Text

Issuance of Amendment No. 181 Regarding Adoption of 10 CFR Part 50, Appendix J, Option B, for Type B and C Testing and for Permanent Extension of Type a, B, and C Leak Rate Testing Frequencies(L-2019-LLA-0256)
	ML20259A512
Person / Time
Site:	Harris
Issue date:	12/08/2020
From:	Michael Mahoney; Plant Licensing Branch II
To:	Maza K; Duke Energy Progress
	Mahoney M
References
	EPID L-2019-LLA-0256
	Download: ML20259A512 (57)
	v • d • e

December 8, 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. 181 REGARDING ADOPTION OF 10 CFR PART 50, APPENDIX J, OPTION B, FOR TYPE B AND C TESTING AND FOR PERMANENT EXTENSION OF TYPE A, B, AND C LEAK RATE TEST FREQUENCIES (EPID L-2019-LLA-0256)

Dear Ms. Maza:

The U.S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 181 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 November 8, 2019, as supplemented by letter dated April 16, 2020.

The amendment revises Technical Specification 6.8.4.k, Containment Leakage Rate Testing Program, by replacing the reference to Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, with a reference to Nuclear Energy Institute (NEI) 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, and the conditions and limitations specified in NEI 94-01, Revision 2-A, of the same name, dated October 2008. The amendment would allow a permanent extension of the Type A test interval from 10 years to 15 years, a more conservative allowable test interval extension of 9 months for Type A, Type B and Type C leakage rate tests, and an extension of the Type C test interval up to 75 months.

K. Maza 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. 181 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. 181 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 November 8, 2019, as supplemented by letter dated April 16, 2020, 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. 181, 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 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Undine Undine S. S. Shoop Date: 2020.12.08 Shoop 08:28:00 -05'00' Undine S. Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility License No. NPF-63 and Technical Specifications Date of Issuance: December 8, 2020

ATTACHMENT TO LICENSE AMENDMENT NO. 181 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 3/4 6-1 3/4 6-1 3/4 6-2 3/4 6-2 3/4 6-4 3/4 6-4 3/4 6-4a 3/4 6-4a 3/4 6-5 3/4 6-5 3/4 6-8 3/4 6-8 6-19c 6-19c

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. 181, 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. 181

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

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

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At the frequency specified in the Surveillance Frequency Control Program by verifying that all penetrations*# not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3;

b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and

c. By performing required visual examinations and leakage rate testing, except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program.

Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

Valves CP-B3, CP-B7, and CM-B5 may be verified at the frequency specified in the Surveillance Frequency Control Program by manual remote keylock switch position.

SHEARON HARRIS - UNIT 1 3/4 6-1 Amendment No. 181

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be within the limits specified in the Containment Leakage Rate Testing Program.

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

ACTION:

With the containment leakage rate not within the limits specified in the Containment Leakage Rate Testing Program, restore the leakage rate to within the limits specified in the Containment Leakage Rate Testing Program prior to increasing the Reactor Coolant System temperature above 200°F.

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rate tests shall be performed in accordance with the Containment Leakage Rate Testing Program described in Technical Specification 6.8.4.k.

PAGE 3/4 6-3 WAS DELETED BY AMENDMENT NO. 181 SHEARON HARRIS - UNIT 1 3/4 6-2 Amendment No. 181

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Two containment air locks shall be OPERABLE:

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

ACTION:

............................................................................. Notes .................................................................

1. Entry and exit is permissible to perform repairs on the affected air lock components.

2. A separate ACTION is allowed for each air lock.

3. Enter 3.6.1.1 LCO for "Containment Integrity" when the air lock leakage results in exceeding the containment leakage rate, Specification 3.6.1.2.

4. Locking a Personnel Air Lock door shut consists of locking the associated manual pumping stations and deactivating the electronic mechanisms used to open a Personnel Air Lock door once the associated air lock door is shut. Locking an Emergency Air Lock door shut consists of locking the mechanical operator.

a. One or more containment air locks with one containment air lock door inoperable:#

1. Within one hour, verify the OPERABLE door is closed in the affected air lock, and

2. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , lock the OPERABLE door closed in the affected air lock, and

3. Once per 31 days, verify the OPERABLE door is locked closed in the affected air lock*, or

4. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1. ACTIONS 3.6.1.3.a.1, 3.6.1.3.a.2, 3.6.1.3.a.3, and 3.6.1.3.a.4 are not applicable if both doors in the same air lock are inoperable and ACTION 3.6.1.3.c is entered.

2. Entry and exit is permissible for 7 days under administrative controls if both air locks are inoperable.

Air lock doors in high radiation areas may be verified closed by administrative means.

SHEARON HARRIS - UNIT 1 3/4 6-4 Amendment No. 181

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION

b. One or more containment air locks with containment air lock interlock mechanism inoperable.##

1. Within one hour, verify an OPERABLE door is closed in the affected air lock, and

2. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , lock an OPERABLE door closed in the affected air lock, and

3. Once per 31 days, verify the OPERABLE door is locked closed in the affected air lock*, or

4. Otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. One or more containment air locks inoperable for reasons other than 3.6.1.3.a or 3.6.1.3.b.

1. Immediately initiate action to evaluate containment leakage rate per LCO 3.6.1.2, and

2. Within one hour, verify a door is closed in the affected air lock, and

3. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , restore air lock to OPERABLE status, or

4. Otherwise be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1. 1. ACTIONS 3.6.1.3.b.1, 3.6.1.3.b.2, 3.6.1.3.b.3, and 3.6.1.3.b.4 are not applicable if both doors in the same air lock are inoperable and ACTION 3.6.1.3.c is entered.

2. Entry and exit of containment is permissible under the control of a dedicated individual.

Air lock doors in high radiation areas may be verified closed by administrative means.

SHEARON HARRIS - UNIT 1 3/4 6-4a Amendment No. 181

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE by:

a. Performing required air lock leakage rate testing in accordance with the Containment Leakage Rate Testing Program, as modified by the approved exemption###.

b. At the frequency specified in the Surveillance Frequency Control Program by verifying that only one door in the air lock can be opened at a time**.

1. 1. An inoperable air lock door does not invalidate the previous successful performance of the overall airlock leakage test.

- Only required to be performed upon entry or exit through the containment air lock. (If Surveillance Requirement 4.6.1.3.b has not been performed in the interval specified by the Surveillance Frequency Control Program, then perform Surveillance Requirement 4.6.1.3.b during the next containment entry through the associated air lock.)

SHEARON HARRIS - UNIT 1 3/4 6-5 Amendment No. 181

CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.1.

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

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.6.1 Containment Vessel Surfaces. The structural integrity of the exposed accessible interior and exterior surfaces of the containment vessel, including the liner plate, shall be determined, during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.1.c), by a visual inspection of these surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnormal degradation. Additional inspections shall be conducted in accordance with Subsections IWE and IWL of the ASME Boiler and Pressure Vessel Code,Section XI.

SHEARON HARRIS - UNIT 1 3/4 6-8 Amendment No. 181

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

k. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54 (o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical Report (TR)

NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, with the following exception noted:

Visual examination of the containment system shall be in accordance with Specification 4.6.1.6.1.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident is 41.8 psig. The calculated peak containment internal pressure related to the design basis main steam line break is 41.3 psig. Pa will be assumed to be 41.8 psig for the purpose of containment testing in accordance with this Technical Specification.

The maximum allowable containment leakage rate, La at Pa, shall be 0.1 % of containment air weight per day.

Leakage rate acceptance criteria:

1) The containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La for Type A tests.

2) Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is 0.05 La when tested at Pa.

b) For each door, leakage rate is 0.01 La when pressurized to Pa.

The provisions of Surveillance Requirement 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

Nothing in these Technical Specifications shall be construed to modify the testing frequencies required by 10 CFR 50, Appendix J.

SHEARON HARRIS - UNIT 1 6-19c Amendment No. 181

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 181 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 application dated November 8, 2019 (Reference 1), as supplemented by letter dated April 16, 2020 (Reference 2) Duke Energy Progress, LLC (the licensee), requested changes to the technical specifications (TSs) for the Shearon Harris Nuclear Power Plant (Harris or HNP),

Unit 1. The license amendment request (LAR) proposes changes to Technical Specification (TS) 6.8.4.k, Containment Leakage Rate Testing Program, by replacing the reference to Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program (Reference 3), with a reference to Nuclear Energy Institute (NEI) 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, (Reference 4) and the conditions and limitations specified in NEI 94-01, Revision 2-A, of the same name, dated October 2008 (Reference 5).

The proposed amendment would allow a permanent extension of the Type A test interval from 10 years to 15 years, a more conservative allowable test interval extension of 9 months for Type A, Type B and Type C leakage rate tests, and an extension of the Type C test interval up to 75 months, based on acceptable performance history as defined in NEI 94-01, Revision 3-A.

The proposed amendment also adopts Title 10 of the Code of Federal Regulations, Part 50 (10 CFR), Domestic Licensing of Production and Utilization Facilities, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B, Performance-Based Requirements, subject to certain NRC-approved exemptions, for the performance-based testing of Type B and C tested components and the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8-2002, Containment System Leakage Testing Requirements.

The supplement dated April 16, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staffs initial proposed no significant hazards consideration determination as published in the Federal Register on January 28, 2020 (85 FR 5052).

2.0 REGULATORY EVALUATION

2.1 System Descriptions Section 3.1, Description of Primary Containment System, of the licensees letter dated November 8, 2019, reads, in part:

The Concrete Containment Structure (CCS) is a steel lined reinforced concrete structure in the form of a vertical right cylinder with a hemispherical dome and a flat base with a recess beneath the reactor vessel. No pre-stressing tendon system is employed in the containment design and construction.

The structure consists of a cylindrical wall measuring 160 ft. in height from the liner on the base to the springline of the dome and has an inside diameter of 130 ft. The cylinder wall is 4 ft.-6 in. thick. The inside radius of the 2 ft.-6 in.

thick dome is equal to that of the cylinder so that the discontinuity at the spring line due to the change in thickness is on the outer surface.

The basic structural elements considered in the design of the containment structure are the basemat, cylinder wall, and dome. These act essentially as one structure under all loading conditions. The nominal liner plate is 3/8 in. thick in the cylinder, 1/4 in. thick on the bottom, and 1/2 in. thick in the dome. The liner is anchored to the concrete shell by means of anchor studs fusion welded to the liner plate so that it forms an integral part of the containment structure. The liner functions primarily as a leaktight membrane.

The base mat consists of a 12 ft. thick structural concrete slab and a metal liner.

The liner is welded to inserts embedded in the concrete slab. The base liner is covered with concrete, the top of which forms the floor of the containment. The base mat is supported by sound rock. The entire mat is structurally independent of adjacent Seismic Category I foundations.

An impervious plastic waterproofing membrane is located between the containment foundation mat and the ground. Before laying the membrane, a concrete leveling surface was placed on the rock. After installing the membrane, a concrete protective layer was installed before placing reinforcement for the foundation mat. The waterproofing membrane for the Containment Building is continuous under the containment foundation mat and terminates into waterstops at the joints with adjacent structures.

Containment Penetrations The equipment hatch is a welded steel assembly having an inside diameter of 24 ft. 0 in. with a weld-on cover with sufficient material to initially allow for six removals and re-welding. A 15 ft.-0 in. inside diameter (ID) bolted cover is provided in the equipment hatch cover for passage of smaller equipment during plant operation. Provision is made to pressurize the space between the gaskets of the bolted hatch cover to meet the requirements of 10 CFR Part 50, Appendix J.

One breech-type personnel air lock and one personnel emergency air lock are provided. Each lock is a welded steel assembly having two doors which are double-gasketed with material resistant to radiation. Provisions are made to pressurize the space between the gaskets.

Mechanical penetrations are divided into two general types:

Type I - High pressure, high-temperature piping (above 200°F).

The process pipe is connected to a containment penetration sleeve (which is partially embedded in the concrete wall) by a forged flued head fitting. The flued head fittings are designed to carry the forces and moments due to the normal operating conditions and due to the postulated pipe rupture loads by transferring these forces to the containment penetration sleeves and further into the concrete containment wall.

Type II - General piping (penetrations which are subject to only relatively small pipe rupture forces and temperatures up to 200°F).

The process pipe passes through a containment penetration sleeve which is partially embedded and anchored into the concrete wall. The annular gap between the process pipe and the sleeve is sealed on both the inside and outside faces of the concrete wall. The inside plate is designed to withstand the internal pressure and to transfer all of the normal operating loads and/or the postulated accident piping rupture loads from the piping system to the penetration sleeve and then into the concrete wall. The outside seal is flexible to accommodate thermal expansion movements.

Type II penetrations also include heating, ventilation, and air conditioning (HVAC) penetrations and groups of small diameter lines (instrument, sampling lines) which incorporate socket weld couplings welded to closure plates. Two categories of penetrations are included in Type II penetrations: Type IIA for single tubing or multiple pipes and/or tubings, and Type IIB for single pipe.

HVAC penetration sleeves, 48 in. and 24 in. diameter, are mechanical Type II penetration sleeves.

A fuel transfer penetration is provided to transport fuel assemblies between the refueling cavity in the containment and the fuel transfer canal in the Fuel Handling Building. This penetration consists of a 20 in. diameter stainless steel pipe installed inside a 26 in. pipe. The inner pipe acts as the transfer tube and is fitted with a double-gasketed blind flange in the refueling cavity and a standard gate valve in the fuel transfer canal. The penetration sleeve is welded to the steel liner and anchored into the concrete wall. Provision is made for testing welds essential to the integrity of the liner. Bellows expansion joints are provided to compensate for any differential movement between the structures due to operating thermal expansion and seismic movements. The fuel transfer tube expansion joints are not part of the containment pressure boundary; rather, the transfer tube is rigidly attached to the containment penetration sleeve.

Electrical penetrations are included within the Type III penetrations. Modular type penetrations are used for all electrical conductors passing through the

containment wall. Each penetration assembly consists of a stainless-steel header plate attached to a carbon steel welded ring, which is in turn welded to the pipe sleeve.

2.2 Description of Changes The licensee proposed to revise Harris Technical Specifications, as follows:

TS LCO 4.6.11c currently states, as follows:

After each closing of each penetration subject to Type B testing, except the Z containment air locks, if opened following a Type A or B test, by leak rate testing Z the seal with gas at a pressure not less than Pa, and verifying that when the Z measured leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2a. for all other Type B and C penetrations, the Z combined leakage rate is less than 0.60 La.

TS LCO 4.6.11c will state, as follows:

By performing required visual examinations and leakage rate testing, except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program.

TS LCO 3.6.1.2 currently states, as follows:

Containment leakage rates shall be limited to:

a. An overall integrated leakage rate within limits specified in the Containment Leakage Rate Testing Program.

b. A combined leakage rate of less than or equal to 0.60La for all penetrations and valves subject to Type B and C tests when pressurized to Pa.

TS LCO 3.6.1.2 will state, as follows:

Containment leakage rates shall be within the limits specified in the Containment Leakage Rate Testing Program.

TS LCO 3.6.1.2 Action statement currently states, as follows:

With either the measures overall integrated containment leakage rate exceeding 0.75 La or the measures combined leakage rate for all penetrations and valves subject to Types B and C tests exceeded 60 La, restore the overall integrated leakage rate to less than 0.75 La, and the combined leakage rate for all penetrations subject to Type B and C tests to less than 0.60 La prior to increasing the Reactor Coolant System temperature above 200 °F.

TS LCO 3.6.1.2 Action statement will state, as follows:

With the containment leakage rate not within the limits specified in the Containment Leakage Rate Testing Program, restore the leakage rate to within the limits specified in the Containment Leakage Rate Testing Program prior to increasing the Reactor Coolant System temperature above 200°F.

Surveillance Requirement 4.6.1.2 currently states, as follows:

The Type A containment leakage rate tests shall be performed in accordance with the Containment Leakage Rate Testing Program described in Technical Specification 6.8.4.k. The Type B and Type C containment leakage rate tests shall be demonstrated at the test schedule and shall be determined in conformance with the criteria specified in 10 CFR Part 50, Appendix J, Option A.

a. Type B and C tests shall be conducted with gas at pressure not less than Pa, at intervals no greater than 24 months except for test involving:

1. Air locks.

2. Containment purge makeup isolation valves with resilient material seals.

b. Air locks shall be tested and demonstrated OPERABLE by the requirement of Specification 4.6.1.3.

c. Purge makeup and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE by the requirements of Specifications 4.6.1.7.2.

d. The provisions of Specification 4.0.2 are not applicable.

Surveillance Requirement 4.6.1.2 will state, as follows:

The containment leakage rate tests shall be performed in accordance with the Containment Leakage Rate Testing Program described in Technical Specification 6.8.4.k.

TS 3.6.1.3 Action statement, Note 3, currently states, as follows:

Enter 3.6.1.1 LCO for Containment Integrity when the air lock leakage results in exceeding the overall containment leakage rate, Specification 3.6.1.2.a.

TS 3.6.1.3 Action statement, Note 3, will state, as follows:

Enter 3.6.1.1 LCO for "Containment Integrity" when the air lock leakage results in exceeding the containment leakage rate, Specification 3.6.1.2.

TS 3.6.1.3 Action statement c.1, currently states, as follows:

Immediately initiate action to evaluate overall containment leakage rate per LCO 3.6.1.2, and TS 3.6.1.3 Action statement c.1, will state, as follows:

Immediately initiate action to evaluate containment leakage rate per LCO 3.6.1.2, and Surveillance Requirement (SR) 4.6.1.3 currently states, as follows:

Each containment air lock shall be demonstrated OPERABLE by:

a. Performing required air lock leakage rate testing in accordance with 10 CFR Part 50, Appendix J, as modified by approved exemptions###. The acceptance criteria for air lock testing are:

1. Overall air lock leakage rate is .05 La when tested at Pa.

2. For each door, leakage rate is .01 La when tested at Pa.

b. At the frequency specified in the Surveillance Frequency Control Program by verifying that only one door in the air lock can be opened at a time**.

SR 4.6.1.3 will state, as follows:

Each containment air lock shall be demonstrated OPERABLE by:

Performing required air lock leakage rate testing in accordance with the Containment Leakage Rate Testing Program, as modified by the approved exemption###.

The licensee proposed to delete footnote ### 2 to on the bottom of page 3/4 6-5 and the numbering to the remaining footnote 1 is deleted, as there is only one footnote remaining.

The licensee proposed to revise the reference in SR 4.6.1.6.1 from 4.6.1.2 to 4.6.1.1.c.

TS 6.8.k currently states, as follows:

k. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54 (o) and 10 CFR Part 50, Appendix J. Option B, as modified by approved exemptions. This program shall be in conformance with the NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, with the following exceptions noted:

1) The above Containment Leakage Rate Testing Program is only applicable to Type A testing. Type B and C testing shall continue

to be conducted in accordance with the original commitment to 10 CFR Part 50, Appendix J, Option A.

2) The first Type A test performed after the May 23. 1997 Type A test shall be performed no later than May 23. 2012.

3) Visual examination of the containment system shall be in accordance with Specification 4.6.1.6.1.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident is 41.8 psig. The calculated peak containment internal pressure related to the design basis main steam line break is 41.3 psig. Pa will be assumed to be 41.8 psig for the purpose of containment testing in accordance with this Technical Specification.

The maximum allowable containment leakage rate, La at Pa, shall be 0.1 % of containment air weight per day.

The containment overall leakage rate acceptance criterion is 1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La for Type A tests.

The provisions of Surveillance Requirement 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. However. test frequencies specified in this Program may be extended consistent with the guidance provided in Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR [Code of Federal Regulations]

Part 50, Appendix J, as endorsed by Regulatory Guide 1.163.

Specifically, NEI 94-01 has this provision for test frequency extension:

1) Consistent with standard scheduling practices for Technical Specifications Required Surveillances, intervals for recommended Type A testing may be extended by up to 15 months. This option should be used only in cases where refueling schedules have been changed to accommodate other factors.

The provisions of Surveillance Requirement 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

TS 6.8.k will state, as follows:

k. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54 (o) and 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical

Report (TR), NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, with the following exception noted:

Visual examination of the containment system shall be in accordance with Specification 4.6.1.6.1.

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident is 41.8 psig. The calculated peak containment internal pressure related to the design basis main steam line break is 41.3 psig. Pa will be assumed to be 41.8 psig for the purpose of containment testing in accordance with this Technical Specification.

The maximum allowable containment leakage rate, La at Pa, shall be 0.1 % of containment air weight per day.

Leakage rate acceptance criteria:

1) The containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La for Type A tests.

2) Air lock testing acceptance criteria are:

a) Overall air lock leakage rate is 0.05 La when tested at Pa.

b) For each door, leakage rate is 0.01 La when pressurized to Pa.

The provisions of Surveillance Requirement 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

Nothing in these Technical Specifications shall be construed to modify the testing frequencies required by 10 CFR Part 50, Appendix J.

2.3 Applicable Regulatory Requirements and Guidance Regulations Section 50.36(c)(2) to 10 CFR Part 50 states, in part, that the limiting conditions for operation (LCOs) are the lowest functional capability or performance level of equipment required for safe operation of the facility, and when LCOs are not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the LCO can be met.

Section 50.36(c)(3) to 10 CFR Part 50 requires, in part, that TSs include surveillance requirements (SRs), which are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

Section 50.36(c)(5) to 10 CFR Part 50 requires, in part, the inclusion of administrative controls in TSs that are necessary to ensure operation of the facility in a safe manner.

Section 50.54(o) to 10 CFR Part 50 requires, in part, that primary reactor containments for water-cooled power reactors be subject to the requirements in 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.

Appendix J contains two options: Option A - Prescriptive Requirements and Option B -

Performance-Based Requirements, either of which can be used to meet Appendix J requirements. The testing requirements in Appendix J ensure that (1) leakage through containments or systems and components penetrating containments does not exceed allowable leakage rates specified in the TSs and (2) integrity of the containment structure is maintained during the service life of the containment. Harris adopted 10 CFR Part 50, Appendix J, Option B, Performance-Based Requirements for Type A (integrated leak rate test (ILRT)) by Amendment No. 91, dated September 17, 1999 (ADAMS Accession No. ML020600010). Harris testing requirements for 10 CFR Part 50, Appendix J, Type B primary reactor containment penetrations and Type C containment isolation valves both remain governed by 10 CFR Part 50, Appendix J, Option A, Prescriptive Requirements.

Section 50.55a(g)(4) to 10 CFR Part 50 requires, in part, containment inservice inspection requirements, which, in conjunction with the requirements of 10 CFR Part 50, Appendix J, ensure the continued leak-tight and structural integrity of the containment during its service life.

Section 50.65 to 10 CFR Part 50 requires, in part, that the licensee:

shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components, are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and where practical, take into account industrywide operating experience.

Appendix J, Option B,Section V.B.3 to 10 CFR Part 50 requires, in part, that the TSs include, by general reference, the RG or other implementation document used by the licensee to develop a performance-based leakage-testing program. The submittal for TS revisions must also contain justification, including supporting analyses, if the licensee chooses to deviate from methods approved by the NRC.

Appendix J, Option B to 10 CFR Part 50 specifies performance-based requirements and criteria for preoperational and subsequent leakage rate testing. These requirements are met by:

1. Type A tests to measure the containment system overall integrated leakage rate,

2. Type B pneumatic tests to detect and measure local leakage rates across pressure retaining leakage-limiting boundaries such as penetrations, and

3. Type C pneumatic tests to measure containment isolation valve (CIV) leakage rates.

After the containment system has been completed and is ready for operation, Type A tests are conducted at periodic intervals based on the historical performance of the overall containment system to measure the overall integrated leakage rate. The leakage rate test results must not exceed the maximum allowable leakage rate (La) at design-basis loss-of-coolant accident (DBLOCA) pressure (Pa) with margin, as specified in the TSs. Option B also requires that a general visual inspection for structural deterioration of the accessible interior and exterior surfaces of the containment system, which may affect the containment leak-tight integrity, be

conducted prior to each Type A test and at a periodic interval between tests based on the performance of the containment system.

Appendix J, Option A, III.D.2 to 10 CFR Part 50, reads, in part, Type B tests, except tests for air locks, shall be performed during reactor shutdown for refueling, or other convenient intervals, but in no case at intervals greater than 2 years. Air locks shall be tested prior to initial fuel loading and at 6-month intervals thereafter at an internal pressure not less than Pa.

Appendix J, Option A, III.D.3 to 10 CFR Part 50, reads, Type C tests shall be performed during each reactor shutdown for refueling but in no case at intervals greater than 2 years.

Regulatory Guidance NEI 94-01, Revision 0 (Reference 6), provides methods for complying with Option B of 10 CFR Part 50, Appendix J, and allows for the extension of the performance-based Type A test interval for up to 10 years, based upon two consecutive successful tests. NEI 94-01, Revision 0 was endorsed by the NRC in RG 1.163, with conditions.

NEI 94-01, Revision 2-A, incorporated the NRC conditions in RG 1.163 and added provisions for extending Type A test intervals up to 15 years. This revision of NEI 94-01 was supported by Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, dated August 2007 (Reference 7). The EPRI report provides a generic assessment of the risks associated with permanently extending the ILRT interval to 15 years, and it provides a risk-informed methodology to be used to confirm the risk impact of the ILRT extension on a plant-specific basis. Probabilistic risk assessment (PRA) methods are used in combination with ILRT performance data and other considerations to justify the extension of the ILRT interval. This is consistent with guidance provided in RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference 8), and RG 1.177, An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications (Reference 9), to support changes to test intervals.

The NRC staffs review of both NEI 94-01, Revision 2 (Reference 36), and EPRI Report No. 1009325, Revision 2, is described in an NRC safety evaluation dated June 25, 2008 (Reference 10). This NRC safety evaluation states that NEI 94-01, Revision 2, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR Part 50, Appendix J, Option B. The NRC staff concluded that NEI 94-01, Revision 2-A, is acceptable for referencing by licensees proposing to amend their containment leakage rate testing TSs, subject to the conditions listed in Section 4.1 of the safety evaluation.

NEI 94-01, Revision 3-A, added guidance for extending Type C local leak rate test (LLRT) intervals beyond 60 months. In a June 8, 2012 (Reference 11) safety evaluation, the NRC staff concluded that NEI 94-01, Revision 3, describes an acceptable approach for implementing the optional performance-based requirements of Appendix J, and is acceptable for reference by licensees proposing to amend their containment leakage rate testing TSs, subject to two conditions related to Type C testing. The safety evaluation was incorporated into Revision 3 and subsequently issued as NEI 94-01, Revision 3-A, on July 31, 2012.

RG 1.174, Revision 3, describes an acceptable risk-informed approach for assessing the nature and impact of proposed permanent licensing basis changes by considering engineering issues

and applying risk insights. This RG also provides risk acceptance guidelines for evaluating the results of such evaluations.

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference 12), describes an acceptable approach for determining whether the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision making for light water reactors. RG 1.200 provides guidance for assessing the technical adequacy of a PRA. Revision 2 of RG 1.200, endorses, with clarifications and qualifications, the use of the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Standard, RA Sa 2009, Addenda to ASME RA S 2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (i.e., the PRA Standard).

Regulatory Issue Summary (RIS) 2007-06, Regulatory Guide 1.200 Implementation (Reference 13), describes how the NRC staff will use Revision 2 of RG 1.200 to assess technical adequacy of the PRA used to support risk informed applications received after March 2010.

3.0 TECHNICAL EVALUATION

3.1 Type A Integrated Leak Rate Test Harris, Unit 1, TS 6.8.4.k, currently requires Type A testing in accordance with RG 1.163, which endorses the methodology for complying with 10 CFR Part 50, Appendix J, Option B. Since the adoption of Option B in September 1999, the performance leakage rates have been calculated in accordance with NEI 94-01, Revision 0, Section 9.1.1 for Type A testing.

Per TS 6.8.4.k, the Harris containment has a maximum allowable leakage rate, La, at the peak design containment internal accident pressure, Pa, of 0.1 percent of containment air weight per day; here, Pa equals 41.8 pounds per square inch, gauge (psig).

Since 1986, a total of five ILRTs have been performed on the Harris containment. All five ILRTs had satisfactory leakage rate results. These five ILRT results were documented in the two tables contained in LAR Section 3.2.4 Integrated Leakage Rate Testing History. These test results are consolidated in Table 3.1.1-1 below:

Table 3.1.1-1 Harris, Unit 1 Type A ILRT History Test Upper 95% Level As-Left Min Adjusted As Found As Left Date Confidence Corrections Pathway As Left Acceptance Acceptance Level (wt.%/day) Penalty for Leak Rate Criteria, La Criteria (wt.%/day) Isolated (Wt.%/day) (Wt.%/day) (Wt.%/day)

(Test Pathways Pressure) (wt.%/day)

Feb. 0.05199(1) (3) (3) (3) 0.100 0.075 1986(6) (0.75La)

Oct. 0.0406(1) (3) (3) (3) 0.100 0.075 1989 (0.75La)

Sept. 0.0701(1) (3) (3) (3) 0.100 0.075 1992 (0.75La)

May 0.0265(1)(4) -0.0000 0.0004 0.0269 0.100 0.075 1997 (44.0 psig) (0.75La)

May 0.0605(2)(4) -0.0000 0.0008 0.0613 0.100 0.075 2012 (43.3 psig) (0.75La)

Table 3.1.1-1 Notes:

(1) As was specified in TS 6.8.4.k, the maximum allowable containment leakage rate La, at Pa of 41.2 psig, is 0.1% of primary containment air weight per day (prior to Amendment No. 107 as issued October 2001).

(2) As specified in TS 6.8.4.k, the maximum allowable containment leakage rate L , at P of 41.8 a a psig, is 0.1% of primary containment air weight per day (reference Amendment No. 107 for Steam Generator Replacement and Power Uprate).

(3) Data not provided in LAR (4) Test Method / Data Analysis Techniques: Absolute Method / Mass Point Analysis (5) Preoperational The NRC issued Harris License Amendment No. 122 on March 30, 2006 (Reference 14). This amendment authorized a one-time extension of the 10 CFR Part 50, Appendix J, Option B, Type A, Containment ILRT interval from once in 10 years to once in 15 years. The ILRT interval for Harris reverted to 10 years after the ILRT completion in May 2012.

The NRC staff notes that the last sentence of Section 9.2.3, Extended Test Intervals of NEI 94-01, Revision 3-A states In the event where previous Type A tests were performed at reduced pressure (as described in 10 CFR 50, Appendix J, Option A), at least one of the two consecutive periodic Type A tests shall be performed at peak accident pressure (Pa). Section 9.1.2 of NEI 94-01 states, in part, The elapsed time between the first and the last tests in a series of consecutive passing tests used to determine performance shall be at least 24 months.

The NRC staff confirmed that the Pa requirement of Section 9.2.3 of NEI 94-01, Revision 3-A, has been satisfied, as both ILRTs of May 1997 and May 2011, were performed at a test pressure within the limitations of ANSI 56.8-1994, Section 3.2.11, Type A Test Pressure.

More specifically, the ILRT Pressure, Pt, can range from 0.96Pac<Pt<Pd where:

Pa = 41.8 psig Pac= the calculated peak containment internal accident pressure related to the design-basis Accident (DBA)

Therefore Pa = Pac 0.96Pac = 0.96Pa = 40.1 psig Pd = Containment Design Pressure Section 3.1.3, Relief Request I3R-18, Regarding Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, Inservice Inspection Program for Containment, Third Ten-Year Interval (Page 18 of 106) of the licensees letter dated November 8, 2019, lists the containment design pressure, Pd, as 45 psig. As can be seen in Table 3.1.1-1 (above), Type A tests of May 1997 and May 2011 were performed within the limitations of ANSI 56.8-1994. Therefore, the above requirements Sections 9.1.2 and 9.2.3 of NEI 94-01, Revision 3-A, have been satisfied.

Currently, TS 6.8.4.k references RG 1.163. Regulatory Position C of RG 1.1.63 in turn, states that NEI 94-01, Revision 0, provides methods acceptable to the NRC staff for complying with the provisions of Option B in Appendix J to 10 CFR Part 50. The third paragraph of Section 9.2.3, Extended Test Intervals, of NEI 94-01, Revision 0 states, in part:

In reviewing past performance history, Type A test results may have been calculated and reported using computational techniques other than the Mass Point method from ANSI/ANS-56.8-1994 (e.g., Total Time or Point-to-Point).

Reported test results from these previously acceptable Type A tests can be used to establish the performance history. Additionally, a licensee may recalculate past Type A Upper Confidence Limit (UCL) (using the same test intervals as reported) in accordance with ANSI/ANS-56.8-1994 Mass Point methodology and its adjoining Termination criteria in order to determine acceptable performance history.

NEI 94-01, Revision 3-A, reiterates this guidance, and identifies the test standard as ANSI/ANS-56.8-2002.

The NRC staff notes that Section 9.2.3 of NEI 94-01, Revision 0, does not mandate that a licensee recalculate past Type A test results to demonstrate conformance with the definition of performance leakage rate contained in NEI 94-01, Revision 3-A. The NRC staff also notes that the IRLT results since May 1997 demonstrated ample margin (i.e., > 39 percent) between each As-found Leakage Rate value and La. Accordingly, the NRC staff did not request that the licensee reconstitute the Type A test results from earlier than the ILRT of May 1997.

TS 6.8.4.k establishes the maximum limit for the As-Left Leakage Rate for Unit startup following completion of Type A testing at 0.75 La, which equals 0.075 percent of containment air weight per day.

The past two ILRTs results dating back to 1997 have confirmed that the containment leakage rates are acceptable with respect to the allowable leakage criterion of percent containment air weight (La) per day. Since the last two Type A tests for Harris had As Found Leakage Rate test results of less than 1.0 La at the peak design containment internal accident pressure (Pa), a test frequency of 15 years in accordance with NEI 94-01, Revision 3-A, and the conditions and limitations of NEI 94-01, Revision 2-A, would be acceptable. Based on the last two Harris ILRT

results, the NRC staff concludes that the requirements of Sections 9.1.2 and 9.2.3 of NEI 94-01, Revision 3-A, have been satisfied.

3.2 Type B and Type C Leak Rate Testing Program The Type B and C testing program requires testing of electrical penetrations, airlocks, hatches, flanges and CIVs in accordance with 10 CFR Part 50, Appendix J, Option A, Prescriptive Requirements.

Currently, Harris TS 6.8.4.k, Containment Leakage Rate Testing Program states, in part:

The calculated peak containment internal pressure related to the design basis loss-of-coolant accident is 41.8 psig. The calculated peak containment internal pressure related to the design basis main steam line break is 41.3 psig. Pa will be assumed to be 41.8 psig for the purpose of containment testing in accordance with this Technical Specification.

The maximum allowable containment leakage rate, La at Pa, shall be 0.1% of containment air weight per day.

The containment overall leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the combined Type B and Type C tests, and 0.75 La for Type A tests.

Section 3.4.5, Containment Leakage Rate Testing Program - Type B and Type C Testing Program of the licensees letter dated November 8, 2019, contains Table 3.4.5-1 HNP [Harris Nuclear Plant] Unit 1 Types B and C LLRT Combined As-Left Trend Summary. The NRC staff reviewed the local leak rate summaries from the last seven refueling outages contained therein.

The licensee states, in part, In accordance with TS 6.8.4.k, the allowable maximum pathway total Type B and C leakage is 0.60 La (101,200 standard cubic centimeters per minute (sccm))

where La equals 168,800 sccm.

With the use of these La values, and the data contained in subject table; the NRC staff confirmed the accuracy of the Percentage of 0.6La values contained in the table. The NRC staff concluded that the As-Left maximum pathway leakage rates for the last seven refueling outages since 2009 had an average value of 21.6 percent of 0.6La with a high of 33.6 percent 0.6La.

Section 3.4.6 Type B and Type C Local Leak Rate Testing Program Implementation Review of the licensees letter dated November 8, 2019, contains Table 3.4.6-1 HNP Unit 1 Type B and C LLRT Program Implementation Review, which provides a listing of CIVs that did not demonstrate acceptable performance during the two most recent refueling outages 2016-H1R20 and 2018-H1R21.

The NRC staff observed that of all the Harris Type B and Type C containment penetrations, only Penetration M-88 recorded failures on two different isolation valves during each of the last two outages. No other penetration or component showed failures in successive outages.

CIV 1SP-200, which caused the failure of the Type C LLRT during 2016-H1R20, was reworked during 2018-H1R21 and passed its retest. CIV 1SP-201 caused the failure of the Type C LLRT

for Penetration M-88 during 2018-H1R21. Valve 1SP-201 exceeded its administrative limit acceptance criteria. A work request (WR) was issued and the valve was reworked. However, the valve exhibited a higher than the allowed leakage rate (i.e., above the administrative limit) during its retest. The higher leak rate was evaluated against design requirements and was accepted based on Unit 1 Types B and C LLRT Combined As-Left margin (i.e., > 66%) to the TS 6.8.4.k limit of 0.6 La. The reworking of Valve 1SP-201 was planned for the subsequent refueling outage, 2020-H1R21. The rework and retesting of Valve 1SP-201 had not yet occurred when this safety evaluation was prepared. However, the NRC staff determined that the large amount of margin (> 66%), provided reasonable assurance that public health and safety will be adequately protected upon implementation of this amendment.

The regulations in 10 CFR Part 50, Appendix J, Option A, regarding test programs for Type B and C LLRTs, require that every penetration be tested each refueling outage. The lack of repeat failures exhibited in the licensees Table 3.4.6-1 and the exhibited leakage summary and margins to the allowable leakage limit of TS 6.8.4.k contained in LAR Table 3.4.5-1, indicates that the Harris maintenance program has been sufficiently effective in providing a leak-tight containment.

The NRC staff notes that 10 CFR Part 50, Appendix J, Option A, does not require as-found testing for Type B and Type C penetrations. Given the industry performance since Appendix J, Option B was codified into 10 CFR Part 50, the staffs expectation is that the as-found test totals of the licensees Table 3.4.5-1 would be acceptable based on historical industry test results from plants that have adopted Option B for the testing of Type B and Type C penetrations.

The licensee stated that Upon implementation of the proposed amendments to the HNP TS, as-found LLRT testing will be required in accordance with the requirements of NEI 94-01, Revision 3-A, Section 10.2.1 for Type B Test Intervals, and Section 10.2.3 for Type C Test Intervals.

Based on the NRC staffs review of the historical information provided in Section 3.4.6 of the licensees letter dated November 8, 2019, the NRC staff observed that there was no indication of the licensees failure to adequately implement the prescriptive requirements of its 10 CFR Part 50, Appendix J, Option A program. Moreover, the licensee provided sufficient corrective actions for the CIV administrative limit failures associated with the sole repetitive LLRT Type C penetration (i.e., M-88) failure experienced during the last two refueling outages.

From its review of the information contained in Section 3.4.6, the NRC staff finds there is substantial evidence that the licensee has been compliant in satisfying the prescriptive requirements of 10 CFR Part 50, Appendix J, Option A.

Furthermore, the licensee acknowledged that with the adoption of the proposed TS amendment, the recordkeeping for Harris will reflect the following requirements:

10 CFR Part 50, Appendix J Option B, Section IV:

The results of the preoperational and periodic Type A, B, and C tests must be documented to show that performance criteria for leakage have been met. The comparison to previous results of the performance of the overall containment system and of individual components within it must be documented to show that the test intervals established for the containment system and components within it are adequate. These records must be available for inspection at plant sites.

NEI 94-01, Revision 3-A, Section 12.1, Report Requirements:

A post-outage report shall be prepared presenting results of the previous cycles Type B and Type C tests, and Type A, Type B, and Type C tests, if performed during that outage. The technical contents of the report are generally described in ANSI/ANS-56.8-2002 and shall be available on-site for NRC review. The report shall show that the applicable performance criteria are met and serve as a record that continuing performance is acceptable. The report shall also include the combined Type B and Type C leakage summation, and the margin between the Type B and Type C leakage rate summation and its regulatory limit. Adverse trends in the Type B and Type C leakage rate summation shall be identified in the report and a corrective action plan developed to restore the margin to an acceptable level.

Based on the information discussed above, NRC staff concludes that the licensee has demonstrated a history of adherence to the prescriptive requirements of Appendix J Option A.

Furthermore, in the LAR the licensee has acknowledged that in adopting the performance-based requirements of 10 CFR Part 50, Appendix J, Option B, for the testing of Type B and Type C penetrations:

the guidance of NEI 94-01, Revision 3-A and ANSI/ANS-56.8-2002 replaces the guidance of ANSI/ANS 56.8-1987; the Type B penetration test intervals will be subject to the requirements NEI 94-01, Revision 3-A Section 10.2.1; the Type C penetration test intervals will be subject to the requirements NEI 94-01, Revision 3-A Section 10.2.3; and post-outage reports shall be prepared presenting results of the previous cycles Type B and Type C tests, and Type A, Type B, and Type C tests in accordance with the requirements of NEI 94-01 Revision 3-A, Section 12.1, Report Requirements.

For the Harris containment, the conclusion above supports allowing an extended test interval of up to 120 months for Type B tested penetrations and of up to 75 months for the Type C tested CIVs in accordance with the guidance of NEI 94-01, Revision 3-A.

3.3 Containment Inspection and Testing Programs 3.3.1 Containment Inservice Inspection Program In Section 3.0, Technical Evaluation, of the licensees letter dated November 8, 2019, the licensee described the physical characteristics of the cylindrical reinforced concrete containment structure, with a hemispherical dome and a flat basemat with a recess at the center to house the reactor pressure vessel as well as a welded liner plate anchored on the entire inner surface of the containment structure. The licensee also described the portion of welded containment liner plate that is embedded in the slab and is inaccessible for visual inspections. The licensee described the containment building penetrations of the Equipment Hatch, Personnel Air Locks, Mechanical/Electrical Penetrations, Fuel Transfer Tube and Sump Line Valve Chambers.

As required by 10 CFR 50.55a(g)(4)(ii), periodic visual inspections of accessible interior and exterior surfaces of the reinforced concrete containment structure shall be performed in accordance with the American Society of Mechanical Engineers (ASME Boiler and Pressure Vessel Code (B&PV Code)),Section XI, Subsections IWE and IWL to identify problems affecting the performance of pressure tests performed prior to initiating internal pressure tests to maintain the defense-in-depth safety margins of the containment during its service-life.

In Subsection 3.4.2, Containment Inservice Inspection (CISI) Program, of its letter dated November 8, 2019, the licensee provided information related to the inservice inspections (ISI) performed at Harris, Unit 1. The licensee stated that Harris complies with the 2007 Edition of ASME Section XI, through the 2008 Addenda, Rules for Inservice Inspection of Nuclear Power Plant Components, Subsections IWE and IWL Programs and to the conditions specified in 10 CFR 50.55a(b)(2). The licensee also credited CISI programs as described in the license renewal Safety Evaluation Report (SER), NUREG-1916, Safety Evaluation Report Related to the License Renewal of Shearon Harris Nuclear Power Plant, Volume 1, published November 2008 (Reference 15).

In Subsection 3.4.2 of its November 8, 2019, letter, the licensee described historical dates for intervals and periods for CISI in Tables 3.4.2-1 and 3.4.2-2 for Metal Containment (Class MC) components (IWE) and Tables 3.4.2-3 and 3.4.2-4 for Concrete Containment (Class CC) components (IWL). The licensee confirmed that the ASME Subsection XI inspections are performed per the requirements of 10 CFR 50.55a(g)(4)(ii) and described the implementation of the following NRC conditions for ASME Section XI in 10 CFR 50.55a:

10 CFR 50.55a(b)(2)(vi): use of applicable editions and addenda of ASME Section XI, Subsections IWE and IWL.

10 CFR 50.55a(b)(2)(viii)(E): requirements for the inspection of inaccessible areas of concrete containment.

10 CFR 50.55a(b)(2)(ix)(A): requirements for the inspection of inaccessible areas of metal containment.

The licensee used portions of the NRC-endorsed ASME Code Cases in NRC RG 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 17, dated August 2014, (Reference 16), even though they are not in the program plan.

In Subsection 3.4.3, Supplemental Inspection Requirements, of the November 8, 2019 letter, the licensee described the applicability of NRC guidance of visual inspection scheduling in SER Section 3.1.1.3, Adequacy of Pre-Test Inspections (Visual Examinations), of NEI 94-01, Revision 2-A.

The licensee also referred to the TS Surveillance Requirement (SR) 4.6.1.6.1, Containment Vessel Surface, for assuring that additional inspections be conducted in accordance with Subsections IWE and IWL and the visual inspection requirements of NEI 94-01, Revision 3-A.

In Subsection 3.4.4, Results of Recent Containment Examinations, of the November 8, 2019, letter, the licensee presented the results of recent visual inspections for IWE (VT-3) in Tables 3.4.4-1 and 3.4.4-3, and for IWL (VT-3C) in Table 3.4.4-2. The licensee evaluated the inspection results and concluded that they were acceptable. Based on this information, the

NRC staff concluded that the licensee had no reportable degradation and moisture intrusion in the containment liner plate.

In Section 3.5, Operating Experience (OE), of the November 8, 2019 letter, the licensee evaluated the following site specific and industry events for applicability to the Subsection IWE Program at Harris, Unit 1:

In Subsection 3.5.2 of the November 8, 2019, letter, the license concluded that NRC Information Notice (IN) 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner (Reference 17), is not applicable to Harris, Unit 1, since there are no leak-chase channel systems through the basemat or at floor-elevation at Harris, Unit 1. However, the licensee described the possible degradation path for the concrete-inaccessible part of the containment liner plate is around the edges of containment, which is protected via the moisture barrier that is in-place and is inspected every period.

In Subsection 3.5.3 of the November 8, 2019, letter, the licensee indicated that IN 2004-09, Corrosion of Steel Containment and Containment Liner (Reference 18), is applicable to Harris, Unit 1, and performed an extensive evaluation to identify corrosion on the containment steel liner plate locations near its interface with the concrete basemat during refueling outage (RFO) 7. The licensee collected inspection data during RFO7 and RFO8 in accordance with the ASME Section XI, Subsection IWE program. The licensee also found the thickness of the containment liner plate is above the minimum required thickness and identified no evidence of corrosion progression near the seal. The licensee also concluded the IWE Program, the SL1 Coatings Program, and the required examination prior to Type A testing are sufficient to identify initiating degradation in the containment liner plate.

In Subsection 3.5.4 of the November 8, 2019, letter, the licensee indicated that IN 2010-12, Containment Liner Corrosion (Reference 19), is applicable to Harris, Unit 1, and concluded that the existing visual inspection procedure within the IWE program has provisions to identify corrosion and bulging of the containment liner plate.

In Subsection 3.5.5 of the November 8, 2019, letter, the licensee noted that NRC RIS 2016-07, Containment Shell or Liner Moisture Barrier Inspection (Reference 20), identified several instances in which the containment shell or liner plate moisture barrier materials (e.g., caulking, flashing, and other sealants used for this application) were not properly inspected in accordance with the IWE Program. The licensee developed action items for Harris, Unit 1, and resealed the degraded areas during RFO H1R21 (spring 2018).

In Section 3.6, License Renewal Aging Management, of the November 8, 2019, letter, the licensee credited CISI programs for license renewal, as described in NUREG-1916. The licensee also referred to Chapter 18, Final Safety Analysis Report (FSAR) Supplement for License Renewal, of Harris, Unit 1, as required by 10 CFR 54.21(d), to describe the applicable aging management programs (AMPs). In the following subsections of the Harris, Unit 1, FSAR, the licensee described the applicable IWE and IWL programs for the primary containment.

In FSAR Section 18.1.1.26, ASME Section XI, Subsection IWE Program, the licensee indicated that the existing IWE program is implemented in accordance with the requirements of 10 CFR 50.55a and the applicable Edition and Addenda of the ASME Code,Section XI, as required by 10 CFR 50.55a(g)(4)(ii).

The licensee stated that prior to the period of extended operation, its ASME Section XI, Subsection IWE Program implementing procedure will be enhanced to include:

additional recordable conditions,

moisture barrier and applicable aging effects,

pressure retaining bolting and aging effects, and

augmented examinations.

In FSAR Section 18.1.1.27, ASME Section XI, Subsection IWL Program, the licensee stated that the existing IWL program is also used for the aging management of accessible and inaccessible reinforced concrete of the Harris, Unit 1, containment structure and is implemented in accordance with the requirements of 10 CFR 50.55a and the applicable Edition and Addenda of ASME Code,Section XI, as required by 10 CFR 50.55a(g)(4)(ii).

The NRC staff reviewed the information summarized above and finds that the licensee acceptably addressed the relevant regulatory requirements, guidance, and operating experience described above through inspection and aging management programs. Therefore, the licensees containment inservice inspection program provides reasonable assurance that the containment will maintain its capability to perform its safety related function.

3.3.2 Coating Inspections In Subsection 3.4.1, Nuclear Coating Program, of the November 8, 2019, letter, the licensee described the protective coatings program of Service Level 1 (SL1) coatings applied to the structures, systems and components (SSCs) located inside the primary containment that are performed during every RFO. The regulatory requirements for the coatings program are based on RG 1.54 (Reference 21) and defines the SL1 as applicable to coating failures (detached coatings) that adversely affect the operation of post-accident fluid systems and impair safe-shutdown (e.g., emergency core cooling system (ECCS)). The licensee described the tracking process to quantify unqualified coatings to ensure that the documented quantity does not exceed the postulated maximum allowable quantity in the Harris, Unit 1, Containment Building. In this section, the licensee also provided information from the SL1 Coatings Exempt Logs, and Unqualified Vendor Coatings Exempt Logs, which show the documented quantities are below the limits of 10,000 square feet and 4,350 square feet, respectively, from the condition monitoring walkdowns during RFOs in 2016 and 2018.

Based on its review, the NRC staff finds the licensee effectively manages the existing CISI programs of IWE and IWL in accordance with the requirements of ASME Section XI, and the conditions in 10 CFR 50.55a and the Protective Coatings Program and has responsively addressed industry events for applicability to Harris, Unit 1. Therefore, the NRC staff concludes that the periodic effective visual inspection programs as part of the Harris containment leakage rate testing provides reasonable assurance that public health and safety will be protected upon the permanent extension of Type A, B, and C leak rate test frequencies at Harris.

3.4 NEI 94-01, Revision 2-A, Conditions As required by 10 CFR 50.54(o), the Harris, Unit 1, containment is subject to the requirements set forth in 10 CFR Part 50, Appendix J. For type A tests, Harris has elected to use Option B, which requires that test intervals for Type A testing be determined by using a performance-based approach. Currently, the Harris 10 CFR Part 50, Appendix J, Containment Leakage Rate Testing Program, invokes RG 1.163 as the plan implementation document. The licensee proposes to revise the Harris, Unit 1, Containment Leakage Rate Testing Program, by replacing this implementation document with the guidance contained in NEI 94-01, Revision 3-A, and the conditions and limitations of NEI 94-01, Revision 2-A.

By letter dated June 25, 2008 (Reference 10), the NRC staff published a safety evaluation (SE) with limitations and conditions, for NEI 94-01, Revision 2. In this SE, the NRC staff concluded that NEI 94-01, Revision 2, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR Part 50, Appendix J, and is acceptable for referencing by licensees proposing to amend their TS pertaining to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.0 of the SE.

Section 4.1 of the June 25, 2008, SE establishes limitations and conditions pertaining to deterministic requirements, while Section 4.2 establishes limitations and conditions pertaining to the plants PRA analysis. More explicitly, the SE included provisions for extending the ILRT Type A interval to a maximum of 15 years subject to the six limitations and conditions provided in the SE. The NRC staff noted in the SE that NEI 94-01, Revision 2, incorporates the regulatory positions stated in RG 1.163. The accepted version of NEI 94-01, Revision 2, was subsequently issued as Revision 2-A. NEI issued Revision 2-A to NEI 94-01 on November 19, 2008. With Revision 2-A, the NEI 94-01 was revised to incorporate the June 25, 2008 NRC SE.

The NRC staffs review of LAR Section 3.7 NRC SER Limitations and Conditions which contains Table 3.7.1-1, entitled NEI 94-01, Revision 2-A Limitations and Conditions noted that Harris, Unit 1, intends to satisfy the limitations and conditions of NEI 94-01 Revision 2, Section 4.1. Accordingly, as previously noted, Harris, Unit 1, intends to adopt the testing methodology of ANSI/ANS 56.8-2002 in place of the methodology of ANSI/ANS 56.8-1994.

The leakage rate testing requirements of 10 CFR Part 50, Appendix J, Option B for Type A tests and Option A, for Type B, and Type C tests, and the CISI requirements mandated by 10 CFR 50.55a, together, have ensured the continued leak-tight and structural integrity of the Harris, Unit 1, containment during its service life to date.

Type B testing ensures that the leakage rate of individual containment penetration components is acceptable. Type C testing ensures that individual CIVs are essentially leak tight. In addition, aggregate Type B and Type C leakage rates support the leakage tightness of the Harris, Unit 1, containment by minimizing potential leakage paths.

The licensee proposes that Harris, Unit 1, invoke NEI 94-01, Revision 3-A, along with the conditions and limitations of NEI 94-01, Revision 2-A, as the reference documents for the Harris, Unit 1, Containment Leakage Rate Testing Program in TS 6.8.4.k. Therefore, the licensee is also applying to extend the frequencies of both the Type B and the Type C prescriptive based test intervals of 24 months, to the maximum permitted performance-based test intervals of 120 months and 75 months, respectively.

The NRC staff has found that the use of NEI 94-01, Revision 2-A, is acceptable for referencing by licensees proposing to amend their TS to permanently extend the ILRT surveillance interval to 15 years, provided the following applicable six conditions are satisfied; the NRC staffs evaluation of the licensees satisfaction of those six conditions follows:

3.4.1 Limitation and Condition 1 Limitation and Condition 1 specifies that for calculating the Type A leakage rate, licensees should use the definition in NEI 94-01, Revision 2, in lieu of that in ANSI/ANS-56.8-2002.

Licensees Response to Limitation and Condition 1 In Table 3.7.1-1 NEI 94-01, Revision 2-A Limitations and Conditions, contained in LAR Section 3.7, on Enclosure Page 90 of 106 of the Duke Energy Serial Letter RA-19-0067, the licensee stated:

HNP will utilize the definition in NEI 94-01, Revision 3-A, Section 5.0. This definition has remained unchanged from Revision 2-A to Revision 3-A of NEI 94-01.

Staff Assessment of Licensees Response to Limitation and Condition 1 Section 3.2.9 Type A test performance criterion of ANSI/ANS-56.8-2002 defines the performance leakage rate and states, in part:

The performance criterion for a Type A test is met if the performance leakage rate is less than La. The performance leakage rate is equal to the sum of the measured Type A test UCL and the total as-left MNPLR of all Type B or Type C pathways isolated during performance of the Type A test.

The NRC staffs SE, Section 3.1.1.1, Enclosure, page 6, for NEI 94-01, Revision 2, states in part:

Section 5.0 of NEI TR 94-01, Revision 2, uses a definition of performance leakage rate for Type A tests that is different from that of ANSI/ANS-56.8-2002.

The definition contained in NEI TR 94-01, Revision 2, is more inclusive because it considers excessive leakage in the performance determination. In defining the minimum pathway leakage rate, NEI TR 94-01, Revision 2, includes the leakage rate for all Type B and Type C pathways that were in service, isolated, or not lined up in their test position prior to the performance of the Type A test. Additionally, the NEI TR 94-01, Revision 2, definition of performance leakage rate requires consideration of the leakage pathways that were isolated during performance of the test because of excessive leakage in the performance determination. The NRC staff finds this modification of the definition of performance leakage rate used for Type A tests to be acceptable.

Section 5.0 Definitions of NEI 94-01, Revision 3-A states in part:

The performance leakage rate is calculated as the sum of the Type A upper confidence limit (UCL) and as-left minimum pathway leakage rate (MNPLR) leakage rate for all Type B and Type C pathways that were inservice, isolated, or

not lined up in their test position (i.e., drained and vented to containment atmosphere) prior to performing the Type A test. In addition, leakage pathways that were isolated during performance of the test because of excessive leakage must be factored into the performance determination. The performance criterion for Type A tests is a performance leak rate of less than 1.0La.

The NRC staff reviewed the definitions of performance leakage rate contained in NEI 94-01, Revision 2, and Revision 3-A. The NRC staff concluded that the definitions contained in both documents are identical.

Therefore, the NRC staff concludes that Harris, Unit 1, will use the definition found in Section 5.0 of NEI 94-01, Revision 2, for calculating the Type A leakage rate in the Harris, Unit 1, Containment Leakage Rate Testing Program.

Based on the above review, the NRC staff finds that the licensee has adequately addressed Condition 1.

3.4.2 Limitation and Condition 2 Limitation and Condition 2 stipulates that licensees submit a schedule of containment inspections to be performed prior to and between Type A tests.

Licensees Response to Limitation and Condition 2 In Table 3.7.1-1 NEI 94-01, Revision 2-A Limitations and Conditions, contained in LAR Section 3.7, on Enclosure Page 90 of 106 of the Duke Energy Serial Letter RA-19-0067, the licensee stated:

Reference Section 3.4.2 (Tables 3.4.2-1, 3.4.2-2, 3.4.2-3 and 3.4.2-4) of this LAR submittal.

Staff Assessment of Licensees Response to Limitation and Condition 2 Section 3.4.2, Containment Inservice Inspection (CISI) Program, of the November 8, 2019, letter, indicates that the general visual examinations of the accessible surfaces of the Harris, Unit 1, containment are performed in accordance with the CISI. Specifically, the licensee indicated that the Inservice Inspection Examination Plan (ISI Plan) provides requirements for examination, testing, and inspection of Class 1, 2, 3, MC, and Concrete Containment (CC) components and systems, and their supports.

Section 3.4.2 indicates that the first CISI interval for Harris, Unit 1, was effective from September 9, 1998, through September 8, 2008. As allowed by IWA-2430(c)(1), a one-year extension was taken for the first CISI interval until September 8, 2009. This extension did not affect the start of the second CISI interval. Unit 1 extended the first CISI interval in order to complete the first CISI interval examinations. The second CISI interval began on September 9, 2008, and was effective through September 8, 2018. Harris is currently in in its third 10-year inspection interval, from May 20, 2018, to September 8, 2027. Based on the third interval start date, the latest edition and addenda of ASME Section XI referenced in 10 CFR 50.55a(b)(2) is the 2007 Edition through the 2008 Addenda.

The licensees Table 3.4.2-1, Third CISI Interval/Period/Outage Matrix (For CISI Class MC Component Examinations - IWE Only), details the schedule for the inspection of the IWE Containment Liner components. From Table 3.4.2-1, the third 10-year interval contains three IWE periods. The IWE schedule requires a 100 percent general visual inspection of the accessible containment surface areas, pressure retaining bolting and moisture barriers during each period of each 10-year IWE interval. The third period of the second CISI interval would have run in its entirety subsequent to the completion of the last ILRT during May 2012.

Therefore, it can be concluded that adoption of a 15-year Type A test interval for Harris, Unit 1, will encompass at least four IWE periods and four 100% IWE containment liner inspections before the next proposed ILRT of 2027.

In Section 3.4.2 of the November 8, 2019, letter, the licensee states, The IWL examinations are based upon the requirements of ASME Section XI. Specific examinations are based on the requirements of the ASME B&PV Code,Section XI, Table IWL-2500-1 and the Third Ten-Year IWE/IWL Schedule.

The licensees Table 3.4.2-7, IWL-2500-1, Examination Category L-A, Concrete indicates that the scope of the IWL program includes 100% inspection of accessible concrete surface areas.

These Subsection IWL for Class CC examinations have a frequency of every 5 years, based on the initial inspection date of September 7, 2001 and every 5 years thereafter.

The licensees Table 3.4.2-3, Third CISI Interval/Period/Outage Matrix (For CISI Class CC Component Examinations - IWL Only) details the schedule for the inspection of the IWL Containment concrete components. Note 1 to Table 3.4.2-3 reads, as follows:

The CISI Interval for Class CC components is the same as the CISI Interval for Class MC components. The actual inspection schedule for Class CC components is based on a rolling 5-year frequency (+/- 1 year) from the date of completion of the original examinations (09/07/2001) performed during the initial September 9, 1996 - September 8, 2001, rulemaking implementation period. The rolling 5-year inspection schedule for CC is in accordance with the ISI schedule specified in IWL-2400 as modified by the initial regulatory rulemaking.

From Table 3.4.2-3, the third 10-year interval contains two IWL inspection periods. The IWL schedule requires a 100% inspection of accessible concrete surface areas during each period of each 10-year IWL interval. The second period of the second CISI interval would have run in its entirety subsequent to the completion of the last ILRT during May 2012. Therefore, it can be concluded that adoption of a 15-year Type A test interval for Harris, Unit 1, will encompass at least three IWL periods and three100% IWL containment concrete component inspections before the next proposed ILRT of 2027.

The NRC staff finds that the licensees statements in section 3.4.2 of the LAR satisfy the guidance in NEI 94-01, Revision 2 and Revision 3-A, with respect to Limitation and Condition 2.

The NRC staffs SE Section 3.1.1.3, Enclosure, page 7, for NEI 94-01, Revision, 2, states, in part:

NEI TR 94-01, Revision 2, Section 9.2.3.2, states that: To provide continuing supplemental means of identifying potential containment degradation, a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior

to each Type A test and during at least three other outages before the next Type A test if the interval for the Type A test has been extended to 15 years. NEI TR 94-01, Revision 2, recommends that these inspections be performed in conjunction or coordinated with the examinations required by ASME Code,Section XI, Subsections IWE and IWL. The NRC staff finds that these visual examination provisions, which are consistent with the provisions of regulatory position C.3 of RG 1.163, are acceptable considering the longer 15-year interval. Regulatory Position C.3 of RG 1.163 recommends that such examination be performed at least two more times in the period of 10 years. The NRC staff agrees that as the Type A test interval is changed to 15 years, the schedule of visual inspections should also be revised. Section 9.2.3.2 in NEI TR 94-01, Revision 2, addresses the supplemental inspection requirements that are acceptable to the NRC staff.

Page 12 of NEI 94-01, Revision 2, and Revision 3-A, Section 9.2.3.2, Supplemental Inspection Requirements both read:

To provide continuing supplemental means of identifying potential containment degradation, a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior to each Type A test and during at least three other outages before the next Type A test if the interval for the Type A test has been extended to 15 years. It is recommended that these inspections be performed in conjunction or coordinated with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE/IWL required examinations.

Page 10 of NEI 94-01, Revision 3-A, Section 9.2.1, Pretest Inspection and Test Methodology states, in part:

Prior to initiating a Type A test, a visual examination shall be conducted of accessible interior and exterior surfaces of the containment system for structural problems that may affect either the containment structure leakage integrity or the performance of the Type A test. This inspection should be a general visual inspection of accessible interior and exterior surfaces of the primary containment and components. It is recommended that these inspections be performed in conjunction or coordinated with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE/IWL required examinations.

Accordingly, consistent with NEI 94-01, Revision 2, Section 9.2.3.2, and NEI 94-01, Revision 3-A, Section 9.2.1, the requisite number of 100 percent examinations of CISI Class MC components and Class CC components are scheduled for completion between Containment Appendix J Option B Type A test of May 2012 and the next proposed Type A test of Spring 2027.

Based on the above, the NRC staff concludes that the requirements of SE, Section 3.1.1.3, for NEI 94-01, Revision 2, are satisfied.

3.4.3 Limitation and Condition 3 Limitation and Condition 3 stipulates that licensees address the areas of the containment structure potentially subjected to degradation.

Licensees Response to Limitation and Condition 3 In Table 3.7.1-1, NEI 94-01, Revision 2-A Limitations and Conditions, contained in LAR Section 3.7, on Enclosure Page 90 of 106 of the Duke Energy Serial Letter RA-19-0067, the licensee stated:

Reference Section 3.4.2 (Tables 3.4.2-5, and 3.4.2-6) of this LAR submittal.

Further, in Section 3.7.1, Limitations and Conditions Applicable to NEI 94-01, Revision 2-A, of the November 8, 2019, letter, the licensee referred to Table 3.7.1-1, NEI 94-01 Revision 2-A Limitations and Conditions, which referred to Tables 3.4.2-5, IWE-2500-1, Examination Category E-A, Containment Surfaces, and 3.4.2-6, IWE-2500-1, Examination Category E-C1, Containment Surfaces Requiring Augmented Examination. Subsection 3.4.2, Containment Inservice Inspection (CISI) Program, of the licensees November 8, 2019, letter, discusses areas of the Harris, Unit 1, containment building that may be subjected to degradation for Concrete Containment (Class CC) and Metal Containment (Class MC) examinations.

The licensee provided examination results for IWE surfaces (VT-3) in Tables 3.4.4-1, 3.4.4-3 and for IWL surfaces (VT-3C) in Table 3.4.4-2. Welded electrical junction boxes attached to the containment wall are considered inaccessible for general visual examination in accordance with Item E1.11 of Examination Category E-A, as stated in Note 1 of Table 3.4.2-5.

In Table 3.4.2-6, the licensee added one augmented examination for Item E4.11 of Examination Category E-C for the third containment ISI interval. This item is normally considered to be in an inaccessible area due to extensive lead shielding near fuel transfer tube Penetration S-65.

Note 3 of Table 3.4.2-6 noted that if no conditions were to be observed during the augmented examination, the licensee may discontinue performing follow-up augmented examination.

Even though the licensee did not refer to Note 2 of Table 3.4.2-5 in Table 3.7.1-1, the NRC staff recognized that the licensee scheduled additional visual examinations of moisture barriers in the third interval due to RIS 2016-07, since the only possible degradation path for the inaccessible part of the embedded containment liner plate is around the edges of containment protected by the moisture barrier. As mentioned previously, the licensee indicated that IN 2014-07 is not applicable to Harris, Unit 1, since there is no leak-chase channel system in place in the concrete basemat at Harris, Unit 1.

The NRC staff also performed a review of the discussions of recent containment examinations associated with the moisture barriers under the Comments column in Tables 3.4.4-1, 3.4.4-2 and 3.4.4-3 in Subsection 3.4.4, where the licensee did not report degradation and moisture intrusion in containment liner plate.

Staff Assessment of Licensees Response to Limitation and Condition 3 Based on the information above, the NRC staff finds that the licensee provided an acceptable level of information regarding the implementation of the Examination Categories of E-A and E-C of ASME Section XI, Subsection IWE, that exhibit and/or would indicate the presence of potential degraded conditions in the accessible and inaccessible areas of the containment concrete and liner plate.

Based on the above, the NRC staff concludes the licensee has established a CISI program that satisfies the intent of the issues presented in NRC SE Section 3.1.3. Therefore, the NRC staff finds that the licensee has adequately addressed Condition 3.

3.4.4 Limitation and Condition 4 Limitation and Condition 4 specifies that licensees address any tests and inspections performed following major modifications to the containment structure, as applicable.

Licensees Response to Limitation and Condition 4 In Table 3.7.1-1 NEI 94-01, Revision 2-A Limitations and Conditions, contained in Section 3.7 of the November 8, 2019, letter, the licensee stated:

HNP removed and re-welded the equipment hatch in 2001 to support SGR. HNP is removing and re-welding the equipment hatch in support of reactor pressure vessel head replacement during the fall 2019 refueling outage. Reference Section 3.1.3 of this submittal.

In Section 3.7.1, of the November 8, 2019, letter, the licensee referred to Table 3.7.1-1 which referred to Section 3.1.3, Relief Request (RR) 13R-18, that stated the licensee has removed and rewelded the equipment hatch body temporarily to the containment penetration sleeve in support of the reactor vessel head replacement activities performed during RFO 2019.

The licensee submitted RR I3R-18 (Reference 22), to the NRC on June 4, 2018, which describes alternative repair and replacement testing requirements for the containment building equipment hatch sleeve weld portion of Harris, Unit 1, reactor vessel head replacement.

Pursuant to 10 CFR 50.55a(z)(1), the licensee proposed to use a bubble test in accordance with the requirements of the ASME Code,Section V, Article 10, Appendix I, Bubble Test-Direct Pressure Technique, instead of the 10 CFR 50.55a requirement to perform a containment leak rate test. The licensee also stated that the bubble test meets the equipment supplier requirement since, if the leakage occurs, the reweld is repaired and retested to ensure zero leakage at the reweld-area. After the equipment hatch body was reinstalled and rewelded to the existing penetration sleeve, the licensee also performed 100-percent volumetric post-weld examinations with 100-percent radiographic examinations and 100-percent magnetic-particle testing to ensure the post-weld integrity per the requirements of the 2007 Edition with 2008 Addenda of ASME Section Ill, Subsection NE-5000, as conditioned by 10 CFR 20.55a(b)(2)(ix)(j). The NRC authorized the proposed alternative for the one-time reactor vessel head replacement activity on February 26, 2019 (Reference 23).

Staff Assessment of Licensees Response to Limitation and Condition 4 Based on the above, the NRC staff finds that that the licensee provided sufficient information related to rewelding of the equipment hatch to the existing penetration sleeve following the reactor pressure head replacement activities performed during the refueling outage in fall 2019.

Therefore, the staff concludes that the licensee has adequately addressed Condition 4.

3.4.5 Limitation and Condition 5 Limitation and Condition 5 specifies that the normal Type A test interval should be less than 15 years. If a licensee utilizes the provision of Section 9.1 of NEI 94-01, Revision 2, related to

extending the ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition.

Licensees Response to Limitation and Condition 5 In Table 3.7.1-1, NEI 94-01, Revision 2-A Limitations and Conditions, contained in Section 3.7, of the November 8, 2019, letter, the licensee stated:

HNP will follow the requirements of NEI 94-01, Revision 3-A, Section 9.1. This requirement has remained unchanged from Revision 2-A to Revision 3-A of NEI 94-01.

In accordance with the requirements of NEI 94-01, Revision 2-A, SER Section 3.1.1.2, HNP will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.

Staff Assessment of Licensees Response to Limitation and Condition 5 The NRC staffs SE, Section 3.1.1.2, Enclosure, page 6, for NEI 94-01, Revision 2, states:

As noted above, Section 9.2.3, NEI TR 94-01, Revision 2, states, Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per 15 years based on acceptable performance history. However, Section 9.1 states that the required surveillance intervals for recommended Type A testing given in this section may be extended by up to 9 months to accommodate unforeseen emergent conditions but should not be used for routine scheduling and planning purposes. The NRC staff believes that extensions of the performance-based Type A test interval beyond the required 15 years should be infrequent and used only for compelling reasons. Therefore, if a licensee wants to use the provisions of Section 9.1 in TR NEI 94-01, Revision 2, the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists.

The licensee stated that HNP will follow the requirements of NEI 94-01, Revision 3-A, Section 9.1. The NRC staff notes that NEI 94-01, Revision 3-A, Section 9.1, Introduction, contains the relevant passage from the NRC staff SE for NEI 94-01, Revision 2, and states, in part:

Required surveillance intervals for recommended Type A testing given in this section may be extended by up to 9 months to accommodate unforeseen emergent conditions but should not be used for routine scheduling and planning purposes.

The NRC staff concludes that the licensee has demonstrated its understanding that any extension of the Type A test interval beyond the upper-bound performance-based limit of 15 years should be infrequent and that any requested permission (i.e., for such an extension) will demonstrate to the NRC staff that an unforeseen emergent condition exists.

Based on the above review, the NRC staff finds that the licensee has adequately addressed the requirements of SE Section 3.1.1.2 for NEI 94-01, Revision 2 and Condition 5.

3.4.6 Limitation and Condition 6 Limitation and Condition 6 specifies that for plants licensed under 10 CFR Part 52, applications requesting a permanent extension of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past containment ILRT data. The NRC staff concludes that Condition 6 is not applicable to Harris, Unit 1, because Harris was licensed under 10 CFR Part 50.

3.4.7 Conclusion Related to the Six Limitations and Conditions Listed in NEI 94-01, Revision 2-A, Section 4.1, of the NRC Staffs Safety Evaluation for Revision 2-A The NRC staff evaluated each of the six limitations and conditions listed above and determined that the licensee adequately satisfied all six limitations and conditions identified in NEI 94-01, Revision 2-A, Section 4.1, of the NRC staffs safety evaluation for Revision 2-A. Therefore, the NRC staff finds it acceptable for Harris, Unit 1, to adopt the conditions and limitations of NEI 94-01, Revision 2-A, as part of the implementation documents listed in TS 6.8.4.k.

3.5 NEI 94-01, Revision 3-A, Conditions In the NRC staffs SE dated June 8, 2012 (Reference 11), the NRC staff concluded that the guidance in NEI 94-01, Revision 3 (Reference 24), is acceptable for reference by licensees proposing to amend their TSs with regard to containment leakage rate testing, subject to two conditions. The NRC staffs evaluation of the LARs satisfaction of these conditions follows.

3.5.1 NEI 94-01, Revision 3-A - Condition 1 The June 8, 2012, NEI 94-01, Revision 3, Safety Evaluation, Section 4.0, Condition 1, stipulates that:

NEI TR 94-01, Revision 3, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The staff is allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84 months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g., BWR MSIVs

[Boiling Water Reactor Main Steam Isolation Valves]), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.

Condition 1 identifies three issues that are required to be addressed:

(1) The allowance of an extended interval for Type C LLRTs of 75 months requires that a licensees post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit; (2) A corrective action plan is to be developed to restore the margin to an acceptable level; and (3) Use of the allowed 9-month extension for eligible Type C valves is only allowed for non-routine emergent conditions, but not for valves categorically restricted and other excepted valves.

Licensees Response to Condition 1 In LAR Section 3.7.2 Limitations and Conditions Applicable to NEI 94-01, Revision 3-A, (Enclosure, Page 92 of 106), the licensee stated:

Response to Condition 1, ISSUE 1 The post-outage report shall include the margin between the Type B and Type C MNPLR summation value, as adjusted to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.60 La.

Response to Condition 1, ISSUE 2 When the potential leakage understatement adjusted Type B and C MNPLR total is greater than the HNP administrative leakage summation limit of 0.5 La, but less than the regulatory limit of 0.6 La, then an analysis and determination of a corrective action plan shall be prepared to restore the leakage summation margin to less than the HNP leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues so as to maintain an acceptable level of margin.

Response to Condition 1, ISSUE 3 HNP will apply the 9-month allowable interval extension period only to eligible Type C components and only for non-routine emergent conditions. Such occurrences will be documented in the record of tests.

NRC Staff Assessment of Licensees Response to Condition 1 The NRC staff has reviewed the requirements of NEI 94-01, Revision 3, against the licensees responses for Issues (1), (2) and (3) of NEI 94-01, Revision 3-A, Condition 1. Based on this review, the NRC staff concludes that the licensee has acknowledged all the requirements of Condition 1 and has satisfactorily established its intent for Harris, Unit 1, to comply with these requirements.

3.5.2 NEI 94-01, Revision 3-A - Condition 2 The NRC staffs Safety Evaluation dated June 8, 2012, Section 4.0, Condition 2, stipulates that:

The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time. The containment leakage condition monitoring regime involves a portion of the penetrations being tested each refueling outage, nearly all LLRTs being performed during plant outages. For the purposes of assessing and monitoring or trending overall containment leakage potential, the as-found minimum pathway leakage rates for the just tested penetrations are summed with the as-left minimum pathway leakage rates for penetrations tested during the previous 1 or 2 or even 3 refueling outages. Type C tests involve valves which, in the aggregate, will show increasing leakage potential due to normal wear and tear, some predictable and some not so predictable. Routine and appropriate maintenance may extend this increasing leakage potential. Allowing for longer intervals between LLRTs means that more leakage rate test results from farther back in time are summed with fewer just tested penetrations and that total used to assess the current containment leakage potential. This leads to the possibility that the LLRT totals calculated understate the actual leakage potential of the penetrations. Given the required margin included with the performance criterion and the considerable extra margin most plants consistently show with their testing, any understatement of the LLRT total using a 5-year test frequency is thought to be conservatively accounted for. Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI 94-01, Revision 3, Section 12.1.

When routinely scheduling any LLRT valve interval beyond 60 months and up to 75 months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B & C total, and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Condition 2 identifies two issues that are required to be addressed:

(1) Extending the Type C LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative, provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI 94-01, Revision 3, Section 12.1; and (2) When routinely scheduling any LLRT valve interval beyond 60 months and up to 75 months, the Primary Containment Leakage Rate Testing Program trending or monitoring must include an estimate of the amount of understatement in the Type B and Type C total, and must be included in a

licensees post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Licensees Response to Condition 2 In Section 3.7.2 Limitations and Conditions Applicable to NEI 94-01, Revision 3-A, of the November 8, 2019, letter, the licensee stated:

Response to Condition 2, ISSUE 1 The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94-01, Revision 3-A, represents an increase of 25% in the LLRT periodicity. As such, HNP will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the actual as-left leak rate, which will increase the as-left leakage total for each Type C component currently on greater than a 60-month test interval up to the 75-month extended test interval. This will result in a combined conservative Type C total for all 75-month LLRTs being carried forward and will be included whenever the total leakage summation is required to be updated (i.e., either while on-line or following an outage).

When the potential leakage understatement adjusted leak rate total for those Type C components being tested on greater than a 60-month test interval up to the 75-month extended test interval is summed with the non-adjusted total of those Type C components being tested at less than or equal to a 60-month test interval, and the total of the Type B tested components, results in the MNPLR being greater than the HNP administrative leakage summation limit of 0.50 La but less than the regulatory limit of 0.6 La, then an analysis and corrective action plan shall be prepared to restore the leakage summation value to less than the HNP leakage limit. The corrective action plan should focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.

Response to Condition 2, ISSUE 2 If the potential leakage understatement adjusted leak rate MNPLR is less than the HNP administrative leakage summation limit of 0.50 La, then the acceptability of the greater than a 60-month test interval up to the 75-month LLRT extension for all affected Type C components has been adequately demonstrated and the calculated local leak rate total represents the actual leakage potential of the penetrations.

In addition to Condition 1, ISSUES 1 and 2, which deal with the MNPLR Type B and C summation margin, NEI 94-01, Revision 3-A, also has a margin-related requirement as contained in Section 12.1, Report Requirements.

A post-outage report shall be prepared presenting results of the previous cycles Type B and Type C tests, and Type A, Type B and Type C tests, if performed

during that outage. The technical contents of the report are generally described in ANSI/ANS-56.8-2002 and shall be available on-site for NRC review. The report shall show that the applicable performance criteria are met and serve as a record that continuing performance is acceptable. The report shall also include the combined Type B and Type C leakage summation, and the margin between the Type B and Type C leakage rate summation and its regulatory limit. Adverse trends in the Type B and Type C leakage rate summation shall be identified in the report and a corrective action plan developed to restore the margin to an acceptable level.

At HNP, in the event an adverse trend in the aforementioned potential leakage understatement adjusted Type B and C summation is identified, then an analysis and determination of a corrective action plan shall be prepared to restore the trend and associated margin to an acceptable level. The corrective action plan shall focus on those components which have contributed the most to the adverse trend in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.

At HNP, an adverse trend is defined as three consecutive increases in the final pre-mode change Type B and C MNPLR leakage summation values, as adjusted to include the estimate of applicable Type C leakage understatement, as expressed in terms of La.

NRC Staff Assessment of Licensees Response to Condition 2 The NRC staff has reviewed the requirements of NEI 94-01, Revision 3, against the Duke Energy responses to Issues (1) and (2) of NEI 94-01, Revision 3-A, Condition 2. Based on this review, the NRC staff concludes that Duke Energy has acknowledged all the requirements of Condition 2 and has satisfactorily established its intent for Harris, Unit 1, to comply with these requirements.

3.5.3 NEI 94-01, Revision 3-A, Safety Evaluation Section 4.0 Based on the above evaluation of each condition, the NRC staff concludes that the licensee has adequately addressed both conditions in Section 4.0 of the NRC SE for NEI 94-01, Revision 3-A. Therefore, the NRC staff finds it acceptable for Harris, Unit 1, to adopt NEI 94-01, Revision 3-A, as the implementation document in TS 6.8.4.k Containment Leakage Rate Testing Program.

3.6 Deletion of Exception in TS 6.8.4.k The licensee proposed an administrative change to delete information in TS 6.8.4.k, exception 2, related to performance of the next Type A test, to be performed no later than May 23, 2012. This referenced Type A test requirement had been previously approved by the NRC in Amendment No. 122 dated March 30, 2006 (Reference 14), and is no longer applicable, since the testing has already occurred. Therefore, the NRC staff concludes that the deletion of TS 6.8.4k, exception 2, is acceptable.

3.7 Probabilistic Risk Assessment of the Proposed Extension of the ILRT Test Intervals The licensee provided a plant-specific risk assessment for permanently extending the currently allowed containment Type A ILRT interval from 10 years to 15 years in Attachment 3 to the LAR.

The licensee states that its plant-specific risk assessment follows the guidance in NEI 94-01 and EPRI TR-1018243 (also identified as EPRI TR-1009325, Revision 2-A). In the NRC staffs SE, dated June 25, 2008, the NRC staff found the methodology in NEI 94-01, Revision 2-A, and EPRI TR-1009325, Revision 2-A, acceptable for referencing by licensees proposing to amend their TS to permanently extend the ILRT interval to 15 years, provided certain conditions are satisfied. The licensee addressed each of the four conditions for the use of EPRI TR-1009325, Revision 2-A, which are listed in Section 4.2 of the NRC staffs SE dated June 25, 2008.

These four conditions are:

Condition 1 - The licensee submits documentation indicating that the technical adequacy of their PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension application.

Condition 2 - The licensee submits documentation indicating that the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years is small and consistent with the clarification provided in Section 3.2.4.5 of this SE. Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive. In addition, a small increase in CCFP should be defined as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage point.

Condition 3 - The methodology in EPRI Report No. 1009325, Revision 2, is acceptable except for the calculation of the increase in expected population dose (per year of reactor operation). In order to make the methodology acceptable, the average leak rate for the pre-existing containment large leak rate accident case (accident case 3b) used by the licensees shall be 100 La instead of 35 La.

Condition 4 - A LAR is required in instances where containment over-pressure is relied upon for ECCS performance.

A summary of how each condition is met in the LAR is provided in Sections 3.7.1 through 3.7.4 below. Additionally, the licensee applied the methodology from Calvert Cliffs Nuclear Power Plant (Reference 37), to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval.

3.7.1 Condition 1 (PRA Quality)

The first condition in Section 4.2 of the NRC staffs SE for EPRI TR-1009325, Revision 2-A stipulates that the licensee submit documentation indicating that the technical adequacy of its PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension application.

RG 1.200 describes one acceptable approach for determining whether the technical adequacy

of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors.

Consistent with RIS 2007-06, the NRC staff uses Revision 2 of RG 1.200 to assess the technical adequacy of the PRA used to support risk-informed applications. In Section 3.2.4.1 of the NRC staffs SE for NEI 94-01, Revision 2 and EPRI TR-1009325, Revision 2, the NRC staff stated that Capability Category (CC) I of the ASME PRA standard shall be applied as the standard for assessing PRA quality for ILRT extension applications, since approximate values of core damage frequency (CDF) and large early release frequency (LERF) and their distribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies.

The NRC staffs review of the internal events PRA was based on the results of the peer review of the licensees internal events PRA, the associated facts and observations (F&Os) closure review described in Section 3.3.2 of the licensees November 8, 2019, letter, and previously docketed information on PRA quality submitted to the NRC for relocation of surveillance frequencies to the licensee control program, by means of License Amendment No. 154, dated November 29, 2016 (Reference 25), and the licensees request to adopt National Fire Protection Association (NFPA) Standard 805 (Reference 26).

The internal events model was subject to a full scope peer review in 2002 in accordance with the guidance in NEI 00-02, Industry Probabilistic Risk Assessment (PRA) Peer Review Process Guidelines, (Reference 27), and Appendix B of RG 1.200, Revision 2. The licensee stated that in March 2017, an F&O closure review was performed by an independent team on all internal events and internal flooding finding-level F&Os. This F&O closure review was performed as detailed in Appendix X to the guidance in NEI 05-04, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, November 2008 (Reference 28),

NEI 07-12 Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Revision 1, June 2010 (Reference 29), and NEI 12-13 External Hazards PRA Peer Review Process Guidelines, NEI 12-13, dated August 2012 (Reference 30), concerning the process Close Out of Facts and Observations.

The NRC staff accepted, with conditions, a final version of Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13, in an NRC letter dated May 3, 2017 (Reference 31). The March 2017 closure review closed out all open internal events F&Os. Although the Harris, Unit 1, F&O closure review was performed prior to the NRC staff's acceptance of the F&O closure process, Harris, Unit 1, was one of the pilot plants for the F&Os closure process, and the NRC staff observed the Harris, Unit 1, F&Os closure process. Based on the NRC staff's observations, the NRC staff finds the March 2017 F&Os closure at Harris, Unit 1, was performed in accordance with the NRC-accepted version of Appendix X.

The internal flooding model was subject to a self-assessment and a full scope peer review in August 2014. In March 2017, an F&O closure review was performed by an independent team on all internal events finding-level F&Os. The licensee submitted a list of all F&Os from the peer reviews that remained open after the F&O closure review in Attachment 3 of the licensees November 8, 2019, letter. In Attachment 3, the licensee provided a disposition for each of the open F&Os for this application. The NRC staff reviewed the licensee's resolution of all the peer review findings and assessed the potential impact of the open findings on the ILRT extension and found them to be acceptable.

The NRC staff reviewed the results of the peer review of the fire PRA and associated F&O closure review described in Attachment 3 of the LAR. The licensee's fire PRA was subject to both an NRC staff review and a full-scope industry peer review in 2008 during the review associated with a license amendment seeking to adopt NFPA Standard 805. The 2008 review was conducted using the ANSI/ANS-58.23-2007 standard, whereas RG 1.200, Revision 2, references the 2009 ASME/ANS RA-Sa-2009 standard. The ASME/ANS-RA-Sa-2009 standard states that it was assembled from the ANSI/ANS-58.23-2007 fire PRA standard. The licensee stated that it assessed the differences between the ANSI/ANS-58.23-2007 standard and the current version of the fire PRA standard in ASME/ANS RA-Sa-2009 and confirmed there were no technical differences between the two versions of the standard. The NRC staff reviewed the licensee's resolution of the peer review findings and assessed the potential impact of the findings on the ILRT extension and found them to be acceptable.

With respect to external events, RG 1.174 stipulates that established acceptance guidelines are intended for comparison with a full-scope assessment of the change in the applicable risk metrics and recognizes that many PRAs are not full scope and PRA information of less than full scope may be acceptable. The methodology described in EPRI TR-1009325, Revision 2-A, which the NRC staff found to satisfy the key principles of risk-informed decision-making of RG 1.174, states that if the external event analysis is not of sufficient quality or detail to allow direct application of the methodology, the quality or detail will be increased or a suitable estimate of the risk impact from the external events should be performed. This assessment can be taken from existing, previously submitted and approved analyses or another alternate method of assessing an order-of-magnitude estimate for contribution of the external event to the impact of the changed interval. Based on the RG 1.174 stipulation regarding external events, the licensee used the seismic risk estimates from NRC Generic Issue 199 (GI-199),

Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants: Safety/Risk Assessment (Reference 32). The licensee used a simple average seismic CDF of 1.4E-06 per year, reported in Table D-1 of Appendix D: Seismic Core-Damage Frequencies using the 2008 USGS Seismic Hazard Curves (Reference 33).

Further, the licensee referenced an NEI letter dated March 12, 2014 (Reference 34), stating that the conclusions reached in 2010 by GI-199 remain valid for estimating seismic CDF at plants in the Central and Eastern United States which includes Harris, Unit 1. The NEI letter provided a fleetwide seismic risk study which was conducted by EPRI for plants located in the Central and Eastern United States. In a letter dated May 9, 2014 (Reference 35), the NRC staff stated that the results of the NRC staffs independent review confirm that EPRI-fleetwide seismic risk estimates are consistent with the approach and results used in the Gl-199 safety/risk assessment and that the conclusions reached in Gl-199 remain valid. Therefore, the NRC staff finds the use of the seismic CDF from GI-199 acceptable.

To account for the seismic LERF, the licensee assumed that the LERF to CDF ratio will be similar for seismic risk as for internal events risk. The NRC staff finds that the licensees approach is sufficient to support an order-of-magnitude external events risk assessment for the Harris ILRT extension.

Based on the NRC staffs review of the above information, the NRC staff finds the licensee has addressed the relevant findings and gaps from the peer reviews and that they have no impact on the results of this application. Therefore, the NRC staff concludes that the PRA model used by the licensee is of sufficient quality to support the evaluation of changes to ILRT frequencies.

Accordingly, the first condition for the use of EPRI TR-1009325, Revision 2-A, is met.

3.7.2 Condition 2 (Estimated Risk Increase)

The second condition for the use of EPRI TR-1009325, Revision 2-A, stipulates that the licensee submit documentation indicating that the estimated risk increase associated with permanently extending the ILRT interval to 15 years is small, consistent with the guidance in RG 1.174 and the clarification provided in the NRC staffs SE for NEI 94-01, Revision 2-A, and EPRI TR-1009325, Revision 2-A. Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive. In addition, a small increase in conditional containment failure probability (CCFP) should be defined as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests.

This would require that the increase in CCFP be less than or equal to 1.5 percentage points.

Lastly, for plants that rely on containment over-pressure for net positive suction head (NPSH) for ECCS injection, both CDF and LERF will be considered in the ILRT evaluation and compared with the risk acceptance guidelines in RG 1.174. The licensee stated in Section 3.1.2 of the November 8, 2019, letter, that no reliance is placed on the containment pressure for meeting the NPSH requirements. Since Harris, Unit 1, does not rely on containment accident pressure for ECCS performance, extending the ILRT interval does not impact CDF. Thus, the associated risk metrics for this application include LERF, population dose, and CCFP.

Section 3.3.3 of the licensees November 8, 2019, letter summarizes the results of the plant-specific risk assessment. The licensee drew the following conclusions from its analysis, associated with extending the Type A ILRT frequency:

1. RG 1.174 defines small changes in LERF to be between 1E-7/year and 1E-6/year with a total LERF less than 1E-5. RG 1.174 considers very small changes in LERF to be less than 1.0E-07/year. The increase in LERF resulting from a change in the Type A ILRT test interval from 3-in-10 years to 1-in-15 years is estimated as 6.52E-08/year using the EPRI guidance. This value increases negligibly if it includes the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval. As such, the estimated change in LERF is determined to be very small using the acceptance guidelines of RG 1.174. When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 3-in-10 years to 1-in-15 years is estimated as 4.57E-07/year using the EPRI guidance, and total estimated LERF is 5.46E-06/year. As such, the estimated change in LERF is determined to be small using the acceptance guidelines of RG 1.174. The risk change resulting from a change in the Type A ILRT test interval from 3-in-10 years to 1-in-15 years bounds the 1-in-10 years to 1-in-15 years risk change.

2. The result from changing the Type A test frequency to 1-per-15 years measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing is 0.038 person-rem/year. NEI 94-01 Revision 2-A, states that a small total population dose is defined as an increase of 1.0 person-rem/year, or 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The reported increase in total population dose is below the acceptance criteria provided in EPRI TR-1009325, Revision 2-A, and defined in Section 3.2.4.6 of the NRC staffs SE for NEI 94-01, Revision 2-A. Thus, the increase in the total integrated plant risk for the proposed change is considered small and supportive of the proposed change.

3. The increase in the CCFP due to the change in test frequency from 3-in-10 years to 1-in- 15 years is 0.753%. NEI 94-01, Revision 2-A, states that increases in CCFP of 1.5% is small. This value is below the acceptance guidelines in Section 3.2.4.6 of the NRC SER for NEI 94-01, Revision 2-A, and supportive of the proposed change.

Based on the licensees risk assessment results, the NRC staff concludes that the increase in LERF is small and is consistent with the acceptance guidelines of RG 1.174, and the increase in the total population dose and the magnitude of the change in the CCFP for the proposed change are small. Additionally, the NRC staff confirmed that the licensee accounted for the risk-implications of corrosion-induced leakage of steel liners going undetected during the extended test interval. The licensee applied the NRC staff-approved methodology described in EPRI TR-1009325, Revision 2-A, and showed that the estimated risk increase due to this effect is negligible.

The defense-in-depth philosophy is maintained as the independence of barriers will not be degraded because of the requested change, and the use of the quantitative risk metrics collectively ensures that the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved. Accordingly, the second condition for the use of EPRI TR-1009325, Revision 2-A, is met.

3.7.3 Condition 3 (Leak Rate for the Large Pre-Existing Containment Leak Rate Case)

The third condition for the use of EPRI TR-1009325, Revision 2-A, stipulates that to make the methodology in EPRI TR-1009325, Revision 2 acceptable, the average leak rate for the pre-existing containment large leak rate accident case (i.e., accident case 3b) used by licensees shall be 100 La instead of 35 La. As noted by the licensee in Section 3.3.1 of the November 8, 2019, letter, the methodology in EPRI TR-1009325, Revision 2-A, incorporated the use of 100 La as the average leak rate for the pre-existing containment large leakage rate accident case (accident case 3b), and this value has been used in the Harris, Unit 1, plant-specific risk assessment. Accordingly, the third condition for the use of EPRI TR-1009325, Revision 2-A, is met.

3.7.4 Condition 4 (Containment Overpressure Reliance for ECCS Performance)

The fourth condition for the use of EPRI TR-1009325, Revision 2-A, stipulates that in instances where containment over-pressure is relied upon for ECCS performance, a LAR is required to be submitted. In Section 3.1.2 of the November 8, 2019, letter, the licensee stated that no reliance is placed on the containment pressure for meeting the NPSH requirements. Accordingly, the fourth condition is not applicable at Harris.

3.7.5 Probabilistic Risk Assessment Summary Based on the above, the NRC staff concludes that the licensee has met the four limitations and conditions for the use of EPRI TR-1009325, Revision 2-A. Accordingly, the NRC staff concludes that the PRA used by the licensee is sufficient for this application and that the risk impact for extending the ILRT intervals as proposed by the licensee is within the acceptance guidelines of RG 1.174.

3.8 Technical Conclusion As discussed above, the licensee proposed to extend the Harris, Unit 1, current performance-based Type A test interval to no longer than 15 years by adopting NEI 94-01, Revision 3-A and the conditions and limitations of NEI 94-01, Revision 2-A, as the implementation documents in TS 6.8.4.k. This change would allow the licensee to conduct the next Type A test no later than May 2027, in lieu of the current requirement of no later than May 2022.

Based on the preceding regulatory and technical evaluations, the NRC staff finds that the licensee has adequately implemented its existing primary containment leakage rate testing program consisting of performance-based ILRTs and prescriptive based LLRTs. The results of the recent ILRTs and of the LLRTs combined totals demonstrate acceptable performance and support a conclusion that the structural and leak-tight integrity of the primary containment is adequately managed and will continue to be periodically monitored and managed effectively with the proposed changes.

The NRC staff has determined that the licensees containment inspection programs support an extension of the ILRT frequency as requested in the licensees submittal of November 8, 2019.

The NRC staff finds that there is reasonable assurance that the structural integrity of the Harris, Unit 1, primary containment will continue to be monitored and maintained with the current performance-based Type A test interval permanently extended to require one test in 15 years without undue risk to public health and safety because: (1) the licensee has adequately implemented its CISI program to periodically examine, monitor, and manage the condition of its containment structure; and (2) the results of past containment concrete and liner visual inspections demonstrate acceptable performance of the containment and demonstrate that the structural integrity of the containment structure is adequate.

The NRC staff finds that the PRA used by the licensee is sufficient for this application and that the risk impact for extending the ILRT intervals is within the acceptance guidelines of RG 1.174.

Consistent with the guidance in NEI 94-01, Revision 3-A, and the conditions and limitations of NEI 94-01, Revision 2-A, the licensee justified the proposed change by demonstrating the adequate performance of the Harris, Unit 1, primary containment based on: (a) plant-specific containment leakage testing program results; (b) CISI results; and (c) a plant-specific risk assessment.

Based on the NRC staffs review of the licensees submittals of November 8, 2019, and April 16, 2020, and the regulatory and technical evaluations above, the NRC staff concludes that the licensee has addressed the NRC staffs conditions to demonstrate the acceptability of adopting NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, as the 10 CFR Part 50, Appendix J, Option B, implementation documents for Harris, Unit 1.

The NRC staff finds that the licensee has adequately implemented its Containment Leakage Rate Testing Program (i.e., Types A, B, and C leakage tests), for the Harris, Unit 1, primary containment. The results of past ILRTs and recent LLRTs demonstrate acceptable performance of the Harris, Unit 1, primary containment and demonstrate that the structural and leak-tight integrity of the containment structures are being adequately maintained. The NRC staff also finds that the structural and leak-tight integrity of the Harris, Unit 1, primary containment will continue to be monitored and maintained if Harris, Unit 1, adopts NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, as the 10 CFR Part 50, Appendix J, Option B, implementation documents for Harris, Unit 1. Accordingly, if the current

Type A test intervals are extended to 15 years and if the current Type B and Type C test intervals for qualifying penetrations and CIVs are extended up to a maximum of 120 months and 75 months, respectively, the NRC staff concludes that there is reasonable assurance that the structural and leak-tight integrity for the Harris, Unit 1, primary containment will continue to be maintained, without undue risk to public health and safety.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of North Carolina official was notified of the proposed issuance of the amendment on August 21, 2020 (ADAMS Accession No. ML20234A254). The State of North Carolina official responded on August 27, 2020, with no comments (ADAMS Accession No. ML20241A010).

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes the requirements with respect to installation or use of a facilitys components located within the restricted area as defined in 10 CFR Part 20 and 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 (85 FR 5052, dated January 28, 2020), and there has been no public comment on such finding. Accordingly, the amendments meet 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.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1. Hamilton, Tanya M., Duke Energy Progress, LLC., letter to U.S. Nuclear Regulatory Commission (NRC), License Amendment Request to Revise Technical Specifications for the Adoption of 10 CFR Part 50, Appendix J, Option B for Type B and C Testing and for Permanent Extension of Type A, B and C Leak Rate Test Frequencies, November 8, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19312C931).

2. Hamilton, Tanya M., Duke Energy Progress, LLC., letter to NRC, Supplement to License Amendment Request to Revise Technical Specifications for the Adoption of 10 CFR Part 50, Appendix J, Option B for Type B and C Testing and for Permanent Extension of Type A, B and C Leak Rate Test Frequencies, April 16, 2020 (ADAMS Accession No. ML20107F843).

3. U.S. Nuclear Regulatory Commission, Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, September 1995 (ADAMS Accession No. ML003740058).

4. Nuclear Energy Institute (NEI) 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 2012 (ADAMS Accession No. ML122210254 (package); ML12221A202 (report)).

5. Nuclear Energy Institute 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, November 19, 2008 (ADAMS Accession No. ML100620847).

6. Nuclear Energy Institute 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 21, 1995 (ADAMS Accession No. ML11327A025).

7. Electric Power Research Institute (EPRI) Report No. 1009325, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2, Final Report, August 2007 (ADAMS Accession No. ML072970204 (package); ML072970208 (report)).

8. U.S. Nuclear Regulatory Commission, RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, May 2011 (ADAMS Accession No. ML100910006).

9. U.S. Nuclear Regulatory Commission, RG 1.177, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, Revision 1, May 2011 (ADAMS Accession No. ML100910008).

10. Maxin, M.J., NRC, letter to John C. Butler, NEI, Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 94-01, Revision 2, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, and Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (TAC No. MC9663), June 25, 2008 (ADAMS Accession No. ML081140105).

11. Bahadur, S., NRC, letter to Biff Bradley, NEI, Final Safety Evaluation of Nuclear Energy Institute (NEI) TR 94-01, Revision 3, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J (TAC No. ME2164), June 8, 2012 (ADAMS Accession No. ML121030286).

12. U.S. Nuclear Regulatory Commission, RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009 (ADAMS Accession No. ML090410014).

13. U.S. Nuclear Regulatory Commission, Regulatory Issue Summary (RIS) 2007-06, Regulatory Guide 1.200 Implementation, March 22, 2007 (ADAMS Accession No. ML070650428)

14. Patel, Chandu P., NRC to Gannon, Cornelius J., Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Regarding Containment Integrated Leak Rate Test (TAC No. MC6722), March 30, 2006 (ADAMS Accession No. ML060720228)

15. U.S. Nuclear Regulatory Commission, NUREG-1916, Volume 1, Safety Evaluation Report Related to the License Renewal of Shearon Harris Nuclear Power Plant, Unit 1, Docket No.

50-400, Published November 2008 (ADAMS Accession No. ML090060733)

16. U.S. Nuclear Regulatory Commission, RG 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 17, August 30, 2014 (ADAMS Accession No. ML13339A689).

17. U.S. Nuclear Regulatory Commission, Information Notice (IN) 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner, May 5, 2014 (ADAMS Accession No. ML14070A114).

18. U.S. Nuclear Regulatory Commission, IN 2004-09, Corrosion of Steel Containment and Containment Liner, April 27, 2004 (ADAMS Accession No. ML041170030).

19. U.S. Nuclear Regulatory Commission, IN 2010-12, Containment Liner Corrosion, June 18, 2010 (ADAMS Accession No. ML100640449).

20. U.S. Nuclear Regulatory Commission, RIS 2016-07, Containment Shell or Liner Moisture Barrier Inspection, May 9, 2016 (ADAMS Accession No. ML16068A436).

21. U.S. Nuclear Regulatory Commission, RG 1.54, Service Level I, II, and III Protective Coatings Applied to Nuclear Power Plants, Revision 0, June 30,1973 (ADAMS Accession No. ML003740187).

22. Hamilton, Tanya M., Duke Energy Progress, LLC., letter to NRC, Relief Request I3R-18, Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, lnservice Inspection Program for Containment, Third Ten-Year Interval, June 4, 2018 (ADAMS Accession No. ML18156A026).

23. Shoop, Undine, NRC letter to Hamilton, Tanya M., Shearon Harris Nuclear Power Plant, Unit 1 Relief Request I3R-18, Regarding Alternative Repair and Replacement Testing Requirements for the Containment Building Equipment Hatch Sleeve Weld, Inservice Inspection Program for Containment, Third Ten-Year Interval (EPID L-2018-LLR-0081),

February 26, 2019 (ADAMS Accession No. ML19018A025).

24. Nuclear Energy Institute 94-01, Revision 3, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, June 9, 2011 (ADAMS Accession No. ML112920567 (report); ML112920561 (transmittal)).

25. Barillas, Martha, NRC, letter to Hamilton, Tanya M., Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Regarding Risk-Informed Justifications for the Relocation of Specific Surveillance Frequency Requirements to a Licensee-Controlled Program (CAC NO. MF6583), November 29, 2019 (ADAMS Accession No. ML16200A285).

26. Duncan, R. J., Carolina Power & Light Company, letter to NRC, Request for License Amendment to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), May 29, 2008 (ADAMS Accession No. ML081560641).

27. Nuclear Energy Institute 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidelines, Revision A3, March 20, 2000 (ADAMS Accession No. ML003728023).

28. Nuclear Energy Institute 05-04, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, Revision 2, November 30, 2008 (ADAMS Accession No. ML083430462)

29. Nuclear Energy Institute 07-12 Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, Revision 1, June 12, 2010 (ADAMS Accession No. ML102230070)

30. Nuclear Energy Institute 12-13 "External Hazards PRA Peer Review Process Guidelines,"

NEI 12-13, August 27,2012 (ADAMS Accession No. ML122400044).

31. Giitter, Joseph, and Ross-Lee, Mary Jane, NRC, letter to Kruger, Greg, NEI, U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 07-12, and 12-13, Close-Out of Facts and Observations (F&Os),

May 3, 2017 (ADAMS Accession No. ML17079A427)

32. U.S. Nuclear Regulatory Commission, NRC Results of Safety/Risk Assessment of Generic Issue 199 Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants: Safety/Risk Assessment September 2, 2010 (ADAMS Accession No. ML100270582).

33. U.S. Nuclear Regulatory Commission, NRC Results of Safety/Risk Assessment of Generic Issue 199 Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants: Safety/Risk Assessment, Appendix D, Seismic Core-Damage Frequencies, September 2, 2010 (ADAMS Accession No. ML100270756).

34. Pietrangelo, Anthony, NEI letter to Leeds, Eric J., NRC Seismic Risk Evaluations for Plants in the Central and Eastern United States, March 12, 2014 (ADAMS Accession No. ML14083A596).

35. Leeds, Eric J., NRC letter to All Power Reactor Licensees, Seismic Screening and Prioritization Results Regarding Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review of Insights, May 9, 2014 (ADAMS Accession No. ML14111A147).

36. Nuclear Energy Institute 94-01, Revision 2, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, August 31, 2007 (ADAMS Accession No. ML072970206).

37. Cruse, Charles H., Constellation Nuclear, Calvert Cliffs Nuclear Plant, letter to NRC, Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, March 27, 2020 (ADAMS Accession No. ML020920100).

Principal Contributors: D. Nold, NRR T. Nakanishi, NRR A. Istar, NRR M. Mahoney, NRR Date of Issuance: December 8, 2020

ML20259A512 *by memorandum **by e-mail OFFICE DORL/LPL2-2/PM** DORL/LPL2-2/LA** DSS/SCPB/BC* DEX/ESEB/BC*

NAME MMahoney BAbeywickrama BWittick SKrepel (A)

DATE 09/24/2020 09/19/2020 03/31/2020 04/15/2020 OFFICE DRA/ARCB/BC** DSS/STSB/BC** DEX/EMIB/BC** OGC - NLO**

NAME KHsueh VCusumano ABuford STurk DATE 08/20/2020 08/18/2020 08/26/2020 10/21/2020 OFFICE DORL/LPL2-2/BC** DORL/LPL2-2/PM**

NAME UShoop MMahoney DATE 12/04/2020 12/08/2020

v t e Letter Sequence Approval
EPID:L-2019-LLA-0256, (Approved, Closed)
Initiation Acceptance... Supplement Administration Questions Answers Results Approval Other:
MONTHYEAR2020-12-08 [Table View]

ML20259A512

Text

SUBJECT:

Dear Ms. Maza:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

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