer (EPT) elastomer seats of SSW isolation butterfly valves (I 1341FOUAIS &

I P41 F068A/B), and KEL-1181 plastic seats of the Anderson Greenwood relief valves I P41 F1 MAIM The corrosion impact on carbon steel and 70-30 CuNi by a 9-hour acid recirculating/soaking following less than 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> of mixinglinjecting of the acid solution has been determined to be insignificant and well within the corrosion allowances . The EPT elastomer is quite compatible with citric acid solution. No detailed information is available on the KEL-M plastic in the vendor manual . However, no adverse effect on these non-metallic materials has been observed from numerous 4% citric acid flushing of the ESF Switchgear Room Coolers or a previous flush of the entire SSW system using a 2.5 % citric acid solution for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A recent 24-hour soaking of the EPT seat material from a similar butterfly valve in the proposed acid flush solution also resulted in no significant visible changes and no noticeable change in its Shore "A" Hardness value (approximately 65) . Since the safety positions of these safety-related valves are "OPEN" so as to allow SSW flow under postulated accident conditions, leak-tightness is not required for their safety function . Therefore, even if unlikely, slight degradation in the EPT material should result in minor leakage after such exposure, it would not pose a safety concern . Also, even with slight leakage through the EPT packing, there would still be sufficient torque for the valve operators to open the valves were they found to be closed unexpectedly . In reality, these butterfly valves as well as the relief valves, being leak-tight and for away from the junctions with the main piping, are not expected to be affected by the cleaning solution during the short cleaning duration . Therefore, no components will be degraded by the acid flush as to increase the frequency of occurrence of an accident previously evaluated in the FSAR.

In the beginning of the proposed acid cleaning, the HX outboard isolation drain valves (1P41 F1 65A/8 & 1P41 F1 67A/13) of the affected train would be temporarily converted into inj and collection ports for the acid solution. The conversion is to replace the valve stem and other internals with a quill, which will be held in place to seat off the outlet opening and allow the acid leakage solution to go through the inlet opening . The total owe from the four inboard drain Lion valves of the affected train is to be verified w be less than 2,35 gpm to ensure that the inboard drain valves are functional before commencing any work on the outboard drain valves.

When cutting out outboard drain valve IP41FI67B to replace it like-for-like, the cut is to be within 22" from the nearest "three-dimension" support in order to maintain the piping structural integrity for all loading conditions including a seismic event. The inboard drain isolation valves (1P41F166A/B, 1P41F214A/B, 1P41F158A/B, and 1P41F164A/B) would serve the function of LI-101-01, Rev. 8; Effective Date' 6123105

Page 12 of 13 citation during the valve conversion work and the setup work connecting hoses with fittings, valves and pump to the quills. The RHR and SSW systems will remain operable as long as the inboard drain valves are closed . During the acid mixing and injecting as well as the acid recirculating/soaking, the SSW inlet and outlet isolation butterfly valves would be closed but not tag-closed so that they could open automatically in case of an initiation of the SSW system . The hoses connected to the quills installed on the outboard drain valves will be taped down on the floor in the immediate vicinity of the quills so that the weights of the whole assemblies will not be placed on the piping and would have negligible impact on the piping in case of a seismic event.

The inboard isolation valves IP41F214A/B and IP41F164At8, and the temporary injection and collection ports would be open, and the RHR and SSW systems will be declared INOP but they will remain functional by posting a dedicated Operator to close the inboard drain isolation valves in case of an initiation of the SSW Pump . The only hazard that this could present in case of an accident resulting in actuation of the SSW Pump during these periods would be a potential for SSW coming out of the collection port or even the injection port and briefly overflowing the Mixing Drums. Caution would be placed in the work package to minimize this hazard, which would not be a plant safety concern. Thus, the inboard isolation valves would serve the function of isolation during these periods of acid cleaning work and the frequencies of occurrence of all accidents previously evaluated in the FSAR would not increase as a result .

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a (Q Yes structure, system, or component important to safety previously evaluated in the FSAR? No BASIS:

The corrosion impact on limiting components within the chemical cleaning boundary, namely carbon steel piping and 70-30 CuNi tubes, by the proposed acid flush has been determined to be well within the corrosion allowances. The safety function of the isolation valves, butterfly valves or otherwise, would not be affected by the acid flush. Therefore, no structures, systems, or components important to safety within the chemical cleaning boundary would be affected by the acid flush as to increase the likelihood of occurrence of a malfunction. The inboard drain valves are about 95' from each other; therefore, a single dedicated Operator can isolate them In case of accident within a relatively short time .

3. Result in more than a minimal increase in the consequences of an accident previously _n Yes evaluated in the FSAR? No BASIS:

The Integrity of the RHR Heat Exchangers and associated piping would not be compromised by the acid flush since the corrosion effect would be well within the allowance. The safety function of all isolation valves would not be impaired by the acid flush. Therefore, no systems or components within the chemical cleaning boundary would be prevented from performing their safety function during an accident previously evaluated in the FSAR as to cause any Increase in the consequences of the accident . The dedicated Operator would be able to close the inboard drain isolation valves in case of an accident within a relatively short time so that they could perform their safety function during the accident and would not increase the consequences of the accident.

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, [l Yes system, or component important to safety previously evaluated In the FSAR? No BASIS:

The acid flush would not compromise the integrity of the SSW system boundary or degrade the heat exchanger function or affect the operation of any other safety system/component within the chemical cleaning boundary required for mitigating the consequences of an accident; therefore, it would not cause any increase in the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the FSAR .

5. Create a possibility for an accident of a different type than any previously evaluated in the © Yes FSAR? No LI-101-01, Rev. 8; Effective Date : 6123105

50 .59 REVIEW FORM Page 13 of 13 BASIS:

The impact of the proposed acid flush on all exposed components within the chemical cleaning boundary is insignificant and well within the corrosion allowances. Leaving the inboard drain valves and the injection and collection ports open during the acid cleaning work would only add the need to post a dedicated Operator to close the Inboard drain valves but would not create a possibility for an accident of a different type to occur since the inboard drain valves would serve the isolation function . Therefore, no possibility for an accident of a different type than any previously evaluated in the FSAR could be created .

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

The impact of the proposed acid flush on all exposed components within the chemical cleaning boundary is insignificant and well within the corrosion allowances . The inboard drain valves would be closed by the posted dedicated Operator in case of a postulated accident resulting in actuation of the SSW System to serve the isolation function as usual . The results of any malfunction of a structure, system, or component Important to safety previously evaluated would not be made different by the acid flush . Therefore, no possibility for a malfunction with a different result than any previously evaluated in the FSAR could be created.

7. Result in a design basis limit for a fission product barrier as described in the FSAR being Yes exceeded or altered?

BASIS:

The impact of the proposed acid flush on all exposed components within the chemical cleaning boundary is insignificant and well within the corrosion allowances . The inboard drain isolation valves would be closed within a relatively short time by the posted dedicated Operator in case of a postulated accident resulting in actuation of the SSW System, causing negligible loss in the SSW.

Therefore, the acid flush would not change any result of accidents previously analyzed In the FSAR. Hence, it could not result in a design basis limit for a fission product barrier as described in the FSAR being exceeded or altered.

8. Result in a departure from a method of evalu in the FSAR used in establishing [l Yes the design bases or in the safety analyses? No BASIS:

The proposed acid flush is to chemically clean the RHR Heat Exchangers and associated piping to be better prepared for the Eddy Current Testing and possibly improve the heat exchanger thermal performance . The heat removal capability of the RHR Heat Exchangers used in safety analyses was based on the design fouling level and would not be affected by the acid flush . Therefore, the proposed acid flush would not affect any method of evaluation described In the FSAR used in establishing the design bases or in the safety analyses.

If any of the above questions is checked "YES," obtain NRC approval prior to implementing the change by initiating a change to the Operating License In accordance with NMM Procedure LM 13 .

LI-101-01, Rev. 8; Effective Date : 6123/05

GGNS 50.59 Safety Evaluation Number SE 2005-0005-R00

50 .59 REVIEW FORM Page 1 of 12 r

I. OVERVIEW/ SIGNATURES 1

Document Reviewed : Core Operating Limits Report (COLR) Change/Rev .: LDC 2005-060 System Designator(s)/Description : J11 Description of Proposed Activity:

This evaluation addresses the Cycle 15 reload changes and operation of the Cycle 15 reload core as given in the Core Operating Limits Report (COLR)

Check the applicable review(s): (Only the sections indicated must be included in the Review.)

© EDITORIAL CHANGE of a Licensing Basis Document Section 1 -

[] SCREENING Sections I and 11 required 50 .59 EVALUATION EXEMPTION Sections 1, 11, and III required 50 .59 EVALUATION (# : 2005 - ) Sections 1, 11, and IV required a

Preparer : Guy B Spikes / !t-, ?, tea. / e o .t t At) u- S A / 1011010 S Name (print) / Signature / Company / Department / Date f ',

(

Reviewer ,

 ('t (pQ~ '~,~ t o 1 O -°O ~a Name (print) / Signatu e / C any / De ment / Dam -j -i OSRC : o t u nS' Chairman's Name (print) / Signature / Date (Required only for Programmatic Exclusion Screenings and 50 59 Evaluations)

List of Assisting/Contributing Personnel:

Name : Scope of Assistance :

J. A. Elam (Central Enaineerina BWR Fuels) Core design and neutronic input Shen G. Shue (Central Eno. BWR Fuels) Core design and neutronic input D. L. Smith Central Enaineerina BWR Fuels) Fuel mechanical input J. P. Head (Central Enalneering BWR Fuels) Core stability and hydraulic input G. W. Smith (GGNS-PSA) EOP Input LI-101-01, Rev. 8; Effective Date : 6/23/05

50 .59 REVIEW FORM Page 2 of 12 It. SCREENINGS A. Licenshna Basis Document Review

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

Operating License YES NO CHANGE # and/or SECTIONS IMPACTED Operating license 13 A041 TS NRC Orders El 91 If "YES," obtain NRC approval prior to implementing the change by initiating an LSD change in accordance with NMM LI-1 13. (See LI-I 01 for exceptions .)

LBDs controlled under 60.59 YES NO CHANGE # (if applicable) and/or SECTIONS IMPACTED FSAR ED En LDC 2W5061 TS Bans El ED Technical Requirements Manual Core Operating omits Report ED E3 LDC 2005-060 NRC Safety Evaluation Report and Cl ED supplements for the initial FSAR 1 NRC Safety Evaluations for El 0 amendments to the Operating License' F1 If "YES," perform an Exemption Review per Section III OR perform a 50 .59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change by initiating an LSD change in accordance with NMMLI-113. If obtaining NRC approval, document the LSD change in Section lLA.5. However, the change cannot be implemented until approved by the NRC. Complete Section 11.

LSDs controlled under other YES NO CHANGE # (if applicable) and/or regulations SECTIONS IMPACTED Quality Assurance Program ManUal 2 F1 ED Emergency Plan 2,3 E3 ED Fire Protection Program" Ej to (includes the Fire Hazards Analysis)

Offsite Dose Calculations If "YES," evaluate any changes in accordance with the appropriate regulation AND initiate an LSD change in accordance with NMM LI-113 .

l if"YES,"seeLl-101 No LBD change is required 2 if "YES," notify the responsible department and ensure a 50 54 evaluation is performed Attach the 50 54 evaluation 3

Changes to the Emergency Plan, Fire Protection Program, and Ciffsite Dose Calculation Manual must be approved by the OSRC in accordance with NMM OM-119 4 4 "YES," evaluate he change m accordance with the requirements a the facility's Operating license Condition or under 50 59, as appropriate LI-101-01, Rev. 8; Effective Date : 6123105

50.59 REVIEW FORM Page 3 of 12

2. Does the proposed activity involve a test or experi nt not described in the FSAR? [l Yes No If "YES," perform a 50.59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change AND initiate an LBD change in accordance with NMM LI-113, If applicable. If obtaining NRC approval, document the change in Section II.A.S. However, the change cannot be implemented until approved by the NRC. Complete Section If.

3. Basis Explain why the proposed activity does or does not impact the Operating License/Technical Specifications and/or the FSAR If the proposed activity involves a potential test or experiment not previously described m the FSAR also include an explanation Discuss other LBDs if impacted Adequate basis must be provided within the Screening such that a third-party reviewer can reach the same conclusions Simply stating that the change does not affect TS or the FSAR is not an acceptable basis This evaluation addresses the reload-related changes associated with the Cycle 15 reload and operation of the Cycle 15 reload core as given in the Core Operating Limits Report (COLR) located in the Operating License Manual (OLM) Cycle 15 has been designed for 511 Effective Full Power Days with a core consisting of 232 fresh, 244 once-burnt, 239 twice-burnt, and 85 thrice burnt ATRIUM-10 assemblies There are no TS or TS Bases changes required to operate with this new core, however, the FSAR does require updates The Cycle 15 core has been designed and analyzed for a rated thermal power of 3898 MWt Attachment 1 provides a detailed description of the Cycle 15 reload analysis and the issues considered in this evaluation Control rod behavior indicative of increased (abnormal) channel bow was observed in some control cells during Cycle 14 operation (CR-GGN-2005-3287) The Cycle 15 reload safety analyses includes abnormal channel bow that bounds the bow measured during RF14 fuel channel inspections and the expected bowing during Cycle 15 operation The channels of seventeen ATRIUM-10 bundles expected to experience the worst bowing during Cycle 15 have been replaced with fresh (unirradiated) channels During RF14, two Cycle 15 thrice-burnt ATRIUM-10 bundles were discharged and replaced with two similar thrice-burnt bundles expected to experience less bowing during Cycle 15 This change to the original Cycle 15 core reference loading pattern is also considered in this evaluation Operating License/Technical Specifications (OUTS)

The current MCPR Safety Limit has been shown to be applicable to the Cycle 15 core As such, Tech Spec 2 1 1 2 does not need to be revised There are no other Tech Specs, LCO's, TS Bases, surveillances or other controls in the GGNS OUTS affected by the Cycle 15 reload TRM The Cycle 15 reload does not affect any TRM requirements As such, the TRM is not impacted by the Cycle 15 reload evaluation FSAR The Cycle 15 core will contain fuel types currently described m the FSAR However, the core characteristics and response will be different than that currently described in the FSAR As such, Cycle 15 analyses have been performed for the new core and the FSAR will be updated to reflect these analyses and operation of the Cycle 15 core COLR Cycle 15 operation will require new core operating limits and the Core Operating Limits Report has been revised to include these new limits These limits include flow-, power-, and exposure-dependent LHGR, MAPLHGR, and MCPR limits Test or Experiment The Cycle 15 reload does not involve any tests or experiments There are no NRC orders applicable to the Cycle 15 reload campaign The evaluation does not affect the FHA, ODCM, QAPM, E-Plan, or NRC SERs LI-101-01, Rev. 8; Effective Date: 6123105

50.59 REVIEW FORM Page 4 of 12

4. References Discuss the methodology for performing LBD searches State the location of relevant licensing document 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 NOTE: Ensure that manual searches are performed using controlled copies of the documents. If you have any questions, contact your site Licensing department.

Electronic search method used Keywords .

GGNS Autonomy LBDs OLM, FSAR, COLR, Fuel, reload, channel, COLR TS Bases, TRM LBOs reviewed manually COLR

5. Is the validity of this Review dependent on any other change?

If "YES," list the required changes/submittals. The changes covered by this 50.59 Yes Review cannot be implemented without approval of the other identified changes (e.g .,

No license amendment request). Establish an appropriate notification mechanism to ensure this action is completed .

An acceptable final core loading For a final core loading not exactly as provided m -

JLR 05 131, an evaluation of the as-loaded core must be performed to ensure that the Cycle 15 reload analyses continues to be acceptable Core loading verification is accomplished IAW procedure 17-S-02-108, Core Loading Verification LI-101-01, Rev . 8; Effective Date: 6123105

50.59 REVIEW FORM Page 5 of 12 B. ENVIRONMENTAL SCREENING If any of the following questions is answered "yes," an Environmental Review must be performed in accordance with NMM Procedure EV-115 and attached to this 50.58 Review. Consider both routine and non-routine (emergency) discharges when answering these questions.

Will the proposed activity being evaluated :

YES NO 1 Involve a land disturbance equal to or in excess of one acre (i e , grading activities, construction of buildings, excavations, reforestation, creation or removal of ponds)?

2 M Involve any land disturbance of undisturbed land areas (i e, grading activities, construction, excavations, reforestation, creating, or removing ponds)?

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

4 [3 0 increase the amount of thermal heat being discharged to the river or lake?

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

6. Discharge any new or different chemicals that are currently not authorized for use by the state regulatory agency?

7 C3 Change the design or operation of the intake or discharge structures?

S Modify the design or operation of the cooling tower that will change water or air flow characteristics?

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

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

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

Involve the installation or use of equipment that will result in a new or additional air emission discharge?

13 [] Involve the installation or modification of a stationary or mobile tank?'

14 (~ Involve the use or storage of oils or chemicals that could be directly released into the environment?

15 (~ 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 for guidance m answering this question LI-101-01, Rev . 8; Effective Date: 6123105

50.59 REVIEW FORM Page 6 of 12 C. SECURITY PLAN SCREENING If any of the following questions is answered "yes," a Security Plan Review must be performed by the Security Department to determine actual impact to the Plan and the need for a change to the Plan.

Could the proposed activity being evaluated :

YES NO 1 [] Add, delete, modify, or otherwise affect Security department responsibilities (e g ,

including fire brigade, fire watch, and confined space rescue operations)?

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

3 [] Cause materials or equipment to be placed or installed within the Security Isolation Zone?

4 [] 2 Affect (block, move, or alter) security lighting by adding or deleting lights, structures, buildings, or temporary facilities?

11 ED Modify or otherwise affect the intrusion detection systems (e g , E-fields, microwave, fiber optics)?

a OR Modify or otherwise affect the operation or field of view of the security cameras?

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

8 [] Modify or otherwise affect primary or secondary power supplies to access control equipment, intrusion detection equipment, other security equipment, or to the Central Alarm Station or the Secondary Alarm Station?

[] Modify or otherwise affect the facility's security-related signage or land vehicle barriers, including access roadways?

10 [j Modify or otherwise affect the facility's telephone or security radio systems?

The Security Department answers the following question if one of questions C.1 through C.10 above was answered "yes."

Is a change to the Security Plan required? El Yes No Attach to this 50.59 Review or reference below documentation for accepting a "yes" answer for any of Questions C.1 through C.10, above.

Name of Security Plan reviewer (print 1 Signature J Data LI-101-01, Rev . 8 ; Effective Date: 6123105

50.59 REVIEW FORM Page 7 of 12 D. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI} SCREENING (NOTE. This section is not applicable to Grand Gulf or Waterford 3 and may be removed from 50 59 Reviews performed for Waterford 3 proposed activities )

If any of the following questions is answered "YES," a 72.48 Review must be performed in accordance with NMM Procedure LI-112 and attached to this 50.59 Review.

Will the proposed activity being evaluated:

YES NO 1 n Any activity that directly impacts spent fuel cask storage or loading operations?

2 0 Involve the ISFSI including the concrete pad, security fence, and lighting?

3 Involve a change to the on-site transport equipment or path from the Fuel Building to the ISFSI?

4 Involve a change to the design or operation of the Fuel Building fuel bridge including setpoints and limit switches?

5 1O" Involve a change to the Fuel Building or Control Room(s) radiation monitoring?

6 O Involve a change to the Fuel Building pools including pool levels, cask pool gates, cooling water sources, and water chemistry?

Involve a change to the Fuel Building handling equipment (e g , bridges and cask cranes, structures, load paths, lighting, auxiliary services, etc)?

8 Involve a change to the Fuel Building electrical power that could potentially impact cask loading or storage activities?

a Involve a change to the Fuel Building ventilation that could potentially impact cask loading or storage activities?

10 Involve a change to the ISFSI security?

11 Involve a change to off-site radiological release projections from non-ISFSI sources?

12 (~ Involve a change to spent fuel characteristics?

13 0 Redefine/change heavy load pathways?

14 0 Involve fire and explosion protection near or in the on-site transport paths or near the ISFSI?

15 Involve a change to the loading bay or supporting components power that could potentially impact cask loading or storage activities?

16 [~ New structures near the ISFSI?

17 [~ Modifications to any plant systems that support dry fuel storage activities?

18 [~ Involve a change to the nitrogen supply, service air, demineralized water or borated water system in the Fuel Building?

LI-101-01, Rev . 8; Effective Date: 6123105

50.59 REVIEW FORM Page 8 of 12 IV. 50 .59 EVALUATION License Amendment Determination Does the proposed Change being evaluated represent a change to a method of evaluation [l Yes ONLY? If "Yes," Questions 1- 7 are not applicable; answer only Question 8. If "No," answer No all questions below.

Does the proposed Change 1 Result in more than a minimal increase in the frequency of occurrence of an accident [] Yes previously evaluated m the FSAR? No BASIS :

The Cycle 15 core loading and cycle operation will not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the FSAR The precursors to these events are independent of the core design and the frequency classifications reported in FSAR Chapter 15 are unaffected by the core parameters The following considerations support this conclusion Mechanical The ATRIUM-10 mechanical design has been reviewed for use at Grand Gulf No unusual failure modes or increased failure frequency have been identified for this fuel design This is the fourth reload at GGNS with ATRIUM-10 fuel and this fuel design has accumulated operational experience at GGNS and other plants with no significant problems The Cycle 15 bundles will operate within the power history assumptions in the fuel mechanical analyses and will result in exposures within the analyzed burnup limits of the ATRIUM-10 mechanical design, including those bundles that will be irradiated for a fourth cycle The re-channeled fuel bundles continue to satisfy all mechanical design criteria Although an increased channel bow condition can result in increased friction between the control blade and its corresponding fuel assemblies, control rod settle and insertion testing (EPI 04 03-C11-7) will continue to be performed during Cycle 15 to ensure that the increased axial friction loads on the channel and fuel assembly load chain remain below acceptable limas Nuclear The neutronic characteristics of the Cycle 15 ATRIUM-10 core design have been considered in the safety analysis Adequate shutdown margin has been predicted by analysis and will be confirmed during startup tests In addition, the hold-down capability of the standby liquid control system and the subcriticality of Cycle 15 fuel in the spent fuel storage racks have been confirmed Therefore, the probability of inadvertent criticality has not been increased by the introduction of the Cycle 15 reload fuel The neutronic characteristics of the ATRIUM-10 bundles are not affected by channel replacement or by abnormal channel bow Thermal-Hydraulic Cycle 15 is an all ATRIUM-10 core Therefore, considerations of the thermal-hydraulic compatibility of the ATRIUM-10 with co-resident fuel types do not apply Analyses have been performed to demonstrate that Cycle 15 meets all Enhanced-1A stability performance criteria without changes to the EIA hardware or power-flow map region boundaries The thermal-hydraulic performance of the ATRIUM-10 bundles is not affected by channel replacement or by abnormal channel bow Therefore, the probability of thermal-hydraulic instabilities has not increased Analyzed Events The probability of the occurrence of anticipated operational events is not dependent on the core configuration No changes to the plant design are required for the Cycle 15 core The Cycle 15 core loading will not affect the precursors to any of the Chapter 15 events The probability of an analyzed event therefore has not increased As described in FSAR Section 15A 6 5 3, the Control Rod Drop Accident (CRDA) results from a failure of the control rod-to-drive mechanism coupling after the control rod becomes stuck in its fully LI-101-01, Rev . 8; Effective Date: 6/23105

50.59 REVIEW FORM Page 9 of 12 inserted position Although an increased channel bow condition can result in increased friction between the control blade and its corresponding fuel assemblies, analyses have shown that there would not be sufficient friction to result in a mechanical failure of the coupling Additionally, the control rod drive mechanism would not produce enough force to result in a mechanical failure of the coupling even if the channel bow was so severe that the assemblies would preclude blade movement As such, channel bow is not considered a precursor to the CRDA, and any increased bow associated with the high exposure ATRIUM-10 bundles would not increase the probability of this event On these bases, the probability of occurrence of accidents previously identified in the FSAR is not increased for the Cycle 15 core with increased channel bow 2 Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a 0 Yes structure, system, or component important to safety previously evaluated in the FSAR? No BASIS :

No plant modifications are required to accommodate the new all ATRIUM-10 core design The mechanical design and neutronic, and thermal-hydraulic characteristics of the ATRIUM-10 fuel bundles have been shown to be unaffected by channel replacement The only additional loads placed on plant equipment would be due to increased friction between the control blades and excessively bowed ATRIUM-10 bundles This probability has been reduced by re-channeling 17 ATRIUM-10 fuel bundles and replacing two others considered most susceptible to abnormal bow Based on previous experience with bowed fuel at GGNS and other BWR-6's, increased control blade friction can result in increased control rod settle times but is not expected to significantly impact scram times Technical Specification scram time testing and control rod settle and insertion testing (EPI 04-1-03-C11-7) will continue to be performed during Cycle 15 These actions would identify any potential scram time or other impacts and such that appropriate corrective actions are taken These actions will ensure that the increased control blade friction loads are not sufficient to cause any failures associated with the control blades or the control blade drive system, the fuel assembly load chain, or the vessel internals A conservative vessel overpressurization analysis has been performed, which shows that the vessel pressure limit is not exceeded The precursors to any malfunction of equipment important to safety are not affected by this the Cycle 15 reload core Therefore, there is not more than a minimal increase in the likelihood of an occurrence of a malfunction of a SSC important to safety previously evaluated in the FSAR 3 Result in more than a minimal increase in the consequences of an accident previously [] Yes evaluated in the FSAR? 14 No BASIS :

As reported in Attachment 1, the acceptance criteria reported in FSAR Section 15 0 3 1 and the Technical Specifications are satisfied for each event classification Core operating limits have been developed to ensure that moderate frequency events do not violate the MCPR safety limit or fuel cladding strain limits The consequences of infrequent events have been shown to meet the appropriate acceptance criteria while the individual acceptance criteria for the limiting faults have been demonstrated to be satisfied As such, the consequences of infrequent events and limiting faults described in the FSAR are unchanged for the Cycle 15 reload core The following considerations support these conclusions Moderate Frequency Events The Cycle 15 core operating limits have been developed with NRC-approved methodologies such that the MCPR safety limit and the fuel cladding strain limit will not be violated by any analyzed moderate frequency transient initiated from any statepoint available to GGNS As such, no fuel failures are expected to result from any moderate frequency event These analyses considered GGNS-specific operational modes such as MEOD, SLO, FHOOS, and EOC-RPT inoperable These core operating limits consist of MCPR, MAPLHGR and LHGR curves that are functions of flow, power, and exposure These limits consider conservative channel bow assumptions that bound the LI-101-01, Rev . 8; Effective Date: 6123105

0.59 REVIEW FOR Page 10 of 12 current measured bow data and the expected increased bow associated with the highly exposed ATRIUM-10 fuel These core operating limits will be incorporated into the core monitoring system Infrequent Events The consequences of the limiting infrequent events have been evaluated and shown to meet their respective acceptance criteria These events include the pressure regulator failure downscale, misplaced (i e, misonented and mislocated) bundle and single loop operation pump seizure accidents Radiological analyses using the alternative source term (AST) have been performed to ensure that these events will not result in an increase m offsite or control room doses or doses greater than their respective acceptance criteria These evaluations include conservative channel bow assumptions that bound the current measured bow data and the expected increased bow associated with the highly exposed ATRIUM-10 fuel Limiting Faults The limiting faults at GGNS include the fuel handling accident, the control rod drop accident, and the design basis LOCH The radiological analyses for these events have been developed as part of the GGNS AST effort and bound the Cycle 15 core parameters For the LOCA, MAPLHGR operating limits and single-loop multipliers have been developed for the Cycle 15 core configuration such that the requirements of 10CFR50 46 are satisfied The containment response for the Cycle 15 core was found to be bounded by previous cycles as is the hydrogen analysis The seismicILOCA response of the Cycle 15 core has been confirmed to be acceptable The Cycle 15 core design results in minor changes to three EP parameters (Mclad, Mfuei, Fafl-18), however, the existing EP's remain applicable to Cycle 15 Therefore, the proposed change does not result in more than a minimal increase in the 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 structure, F1 Yes system, or component important to safety previously evaluated in the FSAR? No BASIS :

The Cycle 15 ATRIUM-10 reload fuel design has been shown to be compatible with the co-resident ATRIUM-10 fuel inserted in previous cycles Channel replacement has been shown to have no affect on the ATRIUM-10 fuel bundle envelope or mechanical design The malfunctions of key plant components are analyzed as part of the reload process with the results reported in various sections of the FSAR The consequences of these malfunctions have been shown to remain unchanged for Cycle 15 operation Therefore, Cycle 15 operation will not result in more than a minimal increase m the consequences of a malfunction of a structure, system, or component important to safety previously evaluated in the FSAR 5 Create a possibility for an accident of a different type than any previously evaluated in the El Yes FSAR No BASIS :

The Cycle 15 ATRIUM-10 reload fuel is similar to and compatible with the ATRIUM-10 fuel that was inserted in previous cycles The details of this design have been specifically considered in the safety analysis and the core monitoring system Channel replacement has been shown to have no affect on the ATRIUM-10 fuel bundle envelope or mechanical design No plant modifications are required to accommodate the new core design or Cycle 15 operation The GGNS Cycle 15 fuel has been approved for the Cycle 15 reactor chemistry conditions The GGNS operational parameters (water chemistry requirements, spectral-shift core designs, and MEOD rod-lines) have been reviewed and are not expected to result in unusual crud buildup like that observed on the high-power GE11 bundles at River Bend Inspection of a high-power, once-burnt representative fuel bundle during GGNS RF10 has confirmed that the high-power GGNS Cycle 10 fuel bundles have no unusual crud buildup LIA01-01, Rev . 8; Effective Date : 6/23/05

50.59 REVIEW FORM Page 11 of 12 Therefore, Cycle 15 operation will not create a possibility for an accident of a different type than any previously evaluated in the FSAR 6 Create a possibility for a malfunction of a structure, system, or component important to safety El Yes with a different result than any previously evaluated in the FSAR No BASIS :

The Cycle 15 ATRIUM-10 reload fuel design has been shown to be mechanically, neutroncally, and thermal-hydraulically compatible with the co-resident ATRIUM-10 fuel Cycle 15 is an all ATRIUM-10core As such, the reload fuel will not introduce any adverse flow distribution effects The mechanical design and neutronic, and thermal-hydraulic characteristics of the ATRIUM-10 fuel bundles have been shown to be unaffected by channel replacement No plant modifications are required to accommodate the new core design and no additional loads will be imposed on any existing equipment The ATRIUM-10 bundles provide sufficient clearance for proper control blade operation and allow sufficient bypass flow in the bypass region to provide adequate cooling for control blades and in-core detectors There are no special operational considerations associated with the Cycle 15 core other than those associated with the increased bow condition Control rod settle and insertion testing (EPI 04-1-03-Cl 1-7) will continue to be performed during Cycle 15 to ensure that the increased control blade fraction is not sufficient to cause any failures associated with the control blades or the control blade drive system, the fuel assembly load chain, or the vessel internals Therefore, Cycle 15 operation will not create the possibility for a malfunction of an SSC 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 being n Yes exceeded or altered? 04 No BASIS :

Mechanical analyses have been performed to ensure that all fuel m the Cycle 15 core meet the mechanical design limits for steady-state operation as well as transient conditions including fatigue damage, creep collapse, corrosion, fuel rod internal pressure, rod bow, internal pressure, etc The re-channeled ATRIUM-10 bundles have been shown to meet the applicable mechanical design limits for steady-state and transient operation Additionally, no Cycle 15 fuel will exceed the applicable burn-up limits Core operating limits have been developed using N RC approved codes in order to ensure that the Cycle 15 fuel will not exceed the MCPR safety limits for steady-state operation and anticipated operation occurrences Similarly, operating limits have been developed to ensure that the Cycle 15 fuel will not exceed the 1% cladding strain limit or experience core-wide fuel melt during steady-state operation or AOO's Although some vessel blowdown to the suppression pool may be experienced during some AOO's, which would increase the suppression pool temperature, the bulk containment pressure increase is negligible and would not exceed the design limit As described in Attachment 1, a bounding pressurization event with a failure of the direct scram has been analyzed for Cycle 15 to ensure compliance with ASME code requirements This analysis indicates that the vessel pressure safety limit is not exceeded for Cycle 15 A design basis limit for the peak fuel enthalpy of 280 cal/gm has been established for the control rod drop accident (CRDA) to preclude significant fuel cladding failure such that core geometry and cooling may be impacted The CRDA has been evaluated for Cycle 15 This evaluation considers all potential withdrawal sequences and 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 280 cal/gm, there is no impact on the vessel or containment pressures As such their respective limits are not exceeded LIA01-01, Rev. 8; Effective Date: 6123105

50.59 REVIEW FORM Page 12 of 12 10CFR50 46 provides limits associated with the ECCS performance analysts (LOCA analysis) Two such limits are Peak Clad Temperature (PCT) and local clad oxidation Although these limits are not subject to 10CFR50 59, they are discussed in this evaluation for completeness Grand Gulf specific analyses have been performed for ATRIUM-10 fuel m accordance with 10CFR50 46 These analyses, which are applicable to Cycle 15, show that the PCT and local oxidation are well below the limits set forth in 10CFR50 46 These analyses also show that the core-wide metal water reaction, which is used to evaluate compliance with the containment design limit, is less than the 10CFR50 46 limit The remainder of the existing containment analysis associated with LOCA events is applicable to Cycle 15 as described in Attachment 1 As such, the containment pressure design limit will not be exceeded in Cycle 15 An ATWS evaluation has also been performed for Cycle 15 As described in Attachment 1, the resulting vessel pressure remains below the ASME emergency vessel pressure limit of 1500 prig and the temperature response used m the existing ATWS containment analysis is applicable to Cycle 15 Thus, the containment pressure design limit will not be exceeded for the ATWS event Additional evaluations have been performed for Cycle 15 including Appendix R (Fire Protection),

hydrogen analyses, and SBO as described in Attachment 1 These evaluations show that the existing evaluations are applicable to Cycle 15 and that their respective limits are not exceeded Therefore, Cycle 15 operation will not result in a design basis limit for a fission product 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 establishing Q Yes the design bases or in the safety analyses? No BASIS :

The reload analyses performed by the fuel vendor utilized NRC approved methods as listed in Technical Specification 5 6 5 and described throughout the FSAR These methods are consistent with those used for Cycle 15 As described in Attachment 1, uncertainty applied in the Safety Limit calculation associated with each of the equipment out of service combinations was calculated in accordance with Framatome-ANP's NRC approved methodology The pellet exposure based LHGR limit (PEBLL) was developed in accordance with Framatome-ANP's NRC approved methodology for analyzing the Fuel Design Limit The abnormal channel bow data assumed in the safety analyses is within the bow database of Framatome-ANP's approved methodology Framatome-ANP recently revised the methodology used to calculate the fuel channel stresses due to channel wall differential pressure This revised methodology, which is applied to all channels in the GGNS Cycle 15 core, was recently approved by the NRC All remaining GGNS evaluations currently described in the FSAR have been shown to be applicable to Cycle 15 As such, no new methods were used Finally, the GGNS EP calculation has been updated to consider the minor changes to two fuel parameters This revision did not incorporate any new or different methods Therefore, Cycle 15 operation will not result in a departure from a method of evaluation described in the FSAR used m establishing the design bases or in the safety analyses LI-101-01, Rev . 8; Effective Date: 6123105

GGNS 50.59 Safety Evaluation Number SE 2005-0006-R00

50.59 REVIEW FORM Page 1 of 10 I. OVERVIEW 1 SIGNATURES Facility :

Document Reviewed : Calculation XC-Q1 P53-05011 Change/Rev.: 0 Calculation XC-Q1 M46-04004 Change/Rev.: 1 System Designator(s)/Description : M46, P52, P53 Description of Proposed Activity:

Calculation XC-Q1 P53-05011 determines the offsite and control room doses associated with secondary containment bypass leakage through the instrument air and service air piping This analysis was necessary considering the potential post-accident unavailability of the active venting systems for these lines as described in CR-GGN-2005-02334 Calculation XC-Q1 M46-04004 determines the offsite and control room doses associated with water leakage through the fuel transfer tube door of the Horizontal Fuel Transfer System (HFTS) This calculation was necessary since this leakage path is not currently considered in the LOCA dose analysis These leak paths result m very small increases to the LOCA doses at all locations. The proposed change will therefore add the radiological impacts of secondary containment bypass through the service and instrument air piping and water leakage through the fuel transfer tube door to the current doses associated with the Loss of Coolant Accident (LOCA)

Check the applicable review(s): (Only the sections indicated must be included in the Review.)

EDITORIAL CHANGE of a Licensing Basis Document Section I

© SCREENING Sections I and II required

[] 50.59 EVALUATION EXEMPTION Sections I, 11, and III required "Z1' t 50 .59 EVALUATION (#: ,~ J' d0a~:_. I dIJ r

'-'N if "ThT r .al_ .r" Sections 1, 11, and IV required Reviewer: V, llk~, e, L.OT-Name (print) I Si atu  ! Company I OSRC : "s Chairman's Name (print) / Signature t Date (Required only for Programmatic Exclusion Screenings and 50 59 Evaluations )

LIA01-01, Rev. 8 ; Effective Date : 6/23105

50.59 REVIEW FORM Page 2 of 10

11. SCREENINGS A. Licensing Basis Document Review

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

Operating License YES NO CHANGE # and/or SECTIONS IMPACTED Operating License F1 TS 11 0141 NRC Orders 0 00 If "YES," obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMM LI-113. (See LI-101 for exceptions.)

LBDs controlled under 50.59 YES NO CHANGE # (if applicable) and/or SECTIONS IMPACTED FSAR Sections 6 2 and 15 6 5 TS Bases '1 F Section 3 6 4 2 Technical Requirements Manual ISO Core Operating Limits Report 0 NRC Safety Evaluation Report and supplements for the initial FSAR' NRC Safety Evaluations for amendments to the Operating License' If YES,  perform an Exemption Review per Section 111 - OR perform a 50.55 Evaluation pe Section IV OR obtain NRC approval prior to implementing the change by Initiating an LBD change in accordance with NMM LI-113 . If obtaining NRC approval, document the LBD change in Section II.A.S. However, the change cannot be implemented until approved by the NRC . Complete Section II .

LBDs controlled under other YES NO CHANGE # (if applicable) and/or regulations SECTIONS IMPACTED Quality Assurance Program Manual2 [7 3

Emergency Plane' Fire Protection Program3 ' 4 © I'S (includes the Fire Hazards Analysis)

Offsite Dose Calculations Manual3' 4 If "YES," evaluate any changes in accordance with the appropriate regulation AND initiate an LBD change in accordance with NMM LI-113.

' lf "YES," see t-1-101 No LBD change is required 2 If "YES," notify the responsible department and ensure a 50 54 evaluation is performed Attach the 50 54 evaluation 3 Changes to the Emergency Plan, Fire Protection Program, and Offsde Dose Calculation Manual must be approved by the OSRC in accordance with NMM OM-119 If "YES," evaluate the change in accordance with the requirements of the facility's Operating License Condition or under 50 59, as appropriate LI-101-01, Rev . 8; Effective Date: 6/23/05

50.59 REVIEW FORM Page 3 of 10

2. Does the proposed activity involve a test or experiment not described in the FSAR? El Yes No If "YES," perform a 50.59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change AND initiate an LBD change in accordance with NMM LI-113, if applicable . If obtaining NRC approval, document the change in Section II.A.S. However, the change cannot be implemented until approved by the NRC . Complete Section II.

3. Basis Explain why the proposed activity does or does not impact the Operating LicenseRechnical Specifications and/or the FSAR If the proposed activity involves a potential test or experiment not previously described in the FSAR also include an explanation Discuss other LBDs if impacted Adequate basis must be provided within the Screening such that a third-party reviewer can reach the same conclusions Simply stating that the change does not affect TS or the FSAR is not an acceptable basis.

Tech Specs/Operating License The current GGNS Tech Specs and Operating License are inputs into these dose analyses No changes were identified or proposed by these analyses FSAR The LOCA dose analysis is reported in SAR 15 6 5 Several changes to this section have been identified as noted m Section 11 of this 50 59 review LDC 2005-065 makes the applicable changes Test or Experiment not Described in the SAR These calculations only quantify the radiological impact of certain post-LOCA leakage paths This calculation does not call for any action in the plant or changes to plant procedures, other than limiting the LLRT leakage values to the current admen limits applied m the radiological analysis (as described in detail in Section tl A 5)

4. References Discuss the methodology for performing LBD searches State the location of relevant licensing document information and explain the scope of the review such as electronic search crrtena used (e g , key words) or the general extent of manual searches NOTE: Ensure that manual searches are performed using controlled copies of the documents . If you have any questions, contact your site Licensing department.

Electronic search method used Keywords Tech Specs, Operating License, FSAR, "LOCH Dose". "LOCA Radiological"; "bypass COLR, ODCM, Emergency Plan, SER leakage". "secondary containment bypass" LBDs reviewed manually SAR 15.6.5

5. Is the validity of this Review dependent on any other change? 0163 Yes 0 No If "YES," list the required changes/submittals . The changes covered by this 50.59 Review cannot be implemented without approval of the other identified changes (e.g., license amendment request) . Establish an appropriate notification mechanism to ensure this action is completed .

These calculations apply containment penetration leak rates that are based on the LLRT administrative limits reported in SEP-APJ-001 Although the current (past-RF14) LLRT results for these penetrations have been confirmed to be well below these adman limits, Section 3 2 of Appendix C to SEP-APJ-001 allows GGNS the flexibility to exceed the admen limit if the total Type B and C leak rates do not exceed their respective allowable limits The HFTS leakage rate is not included in the Type A, B and C leak rates CR-GGN-2005-02334, CA#7 and WT-GGN-2005-0000, CA#818 have been issued to Engineering Programs to ensure that the adman limits for Penetrations 4, 41, 42, and 70 are not exceeded without supporting dose evaluations since they are direct inputs into the safety analysis LI-101-01, Rev . 8; Effective Date: 6/23/05

50.59 REVIEW FORM Page 4 of 10 B. ENVIRONMENTAL SCREENING If any of the following questions is answered "yes," an Environmental Review must be performed in accordance with NMM Procedure EV-115 and attached to this 50.59 Review. Consider both routine and non-routine (emergency) discharges when answering these questions.

Will the proposed activity being evaluated :

YES 1 Involve a land disturbance equal to or in excess of one acre (i e, grading activities, construction of buildings, excavations, reforestation, creation or removal of ponds)?

2 © Involve any land disturbance of undisturbed land areas (i e, grading activities, construction, excavations, reforestation, creating, or removing ponds)?

3 E] involve dredging activities in a lake, river, pond, ditch, or stream?

4 C] Increase the amount of thermal heat being discharged to the river or lake?

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

6 Q Discharge any new or different chemicals that are currently not authorized for use by the state regulatory agency?

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

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

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

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

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

12 involve the installation or use of equipment that will result in a new or additional air emission discharge?

100 Involve the installation or modification of a stationary or mobile tank?'

14 M involve the use or storage of oils or chemicals that could be directly released into the environment?

OR 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 for guidance in answering this question LI-101-01, Rev . 8; Effective Date: 8123105

50 .59 REVIEW FORM Page 5 of 10 C. SECURITY PLAN SCREENING If any of the following questions is answered "yes," a Security Plan Review must be performed by the Security Department to determine actual Impact to the Plan and the need for a change to the Plan .

Could the proposed activity being evaluated:

1 (~ Add, delete, modify, or otherwise affect Security department responsibilities (e g ,

including fire brigade, fire watch, and confined space rescue operations)?

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

3 02 Cause materials or equipment to be placed or installed within the Security Isolation Zone?

4 [~ Affect (block, move, or alter) security lighting by adding or deleting lights, structures, buildings, or temporary facilities?

1:" Modify or otherwise affect the intrusion detection systems (e g, E-fields, microwave, fiber optics)?

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

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

8 0.3 Modify or otherwise affect primary or secondary power supplies to access control equipment, intrusion detection equipment, other security equipment, or to the Central Alarm Station or the Secondary Alarm Station?

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

10 C] 14 Modify or otherwise affect the facility's telephone or security radio systems?

The Security Department answers the following question if one of questions C.1 through C.10 above was answered "yes."

Is a change to the Security Plan required? 0 Yes No Attach to this 50.59 Review or reference below documentation for accepting a "yes" answer for any of Questions C.1 through C.10, above.

Name of Security Plan reviewer (print I Signature 1 Data LI-101-01, Rev. 8; Effective Date : 6/23/05

50.59 REVIEW FORM Page 6 of 10 D. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) SCREENING (NOTE: This section is not applicable to Grand Gulf or Waterford 3 and may be removed from 50.59 Reviews performed for Waterford 3 proposed activities.)

If any of the following questions is answered "YES," a 72.48 Review must be performed in accordance with NMM Procedure LI-112 and attached to this 50.59 Review.

Will the proposed activity being evaluated :

1 n Any activity that directly impacts spent fuel cask storage or loading operations?

2 © Involve the ISFSI including the concrete pad, security fence, and lighting?

3 n Involve a change to the on-site transport equipment or path from the Fuel Budding to the ISFSI?

Involve a change to the design or operation of the Fuel Budding fuel bridge including setpoints and limit switches?

5 M Involve a change to the Fuel Budding or Control Room(s) radiation monitoring?

6 © Involve a change to the Fuel Budding pools including pool levels, cask pool gates, cooling water sources, and water chemistry?

Involve a change to the Fuel Budding handling equipment (e g , bridges and cask cranes, structures, load paths, lighting, auxiliary services, etc)?

8 (] Involve a change to the Fuel Budding electrical power that could potentially impact cask loading or storage activities?

9 FOR Involve a change to the Fuel Budding ventilation that could potentially impact cask loading or storage activities?

10 [] Involve a change to the ISFSI security?

11 Involve a change to off-site radiological release projections from non-ISFSI sources?

12 Q Involve a change to spent fuel characteristics?

13 © Redefinetchange heavy load pathways?

14 F,,4 Involve fire and explosion protection near or in the on-site transport paths or near the ISFSI?

15 F Involve a change to the loading bay or supporting components power that could potentially impact cask loading or storage activities?

16 M New structures near the ISFSI?

17 [] 024 Modifications to any plant systems that support dry fuel storage activities?

18 [l involve a change to the nitrogen supply, service air, demineralized water or borated water system in the Fuel Building?

t-1-901-01, Rev. 8; Effe e Date: 6123105

50.59 REVIEW FORM Page 7 of 10 Iii. 50.59 EVALUATION EXEMPTION A. Check the applicable box below. If a box is checked, clearly document the basis in Section 111.13, below. If none of the boxes are appropriate, perform a 50.59 Evaluation in accordance with Section IV. Provide supporting documentation or references as appropriate.

C] The proposed activity meets all of the following criteria regarding design function*

The proposed activity does not adversely affect the design function of an SSC as described in the FSAR, AND The proposed activity does not adversely affect a method of performing or controlling a design function of an SSC as described in the FSAR, AND The proposed activity does not adversely affect a method of evaluation that demonstrates intended design function(s) of an SSC described in the FSAR will be accomplished An approved, valid 50 59 Review(s) covering associated aspects of the proposed activity already exists Reference 50 59 Evaluation # (if applicable) or attach documentation Verify the previous 50 59 Review remains valid .

a The NRC has approved the proposed activity or portions thereof Reference B. Basis Provide a clear, concise basis for determining the proposed activity may be exempted such that a third-party can reach the same conclusions LI-901-01, Rev . 8; Effective Date: 6123105

50.59 REVIEW FORM Page 8 of 10 N. 50.59 EVALUATION License Amendment Determination Does the proposed Change being evaluated represent a change to a method of evaluation [] Yes ONLY? If "Yes," Questions 1- 7 are not applicable ; answer only Question 8. If "No," answer No all questions below.

Does the proposed Change 1 Result in more than a minimal increase in the frequency of occurrence of an accident [] Yes previously evaluated in the FSAR? No BASIS The proposed change does not physically modify any structure, system, or component (SSC) The proposed change therefore does not affect any accident initiators . Deleting the credit for the instrument air system venting in the dose analysis does not affect the overall system performance or reliability and cannot change the likelihood of a loss of instrument air event in SAR 15 2 10 2 Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a © Yes structure, system, or component important to safety previously evaluated in the FSAR? No BASIS The proposed change only updates the LOCA dose analysis and makes no physical modifications to the plant such that important-to-safety SSCs will not be impacted by the proposed change This change does slightly increase the source term release into the Auxiliary Budding with the addition of the HFTS leakage rate The environmental qualification analyses are not impacted by the proposed change If the service air and instrument air lines become a path for secondary containment bypass leakage due to the failure of the active vent function and continued integrity of the piping runs, the source term inventory in the Auxiliary Building would actually decrease due to the bypass of these source terms For the HFTS leakage, the primary release into the Auxiliary Building is from noble gases that evolve from the decay of dissolved iodine in the spent fuel pool Per Calculation 5 6 7-N, these airborne source terms are neglected in the general area dose rate evaluations Also, as noble gases, these source terms would not be removed by the SGTS and would not contribute to doses in the SGTS room The small amount of iodine released from the pool and collected by the SGTS filter tram is negligible compared to the overall LOCA iodine source term currently modeled on the filter train LI-101-01, Rev . 8; Effective Date: 6123105

50.59 REVIEW FORM Page 9 of 10 3 Result m more than a minimal increase in the consequences of an accident previously [] Yes evaluated in the FSAR? No BASIS The only accident that is affected by these leakage paths is the LOCA dose analysis, which is documented in FSAR Section 15 6 5 per Calculation XC-Q1111-98017, Rev 2 The current results are documented in Table 15 6-14, which was recently updated by LDC 2005-037 This change adds the impacts of the secondary containment bypass leakage through the service air and instrument air lines and the water leakage through the HFTS to the current SAR results The updated LOCA doses are compared to the current SAR values in the table below Dose Rem TEDE Location Current SAR New SAR Table 15 6- Regulatory Table 15 6-14 14 with these leak Limit LDC 2005-037 paths 10CFR50 63 Exclusion Area Boundary 841 845 25 Low Population Zone Control Room L 446 364 456 369 25 5

A minimal increase in consequences is defined as 10% of the difference between the current calculated dose value and the regulatory limit . As shown below, these increases are less than "minimal" increases at all dose locations Therefore, these changes do no result in more than a minimal increase in the consequences of an accident previously evaluated in the FSAR Dose Re TETEDE Location "Minimal" Actual Increase Increase Exclusion Area Boundary 166 004 Low Population Zone 205 010 Control Room 014 005 4 Result m more than a minimal increase in the consequences of a malfunction of a structure, 0 Yes system, or component important to safety previously evaluated in the FSAR? No BASIS The active venting function of the instrument and service air systems is an important-to-safety SSC that is currently credited in the FSAR with mitigating the doses from a LOCA This change determines the consequences of the failure of the active vent system in the event of a LOCA such that there is no reliance on this SSC to function post-LOCA As shown m the response to Question 3, this increase in consequences is not more than a minimal increase, even when combined with the doses from the HFTS leakage 5 Create a possibility for an accident of a different type than any previously evaluated in the [] Yes FSAR' 2 No BASIS This change does not physically modify any SSC and cannot create any accident of a different type than evaluated in the FSAR 6 Create a possibility for a malfunction of a structure, system, or component important to safety 0 Yes with a different result than any previously evaluated in the FSAR? No BASIS The proposed change only updates the LOCA dose analysis and makes no physical modifications to the plant such that important-to-safety SSCs will not be impacted by the proposed change Therefore, this change will not create the possibility for a malfunction will a result different than evaluated m the FSAR LI-101-01, Rev . 8; Effective Date: 6123105

50.59 REVIEW FORM Page 1 0 of 10 7 Result in a design basis limit for a fission product barrier as described in the FSAR being j] Yes exceeded or altered' No BASIS This change does not result in a design basis fission product barrier being exceeded or altered The LOCA dose analysis is performed based on the failure of the fuel cladding barrier and the RCS pressure boundary The LOCA dose analysis credits the containment and secondary containment and these changes do not alter or degrade the effectiveness of these boundaries. This change does not physically modify any SSC such that the fission product barriers are not exceeded or altered 8 Result in a departure from a method of evaluation described in the FSAR used in establishing © Yes the design bases or in the safety analyses? No BASIS The radiological analyses evaluated in these calculations are leak paths that may exist after a LOCA The radiological computer code applied in these calculations is called RAPTOR and has recently been approved for use per Safety Evaluation 2005-0002-R00 based on extensive benchmarks to the previous GGNS methodologies and the NRC's own methods Calculation XC-+Q1 P53-05011 for the instrument and service air leakage paths credits aerosol settling and halogen deposition to reduce the source term release to the environment These models have not been applied at GGNS but have been endorsed by the NRC for other BWR applications Specifically, the aerosol settling model was developed by the NRC in Accident Evaluation Branch (AEB) 98-03 for the main steam line at Perry and has been accepted at other plants besides Perry The elemental and organic halogen deposition model was developed by Cline [J E Cline, "MSIV Leakage Iodine Transport Analysis,"

Letter Report dated March 26, 1991] and is endorsed in Appendix A to Reg Guide 1 183 It is important to note that these models were initially developed to model source term transport in the main steam line piping, which is a significant leakage path for many BWRs and this application applies these same models to the instrument air and service air piping The extension of this methodology to the smaller diameter and lower temperature piping associated with the air systems is provided in the methodology Calculation XC-Q1 M46-04004 documents the impact of leakage through the HFTS This is not a secondary containment bypass leakage path like instrument and service air but is specifically evaluated since it is not part of the containment La calculation or the analyzed 1 12 gpm of suppression pool leakage in the LOCA dose analysis This calculation applies the NRC-approved assumptions documented in Reg Guide 1 183 with the RAPTOR methodology Therefore, these changes apply the relevant methodologies approved for use at GGNS and do not represent a departure from a method of evaluation described in the FSAR If any of the above questions is checked "YES," obtain NRC approval prior to implementing the change by initiating a change to the Operating License In accordance with NMM Procedure LI-113.

LI-101-01, Rev . 8; Effective Date: 6123105

GGNS 50.59 Safety Evaluation Number SE 2005-0007-R00

el r--

J It Mo 7 _X 50.69 REVIEW FORM Page I of 9

1. OVERVIEW / SIGNATURES Facility:

Document Reviewed : LBDC 2004-0095 Change/Rev . :

System Designator(s)/Description

N71 Circulating Water System Description of Proposed Activity : The proposed changes modify the TRM and ODCM the required act and operability requirements of ODCM/TRM 6.3 .9 applicable to Discharge Canal flow monitoring instrumentation . The change wilt affect administrative requirements only, and no physical modification is being made . The affected instrumentation is non-safety related and has no automatic functions. This change will make Circulating Water Slowdown the primary source of dilution flow for liquid radwaste discharges, and allow use of Discharge Canal flow instrumentation only as a means of estimating dilution flow when Circulating Water Slowdown flow instrumentation is inoperable . Only the Circulating Water Slowdown flow instrumentation will be considered a ODCMITRM required channel.

FSAR Section 11 .2 describes the Liquid Radwaste discharge system . Prior to being released to the environment, liquid radwaste is processed on a batch bats and sampled to determine radioactivity. Based on the results of the sample analysis, the waste may be released under controlled conditions to the environment.

Liquid radwaste is only released via the discharge basin, after being diluted by Circulating water blowdown or Plant Service Water.

CM 6.3 .9 specifies the requirements for radioactive liquid effluent monitoring instrumentation. The ng conditions and actions associated with this TRM are applicable at all times. Required instrumentation es one channel of radiation monitoring on the liquid radwaste effluent monitoring line . The radiation monitor provides alarm and termination of the release. In addition to radiation monitoring, flow rate measurement devices are provided on the liquid radwaste effluent line, and on two dilution flow paths.

TRM 6.3 .9/ODC M Table 2.b currently allows use of flow instrumentation on either the Discharge Canal or Circulating Water Slowdown line for measuring dilution flow when discharging liquid radwaste to the environment. The two instrumentation channels are independent and provide operational flexibility for performing discharges. The Circulating Water Slowdown flow instrumentation measures flow into the discharge basin. This channel cannot be used for dilution with PSW. The Discharge Canal flow element is located between the discharge basin and the outflow to the river. It can be used to measure dilution flow from both Circulating water blowdown and Plant Service Water.

Inherent limitations in the design and application of the discharge canal flow monitor have adversely affected the availability of the instrument channel. Operating experience at GGNS has shown this instrumentation to be difficult to maintain within an acceptable level of accuracy . Circulating water blowdown flow instrument channel has proven much more reliable . Also, Circulating water blowdown flow rate instrumentation provides an automatic isolation of the radwaste discharge on low dilution flow, whereas no automatic functions are associated with the Discharge Canal flow instrumentation . Consequently, circulating water blowdown is preferred and generally used to provide dilution flow and associated flow monitoring during release.

The proposed change will result in the Circulating water blowdown flow instrumentation being the only required dilution flow channel. Canal discharge flow will no longer be a ODCM/TRM qualified instrument . The canal discharge flow rate instrumentation currently provides operational flexibility as an alternative to circulating water blowdown flow instrumentation or W PSW 4 used for dilution flow . As a result of the change, use of PSW flow for dilution will be allowed only a%* LCO &19 is entered, since there will be no qualified instrument channel capable of measuring the PSW flow . This is acceptable, since circulating water is the preferred channel. As discharge flow not previously discussed, cwt 44 normally used for monitoring discharges.

L1401-01, Rev. 8 ; Effective Date : 6/23105

50 .59 REVIEW FORM Page 2 of 9 There is no safety significance to this change since radwaste discharges can be still be performed as usual with circulating water blowdown . There is no additional level of safety provided by the canal discharge flow instrumentation . There is no automatic isolation associated with the canal discharge flow instrumentation. Also, since discharges are performed on a batch basis, unavailability of dilution flow monitoring can normally be Action corrected while discharge is secured in accordance with TRM 019 AA . In situations where blowdown flow instrumentation cannot be restored before a batch discharge is necessary, Action A.2 allows entry into Condition C which requires dilution flow to be estimated once per four hours. Use of the canal discharge flow instrumentation will only be allowed after entering LCO 530. It could then be used if available for estimating dilution flow per Acton C.I . Estimating flow is already required by Action CA, and the canal discharge flow will provide an additional means of estimating flow. Also, per Action C.2 the required channel must be restored to operable status within 30 days The changes to ODCM/TRM 6.3 .9 constitute a change to the ODOM since this TRM LCO is also contained in the ODCM . However, no ODOM calculation methodologies or other information is affected . Changes to the Offsite Dose Calculation Manual are controlled by Grand Gulf Nuclear Station Technical Specification (TS),

Administrative Controls Section 5.5 .1 . In accordance with TS Section 5 .5 .1 an ODCM change shall contain:

1 . sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change, and

2. a determination that the change(s) maintain the levels of radioactive effluent control required by IOCFR20.1302, 40CFRI90, IOCFR50.36a, and 10CFR50, Appendix 1, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations .

Regulations 40CFR190, 10CFR50.36a, and lOGFR50 Appendix I deal with dose calculations in the ODOM .

None of the dose calculation methodologies in the ODOM are affected by this change . Therefore, these regulations are not affected .

with radioactive Regulation I OCFR20 .1302 deals releases to unrestricted areas. TRM LCO 6.11 .1 is the technical requirement for 10CFR2a1302. The proposed changes only affect liquid radwaste discharges . No liquid or gaseous release points, parameters, or requirements are affected . Therefore the requirements of I OCFR20 .1302 (TRM LCO 6.11 .1) are met.

Check the applicable review(s): (only the sections indicated must be included in the Review.)

EDITORIAL CHANGE of a Licensing Basis Document Section I El SCREENING Sections I and 11 required n 50.59 EVALUATION EXEMPTION Sections 1, 11, and III required ED 5159 EVALUATION (# : 4IQ 6 3 - I - R n ?) __J Sections 1, 11, and IV required 1 I Ricky M Uddell Name (print) I Signature Chairman's Name (print) 1 ignature / Date (Required only for Programi Itic Exclusion Screenings and 50 .59 Evaluations .)

LI-101-01, Rev. 8; Effective Date : 6/23/05

50.59 REVIEW FORM Page 3 of 9 If. SCREENINGS A. Licensing Basis Document Review

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

Operating License YES NO CH NGE # and/or SECTIONS IMPACTED Operating License 0 EN TS Ej 01 NRC Orders 1 0 1 ~Q I If "YES," obtain NRC approval prior to implementing the change by initiating an LBD change in a ccordance with NMMLI-113 . (See 1-1-101 for exceptions .)

LBDs controlled under 50.59 YES NO CHANGE # (if applicable) and/or SECTIONS IMPACTED FSAR Ej 401 TS Bases 0 Aral Technical Requirements Manual OR El TRM 6.3.9 Action C.1 (Pg 6.3-20), and Table 6.3.9-1, Section Zb (Pg 6.3-23)

Core Operating Limits Report NRC Safety Evaluation Report and supplements for the initial FSAR 1 NRC Safety Evaluations for amendments to the Operating License' Section OR If "YES," perform an Exemption Review per III perform a 50.59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMM Lt-113 . If obtaining NRC approval, document the LSD change in Section II.A.5. However, the change cannot be implemented until approved by the NRC . Complete Section 11.

LBDs controlled under other YES NO CHANGE # (if applicable) and/or regulations SECTIONS IMPACTED Quality Assurance Program Manua, 2 ri 0 3

Emergency Plane' El 1101 3,4 1 

Fire Protection PrograM El

{includes the Fire Hazards Analysis)

Offsite Dose Calculations ManUap 4 N El I ODcmIrRM 6.3.9 Action C.1 (Pgr A-10, and Table I 6.3.9-1, Section 2.b (Pg A-14)

If "YES," evaluate any changes in accordance with the appropriate regulation AND initiate an LBD change in accordance with NMM LI-I 13.

"2 if "YES," see LIA01 . No LSD change is required .

If "YES," notify the responsible department and ensure a 50.54 evaluation is performed. Attach the 50 .54 evaluation.

3 Changes to the Emergency Plan, Fire Protection Program, and Offshe Dose Calculation Manual must be approved by the OSRC in accordance with NMM OM-119 .

4 K YES : evaluate he change in accordance with the requirements of the facility's Operating License Condition or under 50.59, as appropriate, LI-101-01, Rev . 8; Effective Date: 6/23105

50 .59 REVIEW FORM Page 4 of 9

2. Does the proposed activity involve a test or experiment not described in the FSAR? 0 Yes No If "YES," perform a 50 .59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change AND initiate an LBD change in accordance with NMM LI-113, if applicable . If obtaining NRC approval, document the change in Section II .A .5 . However, the change cannot be implemented until approved by the NRC. Complete Section 11 .

3. Basis Explain why the proposed activity does or does not impact the Operating License/Technical Specifications and/or the FSAR . If the proposed activity involves a potential test or experiment not previously described in the FSAR also include an explanation . Discuss other LBDs if impacted . Adequate basis must be provided within the Screening such that a third-party reviewer can reach the same conclusions . Simply stating that the change does not affect TS or the FSAR is not an acceptable basis.

TRM 6.3.9, Radioactive Liquid Effluent Monitoring Instrumentation, is the only LBD that addresses the affected instrumentation. This instrumentation is not discussed in the Operating License or Technical Specifications. The proposed change to TRM 6.3.9 will remove the canal discharge flow as a required Instrumentation channel for dilution flow for liquid radioactive waste discharges.

This change involves no test or experiment, and will in no way affect the normal method of monitoring liquid radwaste discharge to the environment. The primary channel used for dilution flow rate monitoring is circulating water blowdown flow, which is unaffected by this change. Only the use of the alternate channel, discharge canal flow, is affected by the change. The physical instrumentation for discharge canal flow will remain and will be unaffected by the TRM change.

Only the TRM operability and surveillance requirements for the subject instruments are affected.

4. References Discuss the methodology for performing LBD searches . State the location of relevant licensing document 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 . NOTE : Ensure that manual searches are performed using controlled copies of the documents. If you have any questions, contact your site Licensing department.

Electronic search method used : Keywords :

Keyword search of UFSAR, Operating Radwaste discharge, radioactive release, License Manual, TSITRM, TS Bases, ODCM effluent, dilution flow LBDs reviewed manually :

UFSAR Section 99 .2.9, 19 .2.3, 15.7.2, 15.7.3 ODCM, QAPM, FPP, Emergency Plan

5. Is the validity of this Review dependent on any other change? El Yes No If "YES," list the required changes/submittals . The changes covered by this 50 .59 Review cannot be implemented without approval of the other identified changes (e.g ., license amendment request) . Establish an appropriate notification mechanism to ensure this action is completed.

LI-10'0-01, Rev. 8; Effect Date : 6123105

50 .59 REVIEW FORM Page 5 of 9 B. ENVIRONMENTAL SCREENING If any of the following questions is answered "yes," an Environmental Review must be performed in accordance with NMM Procedure EV-115 and attached to this 50 .59 Review . Consider both routine and non-routine (emergency) discharges when answering these questions.

Will the proposed activity being evaluated:

YES NO Involve a land disturbance equal to or in excess of one acre (i .e ., grading activities, construction of buildings, excavations, reforestation, creation or removal of ponds)?

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

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

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

5. R Increase the concentration or quantity of chemicals being discharged to the river, lake, or air?

6. [j Discharge any new or different chemicals that are currently not authorized for use by the state regulatory agency?

7. r_1 Change the design or operation of the intake or discharge structures?

0.0 Modify the design or operation of the cooling tower that will change water or air flow characteristics?

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

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

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

12 . Involve the installation or use of equipment that will result in a new or additional air emission discharge?

13 . n Involve the installation or modification of a stationary or mobile tank?'

14 . [~ Involve the use or storage of oils or chemicals that could be directly released into the environment?

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 for guidance in answering this question .

LI-101-01, Rev. 8; Effective Date : 6123/05

50 .59 REVIEW FORM Page 6 of 9 C. SECURITY PLAN SCREENING If any of the following questions is answered "yes," a Security Plan Review must be performed by the Security Department to determine actual impact to the Plan and the need for a change to the Plan .

Could the proposed activity being evaluated:

YES NO

1. [l Add, delete, modify, or otherwise affect Security department responsibilities (e .g .,

including fire brigade, fire watch, and confined space rescue operations)?

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

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

4. [] 11 Affect (block, move, or alter) security lighting by adding or deleting lights, structures, buildings, or temporary facilities?

5. [] Modify or otherwise affect the intrusion detection systems (e .g ., E-fields, microwave, fiber optics)?

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

7. [] 141 Modify or otherwise affect (block, move, or alter) installed access control equipment, intrusion detection equipment, or other security equipment?

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

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

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

The Security Department answers the following question if one of questions C.1 through C.10 above was answered "yes ."

Is a change to the Security Plan required? E][] Yes No Attach to this 50 .59 Review or reference below documentation for accepting a "yes" answer for any of Questions C.1 through C.10, above.

Name of Security Plan reviewer (print l Signature I Data LI-101-01, Rev. 8; Effective Date : 6123105

50 .59 REVIEW FORM D. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) SCREENING (NOTE: This section is not applicable to Grand Gulf or Waterford 3 and may be removed from 50.59 Reviews performed for Waterford 3 proposed activities .)

If any of the following questions is answered "YES," a 72A8 Review must be performed in accordance with NMM Procedure LI-i 12 and attached to this 50.59 Review.

Will the proposed activity being evaluated : NIA FOR GGNS PER STEP 5.4.2.4 YES NO

1. El [] Any activity that directly impacts spent fuel cask storage or loading operations?

2. F1 [:] Involve the ISFSI including the concrete pad, security fence, and lighting?

3. E] E] Involve a change to the on-site transport equipment or path from the Fuel Building to the ISFT?

Involve a change to the design or operation of the Fuel Building fuel bridge including setpoints and limit switches?

the

5. F1 n Involve a change to Fuel Building or Control Room(s) radiation monitoring?

6. Involve a change to the Fuel Building pools including pool levels, cask pool gates, cooling water sources, and water chemistry?

7. involve a change to the Fuel Building handling equipment (e.g, bridges and cask cranes, structures, load paths, lighting, auxiliary services, etc)?

Involve a change to the Fuel Building electrical power that could potentially impact cask loading or storage activities?

9. M D Involve a change to the Fuel Building ventilation that could potentially impact cask loading or storage activities?

10. F-1 El Involve a change to the ISFSI security?

11 . El n Involve a change to off-site radiological release projections from non-ISFSI sources?

12 . F] n Involve a change to spent fuel characteristics?

13. F1 0 Redefine/change heavy load pathways?

14. r_1 n Involve fire and explosion protec n near or in the on-site transport paths or near the ISF&?

15. El involve a change to the loading bay or supporting components power that could potentially impact cask loading or storage activities?

16. M 0 New structures near the ISFSI?

17. F M Modifications to any plant systems that support dry fuel storage activities?

18. El [] Involve a change to the nitrogen supply, service air, dernineralized water or borated water system in the Fuel Building?

LI-101-01, Rev . 8; Effective Date: 6/23/05

50 .59 REVIEW FORM Page 8 of 9 IV . 50 .59 EVALUATION License Amendment Determination Does the proposed Change being evaluated represent a change to a method of evaluation [~ Yes ONLY ? If "Yes," Questions 1 - 7 are not applicable ; answer only Question 8. If "No," answer No all questions below.

Does the proposed Change :

1. Result in more than a minimal increase in the frequency of occurrence of an accident © Yes previously evaluated in the FSAR? No BASIS:

The affected instrumentation is non-safety related, and no physical modification is being made to any plant equipment. Accidents associated with liquid radioactive waste releases that are evaluated in FSAR 15.7 are bounded by liquid radwaste tank failures. There is no accident analysis associated with a failure of the discharge canal flow instrumentation, nor is there any credit given for this function . This change will not affect the bases or results of any accident analyses . Therefore, this change will not increase the frequency of occurrence of an accident.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a M Yes structure, system, or component important to safety previously evaluated in the FSAR? 0 No BASIS:

The affected instrumentation is non-safety related and is not credited in any safety analysis. No physical modification is being made, therefore this change will have no effect on any structure, system or component important to safety. The requirements of TRMIODCM 6.3.9 will be maintained for the liquid radwaste effluent line flow measurement, radiation monitor, and dilution flow (i e. circulating water blowdown) measurement. Only the requirements associated with the discharge canal flow instrumentation are affected . Therefore this change will not increase the likelihood of occurrence of a malfunction of a structure, system or component important to safety.

3. Result in more than a minimal increase in the consequences of an accident previously [] Yes evaluated in the FSAR? No BASIS:

The affected instrumentation is non-safety related. There are no automatic functions associated with discharge canal flow instrumentation and no credit is taken for it in any safety analysis. Accidents associated with liquid radioactive waste releases are bounded by liquid radwaste tank failures. No analyzed accidents or equipment used to mitigate an accident are affected by this change. Therefore, this change will not increase the consequences of any accident.

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, [] Yes system, or component important to safety previously evaluated in the FSAR? No BASIS:

There is no credit taken for discharge canal flow instrumentation in any safety analysis, and no other equipment is affected in any way by this change . No equipment modification is being made and no requirements associated with equipment important to safety are affected. No structure, system or component important to safety is in any way affected by this change . The requirements of TRMIODCM 6.3.9 will be maintained for the liquid radwaste effluent line flow measurement, radiation monitor, and dilution flow (i.e . circulating water blowdown) measurement. Only the requirements associated with the discharge canal flow instrumentation are affected. Therefore, this change will not increase the consequences of a malfunction of a structure, system or component important to safety.

LI-101-01, Rev. 8; Effective Date : 6/23/05

50 .59 REVIEW FORM Page 9 of 9 Create

5. a possibility for an accident of a different type than any previously evaluated in the Yes FSAR? No BATS :

No physical change is being implemented. Only the operability and surveillance requirements for discharge canal flow instrumentation is affected. No new failure modes are created for any structure, system or component as a result of this change. Therefore, this change will not create the possibility of a different type of accident than previously evaluated.

6. Create a possibility for a malfunction of a structure, system, or component important to safety n Yes with a different result than any previously evaluated in the FSAR? 140 BATS :

There is no physical change being made to any structure, system or component Only the operability and surveillance liquid requirements for discharge canal flow instrumentation is affected. Accidents associated with radioactive waste releases are bounded by liquid radwaste tank failures. No new potential for a malfunction of equipment is created, and no potential for any different results of malfunctions previously evaluated. This change will not create the possibility of a different type of accident than previously evaluated.

Result

7. in a design basis limit for a fission product barrier as described in the FSAR being F-1 Yes exceeded or altered? Eg No BASIS:

There is no credit for this instrumentation in any safety analysis . There is no physical change being made to any structure, system or component Fission product barriers, i. e. fuel cladding, reactor coolant pressure boundary, primary and secondary containment, are in no way affected. Analyzed accidents associated with liquid radioactive waste releases occur outside containment and do not involve any containment barrier integrity aspects. Therefore, this change will not affect any fission product barriers.

8. Result in a departure from a method of evaluation described in the FSAR used in establishing n Yes the design bases or in the safety analyses? Z No BASIS:

The requirements of TRMIODCM 6.19. t 6A AZ and 6-11 .3 for liquid radwaste effluent concentration, dose limits, and treatment systems are unchanged. All the basic requirements for radioactive liquid effluent monitoring instrumentation as discussed in the ODCM Bases for TRM 6.3.9 are maintained. The applicable General Design Criteria 60, 63, and 64 of 10CFR Appendix A will continue to be met. The maintained requirements of TRMIODCM 6.3 .9 wV be for the liquid radwaste effluent line flow a water measurement, radiation monitor, and dilution flow (L . circulating b/ow,doval) measurement. Only the requirements associated with the discharge canal flow instrumentation are affected. Only the operability and surveillance requirements for discharge canal flow instrumentation is affected. There is no credit for this instrumentation in any safety analyses, and existing methods of evaluations for accident analyses as described in the FSAR are unchanged. Therefore, this change will not affect any methods of evaluation for design bases or safety analyses.

If any of the above questions is checked "YES,' obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure LI-113 .

LI-1 01 -01, Rev. 8 ; Effective Date : 6123/05

GGNS 50 .59 Safety Evaluation Number SE 2005-0008-R00

ge I of 13

1. OVERVIEW I SIGNATURES Facility :

Document Reviewed : ER-GG-2003-0234-001 Change/Rev.: QQ System Designator(syDescription :

P75 Standby Diesel Generator Description of Proposed Activity :

ER-GG-2003-0234-001 extends the frequency of the inspection of the Division 2 Diesel generator fuel oil storage tank by three months, from December 2006 until March 2006.

NOTE: ER-GG-2003-0234-000 approved extending the inspection for the Division I and Ii Diesel generator fuel oil storage tanks until December 2005. Reference Safety Evaluation number 2004-0004-ROO approved August 09, 2004.

Check the applicable review(s): (Only the sections indicated must be included in the Review.)

EDITORIAL CHANGE of a Licensing Basis Document Section I SCREENING Sections I and 11 required E] 50.59 EVALUATION EXEMPTION Sections 1, 11, and III required N 60.59 EVALUATION (#: Z100S---0009-'RQ6 Sections l, ll, and IV required Preparer K. M. Black/ .1010*01 &Xwn- /Entergy/Engineering/

Name (print) / Signat5/ Company / Department! Date Reviewer: R. W. Fuller! /Entergy/Engineering/

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

Chairman's Name (print%/ Signature / Date (Required only far Programmatic Exclusion Screenings and 50.59 Evaluations.)

Coll

50 .59 REVIEW FORM Page 2 of 13

11. SCREENINGS A. Licensina Basis Document Review

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

Operating License YES NO CHANGE # and/or SECTIONS IMPACTED Operating License Ej IS TS EJ 190V NRC Orders I IT If "YES," obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMM LI-113. (See LI-101 for exceptions.)

I-BDs controlled under 50 .59 YES NO CHANGE # (if applicable) and/or SECTIONS IMPACTED FSAR Q El I U FSAR Appendix 3A, Reg. Guide 1 .137, LBD 2005-0082

-7 TS Bases 1-3 G] i .0 Technical Requirements Manual 101 El TRM SR TR 3.8.3.6, LBD 2005-0082 Core Operating Limits Report 1:1 R NRC Safety Evaluation Report and supplements for the initial FSAR' NRC Safety Evaluations for amendments to the Operating License' If "YES," perform an Exemption Review per Section III OR perform a 50 .59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMM U-113. If obtaining NRC approval, document the LBD change in Section II.A.5. However, the change cannot be implemented until approved by the NRC. Complete Section [I.

LBDs controlled under other YES NO CHANGE # (if applicable) and/or regulations SECTIONS IMPACTED 2

Quality Assurance Program ManUal Emergency Plan 2* 3 Fire Protection PrograM3' 4 E3 N (includes the Fire Hazards Analysis)

Offsite Dose Calculations Manua13, 4

0 Z If "YES," evaluate any changes in accordance with the appropriate regulation AND initiate an LBD change In accordance with NMM Lt-113 .

' if "YES," see LIAI)i . No LBD change is required .

2 If "YES," notify the responsible department and ensure a 50 .54 evaluation is performed . Attach the 50.54 evaluation .

3 Changes to the Emergency Plan, Fire Protection Program, and Offsite Dose Calculation Manual must be approved by the OSRC in accordance with NMM OM-119 .

4 If 'YES,* evaluate the change in accordance with the requirements of the facility's operating license Condition or under 50.59, as appropriate, LI-101-01, Rev. 8; Effective Date : 6123105

50 .59 REVIEW FORM Page 3 of 13

2. Does the proposed activity Involve a test or experiment not described in the FSAR? n Yes No If "YES," perform a 50 .59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change AND initiate an LBD change in accordance with NMM LI-113, if applicable. If obtaining NRC approval, document the change in Section II.A.5. However, the change cannot be implemented until approved by the NRC. Complete Section II.

3. Bas LI-101-01, Rev . 8; Effective Date: 6/23/05

50.59 REVIEW FORM Page 4 of 13 Explain why the proposed activity does or does not impact the Operating License/Technical Specifications and/or the FSAR . If the proposed activity involves a potential test or experiment not previously described in the FSAR also include an explanation . Discuss other LBDs if impacted. Adequate basis must be provided within the Screening such that a third-party reviewer can reach the same conclusions. Simply stating that the change does not affect TS or the FSAR is riot an acceptable basis.

The purpose of the evaluation is to provide the rationale for extending the Division 11 inspection to March 2006. The one time inspection extension will be documented in the TRM requirement SR TR3 .8.3.6 and FSAR Appendix 3A, Reg . Guide 1 .137. The change is based on previous Diesel Generator Fuel tanks inspections where only minor wall wear and degradation was observed (ref. MNCR 108-92, MNCR 174-92, MAI 327093 and WO 00056003). The TRM and FSAR revision will be to take credit for the minor wall wear and wall degradation to the Diesel Generator Fuel Storage tank. The wall degradation is due to the sample element . The sample element is the device used to measure the tank volume and the degradation is due to monthly use. The Division I fuel oil storage tank was inspected in Feburary of 2005 and no anomalies were noted.

Operating License :

The Grand Gulf Nuclear Station (GGNS) operating license does not affect Diesel Generator Fuel tank inspections . The Technical Specifications and the Environmental Protection Plan are not impacted by this ER. Therefore, the proposed activity does not impact the GGNS operating license .

Technical Specifications :

The Diesel Generator Fuel tank inspection is not covered by Technical Specifications. However, Technical Requirement Manual Surveillance Requirement SR TR3.8.3 .6 has requirements for Fuel tank inspections . The evaluation will not create a system configuration or operating condition such that a Technical Specifications LCO or surveillance requirement is no longer adequate. Likewise, the evaluation will not bypass or invalidate features required to be operable by the Technical Specifications or exceed any limits specified in the Operating License and Technical Specifications . Therefore, no Technical Specifications change is required for the issuance of this evaluation .

UFSAR:

The LIFSAR is affected by this evaluation because it is a one time extension of the Division 11 Fuel Oil Storage tank inspection to March 2006. UFSAR section for Regulatory Guide 1 .137 on page 3A/1 .137 identifies the Fuel oil system for Standby Diesel Generators . This part of the FSAR will be changed for the one time extension of the Diesel Fuel Oil Storage Tank inspection to March 2006. The one time exception to the inspection will allow the Fuel Oil Storage Tank inspection to be extended to March 2006 .

This 50.59 provides a basis for the Diesel Generator Fuel Storage Tank inspection extension to March 2006.

NRC Orders :

The NRC Orders issued at Grand Gulf are not affected by this evaluation because this evaluation deals with Diesel Generator Fuel Oil Storage tank inspection and this evaluation is not to be used for security reasons .

Technical Specification Bases :

There are no Technical Specifications or Bases impacted by this activity. The Technical Specification for Diesel Fuel Oil is 3.8.3 and the surveillance requirement is under Technical Requirement Manual is TR3531 for Diesel Generator Fuel Oil Storage Tank inspection . These items will remain the same.

This is an evaluation for increasing the inspection to March 2006 which is not part of the Technical Specification Bases .

LI-101-01, Rev . 8; Effective Date: 6/23/05

50.59 REVIEW FORM Page 5 of 13 Technical Requirements Manual (TRM):

Technical Requirements Manual SR TR3.8.3.6 is affected by this activity. This section is revised to indicate the inspection extension for Division 2 DG fuel oil storage tank until March 2006. This section mentions that the fuel storage tank inspection is in conjunction with of ASME Boiler and Pressure Vessel Section Xi inspection . The only ASME B&PV Section XI requirement is pressurizing the tank with the fuel still in the tank. This 50.59 clarifies that Diesel Generator Fuel Oil storage tank inspection will be extended one time to March 2006. The reason is that previous diesel generator fuel oil storage tank inspections discovered only minor wear and wall degradation to the fuel oil tank and that increasing the inspection to March 2006 will be acceptable.

Core Operating Limits Report :

This activity does not impact the COLR (GGNS Core Operating Limits Report) . This evaluation explains extending the Diesel Fuel Oil Storage tank inspection to March 2006. It does not have any impact on the COLR and does not affect any licensing activities.

Offsite Dose Calculations Manual :

This activity does not impact any equipment required to monitor offsite dose. Therefore, no changes to the ODCM is required .

NRC Safety Evaluation Reports :

There is no impact to any SERs by providing an evaluation for evaluating extending the diesel fuel oil storage tank inspection to March of 2006.

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.

Emergency Plan:

There is no impact to the Emergency Plan for evaluating extending the diesel generator fuel oil storage tank inspection to March of 2006 .

Fire Protection Program:

This activity does not change any commitments contained in the Fire Protection Program . Therefore, this activity does not require a change to the Fire Protection Program .

Test and Experiment:

Evaluating extending the diesel fuel oil storage tank inspection to March of 2006 does not constitute a test or experiment .

4. References Discuss the methodology for performing LBD searches. State the location of relevant licensing document 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. NOTE: Ensure that manual searches are performed using controlled copies of the documents. If you have any questions, contact your site Licensing department.

LI-101-01, Rev. 8; Effective Date: 6123105

50 .59 REVIEW FORM Page 6 of 13 Electronic search method used : Keywords:

Autonamy Fuel oil storage tank LBDs reviewed manually :

TRM SR TR3.8.3.6, UFSAR Appendix 3A page 3A/1 .137-1 & 2, UFSAR Sections 8.3 and 9.5.4 and Technical Specification Bases 3.8.3

5. Is the validity of this Review dependent on any other change? El Yes If "YES," list the required changes/submittals . The changes covered by this 50 .59 Review cannot be implemented without approval of the other identified changes (e.g ., license amendment request). Establish an appropriate notification mechanism to ensure this action is completed.

t-1-101-01, Rev. 8; Effective Date : 6/23105

50.59 REVIEW FORM Page 7of13 B. ENVIRONMENTAL SCREENING If any of the following questions is answered "yes," an Environmental Review must be performed in accordance with NMM Procedure EVA15 and attached to this 50 .59 Review. Consider both routine and non-routine (emergency) discharges when answering these questions.

Will the proposed activity being evaluated :

YES N a Involve a land disturbance equal to or in excess of one acre (i.e., grading activities, construction of buildings, excavations, reforestation, creation or removal of ponds)?

2. [~ Involve any land disturbance of undisturbed land areas (i.e., grading activities, construction, excavations, reforestation, creating, or removing ponds)?

3. F] involve dredging activities in a lake, river, pond, ditch, or stream?

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

5. [1 23 Increase the concentration or quantity of chemicals being discharged to the river, lake, or air?

6. Discharge any new or different chemicals that are currently not authorized for use by the state regulatory agency?

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

Modify the design or operation of the cooling tower that will change water or air flow characteristics?

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

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

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

12. (_ Involve the installation or use of equipment that will result in a new or additional air emission discharge?

OR Involve the installation or modification of a stationary or mobile tank?'

Involve the use or storage of oils or chemicals that could be directly released into the environment?

15. (] 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 for guidance in answering this question.

LI-101-01, Rev . 8; Effective Date: 6123105

50.59 REVIEW FORM Page 8 of 13 C. SECURITY PLAN SCREENING If any of the following questions is answered "yes," a Security Plan Review must be performed by the Security Department to determine actual impact to the Plan and the need for a change to the Plan.

Could the proposed activity being evaluated :

YES NO

1. D Add, delete, modify, or otherwise affect Security department responsibilities (e.g.,

including fire brigade, fire watch, and confined space rescue operations)?

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

3. 0 1a Cause materials or equipment to be placed or installed within the Security Isolation Zone?

4. Affect (block, move, or alter) security lighting by adding or deleting lights, structures, buildings, or temporary facilities?

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

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

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

11 Modify or otherwise affect primary or secondary power supplies to access control equipment, intrusion detection equipment, other security equipment, or to the Central Alarm Station or the Secondary Alarm Station?

9. F1 Modify or otherwise affect the facility's security-related signage or land vehicle barriers, including access roadways?

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

The Security Department answers the following question if one of questions C.1 through C.10 above was answered "yes."

Is a change to the Security Plan required? [l Yes

[] No Attach to this 50.59 Review or reference below documentation for accepting a "yes" answer for any of Questions C.1 through C .10, above.

Name of Security Plan reviewer (print t Signature I Data LI-101-01, Rev. 8; Effective Date: 6123/05

50.59 REVIEW FORM Page 9 of 13 D. INDEPENDENT SPENT FUEL STORAGE INSTALLATION {ISFSI) SCREENING NOTE: This section is not applicable to Grand Gulf or Waterford 3 and may be removed from 50.59 Reviews performed for Waterford 3 proposed activities.)

If any of the following questions is answered "YES," a 72.48 Review must be performed in accordance with NMM Procedure LI-112 and attached to this 50.59 Review.

Will the proposed activity being evaluated :

YES NO

1. F-1 Any activity that directly impacts spent fuel cask storage or loading operations?

2. El 0 Involve the ISFSI including the concrete pad, security fence, and lighting?

3. n Involve a change to the on-site transport equipment or path from the Fuel Building to the ISFSI?

Involve a change to the design or operation of the Fuel Building fuel bridge including setpoints and limit switches?

5. El 0 Involve a change to the Fuel Building or Control Room(s) radiation monitoring?

6. Involve a change to the Fuel Building pools including pool levels, cask pool gates, cooling water sources, and water chemistry?

7. Involve a change to the Fuel Building handling equipment (e.g ., bridges and cask cranes, structures, load paths, lighting

• auxiliary services, etc)?

11 Involve a change to the Fuel Building electrical power that could potentially impact cask loading or storage activities?

9. © Involve a change to the Fuel Building ventilation that could potentially impact cask loading or storage activities?

10. ~Zv Involve a change to the ISFSI security?

11 . "5 Involve a change to off-site radiological release projections from non-ISFSI sources?

12. 100 Involve a change to spent fuel characteristics?

13. El Redefine/change heavy load pathways?

14. [] OR Involve fire and explosion protection near or in the on-site transport paths or near the ISFSI?

15. E] Z Involve a change to the loading bay or supporting components power that could potentially impact cask loading or storage activities?

16. [:1 New structures near the ISFSI?

17. 0 1~4 Modifications to any plant systems that support dry fuel storage activities?

18. El Involve a change to the nitrogen supply, service air, demineralized water or borated water system in the Fuel Building?

LI-101-01, Rev . 8; Effective Date: 6/23105

50.59 REVIEW FORM Page 10 of 13 Ill . 50.59 EVALUATION EXEMPTION A. Check the applicable box below. If a box is checked, clearly document the basis to Section III .B, below. If none of the boxes are appropriate, perform a 50.59 Evaluation in accordance with Section IV. Provide supporting documentation or references as appropriate .

[] The proposed activity meets all of the following criteria regarding design function :

The proposed activity does not adversely affect the design function of an SSC as described in the FSAR; AND The proposed activity does not adversely affect a method of performing or controlling a design function of an SSC as described in the FSAR; AND The proposed activity does not adversely affect a method of evaluation that demonstrates intended design function(s) of an SSC described in the FSAR will be accomplished .

An approved, valid 50.59 Review(s) covering associated aspects of the proposed activity already exists. Reference 50.69 Evaluation # (if applicable) or attach documentation . Verify the previous 50.59 Review remains valid.

The NRC has approved the proposed activity or portions thereof .

Reference :

B. Basis Provide a clear, concise basis for determining the proposed activity may be exempted such that a third-party reviewer can reach the same conclusions .

LI-101-01, Rev. 8; Effective Date : 6123105

50.59 REVIEW FORM Page 1 1 of 13 IV. 50.59 EVALUATION License Amendment Determination Does the proposed Change being evaluated represent a change to a method of evaluation M Yes Ly? If "Yes," Questions I - 7 are not applicable; answer only Question 8. lf"No,"answer L_<j No all questions below.

Does the proposed Change :

1 Result in more than a minimal increase in the frequency of occurrence of an accident EJ Yes previously evaluated in the FSAR? 0 No BASE:

The frequency of occurrence of an accident is not affected by extending the Division 11 Diesel Fuel Oil Storage Tank inspection to March 2006. There have been previous inspections of the Division 1, Division 11 and Division III fuel oil storage tanks . The inspections have resulted in discovery of minor areas of degradation of the coating of the sample probes . The most recent inspection of the Division I tank in Feburaryof 2005 resulted in no anomilies being discovered . UFSAR section 3A/1 .137 is affected by this evaluation because it is a one time extension of the Division 11 Diesel Fuel Oil Storage Tank inspection to March 2006. UFSAR section for Regulatory Guide 1 .137 on page 3A/1 .137 addresses the Fuel Oil Systems for Standby Diesel Generators . Regulatory Guide 1 .137 requires the draining of the fuel oil stored in the supply tanks, removal of accumulated sediment, and tank cleaning at a 10 year intervals . As stated above, previous inspections noted that degradation being minimal and the last inspection of Div I showed no increase, therefore an extension to Div 11 can be applied since they are subjected to the same conditions.

This part of the UFSAR will be changed to reflect the one time extension of the Diesel Fuel Oil Storage Tank inspection . The one time exception to the scheduled inspection will allow the Fuel Oil Storage Tank inspection to be extended to March 2006. The frequency of occurrence of an accident is not affected by extending the Division 11 Diesel Fuel Oil Storage Tank inspection to March 2006. As there are no indications that tank degradation is beyond the minimal amount noted previously and the design of the tank is not challenged, thus there is a very low probability that the tank will fail prior to being inspected . As tank failure is not expected, there is no increase in the frequency of occurrence of an accident previously evaluated in the FSAR by extending the inspection time until March 2006.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a El Yes structure, system, or component important to safety previously evaluated in the FSAR? 2 No BASIS:

UFSAR 3AI A 37 is affected by this evaluation because it is a one time extension of the Division 11 Fuel Oil Storage tank inspection to March 2006. UFSAR section for Regulatory Guide 1 .137 on page 3A/1 .137 addresses the Fuel oil system for Standby Diesel Generators . This part of the UFSAR will be changed to reflect the one time extension of the Diesel Fuel Oil Storage Tank inspection .

The extension is based on previous ten year inspections showing minor wear and wall degradation to the Diesel Generator Tank walls and no serious deterioration of the diesel generator fuel oil storage tanks.

The wall degradation is due to the sample element probes in the tank and this is minor. Additionally, The Division I fuel oil storage tank was inspected in Feburary of 2005 and no anomalies were noted . These inspections are documented in MNCR 108-92, MNCR 174-92, MAI 327093 and WO 00056003. The proposed activity does not adversely affect the design function of the Diesel Fuel Oil storage tank as described in the FSAR. Inspection of the tanks will still occur. The inspection schedule extension will be based on the minor wear discovered in the Diesel Generator Fuel Oil tanks from previous inspections .

Therefore, proposed activity does not result in more than a minimal increase in the likelihood of occurrence the of a malfunction of a structure, system, or component important to safety previously evaluated in FSAR.

LI-101-01, Rev . 8; Effective Date: 6123/05

50.59 REVIEW FORM Page 1 2 of 13

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

UFSAR3AMA37 is affected by this evaluation because it is a onetime extension of the Division 11 Fuel Oil Storage tank inspection to March 2006. UFSAR section for Regulatory Guide 1 .137 on page 3A/1 .137 i addresses the Fuel oil system for Standby Diesel Generators. This part of the UFSAR will be changed to reflect the one time extension of the Diesel Fuel Oil Storage Tank inspection. The one time exception to the scheduled inspection will allow the Fuel Oil Storage Tank inspection to be extended to March 2006.

Diesel failure or Diesel Fuel Oil storage tank failure are unaffected by extending The consequences of a the frequency of the tank inspection . The proposed activity does not adversely affect the design function of the Diesel Fuel Oil storage tank as described in the FSAR. Inspection of the tank will still occur. The scheduled inspection extension is based on the minor wear discovered in the Diesel Generator Fuel Oil tanks from previous inspections . As the design function is not affected, there is no increase to the chance of failure, thus there is no adverse affect to the consequences of any of the accidents previously evaluated in the FSAR.

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, El Yes system, or component important to safety previously evaluated in the FSAR? 0 No BAST:

UFSAR 3A11 .137 is affected by this evaluation because it is a one time extension of the Division 11 Fuel Oil Storage tank scheduled inspection . UFSAR section for Regulatory Guide 1 .137 on page 3A/1 .137 addresses the Fuel oil system for Standby Diesel Generators . This part of the UFSAR will be changed to reflect the one time extension of the Diesel Fuel Oil Storage Tank scheduled inspection. The one time exception to the inspection will allow the Fuel Oil Storage lank inspection to be extended to March 2006.

The consequences of a Diesel failure or Diesel Fuel Oil storage tank remained unchanged . The proposed activity does not adversely affect the design function of the Diesel Fuel Oil storage tank as described in the FSAR. Inspection of the tank will still occur. It will be extended based on the minor wear discovered in the Diesel Generator Fuel Oil tanks from previous inspections . The proposed activity does not adversely affect the consequences of component malfunction previously evaluated in the FSAR.

different type than any previously evaluated in the

5. Create a possibility for an accident of a Yes FSAR? No BAST:

The possibility of a different type of accident is not affected by extending the Diesel Fuel Oil Storage Tank inspection to March 2006. There are no new components being added to the tank and the tank is not being modified or changed . The UFSAR is affected by this evaluation because it is a one time extension of the Division 11 Fuel Oil Storage tank scheduled inspection . UFSAR section for Regulatory Guide 1 .137 on page 3A/1 .137 addresses the Fuel oil system for Standby Diesel Generators. This part of the UFSAR will be changed to reflect the one time extension of the Diesel Fuel Oil Storage Tank scheduled inspection.

The one time exception to the inspection will allow the Fuel Oil Storage Tank inspection to be extended to March 2006. This 50.59 provides a basis for the Diesel Generator Fuel Storage Tank inspection extension to March 2006 .

LI-101-01, Rev . 8; Effective Date: 6123/05

50.59 REVIEW FORM Page 13 of 13

6. Create a possibility for a malfunction of a structure, system, or component important to safety Yes with a different result than any previously evaluated in the FSAR? No BASIS:

The UFSAR is affected by this evaluation because it is a one time extension of the Division 11 Fuel Oil Storage tank scheduled inspection to March 2006. UFSAR section for Regulatory Guide 1 .137 on page 3A/1 .137 addresses the Fuel oil system for Standby Diesel Generators . This part of the UFSAR will be changed to reflect the one time extension of the Diesel Fuel Oil Storage Tank inspection to March 2006.

The extension is based on previous ten year inspections showing minor wear and wall degradation to the Diesel Generator Tank walls and no serious deterioration of the diesel generator fuel oil storage tanks.

The wall degradation is due to the sample element probes in the tank and this is minor. Additionally, The Division I fuel oil storage tank was inspected in Feburary of 2005 and no anomalies were noted. These inspections are documented in MNCR 108-92, MNCR 174-92, MAI 327093 and WO 00056003. The proposed activity does not adversely affect the design function of the Diesel Fuel Oil storage tanks as described in the FSAR. Inspection of the tanks will still occur . The inspection extension will be based on that previous inspections indicated only minor wear being discovered in the Diesel Generator Fuel Oil tanks. The proposed activity does not produce a different result for the malfunction of the Diesel Fuel Oil storage tank as described in the FSAR.

7. Result in a design basis limit for a fission product barrier as described in the FSAR being Yes exceeded or altered? No BASIS :

The UFSAR is affected by this evaluation because it is a one time extension of the Division 11 Fuel Oil Storage tank inspection to March 2006. UFSAR section for Regulatory Guide 1 .137 on page 3A/1 .137 addresses the Fuel oil system for Standby Diesel Generators . This part of the UFSAR will be changed to reflect the one time extension of the Diesel Fuel Oil Storage Tank scheduled inspection.

The extension is based on previous ten year inspections showing minor wear and wall degradation to the Diesel Generator Tank walls and no serious deterioration of the diesel generator fuel oil storage tanks .

The wall degradation is due to the sample element probes in the tank and this is minor. Additionally, The Division I fuel oil storage tank was inspected in Feburary of 2005 and no anomalies were noted. These inspections are documented in MNCR 108-92, MNCR 174-92, MAI 327093 and WO 00056003. The proposed activity does not adversely affect the design function of the Diesel Fuel Oil storage tanks as described in the FSAR. Inspection of the tanks will still occur. The scheduled inspection extension will be based on the minor wear discovered in the Diesel Generator Fuel Oil tanks from previous inspections.

There are no fission barriers affected by extending the inspection to March 2006 of the Diesel Fuel Oil storage tank as described in the FSAR .

8. Result in a departure from a method of evaluation described in the FSAR used in establishing 0 Yes the design bases or in the safety analyses? No BASIS :

The UFSAR is affected by this evaluation because it is a one time extension of the Division 11 Fuel Oil Storage tank inspection to March 2006. UFSAR section for Regulatory Guide 1 .137 on page 3A/1 .137 addresses the Fuel oil system for Standby Diesel Generators . This part of the UFSAR will be changed to reflect the one time extension of the Diesel Fuel Oil Storage Tank inspection to March 2006.

There is no change in method of inspection of the Diesel Fuel Oil Storage tank. Therefore, this does not result in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses .

If any of the above questions is checked "YES," obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure LI-113.

LI-101-01, Rev . 8; Effective Date: 6/23105

GGNS 50.59 Safety Evaluation Number SE 2006-0001-R00

50.69 REVIEW FORM Page I of 13

1. OVERVIEW I SIGNATURES Facility:

Document Reviewed : ER-2005-0197-000 Change/Rev .: 0 System Designator(s)/Description : G41 Description of Proposed Activity :

Change the decay heat analytical method used in the thermal-hydraulic analysis of the spent fuel pool and fuel pool cooling system from Branch Technical Position ASB 9-2 to the Oak Ridge isotope Generation and Depletion Code (ORIGEN V2.1). This proposed activity is a methods change as defined in 10 CFR 50.59 paragraph (a)(2) . The proposed activity does not involve any physical changes to the facility .

Check the applicable review(s) : (Only the sections indicated must be included in the Review.)

fc 1 EDITORIAL CHANGE of a Licensing Bats Document Section 1 F-1 SCREENING Sections I and 11 required 50.59 EVALUATION EXEMPTION Sections 1, 11, and III required Z I 50.59 EVALUATION {#: 2 QOAr - 0 0 0 1 - It 410 us 0&

Sections l,11, and IV required Preparer: Guy B. Spikes / ~,, 6 . ~ E 01 JN E ~ 1 /3 1Z QoC Name (print) / Signature I Company / Department / Date Reviewer : William E. Long mar ;I~OVI - 4/ - 6 (a Name (print) / Signature / Comp-anT/ Aera I Me OSRC:

Chairman's Name (print) / signature I Date

{Required only for Programmatic Exclusion Screenings and 50,59 Evaluations.)

LI-1 01 -01, Rev . 8; Effective Date: 6123105

50 .59 REVIEW FORM Page 2 of 13 II. SCREENINGS A. Licensing Basis Document Review 1 . Does the proposed activity impact the facility or a procedure as described in any of the following Licensing Basis Documents?

Operating License YES NO CHANGE # and/or SECTIONS IMPACTED Operating License E3 N TS El Ava NRC Orders 1 11 1 Z I If "YES," obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMM LI-113. (See LI-101 for exceptions.)

LBDs controlled under 50 .59 YES NO CHANGE # (if applicable) and/or SECTIONS IMPACTED FSAR Z 0 LBDC 2005-083 TS Bases El to Technical Requirements Manual Q to Core Operating Limits Report El N NRC Safety Evaluation Report and 0 041 supplements for the initial FSAR 1 NRC Safety Evaluations for El -014 amendments to the Operating License' If "YES," perform an Exemption Review per Section III OR perform a 50.69 Evaluation per Section IV 2R obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMM Lt-113 . If obtaining NRC approval, document the LBD change in Section 11 . However, the

.5

.A change cannot be implemented until approved by the NRC. Complete Section II.

LBDs controlled under other YES NO CHANGE # (if applicable) and/or regulations SECTIONS IMPACTED Program Quality Assurance Manual 2 EN] a Emergency Plan 2,3 0 to I

3' 4 Fire Protection PrograM (includes the Fire Hazards Analysis)

Offsite Dose Calculations Manual' 4 If "YES," evaluate any changes in accordance with the appropriate regulation AND Initiate an LBD change in accordance with NMM LI-113 .

' If "YES," see 1-1-101 . No LBD change is required .

2 If "YES," notify the responsible department and ensure a 50 .54 evaluation is performed. Attach the 50 .54 evaluation, 3

Changes to the Emergency Plan, Fire Protection Program, and Oftite Dose Calculation Manual must be approved by the OSRC in accordance with NMMthe OM-419 .

4 If "YES," evaluate change in accordance with the requirements of the facility's Operat nse Condition or under 50 .59, as appropriate.

LI-101-01, Rev. 8; Effective Date : 6123105

5019 REVIEW FORM Page 3 of 13

2. Does the proposed activity involve a test or experiment not described in the FSAR? F1 Yes ED No If "YES," perform a 50.59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change AND initiate an LBD change in accordance with NMM LI-113, it applicable. If obtaining NRC approval, document the change In Section II.A.S. However, the change cannot be implemented until approved by the NRC . Complete Section IL

3. Basis Explain why the proposed activity does or does not impact the Operating Ucense[Technical Specifications and/or the FSAR. If the proposed activity involves a potential test or experiment not previously described in the FSAR also include an explanation . Discuss other LBDs if impacted. Adequate basis must be provided within the Screening such that a third-party reviewer can reach the same conclusions . Simply stating that the change does not affect TS or the FSAR is not an acceptable basis.

The proposed activity involves changing the current analytical method used to calculate the decay heat from spent fuel bundles in the fuel storage pools from that described in Branch Technical Position ASB 9-2 to the Oak Ridge Isotope Generation and Depletion Code (ORI GEN V2.11). The change is applicable only to calculating spent fuel decay heat used in the thermal-hydraulic analysis of the spent fuel pool and Fuel Pool Cooling and Cleanup (FPCC) system . The proposed activity is a methods change as defined in 10 CFR 50.59 paragraph (a)(2) and does not involve any physical changes to the facility. The scope of this evaluation is limited to demonstrating that the proposed methodology change does not constitute a departure from a method of evaluation described in the FSAR in accordance with 10 CFR 50.59 paragraph (c)(2)(viii) . Implementation of this new methodology will be performed subsequent to this evaluation and associated changes to affected LBDs implemented under 10 CFR 50.59 at that time.

Operating Licenserrechnical Specifications (OUTS)

The OL/TS and TS Bases include several references to reactor core and fuel pool decay heat.

However, the methods used to calculate fuel storage pool decay heat are not described in the OLM or in any Tech Spec, LCO, or TS Bases . As such, no TS, LCO, TS Bases, surveillances or other controls in the GGNS OUTS are affected by the proposed activity.

TRM The TRM is not impacted by the proposed activity. TRM requirements do not describe the method used to calculate decay heat. As such, changing the method used to calculate fuel storage post decay heat does not affect any TRM requirements .

FSAR The methodology currently used to calculate the spent fuel pool design normal maximum and abnormal maximum decay heat loads (ASB 9-2) used in the spent fuel pool cooling (FPCC) system performance analysis is described in FSAR Section 9.1 .3.3. The resulting normal maximum and abnormal maximum decay heat loads are shown in FSAR Table 9.1-12 . The scope of this evaluation is limited to demonstrating that changing the methodology used to calculated the spent fuel pool decay heat load from ASB 9-2 to ORIGEN V2.1 (ORIGEN2) does not constitute a departure from a method of evaluation described in the FSAR . Implementation of the new methodology is not included in the scope of this evaluation. Therefore, the description of the decay heat calculation method in FSAR Section 9.1 .3.3 is revised to include the ORIGEN2 code as an approved method. However, the decay heat values in the FSAR tables are not changed.

FSAR Section 9.2.5.3 describes ASB 9-2 (APCSB 9-2) as the method used to calculate the post-accident spent fuel pool heat rate input to the Standby Service Water (SSW) Ultimate Heat Sink (UHS) capability analysis. This value is reported in FSAR Tables 9 .2-16 and 9.2-17. The proposed change to decay heat methods applies only to calculating the spent fuel pool heat load for the thermal-hydraulic analysis of the spent fuel pool and Fuel Pool Cooling and Cleanup (FPCC) system . This evaluation does not consider changing the methodology applied in the UHS analysis.

Therefore, this proposed activity does not affect FSAR Section 9.2.5.3.

LI-101 -01, Rev . 8; Effective Date: 6/23/05

60.59 REVIEW FORM Page 4 of 13 FSAR Table 12.3-2 is a list of computer codes used in radiation shielding design . The ORIGEN code is included in this table and in FSAR Section 12.3.5 (References) . This description of the ORIGEN methodology refers to an application (radiation shielding) different from that considered in this evaluation (fuel bundle decay heat) . Therefore, the proposed method change does not affect this description .

COLR Decay heat or decay heat methods are not described in the COLR. As such, the proposed activity does not impact the GGNS COLR.

NRG SERB Various NRC Safety Evaluation Reports associated with licensing the high density spent fuel storage racks (HDSFR) describe ASB 9-2 (either directly or by reference) as the method for calculating pool decay heat for the pool thermal-hydraulic analyses . These SER's include MAEC 86/0264 {interim HDSFR SERI and GNRI 92/00163 (final SER). This evaluation determines whether or not replacing ASB 90 with a new methodology constitutes a departure from a method of evaluation described in the FSAR in accordance with 10 CFR 50.59 paragraph (c)(2)(viii) . The outcome of this evaluation does not affect the descriptions in SER's previously issued by the NRC .

Test or Experiment The proposed activity changes the method of calculating fuel decay heat from ASS 9-2 to ORIGEN2 .

This change does not involve any tests or experiments.

There am no NRC orders applicable to decay heat methods . The proposed activity does not affect the FHA, ODCM, QAPM, or E-Plan .

4. Reference Discuss the methodology for performing LBD searches. State the location of relevant licensing document informa and explain the scope of the review such as electronic search criteria used (e.g., key words) or the general extent of manual searches. NOTE: Ensure that manual searches are performed using controlled copies of the documents. If you have any questions, contact your site Licensing department Electronic search method used: Keywords:

GGNS Autonomy. LBDs: OLM, FSAR, TS, TS ORIGEN, ASB 9-2, APCSB 9-2, Branch Position, Bases, TRK NRC SERs. Branch Technical Position, decay heat, spent fuel pool.

LBDs reviewed manually .

FSAR Sections 9.1 .2, 9.1.3. 9.2.1, 9.2-5, 12.3, FSAR Tables 9.1-12, 9.2-16, 9.2-17, 12.3-2.

NRC SERs MAEC 86/0264, GNRI 92/00163 .

6. Is the validity of this Review dependent on any other change?

If "YES," list the required changesisubmittals . The changes covered by this 50.59 n Yes Review cannot be implemented without approval of the other identified changes (e.g.,

license amendment request). Establish an appropriate notification mechanism to C9 No ensure this action is completed .

LI-101-01, Rev . 8 ; Effective Date: 6123105

50.59 REVIEW FORM Page 5 of 13

8. ENVIRONMENTAL SCREENING If any of the following questions is answered "yes," an Environmental Review must be performed in accordance with NMM Procedure EV-115 and attached to this 50 .59 Review. Consider both routine and non-routine (emergency) discharges when answering these questions .

Will the proposed activity being evaluated :

involve a land disturbance equal to or in excess of one acre (i.e., grading activities, construction of buildings, excavations, reforestation, creation or removal of ponds)?

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

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

4. F1 Increase the amount of thermal heat being discharged to the river or lake?

5. Q Increase the concentration or quantity of chemicals being discharged to the river, take, or air?

6. El Discharge any new or different chemicals that are currently not authorized for use by the state regulatory agency?

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

Modify the design or operation of the cooling tower that will change water or air flow characteristics?

9. D 19 Modify the design or operation of the plant that will change the path of an existing water discharge or that will result in a new water discharge?

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

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

Involve the installation or use of equipment that will result in a new or additional air emission discharge?

Involve the installation or modification of a stationary or mobile tank?'

Forl Involve the use or storage of oils or chemicals that could be directly released into the environment?

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

See NMM vroceaure cv- i i i Tor guidance in answering inis quesuun .

LI-101-01, Rev . 8; Effec Date: 6123105

50.59 REVIEW FORM Page 6 of 13 C. SECURITY PLAN SCREENING If any of the following questions swered "yes," a Security Plan Review must be performed by the Security Department to determ al impact to the Plan and the need for a change to the Plan.

Could the proposed activity being evaluated:

1. Add, delete, modify, or otherwise affect Security department responsibilities (e.g.,

including fire brigade, fire watch, and confined space rescue operations)?

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

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

Affect (block, move, or alter) security lighting by adding or deleting lights, structures, buildings, or temporary facilities?

5. [] Modify or otherwise affect the intrusion detection systems (e.g., E-fields, microwave, fiber optics)?

6. F-1 Modify or otherwise affect the operation or field of view of the security cameras?

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

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

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

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

The Security Department answers the following question if one of questions C.1 through C.10 above was answered "yes."

Is a change to the Security Plan required? 0 Yes 0 No Attach to this 50 .59 Review or reference below documentation for accepting a "yes" answer for any of Questions C .1 through C.10, above.

Name of Security Plan reviewer (print I Signature t Data LI-101-01, Rev. 8; Effective Date: 6123105

50.59 REVIEW FORM Page 7 of 13 D. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) SCREENING (NOTE : This section is not applicable to Grand Gulf or Waterford 3 and may be removed from 50.59 Reviews performed for Waterford 3 proposed activities.)

If any of the following questions is answered "YES," a 72,48 Review must be performed in accordance with NMM Procedure LI-112 and attached to this 50.59 Review .

Will the proposed activity being evaluated :

YES NO

1. C1 Any activity that directly impacts spent fuel cask storage or loading operations?

2. © Involve the ISFSI including the concrete pad, security fence, and lighting?

3. [~ Involve a change to the on-site transport equipment or path from the Fuel Building to the ISFSI?

4. [J Involve a change to the design or operation of the Fuel Building fuel bridge including setpoints and limit switches?

5. M 0 Involve a change to the Fuel Building or Control Room(s) radiation monitoring?

6. [] Involve a change to the Fuel Building pools including pool levels, cask pool gates, cooling water sources, and water chemistry?

7. [ h" Involve a change to the Fuel Building handling equipment (e.g., bridges and cask cranes, structures, load paths, lighting, auxiliary services, etc)?

8. © Involve a change to the Fuel Building electrical power that could potentially impact cask loading or storage activities?

9. [] Involve a change to the Fuel Building ventilation that could potentially impact cask loading or storage activities?

10. [] Involve a change to the ISFSI security?

11 . [1 Involve a change to off-site radiological release projections from non-ISFSI sources?

12. 0 Involve a change to spent fuel characteristics?

13. [] Redefine/change heavy load pathways?

14. [~ Involve fire and explosion protection near or in the on-site transport paths or near the ISFSI?

15 . 00 Involve a change to the loading bay or supporting components power that could potentially impact cask loading or storage activities?

16. New structures near the ISFSI?

17. [1 Modifications to any plant systems that support dry fuel storage activities?

Involve a change to the nitrogen supply, service air, demineralized water or borated water system in the Fuel Building?

LI-101-01, Rev . 8; Effective Date: 6123105

0.59 REVIEW FORM Page 8 of 13 IV. 50.59 EVALUATION License Amendment Determination Does the proposed Change being evaluated represent a change to a method of evaluation 21 Yes ONLY? If "Yes," Questions I - 7 are not applicable; answer only Question 8. If "No," answer ED No all questions below.

Does the proposed Change :

I Result in more than a minimal increase in the frequency of occurrence of an accident ED Yes previously evaluated in the FSAR? ED No BASIS :

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a ED Yes structure, system, or component important to safety previously evaluated in the FSAR? n No BASIS :

3. Result in more than a minimal increase in the consequences of an accident previously ED Yes evaluated in the FSAR? C3 No BASIS :

4, Result in more than a minimal increase in the consequences of a malfunction of a structure, 0 Yes system, or component important to safety previously evaluated in the FSAR? El No BASIS :

5, Create a possibility for an accident of a different type than any previously evaluated in the ED Yes FSAR? ED No BATS:

6. Create a possibility for a malfunction of a structure, system, or component important to safety F1 Yes with a different result than any previously evaluated in the FSAR? 0 No BASIS :

7. Result in a design basis limit for a fission product barrier as described in the FSAR being Yes exceeded or altered? No BASIS:

8. Result in a departure from a method of evaluation described in the FSAR used in establishing F1 Yes the design bases or in the safety analyses? (Z No BASIS :

The proposed change to the current fuel storage pool decay heat analytical method from Branch Technical Position ASS 9-2 to the Oak Ridge Isotope Generation and Depletion Code (ORIGEN V2.1) does NOT result in a departure of a method of evaluation . The definition of "departure from a method of evaluation .. ." provides flexibility to adopt a completely new methodology without prior NRC approval provided that the new method is approved by the NRC for the intended application. A new method is "approved by the NRC for the intended application" if it is approved for the type of analysis being conducted and the licensee satisfies the terms and conditions for its use . The NRC L1-101-01, Rev . 8; Effective Date: 6/23106

50.59 REVIEW FORM Page 9 of 13 has approved the use of ORIGEN V2.1 (ORIGEN2) for spent fuel pool applications through the issuance of Safety Evaluation Reports (SERB}. This evaluation reviews these SERB and demonstrates that the ORIGEN2 applications approved by the NRC are entirely consistent with the proposed application of the ORIGEN2 methodology at GGNS. The criteria in the NEI Guidelines for 50.59 Implementation (NEI 965-07 Rev . 1) and the Entergy 10CFR50.59 Program Guidelines (ENS-Li-101 Attachment 9.3) are also used to ensure that the important considerations for determining that the proposed application of ORIGEN2 is technically appropriate for the intended application, within the limitations of the applicable SERB, consistent with the GGNS licensing basis, and does not require NRC approval.

BACKGROUND The current ASS 9-2 and proposed ORIGEN V2.1 (ORIGEN2) methodologies are briefly described below .

ASS 9-2 models the energy release from the fission products of U-235 and heavy elements U-239 and Np-239 using a summation of exponential terms with empirical constants . ASS 9-2 is based on experimental data relating to energy release from the decay of fission products published from 1958 to 1973. It draws heavily on an ANS decay heat standard proposed in 1971 . This proposed standard was simplistic in that a single curve (fission product decay heat versus cooling time) was chosen to represent the decay heat power of uranium-fueled thermal reactors. Many phenomena that make the decay heat power unique to each case were ignored and assumed to be included within the appropriately large uncertainties that were adopted . The actual ASB 9-2 equation resulting from the curve fit is more complex than that in the proposed ANS standard; however, the results of the curve fit equations agree with each other reasonably well . In addition, the exponential terms and empirical constants for decay heat generation due to heavy elements and the uncertainty factors in ASS 9-2 were taken directly from the proposed ANS standard . ASS 9-2 acknowledges the lack of consistent experimental data and the differing results of various calculations available at the time and concludes that "...the effect of all uncertainties can be treated .. . by a suitably conservative multiplying factor."

This factor is 20% for decay times less than 103 seconds and 10% for decay times between 103 and 107 seconds. While the experimental data bases for ASS 9-2 extend to shutdown times up to 107 seconds (-118 days), the NRC Standard Review Plan for spent fuel pool cooling systems (NUREG-0800, SRP 9.1 .3) states that, for calculating the amount of heat to be removed by the spent fuel pool cooling system, ASB 9-2 can be extended to times >107 seconds. For these long-term fuel pool cooling calculations, the SRP 9.1 .3 methodology specifies an uncertainty factor of 10%. These uncertainties and the empirical constants are built into the ASS 9-2 methodology .

ORIGEN2 is a more rigorous and precise method of calculating decay heat than the empirically based ASS 9-2 methodology . ORIGEN2 explicitly models fissile material behavior during periods of irradiation and decay by computing time-dependent concentrations and source terms of a large number of isotopes which are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, and physical or chemical removal rates . ORIGEN2 was released in 1980 with the primary objective of providing a code that can perform a broad range of fuel cycle analyses with simple input specifications and a few select cross-section data libraries . ORIGEN2 and its predecessor, ORIGEN, are the most widely used computer codes for predicting the characteristics (isotopic inventory, radiation source terms and decay power) of spent nuclear fuel. The required input for ORIGEN2 consists of data relating to the specific problem to be analyzed, including fissile isotope concentrations (i.e., bundle enrichments and uranium weight), bundle power during irradiation, length of irradiation, and length of decay period . Thus ORIGEN2 provides a rigorous treatment of the decay heat calculation . This assessment is supported by the NRC, which states in Information Notice 96-39:

"ORIGEN does not use empirical methods to calculate decay heat but tracks the buildup and decay of the individual fission products within the reactor core during operation and shutdown .

ORIGEN also includes the effect of element transmutation from neutron capture, both in fissi)e isotopes and fission products. Because ORIGEN is a rigorous calculation of all decay heat inputs, it was used in the calculations for decay heat .. ."

In addition to the empirical constants and uncertainty terms discussed above, the ASS 9-2 methodology includes techniques for selecting input parameter values which ensure conservative parameter selection. The affected input parameters include bundle irradiation time, outage time, and 1-1-1011-011, Rev . 8; Effective Date: 6123106

60 .59 REVIEW FORM Page 1 0 of 13 bundle specific power. Since these techniques are described in ASS 9-2, they are considered part of the ASS 9-2 methodology . The ORIGEN2 code manual describes code inputs and formats but does not describe the method of selecting values of input parameters. The ORiGEN2 code also does not explicitly account for code biases or uncertainties .

EVALUATION The NRG has previously approved the use of ORIGEN2 for calculating fuel bundle decay heat in spent fuel pool thermal-hydraulic analyses. Three approvals, one for a PWR (V. C. Summer), one for an older BWR14 (Duane Arnold), and a more recent approval for a newer BWR/6 (Clinton Power Station), are discussed in this evaluation.

In a letter dated September 21, 2001, the NRC issued a license amendment to the Duane Arnold Energy Center (DAEC) for a revised thermal-hydraulic analysis of the spent fuel pool. In discussing the methodology for determining bundle decay heat, the Technical Evaluation Report referenced by me SER states:

"This program can perform decay heat calculations using either Branch Technical Position ASS 9-2, or the ORIGIN2 [sic] computer code. For both analyses .. . the ORIGIN2 [sic] option was used. All fuel assemblies were assumed to have been irradiated to the appropriate maximum bumup level. Based on this review, BNL [Brookhaven National Laboratory] concurs that the methodology and assumptions the licensee used to calculate the decay heat loads meet the intent of the applicable NRC guidelines."

In the SER, the NRC echoed this conclusion, stating that:

"Based on its review, the NRC staff concluded that the methodology and assumptions used by the licensee to calculate the decay heat loads and to calculate the SFP bulk temperatures met the intent of the applicable NRC guidelines."

In an SER dated August 30, 2002, the NRC issued an amendment to V. C. Summer for spent fuel pool re-racking . In discussing the analysis of the spent fuel pool decay heat removal capability, the SER states :

"The decay heat is calculated using the ORIGEN2 code assuming a 2-percent thermal power uncertainty and using the licensed thermal power at the time of discharge for historical discharges. .. .The staff performed independent calculations of decay heat load and heat exchanger performance to verify the accuracy of the analyses provided by SCE&G . The decay heat load calculations used the method described in Branch Technical Position ASS 9-

2. . .. These calculations, with consideration for the differing analytical methods and assumptions, confirmed the results provided by SCE&G were accurate ."

In a more recent application, AmerGen Energy Company (AmerGen) submitted a license amendment request (AR) to the NRC to increase the fuel storage capacity in the spent fuel pool at Clinton Power Station . The associated licensing analysis included a comprehensive thermal-hydraulic evaluation of the spent fuel pool. The calculation of long-term decay heat was performed using the ORIGEN2 code. In a subsequent Request for Additional Information (RAI), the NRC questioned the decay heat loads calculated in the licensing analysis. Specifically, the staff noted that the maximum decay heat load to the pool and the peak bulk pool temperature calculated in the licensing analysis, which included the additional fuel due to proposed fuel storage expansion, was less than the heat load and peak temperature from the existing analysis as reported in the USAR. In their response, ArnerGen stated that the licensing analysis decay heat evaluation:

". ..employs the precision computer code ORIGEN2 to compute the radioactive energy release from irradiated spent nuclear fuel. This procedure avoids the empirical methods (i.e., Branch Technical Position ASS 9-2 "Residual Decay Energy for LightWater Reactors for Long-Term Cooling") deployed in the Clinton Power Station (CPS) Updated Safety Analysis Report (USAR) that provided conservative estimates of decay heats . Although the quantity of fuel to be stored in [the] storage expansion application is increased, the calculated decay heat load and maximum bulk temperature that results from the increased quantity of spent fuel is more than offset by removal of excessive conservatisms,"

The NRC issued the requested license amendment to AmerGen Energy Company (AmerGen) on October 31, 2005. In discussing the spent fuel pool thermal-hydraulic analysis, the NRC SER states:

Lt-101-01, Rev . 8; Effective Date: 6123/05

50.59 REVIEW FORM Page 1 1 C413 "The licensee evaluated the SFP maximum bulk water temperature for this case, incorporating of the into the analysis . .. a more precise treatment deny heat generated by the spent fuel by using ORIGEN2 calculations . The staff has reviewed the licensee's submittal and finds the heat load calculation is acceptable."

Based on the above examples, the ORIGEN2 methodology has been previously approved by the NRC for the calculation of decay heat loads in spent fuel pool thermal-hydraulic applications . These applications are entirely consistent with the proposed application of the ORIGEN2 methodology at GGNS. Further, in reviewing the pool storage expansion request for Clinton Power Station, the NRC recognized that spent fuel decay heat loads calculated by ORIGEN2 are more precise and less conservative than those calculated using AS13 9-2 methods .

In addition to the previous NRC approvals discussed above, Energy Northwest evaluated and approved changing the methodology for calculating spent fuel pool bundle decay heat at the Columbia Generating Station from ASB 9-2 to ORIGEN2 and ORIGEN-ARP methods in accordance with 10 CFR 50,59 paragraph (c)(2)(viii) .

Section 4.3.8.2 of NEI 96-07 (50 .59 Implementation Guidelines) provides specific guidance for determining when changing from one method of evaluation to another is not considered a departure from a method of evaluation described in the FSAR. The use of a new NRC-approved methodology (e.g., new or upgraded computer code) to reduce uncertainty, provide more precise results or other reasons, is acceptable provided that such use is:

1 . Based on sound engineering practice. The ORIGEN methodology provides a rigorous calculation of the physical phenomenon involved in predicting the decay heat associated with irradiated spent nuclear fuel. ORIGEN2 computes time-dependent concentrations and source terms of a large number of isotopes, which are simultaneously generated or depleted through neutronic transmutation, fission, radioactive decay, and physical or chemical removal rates. As discussed above, the NRC has acknowledge that the rigorous methodology in ORIGEN2 is superior to the empirically-based ASB 9-2 methodology .

2. Appropriate for the intended application . ORIGEN2 and its predecessor, ORIGEN, are the most widely used computer codes for predicting the characteristics (isotopic inventory, radiation source terms, and decay power) of spent nuclear fuel, fissile material, and other radioactive materials.

The ORIGEN code series was developed to specifically address problems associated with out-of-reactor applications, such as the characterization of spent nuclear fuel. ORIGEN2 computes time-dependent material concentrations based on point (i.e., no spatial dependence) depletion/decay methods and is able to capture the build-up and decay of a large number of nuclides needed for this class of problem . Thus, the ORIGEN2 code is appropriate, and widely used, for calculating the physical characteristics of spent fuel, including isotopic inventory, radiation source terms, and decay heat. Like the original ORIGEN code, ORIGEN2 is designed to operate as a stand-alone calculational tool with fixed cross-section data libraries provided for several reactor models.

3. Within the limitations of the applicable SER . The NRC has previously approved the use of ORIGEN2 for calculating fuel bundle decay heat in the spent fuel pool thermal-hydraulic analyses at two BWRs (Duane Arnold, Clinton Power Station) and a PWR (V. C. Summer) . Each of these applications accounted for the existing spent fuel in the pools and the projected pool heat load based on filling the pool to the limit of storage rack capacity considering conservative bounding equilibrium fuel cycles . The ORIGEN2 code will be used at GGNS to calculate the decay heat of spent fuel stored in the spent fuel pool in order to model FPCC system performance and estimate pool temperatures . Bounding analyses of the normal maximum and abnormal maximum decay heat loads will be calculated using a combination of actual data for the existing fuel stored in the pool and projected data based on equilibrium cycle estimates . The proposed application of ORIGEN2 at GGNS is entirely consistent with the applications in the referenced SERB and is within the limitations of these SERB as discussed below.

GGNS, Clinton Power Station, and Duane Arnold are currently using advanced BWR fuel designs (e.g., GE14, ATRIUM-10). The ORIGEN2 cross-section libraries contain a file corresponding to a generic extended bumup BWR fuel assembly. This generic library file conservatively maximizes the decay heat calculated by ORIGEN2 for the advanced fuel designs used at GGNS. The generic cross-section libraries originally issued with ORIGEN2 were used in the applications of ORIGEN2 in the T/H analyses supporting the DAEC and V. C.

LI-101-01, Rev . 8; Effective Date: 6123105

60.59 REVIEW FORM of 13 Summer submittals approved by the NRC .

The license submittals for DAEC and V. C. Summer contain the relevant spent fuel parameter inputs to ORIGEN2 (e.g., bumup, cooling times) that are typical of the inputs for the fuel stored in the GGNS spent fuel pool.

Both the DAEC and V. C. Summer submittals apply a power measurement uncertainty of 2%

to the ORIGEN2 decay heat load calculations . The proposed activity to change the GGNS decay heat load methodology from ASB 9-2 to ORIGEN2 will therefore add a requirement to either directly use this 2% uncertainty factor or otherwise account for power measurement uncertainties in calculating the design basis pool decay heat loads,

4. Consistent with the facility's licensing basis and relevant industry standards. Section 9.1 .3 of the fuel pool cooling system Standard Review Plan endorses the use of the ASB 9-2 methodology for calculating the decay heat of irradiated fuel stored in the spent fuel pool. The GGNS FSAR and HDSFR licensing submittals, while discussing AS13 9-2, do not contain a specific commitment to comply with SRP 9.1 .3 and the GGNS FSAR does not contain a commitment to comply with the NRC's Standard Review Plan. In addition, there are no 10 CFR Part 50 requirements that specify the heat load methodology for the fuel pool cooling (FPCC) system. There are no GGNS Core Operating Limits Report (COLR) or Regulatory Guide commitments that specify the decay heat load method for the FPCC system . The use of ORIGEN2 for calculating spent fuel pool decay is entirely consistent with relevant industry standards . As described in this evaluation, ORIGEN2 and its predecessor, ORIGEN, are the most widely used computer codes for predicting the characteristics of spent nuclear fuel. ORIGEN2 has been applied at other plants in performing decay heat calculations similar to those proposed for GGNS. As such, application of this methodology does not require exemptions to regulations, exceptions to industry standards and guidelines, or is otherwise inconsistent with the GGNS licensing basis .

5. If a computer code is involved, has the code been installed in accordance with applicable software quality assurance requirements? The ORIGEN2 code package was procured by Entergy from the Oak Ridge Radiation Shielding Information Center (RISC). The ORIGEN2 software has been installed and verified in accordance with applicable Entergy software QA procedures . These procedures require that code installation, verification and validation be formally documented . Verification and validation is accomplished by execution of sample problems and comparison of results to those provided by the code developer. The procedures also delineate qualification requirements for users and tracking of code error notices supplied by the code developer.

6. Has the code been qualified through benchmark comparisons against test data, plant data or approved engineering analyses? The accuracy of ORIGEN2 fuel bundle decay heat predictions has been demonstrated in two benchmark studies. The first study compared ORIGEN2 decay heat predictions to those from an ANS decay heat standard (ANSI/ANS-5 .1-1978). The second study compared ORIGEN2 decay heat to measured decay heat data from three PWR spent fuel assemblies. Results from these benchmarks showed excellent agreement between the ORIGEN2 and calculated (ANS) and measured decay heat data.

7. The design and operation of the facility for which the methodology has been approved is consistent with the facility to which the methodology is to be applied . The NRC has approved applications of ORIGEN2 for calculating spent fuel pool decay heat loads for a BWR/4 (Duane Arnold), a BWR/6 (Clinton Power Station) and a PWR (V. C. Summer) . These applications use ORIGEN2 to calculate spent fuel decay heat for evaluations of spent fuel pool heat loads and temperatures assuming bounding fuel pool inventory and various fuel pool cooling system alignments . Spent fuel decay heat is a function of the specific power of the core, initial bundle enrichment, bundle exposure, operating cycle length, and cooling time. The values of these parameters in the NRC approved applications are not significantly different form the proposed application at GGNS. The proposed change in decay heat methodology does not introduce or exclude any design basis accident. This change is applicable only to calculating spent fuel decay heat to evaluate the performance of the FPCC system under design conditions. Therefore, there are no identified differences in configuration and licensing bases that impact the use of ORIGEN2 as a method for determining the spent fuel heat load.

In summary, the above evaluation demonstrates that the proposed change to the fuel storage pool decay heat analytical method from NRC 1s Branch Technit.;al Position ASB 9-2 to a more realistic (and L1-101-01, Rev . 8; Effective Date: 6/23105

$0.59 REVIEW FORM Page 13 of 13 less conservative) methodology based on the ORIGEN2 code is not a departure from a method of evaluation under 10 CFR 60.59 (c)(2)(viii) because the new method and proposed application:

Is technically appropriate for the intended application, Has been previously approved by the NRC for the intended application, The design and operation of the facilities for which the methodology has been approved (spent fuel storage pools and FPCC systems) is consistent with the proposed application .

The NRC SERs do not include any special restrictions or limitations on the use of the ORIGEN2 methodology. The application of ORIGEN2 considered in the SERs and in this evaluation is limited to calculating the decay heat of fuel bundles stored in the spent fuel pool for use in the thermal-hydraulic analysis of the spent fuel pool and fuel pool cooling system. This application will use plant-specific code inputs and appropriately account for uncertainties in core thermal power measurement.

REFERENCES The following references were used in preparing this evaluation.

1 . ER-2005-0197-000.

2. MAEC 86/0264, L. L. Kintner (USNRC) to 0. D. Kingsley, Jr., "Revision to Technical Specifications - Fuel Storage and Spent Fuel Storage Pool Temperature," dated August 18, 1986.

3. GNRI 92/00163, P. W. O'Conner (USNRC) to W. T. Coftle, "Proposed Method to Provide Augmented Spent Fuel Pool Cooling," dated July 30, 1992.

4. Branch Technical Position ASB 9-2, "Residual Decay Energy for Light-Water Reactors for Long-term Cooling," included in NUREG-0800, Standard Review Plan Section 9.2.5 "Ultimate Heat Sink," Revision 2, dated July, 1981 .

5. ORNL/TM-7175, "A User's Manual for the ORIGEN2 Computer Code," A. G. Croff, Union Carbide Corporation, Oak Ridge National Laboratory, Oak Ridge, TN, July, 1980.

6, NUREG-0800, Section 9.1 .3, Rev. 2, *Spent Fuel Pool Cooling and Cleanup System," July, 1981 .

7. NEI 96-07, Rev . 1, "Guidelines for 10 CFR 50 .59 Implementation," NEI, November, 2000.

8. ANS Proposed Standard, ANS 5.1 "Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors,* American Nuclear Society Subcommittee ANS-5, October, 1971, Revised October, 1973.

9. NRC Information Notice 96-39, "Estimates of Decay Heat using ANS 5.1 Decay Heat Standard May Vary Significantly," USNRC, July 5, 1996.

10. B. L, Mozafari (USNRC) to G . V. Middlesworth (DAEC), "Duane Arnold Energy Center-Issuance of Amendment for Revised Thermal-Hydraulic Analysis for Spent Fuel Pool (TAC No. MB0596)"

dated September 21, 2001 (ML012500246) .

11 . K. R. Cotton (USNRC) to S. A. Byrne (SCE&G), "Virgil C. Summer Nuclear Station, Unit 1-Issuance of Amendment Re: Spent Fuel Pool Expansion (TAC No. MB2475)," dated August 30, 2002 (ML022330203) .

12. K. R. Jury (AmerGen) to USNRC, "Additional Information Supporting the Request for License Amendment Related to Onsite Spent Fuel Storage Expansion," dated June 14, 20()5 (ML051730431) .

13. K. N, Jabbour (USNRC) to C. M. Crane (AmerGen), "Clinton Power Station, Unit 1-Issuance of Amendment - Re: Onsite Spent Fuel Storage Expansion (TAC No. MC4202)," dated October 31, 2005 (ML053070593) .

14. Croff, Allen G, "ORIGEN2: A Versatile Computer Code for Calculating the Nuclide Compositions and Characteristics of Nuclear Materials," Nuclear Technology, Vol. 62, September, 1933,

15. 10CFR50 .59 Evaluation 5059-05-0001, Rev. 0, "Methodology Change to Use ORIGEN-ARP to Calculate Decay Heat of Spent Fuel Stored in the Fuel Pool," Energy Northwest, April 6, 2005.

16. CPDP X-95/0002, OR)GEN2 Ver. 2.1 Computer Program Documentation Package .

17. CDP QR-016-26 .01, -ORIGEN2.IR .POG, Isotopic Generation and Depletion Code for the IBM RISC Computer."

LI-101-01, Rev . 8; Effective Date: 6/23105

GGNS 50.59 Safety Evaluation Number SE 2006-0002-R00

60.59 REVIEW FORM Page I of 8

1. OVERVIEW I SIGNATURES Facility: Grand Gulf Nuclear Station Document Reviewed: LDC 2006-002 Chang V.: -

System Designator(s)/Description :

TURBINE OVERSPEED PROTECTION SYSTEM - TRM 6.3.8 Description of Proposed Activity :

This change involves a relaxation of Technical Requirements Manual (TRM) 6.3.8, Turbine Overspeed Protection System required actions and completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore operability of inoperable stop or control valves. To allow this change, when a stop or control valve is inoperable, detail is added to TRM Bases 6.3.8 to require an evaluation to ensure the validity of the assumptions to the turbine missile discussion in UFSAR Section 3.5.1 .3. The following changes are being made.

TRU 6.3.8 - Relaxation of 72 Hours Completion Time The change evaluated involves modification of the TRM LGO 6.3.8 Required Action A.1 and A.2 which required a restoration of inoperable stop or control valves to OPERABLE status or close one valve in the affected steam line - either action had a Completion, Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time is changed to "immediately" and the required action to restore operability or close one valve in the affected steam line is changed to enter the actions of TRM 6.0.1 . TRM 6.0.1 requires the following.

1, Develop and implement compensatory actions as needed .

2. Verify that a required safety function is not compromised by the inoperabilities .

3. Develop a plan for exiting LCO 6.0.1.

4. Obtain Duty Manager approval of the compensatory actions and a plan for exiting LCO &0.1 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

TRM Saws 118 - Addition of detail to Support Relaxation This change also involves addition of detail to TRM Bases B 6.3.8 to support the relaxation to TRM 6.3.8. The new detail added to TRM Bases 6.3.8 will now require a review of the Turbine Missile discussion in UFSAR section 3.5.1 .3 for affect on the probability analysis to ensure risk is appropriately addressed should a stop or control valve become inoperable .

BACKGROUND UCENSING BASIS INFORMATION The Turbine Overspeed Protection System previously resided in the Grand Gulf Technical Specifications and was relocated to the TRM via Technical Specification Amendment 120. The basis for relocation was an amendment application (GNRO93-00109, Enclosure 2, Section 3.3 page 167) as follows :

The turbineTheoverspeed protection system is not considered in any design basis accident or transient. system is used to prevent overspeed which may result in the generation of missiles which could impact safety related equipment . However, the system performs no functions to mitigate the effects of the subsequent transient. Further, the evaluation summarized in NEDO-31466 determined the loss of this instrumentation to be a non-significant risk contributor to core damage frequency and offsite release . Therefore, the requirements specified for this function did not satisfy the NRC Interim Policy Statement technical specification screening criteria as documented in the application of Selection Crite to the GGNS TS and have been relocated to plant documents controlled in accordance with 10CFR 50.59.

LI-101-01, Rev . 8; Effective mate : 6123105

50 .59 REVIEW FORM

-e INFORMATION The NRC then issued Technical Specification Amendment 120(GNR195-00044) allowing relocation of the Turbine Overspeed Protection system to the TRM. The Basis for approval is as follows:

The existing TS X4 .3 .9 conditions, RAs, and SRs for the turbine overspeed protection system instrumentation have been relocated . to other plant documents. The turbine overspeed protection system instrumentation is not considered to prevent or mitigate any design b accident or transient. Although the design basis accidents and transients include a variety of system failures and conditions which might result from turbine missiles striking various plant systems and equipment, the system failures and plant conditions could be caused by other events as well as turbine failures . In view of the low likelihood of turbine missiles, this scenario does not constitute a part of the primary success path to prevent or mitigate such design basis accidents and transients . Similarly, the turbine overspeed control is not part of an initial condition of a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The requirements associated with these instrumentation functions will be relocated to the UFSAR and will be controlled in accordance with 10 CFR 50 .59.

Conclusion A thorough search of Licensing Basis and Commitment documents did not reveal a documented basis for the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time listed in TRM 6.3.8 . Replacement of the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time to isolate the affected steam line or restore operability with an allowance to enter TRM 6.0 .1 will impose a risk based approach to any inoperability. The requirements in TRM 6.0.1 require definitive action in regard to safety function, therefore there is no impact on safety. The addition of detail to TRM Bases 6.3 .8 to require a review of the turbine missile analysis will help avoid an error trap by referencing UFSAR 3.5 .1 .3 .

The underlying basis for the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowance as specified is not documented and no basis can be found. Therefore this change is evaluated via 10CFR50.59 process. Based on responses to Section IV 50 .59 questions this change is acceptable .

Check the applicable reviews): (C?nly the sections indicated must be included in the Review .)

EDITORIAL CHANGE of a Licensing Basis Document Section I

© SCREENING Sections I and 11 required 0 50.59 EVALUATION EXEMPTION Sections 1, It, and If required

~ I 50.59 EVALUATION (#: f ~0 Sections 1, Il, and IV required 1 /V7 1, Preparer: NIEOIINSA PLANT LICENSING e (print) I Signature I Company / Department 1 Date Reviewer: P lI Name (print) t Sig OSRG :

Chairman's Name (print) 7 Signature / Date (Required only for Programmatic Exclusion Seree and 50.59 Evaluations .)

L1-101-01, Rev. 8; Effective Date : 6123/05

.59 REVIEW FORM Page 3 o tl. SCREENINGS A. Licensing Basis Document Review 1 . Does the proposed activity impact the facility or a procedure as described in any of the following Licensing Basis Documents?

Operating License YES N CHANGE # and/or SECTIONS IMPACTED Operating License 1:1 1 TS NRC Orders 0 O i l

If "YES," obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMM L1413 . (See LI-101 for exceptions.)

LBDs controlled under 50.59 YES NO CHANGE # (If applicable) and/or SECTIONS IMPACTED FSAR TS Bases Technical Requirements Manual N En LDCZX6002 Core Operating Limits Report El NRC Safety Evaluation Report and supplements for the initial FSAR' NRC Safety Evaluations for 0 1100 amendments to the Operating License' If "YES," perform an Exemption Review per Section III QR perform a 50.59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMM LI-113. If obtaining NRC approval, document the LBD change in Section ILA .6. However, the change cannot be implemented until approved by the NRC. Complete Section 11.

L Ds controlled under other YES NO CHANGE # (if applicable) andfor regulations SECTIONS IMPACTED Quality Assurance Program Manuae [1 Emergency Plan" Fire Protection Program3' 4 )D (includes the Fire Hazards Analysis)

Offqe Dw ,~Ge Calculations ManUae ' 4 If "YES," evaluate any changes in accordance with the appropriate regulation AND initiate an LBD change in accordance with NM1M LI-1 13.

,Af'YES, - seeLl-101. No LBO change is required.

' If - YES." notify the responsible department and ensure a 50.54 evaluation is perforarted. Attach the 50.54 ev2iuation 3

Changes to the Emergency Plan . Fire Protection Program, and Offsite Elose Calculation Manual must be approved by the OSRC in acccrdano. -with NUM OWAt G_

if -YES,'evaluate the change in the requiroanaris oftl,,e facifily's Cpeiadng Uccnsa Condivor, or under 50_59. 3s appropriate.

LI-101-01, Rev. 8 ; Effective Date: 612310-5

50 .59 REVIEW FORM

2. Does the proposed activity involve a test or experiment not described in the FSAR? El Yes 0 No If "YES," perform a 50 .59 Evaluation per Section IV all obtain NRC approval prior to implementing the change AND initiate an LSD change in accordance with NMM LI-413, if applicable . If obtaining NRC approval, document the change in Section II.A.5. However, the change cannot be implemented until approved by the NRC. Complete Section 11.

3.

Explain why the proposed activity does or does not impact the Operating Licenserrechnical Specifications and/or the FSAR. If the proposed activity involves a potential test or experiment not previously described in the FSAR also include an explanation. Discuss other U3Ds if impacted. Adequate basis must be provided within the Screening such that a third-party reviewer can reach the same conclusions. Simply stating that the change does not affect TS or the FSAR is not an acceptable basis .

This change involves relaxation of required action and completion times as described in TRM Section 6.3.8 . The changes are administrative in nature and no new tests or experiments are imposed as a result of these changes. The requirement to ensure UFSAR Section 3.5 .1 .3 remains valid will require a risk based approach is taken when evaluating turbine stop or control valve inoperabilities. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> requirement has no written technical basis_ The affected TRM was previously located in the Technical Specifications . The NRC allowed relocation in Technical Specification Amendment 120; therefore there is no impact on the Operating License or Technical Specifications .

4. References Discuss the methodology for performing t_BD searches . State the location of relevant licensing document information and explain the scope of the review such as electronic search criteria used (e .g ., key wads) or the general extent of manual searches. NOTE: Ensure that manual searches are performed using controlled copies of the documents. If you have any questions, contact your site Licensing department.

Electronic search method used : Keywords:

All of the documents in Section ILA.1 TURBINE, MISSILE LBDs reviewed manually :

5. is the validity of this Review dependent on any other change? 0 Yes 0 No If "YES," list the required changeslsubmittals. The changes covered by this 54 .59 Review cannot be implemented without approval of the otter identified changes (e.g ., license amendment request). Establish an appropriate notification mechanism to ensure this action is completed.

LI-141-01, Rev. 8; Effective Date : 6123106

Page 5 of 8 B. ENVIRONMENTAL SCREENING If any of the following questions is answered "yes," an Environmental Review must be performed in accordance with NMM Procedure EV-115 and attached to this 50.59 Review. Consider both routine and non-routine (emergency) discharges when answering these questions .

Will the proposed activity being evaluated :

YES NO C/ Involve a land disturbance equal to or in excess of one acre (i.e., grading activities, construction of buildings, excavations, reforestation, creation or removal of ponds)?

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

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

4. [] Increase the amount of thermal heat being discharges! to the river or lake?

5. [l [9 increase the concentration or quantity of chemicals being discharged to the river, take, or air?

6. 0 Discharge any new or different chemicals that are currently not authorized for use by the state regulatory agency?

?, 11 Change the design or operation of the intake or discharge structures?

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

9. [] g( Modify the design or operation of the plant that will change the path of an existing water discharge or that will result in a new water discharge?

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

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

12. El Involve the installation or use of equipment that will result in a new or additional air emission discharge?

13. [l E Involve the installation or modification of a stationary or mobile tank?'

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

15. 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 for guidance i; srFering this question .

t.i-101-01, Rev. 8; Effective Date: 6123105

Page 6 of 8 C. SECURITY PLAN SCREENING If any of the following questions is answered "yes," a Security Plan Review must be performed by the Security Department to determine actual impact to the Plan and the need for a change to the Plan.

Could the proposed activity being evaluated :

YES NO

1. 11 i4 Add delete, modify, or otherwise affect Security department responsibilities (e.g.,

including fire brigade, fire watch, and confined space rescue operations)?

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

3. (1 02 Cause materials or equipment to be placed or installed within the Security Isolation Zone?

4. Affect (block, move, or after) security lighting by adding or deleting lights, structures, buildings, or temporary facilities?

S. (l Modify or otherwise affect the intrusion detection systems (e.g., E-fields, microwave, fiber optics)?

s. G Modify or otherwise affect the operation or field of view of the security cameras?

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

8. © Modify or otherwise affect primary or secondary power supplies to access control equipment, intrusion detection equipment, other security equipment, or to the Central Alarm Station or the Secondary Alarm Station?

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

10. F1 Modify or otherwise affect the facility's telephone o security radio systems?

The Security Department answers the following question if one of questions C.1 through C.10 above was answered "yes."

Is a change to the Security Plan required? [1 Yes Q No Attach to this 50,59 Review or reference below documentation for accepting a "yes" answer for any of Questions C.1 through C.10, above.

Name of Security Plan reviewer (print/ Signature 1 Data LI-101-01, Rev. 8; Effective Date: 6123105

58.59 REVIEW FORM Page 7of8 IV. 50.59 EVALUATION License Amendment Determination Does the proposed Change being evaluated represent a change to a method of evaluation "No," answer 0 Yes ONLY ? If "Yes," Questions 1-7 are not applicable ; answer only Question 8. If No all questions below.

Does the proposed Change :

1. Result in more than a minimal increase in the frequency of occurrence of an accident [l Yes previously evaluated in the FSAR? ED No BASIS:

The turbine overspeed protection system is not considered in any design basis accident or transient. The system is used to prevent overspeed of the turbine which may result in the generation of missiles which could impact safety related equipment. Turbine failure and resulting missile damage is not an evaluated accident nor is it an initiator to any accident described in the UFSAR.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a [~ Yes structure, system, or component important to safety previously evaluated in the FSAR? No BASIS:

Turbine failure and resulting missile damage to structures, systems, and components important to safety has been evaluated in UFSAR section 3.5 .1 .3 . The new detail added to TRM Bases 6.3.8 will now require a review of the Turbine Missile discussion in UFSAR section 3-5.1-3 for affect on the probability analysis to ensure risk is appropriately addressed should a stop or control valve become inoperable . This addition will ensure there is no minimal increase in the likelihood of a malfunction .

3. Result in more than a minimal increase in the consequences of an accident previously [] Yes evaluated in the FSAR? No BASIS:

Since the turbine overspeed protection system is not considered in any design basis accident or transient there will not be any increase in any consequences of an accident previously evaluated.

LI-101-£i1, Rev . 8; Effective Date : 6123105

60 .69 REVIEW FORM Page 8 of 8

4. Result in more than a minimal increase in the consequences of a malfunction of a structure, [l Yes system, or component important to safety previously evaluated in the FSAR? No BASIS:

The wording addition to TRM Bases 6.3.8 will impose a new requirement to verify the continued validity of the turbine missile analysis described in UFSAR 3.5.1 .3. This will ensure the probability of a turbine overspeed event and any associated missile damage which could possibly cause radiation dose release caused by damage to a structure, system, or component is bounded by the UFSAR analysis . Therefore the consequences of a malfunction previously evaluated in the UFSAR is not increased .

5. Create a possibility for an accident of a different type than any previously evaluated in the Yes FSAR? No BASIS:

Based on review of Chapter 15 of the UFSAR, turbine overspeed and subsequent turbine missiles is not an analyzed accident. The turbine overspeed system minimizes the probability of damage occurring to any safety related structure as discussed in UFSAR Section 3.5.1 .3, "Probability Analysis for High Trajectory Missiles .' No new accidents are created as a result of this change since any inoperable turbine stop or control valve will require a validity check of the assumptions of the turbine missile analysis discussed in UFSAR Section 3.5.1 .3. Therefore, no new accidents of a different type are introduced.

6. Create a possibility for a malfunction of a structure, system, or component important to safety © Yes with a different result than any previously evaluated in the FSAR? No BASIS:

With the new requirements imposed in TRM Bases 6.3.8, there will not be a malfunction of any structure, system, or component as long as a stop or control valve inoperability is evaluated against the turbine missile assumptions described in UFSAR section 3.5.1 .3. Evaluatio n against the assumptions specified in UFSAR 3.5.1 .3 ensures there is not a different result than previously evaluated .

Result in a design basis fission product barrier as described in the FSAR being Q Yes exceeded or altered? No BASIS:

This change does not affect fuel cladding, reactor coolant system boundaries, or containment since the probability of damage to a structure is kept within allowable values. The missile analysis discussed in UFSAR Section 3 .5.1 .3 discusses probability of damage to containment ;

however with TRM 6.0.1 controls imposed by this change and addition of wording to the TRM Bases 6.3.8 there will be no affect on containment . Operation within the specified probability of damage values speed in UFSAR Section 3.5.1 .3 ensures there is no affect on a design basis limit for fission product barriers.

8. Result in a departure from a method of evaluation described in the FSAR used in establishing (l Yes the design bases or in the safety analyses? ED No BASIS:

The proposed change does not change any analysis or methods used for event evaluation described in the FSAR. Therefore this change does not depart from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses .

If any of the above questions is checked "YES," obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NNIM Procedure L#-113 .

1.1-101-01, Rev. 8; Effective Date : 8123105

GGNS 50.59 Safety Evaluation Number SE 2006-0003-R00

50 .59forERNO . 2005-0110-00-00 Page I of 20 Boo Q L OVERVIEW / SIGNATURES Facility : Grand Gulf Nuclear Station Document Reviewed : ER-GGN-2005-0110-00-00, Deletion of Low Control Air Pressure Pip during a LOCA Change/Rev.:

System Designator(s)/Description: P75 - Standby Diesel Generator System Description of Proposed Change:

ER GG-2005-01 10 requested that Engineering evaluate moving the low control air pressure (< 40 prig} sensor from the Diesel Generator control LOCA logic. This is to make the diesel more reliable CAIduring post LOCAbeing conditions if a loss of control air should occur. The associated Low Lube Pressure Trip A relocated from the Emergency Mode logic over to the Normal Mode logic. This removes the reliance on operator actions and non-safety related equipment during Emergency Diesel mode of operation. The action would have been to replenish air. The sensing of low control air pressure and a Low Lube CAI Pressure trip will also be available during a Diesel Start in Normal Mode. The Low Lube Oil Pressure trip and High Crankcase pressure trip are going from 2 out of 3 transmitters being used to a single sensor.

Check the applicable review(s): (only the sections indicated must be included in the Review .)

EDITORIAL CHANGE of a Licensing Basis Document Section I SCREENING Sections I and 11 required 50.69 EVALUATION EXEMPTION Sections 1, 11, and III required 50.59 EVALUATION (#: 2606-6003-ROO Sections 1, 11, and IV required Preparer: Robert W. Fuller I r4taAQW f%0J_P,/ E01 I Design Eng - Meth 26 - 6 (a Name (print) / Signature / Company / Department I Date Reviewer: Jul / E01 / Design Engin ering Name (print) / 8`ignature, / Company / D rtment / Date OSRQ ,~ .4 4"An LQ3 TDR - L-~--N 11~ fv~

Chairman's Name (print) / Signature / Date

[Required only for ProgrammaticExclosion Screenings and 50M Evaluations .]

50 .59forERNO. 2005-0110-00-00 Page 2 of 20 Rev. -A

11. SCREENINGS A. LkertsMv BasisDocument Reuiew

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

Operating License YES NO CHANGE 4 and/or SECTIONS "ACTED Operating License El 041 TS F] GNR02005-00016,TSTIF400,GNRI-2006-0006 NRC Orders El 0 If "YES", obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMM ENS-1-1-1 13. (See Section 6.2[13] for exceptions.)

LBDs controlled under 50.59 YES NO CHANGE # (if applicable) and/or SECTIONS IMPACTED WR to F] LDC-2005-081 TS Bases [My n LDC-200=81 Technical Requirements Manual EJ QN Core Operating Limits Report F]

NRC Safety Evaluation Report El 0141 and supplements for the initial FSAR1 NRC Safety Evaluations for [~ 1f amendments to the Operating License' If "YES", perform an Exemption Review per Section III OR perform a 50.59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change. If obtaining NRC approval, document the LBD change in Section II.A.5; no further 50.59 review is required .

However, the change cannot be implemented until approved by the NRC. AND initiate an LBD change in accordance with NMM ENS-1-1-1113 .

GE 4 (if applicable) and/or SECTIONS WACTED

' lf"YES,' see Section 5.215). No LBD change is required .

50.59 for ER NO. 2005-0110-00-00 Page 3 of 20 Rev . A Quality Assurance Program Manual2 Emergency Plan2,3 Fire Protection PrograM3,4 (includes the Fire Hazards Analysis)

Offsite Dose Calculations El M Manua1 3.4 I If "YES,", evaluate any changes in accordance with the appropriate regulation AND initiate an 1-131) change in accordance with NMM ENS-1-11-1113 . No further 50.69 review is required .

2 VYES," notify the responsible department and ensure a 50.54 Evaluation is performed APO the 50.54 Review.

Changes to the Emergency Plan, Fire Protection Program, and Offsfte Dose Calculation Manual must 6e approve(

3 the OSRC in accordance with NMMOM-1 19 .

If 'YES," evaluate the change in accordance with the requirements of the facility's Operating License Condition or under 50.59, as appropriate .

50.59forERNO. 2005-0110-00-00 Page 4 of 20 Rev. .9

2. Does the proposed activity involve a test or experiment not described in the FSAR?

If "yes," perform a 5159 Evaluation per Section IV OR obtain NRC approval prior to implementing the change AND initiate an LBD change in accordance with NMM LI-I 13.

If obtaining NRC approval, document the change in Section II.A.5; no further 50.69 review is required . However, the change cannot be implemented until approved by the NRC.

3. Basis

50 .59 for ERNO. 2005-0110-00-00 Page 5 of 20 Rev. 0 lain why the proposed activity does or does not impact the Operating License/Technical Specifications and/or the FSAR and why activity does or does not involve a new test or experiment not previously described in the FSAR . Discuss other LBDs if impacted . Adequate basis must be provided within the Screening such that a third-party reviewer can reach the same conclusions.

Simply stating that the change does not affect TS or the FSAR is not an acceptable basis.

ER GG-2005-0110 requested that Engineering evaluate moving the low control air pressure (< 40 psig) sensor from the Diesel Generator Start in Emergency Mode logic to a start in Normal Mode.

This is to make the diesel more reliable during post LOCA conditions if a loss of control air should occur. This removes the reliance on operator actions and non-safety related equipment. The required action would have been to replenish air.

Operating License:

The Grand Gulf Nuclear Station (GGNS) Operating License (OL) does discuss the reliability of the diesel generators (#25), but it deals with installing a turbo charger on the diesels and not control air. The Operating License, and the Environmental Protection Plan are not impacted by ER GG-2005-0110 . Therefore, the proposed activity does not impact the GGNS operating license.

Technical Specifications :

The scope of this ER does affect the Technical Specifications . Technical Specification SR 3.8.3.4 deals with starting air which supplies control air. Modifying the control air logic will not impact or impede starting air and it's Technical Specification requirements . Technical Specification 3 .8.1.13 lists a Low Lube Oil Pressure trip but this critical trip will be relocated as a non-critical trip. An NRC evaluation per TSTIF 400 addresses the removal of the surveillance requirement.

This 50.59 evaluates relocating the Low Lube Oil Pressure as a critical trip to a non-critical trip.

Relocating the Low Lube Oil Pressure Critical trip to a non-critical trip removes the signal from the Emergency Mode of the Shutdown Logic to the Normal Mode of the Shutdown logic . This removes the two out of three signal requirement (Reg Guide 1 .9) for the Low Lube Oil pressure .

The results and conclusions do not adversely affect the mode of operation of any important to safety equipment or Technical Specification associated equipment. In addition, the moving the low control air pressure sensor from the Emergency Start logic to the Normal Start logic does not create a system configuration or operating condition such that a Technical Specifications LCO or surveillance requirement is no longer adequate . Likewise, ER GG-2005-0110 will not bypass or invalidate automatic actuation features required to be operable by the Technical Specifications or exceed any limits specified in the Operating License and Technical Specifications . There is a Technical Specifications change is required for the issuance of this ER. It removes the surveillance requirement from Technical Specification 3 .8.1 .13.

50 .59 for ER NO. 2005-0110-00-00 Page 6 of 20 Rev. 0 UFSAR:

UFSAR sections 8 .3 .1.1 .4.1 f(2)(f), 8.3.1 .2 .1 b 5(g) and FSAR Figure 8.3-008 are affected by this ER response . The requirement for the low control air pressure sensor is not needed post-LOCA (Emergency Start) . It serves no safety related function and removing it will increase the reliability of the diesel during Emergency Mode Operation. The lobe oil trip is moved from the Diesel in Emergency Mode to Normal Mode. This makes the Low Lube Oil Pressure trip non-critical/ FSAR Figure 8.3-008 will be updated to show this. An NRC evaluation per TSTIF 400 evaluates deleting the requirements for verifying the trip surveillance . A survey of the diesel owner's group discovered that many do not have the Lube Oil Pressure Low trip for the Emergency Mode of a Diesel start. Moving the Low Lube Oil Pressure Trip during Emergency mode to Normal mode will improve the reliability of the diesel operation post LOCA.

NRC Orders :

The NRC Orders issued at Grand Gulf are not affected by this ER because it deals with moving the low control air pressure sensor (<40 psig) and associated low lube oil pressure trip and its affect on Diesel operation for a Diesel Start in Emergency Mode and ER-GG-2005-0110 is not to be used for security reasons which is what Grand Gulf's current NRC Orders deal with.

Technical Specification Bases:

The Technical Specifications Bases are impacted by this activity. LDC-2005-081 is issued identifying that the Division I and li Low Lube Oil Pressure trip are non-critical trips.

Technical Requirements Manual (TRM):

There are no impacts to the Technical Requirements Manual affected by this activity .

Core Operating Limits Report :

This activity does not impact the COLR (GGNS Core Operating Limits Report). ER-GG-2005-0110 evaluates the acceptability of moving the diesel control air low pressure sensor (<40 psig) and associated low Tube oil pressure trip from the Emergency Mode to the Normal Mode . It does not have any impact on the COLR and does not affect any licensing activities.

Offsite Dose Calculations Manual:

This activity does not impact any equipment required to monitor offsite dose. Therefore, no changes to the ODCM is required .

NRC Safety Evaluation Reports:

There is no impact to any SERs for evaluating deleting the diesel control air low pressure trip (<

40 psig) and its affect on Diesel operation and operability. However, SER Supplement 7 documents that a Low Lube Oil Pressure Trip is present during a LOCA. This is being deleted from the FSAR and Technical Specification Bases. An NRC evaluation per TSTIF 400 evaluated the removal from Technical Specification 3 .8.1 .13.

50 .59 for ER NO. 2005-0110-00-00 Page 7 of 20 Rev. .9 Quality Assurance Program Manual:

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

Emergency Plan :

This ER does not e interaction of GGNS personnel and offsite agencies in response to an emergency .

Security Plan :

This ER does not impact the Security Plan since it does not require the breaching of security Fire Protection Program (includes the Fire Hazards Analysis):

This calculation does not impact the Fire Protection Program.

4. References Discuss the methodology for performing LBD searches. State the location of relevant licensing document 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 .5 .1[5](d) of LI-101 . NOTE : Ensure that manual searches are performed using controlled copies of the documents. If you have any questions, contact your site Licensing department.

LBD&Clocumerlts reviewed via keyword Keywords :

search :

Control Air, Low Lube Oil Pressure Trip Operating License, UFSAR, Technical Specification, TRM, NRC Orders, Technical Specification Bases, Technical Requirements Manual, Core Operating Limits Report, NRC Safety Evaluation Reports, QAPM, Emergency Plan, Security Plan, Fire Protection Program LBDs/Documents reviewed manually :

UFSAR sectionsS.3.1 .1 .4 .1, 8.3.1 .2.1 and TRM Bases 3.8.1 .14.

5. Is the validity of this Review dependent on any other change? Z yes EEI No

50 .59 for ER NO. 2005-0110-00-00 Page 8 of 20 Rev. 0 If "YES", list the required changes/submittals. The changes covered by this 50 .59 Review cannot be implemented without approval of the other identified changes (e.g., license amendment request). Establish an appropriate notification mechanism to ensure this action is completed.

(List the required chan~-es 1 submittals GNRO 2005-00016. TSTIF 400

50 .59 for ER NO . 2005-0110-00-00 Page 9 of 20 Rev. 0 B. ENVIRONMENTAL SCREENING If any of the following questions is answered "yes," an Environmental Review must be performed in accordance with NMM Procedure ENS-EV-115, "Environmental Evaluations," and attached to this 50.59 Review. Consider both routine and non-routine (emergency) discharges when answering these questions.

Will the proposed Change being evaluated:

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 of ponds)?

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

involve dredging activities in a lake, river, pond, or stream?

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

F4 Increase the concentration or quantity of chemicals being discharged to the river, lake, or air?

Discharge any chemicals new or different from that previously discharged?

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

Modify the design or operation of the cooling tower that will change water or air flow characteristics?

9. [:1 Modify the design or operation of the plant that will change the path of an existing water discharge or that will result in a new water discharge?

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

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

Involve the installation or use of equipment that will result in a new or additional air emission discharge?

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

50 .59 for ER No. 2005-0110-00-00 Page 10 of 20 Rev. 0

13. [] Involve the installation or modification of a stationary or mobile tank?

Involve the use or storage of oils or chemicals that could be directly released into the environment?

15. F1 Involve burial or placement of any solid wastes in the site area that may affect runoff, surface water, or groundwater?

50 .59 for ER NO . 2005-0110-00-00 Page 11 of 20 Rev. 0 C. SECURITY PLAN SCREENING If any of the following questions is answered "yes," a Security Plan Review . must be performed by the Security Department to determine actual impact to the Plan and the need for a change to the Plan.

Could the proposed activity being evaluated ;

Yes

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

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

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

4. C} Affect (block, move, or alter) security lighting by adding or deleting lights, structures, buildings, or temporary facilities?

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

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

7. [ :1 Modify or otherwise affect (block, move, or alter) installed access control equipment, intrusion detection equipment, or other security equipment?

8. F] Modify or otherwise affect primary or secondary power supplies to access control equipment, intrusion detection equipment, other security equipment, or to the Central Alarm Station or the Secondary Alarm Station?

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

1,541 Modify or otherwise affect the facility's telephone or security radio systems?

Documentation for accepting any'" statement for these reviews will be attached to this 50.59 Review or referenced below.

50.59 for ER NO. 2005-0110-00-00 Page 12 of 20 Rev . 0 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) SCREENING (NOTE- This section is not applicable to Waterford 3 and may be removed from 50.59 Reviews performed for Waterford 3 proposed activities.)

If any of the following questions is answered "yes," an ISFSI Review must be performed in accordance with NMM Procedure ENS-L1-112, "72,48 Review," and attached to this Review.

Will the proposed Change being evaluated :

Yes No

1. Any activity that directly impacts spent fuel cask storage or loading operations?

2. M Involve the Independent Spent Fuel Storage Installation (ISFSI) including the concrete pad, security fence, and lighting?

3. Fj involve a change to the on-site transport equipment or path from the Fuel Building to the ISFSI?

4. [~ Involve a change to the design or operation of the Fuel Building fuel bridge including setpoints and limit switches?

5. 0 ;,1 Involve a change to the Fuel Building or Control Room(s) radiation monitoring?

6. F-1 Z Involve a change to the Fuel Building pools including pool levels, cask pool gates, cooling water sources, and water chemistry?

7. [l Involve a change to the Fuel Building handling equipment (e.g., bridges and cask cranes, structures, load paths, lighting, auxiliary services, etc)?

8. [~ Involve a change to the Fuel Building electrical power?

9. [~ Involve a change to the Fuel Building ventilation?

© involve a change to the ISFSI security?

10 .

11 . Involve a change to off-site radiological release projections from non-ISFSI sources?

12. o Involve a change to spent fuel characteristics?

13 . a Redefine/change heavy load pathways?

50 .59 for ER NO. 2005-0110-00-00 Page 13 of 20 Rev. 0

14. 0 Fire and explosion protection near or in the on-site transport paths or near the ISFSI?

15. US Involve a change to the loading bay or supporting components?

16. a IAS New structures near the ISFSI?

17. F~ Modifications to any plant systems that support dry fuel storage activities?

18 . o 63. Involve a change to the nitrogen supply, service air, demineralized water or borated water system in the Fuel Building?

50 .59 for ER NO. 2005-0110-00-00 Page 14 of 20 Rev. -2 IV. 50 .59 EVALUATION License Amendment Determination Does the proposed Change being evaluated represent a change to a method of evaluation Yes answer only Question it If "No,"

ONLY? If "Yes," Questions I - ?are not applicable; NO answer all questions below.

Does the proposed Change:

1. Result in more than a minimal increase in the frequency of occurrence of an Yes accident previously evaluated in the FSAR? NO

50 .54 for ER NO. 2005-0110-00-00 Page 15 of 20 Rev. 0 BASIS:

This modification relocates the pneumatic Division I and II Low Control Air pressure sensor and associated Low Lube Oil Pressure trip from the Emergency Shutdown Logic to the Normal Shutdown logic. The tubing and fittings are designed and installed to the same requirements as the interfacing pressure boundaries in accordance with J-702.0, Analysis for installing Tubing (Piping Input) . These requirements include ASME Code Section 111, Code Class 2, Safety Class 2, Seismic Category I and Tornado Protection requirements . The Diesel system modes of operation are not changed or affected by this modification . The Diesel will continue to initiate on LOCA, LOP and LOPILOCA. The sensing of low control air pressure and low Lube Oil Pressure trip will be moved to the Normal mode and deleted from the Emergency Mode to improve the reliability of the diesel during post Accident conditions. An NRC evaluation per TSTIF 400 deleted the surveillance requirement for the Low Lube Oil Pressure trip from Technical Specification 3 .8.1 .13 . A survey of the Diesel Owner's group discovered that many in the owner's group including Riverbend do not have the Low Lube Oil Pressure Trip during an Emergency Start. Moving the Low tube Oil Pressure trip to the normal circuit of the shutdown logic will make it a non-critical trip that would alarm during surveillance testing if the condition existed. During an Emergency Start, the low Lube Oil Pressure trip would be bypassed . This would make the diesel more reliable during an Emergency Mode of Operation. There would only be two critical trips (Overspeed and Generator Differential) not bypassed during an Emergency Start which meets the requirements of Regulatory Guide 1 .4. The Low Lube Oil Pressure trip and High Crankcase pressure trip are going from 2 out of 3 transmitters being used to a single sensor.

The sections potentially impacted are FSAR 8.3.1 .2.1, and 8.3.1 .1 .4.1 . The loss of control air event is evaluated herein . The low tube oil pressure trip will be relocated from the Emergency Shutdown Logic to the Normal shutdown logic and the diesel will not trip on low tube oil pressure or low control air pressure during Emergency conditions . The diesel will continue to run even if a valid diesel protection trip should happen to come in. Critical trips (Overspeed and Generator Differential) are unaffected and would trip the diesel during an Emergency Start. This is the preferred method of operating the diesel post accident . It provides a much more reliable diesel during accident conditions . Moving the Low Lube Oil Pressure Trip to the normal mode has no effect on the frequency of occurrence of an accident described in the FSAR. Removing this trip would improve the reliability of the diesel and removal a failure mechanism from the diesel trip logic. Per the discussion in USAR section Table 3.2-1 XLI, the pneumatic tubing would be 1331 .1 designed equipment. There is no change to these design requirements and no impact on the frequency of occurrence of a FSAR accident .

2. Result in more than a minimal increase in the likelihood of occurrence of a [] Yes malfunction of a structure, system, or component important to safety previously F" No evaluated in the FSAR?

50 .59 for ER NO. 2005-0110-00-00 Page 1 6 of 20 Rev. 0 BASIS:

This modification relocates the Division I and 11 Low Lube Oil Pressure trip and assocated low Control Air pressure sensor to the Normal mode of the start circuit. The tubing and fittings are designed and installed to the same requirements as the interfacing pressure boundaries in accordance with J-702.0, Analysis for installing Tubing (Piping Input) or Specification M-018 .0.

These requirements include ASME Code Section 111, Code Class 2, Safety Class 2, Seismic Category I and Tornado Protection requirements. The Diesel system modes of operation are not changed or affected by this modification . The Diesel will continue to initiate on LOCA, LOP and LOPILOCA . The low lube oil pressure trip will be moved to the Normal mode and deleted from the Emergency Mode to improve the reliability of the diesel during post Accident conditions. An NRC evaluation per TSTIF deleted the surveillance requirement for the Low Lube Oil Pressure trip from Technical Specification 3.8.1 .13 . A survey of the Diesel Owner's group discovered that many in the owner's group including Riverbend do not have the Low Lube Oil Pressure Trip during an Emergency Start . Moving the Low lube Oil Pressure trip to the Normal Mode of the shutdown logic will make it a non-critical trip that would alarm during surveillance testing if the condition existed. During an Emergency Start, the low Lube Oil Pressure trip would be bypassed. This will make the diesel more reliable during a LOCA with low Lube Oil Pressure trip bypassed . The reliability comes with the current configuration, one Low Lube Oil pressure signal (PS-24C) is processed two out of three times (Reg Guide 1 .9 requirement). ER-GG-2005-0110 will have only one pressure signal and switch and move it over to the Normal mode . This removes the potential for a malfunction ofthe Low Lube oil pressure switch during a LOCA.

There will be two critical trips (Overspeed and Generator Differential) not bypassed during an Emergency Start which still meets the requirements of Regulatory Guide 1.9. The Low Lube Oil Pressure trip and High Crankcase pressure trip are going from 2 out of 3 transmitters being used to a single sensor.

The sections potentially impacted are FSAR 8.3.1 .2.1, and 8.3.1 .1 .4.1 . The loss of control air event is evaluated herein. The low lobe oil pressure trip will be relocated to the Normal shutdown logic and the diesel will not trip on low control air pressure during an Emergency Start.

The diesel will, continue to run even if a valid diesel protection trip should happen to come in.

This is the preferred method of operating the diesel post accident. It provides a much more reliable diesel during accident conditions . Moving the diesel low control air pressure sensor and associated Low Lube Oil Pressure Trip to the normal -made has no effect on the frequency of occurrence of an accident described in the FSAR. Moving this trip would improve the reliability of the diesel and removal a failure mechanism from the diesel trip logic. The malfunction of the diesel generator control air system and low tube oil pressure switch during Emergency mode is decreased. This enhances the operation and reliability of the diesel.

3. Result in, more than a ,minimal increase in the consequences of an accident Yes previously evaluated in the FSAR? No

50 .59 for ER NO. 2005-0110-00-00 Page 1 7 of 20 Rev. 0 This modification moves the diesel generator low control air pressure sensor from the Emergency mode shutdown pneumatic circuit to the Normal mode shutdown pneumatic circuit. In addition to moving the low control air pressure sensor, the two out of three Low Lube oil pressure trip is moved from the Emergency Mode shutdown circuit to the Normal mode shutdown circuit. All remaining affected tubing and valve pressure boundaries are qualified to the appropriate operational conditions and meet the design and licensing requirements for pressure boundary integrity .

Accidents with consequences would be such as high energy line breaks . There are no new high energy line breaks as a result of this modification . Therefore, high energy pipe beak and moderate energy line crack evaluations in accordance with USAR Section 3C.2.5 are not affected by this modification .

USAR Chapter 9.0 Section 9.5, Appendix A, Fire Hazards Analysis Report, evaluates the affects of fires involving combustible materials, both fixed and transient, on the ability to safely shutdown the plant and minimize radioactive releases. This modification is located within the diesel generator building (panels IH22P400 and IH22P401) and does not penetrate any structural wall or barriers . Therefore there is no affect to the boundary integrity of any fire area. This modification uses copper tubing which adds a negligible combustible loading to plant fire areas.

Also this modification is deleting two pneumatic valves associated with high crankcase pressure

. This two out of three logic was previously moved from the Emergency mode of the shutdown circuit to the Normal mode shutdown circuit. There is no requirement for two out of three logic in the Normal mode . Therefore, this modification will not compromise the function nor integrity of structures, systems or components important to safety and has no effect on the Fire Hazards Analysis Report .

A review of USAR Chapter 15 Accident Analysis was performed. The proposed modification was evaluated against the existing safety analyses to determine if any of the analyses are impacted. The criteria used in this evaluation is that the change shall not impact the ability of Division I and If Diesel Generators to provide backup power to ECCS equipment, shall not create an event of a type not previously analyzed, and previous component analyses shall not be negatively impacted. The proposed modification satisfies the evaluation criteria, and therefore, the modification is within the bounds of the existing safety analyses.

Specifically, the consequences of the transients and accidents evaluated in USAR Chapters 6 and 15 are unaffected by the moving the Low Control pressure sensor from the Emergency mode shutdown circuit to the Normal Mode shutdown circuit. Likewise for moving the Low Lube Oil pressure trip from Emergency mode to Normal mode shutdown pneumatic circuit will not impact the consequences of an accident. A review of Chapter 15 reveals that the Division I and II Diesel Generators will continue to meet its design basis function to mitigate the consequences ofthese ts. Moving the Loss of Control Air sensor and associated Low Lube Oil Pressure trip over to the Normal shutdown pneumatic circuit will make the Diesels more reliable in Emergency mode (LOCA, LOP, LOPILOCA) and will have no impact on accident consequences .

50 .59 for ER NO. 2005-0110-00-00 Page 1 8 of 20 Rev. 0 The other potentially impacted accidents are a FSAR Chapter 6 analysis . These accidents are considered limiting faults. For the case of the recirculation line break inside containment (i.e.,

drywell) coincident with a Loss of Power, the Division I and II diesel generator would continue to initiate and backup power would be available due to LOCA initiation signal (i.e., high drywell pressure or reactor vessel level low) and LOP. Thus, the evaluation of the consequences of this event are not changed by moving the Low Control Air pressure and Low Lube Oil Pressure from the Emergency Shutdown pneumatic circuit to the Normal Shutdown circuit. The Low Lube Oil Pressure trip and High Crankcase pressure trip are going from 2 out of 3 transmitters being used to a single sensor.

All essential plant systems and equipment will function as assumed in the Accident Analysis.

There is no increase in offsite dose due to any accident previously evaluated. Therefore the proposed activity does not increase the consequences of an accident evaluated previously in the USAR.

4. Result in more than a minimal increase in the consequences of a malfunction of a [~ Yes structure, system, or component important to safety previously evaluated in the No FSAR?

BASIS:

This modification meets the current design and licensing basis such that all affected and nonaffected systems, structures, and components, including the RPY and its internals that are important to safety meet all required operational modes and will function as assumed in the Accident Analysis. The function of the Division I and 11 during an accident (LOCA, LOP, and LOP/LOCA) will continue to provide backup power to it's ESF (Engineering Safety Features)

Buses. The ECCS systems associated with each diesel are unaffected by moving the low control air pressure sensor and low lube oil pressure trip to the normal mode of the Diesel Shutdown logic. Failure of a diesel generator due to the engine itself or one of it's remaining two critical trips (Generator Differential and Overspeed) and/or malfunctions of safety related or important to safety equipment and the mitigating actions for these failures or malfunctions remain the same.

As such, there is no change in the radiological consequences at the site boundary . Therefore, this modification will not increase the consequences of a malfunction of equipment important to safety evaluated previously in the USAR.

5 . Create a possibility for an accident of a different type than any previously (~ Yes evaluated in the FSAR? No

50.59 for ER NO. 2005-0110-00-00 Page 19 of 20 Rev. 0 BASIS:

This modification meets and does not invalidate the current design and licensing basis for the following:

" Drywell and Containment isolation provisions.

" Fire hazards analysis .

" USAR Chapter b and 15 - Accident Analysis .

" Loss of Offsite and Onsite Power (LOP)

There are no other events postulated as a result of this modification which could create the possibility of an accident of a different type than any evaluated previously in the USAR.

Therefore this modification as previously described will not create the possibility of an accident of a different type than evaluated previously in the USAR.

6. Create a possibility for a malfunction of a structure, system, or component [] Yes important to safety with a different result than any p evaluated in the "f No FSAR?

BASIS:

This modification meets the current design and licensing basis such that all affected and nonaffected systems, structures, and components including the Diesel and, Diesel Generator that are important to safety meet all required plant operational modes and events . This includes Loss of Power (Offsite and Onsite) concurrent with an accident such as a LOCA (Loss of Coolant Accident) and USAR Chapter 15 Accident Analysis.

With the moving of the low control air pressure sensor and Low Lube Oil Pressure trip from the Emergency to the Normal mode of the shutdown logic, the possibility of a malfunction of equipment (i.e., a diesel trip) during an accident is decreased.

There are no other postulated events which could create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the USAR.

Therefore this modification as previously described will not create the possibility of a malfunction of equipment important to safety of a different type than any evaluated previously in the USAR.

7. Result in a design basis limit for a fission product barrier as described in the F1 Yes FSAR being exceeded or altered?

50 .59 for ER NO. 2005-0110-00-00 Page 20 of 20 Rev. 0 BASIS :

This modification meets and does not invalidate the current design and licensing basis for the FSAR Chapter 15 Accident Analysis. There are no other events postulated as a result of this modification which could create the possibility of a Diesel failure than any evaluated previously in the USAR.

Therefore this modification as previously described will not result in a design basis limit for a fission product barrier being altered or exceeded.

8. Result in a departure from a method of evaluation described in the FSAR used in F-1 Yes establishing the design bases or in the safety analyses? No BASIS:

Moving the Low Lube Oil Pressure Trip and Low Control Air Pressure sensor from the Emergency to Normal Mode Shutdown logic does not result in a departure from a method of evaluation. All evaluations utilized that are described in the FSAR are still being used for establishing the design bases and safety analyses .

If any of the above questions is checked "YES," obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure LI-143 .

GGNS Commitment Change Evaluation Number CCE 2005-002

COMMITMENT CHANGE EVALUATION FORM Commitment Number : P-24[91, P-24192 Plant Licensing Tracking Number: 00 00 0 ~,

11; Source Document : P-24191 : AECM-8610395, Attachment l, Page 3, Item E, third sentence P-24192: AECM-84/0395, Attachment 1, Page 3, Item E, fourth sentence Commitment: Deletion? _F-4 1 Revision? L]

Has the original commitment been implemented?_[ YES NO, Notify Plant Licensing Original Commitme nt Descrip P-24191 : Independent verification of amendments implementation checklist developed for each TS amendment.

P-24192: Hold points and final verification will be established on the checklist prior to declaring a system operable .

ised Cam cription "

Delete P-24191  I "L, Summary of Justification for Change or Deletion:

P-24191and P-24192 are continuing compliance commitments entered to track items identified in the source document, AECM-86/0395 . This letter documented an LER written for a reportable event that resulted in the failure to revise the Daily Operating Log needed to implement a TS amendment. In the LER, MP&L informed the NRC that the plant licensing procedure "is being revised to require an independent verification of the amendment checklist" (P-24191) and that "hold points and final verifications will be established on the checklist to ensure actions are completed prior to declaring a system operable" (P-24192).

Although responsive to the event and a useful tool for implementing TS amendments, there is no regulation that req licensee to use an implementation checklist. The use of such a tool should be left up to the discretion of the licensee and should not be tracked as a continuing compliance commitment. Therefore, P-24191 and P-24192 should be deleted.

(Attach additional sheets if necessary)

Refer to Attachment 9.4 for a flow diagram that outlines the commitment change evaluation process.

Prepared By : Guy Davant ,

VY - //- as -

f Name/Si nature Date Management Jerry Burford Approval :

LJ tnt Name/Signature Date Plant Licensing v'"

Management G ~'~?t7 V t-~ " Z _ d Concurrence: ---

(- W " I IIII~11 " IWlolly" PrintIIWq~~IpIj~IW~Y~I IWII Name/Signature Date REP tl" 110

PART I ment located in the Updated Final Safety Analysis Report, Emergency Plan, Quality e Program, Fire Protection Program, or Security Plan?

(3 YES STOP . Do not proceed with this evaluation. Instead use appropriate codified process (e .g ., 10 CFR 50.71(e),10 CFR 30.54) to evaluate commitment.

_ NO Go to Part 11 .

PART t1 2 .1 Could the change negatively impact the ability of a System, Structure, or Component (SSC) to perform its safety function or negatively impact the ability of plant personnel to ensure the SSC is capable of performing its intended safety function?

0 YES Go to Question 2 .2.

Z NO Continue with Part III. - Brieflydescribe rationale:

commitments does not involve operation of any plant equipment.

2.2 Perform a safety evaluation using the following 10 CFR 50 .42 criteria to determine if a significant hazards consideration exists :

Does the revised commitment involve a sig the probability or consequences of an accident previously evaluated?

C) YES C] NO Basis:

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

YES C] NO Basis :

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

C3 YES CT NO Basis:

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

(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)?

© YES Go to question 3.2.

0 NO Go to Part IV.

3.2 Is the proposed revised commitment date necessary and justified?

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

Rationale:

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

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

YES Go to Question 4.2.

NO Go to Part V.

4.2 Has the original commitment been implemented?

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

0 NO Go to Question 5.1.

PART V 5.1 Was the original commitment made to minimize recurrence of a condition adverse to quality (e.g., a long-term corrective action stated in an LER)?

YES Go to Question 5.2.

0 NO STOP. You have completed this evaluation. Rev the commitment. No NRC notification required .

REF : Lt-110

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

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

no NRC notification is required:

REFERENCES List documents (e.g., rocedures, NRC submittals, etc.) affected !?I this change.

Doe. Number Description NMM Procedure ENS-Ll-113 1 Licensing Basis Document (LBD) Control Program REFS U"110

GGNS Commitment Change Evaluation Number CCE 2005-003

COMMITMENT CHANGE EVALUATION FORM Commitment Number : P-24106 Plant Licensing Tracldng Number : C,]-- ;L()05'-0005 Source Document: AECM-86/0077.ATT.l,PGAIT.3 Deletion? L) Revision?___

Has the original commitment been implemented? I Z YES I El NO, Notify Plant Licensing Original Commitment

Description:

lement dose related restricted locations in the spent fuel pool per analysis documented in the source document .

Revised Commitment Description :

Implement revised dose restricted locations performed in ER-2003-0018-019 . AECM 86/0077 specifically allowed in Item 5 ofAttachment that deviation from the prescribed guidelines could be done as long as an evaluation of the impact was performed . The evaluation of the impact has been evaluated and documented in ER-2003-0018-019.

Summary of Justification for Change or Deletion :

Additional locations in rack B I and C 1 (east and south walls of SFP) that had been previously inaccessible are now accessible after removal of the underwater work table . These locations had not been analyzed for dose considerations previously due to being inaccessible at the time of the original analysis documented in AECM-86/0077 . A new analysis has been performed and documented in ER-2003-0018-019 and the new dose restrictions are different than previously analyzed.

Hence, the ER guidelines supersedes the restrictions listed in AECM 86/0077.

(Attach additional sheets if necessary)

Refer to Attachment 9.4 for a flow diagram that outlines the commitment change evaluat process.

Prepared By: Paul M. Different/

&JA Print Name/Sin a e 7/7

/0.

Date Management Approval:

t? h W I kf Nf rint Name/Si nature Date Plant Licensing Management Concurrence : `lo'tN " 6 Print Name/Si nature Date REF: U-1 10

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

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 II.

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

YES Go to Question 2.2.

NO Continue with Part III . Briefly describe rationale:

Intent of the commitment is not being changed - that is 2 .5 mR1hr dose rates in areas adjacent to the spent fuel pool. The updated analysis changes the locations required to be dose restricted to meet this intent but it does not change the intent of the commitment.

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

Does the revised commitment involve a significant increase robability or consequences of an accident previously evaluated?

YES El NO Basis:

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

YES 0 NO Basis:

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

YES (1 NO Basis:

If any of the above questions are answered Yes, STOP. Do not proceed with the revision, OR discuss change with NRC and obtain necessary approvals prior to implementation of the proposed change. If all three 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 (i.e., rule, regulation, order or license condition)?

(:1 YES Go to question 3.2.

P4 NO Go to Part IV.

3.2 Is the proposed revised commitment date necessary and justified?

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

Rationale :

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

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

YES Go to Question 4.2.

El NO Go to Part V.

4.2 Has the original commitment been ed?

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

El NO Go to Question 5.1.

PART V 5.1 Was the original commitment made to minimize recurrence of a condition adverse to quality (e.g., a long-term corrective action stated in an LER)?

YES Go to Question 5.2.

F1 NO STOP. You have completed this evaluation. Revise the commitment. No NRC notification required .

REF : u-a 10

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

0 YES Revise the commitment and notify NRC of revised commitment in next annua11RFO interval summary report.

El NO Revise commitment: no NRC notification is required :

REFERENCES List documents e.g., procedures, NRC submittals, etc. affected b this change.

Doc. Number Description and Inventory Control REF: U-1 10

GGNS Commitment Change Evaluation Number CCE 2005-004

COMMITMENT CHANGE EVALUATION FORM Commitment Number: P-24107 Plant Licensing Tracking Number :

Source Document: AECM-86/0077 .ATT.l,PG.4,IT.4.B Commitment: Deletion? C] Revision ?

Has the original commitment been im lemented? [II©1 NO, Notify Plant Licensing Revised Commitment Description :

Implement revised dose restricted locations and shield assemblies with 5 years of decay performed in ER-2003-0018-019 .

AECM 86/0077 specifically allowed in Item 5 of Attachment that deviation from the prescribed guidelines could be done as long as an evaluation of the impact was performed. The evaluation of the impact has been evaluated and documented in ER-2003-0018-019 .

Summary of Justification for Change or Deletion:

At the time of the original dose analysis the only fuel available to use as shield assemblies was cycle discharge fuel.

Obviously there are many more discharged assemblies available now that can be used for shield assemblies. ER-2003-0018-019 analyzed this and requires a shield assembly to have 5 years of decay to be placed in a dose restricted location (Attach additional sheets if necessary)

Refer to Attachment 9.4 for a flow diagram that outlines the commitment change evaluation process .

Prepared By: I Paul M. Different/

Print Name/Signat Date l1~w Pr t NameISignature Date Plant Licensing Management Concurrence : G .fps 6,,-A c 0'-

Print Name/Signature Date REF: U-110

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

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 Go to Part II.

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

YES Go to Question 2.2.

NO Continue with Part III . Briefly describe rationale :

t of the commitment is not being changed - that is 2 .5 mRthr dose rates in areas adjacent to the spent fuel pool. The updated analysis changes the time of decay from 1 year to 5 years required to meet this intent but it does not change the intent of the commitment .

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

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

YES NO Basis:

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

Q YES F] NO Basis :

Does the revised commitment volve a significant reduction in a margin of safet YES F1 NO Basis :

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

(Attach additional sheets as necessary.)

REF: CI-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)?

© YES Go to question 3.2.

NO Go to Part IV.

3.2 Is the proposed revised commitment date necessary and justified?

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

Rationale:

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

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

YES Go to Question 4.2.

El NO Go to Part V.

4.2 Has the original commitment been implemented?

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

ONO Go to Question 5.1 .

PART V 5 .1 Was the original commitment made to minimize recurrence of a condition adverse to quality (e.g., a long-term corrective action stated in an LER)?

YES Go to Question 5.2.

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

REF: Lt-110

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

YES Revise the commitment and notify NRC of revised commitment in next annual"O interval summary report.

El NO Revise commitment : no NRC no tion required :

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

Doc. Number Description I7-s-o2-3ao SNM Movement and Inventory Control REF: Lt-110

GGNS Commitment Change Evaluation Number CCE 2005-005

COMMITMENT CHANGE EVALUATION FORM Commitment Number : P-24109, P-29287, P-29288, Plant Licensing Tracking Number:

P-29289 n CG 7-00 5,40 6 5' Source Document : AECM-86/0089, Attachment 1,Section IV.B, paragraphs 1 and 2 Commitment: Deletion?

Has the original commitment been implemented? YES I [I NO, Notify Plant Lic ~ _,

Original Commitment Description :

Guidance for justification for UFSAR commitment deletion Revised Commitment

Description:

Delete P-24109, P-29287, P-29288, and P-29289 .

Summary ofJustification for Change or Deletion :

The identified commitments are continuing compliance commitments entered to track items identified in AECM-86/0089.

This letter documents a response to a Notice of Violation involving "the failure to conduct 10 CFR 50.59 safety evaluations of changes incorporated into the Updated FSAR." In the response, Entergy (MP&L) agreed to perform full 50.59 Evaluations for FSAR changes and also agreed to provide "more explicit criteria on what constitutes adequate justification for commitment deletion."

Performing 50.59 Evaluations for FSAR changes, as committed in the letter, goes beyond the requirements of the current 50.59 rule and imposes additional burden on the licensee with no significant increase in safety .

Regarding the use of 50.59 to delete commitments, since the issuance of AECM-8610089 (4/10/86), NEI developed and published NEI 99-04, Guidelinesfor Maintaining Commitments, which was endorsed by the NRC . With the advent of this document, the industry no longer uses the 50.59 process to delete commitments, whether contained in the FSAR or not.

LI-110, Commitment Management Program, reflects information and guidance contained in NEI 99-04 and is the procedure that controls and manages commitments .

Based_ on the above discus sion, the identified commitments are no longer valid and should be deleted.

(Attach additional sheets if necessary)

Refer to Attachment 9.4 for a flow diagram that outlines the commitment change evaluation process.

Prepared By: Guy Davant r rte_ /Z p s-Print Name/Signature Date Management Jerry Burford Approval:

i2- 1'tf =a ~

Pr' me/Signature Date

°"

Plant Licensing Management Concurrence : t r' Print Name/Si nature Date REF: 1.1-11 0

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

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 11.

PART I1 2.1 Could the change negatively impact the ability of a System, Structure, or Component (SSC) to perform its safety function or negatively impact the ability of plant personnel to ensure the SSC is capable of performing its intended safety function?

M YES Go to Question 2.2.

0 NO Continue with Part III . Briefly describe rationale:

These comm s do not involve operation of any plant equipment .

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

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

[] YES NO Basis:

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

YES NO Basis:

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

[I YES [Q NO Basis:

If any of the above questions are answered Yes, STOP . Do not proceed with the revision, OR discuss change with NRC and obtain necessary approvals prior to implementation of the proposed change. If all three questions are answered NO, go to Part 111 .

(Attach additional sheets as necessary.)

REF: X1-110

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)?

YES Go to question 3.2.

NO Go to Part IV.

3.2 Is the proposed revised commitment date necessary and justified?

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

Rational

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

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

0 YES Go to Question 4.2.

Z NO Go to Part V.

4.2 Has the original commitment been implemented?

C] YES STOP, You have completed this evaluat revised commitment in summary report .

. Revise the commitment and notify NRC of

© NO Go to Question 5.1.

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

term corrective action stated in an LER)?

0 YES Go to Question 5.2.

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

REF: U-110

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

(] YES Revise the commitment and notify NRC of revised commitment in next annuaVRFO Interval summary report.

n NO Revise commitment: no NRC notification is required :

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

Doe. Number I- Description NMM Procedure ENS-Ll-113 Licensing Basis Document (LBD) Control Program REF: U-110