ML14325A752

License Amendment Request Application to Revise Current Fluence Methodology

ML14325A752
Person / Time
Site:	Grand Gulf
Issue date:	11/21/2014
From:	Mulligan K J Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML14325A758	List: ML14325A752
References
GNRO-2014/00080
Download: ML14325A752 (16)

Text

Q Entergy Operations, Inc.P.O.Box 756 Port Gibson, MS 39150 Kevin Mulligan Site Vice PresidentAh 2&3C.0.Grand Gulf Nuclear Station ttac ments ontain PR PRIET ARY Information Tel.(601)437-7500 GNRO-2014/00080 Nov 21,2014 u.s.Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

REFERENCES:

License Amendment Request Application to Revise Grand Gulf Nuclear Station Unit 1's Current Fluence Methodology from 0 Effective Full Power Years (EFPY)Through the End of Extended Operations to a Single Fluence Method Grand Gulf Nuclear Station, Unit 1 Docket No.50-416 License No.29 1.Severity Level IV non-cited violation of 10 CFR 50.59,"Changes, Tests, and Experiments involving the licensee's failure to obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a new method of evaluation for determining reactor vessel neutron fluence;Grand Gulf Nuclear Station-NRC Integrated Inspection Report 05000416/2013004, dated November 27,2013 2.U.S.Nuclear Regulatory Commission Letter,"Requests for Additional Information for the Review of the Grand Gulf Nuclear Station, License Renewal Application," dated August 28, 2013 (Accession No.ML 13227 A394)3.Grand Gulf Nuclear Station Letter,"Response to Requests for Additional Information (RAI)set 47," dated September 23,2013 (Accession No.ML 13266A368) 4.U.S.Nuclear Regulatory Commission Regulatory Guide, Regulatory Guide 1.190, dated March 2001 (Accession No.ML01 0890301)

Dear Sir or Madam:

In accordance with the provisions of Section 50.90 of Title 10 Code of Federal Regulations (10 CFR), Entergy Operations, Inc.(Entergy)is submitting a request for an amendment to revise our existing license basis for Grand Gulf Nuclear Station (GGNS), Unit 1.The proposed amendment is to revise Grand Gulf Nuclear Station, Unit 1's license basis to adopt a single fluence method.This change is needed to address a legacy issue in which the GNRO-2014/000BO Page 2 of 3 current method was determined to be utHized without receiving prior Nuclear Regulatory Commission (NRC)approval.Attachment 1 provides a description of the proposed change.Attachment 2 provides the topical report from MP Machinery and Testing, LLC.Attachment 3 provides the report documenting the application of the single fluence method at GGNS.This report also applies to the Maximum Extended Load Line Limit Plus (MELLLA+)License Amendment Request (LAR)in letter2013/00012 (Accession No.ML 13269A140).

Although this request is neither exigent nor emergency, your prompt review is requested.

Once approved, the amendment shall be implemented within 60 days.MP Machinery and Testing, LLC (MPM)considers the information provided in Attachments 2 and 3 to be proprietary and therefore exempt frompublicdisclosure pursuant to 10 CFR 2.390.The proprietary information was provided to Entergy in an MPM transmittal.

Therefore, on behalf of MPM, Entergy requests to withhold Attachments 2 and 3 from public disclosure in accordance with 10 CFR 2.390(b)(1).

This letter contains no new commitments.

If you have any questions or require additional information, please contact Mr.James Nadeau at (601)437-2103.I declare under penalty of perjury that the foregoing is true and correct.Executed on November 21,2014.Sincerely, KJM/ras Attachments:

1.Analysis of Proposed Single Fluence Methodology 2.Topical Report from MP Machinery and Testing, LLC: MPM-614993,"Benchmarking of MPM Methods for BWR Neutron Transport Calculations," November, 2014.3.Single Fluence Method Applied to GGNS: MPM-B14779 Revision 1,"Neutron Transport Analysis for Grand Gulf Nuclear Station," November, 2014.cc: See next page GNRO-2014/00080 Page 3 of 3 cc: with Attachment and Enclosures U.S.Nuclear Regulatory Commission ATTN: Mr.John Daily, NRRlDLR Mail Stop OWFN/11 F1 11555 Rockville Pike Rockville, MD 20852-2378 cc: without Attachment and Enclosures U.S.Nuclear Regulatory Commission ATTN: Mr.Mark Dapas, (w/2)Regional Administrator, Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150 U.S.Nuclear Regulatory Commission ATTN: Mr.Alan Wang, NRRlDORL (w/2)Mail Stop OWFN 8 B1 Washington, DC 20555-0001 Dr.Mary Currier, M.D., M.P.H State Health Officer Mississippi Department of Health P.O.Box 1700 Jackson, MS 39215-1700 Attachment 1 GNRO-2014/00080 Analysis of Proposed Single Fluence Methodology Attachment 1 to GNRO-2014/000BO Page 1 of 12

1.0 DESCRIPTION

The proposed amendment revises the Grand Gulf Nuclear Station, Unit 1 (GGNS)license bases to adopt a single fluence methodology from 0 Effective Full Power Years (EFPY)through the end of extended operations.

This change is needed to address a legacy issue in which the current method was determined to have required Nuclear Regulatory Commission (NRC)approval prior to being utilized, and also applies to the Maximum Extended Load Line Limit Plus Analysis Plus (MELLLA+)License Amendment Request (LAR)(Accession No.ML 13269A140).

The new fluence methodology is referred to below as the"single fluence method" to indicate that all calculations are performed in a consistent manner.The methodology is described in detail in References 1 and 2 which give the results ofthemethodology benchmarking analysis and the application of the methodology to the GGNS fluence analysis covering operation for the first 21 fuel cycles.

2.0 BACKGROUND

During the GGNS License Renewal process, it was determined that the current fluence methodology should have received NRC approval prior to being utilized.This resulted in a Severity Level IV Non-Cited Violation (NCV)of 10 CFR 50.59,"Changes, Tests, and Experiments" involving failure to obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a new method of evaluation for determining reactor vessel neutron fluence, as documented in the Grand Gulf Nuclear Station-NRC Integrated Inspection Report 05000416/2013004, dated November 27,2013.License Renewal Application (LRA)License Amendment Request Response to Request for Additional Information (RAI), set 47, question 5 response, stated, in part: GGNS RAI4.2.1-2c Response 5)A supplemental response to this RAI is being developed and will be provided to the NRC.The supplemental response will provide the answers to the following questions:

a)Pleaseprovide fluence values that have been determined from Beginning Of Life to End of Life Extended in accordance with a single method.i)If the method is NRC-approved insofar as it applies to vessel fluence calculations, provide the reference to the staff-accepted methodology.

ii)If the method is not NRC-approved, provide the plant-specific calculations and documentation, and include sufficient information, to enable the NRC staff to determine whether the calculation adheres to NRC Regulatory Guide (RG)1.190,"Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," or other justification as required to establish that the fluence calculation is acceptable.

iii)Refer to Regulatory Position 3,"Reporting," for the specific documentation required to establish adherence to NRC RG 1.190.

Attachment 1 to GNRO-2014/00080 Page 2 of 12

3.0 TECHNICAL ANALYSIS

3.1 Applicability

Entergy has reviewed the U.S.Nuclear Regulatory Commission Regulatory Guide, RG 1.190 (Reference 3).This guide describes the application and qualification of a methodology acceptable to the NRC staff fordeterminingthe best-estimate neutronfluenceexperienced by materials in the beltline region of light water reactor (LWR)pressure vessels, as well as for determining the overall uncertainty associated with those best-estimate values.Entergy has concluded that the proposed fluence methodology complies with the guidance of RG 1.190, and therefore is appropriate to be utilized at GGNS.3.2 Determination of compliance with RG 1.190 3.2.1 Regulatory Position (RP)Compliance Discussion.

FLUENCE CALCULATION METHODS RP 1.3 Fluence Determination.

Absolute fluence calculations, rather than extrapolated fluence measurements, must be used for the fluence determination.

GGNS Response: All calculations performed using the GGNS proposed methodology use absolute fluence calculations.

This meets the requirement of RP 1.3.

Attachment 1 to GNRO-2014/00080 Page 3 of 12 RP 1.1.1 Modeling Data.The calculation modeling (geometry, materials, etc.)should be based on documented and verified plant-specific data.GGNS Response: The GGNS proposed methodology is based on documented and verified plant-specific data.Further, the calculations use as-built data for plant structures and material compositions whenever these data are available.

The fuel data is specific for each fuel cycle and includes results for power distributions and water densities taken from the fuel depletion analysis.This meets the requirement of RP 1.1.1.RP 1.1.2 Nuclear Data.The latest version of the Evaluated Nuclear Data File (ENDF/B)should be used for determining nuclear cross-sections.

Cross-section sets based on earlier or equivalent nuclear-data sets that have been thoroughly benchmarked are also acceptable.

When the recommended cross-section data change, the effect of these changes on the licensee-specific methodology must be evaluated and the fluence estimates updated when the effects are significant.

GGNS Response: The calculations use the BUGLE-96 cross section set which is based on the latest version (VI)of the Evaluated Nuclear Data File (ENDF/B).This meets the requirement of RP 1.1.2.RP 1.1.2 Cross-Section Angular Representation.

In discrete ordinates transport calculations, a P3 angular decomposition of the scattering cross-sections (at a minimum)must be employed.GGNS Response: All calculations use a P3 angular decomposition in accordance with regulatory guide.This meets the requirement of RP 1.1.2.RP 1.1.2 Cross-Section Group Collapsing.

The adequacy of the collapsed job library must be demonstrated bycomparingcalculations for a representative configuration performed with both the master library and the job library.GGNS Response: All calculations are performed with the BUGLE-96 library which is collapsed to 47 neutron groups.This library has been thoroughly benchmarked.

This meets the requirement of RP 1.1.2.RP 1.2 Neutron Source.The core neutron source should account for local fuel isotopics and, where appropriate, the effects of moderator density.The neutron source normalization and energy dependence must account for the fuel exposure dependence of the fission spectra, the number of neutrons produced per fission, and the energy released per fission.GGNS Response: The neutron source is calculated taking into account changes in neutrons per fission, energy per fission, and the average fission spectrum which develops as the Uranium-235 is burned and other Attachment 1 to GNRO-2014/00080 Page 4 of 12 isotopes, such as Pu239, increase in fission fraction.This meets the requirement of RP 1.2.RP 1.2 End-of-Life Predictions.

Predictions of the vessel end-of-life fluence should be made with a best-estimate or conservative generic power distribution.

If a best estimate is used, the power distribution must be updated if changes in core loadings, surveillance measurements, or other information indicate a significant change in projected f1uence values.GGNS Response: GGNS has calculated the fluence for every cycle up to the present.This gives the most accurate value of the present fluence.Extrapolations to future times are made using best estimate values of future fuel designs in a cycle 21 best estimate projection cycle.As changes in fuel core loadings are made, updated extrapolations will be made.Further, the flux values from the cycle 21 transport calculation were multiplied with a factor of 1.1 applied for projection to exposures after cycle 21.This 10%)conservatism was applied by Entergy to ensure that fluence estimates remain conservative.

Periodiccycle transport updates are planned to ensure the 100/0 factor adequately covers future as-burned fluxes.This meets the requirement of RP 1.2.RP 1.3.1 Spatial Representation.

Discrete ordinates neutron transport calculations should incorporate a detailed radial and azimuthal-spatial mesh of-2 intervals per inch radially.The discrete ordinates calculations must employ (at a minimum)an Sa quadrature and (at least)40 intervals per octant.GGNS Response: The present methodology meets or exceeds these requirements.

All 3D calculations were performed with an S16 quadrature.

The R-8 model included 215 mesh points in the radial direction.

The radial mesh covers the range from the center of the core to about 15 inches into the biological shield.The large number of mesh points was used to accurately calculate the neutron flux transport from the core edge to the outside of the vessel.In the azimuthal direction, 185 mesh points were used to model a quadrant of the reactor.Inspection of the fuel loading patterns indicated that only minor deviations from quarter core symmetry were present, and these were not significant.

However, there are some deviations from octant fuel symmetry and these, together with the quadrant symmetry of the jet pumps, led to the use of the quarter core model.The 185 azimuthal points provided good definition of the variation of the core edge with angle and accurately defined the azimuthal flux variation in the shroud.In the axial direction, the 3D model extends past BAF to about 18 inches below the H7 shroud weld.The model of the region above the top guide extends axially past the top guide weld H1 by an axial extent of 18 inches.This resulted in a total of 246 mesh points in the axial direction, with 105 axial meshes in Attachment 1 to GNRO-2014/00080 Page 5 of 12 the core region.This resulted in a model with over 9.7 million meshes in the 3D model.Thismeets the requirement of RP 1.3.1.RP 1.3.1 Multiple Transport Calculations.

If the calculation is performed using two or more"bootstrap" calculations, the adequacy of the overlap regions must be demonstrated.

GGNS Response: It was not necessary to use bootstrapping for these calculations so this requirement does not apply.RP 1.3.2 Point Estimates.

If the dimensions of the tally region or the definition of the average-flux region introduce a bias in the tally edit, the Monte Carlo prediction should be adjusted to eliminate the calculational bias.The average-flux region surrounding the point location should not include material boundaries or be located near reflecting, periodic, or white boundaries.

GGNS Response: This requirement only applies to Monte Carlo calculations which are not used here.Variance reduction methods were not used.RP 1.3.2 Statistical Tests.The Monte Carlo estimated mean and relative error should be tested and satisfy all statistical criteria.GGNS Response: This requirement only applies to Monte Carlo calculations which are not used here.RP1.3.2 Variance Reduction.

All variance reduction methods should be qualified by comparison with calculations performed without variance reduction.

GGNS Response: This requirement only applies to Monte Carlo calculations which are not used here.RP 1.3.3 Capsule Modeling.The capsule fluence is extremely sensitive to the geometrical representation of the capsule geometry and internal water region, and the adequacy of the capsule representation and mesh must be demonstrated.

GGNS Response: The capsule geometry is modeled using as-built drawings.The water between the capsule and the vessel surface is also modeled usingbuilt drawings.The mesh is refined in the region of the capsule and the adequacy of the mesh has been demonstrated by the accuracy of the CIM ratios for the cycle 1 dosimetry analysis.This meets the requirement of RP 1.3.3.

Attachment 1 to GNRO-2014/00080 Page 6 of 12 RP 1.3.3 Spectral Effects on RTNDT.In order to account for the neutron spectrum dependence of RTNDT, when it is extrapolated from the inside surface of the pressure vessel to the T/4 and 3T/4 vessel locations using the E>1-MeV fluence, a spectral lead factor must be applied to the fluence for the calculation of aRTNDT.GGNS Response: This requirement only applies to extrapolation through the vessel and does not affect the benchmark calculations.

However, when fluence within the vessel is required, the displacement per atom (dpa)extrapolation methodology is applied to vessel calculations as specified in RG 1.99, Revision 2.Data are supplied to enable extrapolation using dpa calculated extrapolation or using RG 1.99 Rev.2 extrapolation to account for the spectral shift.This meets the requirement of RP 1.3.3.RP 1.3.5 Cavity Calculations.

In discrete ordinates transport calculations, the adequacy of the S8 angular quadrature used in cavity transport calculations must be demonstrated.

GGNS Response: No cavity dosimetry work has been performed at GGNS.RP 1.4.1, 1.4.2, 1.4.3 Methods Qualification.

The calculational methodology must be qualified by both (1)comparisons to measurement and calculational benchmarks and (2)an analytic uncertainty analysis.The methods used to calculate the benchmarks must be consistent (to the extent possible)with the methods used to calculate the vessel fluence.The overall calculational bias and uncertainty must be determined by an appropriate combination of the analytic uncertainty analysis and the uncertainty analysis based on the comparisons to the benchmarks.

GGNS Response: An extensive benchmarking program has been carried out to qualify the MPM neutron transport methodology.

All of the requirements of RG 1.190 have been met.In particular, all C/M results fall within allowable limits (+1-20%), and it was determined that no bias need be applied to MPM fluence results.The uncertainty analysis indicates that all fluence results in the beltline region have uncertainty of less than 20%.The results of this analysis are documented in References 1 and 2.This meets the requirement of RP 1.4.1, 1.4.2, and 1.4.3.RP 1.4.3 Fluence Calculational Uncertainty and Bias Estimation.

The vessel fluence (1 sigma)calculational uncertainty must be demonstrated to be S 20 0 k for RTPTS and RTNDT determination.

In these applications, if the benchmark comparisons indicate differences greater than 20%, the calculational model must be adjusted or a correction must be applied to reduce the difference between the fluence prediction and the upper 1-sigma limit to within 200/0.For other applications, the accuracy should be determined using the approach described in Regulatory Position 1.4, and an uncertainty allowance should be included in the fluence estimate as appropriate in the specific application.

Attachment 1 to GNRO-2014/00080 Page 7 of 12 GGNS Response: An extensive evaluation of all contributors to the uncertainty in the calculated fluence was made for the BWR plant calculations performed to date.This evaluation indicated that the uncertainty in calculated fluences in the reactor beltline region is below 20 0 k as specified in the guide.In addition, the comparisons with measurements indicate agreement well within the 20%limit.The agreement of calculations with measurements to within+/-200/0 uncertainty indicates that the MPM calculations can be applied forfluencedetermination with no bias.This meets the requirement of RP 1.4.3.FLUENCE MEASUREMENT METHODS RP 2.1.1 Spectrum Coverage.The set of dosimeters should provide adequate spectrum coverage.GGNS Response: GGNS does not have dosimetry sets installed to provide spectrum definition.

This is in common with GE BWRs which have only limited dosimetry installed in surveillance capsules.The GGNS dosimetry analyzed to date is from iron wires attached to a surveillance capsule and removed after the first cycle of operation.

Calculated neutron spectra are validated using the detailed measurements for test reactors and the calculational benchmark included in RG 1.190.Results are documented in Reference 1.RP 2.1.1 Dosimeter Nuclear and Material Properties.

Use of dosimeter materials should address melting, oxidation, material purity, total and isotopic mass assay, perturbations by encapsulations and thermal shields, and accurate dosimeter positioning.

Dosimeter half-life and photon yield and interference should also be evaluated.

GGNS Response: The GGNS dosimetry wire material characterization was performed by GE (Reference 4).All nuclear constants and parameters used in the dosimetry counting and analysis follow ASTM standard procedures and use validated nuclear constants.

This meets the requirement of RP 2.1.1.RP 2.1.2 Corrections.

Dosimeter-response measurements should account for fluence rate variations, isotopic burnup effects, detector perturbations, self shielding, reaction interferences, and photofission.

GGNS Response: The fluence rate variations for the dosimetry removed after cycle 1 were explicitly taken into account by dividing the cycle into 9 segments and calculating the capsule fluence rate for each segment.The isotopic burnup effects are accurately accounted for as previously discussed.

Attachment 1 to GNRO-2014/00080 Page 8 of 12 The other effects have all been evaluated to be small or not applicable.

This meets the requirement of RP 2.1.2.RP 2.1.3 Response Uncertainty.

An uncertainty analysis must be performed for the response of each dosimeter.

GGNS Response: Uncertainty analyses for each dosimeter have been reported in all of the surveillance capsule reports performed by MPM.For GGNS, GE performed the cycle 1 dosimetry analysis (Reference 4).The uncertainty for each cycle 1 dosimeter was reported.This meets the requirement of RP 2.1.3.RP2.2 Validation.

Detector-response calibrations must be carried out periodically in a standard neutron field.GGNS Response: Regarding the GGNS cycle 1 dosimetry work done by GE, response calibrations have been carried out in the past using documented procedures.

The GE radiochemistry laboratory participated in past interlaboratory dosimetry measurement calibration programs both for fast reactor irradiations (Interlaboratory LMFBR Reaction Rate Program)(Reference 5)and light water reactor irradiations (LWR Surveillance Dosimetry Improvement Program)(Reference 6).In these programs, dosimeters were irradiated in standard neutron fields and measured by multiple laboratories.

These programs demonstrated the accuracy of measurements and improyed the state-of-the-art.

Regulatory Position 2.2 is dated.The above programs are completed and standard field detector response irradiations are no longer done.These programs served their purpose to demonstrate the validity of measurement uncertainty estimates.

Although the calibration programs envisioned continuing use of standard field irradiations, these are no longer considered necessary and modern counting practices involve the use of NIST traceable mixed gamma sources to provide accurate energy and efficiency spectrometer calibrations.

RP2.3 Fast-Neutron Fluence.The E>1 MeV fast-neutron fluence for each measurement location must be determined using calculated spectrum-averaged cross-sections and individual detector measurements.

As an alternative, the detector responses may be used to determine reaction probabilities or average reaction rates.GGNS Response: Measurement-to-Calculation ratios are determined for each dosimeter measurement using calculated detector responses (References 1 and 2).This meets the requirement of RP 2.3.

Attachment 1 to GNRO-2014/00080 Page 9 of 12 RP 2.3 Measurement-to-Calculation Ratios.The MIC ratios, the standard deviation, and bias between calculation and measurement, must be determined.

GGNS Response: The MIC ratios, standard deviations, and comparisons between calculation and measurement have been done and have been reported as discussed earlier (References 1 and 2).This meets the requirement of RP 2.3.REPORTING PROVISIONS RP 3.1 Neutron fluence and uncertainties details of the absolute fluence calculations, associated methods qualification, and fluence adjustments (if any)should be reported.Justification and a description of any deviations from the provisions of this guide should be provided.GGNS Response: The report on the GGNS calculations meets all RG 1.190 reporting requirements (Reference 2).This meets the requirement of RP 3.1.RP3.2 Calculated multigroup neutron fluences and fluence rates should be reported.GGNS Response: Because of the extensive amount of data, only the fluence distributions for neutrons with energy greater than 1 MeV are reported, except for the vessel where neutrons with energy greater than 0.1 MeV and dpa are also reported.The multigroup fluence rates are permanently stored for future access if required.These are not seen as having any application except for possible future dosimetry analysis.Multigroup flux values are readily available for all dosimetry locations.

This meets the requirement of RP 3.2.RP3.2 The value and basis of any bias or model adjustment made to improve the measurement-to-calculation agreement must be reported.GGNS Response: No bias is observed and none is applied (References 1 and 2).This meets the requirement of RP 3.2.RP3.3 Calculated integral fluences and f1uence rates for E>1 MeV and their uncertainties should be reported.GGNS Response: These are all reported (Reference 2).This meets the requirement of RP 3.3.RP3.4 Measured and calculated integral E>1 MeV fluences or reaction rates and uncertainties for each measurement location should be reported.The MIC ratios and the spectrum averaged cross-section should also be reported.

Attachment 1 to GNRO-2014/000BO Page 10 of 12 GGNS Response: These are reported for all dosimetry locations (References 1 and 2).This meets the requirement of RP 3.4.RP3.5 The results of the standard field validation of the measurement method should be reported.GGNS Response: Not applicable.

Specific Activities and Average Reaction Rates RP3.5 The specific activities at the end of irradiation and measured average reaction rates with uncertainties should be reported.GGNS Response: This is contained in all reports with measured dosimetry data.This meets the requirement of RP 3.5.RP3.5 All corrections and adjustments to the measured quantities and their justification should be reported.GGNS Response: No corrections made.This meets the requirement of RP 3.5.CONCLUSIONS In conclusion, the proposed single fluence method from 0EFPYthrough the end of extended operations is in full compliance with RG 1.190, and its referenced regulations, and ensures the fluence inputs to the analysis are appropriate.

Therefore, the adoption of this single fluence method is considered acceptable and maintains applicable safety margins.4.0 REGULATORY SAFETY ANALYSIS NRC RG 1.190 describes the application and qualification of a methodology acceptable to the NRC staff for determining the best-estimate neutron fluence experienced by materials in the beltline region of light water reactor (LWR)pressure vessels, as well as for determining the overall uncertainty associated with those best-estimate values.This request for license amendment provides the GGNS-specific actions to resolve the nonconforming condition.

GGNS has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, and do not affect conformance with any draft General Design Criteria differently than described in the GGNS UFSAR, as described below.4.1 Applicable Regulatory Requirements/Criteria Regulatory requirement 10 CFR Part 50,"Domestic Licensing of Production and Utilization Facilities" Appendix G,"Fracture Toughness Requirements," and 10 CFR 50.61,"Fracture Attachment 1 to GNRO-2014/00080 Page 11 of 12 Toughness Requirements for Protection Against Pressurized Thermal Shock Events", are regulations that ensure the structural integrity of the reactor pressure vessel for cooled power reactors.Chapter 15, Accident Analysis, of the Standard Review Plan (NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants)assumes the pressure vessel does not fail.The proposed single fluence method from 0EFPYthrough the end of extended operations is in full compliance with RG 1.190 and its referenced regulations and ensures the fluence inputs to the analysis are correct.In conclusion, based on the considerations discussed above, (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 amendment will not be inimical to the common defense and security or to the health and safety of the public.4.2 No Significant Hazards Consideration Determination Entergy requests adoption of the proposed single fluence method from 0 EFPY through the end of extended operations.

Entergy 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 Title 10 Code of Federal Regulations (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.The proposed change adopts a single flux methodology.

While Chapter 15, Accident Analysis, of the Standard Review Plan (NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants)assumes the pressure vessel does not fail, the flux methodology is not an initiator to any accident previously evaluated.

Accordingly, the proposed change to the adoption of the flux methodology has no effect on the probability 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 change adopts a flux methodology.

The change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operations.

The change does not alter assumptions made in the safety analysis regarding fluence.

Attachment 1 to GNRO-2014/00080 Page 12 of 12 Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.Does the proposed change involve a significant reduction in a margin of safety?Response: No.The proposed change adopts a single fluence methodology.

The proposed change does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined.

The proposed change ensures that the methodology used for fluence is in compliance with RG 1.190 requirements.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.Based on the above, Entergy concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of"no significant hazards consideration" is justified.

5.0 ENVIRONMENTAL

CONSIDERATION The proposed change would not change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, and would not change an inspection or surveillance requirement.

The proposed change 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 change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

6.0 REFERENCES

1.MPM-614993,"Benchmarking of MPM Methods for BWR Neutron Transport Calculations," November, 2014.2.MPM-814779,"Neutron Transport Analysis for Grand Gulf Nuclear Station", November 2014.3.U.S.Nuclear Regulatory Commission, Regulatory Guide 1.190, dated March 2001 (Accession No.ML01 0890301)4."Flux Wire Dosimeter Evaluation for Grand Gulf Nuclear Power Station," GE Report35-0387, April 1987.5.W.N.McElroy,"Data Development and Testing for Fast Reactor Dosimetry", Nucl.Tech.25, 177 (1975).6.R.Gold and W.N.McElroy,"The Light Water Reactor Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP):

Past Accomplishments, Recent Developments, and Future Directions," Reactor Dosimetry:

Methods, Applications, and Standardization, ASTM STP 1001, Harry Farrar and E.P.Lippincott, eds., American Society for Testing and Materials, Philadelphia, 1989, pp 44-61.