ML082770406

Issuance of Amendment No. 179, Regarding Changes to Technical Specification 5.6.5, Core Operating Limits Report (Colr)

ML082770406
Person / Time
Site:	Grand Gulf  (NPF-029)
Issue date:	10/16/2008
From:	Lyon C Plant Licensing Branch IV
To:	Entergy Operations
Lyon C Fred, NRR/DORL/LPL4, 301-415-2296
References
TAC MD7493
Download: ML082770406 (20)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 16, 2008 Vice President, Operations Entergy Operations, Inc.

Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150

SUBJECT:

GRAND GULF NUCLEAR STATION, UNIT 1 -ISSUANCE OF AMENDMENT RE: CHANGES TO TECHNICAL SPECIFICATION 5.6.5, "CORE OPERATING LIMITS REPORT (COLR)" (TAC NO. MD7493)

Dear Sir or Madam:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 179 to Facility Operating License No. NPF-29 for the Grand Gulf Nuclear Station, Unit 1. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated December 5,2007, as supplemented by letters dated July 21 and August 28,2008.

The amendment changes 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 Operations, Inc., to use a Global Nuclear Fuel method to determine fuel assembly critical power of AREVA ATRIUM-10 fuel. Additionally, the amendment makes an administrative change to an existing reference in TS 5.6.5.

A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely, Carl F. Lyon, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-416

Enclosures:

1. Amendment No. 179 to NPF-29

2. Safety Evaluation cc w/encls: See next page

Grand Gulf Nuclear Station (9/24/2008) cc:

Attorney General Department of Justice State of Louisiana P.O. Box 94005 Baton Rouge, LA 70804-9005 Additional Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY OPERATIONS, INC.

SYSTEM ENERGY RESOURCES, INC.

SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION ENTERGY MISSISSIPPI. INC.

DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 179 License No. NPF-29

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Entergy Operations, Inc. (the licensee), dated December 5,2007, as supplemented by letters dated July 21 and August 28, 2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

- 2

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Facility Operating License No. NPF-29 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 179 are hereby incorporated in the license.

Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License No. NPF-29 and the Technical Specifications Date of Issuance: October 16, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 179 FACILITY OPERATING LICENSE NO. NPF-29 DOCKET NO. 50-416 Replace the following pages of the Facility Operating License No. NPF-29 and the Appendix A, Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Operating License Remove -4 Technical Specifications Remove 5.0-21 5.0-21

(b)

SERI is required to notify the NRC in writing prior to any change in (i) the terms or conditions of any new or existing sale or lease agreements executed as part of the above authorized financial transactions, (ii) the GGNS Unit 1 operating agreement, (iii) the existing property insurance coverage for GGNS Unit 1 that would materially alter the representations and conditions set forth in the Staff's Safety Evaluation Report dated December 19, 1988 attached to Amendment No. 54.

In addition, SERI is required to notify the NRC of any action by a lessor or other successor in interest to SERI that may have an effect on the operation of the facility.

C.

The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level Entergy Operations, Inc. is authorized to operate the facility at reactor core power levels not in excess of 3898 megawatts thermal (100 percent power) in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.179 are hereby incorporated into this license.

Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance Requirements (SRs) for Diesel Generator 12 contained in the Technical Specifications and listed below, are not required to be performed immediately upon implementation of Amendment No. 169.

The SRs listed below shall be successfully demonstrated at the next regularly scheduled performance.

SR 3.8.1.9, SR 3.8.1.10, and SR 3.8.1.14 4

Amendment No. 179

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 Core 0Reratinq Limits ReRort (COLR) (continued)

21. NEDE-33383-P, "GEXL97 Correlation Applicable to ATRIUM 10 Fuel," Global Nuclear Fuel.

22. EMF-CC-074(P)(A), Volume 4, "BWR Stability Analysis Assessment of STAIF with Input from MICROBURN-B2",

Siemens Power Corporation, Richland, WA.

23. EMF-2292(P)(A), "ATRIUM-I0 Appendix K Spray Heat Transfer Coefficients", Siemens Power Corporation, Richland, WA.

24. NEDE-24011 -P-A, General Electric Standard Application for Reactor Fuel (GESTAR-II).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle reV1Slons or supplements, shall be provided upon issuance for each reload cycle to the NRC.

GRAND GULF 5.0-21 Amendment No. ++J, J~

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 179 TO FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS, INC.. ET AL.

GRAND GULF NUCLEAR STATION. UNIT 1 DOCKET NO. 50-416

1.0 INTRODUCTION

By application dated December 5, 2007 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML073440113) (Reference 1), as supplemented by letters dated July 21 and August 28.2008 (ADAMS Accession Nos. ML082070087 and ML082470361, respectively) (References 2 through 4), Entergy Operations, Inc. (Entergy, the licensee),

requested changes to the Technical Specifications (TSs) for Grand Gulf Nuclear Station, Unit 1 (GGNS).

The proposed changes would revise 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 the licensee to use a Global Nuclear Fuel (GNF) method to determine the fuel assembly critical power of AREVA (formerly Framatome ANP) ATRIUM-10 fuel. Additionally, the licensee proposes an administrative change to an existing reference in TS 5.6.5.

GGNS currently operates with a full core of ATRI UM-1 0 fuel. The licensee intends to load GE14 fuel during its upcoming fall 2008 refueling outage. Entergy plans to use the GEXL97 correlation to determine overall core operating limits for GGNS operating Cycle 17, which begins with the fall 2008 refueling outage. The proposed changes support the licensee's tr.ansition from ATRI UM-1 0 to GE14 fuel and associated analytical methods.

The supplements dated July 21 and August 28. 2008, provided additional information that clarified the application. did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on December 31. 2007 (72 FR 74358).

- 2

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 50.34(b), "Final safety analysis report," requires, in part, that:

The final safety analysis report shall include...the following: (4) A final analysis and evaluation of the design and performance of structures, systems, and components with the objective [of assessing... the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.]

As part of the reload design process, the licensee (or its vendor) performs reload safety analyses with approved methodologies to ensure that the design cycle will continue to meet the applicable regulatory criteria. To confirm that the analyses remain acceptable, the licensee confirms that key results of the safety analyses, such as the critical power ratio (CPR), are conservative with respect to the current design cycle. If key safety analysis results are not acceptable, a re-analysis or reevaluation of the affected transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied.

GGNS TS 5.6.5, "Core Operating Limits Report (COLR)," requires, in part, that "[t]he analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC...."

1a CFR Part 50, Appendix A, General Design Criterion (GDC)-10, "Reactor design," requires that:

The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

GDC-12, "Suppression of reactor power oscillations," requires that:

The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

GDC-14, "Reactor coolant pressure boundary," requires that:

The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

GDC~35, "Emergency core cooling," requires that:

A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following

- 3 any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal water reaction is limited to negligible amounts.

In its review, the NRC staff used the guidance of applicable sections of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants."

3.0 TECHNICAL EVALUATION

The licensee proposes to include a new correlation for analyzing the critical power performance of the transition cores from ATRIUM-1 0 fuel to GE14 fuel. In addition, with the introduction of a new vendor fuel, the licensee elected to transition their analytical methods from its vendor AREVA to General Electric (GE).

To justify these changes, the licensee submitted information supporting the use of GEXL97. In addition, the licensee provided information to demonstrate the compatibility of GE14 fuel with the co-resident fuel. The licensee provided information showing the applicability of the analytical methods used in the licensing calculations. The licensee analyzed the affected licensing basis events based on the GE methods and showed that the results meet the applicable acceptance criteria.

The NRC staff has reviewed the licensee's application (Reference 1) in conjunction with the supplemental information (References 2 through 4) to (1) evaluate the acceptability of the GEXL97 correlation, (2) evaluate the use of the associated GE analytical methods for licensing applications, and (3) confirm that the effects of the proposed changes on the licensing basis analyses are acceptable.

3.1 GEXL97 The data for the GEXL97 development specific to ATRIUM-1 0 fuel was generated using the NRC-approved SPCB correlation developed by AREVA (Reference 5; SPCB is the AREVA critical power correlation for ATRIUM-10 fuel). The database consisted of ATRIUM-10 sub bundle and full bundle critical power data generated by the SUb-channel code XCOBRA, incorporating the NRC-approved SPCB correlation. The objective of this data collection was to obtain ATRIUM-1 0 quality data appropriate for GEXL analysis.

The span of the data collection encompasses cosine, top-peaked, bottom-peaked, and double-humped axial power shapes. This data was generated to cover the complete range of expected operation of the ATRIUM-10 fuel in the GGNS boiling-water reactor (BWR) core. The data was used to develop a new GEXL correlation for the ATRIUM-10 design, designated as GEXL97. The GEXL97 correlation uses the same functional form as previous GEXL correlations with different constants for the GEXL correlation coefficient parameters.

The GE critical quality-boiling length correlation (GEXL) was developed to accurately predict the onset of boiling transition in BWR fuel assemblies during both steady-state and reactor transient conditions. The GEXL correlation is necessary for determining the minimum critical power ratio (MCPR) operating limits resulting from transient analysis, the MCPR safety limit analysis, and the core operating performance and design. The GEXL correlation is an integral part of the

- 4 transient analysis methodology. It is used to confirm the adequacy of the MCPR operating limit, and it can be used to determine the time of onset of boiling transition in the analysis of other events. The NRC staff's review considered the following: 1) adequacy of the database generated with the sub-channel code XCOSRA, 2) proper determination of the uncertainty in GEXL97 correlation predictions for the ATRIUM-1 0 fuel design, and 3) applicability of the proposed operating range of GEXL97 correlation to the ATRIUM-10 fuel.

3.1.1 Validity of the Database and Associated Uncertainties ATRIUM-10 fuel is a 10x10 fuel bundle with a water channel design that displaces nine fuel rods. It contains a total of 83 full-length fuel rods and 8 part-length rods. It has 27 unique fuel rod locations within the 1Ox1 0 lattice for which dryout data was collected.

The SPCS correlation for the ATRIUM-10 fuel, as encoded in the sub-channel computer code XCOSRA, is used to generate a database of predicted critical power values for a range of operating conditions corresponding to the range of the ATRIUM-10 correlation. This database was then treated in the same way as an experimental database, using the approved methodology for GEXL correlation development. Utilizing this approach, GNF produced a new form of the GEXL correlation, namely GEXL97, applicable only to GGNS and ATRIUM-10 fuel design.

The data for the GEXL97 development specific to ATRIUM-10 fuel was generated using NRC-approved AREVA SPCS correlation encoded in the above stated sub-channel code.

Specified rod-to-rod peaking factors, axial power shapes, pressure, mass flux and sub-cooling were used with the AREVA SPCS correlation to determine critical power at dryout.

Therefore, the NRC staff concludes that generating the hypothetical databases using the SPCS correlation encoded in the subchannel code XCOSRA is a reasonable engineering approach to dealing with mixed core fuel, where the experimental database and critical power correlation for the previous vendor's fuel is not available to the new vendor.

3.1.2 Determination of Uncertainties The database used in the development of the GEXL97 correlation for ATRIUM-10 fuel was provided by the licensee in Table 2-1 of NEDC-33383-P (Attachment 4 of Reference 1). This table shows the number of calculated critical power data points obtained using the AREVA critical power correlation for cosine, inlet, outlet, and double-humped axial power distributions.

It also shows the fuel pin dryout location that formed the basis of the 28 different sets of AREVA-calculated critical power data. Table 2-2 of the same document provides additional information by further dividing the data collected into SUbgroups of pressure, mass flux, and inlet sub-cooling.

The GEXL97 database generated in this manner is artificial in construct, created with a computer code which has implemented the SPCS correlation, and can at best only approximate the actual critical power raw data behavior of the ATRIUM-10 fuel. However, with reasonable engineering practices, and proper statistical accountability, the database can predict the critical power behavior with acceptable uncertainties. Testing the hypothetical databases as if it were

- 5 real data in the regression analysis, therefore, introduces unavoidable error into the correlation being derived from it.

As stated earlier, the database for the GEXL97 development specific to ATRIUM-10 fuel was generated using the NRC-approved SPCS correlation. The database consisted of ATRIUM-10 sub-bundle and full bundle critical power data generated by the sub-channel code XCOSRA, incorporating the NRC-approved SPCS correlation. The local critical power values predicted with the approved SPCS correlation can be expected to vary over the range of the database.

Since the GEXL97 correlation is fitted to this "hypothetical" database, the error in the critical power prediction of the GEXL97 correlation for a given set of conditions will have some additional error relative to the real critical power value for those conditions, over and above the uncertainty of the correlation's fit to the hypothetical database. Therefore, the approach of the correlation procedure can be valid only if the overall uncertainty in the new GEXL97 correlation is appropriately characterized in terms of the uncertainty in its fit to the hypothetical database and the uncertainty of the critical power values in the hypothetical database itself.

The treatment of the overall uncertainty of the GEXL97 correlation for ATRIUM-10 fuel, as originally presented in the licensee's submittal, is complete, in that GE used a statistical combination of uncertainty techniques to appropriately combine the uncertainty of the fit of GEXL97 correlation to the hypothetical database and the uncertainty of the database itself, which is a function of the uncertainty of SPCS correlation.

Therefore, the NRC staff concludes that the total uncertainty in the correlation's critical power predictions appropriately accounts for the uncertainty in the new correlation's fit to the hypothetical database and the uncertainty in the hypothetical database with respect to the underlying experimental data are appropriately treated.

3.1.3 Generation of the GEXL97 Correlation and the Range of Applicability The licensee stated that, in developing the GEXL97, GE took steps to optimize GEXL97 critical power predictions for the ATRIUM-10 fuel design, and to minimize the prediction uncertainty.

This process is identical to that used by GE when developing GEXL correlation coefficients for GNF/GE fuel designs using raw test data, and has been used in the past development of GEXL correlations applicable to other legacy fuel.

The procedure used for development of the GEXL97 correlation is summarized below:

a) First, a range of generated data covering all parameter(s) variations is selected to form a correlation development database. This database consists of the majority of the generated data. A separate dataset is set aside to form a correlation verification database.

b) The GEXL97 correlation coefficients are then chosen to minimize the bias and standard deviation in correlating the development database, and to minimize any trend errors in reference to flow, pressure, sub-cooling, and R-factor (the R-factor is an input to the correlation that accounts for the effects of the fuel rod distributions and the fuel assembly and channel geometry on the fuel assembly critical power).

- 6 c) Once the optimum coefficients were determined, the apparent R-factors are calculated for each assembly. The apparent R-factor is defined as that R-factor which yields an overall ECPR of 1.0 for a given assembly. ECPR is defined as the ratio of the GEXL97 calculated critical power to the SPCS calculated critical power.

d) A final set of additive constants (Table 4-2 of Reference 1) are determined by adjusting the preliminary additive constants subject to minimizing the difference between the R-Factors.

The range of application for the GEXL97 correlation as stated in the submittal (Section 4.2 of Reference 2), is the same as the range of the hypothetical database over which the correlation is derived, and within the AREVA SPCS development database. The application range covers the complete range of expected operation of the ATRIUM-10 fuel during normal steady-state and transient conditions in the GGNS SWR core. Therefore, the licensee's use of the new GEXL97 correlation within the limits of the hypothetical database, bounded by the experimental limits of the ATRIUM-10 database, is acceptable.

3.2 Thermal-Hydraulic Compatibility of the GE14 fuel with the ATRIUM-10 fuel The supplemental information provided by the licensee (References 2 and 3) provided independent verification of the conclusion made by the fuel vendor (GE) that the GE14 and ATRIUM-10 fuels are thermal-hydraulically compatible. The next three cycles at GGNS will be designated as mixed cores with the core comprised of ATRIUM-10 fuel and GE14 fuel.

Cycle 17 will consist of 2/3 core of ATRI UM-1 0 fuel and 1/3 core of GE14 fuel; Cycle 18 will consist of 1/2 core of ATRIUM-10 fuel and 1/2 core of GE14 fuel; and Cycle 19 will consist of 1/3 ATRIUM-10 fuel and 2/3 GE14 fuel. Consequently, GE performed calculations to verify the mixed core calculations results regarding the similarity in thermal-hydraulic performance of the GE14 and ATRIUM-10 fuel designs. Data provided by AREVA was used by GE to develop computer code models to perform the various evaluations.

Specifically, GE investigated the thermal-hydraulic compatibility between GE14 and ATRIUM-10 through a series of mixed cores, progressing from the full core of ATRIUM-10 fuel to a full core of GE14 fuel. The mixed core simulations analyses projected the performance of both fuel types during transition cores, going from a full core of ATRIUM-10 fuel to a full core of GE14 fuel. During the core transition cycles, at least once burned ATRIUM-10 assemblies are placed at the core periphery.

GE also performed evaluations to demonstrate compliance with safety and performance criteria, including core nuclear design and the thermal-hydraulic critical power correlations for the ATRIUM-10 fuel. GE calculations provided confirmation that the thermal-hydraulic performance characteristics applied in the calculations met specific acceptance criterion associated with thermal-hydraulic compatibility of GE14 fuel and the legacy fuel.

3.3 Use of Approved Analytical Methods Analytical methods (e.g., computer codes, correlations) used to support licensing calculations are generally documented in topical reports (TRs), which may be reviewed by the NRC staff on a generic basis. In the NRC staff safety evaluation (SE) approving the TR, the staff defines the

- 7 basis for acceptance in conjunction with any limitations and conditions on use of the TR, as appropriate. In Reference 4, the licensee listed the topical reports and analytical methods used for each affected analysis and stated that such use is consistent with the corresponding NRC staff approval. The analytical methods used for GGNS are summarized in Table 1. The NRC staff "finds that the topical reports and analytical methods used by the licensee are acceptable.

3.4 Licensing Basis Analyses In Reference 4, the licensee provided plant-specific information to show and justify the effect of the fuel and methodology change for GGNS. Specifically, the licensee performed analyses of the limiting updated final safety analysis report (UFSAR) events, using the GE methods, to demonstrate that the results of the analyses meet the applicable acceptance criteria. The information provided by the licensee is summarized in Table 1.

The events analyzed included:

Limiting Anticipated Operational Occurrences (AOOs; Turbine Trip with no Bypass, Load Rejection with no Bypass, Feedwater Controller Failure)

American Society of Mechanical Engineers (ASME) Overpressure (Main Steam Isolation Valve Closure with Flux Scram)

Stability Loss-of-Coolant Accident (LOCA)

Anticipated Transient without Scram (ATWS; Main Steam Isolation Valve Closure, Pressure Regulator Failure Open)

The plant responses to the limiting AOOs are analyzed for each reload cycle to establish the operating limit minimum critical power ratio (OLMCPR). The ASME overpressure analysis is performed to ensure that vessel pressurization following a limiting transient is within an acceptable limit. The licensee performed a reload analysis to cover the projected operating conditions within the licensed power-to-flow map, the expected core exposures, and equipment out-of-service options. In Reference 4, the licensee provided the OLMCPR limits for GE14 and co-resident ATRIUM-1 0 fuel and showed that the ASME overpressure results are acceptable.

GGNS is currently operating under the requirements of the reactor stability Long-Term Solution Enhanced Option I-A (E1A) approved by the NRC staff in GE Licensing Topical Report NEDO-32339-A and its supplements (References 10 through 14). Option E1 A is a solution based exclusively on prevention of instabilities. A conservative exclusion region where instabilities are unlikely is defined or validated by cycle-specific calculations based on an approved procedure. This exclusion region is enforced automatically by the reactor protection system. The licensee validated the exclusion region by showing that core and channel decay ratios are within acceptable levels based on the approved ODYSY methodology (Reference 9).

The emergency core cooling system (ECCS) is designed to mitigate postulated LOCAs due to ruptures in the primary system piping. The ECCS performance under all LOCA conditions and

- 8 the analysis models must satisfy the requirements of 10 CFR 50.46 and 10 CFR Part 50, Appendix K. The analysis methodology used for the GGI\\JS LOCA analysis is the SAFER/GESTR-LOCA evaluation model (References 15 through 19), which has been approved by the NRC. The licensee stated in Reference 4 that the limiting power to flow condition is the Maximum Extended Loadline Limit Analysis (MELLLA) point, having the limiting power with lowest core flow. The licensee evaluated both mid-peaked and top-peaked axial power shapes.

The limiting axial power shape is identified as mid-peaked for large breaks and the limiting axial power shape for small breaks is identified as top-peaked. Both mid-peaked and top-peaked axial power shapes were considered. The licensee recalculated the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for the co-resident fuel applicable to the transition cycles based on the SAFER/GESTR method. The licensee showed that the ECCS LOCA acceptance criteria are met by considering the potentially limiting initial conditions.

ATWS is defined as an AOO followed by the failure of the reactor protection system. To demonstrate acceptability, the ATWS analysis must show that (1) the peak vessel bottom pressure is less than the ASME service level C limit of 1500 pounds per square inch gauge (psig); (2) the peak clad temperature (PCT) is within the 10 CFR 50.46(b)(1) limit of 2200 degrees Fahrenheit (OF); (3) the peak suppression pool temperature is less than the design limit (185 of for GGNS); and (4) the peak containment pressure is less than the containment design pressure (15 psig for GGNS). The licensee performed a plant-and cycle specific ATWS analysis for GGNS Cycle 17 conditions. The analysis results meet the acceptance criteria, as summarized in Table 1.

Based on the information submitted by the licensee, the NRC staff finds that the results of the analysis meet the applicable acceptance criteria. The licensee confirmed that the CPR safety analyses remain bounding, and that key inputs to the safety analyses are conservative with respect to the current design cycle. Therefore, the staff concludes that the impact of the fuel and methodology change on the safety analysis for GGNS is acceptable.

Table 1 - Limiting Analysis Results and Computer Codes Used in the Analysis Analysis Code{s)

Used Staff Approval Key Parameter(s)

Result vs.

Acceptance Criteria Anticipated Operational Occurrence (AOO)

ISCOR09 PANAC11 ODYN09V TASC03 Refs. 6, 7, and 8 MCPR All equipment in service condition OLMCPR range of 1.18 to 1.33 (See Note 1)

ASME Overpressure ISCOR09 PANAC11 ODYN09V Refs. 6, and 7 Peak Dome Pressure (psig)

Peak Vessel Pressure (psig) 1276 (S1325) 1308 (S1375)

Stability ISCOR09 PANAC11 ODYSY05 Refs. 9,10, 11, 12, 13, and 14 Channel Decay Ratio Core Decay Ratio 0.307 <<0.8 ODYSY Criteria) 0.754 <<0.8 ODYSY Criteria)

ECCS-LOCA ISCOR09 LAMB08 SAFER04/GE Refs. 15, 16, 17,18,and 19 PCT (OF)

Max local oxidation (%)

Core wide Metal-water reaction (%)

1880 (S2200) 3 (S17) 0.1 (S1.0)

- 9 Analysis Code(s)

Used Staff Approval Key Parameter(s)

Result vs.

Acceptance Criteria STR08 TASC03 ATWS ISCOR09 PANAC11 ODYN09V STEMP04 TASC03 Ref. 6, 7, and 8

Peak Vessel Pressure (psig)

Peak Suppression Pool Temperature (OF)

Peak Containment Pressure (psig)

Peak Cladding Temperature (OF)

Peak Local Cladding Oxidation (%)

1299 (:51500) 170 (:5185) 8.2 (:515) 1509 (:52200)

Insignificant (:517)

Note 1: The AOO analysis determines the OLMCPR such that a worst case transient would not violate the safety limit MCPR (SLMCPR). In Reference 4, the licensee showed that, with all equipment in service and an OLMCPR range from 1.18 to 1.33, spanning the GE14 and ATRIUM-10 fuels from beginning-to end-of-cycle exposure conditions, the SLMCPR would not be violated.

3.5 Technical Specification Changes TS 5.6.5.b provides a list of topical reports (TRs) documenting the NRC-approved methodologies used to determine the values of cycle-specific parameters included in the COLA.

The licensee proposes to add the following TR to the reference list:

NEDC-33383P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," Globall'Juclear Fuel As discussed above, the NRC staff finds that the GEXL97 method documented in the referenced TR is acceptable for use in support of GGNS licensing applications. Therefore, the staff finds that the addition of the TR is acceptable.

The licensee also proposes to revise Reference 24 of TS 5.6.5.b to remove the exceptions to the use of NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR-II)." The exceptions were previously approved by the NRC to address conditions that no longer exist at GGNS. The removal of the exceptions does not result in any change to the licensee's method of performing the analyses, which is consistent with the NRC-approved GESTAR-II. Therefore, the staff finds the change acceptable.

3.6 Conclusion The NRC staff reviewed the analyses and results provided by the licensee in References 1 through 4 and determined that the analyses and results comply with 10 CFR 50.34(b), and are consistent with the guidance of applicable sections of NUREG-0800.

The staff concludes that the use of GEXL97 is acceptable for the following reasons:

a) the total uncertainty in the correlation's critical power predictions appropriately accounts for the uncertainty in the new correlation's fit to the hypothetical database and the uncertainty in the hypothetical database with respect to the underlying experimental data are appropriately treated;

- 10 b) generating the hypothetical databases using the SPCS correlation encoded in the subchannel code XCOSRA is a reasonable engineering approach to dealing with mixed core fuel, where the experimental database and critical power correlation for the previous vendor's fuel is not available to the new vendor; c) the licensee intends to utilize the new GEXL97 correlation within the limits of the hypothetical database, bounded by the experimental limits of the ATRIUM-10 database; and d) the licensee confirmed that the CPR safety analyses remain bounding, and that key inputs to the safety analyses are conservative with respect to the current design cycle.

In addition, the staff finds that the introduction of the GE14 fuel will not adversely impact the performance of the ATRIUM-10 fuel, and that the two distinct fuel designs are thermal hydraulically compatible. The staff finds that the use of the proposed analytical methods for GGNS is consistent with the corresponding NRC staff approval and that the results of the analyses meet the applicable acceptance criteria. Therefore, the staff concludes that the safety analysis impact of the fuel and methodology change for GGNS is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Mississippi State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on December 31, 2007 (72 FR 74358). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

- 11

7.0 REFERENCES

1.

Letter from W. Brian (Entergy) to NRC, dated December 5,2007 (ADAMS Accession No. ML073440113).

2.

Letter from M. Krupa (Entergy) to NRC, dated July 21,2008 (ADAMS Accession No. ML082070087).

3. to Reference 2 (ADAMS Accession No. ML082070089, public version).

4.

Letter from M. Krupa (Entergy) to NRC, dated August 28, 2008 (ADAMS Accession l'Jo.

ML082470361 ).

5.

EMF-2209(P)(A) Revision 2, "SPCB Critical Power Correlation," Framatome ANP, September 2003.

6.

NEDE-30130-P-A, "Steady-State Nuclear Methods," April 1985, and for PANACEA11, Letter from S. Richards (NRC) to G. Watford (GE), "Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, GESTAR II Implementing Improved GE Steady-State Methods," (TAC NO. MA6481), November 10,1999.

7.

NEDC-24154P-A, "Qualification of the One-Dimensional Core Transient Model (ODYN) for Boiling Water Reactors, Supplement 4, Volume 1," January 1998.

8.

NEDC-32084P-A, "TASC-03A, A Computer Program for Transient Analysis of a Single Channel, Revision 2," July 2002.

9.

NEDC-32992P-A, "ODYSY Application for Stability Licensing Calculations," July 2001.

10.

NEDO-32339-A, "Reactor Stability Long-Term Solution: Enhanced Option I-A, Revision 1," April 1998.

11.

NEDO-32339-A Supplement 3, "Reactor Stability Long-Term Solution: Enhanced Option I-A, Flow Mapping Methodology, Revision 1," April 1998.

12.

NEDO-32339-A Supplement 4, "Reactor Stability Long-Term Solution: Enhanced Option I-A, Generic Technical Specifications, Revision 1," April 1998.

13.

NEDC-32339P-A Supplement 1, "Reactor Stability Long-Term Solution: Enhanced Option I-A, ODYSY Application to E1A," December 1996.

14.

NEDC-32339P-A Supplement 2, "Reactor Stability Long-Term Solution: Enhanced Option I-A, Solution Design, Revision 1," April 1998.

15.

I'JEDE-20566P-A, "General Electric Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K," September 1986.

- 12

16.

NEDE-23785-1-P-A Rev. 1, "The GESTR-LOCA and SAFER Models tor the Evaluation at the Loss-at-Coolant Accident, Vol. 1, GESTR-LOCA - A Model tor the Prediction at Fuel Rod Thermal Performance," October 1984.

17.

I\\lEDE-23785-1-P-A Rev. 1, "The GESTR-LOCA and SAFER Models tor the Evaluation at the Loss-at-Coolant Accident, Vol. 2, SAFER - Long Term Inventory Model tor BWR Loss-at-Coolant Analysis," October 1984.

18.

NEDE-23785-1-P-A Rev. 1, "The GESTR-LOCA and SAFER Models tor the Evaluation at the Loss-at-Coolant Accident, Vol. 3, SAFER/GESTR Application Methodology,"

October 1984.

19.

I\\IEDE-23785P-A Rev. 1, "The GESTR-LOCA and SAFER Models tor the Evaluation at the Loss-at-Coolant Accident, Vol. 3 Supplement 1, Additional Information tor Upper Bound PCT Calculation," March 2002.

Principal Contributors: T. Nakanishi A. Attard Date:

October 16, 2008

ML082770406

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DATE 10/7/08 9/11/08 10/8/08 10/15/08 10 (1C.(oi lD lib Cii?