ML25087A128
| ML25087A128 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 04/24/2025 |
| From: | Bryant J Plant Licensing Branch II |
| To: | Snider S Duke Energy Carolinas |
| Williams S | |
| References | |
| EPID L-2024-LLA-0103 | |
| Download: ML25087A128 (1) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 24, 2025 Steven M. Snider Vice President, Oconee Nuclear Station Duke Energy Carolinas, LLC 7800 Rochester Highway Seneca, SC 29672-0752
SUBJECT:
OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 - ISSUANCE OF AMENDMENT NOS. 431, 433, 432, RE: REVISE TECHNICAL SPECIFICATION 3.7.7, LOW PRESSURE SERVICE WATER (LPSW)
SYSTEM, TO EXTEND THE COMPLETION TIME FOR ONE REQUIRED LPSW PUMP INOPERABLE ON A ONE-TIME BASIS (EPID L-2024-LLA-0103)
Dear Mr. Snider:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment Nos. 431, 433, and 432 to Subsequent Renewed Facility Operating Licenses DPR-38, DPR-47, and DPR-55, for the Oconee Nuclear Station, Units 1, 2, and 3, respectively.
The amendments are in response to the application from Duke Energy Carolinas, LLC, dated July 29, 2024.
The amendments revise Technical Specification (TS) 3.7.7, Low Pressure Service Water (LPSW) System, to extend the Completion Time for one required inoperable LPSW Pump on a one-time basis for Oconee Nuclear Station, Units 1, 2, and 3. Specifically, the amendments would replace an expired Completion Time Note associated with TS 3.7.7, Condition A, Required Action A.1, to now allow 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> to be used once during a subsequent Unit 2 refueling outage to allow for the tie-in and testing of an alternate suction source to the shared Unit 1 and Unit 2 A and B LPSW pumps. The Completion Time Note expires December 31, 2027.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice
If you have any questions, please call me at 301-415-0610, or by email at jack.minzerbryant@nrc.gov.
Sincerely,
/RA/
Jack Minzer Bryant, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287
Enclosures:
- 1. Amendment No. 431 to DPR-38
- 2. Amendment No. 433 to DPR-47
- 3. Amendment No. 432 to DPR-55
- 4. Safety Evaluation cc: Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 431 Subsequent Renewed License No. DPR-38
- 1.
The Nuclear Regulatory Commission (NRC, the Commission) has found that:
A.
The application for amendment to the Oconee Nuclear Station, Unit 1 (the facility), Subsequent Renewed Facility Operating License No. DPR-38, filed by Duke Energy Carolinas, LLC (the licensee), dated July 29, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 hereby amended by page changes to the Subsequent Renewed Facility Operating License and Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Subsequent Renewed Facility Operating License No. DPR-38 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 431, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Subsequent Renewed Facility Operating License No. DPR-38 and the Technical Specifications Date of Issuance: April 24, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.04.24 12:52:00 -04'00'
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 433 Subsequent Renewed License No. DPR-47
- 1.
The Nuclear Regulatory Commission (NRC, the Commission) has found that:
A.
The application for amendment to the Oconee Nuclear Station, Unit 2 (the facility), Subsequent Renewed Facility Operating License No. DPR-47, filed by Duke Energy Carolinas, LLC (the licensee), dated July 29, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 hereby amended by page changes to the Subsequent Renewed Facility Operating License and Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Subsequent Renewed Facility Operating License No. DPR-47 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 433, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Subsequent Renewed Facility Operating License No. DPR-47 and the Technical Specifications Date of Issuance: April 24, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.04.24 12:52:38 -04'00'
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT 3 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 432 Subsequent Renewed License No. DPR-55
- 1.
The Nuclear Regulatory Commission (NRC, the Commission) has found that:
A.
The application for amendment to the Oconee Nuclear Station, Unit 3 (the facility), Subsequent Renewed Facility Operating License No. DPR-55, filed by Duke Energy Carolinas, LLC (the licensee), July 29, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 hereby amended by page changes to the Subsequent Renewed Facility Operating License and Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 3.B of Subsequent Renewed Facility Operating License No. DPR-55 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 432, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Subsequent Renewed Facility Operating License No. DPR-55 and the Technical Specifications Date of Issuance: April 24, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.04.24 12:53:16 -04'00'
Attachment ATTACHMENT TO LICENSE AMENDMENT NOS. 431, 433, AND 432 OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NOS. DPR-38, DPR-47, AND DPR-55 DOCKET NOS. 50-269, 50-270, AND 50-287 Replace the following pages of the Operating Licenses and 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 Pages Insert Pages Operating Licenses Operating Licenses License No. DPR-38, page 3 License No. DPR-38, page 3 License No. DPR-47, page 3 License No. DPR-47, page 3 License No. DPR-55, page 3 License No. DPR-55, page 3 Technical Specifications Technical Specifications 3.7.7-1 3.7.7-1
Subsequent Renewed License No. DPR-38 Amendment 431
- 3.
This subsequent renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I, Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50 and Section 70.32 of Part 70; 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:
A.
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 431, are hereby incorporated in the subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
C.
This subsequent renewed operating license is subject to the following antitrust conditions:
Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.
Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in ¶1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.
- 1.
As used herein:
(a)
Bulk Power means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.
(b)
Neighboring Entity means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and Subsequent Renewed License No. DPR-47 Amendment No. 433
- 3.
This subsequent renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I, Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50 and Section 70.32 of Part 70; 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:
A.
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 433 are hereby incorporated in the subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
C.
This subsequent renewed operating license is subject to the following antitrust conditions:
Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.
Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in 1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.
- 1.
As used herein:
(a)
Bulk Power means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.
(b)
Neighboring Entity means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and Subsequent Renewed License No. DPR-55 Amendment No. 432
- 3.
This subsequent renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I, Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50 and Section 70.32 of Part 70; 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:
A.
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2610 megawatts thermal.
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 432 are hereby incorporated in the subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.
C.
This subsequent renewed operating license is subject to the following antitrust conditions:
Applicant makes the commitments contained herein, recognizing that bulk power supply arrangements between neighboring entities normally tend to serve the public interest. In addition, where there are net benefits to all participants, such arrangements also serve the best interests of each of the participants. Among the benefits of such transactions are increased electric system reliability, a reduction in the cost of electric power, and minimization of the environmental effects of the production and sale of electricity.
Any particular bulk power supply transaction may afford greater benefits to one participant than to another. The benefits realized by a small system may be proportionately greater than those realized by a larger system. The relative benefits to be derived by the parties from a proposed transaction, however, should not be controlling upon a decision with respect to the desirability of participating in the transaction. Accordingly, applicant will enter into proposed bulk power transactions of the types hereinafter described which, on balance, provide net benefits to applicant. There are net benefits in a transaction if applicant recovers the cost of the transaction (as defined in 1 (d) hereof) and there is no demonstrable net detriment to applicant arising from that transaction.
- 1.
As used herein:
(a)
Bulk Power means electric power and any attendant energy, supplied or made available at transmission or sub-transmission voltage by one electric system to another.
(b)
Neighboring Entity means a private or public corporation, a governmental agency or authority, a municipality, a cooperative, or a lawful association of any of the foregoing owning or operating, or proposing to own or operate, facilities for the generation and
LPSW System 3.7.7 (continued)
OCONEE UNITS 1, 2, & 3 3.7.7-1 Amendment Nos. 431, 433, & 432 3.7 PLANT SYSTEMS 3.7.7 Low Pressure Service Water (LPSW) System LCO 3.7.7 For Unit 1 or Unit 2, three LPSW pumps and one flow path shall be OPERABLE.
For Unit 3, two LPSW pumps and one flow path shall be OPERABLE.
The LPSW Waterhammer Prevention System (WPS) shall be OPERABLE.
NOTE---------------------------------------------
With either Unit 1 or Unit 2 defueled and appropriate LPSW loads secured on the defueled Unit, such that one LPSW pump is capable of mitigating the consequences of a design basis accident on the remaining Unit, only two LPSW pumps for Unit 1 or Unit 2 are required.
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One required LPSW pump inoperable.
A.1 Restore required LPSW pump to OPERABLE status.
NOTE------------
During a Unit 2 refueling outage with Unit 2 defueled, appropriate LPSW loads secured, and contingent on implementation of the compensatory measures described in Attachment 1 of letter RA-22-0089 dated April 14, 2022, the Completion Time is 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> for the tie-in and testing of an alternate suction source to the shared Unit 1/2 LPSW Pumps A and B. Only applicable one time and expires December 31, 2027.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR LICENSE AMENDMENT NOS. 431, 433, AND 432 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NOS. DPR-38, DPR-47, AND DPR-55 DUKE ENERGY CAROLINAS, LLC OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287
1.0 INTRODUCTION
By application dated July 29, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24212A048), Duke Energy Carolinas, LLC (Duke Energy or the licensee), requested changes to the technical specifications (TSs) for the Oconee Nuclear Station, Units 1, 2, and 3 (Oconee).
The proposed change would revise TS 3.7.7, Low Pressure Service Water (LPSW) System, to extend the Completion Time for one required inoperable LPSW pump on a one-time basis for Oconee, Units 1, 2, and 3. Specifically, the amendments would replace an expired Completion Time Note associated with TS 3.7.7, Condition A, Required Action A.1, to now allow 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> during a subsequent Unit 2 refueling outage to allow for the tie-in and testing of an alternate suction source to the shared Unit 1 and Unit 2 A and B LPSW pumps. The proposed one-time extended Completion Time would expire December 31, 2027. The alternate suction source to the shared Units 1 and 2 A and B LPSW pumps is needed to permit draining of the condenser circulating water (CCW) system cross-connect header for the replacement of three CCW valves.
The normal Completion Time for TS 3.7.7, Condition A, Required Action A.1 is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
2.0 REGULATORY EVALUATION
2.1 Plant and System Description Each Oconee unit is a two-loop pressurized water reactor (PWR) supplied by Babcock and Wilcox with a large dry containment. As stated in Section 1.2.2.2 of the Oconee Updated Final Safety Evaluation Report (UFSAR), Revision 30, (ML24179A111) the steam generators (SGs) in the two loops are vertical, straight tube units producing superheated steam at constant pressure. With the once-through design of the SGs, natural circulation flow is adequate to remove full decay heat without the use of reactor coolant pumps (RCPs). Each loop has two vertical, single speed centrifugal RCPs equipped with controlled leakage shaft seals. An electrically heated pressurizer establishes and maintains the reactor coolant system (RCS) pressure and provides a surge chamber and a water reserve to accommodate RCS volume changes during plant operation.
As described in the UFSAR, Section 6.2.1.1.1, the reactor building (RB) completely encloses the RCS to minimize release of radioactive material to the environment should a serious failure of the RCS occur. The RB structure designed for an internal pressure of 59 psig provides adequate biological shielding for both normal operation and accident situations.
The LPSW system description can be found in the UFSAR, Section 9.2.2.2.3. The LPSW system provides a heat sink for safety related components during a transient or accident. The LPSW system also provides this heat sink function during normal operation and normal shutdown for various components, including the reactor building cooling units (RBCU), low pressure injection (LPI) coolers, high pressure injection (HPI) pump motor coolers, and the motor driven emergency feedwater (EFW) pumps. In the license amendment request (LAR), the licensee stated that:
The LPSW [S]ystem for Unit 1 and Unit 2 is shared and consists of three LPSW pumps (i.e., A, B and C) which can supply multiple combinations of pathways to supply required components. The shared Unit 1 and 2 pumps take suction from the 42-inch cross-connection between the condenser inlet headers of all three units; two LPSW pumps (A and B) are supplied by one suction branch line and the other pump (C) is supplied by the other suction branch line.
The LPSW System for Unit 3 consists of two LPSW pumps and like the Unit 1 and 2 pumps, also take their suction from the CCW 42-inch cross-connection header.
The CCW system description can be found in the UFSAR, Section 9.2.2.2.1. The CCW system provides for cooling of the condensers during normal operation. The system generally uses lake water as the ultimate heat sink for decay heat removal during cooldown of the plant. The CCW system is the suction source for other service water systems, including high pressure service water (HPSW), LPSW, protected service water (PSW), and standby shutdown facility (SSF)
Auxiliary Service Water (ASW). In addition, CCW provides a heat sink for the recirculated cooling water system (RCW). The CCW system is designed to supply suction to the LPSW pumps from the 42-inch crossover header during normal operation and emergencies.
2.2 Licensee Proposed Changes Current TS 3.7.7 The Limiting Condition for Operation (LCO) for TS 3.7.7 specifies that for the shared Units 1 and 2 LPSW system, three LPSW pumps are required to be operable. The LCO is modified by a Note which requires only two LPSW pumps to be operable for the shared Units 1/2 LPSW system if either unit is defueled. The shared Units 1 and 2 LPSW system requires only two pumps to meet the single failure criterion provided that one of the units has been defueled and the following LPSW system loads on the defueled unit are isolated: RB cooling units, RB auxiliary coolers, component cooling, main turbine oil tank, RCPs and LPI. The LCO further requires that one flow path be operable for Unit 1 and one flow path be operable for Unit 2.
TS 3.7.7, Condition A, is the condition for one required LPSW pump inoperable. For Units 1 and 2, Condition A is entered when one of the three required LPSW pumps is inoperable unless Unit 1 or Unit 2 is defueled with the appropriate loads isolated. With Unit 1 or Unit 2 defueled with the appropriate loads isolated, Condition A would only be entered for Unit 1 or Unit 2 if two LPSW pumps are inoperable. Required Action A.1 specifies that action must be taken to restore the required LPSW pump to operable status within the completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The completion time was modified by a now expired note indicating that during Unit 2, Refuel 31 with Unit 2 defueled, appropriate LPSW loads secured, and contingent on implementation of contingency measures described in Attachment 1 of the Emergency Feedwater System Due to Main Feedwater Pump Malfunction|April 14, 2022 letter]] (RA-22-0089)
(ML22104A010), the completion time of 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br /> can be used for the tie-in and testing of an alternate suction source to the shared Unit 1/2 LPSW pumps A and B. However, the licensee stated in the current LAR that it was unable to perform the modification during Unit 2, Refuel 31 due to valve vendor delivery delays. Therefore, the Completion Time note expired without being used and the licensee has requested a new note to complete the work during a subsequent Unit 2 refueling outage.
Current TS 3.7.7, Condition A, Completion Time, Note states:
During Unit 2, Refuel 31with Unit 2 defueled, appropriate LPSW loads secured, and contingent on implementation of the compensatory measures described in Attachment 1 of letter RA-22-0089 dated April 14, 2022, the Completion Time is 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br /> for the tie-in and testing of an alternate suction source to the shared Unit 1/2 LPSW Pumps A and B.
Revised TS 3.7.7, Condition A, Completion Time, Note would state:
During a Unit 2 refueling outage with Unit 2 defueled, appropriate LPSW loads secured, and contingent on implementation of the compensatory measures described in Attachment 1 of letter RA-22-0089 dated April 14, 2022, the Completion Time is 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> for the tie-in and testing of an alternate suction source to the shared Unit 1/2 LPSW Pumps A and B. Only applicable one time and expires December 31, 2027.
The proposed Completion Time extension is requested to install a temporary alternate LPSW suction source to the shared Units 1 and 2 LPSW system.
The alternate LPSW suction source will permit Oconee to operate Unit 1 while Unit 2 is defueled. In turn, this will support draining the CCW cross-connect header to replace three CCW valves without requiring Unit 1 to be shut down while Unit 2 is defueled.
2.3 Applicable Regulatory Requirements and Guidance 2.3.1 Regulatory Requirements The regulation in Title 10 of the Code of Federal Regulations (10 CFR) Part 50.36(b) states, in part, that The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto... In determining whether the proposed TS remedial actions should be granted, the Commission will apply the reasonable assurance standards of 10 CFR 50.40(a) and 50.57(a)(3).
The regulation in 10 CFR 50.36(c)(2) requires that TSs contain LCOs. The regulations in 10 CFR Part 50.36(c)(2)(i) states, in part, that:
Limiting conditions for operation [LCO] 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 regulations in 10 CFR 50.65(a)(4) states:
Before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of the assessment may be limited to structures, systems, and components that a risk-informed evaluation process has shown to be significant to public health and safety.
The regulations in 10 CFR 50.46(b) require that during a loss-of-coolant accident (LOCA) event, the following criteria are satisfied:
(1) Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200° F.
(2) Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
(3) Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
(4) Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
(5) Long-term cooling. After any calculated successful initial operation of the ECCS
[emergency core cooling system], the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
The regulations in 10 CFR 50.40(a) state that in determining whether to grant a license, the Commission will be guided by, among other things, consideration about whether the processes to be performed, the operating procedures, the facility and equipment, the use of the facility, and other technical specifications, or the proposals, in regard to any of the foregoing collectively provide reasonable assurance that the applicant will comply with the regulations in this chapter, including the regulations in part 20 of this chapter, and that the health and safety of the public will not be endangered.
The regulations in 10 CFR 50.57(a)(3) state that the Commission may issue an operating license upon finding that, among other things, There is reasonable assurance (i) that the activities authorized by the operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the regulations in this chapter.
The Atomic Energy Commission (AEC) issued the construction permits for Oconee on November 6, 1967. The plants design approval for the construction phase was based on the proposed general design criteria (GDC) published by the AEC in the Federal Register (32 FR 10213) on July 11, 1967. Thus, the GDC that constitute the licensing basis for Oconee are those described in the UFSAR Chapter 3.1, Conformance with NRC General Design Criteria, and in applicable UFSAR sections. Based on its review of the UFSAR and the licensees submittal, the NRC staff identified the following UFSAR Chapter 3.1, Conformance with NRC General Design Criteria, as being applicable to the proposed amendment:
UFSAR, Section 3.1.6, Criterion 6 - Reactor Core Design (Category A), states, in part, that:
The reactor core shall be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all off-site power.
UFSAR, Section 3.1.10, Criterion 10 - Containment (Category A), states, in part, that:
Containment shall be provided. The containment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary break, without loss of required integrity and, together with other engineered safety features as may be necessary, to retain for as long as the situation requires the functional capability to protect the public.
UFSAR, Section 3.1.41, Criterion 41 - Engineered Safety Features Performance Capability (Category A), states, in part, that:
Engineered safety features such as Emergency Core Cooling and Containment Heat Removal Systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component.
UFSAR, Section 3.1.49, Criterion 49 - Containment Design Basis (Category A), states, in part:
The containment structure, including access openings and penetrations, and any necessary containment heat removal systems shall be designed so that the containment structure can accommodate without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a loss-of-coolant accident, including a considerable margin for effects from metal-water or other chemical reactions that could occur as a consequence of failure of Emergency Core Cooling Systems.
UFSAR, Section 3.1.52, Criterion 52 - Containment Heat Removal Systems (Category A),
states, in part:
Where active heat removal systems are needed under accident conditions to prevent exceeding containment design pressure, at least two systems, preferably of different principles, each with full capacity shall be provided.
2.3.2 Regulatory Guidance Chapter 18, Human Factors Engineering, of NUREG-0800 (ML16125A114), describes the guidance used by NRC staff to conduct regulatory reviews of license amendments that address human factors topics. It directs NRC staff to applicable review criteria such as those included in NUREG-1764, Guidance for the Review of Changes to Human Actions, Revision 1, (ML072640413). NUREG-1764 describes how to assess changes to manual operator actions. It provides a risk-informed process to determine the level of NRC review necessary and provides the acceptance criteria for each level of review.
3.0 TECHNICAL EVALUATION
Section 3.5 of the LAR describes how the licensee utilizes their responses to the requests for additional information (RAIs) in the Letter from Duke Energy to NRC, Response to Request for Additional Information (RAI) Regarding Application to Revise Technical Specification 3.7.7, Low Pressure Service Water (LPSW) System, to Extend the Completion Time for One Required Inoperable LPSW Pump on a Temporary Basis, dated April 14, 2022 (ML22104A010) associated with a previously approved LAR dated September 2, 2021 (ML21245A210). In Section 3.5 of the LAR the licensee further states which RAI responses to the previous LAR are applicable and appropriate to the current LAR or modifies the responses to address any changes due to the differences between the previous and current requests.
3.1 Accident Analysis Section 2.2 of the LAR states that the LPSW system supplies cooling water to the safety-related RBCU system and LPI system coolers. These systems are used to mitigate the consequences of a design basis (DB) loss-of-coolant accident (LOCA) (DBLOCA).
3.1.1 Unit 1 DBLOCA Analysis As described in Oconee UFSAR Section 9.2.2.2.3, the DBLOCA in Unit 1 is a worst case of a large break LOCA in conjunction with loss of offsite power (LOOP). During this scenario, while Unit 1 is in LCO 3.7.7, only Unit 1 and 2 LPSW pump C flow will be available for the mitigation of a LOCA during the proposed LCO Completion Time of 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br />. If the LPSW pump C fails during this time, the licensee has proposed to manually connect one of the Unit 3 LPSW pump to supply Unit 1 RBCU and LPI system safety-related cooling loads through a beyond-design-basis cross-connect piping between the units. In Section 3.5 of the LAR Enclosure, the licensee stated the response to Nuclear Systems Performance Branch (SNSB) Request for Additional Information (RAI) No. 1 (ML22104A010) on the previous LAR (ML21245A210) is applicable and appropriate for the requested 360-hour Completion Time in this LAR.
On reviewing the response to SNSB-RAI No. 1, the NRC staff determines the following:
The time critical operator action time to manually cross-connect Unit 3 LPSW pump with Unit 1 and 2 LPSW system will be 14.72 minutes which includes the dispatch and local task time. The licensee stated that the cross-connect action time of 14.72 minutes has been verified using Oconee procedure PT/0/A/0120/033.
For containment peak pressure and temperature response, an RCS large cold-leg break LOCA gives more limiting (higher) results than an RCS large hot-leg break LOCA. For all 3 cold-leg break LOCA cases R1, R2, and R3 (ML22104A010, response to SNSB RAI No. 1, Part b for analysis results), the peak RB pressure and vapor temperature values are bounded by the Analysis of Record (AOR) values shown in UFSAR Figures 6-36 and 6-37. The analyzed peak temperature remains bounded by the AOR equipment qualification temperature profile. The peak RB wall temperature based on the peak vapor temperature also remains bounded by the AOR RB structural design temperature of 286ºF.
For the RB sump temperature response, an RCS large hot-leg break LOCA produces more limiting (higher temperature) results than an RCS large cold-leg break LOCA. The large hot-leg break LOCA also produces more limiting (lesser) available containment accident pressure (CAP) for the LPI and reactor building spray (RBS) systems pumps Net-Positive Suction Head (NPSH). NPSH requirements for all LPI pumps are met using the saturation pressure of the sump water. The NPSH analysis results shows positive NPSH margin (NPSH available minus NPSH required) for the RBS pumps and therefore pump performance during the LOCA recirculation phase is acceptable.
The NRC staff review finds: (i) the method used to determine the time critical operator action is acceptable because it was determined according to an Oconee procedure, (ii) the containment pressure and temperature response is acceptable because it is bounded by the AOR results, and (iii) the RBS pumps will perform satisfactorily because their NPSH margin is positive.
3.1.2 Mitigation of Transient and Accident Events Listed in Oconee UFSAR Table 15-32 The unavailability of the LPI system for the temporary period from 0 to 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> may impact the analysis results of the transient and accident cases listed in UFSAR Table 15-32. In Section 3.5 of the LAR Enclosure, the licensee states the following:
Duke Energy provided a response to an NRC request for additional information associated with the previously approved LAR by Emergency Feedwater System Due to Main Feedwater Pump Malfunction|letter dated April 14, 2022]] (RA-22-0089). The following addresses each RAI response and states whether it is applicable and appropriate to this LAR or modifies the response to address any changes due to this new request.
In Section 3.5 of the LAR Enclosure, the licensee stated the response to SNSB RAI No. 2 (ML22104A010) in the previous LAR discussed above (ML21245A210) is applicable and appropriate for the requested 360-hour Completion Time in this LAR.
On reviewing the response to SNSB RAI No. 2, the NRC staff determined the following:
Among the events listed in UFSAR Table 15-32, the UFSAR Section 15.6, Locked Rotor, Section 15.9, Steam Generator Tube Rupture, Section 15.12, Rod Ejection, Section 15.13, Steam Line Break, and Section 15.17, Small Steam Line Break events, only credit the LPI system during their cooldown portions of their dose consequences analysis. The NRC staff finds that the unavailability of the LPI coolers for the short time of 14.72 minutes required for cross-connecting Unit 3 LPSW pump with Unit 1 and 2 LPSW system at any time during the 0-360 hours period will not impact their dose consequence analyses.
For the LOCA event, the UFSAR Section 15.14 addresses the ECCS acceptance criteria in 10 CFR 50.46(b)(1) through (b)(5). During the LOCA cold-leg safety injection phase, to satisfy the first four criteria, i.e., (b)(1) through (b)(4), the LPI system is not credited.
To satisfy the 10 CFR 50.46(b)(5) criteria, the LPI system is credited in the LOCA long-term cooling analysis and to maintain the post-LOCA boric acid solubility during the sump recirculation phase. During the unavailability of the LPSW cooling flow through the LPI coolers for 15 minutes in Unit 1, the LPI flow would still continue ensuring adequate circulation to prevent boron concentration buildup.
During small break LOCAs, the LPI fluid temperature is important for the operation of the HPI pumps because the LPI system supplies cooling water to the HPI pump motor coolers. The licensee stated that two of the three HPI pumps may be aligned to take suction from the LPI system during the sump recirculation phase of the LOCA, leaving one HPI pump in reserve. If the sump fluid temperature exceeds 200ºF during the time LPI cooling is lost, operator action will secure the operating HPI pumps until LPI cooling is restored. During this scenario, SG cooling will be used.
The NRC staff finds the 10 CFR 50.46(b)(1) through (b)(4) criteria are not impacted because the LPI flow is not credited prior to the LOCA sump recirculation phase. For satisfying the long-term cooling 10 CFR 50.46(b)(5) criteria, the licensees proposed methods of cooling using the HPI system and SGs are acceptable to the NRC staff.
3.1.3 Decay Heat Removal in Unit 3 Experiencing a LOOP During its Normal Shutdown The licensee considered a DBLOCA in Unit 1, during which the cross-connect is supplying cooling water to Unit 1 from one of the two Unit 3 LPSW pumps, in conjunction with a LOOP in Unit 3, while the one remaining LPSW pump is unavailable due to being cross-connected. In Section 3.5 of the LAR Enclosure, the licensee stated the response to SNSB RAI No. 3 (ML22104A010) on the previous LAR (ML21245A210) is applicable and appropriate for the requested 360-hour Completion Time in this LAR.
On reviewing the response to SNSB-RAI No. 3, the NRC staff noted that Unit 3 pumps 3A and 3B each has a design flow rate of 15,000 gpm (total 30,000 gpm). This flow rate bounds the maximum required flow rate of 27,700 gpm for mitigation of DBLOCA in Unit 1 along with decay heat removal during normal shutdown of Unit 3 with a LOOP. An additional flow rate margin of 3,500 gpm will be available because some of the Unit 3 loads can be isolated at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after LOOP occurs.
The NRC staff finds the LPSW pumps 3A and 3B have adequate capacity for long-term decay heat removal to bring Unit 3 to Mode 5 in the presence of a LOOP while supplying the flow to Unit 1 through the cross-connect tie.
3.1.4 NRC Staff Technical Conclusions Based on the above, the NRC staff concludes:
For mitigation of a DBLOCA conjunction with a LOOP in Unit 1, an operator action time of 14.72 minutes to cross-connect the Unit 3 LPSW pump with Unit 1 and 2 LPSW system is acceptable.
The DBLOCA peak RB pressure and vapor temperature values are bounded by the AOR values shown in UFSAR Figures 6-36 and 6-37. The peak RB vapor temperature remains bounded by the AOR equipment qualification temperature profile. The peak RB wall temperature is bounded by the AOR RB structural design temperature.
The NPSH margin for the RBS pumps remains positive, and therefore pump performance is acceptable.
To bring Unit 3 to Mode 5 during a LOOP, the LPSW pumps 3A and 3B have adequate capacity for Unit 3 decay heat removal and supply flow to Unit 1 through the cross-connect tie.
3.1.5 NRC Staff Regulatory Conclusions Based on the above, the NRC staff concludes the following regarding the applicable regulatory requirements in Section 2:
10 CFR 50.46(b)(1) through (b)(5) criteria are satisfied, during a Unit 1 LOCA scenario.
UFSAR, Section 3.1.6, Criterion 6 - Reactor Core Design (Category A) is satisfied because in the Unit 1 LOCA scenario, 10 CFR 50.46(b)(1) through (b)(5) criteria are satisfied, and adequate Unit 3 decay heat removal capacity would be available in a concurrent LOOP.
UFSAR, Section 3.1.10, Criterion 10 - Containment (Category A), is satisfied because the Unit 1 RB integrity would be maintained during a LOCA, and the RB leakage would not exceed the AOR value because the peak RB pressure and vapor temperature are bounded by their AOR values.
UFSAR, Section 3.1.41, Criterion 41 - Engineered Safety Features Performance Capability (Category A), is satisfied because during a LOCA in Unit 1, ECCS and RB heat removal system would adequately perform even with loss of LPSW heat removal for 14.72 minutes and would still fulfill the required safety function in the presence of a single failure of an active component.
UFSAR, Section 3.1.49, Criteria 49 - Containment Design Basis (Category A), is satisfied because during a LOCA in Unit 1, RB heat removal system pumps would have adequate NPSH margin and therefore their performance would not be impacted for maintaining the RB pressure, vapor temperature, and leakage below their AOR values.
UFSAR, Section 3.1.52, Criteria 52 - Containment Heat Removal Systems (Category A), is satisfied because during a LOCA in Unit 1, the fan coolers and the RBS system would be available for RB heat removal to maintain its internal pressure and vapor temperature below AOR values.
Therefore, the NRC staff concludes that the proposed one time extension of Completion Time for TS 3.7.7, Condition A, Required Action A.1 from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> is acceptable.
3.2 Plant Systems Review The plant systems review focused on the Unit 3 LPSW cross-connect to the shared Units 1 and 2 LPSW header through LPSW-1095 valve. In its LAR, the licensee stated the use for this cross-connect is defense-in-depth and identified it as Compensatory Measure No. 1 to be used in the event the shared Units 1 and 2 C LPSW pump becomes inoperable during the temporarily extended completion time (360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br />). The NRC staff assessed the capability of the Unit 3 LPSW pumps to provide adequate flow for functions necessary to mitigate accidents along with the ability of LPSW safety functions to be maintained.
3.2.1 Licensees Evaluation The licensee states the following in the LAR:
The alternate suction source described in Section 3.1 above will be placed into service to meet the LCO of TS 3.7.7 while the CCW valves are replaced.
However, prior to placing the alternate suction source into service to meet the LCO, only the shared Units 1 and 2 C LPSW pump will be OPERABLE to support Unit 1 (see Figure 5 of Attachment 4). The required window to complete the tie-in and perform a functional test of LPSW-2 and the alternate suction source is projected to require 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br />, which exceeds the TS 3.7.7 Completion Time for Required Action A.1 of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Although operability of the single Units 1 and 2 C LPSW Pump can provide for the specified safety function of the system for Unit 1, ONS has the capability to procedurally cross connect the ONS
[Oconee Nuclear Station] Unit 3 LPSW pumps to the ONS Unit 1 and 2 LPSW header by opening valve LPSW-1095 should the C LPSW Pump become inoperable. This cross connect is the ACTION/EXPECTED RESPONSE for a loss of Unit 1 and 2 LPSW pumps in ONS procedure AP/1/A/1700/024, Loss of LPSW.... This cross connect capability provides defense-in-depth during the proposed 360-hour temporary Completion Time of Require[d] Action A.1 with Unit 2 defueled and required loads isolated.
As described in Oconee UFSAR Section 9.2.2, Cooling Water Systems, the LPSW system provides cooling water to RB cooling units, decay heat removal coolers, HPI pump motor bearing coolers, and motor-driven emergency feedwater pump motor air coolers. The RB cooling units provide containment heat removal and decay heat removal coolers provide cooling to LPI in RB emergency sump recirculation mode operation of low-pressure injection system following postulated LOCAs as discussed in UFSAR Chapter 6. Of the transients and accidents analyzed in Oconee UFSAR Chapter 15, steam generator tube rupture and small LOCA credit high pressure injection system for inventory control during the initial phase and steam line break accident credits motor-driven emergency feedwater system.
In Section 3.5 of the LAR Enclosure, the licensee stated the response to Containment and Plant Systems Branch (SCPB) RAI No. 1 (ML22104A010) on the previous LAR (ML21245A210) is applicable and appropriate. In this response, the licensee stated the following:
The hydraulic models for the LPSW system used to evaluate the LOCA/LOOP response are separate for the Unit 1/Unit 2 and Unit 3 systems. Within these analyses is an evaluation of various data for the effects of the two systems being cross connected that concludes there is no flow division away from the LOCA unit. The basic supposition is that all TS required pumps are available and a single failure (pump or electrical bus) occurs on the LOCA unit. Aside from flow rates, the analysis predicts the LPSW header pressure for the LPSW system associated with the LOCA unit. Periodic performance testing shows that the normal header pressure is roughly the same for both systems. When Engineered Safety Features Actuation System (ESFAS) actuation occurs with the postulated single failure that results in one LPSW pump failing to start, it is assumed that the LPSW system associated with the LOCA unit will experience a slight lowering in header pressure. Thus, the LPSW system associated with the non-LOCA unit would reasonably be expected to provide some flow to the LOCA unit. This flow is not credited in any analysis since the LPSW systems are modeled independently. Note that engineering judgment is applied here, as the analysis does not assume the Unit 1/Unit 2 A and B LPSW pumps are inoperable, along with a failure of the C LPSW pump.
The licensee acknowledges that the configuration for the proposed change with Unit 1 in TS 3.7.7, Condition A (one required LPSW pump inoperable), and assuming a single failure of the remaining shared Units 1 and 2 C LPSW pump, results in an unanalyzed scenario relative to mitigating a DBA or transient because under normal or accident conditions one LPSW pump per unit must operate to supply loads. However, the licensee states that there is reasonable assurance that starting the non-running Unit 3 LPSW pump, as part of the Loss of LPSW Abnormal Procedure would make up for most of the flow loss when the systems are cross connected because Unit 2 will be defueled in conjunction with the proposed change with necessary loads isolated. The licensee expects the flow rates near those in the existing analyses when two LPSW pumps are operating for two units (in this case Units 1 and 3) with LPSW demands.
In addition, a periodic test is conducted that places the LPSW cross connect in service with the shared Units 1 and 2 LPSW system supplying all three units (i.e., the Unit 3 LPSW pumps are turned off). The licensee indicated that the test is successfully performed with regularity and header pressure is maintained well above what accident analysis predicts would exist during an event and that although during the test the shared Units 1 and 2 LPSW system supplies Unit 3, it would be expected that the two Unit 3 LPSW pumps could supply any two units since the cross connect piping is a 24 line with no check valves in the line because the pressure drop in the cross connect piping would be the same regardless of the direction of flow. Therefore, based on engineering judgment, the licensee expects that two pumps on the Unit 3 LPSW system could carry the loads for any two units, when the third unit is defueled and appropriate loads isolated, as the case for the proposed change.
Furthermore, HPI pumps use LPSW as the normal means to cool the motors upper thrust bearing. The licensee indicated that LPSW is backed up by HPSW, which will be supplied from the Elevated Water Storage Tank (EWST) if the HPSW pumps are not running. The licensee confirmed that enough inventory exists in the EWST to support placing the LPSW cross-connect in service since the time to align is much shorter than the time before the EWST requires refilling.
The licensee states that LPSW also provides cooling water to the motor driven emergency feedwater pump motor air coolers and there is no backup; however, the licensee confirmed that a turbine driven emergency feedwater pump exists that is also backed up by HPSW and would be available until the LPSW cross connect is placed in service. The licensee states that additional emergency feedwater (EFW) system redundancy includes, (a) ability to cross connect EFW systems manually between all three units and (b) full capacity SSF ASW pump capable of feeding all three units SGs simultaneously.
3.2.2 NRC Staff Evaluation As provided in the applicable response to SCPB-RAI No. 1, the NRC staff finds that based on the licensees evaluation, the non-running Unit 3 LPSW pump being started as part of the Loss of LPSW Abnormal Procedure, the periodic tests, and the licensees engineering judgment; if a transient or an accident occurs and the shared Units 1 and 2 C LPSW pump fails, there is reasonable assurance that the cross-connect from Unit 3 LPSW pumps to Unit 1 would provide adequate flow for functions necessary to mitigate accidents. Additionally, the NRC staff finds that if the shared Units 1/2 C LPSW pump fails, there is reasonable assurance that both high pressure injection pumps and EFW pumps would continue to be cooled from EWST water until the LPSW cross connect is placed in service because of the availability of EWST, turbine driven emergency feedwater pump, and other redundancies. The NRC staff determined that the extended completion time to 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> does not change the defense in depth capabilities in place to provide adequate flow and that the LPSW safety function will be maintained during this period.
3.
2.3 NRC Staff Conclusion
The NRC staff concludes that if the shared Units 1/2 C LPSW pump becomes inoperable, the compensatory measure provided to procedurally cross-connect the Unit 3 LPSW pumps to the shared Units 1/2 LPSW header and cooling from EWST water until the cross connect is placed in service would enable Unit 3 LPSW pumps to provide sufficient flow for functions necessary to mitigate accidents for Unit 1 when Unit 2 is defueled. Therefore, the NRC staff determines that there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change. Therefore, in accordance with 10 CFR 50.40(a) and 10 CFR 50.57(a)(3),
the NRC staff concludes that the proposed one-time extension of Completion Time for TS 3.7.7, Condition A, Required Action A.1 from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> is acceptable.
3.3 Human Factors Review 3.4.1 Description of Personnel Actions The activities which are scheduled during the proposed extended LCO will include procedural actions which differ from the current TS and LCO. These proposed changes include work performed under this LCO. The submittal also includes a compensatory measure for defense-in-depth to cross-connect Unit 3 to the shared Units 1 and 2 LPSW system.
The submittal provided the following list of activities with the projected allotted times that are scheduled to take place during the proposed temporary 360-hour TS 3.7.7 Action statement:
Tag Out of LPSW Pumps A and B (enter TS 3.7.7 Action statement) and make inflatable line stop wet tap - 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Install Inflatable line stop and Red Tag - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Drain A and B LPSW Header section of pipe, sever header at new CCW-522 location, and remove LPSW 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Prepare and weld flanged spool to A and B LPSW Header for CCW-522, install new LPSW 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Install CCW-522 to new A and B LPSW header flange - 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Install new header section to CCW-522 and LPSW-2 with associated supports - 48 Hours Install CCW-518 valve to A and B LPSW header - 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Install 36-inch tee connecting alternate suction source to A and B LPSW header - 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Clear Tags, fill and vent A and B LPSW Header and alternate suction source line - 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> Test LPSW A and B Pumps and exit TS 3.7.7 Condition A - 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> The allotted times presented in the submittal are each rounded to the nearest six-hour shift.
These actions facilitate the replacement of three CCW valves (2CCW-41, CCW-72, CCW-73).
The proposed change described affected systems and the associated actions and indicated that the changes were limited to actions directly affecting the LPSW system.
Additionally, the submittal discussed the availability of the procedural action to cross connect Unit 3 LPSW pumps to the shared Units 1 and 2 LPSW system. The licensee stated that Oconee has the capability to cross connect the Unit 3 LPSW pumps to the shared Units 1 and 2 LPSW header. This connection is completed by opening valve LPSW-1095 should the shared Units 1 and 2 C LPSW pump become inoperable and is done in response to a loss of the shared Units 1 and 2 LPSW pumps in ONS procedures. The submittal stated that this cross connect capability provides defense-in-depth during the proposed extended LCO completion time while Unit 2 is defueled and required loads are isolated.
3.3.2 Risk Assessment to Determine Human Factors Level of Review The NRC staff reviewed the proposed changes and considered the human actions to tie-in and test the alternate suction source and, if needed, to cross-connect Unit 3 LPSW to the shared Units 1 and 2 LPSW header. The proposed human actions occur during the refueling of Unit 2.
The NRC staff reviewed Table A.2, Generic PWR Human Actions That Are Risk Important, in Appendix A of NUREG-1764 and verified that no actions from that table are included in the submittal.
The NRC staff considered available relevant risk information in the submittal. In a previous LAR (ML21245A210) Oconee provided an evaluation of the risk associated with extending the completion time for one required LPSW pump inoperable (TS 3.7.7, Action A) during an Oconee Unit 2 refueling outage (O2R31), from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br /> on a temporary, one-time basis.
This evaluation included probabilistic risk assessment (PRA) models for internal events and fire.
Core Damage Frequency was calculated in these two PRA models and the results and the deltas between the proposed change and the baseline were determined. The methodology and analysis approach remain the same with the only adjustment from the previous LAR to the current LAR being the previously approved Completion Time of 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br /> changed to the currently requested Completion Time of 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br />. The LAR states the following:
For both internal events and fire, Core Damage Frequency (CDF) results were obtained and deltas between the proposed configuration and baseline CDF values were then determined. From these delta values, a total Incremental Conditional Core Damage Probability (ICCDP) was calculated for the 288-hour period beyond the 72-hour allowed outage time that is being requested to perform the work.
The Oconee internal events model resulted in an internal events ICCDP of 6.9E-8.
The NRC staff confirmed the licensees results by running a Standardized Plant Analysis Risk (SPAR) model for the proposed configuration. The NRC staff model results determined a change in risk of 2E-9 which confirms the licensees evaluation that the proposed extension to a 360-hour Completion time is not risk significant. Therefore, the NRC staff determined this approach to be acceptable.
Based on the review of Table A.2, in NUREG-1764 and range in which the ICCDP fall according to NUREG-1764, the NRC staff determined that a Level 3 review, the least stringent, human factors review was appropriate. Therefore, the NRC staff applied the criteria for a Level 3 review.
3.3.3 NRC Staff Evaluation The licensee described the operator actions associated with the extended LCO. The NRC staff reviewed the list of proposed scheduled activities and verified that these actions are in the low-risk category by assessing them against the applicable tables in NUREG-1764. The NRC staff also considered the defense-in-depth actions (cross connecting the units) and based on the licensees results and independent NRC staff results, the NRC staff finds the proposed 360-hour Completion time presents a minimal change in risk.
3.
3.4 NRC Staff Conclusion
The NRC staff confirmed that the operator actions described in the amendment are low-risk, and, therefore, the staff conducted a Level 3 review which focused on ensuring that defense-in-depth is not degraded. The NRC staff finds that the existing operator actions contained in the current licensing basis continue to provide a reasonable means of defense-in-depth during the outage period. The NRC staff concludes there is not a significant increase in risk as a result of approving this amendment. Based on the above, the NRC staff concludes the requested change to TS 3.7.7 acceptable with regard to human factors engineering.
3.4 Technical Specification Review Proposed Changes to TS LCO 3.7.7, Low Pressure Service Water (LPSW) System Condition A of TS LCO 3.7.7 addresses one required LPSW pump inoperable. The required action for Condition A is to restore required LPSW pump to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The licensee proposed to revise the Completion Time Note to extend the completion time to 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> during a Unit 2 refueling outage for the tie-in and testing of an alternate suction source to the shared Units 1 and 2 A and B LPSW pumps. The additional time provided by the note is requested on a one-time temporary basis and is limited to when the Unit 2 is defueled and the appropriate LPSW loads are secured. It is also contingent on implementation of the compensatory measures described in the enclosure to the licensees submittal and expires December 31, 2027.
The NRC staff has reviewed the licensees proposed compensatory measures. The proposed compensatory measures include ensuring the capability to cross-connect the Unit 3 LPSW system should a loss of all LPSW on Unit 1 occur and protecting essential equipment to prevent a loss of all LPSW conditions and mitigate it should one transpire. The proposed compensatory measures ensure adequate protection of public health and safety while the licensee is conducting the modification by providing a heat sink for the removal of process and operating heat from safety related components during a transient or accident. The NRC staff finds that the compensatory measures are sufficient to reduce the risk of unnecessary plant transients, protect systems needed for accident mitigation, and raise operator awareness of necessary recovery actions should a loss of all LPSW on Unit 1 were to occur.
As discussed in the LAR and as reflected in the TS Bases for TS LCO 3.7.7, the licensee explained that the added note expires at 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> or upon completion of the tie-in and satisfactory testing of an alternate suction source to the shared Units 1 and 2 A and B LPSW pumps, whichever comes first. Therefore, at the completion of the final tie-in and testing of the alternate suction source, TS LCO 3.7.7 required action A.1 will be exited, as the LCO for Unit 1 would be considered met at that time. This will reduce the time the licensee is in Condition A to the minimum necessary to complete the modification, thus minimizing the overall risk associated with the modification.
The NRC staff concludes the proposed change is acceptable because the licensee has appropriate compensatory measures in place to prevent and mitigate any unintended consequences resulting from the extended completion time. In addition, with the proposed one-time temporary note, TS LCO 3.7.7 Required Action A.1, continues to meet 10 CFR 50.36(c)(2) as the compensatory measures provide appropriate remedial action until the condition in the LCO can be met.
3.5 Risk Insights In the LAR, the licensee requested a one-time change to extend the Completion Time for one required inoperable LPSW Pump on a temporary basis. Specifically, the licensee requested to modify the Completion Time for TS 3.7.7, Condition A, Required Action A.1, to 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> to facilitate the tie-in and testing of an alternate suction source for the shared Unit 1 and Unit 2 A and B LPSW pumps during the Unit 2 refueling outage. Although the licensee provided risk insights related to the proposed TS CT change, the NRC staff determined that the risk-insights could not be considered in the decision making process because the licensee submitted its request based on deterministic criteria and did not submit all the information normally required for a risk-informed LAR as described in Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.
The NRC staff has previously reviewed the technical adequacy of the Oconee PRA models for Oconee Nuclear Station, Units 1, 2, and 3 - Re: Authorization Of Alternative To Use RR-22-0174, Risk-Informed Categorization And Treatment For Repair/Replacement Activities In Class 2 and 3 SystemsSection XI, Division 1 December 13, 2023 (ML23262A967); Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding Adoption of Technical Specification Task Force (TSTF) -425, Revision 3, Relocate Surveillance Frequencies to Licensee Control Risk Informed Technical Specification Task Force (RITSTF) Initiative 5b, dated March 21, 2011 (ML110470446); and, License Amendment Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding Transition to a Risk-Informed, Performance-Based Fire Protection Program in Accordance with 10 CFR 50.48(c), dated December 29, 2010 (ML103630612).
3.5.1 Licensees Evaluation In Section 3.3 of the LAR enclosure, the licensee described risk insights developed to support the proposed one time Completion Time extension. To obtain these risk insights for the operating Unit 1 in the proposed configuration, the licensee modified their fire and internal events PRA models to reflect the new configuration. The licensee calculated CDF for both fire and internal events, with delta values between the proposed configuration and the baseline CDF used to determine the ICCDP for the requested extension. The licensee concluded that the proposed change would not affect the design and licensing basis mitigation function of the LPSW pump system and that the system would remain capable of performing its safety function during the 360-hour temporary Completion Time, just as it does during the existing 72-hour Completion Time.
In Section 3.3 of the LAR enclosure, the licensee stated that the total risk increase associated with extending the Completion Time from the normal 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 360 hours0.00417 days <br />0.1 hours <br />5.952381e-4 weeks <br />1.3698e-4 months <br /> on a one-time temporary basis was calculated to be an ICCDP of 1.68E-7.
3.5.2 NRC Staff Evaluation This was not a risk-informed application, and the PRA models used to derive risk insights were not reviewed by the NRC staff to determine their technical acceptability as a basis to support this application. As a result, the staff did not rely on the numerical results provided by the licensee. However, the staff noted that the licensees PRA was evaluated for a previous license amendment request and was found acceptable. The staff considered the licensee-provided risk insights to aid in the deterministic review of the proposed change. The staff performed an independent assessment using the SPAR model for Oconee to evaluate the risk contribution.
The staff noted that the licensee-provided ICCDP and the staffs independent assessment of ICCDP were within the Tier 1 acceptance guidelines for one-time technical specification changes, as described in RG 1.177, Rev. 2, Plant-Specific Risk-Informed Decisionmaking:
Technical Specifications.
3.
5.3 NRC Staff Conclusion
The results of the NRC staffs independent assessment were consistent with the results reported by the licensee and below the quantitative change in risk acceptance in RG 1.177, Rev. 2. Because this is not a risk-informed LAR, the internal events and fire PRA models used by the licensee to derive risk insights were not reviewed by NRC staff to determine their technical acceptability to support this SE. However, based on the licensees risk insights, as independently verified by the NRC staff using the SPAR model, the NRC staff concludes that the risk insights support the deterministic review findings made in this SE.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the South Carolina State official was notified of the proposed issuance of the amendments on March 28, 2025. On April 15, 2025, the State official confirmed that the State of South Carolina had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change the requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration published in the Federal Register on October 1, 2024 (89 FR 79968), 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 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 Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
A. Russell, NRR A. Sallman, NRR B. Lee, NRR D. Ju, NRR N. Iqbal, NRR R. Atienza, NRR Y. Wong, NRR K. Martin, NRR Date of Issuance: April 24, 2025
ML25087A128 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/STSB/BC NAME JMinzerBryant KZeleznock SMehta DATE 3/28/2025 4/2/2025 3/5/2025 OFFICE NRR/DSS/SNSB/BC NRR/DEX/EMIB/BC NRR/DSS/SCPB/BC NAME DMurdock SBailey MValentin DATE 2/27/2025 4/2/2025 4/2/2025 OFFICE NRR/DRO/IOLB/BC NRR/DRA/APLB/BC OGC - NLO NAME JAnderson EDavidson SGellen DATE 4/2/2025 3/27/2025 4/18/2025 OFFICE NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME MMarkley JMinzerBryant DATE 4/24/2025 4/24/2025