ML19197A114

Response to Request for Additional Information - License Amendment Request to Revise the Technical Specifications - Permanently Defueled Technical Specifications

ML19197A114
Person / Time
Site:	Pilgrim
Issue date:	07/16/2019
From:	Gaston R Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML19217A300	List: ML19197A114
References
2.19.047
Download: ML19197A114 (34)

Text

Entergy Nuclear Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5573 Ronald W. Gaston Director, Nuclear Licensing 2.19.047 July 16, 2019 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Response to Request for Additional Information - License Amendment Request to Revise the Pilgrim Nuclear Power Station Technical Specifications - Permanently Defueled Technical Specifications Pilgrim Nuclear Power Station NRC Docket No. 50-293 Renewed Facility Operating License No. DPR-35

REFERENCES:

1. Entergy Nuclear Operations, Inc. letter to NRC, "Technical Specifications Proposed Change - Permanently Defueled Technical Specifications," (ADAMS Accession No. ML18260A085), dated September 13, 2018

2. NRC email to Entergy Nuclear Operations, Inc., "Pilgrim: Request for Additional Information (RAI) - Pilgrim Post-Decommissioning Technical Specifications (PDTS) License Amendment Request (LAR)

(EPID: L-2018-LLA-0268)," (ML19154A524), dated June 3, 2019 Entergy Nuclear Operations, Inc. (Entergy) submitted a License Amendment Request (LAR) to the U.S. Nuclear Regulatory Commission (NRC) for approval of the Permanently Defueled Technical Specifications, September 13, 2018 (Reference 1). A request for additional information was received from the NRC on June 3, 2019 (Reference 2).

Entergy is providing a response to the RAI in the enclosure to this letter.

The enclosed documentation contains proprietary information as defined by 10 CFR 2.390.

Global Nuclear Fuels (GNF), as owner of the proprietary information, has executed the enclosed affidavit, which identifies that the enclosed proprietary information has been handled and classified as proprietary, is customarily held in confidence, and has been withheld from public disclosure. The proprietary information was provided to Pilgrim Nuclear Power Station (Entergy) in a GNF transmittal that is referenced in the affidavit. The proprietary information has been faithfully reproduced in the enclosed such that the affidavit remains applicable. GNF hereby requests the enclosed proprietary information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390 and 9.17. A non-proprietary version of the documentation also is provided.

This letter contains Proprietary Information - Enclosure 1 of the Attachment to this letter is withheld from public disclosure per 10 CFR 2.390.

2.19.047 Page 2 of 2 This letter contains no new commitments and no revisions to existing commitments. If you have any questions or require additional information, please contact Mr. Peter J. Miner at (508) 830-7127.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July 16, 2019.

Respectfully, RWG/fxm

Enclosure:

Response to Request for Additional Information - License Amendment Request to Revise the Pilgrim Nuclear Power Station Technical Specifications -

Permanently Defueled Technical Specifications Attachment to

Enclosure:

GNF Letter KGO-ENO-HK1-19-058, "Pilgrim Rev 0 Response on Boraflex CSA RAI," July 1, 2019.

GNF Letter

Enclosures:

1) Pilgrim Nuclear Power Station: Fuel Storage Criticality Safety Analysis of Spent Fuel Storage Racks to Remove Boraflex Credit - RAI Response Proprietary

2) Pilgrim Nuclear Power Station: Fuel Storage Criticality Safety Analysis of Spent Fuel Storage Racks to Remove Boraflex Credit - RAI Response Non-Proprietary

3) KGO-ENO-HK1-19-058 Affidavit cc: USNRC Regional Administrator, Region I USNRC Project Manager, Pilgrim USNRC Resident Inspector, Pilgrim Planning and Preparedness Section Chief, Massachusetts Emergency Management Agency Director, Massachusetts Department of Public Health, Radiation Control Program This letter contains Proprietary Information - Enclosure 1 of the Attachment to this letter is withheld from public disclosure per 10 CFR 2.390.

Enclosure 2.19.047 Response to Request for Additional Information - License Amendment Request to Revise the Pilgrim Nuclear Power Station Technical Specifications -

Permanently Defueled Technical Specifications 2.19.047 Enclosure Page 1 of 4 Response to Request for Additional Information License Amendment Request to Revise the Pilgrim Nuclear Power Station Technical Specifications -

Permanently Defueled Technical Specifications NRC REQUEST FOR ADDITIONAL INFORMATION (RAI)

By letter dated September 13, 2018 (Agency wide Documents Access and Management System (ADAMS) Accession No. ML18260A085), as supplemented by letters dated January 10, February 8, and March 14, 2019 (ML19016A135, ML19044A574, and ML19079A158),

Entergy Nuclear Operations, Inc. (Entergy) submitted a license amendment request to revise Pilgrim Nuclear Power Station (Pilgrim) Renewed Facility Operating License and associated Technical Specifications (TS) to Permanently Defueled Technical Specifications (PDTS) consistent with the permanent cessation of reactor operation and permanent defueling of the reactor.

Background

On April 7, 2016, the NRC issued Generic Letter (GL) 2016-01, "Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools" (ML16097A169), to address the degradation of neutron-absorbing materials (NAMs) in wet storage systems for reactor fuel at power and non-power reactors. The generic letter requested that licensees provide information to allow the NRC staff to verify continued compliance through effective monitoring to identify and mitigate any degradation or deformation of NAMs credited for criticality control in spent fuel pools (SFPs).

By letter dated November 3, 2016 (ML16319A131), as supplemented by letter dated February 8, 2018 (ML18039A843), Entergy responded to GL 2016-01 for Pilgrim. In Entergys response to GL 2016-01, as supplemented, the licensee also identified that 2016 testing on the Boraflex installed in the SFP at Pilgrim showed that some of the Boraflex was no longer bounded by the nuclear criticality safety analysis of record. This resulted in the licensee implementing corrective actions to manage Boraflex degradation and maintain subcriticality in the SFP. On September 26, 2018, the NRC issued a letter to Entergy regarding the closeout of GL 2016-01. The letter states that the NRC staff found interim corrective actions taken to be adequate, and that the licensee-identified non-conservative TS would be resolved per Administrative Letter 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety," dated December 29, 1998 (ML031110108).

Entergys corrective actions for this issue were inspected by the NRC and the results of the inspection are documented in integrated inspection report 05000293/2019001 (ML19133A225) dated May 13, 2019. No deficiencies or safety concerns were noted.

2.19.047 Enclosure Page 2 of 4 Issue The current and the proposed TS 4.3.1.1.a. both state the following:

4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum k-infinity of 1.32 for standard core geometry, calculated at the burnup of maximum bundle reactivity, and an average U-235 enrichment 4.6% average over the axial planar zone of highest average enrichment; and Because of the degraded neutron absorption capability of the Boraflex in the spent fuel pool racks, the TS maximum allowable infinite lattice multiplication factor (k-inf) of 1.32 will no longer bound the effective multiplication factor (k-eff) of 0.95, to ensure spent fuel pool conditions remain sufficiently sub-critical.

NRR-SNPB-01.a Provide the controls that ensure the Pilgrim SFP will meet the regulatory requirements for sub-criticality for the entire service life of the Pilgrim SFP.

ENTERGY RESPONSE The Pilgrim Nuclear Power Station (PNPS) spent fuel pool (SFP) contains fourteen storage racks. Nine of these storage racks are Boraflex, four are Boral and one is Metamic. The Criticality Safety Analysis (CSA) performed to support maintaining the k-effective of fuel stored in the Boraflex SFP racks less than 0.95 does not credit the neutron absorption properties of the boron present in the Boraflex storage racks. The CSA requires that the Boraflex racks be zoned into Region 2, one cell in four of a 2x2 array empty, or Region 3, two cells in four of a 2x2 array empty (checkerboard). In addition, only fuel analyzed for storage in either Region 2 or 3 may be stored in the Boraflex racks. The fuel removed from the reactor vessel following permanent shutdown was stored in the Boral or Metamic racks which have been analyzed to store the most reactive peak reactivity bundle in the PNPS SFP.

The CSA evaluated the adequacy of the Region 2 and Region 3 zones in the Boraflex racks.

A station procedure implements the configuration controls used to ensure that the requirements of the CSA are maintained as follows:

1. 1 Reconfigure the SFP Boraflex Racks into Regions 2 and 3 configurations. Ensure Region 2 is loaded having one empty location out of every 2X2 array. Ensure Region 3 is loaded having two empty locations out of every 2X2 array. Load Regions 2 and 3 in the SFP Boraflex racks with fuel eligible for loading into the respective region in accordance with the CSA. Ensure that locations required to be empty do not contain a fuel bundle by documented verification of the SFP Boraflex rack configuration.

2.19.047 Enclosure Page 3 of 4

1. 2 Block the locations adjacent to Panel RR35 South with blade guides.

Row Column 34 35 Defective Panel RR35 South RR X X X = locations to be blocked by Blade Guides

1. 3 Administrative Controls outlined below are also established to prevent inadvertant installation of an incorrect fuel bundle into the Boraflex racks.

a. Additional review by a qualified reviewer for move sheet preparation for fuel movement within the Boraflex racks,

b. Post Defueling, a Certified Fuel Handler will directly supervise fuel movement in the Spent Fuel Pool.

These administrative controls and the required region configurations have been incorporated into station procedure 4.3, "Fuel Handling." This procedure implements these controls and requires verification of the fuel stored in the Boraflex racks whenever fuel is moved within or into these racks. This verification is to ensure that the correct fuel is stored in the correct configuration as required by the CSA. The updated Special Nuclear Material records, SFP Item Control Area account area form, based upon the completed fuel movement sheets is compared to the as left condition of the fuel in the Boraflex racks and to the region bundle listing and region definitions established in the CSA. This activity ensures that the correct fuel is stored in the correct configuration to maintain keff less than 0.95.

Panel RR35 South is the only Boraflex storage panel found by testing to have experienced large gap growth. As a conservative measure, even though the Boraflex racks have been analyzed without credit for the boron absorber, the cells adjacent to RR35 South are required to be blocked by storing blade guides in these cells. All other measured Boraflex panels were found to be within the bounds of the original CSA.

Prescriptive procedural requirements, extensive use of human performance tools and direct supervisory oversight have been successful at PNPS in avoiding fuel handling errors, such as misplaced fuel bundles. PNPS has successfully moved fuel over the past three years using the above referenced configuration controls to maintain SFP subcritical margin.

Non-Boraflex racks in the spent fuel pool are not susceptible to formation of gaps found in the Boraflex racks. Therefore, these racks are not required to be configured into fuel storage protection regions to ensure SFP criticality requirements. Fuel in cells of any Boral and Metamic racks can be moved as required using the procedure controls identified in station procedure 4.3. These SFP storage racks may be used to store the most reactive peak reactivity bundles in the PNPS SFP. The interface between the Boraflex racks and the Boral/Metamic racks was evaluated based on actual peak reactivity of the PNPS fuel. The reactivities used bound all of the fuel currently stored in the SFP.

Aging management controls applicable to the SFP and associated SFP storage racks consist of the Water Chemistry Control - BWR Program that is described in UFSAR, Appendix S, Section S.2.37 and the Structures Monitoring Program, described in UFSAR, Appendix S, and Section S.2.32. These aging management programs are implemented by station procedure 7.8.1, "Chemistry Sample and Analysis Program," and station procedure P-EN-DC-150, "Condition Monitoring of Maintenance Rule Structures." These programs ensure that the SFP

2.19.047 Enclosure Page 4 of 4 racks and SFP are exposed to a treated water environment and are monitored for potential degradation. NRC staff evaluation of these aging management programs is documented in SER Sections 3.0.3.1.13 and 3.0.3.2.17 respectively (see Enclosure reference 1).

PNPS also maintains a Neutron Absorber Monitoring Program for the Boral and Metamic racks as part of the aging management program. This program is described in UFSAR, Appendix S, Section S.2.41 and involves coupon sampling and sample material evaluation.

The most recent coupons, one Boral and one Metamic, were removed from the SFP in late 2018 and were analyzed in early 2019. The coupons were found to be in good condition with test results meeting identified acceptance criteria. Coupons are removed and analyzed at a frequency established by station procedure 4.8, "Spent Fuel Pool Storage Rack Coupon Retrieval." These surveillance controls ensure that the Boral and Metamic spent fuel racks are monitored and maintained for the expected service life of the SFP. Attachment 1 to station procedure 4.8 identifies that the next coupon sample tests were scheduled to be performed in 2023. However, the transfer of spent fuel to dry cask storage is planned to be completed no later than 2022 based on the current schedule identified in the previously submitted Post Shutdown Decommissioning Activity Reports (see Enclosure references 2 and 3)

NRR-SNPB-01.b Provide the analysis that demonstrates those controls will ensure the Pilgrim SFP will meet the regulatory requirements for sub-criticality for the entire service life of the Pilgrim SFP.

ENTERGY RESPONSE b) See Attached - GNF Letter KGO-ENO-HK1-19-058, "Pilgrim Rev 0 Response on Boraflex CSA RAI," July 1, 2019.

Note: Enclosure 1 to the attached letter contains GNF Proprietary Information and is withheld from public disclosure per 10 CFR 2.390.

References

1. Letter, USNRC to Entergy, "Safety Evaluation Related to the License Renewal of Pilgrim Nuclear Power Station, ", (ML071410455), dated June 28, 2007.

2. Letter, Entergy Nuclear Operations, Inc. to USNRC, "Post Shutdown Decommissioning Activities Report, Pilgrim Station", 2.18.070, (ML18320A034), dated November 16, 2018

3. Letter, Holtec Decommissioning International to USNRC, "Notification of Revised Post Shutdown Decommissioning Cost Estimate for Pilgrim Nuclear Power Station,"

(ML18320A040), dated November 16, 2018

ENCLOSURE2 KGO-ENO-HKI-19-058 Pilgrim Nuclear Power Station:

Fuel Storage Criticality Safety Analysis of Spent Fuel Storage Racks to Remove Boraflex Credit - RAI Response Non-Proprietary Information INFORMATION NOTICE This is a non-proprietary version of Enclosure I of KGO-ENO-HKI-19-058, which has the proprietary information removed. Portions of the document that have been removed are indicated by white space inside open and closed bracket as shown here (( )).

Global Nuclear Fuel Global Nuclear Fuel 005N3339 Revision 0 July 2019 Non-Proprietary Information Pilgrim Nuclear Power Station:

Fuel Storage Criticality Safety Analysis of Spent Fuel Storage Racks to Remove Boraflex Credit - RAI Response Copyright 2019 Global Nuclear Fuel -Americas, LLC All Rights Reserved

005N3339 Revision 0 Non-Proprietary Information INFORMATION NOTICE This is a non-proprietary version of Enclosure I of KGO-ENO-HKI-19-058, which has the proprietary information removed. Portions of the document that have been removed are indicated by white space inside open and closed bracket as shown here (( )).

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document are furnished for the purpose of providing the spent fuel pool criticality analysis and results for Pilgrim Nuclear Power Station (PNPS). The use of this information by anyone other than Entergy, or for any purpose other than that for which it is furnished by GNF is not authorized; and with respect to any unauthorized use, GNF makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

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005N3339 Revision 0 Non-Proprietary Information Revision Status Revision Date Description of Change Number 0 July 2019 Initial Release lll

005N3339 Revision 0 Non-Proprietary Information Table of Contents

1. 0 RAI Response .................................................................................................................................. 1 Appendix A - Knominal and Kmax (95/95) Distributions for Eligible Assemblies in Evaluation 3 ................... 5

((.......................................................... *u ................................................................... 7 Appendix C - Peak Infinite Lattice Reactivity of Pilgrim Bundles .............................................................. 8 Appendix D - Rack Region definitions and Calculation Models ................................................................. 9 Appendix E - Fuel Model B - Bum-up Credit Method .............................................................................. 11 Appendix F - Kmax (95/95) Adder Studies for Evaluation 3 .................................................................... 16 List of Tables Table 1 - Summary of Evaluation Methods and Results .............................................................................. 1

.................................................................... )) 3

(( ...................................................

Table 3 - Maximum In-Core kxi of Bundles at Pilgrim ............... .. ...................... ........ .. ...... .. ...... .. ...... .. ....... 8

((*********************************************************u .................................................................... 11 List of Figures Figure 1 - Nominal In-Rack Reactivities of Eligible Assemblies from Evaluation 3 (BUC + Region 2) .... 5 Figure 2 - Kmax (95/95) of Eligible Assemblies from Evaluation 3 (BUC + Region 2) ................................ 6 Figure 3 - Region 2 Rack Model ................................................................................................................ 10 Figure 4 - Region 3 Rack Model ................................................................................................................ 10 Figure 5 - Illustration of Collapsed Void Fraction versus Explicit Void Fraction History ........................ 13 Figure 6 - MCNP-05P Plots of Fuel Model B Geometry ............................................................ .. ...... .. ..... 14 Figure 7 - Evaluation 3 Method Flow Chart ............................... .. ...... .. ...... .. ...... .. ...... .. ...... .. ...... .. ...... .. ..... 15 IV

005N3339 Revision 0 Non-Proprietary Information 1.0 RAI RESPONSE GNF has performed a full scope criticality safety analysis to define Bundle ID specific storage eligibility in Boraflex racks at Pilgrim Nuclear Power Station (PNPS) without any Boraflex credited in the rack design. The analysis covers all fuel discharged from Pilgrim that is currently in the spent fuel pool. Fuel that was operating in Cycle 22 (C22) at the time of the analysis was also evaluated; however, the decision was subsequently made to preclude Cycle 22 discharged assemblies from storage in any cell in the Boraflex racks. This reduces the number of assemblies with significant residual reactivity from storage consideration, improving margin in the Boraflex rack system. This response provides a summary of the analysis with two purposes:

A. Describe the overall evaluation approach and associated rack/fuel models to provide confidence that the methods utilized provide a conservative representation of system reactivity, and B. Demonstrate significant margin to the subcriticality limit for all eligible assemblies in each Region of the Boraflex racks.

The analysis is divided into three evaluations, with associated models and results summarized in Table 1.

Table 1 - Summary of Evaluation Methods and Results Evaluation # of Eligible Number Fuel Evaluated Fuel Model Rack Model Knominal(l) K max (95/95) Assemblies Single Bounding Model A-1 Lattice Peak Reactivity Region 3 (2/4) ((

Lattices Bounding Model A-2 U4XXX/USXXX Bundle IDs Peak Reactivity Region 2 (3/4)

All Bundle IDs Discharged Model B-3 After (( JJC2) Burn-up Credit Region 2 (3/4) Jl Notes:

1. (( ))

2. See Appendix A for histograms ofBtmdle ID Specific values for eligible assemblies.

In these Evaluations, the calculation of appropriate adder terms and a statistical roll-up of results to define Kmax (95/95) values is performed in a manner consistent with guidance in Reference 1, as summarized in the Equations below.

Kma x (95/95) = Knominal + f1ks ias + f1ku ncer taint y + !1k1nterface Where:

= L !).k Bi n

!).k Bias i=l

005N3339 Revision 0 Non-Proprietary Information Additional details on the fuel and rack models used in each case are provided in the evaluation specific sections that follow. These sections also summarize the manner in which the adder terms were developed and applied in each evaluation. Where appropriate, some of the more technical details of the work are provided as Appendices to allow for a summary of the approach and results to be presented in a more streamlined way. The response closes with a discussion of analysis conclusions.

Evaluation Descriptions Evaluation 1 - Bounding Peak Reactivity Lattice in Checkerboard Loading Pattern The first evaluation leverages a traditional peak, reactivity analysis method. The method utilized is consistent with GNF's typical fuel rack criticality analysis approach that has been described in detail in recent license amendment requests [2]. In these calculations, a single assembly is modeled in every eligible storage location. The assembly is modeled such that its entire axial length is assumed to exist at the exposure dependent, peak reactivity state corresponding to the bounding two-dimensional (2D) lattice of the limiting nuclear design, also referred to in the analysis as "Fuel Model A" . The bounding lattice modeled was selected based on a survey of all fuel at Pilgrim, as discussed in Appendix C.

This approach is used to define eligibility for all bundles in the pool in a checkerboard configuration (i.e. , 2/4 loading pattern or "Region 3"). The rack model definition used in the analysis is further described in Appendix D.

This evaluation calculates biases and uncertainties in a manner that is consistent with guidance in Reference 1 and as described more specifically in Reference 2. ((

))

Evaluation 2 - Low Reactivity Fuel Designs Modeled at Peak Reactivity in 3/4 Loading Pattern The second evaluation is also a traditional , peak reactivity analysis with a fuel modeling approach that is consistent with Evaluation 1 (i.e. , Fuel Model A); however, in this case two different unique nuclear designs 1 are evaluated. The nuclear designs analyzed are consistent with the design used for all Bundle IDs that start with "LJ4" and "LJ5". These designs were selected for additional evaluation based on their low peak reactivity characteristics, as discussed in Appendix C.

The assemblies were evaluated in a more densely packed loading pattern (i .e. , 3/4 loading pattern or Region 2) than was used in Evaluation 1. The rack model definition used in the analysis is also described in Appendix D.

Again, similar to 1, this evaluation calculates biases and uncertainties for each of the two unique nuclear designs evaluated in a manner that is consistent with guidance in Reference 1 and as described more specifically in Reference 2. ((

))

Evaluation 3 - Well Characterized Fuel Modeled with Burn-up Credit in 3/4 Loading Pattern 1A tmique nuclear design" is defined as a btmdle that has a consistent fuel product and fuel loading pattern combination 2

005N3339 Revision 0 Non-Proprietary Information In the third evaluation, assemblies are modeled on a Bundle ID specific basis with explicit nodal discharge exposure profiles that are associated with that unique assembly's local operational conditions. These local conditions are defined by GNF's core simulator code, PANACI 1. The use of such a three-dimensional (3D) model with best-estimate nodal isotopics in a criticality safety analysis is commonly referred to as a "burn-up credit method", or "Fuel Model B" in this analysis.

Additional details on the mechanics behind the construction of these fuel models is provided in Appendix E of this response.

In each case, the assemblies were evaluated in the same loading pattern as Evaluation 2 (i .e. , 3/4 loading pattern or Region 2).

This method requires a unique calculation for every Bundle ID analyzed to define its KnominaI value.

In these calculations, the assembly of interest is modeled in every eligible storage location.

Evaluating the system in this way ensures any combination of eligible bundles stored in a given region will result in a system reactivity that is less than the maximum reactivity calculated in these infinite, homogenous studies. For the PNPS analysis, this required (( )) individual in-rack cases to be calculated.

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Note: (( ))

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Interface Considerations Interfaces between different combinations of Region 2 and Region 3, as well as interfaces between these Regions and adjacent non-Boraflex racks, were explicitly evaluated. These interface evaluations included consideration of a misload assembly in a position to maximize the reactivity of the interface region of the system. Results from these studies demonstrated that interfaces did not result in significant increases in system reactivity compared to the reactivity of their constituent Regions, confirming that no additional restrictions were necessary on loading patterns to maintain reactivity margin in the spent fuel pool.

Analysis Conclusions The results from all three evaluations were used to define Bundle ID specific eligibility in each Region of the Boraflex racks. This eligibility was presented as a simple table in the final report, where each Bundle ID in the Pilgrim spent fuel pool was characterized as either acceptable or not acceptable for storage in Region 2 and Region 3.

As shown in Table 1 and Appendix A, there is significant margin to the regulatory limit of 0.95 for all eligible assemblies in both Regions of Boraflex racks with no credit taken for Boraflex neutron absorption in the analysis.

References

1. NEI 12-16 Revision 3, "Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants," March 2018. (NRC ADAMS Accession Number ML l 8088B400).

2. NEDC-33886P, "River Bend Station: Fuel Storage Criticality Safety Analysis of Spent Fuel Storage Racks with Rack Inserts," Revision 1, October 2018 (NRC ADAMS Accession Number ML18297 Al03).

3. NUREG/CR-7108," An Approach For Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses - Isotopic Composition Predictions", April 2012.

4. Metwally, W. , Sugawara, M. , Mills, V. , Hannah, J. , "TGBLA Spent Fuel Isotopic Predictions And Their Effect on Criticality Calculations", in proceedings of PHYSOR 2010, Pittsburg, Pennsylvania, USA, May 9-14, 2010, (2010).

5. NUREG/CR-7162, " Analysis of Experimental Data for High Burnup BWR Spent Fuel Isotopic Validation - SVEA-96 and GE14 Assembly Designs", March 2013.

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005N3339 Revision 0 Non-Proprietary Information APPENDIX A- Knominal AND Kmax (95/95) DISTRIBUTIONS FOR ELIGIBLE ASSEMBLIES IN EVALUATION 3

((

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Figure 1 - Nominal In-Rack Reactivities of Eligible Assemblies from Evaluation 3 (BUC + Region 2) 5

005N3339 Revision 0 Non-Proprietary Information

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Figure 2 - Kmax (95/95) of Eligible Assemblies from Evaluation 3 (BUC + Region 2) 6

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((

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005N3339 Revision 0 Non-Proprietary Information APPENDIX C - PEAK INFINITE LATTICE REACTIVITY OF PILGRIM BUNDLES Table 3 provides in-core kco results calculated by TGBLA06 for each bundle at Pilgrim. The values shown correspond to the peak reactivity lattice of each bundle at its peak reactivity exposure.

The assembly corresponding to the bold values at the top of the table on the left were selected for modeling in Evaluation 1. The assemblies corresponding to the italicized values at the bottom of the table on the right were selected for modeling in Evaluation 2.

Table 2 - Maximum In-Core kco of Bundles at Pilgrim Bundle Lattice Peak In- Bundle Lattice Peak In-Number Number< 1) Core kco Number Number Core kco rr ll Note 1: ((

))

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005N3339 Revision 0 Non-Proprietary Information APPENDIX D - RACK REGION DEFINITIONS AND CALCULATION MODELS The full scope criticality analysis establishes, on a Bundle ID specific basis, which assemblies are eligible for storage in the following configurations in the plant's Boraflex racks:

1. Region 2: Bundles stored in 3 out of every 4 cells - i.e., 3/4

2. Region 3: Bundles stored in 2 out of every 4 cells - i.e., 2/4 A "fully loaded" Region 1 (i.e., 4/4 loading pattern) is not allowed in the Boraflex racks.

Conservative models of the rack systems were developed that are defined by the following characteristics:

1. An array of 8x8 cells with periodic boundary conditions in the x and y axes to preclude leakage in the radial direction,

2. Twelve inches of water above and below the active fuel to limit axial leakage of neutrons,

3. Explicit modeling of rack structural material in the active fuel region with the minimum pitch between storage cells that can result from the combination of different sub-elements of the rack design,

4. Removal of all Boraflex from the system (replaced with water), and

5. Cells which are precluded from fuel storage are modeled with a physical blocking device,

((

)). This blocking device is modeled based on the cross section of a GE blade guide in the active fuel region . Studies were performed to demonstrate explicit modeling of this device results in an increase in system reactivity; therefore, final configurations that do or do not contain a blocking device are acceptable from a criticality calculation perspective.

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005N3339 Revision 0 Non-Proprietary Information A radial cross-section of the associated Region 2 and Region 3 rack models are provided in Figure 3 and Figure respectively. The empty locations (blue color) are filled with water and the cell blocking device.

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Figure 3 - Region 2 Rack Model Figure 4 - Region 3 Rack Model 10

005N3339 Revision 0 Non-Proprietary Information APPENDIX E - FUEL MODEL B- BURN-UP CREDIT METHOD The burn-up credit fuel model is developed using the following procedure:

1. Exposure Accounting- Using actual plant operating history, perform exposure accounting for Cycles ((

))

2. Nodal History Collapsing - Collapse historical operational data into nodal depletion history on a discharge bundle specific basis.

3. Nodal Isotopic Definition - Reburn each node of the analyzed discharged bundles using its collapsed operational history with TGBLA06.

4. MCNP Input Generation - Combine 2D lattice physics defined pin specific isotopic outputs into 3D MCNP bundle input files on a Bundle ID specific basis.

Additional details on each step in the Fuel Model development process are provided in the sections that follow. An overall flowchart of this process, including details on how it fits into the overall evaluation method, are provided at the end of this Appendix in Figure 7.

Exposure Accounting In order to quantify the reactivity of a spent fuel assembly, it is necessary to account for the operational history that the individual bundle experienced throughout its time in the core. For a large number of historical cycles for a given BWR, the history of a plant's operation is available either in plant records or on GNF's Boiling Water Reactor Engineering Data Bank (BWREDB).

The information is most commonly available on either daily or weekly intervals. This plant operational history typically includes, among other things, core power, core flow rate, and control blade positions as a function of exposure for a given cycle. Using this information, GNF's nodal core simulator PANACl 1 can be used to calculate the following information on a nodal basis for a corresponding fine exposure grid:

1. Exposure

2. Instantaneous Void Fraction

3. Fuel Temperature

4. Control State

5. Power Density As is common for BWR nodal core simulators, each node is defined as an approximately 6-inch cube in the calculation method. For Pilgrim, the active fuel height is 145.24 inches tall divided evenly into 24 nodes that are slightly larger than 6 inches. This process of defining historical nodal conditions as a function of core operation is commonly referred to as exposure accounting and is performed on a regular basis as a part of the core design process, where the local reactivity definitions of reinserted bundles are essential for planning the next core design at the plant.

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Nodal History Collapsing The outcome of exposure accounting is a database which contains the history of nodal conditions for a given bundle in the spent fuel pool. This history often includes hundreds of exposure steps for each node. In order to be able to practically handle this information with a depletion code, this information is collapsed into 1 GWD/STU exposure steps. The collapsing is performed on an exposure-weighted basis for all of the parameters relevant to the node ' s depletion (i .e., items 2-5 in the previous section). Note that, given the control state input is binary (controlled or uncontrolled) for the 2D lattice physics calculations, the control state which defines the majority of the 1 GWD/STU step is used to characterize the entire step. The use of this best estimate approximation has a small effect on local reactivity and is appropriate because there is no a priori conservative control state to select at any given point in a node' s exposure accumulation as it rel ates to discharge reactivity.

Depletion on an exposure grid that is at least this fine is consistent with generally accepted practices for lattice physics codes and provides sufficient detail to adequately account for the changes in isotopics and the corresponding effect on physics parameters. Note that a final step is taken to the calculated discharge exposure (off the 1 GWD/STU grid) in this collapsing process to accurately reflect the actual 3D discharge exposure profile of the bundle. A graphical representation of this collapsing method is provided in Figure 5.

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Figure 5 - Illustration of Collapsed Void Fraction versus Explicit Void Fraction History Nodal Isotopic Definition The collapsed operational history data is used in conjunction with explicit bundle geometry and fuel loading information available on the BWREDB to define lattice physics inputs on the collapsed exposure grid for any unique assembly in the pool. Because any unique assembly can be explicitly modeled, the development of representative geometry or operational history definitions for application to large populations of assemblies is not required.

A single lattice physics input is created for each node (typically 24 or 25) in a bundle using this approach. These lattice physics inputs are executed in GNF's lattice physics code, TGBLA06, to generate pin specific isotopics on a nodal basis at the time of discharge for a given assembly.

Because this process used the operating history of each node to perform the depletions, history parameters, such as void history or control history, are implicitly included in the depletion and do not need to be separately addressed.

The lattice physics code also has the capacity to assess the isotopic content as a function of decay time to allow for an explicit accounting of cooling time. This feature is leveraged in this evaluation such that the nodal isotopics are provided for evaluation in the storage rack assuming bundles have experienced isotopic decay from the time of their discharge to June 1, 2018.

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005N3339 Revision 0 Non-Proprietary Information MCNP Input Generation The TGBLA06 defined geometry and pin specific isotopics for any node are translated to 2D MCNP inputs using a GNF utility code similar to the peak reactivity method. The 2D inputs are combined into 3D bundle inputs with the same nodal structure used in the core simulator. For those regions of the bundle where geometry is non-uniform in a node (i.e., hybrid nodes), the geometric heterogeneity is preserved in MCNP by adding an additional lattice physics simulation and MCNP translation to the process. This treatment results in a node in MCNP that is characterized by the same operational history but is subdivided based on geometry.

A graphical representation of a 3D bundle generated with this method that demonstrates these features is provided in Figure 6. In this figure, each color represents a unique cell and material definition defined in this process. ((

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Figure 6 - MCNP-05P Plots of Fuel Model B Geometry Again, like the peak reactivity method, this fuel model is combined with the 3D MCNP rack definition to allow for each unique bundle to be analyzed as if it existed in every eligible storage location in the pool. Executing these inputs in MCNP allows for the definition of bounding in-rack reactivities for any bundle with sufficiently detailed operational history. ((

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Figure 7 - Evaluation 3 Method Flow Chart

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005N3339 Revision 0 Non-Proprietary Information APPENDIX F- KMAX (95/95) ADDER STUDIES FOR EVALUATION 3 As was done in Evaluation I and Evaluation 2, Evaluation 3 calculates biases and uncertainties in a manner that is generally consistent with guidance in Reference I and as described more specifically in Reference 2, ((

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In this case, however, it is necessary to consider additional uncertainties to account for the increased modeling complexity and higher exposures that are explicitly evaluated. The additional biases and uncertainties considered are described in more detail below.

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ENCLOSURE 3 KGO-ENO-HK1-19-058 Affidavit

Global Nuclear Fuel Americas AFFIDAVIT I, Brian R. Moore, state as follows:

(1) I am the General Manager, Core & Fuel Engineering, Global Nuclear Fuel Americas, LLC paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) -

ENO-HK1 Entergy Nuclear Operations, Inc Pilgrim Nuclear Power Station: Fuel Storage Criticality Safety Analysis of Spent Fuel Storage Racks to Remove Boraflex Credit RAI Response, June 2019. GNF proprietary information in Enclosure 1 is identified by a dotted underline inside double square brackets. ((This sentence is an example.{3}))

used to indicate that the entire content between the double brackets is proprietary. In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GNF-A relies upon the exemption from disclosure set forth in the Freedom of (Exemption 4). The material for which exemption from disclosure is here sought also qualify for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GNF-A's competitors without license from GNF-A constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future GNF-A customer-funded development plans and programs, resulting in potential products to GNF-A; KGO-ENO-HK1-19-058 Enclosure 1 Affidavit Page 1 of 3

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.

(5) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GNF-A, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GNF-A, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GNF-A.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GNF-A are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2) is classified as proprietary because it contains details of GNF- and licensing methodology.

The development of the methods used in these analyses, along with the testing, development and approval of the supporting methodology was achieved at a significant cost to GNF-A or its licensor.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GNF-A's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

KGO-ENO-HK1-19-058 Enclosure 1 Affidavit Page 2 of 3

The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GNF-A.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GNF-A would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GNF-A of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on this 1st day of July 2019.

Brian R. Moore General Manager, Core & Fuel Engineering Global Nuclear Fuel Americas, LLC 3901 Castle Hayne Road Wilmington, NC 28401 Brian.Moore@ge.com KGO-ENO-HK1-19-058 Enclosure 1 Affidavit Page 3 of 3