ML25344A427

License Amendment Request 25-05 Application to Utilize Adopt Fuel for Improved Fuel Performance, Part 1

ML25344A427
Person / Time
Site:	Diablo Canyon
Issue date:	12/03/2025
From:	Rogers J Pacific Gas & Electric Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
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: ML25344A427 (0)

Text

I Enclosures 2 and 4 of this letter contain Proprietary Information - Withhold Under 1 O CFR 2.390 Pacific Gas and Ele'Ctric Company" PG&E Letter DCL-25-087 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 25-05 Justin E. Rogers Station Director Application to Utilize ADOPT' Fuel for Improved Fuel Performance

Dear Commissioners and Staff:

Diablo Canyon Power Plant Mail code 104/5/502 P.O. Box 56 Avila Beach, CA 93424 805.545.3088 Justin.Rogers@pge.com 10 CFR 50.90 Pursuant to 10 CFR 50.90, Pacific Gas and Electric Company (PG&E) hereby requests approval of the enclosed proposed amendment to modify the Diablo Canyon Power Plant (DCPP) Units 1 and 2 Technical Specifications (TS) 2.1.1, "Reactor Core Sls," TS 4.2.1, "Fuel Assemblies," and TS 5.6.5 "Core Operating Limits Report (COLR)" to allow the use of ADOPT' fuel pellets and adoption of improved analysis methods.

This license amendment request (LAR) is part of an advanced fuel feature strategy to optimize DCPP's fuel design and achieve superior fuel reliability and operating margins, resulting in safe, successful, economic operations. ADOPT fuel pellets provide increased thermal stability, enhanced corrosion resistance, and increased pellet density for improved core design flexibility consistent with safely maintaining DCPP generation and thereby supporting electrical grid reliability in California. This change is consistent with the Nuclear Regulatory Commission (NRC) safety evaluation that approved the use of ADOPT fuel pellets (Reference 1 ). provides a description and technical evaluation of the proposed changes, a regulatory evaluation, and a discussion of environmental considerations. to Enclosure 1 provides the existing respective DCPP Unit 1 and Unit 2 TS pages marked up to show the proposed changes. Attachment 2 to ADOPTm is a trademark or registered trademark of Westinghouse Electric Company LLC.

A member o f the STARS Al liance Callaway

• Diablo Ca nyon

• Pa l o Ve rde

• Wolf Creek Enclosures 2 and 4 of this letter contain Proprietary Information - Withhold Under 1 O CFR 2.390 When separa ted from Enclosures 2 and 4, this documen t is decon trolled

Enclosures 2 and 4 of this letter contain Proprietary Information - Withhold Under 10 CFR 2.390 Document Control Desk Page 2 PG&E Letter DCL-25-087 provides the revised (clean) respective DCPP Unit 1 and Unit 2 TS pages. provides the justification for compliance to the Limitations and Conditions from the Safety Evaluation Reports (SERs) which have been integrated into the following approved and verified Topical Reports: WCAP-18482-P-A, WCAP-14882-P-A, WCAP-14565-P-A, WCAP-14565-P-A Addendum 2, WCAP-17642-P-A Revision 1, and WCAP-18240-P-A as provided in References 1 through 6. Enclosure 3 provides the non-proprietary version of the information included in Enclosure 2. provides a plant specific justification for the application of WCAP-15806-P-A (Reference 7) to address Regulatory Guide (RG) 1.236, "Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents." Enclosure 5 provides the non-proprietary version of this information included in Enclosure 4.

Enclosures 2 and 4 contain information proprietary to Westinghouse Electric Company LLC ("Westinghouse"). Accordingly, Enclosure 6 provides the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-25-060 affidavit supporting the proprietary withholding request for and Enclosure 4. The request is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission ("Commission") and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations. Accordingly, PG&E requests that the Westinghouse proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390.

Correspondence with respect to the copyright or proprietary aspects of the application for withholding related to the Westinghouse proprietary information or the Westinghouse affidavit provided in Enclosure 6 should reference Westinghouse Letter CAW-25-060 and be addressed to Jerrod Ewing, Manager, Operating Plants Licensing, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

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.

PG&E requests approval of this LAR in accordance with the complexity of the submittal and established precedents from other Licensees. Approval is requested no later than September 30, 2026, to allow for sufficient time for purchase and A

member of the STARS Alliance Ca l laway

• Diablo Canyo n

• Pa l o Verde

• Wolf Creek Enclosures 2 and 4 of th is letter contain Proprietary Information - Withhold Under 10 CFR 2.390 When separated from Enclosures 2 and 4, this documen t is decon trolled

Enclosures 2 and 4 of this letter contain Proprietary Information - Withhold Under 10 CFR 2.390 Document Control Desk Page 3 PG&E Letter DCL-25-087 delivery of the new ADOPT fuel pellets, and fuel manufacturing lead times to support loading of ADOPT' fuel pellets in support of the DCPP Unit 2 Spring 2027 outage.

Once approved, the amendment shall be implemented by the Spring 2027 Unit 2 Refueling Outage 26 (2R26) and the Spring 2028 Unit 1 Refueling Outage 27 (1 R27).

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

This letter includes no revisions to existing regulatory commitments.

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

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 non-proprietary enclosures to the California Department of Public Health.

If you have any questions or require additional information, please contact Mr. 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 MJR/51298845 cc:

Diablo Distribution 12/3/25 Date cc/enc: Anthony Chu, Branch Chief, California Dept of Public Health Mahdi 0. Hayes, NRC Senior Resident Inspector Samson S. Lee, NRR Project Manager John D. Monninger, NRC Region IV Administrator Enclosures

1. Evaluation of the Proposed Change Attachments

1. Proposed Technical Specification Changes (Markups)

A membe r o f the STARS Alliance Cal laway

• Diablo Canyon

• Pa l o Verde

• Wolf Creek Enclosures 2 and 4 of this letter contain Proprietary Information - Withhold Under 10 CFR 2.390 When separated from Enclosures 2 and 4, this document is decontrolled

Enclosures 2 and 4 of this letter contain Proprietary Information - Withhold Under 10 CFR 2.390 Document Control Desk Page 4 PG&E Letter DCL-25-087

2. Revised Technical Specifications Pages (Clean)

2. Limitations and Conditions Compliance for WCAP-18482-P-A, WCAP-14882-P-A, WCAP-14565-P-A, WCAP-14565-P-A Addendum 2, WCAP-17642-P-A Revision 1, and WCAP-18240-P-A (Proprietary)

3. Limitations and Conditions Compliance for WCAP-18482-NP-A, WCAP-14882-P-A, WCAP-15306-NP-A, WCAP-15306-NP-A Addendum 2-NP-A, WCAP-17642-NP-A Revision 1, and WCAP-18240-NP-A (Non-Proprietary)

4. Plant Specific Justification for Application of WCAP-15806-P-A, Revision 0, "Westinghouse Control Rod Ejection Accident Analysis Methodology Using Multi-Dimensional Kinetics," November 2003, to Address RG-1.236 (Proprietary)

5. Plant Specific Justification for Application of WCAP-15806-P-A, Revision 0, "Westinghouse Control Rod Ejection Accident Analysis Methodology Using Multi-Dimensional Kinetics," November 2003, to Address RG-1.236 (Non-Proprietary)

6. Application for Withholding Proprietary Information from Public Disclosure CAW-25-060

References:

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

2. Westinghouse Topical Report, WCAP-14882-P-A, "RETRAN-02 Modeling and Qualification for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses," May 1999 (ML053050151 ).

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

4. Westinghouse Topical Report, WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PADS),"

November 2017 (ML17334A826).

5. Westinghouse Topical Report, WCAP-18240-P-A, "Westinghouse Thermal Design Procedure (WTDP)," April 2020 (ML20104C042).

6. 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 ).

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

A member of the STARS Al l iance Ca l laway

• Diablo Canyon

• Palo Verde

• Wolf Creek Enclosures 2 and 4 of this letter conta in Prop rietary Information - Withhold Under 10 CFR 2.390 When separated from Enclosures 2 and 4, this document is decontrolled PG&E Letter DCL-25-087 Evaluation of the Proposed Change

Subject:

Application to Utilize ADOPT' Fuel for Improved Fuel Performance.

1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

2.1 Background

2.2 System Design and Operation 2.3 Current Requirements 2.4 Reason for Proposed Change 2.5 Description of Proposed Change

3.0 TECHNICAL EVALUATION

3.1 Nuclear Design 3.2 Core Thermal Hydraulic Design 3.3 Core Thermal Hydraulic Analysis 3.4 Fuel Rod Design and Performance 3.5 Mechanical Compatibility and Performance 3.6 Non-LOCA 3.7 LOCA 3.8 Control Rod Ejection 3.9 Dose Analysis

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration Determination 4.4 Conclusions

6.0 REFERENCES

ADOPrTM, PRIM Em, Optimized ZIRLO', Low Tin ZIRLO', ZIRLO, FULL SPECTRUMrn LOCA, and FSLOCA', are trademarks or registered trademarks of Westinghouse Electric Company LLC.

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ATTACHMENTS:

PG&E Letter DCL-25-087

1.

Proposed Technical Specification Changes (Markups)

2.

Revised Technical Specification Pages (Clean) 2

1.0

SUMMARY

DESCRIPTION PG&E Letter DCL-25-087 Pacific Gas and Electric (PG&E) is proposing to amend Operating Licenses DPR-80 and DPR-82 for Diablo Canyon Power Plant (DCPP) Units 1 and 2, respectively. The proposed Technical Specification (TS) changes and methodology changes described further in Enclosure 1 are needed to integrate specific fuel product changes.

The proposed license amendment identifies changes to DCPP that include a transition to improved fuel materials and adoption of improved analysis methods. These changes provide a more robust fuel product that maximizes fuel reliability with a supporting safety analysis that conforms to current regulatory guidance.

The fuel assembly-related changes include transition from VANTAGE+ fuel (17x17 Optimized Fuel Assembly (OFA)) to VANTAGE+ fuel with an enhanced PRIME' fuel assembly skeleton and use of ADOPT fuel pellets. These changes provide a more robust fuel design with better protection from debris transported in the reactor coolant, better fuel thermal performance, improved cladding corrosion resistance, and improved fuel utilization. Note that DCPP will transition to PRIME fuel features via 10 CFR 50.59.

The improved analysis methods include Westinghouse Performance Analysis and Design Model (PAD5) for improved fuel rod performance modeling, improved thermal hydraulics codes and methodologies including the Westinghouse Thermal Design Procedure (WTDP), and an updated Control Rod Ejection Accident analysis. These changes update the DCPP supporting analyses to conform with current regulatory guidance and align with modern codes and methods.

In support of the above, the following Technical Specifications (TSs) will require revision: TS 2.1.1, "Reactor Core Sls," to reflect the use of updated Thermal Hydraulic inputs; TS 4.2.1, "Fuel Assemblies," to reflect the use of integral dopants, and TS 5.6.5, "Core Operating Limits Report (COLR)," to include updated analytical methods.

2.0 DETAILED DESCRIPTION This section gives a brief overview of the DCPP current plant configuration as it relates to fuel, including cladding type and burnable absorbers. Additionally, this section summarizes the intended changes to the fuel configuration, analysis methodologies impacted by these changes, new analysis methodologies that support the change, and why the change is planned. TSs that are impacted by these changes are identified, along with the proposed changes to align the specifications with the new fuel and analysis methodologies.

2.1 Background

DCPP plans to implement the following fuel-related changes:

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PG&E Letter DCL-25-087 ADOPT fuel pellets (a modified uranium dioxide (UO2) pellet doped with small amounts of chromia (Cr2O3) and alumina (Al2O3)

PRIME advanced fuel features New methodologies are also being applied for Thermal Hydraulic (T/H) analysis and for the analysis of the Rod Ejection Accident.

The NRC was notified of the generic 17x17 fuel design changes for implementation of PRIME fuel features in L TR-NRC-22-7, "Fuel Criterion Evaluation Process (FCEP)

Notification of the 17x17 OFA PRIME Fuel Product Implementation (Proprietary/Non-Proprietary)," February 28, 2022 (Reference 7).

In summary, based on the above information, PG&E is requesting the Nuclear Regulatory Commission's (NRC's) approval of the following licensing actions to improve fuel reliability and update codes and methods to current industry standards:

Requesting a change to the licensing basis to apply the WCAP-15806-P-A methodology as the new method of analyzing Rod Ejection Accidents; Requesting a change to the licensing basis to apply the WCAP-17642-P-A Revision 1 methodology as the new method of analyzing Fuel Rod Design parameters; Requesting a change to TS Figure 2.1.1-1, "Reactor Core Safety Limit" to incorporate the change to the Departure from Nuclear Boiling (DNB) design basis described in WCAP-18240-P-A, WTDP, and WCAP-14565-P-A, VIPRE-01 modeling methodology; Requesting a change to TS 4.2.1 to reflect the use of ADOPT fuel pellets; Requesting changes to TS 5.6.5 "Core Operating Limits Report," Section b, to include WCAP-18240-P-A for WTDP methodology and WCAP-14565 for VIPRE-01 modeling methodology.

2.2 System Design and Operation The DCPP Updated Final Safety Analysis Report (UFSAR), Chapter 4, summarizes the current fuel design and application (Reference 8). Section 4.2.1.1 describes the fuel assembly design bases. The fuel rods are cold worked partially annealed Zircaloy-4, ZIRLO tubes containing enriched UO2 fuel. The core will also contain Optimized ZIRLO fuel rods, which were approved by the NRC in Reference 13. All fuel rods are pressurized with helium during fabrication to reduce stresses and strains and to increase fatigue life.

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PG&E Letter DCL-25-087 The fuel assembly consists of the guide thimbles fastened to the grids and the top and bottom nozzles. The fuel rods are held in this assembly at points along their length by spring-clip grids which provide support for the fuel rods.

DCPP units are loaded with Westinghouse VANTAGE+ fuel assemblies. Full length Rod Cluster Control Assemblies (RCCA), secondary sources, thimble plug devices, and burnable poison rods may be inserted into the guide thimbles of the fuel assemblies.

The absorber sections of the RCCAs are fabricated with silver-indium-cadmium alloy sealed in stainless steel tubes.

Two types of burnable absorbers can be used at DCPP:

Wet Annular Burnable Absorbers (WABA), each consisting of an aluminum oxide-boron carbide burnable absorber material contained within two concentric Zircaloy tubes, which is a discrete burnable absorber suspended and positioned within selected guide thimbles within the fuel assemblies and ;

Integral Fuel Burnable Absorbers (IFBA), consisting of a thin Zirconium diboride coating on the surface of the fuel pellets.

2.3 Current Requirements The following TSs identify the current plant compliance requirements. Implementation of this fuel upgrade will require changes to these areas as discussed throughout this amendment request. Prior to the transition to a full core of PRIME and ADOPT fuel features, the mixed core will be designed in compliance with the limitations and conditions of currently approved methods.

Descriptions of PRIME fuel features, their benefits, and impacts are provided for information only.

TS Figure 2.1.1-1, "Reactor Core Safety Limit" specifies the fuel centerline melt and Departure from Nuclear Boiling Ratio (DNBR) limits for fuel at DCPP.

TS 4.2.1, "Fuel Assemblies" specifies the requirements for the fuel cladding material as well as the fuel pellet materials.

TS 5.6.5, "Core Operating Limits Report," Section b, provides the list of the NRC approved COLR methods.

The current methodology for the Control Rod Ejection Accident is described in UFSAR Section 15.4.6 (Reference 8).

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2.4 Reason for the Proposed Change PG&E Letter DCL-25-087 ADOPT fuel pellets provide a higher density and are more thermally stable than standard UO2 fuel. The ADOPT fuel pellets provide additional reactivity for use in the core and lead to improved core design flexibility as well as consistent, safe DCPP generation, thereby supporting electrical grid reliability in California. PRIME advanced fuel features improve fuel performance in different mechanical design areas, which reduces oxidation, hydrogen absorption, growth, and improves grid to-rod fretting (GTRF) margin.

With the use of updated COLR analysis methodologies, VIPRE-01 and WTDP, TS 5.6.5, "Core Operating Limits Report," Section b is updated to include WCAP-14565-P-A and WCAP-18240-P-A in the list of NRC-approved COLR References.

The proposed changes to the DCPP TSs incorporate the changes summarized above.

2.5 Description of the Proposed Change TS Figure 2.1.1-1, "Reactor Core Safety Limit" is modified to incorporate the change to the DNB design basis for WTDP and VIPRE-W.

TS 4.2.1 "Fuel Assemblies" is modified to recognize the inclusion of dopants into the standard UO2 fuel pellets.

TS 5.6.5, Section b, is modified to add NRC-approved topical reports WCAP-14565-P-A (Reference 3) and WCAP-18240-P-A (Reference 5) to the list of approved topical reports for the development of the COLR. summarizes the changes to UFSAR Section 15.4.6 necessary to reflect the use of a new methodology for analyzing the impact of control rod ejection. The UFSAR changes are summarized for information only and will be incorporated into the DCPP Units 1 and 2 licensing basis in accordance with 10 CFR 50.71 (e) upon implementation of the approved license amendments.

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3.0 TECHNICAL EVALUATION

PG&E Letter DCL-25-087 The following sections provide a summary of the evaluations performed for the plant-specific application of ADOPT fuel pellets. A demonstration of the evaluation methodologies has been developed, as shown in Section 3.6, to provide assurance that the plant licensing bases are met for the continued operation with Westinghouse VANTAGE+ with PRIME Fuel Features and ADOPT fuel pellets.

Note that several other technical evaluations have been performed, which are not discussed below, with adequate margin available in all cases. All evaluations are available for audit.

3.1 Nuclear Design 3.1.1 Introduction and Background The purpose of the DCPP core analysis is to develop, prior to the cycle-specific reload design, the appropriate range of values for the key safety parameters for the plant. This will allow the safety analysis evaluations to be completed prior to the cycle-specific design analysis and will allow the future DCPP core reload analysis to utilize the methodology described in Reference 9.

DCPP will transition from Westinghouse VANTAGE+ fuel to Westinghouse VANTAGE+

with PRIME Fuel Features and ADOPT fuel to maximize fuel reliability. The Westinghouse PRIME fuel design will contain fuel components as described in Section 2.1. Section 3.5 of this report provides a comparison of the two fuel types that will reside in DCPP Units 1 and 2 during the transition.

The specific values of the core safety parameters (e.g. power distribution, peaking factors, rod worths, reactivity coefficients, and kinetic parameters) are primarily loading-pattern dependent. The variations in the loading-pattern-dependent safety parameters are expected to be represented by the transition and equilibrium cores developed and analyzed for this project. NRC approved analytical codes and methods (References 9 through 11) will accurately describe the neutronic behavior of the DCPP core. The NRC approved codes and methods currently used will continue to be used following the transition to the new fuel product.

3.1.2 Input Parameters and Assumptions The licensing basis for the reload core design is defined in subsection 4.1 of the UFSAR. Continued applicability of the nuclear design inputs to the safety analysis will be evaluated or re-analyzed during the reload safety evaluation process for the reload cycles consistent with Reference 9. The reload design methodology includes the evaluation of the key core safety parameters that consist of the nuclear design dependent input to the UFSAR safety evaluation for each reload.

DCPP will utilize the PRIME fuel design and ADOPT fuel pellets, starting with a mixed core of VANTAGE+ fuel and PRIME fuel features. The effect of mixed cores on DNB analyses is assessed in Section 3.5. A cycle-specific DNB penalty on the VANTAGE+

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PG&E Letter DCL-25-087 fuel is applied by the thermal-hydraulic design group consistent with standard reload practice for mixed cores (Reference 9). The DNB penalty will be analytically applied while the Nuclear Enthalpy Rish Hot Channel Factor (Ft.H) surveilled COLR limit will be maintained consistent between the VANTAGE+ fuel and the PRIME fuel.

The maximum total core peaking factor (Fa) limit utilized in the COLR will be consistent with the lower limit of the Loss of Coolant Accident (LOCA) analysis of record (AOR) or the rod ejection analysis of record. All fuel in the core (VANTAGE+ and PRIME fuel) will be monitored to one Fa limit.

The VANTAGE+ fuel and PRIME fuel with ADOPT fuel pellets are confirmed to remain within the safety analysis assumptions consistent with the methods of Reference 9 for each cycle.

3.1.3 Description of Analyses and Evaluations Representative core designs were developed for the first two transition cycles that include VANTAGE+ fuel as well as full cores of PRIME fuel with ADOPT fuel pellet implementation. Note that the core designs utilized are not intended to represent limiting designs but were instead developed with the intent to determine if sufficient margin exists between typical parameter values and the corresponding safety analysis limits to allow flexibility in designing future cores. The margins for the DCPP core designs were compared to the values for recent reload cycles to evaluate the adequacy of margins between typical safety parameter values and the corresponding limits. As mentioned previously, cycle-specific calculations will confirm that the actual values are within the safety analysis limits.

The representative core designs were utilized in the FULL SPECTRUM' LOCA (FSLOCA ') methodology limit setting. These cores exhibit peaking factors which are intended to represent future reload peaking factors. These core designs are representative of the range of core characteristics expected during and after the fuel transition from Westinghouse VANTAGE+ fuel to Westinghouse PRIME fuel with ADOPT fuel pellets.

3.1.3.1 Inadvertent Loading of an Assembly into an Improper Position The inadvertent loading of an assembly into an improper position described in UFSAR Section 15.3.3 and analyzed in Reference 12, is not dependent on fuel product. The accident detection is determined based on the moveable incore instrumentation which is unchanged as part of the transition. The Reference analysis is unimpacted and continues to be applicable during and following the fuel transition.

3.1.3.2 Single Rod Cluster Control Assembly Withdrawal at Hot Full Power The single rod cluster control assembly withdrawal at hot full power analysis described in UFSAR Section 15.3.5 has significant margin and is confirmed on a cycle-specific basis both during the transition and all subsequent reloads. The analysis supporting this amendment confirms that 5% of the fuel or less will experience DNBR during a 8

PG&E Letter DCL-25-087 single rod withdrawal event. Approved methods are used to confirm the event on subsequent reloads.

3.1.3.3 Static Rod Misalignment The acceptance criteria related to the statically misaligned rod analysis is to be confirmed on a cycle-specific basis both during the transition and all subsequent reloads. The analysis confirms that the DNB design basis is met. Approved methods are used to confirm that the acceptance criteria are met.

3.1.4 Acceptance Criteria The nuclear design analysis for transition to ADOPT fuel pellets was performed using standard core reload methodology described in Reference 9 with the PARAGON and Advanced Nodal Code (ANC9) codes described in References 10 and 11, respectively.

These licensed methods and models have been used in DCPP and other Westinghouse plants and have been demonstrated to accurately describe the neutronic behavior of many different reactor cores. No changes to the nuclear design bases or licensed methods are necessary because of the transition.

The reload design methodology includes an evaluation of the reload core key safety parameters that consist of the nuclear design dependent input to the reload fuel safety evaluation for each reload cycle. This methodology is described in Reference 9. These key safety parameters are evaluated for each DCPP reload cycle. If one or more of the key parameters fall outside the bounds assumed in the licensing basis, then the affected transients will be re-evaluated using approved methods, and the results documented in the Reload Evaluation (RE) report for that cycle.

3.1.5 Results The key safety parameters evaluated for the DCPP transition to ADOPT fuel pellets are similar to the current DCPP designs. These parameters vary cycle-to-cycle as fuel loading patterns change each cycle. The margins available afford the flexibility to accommodate normal cycle-to-cycle variations expected in core loading patterns. The normal methods of feed enrichment variations and insertion of fresh burnable absorbers will be employed to control peaking factors. Compliance with the peaking factor technical specifications can be assured using the NRC approved codes and methods.

3.1.6 Conclusions Margin to key safety parameter limits are sufficient to accommodate transition cycle designs and full implementation of ADOPT fuel pellets. COLR peaking factor limits will be set according to the lowest value supported for the resident fuel assembly's safety analysis of record. General methodology used to support DCPP reload analysis will remain consistent with currently utilized methods described in References 9 through 11.

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3.2 Core Thermal Hydraulic Design PG&E Letter DCL-25-087 DCPP currently operates with the 17x17 OFA fuel design with Intermediate Flow Mixer (IFM) grids (Reference 8) and plans to implement the 17x17 OFA PRIME fuel design with IFM grids in conjunction with incorporating the use of ADOPT fuel pellets.

The 17x17 OFA PRIME fuel assembly design incorporates material changes to the mid and IFM grids from ZIRLO to Low Tin ZIRLO', adjustments to the bottom nozzle flow hole geometries to reduce the pressure drop of the fuel assembly inlet region, and a new guide thimble design featuring an external dashpot tube between the bottom grid and the first mid grid to minimize the potential for incomplete rod insertion. The loss coefficient of the OFA PRIME Advanced Debris Filter Bottom Nozzle (ADFBN) is lower than the standard debris filter bottom nozzle (SDFBN) on a bottom nozzle component basis, which has a corresponding reduction for the overall fuel assembly loss coefficient.

The fuel assemblies must be held down with sufficient load on the lower core plate for all normal operating conditions and must not be damaged during a pump overspeed condition. The larger the number of 17x17 OFA PRIME low resistance fuel assemblies loaded in the core, the lower the lift force increase on the 17x17 OFA PRIME fuel assemblies. This transition core effect has been evaluated and continues to provide sufficient margin to the lift force calculational limit.

The thermal-hydraulic design criteria for thimble tube and core components include the maximum core bypass flow limit, surface boiling in the dashpot, bulk boiling in the thimble and the maximum core component rod temperature. Thimble bypass flow is primarily driven by the difference in pressure between the guide thimble flow holes and the top nozzle, guide thimble flow hole geometry and the thimble tube inside diameter.

The 17x17 OFA PRIME fuel design changes are concluded to have no significant impact on thimble bypass flow.

A thermal evaluation of the thimble and dash pot tubes with core components confirms that the RCCA absorber rod in the 17x17 OFA PRIME fuel assembly will not experience bulk boiling in the thimble or surface boiling in the dashpot. It was determined that bulk boiling in the thimble will not occur for the RCCA absorber rod and margin to surface boiling in the dashpot region is maintained.

In the evaluation of the maximum temperature of the core component rods, the rod surface temperature is assumed to be the maximum surface temperature at system pressure with local boiling existing. Therefore, the change in the bottom nozzle loss coefficient has no impact on this criterion.

3.3 Core Thermal Hydraulic Analysis This section describes the thermal-hydraulic (T/H) analysis performed to support the operation of DCPP utilizing the Westinghouse 17x17 OFA fuel design with PRIME fuel features, Optimized ZIRLO fuel cladding, and ADOPT fuel pellets. The thermal-10

PG&E Letter DCL-25-087 hydraulic design methods for the fuel upgrade remain the same as discussed in the UFSAR except for three changes:

1. The NRC-approved VIPRE-W (VIPRE) subchannel analysis code (Reference 3) was used in place of the TH INC-IV subchannel analysis code and the FACTRAN code for DNBR calculations.

2. The NRC-approved ABB-NV and WLOP correlations in Reference 15 were used in place of the W-3 correlation as the secondary ON B correlation for conditions where the primary DNB correlation is not applicable.

3. The NRC-approved WTDP methodology in Reference 5 was used in place of the ITDP methodology to statistically account for uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters, computer codes, and DNB correlation into design DNBR limit on 95/95 basis.

See Enclosure 2 for dispositioning how the Limitations and Conditions specified in the Safety Evaluation Report section of these topical reports are met.

The current T/H design basis for DCPP includes the prevention of DNB on the limiting fuel rod with a 95 percent probability at a 95 percent confidence level (95/95) for Condition I and Condition II events, as specified in the DCPP FSAR Section 4.4 (Thermal and Hydraulic Design), Subsection 4.4.2.1, Departure from Nucleate Boiling Acceptance Criteria. The DNB analysis with these new fuel features is based on this licensing basis.

The T/H DNB analysis of the proposed fuel at DCPP is based on the WTDP (Reference

5) and the WRB-2 DNB correlation (Reference 14) using the Westinghouse version of the VIPRE-01 subchannel analysis code (Reference 3). The Standard Thermal Design Procedure (STOP) is used when WTDP is not applicable. For analyses which are outside of the range of applicability of the WRB-2 correlation, the ABB-NV and WLOP correlations are used (Reference 15).

The DNB analyses demonstrate that the 95/95 DNB design basis is met at various core conditions corresponding to normal operation, operational transients, and transient conditions arising from faults of moderate frequency. See Enclosure 2 for disposition of limitations and conditions specified in the Safety Evaluation Report sections of the VIPRE-01, ABB-NV/WLOP, and WTDP topical reports. Using the previously described methodology, the DNB analyses performed to support the transition to PRIME fuel features with ADOPT fuel pellets are addressed for the following events: locked rotor, loss-of-flow, RCCA drop/mis-operation, steam line break accident, and uncontrolled RCCA bank withdrawal from subcritical. These are defined in UFSAR Sections 15.4.4, 15.3.4, 15.2.3, 15.4.2, and 15.2.1. The acceptance criteria corresponding to accidents that trip on Over Temperature Delta Temperature (OT 11 T), including uncontrolled RCCA bank withdrawal from power and uncontrolled single RCCA withdrawal at power as described in FSAR Sections 15.2.2 and 15.3.5, were also confirmed to be met.

For each of the events listed above, the DNB analyses determined that the DNBR limits are met with sufficient margin to support the DCPP fuel upgrade project. Cycle-specific evaluations will be performed in accordance with Reference 9. For the Locked Rotor 11

PG&E Letter DCL-25-087 event, a portion of the retained plant margin between the safety analysis limit DNBR and the design limit DNBR was allocated to offset a DNBR penalty associated with the minimum DNBR result being below the safety analysis limit. Despite retained plant margin being allocated to preclude a percentage of rods assuming to fail because of being in DNB, a confirmatory VIPRE-W analysis was performed that conservatively assumes that post-Critical Heat Flux (CHF) conditions were reached at the beginning of the transient to maximize the cladding temperature for the rod with the highest radial peaking factor. The result of this analysis confirmed that the peak clad temperature is well below the allowable limit for Optimized ZIRLO cladding to prevent clad embrittlement (2375°F).

3.4 Fuel Rod Design and Performance DCPP currently is approved to use UO2 fuel pellets, ZrB2 IFBA, ZIRLO, and Optimized ZIRLO cladding. As part of the fuel product transition, the following fuel products and fuel performance methodologies will be implemented:

ADOPT Fuel Pellets 17x17 OFA with PRIME Advanced Fuel Features PAD5 Fuel Performance Models Hydrogen-Based Transient Cladding Strain Methodology It should be noted that the PRIME fuel features and the Hydrogen-Based Transient Cladding Strain Methodology discussed herein are for information only. DCPP will transition to the Hydrogen-Based Transient Cladding Strain Methodology in accordance with 10 CFR 50.59. The NRC was notified of the generic 17x17 fuel design changes for implementation of PRIME fuel features in Reference 7, while the Hydrogen-Based Transient Cladding Strain Methodology was approved for use in Reference 16.

ADOPT fuel pellets and the associated fuel performance methodology are NRC-approved as documented in Reference 1. The impacts of fuel rod design (FRO) due to use of ADOPT pellets has been assessed. All FRO criteria are met for both transition and equilibrium cores. ADOPT fuel pellets are therefore shown to be an acceptable change for DCPP. Compliance with the NRC Safety Evaluation Limitations and Conditions for Reference 1 are discussed in Enclosure 2.

A small difference in coolant temperature is expected for transition cycles during implementation of PRIME advanced fuel features due to the loss coefficient differences between the PRIME ADFBN and the co-resident SDFBN, but the difference is negligible and easily accommodated with available margin. Once a full equilibrium core of PRIME advanced fuel features is implemented, there are no effects to fuel performance. Based on this, the use of PRIME advanced fuel features is an acceptable change for DCPP.

All fuel performance evaluations were performed using the latest fuel performance code, PAD5 (Reference 4). This computer code iteratively calculates the interrelated effects of fuel and cladding deformations including fuel densification, fuel swelling, fuel relocation, fuel rod temperatures, fill and fission gas release (FGR), and rod internal pressure (RIP), as a function of time and linear power. PAD evaluates the power history of a fuel rod as a series of steady-state power levels with instantaneous jumps 12

PG&E Letter DCL-25-087 from one power level to another. The length of the fuel rod is divided into several axial segments and each segment is assumed to operate at a constant set of conditions over its length. Fuel densification and swelling, cladding stresses and strains, temperatures, burnup and FGRs are calculated separately for each axial segment and the effects are integrated to obtain the overall FGR and resulting internal pressure for each time step.

The coolant temperature rise along the fuel rod is calculated based on the flow rate and axial power distribution, and the cladding surface temperature is determined with consideration of corrosion effects and the possibility of local boiling. A comprehensive description of all PAD5 models, NRC Requests for Additional Information (RAI), and the NRC's safety evaluation are documented in Reference 4. Compliance with the NRC Safety Evaluation Limitations and Conditions for Reference 4 are discussed in.

An alternate hydrogen-based approach to evaluating transient cladding strain was also utilized, consistent with the NRC-approved methodology outlined in Reference 16. All design limits are also confirmed as part of the standard reload analyses for each cycle consistent with Reference 9.

The existing licensing basis and methods for the resident (outgoing) ZIRLO cladding fuel product will be retained during the transition cycles, and all new methods, as described above, will be used for all new (incoming) fuel products that will include Optimized ZIRLO cladding, PRIME advanced fuel features, and ADOPT fuel pellets.

Note that NRC approval for the use of Optimized ZIRLO cladding at DCPP was provided in Reference 13. The existing AOR will therefore be applied to the outgoing fuel product and the new AOR will be applied to the new fuel product features. Limit confirmations will be conducted on a reload-specific basis consistent with Reference 9.

3.5 Mechanical Compatibility and Performance This section describes the mechanical compatibility and performance evaluation for the transition to the 17x17 OFA PRIME fuel design with ADOPT fuel pellets. The evaluation includes the mixed core configurations of the resident 17x17 VANTAGE+ with Debris Mitigation Features fuel design and the 17x17 OFA PRIME fuel designs with ADOPT fuel pellets. The evaluation addressed the effect on fuel assembly performance and fuel rod performance.

The criteria pertinent to the mechanical compatibility and performance of the fuel assembly that must be satisfied include the fuel assembly top nozzle holddown force, fuel assembly shipping and handling loads, seismic/LOCA analysis, and fuel assembly interfaces, clearances, and compatibility.

Criteria pertinent to the compatibility and performance of the fuel rod that must be satisfied include GTRF wear and the fuel rod shoulder gap.

A discussion of how the various applicable criteria are satisfied is presented below.

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3.5.1 Fuel Assembly Performance PG&E Letter DCL-25-087 There is a slightly lower pressure drop compared to the resident fuel SDFBN, due to the lower loss coefficient associated with the 17x17 OFA PRIME fuel assembly bottom nozzle. This results in slightly higher best estimate Reactor Coolant System (RCS) flow. There is also a slight increase to the fuel assembly weight compared to the resident fuel as a result of the transition to ADOPT fuel pellets and the heavier 17x17 OFA PRIME bottom nozzle. Both changes could have an effect on the fuel assembly top nozzle holddown force. An evaluation was performed that demonstrated these changes were not significant and the top nozzle holddown force criteria are all satisfied.

The fuel mechanical design criteria associated with Shipping and Handling Loads for Condition I and II events and various fuel assembly mechanical structural design criteria associated with the 17x17 OFA PRIME fuel design with Advanced Fuel Features fuel product were specifically evaluated and it was confirmed through the Westinghouse FCEP (Reference 7) that all the design acceptance criteria related to the fuel assembly structural design requirements are satisfied.

Based on seismic/LOCA evaluations performed for other Westinghouse plants of similar design with the 17x17 OFA PRIME fuel assembly design and given current DCPP seismic/LOCA margins, all relevant fuel assembly seismic/LOCA criteria are expected to be met for DCPP with the 17x17 OFA PRIME fuel features and with the ADOPT pellet stack. Results of this analysis are available for audit.

All applicable interfaces and clearances remain valid and bounding as the fuel assembly envelopes and interfacing dimensions are unchanged for the 17x17 OFA PRIME fuel design with ADOPT fuel pellets as compared to the resident 17x17 VANTAGE+ fuel design. Therefore, applicable interfaces and clearances remain acceptable and remain compatible with the existing 17x17 VANTAGE+ fuel design.

3.5.2 Fuel Rod Performance For GTRF, the design criteria is that wall thickness reduction be no greater than 10 percent when evaluating cladding imperfections, including fretting wear marks.

Analyses were performed examining previous VIPRE testing and post irradiation examination (PIE) results for the 17x17 OFA fuel design and the impacts of changing the grid material from ZIRLO to Low Tin ZIRLO. VIPRE testing and PIE inspections confirmed that the 17x17 VANTAGE+ fuel design is robust with respect to GTRF and satisfies the design criteria for fretting wear performance.

The transition to the 17x17 OFA PRIME Low Tin ZIRLO grids will reduce oxide layer thickness on the grids (lower corrosion rate) as well as minimizing grid growth, both of which result in lower levels of GTRF. Overall, it is expected that the presence of the Low Tin ZIRLO grids will lead to increased margin to GTRF when compared to the current 17x17 VANTAGE+ fuel design. The thinner oxide layers associated with the Optimized ZIRLO fuel rod cladding also provides adequate GTRF protection. Based on an extensive review of field experience for Low Tin ZIRLO grids used with Optimized ZIRLO fuel cladding as detailed in Reference 7, wear is expected to be less than the 14

PG&E Letter DCL-25-087 10% criterion for the 17x17 OFA PRIME fuel. Thus, the GTRF failure risk is considered to be low and acceptable.

The Fuel Rod Shoulder Gap criterion requires that the clearance between the top of the fuel rods and the bottom of the top nozzle adapter plate shall be sufficient to preclude contact of these components. There are no changes to either the grid design nor the fuel rod design that would preclude the ability of the grid to accommodate the differential expansion between the fuel rods and the grid locations within the fuel assembly skeleton. Additionally, the presence of the external dashpot design for the guide thimble design helps ensure that the overall fuel stiffness is enhanced. Thus, the criteria related to fuel assembly distortion and buckling of the fuel rod are satisfied as it is demonstrated that there is more than adequate shoulder gap for the 17x17 OFA PRIME fuel assembly design and ADOPT fuel pellets. In conclusion, the criterion related to fuel rod shoulder gap, including fuel assembly distortion and buckling of the fuel rods, is satisfied.

Criterion regarding cladding stress and fatigue impact due to the transition to 17 OFA PRIME fuel design with ADOPT fuel pellets are met and available for audit.

3.5.3 Cumulative Effects of Fuel Changes The evaluation of the change to 17x17 OFA PRIME fuel design with ADOPT fuel pellets considered the integrated and cumulative effects of changes to fuel assembly characteristics /attributes, such as the fuel assembly weight, grid crush values, corrosion rates (oxide layer), component material properties, fuel rod growth rates, fuel assembly/grid dimensional performance, and GTRF performance. All have been cumulatively taken into consideration to ensure that all the fuel mechanical design criteria, as discussed herein, are satisfied. Based on the evaluations presented herein, it is concluded that the cumulative effects of the 17x17 OFA PRIME fuel design with ADOPT fuel pellets on the applicable fuel mechanical design criteria are acceptable, and all applicable design and safety criteria remain satisfied.

3.5.4 Conclusions With respect to the mechanical design changes with the 17x17 OFA PRIME fuel design with ADOPT fuel pellets, it is concluded that this integrated fuel design is structurally and mechanically acceptable and all applicable design and safety criteria remain satisfied. The 17x17 OFA PRIME fuel design with ADOPT fuel pellets product is compatible with the existing 17x17 VANTAGE+ fuel design.

Based on the evaluations presented herein, it is also concluded that the cumulative effects of the 17x17 OFA PRIME fuel design with ADOPT fuel pellets on the applicable fuel mechanical design criteria are acceptable, and all applicable design and safety criteria remain satisfied.

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3.6 Non-LOCA 3.6.1 Non-LOCA Analyses Introduction PG&E Letter DCL-25-087 This section summarizes the evaluations and analyses of the non-loss-of-coolant accident (non-LOCA) events performed to support the proposed fuel upgrades and computer code and methodology updates. The fuel upgrades involve the use of the Westinghouse 17x17 OFA fuel design with PRIME fuel features and ADOPT fuel pellets.

The computer code and methodology updates involve the use of the RETRAN-02W computer code (Reference 2) for non-LOCA events previously analyzed with other computer codes, the use of the VIPRE-W computer code (Reference 3) and the WTDP methodology (Reference 5) for the core thermal hydraulic (T/H) DNB analyses and the use of analysis inputs based on the PAD5 fuel performance analysis methods (Reference 4). A number of the non-LOCA events were reanalyzed to incorporate changes associated with the proposed fuel upgrades and computer code and methodology updates, while the existing analyses of record for other events were determined by evaluation to remain valid.

3.6.1.1 Initial Conditions For most non-LOCA events analyzed for DNB concerns, the WTDP methodology (Reference 5) was applied as a replacement for the ITDP methodology (Reference 17).

With the WTDP methodology, similar to the ITDP methodology, nominal values are modeled for the initial RCS conditions of power, temperature, pressure, and flow, and the corresponding uncertainty allowances are accounted for statistically in defining the DNBR safety analysis limit. The nominal RCS flow modeled in WTDP analyses is the minimum measured flow (MMF).

For DNB analyses in which the WTDP methodology is not employed, or for events not analyzed for DNB concerns, the initial conditions were modeled with the maximum, steady-state uncertainties applied to the nominal values in the most conservative direction; this is known as STOP for DNB analyses or non-WTDP. In these analyses, the RCS flow was set equal to the thermal design flow (TDF), and the following steady-state initial condition uncertainties were applied as applicable.

+2% Nuclear Steam Supply System (NSSS) power allowance for calorimetric measurement uncertainty.

+5.0°F I -5.5°F Tavg allowance for deadband and system measurement uncertainties.

+/-60 psi pressurizer pressure allowance for steady-state fluctuations and measurement uncertainties.

3.6.1.2 Fuel Design Mechanical Features The effects of fuel design mechanical features were accounted for in the non-LOCA analyses in fuel-related input parameters such as fuel and cladding dimensions, fuel 16

PG&E Letter DCL-25-087 and cladding material properties, fuel temperatures, and core bypass flow. The fuel temperatures utilized were based on the use of the PAD5 fuel performance analysis methods (Reference 4). The DNBR analyses accounted for mixed cores of 17x17 OFA and OFA PRIME fuel assemblies, and full cores of 17x17 OFA PRIME fuel assemblies.

3.6.1.3 Reactivity Coefficients The transient response of the reactor core is dependent on reactivity feedback effects, in particular the Moderator Temperature Coefficient (MTC) and the Doppler Power Coefficient (DPC). Depending upon event-specific characteristics, conservatism dictates the use of either minimum or maximum reactivity coefficient values. The use of the reactivity coefficient values was treated on an event-specific basis in the same manner currently applied in the DCPP analyses of the non-LOCA events.

For minimum reactivity feedback, consistent with the maximum upper limit in the current DCPP Technical Specifications (Reference 18), a maximum MTC of +5 pcm/°F was modeled at power levels less than 70%. At 70% power to 100% power, the maximum MTC is ramped linearly to O pcm/°F. For maximum reactivity feedback, a maximum moderator density coefficient (MDC) of 0.43 tik/g/cc was modeled.

The maximum and minimum integrated DPCs modeled in the safety analyses are the same as those currently utilized in the DCPP non-LOCA safety analyses.

3.6.1.4 RCCA Insertion Parameters The negative reactivity insertion following a reactor trip is a function of the acceleration of the RCCAs and the variation in rod worth as a function of rod position. With respect to the non-LOCA safety analyses, the critical parameter is the time from the start of RCCA insertion to when the RCCAs reach the dashpot region. For the safety analyses, the RCCA insertion time from fully withdrawn to dash pot entry was modeled as 2.7 seconds consistent with the current DCPP Technical Specifications (Reference 18).

The applied negative reactivity insertion following reactor trip accounts for the most reactive RCCA being stuck in the fully withdrawn position and is confirmed on a cycle-specific basis as part of the reload safety evaluation process (Reference 9).

3.6.1.5 RTS and ESFAS Functions The DCPP Technical Specifications (Reference 18) define the available reactor trip system (RTS) and engineered safety feature actuation system (ESFAS) functions. The non-LOCA safety analyses support the RTS and ESFAS functions as defined in the current TSs. The difference between a limiting trip setpoint modeled in a safety analysis and the nominal trip setpoint represents allowance for instrumentation channel error and setpoint error.

During startup tests, it is demonstrated that actual instrument errors and response time delays are equal to or less than the applied values.

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3.6.1.6 Other Major Analysis Inputs PG&E Letter DCL-25-087 Other major analysis inputs considered in the non-LOCA safety analyses are discussed as follows.

Three pressurizer safety valves (PSVs) were modeled with opening setpoints based on a nominal lift setting of 2485 psig. Setpoint tolerances of +/-3.0% were conservatively applied in the modeling of the PSVs. All PSVs installed have been converted to a steam seat design and condensate in the loop is now continuously drained back to the pressurizer, thereby eliminating the water loop seal.

Therefore, the set pressure shift and purge delay previously used to account for the presence of water-filled PSV loop seals is no longer modeled.

Five main steam safety valves (MSSVs) per loop were modeled with opening setpoints based on the following nominal lift settings: 1065 psig, 1078 psig, 1090 psig, 1103 psig, and 1115 psig. Each MSSV was modeled with a +3% setpoint tolerance applied to the nominal lift setting and a 5 psi ramp from the initial opening pressure to the full-open pressure to account for accumulation. The opening pressure also conservatively accounts for the pressure drop (6P) from the MSSV header to the MSSV inlet, with 6P values applied that correspond to the 6P at the full accumulated opening pressure of each MSSV. A -2% setpoint tolerance for the first group of MSSVs (i.e., those with a nominal lift setting of 1065 psig) and a -3% setpoint tolerance for all remaining MSSVs is also supported; however, because none of the non-LOCA events are limiting with minimum setpoints, the negative setpoint tolerance has not been explicitly modeled.

The fission product contribution to decay heat modeled in the non-LOCA safety analyses is consistent with the American National Standards Institute/American Nuclear Society standard ANSI/ANS-5.1-1979 for decay heat power in light water reactors (Reference 19), including two standard deviations of uncertainty.

The minimum shutdown margin at hot zero power (HZP) conditions, which accounts for the most reactive RCCA being stuck in the fully withdrawn position,

was set to the value specified in the COLR, 1.6% L1k/k. This was modeled in the HZP steam line break analysis.

3.6.1.7 Analysis Methodology With the exception of the analysis of the RCCA ejection event (UFSAR Section 15.4.6),

the NRC approved analysis methodologies used in the current DCPP non-LOCA analyses were used in the reanalyses identified in Table 3.6-1. As discussed in Section 3.8, PG&E is requesting NRC approval to utilize the 3-0 rod ejection (3DRE) methodology described in Reference 6 to perform the complex modeling of RCCA ejection events.

Note that the core T/H DNB analyses performed in support of the non-LOCA safety analyses were performed using the WTDP methodology (see Section 3.3), and inputs to the non-LOCA safety analyses were based on the use of the PAD5 fuel performance analysis methods (see Section 3.4).

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3.6.1.8 Computer Codes Utilized PG&E Letter DCL-25-087 The principal computer codes used in the non-LOCA reanalyses are the same as those currently used, except as described below:

The RETRAN-02W computer code (Reference 2) is the Westinghouse version of the RETRAN-02 program. RETRAN-02W was previously used for some event analyses described in UFSAR Sections 15.2 and 15.4, and it continues to be used in the reanalyses of those events. RETRAN-02W was also used to replace the LOFTRAN and PG&E RETRAN-02 computer codes in the reanalyses of other non-LOCA events. See Enclosure 2 for dispositioning how the Limitations and Conditions specified in the Safety Evaluation Report section of the RETRAN-02W topical report are met.

The VIPRE-W computer code (Reference 3) was used in place of the THINC code.

Additionally, VIPRE-W was used in place of the FACTRAN code in the analyses of the Complete Loss of Forced Reactor Coolant Flow (UFSAR Section 15.3.4) and Single Reactor Coolant Pump Locked Rotor (UFSAR Section 15.4.4) events. The use of the VIPRE-W code is discussed further in Section 3.3.

Table 3.6-1 identifies all of the changes to the computer codes utilized in the non-LOCA analyses.

3.6.1.9 Classification of Events Since 1970, the American Nuclear Society (ANS) classification of plant conditions has been used to divide plant conditions into four categories in accordance with anticipated frequency of occurrence and potential radiological consequences to the public. The four categories, Conditions I to IV, are defined in UFSAR Sections 15.1 to 15.4, and are referenced as applicable for the analyses in the subsections listed below.

3.6.2 Events Evaluated or Analyzed Each of the non-LOCA safety analyses presented in Chapter 15 of the DCPP UFSAR were evaluated relative to potential impacts due to the change to the Westinghouse 17x17 OFA fuel design with PRIME fuel features and ADOPT fuel pellets. Additionally, these non-LOCA safety analyses were also evaluated relative to potential impacts due to the use of the WTDP methodology for core T/H DNB analysis and the use of the PAD5 fuel performance analysis methods as the basis for analysis inputs. A subset of the non-LOCA events was determined by evaluation to remain bounded by the safety analyses currently presented in UFSAR Chapter 15, while others required reanalysis to incorporate changes associated with the fuel upgrades and computer code and methodology updates. The UFSAR Chapter 15 non-LOCA events were evaluated or reanalyzed as shown in Table 3.6-1. These evaluations and analyses demonstrate that all applicable safety analysis acceptance criteria are satisfied for DCPP.

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PG&E Letter DCL-25-087 3.6.3 Demonstration Chapter 15 Event Analyses The intent of this section is to demonstrate the application of the fuel upgrades and computer code and methodology updates relative to the DCPP non-LOCA safety analyses. Specifically, this section is intended to demonstrate for a DCPP UFSAR Chapter 15 non-LOCA event, the concurrent application of the RETRAN-02W and VIPRE-W computer codes (References 2 and 3) with the WTDP statistical DNB methodology (Reference 5) employed and analysis inputs based on the PAD5 fuel performance analysis methods (Reference 4) and the Westinghouse 17x17 OFA fuel design with PRIME advanced fuel features and ADOPT fuel pellets. To that end, a demonstration safety analysis for the Single Reactor Coolant Pump Locked Rotor event (UFSAR Section 15.4.4) is described in the following sections. The demonstration analysis was performed using a representative core design. Plant specific values for the uncertainties and biases important to this event, as defined for the demonstration analysis, are available for audit.

3.6.3.1 3.6.3.1.1 Single Reactor Coolant Pump Locked Rotor (UFSAR Section 15.4.4)

Introduction The consequences of a reactor coolant pump (RCP) rotor seizure are very similar to those of an RCP shaft break, with both scenarios producing a sudden reduction in reactor coolant flow in the faulted loop and through the core, leading to a reactor trip on a low reactor coolant loop flow signal. Although the flow reduction is slightly faster for a rotor seizure, a broken shaft could result in the RCP impeller being free to spin in the reverse direction, which would ultimately result in lower flow through the core. A bounding analysis that reflects the most limiting aspects of both scenarios is performed, with the faulted loop modeled as instantaneously preventing impeller rotation in the forward direction, while allowing for the impeller to spin in the reverse direction. For convenience, the simulated locked rotor/shaft break event is simply referred to as the locked rotor event.

Following initiation of the low reactor coolant loop flow reactor trip, heat stored in the fuel rods continues to be transferred to the coolant, causing the coolant to expand. At the same time, heat transfer to the shell side of the steam generators is reduced, first because the reduced flow results in a decreased tube side film coefficient, and then because the reactor coolant in the tubes cools down while the shell side temperature increases (turbine steam flow is conservatively reduced to zero upon plant trip due to turbine trip on reactor trip). The rapid expansion of the coolant in the reactor core, combined with reduced heat transfer in the steam generators, causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge into the pressurizer compresses the steam volume, actuates the automatic pressurizer spray system, opens the pressurizer power-operated relief valves (PORVs), and opens the PSVs, in that sequence. Although the pressurizer sprays and PORVs are designed for reliable operation and are expected to function properly during the event, their pressure reducing effects are not credited in the analysis.

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3.6.3.1.2 Input Parameters and Assumptions PG&E Letter DCL-25-087 Consistent with the current licensing basis analysis, two locked rotor cases are analyzed, one designed to maximize the calculated percentage of rods-in-DNB, and one designed to maximize the calculated RCS pressure, fuel cladding temperature, and amount of zirconium-water reaction.

For the rods-in-DNB case, which is used to determine the percentage of fuel failure, if any, to be considered in the locked rotor radiological analysis, the following applies:

The WTDP methodology (Reference 5) is employed with nominal full power values used for the initial conditions of core power, reactor vessel average temperature, pressurizer pressure, and reactor coolant flow (MMF).

Uncertainties in these initial conditions are statistically accounted for in defining the DNBR limit value, per the WTDP methodology.

All other initial conditions (pressurizer water level, steam generator water level, etc.) are set to nominal conditions.

No credit is taken in the DNBR analysis for the increase in RCS pressure, and therefore the pressurizer PORVs and the automatic pressurizer spray system are not modeled.

For the peak RCS pressure/peak cladding temperature (PCT) case, the following applies:

The initial conditions of core power, reactor vessel average temperature, pressurizer pressure, pressurizer water level, and reactor coolant flow include conservative uncertainty allowances.

o +2% uncertainty for the initial core power.

o +5°F combined uncertainty and bias for the initial reactor vessel average temperature.

o +60 psi combined uncertainty and bias for the initial pressurizer pressure.

o +5.7% span uncertainty for the initial pressurizer water level.

o The RCS flow rate is set to the TDF.

The pressure-reducing effects of the pressurizer PORVs and sprays are not credited.

Also consistent with the current licensing basis analysis, no engineered safety systems (e.g., safety injection) are required to function in either case, and no single active failure in any system or component required for mitigation will adversely affect the consequences of this event.

3.6.3.1.3 Descriptions of Analyses and Evaluations Two computer codes are used to analyze the locked rotor cases. First, the RETRAN-02W computer code (Reference 2) is used to calculate the loop and core flows during the transient, the time of reactor trip based on the calculated loop flow of the faulted loop, the nuclear power transient, and the primary system pressure and temperature transients. The VIPRE-W computer code (Reference 3) is then used to calculate the heat flux and DNBR transients based on the nuclear power, RCS temperature 21

PG&E Letter DCL-25-087 (enthalpy), and flow from RETRAN-02W, and assuming the core pressure remains at the initial condition. The calculation of the percentage of rods-in-DNB represents the maximum of the typical or thimble cell for the fuel. The VIPRE-W code is also used in hot rod calculations to determine the fuel cladding temperature and amount of zirconium-water reaction versus time.

Consistent with the current DCPP licensing basis analysis, the locked rotor event is simulated by modeling an immediate halt in the rotational speed of one RCP. The low reactor coolant loop flow reactor trip safety analysis limit setpoint (85% of nominal loop flow) is reached less than 0.1 second into the transient, with rod motion occurring after a trip delay time of 1.0 second. Loss of offsite power is conservatively assumed to occur at the time of reactor trip (control rod release), causing the unaffected RCPs to lose power and coast down freely.

Rods-ln-DNB Case The locked rotor rods-in-DNB case is analyzed at beginning-of-cycle (BOC) full power conditions, with minimum moderator temperature reactivity feedback, maximum Doppler reactivity feedback, and a maximum value for the delayed neutron fraction. The control rods are initially assumed to be at their fully withdrawn position, and a conservatively low trip reactivity value of 4.0% llp is used to minimize the effect of rod insertion following reactor trip.

A maximum, uniform, 10% steam generator tube plugging (SGTP) level is modeled, but since the locked rotor event is an asymmetric flow coastdown scenario, it can be impacted by the effects of loop-to-loop reactor coolant flow asymmetry, e.g., due to SGTP imbalance. Therefore, a 5% loop-to-loop reactor coolant flow asymmetry, consistent with a 10% SGTP imbalance, is conservatively accounted for by applying a 0.85% core flow penalty.

As noted previously, the WTDP methodology is used for this case. Therefore, nominal initial conditions are used and uncertainties in the initial conditions are statistically accounted for in defining the DNBR limit value used in the rods-in-DNB calculation.

Peak RCS Pressure/PCT Case The locked rotor peak RCS pressure/ PCT case is analyzed at BOC full power conditions, with minimum moderator temperature reactivity feedback, maximum Doppler reactivity feedback, and a maximum value for the delayed neutron fraction. The control rods are initially assumed to be at their fully withdrawn position, and a conservatively low trip reactivity value of 4.0% llp is used to minimize the effect of rod insertion following reactor trip.

A maximum, uniform, 10% SGTP level is modeled, but since the locked rotor event is an asymmetric flow coastdown scenario, it can be impacted by the effects of loop-to-loop reactor coolant flow asymmetry, e.g., due to SGTP imbalance. Therefore, a 5% loop-to-loop reactor coolant flow asymmetry, consistent with a 10% SGTP imbalance, is conservatively accounted for in the PCT and zirconium-water reaction 22

PG&E Letter DCL-25-087 calculations by applying a 0.85% core flow penalty. However, a core flow penalty cannot be applied in addressing the effect of loop flow asymmetry on the calculated peak RCS pressure. Therefore, a separate peak RCS pressure sensitivity case is analyzed with a 5% loop-to-loop reactor coolant flow asymmetry, consistent with a 10%

SGTP imbalance, explicitly modeled. The calculated peak RCS pressure of the sensitivity case is slightly more than that of the symmetric flow case, and therefore the asymmetric flow case is bounding relative to the peak RCS pressure.

As noted previously, no credit is taken for the pressure-limiting effects of the pressurizer PORVs or sprays. Although these functions would be expected to operate normally and would result in a lower peak pressure, an additional degree of conservatism is provided by ignoring their effects.

The lift pressure of the PSVs is modeled at 3% above the nominal set pressure of 2485 psig. All PSVs installed in the Diablo Canyon units have been converted to a steam seat design and condensate in the loop is now continuously drained back to the pressurizer, thereby eliminating the water loop seal. Therefore, the set pressure shift and purge delay previously used to account for the presence of water-filled PSV loop seals is no longer modeled.

A separate VIPRE-W hot rod calculation of the fuel cladding temperature is performed assuming that the hot rod is experiencing DNB throughout the flow transient. The rod power at the hot spot is assumed to be 2.58 times the average rod power at the initial core power level, i.e., Fa = 2.58. With DNB assumed, a film boiling coefficient based on the Bishop-Sandberg-Tong film boiling correlation is calculated by the VIPRE-W computer code, and the fluid properties are evaluated at the film temperature. VIPRE-W calculates the film boiling coefficient at every time step, based on the actual heat transfer conditions at the corresponding time step. The nuclear power, system pressure, bulk density, and mass flow rate as a function of time are used as program input. For this analysis, the initial values of pressure and bulk density are used throughout the transient since they are the most conservative with respect to the fuel cladding temperature response.

The magnitude and time dependence of the heat transfer coefficient between the fuel pellet and cladding (gap coefficient) has a pronounced influence on the thermal results.

The larger the value of the gap coefficient, the more heat will be transferred between the pellet and cladding. Based on investigations of the effect of the gap coefficient on the maximum fuel cladding temperature during the transient, the gap coefficient was assumed to increase from a steady-state value consistent with initial maximum fuel temperatures to approximately 10,000 Btu/hr-ft2-°F at the initiation of the transient.

Therefore, a large amount of energy stored in the fuel was released to the cladding at the initiation of the transient.

The zirconium-water (steam) reaction can become significant above a cladding temperature of 1800°F. The Baker-Just parabolic rate equation is used to define the rate of zirconium-water reaction. The effect of the zirconium-water reaction is included in the calculation of the hot spot fuel cladding temperature transient.

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3.6.3.1.4 Acceptance Criteria and Results PG&E Letter OCL-25-087 A locked rotor is classified by the ANS as a Condition IV event, a limiting fault that is not expected to take place but is postulated because its consequences would include the potential for the release of significant amounts of radioactive material. It results in a rapid reduction in forced reactor coolant loop flow that increases the reactor coolant temperature and subsequently causes the fuel cladding temperature and RCS pressure to increase. The following summarize the acceptance criteria for the locked rotor event analysis:

Fuel cladding damage, including melting, due to increased reactor coolant temperatures must be prevented.

This criterion is met by demonstrating that the maximum fuel cladding average temperature at the core hot spot remains less than 2375°F, and the zirconium-water reaction at the core hot spot is less than 16% by weight.

Pressure in the RCS and Main Steam System (MSS) must be maintained less than acceptable design limits, considering potential brittle as well as ductile failures.

The OCPP licensing basis RCS pressure acceptance criterion is that pressure must not exceed that which would cause stresses to exceed the faulted condition stress limits of the RCS components. This translates to Service Level D of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code). For ease of interpretation, the more restrictive criterion of Service Level C of the ASME Code (i.e., peak RCS pressure shall be less than that which would cause stresses to exceed the emergency condition stress limits), was used. An analysis of individual RCS components was performed in support of the Anticipated Transient Without Scram (ATWS) submittal in 1979 to establish the pressure at which emergency condition stress limits (Service Level C) would be reached. ASME Code stress limits do not directly translate into the same percentage of the RCS design pressure for each RCS component. The limiting (minimum) equivalent pressure calculated in the analysis performed in support of the ATWS submittal was 3200 psig (3214.7 psia). Thus, an RCS pressure limit of 3214.7 psia is used.

With respect to the maximum MSS pressure, this event is bounded by the loss of load/turbine trip (LOL/TT) event presented in UFSAR Section 15.2.7 that is analyzed in a manner that conservatively maximizes the calculated MSS pressure.

For a locked rotor event, the turbine trip occurs following reactor trip, but for the LOL/TT event, the turbine trip is the initiating fault. Therefore, the primary to secondary system power mismatch and resultant MSS heatup and pressurization transients are always more severe for the LOL/TT event. For this reason, it is not necessary to explicitly evaluate the maximum MSS pressure for the locked rotor event.

The total percentage of rods-in-DNB is bounded by that used in the radiological dose analysis and a radiological dose analysis rods-in-DNB limit of 10% is used.

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PG&E Letter DCL-25-087 Rods-ln-DNB Case Results The results of the locked rotor rods-in-DNB case analysis show that no rods in DNB are predicted to occur, and therefore the percentage of rods in DNB is below the limit. The low reactor coolant loop flow reactor trip function provides timely mitigation for the event such that the acceptance criteria are satisfied. The effects of loop-to-loop flow asymmetry due to 10% SGTP imbalance are considered in the analysis. Furthermore, the results and conclusions of this analysis will be confirmed on a cycle-specific basis as part of the normal reload safety evaluation process (Reference 9).

Peak RCS Pressure/PCT Case Results The results of the locked rotor peak RCS pressure / PCT case are presented in Table 3.6-2. The results demonstrate that the applicable fuel cladding damage and RCS pressure acceptance criteria are met. The corresponding time sequence of events is presented in Table 3.6-3, and the transient results are presented in Figures 3.6-1 through 3.6-3.

The locked rotor analysis demonstrates that the peak fuel cladding average temperature at the core hot spot during a locked rotor event remains considerably less than the limit value of 2375°F, and the amount of zirconium-water reaction is small. Under such conditions, the core would remain in place and intact with no loss of core cooling capability.

The analysis also confirms that the peak RCS pressure reached during the transient is less than that which would cause stresses to exceed the faulted condition stress limits of the RCS components, and thereby, the integrity of the primary coolant system is maintained. As stated earlier in this subsection, with respect to MSS pressure, the locked rotor event is bounded by the LOL/TT event.

The low reactor coolant loop flow reactor trip function provides timely mitigation for the event such that the acceptance criteria are satisfied. The effects of loop-to-loop flow asymmetry due to 10% SGTP imbalance are considered in the analysis. Furthermore, the results and conclusions of this analysis will be confirmed on a cycle-specific basis as part of the normal reload safety evaluation process (Reference 9).

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