ML24213A052

Issuance of Amendment No. 201 to Extend Completion Time of Inoperable Reactor Coolant System Accumulator Using Consolidated Line Item Improvement Process

ML24213A052
Person / Time
Site:	Harris
Issue date:	08/02/2024
From:	Michael Mahoney Plant Licensing Branch II
To:	Haaf T Duke Energy Progress
Mahoney M
References
EPID L-2024-LLA-0032
Download: ML24213A052 (1)

Text

August 2, 2024

Thomas P. Haaf 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. 201 TO EXTEND COMPLETION TIME OF INOPERABLE REACTOR COOLANT SYSTEM ACCUMULATOR USING CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (EPID L-2024-LLA-0032)

Dear Thomas Haaf:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued Amendment No. 201 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1. This amendment is in response to your application dated March 20, 2024.

Specifically, the amendment extends the completion time of Action a of Technical Specification 3/4.5.1, Accumulators, from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore a Reactor Coolant System accumulator to operable status when declared inoperable due to any reason except not being within the required boron concentration range.

The changes are consistent with NRC-approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-370, Risk-Informed Evaluation of an Extension to Accumulator Completion Times for Westinghouse Plants. The availability of this TS improvement was announced in the Federal Register on March 12, 2003, as part of the consolidated line item improvement process (CLIIP).

T. Haaf

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

Sincerely,

/RA/

Michael Mahoney, 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. 201 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 FA CILITY OPERATING LICENSE

Amendment No. 201 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 March 20, 2024, 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. 201, 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 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. NPF-63 and Technical Specifications

Date of Issuance: August 2, 2024

ATTACHMENT TO LICENSE AMENDMENT NO. 201

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 page of the Appendix A, Technical Specifications, with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change:

Remove Insert 3/4 5-1 3/4 5-1

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. 201, 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 t he 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. 201 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION

LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:

a. The isolation valve open with power supply circuit breaker o pen,

b. A contained borated water volume of between 66 and 96% indicated level,

c. A boron concentration within the limits specified in the COLR, and

d. A nitrogen cover-pressure of between 585 and 665 psig.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

a. With one accumulator inoperable, except as a result of boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. DELETED.

c. With one accumulator inoperable due to boron concentration not within limits, restore the boron concentration within limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

a. At the frequency specified in the Surveillance Frequency Co ntrol Program by:

1. Verifying that the contained borated water volume and nitrogen cover-pressure in the tanks are within their limits, and

2. Verifying that each accumulator isolation valve is open.

• RCS pressure above 1000 psig.

SHEARON HARRIS - UNIT 1 3/4 5-1 Amendment No. 171, 201 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO AMENDMENT NO. 201 TO

RENEWED FACILITY OPERATIN G LICENSE NO. NPF-63

DUKE ENERGY PROGRESS, LLC

SHEARON HARRIS NU CLEAR POWER PLANT, UNIT 1

DOCKET NO. 50-400

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC or the Commission) dated March 20, 2024 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML24080A449), Duke Energy Progress, LLC (Duke Energy or the licensee) requested changes to the technical specifi cations (TSs) for the Shearon Harris Nuclear Power Plant, Unit 1 (Harris).

Specifically, the requested change would extend the completion time (CT) of Action a of TS 3/4.5.1, Accumulators, and its associated Bases from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore a Reactor Coolant System (RCS) accumulator to operable status when declared inoperable due to any reason except not being within the required boron concentration range.

The licensee stated that the proposed changes are consistent with NRC-approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-370, Risk-Informed Evaluation of an Extension to Accumulator Completion Times for Westinghouse Plants (ML003771348). The availability of this TS improvement was announced in the Federal Register on March 12, 2003, as part of the consolidated line item improvement process (CLIIP).

2.0 REGULATORY EVALUATION

2.1 Background

Topical Report WCAP-15049, Risk-Informed Evaluation of an Extension to Accumulator Completion Times, was submitted to the NRC on August 20, 1998, and approved in the NRC letter dated February 19, 1999. The WCAP evaluates the risk associated with extending the accumulator CT from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for reasons other than boron concentration out of specification.

Enclosure 2

Wolf Creek was the lead plant for the Westinghouse Owners Group (WOG) program and received plant specific approval for changes to the TSs on April 27, 1999 (License Amendment No. 124). In the NRC letter of February 19, 1999, the staff indicated that it will not repeat its review of the matters described in Topical Report WCAP-15049 when the report appears as a reference in license applications, except to ensure that the material presented applies to the specified plants involved.

2.2 Description of Changes

Current TS 3.5.1, Action a, states:

With one accumulator inoperable, except as a result of a closed isolation valve or boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig [pounds per square inch gauge] within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Current TS 3.5.1, Action b, states:

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The licensee proposed to revise TS 3.5.1, Action a, as follows (deletions in double strike-through and additions in bold):

With one accumulator inoperable, except as a result of a closed isolation valve or boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1 hour or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Additionally, the licensee proposed to delete TS 3.5.1, Action b, in its entirety.

2.3 Applicable Regulatory Requirements

Title 10 of the Code of Federal Regulations (10 CFR) 50.36(b) states that TSs will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34, Contents of applications; technical information.

The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). As required by 10 CFR 50.36(c)(2), the TSs will include limiting conditions for operation (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the LCO can be met.

2.4 Applicable Regulatory Guidance

Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3

(ML17317A256), describes an approach that is acceptable to the NRC staff for developing risk-informed applications for a licensing bas is change that considers engineering issues and applies risk insights.

RG. 1.177, Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, Revision 2 (ML20164A034), describes an approach that is acceptable to NRC staff for developing risk-informed applications for changes to CTs.

NRC Regulatory Issue Summary 2000-006, Consolidated Line Item Improvement Process for Adopting Standard Technical Specifications Changes for Power Reactors, dated March 20, 2000 (ML003693442), describes the process available to licensees for proposed adoption of generic changes to STS.

3.0 TECHNICAL EVALUATION

3.1 Deterministic Evaluation

The purpose of the emergency core cooling system (ECCS) accumulators is to supply water to the reactor vessel during the blowdown phase of a loss-of-coolant accident (LOCA). The accumulators are large volume tanks, filled with borated water and pressurized with nitrogen.

The cover-pressure is less than that of the RCS so that following an accident, when the RCS pressure decreases below tank pressure, the accumulators inject the borated water into the RCS cold legs. The current deterministic safety analysis has not been changed and, thus, the LCO, that is, the lowest functional capability required for safe operation continues to be:

LCO 3.5.1 Three ECCS accumulators shall be OPERABLE.

Applicability: MODES 1 and 2, MODE 3 with RCS pressure > 1000 psig.

Under Actions, TSs allow for limited deviations fr om the LCO. Historically, these Actions and associated CTs have been set using judgment and are not part of the deterministic safety analysis discussed above. Currently, the Harris TSs allow for one accumulator to be inoperable for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for reasons other than boron concentration not within limits during Modes 1, 2, and in Mode 3 with pressurizer pressure greater than a plant specific pressure. The WCAP, as well as this TSTF, proposes to increase this CT to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The proposed CT of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is an extension of the current ACTION statement. CTs are, by their nature, determined by conditions of risk and the impact of the proposed change on risk is reviewed in the following section.

3.2 Risk Evaluation

A three-tiered approach, consistent with RG 1.177, was used by the staff to evaluate the risk associated with the proposed accumulator CT, or allowed outage time (AOT), extension from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The need for the proposed change was that the current 1-hour CT would be insufficient, in most cases, for licensees to take a reasonable action when an accumulator was found to be inoperable.

Tier 1: Quality of Probabilistic Risk Assessment (PRA) and Risk Impact

Westinghouse used a reasonable approach to assess the risk impact of the proposed accumulator CT extension. The approach is generally consistent with the intent of the applicable

NRC RGs 1.174 and 1.177. The quantitative risk m easures addressed in the topical report included the change in core damage frequency (CDF) and incremental conditional core damage probability (ICCDP) where ICCDP = [(conditional CDF with the subject equipment out-of-service) - (baseline CDF with nominal expected equipment unavailabilities) x (duration of single CT under consideration)] for a single CT.

The change in large early release frequency (LERF) and incremental conditional large early probability (ICLERP) where ICLERP = [(conditional LERF with the subject equipment out-of-service) - (baseline LERF with nominal expected equipment unavailabilities) x (duration of single CT under consideration)] for a single CT was qualitatively addressed. Representative calculations were performed to determine the risk impact of the proposed change.

Various accumulator success criteria were considered in these calculations to encompass the whole spectrum of Westinghouse plants (e.g., two-, three-and four-loop plants). A reasonable effort was also made to address the differences in other components of risk analysis such as initiating event (IE) frequency and accumulator unavailability among Westinghouse plants.

Westinghouse considered a comprehensive range of IEs in the risk analysis. LOCAs in all sizeslarge, medium, and smallwere included, and reactor vessel failure and interfacing system LOCAs were also considered. Modeling of accumulators for mitigation of events other than large, medium and small LOCAs was identifi ed to have insignificant risk impact; therefore, the analysis was performed only on accumulator injection in response to large, medium, and small LOCA events.

The success criteria considered are summarized as follows:

LOCA Category No. of Loops Success Criteria

Large 4 3 accumulators to 3 of 3 intact loops (3/3) 2 accumulators to 2 of 3 intact loops (2/3) no accumulators required (0/3) 3 2 accumulators to 2 of 2 intact loops (2/2) 1 accumulator to 1 of 2 intact loops (1/2) no accumulators required (0/2) 2 1 accumulator to 1 of 1 intact loop (1/1) no accumulators required (0/1)

Medium and Small 4 3 accumulators to 3 of 3 intact loops (3/3) 3 2 accumulators to 2 of 2 intact loops (2/2)

2 1 accumulator to 1 of 1 intact loop (1/1)

The success criteria considered in this analysis were comprehensive and considered conservative in many cases. For example, many plants indicated the accumulator success criteria for medium and small LOCA events resulted from their role in an alternate success path, in which high-pressure injection (HPI) had already failed. Additionally, the NRC staffs review of

a number of the original individual plant examinations (IPEs) indicated that no accumulator was needed at all for many medium LOCA sequences and for most of the small LOCA sequences.

The fault-trees that model accumulator unavailabilities were evaluated. The assumptions made in the fault tree modeling were detailed and were found to be reasonable. For example, the model assumed that the total CT would be used for each corrective maintenance, and this was considered conservative. A comprehensive list of failure mechanisms was considered, and potential common cause failures for check valves and motor-operated valves were also included. Westinghouse used the Multiple Greek Letter technique to determine the common cause failure contributions to the accumulator injection failure.

The component failure rates were taken from the Advanced Light Water Utility Requirements Document, Advanced Light Water Utility Requirements Document, Volume II, ALWR Evolutionary Plant, Chapter 1, Appendix A, PRA Key Assumptions and Ground Rules, Rev. 5, issued December 1992. Accumulator unavailabilities due to boron concentration out-of-limit and due to other reasons were calculated based on a survey of a number of Westinghouse plants.

The values for component failure rates and accumulator unavailabilities were within reasonable range. The common cause factors used were also comparable to those used in other PRAs.

The accumulator fault trees were quantified using the WesSAGE computer code. The code provided information on the unavailability and cutsets related to the component failures and maintenance activities modeled in the fault trees. A separate hand calculation was used to determine the unavailability due to potential common cause failures. Evaluation of some of the cutsets provided in the topical report did not reveal any unexpected results.

The NRC staff examined the accident sequence i dentification for each LOCA category. The probability of the sequence leading to core damage involving accumulator failure is summarized for each LOCA category as follows:

Large LOCA (Large LOCA IE frequency) x (accumulator unavailab ility)

Medium LOCA (Medium LOCA IE frequency) x (unavailability of HPI) x (accumulator unavailability)

Small LOCA (Small LOCA IE frequency) x (unavailability of HPI) x (accumulator unavailability)

The LOCA IE frequencies used for WCAP-15049 are summarized below. Also listed are the LOCA frequencies used in NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Volume 1, Rev. 1, January 1990 (the NUREG-1150 study), for pressurized-water reactors and those in the original IPEs.

WCAP-15049 NUREG-1150 IPE Average (High; Low)

Large LOCA 3x10-4/yr 5x10-4/yr 3.3x10-4/yr (5x10-4/yr; 1x10-5/yr)

Medium LOCA 8x10-4/yr 1x10-3/yr 7.9x10-4/yr (2.6x10-3/yr; 1x10-4/yr)

Small LOCA 7x10-3/yr 1x10-3/yr 8.9x10-3/yr (2.9x10-2/yr; 3.7x10-4/yr)

Westinghouse indicated that the IE frequencies for WCAP-15049 were based on the plant-specific information contained in the WOG Probabilistic Safety Assessment Comparison Database, which documented the PRA modeling methods and results of the updated PRAs for Westinghouse plants. The mean IE frequencies were used for the risk analysis. These were comparable to the values used for the NUREG-1150 study and the average values in the original IPEs. The NRC staff also found that t he IE frequency values in the high range among the original IPEs were not much higher than those used for this topical report. The HPI unavailability values used were 7x10 -3 and 1x10-3/yr for medium and small LOCA events, respectively. The NRC staffs examination revealed that the HPI unavailability values were generally comparable to those used in other PRAs and were generally conservative.

The risk measures calculated to determine the impact on plant risk were based on three different cases. The risk measures considered in each case included the impact on CDF and ICCDP for a single CT, and the impact on LERF and ICLERP for a single CT were qualitatively considered.

The three cases considered were:

Design basis case. This case required accumulator injection only for mitigation of large LOCA events (3/3 for 4-loop, 2/2 for 3-loop, and 1/1 for 2-loop).

Case 1. This case credited realistic accumulator success criteria (2/3 for 4-loop, 1/2 for 3-loop, and 0/1 for 2-loop) for large LOCA events and credited the use of accumulators in responding to medium and small LOCA events (3/3, 2/2, and 1/1 for 4-loop, 3-loop, and 2-loop, respectively) following failure of HPI.

Case 2. This case credited more realistic impr oved accumulator success criteria (no accumulator required) for large LOCA events and credited the use of accumulators in responding to medium and small LOCA events (3/3, 2/2, and 1/1 for 4-loop, 3-loop, and 2-loop, respectively) following failure of HPI.

The results were summarized as follows:

Case LOCA CDF(/yr) LOCA CDF(/yr) CDF ICCDP (Current) (Proposed)

4-loop Design Basis 6.93x10-7 9.24x10-7 2.31x10-7 8.20x10-7 4-loop Case 1 6.23x10-8 7.77x10-8 1.54x10-8 5.53x10-8 4-loop Case 2 4.57x10-8 6.09x10-8 1.52x10-8 5.41x10-8

3-loop Design Basis 4.62x10-7 6.18x10-7 1.56x10-7 8.21x10-7 3-loop Case 1 4.27x10-8 5.31x10-8 1.04x10-8 5.48x10-8 3-loop Case 2 3.05x10-8 4.08x10-8 1.03x10-8 5.42x10-8

2-loop Design Basis 2.31x10-7 3.09x10-7 7.80x10-8 8.21x10-7 2-loop Case 1 1.52x10-8 2.04x10-8 5.20x10-9 5.42x10-8 2-loop Case 2 1.52x10-8 2.04x10-8 5.20x10-9 5.42x10-8

For both realistic cases, the CDFs and ICCDPs were very small for 2-loop, 3-loop, and 4-loop plants, and were much below the numerical guidelines in RGs 1.174 and 1.177. The NRC staff also noted that the values were considered still bounding in the sense that the risk analysis used a multitude of conservative assumptions and data in the modeling. For many Westinghouse plants, the realistic impact on risk would be much smaller than the values above.

A set of sensitivity cases were also calculated using higher IE frequencies for small and medium LOCAs. The results of the sensitivity calculations did not cause the overall risk impact to increase significantly.

Westinghouse indicated that accumulator success or failure has no direct impact on the containment performance, and that the LERF would, therefore, increase only in direct proportion to the increased CDF due to accumulator failures. Westinghouse concluded that, since the impact on CDF was small, the impact on LERF would also be small. The NRC staff found the Westinghouse argument to be acceptable; therefore, the impact on LERF and ICLERP for a single CT was very small.

One of the potential benefits of the proposed extended CT was the averted risk associated with avoiding a forced plant shutdown and startup. The risk associated with a forced plant shutdown and ensuing startup due to the inflexibility in current TSs could be significant in comparison with the risk increase due to the proposed accumulator CT increase.

Based on the NRC staffs Tier 1 review, the qualit y of risk analysis used to calculate the risk impact of the proposed accumulator CT extension was reasonable and generally conservative.

It was also found that the risk impact of the proposed change was below the NRC staff guidelines in RGs 1.174 and 1.177.

Tiers 2 and 3: Configuration Risk Control

Tier 2 of RG 1.177 addresses the need to preclude potentially high-risk configurations which could result if certain equipment is taken out-of-service during implementation of the proposed TS change (in this case accumulator CT). If such configurations are identified, the licensee should also identify appropriate measures to avoid them.

The accumulators are always needed to mitigate large size LOCAs. Large LOCAs require accumulators to inject as analyzed under Tier 1 in order to avoid core damage. This means that if a large LOCA occurs without the accumulator function, the core will be damaged independently of whether other systems, such as HPI, function properly or not. However, the probability that a large LOCA occurs in the 24-hour CT is extremely small (in the order of 1E-7 or less). Furthermore, no compensatory or other measures are possible. Due to the negligible risk increase associated with this scenario and the fact that there are no measures to take once a large LOCA occurs, no high risk configurations are associated with this scenario.

In general, medium LOCAs do not require accumulators if at least one HPI train is available.

This means that if a medium LOCA occurs when minimum accumulator functionality is unavailable and at the same time HPI is unavailable, the core will be damaged. However, the probability that a medium LOCA occurs in the 24-hour CT and at the same time both trains of HPI are unavailable is extremely small (in the order of 1E-8 or less), because it is assumed that the plant is not operating at power with both HPI trains out-of-service. This assumption is based on current STSs that limit operation at power with no HPI capability. Therefore, no Tier 2 restrictions beyond those currently in the STSs are deemed necessary.

Tier 3 calls for a program to identify risk significant configurations beyond those identified in Tier 2 resulting from maintenance or other operational activities and take appropriate compensatory measures to avoid such configurations. Because the accumulator sequence modeling is relatively independent of that for other systems, the Tier 2 analysis by itself is sufficient.

Furthermore, 10 CFR 50.65(a)(4) (Maintenance Rule) requires that licensees assess the risk any time maintenance is being considered on safety-related equipment. This requirement serves the objectives of Tier 3.

In summary, the Tier 2 evaluation did not identify the need for any additional constraints or compensatory actions that, if implemented, would avoid or reduce the probability of a risk-significant configuration. The current TS provisions were found to be sufficient to address the Tier 2 issue. Because the accumulator sequence modeling is relatively independent of that for other systems and the implementation of the Maintenance Rule, the NRC staff concluded that application of Tier 3 to the proposed accumulator CT was not necessary.

The NRC staff finds that the proposed changes will allow safe operation with the changes in CT from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Action (a) of TS LCO 3.5.1. The NRC staff also finds that the proposed changes are consistent with the ICCDPs calculated in WCAP-15049 for the accumulator AOT increase and meet the criterion of 5E-07 in RGs 1.174 and 1.177. The analysis and acceptance provided in this SE, as demonstrated by WCAP-15049, covers all Westinghouse nuclear steam system supplier plants regardless of plant vintage and number of loops. The NRC staff, therefore, concludes that the proposed TSTF-370, Revision 0 changes are acceptable.

3.3. NRC Evaluation of Variation from TSTF-370

In its application, the licensee states the following:

The HNP [Harris] TS are based upon the format and content of the NUREG-0452, Standard Technical Specifications for Westinghouse Pressurized Water Reactors, series (ADAMS Accession No. ML102590431). As a result, the HNP

TS utilize different numbering and titles than NUREG-1431, Standard Technical Specifications -Westinghouse Plants (ADAMS Accession No. ML21259A155) and has a separate action for an inoperable accumulator due to the isolation valve being closed. TSTF-370, which is based on NUREG-1431, addresses extending the completion time for one accumulator being inoperable for reasons other than the boron concentration not within limits, which would apply to both HNP TS Actions a and b. For consistency with NUREG-1431 and TSTF-370, this variation addresses the revision to HNP TS 3.5.1 Action a to delete the exclusion of its applicability to a closed isolation valve and delete Action b in its entirety. The applicability of TSTF-370 and the supporting evaluation from WCAP-15049-A, Risk-Informed Evaluation of an Extension to Accumulator Completion Times, for a 24-hour completion time is for any reason other than boron concentration not within limits; therefore, this would include a closed isolation valve. There is no technical reason to have a separate action or different completion time for the accumulator being inoperable due to a closed isolation valve.

The NUREG-1431 standard TSs (which TSTF-370 is compared to), do not contain a specific condition for an accumulator inoperable due to a closed isolation valve. The licensees proposed Action a provides a CT for an inoperable accumulator for any reason other than boron concentration not within limits, which would include a closed isolation valve (which is the purpose of current Action b). WCAP-15049 evaluates the risk associated with extending the accumulator CT from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for reasons other than boron concentration out of specification, which includes a closed isolation valve. Volume 2 of NUREG-1431, Standard Technical Specifications - Westinghouse Plants: Bases, Revision 5.0 (ML21259A159), states that the 24-hour CT to open the isolation valve or remove power to the valve ensures that prompt action will be taken to return the inoperable accumulator to OPERABLE status.

The licensees proposed Action a also includes the same remedial action as its current Action b to be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to be less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The NRC staff has reviewed this variation and determined that the licensees Action a, as revised, is inclusive of a closed isolation valve and the same remedial action as current Action b. Therefore, Action b is not needed. As such, the NRC staff has determined that the variation from TSTF-370, as described by the licensee, in deleting TS 3.5.1, Action b, is acceptable.

Based on its review noted above, the NRC staff finds that Harris TS 3.5.1, as revised, meets the requirement in 10 CFR 50.36(c)(2)(i) because the LCO will continue to require the lowest functional capability or performance levels of equipment required for safe operation of the facility, and when the LCO is not met, the licensee will shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

An evaluation of the no significant hazards consideration is presented below, which was previously published in the Federal Register on July 15, 2002 (67 FR 46542).

Criterion 1 - The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated

The basis for the accumulator limiting condition for operation (LCO), as discussed in Bases Section 3.5.1, is to ensure that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators, thereby providing the initial cooling mechanism during large RCS pipe ruptures. As described in Section 9.2 of the WCAP-15049, Risk-Informed Evaluation of an Extension to Accumulator Completion Times, evaluation, the proposed change will allow plant operation in a configuration outside the design basis for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, instead of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, before being required to begin shutdown. The impact of the increase in the accumulator CT on core damage frequency for all the cases evaluated in WCAP-15049 is within the acceptance limit of 1.0E-06/yr for a total plant core damage frequency (CDF) less than 1.0E-03/yr. The incremental conditional core damage probabilities calculated in WCAP-15049 for the accumulator CT increase meet the criterion of 5E-07 in Regulatory Guides (RG) 1.174 and 1.177 for all cases except those that are based on design basis success criteria. As indicated in WCAP-15049, design basis accumulator success criteria are not considered necessary to mitigate large break loss-of-coolant accident (LOCA) events, and were only included in the WCA-15049 evaluation as a worst case data point. In addition, WCAP-15049 states that the NRC has indicated that an incremental conditional core damage frequency (ICCDP) greater than 5E-07 does not necessarily mean the change is unacceptable.

The proposed technical specification change does not involve any hardware changes nor does it affect the probability of any event initiators. There will be no change to normal plant operating parameters, engineered safety feature (ESF) actuation setpoints, accident miti gation capabilities, accident analysis assumptions or inputs.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Criterion 2 - The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of the proposed change. As described in Section 9.1 of the WCAP-15049 evaluation, the plant design will not be changed with this proposed technical specification CT increase. All safety systems still function in the same manner and there is no additional reliance on additional systems or procedures. The proposed accumulator CT increase has a very small impact on core damage frequency. The WCAP-15049 evaluation demonstrates that the small increase in risk due to increasing the accumulator allowed outage time (AOT) is within the acceptance criteria provided in RGs 1.174 and 1.177. No

new accidents or transients can be introduced with the requested change and the likelihood of an accident or transient is not impacted.

The malfunction of safety related equipment, assumed to be operable in the accident analyses, would not be caused as a result of the proposed technical specification change. No new failure mode has been created and no new equipment performance burdens are imposed.

Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3 - The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety

The proposed change does not involve a significant reduction in a margin of safety. There will be no change to the departure from nucleate boiling ratio (DNBR) correlation limit, the design DNBR limits, or the safety analysis DNBR limits.

The basis for the accumulator LCO, as discussed in Bases Section 3.5.1, is to ensure that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators, thereby providing the initial cooling mechanism during large RCS pipe ruptures. As described in Section 9.2 of the WCAP-15049 evaluation, the proposed change will allow plant operation in a configuration outside the design basis for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, instead of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, before being required to begin shutdown. The impact of this on plant risk was evaluated and found to be very small. That is, increasing the time the accumulators will be unavailable to respond to a large LOCA event, assuming accumulators are needed to mitigate the design basis event, has a very small impact on plant risk.

Since the frequency of a design basis large LOCA (a large LOCA with loss of offsite power) would be significantly lower than the large LOCA frequency of the WCAP-15049 evaluation, the impact of increasing the accumulator CT from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on plant risk due to a design basis large LOCA would be significantly less than the plant risk increase presented in the WCAP-15049 evaluation.

Therefore, this change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, the NRC staff concludes that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.

5.0 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 30, 2024. The State of North Carolina official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to installation or use of a facilitys components 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 (89 FR 49241, dated June 11, 2024), and there has been no public comment on such finding. Accordingly, th e 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 to be prepared in connection with the issuance of the amendments.

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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: R. Grover, NRR M. Mahoney, NRR

Date of Issuance: August 2, 2024

ML24213A052 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DSS/STSB/BC NAME MMahoney ABaxter SMehta DATE 07/31/2024 07/31/2024 08/01/2024 OFFICE NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME DWrona (KGreen for) MMahoney DATE 08/02/2024 08/02/2024