ML18344A452

Issuance of Amendment No. 234 Change to Remove Boraflex Credit from Spent Fuel Racks

ML18344A452
Person / Time
Site:	Nine Mile Point
Issue date:	01/11/2019
From:	Marshall M Plant Licensing Branch 1
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
Marshall M, NRR/DORL/LPL, 415-2871
References
EPID L-2018-LLA-0039
Download: ML18344A452 (32)

Text

OFFICIAL USE ONLY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 11, 2019 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNIT 1 - ISSUANCE OF AMENDMENT NO. 234 RE: CHANGE TO REMOVE BORAFLEX CREDIT FROM SPENT FUEL RACKS (EPID L-2018-LLA-0039)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 234 to Renewed Facility Operating License No. DPR-63 f or the Nine Mile Point Nuclear Station, Unit 1. The amendment consists of changes to the Technical Specifications in response to your application dated February 9, 2018, as supplemented by letter dated August 17, 2018.

The amendment removes the Boraflex credit from the two remaining Boraflex storage racks located in the spent fuel pool. The changes eliminate reliance on Boraflex f or spent fuel pool reactivity control.

A copy of the related Safety Evaluation is enclosed. Notice of issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Michael L. Marshall, Jr., Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-220 NOTICE: Enclosure 3 to this letter contains

Enclosures:

Proprietary Information. Upon separation from

1. Amendment No. 234 to DPR-63 Enclosure 3, this letter is DECONTROLLED.

2. Safety Evaluation (Non-Proprietary)

3. Safety Evaluation (Proprietary) cc w/o Enclosure 3: Listserv OFFICIAL USE ONLY PROPRIETARY INFORMATION

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC NINE MILE POINT NUCLEAR STATION, LLC DOCKET NO. 50-220 NINE MILE POINT NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 234 Renewed License No. DPR-63

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (Exelon, the licensee) dated February 9, 2018, as supplemented by letter dated August 17, 2018, 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. 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 Renewed Facility Operating License No. DPR-63 is hereby amended to read as follows:

Enclosure 1

(2) Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 234, is hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented during the spent fuel pool cleanup plan scheduled to begin after the 2019 refuel outage.

FOR THE NUCLEAR REGULATORY COMMISSION g~~

James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: January 11, 201 9

ATTACHMENT TO LICENSE AMENDMENT NO. 234 NINE MILE POINT NUCLEAR STATION, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-63 DOCKET NO. 50-220 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Page Insert Page 3 3 Replace the following page of the Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Page Insert Page 346 346

(1) Exelon Generation pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (2) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components.

(5) Exelon Generation pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:

Part 20, Section 30.34 of Part 30; Section 40.41 of Part 40; Section 50.54 and 50.59 of Part 50; and Section 70.32 of Part 70. This renewed license 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 also subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 1850 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, which is attached hereto, as revised through Amendment No. 234, is hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.

(3) Deleted Renewed License No. DPR-63 Amendment No.191 tAF8l-l1R 210,211,213,214,215, 21e, 217,218,220,222,223,224,225,227,229,231,233,234 COFrection Letter DateEI Al-lgl-lst 7, 2012 Correction Letter DateEI Marsll 17, 2015 Correction Letter ElateEI Jyly 29, 201 e

5.5 Storage of Unirradiated and Spent Fuel Unirradiated fuel assemblies will normally be stored in critically safe new fuel storage racks in the reactor building storage vault. Even when flooded with water, the resultant keff is less than 0.95. Fresh fuel may also be stored in shipping containers. The unirradiated fuel storage vault is designed and shall be maintained with a storage capacity limited to no more than 200 fuel assemblies.

The north and south half of the spent fuel pool is analyzed to store 1840 and 2246 spent fuel assemblies, respectively, in storage racks containing the neutron absorber material Boral. The spent fuel pool is analyzed to store a total of 306 spent fuel assemblies in storage racks containing Boraflex. Both types of storage racks will maintain a keff of less than 0.95 under normal, abnormal and accident conditions. The spent fuel stored in both types of storage racks must have a peak lattice enrichment of 4.6% or less and the k-inf in the standard cold core geometry must be less than or equal to 1.31.

5.6 (Deleted)

Amendment No. 142, 167,189,234 346

Of'f'ICIAL USE ONLY PROPRIETARY INf'ORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 234 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-63 NINE MILE POINT NUCLEAR STATION, LLC EXELON GENERATION COMPANY, LLC NINE MILE POINT NUCLEAR STATION, UNIT 1 DOCKET NO. 50-220 Proprietary information pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations has been redacted from this document.

Redacted information is identified by blank space enclosed within bold double brackets ([[ ]]).

1.0 INTRODUCTION

By letter dated February 9, 2018, as supplemented by letter dated August 17, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML18040A636 and ML18229A060, respectively), Exelon Generation Company, LLC (the licensee) submitted a license amendment request (LAR) for changes to the Nine Mile Point Nuclear Station, Unit 1 (Nine Mile Point 1), Technical Specifications (TSs). The LAR proposes modifications to Nine Mile Point 1 TS Section 5.5, "Storage of Unirradiated and Spent Fuel," in support of a new nuclear criticality safety analysis methodology. The purpose of the methodology and the associated analysis presented in the LAR is to qualify the two Boraflex storage racks in the Nine Mile Point 1 spent fuel pool (SFP) with the same TS requirements as the Boral racks, while also removing the credit for Boraflex.

The supplement dated August 17, 2018, 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 or the Commission) staff's initial proposed no significant hazards consideration determination as published in the Federal Register on July 3, 2018 (83 FR 31184). On July 10, 2018, a correction notice was issued in the Federal Register (83 FR 31981) to correct the hearing date in the original notice. The correction notice did not Of'f'ICIAL USE ONLY PROPRIETARY INf'ORMATION Enclosure 2

OFFICIAL USE ONLY PROPRIETARY INFORMATION change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on July 3, 2018.

2.0 REGULATORY EVALUATION

The regulatory requirements and guidance that the NRC staff considered in its review of the LAR are described below.

2.1 Regulatory Requirements Title 10 of the Code of Federal Regulations (1 O CFR) Section 50.36, "Technical specifications,"

establishes the regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plant's TSs.

Section 50.36(c)(3) to 10 CFR requires surveillance requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCO for operation will be met.

Section 50.36(c)(4) to 10 CFR requires that design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1 ), (2), and (3) of Section 50.36.

Section 50.68(b)(1) to 10 CFR requires that plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

Section 50.68(b)(4) to 10 CFR requires, in part, that if no credit for soluble boron is taken, the k-effective (keff) of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95 at a 95 percent probability and 95 percent confidence level, if flooded with unborated water.

The regulations in Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 (hereinafter referred to as GDC), establish the minimum requirements for the principal design criteria for water-cooled nuclear power plants. The principal design criteria establish the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety.

The construction permit for Nine Mile Point 1 was issued by the Atomic Energy Commission on April 12, 1965, and the operating license was issued on December 26, 1974. The Atomic Energy Commission published the final rule that added Appendix A to 1O CFR Part 50 in the Federal Register (36 FR 3255) on February 20, 1971, with the rule effective on May 21, 1971.

In accordance with NRC staff requirements memorandum from S. J. Chilk to J.M. Taylor, "SECY-92-223 - Resolution of Deviations Identified During the Systematic Evaluation Program,"

dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971, which includes Nine Mile Point 1.

OFFICIAL USE ONLY PROPRl&TARY INFORMATION

OFFICIAL US& ONLY PROPRIETARY INFORMATION Nine Mile Point 1 was not licensed to the regulations in Appendix A to 10 CFR Part 50. The plant design criteria for Nine Mile Point 1 are listed in Table 1-1 of the Updated Final Safety Analysis Report (UFSAR),Section I, "Principal Design Criteria." This UFSAR table refers to the Nine Mile Point 1 Technical Supplement to Petition for Conversion from Provisional Operating License to Full-Term Operating License, dated July 1972, for the details of the assessment against the current GDC at that time. A review has determined that the plant-specific requirements for Nine Mile Point 1 are sufficiently similar to the regulations in Appendix A to 10 CFR Part 50, as related to the proposed changes.

Nine Mile Point 1 was designed and constructed to meet the 27 criteria issued by the Atomic Energy Commission on November 22, 1965 (USAEC Press Release H-252, "General Design Criteria for Nuclear Power Plant Construction Permits," dated November 22, 1965). With respect to the 27 criteria, Criteria 24 and 25 are applicable.

Criterion 24 states:

All fuel storage and waste handling systems must be contained if necessary to prevent the accidental release of radioactivity in amounts which could affect the health and safety of the public.

Criterion 25 states:

The fuel handling and storage facilities must be designed to prevent criticality and to maintain adequate shielding and cooling for spent fuel under all anticipated normal and abnormal conditions, and credible accident conditions.

Variables upon which health and safety of the public depend must be monitored.

Criteria 24 and 25 have equivalents in GDC 61, "Fuel storage and handling and radioactivity control," and GDC-62, "Prevention of criticality in fuel storage and handling."

GDC 61 requires, in part, that these systems shall be designed (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.

GDC 62 requires that criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

Therefore, the NRC staff reviews amendment requests for the Nine Mile Point Renewed Facility Operating License using the 10 CFR Part 50, Appendix A, GDC, unless there are specific criteria identified in the UFSAR.

2.2 Applicable Guidance and Industry Documents The guidance that the NRC staff considered in its review of this LAR includes the following from NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition" (hereinafter referred to as the SRP).

Relevant sections of the SRP used in the review of this LAR include the following:

SRP Section 3.8.4, Revision 4, "Other Seismic Category I Structures," dated September 2013 (ADAMS Accession No. ML13198A258), including Appendix D, "Guidance on Spent Fuel Pool OFFICIAL US& ONLY PROPRIETARY INFORMATION

OFFICIAb USE ONbY PROPRIETARY INFORMATION Racks," provides guidance to the NRG staff for review of SFP racks and the associated structures, which would meet the acceptance criteria specified in Subsection II of this SRP section.

SRP Section 9.1.1, Revision 3, "Criticality Safety of Fresh and Spent Fuel Storage and Handling," dated March 2007 (ADAMS Accession No. ML070570006), provides guidance to the NRG staff to verify that storage facilities maintain the new and spent fuel in subcritical arrays during all credible storage conditions and during fuel handling, in accordance with GDC 62 and 10 CFR 50.68.

SRP Section 9.1.2, Revision 4, "New and Spent Fuel Storage," dated March 2007 (ADAMS Accession No. ML070550057), provides guidance to the NRG staff that addresses the capability of the new and spent fuel storage facilities to maintain the fuel in a safe and subcritical array during all anticipated operating and accident conditions.

The NRG staff issued a memorandum dated August 19, 1998 (ADAMS Accession No. ML003728001 ), also known as the "Kopp memo," containing NRG staff guidance for performing the review of SFP nuclear criticality safety analyses. This guidance supports determining compliance with GDC 62 and existing SRP Sections 9.1.1 and 9.1.2. The principal objective of this guidance is to clarify and document NRG staff positions that may have been incompletely or ambiguously stated in previously issued safety evaluations and other staff documents. A second purpose is to state NRG staff positions on recently proposed storage configurations and characteristics in spent fuel rerack or enrichment upgrade requests, multiple region spent fuel storage racks, checkerboard loading patterns for new and spent fuel storage, credit for burnup in the spent fuel to be stored, and credit for non-removable poison inserts.

The Division of Safety Systems (DSS) Interim Staff Guidance (ISG) DSS-ISG-2010-01, Revision 0, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools," dated October 13, 2011 (ADAMS Accession No. ML110620086), provides updated guidance to address the increased complexity of recent SFP nuclear criticality analyses and operations. The guidance is intended to reiterate existing guidance, clarify ambiguity in existing guidance, and identify lessons learned based on recent submittals. Similar to the Kopp memo, this guidance supports determining compliance with GDC 62 and existing SRP Sections 9.1.1 and 9.1.2.

DSS-ISG-2010-01 provides guidance that depletion simulations should be performed with parameters that maximize the reactivity of the depleted fuel assembly. It references NUREG/CR-6665, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," dated February 2000 (ADAMS Accession No. ML003688150), which discusses the treatment of depletion parameters.

DSS-ISG-2010-01 references NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology," dated January 2001 (ADAMS Accession No. ML050250061 ).

NUREG/CR-6698 states, in part, that:

In general, the critical experiments selected for inclusion in the validation must be representative of the types of materials, conditions, and operating parameters found in the actual operations to be modeled using the calculational method. A sufficient number of experiments with varying experimental parameters should be OFFICIAb USE ONbY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION selected for inclusion in the validation to ensure as wide an area of applicability as feasible and statistically significant results.

The NRC staff used NUREG/CR-6698 as guidance for review of the code validation methodology presented in the application. The basic elements of validation are outlined in NUREG/CR-6698, including identification of operating conditions and parameter ranges to be validated, selection of critical benchmarks, modeling of benchmarks, statistical analysis of results, and determination of the area of applicability.

NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic Technical Activity A-36," dated July 1980 (ADAMS Accession No. ML070250180), provides guidance to systematically examine staff licensing criteria and the adequacy of measures in effect at operating plants, and to recommend necessary changes to assure the safe handling of heavy loads.

NUREG/CR-7109, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (kett) Predictions," dated April 2012 (ADAMS Accession No. ML12116A128), describes how model-specific sensitivity data can be used to translate nuclear data uncertainties into uncertainty in the model kett value.

NRC Generic Letter 78-11, "Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978 (ADAMS Accession No. ML031280383), provides additional guidance for the type and extent of information needed by the NRC staff to perform the review of licensee proposed modifications of an operating reactor spent fuel pool.

NRC Generic Letter 79-04, "Referencing 4/14/78 Letter- Modifications to NRC Guidance

'Review and Acceptance of Spent Fuel Pool Storage and Handling Applications,"' dated January 18, 1979, provides additional guidance on modifications to Generic Letter 78-11. The modifications involve pages IV-5 and IV-6 of the document and comprise modified rationale and corrections.

NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage" dated June 29, 1979 (ADAMS Accession No. ML031110008), provides guidance regarding the issues concerning the use of Boraflex in spent fuel storage racks.

NRC Generic Letter 2016-01, "Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools,"

dated April 7, 2016 (ADAMS Accession No. ML16097A169), provides guidance regarding degradation of neutron-absorbing materials in wet storage systems for reactor fuel at power and non-power reactors. In the context of this generic letter, "credited" means that the neutron-absorbing material is necessary to limit the maximum effective multiplication factor (kett), under optimum conditions of moderation and reflection, to less than that assumed in the licensing and design basis.

Holtec International (HI) Report Hl-2178001, Revision O (criticality analysis), was provided as an attachment to the LAR. Hl-2178001 presents the nuclear criticality safety analysis for the Nine Mile Point 1 SFP racks.

Hl-2104790, Revision 1, "Nuclear Group Computer Code Benchmark Calculations" (benchmarking analysis and validation report), was provided to the NRC staff as an attachment to a letter dated December 30, 2014 (ADAMS Accession No. ML14364A100). The changes made in Revision 3 were assessed by the NRC staff in a prior LAR review for the Shearon Harris Nuclear OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAb USE ONbY PROPRl&TARY INFORMATION Power Plant, Unit 1, dated June 28, 2017 (ADAMS Accession No. ML17193B165). Hl-2104790 presents the benchmarking evaluation performed for the Monte Carlo N-Particle, Version 1.51 (MCNP5-1.51), used for the nuclear criticality safety analysis to demonstrate the applicability of the code to geometries and compositions being analyzed and to determine the code bias and uncertainty.

3.0 TECHNICAL EVALUATION

The Nine Mile Point 1 SFP contains two Boraflex fuel storage SFP racks in two different module sizes, designated "rack type 198" and "rack type 218," after the number of usable locations.

Nine Mile Point 1 also contains 14 Boral storage racks of various sizes. The Boraflex fuel storage SFP racks are constructed with two Boraflex panels encased in stainless steel cladding separated by a water gap, all within a stainless steel Boraflex box. Adjacent to each Boraflex box are two stainless steel fuel storage cells separated by a stainless steel wall. Thus, each row of Boraflex boxes separates rows of fuel storage cells.

The Boral fuel storage SFP racks are high density racks constructed using formed and fabricated cells with a Boral neutron absorbing panel encased in stainless steel sheathing on each of the four sides of the fabricated cells. The Boral storage racks credit the Boral neutron absorbing panels and have the same TS requirements as the Boraflex racks.

While the dimensions of the Boral storage racks are presented and discussed in the LAR, the nuclear criticality safety analysis methodology submitted for review is applicable to only the Boraflex storage racks. A separate nuclear criticality safety analysis methodology applicable to the Boral storage racks was previously reviewed and approved by the NRC. However, within the present LAR, the licensee does consider the potential impact that interface between the two rack designs may have on nuclear criticality safety.

The credited neutron absorber material installed in Nine Mile Point 1 SFP Boraflex storage racks ensures that the effective multiplication factor kett does not exceed the values and assumptions used in the nuclear criticality safety analyses of record and other licensing basis documents. Neutron absorber materials utilized in SFP racks exposed to treated water or treated borated water may be susceptible to reduction of neutron-absorbing capacity, changes in dimension that increase kett, and loss of material. A monitoring program is implemented at Nine Mile Point 1 to ensure that degradation of the neutron-absorbing material used in the SFPs, which could compromise the nuclear criticality safety analysis, will be detected. With the present application, the licensee intends to discontinue crediting the neutron absorber material installed in Nine Mile Point 1 SFP Boraflex storage racks.

The NRC staffs review included the benchmarking analyses (in Hl-2104790) and the nuclear criticality safety analyses (Hl-2178001) for the Nine Mile Point 1 SFP racks. The Hl-2178001 report describes the methodology and analytical models used in the nuclear criticality safety analysis to show that the maximum k-effective (kett) of the SFP rack will be no greater than 0.95 when flooded with unborated water.

The related TS changes proposed in the licensee's February 9, 2018, submittal are addressed below. The nuclear criticality safety analysis analyzes the existing Nine Mile Point 1 Boraflex SFP storage racks without accounting for the presence of the Boraflex neutron absorbing material in order to remove the credit currently being taken for it.

OFFICIAb USE ONbY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.1 Description of Proposed Technical Specification Changes The proposed changes will eliminate reliance on Boraflex for SFP reactivity control. Nine Mile Point 1 will install 108 permanent cell blockers in predetermined spent fuel rack cells that will prevent the placement of spent fuel assemblies within the blocked cells. The cell blockers will be arranged in spent fuel racks such that an analyzed geometric pattern will be created for the placement of the spent fuel assemblies. As a result, continued performance of the Boraflex monitoring program and the license renewal commitment to enhance the program will no longer be necessary. It is the licensee's intent to terminate the monitoring program with implementation of the approved license amendment.

The licensee is proposing to revise TS Section 5.5. The proposed changes will read as follows:

The north and south half of the spent fuel pool is analyzed to store 1840 and 2246 spent fuel assemblies, respectively, in storage racks containing the neutron absorber material Boral. The spent fuel pool is analyzed to store a total of 306 spent fuel assemblies in storage racks containing Boraflex. Both types of storage racks will maintain a kett of less than 0.95 under normal, abnormal and accident conditions. The spent fuel stored in both types of storage racks must have a peak lattice enrichment of 4.6% or less and the k-inf in the standard cold core geometry must be less than or equal to 1.31.

This approach ensures that all fuel assemblies loaded in the Nine Mile Point 1 Boraflex SFP storage racks are bounded by the nuclear criticality safety analysis submitted as part of this LAR.

3.2 Spent Fuel Pool Nuclear Criticality Safety Analysis Review 3.2.1 Spent Fuel Pool Nuclear Criticality Safety Analysis Methodology The methods used for the nuclear criticality safety analysis for fuel in the Nine Mile Point 1 SFP are described in Hl-2178001. The computer code benchmarking analyses supporting use of MCNP5-1.51 for nuclear criticality safety analyses are described in Hl-2104790. Although some SFP analysis deficiencies were identified during the review and will be discussed below, sufficient margin is built into the analysis methodology to offset the deficiencies for the existing fuel. Consequently, the methodology is specific to this analysis and, without further revision, is not appropriate for other applications. This is acceptable for the limiting fuel currently stored in the SFP.

3.2.1.1 Computational Methods For the criticality calculation, the licensee used MCNP5-1.51 with continuous energy cross-section data based on the Evaluated Nuclear Data File, Version 7 (ENDF/B-VII), neutron cross-section library. MCNP5-1.51 is a state-of-the-art Monte Carlo criticality code developed and maintained by Los Alamos National Laboratory for use in performing reactor physics and criticality safety analyses for nuclear facilities and transportation/storage packages. The code and its accompanying nuclear data sets have been extensively validated by the Los Alamos National Laboratory for various neutron transport calculations, including criticality calculations.

MCNP5-1.51 has been used in many nuclear criticality safety analyses and is an industry standard. Therefore, the NRC staff finds the underlying neutron transport methodology to be acceptable, but the code needs to be validated for specific applications. The NRC staff's review OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONbY PROPRIETARY INFORMATION of the licensee's validation of MCNP for its SFP nuclear criticality safety application is discussed in Section 3.2.1.2 of this safety evaluation.

For the depletion calculation used to determine the spent fuel isotopic compositions, the licensee used the two-dimensional (2-D) transport theory code CASM0-5 (CASM0-5),

Version 2.08.00, computer code with a 586-group cross-section library, mainly derived from the ENDF/B-VII neutron cross-section library. In some cases, the ENDF data has been supplemented by other data sources. CASM0-5 was approved by the NRC in a letter dated September 15, 2017 (ADAMS Accession No. ML17236A419), for depletion analysis with a wide range of pressurized-water reactor (PWR) fuel assembly designs. At the time, the NRC review of CASM0-5 did not extend to boiling-water reactor (BWR) fuel assembly designs, and as a result, CASM0-5 has not yet been approved for BWR depletion analyses.

From prior experience, the NRC staff has found that the limiting depletion conditions for BWR SFP nuclear criticality safety analysis using the peak reactivity method typically occur at 0 percent void at relatively low burnups. Zero percent void accelerates the gadolinium (Gd) burnout (depletion of Gd isotopes from gadolinia) for a given lattice, resulting in a higher maximum kint peak. The peak reactivity results provided in the LAR demonstrate that this is the case for the Nine Mile Point 1 nuclear criticality safety analyses (i.e., all the highest reactivity calculations occur under conditions with O percent void). The conditions of relatively low burn up at O percent void in the presence of Gd are similar to the conditions associated with the PWR validation data examined by the NRC staff during the review of CASM0-5. As a result, the PWR approval of CASM0-5 may apply to Nine Mile Point 1, a BWR, in this narrow set of conditions, provided there is reasonable assurance of similar isotopic composition predictions.

Therefore, the NRC requested information justifying the application of CASM0-5 to Nine Mile Point 1 for the narrow set of depletion conditions used to generate the peak reactivity predictions of the nuclear criticality safety analyses. Therefore, in order for the NRC staff to complete its technical review of the licensee's proposed changes, the licensee was requested by NRC e-mail dated July 3, 2018 (ADAMS Accession No. ML18184A288), to justify the application of CASM0-5 to Nine Mile Point 1 for the narrow set of depletion conditions used to generate the peak reactivity predictions of the nuclear criticality safety analyses.

In its response to Request for Additional Information (RAl)-1, provided in the supplemental letter dated August 17, 2018, the licensee identified those core operating parameters that are different between BWR and PWR fuel depletion and have an important impact on spent fuel isotopic compositions: void fraction, control blade insertion, and peak reactivity of BWR fuel due to the use of Gd. The licensee then assessed CASM0-5's ability to model the impact of these parameters. For both void fraction and control blade insertion, the manner in which these impact fuel depletion is a change in the energy of the neutron spectrum. Increased void fraction and increased control blade insertion result in a harder neutron spectrum. Since CASM0-5 utilizes a 586 energy group structure for microscopic cross-sections, 128 groups of which are at fast neutron energies, and the neutron energies span a range of 10-5 eV to 20MeV, the changes in the neutron energy spectrum due to void fraction and control blade insertion are well covered by the code. For peak reactivity of BWR fuel due to Gd, the licensee identifies that CASM0-5 was validated with cases of PWR fuel utilizing Gd as an integral absorber. Thus, CASM0-5 is also capable of modeling the inclusion of Gd in BWR fuel.

In addition to examining pertinent core parameters important to BWR fuel depletion, the licensee's response to RAl-1 also compared the depletion method differences between CASM0-5 and CASM0-4, which is a NRG-approved codes for both PWR and BWR applications. The comparison indicates the depletion methods between the two codes are very OFFICIAL USE ONbY PROPRIETARY INFORMATION

OFFICIAL US& ONLY PROPRl&TARY INFORMATION similar, and modeling improvements have been made to CASM0-5 in the areas of Gd depletion and fission product tracking. Specifically, CASM0-5 explicitly treats every isotope, which enhances the code's ability to predict depletion isotopic compositions more accurately and provides a finer resolution of inputs to MCNP5 for an improved determination of kett- Based on the RAl-1 response, the NRC staff finds the CASM0-5 computational methods are applicable to nuclear criticality safety analyses of BWR fuel.

The licensee also described its treatment of short-lived, volatile, and gaseous isotopes. [[

]]

For all MCNP5-1.51 calculations, the licensee used reasonable values for the following calculational parameters: number of histories per cycle, number of cycles skipped before averaging, total number of cycles, and the initial source distribution. More importantly, the licensee confirmed that all calculations converged using appropriate checks.

3.2.1.2 Computer Code Validation Since the nuclear criticality safety analysis credits fuel burnup, it is necessary to consider validation of the computer codes and data used to calculate burned fuel compositions and the computer code and data that use the burned fuel compositions to calculate kett for systems with burned fuel.

The purpose of the criticality code validation is to ensure that appropriate code bias and bias uncertainty are determined for use in the criticality calculation. The NRC staff used NUREG/CR-6698 as guidance for review of the criticality code validation methodology presented in the licensee's submittal. The basic elements of validation are outlined in NUREG/CR-6698, including identification of operating conditions and parameter ranges to be validated, selection of critical benchmarks, modeling of benchmarks, statistical analysis of results, and determination of the area of applicability.

In Hl-2104790, Revision 3, the validation of MCNP5-1.51 was performed by comparing calculated kett values with several different sets of critical configurations. A total of 562 critical configurations was included. The licensee determined that it would be appropriate to treat all experiments as a single set, but the licensee applied the distribution-free statistical approach to determine the bias and bias uncertainty because the data was not normally distributed. In addition, the licensee evaluated separate sets of bias and bias uncertainty based on subsets of the critical experiments that exhibited specific storage characteristics.

However, few relevant critical experiments are publicly-available for fission products and minor actinides, and therefore, direct validation is not feasible and cannot be directly included in the MCNP5-1.51 benchmark bias and bias uncertainty. Based on similar nuclear criticality safety OFFICIAL US& ONLY PROPRl&TARY INFORMATION

OFFICIAb USE ONbY PROPRIETARY INFORMATION analyses reviewed by the NRC, an additional uncertainty may be necessary to account for the lack of validation for keff calculations of burned fuel systems containing minor actinides and fission products. Work published in NUREG/CR-7109 indicates that a bias of 1.5 percent of the minor actinide and fission product worth should be sufficient to conservatively bound the uncertainty in fission product number densities from the depletion code (i.e., CASM0-5) and the resulting effect on the keff predicted by the criticality code (i.e., MCNP5-1.51 ). While the licensee indicates in the Criticality Analysis Checklist table of Appendix C of Attachment 3 to the LAR that such an uncertainty is considered, the magnitude of the uncertainty and how it was determined is not discussed in the LAR. Therefore, in RAl-2, the NRC staff asked the licensee to address the magnitude of the uncertainty and bias.

In its response to RAl-2, the licensee indicated that while the NUREG/CR-7109 recommendation of a bias of 1.5 percent of the reactivity worth of the fission product isotopes is applicable, it was not included for several reasons. First, the uncertainty in fission product number densities applied in prior nuclear criticality safety analyses was due in part to the nature of the depletion code lumping all the fission products together. CASM0-5 explicitly tracks individual isotopes, and therefore, a degree of high confidence exists in the code's ability to predict spent fuel number densities. While the NRC staff agrees that an increased accuracy in fission product number densities ought to result from explicit tracking, no information was provided by the licensee to quantify this impact. Second, a 5 percent depletion uncertainty has been included in the analysis, which, according to DSS-ISG-2010-01, is to cover the uncertainty in isotopic number densities generated during depletion simulations. This includes some consideration of the potential impact due to fission products and minor actinides. The NRC staff agrees with this statement. The application of the 5 percent uncertainty is discussed further below. Third, if the 1.5 percent bias were included, the resulting change in calculated keff would be very small. To demonstrate this, the licensee performed reactivity calculations on the limiting design-basis lattice with and without fission products present (except those considered allowable by NUREG/CR-7109). A 1.5 percent of the reactivity difference between these calculations is approximately 0.00015 ~k. This change in reactivity is readily accommodated by the margin to the regulatory limit as discussed in Section 3.2.5.1 of this safety evaluation.

The MCNP5-1.51 bias and uncertainty values for the Nine Mile Point 1 analyses, which are obtained from Revision 3 of validation report Hl-2104790, are slightly more conservative than Revision 1 of validation report Hl-2104790. Consequently, the validation report was not reevaluated as part of the review of this LAR. However, the licensee was still required to demonstrate how the validation report's findings were applied for this specific analysis.

In Table 2.1 of the LAR, the licensee compared the spectrum type, fuel density, lattice geometry, and material properties between the benchmarks and the Nine Mile Point 1 SFP conditions to demonstrate that the selected benchmarks are applicable. The separate bias/uncertainty sets were examined in Hl-2178001 to confirm that none of them were more limiting than the bias and bias uncertainty, as determined, based on all experiments for the conditions expected in the SFP racks.

The licensee did not perform any code validation for CASM0-5, although a proprietary report performed by Studsvik was referenced, which performed comparisons for different fuel assembly designs. Consistent with the guidance provided in the Kopp memo and DSS-ISG-2010-01, the nuclear criticality safety analysis incorporates a "5 percent of the reactivity decrement" uncertainty to cover lack of validation of spent fuel isotopic composition calculations. This uncertainty is calculated as 0.05 times the change in keff from the fresh fuel OFFICIAb USE ONbY PROPRIETARY INFORMATION

OFFICIAL USE! ONLY PROPRIETARY INFORMATION without gadolinia to the credited final fuel burnup at peak reactivity, including credit for residual gadolinia. This uncertainty was calculated by the licensee and applied correctly.

3.2.2 Spent Fuel Pool and Fuel Storage Racks 3.2.2.1 Spent Fuel Pool Water Temperature The SFP water temperature was treated in a bounding manner. The design-basis calculations were run using the maximum SFP temperature, and additional calculations were performed to verify that the minimum SFP temperature did not result in a higher keff value. This result is expected because the Boraflex SFP storage racks are being modeled without the presence of Boraflex. Configurations without strong neutron absorbers typically show a higher reactivity at a higher water temperature. The design-basis calculations were performed using the bounding normal condition moderator temperature of-65.55 degrees Celsius (°C) (338.7 Kelvin (K)).

Temperatures above this amount are considered to be accident conditions, and the licensee performed calculations to determine the reactivity effects up to a maximum temperature of 100 °C (373.15 K).

The NRC staff found that the licensee's use of SFP water temperature in the criticality analysis was treated in a bounding manner. As the approach used supports more reactivity in the modeled depleted fuel assembly, the NRC staff finds the licensee's approach to be acceptable.

3.2.2.2 Spent Fuel Pool Storage Rack Models Boraflex SFP fuel storage racks are situated north/south of each other in the southwest corner of the SFP. They are also bordered on the south and west by the SFP walls. To the north and east, the Boraflex racks are bordered by Boral racks, but separated by a 2-inch water gap on the north side and a inch water gap on the east side. These gaps form flux traps. The licensee will install 108 permanent cell blockers in a partial checkerboard pattern within the two Boraflex racks. The partial checkerboard pattern is an arrangement wherein [[

]].

The cell blockers are discussed in greater detail in Section 3.2.2.5 of this safety evaluation.

To facilitate the new nuclear criticality safety analysis and remove the Boraflex credit, the Boraflex panels are modeled as water (i.e., the empty location is filled with water). The modeling of the Boraflex panels as water is an acceptably conservative approach in that the essential geometry for neutron transport through the Boraflex is captured while removing an absorption mechanism for neutrons traveling through that region. The cell blockers are also modeled as water. The NRC staff finds the cell blockers modeled as water acceptable because modeling the cell blockers will displace water in the model, which is nonconservative when the storage racks are unpoisoned.

In the criticality analysis, the licensee chose to use a bounding approach in which the most reactive fuel assembly lattice is identified and then used as the design basis lattice. The SFP is then assumed to be fully loaded with this lattice at the exposure at which its in-rack reactivity reaches a maximum. A variety of 2x2 array fuel configurations (in which one of the positions always contains a cell blocker) are examined to identify the most reactive conditions. The fuel configurations examined include eccentrically positioned fuel, rotated fuel, and channeled and unchanneled fuel configurations. This ignores the presence of water gaps between the two Boraflex SFP storage rack modules. The potential for any water gaps would occur at the OFFICIAL USE! ONLY PROPRl6T.'\.RY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION interface between the Boraflex and Boral racks; therefore, the reactivity considerations associated with water gaps are discussed in Section 3.2.2.4 of this safety evaluation.

3.2.2.3 Spent Fuel Pool Storage Rack Models Manufacturing Tolerances and Uncertainties The manufacturing tolerances of the SFP racks contribute to SFP reactivity. DSS-ISG-2010-01 does not explicitly discuss the approach to be used in determining manufacturing tolerances; however, past practice has been consistent with the Kopp memo that determination of the maximum ketr should consider either: (1) a worst-case combination with mechanical and material conditions set to maximize ketr or (2) a sensitivity study of the reactivity effects of tolerance variations. If used, a sensitivity study should include all possible significant tolerance variations in the material and mechanical specifications of the SFP racks. The licensee chose to utilize the latter approach for addressing the manufacturing tolerances.

The licensee's evaluation of the tolerance variations included the following components: SFP cell inner width, SFP cell pitch, SFP cell wall thickness, and Boraflex panel width. Calculations were performed using the design-basis model in MCNP5-1.51, which combined the most limiting design-basis lattice with the limiting depletion condition set and SFP water temperature.

For each set of calculations associated with varying a specific parameter, the maximum reactivity increase was identified. These uncertainties were statistically combined with the other uncertainties and included in the final estimation of ketr- This is consistent with past precedent for criticality analyses and with the guidance provided in the Kopp memo, and thus, is acceptable.

The assumption made for the SFP cell wall thickness tolerance was that it was approximately 9 percent of the SFP cell wall thickness value. A review of data from previously submitted SFP nuclear criticality safety analyses associated with NRG-approved LARs from the past decade shows that a tolerance of approximately 9 percent would reasonably bound the tolerances used for BWR SFP storage racks. Based on this, the assumptions used for the unknown SFP cell manufacturing tolerances are acceptable.

The tolerance on the Boraflex sheathing thickness was not evaluated. An increase in the sheathing thickness would be expected to have a similar effect to an increase in the SFP cell wall thickness. If the reactivity effect of an increase in the sheathing thickness is exactly the same as that of the SFP cell wall thickness, then the result would be a 0.00026 ilk increase in reactivity. This increase can be accommodated by the available margin to the regulatory limit.

In addition to evaluating the manufacturing tolerances, the normal condition also includes many permutations of how fuel assemblies could be positioned in the SFP cells. The design-basis calculations were performed assuming a cell-centered loading with all fuel assemblies oriented similarly. Further calculations were performed using an infinite array of 2x2 cell configurations.

One series of calculations examined the impact of eccentric loading where fuel assemblies were placed in different positions within the SFP cell with and without channels, in an effort to determine which positions may increase reactivity. Another series of calculations was performed to determine if the reactivity would increase if the most reactive quadrants of adjacent fuel assemblies in a 2x2 array were oriented towards each other. The maximum positive reactivity differences, if any, for the eccentric positioning and eccentric rotation configurations were applied as biases, and the applicable bias uncertainties were statistically combined with the other uncertainties and included in the final estimation of keff.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.2.2.4 Spent Fuel Pool Storage Rack Interfaces The criticality analysis assumes an infinite array of 2x2 storage cell configurations with the most limiting fuel lattice loaded into three of the four cells (the fourth cell being the location of the cell blocker, which is modeled as water). As a result, the water gap of 0.375 inches separating the two SFP Boraflex racks appears to be neglected in the modeling. For situations where the fuel lattice is under-moderated, it is possible that additional moderator between adjacent fuel assemblies (such as the existence of a water gap between SFP racks) may result in higher reactivity. However, the results from the calculations examining eccentric loading configurations (where the fuel assemblies in each of the storage cells are closer together) indicate that removing the water gap is more conservative. Therefore, including the water gap would be expected to reduce the reactivity. This conclusion is supported by multiple uncertainty analyses that indicate that reducing the amount of moderator and distance between fuel assemblies will result in higher reactivities. Therefore, reasonable assurance exists that the water gap has been conservatively modeled in the determination of the interface bias.

The SFP Boraflex rack modules are manufactured by welding together a checkerboard arrangement of steel boxes containing Boraflex in every cell along the periphery of the SFP Boraflex rack module and bounded by a Boraflex box on only one side. The opposite side along the SFP Boraflex rack module edge is covered by a steel plate. As a result, initial concerns existed that it may be possible for two SFP racks to be adjacent to each other in such a way that the gaps in the Boraflex lined up, leaving only steel filler plates between two fuel assemblies.

However, examination of the orientation of the Boraflex racks indicates this configuration is not possible.

Another interface that may exist is along the edge of the SFP Boraflex rack modules adjacent to the west and south SFP walls. The water and concrete act as reflectors, returning a certain percentage of neutrons back into the SFP cells without Boraflex boxes installed on the sides of the rack modules facing the SFP walls. However, this configuration is bounded by the above interface evaluation between two SFP rack modules. The symmetry of the interface between two SFP rack modules, with no gap between the modules, is such that the configuration emulates a perfect reflection of neutrons back into the SFP cell for each wall with no Boraflex box installed. At the SFP wall, the reflection will be imperfect, and some neutrons are expected to be lost due to leakage.

Of special consideration are the interfaces that exist along the north and east sides of the Boraflex rack modules. These sides are adjacent to high density Boral storage racks. Along the northern interface, a water gap of 2 inches exists between the Boraflex and Boral storage racks, while along the eastern interface, a water gap of 3.15 inches exists between the two rack types. The licensee evaluated each of these interfaces.

The licensee indicates that, along the east-west interface, the Boraflex and Boral racks share very similar geometry (i.e., cell inner diameter and cell pitch are nearly identical), and the two racks possess nearly identical calculated keff values of approximately 0.94 (determined by including all biases and uncertainties). As a result, from the perspective of the Boral storage racks, the Boraflex racks are a continuation of the Boral storage configuration, but a continuation where the water gap is acting as an additional flux trap and the fuel assemblies are separated by empty cells (cell blockers) instead of neutron absorbers (effectively performing the same function). From the perspective of the Boraflex storage racks, the Boral racks are a continuation of the Boraflex storage, just with a larger flux trap (the water gap) and added neutron absorbers. It is, therefore, anticipated that along the east-west interface, neither the OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Boral nor the Boraflex racks will increase the reactivity of the other. The NRC staff finds this assessment to be reasonable.

Along the north-south interface, the licensee indicates that the same assessments apply, but with the additional observation that the presence of the Boraflex boxes increases the cell pitch, further reducing reactivity. It is, therefore, anticipated that along the north-south interface, the Boraflex and Boral racks will not increase the reactivity of the other. The NRC staff finds this assessment to be reasonable.

Therefore, the evaluation of the interface between SFP rack modules performed by the licensee is sufficient to bound all possible interfaces that may exist in the SFP. The possible reactivity impacts of the SFP rack interfaces was qualitatively assessed and found to be negligible. The NRC staff concludes that this is acceptable.

3.2.2.5 Spent Fuel Pool Storage Rack Cell Blockers To ensure that fuel assemblies are not inadvertently placed into the designated unusable cell in each 2x2 array of storage locations, the licensee will block the unusable cell by installing a structure that occupies the space and physically prevents placement of a fuel assembly. The licensee refers to these structures as "cell blockers."

The cell blockers are fabricated from austenitic stainless steel and are dimensionally sized to accomplish two purposes: to prevent insertion of a fuel assembly in a designated, unusable fuel storage location, and to fit within a fuel storage rack cell with sufficient clearance on all sides to maintain the free movement of water into and out of the cell. The clearance on all sides is enough to allow for the free movement of water, but not so large that the cell blocker can be significantly tilted within the cell.

Nuclear Energy Institute (NEI) 12-16, Revision 3, "Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants," dated March 2018 (ADAMS Accession No. ML18088B400), Section 6.3.3.4, provides recommendations on the design of blocking devices such as the cell blockers proposed for use by the licensee at Nine Mile Point 1. At the time of the writing this safety evaluation, NEI 12-16 is not officially endorsed by the NRC.

However, the NRC staff has examined the recommendations made within NEI 12-16, Section 6.3.3.4, and found the recommendations speak to the NRC staffs concerns regarding blocking devices for the present review. The recommendations cover the areas of inadvertent installation/removal and movement, allowing continued water flow through the storage cell, and structural soundness during nominal conditions and in case of accidental assembly drop.

For the first of the recommendation areas, the licensee identified that the cell blockers are designed such that specialized handling tools are required for their installation and removal.

The cell blocker lift point is designed such that it cannot be engaged by any traditional or readily available tooling on the refuel bridge (i.e., j-hooks, fuel grapples, etc.). Additionally, the cell blockers shall be locked-in-place using a lock-out cap installed over the lift point, creating a physical barrier preventing any unintended engagement of the lift point with any tool. With regard to the recommended area of maintaining water flow through the storage cell, the interior of the cell blockers is not sealed, and flow holes exist in the top and base plates of the cell blocker. The cell blocker lock-out cap also contains flow holes, and the overall design of the cell blocker does not form a seal with the cell walls. These features allow for continued and sufficient flow of water through the blocked storage cell.

OFFICIAL lJSli ONLY PROPRlliTARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION With regard to the NEI 12-16 recommendations concerning structural soundness, the licensee examined three conditions: limiting normal conditions, limiting upset conditions, and limiting accident conditions. The licensee identified that the limiting normal condition for a cell blocker is handling, and the licensee evaluated the condition with stress design factors for noncritical lifts from NUREG-0612. The licensee indicated that, assuming a hoist load factor of 0.15, all stress design factors are satisfied for the cell blocker, and therefore, the cell blocker is structurally adequate to support its own weight during lifting.

The limiting upset condition for a cell blocker was identified as a static vertical compressive load onto the top of the cell blocker such that the fuel underload trip sensor would activate, which will occur when the fuel handling machine load is :s; 700 pounds. The acceptance criterion for this evaluation is that no component of the cell blocker shall experience a stress exceeding 80 percent of the material yield strength. Strength of material and ANSYS evaluation were performed by the licensee for the cell blocker. These evaluations showed the lock-out cap plate is structurally sound to support the load and that the cell blocker does not exceed the 80 percent material yield strength criterion. The results from the buckling evaluation performed also confirm the cell blocker is structurally adequate. Therefore, the cell blocker will support the 700 pound compressive load and activate the underload trip sensor.

The limiting accident condition for the cell blocker was identified as an impact from above by a dropped fuel assembly. The licensee evaluated this condition using the LS-DYNA based finite element impact analysis method. LS-DYNA is a general-purpose finite element program capable of simulating complex real-world problems. This method has been used by Holtec International in previous spent fuel and dry fuel storage analyses reviewed by the NRC. The acceptance criterion was taken to be that no portion of the active length of the dropped fuel assembly may be axially-positioned adjacent to any portion of the active length of a fuel assembly stored in a neighboring cell. To achieve this, the cell blocker would need to be compressed by no less than 13.375 inches during the accidental assembly drop. The finite element analyses demonstrated that the maximum vertical deflection of the cell blocker is 3 inches, which is less than the limit of 13.375 inches.

In addition to addressing the NEI 12-16 recommended criteria for cell blockers, the licensee also examined the SFP storage rack qualification in light of the installation of cell blockers. This evaluation found that, given the lower weight of the cell blocker as compared to a fuel assembly, the loaded weight of the rack with cell blockers and fuel assemblies is bounded by the rack seismic analysis where every storage cell is loaded with a fuel assembly. Therefore, the stresses of the rack under the loaded weight of fuel assemblies and cell blockers will remain within allowable limits.

3.2.3 Fuel Assembly 3.2.3.1 Bounding Fuel Assembly Design Section 2.3.2 of Hl-2178001 provides information on the process of selecting the design-basis lattice that is used to perform the final, detailed nuclear criticality safety calculations. This is essentially a four-part process that uses different methods to progressively narrow the field of candidate lattices and then, for additional conservativeness, bias several lattice parameters to generate a more reactive, bounding "super lattice." The licensee considered all current and legacy fuel assembly designs, including 7x7, 8x8, 9x9, and the Global Nuclear Fuel (GNF) 2 1Ox10 fuel design. The design-basis fuel assembly selection was evaluated consistent with Section IV.1 of DSS-ISG-2010-01.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION The process of narrowing the field of candidate lattices is based on determining the burnup at which the lattice reactivity will peak (the most reactive burnup) within the SFP rack model. The in-rack model is used because the reactivity burnup of a lattice is dependent upon the SFP geometry and neutronic environment. Within the LAR, this is referred to as the "peak reactivity method." In contrast, the licensee's relevant TS states that each lattice must have a standard cold core geometry (SGGG) reactivity less than or equal to 1.31. The SGGG approach determines the most reactive burn up of the lattice in an uncontrolled in-core geometry.

Typically, some maximum lattice SGGG kint is determined such that the same lattice also meets the in-rack regulatory requirement of kett < 0.95. The approach includes a bias to account for differences between fuel vendor codes and the criticality analysis codes. For Nine Mile Point 1, the fuel vendor bias is 0.01. The fuel vendor SGGG kint plus code bias is then used to ensure all future lattices within the current fuel design with an SGGG kint less than the TS limit of 1.31 are qualified for storage in the SFP. In the licensee's nuclear criticality safety analysis, the peak reactivity lattice that is used to show compliance with regulatory requirements (i.e., the in-rack kett) is also used to show compliance with the TS SGGG limit (i.e., the in-core kint). In other words, the peak reactivity methodology and the SGGG methodology are combined by analyzing both the in-rack and in-core models to find the most reactive lattices. The bounding lattices are then ultimately used in the in-rack model criticality analysis of record.

The first step in finding the bounding lattice is a series of screening calculations, which is discussed in Section 2.3.2 of Hl-2178001. All unique lattices (with the exception of the natural uranium blankets, which are extremely unlikely to be limiting) of the candidate fuel assemblies were evaluated with GASM0-5. Each lattice was depleted separately in a 2-D GASM0-5 calculation using periodic boundary conditions (consistent with typical 2-D depletion methodologies) for in-rack and in-core models. Four different sets of core operating parameters are used for each lattice to ensure that the impact of the most important parameters on each lattice is considered, as described in Hl-2178001, Table 5.6. At points near the peak reactivity for the fuel lattice, the model (including isotopic compositions) is reinitialized in both the SFP rack and in-core environments to calculate its in-rack and in-core kinf. This approach ensures that the peak kint value will be identified for each lattice in both environments by clearly defining the kint trend and its peak. The set of actual lattices having the highest peak reactivity (either in the SGGG or in-rack configurations) is selected for the design-basis analysis performed in the next steps.

The "screen out" criterion used to determine those lattices for which further evaluations will be performed was not presented by the licensee in the LAR. In prior reviews, the NRG staff has observed an acceptance criterion based on the in-rack kint- The primary consideration for whether the value for this criterion is acceptable is whether the kint threshold is set sufficiently low such that reasonable assurance exists that variability in the bias of GASM0-5 compared to MGNP5-1.51 will not change the final selection of the design-basis lattice. The NRG staff observation is that several of the lattices that were screened out possess predicted in-rack kint reactivities greater than two of the seven lattices that were screened in. To assess whether the screening criterion used is sufficient to avoid a change in the selection of the design-basis lattice, the NRG staff asked the licensee in RAl-3 about the adequacy of the screening criterion.

The licensee's response to RAl-3 indicated that the number of lattices and which physical design to use is arbitrary, with the results of the screening calculations used to inform the decision. The licensee also provided the results of the screening calculations in the RAI response along with the thought process underlying the engineering judgement used to screen-in lattices. The NRG staff noted that the thought process contains three key elements.

First, there is the discernment of a reactivity "break point" above which there is a tight grouping OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION of predicted in-rack reactivities, and below which there is a wide spread of predicted in-rack reactivities. Second, there is the observation that the tight grouping of predicted in-rack reactivities above the break point is all attributed to two bundle designs. Third, there is an effort to select lattices representative of all locations along the bundle axial length.

It is this third element that results in the NRC staff's initial observation of several screened-out lattices having predicted in-rack reactivities greater than two of the screened-in lattices. Five of the seven screened-in lattices comprise the full list of lattices above the reactivity break point, but these five lattices do not represent every location along the bundle axial length. The licensee, therefore, supplemented them with the two most reactive lattices representative of the missing locations along the bundle axial length. Additionally, based on the results provided in Hl-2178001, Appendices A and B for the highest reactivity lattices, the range of variability observed in the bias of CASM0-5 relative to MCNPS-1.51 is [[ ]]. The difference between the maximum in-rack kint value calculated for all lattices and the bias variability suggests a minimum acceptable screening criterion would be [[ ]]. This value is approximately that of reactivity break point, indicating there is reasonable assurance the final selection of the design-basis lattice from among the seven screened-in lattices will not change due to the variability in bias between the codes. Given this result, the identification of a predicted in-rack reactivity break point, and the diligence in including lattices representative of the entire bundle axial length, the NRC staff finds that, for the present nuclear criticality safety analysis, the RAI response is acceptable.

The second step in finding the bounding lattice involves performing design-basis depletion calculations with the most reactive lattices from the previous step. In these calculations, CASM0-5 is used to determine pin-specific isotopic compositions with radial Gd depletion. As in the first step, the four sets of core operating parameters are utilized to determine the bounding set. Once the bounding set of core operating parameters is determined, the CASM0-5 isotopic compositions are utilized in SCCG and in-rack MCNPS-1.51 calculations to confirm the set of bounding parameters is consistent between both methods. The set of bounding core operating parameters is then used in all further analyses.

The third step in finding the bounding lattice involves the development of "super lattices." Super lattices are developed by increasing the enrichment of each fuel pin in the most reactive lattices to 4.6 weight (wt) percent U-235 (to be consistent with the relevant TS) and reducing the loading of the Gd rods. This results in large reactivity increases for the set of most reactive lattices.

Thus, super lattices bound actual lattices. The bounding Gd loading is found by adjusting the loading in increments of 0.5 wt percent Gd and confirmed by varying the loading by 0.1 wt percent.

The fourth and final step to determine the most limiting lattice is to perform MCNPS-1.51 calculations on the group of super lattices for both in-rack and SCCG models over the range of Gd loadings examined in the previous step. The CASM0-5 depletion isotopic compositions are used by MCNPS-1.51 in these calculations. For a given super lattice, SCCG calculations are performed to determine the Gd that generates an SCCG kint of at least 1.31. In-rack calculations are then performed on the same super lattice to find the in-rack kett associated with that Gd loading. The super lattice with the highest in-rack kcarc (kett before including biases and uncertainties) at the Gd loading that produces an SCCG kint of at least 1.31 is taken as the most reactive bounding super lattice. For additional conservatism, the licensee further reduces the Gd loading of the most reactive bounding super lattice by 0.4 wt percent, thereby making it more reactive, and uses this super lattice to perform the nuclear criticality safety analysis.

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OfflCIAL USE! ONLY PROPRIET.'\RY INFORMATION This result of the above methodology is a design-basis super lattice with a uniform enrichment of 4.6 wt percent U-235 and some Gd loading that yields an SCCG kint conservatively greater than 1.31. Utilizing this super lattice in the full nuclear criticality safety analysis should yield an SCCG kint greater than the TS limit of 1.31, yet have a kett less than the regulatory limit. In part, the purpose of this approach is to allow for the qualification of future lattices within the current fuel design. Therefore, since the reactivities of all the legacy fuel designs were bounded by the reactivity of the GNF2 fuel design, and the nuclear criticality safety analyses were performed using GNF2 fuel design tolerances, any lattices from fuel designs other than these will need to be evaluated in the SFP rack configuration. In this manner, any impacts from how the nuclear composition affects the burnup characteristics of the fuel will be explicitly considered.

Therefore, the approach of using a design-basis super lattice with a nuclear composition that is significantly different (and more reactive) than existing lattices is acceptable. The resulting bounding super lattice was used in all design-basis calculations. The bounding parameter set was confirmed for use in the design-basis calculations. These studies are discussed in Section 3.2.3.3.3 of this safety evaluation.

Part of the purpose of combining the peak reactivity method and the SCCG method to create a bounding super lattice is to allow for the qualification of future lattices within the current fuel design for storage in the SFP, provided their vendor maximum SCCG kint is less than or equal to the TS requirement of 1.31. However, Section 7.1 of Hl-2178001 states that "the SCCG methodology is not an exact methodology since it cannot be concluded that all lattices with an SCCG kint of 1.31 correspondingly have the same reactivity in the [[ ]]." The NRC staff examination of Hl-2178001, Figures B.1 and B.2, which show MCNP5-1.51 design-basis reactivity predictions for lattices in both the SCCG and in-rack geometry, confirms this statement; relative to each other, some lattices are less reactive in the in-rack geometry than the SCCG, while others are more reactive. In light of this, it may be possible that future lattices for which nuclear criticality safety analyses have not been performed (e.g., as a result of a new number of Gd rods or new Gd rod locations) may not be bounded by the lattices developed in the peak reactivity method, even if the future lattices are within the current fuel design.

Therefore, with these future lattices in place, there is a possibility that the actual kett of the spent fuel storage racks may not meet the 0.95 regulatory requirement on reactivity.

The NRC staff requested the licensee in RAl-5 to provide assurance of meeting the regulatory requirement regarding the qualification of future lattices. In its response to RAl-5, the licensee indicated that the reactivity difference between super lattices that has essentially the same SCCG reactivity but different in-rack reactivity is due primarily to the number of fuel rods, and not the number and location of Gd rods, which has a considerably small effect. This is demonstrated within the results the licensee presented in the RAI response. These results show the effect due to changing from 94 full-length fuel rods to 84 full-length and 6 part-length rods is approximately 0.006 ilk, while the effect due to changes in Gd rod location is

.0.0005 ilk. Additionally, the licensee indicated the reactivity difference between the bounding super lattice and the most reactive actual lattice is about 0.04 ilk, demonstrating the conservatism in the super lattice approach.

Although the demonstrated change between SCCG and in-rack reactivities due to the number of fuel rods is not negligible, any change in the number of fuel rods constitutes a change in fuel design. For a future lattice of a different fuel design, the current analysis would not be bounding, and the future lattice would not be qualified for storage within the SFP. The licensee would need to redo the criticality analyses to demonstrate that the new fuel design will meet regulatory requirements, given the TS limit. Thus, the only viable concern regarding the observed differences between SCCG and in-rack reactivities amounts to the small 0.0005 ilk, OFFICIAL USE ONLY PROPRIET.'\RY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION due to the number and location of Gd rods. This change in reactivity is readily covered by the available margin, as discussed in Section 3.2.5 of this safety evaluation. Therefore, the NRC staff finds the response to RAl-5 acceptable.

3.2.3.2 Fuel Assembly Manufacturing Tolerances and Uncertainties The manufacturing tolerances of the storage racks and fuel assemblies contribute to SFP reactivity. DSS-ISG-2010-01 does not explicitly discuss the approach to be used in determining manufacturing tolerances, but past practice has been consistent with the Kopp memo that determination of the maximum kett should consider either: (1) a worst-case combination with mechanical and material conditions set to maximize kett or (2) a sensitivity study of the reactivity effects of tolerance variations. If used, a sensitivity study should include all possible significant tolerance variations in the material and mechanical specifications of the fuel and racks. The licensee chose to use the latter approach for the fuel assembly manufacturing tolerances.

The licensee's evaluation of the tolerance variations included the following components: fuel pellet density, fuel pellet outer diameter, fuel cladding thickness, fuel pin pitch, channel wall thicknesses, and water rod thicknesses. Calculations were performed using the design-basis model in MCNP5-1.51, which combined the most limiting design-basis lattice with the limiting depletion condition set and limiting SFP water temperature. Boraflex was not modeled so that the credit for it may be removed. For each set of calculations associated with varying a specific parameter, the maximum reactivity increase was identified, and the uncertainty was determined.

These uncertainties were statistically combined with the other uncertainties and included in the final estimation of kett. The NRC staff's evaluation of the licensee's methodology to adequately analyze uncertainties associated with this submittal is consistent with other criticality analyses completed. Additionally, the licensee's analysis is consistent with the guidance as provided in the Kopp memo and is, therefore, acceptable.

The licensee considered various irradiation-caused fuel assembly geometry changes, including fuel rod growth, cladding creep, and channel bulging/bowing. For each set of calculations associated with varying a specific parameter, the maximum reactivity increase was identified, and the bias and bias uncertainty were determined. The bias uncertainties were statistically combined with the uncertainties from other calculations (e.g., depletion uncertainty), while the maximum bias of the results was combined with the biases from other sources (e.g., MCNP code bias). Both were included in the final estimation of kett. The NRC staff's evaluation of the licensee's methodology to adequately analyze uncertainties and biases associated with the licensee's submittal is consistent with other criticality analyses completed. Additionally, the licensee's analysis is consistent with the guidance as provided in the Kopp memo and is, therefore, acceptable.

The licensee only considered the GNF2 design in the full evaluation of the biases and uncertainties due to manufacturing tolerances, fuel assembly geometry changes due to irradiation, and fuel assembly positioning within the SFP cell. The tolerances for other fuel assembly designs were not evaluated because the reactivity of the GNF2 design bounded all the reactivities of the legacy fuel designs. The reactivity difference between the design-basis lattice and the most limiting fuel lattice currently being used or stored at Nine Mile Point 1 is 0.0477 ~k. which is much more than any potential variation in the manufacturing tolerances from use of a different fuel assembly design. Therefore, explicit calculation of the manufacturing tolerances for other fuel assembly designs is not necessary.

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OfflCIAb US& ONbY PROPRll!TARY INFORMATION The NRC staff evaluated the treatment of various fuel manufacturing tolerances, uncertainties, and other potential differences between the fuel lattice model and "real-world" 1\Jel lattices likely to be found at Nine Mile Point 1 and found the treatment to be acceptable.

3.2.3.3 Spent Fuel Characterization Characterization of fresh fuel is based primarily on U-235 enrichment, fuel rod gadolinia content and distribution, and various manufacturing tolerances. The manufacturing tolerances are typically manifested as uncertainties or are bounded by values used in the analysis. These tolerances and bounding values would also carry through to the spent fuel; common industry practice has been to treat the uncertainties as unaffected by the fuel depletion. The characterization of spent nuclear fuel is more problematic. Its characterization is based on the specifics of its initial conditions and its operational history in the reactor. That characterization has three main areas: ( 1) a burn up uncertainty, (2) the axial apportionment of the burn up, and (3) the core operation that achieved that burnup.

3.2.3.3.1 Burnup Uncertainty In the Kopp memo, the NRC staff provided its recommended method for evaluating burnup uncertainty:

A reactivity uncertainty due to uncertainty in the fuel depletion calculations should be developed and combined with other calculational uncertainties. In the absence of any other determination of the depletion uncertainty, an uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest is an acceptable assumption.

The licensee used this approach to address the uncertainty in the burned fuel compositions.

This uncertainty was statistically combined with the other uncertainties and included in the final estimation of kett.

The licensee used CASM0-5 to calculate the isotopic composition of the spent fuel as a function of fuel burn up, initial feed enrichment, and decay time. As discussed in Section 3.2.1.1 of this safety evaluation, the NRC staff has approved CASM0-5 for PWR depletion, but it has not been approved for BWR depletion calculations. From the discussion in Section 3.2.1.2 of this safety evaluation, the NRC staff has found the application of CASM0-5 to Nine Mile Point 1 for the set of depletion conditions used to generate the peak reactivity predictions of the nuclear criticality safety analyses to be acceptable. Therefore, the licensee's application of a 5 percent reactivity decrement is acceptable. The reactivity uncertainty from the 5 percent decrement was determined based on comparison of the reference case at peak reactivity with a fresh, no-gadolinia calculation in a manner consistent with the approach previously accepted by the NRC.

3.2.3.3.2 Axial Apportionment of the Burn up or Axial Burn up Profile The standard BWR peak kint analysis technique uses either a 2-D model or a 3-dimensional (3-D) model with uniform axial burnup distributions. Generally, this is appropriate because the peak in limiting assembly reactivity occurs at lower burnups where the uniform axial burnup distribution is conservative. If one were to credit assembly burnup beyond the limiting peak reactivity burnup, at some assembly burnup value, the use of the uniform axial burnup would become nonconservative.

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OfflCl.6.b USE ONLY PROPRIETARY INFORMATION The licensee chose to adopt the standard approach for dealing with the axial burnup distribution using a 3-D model, with the limiting lattice modeled along the entire active fuel length. Twelve inches of water was modeled above and below the fuel to act as an axial reflector. Review of the Nine Mile Point 1 SFP configuration, as described in the UFSAR, shows that there is at least that much flooded space above and below the active fuel length. Neglecting the structural materials is conservative because the stainless steel structural material will absorb some neutrons rather than reflecting them back into the active fuel. Twelve inches is sufficiently large compared to the diffusion length for water at room temperature such that the reflection of neutrons back into the active fuel is comparable to that of an infinite thickness of water above and below the active fuel.

Since the licensee applied a conservative approach in the use of standard peak reactivity analysis methods for modeling the axial burn up distribution, the NRC staff finds this approach acceptable.

3.2.3.3.3 Burnup History/Core Operating Parameters The reactivity of light-water reactor fuel varies with the conditions the fuel experiences in the reactor. This is particularly true for BWR fuel nuclear criticality safety analyses using the peak kint analysis method. As a result of using gadolinia in fuel rods, fuel assembly reactivity initially increases as the gadolinium isotopes are depleted. The value of the in-rack kett at peak reactivity is affected by the reactor depletion parameters in several ways. Factors that lead to a more thermal neutron energy distribution cause the Gd isotopes (specifically, gadolinium-155 and gadolinium-157) and fission products to deplete more quickly and reduce plutonium generation. This causes the peak reactivity condition to be reached earlier, achieving a higher in-rack kett value. Increased water density and decreased void fraction lead to a more thermal neutron energy distribution and to lower fuel rod temperatures due to improved fuel rod cooling.

Factors that lead to a less thermal neutron energy distribution cause the Gd and fission products to be depleted more slowly and result in increased plutonium generation from neutron energy spectrum hardening. Decreased water density, increased void fraction, and control rod usage all result in neutron energy spectrum hardening. For fuel and moderator temperatures, NUREG/CR-6665 recommends using the maximum operating temperatures to maximize plutonium production. This recommendation is also applied to the moderator density for BWRs, but in practice, the high-void state is typically not the limiting condition for peak reactivity analyses. The limiting lattice k;nt value is established as the maximum value for a given fuel lattice under all possible operating conditions. The higher moderation that occurs in the no-void condition results in an elevated depletion rate of the gadolinia, causing the k;nt to peak earlier and higher. A lower moderator density results in a harder neutron spectrum and increased plutonium production, but this effect is generally not large enough in BWRs to compete with the U-235 depletion that occurs prior to the later peak in k;nt.

Typically, sensitivity studies are performed where each core operating parameter is varied to determine a bounding value that will maximize the impact to the reactivity of the design-basis lattice in the SFP. The reactivity impact of specific core operating parameters may vary for different fuel lattice designs. Therefore, this should also be considered in the design-basis lattice selection. The licensee considered the impact of core operating parameters on the design-basis lattice selection by using four different sets of core operating parameters in the screening calculations that were selected to consider the most likely candidates for limiting conditions. When taking into consideration prior experience with nuclear criticality safety analyses and how the final set of core operating parameters selected compares to the other OfflCIA-b USE ONLY PROPRIETARY INFORMATION

OfflCIAb USE ONbY PROPRIETARY INFORMATION three, the licensee's consideration of how the core operating parameters may impact the design-basis lattice selection was acceptable. However, the licensee did not do specific sensitivity studies that vary individual core operating parameters to verify that the selected value would result in the most positive SFP reactivity.

The NRG staff requested the licensee in RAl-4 to justify the lack of parameter-specific sensitivity studies. In its response to RAl-4, the licensee indicated that bounding core operating parameters were selected when determining the design-basis lattice selection by performing additional sensitivity studies wherein individual core operating parameters were set to extremes.

Additionally, the licensee's response discussed the lack of necessity in evaluating multiple intermediate steps between extremes, indicating the only core operating parameter that may potentially result in a higher reactivity at an intermediate step would be void fraction, but that the increase in reactivity would be slight. The NRG staff agrees with the presented discussion, and notes that the potential increase in reactivity due to an intermediate void fraction would be so slight as to be negligible. Higher void fractions promote plutonium buildup, while lower void fractions more rapidly deplete gadolinia. Both effects cause an increase in reactivity, but the increase in reactivity due to gadolinia depletion dominates. While an intermediate void fraction could exist wherein some plutonium buildup occurs, its contribution to the overall reactivity increase would be trivialized by the gadolinia depletion. Therefore, the selected core operating parameters for the design-basis lattice are acceptable.

NUREG/CR-6665 does riot have a specific recommendation for specific power and operating history. NUREG/CR-6665 estimated this effect to be about 0.002 ~kett, using operating histories it considered. Based on the difficulty of reproducing a bounding or even a representative power operating history, NUREG/CR-6665 merely recommends using a constant power level and retaining sufficient margin to cover the potential effect of a more limiting power history.

The licensee chose to perform the design-basis calculations at a maximum value for the power density, and prior NRG experience is that a modest decrease in power density does not result in a reactivity increase. The moderator temperature cannot increase above saturation, and the impact on void fraction is already accounted for by the conservatively high void fraction used in the sensitivity studies on the core operating parameters. Consequently, no power history uncertainty is applied in the final design-basis calculation. While the licensee has not demonstrated that the use of a constant maximum power density bounds all possible operating histories, the final margin to the regulatory limit is sufficiently large to accommodate the estimated 0.002 ~kett from NUREG/CR-6665. Therefore, the NRG staff does not consider it to be necessary to perform a more detailed sensitivity study or otherwise justify the lack of a power history uncertainty.

The licensee considered the impact of operation with control blades inserted as it affects the reactivity of the discharged assembly by establishing two broad bounding conditions: operation with control blades inserted and operation with control blades withdrawn. The evaluation of the operating scenario where control blades are inserted is a somewhat artificial situation, because it does not account for the reduction in power density resulting from insertion of the adjacent control blade. However, the change in power density is small compared to the reactivity impact of the harder neutron spectrum due to the presence of the control blade.

As such, this approach is acceptable, especially since a fuel assembly would be unlikely to be controlled during its entire depletion to peak reactivity. The results show that operation with control blades inserted results in a more limiting peak in-rack reactivity. Although no discussion was presented in the LAR evaluating the effect different control blade designs may have on OFFICIAb USE ONbY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION operation with the control blades inserted, the NRC staff concludes the results are not expected to differ from one control blade design to another due to the similar impact the blades have on the neutron flux during core operation.

3.2.3.3.4 Integral Burnable Absorbers As is typical for BWR plants, Nine Mile Point 1 utilizes gadolinia poison to help control reactivity and peaking within fuel assemblies. The specific characteristics of the gadolinia poison loading may affect the relationship between the SCCG kint and the in-rack kinf. The licensee chose to perform an evaluation of the in-rack kint and the SCCG kint as part of the screening process for all lattices considered in this nuclear criticality safety analysis. As a result, the various gadolinia loading patterns and their impact on the depletion has been explicitly captured in the calculations. No removable burnable absorbers are used, so there is no need for any further evaluation of the burnable absorbers. Based on the incorporation of gadolinia patterns and impacts in the depletion analysis, the NRC staff finds that the treatment of burnable absorbers is appropriate for the specific conditions at Nine Mile Point 1.

3.2.4 Analysis of Abnormal Conditions Section 2.3.15 of Hl-2178001 presents the abnormal conditions considered in the analysis. The licensee considered the following abnormal conditions:

• SFP temperature exceeding the normal range

• Dropped fuel assembly (horizontal and vertical)

• SFP rack movement due to seismic activity

• Mislocated fuel assembly (fuel assembly positioned outside the SFP rack)

• Misloaded fuel assembly (fuel assembly positioned in wrong location within SFP rack)

Explicit calculations were performed for high and low SFP temperatures, SFP rack movement, and a mislocated fuel assembly (for the only possible location outside the SFP, where the two Boraflex racks meet along the west SFP wall). The limiting condition was found to be when rack movement occurred to minimize the distance between the rack and SFP wall, concurrent with rack-centered eccentric positioning of the fuel.

The same uncertainties and biases from the normal condition calculation were then applied to obtain the kett for the accident condition. The licensee considered it to be unnecessary to perform calculations for the rest of the scenarios for the following reasons:

• Horizontally dropped fuel assembly - The active fuel length in the dropped assembly would be separated from the active fuel length in other assemblies by at least 12 inches by the SFP rack; this configuration is not considered to be more reactive than the mislocated fuel assembly assessments.

• Vertically dropped fuel assembly - A drop onto a stored fuel assembly would cause a small compression of the assembly, but result in no increase in reactivity due to 12 inches of separation between active areas of fuel within the assemblies. A drop onto a cell blocker would result in minimal damage to the cell blocker (a 3-inch compression at most) and no increase in reactivity due to the distance maintained between the active fuel length of the dropped assembly and the active fuel length of any neighboring assembly (as discussed in Section 3.2.2.5 of this safety evaluation).

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• Misloaded fuel assembly - Since the design-basis model evaluates the storage rack locations fully loaded with the most reactive lattice, and the blocked cells do not allow a misplaced bundle, the mislead accident is not a credible accident.

With respect to the scenarios that were not explicitly evaluated, the NRC staff agrees that the horizontally dropped fuel assembly configuration is not likely to be limiting relative to the mislocated fuel assembly scenario. The most limiting mislocated fuel assembly scenario has a fuel assembly almost directly faced adjacent to two fuel assemblies concurrent with rack-centered eccentric positioning of the fuel. This configuration is surrounded, for the most part, by fuel. By contrast, the horizontal dropped fuel assembly would be expected to exhibit weak neutronic coupling with the fuel in the SFP due to the distance, and most of the neutrons associated with the assembly would be lost to leakage. The modeling approach in which the water gap between SFP racks is neglected is discussed in Section 3.2.2.4 of this safety evaluation, and includes the NRC staffs assessment that the decision to neglect the water gap appears to be conservative.

3.2.5 Margin Analysis and Comparison with Remaining Uncertainties This section provides evaluation of additional conservatism in the analysis and evaluation of items that may have been treated nonconservatively.

3.2.5.1 Potential Nonconservatisms Several potentially nonconservative assumptions were identified as part of the NRC staffs review of this LAR. A bounding estimate of the reactivity impact for each assumption is listed in the table below, based on NRC staff calculations or studies. In addition, any extra margins to the regulatory limit identified during the review of the nuclear criticality safety analysis are listed.

Based on the comparison below, the NRC staff concludes that the available margins offset the potential nonconservatisms.

Estimated Reactivity Impact (L1k)

Nonconservatisms Neolectinq water qaps between spent fuel pool racks -0.0*

Limited evaluation of power history effects 0.002 Neolectinq 1.5 percent bias of unvalidated fission products 0.00015 Tolerance of Boraflex sheathing thickness was not evaluated 0.00026 SCCG and in-rack reactivity differences 0.0005 Total reactivity impact of nonconservatisms 0.00291 Conservatisms Margin to reoulatorv limit 0.0110 Total reactivity impact of conservatisms 0.0110

• This assumption is conservative for fuel lattices used in this analysis. This may not be true for other fuel lattice designs that are undermoderated, but the reactivity impact would not be expected to be significant due to the very small size of the water gap.

The NRC staffs review of the Nine Mile Point 1 SFP Boraflex rack nuclear criticality safety analysis without Boraflex credit, documented in Hl-2178001, identified some nonconservative OFFICIAL US& ONLY PROPRl&T.'\RY INFORMATION

OFFICIAL US!s ONLY PROPRllsTARY INFORMATION items. Those items were evaluated against the margin to the regulatory limit.

Nonconservatisms are minor and easily accommodated by the conservatisms in the calculations.

3.2.6 Spent Fuel Pool Neutron Absorber Monitoring Program Nuclear industry experience has demonstrated that Boraflex material undergoes gamma radiation-induced degradation in the SFP environment. The current Nine Mile Point 1 SFP Boraflex fuel racks have been in place since the early 1980s. During this timeframe, the licensee has used an SFP Boraflex monitoring program to monitor and manage the condition of the Boraflex in the SFP racks. This monitoring program is described in the Nine Mile Point 1 UFSAR, Section C.1.5. In the UFSAR, the licensee states that the monitoring program activities include the inspection of test coupons to detect dimensional changes, the correlation of measured levels of silica in the SFP with analysis using a predictive code to estimate boron loss from the Boraflex panels, and neutron attenuation testing to measure the boron areal density of the test coupons. These activities are performed to ensure the required 5 percent margin to criticality in the SFP is maintained.

3.3 NRC Staff Summary The methodology evaluated by the NRC staff in the Nine Mile Point 1 nuclear criticality safety analyses submitted for review was found to be acceptable. The nuclear criticality safety analyses demonstrated adherence to the applicable regulatory requirements without the need to credit Boraflex.

The NRC staff found that the proposed language reflecting a change necessary to ensure that the maximum fuel assembly reactivity will not exceed a kett of 0.95, at a 95/95 confidence level, if flooded with unborated water, is an acceptable TS change. With respect to the short-lived isotopes, the licensee provided a study that demonstrated that use of a cooling time of zero hours was conservative. Therefore, the licensee's treatment of the short-lived, volatile, and gaseous isotopes is conservative and acceptable. Based on the pedigree of the computer codes and the methodology used by the licensee to address known uncertainties, the computational methods implicit in the codes used for the nuclear criticality safety analyses are acceptable.

The NRC staff examined the core operating parameters and concluded CASM0-5 acceptably predicts these quantities for the present nuclear criticality safety analyses. Based on the NRC staff's review of the validation database and its applicability to the compositions, geometries, and methodologies used in the licensee's nuclear criticality safety analyses, the code validation was found to be acceptable, and all identified biases and uncertainties were propagated appropriately. The NRC staff evaluated other relevant aspects of the SFP rack modeling and found them to be modeled conservatively or to use appropriate parameters with uncertainties appropriately addressed. As a result, the NRC staff finds the SFP rack modeling acceptable for the specific conditions at Nine Mile Point 1.

As a result of evaluating the treatment of manufacturing tolerances, uncertainties, and other potential differences between the idealized SFP rack model used in the nuclear criticality safety analyses and real-world SFP racks, the NRC staff has determined that all factors were appropriately considered in either a bounding manner or explicitly evaluated, and any reactivity increase was appropriately applied as an uncertainty. The NRC staff, therefore, finds the treatment of SFP rack modeling factors to be acceptable.

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OFFICIAL YS&. ONbY PROPRIETARY INFORMATION Based on the design features built into the cell blockers for configuration control and to maintain water flow, and based on the analyses performed to demonstrate structural soundness during normal and accident conditions, the NRC staff finds the use of the cell blockers adequate for blocking the unusable fuel storage location in each 2x2 cell array of the Boraflex SFP storage racks. The selection and modeling of the bounding super lattice used in the nuclear criticality safety analyses was reviewed by the NRC staff and found to be appropriate. While some of the specific assumptions used may not be appropriate for general application, they were verified to be appropriate for the specific applications utilized in the licensee nuclear criticality safety analyses and are, therefore, acceptable.

The NRC staff determined the licensee's evaluation of the uncertainty in the fuel depletion calculations and the subsequent applications uses a reactivity uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest and an acceptable methodology for performing depletion of nuclear fuel. Therefore, the NRC staff finds the licensee's evaluation of burn up uncertainty to be acceptable.

Based on its review, the NRC staff finds all other core operating parameters were established at the values that were shown to maximize the reactivity of the design-basis lattice in the SFP rack configuration and were confirmed to be consistent with the SCCG configuration. This approach is an appropriate modeling simplification because the limiting conditions will typically only exist in a limited part of high-power fuel assemblies, rather than the entire axial length of the fuel.

The final calculations were run using the limiting depletion conditions, so the variation due to depletion conditions was treated implicitly, rather than explicitly, as a separate bias. Since this approach is conservative, it is acceptable.

The NRC staff found the margin evaluation performed by the licensee as part of the criticality analysis had minor deficiencies. Use of the normal condition biases and uncertainties for accident conditions is common practice. The reactivity impact due to the biases and uncertainties may increase somewhat with increasing overall reactivity of the configuration, but in this situation, there is a fairly moderate margin to the regulatory limit (as seen in Section 3.2.5 of this safety evaluation). It is not expected that a potential increase in reactivity due to biases and uncertainties will result in challenge to the available margin. Given that the licensee performed a thorough evaluation of all potential accident conditions and considered all possible reactivity impacts, the NRC staff finds the calculation of the kett for accident conditions to be acceptable.

The licensee indicates that the combination of the peak reactivity methodology and the SCCG methodology automatically qualifies all future lattices within the current fuel design for storage in the SFP, provided their vendor maximum SCCG k;nf is less than or equal to the TS requirement of 1.31. The licensee's present analysis was limited to various legacy fuel designs (7x7, 8x8, and 9x9), as well as the currently used GNF2 (10x10) fuel design. Therefore, the introduction of a future lattice from a fuel design not considered in the licensee's present analysis would require the licensee to redo the criticality analyses to demonstrate that the new fuel design will meet regulatory requirements, given the TS limit. The licensee would be expected to follow the appropriate process to update their licensing basis.

The NRC staff also identified concerns with the nuclear criticality safety modeling and assumed uncertainties, and evaluated the effect these may have to ensure that the TS value of SCCG k;nf

< 1.31, and thus the regulatory requirement of 10 CFR 50.68(b)(4), continues to be met. Upon implementation of this LAR, the NRC staff finds the SFP Boraflex monitoring program for the Boraflex SFP racks will no longer be necessary. The NRC staff concludes that adherence to OFFICIAL YSE ONbY PROPRIETARY INFORMATION

OFFICIAb US& ONbY PROPRll!TA.RY INFORMATION these points should ensure the regulatory requirements of 10 CFR 50.36(c)(4) will be met, as well as meeting the regulatory requirements of 10 CFR 50.68(b)(4), which require, in part, that the in-rack kett be less than or equal to 0.95 when no credit for soluble boron is taken.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New York State official was notified of the proposed issuance of the amendment on October 22, 2018. 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 (83 FR 31184). Accordingly, the amendment meets 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 NRC staff 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) there is reasonable assurance that 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.

Principal Contributor: K. Heller Date: January 11, 2019 OFFICIAb US& ONbY PROPRIETARY INFORMATION

ML18344A174 (Proprietary)

ML18344A452 (Non-Proprietary) *by memorandum **by e-mail OFFICE DORL/LPL 1/PM DORL/LPL 1/LA DSS/SNPB/BC* DMLR/MCCB/BC NAME THood LRonewicz RLukes SBloom DATE 12/10/2018 12/12/2018 11/09/2018 12/08/2018 OFFICE OGC-NLO* DORL/LPL 1/BC DORL/LPL 1/PM NAME KGamin JDanna MMarshall DATE 1/02/2019 1/10/2019 1/11/2019