ML18204A286

Issuance of Amendment No. 167 Regarding (CAC MF9996; EPID 2017-LLA-0303)

ML18204A286
Person / Time
Site:	Harris
Issue date:	10/22/2018
From:	Martha Barillas Plant Licensing Branch II
To:	Hamilton T Duke Energy Progress
Barillas M, DORL/LPL2-2, 301-415-2760
References
CAC MF9996, EPID 2017-LLA-0303
Download: ML18204A286 (42)

Text

Ms. Tanya M. Hamilton Site Vice President UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 22, 2018 Shearon Harris Nuclear Power Plant Duke Energy Progress, LLC 5413 Shearon Harris Road M/C HNP01 New Hill, NC 27562-0165

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 167 REGARDING (CAC MF9996; EPID 2017-LLA-0303)

Dear Ms. Hamilton:

The U.S. Nuclear Regulatory Commission (NRC) has issued Amendment No. 167 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). This amendment changes your spent fuel storage pool Technical Specifications (TSs) in response to your application dated June 28, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML171936165), as supplemented by letters dated July 20, 2017, September 14, 2017, January 18, 2018, February 16, 2018, and April 13, 2018 (ADAMS Accession Nos. ML17201A035, ML17257A245, ML18018B974, ML18047A730, and ML18108A106}.

The amendment revises HNP TSs for fuel storage criticality to account for the use of Metamic neutron absorbing rack inserts and soluble boron for the purpose of criticality control in the boiling-water reactor storage racks of HNP's spent fuel Pools A and B that currently credit Boraflex. This license amendment request was submitted to address an operable but degraded condition in HNP's A and B spent fuel pools boiling-water reactor storage racks due to Boraflex degradation.

The NRC staff has completed its review of the information provided by the licensee. The NRC staff's safety evaluation (SE) is enclosed. The NRC staff has determined that the enclosed SE (Enclosure 2) does not contain proprietary information or other sensitive information pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 2.390, "Public inspections, exemptions, requests for withholding." However, the NRC will delay placing the enclosed SE in the public document room for a period of 10-working days from the date of this letter to provide Duke Energy the opportunity to comment on any sensitive aspects of the SE. If you believe that any information in Enclosure 2 contains sensitive information, please identify such information line-by-line and define the basis for withholding pursuant to the criteria of 10 CFR 2.390. After 10-working days, the enclosed SE will be made publicly available, unless we hear from you.

The Notice of Issuance will be included in the Commission's regular biweekly Federal Register notice.

If you have any questions concerning this letter, please contact me at 301-415-2760 or by email at Martha.Barillas@nrc.gov.

Docket No. 50-400

Enclosures:

1. Amendment No. 167 to NPF-63

2. Safety Evaluation cc: Listserv Sincerely, Martha Barillas, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 167 Renewed License No. NPF-63

1.

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

A.

The application for amendment by Duke Energy Progress, LLC (the licensee),

dated June 28, 2017, as supplemented by letters dated July 20, 2017, September 14, 2017, January 18, 2018, February 16, 2018, and April 13, 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. NPF-63 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 167, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

Attachment:

Changes to the Renewed License and Technical Specifications FOR THE NUCLEAR REGULATORY COMMISSION

'v~l)v\\,0~(<'f Undine Shoop, Chief Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: October 22, 2018

ATTACHMENT TO LICENSE AMENDMENT NO. 167 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DOCKET NO. 50-400 Replace the following page of the renewed facility operating license with the revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change:

Remove Page 4 Insert Page 4 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove 5-7a Insert 5-7a 5-7e 5-?f C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 1 Q CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.

(1)

Maximum Power Level Duke Energy Progress, LLC, is authorized to operate the facility at reactor Core power levels not in excess of 2948 megawatts thermal ( 100 percent rated core power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 167, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Antitrust Conditions (4)

(5)

Duke Energy Progress, LLC. shall comply with the antitrust conditions delineated in Appendix C to this license.

Initial Startup Test Program (Section 14)1 Any changes to the Initial Test Program described in Section 14 of the FSAR made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.*

Steam Generator Tube Rupture (Section 15.6.3)

Prior to startup following the first refueling outage, Carolina Power & Light Company* shall submit for NRC review and receive approval if a steam generator tube rupture analysis, including the assumed operator actions, which demonstrates that the consequences of the design basis steam generator tube rupture event for the Shearon Harris Nuclear Power Plant are less than the acceptance criteria specified in the Standard Review Plan, NUREG-0800, at 15.6.3 Subparts II (1) and (2) for calculated doses from radiological releases. In preparing their analysis Carolina Power &

Light Company* will not assume that operators will complete corrective actions within the first thirty minutes after a steam generator tube rupture.

Renewed License No. NPF-63 Amendment No. 167

DESIGN FEATURES

3. BWR Storage Racks in Pools "A" and "B"

a. Racks with Metamic neutron absorber inserts

1. kett less than or equal to 0.95 when flooded with water borated to 1000 ppm.

2. kett less than 1.0 when flooded with unborated water.

3. The reactivity margin is assured for BWR racks in pools "A" and "B" by maintaining a nominal 6.25 inch center-to-center distance in the BWR storage racks.

4. The following restrictions are also imposed through administrative controls:

a. Storage of BWR fuel designs limited to GE3, GE4, GE5, GE6, and GE7 fuel designs.

b. Rack insert orientation is limited to that shown in Figure 5.6-3 and Figure 5.6-4.

c. No fuel shall be stored in Storage Location A 11 of Rack C1 in Spent Fuel Pool A.

b. Racks with Baral neutron absorber

1. kett less than or equal to 0.95 when flooded with unborated water.

2. The reactivity margin is assured for BWR racks in pools "A" and "B" by maintaining a nominal 6.25 inch center-to-center distance in the BWR storage racks.

4. PWR and BWR racks in pools "C" and "D"

a. kett less than or equal to 0.95 when flooded with unborated water.

b. The reactivity margin is assured for pools "C" and "D" by maintaining a nominal 9.017 inch center-to-center distance between fuel assemblies placed in the non-flux trap style PWR storage racks and 6.25 inch center-to-center distance in the BWR storage racks.

c. The following restrictions are also imposed through administrative controls:

1. PWR assemblies must be within the "acceptable range" of the burnup restrictions shown in Figure 5.6-1 prior to storage in pools "C" and "D".

2. BWR assemblies are acceptable for storage in pool "C" provided the maximum planar average enrichments are less than 4.6 wt.% U235 and Kintis less than or equal to 1.32 for the standard cold core geometry (SCCG).

5. In each case, kett includes allowances for uncertainties as described in Section 4.3.2.6 of the FSAR.

DRAINAGE 5.6.2 The pools "A", "B", "C" and "D" are designed and shall be maintained to prevent inadvertent draining of the pools below elevation 277.

CAPACITY 5.6.3.a Pool "A" contains six (6 x 10 cell) flux trap type PWR racks and three (11 x 11 cell) BWR racks for a total storage capacity of 723 assemblies. Pool "B" contains six (7 x 10 cell), five (6 x 10 cell), and one (6 x 8 cell) flux trap style PWR racks and seventeen (11 x 11 cell) BWR racks and is licensed for one additional (11 x 11 cell) BWR rack that will be installed as needed. The combined pool "A" and "B" licensed storage capacity is 3669 assemblies.

SHEARON HARRIS - UNIT 1 5-7a Amendment No. 16 7

DESIGN FEATURES PWRRacks BWRRack BWRRack BWRRack PWR Racks With Inserts With Inserts With Inserts Plant North I Insert Orientation...,

FIGURE 5.6-3 POOL A METAMI C RACK INSERT ORIENTATION SHEARON HARRIS - UNIT 1 5-7e Amendment No. 16 7

DESIGN FEATURES BWRRack BWRRack With Inserts With Inserts BWRRack BWRRack With Inserts With Inserts PWRRacks BWR Rack With Inserts BWRRack With Boral Plan t North Insert Orien t atio n FIGURE 5.6-4 BWRRacks With Boral PO OL B ME TAMIC RACK INSERT ORIE NTATIO N SHEARON HARRIS - UNIT 1 5-7f BWR Racks With Boral Amendment No. 1 6 7

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 167 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DUKE ENERGY PROGRESS, LLC SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-400

1.0 INTRODUCTION

By license amendment request (LAR) dated June 28, 2017, as supplemented by letters dated July 20, 2017, September 14, 2017, January 18, 2018, February 16, 2018, and April 13, 2018 (References 1 through 6), Duke Energy Progress, LLC (the licensee) requested to amend Renewed Facility Operating License No. NPF-63 to revise the Shearon Harris Nuclear Power Plant (HNP), Unit 1 Technical Specifications (TSs). Specifically, the proposed amendment will modify the TSs for fuel storage criticality to account for the use of Holtec's Device for Reactivity Mitigation (DREAM) Metamic neutron absorbing spent fuel pool (SFP) rack inserts and soluble boron for the purpose of criticality control in the boiling-water reactor (BWR) storage racks that currently credit Boraflex in SFPs A and B. Upon implementation of this license amendment, the licensee will no longer credit the Boraflex for criticality control in the A and B SFPs. Instead, the licensee will credit DREAM Metamic inserts and soluble boron for criticality control. The HNP SFP storage capacity consists of four pools, A, B, C, and D. The Boraflex BWR fuel storage racks that are affected by the proposed license amendment are only found in Pools A and B.

The license amendment is being requested because Boraflex is known to have degradation issues and is required to resolve a current operable but degraded condition in the SFPs A and B.

The supplements dated July 20, 2017, September 14, 2017, January 18, 2018, February 16, 2018, and April 13, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC or the Commission) staff's initial proposed no significant hazards consideration determination as published in the Federal Register on December 5, 2017 (82 FR 57481 ).

2.0 REGULATORY EVALUATION

The design of HNP incorporates the use of three SFPs, one new fuel pool, and a cask-loading pool, all of which are connected by a fuel transfer canal system. Spent fuel storage is provided by the New Fuel Storage Pool (Pool A) and the three SFPs commonly referred to as Pools B, C, and D. The BWR racks are used for storage of spent fuel assemblies from the Brunswick Steam Electric Plant, as authorized by HNP's Operating License. The SFPs at HNP contain both pressurized-water reactor (PWR) fuel racks and BWR fuel racks. Both the PWR and BWR racks use a neutron absorber material (NAM) in the rack design for reactivity control. Two types of neutron absorbers, Boral and Boraflex, are used in the BWR fuel racks. This proposed license amendment addresses the Westinghouse-designed rack for spent BWR fuel in use in HNP Pools A and B, which utilizes Boraflex as the neutron absorber. This LAR would allow the licensee to credit the Metamic NAM for criticality control in SFPs A and B. Metamic is a NAM that is manufactured by Holtec International. Metamic is the principal material for the proposed DREAM inserts to be installed at HNP SFPs A and B. The use of Metamic for criticality control in SFPs has been previously approved by the NRC staff.

2.1 Regulatory Requirements Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "Domestic Licensing of Production and Utilization Facilities. Section 50.36, "Technical specifications," establishes the regulatory requirements related to the content of TSs.

Section 50.36(c)(4) to 10 CFR Part 50 requires that design features be included that 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 this section.

Section 50.68 of 10 CFR addresses, "Criticality accident requirements." The licensee has elected to demonstrate compliance with the requirements in 10 CFR 50.68(b). Section 50.68(b) of 10 CFR requires each licensee shall comply with the requirements of 10 CFR 50.68(b)(1 )-(8) in lieu of maintaining a monitoring system capable of detecting a criticality as described in 10 CFR 70.24. Section 50.68(b)(1) of 10 CFR requires that 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.

The new HNP SFP nuclear criticality safety (NCS) analysis for the Boraflex BWR racks will take credit for soluble boron. Section 50.68(b)(4) of 10 CFR requires, in part, if credit is taken for soluble boron, the k-effective (kett) of the spent fuel storage racks (SFSRs) loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95-percent probability, 95-percent confidence level, if flooded with borated water, and the kett must remain below 1.0 (subcritical) at a 95-percent probability, 95-percent confidence level, if flooded with unborated water (Note: the kett is defined as the effective neutron multiplication factor).

Per Section 3.1, "Conformance with NRC General Design Criteria [GDC]," of the HNP FSAR, the plant is designed to meet Appendix A to 10 CFR 50 effective May 21, 1971, and subsequently amended July 7, 1971, and October 23, 1978. Accordingly, those criteria were used to evaluate this LAR. Specific criteria used are listed below.

GDC 1, "Quality standards and records," requires, in part, that structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

GDC 2, "Design bases for protection against natural phenomena," requires, in part, that structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions.

GDC 4, "Environmental and dynamic effects design bases," requires that structures, systems, and components important to safety be designed to accommodate the effects of, and be compatible with, environmental conditions associated with normal operation, maintenance, testing, and postulated accidents.

GDC 61, "Fuel storage and handling and radioactivity control," requires, in part, that fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and accident conditions, with a capability to permit appropriate periodic inspection and testing of components important to safety, and to prevent significant reduction in fuel storage coolant inventory under accident conditions.

GDC 62, "Prevention of criticality in fuel storage and handling," 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.

2.2 Regulatory Guidance NUREG-0800, "Standard Review Plan [SRP]," Section 3.8.4, "Other Seismic Category Structures," including Appendix D, "Guidance on Spent Fuel Pool Racks," Revision 4 (Reference 7). This NUREG section provides guidance for applicable loading combinations and stress limits considered in the seismic analysis of the rack structure for service limits based on American Society of Mechanical Engineers (ASME) Code, Section Ill, Subsection NF.

NUREG-0800, "Standard Review Plan," Section 9.1.1, "Criticality Safety of Fresh and Spent Fuel Storage and Handling" (Reference 8). This NUREG section provides guidance regarding the specific acceptance criteria and review procedures to ensure that the proposed changes satisfy the requirements in 10 CFR 50.68 and GDC 62.

NUREG-0800, "Standard Review Plan," Section 9.1.2, "New and Spent Fuel Storage" (Reference 9). This NUREG section provides guidance regarding the specific acceptance criteria and review procedures to ensure that the proposed changes satisfy the requirements in 10 CFR 50.68.

NUREG-0800, "Standard Review Plan," Section 9.1.3, "Spent Fuel Pool Cooling and Cleanup System" (Reference 10). This NU REG section provides guidance on reviewing structures, systems, and components performing their intended safety functions in the environmental temperature conditions during normal operations as it relates to GDC 4.

Nuclear Energy Institute (NEI) Topical Report (TR) NEI 2016-03-A, "Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools," Revision O (Reference 11 ), provides guidance developed by NEI regarding appropriate monitoring programs for neutron-absorbing materials.

The NRC verified by letter on October 5, 2017 (Reference 12), that the approved version referenced above is acceptable for use in future licensing actions. As such, the guidance contained therein constitutes an acceptable approach to satisfy the current NRC staff position on appropriate features associated with an acceptable monitoring program for neutron-absorbing materials.

The NRC staff issued a memorandum dated August 19, 1998 (Reference 13), also known as the "Kopp Memo," containing staff guidance for performing the review of SFP NCS analyses.

This guidance supports determining compliance with GDC 62 and the existing guidance in SRP Sections 9.1.1 and 9.1.2. The principal objective of this guidance was to clarify and document staff positions that may have been incompletely or ambiguously stated in previously issued safety evaluations (SEs) and other staff documents. A second purpose was to state staff positions on a number of strategies used in SFP NCS analyses at that time.

NRC Interim Staff Guidance (ISG) DSS-ISG-[Division of Safety System Interim Staff Guidance]

2010-01, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools," Revision 0, was issued in October 2011 (Reference 14). The purpose of this ISG is to provide updated guidance to the NRC staff reviewer to address the increased complexity of recent SFP nuclear criticality analyses and operations. This ISG is intended to reiterate existing guidance, clarify ambiguity in existing guidance, and identify lessons learned based on past submittals. Similar to the Kopp Memo, this guidance supports determining compliance with GDC 62 and the existing guidance in SRP Sections 9.1.1 and 9.1.2. The NRC staff notes that while Section 9.1.2 of NUREG-0800 is applicable, it is not concerned directly with fuel storage criticality safety considerations. Therefore, Section 9.1.1 contains the primary SRP guidance for reviewing the proposed changes in the LAR as it relates to fuel storage criticality safety considerations.

NRC Office of Technology (OT) Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications, dated April 14, 1978 (Reference 15). The NRC's OT position paper states that conservative methods should be used to calculate the maximum fuel temperature and the increase in temperature of the water in the pool. The maximum void fraction in the fuel assembly and between fuel assemblies should also be calculated. The bases for the analyses should include the established cooling times for both the usual refueling case and the full core off load case. The OT position paper also states that the average amount of water in the fuel pool and the expected heat up rate of this water assuming loss of all cooling systems shall be specified.

2.3 Compliance with the Technical Specifications The licensee is proposing changes to TS 5.6.1.3 and insertion of two corresponding figures, Figures 5.6-3 and 5.6-4.

Currently, TS 5.6.1 contains several groups of design features that are applicable to specific storage racks installed in the four HNP SFPs. In particular, TS 5.6.1.3 contains two design features associated with all of the storage racks in Pools A and B that are designed to store BWR fuel, which currently includes both the Boral BWR racks and the Boraflex BWR racks.

The first design feature is that the kett is less than or equal to 0.95 when flooded with unborated water. The second design feature is a specified nominal SFP storage cell pitch consistent with the fuel assembly separation distance assumed in the SFP NCS analyses for the SFP storage racks.

The proposed change will split TS 5.6.1.3 into two groups of design features. TS 5.6.1.3.b will apply only to the Boral BWR racks, and will contain identical text to the current TS 5.6.1.3.

TS 5.6.1.3.a will apply only to the Boraflex BWR racks and will credit Metamic neutron absorber inserts upon approval of this amendment. In order to accommodate the specific design features associated with the new SFP NCS analysis documented in Attachment 4 of Reference 6, the design features are modified to be consistent with the 10 CFR 50.68(b)(4) requirement for subcriticality when credit is taken for soluble boron in the SFP. An additional group of restrictions is also captured associated with allowed SFP storage configurations, namely, limitations on: (1) allowable fuel designs for storage, (2) rack insert orientation, and (3) prohibition of fuel in a single SFP storage cell. Figures 5.6-3 and 5.6-4 are also added to support the rack insert orientation restrictions by graphically depicting the allowed rack insert orientation.

3.0 TECHNICAL EVALUATION

3.1 Background

The Reference 6, Attachment 4, Section 4.0 presents the NCS analysis for the HNP Boraflex BWR SFSRs. The report describes the methodology and analytical models used in the NCS analysis to show that the Boraflex BWR SFSRs maximum kett will be no greater than 0.95 when soluble boron in the SFP water is credited, and no greater than 1.0 if flooded with unborated water. This attachment also references a benchmarking evaluation performed for the MCNP5 Version 1.51 code used for the NCS analysis, to demonstrate the applicability of the code to geometries and compositions being analyzed and to determine the code bias and uncertainty.

MCNP5 is a Monte Carlo N-.E.article criticality code developed and maintained by the Los Alamos National Laboratory for use in performing reactor physics and criticality safety analyses for nuclear facilities and transportation/storage packages.

The licensee is installing a maximum of 968 Metamic rack inserts into the Boraflex BWR racks within the next year, which is adequate to qualify all cells for storage. In addition to the Boraflex BWR racks, Pools A and B also contain BWR fuel storage racks crediting Boral (herein referred to as "Boral BWR rack") and PWR fuel storage racks that are used to credit Boraflex, but no longer do so (herein referred to as "PWR racks"). Therefore, while the Boral BWR racks and PWR racks are already adequately addressed within the existing licensing basis for HNP, this license amendment does address the effects resulting from the presence of these racks adjacent to the Boral BWR racks (i.e., the "interface effect").

In order to support this TS change, the licensee submitted a new nuclear criticality safety (NCS) methodology and analysis.

HNP has four SFPs that are currently licensed for storage of different fuel designs in different storage rack modules that credit various combinations of Boraflex, Boral, and Metamic neutron-absorbing materials, as well as some which credit no neutron-absorbing materials. This LAR only addresses the Boraflex BWR racks. Boraflex consists of boron carbide suspended in a silicone polymer that has a long history of significant degradation due to dissolution and subsequent loss of B-10. HNP has taken previous actions to address Boraflex degradation in their SFPs, as discussed in their responses to Generic Letters 1996-04 and 2016-01. The SFP NCS analyses supporting fuel storage in the PWR racks containing Boraflex have already been updated to remove credit for the Boraflex. The SFP NCS analysis submitted as Attachment 4 of Reference 6 will remove credit for the Boraflex as a criticality control component for the BWR racks; instead, rack inserts made of Metamic will be used.

Metamic is a metal matrix made from aluminum and boron carbide, and the rack inserts are fashioned by bending Metamic panels along the long edge to create an L-shaped insert for the SFP storage cells. The resulting Metamic rack inserts are inserted in every cell of the Boraflex BWR racks where fuel storage is intended. The rack insert orientation is pre-defined, so all rack inserts are oriented in the same manner in every SFP storage cell. As a result, with the exception of peripheral walls that face the concrete SFP wall, there will be a single panel of Metamic material located on either side of every cell wall in all Boraflex BWR storage locations containing fuel. The licensee states that the nature of the design of the upper aluminum block attached to the Metamic panels for handling purposes means that a fuel assembly cannot be removed from the Boraflex BWR racks without first removing the Metamic rack insert. Site fuel-handling procedures will include administrative controls prohibiting removal of more than one Metamic rack insert. The orientation of any inserts placed back into the SFP would be restricted to that allowed in TS Figures 5.6-3 and 5.6-4.

The Boraflex BWR racks under concern are located in the HNP Pools A and B. In addition to the Boraflex BWR racks, the HNP SFPs also contain Boral BWR racks and PWR racks. The licensing basis for these storage racks will continue to be based on previously approved SFP NCS analyses, but the new SFP NCS analysis must address the presence of these racks under all normal and credible accident conditions. The racks in HNP Pools C and D are not explicitly included in the scope of this LAR because they are neutronically decoupled from Pools A and B, and do not contain any Boraflex BWR racks. However, fuel assemblies can be transferred between SFPs, so the new SFP NCS analysis must consider fuel assemblies stored in the other SFPs when determining candidates for the limiting fuel assembly mislead accident.

3.2 SFP NCS Analysis Method The methods used for the NCS analysis for fuel in the HNP SFP are described in Section 4.0 of of Reference 6. A previous version of the report documenting the computer code benchmarking analyses supporting use of MCNP5 Version 1.51 for this application was reviewed and approved by the NRC (Reference 16). The area of applicability for the benchmarking analyses was verified to be applicable to the HNP SFP. Additional information describing the methods used is provided in the Request for Additional Information (RAI) responses attached to Reference 4. Some potential nonconservatisms were identified during the review, but as will be discussed below, sufficient margin is built into the analysis methodology to offset the potential reactivity impacts for the fuel stored in the HNP SFP.

3.2.1 Computational Methods For the criticality calculation, the licensee used MCNP5 Version 1.51, with continuous energy cross-section data based on the Evaluated Nuclear Data File, Version 7 (ENDF/B-VII) neutron cross section library. The use of MCNP5 Version 1.51 has previously been approved for use in SFP NCS analyses for licensing purposes (Reference 16).

For all MCNP5 Version 1.51 calculations, the licensee used values within an acceptable range for the following calculational parameters: number of histories per cycle, number of cycles skipped before averaging, and total number of cycles. The initial source distribution was only specified as being "in the fueled regions (assemblies)," however, given the relative uniformity of the base models and number of skipped cycles, the exact source distribution is relatively unimportant as long as the initial neutrons are seeded within the active fuel. More importantly, the licensee confirmed that all calculations converged using appropriate checks.

Based on the pedigree of the computer codes and the use of appropriate inputs and verification methods to ensure that an accurate eigenvalue was determined, the NRC staff finds the computational methods implicit in the codes used for the NCS analyses are acceptable.

3.2.2 Computer Code Validation Since the NCS 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 utilize 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 ISG DSS-ISG-2010-01 references NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology" (Reference 17).

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 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 provided 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.

In Version 1 of the criticality benchmarking report, which was previously reviewed by the NRC (Reference 16), MCNP5 Version 1.51 is validated by comparing calculated keff values with the measured keff for several different sets of critical configurations as determined through experiments. A total of 532 critical configurations were included. The licensee determined that it would be appropriate to treat all data as a single set, but 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 data from subsets of the critical experiments that exhibited specific storage characteristics.

The licensee referenced Revision 3 of the criticality benchmarking report, but did not include the report in their submittal. The validation report was not re-evaluated in its entirety as part of the review of this LAR. However, the NRC staff requested Revision 3 of the validation report in order to review the changes relative to Revision 1, which was previously reviewed by the NRC.

In Reference 4, the licensee provided a summary of changes between Revisions 2 and 3 of the validation report. The NRC staff inferred, based on the discussion in Attachment 4 of Reference 6, that 30 new critical experiments were added to the full set as part of the Revision 2 update, however, no information was provided to assess the suitability of the additional critical experiments. The bias and uncertainty values provided for MCNP5-Version 1.51 with the ENDF/B-VII in Attachment 4 of Reference 6 (which are obtained from Revision 3 of the validation report) are slightly higher (more conservative) than the values in Revision 1 of the validation report (reviewed and approved in Reference 6), and the impact of increasing the number of critical benchmarks by about 6 percent is expected to be small unless a significant gap is addressed (which was not identified in the Reference 6 SE). Therefore, engineering judgement was used to accept the conclusions from Revision 3 of the validation report for this LAR, even though the staff did not explicitly review the 30 critical benchmarks that were added as part of the Revision 2 update to the validation report. This conclusion is supported by the fact that the prior NRC staff finding on Revision 1 to the validation report would be applicable to the HNP SFP, and the bias and uncertainty values used in the HNP analyses are slightly more conservative than the values recommended in Revision 1 of the validation report.

The licensee demonstrated how the validation report's findings were applied for this specific analysis. The licensee identified the applicable operating conditions for the validation (e.g., fuel assembly materials and geometry, enrichment of fissile isotope, fuel density, types of neutron absorbers, moderators and reflectors, rack material, and physical configurations). The licensee compared the spectral parameters (e.g., EALF [energy of average lethargy causing fission],

spectrum type) between the benchmarks and the HNP SFP conditions to demonstrate that the selected benchmarks are applicable.

The NRC staff noted that the validation data set contains experiments that include plutonium as a fissile material. The licensee performed their criticality analyses using fresh, unpoisoned fuel, so the data from these experiments are not applicable. However, the selection of the limiting bias and bias uncertainty included consideration of the subsets within the data set that only contain uranium as a fissile material. Therefore, the final values are verified to bind the design-basis fuel assembly utilized by the licensee in the SFP NCS analyses documented in of Reference 6.

Based on the staff's review of the validation database and its applicability to the compositions, geometries, and methodologies used in the licensee's NCS analyses, the code validation was found to be acceptable and all identified biases and uncertainties were propagated appropriately.

3.3 SFP and Fuel Storage Racks 3.3.1 SFP Water Temperature The SFP water temperature was treated in a bounding manner. The design basis calculations were run using the minimum SFP temperature, and follow-up calculations were performed to verify that the maximum SFP temperature did not result in a higher kett value. In order to obtain appropriate cross sections for some of the intermediate temperatures, the NJOY 99.396 code was used to adjust the standard cross sections to a specified temperature and to apply a molecular energy adjustment for certain materials. Other analyses approved by the NRC (Reference 16) use a similar methodology to account for the fact that the standard MCNP5 cross section libraries only contain data at specific temperatures for each nuclide. In these previous analyses, the TMP and S(a,r3) cards were used in MCNP5 to make the adjustments.

However, the capabilities built into MCNP5 are relatively limited, so the licensee elected to use NJOY to perform the adjustments. NJOY is the same cross section processing code used to create the standard cross section libraries used by MCNP5, with a comparable pedigree, so this approach is an acceptable way to ensure that the cross sections used in the evaluation are reasonably accurate for the given temperatures. Since cross sections are used for the minimum SFP temperature MCNP5 calculations without the need to use NJOY, the NRC staff finds the elevated temperature calculations using cross sections generated with NJOY to be acceptable to confirm that a different temperature will not be more limiting. The SFP water temperature was treated in a bounding manner, so the NRC staff finds the licensee's approach to be acceptable.

3.3.2 SFP Storage Rack Models HNP has multiple Boraflex BWR racks clustered in one area of each SFP, with water gaps between adjacent racks. In the criticality analysis, the licensee chose to use a bounding approach in which a fresh, unpoisoned lattice is identified as the design basis lattice. Further discussion about this lattice and how it was determined to bound all fuel stored in the Boraflex BWR racks can be found in Section 3.3.3 of this SE. The SFP is then assumed to be fully loaded with fuel assemblies that have this lattice along its entire axial length. Most of the calculations are performed utilizing an 11 x11 storage rack model with periodic boundary conditions. In effect, the base model is that of an infinite SFP with 11 x11 storage racks separated by the distance between adjacent Boraflex BWR racks in the HNP SFP. The licensee performed sensitivity studies to determine bounding parameters based on the manufacturing tolerances for the relevant SFP storage rack modeling parameters: storage cell inner diameter, storage rack wall thickness, and storage cell pitch. Twelve inches of water is modeled above and below the modeled active fuel, which is sufficiently large relative to the mean free path of neutrons in water to ensure that any further neutron reflection effects from modeling additional water are minimal. The Metamic rack inserts are included in the model in an orientation consistent with the proposed TS change. Some modeling simplifications are discussed in Section 4.2.3.6.1 of Attachment 4 of Reference 6, but they all have a conservative or negligible impact on reactivity.

The licensee performed a series of sensitivity studies to evaluate various combinations of the fuel assembly position within the SFP cell, the Metamic rack insert position between the fuel assembly and SFP wall, and whether the fuel assembly is channeled or not. The combination of these parameters that yielded the highest reactivity was used in the design basis calculations.

The fuel assembly model used in the sensitivity studies was not the same as the design basis calculations, since it used a lower uranium-235 (U-235) enrichment and nominal fuel design parameters. However, these inputs would not be expected to change the direction of the reactivity variation, only the magnitude. Therefore, this approach to determine the bounding positioning parameters for the fuel assembly and Metamic rack insert is acceptable.

In order to analyze situations involving fuel in locations other than the cells in the Boraflex BWR racks, full pool models are used. These models are used to address the criticality impacts due to seismic events (where the racks may move closer together), interface conditions (where a Boraflex BWR rack is adjacent to one of the other rack types in the SFP) and the mislocated fuel assembly event (where a fuel assembly is accidentally placed or dropped into a location outside the SFP racks). Section 4.5.2 of Attachment 4 of Reference 6 briefly describes the modeling inputs used for the other storage rack designs. In general, the geometries and compositions are modeled using nominal specifications in Table 4.5.2. Some of the structural material, such as the baseplate, is neglected. This is acceptable because stainless steel is a weaker neutron reflector than water, so modeling pure water below the active fuel length is conservative. In the full pool model, the storage rack modeling extends to the full height above the active fuel, in order to account for the fact that the fuel assemblies in different racks are of different lengths. Due to modeling challenges, the gaps between storage racks in the model does not necessarily match the installed configuration, but they do bound (i.e., are smaller than) the installed configuration. This combination of nominal and conservative modeling inputs for the other rack designs is acceptable.

The Boraflex poison in the Boraflex BWR racks is modeled as water in all cases. This is conservative because any remaining Boraflex, while not credited, would result in a reduction in reactivity relative to pure water. Metamic rack inserts are modeled in all but one cell in each Boraflex BWR rack based on design specifications. Similarly to the SFP storage rack parameters, bounding values were determined based on the manufacturing tolerances for the relevant Metamic rack insert parameters: boron-10 loading, insert thickness, and insert width.

The assumption of a single missing Metamic rack insert in each Boraflex BWR rack is based on the fact that for fuel to be moved in or out of a storage cell, the Metamic rack insert must first be removed. Therefore, normal fuel movement would include a condition in which one Metamic rack insert is removed.

In its LAR, the licensee did not describe what configuration controls would be in place to ensure that no more than one Metamic rack insert will be removed at any given time. In response to a RAI from the NRC staff (Reference 4), the licensee stated that the site fuel-handling procedures would contain administrative controls to require that no more than one Metamic rack insert be removed at a time. The NRC staff finds this to be an adequate safeguard addressing a missing insert. Further, because a tool change would be required to change between Metamic rack insert manipulation and fuel assembly movement, the chance of an error is minimal. Therefore, a distinct intermediate step would be required to be marked as complete between movements of successive fuel assemblies or rack inserts within the Boraflex BWR racks. As a result, the NRC staff agrees that having more than one Metamic rack insert removed from the HNP SFP is not a credible scenario. The NRC staff has evaluated the relevant aspects of the storage rack modeling and found them to be modeled conservatively or using appropriate parameters. As a result, the storage rack modeling is acceptable.

The licensee has an acceptable monitoring program, as described in Section 3.4 of this SE.

Therefore, reasonable assurance exists that any significant deviation from the parameters as modeled in the criticality analysis will be identified and evaluated by the licensee.

3.3.2.1 SFP Storage Rack Models Manufacturing Tolerances and Uncertainties The manufacturing tolerances of the storage racks contribute to SFP reactivity.

DSS-ISG-2010-01 (Reference 14) does not explicitly discuss the approach to be used in determining manufacturing tolerances, but past practice has been consistent with the Kopp Memo (referenced in DSS-ISG-2010-01) that determination of the maximum ket1 should consider either: (1) a worst-case combination with mechanical and material conditions set to maximize ket1, 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 racks. The licensee chose to utilize the former approach with all the as-built parameters of the Boraflex BWR racks and Metamic rack inserts that would result in a significant reactivity impact. In the case of the cell pitch, the licensee only analyzed the tolerance in the direction that typically results in the most conservative reactivity impact, in order to verify the appropriate extreme of the manufacturing tolerance range to use. In other cases, both extremes for the manufacturing tolerance range were evaluated to determine which value would lead to the maximum reactivity. The resulting combination of worst-case parameters for reactivity were used for all subsequent criticality calculations.

The licensee did not address the uncertainties associated with the other storage rack types in the HNP SFP. However, the Boraflex BWR racks and Metamic rack inserts are modeled using bounding parameters. This approach means that the uncertainties associated with the Boraflex BWR racks and Metamic rack inserts are being treated as biases rather than uncertainties, which is conservative. In order for the uncertainties associated with other SFP storage rack types to have any effect on the results from this analysis, the statistical combination of uncertainties would have to be larger than the bias inherent in this approach for modeling the Boraflex BWR rack and Metamic rack inserts. The NRC staff has determined that this is unlikely to be the case, based on typical values for total storage rack uncertainties and the difference in reactivity between nominal and bounding parameters documented in Attachment 4 of Reference 6.

As a result of evaluating the licensee's treatment of the manufacturing tolerances, the NRC staff has determined that the licensee treated all relevant Boraflex BWR rack and Metamic rack insert parameters in a bounding (i.e., conservative) manner. Other storage rack types are present in the SFP and are included in some of the analyses, but the NRC staff concluded that any potentially larger reactivity impacts from the uncertainties associated with other storage rack types were unlikely to be sufficiently large to overcome the inherent conservatism in the treatment of the Boraflex BWR rack and Metamic rack insert uncertainties.

3.3.2.2 SFP Storage Rack Interfaces There are two other SFP storage rack types co-residing in the HNP SFPs with the Boraflex BWR racks; PWR racks and a different BWR rack type that utilizes Boral as a neutron-absorbing material (i.e., Boral BWR racks). Each storage rack type is qualified using infinite arrays, therefore, the possible interfaces between storage rack types need to be evaluated. This evaluation should consider the following: the proximity of higher reactivity fuel in adjacent racks, any potential increase in neutron current between adjacent racks compared to adjacent cells within the infinite array, and any changes in initial accident conditions that may result in more severe reactivity impacts. Since this LAR only seeks to establish new SFP storage requirements for the Boraflex BWR racks, the licensee did not need to address the interface between the Boral BWR racks and the PWR racks.

In Attachment 4 of Reference 6, the licensee describes the models developed to evaluate the interface between SFP storage racks. First, the licensee performed calculations using infinite arrays of each storage rack type. The PWR racks can store fuel in one of two configurations with a maximum enrichment of 5.0 weight percent (wt%) U-235: an unlimited storage configuration provided that certain burnup limits are met, or a checkerboard of empty cells and fresh fuel assemblies. The Boral BWR racks and Boraflex BWR racks used a uniform loading of the design-basis fuel assembly discussed in Section 3.3.3 of this SE. Based on these calculations, the fuel storage arrangements inherent in the Boral BWR racks and the PWR racks loaded with the checkerboard arrangement were determined to be less reactive than the Boraflex BWR racks.

However, the licensee did not discuss local variations in the installed neutron absorbing material that result when these fuel storage arrangements are stored adjacent to the Boraflex BWR racks. Section 4.5.2.3 of Attachment 4 of Reference 6 explains that the Boral BWR racks have a Baral panel fixed along the exterior, and the neutron absorber areal density for the Baral panels is given in Table 4.5.2 as a minimum of 0.015 g/cm2* This minimum value is 15-percent higher than the minimum areal density assumed for the Metamic rack inserts in the Boraflex BWR racks, so reasonable assurance exists that the neutron absorption effectiveness of the Baral BWR racks is bounded by the Boraflex BWR racks. The PWR racks include no modeled neutron absorbing material, so the neutron absorption credited for the PWR racks is not bounded by the Boraflex BWR racks. The licensee does perform further analysis for an interface between the Boraflex BWR racks and the PWR racks loaded with a uniform storage configuration of burned fuel. The checkerboard arrangement is less reactive than the uniform arrangement, so the interface between the Boraflex BWR racks and the former arrangement in PWR racks is bounded by the evaluation of the same interface with a uniform arrangement.

The postulated accident conditions were considered in light of the possible accident conditions that may exist. The Baral BWR racks continue to be bounded by the analyzed conditions, for similar reasons to those discussed in the previous paragraph.

For the PWR racks, it is not readily apparent that the checkerboard arrangement is bounded by the evaluation of the uniform arrangement. In Reference 4, the licensee clarified that any checkerboard arrangement would necessarily continue across rack-to-rack gaps. This is consistent with HNP TS requirement 5.1.e. Therefore, the arrangements analyzed in Section 4.2.5.5 of Attachment 4 of Reference 6 are sufficient to address the possible checkerboard arrangements coincident with a mislocated fresh PWR fuel assembly, which is the most reactive accident scenario near the interface between the Boraflex BWR racks and the PWR racks.

According to Attachment 4 of Reference 6, PWR racks loaded with a uniform storage configuration of fuel that satisfies certain burnup limits could not be determined to be less reactive than the Boraflex BWR racks based on the infinite array calculations. Therefore, the infinite array calculations were used to identify a burnup which, when combined with a uniform loading, would yield an infinite array reactivity equal to that of the Boraflex BWR racks. The licensee states that simply modeling two racks next to each other would result in the overall k-eff being driven by the PWR racks, without regard for how the interface affects the Boraflex BWR racks. The NRC staff agrees with the licensee's characterization of the results due to this modeling approach. However, reducing the reactivity of the fuel in the PWR racks means that the interface calculation is not capturing the impact of a larger potential neutron fission source from higher reactivity fuel in adjacent SFP storage racks. As such, the approach described in of Reference 6 does not adequately capture the potential impact of the interface due to the higher reactivity PWR fuel in adjacent racks. The PWR racks may be qualified for storage of higher reactivity fuel using a different methodology, but the interface calculation should at least consider the impact of more reactive fuel as a boundary condition that may act as an additional contributor of neutrons for the Boraflex BWR rack storage cells near the interface. In order to provide further information to demonstrate that the reactivity increase due to the interface between the Boraflex BWR racks and the PWR racks was not significant, the licensee provided additional information in Reference 4. According to the licensee response to the NRC RAI, PWR fuel at 5.0 wt% U-235 can be qualified for storage in the PWR racks with a burnup of 3,620 MWd/MTU less than that assumed in the uniform loading configurations for the analyses documented in Attachment 4 of Reference 6. The impact of more reactive PWR fuel as a boundary condition was addressed separately for normal interface conditions and accident conditions.

For normal interface conditions, the licensee provided information in Reference 4 demonstrating that the fission distribution for the racks would be dominated by the PWR fuel if the lower burnup is utilized in the model. Therefore, the calculated k-eff would effectively become that of the normal condition for the PWR racks, which is already part of the HNP SFP licensing basis.

No further information was provided to quantify the local impact on reactivity for the Boraflex BWR racks as a result of the increased reactivity fuel as a boundary condition. However, the licensee did state that the impact of the reduction in reactivity for the PWR fuel for the base infinity array was about 0.01 ~k. This can be considered to be an extreme upper limit on the impact of utilizing lower burnups in the PWR fuel (since at this point, the reactivity is no longer driven by the BWR fuel). Sufficient margin to the regulatory limit exists to accommodate the aforementioned conservative reactivity increment to account for lower burnups in the PWR fuel (see Section 3.3.5 of this SE for further discussion).

For accident conditions, a reduction in the burnup assumed for the uniform loading configuration would be expected to cause an increase in reactivity for the limiting scenario, that of a mislocated fresh PWR assembly adjacent to PWR racks in addition to the Boraflex BWR racks.

In lieu of an explicit analysis of this revised scenario, the licensee identified several other conservatisms in the existing accident scenario:

1. The limiting mislocated fuel assembly case is not physically possible due to the fact that the gap between SFP racks is too small for a PWR fuel assembly. The BWR fuel assemblies stored at HNP have a significantly lower enrichment-the maximum IMPAE

[initial maximum planar average enrichment] for all BWR fuel at HNP is 4.178 wt%

U-235 compared to the modeled enrichment of 5 wt% U-235 in the PWR fuel.

Additionally, the amount of fissile material per fuel assembly for BWR fuel is lower than that for PWR fuel due to the size of the fuel assembly.

2. The analysis only credits 1000 ppm of soluble boron for this scenario, but the HNP TS limit is 2000 ppm (the change in reactivity from O to 1000 ppm is 0.141 ~k; while the boron worth would be lower at higher soluble boron concentrations, the total change in reactivity from 1000 ppm to 2000 ppm would still be expected to be more than 0.1 O ~k).

Although the reactivity impact from use of lower burnup PWR fuel assemblies was not quantified, the magnitude of the combined reactivity impact of the two conservatisms indicated above is expected to be at least an order of magnitude higher. Therefore, even with some of the available margin to the regulatory limit being credited to disposition other potential nonconservatisms in Section 3.3.5, the NRC staff finds that reasonable assurance exists that the regulatory limit is met for accident scenarios.

The licensee explains in Attachment 4 of Reference 6 that the Boraflex BWR racks are located in Pools A and B, however, only Pool B has Boral BWR racks. Since the only interface that needs further evaluation is the one between Boraflex BWR racks and PWR racks, for convenience, a full pool model was developed only for Pool A. The gaps between storage racks varies in the different pools, so the licensee determined the minimum gap for adjacent racks.for use throughout the model. The results from a separate calculation addressing potential rack movement due to a seismic event demonstrates that a reduction in the gap size between adjacent racks leads to an increase in reactivity, so this approach is conservative.

The normal conditions for the Boraflex BWR racks assumes that there is one missing Metamic rack insert, because the rack insert must be physically moved before the fuel loaded in the cell can be removed. This location is modeled in the middle of the Boraflex BWR racks based on sensitivity studies to determine the limiting location. Attachment 4 of Reference 6 does not appear to consider the potential for a missing Metamic rack insert near the interface for conditions not covered by the sensitivity studies. The NRC requested additional information on how the licensee considered the reactivity impact of a missing Metamic rack insert adjacent to the interface for other scenarios. The licensee responded that the sensitivity studies were sufficient to cover normal conditions. The sensitivity studies did not consider the impact of lower burnup PWR fuel being loaded in the PWR fuel adjacent to a missing rack insert. However, the information provided in Attachment 4 of Reference 6 does show that the most limiting alternative insert location along the rack edge, based on the interface between adjacent Boraflex BWR racks, is 0.0076 ~k less reactive than the base model with a center insert missing. The minimum gap between the BWR and PWR racks is 20-percent smaller, however, the interface calculations documented in Section 4.2.3. 7 of Attachment 4 of Reference 6 show that the maximum increase in reactivity due to the geometry of the BWR to PWR rack interface is 0.0011 ~k. Therefore, the impact of geometry (missing insert combined with the rack-to-rack gap) would not be expected to result in a reactivity increase. The impact of the composition of the lower burnup PWR fuel on the interface was discussed earlier in this section of this SE, and an upper limit estimate was determined to be 0.01 ~k. This estimate was already dispositioned against the available margin to the regulatory limit (see Section 3.3.5 for further discussion).

Even though the combination of these effects will not necessarily be purely additive, given the significant margin to the regulatory limit, the NRC staff finds that reasonable assurance exists that the subcriticality limit will not be challenged for normal conditions.

The licensee also stated (Reference 4) that a missing insert concurrent with the limiting scenario of a mislocated fuel assembly was not credible because rack inserts and fuel assemblies cannot be moved at the same time. However, as discussed earlier in this section of this SE, there is significant conservatism in the postulated mislocated fuel assembly accident. The NRC staff also agrees that the administrative prohibition on storing fuel in the Boraflex BWR rack cell closest to the postulated location for the mislocated fuel assembly would prevent the reactivity variation due to a missing insert from becoming significant enough to challenge the aforementioned conservatism in the analysis. Therefore, a mislocated fuel assembly accident concurrent with a missing insert is bounded by the analyses performed by the licensee.

There are other accident scenarios that may be affected by the presence of a missing insert near the rack interface. The misorientation scenario is bounded by a very conservative scenario with multiple missing inserts in adjacent cells, which could only be possible if there was a concurrent failure to follow fuel-handling procedures in combination with a failure by the operator to recognize that multiple adjacent rack inserts are missing in an area where all cells are expected to have inserts. Therefore, engineering judgment was used to conclude that credible misorientation scenarios at the interface are bounded by the analyzed scenario, considering the reduction in reactivity that would be expected to result from the interface gap along with less neutronic coupling between fuel assemblies. The rack insert orientation is a TS requirement, and as such, cannot be changed without prior NRC approval.

In addition to the effect of a missing Metamic rack insert on the initial conditions for a mislocated fuel assembly accident, the initial conditions at the interface may be affected by changes in the spacing between adjacent racks. One of the postulated accident conditions analyzed by the licensee is rack movement due to a seismic event, in which the racks are allowed to move closer together. The calculations performed by the licensee show that there is a clear increase in reactivity resulting from such a scenario. The NRC staff asked what controls are in place to ensure that, should a seismic event occur, the licensee will verify that the spacings between racks in the HNP SFPs are bounded by the spacings assumed in the normal condition analysis, and take any appropriate corrective action to ensure that the normal condition analysis remains valid. The licensee responded in Reference 4 by stating that administrative procedures would require verification of the SFP gaps following a seismic event. Section 4.2.3. 7 of Attachment 4 of Reference 6 states that the rack-to-rack gaps are modeled as if they are at the minimum values allowed by the rack baseplate extensions. The acceptance criteria in the administrative procedures allow for a reduction of up to 25 percent in the rack-to-rack gap widths relative to the as-installed condition. The information provided in Reference 4 and Attachment 4 of Reference 6 do not indicate how this compares to the configuration modeled in the analyses documented in Attachment 4 of Reference 6. However, based on conservative comparisons using the minimum gap sizes and the configuration modeled for the rack movement accident scenario, the NRC staff determined that the reactivity impact from a reduction in gap width should be no more than 0.015 flk, which can be accommodated by the available margin to the regulatory limit. Therefore, the licensee administrative procedures are sufficient to address the small amount of rack movement from nominal conditions that may be tolerated as part of the normal conditions, and subsequently become part of the accident conditions.

The possible reactivity impacts of the SFP storage rack interfaces were explicitly evaluated by the licensee or dispositioned based on available conservatisms and margin. As a result, the NRC staff has determined that credible variations in the configuration near the interface between the Boraflex BWR racks and the other rack designs in the HNP SFP will not result in a reactivity increase sufficiently large to challenge the regulatory limits.

3.3.3 Fuel Assembly 3.3.3.1 Bounding Fuel Assembly Design The BWR fuel stored in the HNP SFPs may be one of multiple General Electric (GE) fuel designs. The licensee selected the most reactive fuel design allowed for storage in Pools A and Bat HNP (GE7), and modeled it as a fresh fuel assembly with no poison and a uniform U-235 enrichment loading at the IMPAE plus the enrichment tolerance for all fuel currently stored in these pools. There are more reactive fuel designs stored in other SFPs at HNP, but they are administratively restricted from storage in Pools A and B. Therefore, the most reactive fuel design from the other SFPs (GE13) is used as the basis for the multiple misloading accident discussed in Section 3.3.4. The licensee did a calculation in order to demonstrate that use of a uniform U-235 enrichment loading at the IMPAE is conservative relative to a typical BWR lattice with pin-by-pin variation in U-235 enrichment to control peaking. The licensee's conclusions with respect to the relative conservatism of the various fuel designs and use of IMPAE were consistent with the NRC staff's prior evaluations of similar studies.

HNP does not currently have any plans to add new BWR fuel for storage in Pools A and B. In the event that they should choose to do so, however, they would have to ensure that the reactivity of any BWR fuel added to Pools A and B is bounded by the criticality analysis. The standard TSs for BWRs typically includes a limit on the k-infinity, enrichment, or similar parameter for this purpose. The licensee did not provide any such proposed change to their TSs, so the NRC staff asked how they will ensure that one of the most important parameters for criticality safety is controlled. The licensee responded in Reference 4 that the TS restriction on which BWR fuel designs may be stored in Pools A and B meets the same purpose, since the BWR fuel population stored at HNP is static and no more BWR fuel assemblies are expected to be shipped to HNP for storage. Therefore, the criticality analyses for normal conditions conservatively bound all BWR fuel designs allowed for storage in the Pools A and B, as if they were fresh and unpoisoned.

Fuel rod reconstitution can have an impact on reactivity, however, the licensee stated that they do not intend for fuel rod reconstitution operations to be performed on the BWR fuel at HNP.

The BWR fuel was transferred from another location and is in the SFP for storage purposes only. Therefore, this LAR did not cover reconstitution and these types of scenarios were not considered by the NRC staff as part of the review of this LAR.

The selection and modeling of the fuel lattice used in the NCS analyses was determined to bound all fuel intended for storage in HNP Pools A and B. This fuel lattice was used along the full active fuel length for a design-basis fuel assembly, with structural materials neglected. This is conservative, and therefore, acceptable for use in a SFP NCS analysis intended to bound all fuel in Pools A and B at HNP.

3.3.3.2 Fuel Assembly Manufacturing Tolerances and Uncertainties The manufacturing tolerances of the fuel assemblies contribute to SFP reactivity.

DSS ISG-2010-01 (Reference 14) does not explicitly discuss the approach to be used in determining manufacturing tolerances, but past practice has been consistent with the Kopp Memo (referenced in DSS-ISG-2010-01) that determination of the maximum keff should consider either: (1) a worst-case combination with mechanical and material conditions set to maximize keff, 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 assemblies. The licensee chose to utilize the former approach with all the as-built parameters of the GE7 fuel design that would result in a significant reactivity impact.

In cases where the limiting direction for variation in the parameter is well-established, such as the U-235 enrichment, fuel pellet density, and fuel pellet outer diameter, the licensee simply used the upper bound for the manufacturing tolerance range in all calculations. In other cases, both extremes for the manufacturing tolerance range were evaluated to determine which value would result in the maximum reactivity. The resulting combination of worst-case parameters for reactivity were used for all subsequent criticality calculations.

The licensee did not address the uncertainties associated with the other fuel designs in the HNP SFP. However, the design fuel assembly is modeled using bounding parameters. This approach means that the uncertainties associated with the GE7 fuel assembly are being treated as biases rather than uncertainties, which is conservative. In order for the uncertainties associated with other fuel designs in Pools A and B of HNP to have any effect on the results from this analysis, the statistical combination of uncertainties would have to be larger than the bias inherent in this approach for modeling the GE7 fuel design. The NRC staff has determined that this is unlikely to be the case, based on typical values for total fuel assembly uncertainties and the difference in reactivity between nominal and bounding parameters documented in of Reference 6.

As a result of evaluating the licensee's treatment of the manufacturing tolerances, the NRC staff determined that the licensee treated all relevant fuel assembly parameters in a bounding conservative manner. Other fuel designs are present in the SFP and are included in some of the analyses, but the NRC staff concluded that any potentially larger reactivity impacts from the uncertainties associated with other fuel designs were unlikely to be sufficiently large to overcome the inherent conservatism in the treatment of the GE7 fuel design uncertainties.

3.3.3.3 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, as discussed above, or are bounded by values used in the analysis. These tolerances and bounding values would also carry through to the spent nuclear fuel; common industry practice has been to treat the uncertainties as unaffected by the fuel depletion. The characterization of spent nuclear fuel is more complex. Its characterization is based on the specifics of its initial conditions and its operational history in the reactor. That characterization has three main areas: a burnup uncertainty, the axial and radial apportionment of the burnup, and the core operation that achieved that burnup.

The licensee elected to utilize a fresh, unpoisoned fuel lattice to construct the design-basis fuel assembly for evaluation of the Boraflex BWR racks. As discussed in Section 3.3.3.1 of this SE, the U-235 enrichment is sufficient to bind all BWR fuel stored in the HNP pool. The cumulative reactivity impact due to burnup related uncertainties will be much smaller than the reactivity decrease due to the depletion of U-235, so no burnup related uncertainties need to be evaluated for the approach that the licensee selected.

There is, however, a reactivity impact associated with the characterization of the PWR fuel stored in adjacent fuel racks, which is included in the evaluation of interfaces and certain accident conditions. The PWR spent fuel characterization methodology has previously been approved by the NRC as part of the initial installation of the PWR storage racks. However, if the evaluation of the interface scenarios in Attachment 4 of Reference 6 result in more conservative results than the interface conditions evaluated as part of the licensing basis for the PWR storage racks, then the base reactivity (i.e., without biases or uncertainties applied) for the interface conditions from Attachment 4 of Reference 6 supersede the current values for interface conditions in the current licensing basis for the PWR storage racks. The spent fuel characterization related to burnup distributions and core operations are already captured in the assumptions used to generate the PWR fuel composition, but the depletion related uncertainties (depletion uncertainty, burnup measurement uncertainty, and fission product uncertainty) are applied separately. Based on data from approved SFP criticality LARs from the past decade, the NRC staff determined that the maximum combination of depletion related uncertainties for configurations of fuel with similar burnups can be expected to be less than 0.018 ~k for unborated conditions and less than 0.015 ~k for 1000 ppm of boron. These values can only be treated as estimates. The discussion in Section 3.3.5 shows that there is still significant remaining margin to the regulatory limit after accounting for this issue as well as other potential nonconservatisms. Furthermore, in Section 3.3.5, the NRC staff adds an extra layer of conservatism by not statistically combining this reactivity impact with other uncertainties from the licensee's analysis. The NRC staff concluded that inclusion of the depletion uncertainties from the PWR fuel would not challenge the regulatory limit.

Since no depletion related characteristics need to be evaluated for the BWR fuel, and the PWR fuel was addressed using a previously NRC approved methodology, the licensee has adequately considered the relevant impacts of fuel depletion on the SFP criticality analysis.

3.3.3.4 Integral Burnable Absorbers The BWR fuel stored in the HNP SFP utilized gadolinia poison to help control reactivity and peaking within fuel assemblies. In addition to gadolinia related uncertainties, the specific characteristics of the gadolinia poison loading (location of rods, gadolinia concentration, etc.)

may affect the reactivity due to its impact on the neutron spectrum experienced by the fuel during depletion. The licensee used a design-basis fuel assembly with no burnup or burnable poison. Since the reactivity of fresh fuel with no poison is much larger than the reactivity of depleted fuel with poison after accounting for any uncertainties related to fuel depletion and gadolinia (whether they are caused by manufacturing or depletion), no further disposition is necessary for the burnable absorbers.

The licensee's treatment of the integral burnable absorbers represents a very conservative approach, and therefore, is acceptable.

3.3.4 Analysis of Abnormal Conditions Section 4.2.5 of Attachment 4 of Reference 6 presents the abnormal conditions considered in the analysis. The licensee considered the following abnormal conditions:

SFP temperature exceeding the normal range Dropped fuel assembly Misloaded fuel assembly (fuel assembly loaded in the Boraflex BWR racks that is not qualified for storage in these racks)

Mislocated fuel assembly (fuel assembly positioned outside the storage rack)

Rack movement Misorientation of Metamic rack inserts The licensee elected to credit soluble boron present in the SFP, so the criticality calculations were performed assuming O and 1000 ppm of soluble boron in the SFP water. The O ppm calculations demonstrate that the boron dilution accident is irrelevant to HNP because the SFP k-eff was shown to remain less than 0.95 with no soluble boron in the SFP water. The HNP TS 3.7.14, Limiting Condition for Operation (LCO) requirement for soluble boron in the SFP is 2000 ppm, and the proposed value of 1000 ppm requirement for criticality control is being incorporated into the SFP storage design features of the TS 5.6.1.3. Therefore, the soluble boron credit will be implemented in a manner that ensures that the SFP boron concentration will be clearly recognized as an essential prerequisite to ensuring continuing compliance with NRC subcriticality requirements based on the SFP NCS analysis of record.

As discussed in Section 3.3.1 of this SE, the SFP temperature was treated in a bounding manner. The licensee evaluated SFP temperatures well outside the normal range, and selected the most limiting temperature for all of the design-basis calculations. Therefore, the design-basis calculations already incorporate the possibility of an off-normal SFP temperature.

The licensee performed a very conservative evaluation of the misloaded fuel assembly event, in which all fuel assemblies in the design basis calculation were replaced with a fresh 5.0 wt%

U-235 GE13 fuel assembly with the most reactive fuel lattice modeled for a length slightly longer than the active fuel length for a BWR. The GE13 fuel design is administratively restricted from storage in Pools A and Bat HNP, and no other more reactive fuel designs are stored at HNP that would fit in the Boraflex BWR racks. In addition, the U-235 enrichment exceeds the IMPAE for fuel stored at HNP by about 19 percent. This loading of the most reactive fuel design, as a fresh fuel assembly with a bounding U-235 enrichment, into every cell of the Boraflex BWR racks represents a very conservative approach that bounds any possible combination of misloaded fuel. In addition, the licensee states that the dropped fuel assembly scenarios are bounded by this evaluation, given that they generally involve only one fuel assembly and the resulting fuel arrangements would be bounded by the modeled misleading scenario. In the horizontal dropped fuel scenario, a fuel assembly lying on top of the Boraflex BWR racks would not be very strongly neutronically coupled to the fuel in the SFP storage cells. Any reactivity variations due to damage of the Metamic rack inserts or SFP cell resulting from a vertical fuel assembly drop are expected to be much smaller than the relative conservatism of the misleading scenario. As a result, the NRC staff agrees that the dropped fuel assembly scenarios are bounded by the misleading evaluation.

The licensee evaluated a mislocated fuel assembly scenario where a fresh 5.0 wt% U-235 17x17 PWR fuel assembly with no poison is located outside the fuel storage racks, along the periphery of the Boraflex BWR racks. This represents the most reactive fuel design allowed for storage in the HNP pools, modeled with a bounding U-235 enrichment with no credit for burnup or integral neutron absorbers. The location where the licensee modeled the mislocated fuel assembly is not physically possible, since the gap between storage racks is slightly smaller than the size of the fuel assembly. However, the licensee removed much of the adjacent PWR rack exterior surface in the model to allow the PWR fuel assembly to fit. Additional calculations performed by the licensee based on various configurations confirmed that this scenario was more conservative than other physically possible locations. The licensee did not reconsider the interface conditions to determine if alternative conditions might be more limiting for the specific scenario being analyzed. A discussion of the possible situations identified by the NRC staff where the mislocated fuel scenario may be made more conservative by use of different interface conditions can be found in Section 3.3.2.2. During the evaluation, the licensee determined that with a soluble boron credit of 1000 ppm and the design basis Boraflex BWR rack model, the calculated k-effective would not meet the regulatory limit. Therefore, the licensee reanalyzed the same mislocated fuel scenario, but without any fuel loaded in the cell located at the corner of the Boraflex BWR racks closest to the mislocated fuel. In order to ensure that this updated evaluation is valid for the HNP SFPs, this restriction on fuel storage in the aforementioned storage cell was captured as a proposed TS requirement. The original evaluation results were included in Attachment 4 of Reference 6, but are not used to support regulatory conclusions.

The licensee addressed the potential impact of rack movement during a seismic event such that the gaps between racks are reduced by taking a full pool design basis model and revising it so the gaps between racks are nearly closed. This configuration is not physically possible due to the fact that baseplate extensions will normally block the racks from moving that close together.

However, this ensures that the reactivity impact will bound all possible physical configurations resulting from the postulated accident. This evaluation is acceptable to address accident conditions, but the licensee must have controls in place to check for rack movement or otherwise disposition the potential impact of gap closure after a seismic event by restoring the nominal gap sizes or addressing the criticality impact for normal conditions. A discussion on how the licensee ensures this accident condition does not become part of the normal condition can be found in Section 3.3.2.2 of this SE.

Finally, the licensee evaluated misorientation of the Metamic rack inserts. Since the orientation of the Metamic rack inserts is controlled by administrative procedures and is part of the design features being incorporated in the TS, misorientation of multiple Metamic rack inserts is an unlikely occurrence. The normal conditions include one missing Metamic rack insert, so this would have to be combined with any misorientation to maximize the potential reactivity increase.

The licensee chose to do a conservative calculation in which five Metamic rack inserts are missing. This includes the cell with a missing insert that is part of the normal conditions, plus all four face-adjacent cells. The resulting configuration includes 11 SFP storage cells in which at least one boundary shared with another SFP storage cell does not include Metamic absorber material between the two fuel assemblies. This would be expected to be more conservative than any postulated arrangement of misoriented Metamic rack inserts plus one missing rack insert given the presence of administrative procedures to control insert orientation. The reactivity impact due to this arrangement was actually 0.0008 ~k more reactive than the limiting mislocated assembly accident used as the basis for the licensee's accident condition k-ett evaluation. However, this postulated scenario goes well beyond credible accidents, since the Metamic rack insert orientations will be incorporated as a TS requirement upon implementation of the proposed license amendment. The impact of interface conditions is discussed in Section 3.3.2.2 of the SE.

The primary controls that the licensee plans to implement in order to ensure that the normal conditions of the HNP SFPs are maintained such that the postulated accidents do not result in a configuration more reactive than the ones analyzed in Attachment 4 of Reference 6 will be incorporated into the HNP TS as part of the proposed changes. They include: (1) an administrative restriction on loading fuel in the corner cell face adjacent to the limiting mislocated fuel location; (2) an administrative restriction limiting BWR fuel storage in Pools A and B to specific fuel designs; and (3) an administrative requirement controlling the orientation of all Metamic rack inserts.

The licensee performed a thorough evaluation of all potential accident conditions and considered all possible reactivity impacts, so the calculation of the k-infinity for accident conditions is acceptable. Any administrative restrictions necessary to ensure that the normal conditions are controlled in a way to preclude more reactive accident conditions are being incorporated into the TS to provide assurance that these restrictions will remain in place.

3.3.5 Disposition of Non-conservatisms Several potentially nonconservative assumptions were identified as part of the NRC staff review of this LAR. A conservative bounding estimate of the reactivity impact for each assumption is listed in the table below, based on NRC staff independent calculations, studies, or other available data. In addition, the margin to the regulatory limit for each of the relevant limiting analysis conditions are listed. These nonconservative reactivity impacts may or may not exist, but if they do, based on the below comparison, the NRC staff concludes that the available margins would offset the potential nonconservatisms.

Estimated Estimated Reactivity Impact Reactivity Impact

- Normal,

-Accident, Unborated (llk)

Boron Credit (llk)

Potential Non-conservatisms*

Use of lower reactivity fuel than allowed by TS in 0.010 0.010 PWR racks Gap width reduced by up to 25%

0.015 0.015 Depletion related uncertainties for PWR fuel 0.018 0.015 Total reactivity impact of nonconservatisms 0.043 0.040 Identified Margin to Regulatory Limit Margin to regulatory limit identified in licensee 0.0481 0.0315 calculation Margin due to credit for 1000 ppm soluble boron N/A

>0.10 instead of TS limit of 2000 ppm Total identified margin to regulatory limit 0.0481

>0.1315

• Additional conservatisms exist in the licensee analysis, such as the treatment of manufacturing tolerances as bounding parameters and conservatisms related to the mislocated fuel assembly accident (see Section 3.3.2.2 of this SE for more discussion). The NRC staff did not attempt to quantify these conservatisms, but the staff recognized that they would be expected to be significant enough to preclude the need to assess the above potential nonconservatisms to a higher level of certainty.

The NRC staff's review of the analysis documented in Attachment 4 of Reference 6 identified some potentially nonconservative items. In all cases, the NRC staff was able to estimate the reactivity impact of the potential nonconservatisms. As shown in the above table, those reactivity impacts were evaluated against the margin to the regulatory limit described in (10 CFR) 50.68(b)(4) (i.e., margin to a keff of 0.95) and the significant conservatism inherent in crediting less soluble boron than required by the LCO in TS 3. 7.14. As a result, the NRC staff determined that the potential nonconservatisms can be accommodated by the available margin to the regulatory limits described in 10 CFR 50.68(b)(4). Since the licensee has demonstrated that 10 CFR 50.68(b)(4) requirements are met and the proposed TS changes will incorporate that the key parameters ensuring continued safety, 10 CFR 50.36(c)(4) is also satisfied with respect to ensuring that the key design features relevant to criticality safety will be maintained as part of the operating license for HNP. Title 10 CFR 50.68(b)(1) is a general requirement with respect to SFP operations including movement of fuel assemblies outside of the SFP racks.

Therefore, demonstrating adequate subcriticality (i.e., 10 CFR 50.68(b)(4) limits are met) when considering all normal and abnormal SFP operating conditions is sufficient to show that 10 CFR 50.68(b)(1) is satisfied. GDC 62 is implicitly shown to be satisfied through compliance with 10 CFR 50.68(b)(4) requirements for all normal and abnormal SFP operating conditions, in that the existing physical systems or processes have been shown to prevent criticality through maintenance of adequate subcriticality margin.

3.4 Metamic Surveillance Program The licensee has proposed a Metamic Surveillance Program. The purpose of this program is to detect potential degradation of the Metamic material, ensure that the proper monitoring and trending occur, and that the appropriate corrective actions are implemented if degradation is detected. The licensee has proposed a coupon-testing program that will be consistent with the guidance provided in TR NEI 16-03-A (Reference 11 ). The coupons are small sections of Metamic from the same manufacturing lots that were used to fabricate the rack inserts. These coupons would be placed on a "tree," which is then placed in a cell in the SFP so that the coupons experience environmental conditions that bound those of the Metamic panels credited for spent fuel criticality control.

The licensee has proposed certain aspects to be included in the coupon testing program.

These are described in detail in the LAR. Aspects of the coupon testing program include neutron attenuation testing, visual observation, dimensional measurements, and weight and specific gravity measures starting at intervals every 5 years and extending to 10 years provided acceptable performance is observed during the initial test intervals.

The results of the coupon-testing program will be evaluated in accordance with the acceptance criteria described in the licensee's submittal (Reference 1) and request for additional information response (Reference 4). The acceptance criteria will help to identify if unanticipated changes are occurring in the condition of the material. If such changes are identified, the licensee will address potential degradation through evaluation, trending, and the site corrective action program. The licensee stated that the Metamic Surveillance program will be incorporated into a plant procedure that is incorporated by reference into the HNP Final Safety Analysis Report.

The licensee also stated that the Metamic Surveillance program will be used to verify that the Metamic continues to provide the credited criticality control assumed in the SFP criticality analysis.

3.4.1 Metamic Surveillance Program NRC Staff Evaluation The NRC staff has reviewed the contents of the LAR related to the properties of Metamic NAM, compatibility of the Metamic NAM in the SFP environment, and the Metamic Surveillance Program. The staff's review was conducted using guidance from SRP Sections 9.1.1 (Reference 8), and 9.1.2 (Reference 9), as well as the content from TR NEI 16-03-A (Reference 11 ). In addition, the staff considered the precedent provided by the licensee. The staff used this information to determine if the Metamic NAM, and the associated Metamic Surveillance Program, provide reasonable assurance that the appropriate parts of 10 CFR 50.68, GDC 61, and GDC 62 are met.

Metamic has been previously approved for use by the staff in several SFP applications, including SFPs C and D at HNP. Due to the composition of Metamic, and operating experience from previously approved and installed Metamic in other SFPs, the staff has reasonable assurance that the Metamic will be compatible with the SFP environment in SFPs A and B at HNP. The NRC staff also reviewed the details of the coupon monitoring program. The Metamic coupons will be fabricated from the same lots of the Metamic NAM used for the rack inserts.

The coupons will be located in a spent fuel cell that is surrounded with fuel assemblies that will bound the operating conditions of the installed inserts with respect to flow, temperature, and radiation dose. The licensee stated that there will be 14 Metamic coupons in order to provide enough samples to support the surveillance program past the end of the HNP operating license.

These coupons will be pre-characterized in order to compare the pre-and post-duty characteristics. The licensee provided a detailed description of the coupon-testing program in Section 3.6.2, "Coupon Testing Program Details," in the LAR (Reference 1). By manufacturing the coupons from the same lots as the Metamic in the inserts and placing the coupons in a cell surrounded by bounding fuel assemblies, the NRC staff has reasonable assurance that the coupons will be representative, or bounding of the in-service material.

The staff finds the details of the coupon testing program acceptable because the coupons are pre-characterized so that any test data may be monitored and trended against the as-installed data. In addition, the coupon program is acceptable because the licensee conducts neutron attenuation testing, visual examinations, dimensional measures, and weight and specific gravity measurements, on an appropriate interval (maximum of 10 years) given the NAM used. This provides the staff reasonable assurance that the coupon monitoring program will be able to detect signs of potential degradation.

With regards to evaluating the results of the coupon monitoring program the license stated that the evaluation will determine if no material changes occurred, confirmation that anticipated changes occurred, or confirmation that unanticipated changes occurred. If unanticipated changes occurred then the issues will be entered into the site Corrective Action Program (CAP) for further evaluation. Additionally, the licensee will determine if these changes could result in a loss of 8-10 areal density (AD), or impact the criticality analysis. The licensee stated that the surveillance program will include an acceptance criterion for the 8oron-10 (108) AD. This criterion will compare the pre-irradiated 108 AD value of the coupons to the measured 108 AD value of the in-service coupons. The acceptance criterion must be greater than or equal to the pre-irradiation value, while considering measurement uncertainty. The NRC staff finds the approach to evaluating the coupon monitoring program results acceptable because the licensee will evaluate the results of the program and enter an issue into the CAP if material degradation is found. In addition, the staff finds this approach acceptable because the licensee will evaluate impacts of potential material degradation on the 8-1 O AD content of the NAM, and the SFP criticality analysis to ensure 10 CFR 50.68, GDC 61 and GDC 62 continue to be met.

The licensee also stated that the Metamic Surveillance Program would follow the guidance provided in TR NEI 16-03-A (Reference 11 ). The NRC staff finds this acceptable, as the details of the Metamic Surveillance Program provided in the LAR are consistent with the guidance of TR NEI 16-03-A (Reference 11 ), which has been previously approved and endorsed by NRC as a means to meet the regulatory requirements.

In Reference 6, the licensee stated it will incorporate the Metamic Surveillance Program into the licensee-controlled procedure PLP-106, "Technical Specification Equipment List Program and Core Operating Limits Report." The PLP-106 procedure is incorporated by reference into the HNP FSAR. As such, the program will be incorporated into HNP's licensing basis and is subject to the update and reporting requirements for the information in the HNP FSAR.

The NRC staff reviewed information from the licensee's Metamic surveillance program, and has determined that the surveillance program as described in the LAR will provide reasonable assurance that the licensee will be able to detect degradation of the neutron absorbing material before its ability to perform its intended safety function is impacted. Additionally, the staff has reviewed the information provided regarding material compatibility with the SFP environment, as well as the continuing surveillance program, and has reasonable assurance that Metamic will be chemically compatible with the SFP environment. On this basis, the staff concluded that the proposed addition of a SFP neutron absorber monitoring program, the contents of the program, and the use of Metamic in the SFP, meet the applicable requirements of 10 CFR 50.68(b)(4) to maintain SFP storage racks below a keff of 0.95, and GDC 61 to design the system to permit appropriate periodic inspection and testing, and 62 to prevent criticality by physical means, in part by the use of Metamic inserts, and are, therefore, acceptable.

3.5 Thermal Hydraulic Evaluation The licensee performed the following thermal-hydraulic analyses to evaluate the impact of adding the DREAM inserts in the Westinghouse-supplied SFSRs:

1.

Assessment of the plant's current SFP bulk thermal evaluation to determine if it will continue to apply following installation of the DREAM inserts.

2.

Assessment of the plant's current SFP time-to-boil evaluation to determine if it will continue to apply following installation of the DREAM inserts.

3.

A rigorous Computational Fluid Dynamics (CFO) based study to conservatively quantify the peak local water temperatures in the SFSRs following installation of the DREAM inserts.

4.

Determination of a bounding maximum fuel cladding temperature following installation of the DREAM inserts.

These analyses are described in detail in Sections 5.3 through 5.5 of Attachment 4 to Reference 1. A single scenario is postulated and analyzed, with all Westinghouse BWR SFSRs loaded with fuel assemblies having the maximum decay heat per assembly permitted in the Vectra IF-300 shipping cask used to transport them to HNP, and the SFP bulk temperature set to the bulk temperature limit of 150 °F. The licensee identified the following design criteria:

1.

Following the planned offload fuel assemblies from the Harris reactor and with forced cooling available, the bulk SFP temperatures shall be limited to 150 °F. This criterion can be met by demonstrating that the existing licensing basis remains applicable.

2.

Under a complete failure of active cooling during the limiting fuel offload scenario, the water surface is allowed to reach saturation. Sufficient time must be available before the onset of bulk boiling to implement corrective measures. This criterion can be met by demonstrating that the existing licensing basis remains applicable.

3.

Local water and fuel cladding temperatures for the fuel assemblies within the Westinghouse BWR SFSRs shall not exceed the local saturation temperature of water.

The licensee applied the following analytical assumptions, as applicable:

1.

Heat loss by natural convection, mass diffusion and thermal radiation from the surface of the SFP water is neglected, as is conduction heat transfer through the SFP structure. Thus, all decay heat loads are considered to be removed by the SFP cooling system alone, maximizing computed temperatures.

2.

No downcomer flow is assumed to exist between the SFSR modules in the SFPs, minimizing the ability of cooled water to enter the bottom of the rack cells.

3.

All SFSR cells are assumed to have the inlet flow holes geometry of the pedestal cells.

This conservatively reduces the water flow area into the storage cells, thereby increasing the hydraulic resistance.

4.

An additional heat transfer resistance is imposed on the outside of the fuel rods to account for any crud layer, thereby increasing the calculated fuel cladding superheat.

5.

The maximum local water temperature and the peak heat flux are assumed to occur at the same location, which ensures that the calculated peak fuel cladding temperature bounds the fuel cladding temperature anywhere along the length of the fuel assembly.

Regarding the equilibrium SFP temperature evaluation, the licensee reviewed the current licensing basis evaluation to identify whether the SFP water volume or thermal inertia is credited in performing the calculation. This review indicated that steady-state heat balances are used to determine the equilibrium bulk temperatures with no credit taken for heat energy storage by the SFP water. Because no thermal inertia is credited, the displacement of water by the DREAM inserts has no impact and the existing licensing basis evaluation remains applicable following addition of the inserts.

Regarding the time-to-boil evaluation, the addition of the DREAM inserts displaces a quantity of SFP water, which slightly reduces the thermal inertia of the SFP. The current licensing basis time-to-boil cafoulation is a transient evaluation that credits the SFP thermal inertia. The licensee reviewed the current licensing basis and determined that there is sufficient margin in the credited water volume to bind the water displaced by the addition of the DREAM inserts.

Regarding the local water and fuel cladding temperatures, the licensee performed a detailed analysis using three-dimensional CFO. The licensee considered several significant geometric and thermal-hydraulic features of the Westinghouse BWR SFSRs in each SFP. The licensee considered two regions in the SFP: (1) the bulk region outside the Westinghouse BWR SFSRs, where the classical Navier-Stokes equations are solved with turbulence effects included, and (2) the heat-generating zone of Westinghouse BWR SFSRs loaded with fuel assemblies, where water flow is directed vertically upwards by the buoyancy forces through relatively small flow channels formed by the fuel assembly rod arrays in each rack cell. The Westinghouse BWR SFSRs were modeled as porous medium regions in which Darcy's Law governs fluid flow.

Turbulence effects were modeled by relating time-varying "Reynolds' Stresses" to the mean bulk flow quantities by the standard k-£ turbulence model. The licensee included fuel peaking factors. The licensee concluded that local water and fuel cladding temperatures remain below local saturation temperature, which is approximately 240 °F, at the top of the active fuel length.

This is consistent with the assumptions in the criticality analysis, which relies on a certain moderator density range with a no voids condition.

The staff asked for additional information regarding the licensee's modelling of DREAM inserts.

The licensee noted that increased hydraulic resistance can result in elevated fuel cladding temperature and impact the time-to-boil evaluation. The staff asked whether the licensee's evaluation considered the increase in hydraulic resistance that would result from installation of the DREAM inserts. The licensee responded that the local hydraulic resistance does not impact the onset of bulk boiling: the only effect of the DREAM inserts on the time-to-boil is due to displacement of "a quantity of SFP water, which slightly reduces the thermal inertia of the SFP,"

which was evaluated as described above. The licensee also stated that, with respect to fuel cladding temperature, the increase in hydraulic resistance was considered by reducing the flow area in the storage cell to reflect the presence of the DREAM insert. Reducing the flow area in the cell decreases the hydraulic diameter, which (1) increases the local water temperature computed in the CFO by increasing the viscous resistance term, and (2) decreases the clad-to-water superheat by increasing the convective heat transfer coefficient. The licensee stated that they considered both effects.

The NRC staff reviewed the above assumptions and concludes that they will provide conservative analytical results. The staff also reviewed the design data presented in Section 5 of Attachment 4 (Reference 1) and concludes that the data is consistent with the current licensing basis with the addition of the DREAM inserts. With regard to the equilibrium SFP temperatures, the staff concludes that "no change" is the expected result because there is no change in decay heat load or SFP cooling system capability. With regard to time-to-boil evaluation, the staff concludes that the licensee used an acceptable method of demonstrating continued applicability of the current design basis. The licensee performed a CFO to show that the liquid in the rack channels remained below the saturation temperature locally throughout the entire elevation. With regard to local water and cladding temperatures, the results appear reasonable. The staff concludes that the licensee has used acceptable methodologies.

The NRC staff reviewed Chapter 5 of the Holtec report in Attachment 4 of the LAR (Reference 1 ), which supports statements in Section 3.4 of the LAR. The staff concludes that the HNP SFPs will continue to meet the thermal-hydraulic criteria for safe storage of spent fuel following installation of DREAM inserts in the Westinghouse-supplied BWR SFSRs. The licensee used acceptable evaluation methods and conservative assumptions. Bulk SFP temperature and time-to-boil remain bounded by previous analysis, and local water and cladding temperatures meet the guidance in SRP Section 9.1.3 and the NRC's OT Position Paper on spent fuel storage. The CFO analysis assured that, locally, the liquid remains below the saturation temperature, which satisfied the OT and supports the criticality analysis. The staff finds since there is no change in the equilibrium SFP temperature or time to boil, there is no impact on GDC 4 or GDC 61 and they continue to be met.

3.6 Structural/Seismic Evaluation of the Existing Spent Fuel Storage Racks 3.6.1 Duke Energy Analysis of Racks Equipped with DREAM Inserts The LAR Section 3.5, "Rack Structural Evaluation/Seismic Considerations," references Section 6.0, "Existing Rack Structural/Seismic Considerations," of Holtec International Report Hl-2177590, "Licensing Report for Use of DREAM Neutron Absorber Inserts in the Spent Fuel Pools 'A' and 'B' at Shearon Harris NPP," Revision 1 (non-proprietary) (Reference 1 ), which describes the structural evaluation of the HNP SFP racks after DREAM inserts have been added to the existing Westinghouse BWR racks located in SFPs A and B. The Holtec report observes the negligible weight of the inserts when compared to the overall deadweight of the SFP racks and pool structures and points out the conservatism of the analysis. The report also states that the structural design bases for the existing BWR Boraflex racks in Harris Pools A and B, as well as the structural qualification of the pool structures, are not adversely affected by the planned installation of neutron absorber inserts. In its LAR, Duke Energy stated that a single Metamic insert weighs only a small fraction of the total weight of a loaded rack, and that the structural integrity of the DREAM inserts under normal and accident (including seismic) loading conditions was found to be adequate to perform their intended design function.

Although the evaluation of the structural design bases for SFPs A and B considered the seismic qualification of the existing BWR Boraflex racks, the SFP and liner structural qualification, and mechanical accident evaluation, the report does not provide information regarding analysis and evaluation of such. Furthermore, Section 6.1 of Revision 1 of the Holtec report states, in part, that the effects of the DREAM inserts on the structural design bases are evaluated by reviewing the existing analysis reports. The report states DREAM inserts are found to be structurally adequate to perform their intended function under both normal and seismic conditions. Since these reports were not provided in the original LAR submittal, the staff determined that additional information was needed to complete its review. The staff requested additional information from the licensee regarding the structural analysis and design adequacy of the existing SFP racks and pool structure outfitted with Metamic inserts designed to meet NRC regulatory requirements. In its RAI 1 response (Reference 4), the licensee stated that the racks are classified as American Nuclear Society Safety Class 3, Seismic Category I structures, and are designed to withstand normal and postulated dead loads, live loads, loads due to thermal effects, and loads caused by the operating basis earthquake (QBE) and safe shutdown earthquake (SSE) events in accordance with NRC Regulatory Guide 1.29, and allowable stresses defined by ASME Code Section Ill, Subsection NF, and Appendix XVII. Additionally, the design and SE of the racks is in accordance with the NRC's OT Position Paper, including the modifications provided in NRC Generic Letter 79-04, and are designed to withstand an uplift force equal to the maximum uplift capability of the spent fuel bridge crane.

The licensee also stated that the original structural analysis for the Westinghouse Boraflex BWR spent fuel racks was performed by Westinghouse, as documented in their proprietary design report. The report describes the analysis and provides details of the calculation methods and results for the racks. The licensee further stated that because of the structural nonlinearities of the rack support boundaries and the gaps between the fuel assembly and the fuel cell, a nonlinear time-history seismic analysis was performed.

In addition, the licensee stated that the analysis included finite element modeling of an equivalent single cell, which has the structural characteristics of a submerged rack assembly, where the impact behavior between the fuel assembly and cell was represented by a three-dimensional dynamic gap element. The hydrodynamic mass between the fuel assembly and cell was modeled by three-dimensional mass matrix elements. The analysis was performed for a range of friction coefficients to assess both the rocking and sliding behavior of the racks, with the highest coefficient producing the highest impact loads and rocking displacements, and the lowest coefficient producing the highest sliding displacements. From this model, acceleration time-histories were generated and a dynamic transient analysis was performed to determine the maximum seismic loads and displacements based on SSE loading. Loads were factored based on site response spectrum values to OBE for the purpose of structural evaluation. The loads obtained from the seismic analysis of the single cell model were converted into overall rack loads and combined with other loading conditions. Once these loads were determined, classical methods were used to determine the resulting stresses in the fuel rack structural members. The resulting stresses were then compared to ASME Code Section Ill, Subsection NF and Appendix XVII, 1977 Edition, Winter 1979 Addenda criteria to determine minimum stress margins associated with the structural elements of the fuel rack. The addition of the Metamic insert reduces the amount of water that can contribute to the effects of the hydrodynamic mass.

The licensee performed an evaluation to assess the impact of the inserts on the existing rack structure and SFP structure. The insert was assumed to behave similar to the fuel assembly, and, therefore, was considered as a part of the fuel assembly due to the small gaps between the cell and the fuel assembly. In addition, due to the minimal incremental weight between the insert and the fuel assembly, the inserts are not expected to have an appreciable effect on the fundamental frequency of the seismic model. Consequently, to assess their impact to fuel rack structures, it is reasonable to add their weight to the fuel assembly weight. Therefore, assuming the slight increase in weight of a fuel assembly (accounting for the rack insert contribution to the assembly weight), there remains sufficient margin in the original design as documented in Westinghouse Report WNEP-9014 to accommodate the inclusion of the Metamic inserts without impacting the structural integrity of the fuel racks. The licensee also performed a verification check for the components with the smallest safety margin from the original design (i.e., the support block to plate structural weld and cell load buckling) utilizing the original method of analysis documented in WNEP-9014. This check confirmed the above conclusion as compared to ASME Code Section 111, Subsection NF criteria. Of particular note is that the allowable stresses for SSE conditions are twice those of OBE allowable stresses and, thus, are enveloped by QBE. Therefore, the original rack design is based on the OBE stress conditions.

In its supplement dated April 13, 2018 (Reference 6), the licensee provided Revision 2 of Holtec International Report No. HI-2177590, "Licensing Report for Use of DREAM' Neutron Absorber Inserts in the Spent Fuel Pools 'A' and 'B' at Shearon Harris NPP." The report was revised to improve installation and to avoid interference issues. Based on review of the supplement, the staff did not identify any changes that would impact the technical or regulatory portions of the structural evaluations provided in the original submittal.

3.6.2 Acceptance Criteria for Rack Re-analysis Structural Evaluation The applicable loads and loading combinations considered in the seismic analysis of the rack modules and acceptance criteria are based on the OT Position Paper and SRP Section 3.8.4.

The licensee stated that the minimum factor of safety to overturn was much greater than the minimum 1.5 required by the OT Position Paper. The acceptance criteria used to assess safety margin against rack overturning under the OBE and SSE events were 1.5 and 1.1, respectively, based on Section 3.8.5-11-5 of the SRP. The acceptance criteria for stress limits on the rack structure for Level A (normal conditions), Level B (upset conditions, including OBE) and Level D (including SSE) service limits were based on the ASME Code, Section Ill, Division 1, Subsection NF and Appendix XVII, 1977 Edition, Winter 1979 Addenda.

3.6.3 Results of Rack Re-analysis and Structural Evaluation In the original Westinghouse analysis, rack stability was evaluated. The analysis determined that the minimum factor of safety to overturn was much greater than the minimum 1.5 for both OBE and SSE loading conditions required by the NRC OT Position Paper and SRP Section 3.8.5-11-5. In evaluating the impact of the inserts, it was assumed the addition of the inserts would have a negligible effect on the rack overturn analysis due to the large margin of safety in the original analysis. The licensee performed a verification check on rack overturning utilizing the original method of analysis documented in the Westinghouse Report WNEP-9014.

This check determined that the minimum factor of safety to overturn remains significantly large, thus still yielding acceptable results. The original Westinghouse analysis also considered rack sliding and deflection during seismic events. Based on the minimal rack displacement and deflection compared to the available rack-to-rack distance and rack-to-wall distance, the addition of the inserts will not result in rack-to-rack or rack-to-wall impact. Also, any decrease in rack gaps resulting in less than or equal to 75 percent of the original installation gaps post-OBE requires repositioning of racks or an analysis to determine acceptability as identified in site procedures.

Based on staff's review of Duke Energy's input, the analysis results demonstrated that the stress levels in the rack modules are well within the ASME Code limits under the load combinations specified in NRC's OT Position Paper.

3.6.4 Structural Evaluation of the Spent Fuel Pool Structure The structural adequacy of SFPs A and B, after DREAM inserts have been added to the existing Westinghouse BWR racks, considered both the fuel pool structural qualification including the liner. Section 6.3, "Fuel Pool Structural Qualification," and Section 6.4, "Pool Liner Qualification," of Attachment 4 (Revision 2 of Holtec Report Hl-2177590) to the April 13, 2018,

$Upplement (Reference 6) stated that the DREAM inserts have a negligible impact on the structural qualifications for both SFPs, since the total weight of the installed inserts is extremely small in comparison to other load contributors such as pool water inventory, and rack and fuel weight. For the liner, during a seismic event the SFP racks transmit vertical and horizontal forces to the pool liner through the support pedestals at the base of the racks; and the loads acting on the liner are proportional to the loaded weight of the racks.

Duke Energy evaluated the SFP structure to affirm the integrity of the steel liner. The existing site structural calculation of record determined that the most limiting rack design for the steel liner is an 8x9 PWR rack array. This calculation neglects the effects of buoyancy and determines a resulting bearing pad stress due to the 8x9 PWR rack array. A similar evaluation of the 11 x11 Westinghouse BWR rack array was performed to determine the effect of inserts on the steel liner that determined the bearing pad stress for the BWR rack with inserts is bounded by the PWR rack results. On this basis, the SFP steel liner can accommodate the rack inserts.

Additionally, a SFP floor slab evaluation was performed that identified that significant margin is available for bending and shear, ensuring that the SFP floor slab structural integrity is maintained. The licensee stated in Section 6.5, "Mechanical Accident Evaluation," of the Holtec report contained in Reference 6 that the accidental drop of a neutron absorber insert plus its handling tool onto the top of a SFP rack is also bounded by the existing drop analysis as the weight of the insert plus the handling tool is much less than a BWR fuel assembly.

The licensee also stated in the LAR (Reference 1) and its response to RAI 1 (Reference 4) that use of the DREAM inserts in the SFP storage racks results in a small increase in weight on the floor of the SFP structure. The weight of the DREAM inserts are negligibly small in comparison to the overall deadweight of the pool structures and their contents. Since this weight increase in comparison to the other load contributors (e.g., pool water, rack weight, fuel height) is negligibly small compared to the total static load on the SFP B pool floor, the design basis loads, load combinations, and existing structural safety margins on the SFP structure and pool liner qualification are not adversely affected by the planned installation of the DREAM inserts. The licensee concluded that the structural design bases for the existing Boraflex racks in SFPs A and B, as well as the structural qualification of the pool structures, are not adversely affected by the planned installation of Metamic neutron absorber inserts in three BWR Boraflex racks in SFP A and five in SFP B. Furthermore, the structural integrity of the Metamic inserts were found to be structurally adequate to perform their intended function under both normal and seismic conditions. Based on staff's review of the licensee's response to RAI 1, the staff considers the response adequate.

Based on the review of the LAR, its supplement, and additional information provided by the licensee in response to the staff's RAI, the staff finds the structural design bases for the existing BWR Boraflex racks in HNP SFPs A and B, as well as the structural qualification of the SFP structures, are not adversely affected by the planned installation of Metamic neutron absorber inserts.

The staff finds that the analysis results demonstrated that the stress levels in the rack modules are well within the limits of ASME Code, Section Ill, Subsection NF under the load combinations specified in NRC's OT Position Paper. Since the racks and the SFP are designed to withstand the effects of seismic and other conditions associated with normal operation without loss of capability to perform their safety functions, GDC 2 and 4 remain satisfied. Additionally, the staff finds since generally recognized codes and standards were used in the analysis (e.g., ASME Code, NRC RG 1.29, NRC GL 79-04, NUREG-0800 (SRP) Section 3.8.4, and NRC's OT Position Paper), GDC 1 remains satisfied. The staff concludes based on its evaluations of the aforementioned analyses and RAI response that the structural integrity of the inserts under normal, accident, and seismic loading conditions is adequate to perform their intended design function, and that the regulatory requirements of GDC 1, 2, and 4, continue to be met.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of North Carolina official was notified of the proposed issuance of the amendment on July 18, 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 (82 FR 57481). 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 Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) 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.

7.0 REFERENCE LIST

1. Duke Energy Progress, LLC letter HNP-17-008, Tanya M. Hamilton, Vice President, Shearon Harris Nuclear Plant Unit 1, to USNRC document control desk, re: "License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses," June 28, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML171938165).

2. Duke Energy Progress, LLC letter HNP-17-061, Tanya M. Hamilton, Vice President, Shearon Harris Nuclear Plant Unit 1, to USNRC document control desk, re: "Supplement to License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses,"

July 20, 2017 (ADAMS Accession No. ML17201A035).

3. Duke Energy Progress, LLC letter HNP-17-073, Bently K. Jones, Director, Organizational Effectiveness, Shearon Harris Nuclear Plant Unit 1, to USNRC document control desk, re:

"Supplemental Information for License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses," September 14, 2017 (ADAMS Accession No. ML17257A245).

4. Duke Energy Progress, LLC letter HNP-18-003, Tanya M. Hamilton, Vice President, Harris Nuclear Plant Unit 1, to USNRC document control desk, re: "Response to Request for Additional Information Regarding License Amendment Request for Spent Fuel Storage Pool Criticality Analyses (CAC No. MF9996; EPID L-2017-LLA-0303)," January 18, 2018 (ADAMS Accession No. ML180186974).

5. Duke Energy Progress, LLC letter HNP-18-020, Tanya M. Hamilton, Vice President, Harris Nuclear Plant, Duke Energy Progress, LLC to USNRC document control desk, re:

"Supplement to Response to Request for Additional Information Regarding License Amendment Request for Spent Fuel Storage Pool Criticality Analyses," February 16, 2018 (ADAMS Accession No. ML18047A730).

6. Duke Energy Progress, LLC letter HNP-18-039, Tanya M. Hamilton, Vice President, Harris Nuclear Plant, Duke Energy Progress, LLC to USNRC document control desk, re:

"Supplement to License Amendment Request Regarding Spent Fuel Storage Pool Criticality Analyses," April 13, 2018 (ADAMS Accession No. ML18108A106).

7. U. S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Section 3.8.4, "Other Seismic Category Structures," including Appendix D, "Guidance on Spent Fuel Pool Racks," Revision 4, September 18, 2013 (ADAMS Accession No. ML13198A258).

8. U. S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Section 9.1.1, "Criticality Safety of Fresh and Spent Fuel Storage Handling," Revision 3, March 2007 (ADAMS Accession No. ML070570006).

9. U. S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Section 9.1.2, "New and Spent Fuel Storage," Revision 4, March 23, 2007 (ADAMS Accession No. ML070550057).

10. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Section 9.1.3, "Spent Fuel Pool Cooling and Cleanup System," Revision 2, March 23, 2007 (ADAMS Accession No. ML063190013)

11. Nuclear Energy Institute letter, "Submittal of NEI 16-03, Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools, Revision 0, dated August 2016," May 26, 2017 (ADAMS Accession No. ML17263A133)*

12. Letter from D. C. Morey, Chief, Licensing Processes Branch, Division of Licensing Projects, USNRC, to K. Cummings, Senior Project Manager, Used Fuel Programs, Nuclear Energy Institute, "Verification Letter of the Approval Version of the Nuclear Energy Institute Topical Report NEI 16-03-A, Revision 0, 'Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools, Revision O,"' October 5, 2017 (ADAMS Accession No. ML 17262AOOO)*

13. NRC Memorandum from L. Kopp to T. Collins, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants, August 19, 1998 (ADAMS Accession No. ML003728001 ).

14. U.S. Nuclear Regulatory Commission, Division of Safety Systems Interim Staff Guidance DSS-ISG-2010-01, Rev. 0, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools," October 13, 2011 (ADAMS Accession No. ML110620086).

15. U. S. Nuclear Regulatory Commission OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications, dated April 14, 1978 (ADAMS Accession No. ML031280383).

16. NRC letter from B. Mozafari, Senior Project Manager, Plant Licensing 111-2 and Planning and Analysis Branch, Division of Operating Reactor Licensing, US NRC to M. J. Pacilio, Senior Vice President, Exelon Generation Company, LLC, re: "Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments Regarding NETCO Inserts (TAC Nos. MF2489 and MF2490) (RS-13-148)," December 31, 2014 (ADAMS Accession No. ML14346A306).

17. J.C. Dean, R.W. Tayloe, Jr., "Guide for Validation of Nuclear Criticality Safety Calculational Methodology," NUREG/CR-6698, U.S. Nuclear Regulatory Commission, Science Applications International Corporation, January 2001 (ADAMS Accession No. ML050250061 ).

Principal Contributors: Scott T. Krepel Michael R. Breach Robert L. Pettis Alexander N. Chereskin Date: October 22, 2018

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 167 REGARDING (CAC MF9996; EPID 2017-LLA-0303)

DATED OCTOBER 22, 2108 DISTRIBUTION:

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