ML24099A219

Issuance of Amendment Nos. 245 and 247 Revision to TSs to Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b

ML24099A219
Person / Time
Site:	Diablo Canyon
Issue date:	05/29/2024
From:	Samson Lee NRC/NRR/DORL/LPL4
To:	Gerfen P Pacific Gas & Electric Co
Lee S, NRR/DORL/LPL4
References
EPID: L-2023-LLA-0100
Download: ML24099A219 (1)

Text

May 29, 2024

Ms. Paula Gerfen Senior Vice President, Generation and Chief Nuclear Officer Pacific Gas and Electric Company Diablo Canyon Power Plant P.O. Box 56, Mail Code 104/6 Avila Beach, CA 93424

SUBJECT:

DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 245 AND 247 RE: REVISION TO TECHNICAL SPECIFICATIONS TO ADOPT TSTF-505, REVISION 2, PROVIDE RISK-INFORMED EXTENDED COMPLETION TIMES - RITSTF INITIATIVE 4b (EPID L-2023-LLA-0100)

Dear Paula Gerfen:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 245 to Facility Operating License No. DPR-80 and Amendment No. 247 to Facility Operating License No. DPR-82 for the Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Diablo Canyon), respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated July 13, 2023, as supplemented by letter dated January 15, 2024.

The amendments revise the Diablo Canyon TSs to permit the use of risk-informed completion times for actions to be taken when limiting conditions for operation are not met. The changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF [Risk-Informed TSTF]

Initiative 4b, dated July 2, 2018. The NRC st aff issued a final model safety evaluation approving TSTF-505, Revision 2 on November 21, 2018.

P. Gerfen

A copy of the related safety evaluation is enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Samson S. Lee, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket Nos. 50-275 and 50-323

Enclosures:

1. Amendment No. 245 to DPR-80

2. Amendment No. 247 to DPR-82

3. Safety Evaluation

cc: Listserv

PACIFIC GAS AND ELECTRIC COMPANY

DOCKET NO. 50-275

DIABLO CANYON NUCLEAR POWER PLANT, UNIT 1

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 245 License No. DPR-80

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Pacific Gas and Electric Company (the licensee), dated July 13, 2023, as supplemented by letter dated January 15, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations;

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows:

(2) Technical Specifications

The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 245 are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3. This license amendment is effective as of its date of issuance and shall be implemented within 365 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Facility Operating License No. DPR-80 and the Technical Specifications

Date of Issuance: May 29, 2024 PACIFIC GAS AND ELECTRIC COMPANY

DOCKET NO. 50-323

DIABLO CANYON NUCLEAR POWER PLANT, UNIT 2

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 247 License No. DPR-82

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Pacific Gas and Electric Company (the licensee), dated July 13, 2023, as supplemented by letter dated January 15, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations;

D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 2

- 2 -

2. Accordingly, the license is amended by changes to the Technical Specif ications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to read as follows:

(2) Technical Specifications (SSER 32, Section 8)* and Environmental Protection Plan

The Technical Specifications contained in Appendix A and th e Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 247, are hereby incorporat ed in the license. Pacific Gas & Electric Company shall operate th e facility in accordance with the Technical Specifications and t he Environmental Protection Plan, except where otherwise stated in specific license conditions.

3. This license amendment is effective as of its date of issuance and shall be implemented within 365 days of the date of issuan ce.

FOR THE NUCLEAR REGULATORY COMMISSION

Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Facility Operating License No. DPR-82 and the Technical Specifications

Date of Issuance: May 29, 2024 ATTACHMENT TO LICENSE AMENDMENT NO. 245

TO FACILITY OPERATING LICENSE NO. DPR-80

AND LICENSE AMENDMENT NO. 247 TO FACILITY OPERATING LICENSE NO. DPR-82

DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2

DOCKET NOS. 50-275 AND 50-323

Replace the following pages of Facility Operating License Nos. DPR-80 and DPR-82, 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.

Facility Operating License No. DPR-80

REMOVE INSERT Facility Operating License No. DPR-82

REMOVE INSERT

Technical Specifications

REMOVE INSERT 1.3-10 1.3-10

--- 1.3-11

--- 1.3-12 3.4-16 3.4-16 3.4-19 3.4-19 3.4-20 3.4-20 3.4-21 3.4-21 3.5-3 3.5-3 3.6-4 3.6-4 3.6-5 3.6-5 3.6-6 3.6-6 3.6-7 3.6-7 3.6-13 3.6-13 3.6-14 3.6-14 3.7-4 3.7-4 3.7-8 3.7-8 3.7-10 3.7-10 3.7-11 3.7-11 3.7-11a ---

3.7-14 3.7-14 3.7-15 3.7-15

Technical Specifications (continued)

REMOVE INSERT 3.7-16 3.7-16 3.8-1 3.8-1 3.8-2 3.8-2 3.8-3 3.8-3 3.8-18 3.8-18 3.8-18a 3.8-18a 3.8-26 3.8-26 3.8-29 3.8-29 5.0-17a 5.0-17a

--- 5.0-17b

- 3 -

(4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to rece ive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

(5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This License shall be deemed to contain and is subject to the conditions specified in the Commissions 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

The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in e xcess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein.

(2) Technical Specifications

The Technical Specifications contained in Appendix A and th e Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 245 are hereby incorporated in the license.

Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

(3) Initial Test Program

The Pacific Gas and Electric Company shall conduct the post-fuel-loading initial test program (set forth in Section 14 of Pacific Gas and Electric Companys Final Safety Analysis Report, as amended), without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:

a. Elimination of any test identified in Section 14 of PG&Es Final Safety Analysis Report as amended as being essential;

Amendment No. 245

- 3 -

(4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to rece ive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and

(5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This License shall be deemed to contain and is subject to the conditions specified in the Commissions 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

The Pacific Gas and Electric Company is authorized to operate the facility at reactor core power levels not in e xcess of 3411 megawatts thermal (100% rated power) in accordance with the conditions specified herein.

(2) Technical Specifications (SSER 32, Section 8)* and Environmental Protection Plan

The Technical Specifications contained in Appendix A and th e Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 247, are hereby incorporated in the license.

Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

(3) Initial Test Program (SSER 31, Section 4.4.1)

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.

• The 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.

Amendment No. 247 Completion Times 1.3

1.3 Completion Times

EXAMPLES EXAMPLE 1.3-7 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One A.1 Verify affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> subsystem subsystem AND inoperable. isolated.

Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter AND A.2 Restore 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> subsystem to OPERABLE status.

B. Required B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and AND associated Completion B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Time not met.

Required Action A.1 has two Completion Times. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time begins at the time the Condition is entered and each "Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter" interval begins upon performance of Required Action A.1.

If after Condition A is entered, Required Action A.1 is not met within either the initial 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or any subsequent 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval from the previous performance (plus the extension allowed by SR 3.0.2),

Condition B is entered. The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired.

(continued)

DIABLO CANYON - UNITS 1 & 2 1.3-10 Unit 1 - Amendment No. 135, 245 Rev 13 Page 21 of 27 Unit 2 - Amendment No. 135, 247 Tab_1!0u3r13.DOC 0703.0855 Completion Times 1.3

1.3 Completion Times

EXAMPLES EXAMPLE 1.3-8 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One A.1 Restore 7 days subsystem subsystem to OR inoperable. OPERABLE status. In accordance with the RICT Program

B. Required B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and AND associated Completion B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Time not met.

When a subsystem is declared inoperable, Condition A is entered. The 7 day Completion Time may be applied as discussed in Example 1.3 -2.

However, the licensee may elect to apply the Risk Informed Completion Time (RICT) Program which permits calculation of a RICT that may be used to complete the Required Action beyond the 7 day Completion Time.

The RICT cannot exceed 30 days. After the 7 day Completion Time has expired, the subsystem must be restored to OPERABLE status within the RICT or Condition B must also be entered.

The RICT requires recalculation of the RICT to reflect changing plant conditions. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

If the 7 day Completion Time clock of Condition A has expired and subsequent changes in plant condition result in exiting the applicability of the RICT Program without restoring the inoperable subsystem to OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start.

(continued)

DIABLO CANYON - UNITS 1 & 2 1.3-11 Unit 1 - Amendment No. 245 Rev 13 Page 22 of 27 Unit 2 - Amendment No. 247 Tab_1!0u3r13.DOC 0703.0855 Completion Times 1.3

1.3 Completion Times

EXAMPLES EXAMPLE 1.3-8 (continued)

If the RICT expires or is recalculated to be less than the elapsed time since the Condition was entered and the inoperable subsystem has not been restored to OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable subsystems are restored to OPERABLE status after Condition B is entered, Condition A is exited, and therefore, the Required Actions of Condition B may be terminated.

IMMEDIATE When "Immediately" is used as a Completion Time, the Required Action COMPLETION should be pursued without delay and in a controlled manner.

TIME

DIABLO CANYON - UNITS 1 & 2 1.3-12 Unit 1 - Amendment No. 245 Rev 13 Page 23 of 27 Unit 2 - Amendment No. 247 Tab_1!0u3r13.DOC 0703.0855 Pressurizer 3.4.9

3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.9 Pressurizer

LCO 3.4.9 The pressurizer shall be OPERABLE with:

a. Pressurizer water level 90%; and

b. Two groups of pressurizer heaters OPERABLE with the capacity of each group 150 kW and capable of being powered from an emergency power supply.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer water level not A.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> within limit. AND

A.2 Fully insert all rods. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND A.3 Place Rod Control 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> System in a condition incapable of rod withdrawal.

AND A.4 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B. One required group of B.1 Restore required group 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> pressurizer heaters of pressurizer heaters to OR inoperable. OPERABLE status.

In accordance with the RICT Program C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion AND Time of Condition B not met.

C.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

DIABLO CANYON - UNITS 1 & 2 3.4-16 Unit 1 - Amendment No. 135, 245 Rev 17 Page 17 of 40 Unit 2 - Amendment No. 135, 247 Tab_3!4u3r17.DOC 0605.1320 Pressurizer PORVs 3.4.11

3.4 REACTOR COOLANT SYSTEM (RCS)

3.4.11 Pressurizer Power Operated Relief Valves (PORVs)

LCO 3.4.11 Each PORV and associated block valve shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

---

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each PORV.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more PORVs A.1 Close and maintain 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable solely due to power to associated excessive seat leakage. block valve.

B. One PORV inoperable for B.1 Close associated block 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> reasons other than valve.

excessive seat leakage. AND

B.2 Remove power from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> associated block valve.

AND B.3 Restore the Class I 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> PORV to OPERABLE OR status.

In accordance with the RICT Program

C. One block valve inoperable. ----------------NOTE------------------- 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Required Actions do not apply when block valve is inoperable solely as result of complying with Required Actions B.2 or E.3.

C.1 Place associated PORV in manual control.

AND (continued)

DIABLO CANYON - UNITS 1 & 2 3.4-19 Unit 1 - Amendment No. 135,169,171, 245 Rev 17 Page 20 of 40 Unit 2 - Amendment No. 135,170,172, 247 Tab_3!4u3r17.DOC 0629.0950 Pressurizer PORVs 3.4.11

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

C. (continued) C.2 If the block valve is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> associated with a Class I OR PORV:

Restore block valve to In accordance with the OPERABLE status. RICT Program

OR C.3 If the block valve is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> associated with the non-Class I PORV:

Close the block valve and remove its power.

D. Required Action and D.1 Initiate action to restore Immediately associated Completion Class I PORV and/or Time of Condition A, B, or C associated block valves(s) not met. to OPERABLE status.

AND D.2 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND D.3 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> E. Two Class I PORVs E.1 Initiate action to restore Immediately inoperable for reasons other Class I PORVs to than excessive seat OPERABLE status.

leakage. AND E.2 Close associated block 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> valves.

AND E.3 Remove power from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> associated block valves.

AND E.4 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND E.5 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (continued)

DIABLO CANYON - UNITS 1 & 2 3.4-20 Unit 1 - Amendment No. 135, 171, 245 Rev 17 Page 21 of 40 Unit 2 - Amendment No. 135, 172, 247 Tab_3!4u3r17.DOC 0629.0950 Pressurizer PORVs 3.4.11

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME F. More than one block valve ------------------NOTE------------------

inoperable. Required Actions do not apply when block valve is inoperable solely as result of complying with Required Actions B.2 or E.3.

---

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> F.1 Place associated PORVs in manual control.

AND F.2 Restore one block valve 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for a Class I PORV to OPERABLE status.

AND F.3 Restore remaining block 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> valve for a Class I PORV OR to OPERABLE status.

In accordance with the RICT Program OR F.4 If the remaining block 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> valve is associated with the non-Class I PORV, close the block valve and remove its power.

G. Required Action and G.1 Initiate action to restore Immediately associated Completion block valve(s) to Time of Condition F not OPERABLE status.

met. AND

G.2 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND G.3 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

DIABLO CANYON - UNITS 1 & 2 3.4-21 Unit 1 - Amendment No. 135, 245 Rev 17 Page 22 of 40 Unit 2 - Amendment No. 135, 247 Tab_3!4u3r17.docx 1213.1547 ECCS - Operating 3.5.2

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS - Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

---

NOTE-------------------------------------------------------------

In MODE 3, both safety injection (SI) pump flow paths may be isolated by closing the isolation valve(s) for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to perform pressure isolation valve testing per SR 3.4.14.1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One or more trains A.1 Restore train(s) to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status. OR AND In accordance with the At least 100% of the ECCS RICT Program flow equivalent to a single OPERABLE ECCS train available.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion AND Time not met.

B.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

DIABLO CANYON - UNITS 1 & 2 3.5-3 Unit 1 - Amendment No. 135, 159, 202, 245 Rev 12 Page 3 of 9 Unit 2 - Amendment No. 135, 146, 160, 203, 247 Tab_3!5u3r12.docx 0513.1100 Containment Air Locks 3.6.2

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

C. One or more containment C.1 Initiate action to evaluate Immediately air locks inoperable for overall containment reasons other than leakage rate per Condition A or B. LCO 3.6.1.

AND C.2 Verify a door is closed in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the affected air lock.

AND C.3 Restore air lock to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status. OR

In accordance with the RICT Program

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion AND Time not met.

D.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

SR 3.6.2.1 -----------------------------NOTES--------------------------

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

2. Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1 Perform required air lock leakage rate testing in In accordance with accordance with the Containment Leakage Rate the Containment Testing Program. Leakage Rate Testing Program

SR 3.6.2.2 Verify only one door in the air lock can be opened In accordance with the at a time. Surveillance Frequency Control Program

DIABLO CANYON - UNITS 1 & 2 3.6-4 Unit 1 - Amendment No. 135, 200, 245 Rev 10 Page 4 of 20 Unit 2 - Amendment No. 135, 201, 247 Tab_3!6u3r10.DOC 0605.1329 Containment Isolation Valves 3.6.3

3.6 CONTAINMENT SYSTEMS

3.6.3 Containment Isolation Valves

LCO 3.6.3 Each containment isolation valve shall be OPERABLE.

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

ACTIONS

---

NOTES----------------------------------------------------------------

1. Penetration flow path(s) except for 48-inch purge valve flow paths, may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by containment isolation valves.

4. Enter applicable Conditions and Required Actions of LCO 3.6.1, "Containment," when isolation valve leakage results in exceeding the overall containment leakage rate acceptance criteria.

CONDITION REQUIRED ACTION COMPLETION TIME

A. --------------NOTE--------------A.1 Isolate the affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Only applicable to penetration flow path by OR penetration flow paths with use of at least one two containment isolation closed and de-activated In accordance with the valves. automatic valve, closed RICT Program

---

manual valve, blind One or more penetration flange, or check valve flow paths with one with flow through the containment isolation valve valve secured.

inoperable except for a AND containment purge supply and exhaust valve or pressure/vacuum relief valve leakage not within limit.

(continued)

DIABLO CANYON - UNITS 1 & 2 3.6-5 Unit 1 - Amendment No. 135, 230, 245 Rev 10 Page 5 of 20 Unit 2 - Amendment No. 135, 232, 247 Tab_3!6u3r10.DOC 0605.1329 Containment Isolation Valves 3.6.3

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. (continued) A.2 -----------NOTES-----------

1. Isolation devices in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or otherwise secured may be verified by administrative means.

Verify the affected Once per 31 days penetration flow path is following isolation for isolated. isolation devices outside containment AND Prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days for isolation devices inside containment

B. --------------NOTE--------------B.1 Isolate the affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Only applicable to penetration flow path by penetration flow paths with use of at least one two containment isolation closed and de-activated valves. automatic valve, closed

---

manual valve, or blind One or more penetration flange.

flow paths with two containment isolation valves inoperable except for a containment purge supply and exhaust valve or pressure/vacuum relief valve leakage not within limit.

(continued)

DIABLO CANYON - UNITS 1 & 2 3.6-6 Unit 1 - Amendment No. 135, 245 Rev 10 Page 6 of 20 Unit 2 - Amendment No. 135, 247 Tab_3!6u3r10.DOC 0605.1329 Containment Isolation Valves 3.6.3

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

C. --------------NOTE--------------C.1 Isolate the affected 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Only applicable to penetration flow path by OR penetration flow paths with use of at least one only one containment closed and de-activated In accordance with the isolation valve and a closed automatic valve, closed RICT Program system. manual valve, or blind

---

flange.

One or more penetration AND flow paths with one C.2 ------------NOTES----------

containment isolation valve inoperable. 1. Isolation devices in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or otherwise secured may be verified by administrative means.

Verify the affected Once per 31 days penetration flow path is following isolation isolated.

D. One or more penetration D.1 Isolate the affected 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> flow paths with one or more penetration flow path by containment purge supply use of at least one and exhaust and closed and de-activated vacuum/pressure relief automatic valve, closed valves not within purge manual valve, or blind valve leakage limits. flange.

AND (continued)

DIABLO CANYON - UNITS 1 & 2 3.6-7 Unit 1 - Amendment No. 135, 245 Rev 10 Page 7 of 20 Unit 2 - Amendment No. 135, 247 Tab_3!6u3r10.docx 0513.1118 Containment Spray and Cooling Systems 3.6.6

3.6 CONTAINMENT SYSTEMS

3.6.6 Containment Spray and Cooling Systems

LCO 3.6.6 The containment fan cooling unit (CFCU) system and two containment spray trains shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One containment spray A.1 Restore containment 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> train inoperable. spray train to OR OPERABLE status.

In accordance with the RICT Program B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion AND Time of Condition A not met. B.2. -----------NOTE-----------

LCO 3.0.4.a is not applicable when entering MODE 4.

Be in MODE 4. 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> C. One required CFCU system C.1 Restore required CFCU 7 days inoperable such that a system to OPERABLE OR minimum of two CFCUs status.

remain OPERABLE. In accordance with the RICT Program

(continued)

DIABLO CANYON - UNITS 1 & 2 3.6-13 Unit 1 - Amendment No. 135,202,215, 219, 245 Rev 10 Page 13 of 20 Unit 2 - Amendment No. 135,173,203,217, 221, 247 Tab_3!6u3r10.docx 0513.1118 Containment Spray and Cooling Systems 3.6.6

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

D. One required containment D.1 Restore one required 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> spray train inoperable and containment spray system OR one required CFCU system to OPERABLE status, inoperable such that a In accordance with minimum of two CFCUs the RICT Program remain OPERABLE. OR

D.2 Restore one CFCU 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> system to OPERABLE OR status such that four CFCUs or three CFCUs, In accordance with each supplied by a the RICT Program different vital bus, are OPERABLE.

E. Required Action and E.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition C or D AND not met.

E.2 -----------NOTE-----------

LCO 3.0.4.a is not applicable when entering MODE 4.

Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

F. Two containment spray F.1 Enter LCO 3.0.3. Immediately trains inoperable.

OR One containment spray train inoperable and two CFCU systems inoperable such that one or less CFCUs remain OPERABLE.

OR One or less CFCUs OPERABLE.

DIABLO CANYON - UNITS 1 & 2 3.6-14 Unit 1 - Amendment No. 135, 219, 245 Rev 10 Page 14 of 20 Unit 2 - Amendment No. 135, 173, 221, 247 Tab_3!6u3r10.DOC 0626.1341 MSIVs 3.7.2

3.7 PLANT SYSTEMS

3.7.2 Main Steam Isolation Valves (MSIVs)

LCO 3.7.2 Four MSIVs shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3 except when all MSIVs are closed and de-activated.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One MSIV inoperable in A.1 Restore MSIV to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> MODE 1. OPERABLE status. OR

In accordance with the RICT Program B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not met.

C. ------------NOTE----------------C.1 Close MSIV. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Separate Condition entry is AND allowed for each MSIV.

One or more MSIVs C.2 Verify MSIV is closed. Once per 7 days inoperable in MODE 2 or 3.

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time AND of Condition C not met.

D.2 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

DIABLO CANYON - UNITS 1 & 2 3.7-4 Unit 1 - Amendment No. 135, 245 Rev 20 Page 4 of 37 Unit 2 - Amendment No. 135. 247 Tab_3!7u3r20.DOC 0606.1547 ADVs 3.7.4

3.7 PLANT SYSTEMS

3.7.4 10% Atmospheric Dump Valves (ADVs)

LCO 3.7.4 Four ADV lines shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required ADV line A.1 Restore required ADV line 7 days inoperable. to OPERABLE status OR

In accordance with the RICT Program B. Two required ADV lines B.1 Restore at least one ADV 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. line to OPERABLE status.

C. Three or more required C.1 Restore at least two ADV 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ADV lines inoperable. lines to OPERABLE status.

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion AND Time not met.

D.2 Be in MODE 4 without 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> reliance upon steam generator for heat removal.

DIABLO CANYON - UNITS 1 & 2 3.7-8 Unit 1 - Amendment No. 135,169, 245 Rev 20 Page 9 of 37 Unit 2 - Amendment No. 135,170, 247 Tab_3!7u3r20.DOC 0606.1547 AFW System 3.7.5

3.7 PLANT SYSTEMS

3.7.5 Auxiliary Feedwater (AFW) System

LCO 3.7.5 Three AFW trains shall be OPERABLE.

---

NOTE---------------------------------------------------

Only one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

ACTIONS

---

NOTE------------------------------------------------------

LCO 3.0.4b is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME

A. Turbine driven AFW train A.1 Restore affected 7 days inoperable due to one equipment to OR inoperable steam supply OPERABLE status.

OR In accordance with the RICT Program

---

NOTE---------------

Only applicable if MODE 2 has not been entered following refueling.

Turbine driven AFW pump inoperable in MODE 3 following refueling.

B. One AFW train inoperable B.1 Restore AFW train to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in MODE 1, 2 or 3 for OPERABLE status. OR reasons other than Condition A. In accordance with the RICT Program (continued)

DIABLO CANYON - UNITS 1 & 2 3.7-10 Unit 1 - Amendment No. 135,169, 215, 236, 245 Rev 20 Page 11 of 37 Unit 2 - Amendment No. 135,170, 217, 238, 247 Tab_3!7u3r20.DOC 0626.1348 AFW System 3.7.5

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. -----------NOTE---------------C.1 Restore the steam 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Only applicable when the supply to the turbine OR remaining OPERABLE driven train to motor driven AFW train OPERABLE status. In accordance with the provides feedwater to the OR RICT Program steam generator with the inoperable steam supply. C.2 Restore the motor driven 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />

---

AFW train to OR Turbine driven AFW train OPERABLE status.

inoperable due to one In accordance with the inoperable steam supply. RICT Program AND One motor driven AFW train inoperable.

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion AND Time for Condition A, B, or 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> C not met. D.2 Be in MODE 4.

OR Two AFW trains inoperable in MODE 1, 2 or 3 for reasons other than Condition C.

E. Three AFW trains E.1 -----------NOTE---------------

inoperable in MODE 1, 2, or LCO 3.0.3 and all other

3. LCO Required Actions requiring MODE changes are suspended until one AFW train is restored to OPERABLE status.

Initiate action to restore Immediately one AFW train to OPERABLE status F. Required AFW train F.1 Initiate action to restore Immediately inoperable in MODE 4. AFW train to OPERABLE status.

-I

DIABLO CANYON - UNITS 1 & 2 3.7-11 Unit 1 - Amendment No. 135,169, 215, 236, 245 Rev 20 Page 12 of 37 Unit 2 - Amendment No. 135,170, 217, 238, 247 Tab_3!7u3r20.DOC 0626.1348 CCW System 3.7.7

3.7 PLANT SYSTEMS 3.7.7 Vital Component Cooling Water (CCW) System LCO 3.7.7 Two vital CCW loops shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One vital CCW loop A.1 ------------NOTE--------------

inoperable. Enter applicable Conditions and Required Actions of LCO 3.4.6, RCS Loops - MODE 4, for residual heat removal loops made inoperable by CCW.

Restore vital CCW loop to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status. OR

In accordance with the RICT Program B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion AND Time of Condition A not met. B.2 -----------NOTE-----------

LCO 3.0.4.a is not applicable when entering MODE 4.

Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

DIABLO CANYON - UNITS 1 & 2 3.7-14 Unit 1 - Amendment No. 135, 200, 219, 245 Rev 20 Page 16 of 37 Unit 2 - Amendment No. 135, 201, 221, 247 Tab_3!7u3r20.DOC 0606.1547 CCW System 3.7.7

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

SR 3.7.7.1 -----------------------------NOTE-------------------------------

Isolation of CCW flow to individual components does not render the CCW System inoperable

---

In accordance with Verify each CCW manual, power operated, and the Surveillance automatic valve in the flow path servicing safety Frequency Control related equipment, that is not locked, sealed, or Program otherwise secured in position, is in the correct position.

SR 3.7.7.2 Verify each CCW automatic valve in the flow path that In accordance with is not locked, sealed, or otherwise secured in position, the Surveillance actuates to the correct position on an actual or Frequency Control simulated actuation signal. Program SR 3.7.7.3 Verify each CCW pump starts automatically on an In accordance with actual or simulated actuation signal. the Surveillance Frequency Control Program

DIABLO CANYON - UNITS 1 & 2 3.7-15 Unit 1 - Amendment No. 135, 200, 245 Rev 20 Page 17 of 37 Unit 2 - Amendment No. 135, 201, 247 Tab_3!7u3r20.DOC 0606.1547

• • • UNCONTROLLED DOCUMENT - DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

ASW 3.7.8

3.7 PLANT SYSTEMS

3.7.8 Auxiliary Saltwater (ASW) System

LCO 3.7.8 Two ASW trains shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One ASW train inoperable. A.1 ------------NOTE-------------- -----------NOTE----------

Enter applicable A Completion Time of Conditions and Required 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> is Actions of LCO 3.4.6, applicable for ASW RCS Loops - MODE 4, pump 2-2 on a one-for residual heat removal time basis, for Unit 2 loops made inoperable by cycle 24.

ASW.

Restore ASW train to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status OR

In accordance with the RICT Program B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion AND Time of Condition A not met. B.2 -----------NOTE-----------

LCO 3.0.4.a is not applicable when entering MODE 4.

Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

DIABLO CANYON - UNITS 1 & 2 3.7-16 Unit 1 - Amendment No. 135, 200, 219, 238, 245 Rev 20 Page 19 of 38 Unit 2 - Amendment No. 135, 201, 221, 246, 247 Tab_3!7u3r20.docx 0513.1028 AC Sources - Operating 3.8.1

3.8 ELECTRICAL POWER SYSTEMS

3.8.1 AC Sources - Operating

LCO 3.8.1 The following AC electrical sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System; and

b. Three diesel generators (DGs) capable of supplying the onsite Class 1E power distribution subsystem(s); and

c. Two supply trains of the diesel fuel oil (DFO) transfer system.

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

ACTIONS

---

NOTE---------------------------------------------------

LCO 3.0.4b is not applicable to DGs.

CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite circuit A.1 Perform SR 3.8.1.1 for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable. required OPERABLE AND offsite circuit.

Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.

AND

A.2 Restore required offsite 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> circuit to OPERABLE OR status.

In accordance with the RICT Program (continued)

DIABLO CANYON - UNITS 1 & 2 3.8-1 Unit 1 - Amendment No. 135,166,169,215, 245 Rev 14 Page 1 of 37 Unit 2 - Amendment No. 135,167,170,217, 247 Tab_3!8u3r14.DOC 0605.1437 AC Sources - Operating 3.8.1

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One DG inoperable. B.1 Perform SR 3.8.1.1 for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the required offsite AND circuit(s).

Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.

AND

---

NOTE-------------

In MODE 1, 2, and 3, TDAFW pump is considered a required redundant feature.

B.2 Declare required 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from discovery feature(s) supported by of Condition B the inoperable DG concurrent with inoperable when its inoperability of required redundant redundant required feature(s) is inoperable. feature(s).

AND

B.3.1 Determine OPERABLE 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DG(s) is not inoperable due to common cause failure.

OR

B.3.2 Perform SR 3.8.1.2 for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE DG(s).

AND

B.4 Restore DG to 14 days OPERABLE status. OR

In accordance with the RICT Program (continued)

DIABLO CANYON - UNITS 1 & 2 3.8-2 Unit 1 - Amendment No. 135,166,215, 245 Rev 14 Page 2 of 37 Unit 2 - Amendment No. 135,167,217, 247 Tab_3!8u3r14.DOC 0605.1437 AC Sources - Operating 3.8.1

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Two required offsite circuitsC.1 Declare required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from inoperable. feature(s) inoperable discovery of Condition when its redundant C concurrent with required feature(s) is inoperability of inoperable. redundant required features.

AND C.2 Restore one required 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> offsite circuit to OR OPERABLE status. In accordance with the RICT Program D. One required offsite circuitD.1 Restore required offsite 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> inoperable. circuit to OPERABLE OR status. In accordance with the RICT Program AND OR One DG inoperable. D.2 Restore DG to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OPERABLE status. OR In accordance with the RICT Program E. Two or more DGs E.1 Ensure at least two DGs 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> inoperable. are OPERABLE.

F. One supply train of the F.1 Restore the DFO transfer 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> DFO transfer system system to OPERABLE OR inoperable. status. In accordance with the RICT Program G. Two supply trains of the G.1 Restore one train of the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> DFO transfer system DFO transfer system to inoperable. OPERABLE status.

(continued)

DIABLO CANYON - UNITS 1 & 2 3.8-3 Unit 1 - Amendment No. 135, 219, 240, 245 Rev 14 Page 3 of 37 Unit 2 - Amendment No. 135, 221, 241, 247 Tab_3!8u3r14.DOC 0605.1437 DC Sources - Operating 3.8.4

3.8 ELECTRICAL POWER SYSTEMS

3.8.4 DC Sources - Operating

LCO 3.8.4 Three Class 1E DC electrical power subsystems shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One battery charger A.1 Restore battery terminal 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> inoperable. voltage to greater than or equal to the minimum established float voltage.

AND A.2 Verify battery float 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> current 2 amps.

AND A.3 Restore battery charger 14 days to OPERABLE status. OR

In accordance with the RICT Program (continued)

DIABLO CANYON - UNITS 1 & 2 3.8-18 Unit 1 - Amendment No. 135,172,190, 245 Rev 14 Page 20 of 37 Unit 2 - Amendment No. 135,174, 247 Tab_3!8u3r14.DOC 0605.1437 DC Sources - Operating 3.8.4

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One battery inoperable. B.1 Restore battery to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> OPERABLE status. OR

In accordance with the RICT Program

C. One DC electrical power C.1 Restore DC electrical 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> subsystem inoperable for power subsystem to OR reasons other than OPERABLE status.

Condition A or B. In accordance with the RICT Program D. More than one full capacity D.1 Restore the DC electrical 14 days charger receiving power power subsystem to a simultaneously from a configuration wherein single 480 V vital bus. each charger is powered from its associated 480 volt vital bus.

(continued)

DIABLO CANYON - UNITS 1 & 2 3.8-18a Unit 1 - Amendment No. 135,172,190, 219, 245 Rev 14 Page 21 of 37 Unit 2 - Amendment No. 135,174, 221, 247 Tab_3!8u3r14.DOC 0605.1437 Inverters - Operating 3.8.7

3.8 ELECTRICAL POWER SYSTEMS

3.8.7 Inverters-Operating

LCO 3.8.7 Four Class 1E Vital 120 V UPS inverters shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required inverter A.1 -----------NOTE---------------

inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.9, "Distribution Systems -

Operating" with any vital 120 V AC bus de-energized.

Restore inverter to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status. OR

In accordance with the RICT Program B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion AND Time not met.

B.2 -----------NOTE-----------

LCO 3.0.4.a is not applicable when entering MODE 4.

Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct inverter voltage and alignment to In accordance with the required AC vital buses. Surveillance Frequency Control Program

DIABLO CANYON - UNITS 1 & 2 3.8-26 Unit 1 - Amendment No. 135, 200, 219, 245 Rev 14 Page 31 of 37 Unit 2 - Amendment No. 135, 201, 221, 247 Tab_3!8u3r14.DOC 0605.1437 Distribution Systems - Operating 3.8.9

3.8 ELECTRICAL POWER SYSTEMS

3.8.9 Distribution Systems-Operating

LCO 3.8.9 The required Class 1E AC, DC, and 120 VAC vital bus electrical power distribution subsystems shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One AC electrical power A.1 Restore AC electrical 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> distribution subsystem power distribution OR inoperable. subsystem to OPERABLE status. In accordance with the RICT Program B. One 120 VAC vital bus B.1 Restore 120 VAC vital 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> subsystem inoperable. bus subsystem to OR OPERABLE status.

In accordance with the RICT Program C. One DC electrical power C.1 Restore DC electrical 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> distribution subsystem power distribution OR inoperable. subsystem to OPERABLE status. In accordance with the RICT Program

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion AND Time not met.

D.2 -----------NOTE-----------

LCO 3.0.4.a is not applicable when entering MODE 4.

Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

E. Two required Class 1E AC, E.1 Enter LCO 3.0.3. Immediately DC, or 120 VAC vital buses with inoperable distribution subsystems that result in a loss of safety function.

DIABLO CANYON - UNITS 1 & 2 3.8-29 Unit 1 - Amendment No. 135, 215, 219, 245 Rev 14 Page 34 of 37 Unit 2 - Amendment No. 135, 217, 221, 247 Tab_3!8u3r14.DOC 0605.1437 Programs and Manuals 5.5

5.5 Programs and Manuals (continued) 5.5.19 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Ventilation System (CRVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.

The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition, including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"

Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CRVS, operating at the flow rate required by the VFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies required by paragraphs c and d for determining CRE unfiltered inleakage and assessing CRE habitability, and measuring CRE pressure and assessing the CRE boundary.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-17a Unit 1 - Amendment No. 201, 230, 245 Rev 38 Page 18 of 28 Unit 2 - Amendment No. 202, 232, 247 Tab_5!0u3r38.DOC 0626.1217 Programs and Manuals 5.5

5.5 Programs and Manuals (continued) 5.5.20 Risk Informed Completion Time (RICT) Program This program provides controls to calculate a RICT and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines."

The program shall include the following:

a. The RICT may not exceed 30 days;

b. A RICT may only be utilized in MODE 1 and 2;

c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required If the plant configuration change would lower plant risk and would result in a longer RICT.

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC.

The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

DIABLO CANYON - UNITS 1 & 2 5.0-17b Unit 1 - Amendment No. 245 Rev 38 Page 19 of 28 Unit 2 - Amendment No. 247 Tab_5!0u3r38.DOC 0626.1217 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO AMENDMENT NO. 245 TO FACILITY OPERATING LICENSE NO. DPR-80

AND AMENDMENT NO. 247 TO FACILITY OPERATING LICENSE NO. DPR-82

PACIFIC GAS AND ELECTRIC COMPANY

DIABLO CANYON NUCLEAR POWER PLANT, UNITS 1 AND 2

DOCKET NOS. 50-275 AND 50-323

1.0 INTRODUCTION

By letter dated July 13, 2023 (Reference 1), as supplemented by letter dated January 15, 2024 (Reference 2), Pacific Gas and Electric Company (PG&E, the licensee) submitted a license amendment request (LAR) for Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Diablo Canyon, DCPP).

The amendments would revise certain technical specification (TS) requirements to permit the use of a risk-informed completion time (RICT) for an action to be taken when a unit does not meet a limiting condition for operation (LCO). The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF [Risk-Informed TSTF] Initiative 4b, dated July 2, 2018 (Reference 3). The U.S. Nuclear Regulatory Commission (NRC, the Commission) issued a final revised model safety evaluation (SE) to be used when preparing a plant-specific SE of a request to adopt TSTF-505 on November 21, 2018 (Reference 4).

The NRC staff participated in a regulatory audit that included a virtual meeting in December 2023 to ascertain the information needed to support its review of the application and to identify any additional information needed to complete its SE. The licensees supplemental letter dated January 15, 2024, included information identified during the audit that the staff needed to complete its evaluation of the LAR. After the staff confirmed that no request for additional information would be needed following the submittal of the licensees supplemental letter, the audit was closed on January 15, 2024, and on March 22, 2024, the staff issued an audit summary (Reference 5).

The licensee has proposed variations from the TS changes approved in TSTF-505, which are described in section 2.3 and evaluated in sections 3.1-3.3 of this SE.

The supplemental letter dated January 15, 2024, 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 original proposed no signific ant hazards consideration determination as published in the Federal Register on October 3, 2023 (88 FR 68163).

Enclosure 3

2.0 REGULATORY EVALUATION

2.1 Regulatory Review

2.1.1 Applicable Regulations

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, includes, in part, the regulatory requirements for amending a license. The NRC staff has identified the followi ng sections within 10 CFR Part 50 applicable to its review of the licensees application to adopt TSTF-505, Revision 2:

10 CFR 50.36, Technical Specifications, paragraphs (c)(2), Limiting conditions for operation, and (c)(5), Administrative controls

10 CFR 50.55a, Codes and standards, paragraph (h), Protection and safety systems

10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (i.e., the Maintenance Rule)

2.1.2 Regulatory Guidance

NRC regulatory guides (RGs) provide one way to ensure that the codified regulations continue to be met. The NRC staff considered the following guidance, along with industry guidance endorsed by the NRC as described in section 2.1.3 of this SE, during its review of the proposed changes:

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed-Activities, March 2009 (Reference 6).

RG 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, May 2011 (Reference 7).1

RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific-Changes to the Licensing Basis, January 2018 (Reference 8).1

RG 1.177, Revision 2, Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, January 2021 (Reference 9).

NUREG-1855, Revision 1, Guidance on the Tr eatment of Uncertainties Associated with PRAs [Probabilistic Risk Assessments] in Risk-Informed Decisionmaking, March 2017 (Reference 10).

1 In this SE, references to RG 1.174 refer to Revision 3, unless specifically noted.

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP), Section 16.1, Revision 1, Risk-Informed Decision Making: Technical Specifications, March 2007 (Reference 11).

2.1.3 NRC-Endorsed Guidance

Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Revision 0-A (NEI 06-09-A), Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, dated October 2012 (Reference 12),

provides guidance for risk-informed TSs. The NRC endorsed NEI 06-09-A for trial use with clarifications in RG 1.201 and issued a final SE approving the industry guidance on May 17, 2007 (Reference 13).

The licensees submittal asserts that internal event, internal flood, fire, and seismic PRA models are consistent with RG 1.174, Revision 3; RG 1.200, Revision 2; and RG 1.177, Revision 2. The NRC staff evaluated the fire PRA against RG 1.174, Revision 2 because this version was used in the peer review of the fire PRA; however, the update to this regulatory guide in Revision 3 does not include any technical changes that would affect the conformance of NEI 06-09-A to its guidance. Therefore, the NRC staff finds that RG 1.174, Revision 3, is also applicable for use in the licensees adoption of TSTF-505 and will be the revision referred to in the remainder of this SE. Similarly, the staff found that the updates to RG 1.200 in Revision 2 and RG 1.177 in Revision 2 do not include any technical changes that would affect the conformance of NEI 06-09-A to their guidance, and were applicable for peer review and use in the adoption of TSTF-505.

2.2 Description of the RICT Program

The TS LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO is not met, the licensee must shut down the reactor or follow any remedial action (e.g., testi ng, maintenance, or repair activity) permitted by the TSs until the condition can be met. The remedial actions associated with an LCO contain conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are one or more Required Actions, each with an associated Completion Time (CT). Specified CTs are referred to as the front stops in the context of this SE. For certain conditions, the TSs require the licensee to exit the mode of applicability of an LCO (usually, to shut down the reactor).

The licensees submittal requested approval to add a RICT Program to the Administrative Controls section of the TSs, and to modify selected CTs to permit extending the CTs, provided risk is assessed and managed as described in NEI 06-09-A.

The licensees proposed changes to the TSs do not involve changes to the plants design, design basis, or any operating parameter. The effect of the proposed changes, when implemented, will allow CTs to vary based on the risk that is associated with the plant configuration. This depends on what equipment is out of service at that time and assumes no additional failures. It is important to note that RICTs may be used only if the affected system or systems retain the capability to perform the applicable safety function (e.g., one train of a two-train system remains operable and can perform the safety function). This restriction ensures that defense in depth is maintained.

The proposed RICT Program uses plant-specific operating experience for component reliability and availability data. Thus, the allowances permitted by the RICT Program reflect actual component performance in conjunction with component risk significance.

In addition, Example 1.3-8 will be added to TS 1.3, Completion Times, and reads as follows:

EXAMPLE 1.3-8

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

A. One subsystem A.1 Restore 7 days inoperable. subsystem to OR OPERABLE status. In accordance with the RICT Program

B. Required Action B.1 Be in MODE 3. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and associated AND Completion Time not met. B.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />

When a subsystem is declared inoperable, Condition A is entered. The 7 day Completion Time may be applied as discussed in Example 1.3-2. However, the licensee may elect to apply the Risk Informed Completion Time (RICT) Program which permits calculation of a RICT that may be used to complete the Required Action beyond the 7 day Completion Time. The RICT cannot exceed 30 days.

After the 7 day Completion Time has expired, the subsystem must be restored to OPERABLE status within the RICT or Condition B must also be entered.

The RICT requires recalculation of the RICT to reflect changing plant conditions.

For planned changes, the revised RICT must be determined prior to implementation of the change in configuration. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

If the 7 day Completion Time clock of Condition A has expired and subsequent changes in plant condition result in exiting the applicability of the RICT Program without restoring the inoperable subsystem to OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start.

If the RICT expires or is recalculated to be less than the elapsed time since the Condition was entered and the inoperable subsystem has not been restored to OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable subsystems are restored to OPERABLE status after Condition B is entered, Condition A is exited, and therefore, the Required Actions of Condition B may be terminated.

2.3 Variations

While the proposed amendments are consistent with TSTF-505, they do not include all of the items (required actions) that were modified in that document. In attachment 4, Diablo Canyon Unit 1 and 2 Scope for Adopting TSTF-505, Revision 2, to the LAR, the licensee identified proposed variations from the TS changes descr ibed in TSTF-505 or applicable parts of the NRC staffs model SE, including editorial variations. Attachment 4 also includes some plant-specific required actions that vary from the traveler due to plant-specific design and associated TSs. The staff review and evaluation of these variations is documented in sections 3.1-3.3 of this SE.

2.4 Additional Changes

In section 2.4, Additional Changes, of the enclosure to the LAR, the licensee proposed to delete two expired one-time TS extended CTs. The proposed changes would remove notes and required actions from TS 3.7.5, Auxiliary Feedwater (AFW) System, Condition G (and references to it in Conditions B and D), which allowed a 7-day CT to restore one or two inoperable AFW trains due to piping repair during Cycle 22 for Unit 1. Similarly, a one-time change to TS 3.8.4, DC [Direct Current] Sources - Operating, Required Actions B.2.1.1, B.2.1.2, and B.2.2 allowed a 4-hour CT to restore one inoperable battery during Cycle 14 for Unit 1. In attachment 4 of the LAR, the licensee also proposed to delete the expired one-time TS extended completion time from TS 3.7.8, Auxiliary Saltwater (ASW) System, Condition A, which allowed a 144-hour CT to restore ASW pump 1-1 during Cycle 23 for Unit 1, and from TS 3.8.1, AC [Alternate Current] Sources - Operating, Condition F, which allowed a 7-day CT for planned maintenance of each diesel fuel oil transfer pump in 2022. The staff review and evaluation of these variations is documented in section 3.2 of this SE.

3.0 TECHNICAL EVALUATION

An acceptable approach for making risk-informed decisions about proposed TS changes, including both permanent and temporary changes, is to demonstrate that the proposed licensing basis changes meet the five key principles of risk-informed decision-making provided in RG 1.174 and quoted in RG 1.177 as well as the three-tiered approach outlined in RG 1.177.

Each of the key principles and tiers are addressed in NEI 06-09-A and approved in the final model SE issued by the NRC for TSTF-505. NEI 06-09-A provides a methodology for extending existing CTs, and to thereby delay exiting the operational mode of applicability or taking required actions if risk is assessed and managed within the limits and programmatic requirements established by a RICT Program. The NRC staffs evaluation of the licensees proposed use of RICTs against the key safety principles of RG 1.174 and RG 1.177 is discussed below.

3.1 Review of Key Principles

3.1.1 Key Principle 1: Evaluation of Compliance with Current Regulations

Paragraph 50.36(c)(2) of 10 CFR requires that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any required action permitted by the TS until the condition can be met.

The CTs in the current TSs were established using experiential data, risk insights, and engineering judgment. In the LAR, the licensee proposed to add a new program (TS 5.5.20),

Risk Informed Completion Time (RICT) Program, in TS section 5.0, Administrative Controls, of the Diablo Canyon TSs. The proposed program would require adherence to NEI 06-09-A, Revision 0. The RICT Program provides the necessary administrative controls (evaluated in sections 3.1.4 and 3.1.5 of this SE) to permit extension of these CTs. In this program, if (1) risk is assessed and managed within specified limits and (2) programmatic requirements maintain adequate safety margins and sufficient defense in depth, reactor shutdown or completion of other required actions may be delayed. The option to determine the extended CT in accordance with the RICT Program allows the licensee to perform an integrated evaluation in accordance with the methodology prescribed in NEI 06-09-A and proposed TS 5.5.20. The RICT is limited to a maximum of 30 days (termed the back stop).

The typical CT is modified by the application of the RICT Program as shown in the following example. The changed portion is indicated in italics.

ACTIONS

CONDITION REQUIRED ACTION COMPLETION TIME

A. One subsystem A.1 Restore 7 days inoperable. subsystem to OR OPERABLE status. In accordance with the RICT Program

In attachment 1, Proposed Technical Specification Changes (Mark-Up), and enclosure 1, List of Revised Required Actions to Corresponding Probabilistic Risk Assessment (PRA) Functions, to the LAR, as supplemented, the licensee provided a list of the TSs, associated LCOs, and required actions for the CTs that included proposed modifications based on the approved TSTF-505. Attachment 4 to the LAR identified variations from the approved TSTF-505, including reductions in the scope of the RICT Program and provided plant-specific adjustments to the required actions and CTs.

The NRC staff reviewed the proposed changes to the TSs, associated LCOs, required actions, and CTs provided by the licensee and concluded that, with the incorporation of the RICT Program, the required performance levels of equipment specified in LCOs are not changed.

Only the required CTs for the required actions are modified; therefore, the TS modified as proposed in the LAR would continue to meet 10 CFR 50.36(c)(2). In addition, the incorporation of the RICT Program in TS 5.5.20 would ensure that the administrative controls section continues to include 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; therefore 10 CFR 50.36(c)(5) would continue to be met. Based on the discussion provided above, the staff finds that the TS program, LCOs, required actions, CTs, and administrative controls meet the first key principle of RG 1.174 and RG 1.177.

3.1.2 Key Principle 2: Evaluation of Defense in Depth

In RG 1.174, the NRC identified the following considerations used for evaluation of defense in depth and how the defense-in-depth philosophy is maintained as the licensing basis is changed:

Preserve a reasonable balance among the layers of defense.

Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

Preserve adequate defense against potential CCFs [common-cause failures].

Maintain multiple fission product barriers.

Preserve sufficient defense against human errors.

Continue to meet the intent of the plants design criteria.

The licensee requested to use the RICT Program to extend the existing CTs for the LCOs described in attachment 2, Revised Technical Specification Pages, to the LAR, as supplemented. The NRC staffs evaluation of the proposed changes for these LCOs assessed redundant or diverse means to mitigate accidents to ensure consistency with the plant licensing basis requirements. The staff used the guidance in RG 1.174, RG 1.177, and TSTF-505 to ensure that if operated in the proposed manner, adequate defense in depth for each safety function is maintained (i.e., the changes are consistent with the defense-in-depth criteria).

to the LAR, as supplemented, provides information supporting evaluation of the redundancy, diversity, and defense-in-depth for each LCO and required action as it relates to electrical power systems.

3.1.2.1 Proposed Changes to Electrical System TSs

The licensee proposed to apply the RICT Program to various LCOs specific to electrical power systems. If risk remains within the guidance of NEI 06-09-A, a RICT may be applied to the following LCOs.

LCO 3.8.1.AOne required offsite circuit inoperable LCO 3.8.1.BOne DG [diesel generator] inoperable LCO 3.8.1.CTwo required offsite circuits inoperable LCO 3.8.1.DOne required offsite circuit inoperable AND One DG inoperable LCO 3.8.1.FOne supply train of the DFO [diesel fuel oil] transfer system inoperable LCO 3.8.4.AOne battery charger inoperable LCO 3.8.4.BOne battery inoperable LCO 3.8.4.COne DC electrical power subsystem inoperable for reasons other than Condition A or B LCO 3.8.7.AOne required inverter Inoperable LCO 3.8.9.AOne AC electrical power distribution subsystem inoperable LCO 3.8.9.BOne 120 VAC [volt alternating current] vital bus subsystem inoperable LCO 3.8.9.COne DC electrical power distribution subsystem inoperable

3.1.2.2 Evaluation of Proposed Changes to Elec trical System Technical Specifications

According to the Diablo Canyon Final Safety Analysis Report Update (UFSAR) section 8.2.2 (Reference 14), offsite AC electrical power from the 230-kilovolt (kV) switchyard connects to each unit through its startup transformer, and similarly, the 500 kV switchyard provides a backfeed to each unit through that units main transformer and unit auxiliary transformers (associated with proposed changes to LCO 3.8.1 Conditions A, C, and D for offsite circuits and to LCO 3.8.9 Condition A for AC electrical power distribution subsystems).

According to Diablo Canyon UFSAR section 8.3.1.1.3.2, the onsite 4.16 kV Class 1E system for each unit consists of three buses with each supplied by the 230 kV and 500 kV switchyards, respectively, and its own DG. Proposed changes to LCO 3.8.1 Conditions B and D, are associated with DGs and LCO 3.8.9 Condition A. According to UFSAR section 8.3.1.1.3.3.5, only two of three Class 1E 4.16 kV buses are required to safely shut down a unit. Section 9.5.4 of the Diablo Canyon UFSAR discusses the configuration: there are two trains of the DFO system with only one required for safe shutdown, which is related to LCO 3.8.1, Condition F.

According to Diablo Canyon UFSAR section 8.3.2.3.9, General Design Criterion 24, 1967 -

Emergency Power for Protection Systems, the Class 1E 125 Vdc (volt direct current) power system has three load groups with only two required for safe shutdown. The batteries and battery chargers for those DC load groups are associated with proposed changes to LCO 3.8.4 Conditions A, B, and C and to LCO 3.8.9 Condition C for DC electrical power distribution subsystems.

According to Diablo Canyon UFSAR sections 8.3.1.1.5.2 and 8.3.1.1.5.3, the 120 VAC instrument power supply system has four inverters with only two associated with one of the two full engineered safety features trains required for safe shutdown (associated with LCO 3.8.7 Condition A for inverters and LCO 3.8.9 Condition B for 120 VAC vital bus subsystems).

The NRC staff evaluated each proposed RICT for the TSs associated with electrical systems to determine whether each RICT created a potential loss of function (LOF). The staff determined that there is no LOF for any of the RICTs since the electrical systems for each LCO had sufficient redundant and independent electrical sources/train available so that safety function is maintained and viable. The staff verified, for the LAR as supplemented, that each electrical LCO can be entered voluntarily or involuntarily based on NEI 06-09-A. The staff reviewed the design success criteria in enclosure 1 to the LAR, table E1-1, In-Scope TS/LCO conditions to the Corresponding PRA Functions. For each electrical TS LCO, the staff verified that the minimum electrical power sources would remain available to mitigate postulated design-basis accidents (DBAs); safely shutdown the reactor and maintain the reactor in a safe shutdown condition.

With regard to inconsistency in terminology between table E1-2, Unit 1/Unit 2 In-Scope TS/LCO Conditions RICT Estimate, in enclosure 1 to the LAR and some terms used in the Diablo Canyon UFSAR, the licensee stated that the terms subsystem, load groups, groups, channels, and trains are based on terms used in the standard technical specifications for Westinghouse plants, that may differ. The NRC staff found that the variation in terminology had minimal impact on its evaluation of the proposed TS LCO changes.

UFSAR section 8.3.1.1.2.3.2, General Design Criterion 4, 1967 - Sharing of Systems, discusses the 12-kV cross-tie and its alignment to allow one startup transformer to provide power to both units using the cross-tie breaker. Operation in this configuration is restricted by TS. The operating procedure OP J-2: VIII, Guidelines for Reliable Transmission Service for

DCPP, identifies that if the 12 kV cross-tie breaker is closed then one startup offsite power circuit to one unit is inoperable. This condition requires entry into LCO 3.8.1, Condition A, for one required offsite circuit inoperable. The time to provide power to both units is limited to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> by the current LCO 3.8.1 Condition A.2 CT. The NRC staff reviewed the licensees supplemental letter, which addressed when the 12 kV cross-tie is used and how that use is controlled. The licensee stated that the cross-tie is used when one startup transformer is unavailable and both units need access to offsite power. As described in the UFSAR, operation in this configuration is restricted by the TS. Use of the cross-tie is controlled by LCO 3.8.1 Condition A. The staff concluded that this is acceptable because it provides sufficient capacity to deal with a DBA (or unit trip) on one unit, with a concurrent safe shutdown of the second.

The NRC staff also verified that RICT estimates are provided for each of the electrical TS LCOs in LAR table E1-2, consistent with NEI 06-09A and the licensing basis. Based on the above evaluation, the staff finds that the Diablo Canyon electrical power systems would continue to provide safety functions as intended with the proposed TS changes.

In enclosure 12, Risk Management Action Examples, to the LAR, the licensee provided examples of risk management actions (RMAs) that may be considered during a RICT Program entry for the above required conditions to reduce the risk impact and ensure adequate defense in depth. The NRC staff evaluated the RMA examples, including the electrical examples for an inoperable DG and a battery. The staff determined that RMAs in the electrical examples had captured an acceptable level of detail, which describe actions that would reduce risk impact and provide adequate defense in depth. Based on the review, the staff determined that those examples provide reasonable assurance that the actual RMAs selected to monitor and control the risk for each Condition will be of similar quality and appropriate for the actual plant condition.

3.1.2.3 Conclusion with Respect to Defense in Depth

The NRC staff reviewed the licensees proposed electrical TS LCO changes and supporting documentation. Based on the evaluations above, the staff finds that given reduced redundancy in various LCOs, the CT extensions, as allo wed by the RICT Program, are acceptable because (a) the capacity and capability of the remaining operable electrical systems to perform their safety functions (assuming no additional failures) would remain adequate, and (b) the licensees identification and implementation of RMAs as co mpensatory measures, in accordance with the RICT Program, would provide adequate defense in depth.

In addition to the electrical technical specifications evaluated above, the licensee proposed other changes to the LCOs as described in enclosure 1 to the LAR. Each of these is consistent with the NRC-endorsed guidance prescribed in NEI 06-09 and satisfy the second key principle in RG 1.174 and RG 1.177. On this basis, the staff concludes that the changes are consistent with the defense-in-depth philosophy as described in RG 1.174.

3.1.3 Key Principle 3: Evaluation of Safety Margins

Paragraph 50.55a(h) of 10 CFR requires, in part, that [p]rotection systems of nuclear power reactors of all types must meet the requirements specified in this paragraph. Section 2.2.2, Technical Specification Change Maintains Sufficient Safety Margin (Principle 3), of RG 1.177 states, in part, that sufficient safety margins are maintained when:

a. Codes and standards or alternatives approved for use by the NRC are met.

b. Safety analysis acceptance criteria in the final safety analysis report (FSAR) are met, or proposed revisions provide sufficient margin to account for analysis and data uncertainties.

The licensee is not proposing to change any quality standard, material, or operating specification in this application. In the LAR, as supplemented, the licensee proposed to add a new program, Risk Informed Completion Time Pr ogram, in section 5.0, Administrative Controls, of the Diablo Canyon TS, which requires adherence to NEI 06-09-A.

The NRC staff evaluated the effect on safety margins when the RICT is applied to extend the CT up to a backstop of 30 days in a TS condition with sufficient trains remaining operable to fulfill the TS safety function. Although the licensee is able to have design-basis equipment out of service longer than the current TSs allow, any increase in unavailability is expected to be insignificant and is addressed by the consideration of the single failure criterion in the design-basis analyses. Acceptance criteria for operability of equipment are not changed and, if sufficient trains remain operable to fulfill the TS safety function, the operability of the remaining train(s) ensures that the current safety margins are maintained. The NRC staff finds that if the specified TS safety function remains operable, sufficient safety margins would be maintained during the extended CT of the RICT Program.

Safety margins are also maintained if PRA functionality is determined for the inoperable train, which would result in an increased CT. Credit for PRA functionality, as described in NEI 06-09-A, is limited to the inoperable train, subsystem, or component.

Based on the above, the NRC staff finds that the design-basis analyses for Diablo Canyon remains applicable and unchanged, sufficient safety margins are maintained during the extended CT, and the proposed changes to the TS do not include any change in the standards applied or the safety analysis acceptance criteria. The staff concludes that the proposed changes meet 10 CFR 50.55a(h), and therefore, the third key principle of RG 1.174 and RG 1.177.

3.1.4 Key Principle 4: Change in Risk Consistent with the Safety Goal Policy Statement

NEI 06-09-A provides a methodology for a licensee to evaluate and manage the risk impact of extensions to TS CTs. Permanent changes to the fixed TS CTs are typically evaluated by using the three-tiered approach described in SRP section 16.1 and RG 1.177. This approach addresses the calculated change in risk as measured by the change in core damage frequency (CDF) and large early release frequency (LERF), as well as the incremental conditional core damage probability and incremental conditional large early release probability, the use of compensatory measures to reduce risk, and t he implementation of a configuration risk management program (CRMP) to identify risk-significant plant configurations.

The NRC staff evaluated the licensees processes and methodologies for determining that the change in risk from implementation of the RICT Pr ogram is small and consistent with the intent of the Commissions Safety Goal Policy Statement. 2 In addition, the staff evaluated the licensees proposed changes against the three-tiered approach in RG 1.177 for the licensees

2 Commissions Safety Goal Policy Statement, Safety Goals for the Operations of Nuclear Power Plants; Policy Statement, published in the Federal Register on August 4, 1986 (51 FR 28044), as corrected, and republished, on August 21, 1986 (51 FR 30028).

evaluation of the risk associated with a proposed TS CT change. The results of the staffs review are discussed below.

3.1.4.1 Tier 1: PRA Capability and Insights

Tier 1 evaluates the impact of the proposed changes on plant operational risk. The Tier 1 review involves two aspects: (1) scope and acceptability of the PRA models and their application to the proposed changes and (2) a review of the PRA results and insights described in the licensees application.

In enclosures 2, Information Supporting Consistency with Regulatory Guide 1.200, Revision 2, and 4, Information Supporting the Justification of Excluding Sources of Risk Not Addressed by the Diablo Canyon Probabilistic Risk Assessment (PRA) Models, to the LAR, the licensee identified the PRA models that it will use to assess the risk contribution for extending the CT of an LCO.

internal events PRA model internal flood PRA model internal fire PRA model seismic PRA model

Other external hazards are screened from inclusion in the RICT Program based on Nonmandatory Appendix 6-A of ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008 Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (Reference 15, the 2009 PRA Standard, identified in the LAR as Addendum A).

3.1.4.1.1 PRA Scope

In the LAR, the licensee states that the Diablo Canyon PRA uses four at-power models: internal events, internal flooding, internal fire, and seismic events. The flooding, fire, and seismic models are based on the internal events model. Each PRA model is used to evaluate the CDF and LERF.

3.1.4.1.2 Evaluation of PRA Acceptability

In the LAR, the licensee states that the PRA models have been peer-reviewed and assessed using the 2009 PRA Standard and RG 1.200. The peer reviews and internal assessments of the internal events PRA (including internal floods), fire PRA, and seismic PRA models supporting the RICT Program are discussed in enclosure 2 to the LAR, as supplemented.

3.1.4.1.2.1 Internal Event and Internal Flood PRA

As stated in the LAR, an internal event and internal flood PRA full-scope peer review was conducted in December 2012, consistent with RG 1.200 and the 2009 PRA Standard. An independent assessment of the finding-level facts and observations (F&Os) was conducted in June 2023, using the Appendix X process documented in the NEI letter to the NRC Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-out to Facts and Observations, dated February 21, 2017 (Reference 16).

There were 12 finding-level F&Os that were identified as upgrades. A focused-scope peer review was therefore conducted for these 12 F&Os, in conjunction with the closure review. No other F&Os were determined to constitute an upgrade, and the assessment team identified no use of new methods. There are no remaining open peer review finding-level F&Os.

The NRC staff finds that the peer review of the internal event and internal flood PRA models was consistent with RG 1.200 and that all finding-level F&Os have been closed consistent with Appendix X. Therefore, the staff concludes that the internal event and internal flood PRA models are acceptable for use in the RICT Program.

In enclosure 9, Key Assumptions and Sources of Uncertainty, of the LAR, as supplemented, the licensee provided table E9-1, Disposition of Key Assumptions/Sources of Uncertainty Impacting Configuration Risk Calculations. The table includes a discussion of how the key assumptions and sources of uncertainty for each of the PRA models used for the RICT Program were identified, assessed, and dispositioned. In enclosure 9, the licensee stated that the guidance of NUREG-1855 was used. In the LAR supplement, the licensee identified the larger set of key documents it reviewed for uncertainties but did not find any additional sources of uncertainty to be addressed. In the LAR supplement, the licensee also clarified the disposition of specific potential sources of uncertainty indicated in LAR table E9-1. The licensee explained that the model used in the CRMP tool for the RICT Program would implement adjustments to address particular configurations and cases, reflecting the real-time availability of systems, treatment of shared systems and components, removing recovery credit, adjusting initiating event frequencies, applying a 24-hour mission time instead of a 6-hour mission time for the emergency DGs, and adjusting availability of the ASW System when one of two vacuum breakers is unavailable.

The NRC staff reviewed the licensees dispositions and found them to be an acceptable way to ensure that the CRMP tool would calculate realistic completion times with adequate conservatism, consistent with NEI 06-09-A. On this basis, the staff determined that the licensees treatment of key assumptions and uncertainties in the internal event and internal flood PRA was conservative and, therefore, acceptable for the RICT Program.

3.1.4.1.2.2 Internal Fire PRA

The fire PRA was reviewed in January 2008 as part of the pilot application of the fire PRA peer review process of NEI 07-12. The review was conducted against the requirements of the ANS Standard ANSI/ANS-58.23-2007, Fire PRA Methodology (Reference 17), which was withdrawn in 2009. In 2010, a peer review was conducted against the requirements of the endorsed 2009 PRA Standard. The scope of the 2010 review included reexamination of the elements that had not, in 2008, met Capability Category II (CC-II) of the PRA Standard.

All finding-level F&Os were resolved and in August through September 2018, were subjected to an independent assessment in accordance with the Appendix X process. This included two F&Os identified as upgrades, for which a focused-scope peer review was conducted in conjunction with the closure review. At the end of the independent assessment and focused peer review, no finding-level F&Os remained open. Because all finding-level F&Os have been closed consistent with Appendix X, the NRC staff c oncludes that the fire PRA is acceptable for use in the RICT Program.

In enclosure 9 to the LAR, as supplemented, the licensee provided a discussion of how the key assumptions and sources of uncertainty for each of the PRA models used for the RICT Program

were identified, assessed, and dispositioned. The licensee discussed each key assumption and source of uncertainty in LAR table E9-1. For the fire PRA model, the licensee identified the designation of systems and components as always failed in the fire PRA model as a key assumption and source of uncertainty. The licensee stated that those systems that are within the RICT Program and assumed failed are assumed always successful in the baseline PRA model used to calculate the RICT, so the resulting RICT is conservatively bounded.

The NRC staff determined that the licensees treatment of systems in the fire PRA was conservative and is consistent with NEI 06-09-A. Therefore, it is acceptable for the RICT Program.

3.1.4.1.2.3 Seismic PRA

In enclosure 2 to the LAR, the licensee stated that its seismic PRA model received a full-scope peer review in September 2017. The licensee explained that the peer review included a review of the seismic hazard and fragility analyses and was performed consistent with RG 1.200, using PRA Standard ASME/ANS RA-Sb-2013, Addenda to ASME/ANS RA-S2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (Reference 18) (the 2013 PRA Standard, identified as Addendum B in the LAR).

The licensee stated that an independent assessment of the finding-level F&Os was conducted in October through December 2017. The licensee further stated that three F&Os were identified by the licensee as upgrades and two additional F&Os were identified by the assessment team as upgrades. Consequently, a focused-scope peer review was conducted in conjunction with the closure review. The licensee concluded that (a) all applicable supporting requirements of Addendum B were met and (b) that at the conclusion of the independent assessment and focused-scope peer review, supporting r equirements satisfied at least CC-II.

In RG 1.200, the NRC staff endorsed the 2009 PRA Standard (Addendum A). However, the staff has not endorsed the 2013 PRA Standard (Addendum B). Since the peer review was performed using a standard that was not endorsed by the staff, the staff requested that the licensee provide a comparison of the supporting requirements in PRA Standard Addendum B with the supporting requirements in PRA Standard Addendum A to demonstrate that the supporting requirements of the PRA Standard have been met for instances where the criteria between the two standards differ.

In the supplement to the LAR, the licensee compared the supporting requirements in PRA Standard Addendum A with those in PRA Standard Addendum B. The licensee explained that it performed a gap assessment between them, consistent with the Vogtle Electric Generating Station, Units 1 and 2 (Vogtle) assessment described in a letter dated July 11, 2017, from Southern Nuclear Operating Company, Inc. (SNC) (Reference 19). The licensee stated that NRC acceptance of the assessment for Vogtle was documented in an NRC letter to SNC, dated August 10, 2018 (Reference 20).

The licensee stated that the Vogtle assessment showed that all but six of the supporting requirements in PRA Standard Addendum B (SHA-B3, SHA-C3, SFR-C3, SFR-C6, SFR-G3, and SPR-B1) are either equal to or envelop the corresponding supporting requirements of PRA Standard Addendum A. In its supplemental letter to the LAR, the licensee evaluated the six supporting requirements in table A2-2, Comparison of Supporting Requirements of Addendum A and Addendum B. For supporting requirements SHA-B3 and SHA-C3, the licensee identified that the seismic PRA met the requirements of Addendum A at CC-II or

higher. For supporting requirements SFR-C3 and SFR-C6, the licensee stated that the supporting requirements were addressed in the review of supporting requirement SFR-C2 during the peer review. For supporting requirements SFR-G3 and SPR-B1, the licensee stated that the seismic PRA conformed to Addendum A. The NRC staff notes that Addendum A does not provide a distinction between the capability categories for supporting requirements SFR-G3 and SPR-B1.

The letter from the NRC to SNC dated August 10, 2018, stated that the use of Addendum B was an acceptable alternative to the NRC-endorsed approach for Vogtles seismic PRA used to support that application. In doing so, the NRC staff did not endorse the use of Addendum B for other risk-informed licensing actions.

The NRC staff reviewed the evaluation containe d in the letter from SNC to the NRC dated July 11, 2017, and confirmed the relevance of that evaluation to the Diablo Canyon LAR.

Specifically, the staff examined the comparison of the supporting requirements in the endorsed Addendum A and the more recent Addendum B. The staff also reviewed the licensees comparison of the six supporting requirements in its supplement to the LAR. Based on these reviews, the staff finds that the licensees use of Addendum B adequately addresses the technical elements for the development of a seismic PRA for this application.

In enclosure 9 to the LAR, the licensee provided a discussion of how the key assumptions and sources of uncertainty for each of the PRA models used for the RICT Program were identified, assessed, and dispositioned. The licensee discussed each key assumption and source of uncertainty in table E9-1 of enclosure 9 to the LAR. For the seismic PRA model, the licensee identified the designation of systems and components as always failed in the seismic PRA model as a key assumption and source of uncertainty. The licensee stated that those systems that are within the RICT Program and assumed failed are assumed always successful in the baseline PRA model used to calculate the RICT, so the resulting RICT is conservatively bounded.

The NRC staff determined that the licensees treatment of systems in the seismic PRA was conservative and is consistent with NEI 06-09-A. Therefore, it is acceptable for the RICT Program.

In enclosure 2 to the LAR, the licensee stated that the seismic PRA does not credit any diverse and flexible mitigation capability (FLEX) equipment. The licensee stated that operator actions that model FLEX strategies to shed vital DC loads and to manually control the turbine-driven AFW pump are included in the seismic PRA model for a seismically induced station blackout and a station blackout with the loss of all DC power. The licensee noted that these actions include credit for FLEX strategies to monitor steam generator level at the hot shutdown panel without instrument AC power available.

Based on its review, the NRC staff finds that the licensees decision not to take further credit for FLEX equipment in the seismic PRA is conservative and is consistent with NEI 06-09-A.

Therefore, it is acceptable for the RICT Program.

Based on its review, the NRC staff finds that the use of the 2013 PRA Standard is an acceptable alternative to the NRC-endorsed approach for the licensees seismic PRA used to support this application, all finding-level F&Os were closed, and the licensees evaluation of the key assumptions and sources of uncertainty for its seismic PRA is consistent with RG 1.200.

Therefore, the Diablo Canyon seismic PRA is acceptable for use in the RICT Program.

3.1.4.1.2.4 Evaluation of Other External Hazards

Besides the seismic hazard discussed above, the licensee confirmed that other external hazards for Diablo Canyon have an insignificant contribution and proposed that these hazards be screened from the RICT Program.

The licensee provided its assessment of risk from other external hazards in enclosure 4 to the LAR. The hazards assessed in the LAR are those identified for consideration in nonmandatory Appendix 6-A of the 2009 PRA Standard, and they provide a guide for identification of most of the possible external events for a plant site.

In its supplemental letter to the LAR, the licensee provided additional information including the assumptions, data sources, methodology, and results for its assessment of the aircraft impact, extreme wind and tornado, hurricane, and tsunami external hazards. For each of these external hazards, the licensee performed a conservative analysis demonstrating the CDF has a mean frequency of less than 1E-06 per year.

The NRC staff reviewed the licensees assessment in the LAR, as supplemented. Based on its review, the staff finds that the contributions from aircraft impact, extreme wind and tornado, hurricane, tsunami, and other external hazards have an insignificant contribution to configuration risk and can be excluded from the calculation of the proposed RICTs because they either do not challenge the plant or they are bounded by the external hazards analyzed for the plant. The staffs review notes that the preliminary screening criteria and progressive screening criteria used and presented in enclosure 4 to the LAR are the same criteria presented in supporting requirements for screening external hazards EXT-B1 and EXT-C1 of the NRC-endorsed 2009 PRA Standard.

3.1.4.1.3 Application of PRA Models, Results, and Insights in the RICT Program to the LAR states that Diablo Canyon PRA models do not credit any FLEX equipment. Operator actions that model FLEX strategies, such as load shedding and manual control of the turbine-driven AFW pump are included in the seismic PRA model for a seismically induced station blackout or a station blackout with loss of all DC power. These actions include credit for FLEX strategies to monitor steam generator level at the hot shutdown panel without instrument AC power available.

The NRC staff noted that FLEX equipment was not modeled and concluded that the modeling of actions related to FLEX strategies was consistent with the NRC-endorsed 2009 PRA Standard and therefore acceptable.

The CRMP tool incorporates the peer-reviewed PRA models for internal events including internal floods, for internal fires, and for seismic events. The CRMP tool is used to analyze the risk for an extended CT. It produces results in the form of risk metrics that are consistent with the guidance of NEI 06-09-A.

In the LAR, the CRMP tool was described in enclosure 8, Attributes of the Configuration Risk Management Program (CRMP) Model. In the LAR, as supplemented, the licensee describes the changes to the peer-reviewed baseline PRA models made in the configuration risk software.

The adjustments support RICT calculations that preserve the CDF and LERF quantitative results, maintain the quality of the peer-reviewed PRA models, and correctly accommodate changes in risk due to configuration-specific considerations.

In table E1-1 of enclosure 1 to the LAR, as supplemented, the licensee identified several instances where a system, structure, or component (SSC) was not modeled in the PRA.

Instead, the licensee represented them using a surrogate, in a manner described as conservative. In the LAR supplement, the licens ee provided further justification of system success criteria and assumptions in the PRA model to capture the impact on the RICT.

Furthermore, in the LAR, as supplemented, the licensee clarified how the baseline PRA model is adjusted to support the CRMP tool and to disposition assumptions and sources of uncertainty in the baseline PRA model, as described in the previous paragraph.

The NRC staff reviewed the licensees proposed modeling surrogates and adjustments to the CRMP tool and found them to be consistent with NEI 06-09-A, and conservative, as discussed in the following. Therefore, they are acceptable to support the RICT Program.

In the LAR supplement, the licensee clarified how model adjustments are to be accomplished.

When surrogates are used in the PRA models, the baseline risk metrics (CDF and LERF) are calculated on the basis of assumed success of the surrogate, reducing the estimated values for CDF and LERF. These metrics are used as input to the CRMP application model as the baseline for all RICT calculations. When the CRMP tool is used, the reported values of delta CDF (CDF) and delta LERF (LERF) are maximized for all calculations, yielding a shorter and more conservative RICT.

In enclosure 9 to the LAR, the licensee identified that the single-unit PRA base model does not account for trips at dual-unit plants (except in seismic events). This was identified as nonconservative. In the LAR supplement, the licensee explained that shared components are included in the CRMP tool. It will model the actual unavailability of shared components and the impact on the opposite unit will be assessed. Additional risk management actions will be identified and selected when shared components are involved, using the process described in enclosure 12 to the LAR.

Certain modeling choices were described in the LAR as conservative, with adjustments to offset this conservatism. In certain conditions, this adjustment is not appropriate. In the LAR supplement, the licensee explained that the CRMP tool is programmable to the degree that when those particular conditions exist, the adjustment is automatically turned off.

Moreover, the RICT Program ensures that PRA acceptability is maintained for the models on which the CRMP tool is based and for the tool itself. Changes to the as-built, as-operated plant to reflect the operating experience at the plant are discussed in enclosure 7, Probabilistic Risk Assessment (PRA) Model Update Process, to the LAR. The licensee described its PRA model update process that ensures the PRA models that support the RICT Program are maintained consistent with the as-built, as-operated plant. The licensee has established a periodic update and review process for the PRA models and for the CRMP tool. The licensee explained that its process is consistent with NEI 06-09-A because it includes (1) reviewing plant changes and discovered conditions for potential impact on the PRA models and the CRMP tool, (2) establishing criteria for when to incorporate changes into the PRA models and tools immediately or waiting until the next periodic update, (3) updating the PRA models and CRMP tool at least once every two refueling cycles, and (4) performing interim analyses or imposing administrative restrictions if significant plant changes or discovered condi tions cannot be implemented immediately.

The NRC staff finds that the Diablo Canyon PRA models and the CRMP tool that will be used will reflect the as-built, as-operated plant consistent with RG 1.200. Therefore, the staff

concludes that the proposed application of the CRMP tool in the RICT Program is appropriate for use in the adoption of TSTF-505 for performing RICT calculations.

In enclosure 5, Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF), to the LAR, the licensee presented estimated mean total CDF and LERF of the base PRA models. These results demonstrate that Diablo Canyon meets the criteria of RG 1.174, which allow the licensee to incur small changes in risk such as those that may occur when CTs are adjusted under the RICT Program. Therefore, the NRC staff finds that the PRA results and insights used by the licensee in the RICT Program are acceptable.

3.1.4.1.4 Tier 1 Conclusions

Based on the above conclusions, the NRC staff finds that the licensee has satisfied the intent of RGs 1.174 and 1.177 for determining the PRA acceptable. Both the scope of the PRA models (internal events including flooding, fire, and seismic) and evaluation of risk from modeled hazards and other external hazards are appropriate for this application.

3.1.4.2 Tier 2: Avoidance of Risk-Significant Plant Configurations

As described in RG 1.177, the second tier evaluates the capability of the licensee to identify and avoid risk-significant plant configurations. Such configurations could result if equipment, other than that associated with the proposed change, is taken out of service. Other risk-significant operational factors, such as concurrent system or equipment testing, could have a similar effect.

In enclosure 12 to the LAR, the licensee identifies three kinds of RMAs (actions to provide increased risk awareness and control, actions to reduce the duration of maintenance activities, and actions to minimize the magnitude of the risk increase). In enclosure 12, the licensee also explains that RMAs are implemented in accordance with current plant procedures and will be put in place no later than at the point in time when the 1E-06 incremental core damage probability or 1E-07 incremental large early release probability threshold is reached. Under emergent conditions, the RMAs will be implemented when the instantaneous CDF and LERF thresholds are exceeded. The risk monitoring under 10 CFR 50.65(a)(4) will also limit the risk associated with the use of the RICT Program.

The NRC staff concludes that the Tier 2 attri butes of the proposed RICT Program, including limits established for entry into a RICT and implementation of RMAs, are consistent with NEI 06-09-A. The staff finds that the proposed changes are consistent with the Tier 2 guidance of RG 1.177. Therefore, the licensees Tier 2 program is acceptable and supports the proposed implementation of the RICT Program.

3.1.4.3 Tier 3: Risk-Informed Configuration Risk Management

The third tier stipulates that a licensee should develop a program that ensures the risk impact of out-of-service equipment is appropriately evaluated prior to performing any maintenance activity.

The proposed RICT Program uses the CRMP tool based on the underlying PRA models for internal events and floods, internal fires, and seismic events. The licensee uses the CRMP tool to evaluate configuration-specific risk for planned activities associated with the extended CT as well as emergent conditions that may arise during an extended CT. This required assessment of configuration risk, along with the implement ation of compensatory measures and RMAs, is

consistent with the principle of Tier 3 for assessing and managing the risk impact of out-of-service equipment.

Paragraph 50.36(c)(5) of 10 CFR identifies administrative controls as the provisions relating to organization and management, procedures, thereby assuring operation of the facility in a safe manner. In enclosure 8 to the LAR, the licensee confirmed that future changes made to the baseline PRA models and changes made to the online model (i.e., in the CRMP tool) are controlled and documented by plant procedures. In enclosure 10, Program Implementation, to the LAR, the licensee described the attributes that the RICT Program procedures will address, which are consistent with NEI 06-09-A. The NRC staff finds that the licensee has identified appropriate administrative controls consiste nt with NEI 06-09-A and 10 CFR 50.36(c)(5).

The licensee proposed a new TS (TS 5.5.20) to require implementation of the RICT Program in accordance with NEI 06-09-A and to use RMAs as discussed above. This is consistent with the Tier 3 guidance of RG 1.177. Therefore, the NRC staff finds the licensees Tier 3 program is acceptable and supports the proposed implementation of the RICT Program.

3.1.4.4 Key Principle 4: Conclusions

The licensee has demonstrated the technical acceptability and scope of the PRA models it will use to support implementation of the RICT Program. The impact of other external hazards has been considered and addressed appropriately, as discussed in section 3.1.4.1.2.4, above. The licensee has made proper consideration of key assumptions and sources of uncertainty. The risk metrics are consistent with the approved methodology of NEI 06-09-A and the guidance in RG 1.174 and RG 1.177. The RICT Program is controlled administratively through plant procedures and training and also follows the NRC-approved methodology in NEI 06-09-A. The NRC staff finds that the RICT Program satisfies the fourth key principle of RGs 1.174 and 1.177 and is, therefore, acceptable.

3.1.5 Key Principle 5: Performance Measurem ent Strategies - Implementation and Monitoring

The guidance in RGs 1.174 and 1.177 establishes the need for an implementation and monitoring program to ensure that extensions to TS CTs do not degrade operational safety over time and that no adverse degradation occurs due to unanticipated degradation or common cause mechanisms. In enclosure 11, Monitoring Program, to the LAR, the licensee states that the SSCs in the scope of the RICT Program are also in the scope of 10 CFR 50.65 for the Maintenance Rule. The Maintenance Rule monitoring programs provide for evaluation and disposition of unavailability impacts, which may be incurred from implementation of the RICT Program. Furthermore, in enclosure 11 to the LAR, the licensee confirmed that the cumulative risk is calculated at least once every refueling cyc le, not to exceed 24 months. This is consistent with NEI 06-09-A.

The NRC staff concludes that the RICT Program satisfies the fifth key principle of RGs 1.174 and 1.177 because: (1) the RICT Program monitors the average annual cumulative risk increase as described in NEI 06-09-A, thereby providing reasonable assurance that the program, as implemented, continues to meet RG 1.174 guidance for small risk increases; and (2) all affected SSCs are within the Maintenance Rule program, which monitors changes to the reliability and availability of these SSCs.

3.2 Removal of Expired Information

In section 2.4 of the enclosure to the LAR, the licensee proposed to delete one-time changes to TS 3.7.5 Condition G; and TS 3.8.4 Condition B, which allowed a one-time extension of the associated completion times from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days for inoperable AFW trains and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for an inoperable battery, respectively.

In attachment 4 to the LAR, the licensee also proposed to delete one-time changes to TS 3.8.1 Condition F, which allowed a 7-day completion time for planned maintenance of each DFO transfer pump in 2022.

The NRC staff reviewed the proposed deletions in 3.7.5, 3.8.1 and 3.8.4 and determined that because the requirements have expired, the proposed changes would be administrative and nontechnical in nature. The staff finds the changes acceptable because they would clarify the TS requirements for TSs 3.7.5, 3.8.1, and 3.8.4. Their removal has no effect on the adoption of TSTF-505.

In attachment 4, to the LAR, the licensee proposed to delete one-time changes to TS 3.7.8, Condition A, which allowed a 144-hour complete time to restore ASW pump 1-1 during Cycle 23 for Unit 1. Since the note had expired, the licensee concluded the change was editorial and had no effect on the adoption of TSTF-505. However, after the licensee submitted the LAR, the licensee requested, in a separate licensing action, to revise the Note in TS 3.7.8 Condition A to be applicable to ASW pump 2-2 on a one-time basis during Cycle 24 for Unit 2, instead of Unit 1; the NRC staff found the revision acceptable and issued Amendment No. 246 (Reference 21). Because the Note for Unit 1 has been superseded by Amendment No. 246, the NRC staff finds that the licensees request to delete the original Note is no longer applicable to this amendment request and the revised note will be retained, as per Amendment No. 246. The retention of this note is unrelated to and has no bearing on the licensees adoption of TSTF-505, and therefore, is acceptable.

The NRC staff concludes that, as amended by changes proposed in the LAR, TSs 3.7.5, 3.7.8, 3.8.1, and 3.8.4 continue to meet the requirements of 10 CFR 50.36(c)(2) because the LCOs continue to state the lowest functional capability or performance levels of equipment required for safe operation of the facility. The staff concludes that the required actions, as amended by the proposed changes, provide reasonable assurance that operation of the facility in accordance with the proposed changes would continue to support safe operation during times when the LCOs are not met.

3.3 Technical Evaluation Conclusion

The NRC staff evaluated the proposed changes against each of the five key principles of risk-informed decision-making, including the proposed variations from the approved TSTF-505, as discussed in sections 3.1.1 through 3.1.5 of this SE. The staff concludes that the changes proposed by the licensee satisfy the key principles of risk-informed decision-making identified in RG 1.174 and RG 1.177, and therefore, the requested adoption of the proposed changes to the TSs and associated guidance are an acceptable way to assure that the regulatory requirements identified in section 2.1.1 of this SE continue to be met.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the California State official was notified of the proposed issuance of the amendments on March 18, 2024. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change 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, as published in Federal Register on October 3, 2023 (88 FR 68163), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for catego rical 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.

7.0 REFERENCES

1. Petersen, D. B., Pacific Gas and Electric Company, letter to the U.S. Nuclear Regulatory Commission, License Amendment Request 23-01, Revision to Technical Specficiations to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, dated July 13, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23194A228).

2. Petersen, D. B., PG&E, letter to the NRC, Supplement to License Amendment Request 23-01: Revision to Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times, dated January 15, 2024 (ML24016A299).

3. Technical Specifications Task Force, TSTF Comments on Draft Safety Evaluation for Traveler TSTF-505, Provide Risk-Informed Extended Completion Times and Submittal of TSTF-505, Revision 2, TSTF-505, Revision 2, July 2, 2018 (Package ML18183A493).

4. Cusumano, V. G., U.S. Nuclear Regulatory Commission, letter to Technical Specifications Task Force, Final Revised Model Safety Evlaution of Traveler TSTF-505, Revision 2, Provide Risk Informed Extended Competion Times - RITSTF Initiative 4b, dated November 21, 2018 (Package ML18269A041).

5. Lee, S. S, U.S. Nuclear Regulatory Commission, letter to P. Gerfen, Pacific Gas and Electric Company, Diablo Canyon Nuclear Power Plant, Units 1 and 2 - Audit Summary in Support of the License Amendment Request To Adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b (EPID L-2023-LLA-0100), dated March 22, 2024 (ML24081A046).

6. U.S. Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, Revision 2, March 2009 (ML090410014).

7. U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 2, May 2011 (ML100910006).

8. U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 3, January 2018 (ML17317A256).

9. U.S. Nuclear Regulatory Commission, Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications, Regulatory Guide 1.177, Revision 2, January 2021 (ML20164A034).

10. U.S. Nuclear Regulatory Commission, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making, NUREG 1855, Revision 1, Final Report, March 2017 (ML17062A466).

11. U.S. Nuclear Regulatory Commission, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, NUREG-0800; Section 16.1, Revision 1, Risk-Informed Decision Making: Technical Specifications, March 2007 (ML070380228).

12. Bradley, B., Nuclear Energy Institute, letter to S. D. Stuchell, U.S. Nuclear Regulatory Commission, NEI 06-09, Risk-Informed Technical Specifications Initiative 4b:

Risk-Managed Technical Specification (RMTS) Guidelines, Revision 0-A, dated October 2012 (Package ML122860402).

13. Golder, J. M., U.S. Nuclear Regulatory Commission, letter to B. Bradley, Nuclear Energy Insitute, Final Safety Evaluation For Nuclear Energy Institute (NEI) Topical Report (TR)

NEI 06 09, Risk-Informed Technical Spec ifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (TAC No. MD4995), dated May 17, 2007 (ML071200238).

14. Pacific Gas and Electric Company, DCPP Units 1 & 2 FSAR Update, Chapter 8, Electric Power, Revision 27, dated May 2023 (ML23241A089).

15. American Society of Mechanical Engineers / American Nuclear Society Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, PRA Standard ASME/ANS RA-Sa-2009, February 2009, New York, NY (Copyright).

16. Anderson, V. K., Nuclear Energy Institute, letter to S. Rosenberg, U.S. Nuclear Regulatory Commission, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations, dated February 21, 2017 (Package ML17086A431).

17. American Nuclear Society, Fire PRA Methodology, ANSI/ANS 58.23-2007, November 20, 2007.

18. American Society of Mechanical Engineers / American Nuclear Society, Case for ASME/ANS RA-Sb-2013, Standard for Level 1 / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, dated November 22, 2017, New York, NY (Copyright).

19. Hutto, J. J., Southern Nuclear Operating Company, Inc, letter to U.S. Nuclear Regulatory Commission, Vogtle Electric Generating Plant Units 1 & 2 Response to Supplemental Information Needed for Acceptance of Systemat ic Risk-Informed Assessment of Debris Technical Report, dated July 11, 2017 (ML17192A245).

20. Orenak, M., NRC, letter to C. Gayheart, Southern Nuclear Operating Company, Inc.,

Vogtle Electric Generating Plant, Units 1 and 2 - Issuance of Amendments Regarding Application of Seismic Probabilistic Risk Assessment Into the Previously Approved 10 CFR 50.69 Categorization Process, dated August 10, 2018 (ML18180A062).

21. Lee, S. S., U.S. Nuclear Regulatory Commission, letter to P. Gerfen, Pacific Gas and Electric Company, Diablo Canyon Nuclear Power Plant, Unit 2 - Issuance of Amendment No. 246 Re: Revision to Technical Specification 3.7.8, Auxiliary Saltwater (ASW) System (Exigent Circumstances) (EPID L-2023-LLA-0155), dated December 7, 2023 (ML23324A153).

Principal Contributors: Malcolm Patterson Steven Alferink Mihaela Biro Michael Breach Robert Elliott Fred Forsaty Edmund Kleeh Hari Kodali Hanry Wagage Derek Scully Zeechung Wang Khadijah West

Date: May 29, 2024

ML24099A219 NRR-058 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DSS/STSB/BC(A) NRR/DRA/APLA/BC NAME SLee PBlechman SMehta RPascarelli (ABrown for)

DATE 4/4/2024 4/9/2024 4/3/2024 3/28/2024 OFFICE NRR/DRA/APLC/BC NRR/DEX/EEEB/BC NRR/DSS/SCPB/BC NRR/DSS/SNSB/BC NAME SVasavada WMorton BWittick PSahd DATE 2/9/2024 3/1/2024 4/3/2024 4/2/2024 OFFICE NRR/DEX/EMIB OGC NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME SBailey MWoods JRankin SLee DATE 3/29/2024 5/22/2024 5/29/2024 5/29/2024