ML22020A007
| ML22020A007 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 03/10/2022 |
| From: | Michael Mahoney Plant Licensing Branch II |
| To: | Maza K Duke Energy Progress |
| Mahoney M | |
| References | |
| EPID L-2021-LLA 0027 | |
| Download: ML22020A007 (15) | |
Text
March 10, 2022 Ms. Kim Maza Site Vice President Shearon Harris Nuclear Power Plant 5413 Shearon Harris Road Mail Code NHP01 New Hill, NC 27562-9300
SUBJECT:
SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 - ISSUANCE OF AMENDMENT NO. 192 REGARDING REMOVAL OF EXTRANEOUS CONTENT AND REQUIREMENTS FROM THE RENEWED FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS (EPID L-2021-LLA-0027)
Dear Ms. Maza:
The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued Amendment No. 192 to Renewed Facility Operating License No. NPF-63 for the Shearon Harris Nuclear Power Plant, Unit 1 (HNP). This amendment is in response to your application dated February 24, 2021, as supplemented by letter dated July 15, 2021.
The amendment removes, from the renewed facility operating license, License Condition 2.G, Reporting to the Commission, which required the licensee to report any violations of Operating License Section 2.C within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the U.S. Nuclear Regulatory Commission Operations Center via the Emergency Notification System with a written follow-up within 30 days.
Additionally, the amendment deletes HNP TS 3/4.4.10, Structural Integrity, revises Administrative Control TS 6.1.2 to eliminate the annual management directive requirement, revises TS Table 4.3-2, Engineered Safety Features Actuation System Instrumentation Surveillance Requirements, to remove an overly restrictive requirement that impedes the full application of the Surveillance Frequency Control Program for a specific subset of relays, and removes the TS Index and places it under licensee control.
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001
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. 192 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. 192 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 February 24, 2021, as supplemented by letter dated July 15, 2021, 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.
- 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. 192, are hereby incorporated into this license. Duke Energy Progress, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 90 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, 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: March 10, 2022 David J.
Wrona Digitally signed by David J. Wrona Date: 2022.03.10 14:31:26 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 192 SHEARON HARRIS NUCLEAR POWER PLANT, UNIT 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-63 DOCKET NO. 50-400 Replace the following pages of the Renewed Facility Operating License with the revised pages.
The revised pages are identified by amendment number and contain a marginal line indicating the areas of change:
Remove Insert Page 4 Page 4 Page 11 Page 11 Replace (or remove) 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 Index Cover Page i
ii iii iv v
vi vii viii ix x
xi xii xiii xiv xv xvi xvii xviii xix xx 3/4 3-41 3/4 3-41 3/4 3-42 3/4 3-42 3/4 3-46 3/4 3-46 3/4 3-49 3/4 3-49 3/4 4-43 3/4 3-43 6-1 6-1 Renewed License No. NPF-63 Amendment No. 192 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. 192, 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.
- On April 29, 2013, the name 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. 192 (c)
The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above.
G.
Deleted.
H.
The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.
I.
The Updated Safety Analysis Report supplement, as revised, submitted pursuant to 10 CFR 54.21(d), shall be included in the next scheduled update to the Updated Safety Analysis Report required by 10 CFR 50.71(e)(4) following the issuance of this renewed operating license. Until that update is complete, CP&L*
may make changes to the programs and activities described in the supplement without prior Commission approval, provided that CP&L* evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
J.
The Updated Safety Analysis Report supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. Duke Energy Progress, LLC shall complete these activities no later than October 24, 2026, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.
K.
All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of American Society for Testing and Materials E 185-82 to the extent practicable for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation.
All capsules placed in storage must be maintained for future inspection. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.
SHEARON HARRIS - UNIT 1 3/4 3-41 Amendment No. 154, 192 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL FUNCTIONAL UNIT CHANNEL CHECK CHANNEL CALIBRATION ANALOG CHANNEL OPERATIONAL TEST TRIP ACTUATING DEVICE OPERATIONAL TEST ACTUATION LOGIC TEST MASTER RELAY TEST SLAVE RELAY TEST MODES FOR WHICH SURVEILLANCE IS REQUIRED 1.
Safety Injection (Reactor Trip, Feedwater Isolation, Control Room Isolation, Start Diesel Generators, Containment Ventilation Isolation, Phase A Containment Isolation, Start Auxiliary Feedwater System Motor-Driven Pumps, Start Containment Fan Coolers, Start Emergency Service Water Pumps, Start Emergency Service Water Booster Pumps)
- a. Manual Initiation N.A.
N.A.
N.A.
SFCP N.A.
N.A.
N.A.
1, 2, 3, 4
- b. Automatic Actuation Logic and Actuation Relays N.A.
N.A.
N.A.
N.A.
SFCP(1)
SFCP(1)
SFCP 1, 2, 3, 4
N.A.
N.A.
N.A.
1, 2, 3, 4
- d. Pressurizer Pressure
N.A.
N.A.
N.A.
1, 2, 3 e.
Steam Line Pressure
N.A.
N.A.
N.A.
1, 2, 3
SHEARON HARRIS - UNIT 1 3/4 3-42 Amendment No. 154, 192 TABLE 4.3-2 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL FUNCTIONAL UNIT CHANNEL CHECK CHANNEL CALIBRATION ANALOG CHANNEL OPERATIONAL TEST TRIP ACTUATING DEVICE OPERATIONAL TEST ACTUATION LOGIC TEST MASTER RELAY TEST SLAVE RELAY TEST MODES FOR WHICH SURVEILLANCE IS REQUIRED 2.
- a. Manual Initiation N.A.
N.A.
N.A.
SFCP N.A.
N.A.
N.A.
1, 2, 3, 4
- b. Automatic Actuation Logic and Actuation Relays N.A.
N.A.
N.A.
N.A.
SFCP(1)
SFCP(1)
SFCP 1, 2, 3. 4
- c. Containment Pressure--
N.A.
N.A.
N.A.
1, 2, 3 3.
Containment Isolation a.
Phase "A" Isolation
- 1) Manual Initiation N.A.
N.A.
N.A.
SFCP N.A.
N.A.
N.A.
1, 2, 3, 4
- 2) Automatic Actuation Logic and Actuation Relays N.A.
N.A.
N.A.
N.A.
SFCP(1)
SFCP(1)
SFCP 1, 2, 3, 4 3)
Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
b.
Phase "B" Isolation
- 1) Manual Containment Spray Initiation See Item 2.a. above for Manual Containment Spray Surveillance Requirements.
- 2) Automatic Actuation Logic Actuation Relays N.A.
N.A.
N.A.
N.A.
SFCP(1)
SFCP(1)
SFCP 1, 2, 3, 4
SHEARON HARRIS - UNIT 1 3/4 3-46 Amendment No. 154, 192 TABLE 4.3-2 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL FUNCTIONAL UNIT CHANNEL CHECK CHANNEL CALIBRATION ANALOG CHANNEL OPERATIONAL TEST TRIP ACTUATING DEVICE OPERATIONAL TEST ACTUATION LOGIC TEST MASTER RELAY TEST SLAVE RELAY TEST MODES FOR WHICH SURVEILLANCE IS REQUIRED 6.
Auxiliary Feedwater (Continued) f.
Trip of All Main Feedwater Pumps Start Motor-Driven Pumps N.A.
N.A.
N.A.
SFCP N.A.
N.A.
N.A.
1, 2 g.
Steam Line Differential Pressure--High SFCP SFCP SFCP N.A.
N.A.
N.A.
SFCP 1, 2, 3 Coincident With Main Steam Line Isolation (Causes AFW Isolation)
See Item 4. above for all Main Steam Line Isolation Surveillance Requirements.
7.
Safety Injection Switchover to Containment Sump a.
Automatic Actuation Logic and Actuation Relays N.A.
N.A.
N.A.
N.A.
SFCP(1)
SFCP(1)
SFCP 1, 2, 3, 4 b.
RWST Level --Low-Low SFCP SFCP SFCP N.A.
N.A.
N.A.
SFCP 1, 2, 3, 4 Coincident With Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
8.
Containment Spray Switchover to Containment Sump a.
Automatic Actuation Logic and Actuation Relays N.A.
N.A.
N.A.
N.A.
SFCP(1)
SFCP(1)
SFCP 1, 2, 3, 4
SHEARON HARRIS - UNIT 1 3/4 3-49 Amendment No. 154, 187, 192 TABLE 4.3-2 (Continued)
TABLE NOTATION (1)
Each train shall be tested at the frequency specified in the Surveillance Frequency Control Program.
(2)
The Surveillance Requirements of Specification 4.9.9 apply during CORE ALTERATIONS or movement of irradiated fuel in containment.
(3)
Deleted.
(4)
The Steam Line Isolation-Safety Injection (Block-Reset) switches enable the Negative Steam Line Pressure Rate--High signal (item 4.e) when used below the P-11 setpoint.
Verify proper operation of these switches each time they are used.
Setpoint verification not required.
During CORE ALTERATIONS or movement of irradiated fuel in containment.
Trip Function automatically blocked above P-11 and may be blocked below P-11 when safety injection or low steamline pressure is not blocked.
The functions of the Reactor Trip, P-4 interlock required to meet the LCO are:
Trip the main turbine - MODES 1 and 2 Isolate Main Feedwater with coincident low Tavg - MODES 1, 2, and 3 Prevent reactuation of Safety Injection after a manual reset of Safety Injection -
MODES 1, 2, and 3 Prevent opening of Main Feedwater valves if closed on Safety Injection or Steam Generator Water Level - High High - MODES 1, 2, and 3
SHEARON HARRIS - UNIT 1 3/4 4-43 Amendment No. 119, 192 REACTOR COOLANT SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY - DELETED
SHEARON HARRIS - UNIT 1 6-1 Amendment No. 177, 192 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.
6.1.2 The Superintendent-Shift Operations (or, during his absence from the control room, a designated individual) shall be responsible for the control room command function.
6.2 ORGANIZATION 6.2.1 Onsite And Offsite Organization An onsite and an offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.
a.
Lines of authority, responsibility and communication shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. Those relationships shall be documented and updated, as appropriate, in the form of organizational charts. These organizational charts will be documented in the FSAR and updated in accordance with 10 CFR 50.71(e).
b.
There shall be an individual executive position (corporate officer) in the offsite organization having corporate responsibility for overall plant nuclear safety. This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured.
c.
There shall be an individual management position in the onsite organization having responsibility for overall unit safe operation and shall have control over those onsite resources necessary for safe operation and maintenance of the plant.
d.
Although the individuals who train the operating staff and those who carry out the quality assurance functions may report to the appropriate manager onsite, they shall have sufficient organizational freedom to be independent from operating pressures.
e.
Although health physics individuals may report to any appropriate manager onsite, for matters relating to radiological health and safety of employees and the public, the health physics manager shall have direct access to that onsite individual having responsibility for overall unit management. Health physics personnel shall have the authority to cease any work activity when worker safety is jeopardized or in the event of unnecessary personnel radiation exposures.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 192 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 February 24, 2021 (Reference 1), as supplemented by letter dated July 15, 2021 (Reference 6), Duke Energy Progress, LLC (Duke Energy or the licensee), requested changes to the renewed facility operating license and technical specifications (TSs) for the Shearon Harris Nuclear Power Plant (Harris or HNP), Unit 1.
The amendment proposed to delete, from the renewed facility operating license, License Condition 2.G, Reporting to the Commission, which requires the licensee to report any violations of Operating License Section 2.C within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the U.S. Nuclear Regulatory Commission Operations Center via the Emergency Notification System with a written follow-up within 30 days.
Additionally, the amendment proposed to delete HNP TS 3/4.4.10, Structural Integrity, revise Administrative Control TS 6.1.2 to eliminate the annual management directive requirement, revises TS Table 4.3-2, Engineered Safety Features Actuation System Instrumentation Surveillance Requirements, to remove an overly restrictive requirement that impedes the full application of the Surveillance Frequency Control Program for a specific subset of relays, and remove the TS Index from TS and relocate them to a licensee-controlled document.
The supplement dated July 15, 2021, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs initial proposed no significant hazards consideration determination as published in the Federal Register on June 15, 2021 (86 FR 31742).
2.0 REGULATORY EVALUATION
2.1 Description of Changes The licensee proposed the following changes:
Remove License Condition 2.G, Reporting to the Commission, from the renewed facility operating license.
Delete TS 3/4.4.10, Structural Integrity.
Revise Administrative Control TS 6.1.2 to eliminate the annual management directive requirement.
Revise TS Table 4.3-2, Engineered Safety Features Actuation System Instrumentation Surveillance Requirements, to remove an overly restrictive requirement that impedes the full application of the Surveillance Frequency Control Program for a specific subset of relays.
Remove the TS Index.
2.1.1 Removal of License Condition 2.G The licensee proposes to delete, from the renewed facility operating license, License Condition 2.G, Reporting to the Commission, which requires the licensee to report any violations of Operating License Section 2.C within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Nuclear Regulatory Commission Operations Center via the Emergency Notification System with a written follow-up within 30 days.
The current License Condition 2.G, states, as follows:
G.
Reporting to the Commission Except as otherwise provided in the Technical Specifications or Environmental Protection Plan, Duke Energy Progress, LLC, shall report any violations of the requirements contained in Section 2.C of this license in the following manner initial notification shall be made withing twenty four (24) hours to the NRC Operations Center via Emergency Notifications System with written follow up within 30 days in accordance with the procedures described in 10 CFR 50.73 (b), (c) and (e).
The requested change is to delete License Condition 2.G, states, as follows:
G.
Deleted.
2.1.2 Deletion of TS 3/4.4.10, Structural Integrity The licensee proposes to delete the text for TS 3/4.4.10, in its entirety.
The current TS 3/4 4.10, states, as follows:
REACTOR COOLANT SYSTEM 3/4 4.10 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of [American Society of Mechanical Engineers]
ASME Code Class 1, 2, and 3 components shall be maintained in accordance with Specification 4.4.10.
APPLICABILITY: ALL MODES ACTION:
- a.
With the structural integrity of any ASME Code Class 1 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature more than 50F above the minimum temperature required by NDT considerations.
- b.
With the structural integrity of any ASME Code Class 2 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) prior to increasing the Reactor Coolant System temperature above 200F.
- c.
With the structural integrity of any ASME Code Class 3 component(s) not conforming to the above requirements, restore the structural integrity of the affected component(s) to within its limit or isolate the affected component(s) from service.
SURVEIILLANCE REQUIREMENTS 4.4.10 Each reactor coolant pump flywheel shall be inspected per recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. In lieu of Position C.4.b(1) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at 20 year intervals.
The requested change to 3/4 4.10 will state, as follows:
REACTOR COOLANT SYSTEM 3/4 4.10 STRUCTURAL INTEGRITY DELETED 2.1.3 Revision of TS 6.1.2 The licensee proposes to revise Administrative Control TS 6.1.2 to eliminate the annual management directive requirement.
The current TS 6.1.2, states, as follows:
The Superintendent - Shift Operations (or, during his absence from the control room, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the Vice President-Harris Nuclear Plant shall be reissued to all station personnel on an annual basis.
The request change to TS 6.1.2, will state, as follows:
The Superintendent - Shift Operations (or, during his absence from the control room, a designated individual) shall be responsible for the control room command function.
2.1.4 Revision of TS Table 4.3-2 The licensee proposed to revise TS Table 4.3-2, Engineered Safety Features Actuation System Instrumentation Surveillance Requirements to remove an overly restrictive requirement that impedes the full application of the Surveillance Frequency Control Program for a specific subset of relays.
The current Note 3 of TS Table 4.3-2, states, as follows:
(3) Except for relays K601, K602, K603, K608, K610, K615, K616, K617, K622, K636, K739, K740 and K741 which shall be tested at the frequency specified in the Surveillance Frequency Control Program and during each COLD SHUTDOWN exceeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless they have been tested within the previous 92 days.
The licensee proposed to delete Note 3 of TS Table 4.3-2, and will state, as follows:
(3) Deleted.
The surveillance frequency information for Slave Relay Tests of the following Channel Functional Units will be updated to reflect the removal of Note 3, as shown in Table 1 below (deletions are shown in double-strikethrough):
Table 1: Channel Functions in HNP TS Table 4.3-2 That Will Be Updated Function ESFAS Instrumentation SRs of Slave Relay Test
- 1. Safety Injection
- b. Automatic Actuation Logic and Actuation Relays SFCP(3)
- 3. Containment Isolation
- a. Phase A Isolation
- 2) Automatic Actuation Logic and Actuation Relays SFCP(3)
- g. Steam Line Differential Pressure-High Coincident With Main Steam Line Isolation SFCP(3)
- 7. Safety Injection Switchchover to Containment Sump
- a. Automatic Actuation Logic and Actuation Relays SFCP(3)
- 8. Containment Spray Switchover to Containment Sump
- a. Automatic Actuation Logic and Actuation Relays SFCP(3)
Note: ESFAS: Engineered Safety Feature Actuation System SFCP: Surveillance Frequency Control Program SR: Surveillance Requirements 2.1.5 Removal of TS Index The licensee proposes to remove the TS Index from TS and relocate them to a licensee-controlled document. The TS Index will no longer be included in the NRC-issued TS and will no longer be a part of TS.
This proposed change deletes the TS Index cover page and TS pages i through xx.
2.2 Applicable Regulatory Requirements and Guidance Regulations The regulation at Title 10 of the Code of Federal Regulations (10 CFR) 50.36, Technical specifications, establishes the requirements related to the content of the TSs and require that TSs for operation include items in the following categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review, and audit, and reporting necessary to assure operation of the facility in a safe manner. The TS Index is not relied upon to assure safe operation of the facility and is not required to be in the TSs per the requirements of 10 CFR 50.36.
The regulation at 10 CFR 50.36(c)(2), Limiting conditions for operation, states, in part:
(i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.
The regulation at 10 CFR 50.36(c)(3), Surveillance requirements, states:
Surveillance requirements 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 limiting conditions for operation will be met.
The regulation at 10 CFR 50.36(c)(5), Administrative controls, states:
Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review, and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in [10 CFR] 50.4.
The regulations at 10 CFR 50.54, Condition of licenses, specify requirements regarding responsibilities and staffing of license operators. Specifically, 10 CFR 50.54(l) discusses the responsibility of the licensee to designate individuals to be responsible for directing the licensed activities of licensed operators. Additionally, 10 CFR 50.54(m) discusses reactor operators and senior reactor operators licensed under 10 CFR Part 55.
The regulations at 10 CFR 50.55a, Codes and Standards, in part, specify requirements for the design of American Society of Mechanical Engineers (ASME) code class or safety class components and for implementing and conducting required inservice inspections (ISI) and inservice testing (IST) of these components. The rule invokes use of the following ASME Codes for these components: (1) ASME Section III, Division 1 for design requirements (current Edition of Record for the facility), (2) ASME Section XI, Division 1 for ISI requirements (current Edition or Record for the facility), and (3) ASME OM Code for IST requirements (current Edition of Record for the facility).
The regulations at 10 CFR 50.72, Immediate notification requirements for operating nuclear power reactors, and 10 CFR 50.73, Licensee event report system provide requirements for making prompt notifications and submitting written reports to the NRC, respectively.
The following 10 CFR Part 50, Appendix A, General Design Criteria (GDC) are applicable:
Criterion 13, Instrumentation and control, states:
Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
Criterion 20, Protection system functions, states:
The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.
Criterion 21, Protection system reliability and testability, states:
The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed.
Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated.
The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.
Criterion 22, Protection system independence, states:
The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis.
Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.
Regulatory Guidance NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Chapter 13, "Conduct of Operations," Section 13.1.2 - 13.1.3, Revision 7, "Operating Organization," provides guidance for the review of the structure, functions, and responsibilities of the onsite organization established to safely operate and maintain the facility (Reference 3).
NRC Regulatory Guide (RG) 1.14, Reactor Coolant Pump Flywheel Integrity (Reference 4),
provides guidance for the inspection, testing, and structural evaluation of reactor coolant pump (RCP) flywheels. As stated in Chapter 1 of the UFSAR, the licensee indicates that the Sheron Harris Nuclear Power Plant (SHNPP) project follows the recommendations in this RG with some noted exceptions to specified regulatory positions of the RG, as discussed in Chapter 1 of the UFSAR (Reference 5).
3.0 TECHNICAL EVALUATION
3.1 Evaluation of Removal of License Condition 2.G The licensee proposed changes to the Renewed Facility Operating License (RFOL) for HNP, specifically to delete License Condition 2.G, Reporting to the Commission, because the licensee stated that the proposed change was consistent with the consolidated line item improvement process (CLIIP) Notice of Availability, as published in the Federal Register on November 4, 2005 (70 FR 67202). This CLIIP Notice of Availability referenced the model safety evaluation and model no significant hazards considerations, as published in the Federal Register on August 29, 2005 (70 FR 51098).
A section or condition was included in the Facility Operating Licenses issued to some nuclear power plants requiring the licensee to make reports to the Nuclear Regulatory Commission (NRC or the Commission) regarding violations of other sections of the operating license (typically, as License Condition 2.C). In the case of HNP, it was included as License Condition 2.G and reads as follows:
Except as otherwise provided in the Technical Specifications or Environmental Protection Plan, Duke Energy Progress, LLC, shall report any violations of the requirements contained in Section 2.C of this license in the following manner initial notification shall be made within twenty four (24) hours to the NRC Operations Center via Emergency Notifications System with written follow up within 30 days in accordance with the procedures described in 10 CFR 50.73 (b),
(c) and (e).
In addition to the information provided to support licensing decisions, the NRC obtains information about plant operation, licensee programs, and other matters using a combination of inspections and reporting requirements. Routine or scheduled reports that are required to be submitted to the NRC are defined in the related regulations, specific license conditions, TSs, or an NRC-approved program document. The reporting of emergencies, unplanned events or conditions, and other special cases may also be addressed within such documents by the inclusion of reporting thresholds and are also the focus of the reporting requirements in 10 CFR 50.72, Immediate notification requirements for operating nuclear power reactors, and 10 CFR 50.73, Licensee event report system. Changes to the reporting regulations in 10 CFR Sections 50.72 and 50.73 became effective in January 2001 (see Federal Register notice dated October 25, 2000, 65 FR 63769) and included extending the allowable reporting times for licensee event reports (LERs) from 30 days to 60 days.
Section 2.G of the HNP RFOL requires the licensee to report any violations of the requirements of Section 2.C of the RFOL and defines the method and allowable time periods for such reports.
The reporting threshold (i.e., what is considered a violation) for the conditions included in Section 2.C of the RFOL duplicate those defined in 10 CFR 50.72 and 10 CFR 50.73. However, the reporting deadlines in the RFOL do not match those defined in the regulations following a rule change in January 2001. This difference in reporting requirements has led to variations in reporting since many Facility Operating Licenses do not contain the subject license condition.
Licensees with a 30-day reporting requirement in their Facility Operating License obtain less of the regulatory relief provided by the 2001 rulemaking. For those cases where the current Facility Operating License requirement to report violations is also reportable in accordance with the regulations defined in 10 CFR 50.72 and 10 CFR 50.73, the NRC staff finds that the regulations adequately address this issue and the elimination of the duplicative requirement in the Facility Operating License is acceptable.
Some of the conditions addressed in Section 2.G of the Facility Operating License may address the maintenance of particular programs, administrative requirements, or other matters where a violation of the requirement would not result in a report to the NRC in accordance with 10 CFR 50.72 or 10 CFR 50.73. In most cases, there are requirements for reports to the NRC related to these conditions in other regulations, the specific license condition or TS, or an NRC-approved program document. In other cases, there are reports to other agencies or news releases that would prompt a report to the NRC (in accordance with 10 CFR 50.72(b)(2)(xi)).
The NRC staff also assessed violations of administrative requirements that could be reportable under the current License Condition but may not have a duplicative requirement in a regulation or other regulatory requirement. The NRC staff is confident that the information related to such violations that is important to the NRCs regulatory functions would come to light in a time frame comparable to LER requirements. The information would become available to the appropriate NRC staff through the inspection program, updates to program documents, resultant licensing actions, public announcements, or some other reliable mechanism.
The NRC staff finds that elimination of Section 2.G from the HNP RFOL will not result in a loss of information to the NRC that would adversely affect either its goal to protect public health and safety or its ability to carry out its various other regulatory responsibilities. Therefore, elimination of Section 2.G of the HNP RFOL is acceptable.
3.2 Evaluation of Deletion of TS 3/4.4.10, Structural Integrity The licensee proposed changes to TS Section 3/4.4.10 that effectively deletes the TS limiting condition of operation (LCO) for Structural Integrity in TS Section 3.4.10. These changes show up in the LAR as: (1) a proposed deletion of TS Sections 3.4.10 and (2) a proposed deletion of TS Section 4.4.10.
3.2.1 Deletion of TS Sections 3.4.10 The current (LCO) requirements in TS Section 3.4.10 require the licensee to maintain the structural integrity of American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 components in accordance with TS SR Section 4.4.10. For ASME Code Class 1, 2, or 3 components with detected conditions that exceed the applicable ASME Code allowables, the LCO requires the licensee to restore the structural integrity of the affected component(s) to within its/their limits or else to isolate the components in accordance with the isolation provisions for the type of Code Class component in TS Section 3.4.10, subsection a., b., or c. The licensee proposed to delete TS Section 3.4.10 in its entirely, including proposed deletions of the Applicability and Action statement subsections in TS Section 3.4.10. The licensee justified these changes based on its determination that the LCOs in TS Section 3.4.10 are redundant with the ISI requirements and IST requirements for ASME Code Class 1, 2, and 3 components in the Commissions 10 CFR 50.55a regulation.
The NRC staff verified that the Commissions regulation in the 10 CFR that governs performance of ISI of ASME Code Class 1, 2, and 3 components is given in 10 CFR 50.55a, Codes and Standards, with 10 CFR 50.55a(g) providing the requirements for performing ISI of the Code Class 1, 2 and 3 components. The NRC staff observed that the provisions in 10 CFR 50.55a(g) require the licensee to perform ISI of ASME Code Class 1, 2, and 3 components in accordance the ASME Boiler and Pressure Vessel Code,Section XI (ASME Section XI) edition of record that is required for the components in 10 CFR 50.55a(b)(2) or in NRC staff-approved ASME Code Cases that apply to the ASME Section XI, as referenced in the current version of NRC RG 1.147 and allowed for by the Commissions provisions in 10 CFR 50.55a(b)(5). The NRC staff also verified that the applicable AMSE Section XI rules or Code Case alternatives:
(1) provide criteria for inspecting the ASME Code Class components in accordance with specified ASME-defined non-destructive examination (NDE) methods, (2) provide criteria for recording component conditions that are detected using the NDE technology, (3) establish applicable acceptance limits (Code allowables) for the reference condition types, and (4) provide provisions for taking appropriate corrective actions (e.g., performance of supplemental inspections, supplemental component evaluations, or implementation of component-specific repair or replacement activities) if those acceptance limits are not met.
The NRC staff noted that the provisions in TS Section 3.4.10 may be creating requirements for the licensee that actually differ from the type of corrective actions that would be implemented under ASME Section XI because the Code might not necessarily require the licensee to isolate the impacted components if a non-conforming condition (i.e., recordable condition exceeding the ASME Section XI allowable) was detected in the component. In addition, the NRC staff notes that some portions of the reactor coolant pressure boundary (RCPB; ASME Code Class 1 system components) are unisolable and the licensee would be expected to take appropriate corrective action on an unisolable ASME Code Class 1 component if relevant conditions were detected in the components that exceeded the AMSE allowable for the component. In this regard, the requirements in 10 CFR 50.55a and the provisions in ASME Section XI or in NRC staff-approved ASME Section XI Code Cases (as specified in the latest version of NRC RG 1.147 referenced in 10 CFR 50.55a) establish the applicable and appropriate ISI requirements for the current licensing basis (CLB).
The NRC staff also observed that deleting TS LCO Section 3.4.10 from the existing TS may be preferable because the provisions in 10 CFR 50.55a(z) allow the licensee to propose alternatives to the ASME Section XI requirements invoked by 10 CFR 50.55a(g) if the licensee can demonstrate that: (1) the proposed alternative will provide an acceptable level of quality and safety, or (2) compliance with the specified requirements... would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NRC staff noted that if implementation of the ASME ISI requirements were not tied to any TS requirements, the licensee could propose such alternative ISI structural integrity requirements or non-conforming condition corrective action requirements simply through submittal of a 10 CFR 50.55a(z)(1) or (2) based relief request for approval by the Director of the Office of Nuclear Reactor Regulation. However, by linking the ASME ISI requirements for non-conforming conditions to TS LCO Section 3.4.10, such alternatives would need to be proposed as amendments to technical specification under 10 CFR 50.90 if they were to differ from the actions required by the TS Action statement in TS Section 3.4.10.a., b., or c.
Therefore, based on this review, the NRC staff finds the proposed deletion of TS Section 3.4.10 to be acceptable for implementation by the licensee because: (1) the Commissions ISI requirements for ASME Code Class 1, 2, and 3 components are already governed in the CLB by the Commissions requirements in 10 CFR 50.55a, and in ASME Section XI or NRC staff-approved Code Cases for ASME Section XI, as referenced for implementation in the 10 CFR 50.55a rule, and (2) the NRC staff has determined that existing TS Section 3.4.10 requirements are either redundant with the applicable ASME Section XI or Code Case criteria or may be in conflict with them. Thus, the NRC staffs approval of this TS change leaves 10 CFR 50.55a as the Commissions regulation that governs ISI of the applicable ASME Code Class 1, 2, and 3 components in the Shearon Harris plant design.
3.2.2 Deletion of TS Section 4.4.10 The current SR requirements in TS Section 4.4.10 require the licensee to perform ISI of the reactor coolant pump (RCP) flywheels at HNP in accordance with the specified SR criteria:
Each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.
In lieu of Position C.4.b(l) and C.4.b(2), a qualified in-place UT examination over the volume from the inner bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at 20 year intervals.
The licensee proposes to relocate the exact provisions of TS Section 4.4.10 out of the TS SR requirement section. In its letter dated February 24, 2021, the licensee proposed to add a new TS Administrative Controls section, proposed TS Section 6.8.4.s, Reactor Coolant Pump Flywheel Inspection Program, as included in Enclosure 2 of the LAR. However, in its letter dated July 15, 2021, the licensee states Duke Energy no longer requests the addition of TS 6.8.4.s, Reactor Coolant Pump Flywheel Inspection Program, as this content is proposed to be maintained in other accepted exceptions to RG 1.14 in HNP UFSAR Section 1.8.
The NRC staff finds the proposed deletion of TS SR Section 4.4.10 to be acceptable because although the existing TS-required inspections of the RCP flywheels are linked to the current TS LCO provisions for ASME Code Class 1, 2, and 3 components in TS LCO Section 3.4.10, the flywheels do not constitute ASME Code 1, 2, or 3 pressure retaining components. The NRC staff also finds that, upon its initial review of the proposal to delete TS SR Section 4.4.10, it would be acceptable for the licensee to propose a new TS Administrative Controls Section for the RCP flywheels program, given that the licensee discusses its basis for implementing the guidelines of RG 1.14 (with noted exceptions) in Section 1.8 of the updated final safety analysis report (UFSAR) and defines the RCP flywheel program in UFSAR Section 5.4.1. However, the NRC staff finds that the licensees proposed wording for TS Administrative Controls Section 6.8.4.s, Reactor Coolant Pump Flywheel Program (which is identical to the previous wording for the RCP flywheel SRs in TS Section 4.4.10), might not be comprehensive or entirely consistent with the actual RCP flywheel program being implemented by the licensee in accordance with HNP UFSAR Sections 1.8 or 5.4.1 (refer to ADAMS Accession Nos.
ML20147A018 and ML20147A022 for the UFSAR chapters containing those sections).
Specifically, the NRC staff notes that, although the new TS administrative controls program for the RCP flywheels in TS Section 6.8.4.s appropriately accounts for the ISI inspections (with noted exceptions) that the licensee will perform on RCP flywheel discs in accordance with RG 1.14, the TS section did not account for other aspects of the program that is applied to the RCP flywheels, such as the periodic overspeed testing that is performed as part of the program (refer to UFSAR Section 5.4.1.1) or the fracture toughness or fatigue flaw growth analysis/evaluation that is credited for protection against postulated RCP flywheel missiles, as discussed in UFSAR Section 5.4.1.3.6. The NRC staff issued Request for Additional Information (RAI) - 1 to address this gap in information.
The licensee responded to RAI - 1 in its letter dated July 15, 2021 (Reference 6). The licensee stated that it is amending the LAR to move the criteria of TS Section 4.4.10 (as referenced at the beginning of this SE section) into Chapter 1.8 of the UFSAR for HNP, Unit 1, rather than into a newly proposed TS Section 6.8.4.s. The NRC staff finds this amendment of the LAR to be an acceptable alternative to defining the RCP flywheel program in a newly proposed Administrative Controls section in the TS because: (1) the NRC staff has confirmed that the licensee provides a full description of the RCP flywheel inspection, evaluation and testing program in Sections 1.8 and 5.4 of the UFSAR, (2) UFSAR Section 1.8 fully defines all exceptions that the licensee is taking to the regulatory guidance of RG 1.14, and (3) that licensees proposed amendment to delete TS Section 4.4.10, as amended in the letter July 15, 2021, is consistent with the basis credited in the CLB to implement its RCP flywheel program, which calls for the licensee to implement its RCP flywheel in the manner the program, as defined in Sections 1.8 and 5.4 of the UFSAR. In this regard, the NRC staff finds that the licensees basis for deleting TS Section 4.4.10 is supported by the licensees response to RAI -1 and stands on its own merits. The NRC staff also finds that the LAR does not need to include a new TS Administrative Controls Section that would serve as a TS replacement section for the previous requirements in TS Section 4.4.10.
3.3 Evaluation of Revision of TS 6.1.2 The licensee proposed changes to TS 6.1.2 to eliminate the annual management directive reporting requirement. Administrative Control TS 6.1.2 addresses the responsibility of the Superintendent-Shift Operations for the control room command function. The current content of TS 6.1.2 is as follows:
The Superintendent-Shift Operations (or, during his absence from the control room, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the Vice President-Harris Nuclear Plant shall be reissued to all station personnel on an annual basis.
Duke Energy is proposing the elimination of the management directive requirement since it is redundant to the requirements already imposed per HNP TS 6.1.2 and TS Table 6.2-1 for the control room command function. Per Section 10 CFR 50.36(c)(5), the Administrative Controls Section of TS is required to include the provisions relating to organization and management, procedures, record keeping, review and audit, and reporting necessary to ensure safe operation of the facility. The proposed amendment to TS 6.1.2 to remove the management directive requirement does not impact Duke Energys ability to ensure safe operation of the facility since there is no change related to the responsibility of the control room command function or the method in which it is assigned in the absence of the Superintendent-Shift Operations from the control room.
The proposed change to TS 6.2.1 specifies the requirement that the control room command function must be staffed by a senior reactor operator (SRO) qualified individual when the unit is in an operating condition and the associated licensing basis also reflects that the Control Room Supervisor must be SRO qualified. The NRC staff concludes that removing the Chief Nuclear Officers annual management directive from TS 6.2.1 is acceptable, and the revised wording maintains appropriate requirements regarding the control room command function.
The NRC staff finds that the proposed change is acceptable because the control room command function is maintained by licensed SROs who are duly qualified to ensure the safe operation of the plant.
3.4 Evaluation of Revision of TS Table 4.3-2 The licensee proposed to remove the Note 3 from the surveillance frequency of the Slave Relay Test in HNP TS Table 4.3-2, Engineered Safety Feature Actuation System (ESFAS)
Instrumentation Surveillance Requirements. The proposed changes are shown in Section 2.1.5 of this SE.
3.4.1 Slave Relay Test Function The definition of the Slave Relay Test in the Harris TS is as follows:
A Slave Relay Test shall be the energization of each slave relay and verification of Operability of each relay. The Slave Relay Test shall include a continuity check, as a minimum, of associated testable actuation devices.
3.4.2 NRC Staff Evaluation The NRC staff reviewed the LAR to evaluate the licensees proposed revision of the TS ESFAS Instrumentation SRs to determine whether they are considered installed instrumentation intended to continue to meet the requirements of 10 CFR 50.36(c)(3) and GDC 13.
The NRC staff also reviewed:
- 1. The NRC letter dated November 29, 2016, ADAMS Accession No. ML16200A285. The NRC issued the license amendment No. 154 to the Harris Operating License, allowing for the relocation of specific surveillance frequencies to a licensee controlled program.
- 2. NRC-approved Technical Specification Task Force (TSTF) Standard Technical Specifications (STS) change TSTF-425, Relocate Surveillance Frequencies to Licensee Control, RITSTF [Risk Informed TSTF] Initiative 5b, Revision 3, ADAMS Accession No. ML090850627.
- 3. Technical Specifications (TS) Shearon Harris Nuclear Power Plant Unit No. 1 - Docket No. 50-400.
In the LAR, the licensee stated, in part, that:
These relays had 18-month frequencies rather than the standard quarterly frequency due to the potential adverse consequences of testing them while the unit was in operation. Furthermore, these relays were, and currently are, required to be tested during any cold shutdown greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless tested within the previous 92 days. This portion of Note (3) effectively requires performance of the slave relay test for these relays in alignment with the standard quarterly frequency, if the opportunity presents itself (i.e., a cold shutdown greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless tested within the previous 92 days).
However, there is the additional caveat that the relays shall be tested during each COLD SHUTDOWN exceeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless they have been tested within the previous 92 days. This caveat restricts the ability to extend the surveillance frequency beyond the previously established frequency of a refueling cycle (i.e., 18 months), despite the relays being within the purview of the SFCP. In order to be able to utilize the SFCP to extend the surveillance frequency for these relays, Note (3) will need to be revised to remove this restriction.
The NRC staff verified that:
In accordance with the NRC-approved TSTF STS change TSTF-425, Relocate Surveillance Frequencies to Licensee Control, the licensee relocated the SF for performance of ESFAS instrumentation Slave Relay Test in NUREG-1431 (Standard Technical Specifications - Westinghouse Plants) ISTS [Improved Standard Technical Specifications] SR 3.3.2.6 from 92-day [quarterly] to SFCP. Therefore, the requirement to perform additional surveillances during each Cold Shutdown exceeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless tested within the previous 92 days of the Note 3 to the SF of Slave Relay Test is no longer necessary.
Per License Amendment No. 154 to the HNP Operating License, the SF quarterly (92-day) ESFAS Instrumentation Slave Relay Test in the HNP TS SR 4.3.2.1*, was similarly relocated to the SFCP.
- SR 4.3.2.1: Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.
The proposed Note 3 deletion from SF of the Slave Relay Test in TS Table 4.3-2 will not change the requirements of the total number of the channel to trip; the minimum channels operable; and the Modes (in HNP Table 3.3-3, Engendered Safety Features Actuation System Instrumentation) of the functions in Table 1 in Section 2.1.5 of this SE.
The proposed Note 3 deletion from SF of the Slave Relay Test in TS Table 4.3-2 will not change the trip setpoint, allowable value (in HNP Table 3.3-4, Engineered Safety Features Actuation System Instrumentation Trip Setpoints) of the functions in Table 1 in Section 2.1.5 of this SE.
In HNP TS Table 3.3-3, Action 14 (Functions 1.b; 3.a.2; 7.a; and 8.a), Action 16 (Function 7.b), and Action 19 (Function 6.g) state that one channel (or one additional, or one inoperable channel) may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification SR 4.3.2.1. Therefore, the proposed Note 3 deletion would remove this restriction from these actions of the functions in Table 1 in Section 2.1.5 of this SE.
Based on the description above, the NRC staff finds that:
The Slave Relay Test of the functions in Table 1 in Section 2.1.5 of this SE are tested within the SFCP. The proposed Note 3 deletion would remove this restriction for the SF of the Slave Relay Test only. The SRs are maintained to assure that the necessary of quality of systems and components (that include a continuity check of associated required testable actuation devices; therefore, the proposed TS change continues to satisfy the requirement of the 10 CFR 50.36(c)(3).
Furthermore, the proposed change is consistent with the NRC-approved TSTF-425, Revision 3.
Based on the above evaluations, the NRC staff finds that the licensees proposed revision of the TS ESFAS Instrumentation SRs in HNP TS Table 4.3-2 is acceptable.
3.5 Evaluation of Removal of TS Index The TS Index is a list which provides information as to where (i.e., page number) specific TS sections can be found and does not contain any technical information. Inclusion of a TS Index (or table of contents) is not required by 10 CFR 50.36. The licensee states that transferring control of the TS Index from the NRC to Duke Energy does not affect the operation, physical configuration, or function of plant equipment or systems.
Since there is no change to any TS requirement, the NRC staff finds that removal of the TS Index from the TSs is acceptable because the changes are editorial and minor, and the requirements in 10 CFR 50.36 continue to be met.
3.6 Technical Evaluation Summary The NRC staff has determined the following, based on its review:
Elimination of Section 2.G from the HNP RFOL will not result in a loss of information to the NRC that would adversely affect either its goal to protect public health and safety or its ability to carry out its various other regulatory responsibilities.
Deletion of TS 3/4.4.10, Structural Integrity, is acceptable to avoid potential conflicts with the ISI requirements for ASME Code Class 1, 2, and 3 in 10 CFR 50.55a and the licensees ASME Section XI edition of record for the reactor unit.
The NRC staff finds that the proposed change to TS 6.1.2 acceptable because the control room command function is maintained by licensed SROs who are duly qualified to ensure the safe operation of the plant.
Based on the evaluation in Section 3.5 of this SE, the staff finds that the proposed removal Note 3 from SF of the Slave Relay Test in TS Table 4.3-2 is a removal of an extraneous requirement. Therefore, the licensees proposed revision of the TS ESFAS Instrumentation SRs in this table will continue to satisfy the requirement of 10 CFR 50.36(c)(3). Furthermore, this proposed change will be consistent with TSTF-425, Revision 3.
The NRC staff finds that (1) the licensees proposed change to remove the Index pages from the TSs and transfer them to licensee control of the changes, is administrative in nature, and (2) the requirements of 10 CFR 50.36 will continue to be met. As such, the NRC staff finds that proposed change is acceptable.
Based on the considerations discussed above, the NRC staff concludes that the changes proposed by the licensee, as described in their letters dated February 24 and July 15, 2021, are therefore acceptable.
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 November 30, 2021. The State of North Carolina official had no comments.
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 (86 FR 31742 FR, dated June 15, 2021), 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. Duke Energy, License Amendment Request to Remove Extraneous Content and Requirements from the Operating License and Technical Specifications, Serial RA 0252, dated February 24, 2021, (ADAMS Accession No. ML21055A819).
- 2. NRC, Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment No. 184 Regarding Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times - RITSTF Initiative 4B, dated March 31, 2021 (ADAMS Accession No. ML21047A314).
- 3. NRC, Standard Review Plan - NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Chapter 13, Conduct of Operations, Section 13.1.2 - 13.1.3, Revision 7, Operating Organization, dated August 2016 (ADAMS Accession No. ML15007A296).
- 4. NRC, Regulatory Guide 1.14, Revision 1, Reactor Coolant Pump Flywheel integrity, August 1976 (ADAMS Accession No. ML003739936).
- 5. Sheron Harris Nuclear Plant, Unit 1, Amendment 63 to Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant, and Chapter 5, Reactor Coolant System and Connected Systems, May 15, 2020 (ADAMS Accession No. ML20147A018 for UFSAR Chapter 1 and ML20147A022 for UFSAR Chapter 5).
- 6. Duke Energy Serial Letter No. RA-21-0203, Response to Request for Additional Information Regarding License Amendment Request to Remove Extraneous Content and Requirements from the Operating License and Technical Specifications, July 15, 2021 (ADAMS Accession No. ML21196A452).
Principal Contributors: M. Mahoney, NRR J. Medoff, NRR H. Vu, NRR M. Keefe-Forsyth, NRR R. Grover, NRR Date of Issuance: March 10, 2022
ML22020A007 OFFICE DORL/LPL2-2/PM DORL/LPL2-2/LA DSS/STSB/BC(A)
NAME MMahoney RButler NJordan DATE 01/19/2022 1/27/2022 10/17/2021 OFFICE DRO/IOLB/TL DNRL/NVIB/BC DEX/EICB/BC NAME BGreen ABuford MWaters DATE 10/20/2021 01/19/2022 12/01/2021 OFFICE OCG - NLO DORL/LPL2-2/BC DORL/LPL2-2/PM NAME MCarpentier DWrona MMahoney DATE 03/07/2022 03/09/2022 03/10/2022