View source for DCL-24-018, License Amendment Request 24-01 Revision to Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

{{Adams
| number = ML24059A448
| issue date = 02/28/2024
| title = License Amendment Request 24-01 Revision to Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)
| author name = Rogers J
| 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-24-018
| document type = Letter type:DCL, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specifications
| page count = 1
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:PacHic Gas and m Electric Company
* Justin E. Rogers Diablo Canyon Power Plant Station Director Mail code 104/5/502 P.O. Box 56 Avila Beach, CA 93424

805.545.3088 Justin.Rogers@pge.com

PG&E Letter DCL-24-018 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 24- 01 Revision to Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

==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 Diablo Canyon Power Plant (DCPP), Unit 1 and 2 Technical Specification (TS) 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR).

The proposed license amendment request (LAR) revises DCPP TS 5.6.6 to add WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR -M3G and FERRET, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, F luence Determination with RAPTOR -M3G and FERRET - Supplement for Extended Beltline Materials, to permit the use of a newer, more advanced neutron fluence calculational methodology. Calculated fluence is used to determine the RCS pressure and temperature limits.

Approval of the proposed amendment is requested by February 28, 2025. Once approved, the amendment will be implemented within 90 days.

PG&E makes no regulatory commitments (as defined by NEI 99- 04) in this letter.
This letter includes no revisions to existing regulatory commitments.

The enclosure to this letter contains the evaluation of the proposed change.

In accordance with site administrative procedures and the Quality Assurance Program, the proposed amendment has been reviewed by th e Plant Staff Review Committee.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
* Diablo Canyon
* Palo Verde
* Wolf Creek

Document Control Desk PG&E Letter DCL 018 Page 2

Pursuant to 10 CFR 50.91, 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, Regulatory Services, at 805-545-4609.

I state under penalty of perjury that the foregoing is true and correct.

Sincerely,

Justin E. Rogers ____________
Station Director Date

kjse/51205913 Enclosure cc: Diablo Distribution cc/enc: John Monninger, NRC Region IV Administrator Mahdi O. Hayes, NRC Senior Resident Inspector Anthony Chu, Branch Chief, California Dept of Public Health Samson S. Lee, NRR Project Manager

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
* Comanche Peak
* Diablo Canyon
* Palo Verde
* Wolf Creek Enclosure PG&E Letter DCL-23-018

Evaluation of the Proposed Change

==Subject:==
License Amendment Request 24-01, Revision to Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)"
: 1. 

==SUMMARY==
DESCRIPTION
: 2. DETAILED DESCRIPTION

2.1 System Design and Operation 2.2 Current Technical Specifications Requirements 2.3 Reason for the Proposed Change
: 3. TECHNICAL EVALUATION

3.1 WCAP-18124-NP-A, Revision 0 and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0 3.2 Adoption of TSTF-419 3.3 Conclusion
: 4. REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions
: 5. ENVIRONMENTAL CONSIDERATION
: 6. REFERENCES

ATTACHMENTS:
: 1. Proposed Technical Specification Changes (Mark -Up)
: 2. Revised Technical Specification Page

Enclosure PG&E Letter DCL-24-018

EVALUATION
: 1. 

==SUMMARY==
DESCRIPTION

The proposed amendment would revise Diablo Canyon Power Plant (DCPP) Technical Specification (TS) 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," to add WCAP -18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, Fluence Determination with RAPTOR -M3G and FERRET - Supplement for Extended Beltline Materials, (References 1 and 2) to the specified analytical methods used to determine the reactor coolant system (RCS) pressure and temperature (P/T) limits for DCPP, Units 1 and 2.

The proposed change to TS 5.6.6 is requested to permit the use of more recent analytical methods that are approved by the Nuclear Regulatory Commission (NRC) staff when calculating future reactor vessel neutron fluence and RCS P/T limits.
: 2. DETAILED DESCRIPTION

2.1 System Design and Operation

Components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. Pressure and temperature changes are limited during RCS heatup and cooldown to be within the design assumptions and the stress limits for cyclic operation.

P/T limit curves are established for heatup, cooldown, and hydrostatic testing. Each P/T limit curve defines an acceptable region for normal operation. The curves are used during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable P/T limit curve to confirm that operation is within the allowable region. The minimum temperature required for inservice hydrostatic testing is also specified.

To determine the limiting pressurized thermal shock (PTS) reference temperature (RTPTS) for comparison with the applicable screening criteria as defined in 10 CFR 50.61, the maximum neutron exposure levels experienced by each of the reactor vessel materials must be projected to the expiration date of the operating license.

2.2 Current Technical Specifications Requirements

TS 5.6.6.b requires that the analytical methods used to determine the RCS P/T limits and Low Temperature Overpressure Protection System (LTOPS) lift settings and arming temperature shall be those previously reviewed and approved by the NRC, specifically those described in the f ollowing documents:

1 Enclosure PG&E Letter DCL-24-018
* WCAP-14040-NP-A, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.
* Chapter 6.0 of WCAP -15958, Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program.

TS 5.6.6.c states that the PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

2.3 Reason for the Proposed Change

PG&E is proposing to add a new analytical methodology used to determine the neutron fluence for calculating RCS P/T limits. The new analytical methodology will be used to support PG&Es planned updates to the DCPP Unit 1 and 2 PTLR by April 2025. The limits contained in the current Unit 1 and Unit 2 PTLR are valid until 35 effective full power years and will need to be updated to support power operation beyond spring 2025.

The proposed change will add the NRC approved analytical methodology for calculating the neutron fluence contained in WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, Fluence Determination with RAPTOR -M3G and FERRET - Supplement for Extended Beltline Materials to TS 5.6.6.b. The NRC approved neutron fluence methodology contained in References 1 and 2 complies with the guidance in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (Reference 3).

With the proposed neutron fluence methodology addition to TS 5.6.6.b, TS 5.6.6.c will also be revised as a corresponding change to identify the Topical Reports used to prepare PTLR change s so that the specific TS 5.6.6.b methods used to revise the PTLR are identified.
: 3. TECHNICAL EVALUATION

3.1 WCAP-18124-NP-A, Revision 0 and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0

The NRC Staff concluded in the Safety Evaluations contained in References 1 and 2 that the topical report s were acceptable for licensing applications, subject to the two limitations and conditions identified in the Safety Ev aluations being addressed by the licensee. The application of RAPTOR -M3G and FERRET to the extended beltline

2 Enclosure PG&E Letter DCL-24-018

materials for DCPP Units 1 and 2 is compliant with the two Limitations and C onditions from References 1 and 2, as addressed below.

WCAP-18124-NP-A Revision 0 and WCAP -18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0 Limitation and Condition 1:
: 1. Applicability of WCAP-18124-NP, Revision 0, is limited to the reactor pressure vessel region near the active height of the core based on the uncertainty analysis performed and the measurement data provided. Additional justification should be provided via additional benchmarking, fluence sensitivity analysis to response parameters of interest (for example, pressure-temperature limits, material stress and strain), margin assessment, or a combination thereof, for applications of the method to components including, but not limited to, the reactor pressure vessel upper circumferential weld, and RCS inlet and outlet nozzles and reactor vessel internal components.

===Response===
Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, (Reference 4) states that any materials predicted to exceed 1.0 x 1017 n/cm2 (E > 1.0 MeV) at the end of their licensed operating period must be evaluated to determine the changes in fracture toughness. Materials that are not adjacent to the active core yet are predicted to accrue fluence levels greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV) are now commonly referred to as extended beltline materials.

WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials, (Reference 2) provided additional methodology requirements and qualification data necessary to justify the application of the RAPTOR -M3G and FERRET fluence methods to the extended beltline regions which includes materials such as the reactor pressure vessel (RPV) upper circumferential weld and the RCS inlet and outlet nozzles. The NRC staff determined that the fluence methods and qualifications described in WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0 are acceptable for referencing in licensing applications as discussed in the NRC S afety Evaluation.

For DCPP, Limitation and Condition #1 regarding the applicability of the Reference 1 and 2 methodology to the reactor vessel internal components continues to apply.

WCAP-18124-NP-A Limitation and Condition 2:
: 2. Least squares adjustment is acceptable if the adjustments to the measured-to-calculated ratios and to the calculated spectra values are within the assigned

3 Enclosure PG&E Letter DCL-24-018

uncertainties of the calculated spectra, the dosimetry measured reaction rates, and the dosimetry reaction cross sections. Should this not be the case, the user should re-examine both measured and calculated values for possible errors. If errors cannot be found, the particular values causing the inconsistency should be disqualified.

===Response===
For DCPP, a least squares adjustment analysis will not be used to modify the calculated surveillance capsule or reactor pressure vessel neutron exposure when WCAP-18124-NP-A, Revision 0 and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0 methodology is applied. Therefore, Limitation and Condition #2 does not apply.

3.2 Adoption of TSTF-419

The addition of the analytical methods to TS 5.6. 6.b by TR number and title, without date, is consistent with Amendments No. 1 70 and 171 for DCPP Units 1 and 2 (Reference 5), respectively, that added the reference to WCAP-14040-NP-A and revised the referenced TRs to be by title and number only. Amendments No. 1 70 and 171 adopted TSTF-419, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR," and the NRC concluded in the safety evaluation that the proposed change to only list the NRC approved methodology by TR number and title is acceptable.
Additionally, in a letter from the NRC to the Technical Specification Task Force (Reference 6) the NRC indicated that the NRC Staff does not intend to backfit licensees that have Traveler TSTF-419 already in their TSs.

3.3 Conclusion

As cited in the NRC Safety Evaluation included in Reference 2, t he analytical methods used to determine the RCS P/T limits shall be those previously reviewed and approved by the NRC, specifically those described in WCAP -14040- A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004 (Reference 7). WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR -M3G and FERRET, July 2018, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, Fluence Determination with Raptor-M3G and FERRET - Supplement for Extended Beltline Materials, may be used as an alternative to Section 2.2 of WCAP -14040-A.

The proposed change to add the NRC approved methodology for determining neutron fluence values used for calculating RCS P/T limits is acceptable since the L imitations and Conditions contained in the NRC Safety Evaluations have been addressed as discussed above, and the methodology w as approved by the NRC. In addition, the proposed change will require that the methods used to revise the PTLR are identified.

4 Enclosure PG&E Letter DCL-24-018
: 4. REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

Regulations

10 CFR 50.60, Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation, imposes fracture toughness and material surveillance program requirements which are set forth in 10 CFR 50, Appendices G, Fracture Toughness Requirements, and H, Reactor Vessel Material Surveillance Program Requirements.

10 CFR 50 Appendix G specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Compliance with Appendix G is discussed in Section 5.2 of the DCPP Units 1 and 2 Updated Final Safety Analysis Report (UFSAR).

10 CFR 50 Appendix H requires a program to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of light water nuclear power reactors which result from exposure of these materials to neutron irradiation and the thermal environment. Compliance with Appendix H is discussed in Section 5.2 of the DCPP Units 1 and 2 U FSAR.

10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, specifies the calculation of the reference temperature for a reactor vessel material, RT PTS, based on the highest best -estimate neutron fluence projected for a specific material on the expiration date of the operating license.

General Design Criteria

DCPP, Units 1 and 2 were designed to meet the intent of the Atomic Energy Commission (now the NRC) General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, published in July 1967. The DCPP construction permit for units 1 and 2 were issued in April 1968 and December 1970. Conformance with the applicable 1967 General Design Criteria (GDC) discussed below is described in Section 3.1. 7 of the DCPP Units 1 and 2 U FSAR.

GDC 34, Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevention, requires that the RCPB shall be designed to minimize the probability of rapidly propagating type failures. Consideration shall be given: (a) to the notch-toughness properties of materials extending to the upper shelf of the Charpy transition curve, (b) to the state of stres s of materials under static and transient loadings, (c) to the quality control specified for materials and component fabrication to limit flaw sizes, and (d) to

5 Enclosure PG&E Letter DCL-24-018

the provisions for control over service temperature and irradiation effects that may require operational restrictions.

GDC 35, Reactor Coolant Pressure Boundary Brittle Fracture Prevention, requires that under conditions where RCPB system components constructed of ferritic materials may be subjected to potential loadings, such as a reactivity -induced loading, service temperatures shall be at least 120&deg;F above the NDT temperature of the component material if the resulting energy release is expected to be absorbed by plastic deformation, or 60&deg;F above the NDT temperature of the component material if the resulting energy release is expected to be absorbed within the elastic strain energy range.

Regulatory Guidance

Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, contains guidance on methodologies the NRC considers acceptable for determining the effect of neutron radiation on reactor vessel materials. This RG is used as a reference in WCAP -18124-NP-A, Revision 0 (Reference 1).

RG 1.190, Revision 0, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, discusses a methodology acceptable to the NRC staff for determining the best -estimate neutron fluence experienced by materials in the beltline region of light water reactor (LWR) pressure vessels, as well as for determining the overall uncertainty associated with those best -estimate values. In the NRC Safety Evaluations included in References 1 and 2, the NRC stated that the calculational fluence methodology described in References 1 and 2 complies with the guidance in RG 1.190.

RIS 2014-11 clarifies that P/T limit calculations for ferritic RPV materials other than those materials with the highest reference temperature may result in more limiting P/T curves because of higher stresses due to structural discontinuities, such as those in RPV inlet and outlet nozzles. Conformance with RIS 2014-11 is described in Reference 2.

4.2 Precedent

The proposed amendment would allow use of a new method, as described in Topical Report WCAP-18124-NP-A, Revision 0, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, for the determination of P/T limits. The NRC previously approved the WCAP -18124-NP-A, Revision 0 methodology for the calculation of fast neutron fluence in ferritic components of the reactor pressure vessel and the supporting WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0 qualification data as an alternative to the fluence methods described in Section 2.2 of WCAP -14040- A, Revision 4. The addition of WCA P-18124-NP-A, Revision 0, and WCAP -18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0 to TS Section 5.6.6 was previously approved by the NRC in the NRC letter to Prairie Island Nuclear Generating Plant,

6 Enclosure PG&E Letter DCL-24-018

Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Nos.
243 AND 231 RE: Revised Technical Specification 5.6.6, ' Reactor Coolant System (RCS) Pressure and T emperature Limits Report (PTLR)' (EPID L-2022-LLA-0184),
dated January 17, 2024 ( Reference 8).

The Refence 8 precedent included the analytical results associated with application of the References 1 and 2 methodology, including results showing how Reference 1 Limitation and Condition 1 is met, in the LAR. This enclosure does not include the analytical results associated with application of the References 1 and 2 methodology.
Reference 2 provided the justification necessary to narrow the Reference 1 Limitation and Condition 1 and allow licensees to apply the RAPTOR-M3G method in the extended beltline regions of RPVs on a generic basis. The NRC staff determined in the safety evaluation contained in Reference 2 that the fluence methods and qualifications described in WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A are acceptable for referencing in licensing applications. The results of application of the References 1 an d 2 methodology for DCPP will be contained within the submitted PTLR revision when References 1 and 2 are utilized. Approval of the proposed amendment w ill require the identif ication of the specific TS 5.6.6.b methods used to prepare PTLR revisions within the PTLR.

4.3 Significant Hazards Consideration

PG&E proposes to revise the DCPP Units 1 and 2 TS 5.6.6 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), to add WCAP-18124-NP-A, Revision 0 and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0 as a neutron fluence calculational methodology that can be used to determine the RCS P/T limits.

PG&E has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes do not require physical changes to plant systems, structures, or components. There is no interaction with a potential accident initiating mechanism. Implementation of the analytical methods as proposed would continue to provide assurance that appropriate Pressure and Temperature Limits Report (PTLR) limits are established to preserve the integrity of the RCS.
The proposed amendment is based on NRC approved methodolog y. Ensuring appropriate PTLR limits are established will not adversely affect a structure, system, or component of the plant, plant operations, design functions, or analysis that verifies the capability of a structure, system, or component to perform a

7 Enclosure PG&E Letter DCL-24-018

design function. Because there are no adverse effects on systems, structures, or components (SSCs), the likelihood of a malfunction is not increased and consequences of previously evaluated accidents in the Updated Final Safety Analysis Report are not changed.

Therefore, the proposed changes do 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 di fferent accident from any accident previously evaluated?

Response: No.

The proposed changes do not introduce new or different accidents to be postulated and subsequently evaluated, and no changes are being made to the plant that would introduce any new accident causal mechanisms. This license amendment request does not affect any plant systems that are potential accident initiators; nor does it have any significantly adverse effect on any accident mitigating systems.

The proposed change does not cha nge the functional requirements configuration, or method of operation of any system or component. Under the proposed change, no additional plant equipment will be installed. Implementation of the analytical methods as proposed would continue to provide assurance that appropriate PTLR limits are established to preserve the integrity of the RCS. The proposed amendment is based on NRC approved methodology. The proposed amendment would continue to ensure RCS integrity.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed changes do not alter the permanent plant design, nor does it change the assumptions contained in the safety analyses. No safety limits or operating parameters used to establish the safety margin are affected. The safety margins included in analyses of accidents are not affected by the proposed change. Implementation of the analytical methods would continue to provide assurance that appropriate PTLR limits are established in accordance with NRC approved methodology. This ensures that the plant is operated within design limits and the margin of safety in the plant safety analysis is maintained.

8 Enclosure PG&E Letter DCL-24-018

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above evaluation, PG&E concludes that the proposed change does not involve a 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.

4.4 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.
: 5. ENVIRONMENTAL CONSIDERATION

PG&E has evaluated the proposed amendment and has determined that the proposed amendment 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 amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a 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 amendment 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 amendment.
: 6. REFERENCES
: 1. WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR - M3G and FERRET, dated July 2018. [ADAMS Accession No. ML18204A010].
: 2. WCAP-18124-NP-A Revision 0, Supplement 1-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials, dated May 2022. [ADAMS Accession No. ML22153A139].
: 3. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.190, Revision 0, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001. [ADAMS Accession No. ML010890301].

9 Enclosure PG&E Letter DCL-24-018
: 4. NRC Regulatory Issue Summary 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014.
: 5. NRC Issuance of License Amendment Nos. 1 70 and 171 to Facility Operating License Numbers 80 and 82 for Diablo Canyon Power Plant, Uni ts 1 and 2, respectively, May 13, 2004.
: 6. Letter from the NRC to the Technical Specification Task Force (TSTF)
Implementation of Travelers TSTF -363, Revision 0, 'Revise Topical Report References in ITS 5.6.5, COLR [Core Operating Limits Report],' TSTF -408, Revision 1, 'Relocation of LTOP [Low temperature Overpressure Protection]
Enable Temperature and PORV [Power-Operated Relief Valve] Lift Setting to the PTLR [Pressure-Temperature Limits Report],' And TSTF -419, Revision 0,
 'Revise PTLR Definition and References in ISTS [Improved Standard Technical Specification] 5.6.6, RCS [Reactor Coolant System] PTLR',
dated August 4, 2011. [ADAMS Accession No. ML110660285].
: 7. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, dated May 2004. [ADAMS Accession No. ML050120209].
: 8. Letter from NRC to Prairie Island Nuclear Generating Plant, Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Nos. 243 AND 231 RE: Revised Technical Specification 5.6.6, "Reactor Coolant System (RCS) Pressure and T emperature Limits Report (PTLR)" (EPID L-2022-LLA-0184), dated January 17, 2024 [ADAMS Accession No. ML23356A003].

10 Enclosure Attachment 1 PG&E Letter DCL 018

Attachment 1

Proposed Technical Specification Changes (Mark-Up)

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)
: b. The analytical methods used to determine the RCS pressure and temperature and LTOP limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
: 1. WCAP 14040- NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."
: 2. Chapter 6.0 of WCAP-15958, "Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program."
: 3. WCAP-18124-NP-A, "Fluence Determination with RAPTOR-M3G and FERRET," and WCAP-18124-NP-A, Supplement 1-NP-A, "Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials" may be used as an alternative to Section 2.2 of WCAP-14040- NP-A.
: c. The PTLR will identify the Topical Reports used to prepare the PTLR and shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.6.7 Not Used

5.6.8 PAM Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.9 Not Used

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-22 Unit 1 - Amendment No. 135,168,170,198, Unit 2 - Amendment No. 135,169,171,199,

Enclosure Attachment 2 PG&E Letter DCL 018

Attachment 2

Revised Technical Specification Page

Remove Page Insert Page

5.0 - 22 5.0 - 22

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)
: b. The analytical methods used to determine the RCS pressure and temperature and LTOP limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
: 1. WCAP 14040- NP - A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves."
: 2. Chapter 6.0 of WCAP - 15958, "Analysis of Capsule V from Pacific Gas and Electric Company Diablo Canyon Unit 1 Reactor Vessel Radiation Surveillance Program."
: 3. WCAP - 18124-NP - A, " Fluence Determination with RAPTOR - M3G and FERRET, " and WCAP - 18124-NP -A, Supplement 1-NP - A, " Fluence Determination with RAPTOR - M3G and FERRET - Supplement for Extended Beltline Materials " may be used as an alternative to Section 2.2 of WCAP - 14040- NP - A.
: c. The PTLR will identify the Topical Reports used to prepare the PTLR and shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.6.7 Not Used

5.6.8 P AM Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.9 Not Used

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0 -22 Unit 1 - Amendment No. 135,168,17 0,198,
Unit 2 - Amendment No. 135,169,171,199,}}