View source for DCL-23-054, 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 - RITSTF Initiative 4b

{{Adams
| number = ML23194A228
| issue date = 07/13/2023
| title = 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 - RITSTF Initiative 4b
| author name = Petersen D
| author affiliation = Pacific Gas &amp; Electric Co
| addressee name = 
| addressee affiliation = NRC/NRR, NRC/Document Control Desk
| docket = 05000275, 05000323
| license number = DPR-080, DPR-082
| contact person = 
| case reference number = DCL-23-054
| document type = Letter type:DCL, Technical Specification, Amendment
| page count = 1
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Pacific Gas anti Elecmc Company*
Dennis B. Petersen  Diablo Canyon Power Plant Station Director    Mail code 104/5/502 P.O. Box 56 Avila Beach, CA 93424 805.545.4022 Dennis.Petersen@pge.com PG&E Letter DCL-23-054                                                             10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Diablo Canyon Units 1 and 2 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 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 -
RITSTF Initiative 4b

==Dear Commissioners and Staff:==

Pursuant to 10 CFR 50.90, Pacific Gas and Electric Company (PG&E) hereby requests approval of the enclosed proposed amendment to the Technical Specifications for Diablo Canyon Power Plant (DCPP) Units 1 and 2 to implement risk-informed Completion Times.
This license amendment request (LAR) is consistent with the NRC-approved Technical Specifications Task Force (TSTF) traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b.
The changes requested in this LAR will prevent unnecessary unit shutdowns for low-risk scenarios, and is consistent with safely maintaining DCPP generation and thereby supporting electrical grid reliability in California.
Approval of the proposed amendment is requested by July 23, 2024. Once approved, the amendment will be implemented within 365 days.
PG&E makes no regulatory commitments (as defined by NEI 99-04) in this letter.
This letter includes no revisions to any existing regulatory commitments.
The enclosure to this letter contains the evaluation of the proposed change.
In accordance with site administrative procedures and the DCPP Quality Assurance Program, the proposed amendment has been reviewed by the Plant Staff Review Committee.
A  member   of  the STARS   Alliance Callaway  Diablo Canyon  Palo Verde  Wolf Creek

Document Control Desk                                                        PG&E Letter DCL-23-054 Page 2 Pursuant to 10 CFR 50.91 (b)(1), PG&E is notifying the State of California of this LAR by transmitting a copy of this letter and enclosure to the California Department of Public Health.
* If you have any questions or require additional information, please contact James Morris, Manager, Nuclear Regulatory Services, at 8Q5-545-4609.
I state under penalty of perjury that the foregoing is true and correct.
Sincerely, Dennis B. Petersen Station Director Executed on:      7 /t 3/z.oz J Date kjse/51178920 Enclosure cc:        Diablo Distribution cc/enc: Mahdi 0. Hayes, NRC Senior Resident Inspector Samson S. Lee, NRR Project Manager Robert J . Lewis, NRC Acting Region IV Administrator Gonzalo L. Perez, Branch Chief, California Department of Public Health A  mem be r    of  the  STARS   A l l ia nce Callaway
* Di abl o Canyon
* Palo Verde
* Wolf Cr eek

PG&E Letter DCL-23-054 Enclosure Evaluation of the Proposed Change

==Subject:==
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 - RITSTF Initiative 4b
: 1.       

==SUMMARY==
DESCRIPTION
: 2.       ASSESSMENT 2.1      Applicability of Published Safety Evaluation 2.2      Variations and Regulatory Commitments 2.3      Optional Changes and Variations 2.4      Additional Changes
: 3.       REGULATORY EVALUATION
: 4.       ENVIRONMENTAL CONSIDERATION
: 5.       REFERENCES ATTACHMENTS:
: 1.       Proposed Technical Specification Changes (Mark-Up)
: 2.       Revised Technical Specification Pages
: 3.       Proposed Technical Specification Bases Changes (Mark-Up) - For Information Only
: 4.       Diablo Canyon Units 1 and 2 Scope for Adopting TSTF-505, Revision 2 ENCLOSURES:
: 1.       List of Revised Required Actions to Corresponding Probabilistic Risk Assessment (PRA) Functions
: 2.       Information Supporting Consistency with Regulatory Guide 1.200, Revision 2
: 3.       Information Supporting Technical Adequacy of Probabilistic Risk Assessment (PRA) Models Without Standards Endorsed by Regulatory Guide 1.200, Revision 2
: 4.       Information Supporting Justification of Excluding Sources of Risk Not Addressed by the Diablo Canyon Probabilistic Risk Assessment (PRA)
Models 1

PG&E Letter DCL-23-054 Enclosure
: 5. Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)
: 6. Justification of Application of At-Power Probabilistic Risk Assessment (PRA)
Models to Shutdown Modes
: 7. Probabilistic Risk Assessment (PRA) Model Update Process
: 8. Attributes of the Configuration Risk Management Program (CRMP) Model
: 9. Key Assumptions and Sources of Uncertainty
: 10. Program Implementation
: 11. Monitoring Program
: 12. Risk Management Action (RMA) Examples 2

PG&E Letter DCL-23-054 Enclosure EVALUATION
: 1.     

==SUMMARY==
DESCRIPTION The proposed amendment would modify the Technical Specification (TS) requirements related to the Completion Times (CTs) for Required Actions (RAs) to provide the option to calculate a longer, risk-informed completion time (RICT). A new program, the Risk Informed Completion Time Program, is added to TS Section 5, Administrative Controls.
The methodology for using the RICT Program is described in NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0 (Reference 1), which was approved by the NRC on May 17, 2007. Adherence to NEI 06-09-A is required by the RICT Program, and Pacific Gas and Electric Company (PG&E) is not proposing any deviations from the NEI guidance.
The proposed amendment is consistent with Technical Specifications Task Force (TSTF) traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b (Reference 2). However, only those RAs described in  to this Enclosure are proposed to be changed. Attachment 4 does not include all of the RAs that were modified in TSTF-505, Revision 2 and includes some plant-specific RAs that are not included in TSTF-505, Revision 2. The plant-specific RAs are due to the site-specific system design configuration and are Conditions that are not a loss of a safety function; therefore, there is no adverse safety impact from including these plant-specific RAs in the scope of the proposed changes.
: 2. ASSESSMENT 2.1    Applicability of Published Safety Evaluation PG&E has reviewed TSTF-505, Revision 2 and the model safety evaluation dated November 21, 2018 (ADAMS Accession No. ML18253A085). This review included the supporting information provided to support TSTF-505, Revision 2 and the safety evaluation for NEI 06-09-A. As described in the subsequent paragraphs, PG&E has concluded that the technical basis is applicable to Diablo Canyon Units 1 and 2, and supports incorporation of this amendment in the Diablo Canyon TS.
2.2    Variations and Regulatory Commitments In accordance with Section 4.0, Limitations and Conditions, of the safety evaluation for NEI 06-09-A, the following is provided:
3

PG&E Letter DCL-23-054 Enclosure
: 1. Enclosure 1 identifies each of the TS RAs to which the RICT Program will apply, with a comparison of the TS functions to the functions modeled in the probabilistic risk assessment (PRA) of the structures, systems and components (SSCs) subject to those actions.
: 2. Enclosure 2 provides a discussion of the results of peer reviews and self-assessments conducted for the plant-specific PRA models which support the RICT Program, as required by Regulatory Guide (RG) 1.200, Revision 2, Section 4.2.
: 3. Enclosure 3 is not applicable since each PRA model used for the RICT Program is addressed using a standard endorsed by the Nuclear Regulatory Commission.
: 4. Enclosure 4 provides appropriate justification for excluding sources of risk not addressed by the PRA models.
: 5. Enclosure 5 provides the plant-specific baseline core damage frequency (CDF) and large early release frequency (LERF) to confirm that the potential risk increases allowed under the RICT Program are acceptable.
: 6. Enclosure 6 is not applicable since the RICT Program is not being applied to shutdown modes.
: 7. Enclosure 7 provides a discussion of the DCPP programs and procedures that assure the PRA models that support the RICT Program are maintained consistent with the as-built, as-operated plant.
: 8. Enclosure 8 provides a description of how the baseline PRA model, which calculates the average annual risk, is evaluated and modified to assess real-time configuration risk, and describes the scope of, and quality controls applied to, the real-time model.
: 9. Enclosure 9 provides a discussion of how the key assumptions and sources of uncertainty in the PRA models were identified, and how their impact on the RICT Program was assessed and dispositioned.
: 10. Enclosure 10 provides a description of the implementing programs and procedures regarding the plant staff responsibilities for the RICT Program implementation, including risk management action (RMA) implementation.
: 11. Enclosure 11 provides a description of the implementation and monitoring program as described in NEI 06-09-A, Section 2.3.2, Step 7.
: 12. Enclosure 12 provides a description of the process to identify and provide RMAs.
4

PG&E Letter DCL-23-054 Enclosure 2.3     Optional Variations PG&E is proposing variations from the TS changes described in TSTF-505, Revision 2 or the applicable parts of the NRC staffs model safety evaluation dated November 21, 2018. These options were identified as acceptable variations in TSTF-505, Revision 2 and the NRC model safety evaluation due to plant specific design and associated TS, or are otherwise justified in Attachment 4. provides a cross-reference of the TSTF 505, Revision 2, Standard Technical Specification (STS) changes to the Diablo Canyon Units 1 and 2 TS changes proposed in this LAR. Attachment 4 provides individual dispositions of each STS change and Diablo Canyon Units 1 and 2 change. Where the changes are consistent, a disposition of No variation is provided. Where a variation is taken, the disposition provides a justification.
2.4     Additional Changes In addition to the changes associated with the adoption of TSTF-505, Revision 2 described in Attachment 4, PG&E is proposing additional editorial changes as follows:
TS 3.7.5 Condition G was added to TS 3.7.5 and was applicable during Unit 1 cycle 22 during repair of Auxiliary Feedwater (AFW) piping. This Condition has expired and is proposed to be removed from TS 3.7.5, including references to Condition G in Conditions B and D. Condition G and its removal have no impact on adoption of TSTF-505, Revision 2.
TS 3.8.4 RA B.2 was added to TS 3.8.4 and was applicable during Unit 1 cycle
: 14. This RA has expired and is proposed to be removed from TS 3.8.4. RA 3.8.4 B.2 and its removal have no impact on adoption of TSTF-505, Revision 2.
: 3.      REGULATORY SAFETY ANALYSIS 3.1     No Significant Hazards Consideration Analysis PG&E has evaluated the proposed change to the Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has determined that the proposed change does not involve a significant hazards consideration.
DCPP Units 1 and 2 request the adoption of an approved change to the Standard Technical Specifications (STS) and plant-specific TS, to modify the TS requirements related to the Completion Times for Required Actions to provide the option to calculate a longer, risk-informed Completion Time. The allowance is described in a new program 5

PG&E Letter DCL-23-054 Enclosure in Chapter 5, Administrative Controls, entitled Risk Informed Completion Time (RICT)
Program.
As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change permits the extension of Completion Times provided the associated risk is assessed and managed in accordance with the NRC approved Risk-Informed Completion Time Program. The proposed change does not involve a significant increase in the probability of an accident previously evaluated because the change involves no change to the plant or its modes of operation.
The proposed change does not increase the consequences of an accident because the design-basis mitigation function of the affected systems is not changed and the consequences of an accident during the extended Completion Time are no different from those during the existing Completion Time.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not change the design, configuration, or method of operation of the plant. The proposed change does not involve a physical alteration of the plant (no new or different kind of equipment will be installed).
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed change permits the extension of Completion Times provided risk is assessed and managed in accordance with the NRC approved Risk-Informed Completion Time Program. The proposed change implements a risk-informed configuration management program to ensure that adequate margins of safety 6

PG&E Letter DCL-23-054 Enclosure are maintained. Application of these new specifications and the configuration management program considers cumulative effects of multiple systems or components being out of service and does so more effectively than the current TS.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, PG&E concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c),
and, accordingly, a finding of no significant hazards consideration is justified.
3.2      Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
: 4.       ENVIRONMENTAL CONSIDERATION PG&E has reviewed the environmental evaluation included in the model safety evaluation published on November 21, 2018 (ADAMS Accession No. ML18267A259) as part of the Notice of Availability. PG&E has concluded that the NRC staff findings presented in that evaluation are applicable to DCPP Units 1 and 2.
The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
7

PG&E Letter DCL-23-054 Enclosure
: 5. REFERENCES
: 1. NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, Revision 0, November 2006. (ADAMS Accession No. ML12286A322).
: 2. TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times -
RITSTF Initiative 4b. (ADAMS Accession No. ML18183A493, approved at ADAMS Accession No. ML18253A085).
8

PG&E Letter DCL-23-054 Attachment 1 Attachment 1 Proposed Technical Specification Changes (Mark-Up)

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 hour subsystem               subsystem AND inoperable.             isolated.
Once per 8 hours thereafter AND A.2   Restore              72 hours subsystem to OPERABLE status.
B. Required          B.1   Be in MODE 3.        6 hours Action and AND associated Completion        B.2   Be in MODE 5.        36 hours Time not met.
Required Action A.1 has two Completion Times. The 1 hour Completion Time begins at the time the Condition is entered and each "Once per 8 hours 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 hour or any subsequent 8 hour 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.
INSERT 1 IMMEDIATE            When "Immediately" is used as a Completion Time, the Required COMPLETION           Action should be pursued without delay and in a controlled manner.
TIME DIABLO CANYON - UNITS 1 & 2                    1.3-10         Unit 1 - Amendment No. 135, Unit 2 - Amendment No. 135,

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 hours within limit.
                                  -AND A.2      Fully insert all rods.       6 hours
                                  -AND A.3      Place Rod Control            6 hours System in a condition incapable of rod withdrawal.
                                  -AND A.4      Be in MODE 4.                12 hours B. One required group of         B.1      Restore required group       72 hours pressurizer heaters                    of pressurizer heaters to OR inoperable.                            OPERABLE status.
1-In accordance with the RICT Program I
C. Required Action and            C.1      Be in MODE 3.                6 hours associated Completion Time of Condition B not      -AND met.                          C.2       Be in MODE 4.               12 hours DIABLO CANYON - UNITS 1 & 2                   3.4-16          Unit 1 - Amendment No. 135, Unit 2 - Amendment No. 135,

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 hour inoperable solely due to excessive seat leakage.
power to associated block valve.                                                      ,r B. One PORV inoperable for reasons other than excessive seat leakage.
B.1
                                                -AND B.2 Close associated block valve.
Remove power from 1 hour 1 hour
                                                                                                                               ,r associated block valve.
                                                -AND B.3         Restore the Class I                72 hours PORV to OPERABLE OR status.
1-In accordance with the RICT Program                 I C. One block valve inoperable.                  ----------------NOTE-------------------        1 hour 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.
(continued)
                                                -AND DIABLO CANYON - UNITS 1 & 2                                  3.4-19          Unit 1 - Amendment No. 135,169,171, Unit 2 - Amendment No. 135,170,172,

Pressurizer PORVs 3.4.11 ACTIONS CONDITION                           REQUIRED ACTION               COMPLETION TIME C. (continued)                      C.2       If the block valve is        72 hours associated with a Class I PORV:
Restore block valve to OPERABLE status.
                                                                           -I OR I
In accordance with the RICT Program OR-C.3       If the block valve is        72 hours 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 hours AND   -
D.3       Be in MODE 4.                12 hours E. Two Class I PORVs inoperable for reasons other than excessive seat leakage.
E.1 AND   -
Initiate action to restore Class I PORVs to OPERABLE status.
Immediately f
E.2       Close associated block       1 hour valves.
AND   -
E.3       Remove power from            1 hour associated block valves.
AND   -
E.4       Be in MODE 3.                6 hours AND   -
E.5       Be in MODE 4.                12 hours (continued)
DIABLO CANYON - UNITS 1 & 2                         3.4-20       Unit 1 - Amendment No. 135, 171, Unit 2 - Amendment No. 135, 172,

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.
F.1          Place associated                 1 hour PORVs in manual control.
                              -AND F.2         Restore one block valve           2 hours for a Class I PORV to OPERABLE status.                  OR
                              -AND 1-In accordance with the RICT Program I
F.3         Restore remaining block           72 hours valve for a Class I PORV to OPERABLE status.              -OR In accordance with the RICT Program
                              -OR F.4         If the remaining block            72 hours 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 hours
                              -AND G.3         Be in MODE 4.                     12 hours DIABLO CANYON - UNITS 1 & 2                   3.4-21            Unit 1 - Amendment No. 135, Unit 2 - Amendment No. 135,

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 hours 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 hours inoperable.                                         OPERABLE status                      ----------NOTE-----------
AND                                                                                      The Required Action A.1 Completion Time At least 100% of the ECCS                                                                is to be used for flow equivalent to a single                                                              planned maintenance OPERABLE ECCS train                                                                      or inspections. The available.                                                                               Completion Times of OR                                              Required Actions
. - - - - - - - - - - - ,-/ = =                                                                 A.2.1, A.2.2, and OR                                                                                             A.2.3 are for In accordance with the RICT                                                                    unplanned corrective Program                                                                                        maintenance or inspections.
A.2.1      Verify only one                      72 hours subsystem in one ECCS train is inoperable.
AND A.2.2      Determine there is no                72 hours common cause failure in the same subsystem in the OPERABLE ECCS train.
AND A.2.3      Restore train to                     14 days OPERABLE status.
B. Required Action and                      B.1        Be in MODE 3.                        6 hours associated Completion Time not met.                            AND B.2        Be in MODE 4.                        12 hours DIABLO CANYON - UNITS 1 & 2                            3.5-3      Unit 1 - Amendment No. 135, 159, 202, Unit 2 - Amendment No. 135, 146, 160, 203,

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 hour the affected air lock.
                                         -AND C.3       Restore air lock to               24 hours OPERABLE status.
1-OR In accordance with the RICT Program           I D. Required Action and                   D.1       Be in MODE 3.                     6 hours associated Completion Time not met.                        -AND D.2       Be in MODE 5.                     36 hours 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, Unit 2 - Amendment No. -135, -201,

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 hours Only applicable to penetration flow paths with two containment isolation valves.
penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, blind 1-OR In accordance with the RICT Program I
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 containment purge supply                 -AND 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, Unit 2 - Amendment No. 135, 232,

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         Ifollowing isolationIfor 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 hour 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, Unit 2 - Amendment No. 135,

Containment Isolation Valves 3.6.3 ACTIONS (continued)
CONDITION                          REQUIRED ACTION                      COMPLETION TIME C. --------------NOTE--------------      C.1   Isolate the affected              72 hours Only applicable to                          penetration flow path by OR 1-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                                      I
    ------------------------------------        flange.
One or more penetration flow paths with one                  -AND containment isolation valve           C.2   ------------NOTES----------
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          Ifollowing isolationI isolated.
D. One or more penetration               D.1   Isolate the affected              24 hours 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.
(continued)
                                         -AND DIABLO CANYON - UNITS 1 & 2                        3.6-7             Unit 1 - Amendment No. 135, Unit 2 - Amendment No. 135,

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 hours train inoperable.                         spray train to                   ...
                                      ~
OPERABLE status.                   ---------NOTE------------
 -OR                                                                                For planned maintenance or In accordance with the RICT inspections, the Program l---                                               Completion Time is 72 hours. The Completion Times of Required Action A.2 are for unplanned corrective
                                      --OR                                          maintenance or inspections.
A.2      Restore containment                14 days spray train to OPERABLE status B. Required Action and              B.1      Be in MODE 3.                      6 hours associated Completion Time of Condition A not        -AND met.                             B.2.     -----------NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 4.
Be in MODE 4.                     54 hours C. One required CFCU system           C.1      Restore required CFCU              7 days inoperable such that a                    system to OPERABLE minimum of two CFCUs remain OPERABLE.
status.
1-OR In accordance with the RICT Program (continued)
I DIABLO CANYON - UNITS 1 & 2                 3.6-13    Unit 1 - Amendment No. 135,202,215, 219, Unit 2 - Amendment No. 135,173,203,217, 221,

Containment Spray and Cooling Systems 3.6.6 ACTIONS (continued)
CONDITION                 REQUIRED ACTION                    COMPLETION TIME i
D. One required containment      D.1  Restore one required             72 hours spray train inoperable and       containment spray system one required CFCU system         to OPERABLE status,              OR inoperable such that a minimum of two CFCUs remain OPERABLE.           -OR D.2  Restore one CFCU 1-In accordance with the RICT Program 72 hours I
system to OPERABLE status such that four            OR CFCUs or three CFCUs, each supplied by a different vital bus, are OPERABLE.
1-In accordance with the RICT Program I
E. Required Action and         E.1  Be in MODE 3.                    6 hours 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 hours 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, Unit 2 - Amendment No. 135, 173, 221.

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 hours MODE 1.                                     OPERABLE status.
OR 1-In accordance with the RICT Program           I B. Required Action and                   B.1   Be in MODE 2.               6 hours associated Completion Time of Condition A not met.
C. ------------NOTE----------------        C.1   Close MSIV.                 8 hours Separate Condition entry is allowed for each MSIV.                -AND 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 hours associated Completion Time of Condition C not met.               -AND D.2  Be in MODE 4.                12 hours DIABLO CANYON - UNITS 1 & 2                          3.7-4         Unit 1 - Amendment No. 135, Unit 2 - Amendment No. 135,

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 B. Two required ADV lines        B.1    Restore at least one ADV 1-OR In accordance with the RICT Program 72 hours I
inoperable.                          line to OPERABLE status.
C. Three or more required         C.1    Restore at least two ADV     24 hours ADV lines inoperable.                lines to OPERABLE status.
D. Required Action and            D.1    Be in MODE 3.                6 hours associated Completion Time not met.                -AND D.2    Be in MODE 4 without         18 hours reliance upon steam generator for heat removal.
DIABLO CANYON - UNITS 1 & 2                   3.7-8         Unit 1 - Amendment No. -135,169, Unit 2 - Amendment No. -135,170,

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 inoperable steam supply OR -
      -----------NOTE---------------
equipment to OPERABLE status.
                                                                                           -1 OR In accordance with the RICT Program I
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 hours in MODE 1, 2 or 3 for reasons other than Condition A or G.   -
OPERABLE status.
                                                                                           -I OR In accordance with the RICT Program If I (continued)
DIABLO CANYON - UNITS 1 & 2                           3.7-10        Unit 1 - Amendment No. 135,169, 215, 236, Unit 2 - Amendment No. 135,170, 217, 238,

AFW System 3.7.5 ACTIONS (continued)
CONDITION                         REQUIRED ACTION                      COMPLETION TIME C. -----------NOTE---------------        C.1    Restore the steam                48 hours Only applicable when the                    supply to the turbine OR remaining OPERABLE motor driven AFW train provides feedwater to the steam generator with the inoperable steam supply.
driven train to OPERABLE status.
1-In accordance with the RICT Program
                                         -OR                                      48 hours Turbine driven AFW train inoperable due to one inoperable steam supply.
    -AND C.2    Restore the motor driven AFW train to OPERABLE status.                 1-OR In accordance with the RICT Program One motor driven AFW train inoperable.
D. Required Action and                  D.1     Be in MODE 3.                    6 hours associated Completion
                                        -AND Time for Condition A, B, or C,- or G not met.                  D.2     Be in MODE 4.
18 hours
    -OR Two AFW trains inoperable in MODE 1, 2 or 3 for reasons other than Condition C-or G.
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.
(continued)
DIABLO CANYON - UNITS 1 & 2                    3.7-11        Unit 1 - Amendment No. 135,169, 215, 236, Unit 2 - Amendment No. 135,170, 217, 238,

AFW System 3.7.5 ACTIONS (continued)
CONDITION                              REQUIRED ACTION              COMPLETION TIME G. -----------NOTE---------------         G.1       Isolate AFW flow path(s) 2 hours This Condition is only                           to affected steam applicable to Unit 1 once                        generator.
during Unit 1 Cycle 22 during repair of AFW                   AND piping.                                G.2       Restore AFW train(s) to  7 days
    ----------------------------------               OPERABLE status.
One or two AFW trains inoperable in MODE 1, 2, or 3 due to inoperable AFW piping affecting the AFW flow path(s) to one steam generator.
This page that was added in amendments 236 for Unit 1 and 238 for Unit 2 is no longer needed and will be deleted.
DIABLO CANYON - UNITS 1 & 2                  3.7-11a      Unit 1 - Amendment No. 236, Unit 2 - Amendment No. 238,

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
 -OR                                              Actions of LCO 3.4.6, In accordance with the RICT                     RCS Loops - MODE 4, Program                                          for residual heat removal r---            loops made inoperable by CCW.
                                        ~ -.
Restore vital CCW loop to            72 hours OPERABLE status.
B. Required Action and                 B.1    Be in MODE 3.                        6 hours associated Completion Time of Condition A not            -AND met.                                B.2        -----------NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 4.
Be in MODE 4.                    12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                 FREQUENCY SR 3.7.7.1        -----------------------------NOTE----------------------------
Isolation of CCW flow to individual components does not render the CCW System inoperable Verify each CCW manual, power operated, and                        In accordance with the automatic valve in the flow path servicing safety                  Surveillance related equipment, that is not locked, sealed, or                  Frequency Control otherwise secured in position, is in the correct                   Program position.
(continued)
DIABLO CANYON - UNITS 1 & 2                        3.7-14         Unit 1 - Amendment No. 135, 200, 219, Unit 2 - Amendment No. 135, 201, 221,

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 hours is Actions of LCO 3.4.6,               applicable for ASW RCS Loops - MODE 4,               pump 1-1 on a one-for residual heat removal           time basis, for Unit 1 loops made inoperable by            cycle 23.
ASW.
Restore ASW train to                72 hours OPERABLE status B. Required Action and           B.1    Be in MODE 3.
1-OR In accordance with the RICT Program 6 hours I
associated Completion Time of Condition A not      -AND met.                          B.2       -----------NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 4.
Be in MODE 4.                   12 hours DIABLO CANYON - UNITS 1 & 2         3.7-16       Unit 1 - Amendment No. 135, 200, 219, 238, Unit 2 - Amendment No. 135, 201, 221,

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 hour inoperable.                                   required OPERABLE AND offsite circuit.
Once per 8 hours thereafter.
AND A.2         Restore required offsite            72 hours circuit to OPERABLE status.                             OR In accordance with the RICT Program (continued)
DIABLO CANYON - UNITS 1 & 2                      3.8-1       Unit 1 - Amendment No. 135,166,169,215, Unit 2 - Amendment No. 135,167,170,217,

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 hour the required offsite AND circuit(s).
Once per 8 hours thereafter.
AND
                                   ------------NOTE-------------
In MODE 1, 2, and 3, TDAFW pump is considered a required redundant feature.
B.2    Declare required                  4 hours 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 hours DG(s) is not inoperable due to common cause failure.
OR B.3.2  Perform SR 3.8.1.2 for            24 hours OPERABLE DG(s).
AND B.4    Restore DG to                     14 days OPERABLE status.
OR 1-In accordance with the RICT Program (continued)
DIABLO CANYON - UNITS 1 & 2          3.8-2         Unit 1 - Amendment No. 135,166,215, Unit 2 - Amendment No. 135,167,217,

AC Sources - Operating 3.8.1 ACTIONS (continued)
CONDITION                  REQUIRED ACTION                  COMPLETION TIME C. Two required offsite circuits C.1   Declare required              12 hours from inoperable.                       feature(s) inoperable         discovery of Condition when its redundant            C concurrent with OR                                     required feature(s) is        inoperability of In accordance with the RICT            inoperable.                   redundant required Program                                                              features.
AND C.2   Restore one required          24 hours offsite circuit to OPERABLE status.
D. One required offsite circuit  D.1   Restore required offsite      12 hours inoperable.                       circuit to OPERABLE status.
AND                         OR One DG inoperable.          D.2   Restore DG to                 12 hours OPERABLE status.
E. Two or more DGs             E.1  Ensure at least two DGs        2 hours inoperable.                      are OPERABLE.
F. One supply train of the                                         ----------NOTE----------
DFO transfer system A separate one-time inoperable.
use only Completion Time of 7 days is allowed during planned maintenance of each DFO transfer system pump 0-1 and 0-2 in 2022 with the Portable DFO transfer pump staged and available.
F.1  Restore the DFO transfer       72 hours system to OPERABLE status.
G. Two supply trains of the      G.1  Restore one train of the       1 hour 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, Unit 2 - Amendment No. 135, 221, 241,

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 hours inoperable.                          voltage to greater than or equal to the minimum established float voltage.
AND A.2     Verify battery float       12 hours current  2 amps.
AND A.3     Restore battery charger    14 days to OPERABLE status.
OR 1-In accordance with the RICT Program (continued)
DIABLO CANYON - UNITS 1 & 2               3.8-18      Unit 1 - Amendment No. -135,172,190, Unit 2 - Amendment No. -135,174,

DC Sources - Operating 3.8.4 ACTIONS (continued)
CONDITION                      REQUIRED ACTION                COMPLETION TIME B. One battery inoperable. B.1      Restore battery to           2 hours OR--
OPERABLE status.
                                                                       -1 OR In accordance with the B.2.1.1 ------------NOTE------------ RICT Program I
Required Actions B.2.1.1, B.2.1.2, and B.2.2 are applicable, on a one time basis, for Unit 1 cycle 14.                      2 hours Determine OPERABLE batteries are not inoperable due to common cause failure.
OR                           2 hours B.2.1.2 Perform SR 3.8.4.1 and SR 3.8.6.1 for OPERABLE batteries.
4 hours AND--
B.2.2    Restore battery to OPERABLE status.
C. One DC electrical power       C.1      Restore DC electrical        2 hours subsystem inoperable for reasons other than Condition A or B.
power subsystem to OPERABLE status.
                                                                       -I OR In accordance with the RICT Program I
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) f DIABLO CANYON - UNITS 1 & 2              3.8-18a    Unit 1 - Amendment No. 135,172,190, 219, Unit 2 - Amendment No. 135,174, 221,

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 OR                                           Actions of LCO 3.8.9, 1-In accordance with the RICT Program 1--
                                              "Distribution Systems -
Operating" with any vital
                                     ~
120 V AC bus de-energized.
Restore inverter to                  24 hours OPERABLE status.                   .........
B. Required Action and             B.1     Be in MODE 3.                        6 hours associated Completion Time not met.                  -AND B.2       -----------NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 4.
Be in MODE 4.                     12 hours 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, Unit 2 - Amendment No. 135, 201, 221,

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.
                                                               -OR APPLICABILITY:        MODES 1, 2, 3, and 4.                     In accordance with the RICT Program ACTIONS j
CONDITION                      REQUIRED ACTION Ill COMPLETION TIME A.
B.
One AC electrical power distribution subsystem inoperable.
One 120 VAC vital bus A.1 B.1 Restore AC electrical power distribution subsystem to OPERABLE status.
Restore 120 VAC vital I/
8 hours 2 hours subsystem inoperable.                    bus subsystem to OPERABLE status.
I/
C. One DC electrical power distribution subsystem inoperable.
C.1    Restore DC electrical power distribution subsystem to I
2 hours OPERABLE status.
D. Required Action and                D.1    Be in MODE 3.                     6 hours associated Completion Time not met.                    -AND D.2    -----------NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 4.
Be in MODE 4.                    12 hours 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, Unit 2 - Amendment No. 135, 217, 221,

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.
r
: 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.
INSERT 2 DIABLO CANYON - UNITS 1 & 2                5.0-17a       Unit 1 - Amendment No. 201, 230, Unit 2 - Amendment No. 202, 232,

INSERT 1 1.3 Completion Times EXAMPLES          EXAMPLE 1.3-8 (continued)
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 hours and associated AND Completion Time not met.            B.2   Be in MODE 5.       36 hours 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 hours 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.

INSERT 2 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     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 hours 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.

PG&E Letter DCL-23-054 Attachment 2 Attachment 2 Revised Technical Specification Pages Remove Page                                Insert Page 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                                    None 3.7-14                                     3.7-14 3.7-15                                     3.7-15 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

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 hour subsystem               subsystem AND inoperable.             isolated.
Once per 8 hours thereafter AND A.2   Restore              72 hours subsystem to OPERABLE status.
B. Required          B.1   Be in MODE 3.        6 hours Action and AND associated Completion        B.2   Be in MODE 5.        36 hours Time not met.
Required Action A.1 has two Completion Times. The 1 hour Completion Time begins at the time the Condition is entered and each "Once per 8 hours 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 hour or any subsequent 8 hour 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, Rev 13 Page 21 of 27                                          Unit 2 - Amendment No. 135, 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 hours Action and AND associated Completion      B.2   Be in MODE 5.       36 hours 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 hours 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.
Rev 13 Page 22 of 27                                      Unit 2 - Amendment No.
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.
Rev 13 Page 23 of 27                                       Unit 2 - Amendment No.
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 hours within limit.
AND A.2      Fully insert all rods.       6 hours AND A.3      Place Rod Control            6 hours System in a condition incapable of rod withdrawal.
AND A.4      Be in MODE 4.                12 hours B. One required group of         B.1      Restore required group       72 hours pressurizer heaters                    of pressurizer heaters to inoperable.                            OPERABLE status.             OR In accordance with the RICT Program C. Required Action and            C.1      Be in MODE 3.                6 hours associated Completion AND Time of Condition B not met.
C.2      Be in MODE 4.                12 hours DIABLO CANYON - UNITS 1 & 2                   3.4-16          Unit 1 - Amendment No. 135, Rev 17 Page 17 of 40                                          Unit 2 - Amendment No. 135, 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 hour inoperable solely due to                             power to associated excessive seat leakage.                              block valve.
B. One PORV inoperable for                  B.1         Close associated block              1 hour reasons other than                                   valve.
excessive seat leakage.
AND B.2         Remove power from                   1 hour associated block valve.
AND B.3         Restore the Class I                 72 hours PORV to OPERABLE status.                             OR In accordance with the RICT Program C. One block valve inoperable.                 ----------------NOTE-------------------         1 hour 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, Rev 17 Page 20 of 40                                                        Unit 2 - Amendment No. 135,170,172, 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 hours associated with a Class I PORV:                          OR Restore block valve to         In accordance with the OPERABLE status.               RICT Program OR C.3  If the block valve is          72 hours 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 hours AND D.3  Be in MODE 4.                  12 hours 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 hour valves.
AND E.3  Remove power from              1 hour associated block valves.
AND E.4  Be in MODE 3.                  6 hours AND E.5  Be in MODE 4.                  12 hours (continued)
DIABLO CANYON - UNITS 1 & 2                 3.4-20         Unit 1 - Amendment No. 135, 171, Rev 17 Page 21 of 40                                       Unit 2 - Amendment No. 135, 172, 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.
F.1          Place associated                1 hour PORVs in manual control.
AND F.2         Restore one block valve          2 hours for a Class I PORV to OPERABLE status.                 OR In accordance with the RICT Program AND F.3         Restore remaining block          72 hours valve for a Class I PORV to OPERABLE status.              OR In accordance with the RICT Program OR F.4         If the remaining block           72 hours 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 hours AND G.3         Be in MODE 4.                    12 hours DIABLO CANYON - UNITS 1 & 2                     3.4-21            Unit 1 - Amendment No. 135, Rev 17 Page 22 of 40                                              Unit 2 - Amendment No. 135, Tab_3!4u3r17.DOC       0629.0950

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 hours 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 hours inoperable.                                          OPERABLE status.                    OR AND                                                                                      In accordance with the RICT Program At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.
B. Required Action and                      B.1         Be in MODE 3.                       6 hours associated Completion Time not met.                            AND B.2         Be in MODE 4.                       12 hours DIABLO CANYON - UNITS 1 & 2                            3.5-3      Unit 1 - Amendment No. 135, 159, 202, Rev 12 Page 3 of 9                                                Unit 2 - Amendment No. 135, 146, 160, 203, Tab_3!5u3r12.DOC                0605.1325

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 hour the affected air lock.
AND C.3        Restore air lock to               24 hours OPERABLE status.
OR In accordance with the RICT Program D. Required Action and                   D.1        Be in MODE 3.                     6 hours associated Completion AND Time not met.
D.2        Be in MODE 5.                     36 hours 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, Rev 10 Page 4 of 20                                                       Unit 2 - Amendment No. 135, 201, 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 hours 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, Rev 10 Page 5 of 20                                                             Unit 2 - Amendment No. 135, 232, 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 hour 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, Rev 10 Page 6 of 20                                                  Unit 2 - Amendment No. 135, Tab_3!6u3r10.DOC              0605.1329

Containment Isolation Valves 3.6.3 ACTIONS (continued)
CONDITION                         REQUIRED ACTION                     COMPLETION TIME 72 hours C. --------------NOTE--------------      C.1   Isolate the affected 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 containment isolation valve          C.2   ------------NOTES----------
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 hours 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, Rev 10 Page 7 of 20                                                  Unit 2 - Amendment No. 135, Tab_3!6u3r10.DOC              0605.1329

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 hours 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 hours 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 hours 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, Rev 10 Page 13 of 20                              Unit 2 - Amendment No. 135,173,203,217, 221, Tab_3!6u3r10.DOC          0626.1341

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 hours 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 hours 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 hours 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 hours 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, Rev 10 Page 14 of 20                                  Unit 2 - Amendment No. 135, 173, 221, 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 hours MODE 1.                                    OPERABLE status.
OR In accordance with the RICT Program B. Required Action and                  B.1   Be in MODE 2.               6 hours associated Completion Time of Condition A not met.
C. ------------NOTE----------------       C.1   Close MSIV.                 8 hours Separate Condition entry is allowed for each MSIV.                AND 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 hours associated Completion Time of Condition C not met.               AND D.2  Be in MODE 4.                12 hours DIABLO CANYON - UNITS 1 & 2                         3.7-4         Unit 1 - Amendment No. 135, Rev 20 Page 4 of 37                                               Unit 2 - Amendment No. 135.
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 hours inoperable.                          line to OPERABLE status.
C. Three or more required        C.1     Restore at least two ADV     24 hours ADV lines inoperable.                lines to OPERABLE status.
D. Required Action and           D.1     Be in MODE 3.                6 hours associated Completion AND Time not met.
D.2     Be in MODE 4 without         18 hours reliance upon steam generator for heat removal.
DIABLO CANYON - UNITS 1 & 2                   3.7-8         Unit 1 - Amendment No. 135,169, Rev 20 Page 9 of 37                                         Unit 2 - Amendment No. 135,170, 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.
In accordance with the OR 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 hours 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, Rev 20 Page 11 of 37                                                 Unit 2 - Amendment No. 135,170, 217, 238, 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 hours 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                                                  RICT Program steam generator with the           OR inoperable steam supply.           C.2    Restore the motor driven         48 hours
     ----------------------------------        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 hours associated Completion AND Time for Condition A, B, or C not met.                                                                 18 hours 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, Rev 20 Page 12 of 37                                       Unit 2 - Amendment No. 135,170, 217, 238, 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 hours OPERABLE status.
OR In accordance with the RICT Program B. Required Action and         B.1    Be in MODE 3.                       6 hours 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 hours DIABLO CANYON - UNITS 1 & 2              3.7-14         Unit 1 - Amendment No. 135, 200, 219, Rev 20 Page 16 of 37                                    Unit 2 - Amendment No. 135, 201, 221, 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 Verify each CCW manual, power operated, and                           In accordance with automatic valve in the flow path servicing safety                     the Surveillance related equipment, that is not locked, sealed, or                     Frequency Control otherwise secured in position, is in the correct                      Program 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, Rev 20 Page 17 of 37                                                   Unit 2 - Amendment No. 135, 201, Tab_3!7u3r20.DOC      0606.1547

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--------------
Enter applicable Conditions and Required Actions of LCO 3.4.6, RCS Loops - MODE 4, for residual heat removal loops made inoperable by ASW.
Restore ASW train to                  72 hours OPERABLE status OR I
In accordance with the RICT Program B. Required Action and         B.1    Be in MODE 3.                         6 hours associated Completion AND I
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 hours DIABLO CANYON - UNITS 1 & 2               3.7-16         Unit 1 - Amendment No. 135, 200, 219, 238, II II             I Rev 20 Page 18 of 37                                     Unit 2 - Amendment No. 135, 201, 221, II Tab_3!7u3r20.DOC     0606.1547

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 hour inoperable.                                   required OPERABLE AND offsite circuit.
Once per 8 hours thereafter.
AND A.2         Restore required offsite            72 hours 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, Rev 14 Page 1 of 37                                          Unit 2 - Amendment No. 135,167,170,217, 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 hour the required offsite AND circuit(s).
Once per 8 hours thereafter.
AND
                                      ------------NOTE-------------
In MODE 1, 2, and 3, TDAFW pump is considered a required redundant feature.
B.2    Declare required                  4 hours 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 hours DG(s) is not inoperable due to common cause failure.
OR B.3.2  Perform SR 3.8.1.2 for            24 hours 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, Rev 14 Page 2 of 37                                   Unit 2 - Amendment No. 135,167,217, Tab_3!8u3r14.DOC     0605.1437

AC Sources - Operating 3.8.1 ACTIONS (continued)
CONDITION                   REQUIRED ACTION                  COMPLETION TIME C. Two required offsite circuits C.1   Declare required              12 hours 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 hours offsite circuit to            OR OPERABLE status.
In accordance with the RICT Program D. One required offsite circuit  D.1   Restore required offsite      12 hours inoperable.                        circuit to OPERABLE           OR status.
In accordance with the RICT Program AND                          OR One DG inoperable.           D.2   Restore DG to                 12 hours OPERABLE status.              OR In accordance with the RICT Program E. Two or more DGs              E.1  Ensure at least two DGs        2 hours inoperable.                       are OPERABLE.
F. One supply train of the      F.1  Restore the DFO transfer       72 hours 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 hour 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, Rev 14 Page 3 of 37                                        Unit 2 - Amendment No. 135, 221, 241, 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 hours inoperable.                         voltage to greater than or equal to the minimum established float voltage.
AND A.2     Verify battery float       12 hours 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, Rev 14 Page 20 of 37                                 Unit 2 - Amendment No. 135,174, 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 hours OPERABLE status.
OR In accordance with the RICT Program C. One DC electrical power        C.1   Restore DC electrical      2 hours 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, Rev 14 Page 21 of 37                                Unit 2 - Amendment No. 135,174, 221, 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 hours OPERABLE status.
OR In accordance with the RICT Program B. Required Action and             B.1     Be in MODE 3.                      6 hours associated Completion Time not met.                   AND B.2       -----------NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 4.
Be in MODE 4.                   12 hours 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, Rev 14 Page 31 of 37                                        Unit 2 - Amendment No. 135, 201, 221, 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 hours 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 hours 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 hours 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 hours associated Completion Time not met.                    AND D.2     -----------NOTE-----------
LCO 3.0.4.a is not applicable when entering MODE 4.
Be in MODE 4.                    12 hours 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, Rev 14 Page 34 of 37                                         Unit 2 - Amendment No. 135, 217, 221, 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, Rev 38 Page 18 of 28                                          Unit 2 - Amendment No. 202, 232, 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 hours 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.
Rev 38 Page 19 of 28                                         Unit 2 - Amendment No.
Tab_5!0u3r38.DOC         0626.1217

PG&E Letter DCL-23-054 Attachment 3 Attachment 3 Proposed Technical Specification Bases Changes (Mark-Up)
(For Information Only)

Pressurizer B 3.4.9 BASES (continued)
APPLICABILITY    The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature, resulting in the greatest effect on pressurizer level and RCS pressure control. Thus, applicability has been designated for MODES 1 and 2. The applicability is also provided for MODE 3. The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant pump startup.
In MODES 1, 2, and 3, there is need to maintain the availability of pressurizer heaters capable of being powered from either the offsite power source or the emergency power supply, and if necessary, using bus cross-tie to an OPERABLE emergency diesel generator. In the event of a loss of offsite power, the initial conditions of these MODES give the greatest demand for maintaining the RCS in a hot pressurized condition with loop subcooling for an extended period. For MODE 4, 5, or 6, it is not necessary to control pressure (by heaters) to ensure loop subcooling for heat transfer when the Residual Heat Removal (RHR)
System is in service, and therefore, the LCO is not applicable.
ACTIONS          A.1, A.2, A.3, and A.4 Pressurizer water level control malfunctions or other plant evolutions may result in a pressurizer water level above the nominal upper limit, even with the plant at steady state conditions. The upper limit of this LCO is below the Pressurizer Water Level - High Trip at 90% of span.
If the pressurizer water level is not within the limit, action must be taken to bring the unit to a MODE in which the LCO does not apply. To achieve this status, within 6 hours the unit must be brought to MODE 3, with rods fully inserted and the Rod Control System not capable of rod withdrawal (e.g., de-energize all CRDMs by opening the RTBs or de-energizing the motor - generator sets). Additionally, the unit must be brought to MODE 4 within 12 hours. This takes the unit out of the applicable MODES.
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
B.1 If one required group of pressurizer heaters is inoperable, restoration is required within 72 hours or in accordance with the Risk Informed Completion Time Program. The Completion Time of 72 hours is reasonable considering the anticipation that a demand caused by loss of offsite power would be unlikely in this period. Pressure control may be maintained during this time using normal station powered heaters.

Pressurizer PORVs B 3.4.11 BASES ACTIONS A.1 (continued)
(MODE 6) so that maintenance can be performed on the PORVs to eliminate the problem condition. Normally, the PORVs should be available for automatic mitigation of overpressure events and should be returned to OPERABLE and automatic actuation status prior to entering startup (MODE 2).
Quick access to the PORV for pressure control can be made when power remains on the closed block valve. The Completion Time of 1 hour is based on plant operating experience that has shown that minor problems can be corrected or closure accomplished in this time period.
B.1, B.2, and B.3 If one PORV is inoperable and not capable of automatic pressure relief or not capable of being manually cycled, it must be either restored or isolated by closing the associated block valve and removing the power to the associated block valve. The Completion Time of 1 hour for Required Actions B.1 and B.2 is reasonable, based on challenges to the PORVs during this time period, and provides the operator adequate time to correct the situation.
If the inoperable PORV cannot be restored to OPERABLE status, it must be isolated within the specified time of 1 hour. Because at least one Class I PORV remains OPERABLE, an additional 72 hours is provided to restore the inoperable PORV to OPERABLE status if it is Class I. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. If the valve is the non-Class I PORV, there is no required Completion Time.
If the Class I PORV cannot be restored within this additional time, the plant must be brought to a MODE in which the LCO does not apply as required by Condition D.
C.1, C.2, and C.3 If one PORV block valve is inoperable, then it is necessary to either restore the block valve to OPERABLE status within the Completion Time of 1 hour or place the associated PORV in manual control. The PORV control switch has three positions; open, close, and auto.
Placing the PORV in manual control, if required in ACTION C, is accomplished by positioning the switch out of the auto control mode.
The prime importance for the capability to close the block valve is to isolate a stuck open PORV. Therefore, if the block valve cannot be restored to OPERABLE status within 1 hour, the Required Action is to place the associated PORV in manual control.
(continued)

Pressurizer PORVs B 3.4.11 BASES ACTIONS C.1, C.2, and C.3 (continued)
This action is taken to avoid the potential for a stuck open PORV if the valve were to open under automatic control at a time that the block valve is inoperable. The Completion Time of 1 hour is reasonable, based on the small potential for challenges to the system during this time period, and provides the operator time to correct the situation. If the inoperable block valve is associated with a Class 1 PORV, the operator is permitted a Completion Time of 72 hours to restore the inoperable block valve to OPERABLE status. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. The time allowed to restore the Class I PORV block valve is based upon the Completion Time for restoring an inoperable Class I PORV in Condition B, since the PORVs are not capable of mitigating a SGTR or spurious operation of the safety injection system at power event, or a main feedwater line break accident when inoperable. If the block valve is restored within the Completion Time of 72 hours, the PORV will be transferred to the automatic mode of operation. If the block valve cannot be restored within this additional time, the plant must be brought to a MODE in which the LCO does not apply as required by Condition D.
If the inoperable block valve is associated with the non-Class I PORV, the block valve may be closed and the power removed. The 72 hour Completion Time for closing the block valve is the same applied in Required Action C.2. This recognizes that some restoration work may be required since the block valve is inoperable.
Restoration of the non-class I PORV block valve to OPERABLE status is not required because the non-Class I PORV is not required to be available, although having the valve closed impairs the load rejection design capability. Therefore, once the block valve has been closed per Required Action C.3, Completion Time requirements of Condition D do not apply.
If the block valve cannot be placed in the closed position, per Required Action C.3, Condition D applies and the unit must be taken to MODE 4 until the block valve is restored or closed.
The Required Actions are modified by a Note stating that the Required Actions do not apply if the sole reason for the block valve being declared inoperable is as a result of power being removed to comply with other Required Actions. In this event, the Required Actions for inoperable PORV(s) (which require the block valve power to be removed once it is closed) are adequate to address the condition.
(continued)

Pressurizer PORVs B 3.4.11 BASES ACTIONS      F.1, F.2, F.3, and F.4 (continued)
If more than one PORV block valve is inoperable, it is necessary to either restore the block valves within the Completion Time of 1 hour, or place the associated PORVs in manual control and restore at least one block valve within 2 hours and restore the remaining block valve within 72 hours. The PORV control switch has three positions; open, close and auto. Placing the PORV in manual control, if required in ACTION F, is accomplished by positioning the switch out of the auto control mode. The Completion Times are reasonable, based on the small potential for challenges to the system during this time and provide the operator time to correct the situation. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
If the inoperable block valve is associated with the non-Class I PORV, the block valve may be closed and the power removed. The 72 hour Completion Time for closing the block valve is the same time used in Required Action F.3. This recognizes that some restoration work may be required since the block valve is inoperable. Restoration of the non-class I PORV block valve to OPERABLE status is not required because the non-Class I PORV is not required to be available, although having the valve closed impairs the load rejection design capability.
Therefore, once the block valve has been closed per Required Action F.4, Completion Time requirements of Condition G do not apply.
If the block valve can not be placed in the closed position per Required Action F.4, Condition G applies until the block valve is restored or closed.
The required Actions are modified by a Note stating that the Required Actions do not apply if the sole reason for the block valve being declared inoperable is as a result of power being removed to comply with other Required Actions. In this event, the Required Actions for inoperable PORV(s) (which require the block valve power to be removed once it is closed) are adequate to address the condition.
(continued)

ECCS - Operating B 3.5.2 BASES (continued)
ACTIONS          A.1 With one or more trains inoperable and at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available (capable of injection into the RCS, if actuated), the inoperable components must be returned to OPERABLE status within 72 hours or in accordance with the Risk Informed Completion Time Program. The 72 hour Completion Time is based on an NRC reliability evaluation (Ref. 5) and is a reasonable time for repair of many ECCS components.
An ECCS train is inoperable if it is not capable of delivering design flow to the RCS. Individual components are inoperable if they are not capable of performing their safety function or supporting systems are not available.
The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. The intent of this Condition is to maintain a combination of equipment such that 100% of the ECCS flow equivalent to a single OPERABLE ECCS train remains available. (i.e. minimum of one OPERABLE CCP, SI, and RHR pump and applicable flow paths capable of drawing from the RWST and injecting into the RCS cold legs). This allows increased flexibility in plant operations under circumstances when components in opposite trains are inoperable.
The intent of this Condition, to maintain a combination of equipment such that 100% of the ECCS flow equivalent to a single OPERABLE ECCS train remains available, applies to both the injection mode and the recirculation mode.
An event accompanied by a loss of offsite power and the failure of an EDG can disable one ECCS train until power is restored. A reliability analysis (Ref. 5) has shown that the impact of having one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours.
Reference 6 describes situations in which one component, such as an RHR cross-tie valve can disable both ECCS trains. With one or more component(s) inoperable such that 100% of the flow equivalent to a (continued)

ECCS - Operating B 3.5.2 BASES ACTIONS A.1 (continued) single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered.
Opening the containment recirculation sump strainer system access ports, or lower plenum drain valve (SI-1-294 for Unit 1 or SI-2-295 for Unit 2) without pipe cap or inlet strainer (STR-440) installed in MODES 1 through 3 is considered to be a condition which is outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered.
A.2.1, A.2.2, and A.2.3 These Required Actions allow restoring one inoperable ECCS train with no more than one inoperable subsystem to OPERABLE status with a CT of 14 days if it is determined that only one subsystem in one ECCS train is inoperable and that the OPERABLE subsystem is not inoperable due to common cause failure. The common cause failure investigation shall be associated with the subsystem failure that prompts the ECCS subsystem to be declared inoperable originally.
The common cause failure evaluation can be performed by analyses, inspection, and/or testing. The addition of these Required Actions into this TS was per LA 202 for Unit 1 and LA 203 for Unit 2. The 14-day CT is intended to be used for unplanned corrective maintenance or inspections.
The justification to extend the CT to 14 days is based on risk-informed insight where the evaluation would meet the NRC risk informed criteria assuming only one subsystem in one ECCS train is inoperable and with the elimination of conditional failure probability of the redundant ECCS subsystem due to common cause failure. PRA analysis assumes no more than one subsystem in one ECCS train is inoperable. The PRA risk-insignificance thresholds are not met for the14-day Completion Time when a RHR subsystem component is found to be inoperable as a result of a higher conditional failure probability of the redundant component due to common cause failure. To comply with the assumption in the PRA analysis that only one subsystem in one ECCS train is inoperable and to eliminate the common cause failure concerns, the 14-day Completion Time assumes that actions are to be taken within 72 hours to determine that there is only one subsystem in one ECCS train inoperable and there is no common cause failure in the same subsystem in the OPERABLE ECCS train.
(continued)

ECCS - Operating B 3.5.2 BASES ACTIONS      A.2.1, A.2.2, and A.2.3 (continued)
The 72-hour Completion Time in Required Actions A.2.1 and A.2.2 are reasonable and is chosen so that the risk is no worse than the risk associated with the 72 hour Completion Time for Required Action A.1.
The Completion Time is modified by a Note stating that the Required Action A.1 Completion Time is to be used for planned maintenance or inspections. The Completion Times of Required Actions A.2.1, A.2.2, and A.2.3 are for unplanned corrective maintenance or inspections.
This is to prevent accumulating excessive Maintenance Rule unavailability hours.
B.1 and B.2 If the inoperable trains cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours and MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Valve position is the concern and not indicated position in the control room. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves in position by removal of power ensures that they cannot change position as a result of an active failure or be inadvertently misaligned.
The surveillance can be satisfied using indicated position in the control room but may also be satisfied using local observation. These valves are of the type, described in References 6 and 7, that can disable the function of both ECCS trains and invalidate the accident analyses. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. As noted in LCO Note 1, both SI pump flow paths may each be isolated for two hours in MODE 3 by closure of one or more of these valves to perform pressure isolation valve testing.
In addition to the valves listed in SR 3.5.2.1, there are other ECCS related valves that must be appropriately positioned. Improper valve position can affect the ECCS performance required to meet the analysis assumptions. These valves are identified in plant documents and are listed in the following table.
(continued)

Containment Air Locks B 3.6.2 BASES ACTIONS B.1, B.2, and B.3 (continued)
Required Action B.3 is modified by a Note that applies to air lock doors located in high radiation areas and allows these doors to be verified locked closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of the door, once it has been verified to be in the proper position, is small.
C.1, C.2, and C.3 With one or more air locks inoperable for reasons other than those described in Condition A or B, Required Action C.1 requires action to be initiated immediately to evaluate previous combined leakage rates using current air lock test results. An evaluation is acceptable, since it is overly conservative to immediately declare the containment inoperable if both doors in an air lock have failed a seal test or if the overall air lock leakage is not within limits. In many instances (e.g.,
only one seal per door has failed), containment remains OPERABLE, yet only 1 hour (per LCO 3.6.1) would be provided to restore the air lock door to OPERABLE status prior to requiring a plant shutdown. In addition, even with both doors failing the seal test, the overall containment leakage rate can still be within limits.
Required Action C.2 requires that one door in the affected containment air lock must be verified to be closed within the 1 hour Completion Time. This specified time period is consistent with the ACTIONS of LCO 3.6.1, which requires that containment be restored to OPERABLE status within 1 hour.
Additionally, the affected air lock(s) must be restored to OPERABLE status within the 24 hour Completion Time or in accordance with the Risk Informed Completion Time Program. The specified time period is considered reasonable for restoring an inoperable air lock to OPERABLE status, assuming that at least one door is maintained closed in each affected air lock.
D.1 and D.2 If the inoperable containment air lock cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
(continued)

Containment Isolation Valves B 3.6.3 BASES ACTIONS A.1 and A.2 (continued) cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic isolation valve, a closed manual valve (this includes power operated valves with power removed), a blind flange, and a check valve with flow through the valve secured. For a penetration flow path isolated in accordance with Required Action A.1, the device used to isolate the penetration should be the closest available one to containment.
Required Action A.1 must be completed within 4 hours or in accordance with the Risk Informed Completion Time Program. The 4 hour Completion Time is reasonable, considering the time required to isolate the penetration and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3, and 4.
For affected penetration flow paths that cannot be restored to OPERABLE status within the 4 hour Completion Time and that have been isolated in accordance with Required Action A.1, the affected penetration flow paths must be verified to be isolated on a periodic basis. This is necessary to ensure that containment penetrations required to be isolated following an accident and no longer capable of being automatically isolated will be in the isolation position should an event occur. This Required Action does not require any testing or device manipulation. Rather, it involves verification through a system walkdown, which may include the use of local or remote indicators, that those isolation devices outside containment and capable of being mispositioned are in the correct position. The Completion Time of "once per 31 days following isolation for isolation devices outside containment" is appropriate considering the fact that the devices are operated under administrative controls and the probability of their misalignment is low. For the isolation devices inside containment, the time period specified as "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of the inaccessibility of the isolation devices and other administrative controls that will ensure that isolation device misalignment is an unlikely possibility.
Condition A has been modified by a Note indicating that this Condition is only applicable to those penetration flow paths with two containment isolation valves. For penetration flow paths with only one containment isolation valve and a closed system, Condition C provides the appropriate actions.
(continued)

Containment Isolation Valves B 3.6.3 BASES ACTIONS       C.1 and C.2 (continued)
With one or more penetration flow paths requiring isolation following a DBA with one containment isolation valve inoperable, the inoperable valve flow path must be restored to OPERABLE status or the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve (this includes power operated valves with power removed), and a blind flange. A check valve may not be used to isolate the affected penetration flow path. Required Action C.1 must be completed within the 72 hour Completion Time or in accordance with the Risk Informed Completion Time Program. The specified time period is reasonable considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of maintaining containment integrity during MODES 1, 2, 3, and 4 (refer to UFSAR Table 6.2-39, GDC-57, 1971 valves). In the event the affected penetration flow path is isolated in accordance with Required Action C.1, the affected penetration flow path must be verified to be isolated on a periodic basis. This periodic verification is necessary to assure leak tightness of containment and that containment penetrations requiring isolation following an accident are isolated. The Completion Time of once per 31 days following isolation for verifying that each affected penetration flow path is isolated is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low.
Condition C is modified by a Note indicating that this Condition is only applicable to those penetration flow paths with only one containment isolation valve and a closed system. The closed system must meet the requirements of Reference 3. This Note is necessary since this Condition is written to specifically address those penetration flow paths in a closed system.
Required Action C.2 is modified by two Notes. Note 1 applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Note 2 applies to isolation devices that are locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned. Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is small.
(continued)

Containment Spray and Cooling Systems B 3.6.6 BASES (continued)
APPLICABILITY    In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature requiring the operation of the containment spray trains and CFCUs.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Thus, the Containment Spray System and the Containment Cooling System are not required to be OPERABLE in MODES 5 and 6.
ACTIONS          A.1 With one containment spray train inoperable, the inoperable containment spray train must be restored to OPERABLE status within 72 hours or in accordance with the Risk Informed Completion Time Program. In this Condition, the remaining OPERABLE spray and cooling trains are adequate to perform the iodine removal and containment cooling functions. The 72 hour Completion Time takes into account the redundant heat removal capability afforded by the Containment Spray System, reasonable time for repairs, and low probability of a DBA occurring during this period.
The Completion Time is modified by a Note stating that for planned maintenance or inspections, the Completion time is 72 hours. The Completion Times of Required Action A.2 are for unplanned corrective maintenance or inspections.
A.2 With one containment spray train inoperable, the inoperable containment spray train must be restored to OPERABLE status within 14 days. This Required Action applies to unplanned corrective maintenance or inspections. In this Condition, the remaining OPERABLE spray and cooling trains are adequate to perform the iodine removal and containment cooling functions. The 14-day Completion Time is based on PRA analysis and has taken into account the redundant heat removal capability afforded by the Containment Spray System, reasonable time for repairs, and low probability of a DBA occurring during this period.
These Required Action and Completion Time were added to the TS by LA 202 for Unit 1 and LA 203 for Unit 2. The 14-day Completion Time is intended to be used for unplanned corrective maintenance or inspections.
(continued)

Containment Spray and Cooling Systems B 3.6.6 BASES ACTIONS      B.1 and B.2 (continued)
If the inoperable containment spray train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which overall plant risk is reduced. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 54 hours.
Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 9). In MODE 4 the Steam Generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 9, the steam turbine driven Auxiliary Feedwater Pump must be available to remain in MODE 4. Should Steam Generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.
Required Action B.2 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 4. This Note prohibits the use of LCO 3.0.4.a to enter MODE 4 during startup with the LCO not met.
However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 4, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. The extended interval to reach MODE 4 allows 48 hours to restore the containment spray train to OPERABLE status in MODE 3. This is reasonable when considering the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.
(continued)

Containment Spray and Cooling Systems B 3.6.6 BASES ACTIONS      C.1 (continued)
With one CFCU system inoperable such that a minimum of two CFCUs remain operable, restore the required CFCUs to OPERABLE status within 7 days or in accordance with the Risk Informed Completion Time Program. The components in this degraded condition are capable of providing at least 100% of the heat removal needs. The 7 day Completion Time was developed taking into account the redundant heat removal capabilities afforded by combinations of the Containment Spray System and Containment Cooling System and the low probability of DBA occurring during this period.
D.1 and D.2 With one train of containment spray inoperable and the CFCUs system inoperable such that a minimum of two CFCUs remain OPERABLE, restore one required train of containment spray or CFCU system to OPERABLE status within 72 hours or in accordance with the Risk Informed Completion Time Program. The components remaining in OPERABLE status in this degraded condition provide iodine removal capabilities and are capable of providing at least 100% of the heat removal needs after an accident. The 72 hour Completion Time was developed taking into account the redundant heat removal capabilities afforded by combinations of the Containment Spray System and Containment Cooling System, the iodine removal function of the Containment Spray System, and the low probability of DBA occurring during this period.
E.1 and E.2 If the Required Action and associated Completion Time of Condition C or D of this LCO are not met, the plant must be brought to a MODE in which overall plant risk is reduced. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours.
Remaining within the Applicability of the LCO is acceptable to accomplish short duration repairs to restore inoperable equipment because the plant risk in MODE 4 is similar to or lower than MODE 5 (Ref. 9). In MODE 4 the Steam Generators and Residual Heat Removal System are available to remove decay heat, which provides diversity and defense in depth. As stated in Reference 9, the steam turbine driven Auxiliary Feedwater Pump must be available to remain in MODE 4. Should Steam Generator cooling be lost while relying on this Required Action, there are preplanned actions to ensure long-term decay heat removal. Voluntary entry into MODE 5 may be made as it is also acceptable from a risk perspective.
(continued)

MSIVs B 3.7.2 BASES APPLICABLE    c. A break downstream of the MSIVs will be isolated by the closure of SAFETY             the MSIVs.
ANALYSES
: d. Following a steam generator tube rupture, closure of the MSIVs (continued) isolates the ruptured steam generator from the intact steam generators to minimize radiological releases.
: e. The MSIVs are also utilized during other events such as a feedwater line break. This event is less limiting so far as MSIV OPERABILITY is concerned.
The MSIVs satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO           This LCO requires that four MSIVs in the steam lines be OPERABLE.
The MSIVs are considered OPERABLE when the isolation times are within limits, and they close on an isolation actuation signal.
This LCO provides assurance that the MSIVs will perform their design safety function to mitigate the consequences of accidents that could result in offsite exposures comparable to the 10 CFR 50.67 (Ref. 4) limits or the NRC staff approved licensing basis.
APPLICABILITY The MSIVs must be OPERABLE in MODE 1, and in MODES 2 and 3 except when closed and de-activated (vented or prevented from opening), when there is significant mass and energy in the RCS and steam generators. When the MSIVs are closed, they are already performing the safety function.
In MODE 4, the steam generator energy is low, thus OPERABILITY in MODE 4 is not required.
In MODE 5 or 6, the steam generators do not contain much energy because their temperature is below the boiling point of water; therefore, the MSIVs are not required for isolation of potential high energy secondary system pipe breaks in these MODES.
ACTIONS       A.1 With one MSIV inoperable in MODE 1, action must be taken to restore OPERABLE status within 8 hours or in accordance with the Risk Informed Completion Time Program. Some repairs to the MSIV can be made with the unit hot. The 8 hour Completion Time is reasonable, considering the low probability of an accident occurring during this time period that would require a closure of the MSIVs.
The 8 hour Completion Time is greater than that normally allowed for containment isolation valves because the MSIVs are valves that isolate a closed system penetrating containment. These valves differ from other containment isolation valves in that the closed system provides an additional means for containment isolation.
(continued)

ADVs B 3.7.4 BASES (continued)
APPLICABILITY    In MODES 1, 2, and 3, all four ADVs are required to be OPERABLE. In MODE 4, only the ADVs associated with the steam generators being relied upon for heat removal, are required to be OPERABLE.
In MODE 5 or 6, an SGTR is not a credible event.
ACTIONS          A.1 With one required ADV line inoperable, action must be taken to restore OPERABLE status within 7 days or in accordance with the Risk Informed Completion Time Program. The 7 day Completion Time allows for the redundant capability afforded by the remaining OPERABLE ADV lines, a non-safety grade backup in the Steam Bypass System, and MSSVs and is based on a PRA analysis and the low probability of a SGTR and LOOP event occurring during this period that would require the ADV lines.
B.1 With two ADV lines inoperable, action must be taken to restore at least one ADV line to OPERABLE status. This will result in at least three operable ADVs. Since the block valve can be closed to isolate an ADV, some repairs may be possible with the unit at power. The 72 hour Completion Time is reasonable to repair inoperable ADV lines, based on the availability of the Steam Dump System (40% steam dump valves to the condenser) and MSSVs, and the low probability of an event occurring during this period that would require the ADV lines.
C.1 With three or more ADV lines inoperable, action must be taken to restore at least two ADV lines to OPERABLE status. This will result in at least two operable ADVs. Since the block valve can be closed to isolate an ADV, some repairs may be possible with the unit at power.
The 24 hour Completion Time is reasonable to repair inoperable ADV lines, based on the availability of the Steam Dump System (40% steam dump valves to the condenser) and MSSVs, and the low probability of an event occurring during this period that would require the ADV lines.
D.1 and D.2 If the ADV lines cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4, without reliance upon steam generator for heat removal, within 18 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
(continued)

AFW System B 3.7.5 BASES LCO           each powered by a separated Class 1E bus, be OPERABLE in two (continued)  diverse paths, each supplying AFW to separate steam generators. The turbine driven AFW pump is required to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the MSIVs, and shall be capable of supplying AFW to any of the steam generators. The piping, valves, instrumentation, and controls in the required flow paths also are required to be OPERABLE.
The LCO is modified by a Note indicating that one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4. This is because of the reduced heat removal requirements and short period of time in MODE 4 during which the AFW is required and the insufficient steam available in MODE 4 to power the turbine driven AFW pump.
APPLICABILITY In MODES 1, 2, and 3, the AFW System is required to be OPERABLE in the event that it is called upon to function when the MFW is lost. In addition, the AFW System is required to supply enough makeup water to replace the steam generator secondary inventory, lost as the unit cools to MODE 4 conditions.
In MODE 4 the AFW System may be used for heat removal via the steam generators.
In MODE 5 or 6, the steam generators are not normally used for heat removal, and the AFW System is not required.
ACTIONS       A Note prohibits the application of LCO 3.0.4.b to an inoperable AFW train. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an AFW train inoperable and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
A.1 If the turbine driven AFW train is inoperable due to one inoperable steam supply, or if a turbine driven pump is inoperable for any reason while in MODE 3 immediately following refueling, action must be taken to restore OPERABLE status within 7 days or in accordance with the Risk Informed Completion Time Program. The 7 day Completion Time is reasonable, based on the following reasons:
: a. For the inoperability of a steam supply to the turbine driven AFW pump due to one inoperable steam supply, the 7 day Completion Time is reasonable since there is a redundant steam supply line for the turbine driven pump and the turbine driven train is still capable of performing its specified function for most postulated events; (continued)

AFW System B 3.7.5 BASES ACTIONS  A.1 (continued)
: b. For the inoperability of a turbine driven AFW pump while in MODE 3 immediately subsequent to a refueling, the 7 day Completion Time is reasonable due to the minimal decay heat levels in this situation; and
: c. For both the inoperability of a steam supply line to the turbine driven pump due to one inoperable steam supply and an inoperable turbine driven AFW pump while in MODE 3 immediately following a refueling outage, the 7 day Completion Time is reasonable due to the availability of redundant OPERABLE motor driven AFW pumps, and due to the low probability of an event requiring the use of the turbine driven AFW pump.
B.1 With one of the required AFW trains (pump or flow path) inoperable in MODE 1, 2, or 3 for reasons other than Condition A or G, action must be taken to restore OPERABLE status within 72 hours or in accordance with the Risk Informed Completion Time Program. This Condition includes the loss of two steam supply lines to the turbine driven AFW pump. The 72 hour Completion Time is reasonable, based on redundant capabilities afforded by the AFW System, time needed for repairs, and the low probability of a DBA occurring during this time period.
C.1 and C.2 With one of the required motor-driven AFW trains (pump or flow path) inoperable and the turbine-driven AFW train inoperable due to one inoperable steam supply, action must be taken to restore the affected equipment to OPERABLE status within 48 hours or in accordance with the Risk Informed Completion Time Program. Assuming no single active failures when in this condition, the accident (a FWLB or MSLB) could result in the loss of the remaining steam supply to the turbine-driven AFW pump due to the faulted SG.
A note in Condition C limits applicability to only when the remaining OPERABLE motor-driven AFW train provides feedwater to the SG with the inoperable steam supply. This Condition will only apply during the following two scenarios:
: 1)   Motor-driven AFW pump 2 OPERABLE, motor-driven AFW pump 3 inoperable, and steam supply from SG 2 inoperable, or
: 2)   Motor-driven AFW pump 2 inoperable, motor-driven AFW pump 3 OPERABLE, and steam supply from SG 3 inoperable.
(continued)

AFW System B 3.7.5 BASES ACTIONS      C.1 and C.2 (continued)
(continued) This ensures that if a FWLB were to occur affecting the OPERABLE motor driven AFW pump, the turbine-driven AFW pump would still be capable of providing AFW to two intact SGs. If a MSLB were to occur on the SG feeding the remaining OPERABLE steam supply to the turbine-driven AFW pump, the OPERABLE motor-driven AFW pump would still be capable of providing AFW to two intact SGs.
If motor-driven AFW pump 2 and the steam supply from SG 3 are inoperable, or if motor-driven AFW pump 3 and the steam supply from SG 2 are inoperable, then Condition D for two inoperable AFW pumps applies.
The 48 hour Completion Time is reasonable based on the fact that the remaining motor-driven AFW train is capable of providing 100 percent of the AFW flow requirements, and the low probability of an event occurring that would challenge the AFW system.
In MODE 4 with two AFW trains inoperable, operation is allowed to continue because only one motor driven pump AFW train is required in accordance with the Note that modifies the LCO. Although not required, the unit may continue to cool down and initiate RHR.
D.1 and D.2 When Required Action A.1, B.1, C.1, or C.2, G.1, or G.2 cannot be completed within the required Completion Time, or if two AFW trains are inoperable in MODE 1, 2, or 3 for reasons other than Condition C or G, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4 within 18 hours.
The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
In MODE 4 with two AFW trains inoperable, operation is allowed to continue because only one motor driven pump AFW train is required in accordance with the Note that modifies the LCO. Although not required, the unit may continue to cool down and initiate RHR.
(continued)

AFW System B 3.7.5 BASES ACTIONS      G.1 and G.2 (continued) With one or two AFW trains inoperable in MODE 1, 2, or 3 due to inoperable Unit 1 AFW piping affecting the AFW flow path(s) to one steam generator, action must be taken to restore the affected train(s) to OPERABLE status within 7 days. Assuming no single active failures when in this condition, a FWLB could result in the loss of SG cooling in an additional SG due to operator action to isolate flow to the faulted SG.
A note in Condition G limits applicability only to Unit 1 once during Unit 1 Cycle 22 during repair of AFW piping. This condition is not applicable to Unit 2 or to Unit 1 at any other time.
Required Action G.1 to isolate AFW flow path(s) to the affected steam generator is performed by the operator by closing LCV-107 from the turbine driven pump to steam generator 1-2 and/or closing LCV-111 from the motor driven pump to steam generator 1-2 to isolate inoperable AFW piping. The 2 hour Completion time is reasonable based on the time required to complete the Required Action from the control room.
The Required Action G.2 7 day Completion Time to restore AFW train(s) to OPERABLE status is reasonable, based on the capabilities of the two motor driven AFW pumps and one turbine AFW pump to provide adequate AFW cooling flow, the time needed for Unit 1 AFW repairs, and the low probability of a design basis accident occurring during this period.
When Required Action G.1 or G.2 cannot be completed within the required Completion Time, Condition D is entered to place the unit in at least MODE 3 within 6 hours, and in MODE 4 within 18 hours.
(continued)

CCW System B 3.7.7 BASES LCO           A PG&E Design Class I CCW loop is considered OPERABLE when:
(continued)
: a. Two CCW pumps, one CCW heat exchanger, one PG&E Design Class I CCW header and the surge tank are OPERABLE; and
: b. The associated piping, valves, and instrumentation and controls required to perform the PG&E Design Class I function are OPERABLE.
The isolation of CCW from other components or systems may render those components or systems inoperable but does not affect the OPERABILITY of the CCW System, except for isolation of CCW to the CFCUs. Isolation of CCW to the CFCUs could potentially affect the flow balance and requires evaluation to ensure continued operability.
Split loop alignment of the CCW system during normal operation requires Condition A to be entered because the CCW system cannot tolerate a single failure in this configuration.
APPLICABILITY In MODES 1, 2, 3, and 4, the CCW System is a normally operating system, which must be prepared to perform its principal PG&E Design Class I function of removal of accident generated containment heat via the CFCUs and removal of decay heat from the reactor via the Residual Heat Removal (RHR) System.
In MODE 5 or 6, the OPERABILITY requirements of the CCW System are determined by the systems it supports.
ACTIONS       A.1 Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4,"
be entered if an inoperable PG&E Design Class I CCW loop results in an inoperable RHR loop. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.
If one PG&E Design Class I CCW loop is inoperable, action must be taken to restore two PG&E Design Class I CCW loops to OPERABLE status within 72 hours or in accordance with the Risk Informed Completion Time Program. In this Condition, the remaining OPERABLE PG&E Design Class I CCW loop is adequate to perform the heat removal function. The 72 hour Completion Time is reasonable, based on the overall heat transfer capability of ultimate heat sink system, operator action, and the low probability of a DBA occurring during this period.
Split loop alignment of the CCW system during normal operation requires Condition A to be entered because the CCW system cannot tolerate a single failure in this configuration.
(continued)

ASW B 3.7.8 BASES (continued)
ACTIONS          A.1 If one ASW train is inoperable, action must be taken to restore OPERABLE status within 72 hours or in accordance with the Risk Informed Completion Time Program. In this Condition, the remaining OPERABLE ASW train is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE ASW train could result in loss of ASW system function. The Note indicates that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," should be entered if an inoperable ASW train results in an inoperable decay heat removal train. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. The 72 hour Completion Time is based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this time period.
The 72-hour Completion Time is modified by a Note that allows a one-time Completion Time of 144 hours for ASW pump 1-1, for Unit 1 cycle 23 to support emergent replacement of the ASW pump 1-1 motor.
The one-time Completion Time of 144 hours is reasonable considering the capabilities of the other ASW Pump 1-2 to perform the heat removal function, the cross-tie capabilities of ASW from Unit 2, the low probability of a design basis accident occurring during this period, and the one-time use of a 144-hour Completion Time.
(continued)

AC Sources - Operating B 3.8.1 BASES (continued)
APPLICABILITY    The AC sources are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:
: a.       Acceptable fuel design, limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and
: b.       Adequate core cooling is provided and containment OPERABILITY and other PG&E Design Class I functions are maintained in the event of a postulated DBA.
The AC power requirements for MODES 5 and 6 are covered in LCO 3.8.2, "AC Sources - Shutdown."
ACTIONS          A Note prohibits the application of LCO 3.0.4.b to an inoperable DG.
There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable DG and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
A.1 To ensure a highly reliable power source remains with one offsite circuit inoperable, it is necessary to verify the OPERABILITY of the remaining required offsite circuit on a more frequent basis. Since the Required Action only specifies "perform," a failure of SR 3.8.1.1 acceptance criteria does not result in a Required Action not met. However, if a second required circuit fails SR 3.8.1.1, the second offsite circuit is inoperable, and Condition C, for two offsite circuits inoperable, is entered.
A.2 Operation may continue in Condition A for a period that should not exceed 72 hours (Ref. 6). Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. With one offsite circuit inoperable, the reliability of the offsite system is degraded, and the potential for a loss of offsite power is increased, with attendant potential for a challenge to the unit safety systems. In this Condition, however, the remaining OPERABLE offsite circuit and DGs are adequate to supply electrical power to the onsite Class 1E Distribution System.
(continued)

AC Sources - Operating B 3.8.1 BASES ACTIONS      B.3.1 and B.3.2 (continued) Required Action B.3.1 provides an allowance to avoid unnecessary testing of OPERABLE DGs. If it can be determined that the cause of the inoperable DG does not exist on the OPERABLE DGs, SR 3.8.1.2 does not have to be performed. If the cause of inoperability exists on other DGs, the other DGs would be declared inoperable upon discovery and Condition E of LCO 3.8.1 would be entered. Once the failure is repaired, the common cause failure no longer exists, and Required Action B.3.1 is satisfied. If the cause of the initial inoperable DG cannot be confirmed not to exist on the remaining DGs, performance of SR 3.8.1.2 suffices to provide assurance of continued OPERABILITY of those DGs. If a DG has already started and loaded on a bus, it is not necessary to shutdown the DG and perform SR 3.8.1.2. The DG is verified OPERABLE since it is performing its intended function.
In the event the inoperable DG is restored to OPERABLE status prior to completing either B.3.1 or B.3.2, the plant corrective action program will continue to evaluate the common cause possibility. This continued evaluation, however, is no longer under the 24 hour constraint imposed while in Condition B.
According to Generic Letter 84-15 (Ref. 7), 24 hours is reasonable to confirm that the OPERABLE DGs are not affected by the same problem as the inoperable DG.
B.4 Operation may continue in Condition B for a period that should not exceed 14 days. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. This Completion Time was revised from 72 hours to 7 days by License Amendment (LA) 44 for Unit 1 and LA 43 for Unit 2 and from 7 days to 14 days by LA 166 for Unit 1 and LA 167 for Unit 2. In accordance with Reference 17, the 14-day Completion Time is intended to be used for planned maintenance or inspections at a frequency of no more than once per DG per operating cycle for each DG. For all other DG maintenance or inspections, the Completion Time is expected to remain at 7 days. This is consistent with the Completion Times assumed in References 17 and 18.
In Condition B, the remaining OPERABLE DGs and offsite circuits are adequate to supply electrical power to the onsite Class 1E Distribution System. The 14 day Completion Time takes into account the capacity and capability of the remaining AC sources, a reasonable time for repairs, and the low probability of a DBA occurring during this period.
(continued)

AC Sources - Operating B 3.8.1 BASES ACTIONS B.4 (continued)
As in Required Action B.2, the Completion Time allows for an exception to the normal "time zero" for beginning the allowed time "clock." This will result in establishing the "time zero" at the time that the LCO was initially not met, instead of at the time Condition B was entered.
C.1 and C.2 Required Action C.1, which applies when two offsite circuits are inoperable, is intended to provide assurance that an event with a coincident single failure will not result in a complete loss of redundant required safety functions. The rationale for the reduction to 12 hours for Required Action C.1 is provided in Reference 6, which supports a Completion Time of 24 hours for two required offsite circuits inoperable, based upon the assumption that two complete safety trains are OPERABLE. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. When a concurrent redundant required feature failure exists, this assumption is not valid, and a shorter Completion Time of 12 hours is appropriate.
Required features are redundant PG&E Design Class I systems, subsystems, trains, components, and devices that depend on the DGs as a source of emergency power. These features are powered from the three Class 1E AC electrical power distribution subsystems.
Examples of required features would include, but are not limited to, auxiliary saltwater pumps, centrifugal charging pumps, or motor-driven auxiliary feedwater pumps.
The Completion Time for Required Action C.1 is intended to allow the operator time to evaluate and repair any discovered inoperabilities.
This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." In this Required Action the Completion Time only begins on discovery that both:
: a.        All required offsite circuits are inoperable; and
: b.        A required feature is inoperable.
If at any time during the existence of Condition C (two offsite circuits inoperable) a required feature becomes inoperable, this Completion Time begins to be tracked.
(continued)

AC Sources - Operating B 3.8.1 BASES ACTIONS C.1 and C.2 (continued)
Operation may continue in Condition C for a period that should not exceed 24 hours (Ref. 6). This level of degradation means that the offsite electrical power system does not have the capability to effect a safe shutdown and to mitigate the effects of an accident; however, the onsite AC sources have not been degraded. This level of degradation generally corresponds to a total loss of the immediately accessible offsite power sources.
Because of the normally high availability of the offsite sources, this level of degradation may appear to be more severe than other combinations of two AC sources inoperable that involve one or more DGs inoperable. However, two factors tend to decrease the severity of this level of degradation:
: a.      The configuration of the Class 1E AC electrical power system that remains available is not susceptible to a single bus or switching failure; and
: b.      The time required to detect and restore an unavailable offsite power source is generally much less than that required to detect and restore an unavailable onsite AC source.
With both of the required offsite circuits inoperable, sufficient onsite AC sources are available to maintain the unit in a safe shutdown condition in the event of a DBA or transient. In fact, a simultaneous loss of offsite AC sources, a DBA, and a worst case single failure were postulated as a part of the design basis in the safety analysis. Thus, the 24 hour Completion Time provides a period of time to effect restoration of one of the offsite circuits commensurate with the importance of maintaining an AC electrical power system capable of meeting its design criteria.
According to Reference 6, with the available offsite AC sources, two less than required by the LCO, operation may continue for 24 hours. If two offsite sources are restored within 24 hours, unrestricted operation may continue. If only one offsite source is restored within 24 hours, power operation continues in accordance with Condition A.
D.1 and D.2 Operation may continue in Condition D for a period that should not exceed 12 hours (Ref. 6). Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
In Condition D, individual redundancy is lost in the offsite electrical power system and may be lost in the onsite AC electrical power system. Since power system redundancy is provided by two diverse sources of power, however, the reliability of the power systems in this (continued)

AC Sources - Operating B 3.8.1 BASES ACTIONS D.1 and D.2 (continued)
Condition may appear higher than that in Condition C (loss of both required offsite circuits). This difference in reliability is offset by the susceptibility of this power system configuration to a single bus or switching failure. The 12 hour Completion Time takes into account the capacity and capability of the remaining AC sources, a reasonable time for repairs, and the low probability of a DBA occurring during this period. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
E.1 With two or more DGs inoperable, the remaining onsite AC sources are inadequate. Thus, with an assumed loss of offsite electrical power, insufficient standby AC sources are available to power the minimum required ESF functions. Since the offsite electrical power system may be the only source of AC power for this level of degradation, the risk associated with continued operation for a very short time could be less than that associated with an immediate controlled shutdown (the immediate shutdown could cause grid instability, which could result in a total loss of AC power). Since any inadvertent generator trip could also result in a total loss of offsite AC power, the time allowed for continued operation is severely restricted. The intent here is to avoid the risk associated with an immediate controlled shutdown and to minimize the risk associated with this level of degradation.
According to Reference 6, with two or more DGS inoperable, operation may continue for a period that should not exceed 2 hours.
F.1 Condition F corresponds to a level of degradation in which one train of the DFO transfer system is inoperable. The onsite AC electrical power systems are redundant and available to support ESF loads. However, one subsystem required for the onsite AC electrical system operability has lost its redundancy (DFO supply to the DGs).
The 72 hour Completion Time takes into account the capacity and capability of the remaining AC sources, a reasonable time for repairs, and the low probability of a DBA occurring during this period.
Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
(continued)

AC Sources - Operating B 3.8.1 BASES ACTIONS F.1 (continued)
The 72 hour Completion Time is modified by a Note that allows a one-time use only Completion Time of 7 days during the planned maintenance of each DFO transfer system pump 0-1 and 0-2 during the year 2022 with the portable DFO transfer pump staged and available.
The one-time Completion Time of 7 days can only be used once for separate maintenance of each DFO transfer system pump 0-1 and 0-2.
The one-time Completion Time of 7 days is reasonable considering the additional requirement to have the portable DFO transfer pump staged and available during the one-time Completion Time for each DFO transfer system supply train to offset additional risk that is incurred, based on the risk insights obtained from the DCPP Probabilistic Risk Assessment model.
G.1 With both trains of DFO inoperable, the onsite AC sources are inadequate (loss of DFO supply to all DGs). With an assumed loss of offsite electrical power, insufficient AC sources are available to power the minimum required ESF functions. Since the offsite electrical power system is the only source for AC power for this level of degradation, the risk associated with continued operation for a very short time could be less than that associated with an immediate controlled shutdown (the immediate shutdown could cause grid instability, which could result in a total loss of AC power). Since any inadvertent generator trip could also result in a total loss of offsite AC power, the time allowed for continued operation is severely restricted. The intent here is to avoid the risk associated with an immediate controlled shutdown and to minimize the risk associated with this level of degradation.
(continued)

DC Sources - Operating B 3.8.4 BASES ACTIONS      A discharged battery having terminal voltage of at least the minimum (continued) established float voltage indicates that the battery is on the exponential charging current portion (the second part) of its recharge cycle. The time to return a battery to its fully charged state under this condition is simply a function of the amount of the previous discharge and the recharge characteristic of the battery. Thus there is good assurance of fully recharging the battery within 12 hours, avoiding a premature shutdown with its own attendant risk.
If established battery terminal float voltage cannot be restored to greater than or equal to the minimum established float voltage within 2 hours, and the charger is not operating in the current-limiting mode, a faulty charger is indicated. A faulty charger that is incapable of maintaining established battery terminal float voltage does not provide assurance that it can revert to and operate properly in the current limit mode that is necessary during the recovery period following a battery discharge event that the DC system is designed for.
If the charger is operating in the current limit mode after 2 hours that is an indication that the battery is partially discharged and its capacity margins will be reduced. The time to return the battery to its fully charged condition in this case is a function of the battery charger capacity, the amount of loads on the associated DC system, the amount of the previous discharge, and the recharge characteristic of the battery. The charge time can be extensive, and there is not adequate assurance that it can be recharged within 12 hours (Required Action A.2).
Required Action A.2 requires that the battery float current be verified as less than or equal to 2 amps. This indicates that, if the battery had been discharged as the result of the inoperable battery charger, it has now been fully recharged. If at the expiration of the 12 hour period the battery float current is not less than or equal to 2 amps this indicates there may be additional battery problems and the battery must be declared inoperable in accordance with LCO 3.8.6 Required Action B.2.
Required Action A.3 limits the restoration time for the inoperable dedicated battery charger to 14 days. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. This action is applicable if an alternate means of restoring battery terminal voltage to greater than or equal to the minimum established float voltage has been used (e.g., backup charger or non-Class 1E battery charger). The 14 day completion time reflects a reasonable time to effect restoration of the dedicated battery charger to operable status.
(continued)

DC Sources - Operating B 3.8.4 BASES ACTIONS      B.1 (continued)
Condition B represents one DC electrical power subsystem with one battery inoperable. With one battery inoperable, the DC bus is being supplied by the associated OPERABLE battery charger. Any event that results in a loss of the associated 480-V Class 1E bus supporting the normal battery charger will also result in loss of or degraded DC to the associated DC electrical power subsystem. Recovery of the 480-V Class 1E bus, especially if it is due to a loss of offsite power, will be hampered by the fact that many of the components necessary for the recovery (e.g., diesel generator control and field flash, AC load shed and diesel generator output circuit breakers, etc.) likely rely upon the battery. In addition, the energization transients of any DC loads that are beyond the capability of the battery charger and normally require the assistance of the battery will not be able to be brought online. The 2 hour limit allows sufficient time to effect restoration of an inoperable battery given that the majority of the conditions that lead to battery inoperability (e.g., loss of battery charger, battery cell voltage less than 2.07 V, etc.) are identified in Specifications 3.8.4, 3.8.5, and 3.8.6 together with additional specific completion times. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
B.2.1.1, B.2.1.2, B.2.2 The completion time for restoring the inoperable battery to OPERABLE status can be extended to 4 hours, on a one-time basis for Unit 1 Class 1E Battery 1-1 for Unit 1 cycle 14, if additional Required Actions are taken. The 4 hour completion time is based upon Probabilistic Risk Assessment (PRA) calculation of risk given one battery is inoperable.
This PRA assessment makes the assumptions that actions are taken to either determine that the OPERABLE batteries are not inoperable due to common cause failure or SR 3.8.4.1 and SR 3.8.6.1 are performed for the OPERABLE batteries. Taking steps to determine whether the battery condition is a result of a common cause failure will provide assurance that a similar failure will not occur to other OPERABLE batteries. Performing SR 3.8.4.1 and SR 3.8.6.1 will serve the same purpose of ensuring the OPERABLE batteries remain in OPERABLE condition. The 2 hour completion times for Required Actions B.2.1.1, and B.2.1.2 are consistent with completion time to restore a battery to OPERABLE status in Required Action B.1. When Required Actions B.2.1.1 or B.2.1.2 are met, then the inoperable battery can be restored to OPERABLE status in 4 hours.
(continued)

DC Sources - Operating B 3.8.4 BASES ACTIONS      C.1 (continued)
Condition C represents one Class 1E DC electrical power subsystem and associated ESF equipment with a loss of ability to completely respond to an event, and a potential loss of ability to remain energized during normal operation. It is, therefore, imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for complete loss of DC power to the affected subsystem. The 2 hour limit is consistent with the allowed time for an inoperable DC distribution subsystem. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
If one of the required DC electrical power subsystems is inoperable for reasons other than Condition A or B (e.g., inoperable battery charger and associated inoperable battery), the remaining DC electrical power subsystems have the capacity to support a safe shutdown and to mitigate an accident condition. Since a subsequent worst case single failure could, however, result in the loss of the minimum necessary DC electrical power subsystems to mitigate a worst case accident, continued power operation should not exceed 2 hours. The 2 hour Completion Time reflects a reasonable time to assess unit status as a function of the inoperable DC electrical power subsystem and, if the DC electrical power subsystem is not restored to OPERABLE status, to prepare to effect an orderly and safe unit shutdown (Ref. 7).
D.1 The design of the 125-Vdc electrical power distribution system is such that a battery can have associated with it a dedicated full capacity charger powered from its associated 480-V Class 1E bus or a backup full capacity charger powered from another 480-V Class 1E bus. Use of the backup full capacity charger results in more than one full capacity charger receiving power simultaneously from a single 480-V Class 1E bus and causes the requirements of independence and redundancy between subsystems to no longer be maintained. Thus, operation with two chargers powered by the same vital bus is limited to 14 days.
(continued)

Inverters - Operating B 3.8.7 BASES LCO           Maintaining the required inverters OPERABLE ensures that the (continued)  redundancy incorporated into the design of the RPS and ESFAS instrumentation and controls is maintained. The four inverters ensure an uninterruptible supply of AC electrical power to the 120-Vac Class 1E buses even if the 4.16-kV safety buses are de-energized.
Operable inverters require the associated 120-Vac Class 1E bus to be powered by the inverter with output voltage within tolerances, and power input to the inverter from a 125-Vdc station battery. Alternatively, power supply may be from an internal AC source via rectifier as long as the station battery is available as the uninterruptible power supply.
APPLICABILITY The inverters are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:
: a.       Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and
: b.       Adequate core cooling is provided, and containment OPERABILITY and other PG&E Design Class I functions are maintained in the event of a postulated DBA.
Inverter requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.8, "Inverters - Shutdown."
ACTIONS       A.1 With a required inverter inoperable, its associated 120-Vac Class 1E bus becomes inoperable until it is re-energized from its Class 1E constant voltage source transformer.
For this reason a Note has been included in Condition A requiring the entry into the Conditions and Required Actions of LCO 3.8.9, "Distribution Systems - Operating." This ensures that the 120-Vac bus is re-energized within 2 hours.
Required Action A.1 allows 24 hours to fix the inoperable inverter and return it to service. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. The 24 hour limit is based upon engineering judgment, taking into consideration the time required to repair an inverter and the additional risk to which the unit is exposed because of the inverter inoperability. This has to be balanced against the risk of an immediate shutdown, along with the potential challenges to safety systems such a shutdown might entail.
When the 120-Vac Class 1E bus is powered from its constant voltage source, it is relying upon interruptible (continued)

Distribution Systems - Operating B 3.8.9 BASES LCO           In addition, tie breakers between redundant Class 1E AC, DC, and 120-(continued)  Vac bus power distribution subsystems, if they exist, must be open.
This prevents any electrical malfunction in any power distribution subsystem from propagating to the redundant subsystem, that could cause the failure of a redundant subsystem and a loss of essential safety function(s). If any tie breakers are closed, the affected redundant electrical power distribution subsystems are considered inoperable. This applies to the onsite, Class 1E redundant electrical power distribution subsystems. It does not, however, preclude redundant Class 1E 4.16-kV buses from being powered from the same offsite circuit.
APPLICABILITY The electrical power distribution subsystems are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:
: a.       Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and
: b.       Adequate core cooling is provided, and containment OPERABILITY and other PG&E Design Class I functions are maintained in the event of a postulated DBA.
Electrical power distribution subsystem requirements for MODES 5 and 6 are covered in the Bases for LCO 3.8.10, "Distribution Systems -
Shutdown."
ACTIONS       A.1 With one required Class 1E AC electrical power subsystem inoperable, the remaining portions of the AC electrical power distribution subsystems are capable of supporting the minimum safety functions necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in the remaining portions of the power distribution subsystems could result in the minimum required ESF functions not being supported. Therefore, the required Class 1E AC buses, load centers, and motor control centers must be restored to OPERABLE status within 8 hours or in accordance with the Risk Informed Completion Time Program.
Condition A worst scenario is one AC electrical power distribution subsystem without AC power (i.e., no offsite power to the 4.16-kV ESF bus and the associated DG inoperable). In this Condition, the unit is more vulnerable to a complete loss of AC power. It is, therefore, imperative that the unit operator's attention be focused on minimizing the potential for loss of power to the remaining AC electrical power (continued)

Distribution Systems - Operating B 3.8.9 BASES ACTIONS B.1 (continued) constant voltage transformer. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program. The required AC Class 1E bus subsystems must then be re-powered by restoring it's associated inverter to OPERABLE status within 24 hours under LCO 3.8.7. ACTION A.1.
Condition B represents one 120-Vac Class 1E bus without power; potentially both the DC source and the associated AC source are nonfunctioning. In this situation, the unit is significantly more vulnerable to a complete loss of all noninterruptible power. It is, therefore, imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for loss of power to the remaining Class 1E buses and restoring power to the affected 120-Vac Class 1E bus subsystem.
This 2 hour limit is more conservative than Completion Times allowed for the vast majority of components that are without adequate 120-Vac power. Taking exception to LCO 3.0.2 for components without adequate Class 1E 120-Vac power, that would have the Required Action Completion Times shorter than 2 hours if declared inoperable, is acceptable because of:
: a.      The potential for decreased safety by requiring a change in unit conditions (i.e., requiring a shutdown) and not allowing stable operations to continue;
: b.      The potential for decreased safety by requiring entry into numerous Applicable Conditions and Required Actions for components without adequate Class 1E 120-Vac power and not providing sufficient time for the operators to perform the necessary evaluations and actions for restoring power to the affected subsystem; and
: c.      The potential for an event in conjunction with a single failure of a redundant component.
The 2 hour Completion Time takes into account the importance to safety of restoring the 120-Vac Class 1E bus to OPERABLE status, the redundant capability afforded by the other OPERABLE 120-Vac Class 1E buses, and the low probability of a DBA occurring during this period.
(continued)

Distribution Systems - Operating B 3.8.9 BASES ACTIONS      C.1 (continued)
With one DC electrical power distribution subsystem inoperable, the remaining portions of the DC electrical power distribution subsystem are capable of supporting the minimum safety functions necessary to shut down the reactor and maintain it in a safe shutdown condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in the remaining portion of the DC electrical power distribution subsystems could result in the minimum required ESF functions not being supported. Therefore, the DC buses must be restored to OPERABLE status within 2 hours by powering the bus from the associated battery or charger. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.
Condition C represents one DC electrical power distribution subsystem without adequate DC power; potentially both with the battery significantly degraded and the associated charger nonfunctioning for the affected bus. In this situation, the unit is significantly more vulnerable to a complete loss of all DC power. It is, therefore, imperative that the operator's attention focus on stabilizing the unit, minimizing the potential for loss of power to the remaining DC electrical power distribution subsystems and restoring power to the affected subsystem.
This 2 hour limit is more conservative than Completion Times allowed for the vast majority of components that would be without power.
Taking exception to LCO 3.0.2 for components without adequate DC power, which would have Required Action Completion Times shorter than 2 hours, is acceptable because of:
: a.      The potential for decreased safety by requiring a change in unit conditions (i.e., requiring a shutdown) while allowing stable operations to continue; (continued)

PG&E Letter DCL-23-054 Attachment 4 Page 1 of 9 Attachment 4 Diablo Canyon Unit 1 and 2 Scope for Adopting TSTF-505, Revision 2

PG&E Letter DCL-23-054 Attachment 4 Page 2 of 9 Attachment 4: Diablo Canyon Unit 1 and 2 Scope for Adopting TSTF-505, Revision 2 TSTF-505 LCO/RA       DCPP LCO/RA (if different)                                    NOTES Example 1.3-8                                     NO VARIATION 3.3.1 (Various RAs)                               VARIATION: Pacific Gas and Electric Company (PG&E) is not proposing to include the TSTF-505, Revision 2 changes to any Required Actions (RAs) for Technical Specification (TS) 3.3.1, Reactor Trip System (RTS)
Instrumentation.
3.3.2 (Various RAs)                               VARIATION: PG&E is not proposing to include the TSTF-505, Revision 2 changes to any RAs for TS 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation.
3.3.5 (Various RAs)                               VARIATION: PG&E is not proposing to include the TSTF-505, Revision 2 changes to any RAs for TS 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation.
3.3.9 A.1             N/A                         VARIATION: The Diablo Canyon TS do not include Limiting Condition for Operation (LCO) 3.3.9, Boron Dilution Protection System (BDPS).
3.4.5 (Various RAs)                               OPTIONAL CHANGE: PG&E is not proposing to include the TSTF-505, Revision 2 changes to any RAs for TS 3.4.5, Reactor Coolant System (RCS) Loops - MODE 3.
3.4.9 B.1                                         NO VARIATION

PG&E Letter DCL-23-054 Attachment 4 Page 3 of 9 Attachment 4: Diablo Canyon Unit 1 and 2 Scope for Adopting TSTF-505, Revision 2 TSTF-505 LCO/RA    DCPP LCO/RA (if different)                                   NOTES 3.4.11 B.3                                     EDITORIAL: Diablo Canyon LCO 3.4.11 includes both Class 1 power-operated relief valves (PORVs) and the non-Class 1 PORV, and different RAs are applicable. PG&E is not proposing to include changes to the RAs applicable to the non-Class 1 PORV, since these actions do not require restoration of the PORV to OPERABLE status.
3.4.11 C.2         3.4.11 C.2 and C.3          EDITORIAL: Diablo Canyon LCO 3.4.11 has separate Conditions for one inoperable PORV block valve (Condition C) and for more than one 3.4.11 F.2 and F.3          inoperable PORV block valve (Condition F). RA C.2 applies when one PORV block valve associated with a Class 1 PORV is inoperable. RAs F.2 and F.3 apply when both block valves associated with the Class 1 PORVs are inoperable. PG&E is proposing to apply the Risk-Informed Completion Time (RICT) Program to RAs 3.4.11 C.2, F.2, and F.3, which require restoration of the inoperable PORV block valve(s) to OPERABLE status.
PG&E is not proposing to include changes to RAs 3.4.11 C.3 or 3.4.11 F.4, applicable to the PORV block valve associated with the non-Class 1 PORV, since these actions do not require restoration of the inoperable PORV block valve to OPERABLE status.
3.5.2 A.1                                      EDITORIAL: Diablo Canyon Condition A specifies the additional requirement that At least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.
VARIATION: The Diablo Canyon RA A.1 applies to planned maintenance or inspections with a 72-hour Completion Time (CT), while the separate RAs A.2.1, A.2.2, and A.2.3 apply to unplanned corrective maintenance or inspections with a 14-day CT for restoration of the inoperable subsystem.
The plant-specific actions were incorporated into the Diablo Canyon Power Plant (DCPP) TS by License Amendments 203 (Unit 1) and 202 (Unit 2).

PG&E Letter DCL-23-054 Attachment 4 Page 4 of 9 Attachment 4: Diablo Canyon Unit 1 and 2 Scope for Adopting TSTF-505, Revision 2 TSTF-505 LCO/RA    DCPP LCO/RA (if different)                                      NOTES In order to adopt TSTF-505, Revision 2, PG&E is proposing to restructure its plant-specific TS 3.5.2 RA for Condition A to be consistent with TSTF-505, Revision 2 by deleting the Note regarding different RAs for planned and unplanned activities, deleting RAs A.2.1, A.2.2, and A.2.3, and applying the RICT Program to the RA A.1 72-hour Completion Time.
RA A.2.1 is no longer required since the extended CT of 14 days in RA A.2.3 is deleted.
RA A.2.2 to determine common cause impacts on unplanned inoperability within 72 hours is no longer required since this determination is a requirement for using the RICT Program to extend the CT beyond the CT of RA A.1 of 72 hours for unplanned corrective maintenance or inspections.
RA A.2.3 for a 14-day CT is eliminated, since the RICT Program may be applied to any unplanned single subsystem inoperability to establish the extended CT for RA A.1.
3.6.2 C.3                                      NO VARIATION 3.6.3 A.1                                      NO VARIATION 3.6.3 C.1                                      NO VARIATION

PG&E Letter DCL-23-054 Attachment 4 Page 5 of 9 Attachment 4: Diablo Canyon Unit 1 and 2 Scope for Adopting TSTF-505, Revision 2 TSTF-505 LCO/RA    DCPP LCO/RA (if different)                                     NOTES 3.6.6A A.1         3.6.6 A.1                   VARIATION: The Diablo Canyon RA A.1 with a 72-hour CT applies to planned maintenance or inspections, while the separate RA A.2 with a 14-day CT applies to unplanned corrective maintenance or inspections. The plant-specific RAs were incorporated into the DCPP TS by License Amendments 203 (Unit 1) and 202 (Unit 2). In order to adopt TSTF-505, Revision 2, PG&E is proposing to restructure its plant-specific TS 3.6.6 RAs for Condition A to be consistent with TSTF-505, Revision 2 by deleting the Note regarding different RAs for planned and unplanned activities, deleting RA A.2 and the extended 14-day CT, and applying the RICT Program to the RA A.1 72-hour CT.
3.6.6A C.1         3.6.6 C.1                   EDITORIAL: Diablo Canyon TS Condition C includes the additional requirement that a minimum of two containment fan cooling units (CFCU) remain OPERABLE.
3.6.6 D.1 and D.2           VARIATION: The Diablo Canyon plant-specific design has five CFCUs instead of the two-train configuration in the Standard Technical Specification and TSTF-505, Revision 2. The Diablo Canyon TS have a plant-specific Condition D associated with one containment spray train inoperable coincident with one CFCU system inoperable where a minimum of two CFCUs remain OPERABLE. This condition accommodates the plant-specific design (five CFCUs instead of two trains), and does not represent a loss of the containment spray and cooling function. PG&E is proposing to apply the RICT Program to RAs 3.6.6 D.1 and 3.6.6 D.2, which require restoration of the inoperable equipment to OPERABLE status.
3.6.6A D.1         N/A                         VARIATION: The Diablo Canyon TS for LCO 3.6.6, Containment Spray and Cooling Systems, do not include Condition D of TSTF-505, Revision 2 for two inoperable containment cooling trains.

PG&E Letter DCL-23-054 Attachment 4 Page 6 of 9 Attachment 4: Diablo Canyon Unit 1 and 2 Scope for Adopting TSTF-505, Revision 2 TSTF-505 LCO/RA    DCPP LCO/RA (if different)                                     NOTES 3.7.2 A.1                                      NO VARIATION 3.7.4 A.1                                      NO VARIATION 3.7.4 B.1          3.7.4 B.1 and C.1           VARIATION: The Diablo Canyon TS have separate Conditions for two required atmospheric dump valve (ADV) lines inoperable and three or more required ADV lines inoperable. PG&E is not proposing to include the TSTF-505, Revision 2 changes to TS 3.7.4 Conditions B or C since inoperability of more than one required ADV line results in a loss of function.
3.7.5 A.1                                      NO VARIATION 3.7.5 B.1                                      EDITORIAL: Reference to Condition G is being deleted in Conditions B and D since Condition G has expired.
3.7.5 C.1 and C.2           VARIATION: The Diablo Canyon TS have a plant-specific Condition C for the auxiliary feedwater (AFW) system. It applies when the turbine driven AFW train is inoperable due to an inoperable steam supply, one motor driven train is inoperable and the remaining OPERABLE motor driven train provides feedwater to the steam generator with the inoperable steam supply. Entry into Condition C does not result in a loss of function of the AFW system, and PG&E is proposing to apply the RICT Program to RAs 3.7.5 C.1 and C.2.
3.7.7 A.1                                      NO VARIATION 3.7.8 A.1                                      EDITORIAL: The Note included in the CT applicable on a one-time basis for Unit 1 cycle 23 is being deleted. This note and its deletion have no impact on adoption of TSTF-505, Revision 2.

PG&E Letter DCL-23-054 Attachment 4 Page 7 of 9 Attachment 4: Diablo Canyon Unit 1 and 2 Scope for Adopting TSTF-505, Revision 2 TSTF-505 LCO/RA    DCPP LCO/RA (if different)                                   NOTES 3.7.9 A.1          N/A                         VARIATION: The Diablo Canyon TS for LCO 3.7.9, Ultimate Heat Sink (UHS), do not include Condition A since the plant design does not have a cooling tower and cooling tower fans.
3.8.1 A.3          3.8.1 A.2                   EDITORIAL: The Diablo Canyon TS 3.8.1 RA for restoring the required offsite circuit to OPERABLE status is RA A.2.
3.8.1 B.4                                      NO VARIATION 3.8.1 C.2                                      NO VARIATION 3.8.1. D.1                                     NO VARIATION 3.8.1. D.2                                     NO VARIATION 3.8.1 F.1          N/A                         VARIATION: The Diablo Canyon TS do not have a separate Condition for inoperable sequencers.

PG&E Letter DCL-23-054 Attachment 4 Page 8 of 9 Attachment 4: Diablo Canyon Unit 1 and 2 Scope for Adopting TSTF-505, Revision 2 TSTF-505 LCO/RA    DCPP LCO/RA (if different)                                     NOTES 3.8.1 F.1                   VARIATION: The Diablo Canyon plant-specific design for the DGs has a shared diesel fuel oil (DFO) transfer system with two trains. Each DFO transfer system train is capable of providing sufficient fuel oil to all DGs.
The Diablo Canyon TS have a plant-specific Condition F for one inoperable DFO transfer system train. This Condition does not result in a loss of TS function since the remaining OPERABLE DFO transfer system train is capable of providing sufficient fuel oil to all DGs. PG&E is proposing to apply the RICT Program to the RA 3.8.1 F.1, which requires restoration of the inoperable DFO transfer system train to OPERABLE status.
EDITORIAL: Diablo Canyon TS 3.8.1 RA F.1 includes a Note with a plant-specific one-time CT that has expired and is proposed to be deleted. This Note and its deletion have no impact on adoption of TSTF-505, Revision 2.
3.8.4 A.3                                      NO VARIATION 3.8.4 B.1                                      NO VARIATION 3.8.4 C.1                                      NO VARIATION 3.8.7 A.1                                      NO VARIATION 3.8.9 A.1                                      EDITORIAL: Diablo Canyon TS 3.8.9 Condition A does not address more than one inoperable AC electrical power distribution subsystem.
3.8.9 B.1                                      EDITORIAL: Diablo Canyon TS 3.8.9 Condition B does not address more than one inoperable 120 VAC vital bus.
3.8.9 C.1                                      EDITORIAL: Diablo Canyon TS 3.8.9 Condition C does not address more than one inoperable DC electrical power distribution subsystem.

PG&E Letter DCL-23-054 Attachment 4 Page 9 of 9 Attachment 4: Diablo Canyon Unit 1 and 2 Scope for Adopting TSTF-505, Revision 2 TSTF-505 LCO/RA    DCPP LCO/RA (if different)                                    NOTES 5.5.18             5.5.20                      EDITORIAL: Diablo Canyon TS 5.5 already includes a program 5.5.18; the next available number is 5.5.20.
VARIATION: Diablo Canyon is updating the TSTF-505, Revision 2 text for Specification 5.5.18.e to change methods used to support this license amendment, to methods approved for use with this program, to be consistent with other recently approved TSTF-505, Revision 2 applications.
All                All in scope                EDITORIAL: PG&E proposes to use the acronym RICT in each applicable CT instead of spelling out Risk Informed Completion Time.

PG&E Letter DCL-23-054 Enclosure 1 Page 1 of 17 Enclosure 1 List of Revised Required Actions to the Corresponding Probabilistic Risk Assessment (PRA) Functions

PG&E Letter DCL-23-054 Enclosure 1 Page 2 of 17 Section 4.0, Item 2 of the Nuclear Regulatory Commissions (NRCs) Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (Reference 2), identifies the following license amendment request (LAR) content needed on the applicable Technical Specifications (TS), comparison of the TS functions to the probabilistic risk assessment (PRA) functions, and comparison of design basis assumptions to the scope of the PRA:
The LAR will provide identification of the TS Limiting Conditions for Operation (LCO) and action requirements to which the RMTS will apply.
The LAR will provide a comparison of the TS functions to the PRA modeled functions of the structures, systems, and components (SSCs) subject to those LCO actions.
The comparison should justify that the scope of the PRA model, including applicable success criteria such as number of SSCs required, flowrate, etc., are consistent with licensing basis assumptions (i.e., 10 CFR 50.46 emergency core cooling system
[ECCS] flowrates) for each of the TS requirements, or an appropriate disposition or programmatic restriction will be provided.
This enclosure provides confirmation that the Diablo Canyon PRA models include the necessary scope of SSCs and their functions to address each proposed application of the Risk-Informed Completion Time (RICT) Program to the proposed scope of TS LCOs and Conditions, and provides the information requested for Item 2 of the NRC safety evaluation. The scope of the comparison includes each of the TS LCO and Conditions and associated Required Actions (RAs) within the scope of the RICT Program, as identified in Attachment 4 of the LAR.
Table E1-1 below lists each TS LCO and Condition to which the RICT Program is proposed to be applied, and documents the following information regarding the TS with the associated safety analyses, the analogous PRA functions, and the results of the comparison:
Column TS LCO/Condition: Lists all of the LCOs and Conditions within the scope of the TSTF-505, Revision 2, implementation.
Column SSCs Covered by TS LCO/Condition: The SSCs addressed by each LCO/Condition.
Column SSCs Modeled in PRA: Indicates whether the SSCs addressed by the TS LCO/Condition are included in the PRA.
Column Function Covered by TS LCO/Condition: A summary of the required functions in the design basis safety analyses.

PG&E Letter DCL-23-054 Enclosure 1 Page 3 of 17 Column Design Success Criteria: A summary of the success criteria in the design basis safety analyses.
Column PRA Success Criteria: The function success criteria modeled in the PRA.
Column Disposition: Justification or resolution to address any inconsistencies between the TS and PRA functions, regarding the scope of SSCs and the success criteria. Where the PRA scope of SSCs is not consistent with the TS, additional information is provided to describe how the LCO Condition can be evaluated using appropriate surrogate events. Differences in the success criteria for the TS functions are addressed to demonstrate the PRA criteria provide a realistic estimate of the risk of the TS Condition as required by NEI 06-09-A.
The corresponding SSCs for each TS LCO and the associated TS functions are identified and compared to the PRA. This description also includes the design success criteria and the applicable PRA success criteria. Any differences between the scope or success criteria are described in the table. Scope differences are justified by identifying appropriate surrogate events which permit a risk evaluation to be completed using the Configuration Risk Management Program (CRMP) tool for the RICT program.
Differences in success criteria typically arise due to the requirement in the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA Standard to make PRAs realistic rather than bounding, whereas design basis criteria are conservative and bounding. The use of realistic success criteria is necessary to conform to Capability Category II of the ASME/ANS PRA Standard as required by NEI 06-09-A.
The calculated RICT is provided in Table E1-2 for each individual TS Condition to which the RICT Program will be applied (assuming no other SSCs modeled in the PRA are unavailable). Unless stated otherwise, the RICTs presented are based on a Unit 1 model calculation and are applicable to Unit 2 for the purpose of providing an estimate due to the close similarity between the Unit 1 and Unit 2 models. (Actual RICT values will be calculated based on the actual plant configuration using a current revision of the PRA model that represents the as-built/as-operated condition of the plant, as required by NEI 06-09-A and the NRC safety evaluation, and may differ from the RICTs presented.)

PG&E Letter DCL-23-054 Enclosure 1 Page 4 of 17 Table E1-1: In-Scope TS/LCO Conditions to the Corresponding PRA Functions TS LCO/Condition   SSCs Covered by     SSCs       Function Covered by     Design Success Criteria  PRA Success       Disposition TS LCO/Condition    Modeled    the TS LCO/                                      Criteria in the PRA Condition 3.4.9 Pressurizer  2 groups of heaters No         (1) Maintain reactor                                               Note 1 Heaters                                           coolant system (RCS) subcooling margin 3.4.11 Pressurizer 3 PORVs (2 Class    Yes        (1) Depressurize the    (1) 1 of 2 Class I PORVs (1) SAME or more  SSCs are modeled consistent power operated     I, 1 non-Class I)              RCS                     open                     restrictive       with the TS scope and can be relief valves      3 PORV block                                                                                      directly evaluated using the (PORVs)            valves                                                 (2) 1 of 2 Class I PORVs                   CRMP.
(2) Mitigate spurious                            (2) SAME operation of the safety open injection system at                                                The success criteria in the power                                                              PRA are consistent with the design basis criteria, and in (3) Associated block     (3) SAME          some cases are more (3) No excessive seat   valve manually closed leakage                                                            restrictive when the PORVs are credited to mitigate some beyond-design-basis scenarios. The PRA also credits the non-Class I PORV if it is not failed; this is consistent with the TS Bases which identifies that the non-Class I PORV can be used if it is available.

PG&E Letter DCL-23-054 Enclosure 1 Page 5 of 17 Table E1-1: In-Scope TS/LCO Conditions to the Corresponding PRA Functions TS LCO/Condition SSCs Covered by    SSCs       Function Covered by     Design Success Criteria    PRA Success           Disposition TS LCO/Condition   Modeled    the TS LCO/                                        Criteria in the PRA Condition 3.5.2 ECCS -     2 Centrifugal      Yes        (1) Injection from the  (1)                        (1)                   SSCs are modeled consistent Operating        Charging (CH)                 refueling water storage (a) 1 of 2 CH pumps for    (a) small LOCA: 1     with the TS scope and can be pumps (high                   tank into cold legs     small/medium loss of       of 4 CH pumps or      directly evaluated using the pressure)                                             coolant accident (LOCA)    SI pumps into 3       CRMP.
(2) Cold leg            until the RCS is           cold legs; medium 2 Safety Injection            recirculation from the  depressurized to allow     LOCA: 2 of 4 CH       The PRA success criteria (SI) pumps                    containment sumps       injection from 1 of 2 SI   pumps or SI pumps     differ from the design basis (intermediate                                         pumps                      (b) 1 of 2 RHR        in: (1) crediting the centrifugal pressure)                                             (b) 1 of 2 RHR pumps, SI   pumps                 charging pumps or SI pumps (3) Hot leg             pumps, and CH pumps                              where the design basis recirculation from the                             (c) 1 CH pump or for a large LOCA           SI pump for SGTR;     requires one of each to 2 Residual Heat               containment sumps                                                        function; (2) not requiring Removal (RHR)                                         (c) 1 of 2 CH pumps for a  none for MSLB.
steam generator tube                             injection into all RCS loops; pumps (low                                                                                             (3) not crediting mitigation for pressure)                                             rupture (SGTR) or main     (2) SAME except steam line break (MSLB)                          MSLB events; (4) not injection into 2 cold requiring hot leg recirculation.
legs                  The success criteria in the Associated piping, (2) 1 of 2 RHR pumps to                          PRA are based on plant-valves, and heat supply other required                            specific realistic analyses exchangers                                                                       (3) Not required ECCS pump suctions and                           consistent with the PRA injection into the RCS                           standards for capability cold legs                                        category II.
(3) 1 of 2 RHR pumps to                          See Note 2.
supply other required ECCS pump suctions and injection into the RCS hot legs

PG&E Letter DCL-23-054 Enclosure 1 Page 6 of 17 Table E1-1: In-Scope TS/LCO Conditions to the Corresponding PRA Functions TS LCO/Condition  SSCs Covered by     SSCs       Function Covered by Design Success Criteria PRA Success    Disposition TS LCO/Condition    Modeled    the TS LCO/                                 Criteria in the PRA Condition 3.6.2 Containment 2 air locks         No         (1) post-accident                                          SSCs for the containment air Air Locks         (personnel and                 containment leakage                                        locks can be evaluated by a emergency)                     within limits                                              bounding assessment as permitted by NEI 06-09-A.
The PRA model includes an event that involves a large, pre-existing containment leak; this is bounding for the risk on an inoperable air lock and can be used as a bounding surrogate.
See Note 3.

PG&E Letter DCL-23-054 Enclosure 1 Page 7 of 17 Table E1-1: In-Scope TS/LCO Conditions to the Corresponding PRA Functions TS LCO/Condition  SSCs Covered by      SSCs       Function Covered by    Design Success Criteria     PRA Success        Disposition TS LCO/Condition     Modeled    the TS LCO/                                        Criteria in the PRA Condition 3.6.3 Containment 2 active or passive  Yes        (1) Each containment   (1) 1 of 2 isolation        (1) SAME for all   SSCs for those containment Isolation Valves  isolation devices on            penetration isolated   devices per penetration     penetrations       isolation valves not in the each fluid                      within the time limits isolate within the required modeled in the     PRA model can be evaluated penetration line                assumed in the safety  stroke time.                PRA; all other     by a bounding assessment as analyses                                           penetrations were  permitted by NEI 06-09. The evaluated and      PRA model includes an event determined not to  that involves a large, pre-be significant     existing containment leak; sources of fission this is bounding for the risk on product leakage    an inoperable isolation valve and were screened  and can be used as a out.               bounding surrogate.
The PRA does not explicitly model the impact of excessive stroke times. This condition can be addressed in the RICT Program by conservatively assuming that the inoperable containment isolation valve cannot be closed if it is open. Otherwise, the success criteria in the PRA are consistent with the design basis criteria.

PG&E Letter DCL-23-054 Enclosure 1 Page 8 of 17 Table E1-1: In-Scope TS/LCO Conditions to the Corresponding PRA Functions TS LCO/Condition  SSCs Covered by    SSCs       Function Covered by    Design Success Criteria    PRA Success       Disposition TS LCO/Condition   Modeled    the TS LCO/                                       Criteria in the PRA Condition 3.6.6 Containment 2 CS trains        Yes        (1) Containment        (1) 1 of 2 CS trains and 2 (1) SAME          The SSCs in the TS scope Spray (CS) and    5 containment fan             atmosphere cooling to  of 5 CFCUs                                   are modeled in the PRA. The Cooling Systems   cooling units                 limit post-accident    (2) 1 of 2 CS trains       (2) Not modeled   iodine removal function of the (CFCUs)                       pressure and                                                        CS trains is not required for temperature                                                         mitigation of severe accidents (2) Iodine removal to                                               and is not modeled.
reduce the release of fission product                                                     See Notes 4 and 5.
radioactivity from containment to the environment 3.7.2 Main Steam  4 MSIVs            Yes        (1) Isolate steam flow (1) MSIV on the affected   (1) SAME or more  The SSCs are modeled Isolation Valves                                from the secondary     steam line closes, or the  restrictive       consistent with the TS scope (MSIVs)                                         side of the steam      remaining 3 MSIVs on                         and can be directly evaluated generators (SGs)       unaffected steam lines                       using the CRMP.
following a high       close.                                       The success criteria in the energy line break                                                   PRA are consistent with the (HELB)                                                              design basis criteria for a HELB. See Note 6.
The PRA also credits MSIV closure for isolation of a ruptured SG, and on 3 of 4 steam lines to prevent RCS overcooling in the event of a failure of the turbine trip function.

PG&E Letter DCL-23-054 Enclosure 1 Page 9 of 17 Table E1-1: In-Scope TS/LCO Conditions to the Corresponding PRA Functions TS LCO/Condition SSCs Covered by    SSCs       Function Covered by     Design Success Criteria   PRA Success         Disposition TS LCO/Condition   Modeled    the TS LCO/                                       Criteria in the PRA Condition 3.7.4 10%        4 ADV lines (one   Yes        (1) Cool down the unit  (1) 4 of 4 ADVs to cool   (1) 1 of 4 ADVs     The SSCs are modeled Atmospheric Dump per SG, each with             to RHR entry            down the unit at the                          consistent with the TS scope Valves (ADVs)    an ADV and                    conditions, if the      design rate of 100&deg;F per  (2) 1 of 4 ADVs on  and can be directly evaluated associated block              preferred heat sink via hour; 1 of 4 ADVs permits the intact SG lines using the CRMP.
valve)                        steam dump to the       a 25&deg;F per hour cooldown condenser is not        for a natural circulation available.              cooldown event.                               The success criteria in the PRA do not require the maximum rate cooldown (2) Cool down the       (2) 3 of 4 ADVs on the                        capability to mitigate severe RCS following a SGTR    intact SG lines.                              accidents, and therefore more to permit termination                                                 realistic criteria are applicable of primary to                                                         consistent with the PRA secondary break flow.                                                 standards for capability category II.
See Note 7.

PG&E Letter DCL-23-054 Enclosure 1 Page 10 of 17 Table E1-1: In-Scope TS/LCO Conditions to the Corresponding PRA Functions TS LCO/Condition SSCs Covered by     SSCs       Function Covered by   Design Success Criteria     PRA Success          Disposition TS LCO/Condition    Modeled    the TS LCO/                                       Criteria in the PRA Condition 3.7.5 Auxiliary  2 motor-driven      Yes        (1) Supply feedwater  (1) 2 motor-driven pumps    (1) 1 of 3 pumps     SSCs are modeled consistent Feedwater (AFW)  pumps and 1                    to the SGs to remove  or 1 turbine-driven pump                         with the TS scope and can be System           turbine-driven pump            decay heat            for the most limiting event                      directly evaluated using the (loss of main feedwater)                         CRMP.
The success criteria in the PRA are based on a "better estimate" evaluation which demonstrates any one AFW pump can provide 100% of the feedwater flow required for removal of decay heat from the reactor. This is discussed in the plant-specific TS Bases. The use of more realistic success criteria is consistent with the PRA standards for capability category II.
3.7.7 Component  2 vital loops       Yes        (1) Heat sink for the (1) 1 of 2 vital loops with (1) SAME;            The SSCs are modeled Cooling Water                                   removal of process    2 of 3 CCW pumps and 1      successful isolation consistent with the TS scope (CCW) System                                    and operating heat    of 2 heat exchangers        of unnecessary       and can be directly evaluated from safety-related                               CCW heat loads is    using the CRMP.
components                                        also credited and then only 1 of 3 CCW pumps is         The success criteria in the required             PRA are consistent with the design basis criteria, but also include credit for operator action to isolate unnecessary CCW heat loads; in this case, only 1 of 3 CCW pumps is required.

PG&E Letter DCL-23-054 Enclosure 1 Page 11 of 17 Table E1-1: In-Scope TS/LCO Conditions to the Corresponding PRA Functions TS LCO/Condition     SSCs Covered by     SSCs       Function Covered by       Design Success Criteria PRA Success         Disposition TS LCO/Condition    Modeled    the TS LCO/                                       Criteria in the PRA Condition 3.7.8 Auxiliary      2 trains            Yes        (1) Heat sink for the     (1) 1 of 2 trains       (1) SAME; cross-tie The SSCs are modeled Saltwater (ASW)                                     removal of process                                to the unaffected   consistent with the TS scope System                                              and operating heat                                unit is also        and can be directly evaluated from the CCW system                               credited.           using the CRMP.
The success criteria in the PRA are consistent with the design basis criteria.
3.8.1 Alternating    2 offsite circuits  Yes        (1) Source of power to    (1) Automatically power (1) SAME            The SSCs are modeled Current (AC)         3 diesel generators            the engineered safety     the associated ESF                          consistent with the TS scope Sources - Operating  (DG)                           features (ESF)            buses                   (2) SAME            and can be directly evaluated systems                   (2) 1 of 2 trains                           using the CRMP. See Note 8.
2 supply trains of the diesel fuel oil (DFO) transfer                 (2) Source of fuel oil to                                             The success criteria in the system                         the DGs                                                               PRA are consistent with the design basis criteria.
3.8.4 Direct Current 3 Class 1E DC       Yes        (1) Provide control       (1) Aligned to provide  (1) SAME            The SSCs are modeled (DC) Sources -       subsystems                     power to the AC           power to the associated                     consistent with the TS scope Operating                                           emergency power           equipment from the                          and can be directly evaluated system, motive and        battery and associated                      using the CRMP.
control power to          charger selected safety-related equipment, and                                                        The success criteria in the backup 120 volt                                                       PRA are consistent with the alternating current                                                   design basis criteria.
(VAC) vital bus power

PG&E Letter DCL-23-054 Enclosure 1 Page 12 of 17 Table E1-1: In-Scope TS/LCO Conditions to the Corresponding PRA Functions TS LCO/Condition        SSCs Covered by      SSCs       Function Covered by   Design Success Criteria    PRA Success    Disposition TS LCO/Condition     Modeled    the TS LCO/                                      Criteria in the PRA Condition 3.8.7 Inverters -       4 Class 1E           Yes        (1) Provide           (1) Align to the           (1) SAME       The SSCs are modeled Operating               inverters                       uninterruptible power associated 120 VAC vital                  consistent with the TS scope to the Reactor        bus, with input power                     and can be directly evaluated Protection System and from the vital AC and                     using the CRMP.
Engineered Safety     associated battery Features Actuation System                                                          The success criteria in the PRA are consistent with the design basis criteria.
3.8.9 Distribution      Class 1E AC, DC,     Yes        (1) Provide necessary (1) Align to provide power (1) SAME       The SSCs are modeled Systems - Operating     and 120 VAC vital               power to the ESF      to the buses                              consistent with the TS scope bus electrical power            systems                                                         and can be directly evaluated distribution                                                                                    using the CRMP.
subsystems The success criteria in the PRA are consistent with the design basis criteria.
Note 1: The pressurizer heaters will be evaluated for the RICT Program by a bounding assessment as permitted by NEI 06-09-A. Inoperability of the pressurizer heaters will be conservatively bounded by assuming an increase in the frequency of a reactor trip initiating event by a factor of 10; this reflects the adverse impact on pressure control due to inoperable pressurizer heaters. This is conservative since the redundant pressurizer heater group of TS 3.4.9 must be OPERABLE, and additional pressurizer heater groups not required by TS 3.4.9 would typically be available. The safe shutdown of the plant after a reactor trip without pressurizer heaters available is addressed by plant procedures. This surrogate is consistent with recently approved TSTF-505, Revision 2 applications for plants similar in design to Diablo Canyon. This note satisfies the requirements of Table 1 of TSTF-505, Revision 2.
Note 2: TS 3.5.2 Condition A explicitly requires 100 percent of the ECCS flow equivalent to a single OPERABLE ECCS train. Therefore, TS 3.5.2 Condition A meets the requirements for inclusion in the RICT Program.
This note satisfies the requirements of Table 1 of TSTF-505, Revision 2.

PG&E Letter DCL-23-054 Enclosure 1 Page 13 of 17 Note 3: The containment air locks form part of the containment pressure boundary. As such, air lock integrity and leak tightness is essential for maintaining the containment leakage rate within limit in the event of a design basis accident (DBA). Not maintaining air lock integrity or leak tightness may result in a leakage rate in excess of that assumed in the safety analyses. The DBA that results in a release of radioactive material within containment is the LOCA. In the analysis of this accident, it is assumed that containment is OPERABLE such that release of fission products to the environment is controlled by the rate of containment leakage.
Required Action C.1 requires action to be initiated immediately to evaluate previous combined leakage rates using current air lock test results. An evaluation is acceptable, since it is overly conservative to immediately declare the containment inoperable if both doors in an air lock have failed a seal test or if the overall air lock leakage is not within limits. In many instances (e.g., only one seal per door has failed),
containment remains OPERABLE, yet only 1 hour (per LCO 3.6.1) would be provided to restore the air lock door to OPERABLE status prior to requiring a plant shutdown. In addition, even with both doors failing the seal test, the overall containment leakage rate can still be within limits. Required Action (RA) C.2 requires that one door in the affected containment air lock must be verified to be closed within the 1-hour Completion Time. This specified time period is consistent with the ACTIONS of LCO 3.6.1, which requires that containment be restored to OPERABLE status within 1 hour. These Completion Times are well within the existing 24-hour Completion Time for RA C.3. Thus, the containment function is maintained at the point in time when a RICT would be entered.
Therefore, TS 3.6.2 Condition C meets the requirements for inclusion in the RICT Program.
This note satisfies the requirements of Table 1 of TSTF-505, Revision 2.
Note 4: The fire PRA does not credit the CS system nor the CFCUs.
Note 5: The CS and containment cooling systems provide containment atmosphere cooling to limit post-accident pressure and temperature in containment to less than the design values. Reduction of containment pressure and the iodine removal capability of the spray reduces the release of fission product radioactivity from containment to the environment, in the event of a DBA, to within limits. The CS system is modeled in the Diablo Canyon PRA, with the same success criteria as the design success criteria (i.e., one of two CS subsystems).

PG&E Letter DCL-23-054 Enclosure 1 Page 14 of 17 Therefore, TS 3.6.6 Condition A meets the requirements for inclusion in the RICT Program.
This note satisfies the requirements of Table 1 of TSTF-505, Revision 2.
Note 6: The design of the MSIVs precludes the blowdown of more than one SG, assuming a single active component failure (e.g., the failure of one MSIV to close on demand). With one MSIV inoperable in Condition A, the steam line isolation function of TS 3.7.2 is met by the remaining three OPERABLE MSIVs on the other three SGs to prevent blowdown of more than one SG . This note satisfies the requirements of Table 1 of TSTF-505, Revision 2.
Note 7: Condition B of TS 3.7.4 for two inoperable ADV lines is not in the proposed Diablo Canyon TSTF-505, Revision 2 scope; therefore, no additional justification needs to be provided per Table 1 of TSTF-505, Revision 2.
Note 8: The 500 kV offsite circuits are only credited for the mitigation of internal events.

PG&E Letter DCL-23-054 Enclosure 1 Page 15 of 17 Table E1-2: Unit 1/Unit 2 In-Scope TS/LCO Conditions RICT Estimate TS LCO/Condition                               RICT Estimate 3.4.9 Pressurizer Heaters                                                      30 days B. One required group of pressurizer heaters inoperable 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)                        30 days B. One PORV inoperable for reasons other than excessive seat leakage 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)                        30 days C. One block valve inoperable 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)                      19.1 days F. More than one block valve inoperable 3.5.2 ECCS - Operating                                                       21.5 days A. One or more trains inoperable AND at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available 3.6.2 Containment Air Locks:                                                  5.4 days C. One or more containment air locks inoperable for reasons other than Condition A or B 3.6.3 Containment Isolation Valves:                                            30 days A, C. One or more penetration flow paths with one containment isolation valve inoperable 3.6.6 Containment Spray and Cooling Systems:                                   30 days A. One containment spray train inoperable 3.6.6 Containment Spray and Cooling Systems:                                   30 days C. One CFCU inoperable 3.6.6 Containment Spray and Cooling Systems:                                   30 days D. One containment spray train and one CFCU inoperable 3.7.2 Main Steam Isolation Valves (MSIVs):                                     30 days A. One MSIV system inoperable in MODE 1 3.7.4 10% Atmospheric Dump Valves (ADVs):                                      30 days A. One required ADV line inoperable 3.7.5 Auxiliary Feedwater (AFW) System:                                      22.5 days A. One steam supply to turbine-driven AFW pump inoperable 3.7.5 AFW System:                                                             9.2 days B. One AFW train inoperable for reasons other than Condition A

PG&E Letter DCL-23-054 Enclosure 1 Page 16 of 17 Table E1-2: Unit 1/Unit 2 In-Scope TS/LCO Conditions RICT Estimate TS LCO/Condition                              RICT Estimate 3.7.5 Auxiliary Feedwater (AFW) System:                                        1.4 days C. One motor-driven AFW pump and one turbine-driven pump steam supply inoperable 3.7.7 Component Cooling Water (CCW) System:                                     30 days A. One CCW train inoperable 3.7.8 Auxiliary Salt Water (ASW) System:                                        30 days A. One ASW train inoperable 3.8.1 AC Sources - Operating:                                                  9.2 days A. One required offsite circuit inoperable 3.8.1 AC Sources - Operating:                                                   30 days B. One DG inoperable 3.8.1 AC Sources - Operating:                                                  8.3 days C. Two required offsite circuits inoperable 3.8.1 AC Sources - Operating:                                                  1.5 days D. One required offsite circuit inoperable and one DG inoperable 3.8.1 AC Sources - Operating:                                                   30 days F. One DFO supply train inoperable 3.8.4 DC Sources - Operating:                                                 15.6 days A. One battery charger inoperable 3.8.4 DC Sources - Operating:                                                  7.8 days B. One battery inoperable 3.8.4 DC Sources - Operating:                                                 20.6 days C. DC electrical power subsystem inoperable 3.8.7 Inverters - Operating:                                                    30 days A. One required inverter inoperable 3.8.9 Distribution Systems - Operating:                                        4.8 days A. One AC electrical power distribution subsystem inoperable 3.8.9 Distribution Systems - Operating:                                        4.8 days B. One AC vital bus electrical power distribution subsystem inoperable 3.8.9 Distribution Systems - Operating:                                       20.6 days C. One DC electrical power distribution subsystem inoperable

PG&E Letter DCL-23-054 Enclosure 1 Page 17 of 17 References
: 1. Letter from Jennifer M. Golder (USNRC - NRR) to Biff Bradley (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, (TAC No. MD4995), May 17, 2007. (ADAMS Accession No. ML071200238).
: 2. NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, Revision 0, November 2006. (ADAMS Accession No. ML12286A322).

PG&E Letter DCL-23-054 Enclosure 2 Page 1 of 8 Enclosure 2 Information Supporting Consistency with Regulatory Guide 1.200, Revision 2

PG&E Letter DCL-23-054 Enclosure 2 Page 2 of 8 Introduction Section 4.0, Item 3 of the Nuclear Regulatory Commissions (NRCs) Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (Reference 2), requires that the license amendment request (LAR) provide a discussion of the results of peer reviews and self-assessments conducted for the plant-specific probabilistic risk assessment (PRA) models that support the RMTS, including the resolution or disposition of any identified deficiencies (i.e., facts and observations
[F&Os] from peer reviews). This is to include a comparison of the requirements of Regulatory Guide (RG) 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (Reference 3), using the elements of PRA standard American Society of Mechanical Engineers (ASME) RA-Sb-2005, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum B to ASME-RA-S-2002 (Reference 4),
for capability category II for internal events PRA models, and for other models for which RG 1.200, Revision 2 endorsed standards exist.
It is noted that the current accepted revision of RG 1.200 for use with RMTS implementation is Revision 2, which was approved subsequent to the issuance of the NRC safety evaluation for NEI 06-09-A, and which endorsed PRA standard ASME/American Nuclear Society (ANS) RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, (Reference 5).
This enclosure provides information on the technical acceptability of the Diablo Canyon PRA internal event, internal flood, fire, and seismic models that support the Risk-Informed Completion Time (RICT) Program, in support of the LAR to revise the Technical Specifications to adopt Technical Specifications Task Force (TSTF) traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b (Reference 6). This information is consistent with the requirements of Section 4.0, Item 3 of the NRC safety evaluation of NEI 06-09-A, and addresses each PRA model for which a PRA standard endorsed by RG 1.200, Revision 2 exists.
Meeting these requirements satisfies RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3 (Reference 7), requirements for risk-informed plant-specific changes to a plant's licensing basis. This enclosure addresses the clarifications and qualifications found in RG 1.200, Revision 2 as the peer reviews and self-assessment performed included consideration of the clarifications and qualifications.
Note that other external hazards are not addressed by PRA models, and are further discussed in Enclosure 4. Shutdown modes of operation are not in the scope of the RICT Program, and therefore low power and shutdown PRA models are not addressed.
Pacific Gas and Electric Company (PG&E) employs a multi-faceted approach to establishing and maintaining the technical adequacy and fidelity of the PRA models for

PG&E Letter DCL-23-054 Enclosure 2 Page 3 of 8 Diablo Canyon. This approach includes both a PRA maintenance and update process, and the use of self-assessments and independent peer reviews.
The Diablo Canyon PRA models are at-power models consisting of four hazard models:
internal events, internal flooding, internal fire, and seismic events. Each hazard model applies the internal events model as the base model. The models can evaluate both the core damage frequency and large early release frequency.
Peer Review and Peer Review Findings Closure Process All of the PRA models discussed in this Enclosure have been peer reviewed and assessed using PRA Standard ASME/ANS RA-Sa-2009 and RG 1.200, Revision 2.
Each peer review identified F&Os for supporting requirements of the relevant parts of the PRA standard applicable to the scope of the peer review. These included: findings for elements that did not meet at least capability category II of a supporting requirement of the standard, suggestions from the peer review team for elements that met the supporting requirement but could be improved, and best practices.
The review and closure of finding-level F&Os was performed by an independent assessment team using the process documented in Appendix X to NEI 05-04/07-12/12-16, Close-out of Facts and Observations (F&Os) (Reference 8). All of the reviews met the requirements of NEI 17-07, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard, Revision 2 (Reference 9).
Each assessment team (internal events including internal flooding, fire, and seismic) evaluated whether each F&O was closed through the application of a PRA maintenance or upgrade activity, as defined by the PRA standard. If closure of an F&O was identified as an upgrade, a focused scope peer review was conducted. Further, the assessment team re-evaluated any supporting requirements identified by the peer review to be either not met, or met at capability category I, to determine whether closure of the associated F&O(s) resulted in a change in status to either met, or met at least at capability category II.
The PRA scope and technical adequacy is met for this application as the applicable PRA Standard supporting requirements for all models are met at capability category II or higher. There are no remaining open finding-level F&Os for any of the models discussed in this application, and all finding-level F&Os have been independently assessed and closed using the processes discussed above. The resolved findings and the basis for resolution are documented in the Diablo Canyon PRA documentation and the F&O Closure Review reports. The results of the peer reviews and independent assessments have been documented and are available for NRC audit.
Internal Event and Internal Flood PRA An internal event and internal flood PRA peer review was conducted in December 2012, and is documented in LTR-RAM-II-13-002, RG 1.200 PRA Peer Review Against the ASME/ANS PRA Standard Requirements for the Diablo Canyon Nuclear Plant

PG&E Letter DCL-23-054 Enclosure 2 Page 4 of 8 Probabilistic Risk Assessment (Reference 10). The full-scope peer review of these models was performed consistent with RG 1.200, Revision 2, using the current endorsed PRA Standard ASME/ANS RA-Sa-2009.
All F&Os categorized by the peer review team as findings have been resolved by either a PRA model revision or a documentation update.
An independent assessment of the finding-level F&Os was conducted in June 2023 and is documented in PWROG-23015-P, Diablo Canyon F&O Closure and Focused Scope Peer Review Report (Reference 11). The review was conducted in accordance with Appendix X to NEI 05-04/07-12/12-16. The scope of the assessment included all finding-level F&Os resulting from the peer review.
Five suggestion-level F&Os in high-level requirement Large Early Release were identified as upgrades because the supporting requirements were met at capability category I only. Seven internal flooding F&Os were identified by PG&E as upgrades. A focused scope peer review was therefore conducted in conjunction with the closure review for these 12 F&Os. No other F&Os were determined to constitute an upgrade, and the use of any new methods was not identified by the assessment team.
At the conclusion of the independent assessment and focused-scope peer reviews, all applicable supporting requirements of the PRA standard are met, and supporting requirements that distinguish different capability categories satisfy at least capability category II. There are no remaining open peer review finding-level F&Os.
Therefore, the Diablo Canyon internal events and internal flooding PRA model is acceptable for use in the RICT Program.
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 and is documented in LTR-RAM-II-08-019, Pilot Application of the Fire PRA Peer Review Process for the Diablo Canyon Power Plant Fire Probabilistic Risk Assessment (Reference 12). The 2008 peer review was conducted against the requirements of the ANS Standard ANSI/ANS-58.23-2007, FPRA Methodology (Reference 13). At the time of this first peer review, certain technical elements of the fire PRA had not been completed.
The second phase of the peer review was completed in December 2010 and is documented in LTR-RAM-II-11-004, Fire PRA Peer Review Against the Fire PRA Standard SRs From Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the Diablo Canyon Plant Fire Probabilistic Risk Assessment (Reference 14). The 2010 peer review was conducted against the requirements of ASME/ANS RA-Sa-2009. The scope of the 2010 review included re-review of elements from the 2008 review that did not meet at least capability category II of the PRA standard.

PG&E Letter DCL-23-054 Enclosure 2 Page 5 of 8 After the final peer review in 2010, there were 17 identified finding-level F&Os. All F&Os categorized as findings have been resolved by either a PRA model revision or a documentation update.
An independent assessment of the F&Os was conducted in August-September 2018, and is documented in report P3118-004-001, F&O Closeout by Independent Assessment Report for the Diablo Canyon Nuclear Power Plant (DCPP) Fire PRA Model Against the 2009 ASME PRA Standard Requirements and NEI 05-04 Appendix X (Reference 15). The review was conducted in accordance with Appendix X to NEI 05-04/07-12/12-16. The scope of the assessment included all 17 F&Os resulting from the two peer reviews. Two F&Os were identified by PG&E as upgrades, and a focused-scope peer review was therefore conducted in conjunction with the closure review. No other F&Os were determined to constitute an upgrade, and the use of any new methods was not identified by the assessment team.
At the conclusion of the independent assessment and focused-scope peer review, all applicable supporting requirements of the PRA standard are met, and supporting requirements that distinguish different capability categories satisfy at least capability category II. There are no remaining open peer review finding-level F&Os.
Therefore, the Diablo Canyon fire PRA model is acceptable for use in the RICT Program.
It is noted that the Diablo Canyon fire PRA model was reviewed by the NRC as part of the Diablo Canyon NFPA-805 LAR dated June 26, 2013. The NRC review was concluded on April 14, 2016. Based on the staffs review, the NRC staff concluded that the Diablo Canyon fire PRA is of sufficient technical adequacy and that its quantitative results, considered together with the sensitivity studies, can be used to demonstrate that the change in risk due to the transition to NFPA 805 meets the acceptance guidelines in RG 1.174, Revision 2. (It is noted that the current RG 1.174, Revision 3 did not modify the acceptance guidelines found in RG 1.174, Revision 2.)
Seismic PRA A seismic PRA peer review was conducted in June 2017 and is documented in the Pressurized Water Reactor Owners Group (PWROG) report PWROG-17022-P, Peer Review of the Diablo Canyon Units 1 & 2 Seismic Probabilistic Risk Assessment, September 2017 (Reference 16). The full-scope peer review, which also included a review of the seismic hazard and fragility analyses, was performed consistent with RG 1.200, Revision 2, using PRA Standard ASME/ANS RA-Sb-2013 (Reference 17). All F&Os categorized as findings have been resolved by either a PRA model revision or a documentation update.
An independent assessment of the finding-level F&Os was conducted in October-December 2017 and is documented in PWROG-17078-P, Independent Assessment of Facts & Observations Closure and Focused Scope Peer Review of the Diablo Canyon

PG&E Letter DCL-23-054 Enclosure 2 Page 6 of 8 Units 1 & 2 Seismic Probabilistic Risk Assessment, Revision 0 (Reference 18). The scope of the assessment included all finding-level F&Os resulting from the 2013 peer review. Three F&Os were identified by PG&E as upgrades, and two additional F&Os were identified by the assessment team as upgrades; therefore, a focused-scope peer review was conducted in conjunction with the closure review. The use of any new methods was not identified by the assessment team.
At the conclusion of the independent assessment and focused-scope peer review, all applicable supporting requirements of the PRA standard are met, and supporting requirements that distinguish different capability categories satisfy at least capability category II. There are no remaining open peer review finding-level F&Os.
Therefore, the Diablo Canyon seismic PRA model is acceptable for use in the RICT Program.
It is noted that the Diablo Canyon seismic PRA model was submitted to the NRC for review in response to a 10CFR50.54(f) letter regarding lessons learned from the accident at the Fukushima Daiichi Nuclear Power Plant. The NRC review was concluded on January 22, 2019.It concluded that the seismic PRA is of sufficient technical adequacy to support phase 2 regulatory decision-making in accordance with the intent of the 50.54(f) letter.
Additional Information on the Use of FLEX Equipment The Diablo Canyon PRA models do not credit any FLEX equipment.
Operator actions that model FLEX strategies to shed vital DC loads and to manually control the turbine driven auxiliary feedwater pump are included in the seismic PRA model for a seismically induced Station Black Out (SBO) or SBO 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.

PG&E Letter DCL-23-054 Enclosure 2 Page 7 of 8 References
: 1. Letter from Jennifer M. Golder (USNRC - NRR) to Biff Bradley (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines, (TAC No. MD4995), May 17, 2007. (ADAMS Accession No. ML071200238).
: 2. NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, Revision 0, November 2006. (ADAMS Accession No. ML12286A322).
: 3. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009. (ADAMS Accession No. ML090410014).
: 4. ASME RA-Sb-2005, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum B to ASME RA-S-2002, ASME, New York, NY, December 30, 2005.
: 5. ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.
: 6. TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times -
RITSTF Initiative 4b. (ADAMS Accession No. ML18183A493).
: 7. Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"
Revision 3, January 2018. (ADAMS Accession No. ML17317A256).
: 8. NEI Letter to NRC, Final Revision of Appendix X to NEI 05-04/07-12/12-16, "Close-out of Facts and Observations (F&Os), February 21, 2017. (ADAMS Accession No. ML17086A431).
: 9. NEI 17-07, Revision 2, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard, August 2019. (ADAMS Accession No. ML19228A242).
: 10. Westinghouse Letter LTR-RAM-II-13-002, RG 1.200 PRA Peer Review Against the ASME/ANS PRA Standard Requirements for the Diablo Canyon Nuclear Plant Probabilistic Risk Assessment, March 20, 2013.
: 11. PWROG Report PWROG-23015-P, Diablo Canyon F&O Closure and Focused Scope Peer Review Report, July, 2023.
: 12. Westinghouse Letter LTR-RAM-II-08-019, Pilot Application of the Fire PRA Peer Review Process for the Diablo Canyon Power Plant Fire Probabilistic Risk Assessment, October 17, 2008.
: 13. American Nuclear Society (ANS) Standard ANSI/ANS-58.23-2007, FPRA Methodology.

PG&E Letter DCL-23-054 Enclosure 2 Page 8 of 8
: 14. Westinghouse Letter LTR-RAM-II-11-004, Fire PRA Peer Review Against the Fire PRA Standard SRs From Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for the Diablo Canyon Plant Fire Probabilistic Risk Assessment, May 24, 2011.
: 15. Engineering Planning and Management, Inc., P3118-004-001, F&O Closeout by Independent Assessment Report for the Diablo Canyon Nuclear Power Plant (DCPP) Fire PRA Model Against the 2009 ASME PRA Standard Requirements and NEI 05-04 Appendix X, Revision 0, September 2018.
: 16. PWROG Report PWROG-17022-P, Peer Review of the Diablo Canyon Units 1 &
2 Seismic Probabilistic Risk Assessment, September 2017, Revision 0, dated May 1, 2013.
: 17. ASME/ANS RA-Sb-2013, Addendum B to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, dated September 30, 2013.
: 18. PWROG Report PWROG-17078-P, Independent Assessment of Facts &
Observations Closure and Focused Scope Peer Review of the Diablo Canyon Units 1 & 2 Seismic Probabilistic Risk Assessment, Revision 0, March 2018.

PG&E Letter DCL-23-054 Enclosure 3 Page 1 of 1 Enclosure 3 Information Supporting Technical Adequacy of Probabilistic Risk Assessment (PRA) Models without Standards Endorsed by Regulatory Guide 1.200, Revision 2 This enclosure is not applicable to the Diablo Canyon submittal. Pacific Gas & Electric Company is not proposing to use any probabilistic risk assessment models in its Risk-Informed Completion Time Program for which a standard, endorsed by the NRC in Regulatory Guide 1.200, Revision 2, does not exist.

PG&E Letter DCL-23-054 Enclosure 4 Page 1 of 15 Enclosure 4 Information Supporting the Justification of Excluding Sources of Risk Not Addressed by the Diablo Canyon Probabilistic Risk Assessment (PRA)
Models

PG&E Letter DCL-23-054 Enclosure 4 Page 2 of 15 Introduction Section 4.0, Item 5 of the Nuclear Regulatory Commissions (NRCs) Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (Reference 2), requires that the license amendment request provide a justification for excluding any risk sources determined to be insignificant to the calculation of configuration-specific risk, and provide a discussion of any conservative or bounding analyses to be applied to the calculation of risk-informed completion times (RICTs) for sources of risk not addressed by the probabilistic risk assessment (PRA) models.
This enclosure addresses this requirement by discussing the generic methodology used to identify and disposition such risk sources and provides the Diablo Canyon specific results of the application of the generic methodology for impacts on the RICT Program.
NEI 06-09-A does not provide a specific list of hazards to be considered in the RICT Program. In order to identify a comprehensive listing of other external hazards for consideration, the PRA Standard American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS), Addendum A to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009 (Reference 3) supporting requirement EXT-A1 was used which includes a review of the following sources:
NUREG/CR-2300, PRA Procedures Guide (Reference 4)
NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities (Reference 5)
NUREG/CR-5042, Evaluation of External Hazards to Nuclear Power Plants in the United States (Reference 6)
NUREG-1150 Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants (Reference 7)
Appendix 6-A of the ASME/ANS PRA Standard In addition to the above generic sources, the Diablo Canyon Final Safety Analysis Report Update (Reference 8) was reviewed to identify any site-specific hazards consistent with the ASME/ANS PRA Standard supporting requirement EXT-A2. Based on this review, no additional external hazards were identified that are not already listed in the generic references.
Technical Approach The guidance contained in NEI 06-09-A states that all hazards that contribute significantly to the incremental risk of a configuration must be quantitatively addressed in the implementation of the RMTS. The following approach focuses on the risk implications of specific external hazards as they impact the determination of RICTs.

PG&E Letter DCL-23-054 Enclosure 4 Page 3 of 15 The other external hazards are evaluated using a preliminary screening and a quantitative screening using the criteria of the ASME/ANS PRA Standard.
The ASME/ANS PRA Standard identifies external hazard screening criteria (supporting requirements EXT-B1, EXT-B2, and EXT-C2), identified below:
Supporting Requirement EXT-B1:
(1) The hazard would result in equal or lesser damage than the events for which the plant has been designed. This requires an evaluation of plant design bases to estimate the resistance of plant structures and systems to a particular external hazard.
(2) The hazard has a significantly lower mean frequency of occurrence than another event (taking into account the uncertainties in the estimates of both frequencies),
and the hazard could not result in worse consequences than the other event.
(3) The hazard cannot occur close enough to the plant to affect it. Application of this criterion needs to take into account the range of magnitudes of the hazard for the recurrence frequencies of interest.
(4) The hazard is included in the definition of another event.
(5) The hazard is slow in developing, and it can be demonstrated that sufficient time exists to eliminate the source of the threat or to provide an adequate response.
Supporting Requirement EXT-B2 SRP For screening out an external hazard other than seismic events, the design basis for the hazard meets the criteria in the NRC Standard Review Plan (SRP) with justification of the screening if based solely on conformance to SRP.
The ASME/ANS PRA Standard also requires the above qualitative screening to be supported by a review of information on the plants design hazard and licensing basis relevant to the event screened (supporting requirement EXT-B3) as well as a review of any significant changes since the issuance of the original plant operating license for selected events (supporting requirement EXT-B4).

PG&E Letter DCL-23-054 Enclosure 4 Page 4 of 15 For hazards other than internal events, internal flooding, internal fire, and seismic events, the following criteria provide an acceptable basis for a bounding analysis for a demonstrably conservative analysis per the ASME/ANS PRA Standard:
Supporting Requirement EXT-C1 A The current design-basis hazard event has mean frequency less than 10-5/year, and the mean value of the conditional core damage probability (CCDP) is assessed to be less than 10-1.
B The core damage frequency, calculated using a bounding or demonstrably conservative analysis, has a mean frequency of less than 10-6/year.
As allowed in the ASME/ANS PRA Standard supporting requirements EXT-C2 through EXT-C6, the quantitative screening analyses use the mean frequency and other parameters of design-basis hazards. For the remaining other hazards, either realistic or conservative models (i.e., an Internal Event model that meets the system-analysis requirements in Part 2 of the ASME/ANS PRA Standard) identifying those structures, systems, and components (SSCs) vulnerable to the hazard and data (i.e., the hazard analysis and any fragility analysis) were used.
Table E4-1 provides the external hazards evaluated, identifies the applicable screening criteria, summarizes the evaluation, and provides a disposition for the RICT Program.

PG&E Letter DCL-23-054 Enclosure 4 Page 5 of 15 Table E4-1 Evaluation of Risks from External Hazards External     Screening                         Evaluation                               Disposition for RICT Program Hazard        Criteria Aircraft Impact SRP, B     The total core damage frequency (CDF) induced by         Projected air traffic from the small airport an aircraft crash at Diablo Canyon Unit 1 is 7.43-07 per and airways does not pose a significant year. Unit 2 is expected to have a similar risk from     safety impact to Diablo Canyon based on aircrafts due to the shared building structures and      the design of the facility and the low near identical non-shared building structures.           frequency of core damage due to such events. Impacts that were evaluated to cause structural failures result in direct core damage, and therefore have no impact on the RICT Program calculations. The consequences of a lesser impact are bounded by a loss-of-offsite power (LOOP) event, which is already included in the PRA model. It is therefore concluded that no additional PRA model for aircraft impacts is required in order to assess configuration risk for the RICT Program.
Avalanche       3          Location of the site does not support heavy snowfall     Since the event is not credible, no PRA and accumulation that may cause an avalanche.            model for an avalanche is required to assess configuration risk for the RICT Program.
Biological      4, 5       Excessive fouling by slime is not expected in a 24-      Biological events are bounded by the Event                      hour period. Slime buildup occurs over a period of       existing internal events PRA modeling, several weeks and is controlled by chlorination over     therefore, no additional PRA model for the long term. Rapidly occurring biological plugging     biological events is required to assess can occur at Diablo Canyon, such as an intermittent      configuration risk for the RICT Program.
concentration of salp, a gelatinous marine invertebrate at the cooling water intake cove that can cause the plant to ramp down in power, with the

PG&E Letter DCL-23-054 Enclosure 4 Page 6 of 15 External Screening                           Evaluation                              Disposition for RICT Program Hazard     Criteria possibility of a reactor trip or a loss of condenser vacuum. Both of these consequences are currently modeled in the PRA as reactor trip and loss of condenser vacuum initiators. Since the impact of the above biological events is accounted for via the associated internal initiating events, these biological events are screened from further analysis.
Coastal     4, SRP     This is a very slow process; there is a long lead time    Due to the long lead time available to Erosion                to respond by placing the units into cold shutdown.       respond to coastal erosion, the plant would The bedrock beneath the power plant site occupies         be in cold shutdown and therefore no the southerly flank of a major syncline that trends       additional PRA model for the impact of west to northwest. No evidence of a major fault has       coastal erosion is required to assess been recognized within or near the coastal area, and      configuration risk for the RICT Program.
the bedrock relationships in the exploratory trenches positively indicate that no such fault is present within the area of the plant site.
Drought     3          The ultimate heat sink is the Pacific Ocean; the plant    Since the event has no adverse impact, no is not adversely impacted by drought conditions.          PRA model for drought conditions is required to assess configuration risk for the RICT Program.
External    SRP, A, B  It is unlikely that the reservoirs can fail in such a way External flooding scenarios do not pose a Flooding               to pose a threat to the plant; however, a worst case      significant safety impact to Diablo Canyon scenario is still evaluated to conservatively estimate    based on the design of the facility and the hazard. Reservoirs 1-A and 1-B are holding            conformance to the SRP. It is therefore reservoirs (Reservoir 1-B is behind 1-A, and is located   concluded that no PRA model for external nearly 500 feet from the edge of the hillside).           flooding scenarios is required in order to Assuming that both reservoirs lose all their water and    assess configuration risk for the RICT that the entire volume of water flows toward the plant,   Program.
the area covered by the flood is taken to be the triangle formed by the closest point of Reservoir 1-A to the plant (800 feet), and the north and south sides

PG&E Letter DCL-23-054 Enclosure 4 Page 7 of 15 External  Screening                         Evaluation                               Disposition for RICT Program Hazard     Criteria of the plant (800 feet). This area is approximately 320,000 square feet. If the entire reservoir inventory is applied to this area, the depth of flooding will be approximately 2 feet at the back of the plant. The flood will only be temporary and not sustained.
Results of the hydrologic and hydraulic analysis indicate that no safety-related SSCs are inundated by a probable maximum flood (PMF). The 230kV switchyard (non-safety related) would be inundated during the PMF event; however, this type of event is already included in the definition of a severe weather-related LOOP. All other Diablo Canyon facilities and site features remain above the calculated PMF water surface levels, including the intake structure and the entire power block, which consists of the fuel handling building, the auxiliary building, the turbine building, and the two containment buildings.
Extreme Wind B          The conservative strike frequency of a tornado is         A conservative evaluation of an extreme or Tornado              7.0 x 10-5 per year. The CCDP for LOOP due to             wind or tornado event demonstrated an severe weather with no recovery is estimated to be        insignificant contribution to CDF. It is 5.16 x 10-4. The conservatively estimated CDF for a       therefore concluded that no PRA model for tornado event is then 3.92 x 10-8 per year.               extreme winds and tornadoes is required in order to assess configuration risk for the Tornado missile scenarios have been conservatively RICT Program.
evaluated with a CDF of 2.05 x 10-7 per year (Unit 2 has a similar impact).
Fog          4          There is no direct impact to CDF and LERF due to          Since the event has no direct adverse fog; however the indirect impact of fog, such as          impact not already addressed by another impact on aircraft crash frequency, accident data         events, no PRA model for fog conditions is includes the effect of fog.                               required to assess configuration risk for the RICT Program.

PG&E Letter DCL-23-054 Enclosure 4 Page 8 of 15 External Screening                         Evaluation                             Disposition for RICT Program Hazard     Criteria Forest or   1, 4       The area immediately around the plant site boundary    External fire impacts are bounded by the Range Fire             is not heavily wooded, and is adjacent to the Pacific  existing internal events PRA modeling, Ocean. The hazard from external fires to the plant is  therefore, no additional PRA model for remote and the impact of external fires on the offsite external fires is required to assess grid have been accounted for in the LOOP initiating    configuration risk for the RICT Program.
events.
Frost       4          Frost may impact the switchyard and grid. The          Frost impacts are bounded by the existing frequency of a LOOP initiator includes the impact of   internal events PRA modeling; therefore, no frost, and the contribution of frost is judged to be   additional PRA model for frost events is negligible.                                            required to assess configuration risk for the RICT Program.
Hail        4          The impact of hail on offsite power is included in the Hail impacts are bounded by the existing frequency of LOOP analysis. The contribution to the    internal events PRA modeling; therefore, no overall risk is judged to be negligible.               additional PRA model for hail events is required to assess configuration risk for the RICT Program.
Heavy Load  SRP        Maintenance activities are the cause of heavy load     A heavy load drop is not judged to have any Drop                   drops, and as such, are controlled and evaluated       significant impact on the calculation of under the 10 CFR 50.65 (a)(4) risk assessment          RICTs.
process on a case-by-case basis. The DCPP design basis for heavy load drops, single-failure-proof heavy load handling systems and the control of heavy load program at the plant satisfy the SRP Screening Criteria.
High Summer 4          The impact of a high-temperature environment on        Since there is no unique impact on plant Temperature            equipment performance is included in equipment         operation not already considered in the failure data.                                          PRA models, no additional PRA model for high summer temperature events is required to assess configuration risk for the RICT Program.

PG&E Letter DCL-23-054 Enclosure 4 Page 9 of 15 External    Screening                          Evaluation                               Disposition for RICT Program Hazard       Criteria High Tide,     4          The impact is already considered for External            Since the event has no adverse impact not Lake Level, or            Flooding.                                                already addressed by another event, no River Stage                                                                        PRA model for high water conditions is required to assess configuration risk for the RICT Program.
Hurricane      B          Conservatively assuming a hurricane with a wind          The frequency of a hurricane leading to speed of 150 mph leads to core damage, this yields a     core damage is well below 1 x 10-6 per CDF of 5.0 x 10-7 per year, which is below the           year, It is therefore concluded that no PRA screening criterion B. Therefore, it is judged that      model for hurricanes is required in order to hurricane-initiated scenarios are insignificant          assess configuration risk for the RICT contributors to the overall CDF.                         Program.
Ice Cover      4          May impact the switchyard and grid. The frequency of     Ice impacts are bounded by the existing a LOOP initiator includes the impact of ice cover.       internal events PRA modeling, so no additional PRA model for ice events is required to assess configuration risk for the RICT Program.
Industrial or  3          Nearby industrial and military facilities with the       Nearby facility accidents do not pose a Military                  potential to store or use hazardous materials are all    significant safety impact to Diablo Canyon.
Facility                  located at distances greater than five miles from the    It is therefore concluded that no PRA model Accident                  site. Chemicals stored, used, or situated at distances   for nearby facility accidents is required in greater than five miles from the plant do not need to    order to assess configuration risk for the be considered because, if a release occurs at such a     RICT Program.
distance, atmospheric dispersion will dilute and disperse the incoming plume to such a degree that either the toxic limits would never be reached or there would be sufficient time for the control room operators to take appropriate action. In addition, the probability of a plume remaining within a given sector for a long period of time is small. Due to very limited industry within San Luis Obispo County and the distances involved, any hazardous products or materials

PG&E Letter DCL-23-054 Enclosure 4 Page 10 of 15 External   Screening                          Evaluation                               Disposition for RICT Program Hazard      Criteria manufactured, stored, or processed in the areas beyond five miles from the site are not considered to be a significant hazard to the plant and, as such, the explosion, fire, and toxic gas hazards can be screened out.
Intense       4, A       The water depth above the door thresholds and areas       Since there are no adverse impacts from Precipitation            to the west of the turbine and buttress buildings due     intense precipitation events, no PRA model to a local intense precipitation event varied between     is required in order to assess configuration 0.05 ft. and 0.68 ft., with six of the doors/areas        risk for the RICT Program.
showing no inundation. The total force due to hydrostatic and hydrodynamic loading from the event was generally small for all the doors and safety-related structures, varying from 1 to 35 lb/ft for doors and areas experiencing inundation. Forces due to the associated local intense precipitation flood event effects will not adversely impact the doors or power block and surrounding structures. The safety-related fuel oil transfer equipment is elevated six inches above grade, and therefore would not experience any flooding.
Landslide     3          Earthquake loading as a result of an earthquake on        Since there are no adverse impacts from the Hosgri fault zone following periods of prolonged      landslide events, no PRA model is required precipitation will not produce any significant slope      in order to assess configuration risk for the failure that can impact the Class I structures and        RICT Program.
equipment. In addition, potential slope failures under such conditions will not adversely impact other important facilities, including the raw water reservoirs, the 230 kV and 500 kV switchyards, and the intake and discharge structures. Potential landslides may temporarily block the normal paved south access road at several locations. However, there is considerable

PG&E Letter DCL-23-054 Enclosure 4 Page 11 of 15 External    Screening                          Evaluation                            Disposition for RICT Program Hazard        Criteria room adjacent to and north of the paved road to reroute emergency traffic. There is also an unpaved north access road that may be used. Therefore, landslides can be screened out.
Lightning      1, 4       The plant contains lightning protection in the plant   Since the event has no adverse impact not design. The impact on offsite power is included in the already addressed by another event, no LOOP frequency evaluation. The contribution to the     PRA model for lightning events is required overall risk is judged to be negligible.               to assess configuration risk for the RICT Program.
Low Lake       3          Not applicable to Diablo Canyon.                       Since the event is not applicable, no PRA Level or River                                                                   model is required to assess configuration Stage                                                                            risk for the RICT Program.
Low Winter     1, 4       Evaluation performed to assess potential for cold      Since there is no unique impact on plant Temperature               weather to impact Diablo Canyon reliability has        operation not already considered in the concluded the probability for cold weather impact is   PRA models, no additional PRA model for low. The impact on equipment has been included in      low winter temperature events is required to the component (independent and common-cause)           assess configuration risk for the RICT failure rates. Thermal stresses and embrittlement are  Program.
usually insignificant and covered by design codes and standards for the plant design.
Meteorite or   2          The probability of the event is less than 1 x 10-9 per A meteorite or satellite impact is not judged Satellite                 year per Reference 6.                                  to have any significant impact on the Impact                                                                           calculation of RICTs.
Pipeline       3, 4       No natural gas or other pipelines pass within five     There are no pipelines in sufficient Accident                  miles of the plant site. The onsite hazardous buried   proximity to the plant site to cause a piping at the Diablo Canyon plant site are those       significant hazard. Since there is no unique owned by PG&E; which may carry diesel fuel oil,        impact on plant operation not already hydrogen, etc. However, fire and rapid combustion      considered in the PRA models, no resulting from fuel oil and hydrogen are evaluated     additional PRA model for pipeline accidents separately in the Diablo Canyon Fire PRA. As such,

PG&E Letter DCL-23-054 Enclosure 4 Page 12 of 15 External  Screening                          Evaluation                            Disposition for RICT Program Hazard      Criteria rupture of the hydrogen line and the potential rapid   is required to assess configuration risk for combustion that may result are not re-evaluated. The   the RICT Program.
other buried pipes containing hydrocarbons are mainly the diesel fuel oil and waste oil pipes; which are not in the form of toxic gas and have an extremely low likelihood of rapid combustion.
Release of   1          Hazards due to explosion, toxicity, or asphyxiation    There are no chemicals onsite that can Chemicals in            were evaluated and it was concluded that they pose     cause a significant safety hazard. It is Onsite                  no hazard to the control room personnel and PRA        therefore concluded that no PRA model for Storage                 equipment. Any toxic gas that may be generated from    chemical releases is required in order to the accidental release of onsite chemicals would not   assess configuration risk for the RICT impact the control room habitability. Therefore,       Program.
release of chemicals in onsite storage can be screened out.
River        3          Not applicable to the Diablo Canyon site.              Since the event is not applicable, no PRA Diversion                                                                      model is required to assess configuration risk for the RICT Program.
Sandstorm    3, 4       Sandstorms are included in the evaluation for extreme  A sandstorm impact is judged to not have winds and tornados. They are judged to be              any significant impact on the calculation of insignificant in occurrence, frequency, and risk.      RICTs.
Seiche       1, SRP     Seiche effects on intake cove wave heights are less    A seiche has no adverse impact and than 3.2 ft of the maximum crest wave level inside the therefore no PRA model is required to breakwaters of 12.8 ft, and are therefore, not a       assess configuration risk for the RICT concern. The maximum expected water volume loss        Program.
from each of the raw water storage reservoirs is 14,684 gallons. The raw water storage reservoirs are able to perform their design function with only one million gallons per reservoir. As such, loss of 14,684 gallons is not significant.

PG&E Letter DCL-23-054 Enclosure 4 Page 13 of 15 External    Screening                         Evaluation                            Disposition for RICT Program Hazard       Criteria Sink Holes     1, 5       The site suitability evaluation and site development  Sink hole impacts are not a credible event for the plant are designed to preclude the effects of therefore, no PRA model is required to this hazard.                                          assess configuration risk for the RICT Program.
Snow           3          The location of the site is such that it does not     Snow impacts are not a credible event experience heavy snowfall that could impact the       therefore, no PRA model is required to switchyard and grid.                                  assess configuration risk for the RICT Program.
Soil Shrink-   1, 5       This event is a slow process. Contribution to the     A slowly developing soil shrink-swell Swell                     overall risk is judged to be negligible. The site     consolidation event is not judged to have Consolidation             suitability evaluation and site development for the   any significant impact on the calculation of plant are designed to preclude the effects of this    RICTs.
hazard.
Storm Surge    4          The maximum estimated wave height outside the         A storm surge has no adverse impact and breakwaters was found to be 44.6 ft. The maximum      therefore no PRA model is required to crest wave level inside the breakwaters was 12.8 ft. assess configuration risk for the RICT While seiche effects were noted in the intake cove,   Program.
the wave heights were found to be less than 3.2 ft of the maximum estimated wave height, and are, therefore, not a concern.
Toxic Gas      4          See the table entries for Release of Chemical in     See the table entries for Release of Onsite Storage and Industrial Accidents.           Chemical in Onsite Storage and Industrial Accidents.
Transportation 3, B       Various scenarios involving shipping hazards to the   Transportation accidents involving ships Accident                  plant were analyzed. Scenarios involving ship         cannot cause damage to the plant under breakthrough over the breakwater in its normal state  normal conditions, and are shown to have a (not degraded by heavy wave action) were shown to     bounding CDF of significantly less than be not possible due to the speed required to generate 1 x 10-6 per year for degraded conditions. It the kinetic energy needed to physically force a       is therefore concluded that no PRA model passage through the breakwater. Scenarios involving   for transportation accidents is required in

PG&E Letter DCL-23-054 Enclosure 4 Page 14 of 15 External Screening                          Evaluation                                Disposition for RICT Program Hazard   Criteria oil spills and other floating debris were also shown to   order to assess configuration risk for the not have a significant consequence. Analysis              RICT Program.
scenarios involving a degraded breakwater, and therefore greatly increasing the possibility of a ship arriving in the intake cove, result in an estimated CDF of 2.90 x 10-8 per year. Scenarios involving a ship blocking the flow of water into the intake cove result in a conservatively estimated CDF of 2.91 x 10-8 per year. Both of these CDF frequencies are low enough to be screened out.
Tsunami    4, B       Flooding of the intake structure due to a tsunami has     A tsunami is shown to have a bounding an estimated CDF of 2.2 x 10-8 per year, which is low     CDF of significantly less than 1 x 10-6 per enough to be screened out.                                year. It is therefore concluded that no PRA model for tsunamis is required in order to assess configuration risk for the RICT Program.
Turbine-   SRP        Factory test procedures, redundancy in the control        Based on screening due to low probability, Generated             system, and routine testing of the main steam valves      it is concluded that no additional PRA Missiles              and the mechanical emergency over-speed protection        model for turbine missile accidents is system while the unit is carrying load make the           required in order to assess configuration generation of missiles by a turbine runaway that might    risk for the RICT Program.
penetrate the turbine casing highly improbable.
Therefore, turbine missiles can be screened out based upon conformance with the SRP.
Volcanic   3          Not applicable, no active volcanic mountains are near     Volcanic activity is not applicable to Diablo Activity              the plant.                                                Canyon; therefore, no PRA model is required in order to assess configuration risk for the RICT Program.
Waves      4          See the table entry for External Flooding.              See the table entry for External Flooding.

PG&E Letter DCL-23-054 Enclosure 4 Page 15 of 15 References
: 1. Letter from Jennifer M. Golder (USNRC - NRR) to Biff Bradley (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines, (TAC No. MD4995), May 17, 2007. (ADAMS Accession No. ML071200238).
: 2. NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, November 2006. (ADAMS Accession No. ML12286A322).
: 3. ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.
: 4. NUREG/CR-2300, PRA Procedures Guide, A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants, Volume 2, January 1983.
(ADAMS Accession Nos. ML063560440).
: 5. NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, April 1991. (ADAMS Accession No. ML063550238)
: 6. NUREG/CR-5042, Evaluation of External Hazards to Nuclear Power Plants in the United States, December 1987. (ADAMS Accession Nos. ML111950285, ML14196A083).
: 7. NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, December 1990. (ADAMS Accession Nos. ML040140729, 120960691, 16281A233, 16284A005).
: 8. Pacific Gas and Electric Company, Units 1 and 2, Diablo Canyon Power Plant Final Safety Analysis Report Update, (UFSAR), Section 2.

PG&E Letter DCL-23-054 Enclosure 5 Page 1 of 3 Enclosure 5 Baseline Core Damage Frequency (CDF) and Large Early Release Frequency (LERF)

PG&E Letter DCL-23-054 Enclosure 5 Page 2 of 3 Section 4.0, Item 6 of the Nuclear Regulatory Commissions (NRCs) Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (Reference 2), requires that the license amendment request provide the plant-specific total core damage frequency (CDF) and large early release frequency (LERF) to confirm applicability of the limits of Regulatory Guide (RG) 1.174, An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1 (Reference 3). (Note that RG 1.174, Revision 3
[Reference 4], issued by the NRC in January 2018, did not revise these limits.)
This enclosure demonstrates that the Diablo Canyon total CDF and total LERF are below the guidelines established in RG 1.174. RG 1.174 does not establish firm limits for total CDF and LERF, but recommends that risk-informed applications be implemented only when the total plant risk is no more than 1-4/year CDF and 1-5/year LERF. Demonstrating that these limits are met confirms that the risk metrics of NEI 06-09-A may be applied to the Diablo Canyon Risk-Informed Completion Time (RICT)
Program.
Table E5-1 provides the CDF and LERF values that resulted from a quantification of the baseline average annual models (designated DC05) which include contributions from internal events (including internal flooding), fire, and seismic hazards. Other external hazards are below the accepted screening criteria and therefore do not contribute significantly to the totals (see Enclosure 4).
Table E5-1: Total Baseline Average Annual CDF/ LERF Unit 1                                Unit 2 Hazard          CDF (per rx-yr)    LERF (per rx-yr)   CDF (per rx-yr) LERF (per rx-yr)
Internal Events         4.78E-6             9.89E-7           4.78E-6            9.89E-7 Internal Flooding        7.61E-6             2.45E-7           6.48E-6            2.09E-7 Seismic             2.96E-5             5.22E-6           2.96E-5            5.22E-6 Fire              4.60E-5             1.42E-6           3.99E-5            1.31E-6 Total              8.80E-5             7.88E-6           8.08E-5            7.73E-6 NOTE: The ASME/ANS PRA Standard, states the units "per reactor year" (per rx yr in this table) and "per calendar year" are equivalent.
As demonstrated in the table, the total CDF and total LERF for each unit are within the guidelines of RG 1.174 to permit small changes in risk that may occur during RICT Program implementation of extended Completion Times. Therefore, the Diablo Canyon RICT Program is consistent with the NEI 06-09-A guidance.

PG&E Letter DCL-23-054 Enclosure 5 Page 3 of 3 References
: 1. Letter from Jennifer M. Golder (USNRC - NRR) to Biff Bradley (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines, (TAC No. MD4995), May 17, 2007. (ADAMS Accession No. ML071200238).
: 2. NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, Revision 0, November 2006. (ADAMS Accession No. ML12286A322).
: 3. Regulatory Guide 1.174, An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1, November 2002. (ADAMS Accession No. ML023240437).
: 4. Regulatory Guide 1.174, An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, January 2018. (ADAMS Accession No. ML17317A256).

PG&E Letter DCL-23-054 Enclosure 6 Page 1 of 1 Enclosure 6 Justification of Application of At-Power Probabilistic Risk Assessment (PRA)
Models to Shutdown Modes This enclosure is not applicable to the Diablo Canyon submittal. Pacific Gas &
Electric Company is not proposing to apply the Risk-Informed Completion Time Program during shutdown modes.

PG&E Letter DCL-23-054 Enclosure 7 Page 1 of 5 Enclosure 7 Probabilistic Risk Assessment (PRA) Model Update Process

PG&E Letter DCL-23-054 Enclosure 7 Page 2 of 5 Summary Section 4.0, Item 8 of the Nuclear Regulatory Commissions (NRCs) Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (Reference 2), requires that the license amendment request (LAR) provide a discussion of the licensees programs and procedures which assure the probabilistic risk assessment (PRA) models that support the Risk-Informed Completion Time (RICT)
Program are maintained consistent with the as-built/as-operated plant.
This enclosure describes the administrative controls and procedural processes applicable to the configuration control and update of the PRA models used to support the RICT Program, to ensure that these models reflect the as-built/as-operated plant.
Plant changes, including modifications and procedure changes, will be identified, reviewed, and evaluated prior to implementation to determine whether they could impact the PRA models. The Configuration Control Program will ensure these plant changes are incorporated into the PRA models as appropriate. The process will include discovered conditions and errors associated with the PRA models, which will be addressed by the Diablo Canyon Corrective Action Program.
Should a plant change or a discovered condition be identified that has a potential significant impact to the RICT Program calculations as defined by the Configuration Control Program, an interim update of the PRA model will be implemented. Otherwise, the PRA model change is incorporated into a subsequent periodic model update. Such pending changes are considered when evaluating other changes until they are fully implemented into the PRA models.
PRA Model Update Process Internal Event, Internal Flood, Fire, and Seismic Event PRA Models Maintenance and Update The Diablo Canyon risk management process ensures that the applicable PRA model used for the RICT Program reflects the as-built/as-operated plant for each of the two Diablo Canyon units. The PRA configuration control process delineates the responsibilities and guidelines for updating the full power internal event, internal flood, fire, and seismic PRA models, and includes both periodic and interim PRA model updates. Any PRA model change that meets the criteria for an upgrade will be subject to a peer review.
Diablo Canyon Procedures TS3.NR1, Diablo Canyon Power Plant, Departmental Administrative Procedure, Probabilistic Risk Assessment (Reference 3), and AWP E-028, Diablo Canyon Power Plant, Administrative Work Procedure, PRA Model Maintenance and Upgrades (Reference 4), control the update and maintenance of the PRA models. These procedures are in full compliance with ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (Reference 5), for PRA model maintenance and update. Procedure AWP E-028 requires an update to be periodically initiated to reflect

PG&E Letter DCL-23-054 Enclosure 7 Page 3 of 5 industry operating experience, plant design, and procedure changes. An update of the PRA model data, including initiating event, component reliability, and common cause data, is also required. The process includes guidance on identifying, evaluating, and documenting potential impacts to the Diablo Canyon PRA models (e.g., due to plant changes, plant/industry operational experience, or errors or limitations identified in the model). Guidance is also provided for assessing the individual and cumulative risk impacts of pending changes, and controlling the versions of the PRA and Configuration Risk Management Program (CRMP) models.
Review of Plant Changes for Incorporation into the PRA Model (1) Plant changes or discovered conditions are reviewed for potential impact on the PRA models, including the CRMP model and the subsequent risk calculations that support the RICT Program (NEI 06-09-A: Section 2.3.4, Items 7.2 and 7.3, and 2.3.5, Items 9.2 and 9.3).
(2) Plant changes that meet the criteria defined in the PRA Configuration Control Program (including consideration of the cumulative impact of other pending changes) will be incorporated into the applicable PRA model(s) as an interim update, consistent with the NEI 06-09-A guidance. Otherwise, the change is assigned a priority and is incorporated at a subsequent periodic update consistent with procedural requirements. (NEI 06-09-A: Section 2.3.5, Item 9.2)
(3) PRA updates for plant changes are initiated at least once every two refueling cycles (not to exceed 48 months, which is acceptable because (1) a common PRA model is used for both Diablo Canyon units, (2) the Maintenance Rule Program implemented at Diablo Canyon provides assurance that component failures are addressed prior to the development of increasing trends in failure rates, and (3) the Mitigating System Performance Index Program implemented at Diablo Canyon provides assurance that plant changes that have a significant impact on plant risk, including configuration risk, require an interim PRA model/RICT model update). (NEI 06-09-A: Section 2.3.4, Item 7.1, and 2.3.5, Item 9.1).
(4) If a PRA model change is required for the CRMP model, but cannot be immediately implemented for a significant plant change or discovered condition, either:
A. Alternative analyses to conservatively bound the expected risk impact of the change will be performed. In such a case, these alternative analyses become part of the RICT Program calculation process until the plant changes are incorporated into the PRA model during the next update. The use of such bounding analyses is consistent with the guidance of NEI 06-09-A.
B. Appropriate administrative restrictions on the use of the RICT Program will be put in place until the model changes are completed, consistent with the guidance of NEI 06-09-A. (NEI 06-09-A: Section 2.3.5, Item 9.3).

PG&E Letter DCL-23-054 Enclosure 7 Page 4 of 5 Additional Information Regarding Modeling of Reactor Coolant Pump (RCP) Shutdown Seals TSTF-505, Revision 2 (Reference 6) included a recommended additional LAR content for Enclosure 7 to specifically address the reactor coolant pump (RCP) shutdown seals configuration, PRA modeling of these seals, and peer review of the model.
Each RCP in both units have the Westinghouse Low Leakage Shutdown RCP seals (GEN III SDS seals) installed. All hazard models (internal events including internal flooding, fire, and seismic) credit these seals. The PRA model used for the shutdown seals performance is PWROG-14001-P-A, PRA Model for the Generation III Westinghouse Shutdown Seal, Revision 1 (Reference 7). All the limitations and conditions in the NRC safety evaluation for this report were verified to be satisfied; no exceptions were taken.
The shutdown seal modeling was reviewed and determined to be a maintenance activity and not an upgrade during the fire PRA Appendix X closure review in 2018; therefore, no focused-scope peer review is required for the shutdown seal model implementation.

PG&E Letter DCL-23-054 Enclosure 7 Page 5 of 5 References
: 1. Letter from Jennifer M. Golder (USNRC - NRR) to Biff Bradley (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines, (TAC No. MD4995), May 17, 2007. (ADAMS Accession No. ML071200238).
: 2. NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, Revision 0, November 2006. (ADAMS Accession No. ML12286A322).
: 3. TS3.NR1, Diablo Canyon Power Plant Departmental Administrative Procedure, Probabilistic Risk Assessment.
: 4. AWP E-028, Diablo Canyon Power Plant, Administrative Work Procedure, PRA Model Maintenance and Upgrades.
: 5. ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.
: 6. TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times -
RITSTF Initiative 4b. (ADAMS Accession No. ML18183A493).
: 7. PWROG-14001-P-A, PRA Model for the Generation III Westinghouse Shutdown Seal, Revision 1, October 2017. (ADAMS Accession No. ML18019A215 for non-proprietary version of the report).

PG&E Letter DCL-23-054 Enclosure 8 Page 1 of 7 Enclosure 8 Attributes of the Configuration Risk Management Program (CRMP) Model

PG&E Letter DCL-23-054 Enclosure 8 Page 2 of 7

==
Introduction:==

Section 4.0, Item 9 of the Nuclear Regulatory Commissions (NRCs) Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (Reference 2), requires that the license amendment request (LAR) provide a description of probabilistic risk assessment (PRA) models and tools used to support the Risk Informed Completion Time (RICT) Program. This includes identification of how the baseline PRA model is modified for use in the Configuration Risk Management Program (CRMP), consistency of calculated results from the baseline PRA model to the CRMP model, quality requirements applied to the PRA and CRMP, and training and qualification programs applicable to personnel responsible for development and use of the CRMP. This item is also to confirm that the CRMP can be readily applied for each Technical Specification Limiting Condition for Operation within the scope of the plant-specific submittal.
This enclosure describes the necessary changes to the peer-reviewed baseline PRA models for use in the CRMP to support the RICT Program. The process used to adapt the baseline models is demonstrated:
: 1) To preserve the core damage frequency and large early release frequency quantitative results;
: 2) To maintain the quality of the peer-reviewed PRA models; and
: 3) To correctly accommodate changes in risk due to configuration-specific considerations.
Quality controls and training programs applicable for the CRMP are also discussed in this enclosure.
Translation of the Baseline PRA Model for Use in the CRMP:
The baseline PRA models for internal events, including internal floods, internal fires, and seismic events, are the peer-reviewed models, updated when necessary to incorporate plant changes to reflect the as-built/as-operated plant as discussed in . These models are modified to include changes that are needed to facilitate configuration-specific risk calculations to support the RICT Program implementation.
The baseline models, and the changes made to create the CRMP model used in the RICT Program, are controlled using plant calculations, which include all necessary quality controls and reviews.
The changes to the models needed to properly account for configuration-specific issues, which are either not evaluated in the baseline average annual model or are evaluated based on average conditions encountered during a typical operating cycle, are described in Table E8-1.

PG&E Letter DCL-23-054 Enclosure 8 Page 3 of 7 TABLE E8-1 CRMP MODEL CHANGES FOR CONFIGURATION-SPECIFIC RISK DESCRIPTION                                  BASIS FOR CHANGE Plant Availability  The baseline PRA models account for the time the reactor operates at power by using a plant availability factor. This is appropriate for determining the average annual (time-based) risk; however, the factor is not applicable to the configuration-specific risk calculated for the RICT Program. In order to account for the assumption that the plant is always operating in the RICT Program, the frequency of initiating events that include an availability factor are adjusted. This change is necessary to adjust the modeled initiating event frequencies from a per (calendar) year to per reactor (critical) year basis for use in the CRMP.
Maintenance         Maintenance alignment probabilities in the baseline PRA models Alignment           include probabilities that are based on the fraction of the year the Probabilities       equipment is unavailable. For the CRMP model, the actual configuration of equipment is evaluated; therefore, the maintenance alignment probabilities are set to zero. This is also done for the system initiating events that include maintenance contributions.
Excluded            The PRA models do not remove excluded maintenance combinations Maintenance         (i.e., both trains of single safety system being simultaneoulsy Combinations        unavailable); therefore no change to the CRMP model is required.
Room Cooling        The baseline PRA models include conservative success criteria for Success Criteria    room cooling and do not use average annual criteria; therefore no changes to the CRMP model for room cooling success criteria are required.
Unfavorable         The current Diablo Canyon core design reflected in the baseline PRA Exposure Time       model for ATWT events includes a UET for variable success criteria (UET) for           based on time of core life (i.e., the moderator temperature coefficient Anticipated         early in cycle life). The event is set to the fraction of the year for which Transient Without   the UET applies, and will be changed to a probability of 1 or 0 based Trip (ATWT) Events  on operator input using the CRMP tool, depending on the actual time in the operating cycle.
Diesel Generator    The baseline PRA model uses a six-hour mission time for the DGs and (DG) Mission Time   Diesel Fuel Oil Transfer System components. A 24-hour mission time is applied in the CRMP model when there is no OPERABLE offsite AC power source.
Portable Fuel Oil   The baseline PRA model includes credit for a backup portable fuel oil Pump                pump. This is not credited for the CRMP model when a diesel fuel oil transfer system pump is out of service.
Charging Pump       The baseline PRA model includes a recovery factor for a failed Recovery Factor     charging pump. This is not credited for the CRMP model when a charging pump is out of service.

PG&E Letter DCL-23-054 Enclosure 8 Page 4 of 7 Quality Requirements and Consistency of the PRA Model and CRMP The approach for establishing and maintaining the quality of the PRA models, including the CRMP model, includes both a PRA maintenance and update process (described in ), and the use of self-assessments and independent peer reviews (described in Enclosure 2).
The information provided in Enclosure 2 describes the technical adequacy of the Diablo Canyon internal event, internal flood, fire, and seismic PRA models to conform to the associated industry standards endorsed by Regulatory Guide (RG) 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 (Reference 4). This information provides a robust basis for concluding that the PRA models are of sufficient quality for use in the RICT Program.
For maintenance of an existing CRMP model, changes made to the baseline PRA model in translation to the CRMP model will be controlled and documented. Procedures address the process for identification and corrective actions to evaluate and disposition model errors and changes to ensure that the models are accurate, as described in . An acceptance test is performed after every CRMP model update to verify that the software works as intended and that quantification results are reasonable.
These above quality requirements satisfy NEI 06-09-A, Section 2.3.5, Item 9.
Training and Qualification The Diablo Canyon PRA staff is responsible for the development and maintenance of the CRMP model. The PRA staff is trained in accordance with the site Engineering Personnel Training Program. Operations and Work Control staffs will use the CRMP for the RICT Program, and these staff are trained in accordance with a program using National Academy for Nuclear Training documents, which is also accredited by the Institute of Nuclear Power Operations.
Application of the CRMP to the RICT Program Scope Diablo Canyon will use the Electric Power Research Institute (EPRI) software Phoenix, or equivalent. This program is specifically designed by EPRI to support implementation of the RMTS, and is compatible with the PRA model software used at Diablo Canyon.
Phoenix will permit the user to evaluate all configurations within the scope of the RICT Program at Diablo Canyon using appropriate mapping of equipment to PRA model elements.

PG&E Letter DCL-23-054 Enclosure 8 Page 5 of 7 Additional Information Regarding CRMP Modeling Technical Specifications Task Force (TSTF) traveler TSTF-505, Revision 2 (Reference
: 11) included a recommended additional LAR content for Enclosure 8 to specifically address the development and application of the CRMP model.
There are no seasonal variations in any success criteria for the Diablo Canyon baseline PRA models, and therefore there is no need to adjust the CRMP model.
The CRMP model will account for the impacts of out-of-service structures, systems, and components (SSCs) for both initiating events and event mitigation.
The Diablo Canyon model does not solve for cutsets, therefore there is no issue with maintaining the accuracy of pre-solved cutsets.
The process of creating the CRMP model from the baseline PRA models is controlled by procedures and includes a comparison of the results to ensure the fidelity of the CRMP model. The CRMP model is updated on the same schedule as the baseline PRA models.
Additional Information Regarding Common Cause Failure (CCF) Treatment TSTF-505, Revision 2 (Reference 11) included a recommended additional LAR content for Enclosure 8 to specifically discuss the treatment of CCF for planned maintenance and for emergent conditions.
The CCF modeling methodology follows the guidelines in NUREG/CR-4780, Procedures for Treating Common Cause Failures in Safety and Reliability Studies (Reference 6), and NUREG/CR-5485, Guidelines on Modeling Common-Cause Failures in Probabilistic Risk Assessment (Reference 7), for the identification and modeling of CCF events in the systems models. The CCF groups are generally identified from NUREG/CR-6268, Common-Cause Failure Database and Analysis System: Overview (Reference 8), and NUREG/CR-5497, Common-Cause Failure Parameter Estimations (Reference 9).
The alpha factor method is used to calculate CCF rates, then the results are represented in the system fault tree models as individual basic events for each k-out-of-n CCF failure where k represents the number of components failed out of n total components in the CCF group. For example, a two component CCF group would have two basic events, one-out-of-two failures and two-out-of-two failures.
The Diablo Canyon Riskman model is set up so that if train A is out of service, only the independent failure for train B is modeled; this is consistent with Regulatory Position 2.3.3 and Section A-1.3.2 of Appendix A of RG 1.177, Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications (Reference 10), for equipment removed from service for planned maintenance.

PG&E Letter DCL-23-054 Enclosure 8 Page 6 of 7 For emergent equipment issues, if the extent of condition evaluation is not complete or does not demonstrate that redundant components are not impacted, then risk management actions associated with the success of redundant SSCs and with reducing the likelihood of initiating events that rely on the affected function will be considered to address the potential increased likelihood of a common cause event. These requirements of the RICT Program are consistent with the guidance of Reference 2.

PG&E Letter DCL-23-054 Enclosure 8 Page 7 of 7 References
: 1. Letter from Jennifer M. Golder (USNRC - NRR) to Biff Bradley (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines, (TAC No. MD4995), May 17, 2007. (ADAMS Accession No. ML071200238)
: 2. NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, Revision 0, November 2006. (ADAMS Accession No. ML12286A322)
: 3. TS3.NR1, Diablo Canyon Power Plant Departmental Administrative Procedure, Probabilistic Risk Assessment.
: 4. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009. (ADAMS Accession No. ML090410014)
: 5. ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.
: 6. NUREG/CR-4780, Procedures for Treating Common Cause Failures in Safety and Reliability Studies, Volume 1, January 1988, and Volume 2, January 1989.
: 7. NUREG/CR-5485, Guidelines on Modeling Common-Cause Failures in Probabilistic Risk Assessment, June 1998.
: 8. NUREG/CR-6268, Common-Cause Failure Database and Analysis System: Event Data Collection, Classification, and Coding, Revision 1. (ADAMS Accession No. ML072970404)
: 9. NUREG/CR-5497, Common-Cause Failure Parameter Estimations, October 1998.
: 10. Regulatory Guide 1.177, Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, Revision 2. (ADAMS Accession No. ML20164A034)
: 11. TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times -
RITSTF Initiative 4b. (ADAMS Accession No. ML18183A493)

PG&E Letter DCL-23-054 Enclosure 9 Page 1 of 8 Enclosure 9 Key Assumptions and Sources of Uncertainty

PG&E Letter DCL-23-054 Enclosure 9 Page 2 of 8

==
Introduction:==

Section 4.0, Item 10 of the Nuclear Regulatory Commissions (NRCs) Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (Reference 2), requires that the license amendment request provide a discussion of how the key assumptions and sources of uncertainty were identified, and how their impact was assessed and dispositioned.
This enclosure provides a discussion of how the key assumptions and sources of uncertainty for each of the probabilistic risk assessment (PRA) models used for the Risk-Informed Completion Time (RICT) Program were identified, assessed, and dispositioned for the RICT Program.
Process for Identification of Key Assumptions and Sources of Uncertainty:
Sources of model uncertainty and related assumptions, are defined consistent with Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 (Reference 3) and ASME/ANS RA-Sa-2009, American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Probabilistic Risk Assessment (PRA) Standard (Reference 4), and have been identified for the Diablo Canyon baseline PRA models using the guidance of NUREG-1855, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making (Reference 5), and EPRI TR-1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessment (Reference 6). These were reviewed by the respective hazard PRA peer review teams and closure review teams (see Enclosure 2).
The detailed process of identifying, characterizing, and qualitative screening of model uncertainties is in Section 5.3 of Reference 5 and Section 3.1.1 of EPRI TR-1016737.
The process in these references was mostly developed to evaluate the uncertainties associated with the internal events PRA model; however, the approach can be applied to other types of hazard groups. The approach taken is to review the assumptions and sources of uncertainty for each PRA model to identify the items which may be directly relevant to the RICT program calculations.
Disposition of Key Assumptions and Sources of Uncertainty The list of assumptions and sources of uncertainty were reviewed to identify those that would be significant for the evaluation of configuration-specific changes in risk in the RICT Program. If a Diablo Canyon model uses a nonconservative treatment identified as having a non-negligible impact, or uses methods that are not commonly accepted, the underlying assumption or source of uncertainty was reviewed to determine the impact on the RICT Program calculations. This evaluation meets the intent of steps C-1 and E-1 of NUREG-1855. To assess the impact of these sources of uncertainties on the RICT Program application, a review of the base-case sources of uncertainty for the PRA

PG&E Letter DCL-23-054 Enclosure 9 Page 3 of 8 models was performed. Each identified uncertainty was evaluated with respect to its potential to significantly impact the calculated RICTs. This evaluation meets the intent of the screening portion for steps C-2 and E-2 of NUREG-1855.
Key assumptions and sources of uncertainty for the RICT Program application not screened as discussed above are identified and dispositioned in Table E9-1.

PG&E Letter DCL-23-054 Enclosure 9 Page 4 of 8 Table E9-1 Disposition of Key Assumptions/Sources of Uncertainty Impacting Configuration Risk Calculations Assumption/Uncertainty                          Discussion                         Disposition for the RICT Program Dual unit trips (except for seismic        The effects of dual unit trips and events Shared systems and equipment between the events) are not considered in the single   may not be considered in accident         units will be identified in procedures for RICT unit model, and crosstie to the other      sequences. This approach is               Program implementation so that consideration unit's resources may be unavailable.       nonconservative because the plant         of additional risk management actions will be equipment credited may be required by     made.
the second unit and be unavailable for crosstie.
Charging and Safety Injection (SI)         The impact should be minimal for the      This is a conservative approach and should not pumps are credited for inventory make-     baseline PRA model as the                 have a significant impact on the baseline PRA up for a medium loss of coolant            conservatism in the scenario of 1 out of  model. However, whenever a charging pump is accident. It is assumed that 2 out of 4    1 pump available is compensated with a    unavailable, and the SI system fails this high pressure injection pumps (charging    factor of 2 when 2 out of 2 pumps are     recovery factor becomes nonconservative or SI) are required for success; this was  required.                                 since the required 2 out of 2 high head conservatively modeled as 1 out of 2                                                 injection pump criterion is not met. Accordingly, charging pumps and 1 out of 2 SI                                                     the emergency core cooling system charging pumps. To eliminate this modeling                                                    pump recovery factor will not be credited in the conservatism when all support is                                                     RICT Program whenever an emergency core available and when 2 out of 2 charging                                               cooling system charging pump is made pumps are required, a conservative                                                   unavailable.
estimate of the charging system failure fraction is to multiply the split fraction value for 1 of 1 pump train unavailability (CH2) by a factor of 2. Thus, the recovery factor (or conservatism reduction factor) for these conditions is 2*CH2.

PG&E Letter DCL-23-054 Enclosure 9 Page 5 of 8 Table E9-1 Disposition of Key Assumptions/Sources of Uncertainty Impacting Configuration Risk Calculations Assumption/Uncertainty                           Discussion                          Disposition for the RICT Program A 6-hour mission time for the emergency  An EDG mission time of 6 hours is used      The 6-hour mission time of the EDGs does not diesel generators (EDGs) and fuel oil    for initiating events since the probability have a significant impact on the baseline PRA transfer pumps is assumed sufficient for of non-recovery of offsite power after 6    model. However, whenever the 230 kV offsite non-seismic initiators rather than the   hours is sufficiently small. The            power system is unavailable and cannot standard 24-hour mission time.           probabilities of non-recovery values are    reasonably be recovered within 6 hours, the 6-much larger than the fail to run values     hour mission time for the EDGs is of the EDGs and fuel oil transfer pumps     nonconservative. Accordingly, the 24-hour for the remaining 18 hours of the           mission time will be applied to the EDGs and mission time.                               fuel oil transfer pumps in the RICT Program whenever the offsite power 230 kV system is made unavailable.
Containment penetrations that would      Lines that require the failure of three or  Screening these penetrations from the baseline require the failure of three or more     more valves to cause the failure of         model is conservative for RICT calculations.
valves are screened from the             containment isolation have a negligible     The Configuration Risk Management Program containment isolation analysis.          contribution to the containment isolation   model used for the RICT Program will include failure frequency.                          the capability to evaluate these screened penetrations using a bounding surrogate model that assumes the containment boundary is failed, even when redundant isolation valves are available.
Vacuum breakers cannot fail in a         The magnitude of the uncertainty            There are two vacuum relief valves per ASW manner to impact the Auxiliary Salt      attributable to this nonconservative        header. They are mechanical components with Water (ASW) function within the 24-hour  assumption is not known. Although the       a relatively high reliability, thus the random mission time.                            magnitude of the non-conservatism is        failure of an ASW header due to failing both expected to be small, it has not been       vacuum relief valves should not be significant.
quantitatively demonstrated.                The RICT Program will assume inoperability of the ASW train if one or more vacuum breakers are nonfunctional.

PG&E Letter DCL-23-054 Enclosure 9 Page 6 of 8 Table E9-1 Disposition of Key Assumptions/Sources of Uncertainty Impacting Configuration Risk Calculations Assumption/Uncertainty                         Discussion                        Disposition for the RICT Program SI minimum flow valves are not           Failure of the minimum flow valves        SI Recirculation Valves 8974A and 8974B are modeled.                                 could result in flow diversion and impact in series, thus both valves must be impacted, the success criteria non-conservatively. which has a low probability. The operator action to close these valves is also evaluated in the human reliability analysis for switchover to cold leg recirculation. Accordingly, no significant impact is expected to the RICT Program.
Designation of systems/components as     Assuming certain systems and              This conservative assumption is acceptable for always failed in the fire PRA model and  components are always failed for all      the RICT Program. Those systems that are seismic PRA model.                       fires or seismic events is conservative. 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.
Pump runout protection is only modeled   This is a model simplification and        In order for more than one steam generator to for Auxiliary Feedwater Pump 1-2 and is  applies to sequences involving multiple   depressurize in a steam line break event, always successful for pump 1-3           steam generator depressurization.         multiple main steam isolation valves (MSIV) must fail concurrently. Given the low failure probability of the MSIV function and low risk contribution of a steam line break event (approximately 5-9 CDF), the resulting risk contribution is expected to be less than 1-10 CDF. Accordingly, no significant impact is expected to the RICT Program.

PG&E Letter DCL-23-054 Enclosure 9 Page 7 of 8 Table E9-1 Disposition of Key Assumptions/Sources of Uncertainty Impacting Configuration Risk Calculations Assumption/Uncertainty                          Discussion                       Disposition for the RICT Program The recirculation valves to the refueling The recirculation valves are interlocked Two valves must both fail to close; the water storage tank (RWST) are not         to prevent pump discharge to the         combined failure probability would be modeled.                                  RWST during the recirculation phase. If  approximately 5-7 which is insignificant.
these valves failed, long-term cooling   Accordingly, no significant impact is expected during recirculation could be lost.      to the RICT Program.

PG&E Letter DCL-23-054 Enclosure 9 Page 8 of 8 References
: 1. Letter from Jennifer M. Golder (USNRC - NRR) to Biff Bradley (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines, (TAC No. MD4995) May 17, 2007. (ADAMS Accession No. ML071200238).
: 2. NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, Revision 0, November 2006. (ADAMS Accession No. ML12286A322).
: 3. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009. (ADAMS Accession No. ML090410014).
: 4. ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.
: 5. NUREG-1855, Volume 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making. (ADAMS Accession No. ML090970525).
: 6. EPRI TR-1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, December 2008.

PG&E Letter DCL-23-054 Enclosure 10 Page 1 of 4 Enclosure 10 Program Implementation

PG&E Letter DCL-23-054 Enclosure 10 Page 2 of 4 Introduction Section 4.0, Item 11 of the Nuclear Regulatory Commissions (NRCs) Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (Reference 2), requires that the license amendment request provide a description of the implementing programs and procedures regarding the plant staff responsibilities for the RMTS implementation, and specifically discuss the decision process for risk management action (RMA) implementation during a Risk-Informed Completion Time (RICT).
This enclosure provides a description of the implementing programs and procedures regarding the plant staff responsibilities for the RICT Program, including training of the personnel required for implementation of the RICT Program.
Several procedures and processes are detailed in other enclosures that are not repeated in this enclosure: addressing probabilistic risk assessment (PRA) model update (Enclosure 8), Cumulative Risk Assessment and Monitoring Program (Enclosure 11), and RMAs (Enclosure 12). This enclosure provides a description of the implementing programs and procedures regarding the plant staff responsibilities for the RICT Program, including training of plant personnel.
RICT Program Procedures Pacific Gas and Electric Company (PG&E) will develop a program description and implementing procedures for the RICT Program. The program description will establish the management responsibilities and general requirements for risk management, training, implementation, and monitoring of the RICT Program. More detailed procedures will provide specific responsibilities, limitations, and instructions for implementing the RICT Program. The program description and implementing procedures will incorporate the programmatic requirements for the RMTS included in NEI 06-09-A.
Operations, specifically the control room staff, are responsible for compliance with Technical Specification (TS) requirements, and will be responsible for implementation of a RICT and any RMAs determined to be appropriate for the plant configuration. Any use of a RICT and associated RMAs will require management approval prior to entering an extended completion time (CT) for pre-planned activities, and as soon as practicable for an emergent extended CT.
Procedures for the RICT Program will address the following:
* important definitions related to the RICT Program
* departmental and position roles and responsibilities for activities in the RICT Program
* plant management approval for use of a RICT

PG&E Letter DCL-23-054 Enclosure 10 Page 3 of 4
* plant conditions for which the use of the RICT Program for both planned and emergent conditions is authorized
* limitations on implementing a RICT for both planned and emergent conditions
* implementation of the RICT Program 30-day back stop limit
* guidance on plant configuration changes; i.e., recalculating the RICT and risk management action time (RMAT) within the RICT Program time limits
* requirements to identify and implement RMAs when the RMAT is exceeded or is anticipated to be exceeded
* guidance on the use of RMAs, including the conditions under which they may be credited in RICT calculations
* conditions for exiting a RICT
* documentation requirements RICT Program Training The scope of the training for the RICT Program will include rules for the new TS program, Configuration Risk Management Program (CRMP) software, TS Actions included in the program, and procedures. This training will be conducted for the following Diablo Canyon staff positions:
* Operations Manager
* Operations Planning Managers
* Operations Personnel (Licensed and Non-Licensed)
* Work Control Manager
* Work Control Personnel
* Work Week Managers
* Operations Training
* Nuclear Licensing Personnel
* Selected Maintenance Personnel
* Site Engineering
* PRA Engineers
* Other Selected Management Training will be carried out in accordance with established Diablo Canyon training procedures and processes. These procedures were written based on the Institute of Nuclear Power Operations accreditation requirements, as developed and maintained by the National Academy for Nuclear Training. PG&E has planned three levels of training for implementation of the RICT Program. They are described below.

PG&E Letter DCL-23-054 Enclosure 10 Page 4 of 4 Level 1 Training This is the most detailed training. It is intended for the individuals who will be directly involved in the implementation of the RICT Program. This level of training includes the following attributes:
* specific training on the revised TS structure and use
* record-keeping requirements
* case studies
* hands-on time with the CRMP tool for calculating a RMAT and RICT
* identifying appropriate RMAs
* common cause failure RMA considerations in emergent RICTs
* other detailed aspects of the RICT Program Level 2 Training This training is applicable for supervisors, managers, and other personnel who need a broad understanding of the RICT Program. It is significantly more detailed than Level 3 Training (described below), but it is different from Level 1 Training in that hands-on time with the CRMP tool and case studies are not included. The concepts of the RICT Program will be taught; however, this group of personnel will not be qualified to perform the tasks for actual implementation of the RICT Program.
Level 3 Training This training is intended for the remaining personnel who require an awareness of the RICT Program. These employees need basic knowledge of RICT Program requirements and procedures. This training will cover the RICT Program concepts that are important to disseminate throughout the organization.
References
: 1. Letter from Jennifer M. Golder (USNRC - NRR) to Biff Bradley (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines, (TAC No. MD4995), May 17, 2007. (ADAMS Accession No. ML071200238).
: 2. NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, Revision 0, November 2006. (ADAMS Accession No. ML12286A322).

PG&E Letter DCL-23-054 Enclosure 11 Page 1 of 4 Enclosure 11 Monitoring Program

PG&E Letter DCL-23-054 Enclosure 11 Page 2 of 4 Section 4.0, Item 12 of the Nuclear Regulatory Commissions (NRCs) Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (Reference 2), requires that the license amendment request provide a description of the implementation and monitoring program as described in Regulatory Guide (RG) 1.174, An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1, (Reference 3) and NEI 06-09-A. (Note that RG 1.174, Revision 3 [Reference 4], issued by the NRC in January 2018, did not substantively change the requirements for the implementation and monitoring program.)
This enclosure provides a description of the process applied to monitor the calculation of the cumulative risk impact of implementation of the Risk-Informed Completion Time (RICT) Program, specifically the calculation of cumulative risk of extended Completion Times (CTs). Calculation of the cumulative risk for the RICT Program is discussed in Step 14 of Section 2.3.1 and Step 7.1 of Section 2.3.2 of NEI 06-09-A. General requirements for a Performance Monitoring Program for risk-informed applications are discussed in RG 1.174, Element 3.
The calculation of cumulative risk impact is required by the RICT Program at least every refueling cycle, not to exceed 24 months, consistent with the guidance in NEI 06-09-A.
For the assessment period evaluated, plant data is collected to establish the risk increase associated with each application of an extended CT for both core damage frequency (CDF) and large early release frequency (LERF), and the total risk calculated by summing all risk associated with each RICT application. This is the change in CDF or LERF above the zero maintenance baseline levels during the time of operation in the extended CT (i.e., beyond the front-stop CT). The change in risk is converted to average annual values.
The total average annual change in risk for extended CTs is compared to the acceptance guidelines of RG 1.174, Figures 4 and 5, for CDF and LERF changes, respectively. If the actual annual risk increase is acceptable (i.e., not in Region I of the figures), then the RICT Program implementation is acceptable for the assessment period. Otherwise, further assessment of the cause of exceeding the RG 1.174 guidance and implementation of any necessary corrective actions to ensure future plant operation is within the guidance is conducted under the site Corrective Action Program.
The evaluation will identify areas for consideration, including as examples:
RICT applications which dominated the risk increase risk contributions from planned versus emergent RICT applications risk management actions (RMAs) implemented but not credited in the risk calculations risk impact from applying a RICT to avoid multiple shorter outages

PG&E Letter DCL-23-054 Enclosure 11 Page 3 of 4 Based on the evaluation, any necessary corrective actions are developed and approved by the Operations Manager. These may include:
administrative restrictions on the use of RICTs for specific high-risk configurations additional RMAs for specific high-risk configurations rescheduling planned maintenance activities deferring planned maintenance to shutdown conditions use of temporary equipment to replace out-of-service systems, structures or components (SSCs) plant modifications to reduce the risk impact of expected future maintenance configurations In addition to the cumulative impact of the RICT Program implementation, the unavailability of SSCs is also potentially impacted. The existing Maintenance Rule (MR)
Monitoring Programs under 10 CFR 50.65(a)(1) and (a)(2) provide for the evaluation and disposition of the unavailability impacts that may be incurred by implementation of the RICT Program. The SSCs in the scope of the RICT Program are also in the scope of the MR, which allows the use of the MR Program.
The monitoring program for the MR, along with the specific assessment of cumulative risk impact described above, serves as the Implementation and Monitoring Program, defined as Element 3 of RG 1.174 and NEI 06-09-A for the RICT Program.

PG&E Letter DCL-23-054 Enclosure 11 Page 4 of 4 References
: 1. Letter from Jennifer M. Golder (USNRC - NRR) to Biff Bradley (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines, (TAC No. MD4995), May 17, 2007. (ADAMS Accession No. ML071200238).
: 2. NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, Revision 0, November 2006. (ADAMS Accession No. ML12286A322).
: 3. Regulatory Guide 1.174, An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1, November 2002. (ADAMS Accession No. ML023240437).
: 4. Regulatory Guide 1.174, An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 3, January 2018. (ADAMS Accession No. ML17317A256).

PG&E Letter DCL-23-054 Enclosure 12 Page 1 of 5 Enclosure 12 Risk Management Action (RMA) Examples

PG&E Letter DCL-23-054 Enclosure 12 Page 2 of 5 Introduction Section 4.0, Item 13 of the Nuclear Regulatory Commissions (NRCs) Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines (Reference 2), requires that the license amendment request provide a description of the process to identify and provide compensatory measures and risk management actions (RMAs) during extended Completion Times (CTs), including specific examples.
This enclosure describes the process for identification of RMAs applicable during extended CTs and provides examples of RMAs. RMAs will be governed by plant procedures for planning and scheduling maintenance activities. This procedure will provide guidance for the determination and implementation of RMAs when entering the Risk-Informed Completion Time (RICT) Program and is consistent with the guidance provided in NEI 06-09-A.
Responsibilities Work Control is responsible for developing the RMAs with assistance from Operations.
Operations is responsible for the approval and implementation of approved RMAs. For emergent entry into extended CTs, Operations is also responsible for developing RMAs.
Procedural Guidance For planned maintenance activities, implementation of RMAs will be required if it is anticipated that the risk management action time (RMAT) will be exceeded. The RMAs are implemented prior to exceeding the RMAT.
For emergent activities, RMAs must be implemented if the RMAT is reached. If the extent of condition evaluation is not complete or does not demonstrate redundant components are not impacted, then RMAs related to the success of redundant structures, systems, and components (SSCs), and to reducing the likelihood of initiating events relying on the affected function will be considered to address the possible increased likelihood of a common cause event. If an emergent event occurs requiring recalculation of a RMAT that is already in effect, the procedure requires a re-evaluation of the existing RMAs for the new plant configuration to determine whether new RMAs are appropriate. These requirements of the RICT Program are consistent with the guidance of NEI 06-09-A.
RMAs are put in place no later than the point at which an incremental core damage probability of 1-6 is reached, or no later than the point at which an incremental large early release probability of 1-7 is reached. If as the result of an emergent event the instantaneous core damage frequency (CDF) or the instantaneous large early release frequency (LERF) exceeds 1-3 or 1-4 per year, respectively, then RMAs are also required to be implemented. These requirements are consistent with the guidelines of NEI 06 A.

PG&E Letter DCL-23-054 Enclosure 12 Page 3 of 5 By determining which SSCs are most important from a CDF and/or LERF perspective for a specific plant configuration, RMAs may be developed to protect these SSCs.
Similarly, knowledge of the initiating event or sequence contribution to the configuration-specific CDF and/or LERF allows development of RMAs that enhance the capability to mitigate such events.
It is possible to credit RMAs in the RICT calculations. However, such quantification of RMAs is not required by NEI 06-09-A, Revision 0. Crediting RMAs in the RICT calculations is only done consistent with the guidance of NEI 06-09-A.
NEI 06-09-A classifies RMAs into three categories, as described below:
: 1) Actions to increase awareness and control.
Shift brief Pre-job brief Training (formal or informal)
Presence of a system engineer or other expertise related to maintenance activity Special purpose procedure to identify risk sources and contingency plans
: 2) Actions to reduce the duration of maintenance activities.
Pre-staging materials Conducting training on mock-ups Performing the activity around the clock Performing walk-downs on the actual system(s) to be worked on prior to beginning the work
: 3) Actions to minimize the magnitude of the risk increase.
Suspend/minimize activities on redundant systems Suspend/minimize activities on other systems that adversely affect the CDF and/or LERF Suspend/minimize activities on systems that may cause a trip or transient to minimize the likelihood of an initiating event that the out-of-service component is credited to mitigate Use temporary equipment for backup power Use temporary equipment for backup ventilation Reschedule other maintenance activities

PG&E Letter DCL-23-054 Enclosure 12 Page 4 of 5 Specific Examples Example RMAs that may be considered during a RICT Program entry for a diesel generator (DG) or a battery to reduce the risk impact and ensure adequate defense-in-depth include:
A. Diesel Generator:
(1) The condition of the offsite power supply, switchyard, and the grid is evaluated prior to entering a RICT, and RMAs as identified below are implemented, particularly during times of high grid stress conditions, such as during high demand conditions.
(2) Deferral of switchyard maintenance, such as deferral of discretionary maintenance on the main, auxiliary, or startup transformers associated with the unit.
(3) Deferral of maintenance that affects the reliability of the supported equipment associated with the OPERABLE DGs.
(4) Deferral of planned maintenance activities on station blackout mitigating systems, and treating those systems as protected equipment.
(5) Contacting the dispatcher on a periodic basis to provide information on the DG status and the power needs of the facility.
B. Battery:
(1) Limit the immediate discharge of the affected battery, if possible.
(2) Recharge the affected battery to float voltage conditions using a spare battery charger, if possible.
(3) Evaluate the remaining battery capacity and protect its ability to perform its safety function.
(4) Periodically verify that the battery float voltage is equal to or greater than the minimum required float voltage for the remaining OPERABLE batteries.

PG&E Letter DCL-23-054 Enclosure 12 Page 5 of 5 References
: 1. Letter from Jennifer M. Golder (USNRC - NRR) to Biff Bradley (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines, (TAC No. MD4995) May 17, 2007. (ADAMS Accession No. ML071200238).
: 2. NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, Revision 0, November 2006. (ADAMS Accession No. ML12286A322).}}