ML20212L594

Issuance of Amendment No. 179 Regarding Departure from Nucleate Boiling Ratio Safety Limit to Address Transition to New Fuel Type

ML20212L594
Person / Time
Site:	Harris
Issue date:	09/29/2020
From:	Tanya Hood Plant Licensing Branch II
To:	Maza K Duke Energy Progress
Hood T, NRR/DORL/LPL2-2
References
EPID L-2019-LLA-0076, RA-19-0017
Download: ML20212L594 (29)

Text

September 29, 2020 Ms. Kim Maza Site Vice President Shearon Harris Nuclear Power Plant Mail Code NHP01 5413 Shearon Harris Road New Hill, NC 27562-9300

SUBJECT:

SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 179 REGARDING DEPARTURE FROM NUCLEATE BOILING RATIO SAFETY LIMIT TO ADDRESS TRANSITION TO NEW FUEL TYPE (EPID L-2019-LLA-0076)

Dear Ms. Maza:

The U.S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 179 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant (Harris), Unit 1. This amendment revises Technical Specifications (TSs) requirements in response to your application dated April 10, 2019, as supplemented by letters dated June 6, 2019 and December 20, 2019.

The amendment modifies TS 2.1.1.a to add the departure from nucleate boiling ratio safety limit associated with the transition from the high thermal performance fuel to a fuel assembly design with characteristics similar to the GAIA fuel design using the ORFEO-GAIA correlation methodology. The amendment would also add Appendix J for DPC-NE-2005-P, Thermal-Hydraulic Statistical Core Design Methodology, to address the applicability of the ORFEO-GAIA critical heat flux correlation methodology to a fuel assembly design with characteristics similar to the GAIA fuel design at Harris. The TSs would also be revised with minor formatting editorial adjustments to the impacted pages. This safety evaluation does not approve the use of GAIA fuel nor does it approve the use of the generically approved GAIA topical report in ANP-10342P-A, GAIA Fuel Assembly Mechanical Design.

K. Maza A copy of our related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions regular biweekly Federal Register notice.

Sincerely,

/RA/

Tanya E. Hood, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-400

Enclosures:

1. Amendment No.179 to NPF-63

2. Safety Evaluation cc: Listserv

DUKE ENERGY PROGRESS, LLC DOCKET NO. 50-400 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 179 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 April 10, 2019, as supplemented by letters dated June 6, 2019 and December 20, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

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. 179, 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 prior to the startup of Cycle 24.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Undine S. Undine S. Shoop Date: 2020.09.29 Shoop 12:04:40 -04'00' Undine Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed License and Technical Specifications Date of Issuance: September 29, 2020

ATTACHMENT TO LICENSE AMENDMENT NO. 179 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 Insert Page 4 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 Insert 2-1 2-1 6-24a 6-24a 6-24b 6-24b 6-24c 6-24c 6-24d 6-24d

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 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. 179, 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 Duke Energy Progress, LLC. shall comply with the antitrust conditions delineated in Appendix C to this license.

(4) 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.

(5) 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.

1The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

• On April 29, 2013, the name of Carolina Power & Light Company (CP&L) was changed to Duke Energy Progress, Inc. On August 1, 2015, the name Duke Energy Progress, Inc. was changed to Duke Energy Progress, LLC.

Renewed License No. NPF-63 Amendment No. 179

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:

a. The departure from nucleate boiling ratio (DNBR) shall be maintained 1.141 for the HTP DNB correlation for HTP fuel and 1.12 for the ORFEO-GAIA DNB correlation for GAIA fuel.

b. The peak centerline temperature shall be maintained < [4901 - (1.37 x 10-3 x (Burnup, MWD/MTU))] °F.

APPLICABILITY: MODES 1 and 2.

ACTION:

If Safety Limit 2.1.1 is violated, restore compliance and be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig except during hydrostatic testing.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4, and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

SHEARON HARRIS - UNIT 1 2-1 Amendment No. 179

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

c. XN-NF-82-21(P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," approved version as specified in the COLR.

(Methodology for Specification 2.1.1 - Reactor Core Safety Limits, 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - DNB Parameters).

d. XN-75-32(P)(A), "Computational Procedure for Evaluating Fuel Rod Bowing,"

approved version as specified in the COLR.

(Methodology for Specification 2.1.1 - Reactor Core Safety Limits, 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - DNB Parameters).

e. EMF-84-093(P)(A), "Steam Line Break Methodology for PWRs," approved version as specified in the COLR.

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - DNB Parameters).

f. ANP-3011(P), Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis, Revision 1, as approved by NRC Safety Evaluation dated May 30, 2012.

(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

g. XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," approved version as specified in the COLR.

(Methodology for Specification 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 -

Control Bank Insertion Limits, and 3.2.2 - Heat Flux Hot Channel Factor).

h. ANF-88-054(P)(A), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2," approved version as specified in the COLR.

(Methodology for Specification 3.2.1 - Axial Flux Difference, and 3.2.2 - Heat Flux Hot Channel Factor).

i. EMF-92-081(P)(A), "Statistical Setpoint/Transient Methodology for Westinghouse Type Reactors," approved version as specified in the COLR.

(Methodology for Specification 2.1.1 - Reactor Core Safety Limits, 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - DNB Parameters).

j. EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," approved version as specified in the COLR.

(Methodology for Specification 2.1.1 - Reactor Core Safety Limits, 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - DNB Parameters).

SHEARON HARRIS - UNIT 1 6-24a Amendment No. 179

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

k. BAW-10240(P)(A), Incorporation of M5 Properties in Framatome ANP Approved Methods.

(Methodology for Specification 2.1.1 - Reactor Core Safety Limits, 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.1.1.2 - SHUTDOWN MARGIN - MODES 3, 4 and 5, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, 3.2.5 - DNB Parameters, and 3.9.1 - Boron Concentration).

l. EMF-96-029(P)(A), "Reactor Analysis Systems for PWRs," approved version as specified in the COLR.

(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES 3, 4 and 5, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 -

Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.9.1 - Boron Concentration).

m. EMF-2328(P)(A) PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, approved version as specified in the COLR.

(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

n. EMF-2310(P)(A), SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors, approved version as specified in the COLR.

(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - DNB Parameters).

o. Mechanical Design Methodologies XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," approved version as specified in the COLR.

ANF-81-58(P)(A), "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," approved version as specified in the COLR.

XN-NF-82-06(P)(A), "Qualification of Exxon Nuclear Fuel for Extended Burnup,"

approved version as specified in the COLR.

ANF-88-133(P)(A), "Qualification of Advanced Nuclear Fuels' PWR Design Methodology for Rod Burnups of 62 GWd/MTU," approved version as specified in the COLR.

XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results," approved version as specified in the COLR.

EMF-92-116(P)(A), "Generic Mechanical Design Criteria for PWR Fuel Designs,"

approved version as specified in the COLR.

BAW-10231P-A, COPERNIC Fuel Rod Design Computer Code, approved version as specified in the COLR.

SHEARON HARRIS - UNIT 1 6-24b Amendment No. 179

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

(Methodologies for Specification 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

p. DPC-NE-2005-P-A, Thermal-Hydraulic Statistical Core Design Methodology, approved version as specified in the COLR.

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor)

q. DPC-NE-1008-P-A, Nuclear Design Methodology Using CASMO-5/SIMULATE-3 for Westinghouse Reactors, as approved by NRC Safety Evaluation dated May 18, 2017.

(Methodology for Specification 3.1.1.2 - SHUTDOWN MARGIN - MODES 3, 4, and 5, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.9.1 - Boron Concentration).

r. DPC-NF-2010-A, Nuclear Physics Methodology for Reload Design, as approved by NRC Safety Evaluation dated May 18, 2017.

(Methodology for Specifications 3.1.1.1 - SHUTDOWN MARGIN - MODES 1 and 2, 3.1.1.2 - SHUTDOWN MARGIN - MODES 3, 4, and 5, 3.1.1.3 - Moderator Temperature Coefficient, 3.1.2.5 - Borated Water Source - Shutdown, 3.1.2.6 -

Borated Water Sources - Operating, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.5.1 - ECCS Accumulators - Cold Leg Injection, 3.5.4 - ECCS Refueling Water Storage Tank, and 3.9.1 - Boron Concentration).

s. DPC-NE-2011-P-A, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors as approved by NRC Safety Evaluation dated May 18, 2017.

(Methodology for Specification 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 -

Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).

t. DPC-NE-3008-P-A, Thermal-Hydraulic Models for Transient Analysis, as approved by NRC Safety Evaluation dated April 10, 2018.

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor).

u. DPC-NE-3009-P-A, FSAR / UFSAR Chapter 15 Transient Analysis Methodology, as approved by NRC Safety Evaluation dated April 10, 2018.

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor).

SHEARON HARRIS - UNIT 1 6-24c Amendment No. 179

ADMINISTRATIVE CONTROLS 6.9.1.6 CORE OPERATING LIMITS REPORT (Continued)

v. ANP-10341P-A, The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations, approved version as specified in the COLR.

(Methodology for Specification 2.1.1 - Reactor Core Safety Limits, 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - DNB Parameters).

6.9.1.6.3 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.6.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk, with copies to the Regional Administrator and Resident Inspector.

6.9.1.7 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with Specification 6.8.4.l. The report shall include:

a. The scope of inspections performed on each SG,

b. Degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each degradation mechanism,

f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator, and

g. The results of condition monitoring, including the results of tube pulls and in-situ testing.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC in accordance with 10 CFR 50.4 within the time period specified for each report.

6.10 DELETED (PAGE 6-25 DELETED By Amendment No.92)

SHEARON HARRIS - UNIT 1 6-24d Amendment No. 179

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 179 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 letter dated April 10, 20191, as supplemented by letters dated June 6, 20192 and December 20, 20193, Duke Energy Progress, LLC (the licensee) submitted a license amendment request (LAR) for changes to the Shearon Harris Nuclear Power Plant, Unit 1 (Harris), Technical Specifications (TSs). The requested changes would modify TS 2.1.1.a to add the departure from nucleate boiling ratio (DNBR) safety limit associated with the transition from the high thermal performance fuel to a fuel assembly design with characteristics similar to the GAIA fuel design using the ORFEO-GAIA correlation methodology. In addition, TS 6.9.1.6.2 would be revised to include the U.S. Nuclear Regulatory Commission (NRC or the Commission) final safety evaluation report (SER)4 approving topical report, ANP-10341-P-A, The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations, Revision 0,5 as a Core Operating Limits Report (COLR) reference for the NRC-approved departure from nucleate boiling correlation associated with the GAIA fuel design for applicable fuel specifications and characteristics. The GAIA fuel design is limited to the parameters outlined in Attachment 7 to the LAR in Appendix J, Harris Plant Specific Data for GAIA Fuel Application of the ORFEO-GAIA Correlation to the GAIA Fuel Design, to DPC-NE-2005-P, Thermal-Hydraulic Statistical Core Design Methodology, Revision 6. This safety evaluation does not approve the use of GAIA fuel nor does it approve the use of the generically approved GAIA topical report in ANP-10342-P-A, GAIA Fuel Assembly Mechanical Design.6 The amendment would add Appendix J to DPC-NE-2005-P which includes Harris plant specific data of fuel specifications and characteristics related to GAIA fuel using the ORFEO-GAIA critical heat flux correlation. Revision 6 discussed the methodology to address the applicability of the ORFEO-GAIA critical heat flux use at Harris to perform the statistical core design analysis in-house when applying the ORFEO-GAIA critical heat flux correlation at Harris. The TSs would 1

Agencywide Documents Access and Management System (ADAMS) Accession No. ML19100A442 2

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ADAMS Accession No. ML18236A371 5

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ADAMS Accession No. ML19309D913 Enclosure 2

also be revised with minor formatting editorial adjustments to the impacted pages. The NRC staff's initial proposed no significant hazards consideration determination was published in the Federal Register on March 30, 2020 (85 FR 17601).

2.0 REGULATORY EVALUATION

The NRC staff considered the following regulatory requirements, guidance, and licensing and design-basis information during its review of the proposed change.

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

Section 50.36(c)(1)(i)(A) to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, states, in part, that safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity.

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

GDC 10, Reactor design, requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margins to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any conditions of normal operation, including the effects of anticipated operational occurrences. In pressurized water reactors (PWRs), the DNBR safety limit is established to assure compliance with SAFDLs. The DNBR safety limit is the DNBR which corresponds to a 95 percent probability at a 95 percent confidence level that departure from nucleate boiling will not occur.

NRC Generic Letter 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, dated October 4, 19887, outlined a process that a licensee could use to move cycle-specific parameters from the plant-specific TSs to a licensee-controlled document entitled, Core Operating Limits Report (COLR). The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC and documented in the topical report(s). A necessary element of that process was that a licensee must include in applicable TS section the specific topical report(s) with the topical report number, title, and date, or the NRC staff's SER for a plant-specific methodology by NRC letter and date.

Technical Specification Task Force Traveler (TSTF)-363, Revise Topical Report References in ITS [Improved Technical Specification] 5.6.5, COLR (Core Operating Limits Report), dated 7

ADAMS Accession No. ML031150407

August 4, 2011, made a slight modification to Generic Letter 88-16 guidance8 and allowed licensees to omit topical report revision numbers and dates from the TS list of references.

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensees application to determine whether the proposed TS changes are consistent with the regulations, licensing and design basis information, and regulatory guidance discussed in Section 2 of this safety evaluation.

The calculations of DNBRs for design-bases events at PWRs were typically based on two methodologies including: (1) the deterministic methodology that assumed that key input parameters to the core thermal-hydraulic code were simultaneously at their worst level of uncertainty; and (2) the statistical core design methodology that obtained the DNBR limit from statistical analysis of a series of computations as a result of propagation of uncertainties about a statepoint and associated distribution of the DNBR values.

The licensee adopted the statistical core design methodology for DNBR calculations. The statistical core design methodology is documented in the NRC-approved topical report, DPC-NE-2005-P-A, Duke Power Company Thermal-Hydraulic Statistical Core Design Methodology, Revision 3.9 Subsequent revisions added appendices containing plant-specific data that would apply the statistical core design methodology to the licensees individual plants and fuel types.

The proposed TS changes would also add a safety limit of the DNBR being greater than or equal to 1.12 for the ORFEO-GAIA critical heat flux correlation; and TS 6.9.1.6.2 with an addition of topical report, ANP-10341-A, as a COLR reference for the NRC-approved ORFEO-GAIA and ORFEO-NMGRID critical heat flux correlations.

3.1 Revision 6 to DPC-NE-2005 The proposed Revision 6 to DPC-NE-2005-P would add Appendix J, which provides information that includes: application of the ORFEO-GAIA correlation to the GAIA fuel design at Harris; the VIPRE-01 thermal-hydraulic code and model used in applying the statistical core design methodology; the ORFEO-GAIA and ORFEO-NMGRID critical heat flux correlations; and the statepoints, key parameters and uncertainties used in the statistical core design analysis.

Appendix J also includes a discussion of calculation the statistical design limit (SDL) used in safety analyses and allows the licensee to perform the statistical core thermal-hydraulic analysis when applying the ORFEO-GAIA critical heat flux correlation to GAIA fuel at Harris.

The proposed Appendix J indicated that the statistical core design analyses for Harris would be performed with the VIPRE-01 code and the ORFEO-GAIA critical heat flux correlation, both of which were previously reviewed and approved by the NRC. In addition, Revision 6 to DPC-NE-2005-P would be used by the licensee to support operation at Harris with transition from the high thermal performance fuel to a fuel assembly design with characteristics similar to the GAIA fuel design using the ORFEO-GAIA correlation methodology starting in Cycle 24 scheduled for April 2021 and continuing to full-core GAIA fuel in Cycle 26.

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3.1.1 Conditions for Use of DPC-NE-2005 Methodology DPC-NE-2005-P-A described the statistical core design methodology that would allow for computation of a statistical DNBR limit, which explicitly accounted for the uncertainties of the key plant and fuel parameters, including the critical heat flux correlation. The thermal-hydraulic analyses were then performed at nominal plant conditions and compared to the statistical DNBR limit, resulting in a more precise quantification of the margin to departure from nucleate boiling.

The NRC staff's SER on DPC-NE-2005-P-A contained the following three conditions:

1. The statistical core design methodology described in DPC-NE-2005-P-A was acceptable for use in any pressurized-water reactor fuel or reactor, provided that the VIPRE-01 methodology is approved with the use of the core model and correlations including the critical heat flux correlation subject to the conditions in the VIPRE-01 SER.10

2. The licensee must demonstrate that its use of specific uncertainties and distributions are based upon plant data and its selection of statepoints used for generating the SDL are appropriate.

3. The statistical core design methodology was approved only for use in the licensees plants.

Condition 3 has been satisfied because Harris is one of the licensees plants. The NRC staffs evaluation of the licensees compliance with the first two conditions are discussed in subsections 3.1.2 and 3.1.3 of this SE, respectively.

3.1.2 VIPRE-01 Code Model (Condition 1 in the SER Approving DPC-NE-2005-P-A)

Condition 1 in the SER approving DPC-NE-2005-P-A, Revision 0, stated that VIPRE-01 must be approved with the use of the core model and critical heat flux correlations subject to the conditions of the VIPRE-01 SER. There are 5 conditions listed in the VIPRE-01 SER. Only Conditions 2 and 3 are applicable to this application, since the two conditions addressed the limitations on the use of the fuel-specific critical heat flux correlations in VIPRE-01. Condition 2, which allowed the use of previously approved critical heat flux correlations within VIPRE-01 provided that the analysis resulted in a correlation limit that was conservative or the same; and Condition 3, which required discussion and justification of modeling choices for each licensee's application of VIPRE-01 are discussed in the following subsections.

3.1.2.1 DNBR Safety Limit Based on ORFEO-GAIA Critical Heat Flux Correlation and VIPRE-01 (Condition 2 in the SER Approving VIPRE-01)

The NRC staff's condition on the use of DPC-NE-2005-P-A required that the critical heat flux correlation be approved for use with VIPRE-01. The ORFEO-GAIA correlation was originally implemented and approved for use with the COBRA-FLX code and was not previously approved for use with VIPRE-01. However, Condition 2 of the NRC staff's SER discussed in subsection 3.1.1 above for VIPRE-01 allowed the implementation of new critical heat flux correlations.

The licensee provided a discussion in the LAR of its DNBR safety limit analysis using the VIPRE-01 code with the ORFEO-GAIA critical heat flux correlation, along with the ORFEM-10 ADAMS Accession No. ML18033A075

NMGRID correlation for use in the non-mixing grid region of the fuel. The analysis was based on the same methodology and ORFEO-GAIA critical heat flux test data in the NRC-approved topical report ANP-10341-P-A to determine the DNBR safety limit for the statistical core design analysis.

The results of the DNBR safety limit analysis indicated that the measured to predicted critical heat flux in Figure J-3 of the LAR provided reasonable predicted critical heat flux values for the correlated data. The results in Figures J-4 through J-6 of the LAR showed no bias with respect to important measured thermal hydraulic parameters such as local mass flux, pressure, and local quality. The average measured to predicted critical heat flux and the standard deviation of the data correlation presented in Table J-2 of the LAR were both higher than that for the original implementation of the GAIA correlation within COBRA-FLX discussed in ANP-10341-P-A. With consideration of the average measured to predicted critical heat flux and statistical treatment of the standard deviation, the calculated DNBR safety limit was 1.10. The licensee added a bias of 0.01 to the calculated DNBR safety limit, resulting in a conservative DNBR safety limit of 1.11.

The inclusion of a bias of 0.01 complies with Condition 4 of the ANP-10341-P-A for the ORFEO-GAIA correlation to account for variations between the tested fuel assembly and the production fuel assembly (i.e., the fuel assembly installed at Harris) and is acceptable. As shown in Table J-2 of the LAR, the calculated conservative DNBR safety limit of 1.11 was 0.01 less than the value of 1.12 for the original COBRA-FLX prediction. The difference of 0.01 in the prediction was caused by using VIPRE-01 versus COBRA-FLX. To account for the difference predicted by using different codes, the licensee proposed to use the higher DNBR safety limit correlation limit of 1.12 for VIPRE-01 analyses with the ORFEO-GAIA correlation.

The applicable ranges of the fuel and thermal-hydraulic conditions for use of the DNBR safety limit of 1.12 were as follows:

The design of the fuel used in the Harris core was consistent with the fuel specifications and characteristics outlined in Table J-1 of the LAR.

The fuel specifications in Table J-1 of the LAR included the dimensions of the fuel rod outer diameter, thimble tube diameter, instrument guide tube diameter, fuel rod pith, fuel assembly pitch, active fuel length, and fuel rod length.

The fuel characteristics in Table J-1 of the LAR included the specific material, quantity, position and type of the applicable grids, fuel rods, control rod guide tubes, and instrumentation guide tubes.

The applicable ranges of thermohydraulic conditions were within the critical heat flux test conditions listed in Table J-3 of the LAR for the ranges of the pressure, mass flux and thermodynamic quality.

3.1.2.1.1 SER Conditions and Limitations on Use of the ORFEO-GAIA Correlation The NRC staffs final safety evaluation approving the ORFEO-GAIA correlation for topical report ANP-10341-P-A contained three conditions and two limitations. The three conditions are as follows:

Condition 1 required that the inlet subcooling be greater than 0 degrees to ensure that the burnout length would be limited to the fuel region.

Condition 2 required that for the ORFEM-NMGRID correlation, the users must confirm that the reload calculations performed for set points, anticipated operational occurrences, and accidents are far removed from the specific subregion. If the calculations were not far removed from this subregion, then the users must quantify the additional uncertainty of the subregion and apply that increased uncertainty in the analysis.

Condition 3 required additional analysis in quantifying the uncertainty in a low-quality region (i.e., equilibrium qualities below -0.1) when the impact of the low-quality on the limiting minimum DNBR values was not negligibly small. Limiting minimum DNBR was defined as the scenario in which the event was approaching the design limit.

The two limitations are as follows.

Limitation 1 - ORFEO-GAIA was approved for use in predicting the critical heat flux downstream of GAIA and intermediate GAIA mixer mixing grids in GAIA fuel. This prediction must be made in the subchannel code COBRA-FLX with the modeling option as specified in Table 5.1 of the topical report with a design limit of 1.12 over the application domain specified in Table 2-2 of the initial submittal of the topical report. The approved design limit contained a bias of 0.01 which the NRC staff believed was necessary to account for variations between the tested fuel assembly and the production fuel assembly which would be used in the reactor.

Limitation 2 - ORFEO-NMGRID was approved for use in predicting the critical heat flux downstream of W 17x17 HMP non-mixing grids and GAIA and intermediate GAIA mixer mixing grids in GAIA fuel. This prediction must be made in the subchannel code COBRA-FLX with the modeling option as specified in Table 5.1 of the topical report with a design limit of 1.15 over the application domain specified in Table 2-5 of the initial submittal of the topical report.

The NRC staff reviewed the DNBR safety limit analysis discussed above and found that:

1. The licensee indicated that it would adhere to Condition 1 and Condition 3. With respect to Condition 2, the licensee indicated that it would use the ORFEO-NMGRID correlation in the non-mixing span only and the ORFEO-GAIA correlation for the remainder of the assembly, similar to how the BWU-N critical heat flux correlation was used in the non-mixing span with the high thermal performance critical heat flux correlation. The licensee further indicated that at Harris, departure from nucleate boiling was not observed in the non-mixing span region of the fuel assembly and would not be expected in the future applications.

2. The licensees analysis was performed using the acceptable VIPRE-01 code and critical heat flux correlations based on the same methodology and data base documented in NRC-approved ANP-10341-P-A with implementation of the GAIA correlation in COBRA-FLX.

3. The results of the analysis showed that the predicted to measured critical heat flux in Figures J-3 through J-6 of the LAR provided reasonable predicted critical heat flux values for the correlated data.

4. The determined DNBR safety limit of 1.12 adequately includes a bias of 0.01 in meeting Limitation 1 of the ORFEO-GAIA correlation SER to account for variations between the tested fuel assembly and the production fuel assembly and a bias of 0.01 to account for the difference in prediction by using VIPRE-01 versus COBRA-FLX.

5. With respect to Limitation 2, the licensee indicated that it would use the ORFEO-NMGRID correlation in the non-mixing span only and the ORFEO-GAIA correlation for the remainder of the assembly, like how the BWU-N CHF correlation was used in the non-mixing span with the high thermal performance critical heat flux correlation.

6. The applicable fuel specifications and characteristics and the ranges of thermal hydraulic conditions to use the GAIA critical heat flux correlation and its DNBR safety limit of 1.12 were consistent with Table 2-2 and Table 2-3 in the NRC-approved ANP-10341-P-A for use of the ORFEO-GAIA critical heat flux correlation.

In addition, the NRC staff reviewed the licensees compliance with the three SER conditions and two SER limitations above and concluded that the licensee satisfactorily met the required SER conditions and limitations. Therefore, the NRC staff determined that the proposed critical heat flux correlations complied satisfactorily with Condition 2 in the VIRE-01 SER, regarding the licensees appropriate justification of critical heat flux correlations and are acceptable for licensing applications.

3.1.2.2 The Licensees Justification of Modeling Choices for Its Application of VIPRE-01 (Condition 3 in the SER Approving VIPRE-01)

Condition 3 of the NRC staff's SER for VIPRE-01 required justification of modeling choices for each licensee's application of VIPRE-01. The licensee proposed to use the VIPRE-01 core model shown in Figure J-1 of the LAR for performing the statistical core design analysis. The model was based on the high thermal performance model in Figure I-1 of Revision 5 to DPC-NE-2005-P-A11 for high thermal performance fuel at Harris with the following changes:

1. The fuel dimensions were updated to be consistent with the fuel specifications and characteristics listed in Table J-1 of the LAR. The NRC staff found that the updated dimensions were acceptable, since they represented a fuel assembly design with characteristics similar to GAIA fuel at Harris.

2. The model orientation was rotated 90 degrees clockwise around the assembly center instrument tube to put the guide tube type 1 channel in the center for the model. The NRC staff found that the proposed model orientation was acceptable, since the guide tube type 1 channel would have a higher flow loss coefficient and thus a lower minimum DNBR value compared to the guide tube type 0. With the type 1 channel in the center of the VIPRE-01 model, the statistical core design analysis would calculate a lower minimum DNBR and was conservative.

11 ADAMS Accession No. ML15075A221

3. The radial power peaking distribution was specified to be consistent with the distribution in Figure J-2 of the LAR and would be reduced to below the maximum allowable radial peaking (MARP) limits if the cycle-specific statistical core design analysis showed it exceeded the limits.

The NRC staff found that with the VIPRE-01 core model in Figure J-2 rotated 90 degrees back to the original core model applying to the high thermal performance fuel, the radial power peaking distribution was identical to that shown in Figure I-2 of the NRC-approved Revision 5 to DPC-NE-2005-P-A for high thermal performance fuel at Harris. The licensee indicated in Section 5.3 of DPC-NE-3008, Thermal-Hydraulic Models for Transient Analysis,12 that the radial power peaking distribution was intended to be a bounding and conservative radial power peaking distribution for fuel at Harris. To ensure that the licensee clarified how the radial power peaking distribution used in the statistical core design analysis would be bounding throughout the fuel transition from the high thermal performance fuel to a fuel with characteristics similar to the GAIA fuel design using the ORFEO-GAIA correlation methodology, the NRC staff requested additional information in a letter dated November 27, 2019.13 In its response to the NRC staffs request for additional information (RAI)-1, dated December 20, 2019,14 the licensee clarified that the reference power distribution shown in Figure J-2 of the LAR was a flat radial power peaking distribution using a specified proprietary limiting pin radial peaking factor. It was designed for the reference power distribution model such that with the pin powers for all rods surrounding subchannel 5 setting at the same proprietary limiting pin radial peaking factor, the subchannel 5 would yield the most limiting minimum DNBR. Subchannel 5 was a guide tube subchannel surrounded by other subchannels and not any lumped channels. The reference power distribution was tested using various axial peaks and shapes to ensure that subchannel 5 remained the most limiting subchannel. In addition, the licensees process for calculating departure from nucleate boiling margin used MARP limits described in DCP-2011-P, Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors, Revision 2,15 and involved scaling the radial power peaking distribution up and down to obtain a target minimum DNBR value for a given conditions.

The NRC staff found that the combination of the use of the flat radial power peaking distribution surrounding the limiting guide tube channel and MARP calculation process would provide reasonable assurance that the reference radial power peaking distribution would be bounding and conservative for the minimum DNBR calculation for both high thermal performance and ORFEO-GAIA correlation for GAIA fuel throughout the fuel transition, and therefore, the NRC staff determined that the radial power peaking distribution used in the statistical core design analysis was acceptable.

4. The mixing coefficient between subchannels was changed to the GAIA fuel value. The NRC staff found that the value used for the mixing coefficient was acceptable, since it was the same as that specified in Table 5.1 of the NRC-approved topical report ANP-10341-P-A documenting the application of the critical heat flux correlation to GAIA fuel design. Since the VIPRE-01 core model was based on the model approved previously by the NRC staff for high thermal performance fuel with the updated fuel 12 ADAMS Accession No. ML15323A351 13 ADAMS Accession No. ML19331A400 14 ADAMS Accession No. ML19354B380 15 ADAMS Accession No. ML16125A420

dimensions, model orientation, radial power distribution, and mixing coefficient between subchannels representing the ORFEO-GAIA correlation for GAIA fuel design at Harris, the NRC staff determined that the proposed VIPRE-01 model complied satisfactorily with VIPRE-01 SER Condition 3 in the VIPRE-01 SER, and thus was acceptable for use to perform the SCD analysis for the ORFEO-GAIA correlation for GAIA fuel design at Harris.

As discussed above, the NRC staff determined that the proposed use of the ORFEO-GAIA correlation in VIPRE-01 as described in the LAR and the associated DNBR safety limit of 1.12 satisfied Conditions 2 and 3 in the VIPRE-01 SER. Therefore, the NRC staff concludes that the licensees use of the ORFEO-GAIA correlation in VIPRE-01 is acceptable for performing the statistical core design analysis with GAIA fuel design at Harris. Although the licensee did not specifically discuss the option in the Table 5.1 of the topical report for VIPRE-01, corresponding to COBRA-FLX, the NRC staff evaluated this portion of the LAR and found that the licensees proposed use of ORFEO-GAIA with an acceptable VIPRE-01 model would meet the intent of Limitation 1 and was acceptable.

3.1.3 Statistical Design Limit (SDL) for GAIA Fuel As discussed in Section 3.1.1, Condition 2 in the NRC staffs SER approving the DPC-NE-2005-P-A methodology required that the selection of statepoints used for generating the SDL must be justified to be appropriate when it is applied to the licensees plants that would adopt the DPC-NE-2005-P-A methodology.

3.1.3.1 Statepoints Used in the SDL Determination Section 3.3.1 of the LAR indicated that the licensee used the statistical core design methodology contained in the NRC-approved topical report DPC-NE-2005-P-A to determine the SDL for fuel characteristics similar to the GAIA fuel design using the ORFEO-GAIA correlation methodology at Harris. The fifty statepoints used for generating the SDL were identified in Table J-4 of Appendix J to the LAR, which listed for each statepoint specific values of the core exit pressure, core inlet temperature, core inlet flow, core power, axial power peak magnitude and location, and radial power peak. These statepoints represented the range of conditions to which the SDL limit would be applied. Since the statepoints were determined based on the NRC-approved methodology in DPC-NE-2005-P-A to conservatively bound the range of conditions of the applicable analysis of record for Updated Final Safety Analysis Report (UFSAR), Chapter 15 non-loss-of-coolant events, the NRC staff determined that the licensees selection of statepoints complied with Condition 2 on the use of DPC-NE-2005-P-A, discussed in Section 3.1.1 of this SER, (requiring an appropriate justification for the selection of statepoints), and concludes that the proposed changes are acceptable.

3.1.3.2 Key Parameter Uncertainties Used in the Statistical Core Design Analyses Condition 2 discussed in Section 3.1.1 on the use of DPC-NE-2005-P-A also required the use of specific uncertainties and distributions to be justified on a plant-specific basis. The licensee presented in Table J-5 of Appendix J to the LAR, a list of the parameters for uncertainty consideration, the value of the uncertainty used, the distribution associated with the uncertainty of each parameter, and justification for each statistically treated parameter and associated uncertainty. The statistically treated parameters were: core power, coolant flow measurement, bypass flow, core exit pressure, core inlet temperature, radial power measurements, radial power engineering uncertainties, axial power peak prediction uncertainty of the physics code,

axial peak location uncertainty, DNBR correlation uncertainty, and thermal-hydraulic code uncertainties.

The NRC staff found that the values and distributions of the uncertainties in Table J-5 for eight of these 11 parameters were the same as those listed in Table I-5 of DPC-NE-2005-P-A, Revision 5, for use of the SDL analysis based on the high thermal performance fuel at Harris.

These eight parameters were: core power, coolant flow measurement, bypass flow, radial power measurements, radial power engineering uncertainties, axial power peak prediction uncertainty (from the physics code), axial peak location uncertainty, and thermal-hydraulic code uncertainties. However, the values of uncertainties in Table J-5 were changed from the corresponding values in Table I-5 for three parameters. These were: (a) core exit pressure, (b) core inlet temperature, and (c) DNBR correlation uncertainty. The difference of DNBR correlation uncertainty, was caused by using the ORFEO- critical heat flux correlation versus high thermal performance critical heat flux correlation. To ensure that the licensee justified the changes of the uncertainties for the core exit pressure and core inlet temperature, the NRC staff requested additional information in a letter dated November 27, 2019.

In its response to RAI-2, dated December 20, 2019, the licensee indicated that the uncertainties listed in Table I-5 of DPC-NE-2005-P-A, Revision 5, were based on exiting calculations at that time. The changes in Table J-5 of Appendix J to the LAR were not specific to GAIA fuel but were updated to reflect recent revisions to uncertainty calculations. For the core exit pressure and core inlet temperature, the SDL accounted for instrument uncertainty and an operational allowance used by the operators at the plant. The operational allowance was the difference between the +50 psi/+5.00F uncertainty assumed in the SDL analysis and the indication uncertainty. The indication uncertainty was dependent on how the indications are processed and the number of channels available. Table 2-1 and Table 2-2 in the RAI-2 response showed the values of the pressure and temperature uncertainties, respectively, for the control board indicators and the computer channels in three different operational modes, including using one, two, or three board indicators (or computer channels).

The NRC staff evaluated the uncertainty of +50 psi/+5.00F used in SDL bounded instrument uncertainties shown in Table 2-1 and Table 2-2 for core exit pressure and core inlet temperature and found that it would provide operational margin. Therefore, the NRC staff concludes that the proposed use of an uncertainty of +50 psi/+5.0 0F in the SDL analysis is acceptable.

The licensee calculated the statistical DNBR limit for each statepoint based on the approved methodology in DPC-NE-2005-P-A, with the use of the acceptable values for the key parameter uncertainties discussed in Section 3.1.3.2 of this SER. The results for 10,000 case runs were presented in Section 2 of Table J-6 of Appendix J to the LAR which showed that the maximum statistical DNBR limit was less than 1.28. The licensee selected the value of 1.28 as the SDL for fuel characteristics similar to the GAIA fuel design using the ORFEO-GAIA correlation methodology at Harris with the applicable ranges shown in Table J-7 of Appendix J to the LAR for the thermo-hydraulic parameters and physics parameters.

The NRC staff evaluated the SDL of 1.28 and determined that it was conservative and bounded the maximum calculated statistical DNBR limits on Table J-6. The NRC staff also determined that the applicable ranges for the use of the SDL of 1.28 were acceptable, since the ranges of the thermo-hydraulic parameters were within that for the critical heat flux test conditions discussed above in Section 3.1.2.2 and physics parameter were within the maximum allowable peak limit space.

3.1.4 Conclusion for Application of DPC-NE-2005, Revision 6 As discussed in Sections 3.1.1 through 3.1.3 above, the NRC staff evaluated the licensees application of the statistical core design methodology documented in DPC-NE-2005-P, Revision 6, and determined that it appropriately satisfied all three conditions imposed in the NRC's SER for use of DPC-NE-2005-P-A, Revision 0. Therefore, the NRC staff concluded that DPC-NE-2005-P, Revision 6, is acceptable for the use of a fuel assembly design with characteristics similar to GAIA fuel design at Harris.

3.1.5 Mixed Core Penalty (high thermal performance and a fuel assembly design with characteristics similar to GAIA Fuel) 3.1.5.1 Mixed Core Penalty Derivation The licensee indicated in Section 3.4.1 of the LAR that it used MARP limits to assure that the departure from nucleate boiling design basis for UFSAR Chapter 15 non-loss-of-coolant events was met by comparing calculated total peaking factors against MARP limits. The licensee previously calculated the MARP limits for a full-core of high thermal performance fuel using the NRC-approved methodologies for reference by Harris16 and also calculated the MARP limits for a full-core of GAIA fuel design applying the same methodology used to calculate the full-core high thermal performance limits. As indicated in Table 3.2 of the LAR, the coolant flow would divert from a high thermal performance assembly into a GAIA assembly in a mixed core configuration. This flow diversion was caused by higher assembly grid form loss coefficients for high thermal performance fuel and would result in a DNBR penalty for high thermal performance fuel. The power peaking limits for a range of transients were recalculated with the mixed core model and compared to the full-core high thermal performance peaking limits, and a bounding set of penalties was derived based on the differences.

3.1.5.2 Mixed Core VIPRE-01 Models The licensee discussed in Section 3.4.2 of the LAR the nodalization schemes used in the generic mixed core VIPRE-01 model and the expanded mixed core model to determine the mixed core penalty. The generic mixed core model was the same as the previously NRC-approved high thermal performance model in Figures I-1 of Appendix I to DPC-NE-2005-P-A and GAIA model in Figure J-1 of Appendix J to the LAR, with the exception that the hot channel was assumed to be high thermal performance fuel and the surrounding lumped channel was GAIA fuel. The expanded model had a larger number of subchannels than the generic models and assumed that the assembly modeled in subchannel detail was a high thermal performance assembly and that the assemblies immediately surrounding it were set in a checkerboard of high thermal performance and GAIA assemblies to represent the fuel configuration in the first transition core.

3.1.5.3 Mixed Core SDL The licensee indicated in Section 3.4.3 of the LAR that it performed an analysis to determine the SDL for mixed core conditions. The methodology and plant conditions used were the same as that used for high thermal performance fuel, except that the licensee used the VIPRE-01 core model representing a high thermal performance fuel assembly surrounded by GAIA fuel. The licensee indicated that the analysis confirmed that the mixed core penalty derived following the 16 ADAMS Accession No. ML17102A923

methodology discussed in Section 3.1.5.1 of the SER was the only necessary compensation for applying high thermal performance peak limits to high thermal performance assemblies in a mixed core with GAIA fuel.

To clarify the licensees indication that the result of the analysis verified that the SDL calculated for the mixed core was the same as that calculated for a full core with high thermal performance fuel, the NRC staff requested the following additional information in a letter dated November 27, 2019:

(1) Provide the results of the SDL calculation, including similar information in Tables I-4 and I-6 of DPC-NE-2005-P-A, Revision 5, for the mixed core conditions with a high thermal performance fuel assembly surrounded by GAIA fuel.

(2) Provide a clarification to confirm that the SDL analysis was performed for a mixed core representing a fuel assembly design with characteristics similar to a GAIA fuel assembly surrounded by high thermal performance fuel and provide the results of the SDL analysis, or provide a rationale if it was not.

In its response to RAI-3 dated December 20, 2019, the licensee provided the results for item (1) of the SDL calculation for the mixed core conditions with a high thermal performance fuel assembly surrounded by GAIA fuel in Table 3-1 for the state point conditions, Table 3-2 for state point statistical results (500 case runs), and Table 3-3 for state point statistical results (10,000 case run). The referenced Tables in the RAI response provided the information similar to Tables I-4 and I-6 of DPC-NE-2005-P-A, Revision 5, and confirmed that the calculated SDL value for mixed core conditions with high thermal performance fuel surrounded by GAIA assemblies would not exceed the 1.35 value currently used for full-core high thermal performance fuel at Harris. Since the result of the analysis verified that the SDL calculated for the mixed core was the same as that calculated for a full-core with high thermal performance fuel, the NRC staff determined the SDL analysis for the mixed core conditions with an high thermal performance fuel assembly surrounded by fuel with characteristics similar to the GAIA fuel design using the ORFEO-GAIA correlation methodology was acceptable.

In the response for item (2), the licensee confirmed that the SDL analysis was not performed for a mixed core representing a GAIA fuel assembly surrounded by high thermal performance fuel assemblies. The licensees rationale for not performing the SDL analysis was based on its scoping analyses that demonstrated that the minimum DNBR for GAIA fuel surrounded by high thermal performance assembles was slightly higher than the minimum DNBR for the same power distribution in a full-core GAIA model. Figure 3.3 of the LAR showed that the flow flavored the surrounding GAIA assembles and was depressed in the central high thermal performance hot channel. Also, Table 3.2 of the LAR indicated that for a mixed core, the assembly inlet flow would divert to GAIA fuel from the high thermal performance fuel due to the lower assembly pressure drop, resulting in lower DNBR for high thermal performance fuel. The licensee was not crediting the conservatism of the mixed core GAIA configuration and was using the full-core GAIA model to calculate peaking limits for GAIA fuel in a mixed core configuration. For the penalty for the high thermal performance fuel in a mixed core configuration, which was significantly greater than the benefit for GAIA in a mixed core configuration, it would result in the SDL for the worst 10,000 case run increasing by 0.001 from a full-core of high thermal performance to a mixed core high thermal performance configuration.

As a result, the licensee did not expect that the much smaller deviation in GAIA minimum DNBR results would translate into an SDL that would be significant. Therefore, the licensee did not perform an SDL analysis for a high thermal performance and GAIA mixed core. The NRC staff

found that the rationale for not performing the SDL analysis for the high thermal performance and GAIA mixed core was reasonable.

3.1.5.4 DPC-NE-3008-P-A RAI-22 Regarding the Expanded VIPRE-01 Models In Section 3.4.4 of the LAR, the licensee provided parts of its response regarding mixed core applications in RAI 22 concerning DPC-NE-3008. RAI-22 identified the NRC staffs concerns about the use of the expanded VIPRE-01 models for a full-core with high thermal performance fuel at Harris.

3.1.5.4.1 Use of Expanded VIPRE-01 Model (RAI-22a)

RAI-22a identified the NRC staffs concern regarding how the expanded VIPRE-01 model was used for a full high thermal performance core. For a mixed core, the licensee indicated that the expanded VIPRE-01 model could be used to calculate the mixed core penalty factor following the methods discussed in Section 3.4.1 of the LAR. The expanded model could be also used to validate that the DNBR calculations using the generic mixed core model were conservative as discussed in Section 3.4.2 of the LAR, and could be used to assess potential departure from nucleate boiling impacts of fuel assembly design features and potential core issues as discussed in the original response to RAI-22a.

3.1.5.4.2 Crossflow and Turbulent Mixing Coefficients Used in Mixed VIPRE-01 Models (RAI-22e)

RAI-22e identified the NRC staffs concern regarding how the crossflow and turbulent mixing coefficients were selected in the expanded VIPRE-01 model for a full high thermal performance core. The VIPRE-01 energy and momentum equations would involve the turbulent momentum factor (FTM) and turbulent mixing coefficient (ABETA) to account for the turbulent crossflow effects. For the expanded mixed model, the licensee assumed a value of 0.0 for the FTM, and the average of the GAIA and high thermal performance values for the ABETA coefficient for the DNBR analysis.

The NRC staff evaluated the use of a value of 0.0 for the FTM and found that it implied that the turbulent crossflow only mixed enthalpy and not momentum between adjacent subchannels due to turbulent mixing. This assumption neglected the turbulent momentum transfer between subchannels on the assemblys border and more limiting subchannels located towards the center of the hot assembly and would result in a lower flow rate and calculated DNBR, and therefore, was conservative and acceptable.

The NRC staff also evaluated the ABETA coefficient and found that it was a fuel type specific value provided by the fuel vendor. Since the values of ABETA coefficients for GAIA and high thermal performance were similar and from the same fuel vendor, there was virtually no difference in the results of the DNBR analysis when using one fuel assemblys mixing coefficient or the average of the coefficients to account for turbulent crossflow between subchannels. The result would be true for the subchannels on the assembly border and the subchannels located towards the center of the hot assembly. Therefore, the NRC staff determined that the use of the average value for the ABETA coefficient in the expanded mixed models was acceptable.

3.1.5.4.3 Benchmarking the Expanded Model to Generic Model (RAI-22f)

RAI-22f contained the NRC staffs request for the benchmarking analysis of the expanded model against a generic model applying to a full high thermal performance core. In addressing the original RAI 22f, the licensee had benchmarked the high thermal performance expanded model to the generic model from DPC-NE-2005. Both models were applied to the same set of transient statepoints with the same set of code options, correlations and parameters. The licensee repeated a similar benchmark process using a core model with mixed high thermal performance and GAIA. The transients analyzed included: 100% power loss of flow; single uncontrolled rod withdrawal; 100% uncontrolled bank withdrawal at beginning of cycle; and 100% uncontrolled bank withdrawal at end of cycle. The benchmarks were presented in Table 3.3 of the LAR and showed comparable and consistent performance of the generic model and expanded model for the mixed core condition.

The licensee also performed an analysis to determine the flow effects of the expanded mixed model core. This analysis was based on the expanded core model with all pin powers set to 1.0 to assure any observed differences were from hydraulic effects only. The results are shown in 3-dimensional plots representing the calculated channel exit flow for a high thermal performance assembly in the middle with an adjacent GAIA fuel face. The NRC staff found in Figure 3.1 and Figure 3.2 of the LAR that the channel exit flow profiles in the full-core high thermal performance and GAIA remained relatively flat across the hot assembly and neighboring assemblies since the fuel types were identical. The NRC staff also found in Figure 3.3 that the flow would favor the surrounding GAIA assemblies and would be decreased in the central high thermal performance hot channel, which supported the information indicated in Table 3.2 of the LAR stating that coolant flow would divert from a high thermal performance assembly into a GAIA assembly in a mixed core configuration. Therefore, NRC staff concluded that the benchmarks adequately support the use of the expanded model for the core with mixed high thermal performance and GAIA fuel.

The NRC staff evaluated the derivation of the core penalty factor for the high thermal performance fuel at Harris and concluded that it is acceptable since it was based on the acceptable mixed core models with satisfactory benchmark results.

3.2 TS Changes The licensee proposed changes to two TSs (TS 2.1.1.a and TS 6.9.1.6.2) to allow operation at Harris with the transition to a fuel with characteristics similar to the GAIA fuel design using the ORFEO-GAIA correlation methodology from the current high thermal performance fuel. The NRC staffs evaluation of these TS changes is as follows:

3.2.1 TS 2.1.1.a - DNBR Safety Limit For operation at Harris with the current high thermal performance fuel the DNBR safety limit was 1.141 using the high thermal performance DNBR correlation; the DNBR safety limit was included in Harris TS 2.1.1.a, which currently states that the departure from nucleate boiling ratio (DNBR) shall be maintained > 1.141 for the HTP [high thermal performance] DNB correlation.

The licensees LAR would modify Harris TS 2.1.1.a to incorporate the DNBR safety limit for the GAIA fuel utilizing ORFEO-GAIA departure from nucleate boiling correction as follows:

a. The departure from nucleate boiling ratio (DNBR) shall be maintained > 1.141 for the HTP DNB correlation for HTP fuel and >1.12 for the ORFEO-GAIA DNB correlation for GAIA fuel.

As discussed in Section 3.1.2.2 of this SER, the acceptable calculated DNBR safety limit was 1.12 using the ORFEO-GAIA correlation. The NRC staff reviewed the proposed TS 2.1.1.a and found that the added DNBR safety limit used the same acceptable value of 1.12 when applying the ORFEO-GAIA correlation for GAIA fuel. Therefore, the NRC staff concluded that the proposed TS 2.1.1.a was acceptable.

3.2.2 TS 6.9.1.6.2 - Analytical Methods Used to Determine Core Operating Limits The current TS 6.9.1.6.2 includes numbers and titles of topical reports documenting analytical methods that were approved by the NRC for determining the core operating limits, and required that the approved revision number at the time the reload analyses were performed be identified in the COLR. The licensee proposed to add to Harris TS 6.9.1.6.2 the following NRC-approved topical report as:

v. ANP-10341-A, The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations, approved version as specified in the COLR.

(Methodology for Specification 2.1.1 - Reactor Core Safety Limits, 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, and 3.2.5 - DNB Parameters).

After the approval of TSTF-363 in year 2000, the NRC staff had concerns that listing only the topical report number and title in the TS would omit conditions contained in the NRC staffs SER that approved the topical report. To resolve the staffs concerns with TSTF-363, the improved TS was revised to eliminate the provisions of TSTF-363 that allowed licensees to list COLR references in the TS without revisions and dates. The improved TS revision and subsequent withdrawal of TSTF-363 were documented in TSTF-533-T, Transmittal of TSTF-533-T, Revision 0, Remove COLR and PTLR [Pressure-Temperature Limits Report], Revision and Date Relocation Provisions Added by TSTF-363, -408, and -419.17 Because of the large number of previous approvals to implement TSTF-363, the NRC staff considered the requirements of 10 CFR 50.109, Backfitting, and concluded that there was not a substantial increase in the overall protection of the public health and safety to be derived from backfitting licensees that had already adopted TSTF-363. As a result, the NRC staff stated in its letter dated August 4, 2011 that the NRC staff did not intend to backfit licensees whose TS already reflected the flexibility allowed by TSTF-363. For subsequent licensing actions, the NRC staff allowed licensees that had received NRC approval to implement TSTF-363 not to include revision numbers and approval dates in the proposed citations for new NRC-approved topical reports, when the topical reports were proposed for addition to the TS reference list.

As discussed in the licensees submittal, the current Harris TSs had incorporated TSTF-363, which allowed licensees to omit topical report revision numbers and dates from the TS list of COLR references. Following the same TS change practice, the licensee proposed to not include the revision number and date in its reference to ANP-10341-A. The NRC staff evaluated the TS changes and determined that the licensees approach of adding ANP-10341-A without inclusion of the revision and associated date in TS 6.9.1.6.2.v was consistent with the 17 ADAMS Accession No. ML112590444

current NRC review practice aligned with the statement in the NRC staffs letter of August 4, 2011. Therefore, the NRC staff concluded that the proposed TS 6.9.1.6.2.v with the added NRC-approved topical report ANP-10341-A without revision number and date, was acceptable.

4.0 NRC Staff Conclusion

The NRC staff evaluated the proposed changes to modify TS 2.1.1.a to add the DNBR safety limit associated with the transition from the high thermal performance fuel to a fuel assembly design with characteristics similar to the GAIA fuel design using the ORFEO-GAIA correlation methodology. The NRC staff also reviewed the licensee's proposed changes to add topical report, ANP-10341-P-A, to TS 6.9.1.6.2 as a COLR reference for the NRC-approved departure from nucleate boiling correlation associated with the GAIA fuel design for applicable fuel specifications and characteristics. The NRC staff reviewed the information provided by the licensee in DPC-NE-2005-P, Revision 6, that outlined the parameters for Harris plant specific data for GAIA fuel application of the ORFEO-GAIA Correlation to the GAIA Fuel Design. The NRC staff determined that these TS changes are acceptable, based on the acceptable DNBR safety limit with application of the acceptable DPC-NE-2005-P, Revision 6, and ANP-10341-A methodologies. The NRC staff concludes that there is reasonable assurance that the proposed TS changes will have minimal impact on the licensee's ability to continue to comply with the requirements of 10 CFR 50.36 and GDC 10. This safety evaluation does not approve the use of GAIA fuel nor does it approve the use of the generically approved GAIA topical report in ANP-10342-P-A, GAIA Fuel Assembly Mechanical Design.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the State of North Carolina official was notified of the proposed issuance of the amendment on July 20, 2020. The State official had no comments.

6.0 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 and/or changes surveillance requirements. 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 (85 FR 17601). 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.

7.0 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.

Principal Contributors: T. Hood S. Sun Date: September 29, 2020

ML20212L594 *by memorandum **by e-mail OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DSS/SNSB/BC NRR/DSS/STSB/BC NAME THood BAbeywickrama SKrepel VCusumano DATE 07/17/2020 08/05/2020 08/11/2020 08/13/2020 OFFICE NRR/DSS/SFNB/BC OGC - NLO** NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME RLukes STurk UShoop THood DATE 08/10/2020 09/23/2020 09/28/2020 09/29/2020