View source for GNRO-2007/00071, License Amendment Request (Lar), Changes to Technical Specification 5.6.5, Core Operating Limits Report (COLR)

{{Adams
| number = ML073440113
| issue date = 12/05/2007
| title = License Amendment Request (Lar), Changes to Technical Specification 5.6.5, Core Operating Limits Report (COLR)
| author name = Brian W
| author affiliation = Entergy Corp, Entergy Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000416
| license number = NPF-029
| contact person = 
| case reference number = GNRO-2007/00071
| document type = Letter type:GNRO, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 21
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:SEntergy                                                                        Entergy P.O. Box 756 Port Gibson, MIS 39150 Tel 601 437 6409 William R. Brian Vice President - Operations Grand Gulf Nuclear Station GNRO-2007/00071 December 5, 2007 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

==SUBJECT:==
License Amendment Request (LAR)
Changes to Technical Specification 5.6.5, "Core Operating Limits Report (COLR)"
Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29

==REFERENCE:==
Letter from USNRC to Exelon Nuclear, "LaSalle County Station Units 1 and 2, Issuance of Amendments RE: Core Operating Limits Report (TAC Nos. MB9888 and MB9889)," dated July 9, 2004 (ADAMS Accession No. ML033430391)

==Dear Sir or Madam:==

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests a license amendment for Grand Gulf Nuclear Station, Unit 1 (GGNS). A change is proposed to Technical Specification (TS) 5.6.5, "Core Operating Limits Report (COLR)" to add a reference to an analytical method that will be used to determine core operating limits.
Additionally, an administrative change is proposed to an existing reference in TS 5.6.5.
The proposed changes are the result of a decision to insert GE14 fuel during refueling outage RF16 scheduled for fall 2008. GGNS currently operates with a full core of ATRIUM-10 fuel.
The new topical report reference, NEDC-33383P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," will allow Entergy to use a Global Nuclear Fuel (GNF) method for the determination of fuel assembly critical power of AREVA ATRIUM-10 fuel. The NRC previously approved use of GEXL97 for LaSalle County Stations, Units 1 and 2 by the above referenced amendment. NEDC-33383P is provided as Attachment 4.
Some of the information in topical report NEDC-33383P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," is PROPRIETARY to Global Nuclear Fuel - Americas (GNF-A). GNF-A requests that the PROPRIETARY information be withheld from public disclosure in accordance with 10 CFR 9.17(a)(4) and 10 CFR 2.390 (a)(4). A non-proprietary version of the topical report is provided as Attachment 5. An affidavit by the information owner, GNF-A, supporting the request for non-disclosure is provided in Attachment 6.
                                                                                                            *(DI

GNRO-2007/00071 Page 2 of 3 The GEXL97 correlation was developed in part by using the approved AREVA SPCB
'correlation method. AREVA recently identified an error in their SPCB critical power correlation affecting ATRIUM-10 fuel. AREVA is currently evaluating the impact of the error, but it is not expected to significantly impact the GEXL97 topical report. Entergy will provide a supplement to this LAR describing the impact, if any, to the GEXL97 report.
Entergy has not completed analyses to determine whether a change to the TS Minimum Critical Power Ratio (MCPR) safety limits will be required. Entergy is submitting this request in advance of completed core reload analysis to provide sufficient review time. If analyses indicate that a change to the MCPR safety limit or other TS change is required, the TS change will be requested by a separate LAR.
The proposed change has been evaluated in accordance with 10 CFR 5.0.91 (a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards consideration. The bases for these determinations are included in the attached submittal.
The proposed change includes three new commitments.
Entergy requests approval of the proposed amendment by September 10, 2008 to support the RF16 reload. Once approved, the amendment shall be implemented prior to Cycle 17 operation. Although this request is neither exigent nor emergency, your prompt review is requested.
If you have any questions or require additional information, please contact Ron Byrd at 601-368-5792.
I declare under penalty of perjury that the foregoing is true and correct. Executed on December 5, 2007.
Sincerely, WRB/RWB/amm Attachments:
: 1. Analysis of Proposed Technical Specification Change
: 2. Proposed Technical Specification Changes (mark-up)
: 3. List of Regulatory Commitments
: 4. NEDC-33383P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel,"
Proprietary Version
: 5. NEDC-33383P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel,"
Non-Proprietary Version
: 6. Affidavit for Request to Withhold Information cc:      (See Next Page)

GNRO-2007/00071 Page 3 of 3 cc:    U. S. Nuclear Regulatory Commission ATTN: Mr. Elmo E. Collins RegionalAdministrator, Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 U. S. Nuclear Regulatory Commission ATTN: Mr. Bhalchandra Vaidya, NRR/DOLR ATTN: ADDRESSEE ONLY ATTN: U. S. Postal Delivery Address Only Mail Stop OWFN/O-8G14 Washington, DC 20555-0001 Mr. Brian W. Amy, MD, MHA, MPH Mississippi Department of Health P. 0. Box 1700 Jackson, MS 39215-1700 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150

i I ,
Attachment 1 GNRO-2007/00071 Analysis of Proposed Technical Specification Change to GNRO-2007/00071 Page 1 of 12

==1.0      DESCRIPTION==

This letter is a request to amend Operating License NPF-29 for Grand Gulf Nuclear Station, Unit 1 (GGNS).
A change is proposed to Technical Specificatiori (TS) 5.6.5, "Core Operating Limits Report (COLR)" to add a reference to an analytical method that will be used to determine core operating limits. The new reference, NEDC-33383P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," will allow Entergy to use a Global Nuclear Fuel (GNF) 1 method to determine fuel assembly critical power of AREVA ATRIUM-10 fuel. GGNS currently operates with a full core of ATRIUM-10 fuel. Entergy plans to use the GEXL97 correlation for GGNS operating Cycle 17 currently scheduled to begin fall 2008. Additionally, an administrative change is proposed to an existing reference in TS 5.6.5.

==2.0      PROPOSED CHANGE==

TS Section 5.6.5.b lists the analytical methods previously reviewed and approved by the NRC that are used to determine the core operating limits. The proposed change will add the following new reference to TS 5.6.5.b.
: 21. NEDC-33383P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," Global Nuclear Fuel.
While GEXL97 has been approved for use at LaSalle County Station, Units 1 and 2, it has not
-been approved for generic application by licensees. Entergy is proposing that the NRC approve use of this methodology for determining GGNS core operating limits.
Entergy also proposes an administrative change to Reference 24 of TS 5.6.5. Reference 24 is currently listed as:
: 24. NEDE-24011-P-A, General Ele6tric Standard Application for Reactor Fuel (GESTAR-II) with exception to the misplaced fuel bundle analyses as discussed in GNRO-96/00087 and the generic MCPR Safety Limit analysis as discussed in GNRO-96/00100, letters from C. R. Hutchinson to USNRC.
Entergy proposes to revise Reference 24 to:
: 24. NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR-Il).

==3.0      BACKGROUND==

Core operating limits are established each operating cycle in accordance with TS 3.2, "Power Distribution" and TS 5.6.5, "Core Operating Limits Report (COLR)". These operating limits ensure that the fuel design limits are not exceeded during any conditions of normal operation or in the event of any Anticipated Operational Occurrence (AOO).
1 Global Nuclear Fuel (GNF), is a joint venture of General Electric (GE), Hitachi and Toshiba to GNRO-2007/00071 Page 2 of 12 The methods used to determine the operating limits are'those previously found acceptable by the NRC and listed in TS Section 5.6.5.b. The analytical methods currently listed support the determination of core operating limits by using those methods applicable to fuel supplied by General Electric (GE, currently known as Global Nuclear Fuel) or AREVA (formerly known as Framatome Advanced Nuclear Power (FRA-ANP), or Siemens). GGNS has employed fuel supplied by GE'or AREVA since it began commercia; opera.ion but is only using AREVA -
ATRIUM-10 fuel in the current operating cycle.
GGNS has recently decided to load GE14 fuel during its upcoming refueling outage in fall 2008.
GGNS intends to use GE/GNF methodologies to determine overall core operating limits. This change will require the listing of an additional analytical method for analyzing the AREVA ATRIUM-10 fuel. The requested TS change adds the reference needed to determine fuel assembly critical power of ATRIUM-10 fuel.
Entergy also proposes an administrative change to Reference 24 (GESTAR-II) of TS 5.6.5.b to delete information that is no longer needed. Reference 24 was modified in 1996 by Amendment No. 131 (Reference 2) to include twoexceptions to GESTAR-II.. GESTAR-II has since been revised for the misplaced fuel bundle analysis and the generic MCPR safety limit analysis.
These revisions make the exceptions no longer needed..

==4.0      TECHNICAL ANALYSIS==

4.1      GEXL97 Correlation for ATRIUM-10 Fuel TS Safety Limits (SLs) ensure, that specified fuel design limits are not exceeded during steady state operation, normal operational transients, and Anticipated Operational Occurrences (AOOs). The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core do not experience transition boiling. The margin between calculated boiling transition and the MCPR SL is based on a detailed statistical procedure that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty inherent in the critical power correlation. The fuel vendor's critical power correlations are based on data which provide a high degree of assurance that the critical power, as evaluated by the correlation, is within a small percentage of the actual critical power being estimated.
The GEXL correlation is an NRC approved GE method of accurately predicting the occurrence of boiling transition in Boiling Water Reactor (BWR) fuel. The GEXL correlation is necessary for determining the MCPR operating limits resulting from transient analysis, the MCPR safety limit analysis, and the core operating performance and design.
The GGNS reactor core currently contains only AREVA ATRIUM-10 fuel. Entergy plans to insert GE14 fuel in the reactor during the upcoming refueling outage and will begin using GE/GNF's safety analysis methodologies, including GNF's critical power correlation methods.
The GGNS TS currently list NEDE-2401 1-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR-Il)," as a method of determining core operating limits. GESTAR-II describes the use of GEXL or GEXL-PLUS as approved methods for critical power correlations; however, it does not describe a specific approved method of performing critical power correlations for ATRIUM-10 fuel. Therefore, Entergy is proposing to add the GEXL97 reference to TS 5.6.5.b as a correlation method to be used for ATRIUM-10 fuel. In the Safety Analysis process, the GEXL97 correlation is to be applied to the ATRIUM-1 0 fuel in the mixed core while to GNRO-2007/00071 Page 3 of 12 the appropriate approved GEXL correlation will be applied to the GNF fuel (including the determination of an acceptable MCPR safety limit for the mixed core).
Proprietary and hon-proprietary versions of GEXL97 are provided as Attachments 4 and 5 respectively. GEXL97 was previously approved for use at LaSalle County Station, Units 1 and 2 (Refe'rence 1),- but has not been approved generically forother BWRs. The GEXL97 tbpical report approved for LaSalle was NEDC-33106P, "GEXL97 Correlation for ATRIUM-10 Fuel."
The GEXL97 topical report has recently been rewritten as NEDC-33383P, "GEXL97 Correlation Applicable to ATRIUM-10 fuel." The latter version of GEXL97 was rewritten to incorporate resolution of NRC questions raised during the review of the LaSalle amendment request and to make the correlation generically applicable to ATRIUM-10 fuel. The GEXL97 correlation described in NEDC-33383P is based on the same set of critical power data used in NEDC-33106P and consequently is identical to that previously approved for LaSalle, including the correlation uncertainty. Entergy is proposing to use the latest version of GEXL97, NEDE-33383P for the critical power correlation of ATRIUM-1 0 fuel.
Entergy's evaluation of the application of GEXL97 to GGNS reloads considered the LaSalle precedent and specifically the following considerations:
: 1) adequacy of the critical power database generated with the AREVA         sub-channel code XCOBRA thermal hydraulic model in place of an experimental          database;
: 2) proper determination of the uncertainty in the GEXL97 correlation's      predictions for the ATRIUM-10 fuel design; and,
: 3) applicability of the proposed operating range of GEXL97 correlation      to the ATRIUM-10 fuel supported by the range of the database.
Entergy has determined that the use of the GEXL97 correlation is appropriate for GGNS, and provides an equivalent level of protection as that currently provided.
Adequacy of the Critical Power Data Base and Associated Uncertainties The GEXL correlation was developed based on boiling transition data obtained from dryout tests at a GE test facility. However, specific experimental data for AREVA's ATRIUM-10 fuel is not available to the new fuel vendor, GNF. Therefore, a critical power database was generated by using the approved SPCB correlation encoded in the AREVA thermal hydraulic model XCOBRA. This database was then used by GNF to support the development of GEXL97.
The database used in the development of the GEXL97 correlation for the ATRIUM-10 fuel is provided in Table 2-1 of the attached GEXL97 report. This table shows the number of calculated critical power data points obtained using the AREVA SPCB critical power correlation for various axial power distributions. It also shows the fuel pin dryout location that formed the basis of the different sets of AREVA calculated critical power data. Table 2-2 of the report, provides the same information but further divides the data collected into subgroups of pressure, mass flux, and inlet sub-cooling.
The database is treated as real data in the regression analysis to generate the correlation coefficients, which introduces unavoidable error (i.e., uncertainty) into the correlation being derived from it. Since the GEXL97 correlation is fitted to this data, the uncertainty in the critical power prediction of the GEXL97 correlation for a given set of conditions will have some additional uncertainty relative to the real critical power value for those conditions; over and above the uncertainty of the correlation's fit to the database. The GEXL97 correlation

Attachment 1 to GNRO-2007/00071 Page 4 of 12 appropriately determines the overall CPR uncertainty by accounting for both the uncertainty in its fit to the database and for the uncertainty of the critical power values in the database itself. A statistical analysis which demonstrates the ability of the final GEXL97 correlation to predict the ATRIUM-10 simulated critical power data is provided in the attached GEXL97 report.
"Since approval of the GEXL97 correlation methodology for LaSalle, AREVA identified'an error in their SPCB critical power correlation affecting ATRIUM-10 fuel. The error involves the calculation of the local power peaking distribution in the test assemblies used to determine critical power performance. AREVA is currently evaluating the impact of the error, but it is not expected to significantly impact the GEXL97 methodology as described in the topical report.
Entergy will provide a supplement to this LAR describing the impact, if any, to the GEXL97 report.
Generation of the GEXL97 Correlation and the Ranqe of Applicability In developing the GEXL97 correlation, GNF took certain steps to optimize the GEXL97 critical power predictions for the ATRIUM-10 fuel design, and to minimizethe prediction uncertainty.
This process is identical to that used by GNF when developing GEXL correlation coefficients for GNF/GE fuel designs using raw experimental test data, and has been used in past development of GEXL correlations applicable to other co-resident fuel.
The GEXL97 application range is provided in Section 4.2 of the attached report. This application range covers the range of expected operation of the ATRIUM-10 fuel during normal steady state and transient conditions in the GGNS reload cores for pressures down to 800 psia.
The GGNS TS safety limit requires the core thermal power to be < 25% of rated when reactor
 &#xfd;steam dome pressure is < 785 psig (- 800 psia) or core flow is < 10% of rated. This safety limit is intended to provide fuel cladding integrity protection during start-up conditions since the GEXL correlation was not approved at the time as a licensing model for pressures below 785 psig. Entergy is aware that GE has provided a 10 CFR Part 21 Notification, SC05-03 dated March 29, 2005 (Reference 3), which reports that the GEXL correlation lower end of the pressure range (i.e., 800 psia) could temporarily be exceeded during a reactor depressurization transient caused by a Pressure Regulator Failure.-Maximum Demand Open (PRFO). Although GE has recently received NRC approval of a change to the GEXL14 correlation model for pressures down to 700 psia, a mutual agreement on an appropriate TS change needed to resolve the issue has not been reached (see Reference 4). The industry is working with the NRC staff to resolve the TS safety limit issue.
Although a depressurization transient could result in vessel pressures below the range of GEXL97, the transient would not threaten fuel cladding integrity, since it has been determined that the margin to the MCPR safety limit increases with decreasing reactor pressure. Therefore, no immediate actions are required for this TS issue. Entergy is following this effort and will request appropriate TS changes or an extended applicability range for the GEXL97 correlation after an appropriate generic resolution is reached by the industry and NRC staff. Since the depressurization event does not threaten fuel cladding integrity, Entergy believes that final resolution of the Part 21 issue is not necessary prior to the use of GEXL97.
4.2        Administrative Changes to the Existing GESTAR-II Reference The proposed change to Reference 24 (GESTAR-II) of TS 5.6.5.b deletes information that is no longer needed. This change is administrative in nature because it does not alter the method of performing the analyses. Reference 24 was modified in 1996 by Amendment No. 131 to GNRO-2007/00071 Page 5 of 12 (Reference 2) to take two exceptions to the GESTAR-Il version that was current at the time.
The two exceptions involved the generic MCPR safety limit analysis and the misplaced fuel bundle analysis. GESTAR-II has since been revised for the generic MCPR safety limit analysis and the misplaced fuel bundle analysis which makes the exceptions no longer needed.
The exception to the generic MCPR safety limitwas requested iri. 1996 for GGNS Cycle 9-when GGNS began loading GEl 1 fuel. At that time, GESTAR-II used a generic MCPR safety limit which was determined to be non-conservative for some core designs including those with mixed vendor fuels. This issue was the subject of a GE 10 CFR Part 21 notice of May 24, 1996. To address this issue, GGNS took exception to the generic MCPR safety limit in GESTAR-II and calculated a plant cycle-specific MCPR safety limit. Subsequent revisions to GESTAR-II addressed the issue by providing for a cycle-specific calculation of the MCPR safety limit.
Therefore, the exception to GESTAR-II is no longer needed.
The exception to the misplaced fuel bundle analysis was also requested in 1996 due to the differences in GESTAR-II and the GGNS plant licensing basis. GESTAR-II treated the misplaced fuel bundle events as incidents of moderate frequency, whereas the GGNS licensing basis treated the events as infrequent events. The exception allowed GGNS to analyze the misplaced fuel bundle events as infrequent events. GESTAR-II has since been revised to provide for the treatment of the misplaced fuel bundle events as infrequent events. Therefore, the exception to GESTAR-II is no longer needed.
The deletion of these exceptions does not result in any change to the previous method of
&#xfd;performing these analyses. The analyses will continue to be performed in accordance with the NRC approved versions of GESTAR-II.
4.3      Existing TS 5.6.5 Analytical Methods Entergy has reviewed the currently listed analytical methods in TS 5.6.5 and has determined that the remaining references continue to be required for GGNS. A summary of our review is contained in Table 1.
5.0     Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.
Entergy has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the TS, and do not affect conformance with any General Design Criterion (GDC) differently than described in the Updated Final Safety Analysis Report (UFSAR).
* 10 CFR 50.36, Paragraph d(5), states that the TS will include administrative controls that address the provisions relating to organization and management, procedures, record keeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
The COLR is required as a part of the reporting requirements specified in the GGNS TS Administrative Controls section. The TS requires the core operating limits to be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and to be documented in the COLR. In addition, it requires the analytical methods used to determine the core operating limits, to be approved by the NRC and described in the Administrative Controls section of the TS. The proposed TS changes ensure that these requirements are met.
to GNRO-2007/00071 Page 6 of 12 10 CFR 50.34, "Contents of Applications; Technical Information," requires that Safety Analysis Reports be submitted that analyze the design and performance of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents. As part of the core reload design process, reload safety evaluations are performed to ensure that the safety analyses remain. bounding for the design cycle:. To confirm that the analyses remain bounding, key inputs to the safety analyses such as the Critical Power Ratio (CPR) are confirmed to be conservative with respect to the current design cycle. If key safety analysis parameters are not bounded, a re-analysis or re-evaluation of the affected transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied. The proposed TS change is needed to perform reload safety analysis for the.next cycle core reload consisting of fuels from two different fuel vendors.
5.1      No Sigqnificant Hazards Consideration Entergy proposes to use a Global Nuclear Fuels (GNF) analysis method to determine core operating limits for Grand Gulf Nuclear Station (GGNS) beginning with operating Cycle 17. The GNF method of analysis, NEDC-33383P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel,"
will be used to determine fuel assembly critical power of AREVA ATRIUM-1 0 fuel. TS Section 5.6.5.b must be revised to include a reference to the GEXL97 topical report prior to the method being used to establish the core operating limits. Additionally, Entergy proposes to make an administrative change to the existing reference to GESTAR-Il.
'Entergy Operations, Inc. has evaluated whether or not a significant hazards consideration is
:involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
Core operating limits are established each operating cycle in accordance with TS 3.2, "Power Distribution" and TS 5.6.5, "Core Operating Limits Report (COLR)". These core operating limits ensure that the fuel design limits are not exceeded during any conditions of normal operation or in the event of any Anticipated Operational Occurrence (AOO).
The methods used to determine the operating limits are those previously found acceptable by the NRC and listed in TS Section 5.6.5.b.
A change to TS 5.6.5.b is requested to include an additional reference to the list of analytical methods. GGNS currently operates with a full core of AREVA ATRIUM-10 fuel but is scheduled to load GE14 fuel during the next refueling outage. GGNS plans to use the analysis methods of the new fuel vendor, GNF for the analysis of the mixed core.
The GEXL97 correlation accurately models predicted core behavior and appropriately determines the overall critical power uncertainty of the method. In addition, the GEXL97 application range covers the range of expected operation of the ATRIUM-10 fuel during normal steady state and transient conditions in the GGNS reload cores. Although a depressurization transient could result in vessel pressures below the range of GEXL97, the transient would not threaten fuel cladding integrity, sincethe margin to the MCPR safety limit increases with decreasing reactor pressure.
to GNRO-2007/00071 Page7of12 Additionally, Entergy proposes an administrative change to the GESTAR-Il reference in TS 5.6.5.b. The administrative change does not alter any method of analysis as described in the NRC approved versions of GESTAR-Il.
The requested TS changes concern the use of analytical methods and do not involve any planttmodifications or operational changes that could affect any postulated accident precursors or accident mitigation systems and do not introduce any new accident initiation mechanisms. The proposed changes have no effect on the type or amount of radiation released, and has no effect on predicted offsite doses in the event of an accident. Thus, the proposed change does not affect the probability of an accident previously evaluated nor does it increase the radiological consequences of any accident previously evaluated.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed TS changes will not change the design function, reliability, performance, or operation of any plant systems, components, or structures. It does not create the possibility of a new failure mechanism, malfunction, or accident initiators not considered in the design and licensing bases. Plant operation will continue to be within the core operating limits that are established using NRC approved methods that are applicable to the GGNS design and the GGNS fuel.
Therefore, the proposed change does not create the possibility of a new or different kind
* of accident from any previously evaluated.
: 3. Doesthe proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change adds GEXL97 to the list of analytical methods in TS 5.6.5.b that can be used to determine core operating limits. Use of the GEXL97 correlation analytical method provides an equivalent level of protection as that currently provided.
The administrative change does not alter any method of analysis as described in the NRC approved versions of GESTAR-II. The proposed change does not modify the safety limits or setpoints at which protective actions are initiated, and do not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92.
to GNRO-2007/00071 Page 8 of 12 5.2      Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure: Accordingly, the proposed amendment meets the eligibility criterion for'categorical exclusion set forth in 10 CFR51.221(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
6.0      PRECEDENCE The NRC previously approved use of GEXL97 for LaSalle County Stations, Units 1 and 2 (Reference 1). The LaSalle application requested that two methods be added to TS 5.6.5.b.
One of the methods, GEXL96 is for ATRIUM-9B fuel which GGNS does not use. The other method was for ATRIUM-10 fuel and was listed as NEDC-33106P, "GEXL97 Correlation for ATRIUM-10 Fuel." The version of GEXL97 that will be referenced in the GGNS TS is NEDC-33383P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel." NEDC-33383P is an updated version of NEDC-33106P which incorporates resolution of NRC questions that arose during the review of the LaSalle amendment request and makes the correlation generically applicable to ATRIUM-10 fuel. The GEXL97 correlation described in NEDC-33383P is based on the same set of critical power data used in NEDC-33106P and consequently is identical to that previously approved for LaSalle, including the correlation uncertainty.

==7.0      REFERENCES==
: 1. Letter from USNRC to Exelon Nuclear, "LaSalle County Station Units 1 and 2, Issuance of Amendments RE: Core Operating Limits Report (TAC NOS. MB9888 AND MB9889)," dated July 9, 2004 (ADAMS Accession No. ML033430391)
: 2. Letter from USNRC to Mr. Joseph A. Hagan of Entergy, "Issuance of Amendment No. 131 to Facility Operating License No. NPF Grand Gulf Nuclear station, Unit 1 (TAC NO.
M95385), dated November 21, 1996 (ADAMS Accession No. ML021490078)
: 3. Letter from Mr. Jason S. Post of General Electric Company, "10CFR21 Reportable Condition Notification: Potential to Exceed Low Pressure Technical Specification Safety Limit," dated March 29, 2005 (ADAMS Accession No. ML050950428)
: 4. Letter from USNRC to the Technical Specification Task Force (TSTF), "Denial of TSTF-495, Revision 0, "Bases Change to Address GE Part 21 SC05-03." Docket No: PROJ0753 (TAC MD2672), dated August 27, 2007 (ADAMS Accession No. ML072340113) to GNRO-2007/00071 Page 9 of 12 TABLE I ANALYTICAL METHODS CURRENTLY LISTED IN TS 5.6.5 Report                 Applicable               Justification LCO
: 1. "XN-NF-81-58(P)(A), "RODEX2          -,13.2.1  Provides an analyticalcapability to Fuel Rod Thermal-Mechanical            3.2.2  predict BWR fuel thermal and Response Evaluation Model";            3.2.3  mechanical conditions for normal core Exxon Nuclear Company, Inc.,                  operation and to establish initial Richland, WA.                                 conditions for power ramping, non-LOCA and LOCA analyses.
: 2. XN-NF-85-67(P)(A), "Generic             3.2.3  Describes the process used to develop Mechanical Design for Exxon                   linear heat generation rates for fuel Nuclear Jet Pump BWR Reload                   designs.
Fuel", Exxon Nuclear Company, Richland, WA.
: 3. EMF-85-74(P) Supplement 1 (P)(A)        3.2.3  Extends the exposure limit of the and Supplement 2(P)(A),                       RODEX2A code which is a version of "RODEX2A (BWR) Fuel Rod                       RODEX2 that includes a fission gas Thermal-Mechanical Evaluation                 release model specific to BWR fuel Model", Siemens Power                         designs.
Corporation, Richland, WA.
.4. ANF-89-98(P)(A), "Generic               3.2.3  Establishes a set of design criteria Mechanical Design Criteria for                which assures that BWR fuel will BWR Fuel Designs, Advanced                    perform satisfactorily throughout its Nuclear Fuels Corporation",                   lifetime.
Richland, WA.
: 5. XN-NF-80-19(P)(A) Volume 1,             3.2.1  Development of BWR core analysis "Exxon Nuclear Methodology for         3.2.2  methodology which comprises codes Boiling Water Reactors - Neutronic     3.2.3  for fuel neutronic parameters and Methods for Design and Analysis",             assembly burnup calculations, reactor Exxon Nuclear Company,                        core simulation diffusion theory Richland, WA.                                 calculations, core and channel hydrodynamic stability predictions, and producing input for nuclear plant transients. Subsequently approved codes or methodologies have superceded portions of this report.
Applicable portions include CRDA, and methodology to determine neutronic reactivity parameters, void reactivity, Doppler reactivity, scram reactivity, delayed neutron fraction, and prompt neutron lifetime.
to GNRO-2007/00071 Page 10 of 12 Report                  Applicable               Justification LCO
: 6. XN-NF80-19(P)(A) Volume 4,              3.2.1   Summarizes the types of BWR "Exxon Nuclear Methodology for         3.2.2   licensing analyses performed, Boiling Water Reactors: Application    3.2.3   identifies the methodologies used.
of the ENC Methodology to BWR        -
Reloads", Exxon Nuclear Company, Richland, WA.
: 7. EMF-2158(P)(A), "Siemens Power          3.2.1   Provides an advanced neutronics code Corporation Methodology for            3.2.2   package for performing neutronics, Boiling Water Reactors: Evaluation     3.2.3   thermal hydraulics and transient and Validation of                              analyses.
CASMO-4/MCROBU RN-B2",
Siemens Power Corporation, Richland, WA.
: 8. XN-NF-80-19(P)(A) Volume 3,             3.2.2   Provides overall methodology for "Exxon Nuclear Methodology for                 determining a MCPR operating limit.
Boiling Water Reactors, THERMEX Thermal Limits Methodology Summary Description", Exxon Nuclear Company, Richland, WA.
.9. XN-NF-84-105(P)(A) Volume 1,            3.2.2   Provides a capability to perform "XCOBRA-T: A Computer Code for                 analyses of transient heat transfer BWR Transient Thermal-Hydraulic                behavior in BWR assemblies.
Core Analysis", Exxon Nuclear Company. Richland, WA.
: 10. ANF-524(P)(A), "ANF Critical           3.2.2   Provides a methodology for the Power Methodology for Boiling                  determination of thermal margins, Water Reactors", Advanced                      specifically the MCPR safety limit.
Nuclear Fuels Corporation, Richland, WA.
: 11. ANF-913(P)(A), Volume 1,               3.2.2   Provides a computer program for "COTRANSA2: A Computer                         analyzing BWR system transients Program for Boiling Water Reactor Transient Analyses", Advanced Nuclear Fuels Corporation, Richland, WA.
: 12. XN-NF-825(P)(A) Supplement 2,          3.2.2   Extends previously approved topical BWR6 Generic Rod withdrawal                    report for the CRWE transients for Error Analysis, MCPRp for Plant                BWR/6 plants operating in the Operations within the Extended                 extended operating domain.
Operating Domain, Exxon Nuclear Company, Richland, WA.
to GNRO-2007/00071 Page 11 of 12
: 13. ANF-1358(P)(A), "The Loss of          3.2.2 Presents a generic methodology for Feedwater Heating Transient in             evaluating the loss of feedwater Boiling Water Reactors", Advanced          heating event.
Nuclear Fuels Corporation,
   ,,Richl and, WA..
: 14. EMF-1997(P)(A) Revision 0,            3.2.2  Presents an approved critical power ANFB-10 Critical Power                     correlation for ATRIUM-10 fuel.
Correlation", Siemens Power Corporation, Richland, WA.
: 15. EMF-1997(P), Supplement 1             3.2.2  Presents experimental results which (P)(A), "ANFB- 10 Critical Power           justify the local peaking limit approved Correlation: High Local Peaking            for fuel designs.
Results", Siemens Power Corporation, Richland, WA
: 16. EMF-2209(P)(A), "SPCB Critical        3.2.2  Presents an improved critical power Power Correlation", Siemens                correlation for use with the ATRIUM-1 Power Corporation, Richland, WA.           fuel designs.
.17. EMF-2245(P)(A), "Application of      3.2.2  Provides direct and indirect Siemens Power Corporation's                approaches to develop parameters Critical Power Correlations to Co-          necessary to appropriately model co-Resident Fuel, Siemens Power                resident fuel with an approved critical Corporation, Richland, WA.         ,        power correlation.
: 18. EMF-2361(P)(A), "EXEM BWR-            3.2.1  Provides an evaluation model 2000 ECCS Evaluation Model,"                methodology for licensing analyses of Framatome ANP, Richland, Inc.               postulated LOCAs in jet pump BWRs.
The methodology was developed to comply with 10 CFR 50.46 and Appendix K criteria to 10 CFR 50.
: 19. Deleted                                NA    NA
: 20. Deleted                                NA    NA
: 21. Deleted                                NA    NA
: 22. EMF-CC-074(P)(A), Volume 4,           3.2.4  Describes methodology for stability "BWR Stability Analysis                    analysis with input from the Assessment of STAIF with Input              MICROBURN-B2 reactor core from MICROBURN-B2", Siemens                simulator.
Power Corporation, Richland, WA.
: 23. EMF-2292(P)(A), "ATRIUM-10:           3.2.1 Provides measured cladding Appendix K Spray Heat Transfer             temperatures from spray heat transfer Coefficients", Siemens Power               tests to justify the use of Appendix K Corporation, Richland, WA.                 coefficients for ATRIUM-10 fuel LOCA analyses.
to GNRO-2007/00071 Page 12 of 12
: 24. NEDE-24011 -P-A, General             3.2.1  GE neutronic, thermal, hydraulic, Electric Standard Application for    3.2.2  transient analysis and LOCA Reactor Fuel (GESTAR-II) with        3.2.3  analysis methodology.
exception, to the misplaced fuel    3.3.2.1 bundle analyses as discussed in GNRO-96/00087 and the generic     I MCPR Safety Limit analysis as discussed in GNRO-96/00100; letters from C. R. Hutchinson to USNRC.

Attachment 2 GNRO-2007100071 Proposed Technical Specification Changes (mark-up includes affected Operating License page) to GNRO-2007/0007 1 Page 1 of 2 (bl      SERiI      +/-iq rqu-Irec           to notify t~he NMRCin writing prcr &#xfd;        any change in (i, '&#xfd;tW                          or~
cod o-~of an'y an't-; Cr "x' -3t~n                      -a-a enn executed as part of r-,h ab'ove autflna            -ra.cial           trans-ct~on's                     crhe GCNS Uru&#xfd; 1 opeazaing agreOC!Mrot,                                         tile eyisting prpet                        insurance' coeaci              1ftor      c'cils Unit 1 chact              would La&#xfd;eclally ale                  tb
                             . p       ;.nain               and condit'a'-!&#xfd; ;e-.                 i-rh       nr- h Stft!&#xfd;       Sic          - val~nti-n       R~por-    daziad Jr.                           RI ii&#xfd; rte~iured cao ncif SE~c                                             0'tRc t''.t ansfy      eoi-n       by a lessor or cc.,*?r              t-.,ucer~so' in
                             ,nzrer                     EP- tha~t na&#xfd;ve Sa                            an effect o-' uhe C    he                   :,"*-ai~i be dcy-axlid to- coatain and In Slh-t~to ho condizions specAifid in zbo Covmisio's egalationn set forth -In 'IQ1C*F (aotf-..
                   , anld 'is;                      to all applicablo prc'viai-on                        o-f the Act and to the r-ules, rgulaiorns, and <:rdt:z-Y: o' the
(,camis!ao now or 1-eeafter in. eff-aat: and i.'
                          ~u-'rto t'he addizional. conditiOnz sPec f'ed 0 Et~nercy                                         is authori~ed to apra:jce In*-a'sTc.
tnc, tfI(:lity w:. reactory core powier levaeis not in iexceoa-i of &#xfd;39S =gwa                           thermal (1qO p'&#x17d;rcent Pow.er) 7n accordance -wr'-h the condicions !&#xfd;Veifiid hexe+/-in.
The 7ecnical Specificatioris containe~d In All)                                          dx and ther~                     aaa          ProtectiOn P'r-tca       Conayinedn Apond~ Bn               r<,--i Le!thrctzgh A~nn' are herey'c          ~r~               ted into tthi;'Ee~'a                       nrer Opert,ionr'. Inc. si'l.1 Operae t b                                  lt t        lit accordance? wich the Technical Spcif--a!1aoL onn and the Envlror~aiental P'-arectlon Plan,
                  -he S. .-            lance' Raqviremo&#xfd;nts {$P-) fo'r D-esI Gen~ervitor 12 containead in Cna Tachnical-- speimi,Jton-ran listed bolow. aro -not requirec"toN perfrmued iimenite-jupo~n imiplemnentation of. Amendment-a :Mo 1 The SRs liirted belvw s-hall be suc-cessfo ly der.oi'Strited ath ext regularly chdl                             perrorntia SR3..I0                    and krdnon~rcn     1&#xfd;!(..t to GNRO-2007/00071 Page 2 of 2 C        iavr                       Reotn             eremer;
                                      ~ Reo5.6
                                   &#xa3;o~n 22~. EVc*M--074(W~A),        Womun's 4, "SW Stability Analysis         pit Assesmet of STA!F with Input from MCROBURN-62",
Yiemen.. Power Corporation, Richla~nd, WA.
: 23. EMF-2292(P)(A). "ATRIUM-10 Appendix KSpray Heat Transfer Coefficients"..$iemnens Power coroorat ion, R1 ch rid,WK.
20    no. 24E011P.P~AQ,GQnrj1* PlectrkcSa~r         Aj licatia n MOISpl        Je cWON nlan, ys is     a d c'jcsed    1n RV-96/0081   WI thj g~herl 'I CPR odfey Wi~l aria Qs1 as inINRO.9W -~100. 1rtt rs fromi C.
1011                                   Hutchins;on to USN1                                            t2.P 0   The core opera~ting lim'its shall be- determi~ned such tht l applicable limits (e.g., fuel thermal me hanical limis cor therm~al hydraul ic limits , Emeergjency Corn Cool ing System (ECCS) limits, nucear limits such do V>1. transient analysis limits , and AccidIent analysis 1limitst)   0f the safety ainalyi aro e t-.
: d. TeQLRt incl uding any [ntidcyC1 e revi inlr,      or ~ppi mentt, stall be provid~ed i4#n iisuance fo' each relo-adc cyt1 -e tehe

Attachment 3 GNRO-2007/00071 List of Regulatory Commitments to GNRO-2007/00071 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
TYPE (Check one)         SCHEDULED ONE-     CONTINUING      COMPLETION COMMITMENT                         TIME     COMPLIANCE         DATE (If ACTION                      Required)
If analyses indicate that a change to the MCPR             X                         3/14/08 safety limit or other TS change is required, the TS change will be requested by a separate LAR.
AREVA is currently evaluating the impact of the             X                         3/14/08 error, but it is expected to be insignificant and not impact the use of GEXL97. Entergy will provide a supplement to this LAR describing the impact, if any, to the GEXL97 report.
The industry is working with the NRC staff to resolve      X                          After the TS safety limit issue. Entergy is following this                                industry effort and will request appropriate TS changes or an                              resolution of extended applicability range for the GEXL97                                         GE part 21
,correlation after a mutual resolution is reached by                                  report, the industry and NRC staff.                                                         SC05-03}}