DCL-25-087, License Amendment Request 25-05 Application to Utilize Adopt Fuel for Improved Fuel Performance, Part 2

License Amendment Request 25-05 Application to Utilize Adopt Fuel for Improved Fuel Performance, Part 2

ML25344A428
Person / Time
Site:	Diablo Canyon
Issue date:	12/03/2025
From:	Pacific Gas & Electric Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML25344A426	List: DCL-25-087, License Amendment Request 25-05 Application to Utilize Adopt Fuel for Improved Fuel Performance, Part 2 ML25344A427
References
DCL-25-087
Download: ML25344A428 (0)

Text

3.6.3.1.5 Conclusions PG&E Letter DCL-25-087 It is concluded that the analysis of the locked rotor event has adequately accounted for operation of DCPP Units 1 and 2 at conditions supporting the use of the proposed fuel upgrades, and it was performed using acceptable analytical models and methods (including the use of the RETRAN-02W computer code, VIPRE-W computer code, the WTDP statistical DNB methodology, and PAD5 fuel performance analysis methods).

Furthermore, it is concluded that the analysis has demonstrated that the reactor protection and safety systems continue to ensure that the ability to insert control rods is maintained, the reactor coolant pressure boundary pressure limits will not be exceeded, the reactor coolant pressure boundary will behave in a non-brittle manner, the probability of propagating fracture of the reactor coolant pressure boundary is minimized, and adequate core cooling will be provided. Therefore, the proposed fuel upgrades and computer code and methodology updates are acceptable with respect to the locked rotor event.

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Table 3.6-1 Non-LOCA Transients Evaluated or Analyzed UFSAR Event Evaluation or Reanalysis Section 15.2.1 Uncontrolled Rod Cluster Control Assembly Bank Reanalysis Withdrawal from a Subcritical Condition 15.2.2 Uncontrolled Rod Cluster Control Assembly Bank Reanalysis of DNBR cases Withdrawal at Power Evaluation of RCS peak pressure cases 15.2.3 Rod Cluster Control Assembly Misoperation Evaluation of plant transient response Confirmation of DNBR criterion Confirmation of fuel centerline melt criterion 15.2.4 Uncontrolled Boron Dilution Evaluation 15.2.5 Partial Loss of Forced Reactor Coolant Flow Bounded by 15.3.4 15.2.6 Startup of an Inactive Reactor Coolant Loop Precluded by Technical Specifications 15.2.7 Loss of External Electrical Load and/or Turbine Trip Reanalysis 15.2.8 Loss of Normal Feedwater Reanalysis 15.2.9 Loss of Offsite Power to the Station Auxiliaries Reanalysis 27 PG&E Letter DCL-25-087 Changes to Computer Codes Utilized VIPRE-W replaces THINC RETRAN-02W replaces LOFTRAN for DNBR cases VIPRE-W replaces THINC N/A None None RETRAN-02W replaces LOFTRAN for DNBR cases RETRAN-02W replaces PG&E RETRAN-02 for peak pressure cases None None

Table 3.6-1 Non-LOCA Transients Evaluated or Analyzed UFSAR Event Evaluation or Reanalysis Section 15.2.10 Excessive Heat Removal due to Feedwater System Reanalysis Malfunctions 15.2.11 Sudden Feedwater Temperature Reduction Evaluated for temperature reduction less than 73°F Reanalysis for temperature reduction greater than 73°F with load rejection 15.2.12 Excessive Load Increase Incident Reanalysis 15.2.13 Accidental Depressurization of the Reactor Coolant Reanalysis System 15.2.14 Accidental Depressurization of the Main Steam Bounded by 15.4.2.1 System 15.2.15 Spurious Operation of the Safety Injection System Evaluation of DNBR at Power Reanalysis of pressurizer filling 15.3.2 Minor Secondary System Pipe Breaks Bounded by 15.4.2 15.3.3 Inadvertent Loading of a Fuel Assembly into an Evaluation Improper Position 15.3.4 Complete Loss of Forced Reactor Coolant Flow Reanalysis 28 PG&E Letter DCL-25-087 Changes to Computer Codes Utilized None None RETRAN-02W replaces LOFTRAN RETRAN-02W replaces LOFTRAN NIA None N/A None RETRAN-02W replaces LOFTRAN VIPRE-W replaces THINC and FACTRAN

Table 3.6-1 Non-LOCA Transients Evaluated or Analyzed UFSAR Event Evaluation or Reanalysis Section 15.3.5 Single Rod Cluster Control Assembly Withdrawal at Confirmation of rods-in-DNB Full Power criterion 15.4.2.1 Rupture of a Main Steam Line at Hot Zero Power Reanalysis 15.4.2.2 Major Rupture of a Main Feedwater Pipe Reanalysis 15.4.2.3 Rupture of a Main Steam Line at Full Power Evaluation of plant transient response Confirmation of DNBR criterion Confirmation of fuel centerline melt criterion 15.4.2.4 Major Rupture of a Main Feedwater Pipe for Reanalysis Pressurizer Filling 15.4.4 Single Reactor Coolant Pump Locked Rotor Reanalysis 15.4.6 Rupture of a Control Rod Drive Mechanism Reanalysis Housing (Rod Cluster Control Assembly Ejection) 29 PG&E Letter DCL-25-087 Changes to Computer Codes Utilized VIPRE-W replaces THINC VIPRE-W replaces THINC None VIPRE-W replaces THINC None RETRAN-02W replaces LOFTRAN VIPRE-W replaces THINC and FACTRAN See Section 3.8

Table 3.6-2 Locked Rotor Analysis Results Summary Criterion Maximum Fuel Cladding Average Temperature at the Core Hot Spot, °F Maximum Zirconium-Water Reaction at the Core Hot Spot,

% by weight Maximum RCS Pressure, psia PG&E Letter DCL-25-087 Result Limit 1906 2375 0.4 16 2613 3214.7 Table 3.6-3 Locked Rotor Analysis Time Sequence of Events Event Time (seconds)

Rotor locks on one RCP 0.0 Low reactor coolant loop flow reactor trip setpoint reached 0.1 Reactor trip initiated (rods begin to drop) 1.1 Undamaged RCPs lose power and begin coasting down 1.1 Maximum fuel cladding average temperature occurs 3.7 Maximum RCS pressure occurs 4.1 30

PG&E Letter DCL-25-087 Figure 3.6-1 Single RCP Locked Rotor/Shaft Break - Core and Loop Volumetric Flow Rates versus Time 1.20...------------------------,

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PG&E Letter DCL-25-087 Figure 3.6-2 Single RCP Locked Rotor/Shaft Break - Nuclear Power and Core Average Heat Flux versus Time 1.20----------------------------.

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PG&E Letter DCL-25-087 Figure 3.6-3 Single RCP Locked Rotor/Shaft Break - RCS Pressure and Fuel Cladding Average Temperature versus Time

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3.7 LOCA 3.7.1 Background and Scope PG&E Letter DCL-25-087 DCPP Units 1 and 2 received approval to implement the FULL SPECTRUM loss-of-coolant accident (FSLOCA) Evaluation Model (EM) (Reference 20) by Amendment Nos.

234 and 236 dated January 9, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19316A109, Reference 21 ). Changes to the LOCA licensing basis analyses are tracked pursuant to 10 CFR 50.46(a)(3). The effects of the implementation of ADOPT fuel pellets, Optimized ZIRLO cladding, and PRIME fuel features were evaluated on the existing FSLOCA EM Analyses. It is noted that though Optimized ZIRLO cladding was considered with the evaluation of the other fuel features, DCPP Units 1 and 2 received approval to implement Optimized ZIRLO cladding by Amendment Nos. 252 and 254 dated August 13, 2025 (ADAMS Accession No. ML25174A192, Reference 13).

The FSLOCA EM evaluation considered break sizes in which break flow is beyond the capacity of the normal charging pumps, up to and including a double-ended guillotine (DEG) rupture of a RCS cold leg with a break flow area equal to two times the pipe area, including what traditionally are defined as small-break LOCAs (SBLOCAs) and large-break LOCAs (LBLOCAs). The break size spectrum is divided into two regions.

Region I includes breaks that are typically defined as SBLOCAs. Region II includes break sizes that are typically defined as LBLOCAs. Therefore, the impact was evaluated for both Region I/SBLOCA and Region 11/LBLOCA for each Unit.

3.7.2 Method Description In order to provide a robust estimate of effect that considers the uncertainty associated with the highly ranked LOCA phenomena for the fuel transition, full uncertainty analyses for Region I and Region II were executed for each unit modeling the updated fuel type.

The uncertainty analyses were performed in a manner consistent with the NRC-approved FSLOCA EM methodology with the exception of the changes which were previously transmitted to the NRC pursuant to 10 CFR 50.46. The limitations and conditions associated with the FSLOCA EM, as described in Reference 21, continue to be met with this evaluation approach. The uncertainty analysis run sets utilized fuel performance data generated with PAD5 (Reference 4) which modeled the ADOPT fuel pellets and Optimized ZIRLO cladding. None of the major plant parameters and analysis assumptions used in the existing FSLOCA EM analysis were changed.

3.7.3 Estimated Effect The estimated peak cladding temperature (PCT) impact for Unit 1 and Unit 2 are presented in Table 3.7-1 and Table 3.7-3, respectively. The new analysis-of-record 34

PG&E Letter DCL-25-087 PCT resulting from the evaluation of the implementation of the fuel hardware changes is presented in Table 3.7-2 and Table 3.7-4 for Units 1 and 2 respectively. The estimates of effect were determined based on comparison of the current licensing basis FSLOCA EM PCT and the updated FSLOCA EM uncertainty analyses that modeled the updated fuel features. The current licensing basis PCTs remain consistent with the results provided in Attachment 1 to the Enclosure to the letter dated September 23, 2019 (ADAMS Accession No. ML19266A657, Reference 22). For both DCPP Units, the large-break region PCT is limiting, and substantial margin exists to the 10 CFR 50.46(b)(1) PCT limit of 2200 °F when considering the updated fuel type, as seen in Tables 3.7-2 and 3.7-4. Figure 3.7-1 through Figure 3.7-4 show the PCT transients from the updated FSLOCA EM uncertainty analyses and demonstrate that the predicted emergency core cooling system response with the FSLOCA EM is similar with the resident fuel design and the updated fuel design with ADOPT fuel pellets, Optimized ZIRLO cladding, and PRIME fuel features.

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Table 3.7-1 PG&E Letter DCL-25-087 Estimated PCT Impact for DCPP Unit 1 from Evaluation of Fuel Changes, Accounting for Uncertainties Region I Region II DCPP Unit 1

+44 °F

+32 °F Table 3.7-2 New 95/95 PCT Result for DCPP Unit 1 with the FSLOCA EM Region I Region II DCPP Unit 1 1160 °F 1745 °F Table 3.7-3 Estimated PCT Impact for DCPP Unit 2 from Evaluation of Fuel Changes, Accounting for Uncertainties Region I Region II DCPP Unit 2

+31 °F

+58 °F Table 3.7-4 New 95/95 PCT Result for DCPP Unit 2 with the FSLOCA EM Region I Region II DCPP Unit 2 1052 °F 1663 °F 36

PG&E Letter DCL-25-087 Figure 3.7-1: DCPP Unit 1 Peak Cladding Temperature for All Rods for the Region I Analysis PCT Case PCT for Each Rod for Analysis PCT Cose Dummy Rod 1 Dummy Rod 2 Hot Rod Hot Assemb l y Average Rod 1 Average Rod 2 Low Power Rod 1200-,-------------------------,

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3.8 Control Rod Ejection PG&E Letter DCL-25-087 PG&E is also requesting NRC approval to utilize the 3DRE methodology described in WCAP-15806-P-A, "Westinghouse Control Rod Ejection Accident Analysis Methodology Using Multi-Dimensional Kinetics" (Reference 6) to perform the complex modeling of rod ejection events. WCAP-15806-P-A was approved prior to issuance of RG-1.236, "Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents." The current rod ejection methodology is described in UFSAR 15.4.6. A detailed discussion of the application of WCAP-15806-P-A as well as its degree of compliance with RG-1.236, is provided in Enclosures 4 and 5 (Proprietary and Non-Proprietary, respectively).

3.9 Dose Analysis This section discusses the impact of the fuel upgrade and associated changes in the plant licensing basis, on the dose consequences at the Exclusion Area Boundary (EAB),

the Low Population Zone (LPZ), the Control Room (CR) and the Technical Support center (TSC).

Summarized below are the results of the assessments that were performed to address the impact of the following changes in the DCPP design and licensing bases that could affect design inputs used to develop the post-accident dose consequences to the public at off-site locations, and to the operator in the CR and TSC.

a. Fuel Upgrade: Use of ADOPT fuel pellets, PRIME fuel features and Optimized ZIRLO fuel rod cladding could impact the isotopic core activity inventory, and consequently, the coolant isotopic activity concentrations utilized in dose consequence analyses. The assessment concluded that the existing margin (applied to the current isotopic activity inventory in the core to account for future fuel management schemes) would accommodate the impact of the fuel upgrade, and that the current isotopic activity inventory in the core and coolant remains bounding.

b. Implementation of PAD5 methodology to determine the maximum and minimum fuel rod temperatures and core stored energy in plant safety analyses to support fuel rod design could impact the following :

i.

Steam Releases to determine Dose Consequences: A review was performed of the estimated post-accident environmental steam releases currently used to assess the dose consequences following non-LOCA accidents such as the Main Steam Line Break (MSLB), Locked Rotor Accident (LRA), CREA, and Condition II events. The assessment demonstrated that the steam releases for dose analyses are not sensitive to the fuel-to-coolant heat transfer coefficient input used that could change 41

PG&E Letter DCL-25-087 due to the fuel temperatures calculated using the PADS computer code or ADOPT fuel pellets. Thus, the current steam releases used for dose consequences remain valid.

ii.

Hydraulic Evaluation of the Steam Generator Tube Rupture (SGTR): The updated hydraulic evaluation of the SGTR transient and associated time-dependent break flow mass releases into the ruptured Steam Generator (including the associated flash fractions), and the environmental steam releases from the condenser (prior to reactor tip), and from the MSSVs/

PORVs of the ruptured and intact SGs (post-reactor trip) were compared to that used in the current SGTR dose consequence analysis. With the exception of the time interval between reactor trip to when the PORV of the ruptured SG fails open, the differences in mass releases per time interval remained minor in nature. The dose assessment is available for NRC review and audit.

iii.

Containment Response following a Loss-of-Coolant Accident (LOCA): The containment response (specifically the containment temperature, pressure, humidity, condensing rate and sump water temperature transients) associated with the updated post-LOCA mass and energy releases, were compared to that used in the current analyses of record.

The differences in the containment response were determined to be minor, with negligible downstream impact on parameter values used for dose consequences (e.g., elemental iodine and particulate aerosol removal coefficients, containment spray coverage, containment mixing rate, containment pressure relief pathway, sump pH, releases from the Refueling Water Storage Tank (RWST) and the Mechanical Equipment Drains Tank (MEDT). Thus, the current LOCA dose consequence analysis remains bounding.

c. Implementation of 3DRE methodology as described in WCAP-15806-P-A as well as compliance with RG-1.236. The updated analysis does not increase the percentage of fuel failure following a CREA. In addition, there is no expectation for any fuel melt. Thus, the current CREA dose consequence analysis which assumes 10% fuel failure, remains bounding.

d. Transition to the WTDP WCAP-18240-P-A Methodology using VIPRE-W to confirm that departure from nucleate boiling (DNB) design basis is met. For the Locked Rotor Accident (LRA), the updated analysis concluded that the current 10% fuel failure assumption for dose consequences for the LRA dose analysis is acceptable and remains bounding.

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PG&E Letter DCL-25-087 Based on the above, it is concluded that other than the small increase in the dose consequences associated with the SGTR the current dose consequence analyses remain valid for the fuel upgrade and associated changes in the plant licensing basis.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36, 'Technical Specifications," requires that design features be included that described the facility in terms of materials of construction and geometric arrangement, which, if altered or modified, would have a significant effect on safety.

4.1.1 Regulations and Regulatory Guidance NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition" (SRP), Section 4.2, "Fuel System Design,"

provides regulatory guidance to the NRC staff for the review of the fuel pellets and overall fuel system. In addition, the SRP provides guidance for compliance with the applicable General Design Criteria (GDC) of 10 CFR Part 50, Appendix A.

According to SRP Section 4.2, the fuel system safety review provides assurance that:

The fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs).

Fuel system damage is never so severe as to prevent control rod insertion when it is required.

The number of fuel rod failures is not underestimated for postulated accidents.

Coolability is always maintained.

In Reference 1, the NRC approved WCAP-18482-P-A for the use of ADOPT fuel pellets for Westinghouse fuel designs.

4.1.2 General Design Criteria As noted in the UFSAR Section 3.1, DCPP was designed to comply with the Atomic Energy Commission (AEC) (now the NRC) General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, published in July 1967. The DCPP design basis is the 1967 GDCs. Subsequent commitments to GDCs issued later are noted in the discussion of each GDC in the UFSAR and as applicable below.

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Criterion 10 - Reactor Design PG&E Letter DCL-25-087 The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

GDC 10, 1971 supersedes GDC 6, 1967 with respect to the design of the reactor core.

Compliance with GDC 10 is described in UFSAR Section 3.1.3.1.1.

Criterion 27 - Combined Reactivity Control Systems Capability The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the ECCS, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

Compliance with GDC 27, 1971 is described in Appendix 3.1A of the DCPP UFSAR. It should be noted that the DCPP UFSAR Section 3.1.6.1 states that Criterion 27, 1967, is no longer part of the DCPP license basis and has been replaced by GDC 26, 1971.

However, Appendix 3.1A provides information to show the degree to which the DCPP design conforms to the intent of the 1971 GDCs and establishes additional DCPP licensing basis which must be reviewed when evaluating facility changes. Therefore, GDC 27, 1971, is included for completeness.

Criterion 35 - Emergency Core Cooling A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Compliance with GDC 35, 1967, is described in the DCPP UFSAR Section 3.1.7.3 while compliance with GDC 35, 1971, is described in Appendix 3.1A of the DCPP UFSAR.

There will be no changes to the DCPP design such that compliance with any of the regulatory requirements above would come into question. Therefore, DCPP Units 1 and 2 will continue to comply with the applicable regulatory requirements.

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PG&E Letter DCL-25-087 There are no changes being proposed in this amendment application such that conformance or commitments to the regulatory requirements and/or guidance documents above would come into question. The evaluations documented herein confirm that DCPP will continue to comply with all applicable regulatory requirements.

In conclusion, based on considerations discussed herein, (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 Commission's regulations, and (3) issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4.2 Precedent This proposed change is similar in nature to the following LARs approved by the NRC that authorized the use of ADOPT fuel pellets, 3DRE, and supporting codes and methods:

Turkey Point Nuclear Generating Station, Units 3 and 4 (Reference 23, approved by the NRC in Reference 24)

ADOPT fuel pellets have been used previously in Lead Test Assemblies (LTAs).

The most recent examples are for the LTA programs at Vogtle Unit 2 (Reference 26, approved by the NRC in Reference 27) and Byron Unit 2 (Reference 28, approved by the NRC in Reference 29).

PG&E also reviewed the related RAI responses associated with these submittals (References 25 and 30) and incorporated the requested information as appropriate.

4.3 Significant Hazards Consideration Determination Pacific Gas and Electric (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 change to the method of evaluation discussed in Updated Final Safety Analysis Report (UFSAR) Section 15.4.6 uses a Nuclear Regulatory Commission (NRC) approved methodology for determining the 30 core kinetics for the rod ejection accident. The methodology is applicable for the modeling of the rod ejection accident and shows compliance with the requirements of Regulatory Guide (RG) 1.236. The use of a new NRG-approved method of evaluation will not increase the potential for a rod 45

PG&E Letter DCL-25-087 ejection accident. Therefore, the probability of an accident is not increased by the proposed change. Since no fuel failures are predicted, the consequences of an accident are not increased by the proposed change.

The proposed change to Technical Specification (TS) 4.2.1 will allow the use of ADOPT fuel pellets at DCPP. The NRC-approved topical report WCAP-18482-P-A, which addresses ADOPT fuel pellets, demonstrates that ADOPT fuel pellets have essentially the same properties compared to standard UO2 pellets. The use of ADOPT fuel pellets will not result in adverse changes to the operation or configuration of the facility. The pellets are not accident initiators and do not affect accident probability. Use of ADOPT pellets meets the fuel design acceptance criteria and hence does not significantly affect the consequences of an accident.

The proposed change to TS 5.6.5, "Core Operating Limits Report (COLR)," adds the Westinghouse Thermal Design Procedure (WTDP) and VIPRE-01 topical reports, WCAP-18240-P-A and WCAP-14565-P-A, respectively, to the list of COLR References.

The analysis results including ADOPT fuel pellets and Optimized ZIRLO fuel cladding for DCPP meet the technical regulatory requirements of 10 CFR 50.46. Because the reactor core meets the technical regulatory requirements of 10 CFR 50.46 after a postulated Loss of Coolant Accident (LOCA) and the consequences of anticipated operational transients and postulated design basis accidents are not significantly increased, the proposed change does not significantly affect the consequences of an accident.

The proposed changes will not adversely affect accident initiators or precursors nor alter the design assumptions, conditions, and configuration of the facility or the manner in which the plant is operated and maintained. The proposed changes will not alter or prevent the ability of structures, systems, and components (SSCs) from performing their intended functions to mitigate the consequences of an initiating event within the assumed acceptance limits. Applicable design basis accidents and operational occurrences were evaluated using approved methods and codes. Results were within established acceptance criteria, and the plant's assumed source term remains bounding ; therefore, there's no increase in consequences.

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 accident from any accident previously evaluated?

Response: No.

The use of a new analytical methodology for the rod ejection analysis discussed in UFSAR Section 15.4.6 will not create the possibility of a new or different kind of 46

accident from any accident previously evaluated.

PG&E Letter DCL-25-087 The use of ADOPT fuel pellets will not result in adverse changes to the operation or configuration of the facility. The NRC-approved topical report WCAP-18482-P-A, which addresses ADOPT fuel pellets, demonstrates that ADOPT fuel pellets have essentially the same properties compared to standard UO2 pellets. Therefore, the ADOPT fuel pellets will perform similarly to the current fuel pellets, thus precluding the possibility of the fuel pellets becoming an accident initiator and causing a new or different kind of accident.

Therefore, the proposed changes do not cause the initiation of any accident nor create any new failure mechanisms. All equipment important to safety will operate as designed. Component integrity is not challenged. The proposed changes do not result in any event previously deemed incredible being made credible. The Technical Specifications will continue to require operation within the required core operating limits and appropriate actions will be taken when or if limits are exceeded.

Therefore, it is concluded that this 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 to the method of evaluation discussed in UFSAR Section 15.4.6 uses an NRC-approved methodology that shows compliance with the requirements of RG 1.236. Because no fuel failures are predicted, the proposed change does not reduce the margin of safety.

NRC-approved topical report WCAP-18482-P-A, which addresses ADOPT fuel pellets, demonstrates that ADOPT fuel pellets have essentially the same properties compared to standard UO2 pellets. ADOPT fuel pellets are expected to perform similarly to standard UO2 pellets for normal operating and accident scenarios, including both LOCA and non-LOCA scenarios. The use of ADOPT fuel pellets will not result in adverse changes to the operation or configuration of the facility. Operation in accordance with the revised TS 4.2.1 does not impact assumptions made in the safety analyses. This ensures that applicable design and performance criteria associated with the safety analysis will continue to be met and that the margin of safety is not affected.

The proposed Technical Specification (TS) 5.6.5 changes continue to require operation within the core limits that are based on NRC-approved reload design methodologies.

The proposed changes continue to ensure that appropriate actions will be taken if limits are violated. These actions remain unchanged. Future reloads will continue to ensure 47

PG&E Letter DCL-25-087 that operation of the units, within the cycle-specific limits, will not involve a reduction in the margin of safety as defined in the basis for any TS.

Therefore, it is concluded that the proposed change does 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 the proposed change, (2) such activities will be conducted in compliance with the Commission's 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.

5.0 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 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.

6.0 REFERENCES

1. WCAP-18482-P-A, "Westinghouse Advanced Doped Pellet Technology (ADOPT') Fuel," September 2022 (ML22316A013).

2. WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses,"

May 1999 (ML053050151 ).

3. WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999 (ML993160153).

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PG&E Letter DCL-25-087

4. WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis Design Model (PAD5)," November 2017 (ML17334A826).

5. WCAP-18240-P-A, "Westinghouse Thermal Design Procedure (WTDP),"

April 2020 (ML20104C042).

6. WCAP-15806-P-A, "Westinghouse Control Rod Ejection Accident Analysis Methodology Using Multi-Dimensional Kinetics," November 2003 (ML033350166).

7. Westinghouse Letter, L TR-NRC-22-7, "Fuel Criterion Evaluation Process (FCEP) Notification of the 17x17 OFA PRIME Fuel Product Implementation (Proprietary/Non-Proprietary)," dated February 28, 2022 (ML22059B071 ).

8. Diablo Canyon Power Plant Units 1 and 2 Final Safety Analysis Report Update, Revision 28, November 2024.

9. Westinghouse Topical Report, WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985

10. Westinghouse Topical Report, WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004 (ML042250311 ).

11. Westinghouse Topical Report, WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.

12.Westinghouse Topical Report, WCAP-16676-NP, "Analysis Update for the Inadvertent Loading Event," March 2009.

13.ADAMS Accession Number ML25174A192, "Diab lo Canyon Nuclear Power Plant, Units 1 And 2 - Issuance of Amendment Nos. 252 And 254 Re: Utilization of Optimized ZIRLO' for Improved Fuel Rod Cladding Performance (EPID L-2025-LLA-0059)," August 13, 2025.

14. WCAP-10444-P-A, "Reference Core Report VANTAGE 5 Fuel Assembly,"

September 1985 (Westinghouse Proprietary) and WCAP-10445-NP-A Westinghouse Non-Proprietary), Appendix A.2.0, September 1985.

(ML080650257).

15. Westinghouse Topical Report, WCAP-14565-P-A Addendum 2, "Addendum 2 to WCAP-14565-P-A Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications," April 2008. (ML081280711 ).

16. Pressurized Water Reactor Owner's Group Topical Report, PWROG-21001-P-A, "Hydrogen-Based Transient Cladding Strain Limit," October 2023 (ML23311A425).

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