ML24128A277
| ML24128A277 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/04/2024 |
| From: | Richard Guzman NRC/NRR/DORL/LPL1 |
| To: | Carr E Dominion Energy Nuclear Connecticut |
| Shared Package | |
| ML24128A278 | List: |
| References | |
| EPID L-2023-LLA-0074 | |
| Download: ML24128A277 (1) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION
June 4, 2024
Eric S. Carr President - Nuclear Operations and Chief Nuclear Officer Dominion Energy Nuclear Connecticut, Inc.
Millstone Power Station Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
SUBJECT:
MILLSTONE POWER STATION, UNIT NO. 3 - ISSUANCE OF AMENDMENT NO. 290 TO REVISE TECHNICAL SPECIFICATIONS FOR REACTOR CORE SAFETY LIMITS, FUEL ASSEMBLIES, AND CORE OPERATING LIMITS REPORT FOR USE OF FRAMATOME GAIA FUEL (EPID L-2023-LLA-0074)
Dear Eric Carr:
The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 290 to Renewed Facility Operating License No. NPF-49 for the Millstone Power Station, Unit No. 3 (Millstone, Unit 3). This amendment is in response to your application dated May 23, 2023, as supplemented by letter dated December 21, 2023.
The amendment revises the Millstone, Unit 3, technical specifications (TSs) to support the use of Framatome GAIA fuel with M5 TM fuel cladding material, which is currently scheduled for insertion into the Millstone, Unit 3, reactor during the spring 2025 refueling outage. Specifically, the TS changes include updating the reactor core safety limits (TS 2.1.1.2), fuel assembly design features (TS 5.3.1), and the list of approved methodologies for the Core Operating Limits Report (TS 6.9.1.6.b).
The NRC staff has determined that the related safety evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations section 2.390, Public inspections, exemptions, request for withholding. The proprietary information is indicated by bold text enclosed with ((double brackets)). The proprietary version of the SE is provided as enclosure 2. Accordingly, the NRC staff has also prepared a non-proprietary version of the SE, which is provided as enclosure 3.
Enclosure 2 to this letter contains proprietary information. When separated from Enclosure 2, this document is DECONTROLLED.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
E. Carr A copy of the related SE is also enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Docket No. 50-423
Enclosures:
- 1. Amendment No. 290 to NPF-49
- 2. Safety Evaluation (Proprietary)
- 3. Safety Evaluation (Non-Proprietary)
cc: Listserv without enclosure 2
OFFICIAL USE ONLY - PROPRIETARY INFORMATION DOMINION ENERGY NUCLEAR CONNECTICUT, INC., ET AL
DOCKET NO. 50-423
MILLSTONE POWER STATION, UNIT NO. 3
AMENDMENT TO RENEWED FACI LITY OPERATING LICENSE
Amendment No. 290 Renewed License No. NPF-49
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Dominion Energy Nuclear Connecticut, Inc.
(DENC, the licensee), dated May 23, 2023, as supplemented by letter dated December 21, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I;
B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;
C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations;
D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and
E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations, and all applicable requirements have been satisfied.
Enclosure 1
- 2. Accordingly, paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-49 is hereby amended to read as follows:
(2) Technical Specifications
The Technical Specifications contained in Appendix A, as revised through Amendment No. 290 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DENC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3. This license amendment is effective as of the date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Hipólito González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-49 and the Technical Specifications
Date of Issuance: June 4, 2024
ATTACHMENT TO LICENSE AMENDMENT NO. 290
MILLSTONE POWER STATION, UNIT NO. 3
RENEWED FACILITY OPERATING LICENSE NO. NPF-49
DOCKET NO. 50-423
Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Insert 4 4
Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 2-1 2-1 5-5 5-5 6-21 6-21 6-21a 6-21a 6-21b 6-21b 6-21c 6-21c
(2) Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 290 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DENC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3) DENC shall not take any action that would cause Dominion Energy, Inc.
or its parent companies to void, cancel, or diminish DENC's Commitment to have sufficient funds available to fund an extended plant shutdown as represented in the application for approval of the transfer of the licenses for MPS Unit No. 3.
(4) Immediately after the transfer of interests in MPS Unit No. 3 to DNC*, the amount in the decommissioning trust fund for MPS Unit No. 3 must, with respect to the interest in MPS Unit No. 3, that DNC* would then hold, be at a level no less than the formula amount under 10 CFR 50.75.
(5) The decommissioning trust agreement for MPS Unit No. 3 at the time the transfer of the unit to DNC* is effected and thereafter is subject to the following:
(a) The decommissioning trust agreement must be in a form acceptable to the NRC.
(b) With respect to the decommissioning trust fund, investments in the securities or other obligations of Dominion Energy, Inc. or its affiliates or subsidiaries, successors, or assigns are prohibited.
Except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.
(c) The decommissioning trust agreement for MPS Unit No. 3 must provide that no disbursements or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee until the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC.
(d) The decommissioning trust agreement must provide that the agreement cannot be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.
- On May 12, 2017, the name Dominion Nuclear Connecticut, In c. changed to Dominion Energy Nuclear Connecticut, Inc.
Renewed License No. NPF-49 Amendment No. 270-287, 290
DESIGN FEATURES
5.3 REACTOR CORE
FUEL ASSEMBLIES
5.3.1 The core shall contain 193 fuel assemblies. Each fuel assembly shall consist of 264 fuel rods (with zircaloy-4, ZIRLO, Optimized ZIRLO', or M5' cladding) with an initial composition of natural uranium dioxide or a maximum nominal enrichment of 5.0 weight percent U-235 as fuel material. Limited substitutions of zircaloy-4, ZIRLO, M5', or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assembly configurations shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by test or cycle-specific reload analyses to comply with all fuel safety design bases. Each fuel rod shall have a nominal active fuel length of 144 inches. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
CONTROL ROD ASSEMBLIES
5.3.2 The core shall contain 61 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 95.3% hafnium and 4.5% natural zirconium or 80% silver, 15% indium, and 5% cadmium. All control rods shall be clad with stainless steel.
5.4 DELETED
5.5 DELETED
MILLSTONE - UNIT 3 5-5 Amendment No. 12, 37, 81, 212, 253, 290 ADMINISTRATIVE CONTROLS
CORE OPERATING LIMITS REPORT (Cont.)
- 23. DOM-NAF-2-P-A, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code, including Appendix C, Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code, and Appendix D, Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code. Methodology for Specifications:
- 3.2.3.1 RCS Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor
- 3.2.5 DNB Parameters
- 24. EMF-2328-P-A, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, (Framatome Propietary). Methodology for Specification:
- 3.2.2.1 Heat Flux Hot Channel Factor
- 25. EMF-2103-P-A, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, (Framatome Proprietary). Methodology for Specification:
- 3.2.2.1 Heat Flux Hot Channel Factor
- 26. ANP-10349-P-A, GALILEO Implementation in LOCA Methods, (Framatome Proprietary). Methodology for Specification:
- 3.2.2.1 Heat Flux Hot Channel Factor
- 27. ANP-10342-P-A, GAIA Fuel Assembly Mechanical Design, (Framatome Proprietary). Methodology for Specification:
- 3.2.2.1 Heat Flux Hot Channel Factor
6.9.1.6.c The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.
6.9.1.6.d The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
MILLSTONE - UNIT 3 6-21 Amendment No. 24, 40, 50, 69, 104, 173, 212, 215, 229, 238, 245, 249. 252, 255, 256, 279, 290
ADMINISTRATIVE CONTROLS
SPECIAL REPORTS
6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator Region I, and one copy to the NRC Resident Inspector, within the time period specified for each report.
6.10 Deleted.
6.11 RADIATION PROTECTION PROGRAM
6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.
6.12 HIGH RADIATION AREA
As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20:
6.12.1 High Radiation Areas with Dose Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation
- a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
- b. Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent; that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
- c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.
- d. Each individual or group entering such an area shall possess:
MILLSTONE - UNIT 3 6-21b Amendment No. 245, 279, 290 ADMINISTRATIVE CONTROLS
6.12 HIGH RADIATION AREA (cont.)
- 1. A radiation monitoring device that continuously displays radiation dose rates in the area, or
- 2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the devices dose alarm setpoint is reached, with an appropriate alarm setpoint, or
- 3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or
MILLSTONE - UNIT 3 6-21c Amendment No. 245, 279, 290 OFFICIAL USE ONLY - PROPRIETARY INFORMATION
ENCLOSURE 3
(NON-PROPRIETARY)
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
RELATED TO AMENDMENT NO. 290 TO
RENEWED FACILITY OPERATING LICENSE NO. NPF-49
DOMINION ENERGY NUCLEAR CONNECTICUT, INC., ET AL
MILLSTONE POWER STATION, UNIT NO. 3
DOCKET NO. 50-423
Proprietary information pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations has been redacted from this document.
Redacted information is identified by blank space enclosed within (( double brackets )).
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
RELATED TO AMENDMENT NO. 290 TO
RENEWED FACILITY OPERATING LICENSE NO. NPF-49
DOMINION ENERGY NUCLEAR CONNECTICUT, INC., ET AL
MILLSTONE POWER STATION, UNIT NO. 3
DOCKET NO. 50-423
1.0 INTRODUCTION
By letter dated May 23, 2023 (Reference 1), as supplemented by letter dated December 21, 2023 (Reference 2), Dominion Energy Nuclear Connecticut, Inc. (DENC, the licensee),
submitted a license amendment request (LAR) to revise the technical specifications (TS) for Millstone Power Station, Unit 3 (MPS3), to support the use of Framatome GAIA fuel with M5 TM1 fuel cladding material, which is currently scheduled for insertion into the MPS3 reactor during the spring 2025 refueling outage. Specifically, the TS changes include updating the reactor core safety limits (TS 2.1.1.2), fuel assembly design features (TS 5.3.1), and the list of approved methodologies for the Core Operating Limits Report (TS 6.9.1.6.b).
Since the Framatome GAIA fuel assemblies contain fuel rods fabricated with M5 TM cladding material (hereafter referred as M5), the licensee included a request for exemption from Title 10 of the Code of Federal Regulations (10 CFR) section 50.46, Acceptance criteria for emergency core cooling systems [ECCS] for light-water nuclear power reactors, and 10 CFR Part 50, appendix K, ECCS Evaluation Models, as a separate request for approval dated May 2, 2023 (Reference 3). The U.S. Nuclear Regulatory Commission (NRC or Commission) staff approved the request in the exemption dated May 21, 2024 (Reference 4).
The supplemental letter dated December 21, 2023, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no signific ant hazards consideration determination as published in the Federal Register (FR) on August 1, 2023 (88 FR 50186).
1 M5 is a trademark or registered trademark of Framatome or its affiliates, in the United States of America or other countries.
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2.0 REGULATORY EVALUATION
The NRC staff considered the following regulator y requirements and guidanc e during its review of the LAR.
Regulatory Requirements
The regulations under 10 CFR 50.36, "Technical specifications," provide regulatory requirements related to the content of TSs. Section 50.36(b) of 10 CFR requires that each license authorizing the operation of a facility will include TSs and that the TSs will be derived from the safety analysis. Section 50.36(c) of 10 CFR specifies the categories that are to be included in the TSs including (1) safety limits, li miting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs);
(4) design features; and (5) administrative controls. Of relevance to this review,
The regulation under 10 CFR 50.36(c)(1)(i)(A) and 10 CFR 50.36(c)(1)(ii)(A), Safety limits, limiting safety system settings and limiting control settings, states, in part, that:
[s]afety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded.
The regulation under 10 CFR 50.36(c)(4), Design features states that [d]esign features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1), (2), and (3) of this section.
The regulation under 10 CFR 50.36(c)(5), Administrative controls, states, in part, that
[a]dministrative controls are the prov isions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure the operation of the facility in a safe manner.
TSs are derived from a plants safety analysis report, which contains the plants design. As described in Section 3.1 of its updated final safety analysis report (FSAR) (Reference 5), the design of MPS3 satisfies and complies with the design criteria in Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, to 10 CFR Part 50, with exceptions as noted therein.
The staffs review of the LAR considered GDC 10 particularly relevant to the GAIA fuel mechanical design:
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
GDC 10 requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs). SAFDLs ensure fuel is not damaged (i.e., fuel rods do not fail, fuel system dimensions remain within operational tolerances, and functional capabilities are not reduced below those presumed in the safety analysis). The licensee's continued demonstration that the SAFDLs are met would therefore provide assurance that the integrity of the fuel and cladding would be maintained, thus preventing the potential for release of fission products during normal operation or AOOs in compliance with GDC 10.
Regulatory Guidance
The NRC staff relied on the following section of NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [light water reactor]
Edition in its review of this LAR. Review guidance for satisfying GDC 10 and other regulatory requirements is provided in Chapter 4 of the SRP (Reference 6).
SRP Section 4.2 (Fuel System Design) describes known fuel damage criteria and acceptable means for their evaluation. SRP section 4.2 further describes the objective of a fuel system safety review as providing assurance that (1) the fuel system is not damaged as a result of normal operation and AOOs, (2) fuel system damage is never so severe as to prevent control rod insertion when it is required, (3) the number of fuel rod failures is not underestimated for postulated accidents, and (4) coolability is always maintained.
SRP Section 4.3 (Nuclear Design) establishes review criteria for core power distribution, reactivity coefficients, and reactivity control requirements.
SRP Section 4.4 (Thermal and Hydraulic Design) provides specific thermal-hydraulic review criteria for the core and reactor coolant system.
NRC Generic Letter (GL) 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, states, in part, that it is acceptable for licensees to control reactor physics parameter limits by specifying an NRC-approved calculation methodology. Such parameter limits may be removed from the TS and placed in a cycle-specific COLR, which is required to be submitted to the NRC every operating cycle or each time it is revised. (Reference 7)
3.0 TECHNICAL EVALUATION
3.1 Description of Proposed TS Changes
As stated in the LAR,
The proposed change is relevant to the mechanical, thermal-mechanical, and thermal-hydraulic design features of the reactor core fuel assemblies. FSAR Sections 4.2 and 4.4 discuss the current fuel assembly design, fuel rod methodologies, mechanical design limits, fuel centerline temperature melt limit, and applicable cladding for the current fuel product at MPS3. The proposed
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change supports a MPS3 transition to the Framatome GAIA fuel product with M5 fuel cladding. The GAIA fuel assembly design is generically approved by the NRC in Reference [8]. Attachments 3 and 4 summarize the mechanical, thermal-mechanical, and thermal-hydraulic analyses that support the future operation of the Framatome GAIA fuel with M5 cladding at MPS3.
The proposed TS changes are needed to support the transition to Framatome GAIA fuel with M5 cladding at MPS3. DENC and Framatome have entered into an agreement for batch implementation of the GAIA fuel at MPS3. A full reload batch of GAIA fuel assemblies is planned for initial use in Cycle 24, which is currently scheduled to begin operation in the spring of 2025.
A description of the proposed changes is provided below. The proposed additions to the TSs are shown with bold text.
TS 2.1.1 Reactor Core Safety Limits
The LAR would revise TS 2.1.1.2 to clarify that the existing TS is a Westinghouse-specific safety limit for peak fuel centerline temperature. The Westinghouse-specific safety limit for peak fuel centerline temperature is unchanged. This LAR would add the Framatome-specific safety limit for peak fuel centerline temperature, based on the COPERNIC fuel rod performance code (Reference 11), to TS 2.1.1.2. The revised TS proposed by the licensee is shown below.
2.1.1.2 For Westinghouse fuel, the peak fuel centerline temperature shall be maintained less than 5080°F, decreasing by 9°F per 10,000 MWD/MTU of burnup. For Framatome fuel, the peak fuel centerline temperature shall be maintained less than 4901°F, decreasing linearly by 13.7°F per 10,000 MWD/MTU of burnup.
TS 5.3 Reactor Core Design Features
This LAR would revise the fuel assembly description to include M5 as an allowed cladding material. This change is necessary to support the use of M5 fuel cladding material with the GAIA fuel assembly.
5.3.1 The core shall contain 193 fuel assemblies. Each fuel assembly shall consist of 264 fuel rods (with zircaloy-4, ZIRLO, Optimized ZIRLOTM, or M5TM cladding) with an initial composition of natural uranium dioxide or a maximum nominal enrichment of 5.0 weight percent U-235 as fuel material.
Limited substitutions of zircaloy-4, ZIRLO, M5TM, or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assembly configurations shall be limited to those fuel designs that have been analyzed with applicable NRC staff-approved codes and methods, and shown by test or cycle-specific reload analyses to comply with all fuel safety design bases. Each fuel rod shall have a nominal active fuel length of 144 inches. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
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TS 6.9.1.6.b Core Operating Limits Report (COLR)
The LAR would revise the TS 6.9.1.6.b COLR reference list to include the Framatome-developed GAIA fuel Topical Report. TS 6.9.1.6.b requires that the cycle-specific COLR contain the complete identification for each of the TS referenced topical reports used (i.e., report number, title, revision, date, and any supplements); therefore, only the report number and title of the applicable topical reports, along with the associated TS parameter, is provided in the TS 6.9.1.6.b list.
- 10. WCAP-12610, VANTAGE+ Fuel Assembly Report, (W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)
- 19. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLOTM, (W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)
- 27. ANP-10342-P-A, GAIA Fuel Assembly Mechanical Design, (Framatome Proprietary). Methodology for Specification:
3.2.2.1 Heat Flux Hot Channel Factor
The licensee referenced ANP-10342P-A (Reference 8) as the document providing the basis for the NRC-approved SAFDLs that address the acceptance criteria defined in SRP Section 4.2. In addition, the licensee also considered SRP Sections 4.3 and 4.4, and GDC 10 criteria.
Attachments 3 (proprietary) and 4 (non-proprietary) to the LAR submittal provide descriptions of relevant mechanical, thermal-mechanical, and thermal-hydraulic analyses performed for the GAIA fuel design in support of the LAR under review. The licensee analyzed the proposed changes with respect to the current regulatory requirements, to assess whether applicable design criteria are met under normal, upset, and faulted operating conditions.
This SE documents the NRC staff's review of the licensee's proposal for revising the MPS3 TSs to accommodate mechanical design-related features of Framatome GAIA fuel with M5' fuel cladding material. The licensee submitted separate LARs to address other aspects of its planned transition to Framatome GAIA fuel. 2
The sections below provide the NRC staffs evaluation of the following areas:
Technical Specifications Changes GAIA Fuel Assembly Design GAIA Operation Experience Licensing Bases
2 Other items supporting the use of GAIA fuel at MPS3 submitted for NRC approval include the application of Framatome loss-of-coolant accident methodologies (Reference 3); design basis limits for a fission product barrier associated with MPS3-specific application of various NRC-approved methodologies and mixed-core departure from nucleate boiling (DNB) penalties for application to retained DNB margin (Reference 23).
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Design Evaluations o Mechanical Evaluations o Thermal Mechanical Evaluations o Thermal Hydraulic Evaluations
The objectives of the fuel system safety review, as described in SRP Section 4.2, are to provide assurance that (1) the fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs), (2) fuel system damage is never so severe as to prevent control rod insertion when it is required, (3) the number of fuel rod failures is not underestimated for postulated accidents, and (4) coolability is always maintained. A fuel system is "not damaged" when fuel rods do not fail, fuel system dimensions remain within operational tolerances, and functional capabilities are not reduced below those assumed in the safety analyses. Fuel rod failure means that the fuel rod leaks and that the first fission product barrier (the cladding) has been breached. Coolability, which is sometimes termed coolable geometry, means that the fuel assembly retains its rod-bundle geometrical configuration with adequate coolant channels to permit removal of residual heat even after an accident.
3.2 Technical Evaluation of the TS Changes
The LAR proposes to modify the MPS3 TSs to accommodate GAIA fuel assemblies with M5 cladding material, which is currently scheduled for insertion into the MPS3 reactor during the spring 2025 refueling outage. The LAR further references additional topical reports, including the GAIA Mechanical Design (ANP-10342P-A, Reference 8) and COPERNIC (BAW-10231P-A, Reference 11) topical reports, which provide additional descriptions of the GAIA fuel, fuel design criteria, and supporting technical basis for the proposed TS changes.
The licensees analysis of the proposed changes in TS 2.1.1 is supported by the analysis presented in the LAR and its attachments. In particular, the specific centerline melt safety limit and rate of decrease of this safety limit with burnup proposed by the licensee have been generically approved in Chapter 12 of the COPERNIC topical report (BAW-10231P-A, Reference 11). (The NRC staff evaluates the LARs use of BAW-10231P-A below in section 3.5 of this SE.) The NRC staff notes that the structure of the proposed revised safety limit would result in fuel-vendor-specific centerline melt limits applicable to each type of fuel that would be in use at MPS3 The proposed change to TS 2.1.1.2 appropriately adds the Framatome-specific safety limit for peak fuel centerline temperature based on the COPERNIC fuel rod performance code while clarifying that the existing TS is a Westinghouse-specific safety limit for peak fuel centerline temperature which is unchanged. Acco rdingly, the NRC staff finds that the proposed revision to TS 2.1.1.2 is acceptable and satisfies the requirements of 10 CFR 50.36(c)(1).
The proposed revision to TS 5.3.1 would ensure that the reactor core design features are correctly described within the MPS3 TS. The NRC staff notes that the M5 cladding material has been previously reviewed on a generic basis in M5 topical reports (BAW-10227P-A, Revision 1 and 2) and found acceptable for fuel cladding applications. (The NRC staff evaluates LARs use of BAW-10277P-A below in section 3.5 of this SE.) A number of pressurized-water reactor licensees have also been approved for similar applications using M5 cladding. The proposed change to TS 5.3.1 appropriately includes M5 as an allowed cladding material in the GAIA fuel assembly description. Accordingly, the NRC staff fi nds that the proposed revision to TS 5.3.1 is acceptable and satisfies the requirements of 10 CFR 50.36(c)(4).
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TS 6.9.1.6.b identifies the NRC-approved analytical methodologies that are used to determine the core operating limits for MPS3. The licensee stated that TS 6.9.1.6.b will retain the Westinghouse topical reports necessary to support the current MPS3 fuel product during the transition to GAIA. The NRC staff finds the proposed change to TS 6.9.1.6.b acceptable because it would include ANP-10342P-A, which is necessary for the determination of the heat flux hot channel factor for M5 clad GAIA fuel. Additionally, the NRC staff finds it acceptable that the licensee maintain the Westinghouse topical reports necessary to support co-resident Westinghouse fuel assemblies during transition fuel cycles. The licensee's proposal would satisfy guidance in NRC GL 88-16 since the proposed changes will continue to specify NRC-approved methodologies used to determine the core operating limits. The licensee noted that additional TS 6.9.1.6.b COLR references were requested for NRC approval in a separate LAR supporting the evaluation models for analyzing the loss-of-coolant accident (Reference 3).
Based upon the evaluation above, the NRC staff finds that the proposed revision to TS 6.9.1.6.b is acceptable and satisfies the requirements of 10 CFR 50.36(c)(5).
The NRC staff evaluation of the licensees ana lysis supporting the proposed changes is provided in the following sections (3.3 through 3.6).
3.3 GAIA Fuel Assembly Design
The mechanical design for the GAIA fuel assembly design is generically approved by the NRC in the GAIA Mechanical Design topical report (ANP-10342-P-A, Reference 8).
The GAIA fuel assembly design is intended for use in Westinghouse three-loop and four-loop reactors which use a 17x17 fuel rod array, with each assembly containing 264 fuel rods. The SE bound into ANP-10342-P-A and documents the NRC staff basis for approving this specific mechanical design for its intended application.
The GAIA fuel assembly design is a conglomerat e of previous Framatome, Babcock & Wilcox, and Electricité de France fuel designs with some additional new design features focused on thermal efficiency. The GAIA fuel design with M5 cladding consists of a 17x17 array with GAIA and intermediate GAIA mixing (IGM) grids, a lower high mechanical performance (HMP) grid and an upper HMP grid. The fuel assembly includes a MONOBLOC guide tube design, M5 fuel rod design and a GRIP lower nozzle. 3 The fuel is standard uranium dioxide (UO 2) fuel with 2, 4, 6, and 8 weight-percent Gadolinia (Gd 2O3) rods included.
The design uses the M5 advanced alloy, which has been previously approved (Reference 10 and Reference 13) for cladding. Q12 alloy (Reference 9) is used for guide tubes (GT) and instrument tubes (IT), which represents one of the first deployments of this recently approved alloy. Section 2.0 of Attachment 3 to the LAR (i.e., ANP-4040P, Rev. 0, Millstone Unit 3 Mechanical Design Report for GAIA Fuel Transition) describes the GAIA fuel mechanical design. The staff has examined the description and schematic diagrams in ANP-4040P, Rev. 0, which include the fuel assembly, fuel rods, grids, top nozzle, GTs and instrumentation tubing, bottom nozzle, and the materials used for each component, and determined that the description is complete and thorough. Accordingly, the staff finds that a satisfactory description of the fuel
3 MONOBLOC and GRIP, as described further in Framatome topical report ANP-10342, Rev. 0, are trademarks or registered trademarks of Framatome or its affiliates, in the Unites States of America or other countries.
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assembly has been provided to support the proposed application of the GAIA fuel assembly design.
3.4 GAIA Operational Experience
GAIA lead test assembly (LTA) and reload operating experience is provided in Section 3.0 of to the LAR. Table 3-1, GAIA Operating Experience Summary in Attachment 3 provides a summary of the quantity of fuel assemblies in operation.
((
))
The LTA program for confirming the irradiation behavior of the GAIA fuel assembly design placed LTAs in core locations with near-peak power conditions. The GAIA program was a global design effort within Framatome's three main regions of operation (United States of America (U.S.), France, and Germany), in cooperation with two customers to test the design prior to batch implementation. Four GAIA LTAs were inserted in the core of an international reactor in 2012. Eight (8) GAIA LTAs were inserted in the core of a U.S. reactor in 2015. Both plants are Westinghouse 3-loop designs. Four GAIA LTAs were inserted in the core of a Westinghouse 4-loop domestic reactor in 2019. Four GAIA LTAs were inserted in another Westinghouse 4-loop domestic reactor in 2021. Assemblies successfully completed many cycles with leaker-free performance.
Post-irradiation examinations (PIEs) were performed after every irradiation cycle to confirm that the LTAs were operating as predicted. The PIEs performed were appropriate for confirming the performance of the fuel design and the results showed the irradiated fuel continued to meet design criteria, including for material properties. Therefore, the LTA performance is acceptable to ensure sufficient high burnup fuel rods have been examined prior to a full batch of GAIA fuel assemblies having reached end-of-life (EOL).
The licensee stated that a total of three reloads each have been delivered to two Westinghouse 3-loop international reactors, starting in 2020. The initial reload has completed two 12-month cycles of leaker-free operation.
The licensee stated that a total of two reloads have been delivered to a Westinghouse 3-loop domestic reactor, starting in 2021. The initial reload has completed one 18-month cycle of leaker-free operation.
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3.5 Licensing Bases
The approved version of the GAIA Mechanical Fuel Assembly Design topical report (Reference 8), which incorporates the NRC staffs SE including applicable limitations and conditions, is the governing document for the GAIA transition at MPS3. The foundation of Reference 8 is the establishment of NRC-approved SAFDLs that address the acceptance criteria defined in the Section 4.2 of the SRP, NUREG-0800 (Reference 6), for each of the fuel damage mechanisms.
Table 1 provides a summary of the acceptance criteria from References 8 to 19, and the staff has determined these are in general alignment with the analyses that are performed to justify satisfaction of the SAFDLs to address the fuel damage mechanisms in SRP 4.2. For each acceptance criterion and associated analysis, a list of applicable topical reports is provided which the applicant collectively used to demonstrate compliance. Table 1 includes the purpose of each topical report (e.g., criteria, methods, results), and a cross reference to the mechanical, thermal-mechanical, or thermal-hydraulic TR sections where additional details associated with the criteria, methods, and analyses are located. The Table also lists the associated sections of the staff safety evaluation bound into each topical report.
Table 1: Summary of Applicable Criteria and Methods Associated with
Topical Reports Applied to GAIA Fuel
Topical Applicable Analyses Criteria Section Methods Section NRC staff Report (and other applicable (and other applicable SE Section TRs) TRs)
General Component Section 8.1.1.1; Section 8.1.1.2; 3.6.1.1 Stress (Reference 9) (Reference 9) 3.6.1.2 General Component Section 8.1.2.1 Section 8.1.2.2 3.6.1.3 Fatigue Fretting Wear Section 8.1.3.1 Section 8.1.3.2 3.6.1.4 Fuel Rod Growth Section 8.1.5.1 Section 8.1.5.2; 3.6.1.5 (Reference 9, 19)
Fuel Assembly Growth Section 8.1.5.1 Section 8.1.5.2; 3.6.1.5 (Reference 9)
Fuel Rod Internal Section 8.1.6.1; Section 12.1; 3.6.2.7 Pressure (Reference 10, 14) (Reference 11)
Fuel Assembly Lift-Off Section 8.1.7.1; Section 8.1.7.2; 3.6.1.6 (Reference 9, 15) (Reference 9, 15)
ANP-10342 Cladding Internal Section 8.2.1.1 Section 8.2.1.2 3.6.3.9 (Reference 8) Hydriding Cladding Stress (Normal Section 3.3; Section 8.1.1.2 3.6.2.1 Operation, Faulted) (Reference 10)
Cladding Fatigue Section 8.1.2.1 Section 3.6; 3.6.2.5 (Reference 10, 11)
Cladding Oxidation Section 8.1.4.1 Section 12.6; 3.6.2.6 (Reference 11)
Transient Cladding Section 8.1.1.1 Section 12.4; 3.6.2.4 Strain (Reference 11)
Overheating of Fuel Section 8.2.3.1 Section 12.3; 3.6.2.10 Pellets (Reference 11)
Overheating of Cladding Section 8.4.1 Section 8.4.1 N/A Reactivity Coefficients Section 8.4.7 Section 8.4.7 N/A
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Cladding Collapse Section 8.2.2.1; Section 8.2.2.1; 3.6.2.9 (Reference 10, 11, 16) (Reference 10, 11, 16)
Cladding Stress Section 3.3 Section 3.3 3.6.2.1 Cladding Buckling Section 3.3 Section 3.3 3.6.2.2 Cladding Fatigue Section 8.1.2.1 of Section 3.6; 3.6.2.5
Reference 8 (Reference 11)
BAW-10227 Fuel Rod Internal Section 8.1.6.1 of Section 12.1; 3.6.2.7 (Reference Pressure Reference 8; (Reference 11)
- 10) (Reference 14)
Cladding Collapse Section 8.2.2 of Section 8.2.2 of 3.6.2.9 Reference 8; Reference 8; (Reference 11, 16) (Reference 11, 16)
Cladding Structural Section 3.2, 4 and 5 Sections 3.2, 4, and 5; 3.6.2.3 Deformation (Reference 8, 17, 18)
Transient Cladding Section 8.1.1.1 of Section 12.4 3.6.2.4 Strain Reference 8 Cladding Fatigue Section 8.1.2.1 of Section 3.6 of 3.6.2.5 Reference 8 Reference 10 BAW-10231 Oxidation, Hydriding, Section 8.1.4.1 of Section 12.6 3.6.2.6 (Reference Reference 8 3.6.2.8 11)
Fuel Rod Internal Section 8.1.6.1 of Section 12.1 3.6.2.7 Pressure Reference 8 Cladding Collapse Section 8.2.2.1 Section 12.5 3.6.2.9 Reference 8 Overheating of Fuel Section 8.2.3.1 of Section 12.3 3.6.2.10 Pellets Reference 8 XN-75-32 Fuel Rod Bow Section 8.1.5.1 of Section 2.0 - 3.4 3.6.3.1 (Reference Reference 8 12)
BAW-10227 Fuel Rod Bow Section 8.1.5.1 of Section 11.4 3.6.3.1 (Reference Reference 8 13)
General Component Section 8.1.1.1 of Section 8.1.1.2 of 3.6.1.1 Stress Reference 8 Reference 8 3.6.1.2 ANP-10334 Fuel Rod and Assembly Section 10.1 and 12.1; Section 10.1 and 12.1; 3.6.1.5 (Reference 9) Growth Section 8.1.5.1 of Section 8.1.5.2 of Reference 8 Reference 8; (Reference 19)
Fuel Assembly Lift-Off Section 10.1; Section Section 10.1; Section 3.6.1.6 8.1.7.1 of Reference 8; 8.1.7.1 of Reference 8; (Reference 15) (Reference 15)
BAW-10240 Fuel Rod and Assembly Section 8.1.5.1 of Section 8.1.5.2 of 3.6.1.5 (Reference Growth Reference 8 Reference 8;
- 19) (Reference 9)
BAW-10183 Fuel Rod Internal Section 8.1.6.1 of Section 6.1; Section 3.6.2.7 (Reference Pressure Reference 8; 12.1 of Reference 11
- 14) (Reference 10)
ANP-10311 Fuel Assembly Lift-Off Section 8.1.7.1 of Section 8.1.7.1 of 3.6.1.6 (Reference Reference 8; Reference 8;
- 15) (Reference 9) (Reference 9)
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BAW-10084 Cladding Collapse Section 8.2.2 of Section 8.2.2 of 3.6.2.9 (Ref 16) Reference 8; Reference 8; (Reference 10) Reference 10)
ANP-10337 General Component Section 4 Section 7 and 8 3.6.1.7 (Reference Structural Deformation 17, 18) Cladding Structural Section 3.2, 4 and 5 of Section 3.2, 4, and 5 of 3.6.2.3 Deformation Reference 10 Reference 10; (Reference 8)
3.5.1 Justification of Applicability and Compliance with NRC Staff Imposed Limitations and Conditions
The licensees dispositions and NRC staff evaluati on for the limitations and conditions (L&Cs) identified in the NRC staff SEs for the topical reports used in the analysis are provided as follows.
3.5.1.1 GAIA Mechanical Design topical report (ANP-10342P-A)
The GAIA Mechanical Design topical report (ANP-10342-P-A, Revision 0, Reference 8) defines the GAIA fuel design and addresses the acceptance criteria established in Section 4.2 of the Standard Review Plan (NUREG-0800, Reference 6). The NRC staff approved ANP-10342 for use in all Westinghouse 3-loop and 4-loop reactors which use a 17x17 fuel rod array with low-enriched uranium fuel. MPS3 is a Westinghouse 4-loop reactor; therefore, it is applicable to the fuel transition at MPS3.
In Section 4.0 of the SE for ANP-10342 (Reference 8), the NRC staff imposed a total of four (4) L&Cs. The licensee addressed these L&Cs in Table 4-3 of Attachment 3 to the LAR. A summary of each L&C and the NRC staff's assessment of the licensees disposition thereof is provided below:
L&C 1: This GAIA fuel assembly design is approved for use with low enrichment uranium fuel, which has been enriched to less than or equal to 5 percent.
Evaluation: The NRC staff finds this L&C to be satisfied because MPS3 core designs will be limited to low enrichment uranium fuel, which has been enriched to less than or equal to 5 percent.
L&C 2: This GAIA fuel assembly design is licensed for a maximum fuel rod burnup of 62,000 megawatt-days/metric ton of uranium.
Evaluation: The licensee stated that the MPS3 GAIA fuel assembly is designed to support a peak UO2 rod burnup of 62 GWd/MTU. The NRC staff finds this L&C to be satisfied because the methods adopted by the licensee are limited to 62 GWd/MTU and the licensee does not have approval to exceed this burnup limit.
L&C 3: The final LTA program PIE report shall be submitted to NRC staff prior to any reload batch GAIA assemblies reach the third cycle of operation.
Evaluation: The NRC staff finds this L&C to be satisfied because the GAIA LTA PIE report (Reference 20) was submitted to the NRC in July 2021.
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L&C 4: As part of the plant-specific LAR implementing GAIA, the licensee must demonstrate acceptable performance of GAIA under RIA [reactivity-initiated accident]
conditions. The licensee should consider the most up-to-date guidance and analytical limits at the time of submittal.
Evaluation: The governing topical report for the MPS3 rod ejection accident (REA) is ANP-10338 (Reference 21). The NRC staff has determined t hat this L&C is satisfied because MPS3 application of the REA methodology in Reference 21 is compliant with the most up-to-date applicable guidance and analytical limits per RG 1.236, which encompasses RIA conditions. 4
3.5.1.2 M5 topical report (BAW-10227P-A Revision 1)
The M5 topical report (BAW-10227 Revision 1, Reference 10) defines the M5 criteria, methods, and material properties, and is referenced in ANP-10342 (Reference 8). Therefore, the NRC staff finds the proposed implementation of the BAW-10227P-A Revision 1 methodology in the MPS3 LAR in the manner prescribed in ANP-10342 to be consistent with the approved range of applicability for the method. There is no explicit section for L&Cs within the safety evaluation report (SER) for BAW-10227 Revision 1 (Reference 10); however, per Section 4.0 of the SER there is one (1) L&C to consider:
L&C 1: The peak fuel rod burnup is limited to 62,000 MWd/MTU.
Evaluation: The GAIA fuel assembly is designed to support a peak UO 2 rod burnup of 62 GWd/MTU. The NRC staff finds this L&C to be satisfied because the methods adopted by the licensee are limited to 62 GWd/MTU and the licensee does not have approval to exceed this burnup limit.
3.5.1.3 COPERNIC topical report (BAW-10231)
The COPERNIC topical report (BAW-10231, Reference 11) defines the fuel rod computer code used to perform thermal-mechanical analyses and is referenced in ANP-10342 (Reference 8).
Therefore, the NRC staff finds the proposed implementation of the BAW-10231 methodology in the MPS3 LAR in the manner prescribed in ANP-10342 to be consistent with the approved range of applicability for the method.
There is no explicit section for L&Cs within the SER for BAW-10231 (Reference 11); however, per Section 8.0 of the SER there is one (1) L&C to consider for the rod average burnup:
L&C 1: Fuel licensing applications up to a rod average burnup of 62 GWd/MTU.
Evaluation: The GAIA fuel assembly is designed to support a peak UO 2 rod burnup of 62 GWd/MTU. The NRC staff finds this L&C to be satisfied because the methods adopted by the licensee are limited to 62 GWd/MTU and the licensee does not have approval to exceed this burnup limit.
4 The NRC staffs review of the MPS3 compliance with RG 1.236 is discussed in a separate SE associated with the license amendment request dated October 30, 2023 (Reference 23).
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3.5.1.4 Fuel Rod Bow topical report (XN-75-32P-A)
The Fuel Rod Bow topical report (XN-75-32, Reference 12) defines the method for analyzing the effects of fuel rod bow and is referenced in the GAIA Mechanical Design topical report ANP-10342 (Reference 8). Therefore, the NRC staff finds the proposed implementation of the XN-75-32P-A methodology in the MPS3 LAR in the manner prescribed in ANP-10342 to be consistent with the approved range of applicability for the method.
There is no explicit section for L&Cs within the NRC approval letter for XN-75-32 (Reference 12). However, per the NRC approval letter, there is one (1) L&C to consider:
L&C 1: The acceptance is not applicable to fuel designs which exhibit a greater propensity for bowing than that given in data from which the models reviewed were developed.
Evaluation: As stated by the licensee in Section 4.5.1 of Attachment 3 to the LAR, the Fuel Rod Bow topical report is NRC-approved for use with the GAIA fuel design in 17x17 fuel arrays in all Westinghouse 3-loop and 4-loop reactors. As expl ained in Section 3.3.1.6 of the NRC staff's SE on ANP-10342P, the rod bow data for Zircaloy is conservative relative to M5 cladding due to the lower growth rate of M5. Therefore, the NRC staff finds that this L&C is met for the fuel transition at MPS3.
3.5.1.5 M5 topical report (BAW-10227P-A Revision 2)
The latest version of the M5 topical report (BAW-10227 Revision 2, Reference 13) provides updated information, data, and models relative to the original version (Reference 10). Specific to the GAIA fuel transition at MPS3, the gap closure model per Section 11.3 of BAW-10227 Revision 2 will replace that of Section 2.2.1 of the XN-75-32 (Reference 12) TR for the fuel rod bow analysis. Per Section 3.5.4 of Reference 13 SER, the new gap closure model is applicable to Westinghouse fuel designs that meet six key characteristics. Based on the assessments in Table 4-7 of the LAR, Reference 13 is applicable to the fuel transition at MPS3.
Per Section 4.0 of the SER for BAW-10227 Revision 2 (Reference 13), there is a total of one L&C. The limitation and condition states that when applying the methodology described in BAW10227P, Rev. 2, ((
)) licensees shall ensure that changes to expected fatigue cycles are appropriately captured in the fatigue evaluation. For the MPS3 proposed LAR, the GAIA fuel assembly is designed to support a peak UO 2 rod burnup of 62 GWd/MTU and peak gadolinia (GAD) rod burnup of 55 GWd/MTU. The NRC staff finds this L&C to be satisfied because the methods adopted by the licensee are limited to 62 GWd/MTU and the licensee does not have approval to exceed this burnup limit.
3.5.1.6 Q12 topical report (ANP-10334P-A)
The Q12 topical report (ANP-10334P-A, Reference 9) defines the Q12 material properties, specifically the GT growth model, and is referenced in ANP-10342 (Reference 8). Therefore, the NRC staff finds the proposed implementation of the ANP-10334P-A methodology in the MPS3 LAR in the manner prescribed in ANP-10342 to be consistent with the approved range of applicability for the method.
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Per Section 4.0 of the SER for ANP-10334 (Reference 9), there are a total of four L&Cs:
L&C 1: AREVA must follow the model update process as defined by the responses to request for information (RAI) 1 and 2, as described in the AREVA response. 5
Evaluation: L&C 1 is not applicable to the MPS3 transition because L&C 1 is applicable to design-specific Q12 growth models, and not the generic Q12 growth model. The generic Q12 GT growth model was used for the transition analyses.
L&C 2: AREVA must follow the surveillance program as defined by the response to RAI 3, as described in the AREVA response. 6
Evaluation: L&C 2 is not applicable to the MPS3 transition. Per the response to RAI 3 of ANP-10334 (Reference 9), the surveillance program is specific to fuel designs that go to batch loading without an LTA program. Per Section 3.0, the GAIA design has completed several LTA programs.
L&C 3: Burnup limits imposed on fuel and fuel assembly designs will also apply to those assemblies using Q12 structural material.
Evaluation: The NRC staff finds this L&C to be sa tisfied because the GAIA fuel assembly is designed to support a peak UO 2 rod burnup of 62 GWd/MTU. The methods adopted by the licensee are limited to 62 GWd/MTU, and the licensee does not have approval to exceed this burnup limit.
L&C 4: Q12 is only approved for use in PWRs as a structural material. This limits its use to guide tubes, spacer grids, and instrument tubes.
Evaluation: Q12 is only used for the GAIA guide tubes and instrument tubes. This L&C is met for the fuel transition at MPS3.
3.5.1.7 M5 Properties topical report (BAW-10240P-A)
The M5 Properties topical report (BAW-10240P-A, Reference 19) incorporates M5 material properties into several Framatome topical reports that were previously approved only for the Zircaloy-4 material. Specific to the fuel transition at MPS3, the M5 Properties topical report provides the fuel rod growth model and is referenced in ANP-10342P-A (Reference 8).
Therefore, the NRC staff finds the proposed implementation of the BAW-10240P-A methodology in the MPS3 LAR in the manner prescribed in ANP-10342 to be consistent with the approved range of applicability for the method.
Per Section 4.0 of the SER for BAW-10240P-A (Reference 19), there are a total of four L&Cs:
L&C 1: The corrosion limit, as predicted by the best-estimate model will remain below 100 microns for all locations of the fuel.
5 AREVA response to RAI Regarding ANP-10334P, Q12TM Structural Material (Reference 24) 6 AREVA response to RAI 3 Regarding ANP-10334P, Q12TM Structural Material (Reference 25)
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Evaluation: This L&C is not applicable to the MPS3 transition. For the MPS3 transition, BAW-10240 is only used to provide the M5 fuel rod growth model.
L&C 2: All of the conditions listed in the SEs for all [Framatome ANP] FANP methodologies used for M5 fuel analysis will continue to be met, except that the use of M5 cladding in addition to Zircaloy-4 cladding is now approved.
Evaluation: This L&C is not applicable to the MPS3 transition. For the MPS3 transition, BAW-10240 is only used to provide the M5 fuel rod growth model. Therefore, the SEs associated with all the other methodologies that were incorporating M5 material properties are not relevant.
L&C 3: All FANP methodologies will be used only within the range for which M5 data was acceptable and for which the verifications discussed in BAW-10240(P) or Reference 8 was performed.
Evaluation: Per the governing fuel rod growth analysis for the fuel transition at MPS3, the resulting fuel rod fluence value is bounded.
L&C 4: The burnup limit for this approval is 62 GWd/MTU.
Evaluation: GAIA fuel assembly is designed to support a peak UO 2 rod burnup of 62 GWd/MTU.
The NRC staff finds this L&C to be satisfied because the methods adopted by the licensee are limited to 62 GWd/MTU and the licensee does not have approval to exceed this burnup limit.
3.5.1.8 Fuel Rod Gas Pressure Criterion (FRGPC) topical report (BAW-10183P-A)
The Fuel Rod Gas Pressure Criterion (FRGPC) topical report (BAW-10183P-A, Reference 14) defines the allowable steady-state fuel rod gas pressure criterion if it exceeds system pressure, stating in particular that
The internal pressure of the peak fuel rod in the reactor will be limited to a value below that which would cause (1) the fuel-clad gap to increase due to outward cladding creep during steady-state operation and (2) extensive DNB propagation to occur.
BAW-10183 is referenced in Section 12.1.1 of the COPERNIC topical report (Reference 11),
which is referenced in ANP-10342. In addition, per Section 12.1.1 of the COPERNIC topical report, BAW-10183 is extended to M5 material via BAW-10227 Revision 1 (Reference 10).
Therefore, the NRC staff finds the proposed implementation of the BAW-10183P-A methodology in the MPS3 LAR in the manner prescribed in BAW-10231, BAW-10227 Revision 1, and ANP-10342 to be consistent with the approved range of applicability for the method.
There is no explicit section for L&Cs within the SER for BAW-10183 (Reference 14); however, per Section 3.0 of the SER there are two (2) L&Cs to consider:
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L&C 1: If LOCA linear heat generation rate (LHGR) become limiting at extended burnup levels for any BWFC [Babcock & Wilcox Fuel Company] design applications, the LOCA LHGR analyses should be submitted for NRC review.
Evaluation: As stated by the licensee in its LAR, Framatome submits Safety Summary Reports, which cover the LOCA events, to Dominion. Through the adoption of the BAW-10183P-A methodology, Dominion would be required to submit the relevant information to the NRC staff.
Therefore, the NRC staff finds this L&C to be satisfied.
L&C 2: If there are changes involving the power peaking maps or other input parameters for the response surface applications, the core protection analysis should be submitted for review.
Evaluation: The LAR states that if DNB is predicted for any Chapter 15 events, and Reference 14 is used as the method for DNB propagation, the core protection analysis will be submitted for review. The licensee's response to RAI 2 confirmed that there are no events for which DNB is predicted where the BAW-10183 methodology will be applied for the first operating cycle with GAIA fuel at MPS3. Therefore, the information submitted by the licensee confirms satisfaction of L&C 2.
3.5.1.9 COBRA-FLX topical report (ANP-10311P-A)
The COBRA-FLX topical report (ANP-10311, Reference 15) defines a core thermal-hydraulic computer code. Specific to this LAR focused upon mechanical design aspects of the GAIA fuel assembly, COBRA-FLX is predominantly used to generate thermal-hydraulic lift loads in support of downstream analysis, ((
)) Per Section 5.0 of the SER for ANP-10311, the code is suitable for stand-alone application to nuclear core thermal-hydraulic analyses for steady-state and transient conditions.
Therefore, the COBRA-FLX topical report is applicable to the fuel transition at MPS3.
Per Section 4.0 of the SER for ANP-10311 (Reference 15), there are a total of two L&Cs that are paraphrased below:
L&C 1: The fuel rod model in COBRA-FLX and the rewetting model for post-CHF heat transfer will not be used for safety-related analysis and are explicitly excluded from this review. L&C 1 further contains additional detail and specific limitations concerning which empirical correlations may be used in licensing calculations.
Evaluation: As stated by the licensee in its LAR, the use of the COBRA-FLX internal fuel rod model is excluded. The use of the rewetting model is also excluded and the applicable limitations under L&C 1 are addressed. Note that per Reference 15, in a letter dated October 18, 2017, the NRC approved the use of the standard friction factor correlation as part of its response to an Errata; therefore, the standard friction factor correlation is acceptable for use.
Based on the information provided in the LAR, the licensee has proposed to use the COBRA-FLX methodology consistent with L&C 1, and the staff finds this L&C to have been acceptably addressed.
L&C 2: This review examined only the specific models and correlations requested by the applicant, as summarized in Section 2.0 of the SE for ANP-10311 (Reference 15). These
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are the only models and correlations that may be used in licensing calculations with the COBRA-FLX subchannel code. The fuel rod model in COBRA-FLX and the rewetting model for post-CHF heat transfer shall not be used for safety related analysis, and are specifically excluded from this review.
Evaluation: The licensee stated that this L&C is covered by L&C 1. Specifically, the licensee has proposed to use only the models in COBRA-FLX that have been explicitly approved for licensing calculations. The models that have not been approved for use will not be applied. Therefore, the NRC staff finds that L&C 2 has been acceptably addressed.
3.5.1.10 CROV topical report (BAW-10084P-A)
The CROV topical report (BAW-10084P-A, Reference 16) defines the fuel rod creep criteria and methodology for thermal-mechanical analyses, and is referenced in ANP-10342 (Reference 8).
In addition, CROV is extended to M5 material via BAW-10227 Revision 1 (Reference 10). As described in the SER on ANP-10342, the NRC staff approved the use of BAW-10084P-A as applied in ANP-10342. Therefore, the NRC staff finds the proposed implementation of the BAW-10084P-A methodology in the MPS3 LAR in the manner prescribed in ANP-10342 to be consistent with the approved range of applicability for the method.
There is no explicit section for L&Cs within the SER for BAW-10084, Reference 16; however, per Section 5.0 of the SER there is one (1) L&C to consider. The limitation and condition states that the cladding temperature limit shall be less than or equal to 700 °F (for the strain rate equation to be applicable). The MPS3 creep collapse analysis maintained cladding temperatures that were compliant with the licen sing limit. Therefore, the NRC staff finds that all L&Cs are met for the fuel transition at MPS3.
3.5.1.11 PWR Fuel Assembly Structural Response to Externally Applied Dynamic Excitations topical report (ANP-10337P-A)
Topical report ANP-10337P-A, Reference 17, defines the criteria and method for analyzing the fuel assembly structural response to externally applied dynamic excitations (i.e. seismic and LOCA), and is referenced in ANP-10342 (Reference 8). In the NRC SER on ANP-10342, the staff approved the use of ANP-10337P-A as applied in ANP-10342. Therefore, the NRC staff finds the proposed implementation of the ANP-10337P-A methodology in the MPS3 LAR in the manner prescribed in ANP-10342 to be consistent with the approved range of applicability for the method.
Per Section 5.0 of the SER for ANP-10337 (Reference 17), there are a total of nine L&Cs. The L&Cs are as follows, together with the NRC staff evaluation of the information in the LAR, including Table 4-14, to determine if the L&Cs are satisfied:
L&C 1: Dynamic grid crush tests, must be conducted in accordance with Section 6.1.2.1 of ANP-10337P (as amended by RAI 16), and spacer grid behavior must satisfy the requirements in the TR, the key elements of which are:
((
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))
Evaluation: The licensee stated that item a (or part a) is met for the IGM. The licensee further stated that the IGM impact load limit corresponds to an ultimate, or buckling limit; therefore, items b and c (or parts b and c) do not apply. In response to RAI 3, the licensee provided further elaboration to demonstrate that ((
)) Per Section 4.13.2 of Attachment 3 to the LAR, the licensee stated that this L&C is not applicable for the GAIA grid. Considering that the licensee has chosen to apply the deformable grid element model of ANP-10337P, Supplement 1, in lieu of the viscoelastic model in the base methodology, the NRC staff finds this position acceptable.
L&C 2: For fuel assembly designs where spacer grid applied loads are limited based on allowable grid permanent deformation (as opposed to buckling), the following limits from Table 4-1 of the TR (ANP-10337P) apply:
- a. For all OBE analyses, allowable spacer grid deformation is limited to design tolerances and (( ))
))
Evaluation: The licensee stated that the IGM load limit is based on buckling and not an allowable grid permanent deformation; therefore, this L&C is not applicable. The licensee stated that allowable GAIA grid deformation limits for an Operating Basis Earthquake (OBE) have been defined in accordance with item a. The licensee further stated that, per Section 4.13.2 of to the LAR, item b is not applicable for the GAIA grid. Considering the licensee's statement that the IGM load limit is based on buckling and not permanent grid deformation, the NRC staff agrees that L&C 2 is not applicable to the IGM. Regarding the GAIA grid, the licensee has stated its conformance to L&C 2a. Regardi ng L&C 2b, Section 3.2 of the NRC staff's SE on ANP-10337P-A, Supplement 1, states that The numerical values of allowable grid permanent deformation remain unchanged from the base methodology and further cites the 1-mm and 3-mm limits discussed above in L&C 2b. Therefore, the NRC staff considers the purpose of L&C 2b to be satisfied for the GAIA grid also, consistent with the acceptance criteria defined in ANP-10337P-A, Supplement 1, and its SE.
L&C 3: The modification or use of the codes CASAC and ANSYS (or other similar industry standard codes) are subject to the following limitations:
- a. CASAC computer code revisions, necessita ted by errors discovered in the source code, needed to return the algorithms to those described in ANP-10337P (as updated by
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RAls) are acceptable.
- b. Changes to CASAC numerical methods to improve code convergence or speed of convergence, transfer of the code to a different computing platform to facilitate utilization, addition of features that support effective code input/output, and changes to details below the level described in ANP-10337P would not be considered to constitute a departure from a method of evaluation in the safety analysis. Such changes may be used in licensing calculations without NRC st aff review and approval. However, all code changes must be documented in an auditable manner to meet the quality assurance requirements of 10 CFR part 50, appendix B.
- c. ANSYS or other industry standard codes may be used if they are documented in an auditable manner to meet the quality assurance requirements of 10 CFR part 50, appendix B, including the appropriate verification and validation for the intended application of the code.
Evaluation: The licensee stated that the CASAC c ode version used is fully consistent with the requirements. The licensee further stated that the ANSYS code version conforms to Framatomes quality assurance procedures. Bec ause the licensee will be responsible for assuring that purchased services, including use of these codes, conform to procurement documentation (Part 50, Appendix B, Criterion VII), Framatomes application of its QA procedures to the use of these codes will assure the quality called for in L&C 3, and the NRC staff considers this L&C satisfied.
L&C 4: This methodology is limited to applications that are similar to the current operating fleet of PWR reactor and fuel designs. The core geometry should be comparable to the current fleet, in terms of dimensions, dimension tolerances, fuel assembly row lengths, and the gaps between fuel assemblies. Fuel designs should be comparable to the current fleet, in terms of materials, geometry, and dynamic behavior.
Evaluation: The licensee stated that the MPS3 reactor is part of the current fleet of PWR reactors in place at the time of approval of ANP-10337. The NRC staff agrees that this L&C has been acceptably addressed by the licensee because the design of MPS3 and the GAIA fuel design is comparable with the current fleet in terms of the characteristics described in L&C 4.
L&C 5: ANP-10337P established generic fixed damping values intended to be used for all PWR designs. All applications of this methodology to new fuel assembly designs must consider the continued applicability of the fixed damping values of this methodology. If new materials, new geometry, or new design features of a new fuel assembly design may affect damping, additional testing and/or evaluation to determine appropriate damping values may be required.
Evaluation: The licensee stated that damping values per ANP-10337 are applicable to the GAIA fuel design per Section 4.12.1 of Attachment 3 to the LAR. The licensee explained therein that the GAIA fuel assembly mechanical design topical report, ANP-10342, references ANP-10337 in defining the fuel assembly response to externally applied dynamic excitations (i.e., seismic and LOCA loads). The NRC staff finds that the licensee's adoption of the damping values from ANP-10337 for GAIA fuel is supported by the referencing of the ANP-10337 methodology within
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ANP-10342, which has been generically reviewed and approved by the NRC staff. Therefore, the NRC staff considers this L&C to be satisfied.
L&C 6: The ANP-10337P methodology includes the generation of fuel rod loads, but does not provide a means to demonstrate compliance for fuel rod performance under externally applied loads (to applicable acceptance criteria). Applications of this methodology must provide an acceptable demonstration of fuel rod performance.
Evaluation: The licensee stated that GAIA fuel rod performance under faulted conditions was demonstrated per Section 5.2.3 of Attachment 3 to the LAR. The NRC staff has reviewed the licensees description of the relevant analysis below in Section 3.6.2.3 of this SE. Based on the staffs review documented below, the NRC staff finds that this L&C has been satisfied.
L&C 7: As indicated in ANP-10337P when orthogonal deflections from separate core locations are artificially superimposed to calculate component stresses, the component stresses must be compared against the design criteria associated with control rod positions.
Evaluation: The licensee stated that the more limiting component stress allowables associated with rodded locations were used. The NRC staff finds that the analysis performed for the MPS3 GAIA fuel fully considers the actual core location and appropriately considers the guide tube criteria for control rod positions per the requirements of ANP-10337P-A and is therefore acceptable.
L&C 8: In accordance with Regulatory Guide (RG) 1.92, the combination of loads for non-grid component evaluation should ideally be based on three orthogonal components (two horizontal and one vertical). ((
)).
Evaluation: The licensee stated that the analysis performed for MPS3 is in accordance with RG 1.92 and combines loads based on three orthogonal components. The NRC staff finds this L&C to be satisfied because the licensee combined the loads in accordance with it, as the licensee's response confirms.
L&C 9: ((
))
Evaluation: The licensee stated that a linear viscoel astic grid impact model is used for the IGM grid. The model limits the impact force to the buckling strength, ((
)) The licensee further stated that, per Section 4.13.2 of Attachment3 to the LAR, this L&C is not applicable for the GAIA grid. The NRC staff finds this L&C to be satisfied because (1) the licensee stated that impact forces would be less than ((
)) for the IGM grids and (2) because, as elaborated further above in the discussion of L&C 1, the linear viscoelastic grid model from ANP-10337P-A is not applied to the GAIA grids.
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3.5.1.12 Deformable Grid Element (DGE) topical report (ANP-10337, Supplement 1)
The Deformable Grid Element (DGE) topical report (ANP-10337, Supplement 1, Reference 18)
SER defines the criteria and methods for analyzing a spacer grid that deforms significantly before failing. Specific to the GAIA fuel transit ion at MPS3, the DGE topical report is limited to the GAIA structural grids (i.e., not the IGM grids) since the IGM grids do not have this deformation characteristic. Per Section 1.0 of the SER for the DGE topical report, the methodology is intended for use with the ANP-10337P, Revision 0, base methodology (Reference 17) which is referenced in ANP-10342.
Per Section 5.0 of the SER for ANP-10337, Supplement 1 (Reference 18), there are a total of two L&Cs. These L&Cs are used in conjunction with those of the base methodology per Table 4-14 of the LAR, with the exception of L&Cs #1, #2b, and #9 which are only applicable to the linear viscoelastic elements. Based on the assessments of the two (2) additional L&Cs in Table 4-15 of the LAR,
L&C 1: The residual deformation limit must be equal to or less than the smaller of the following two values: (1) the maximum residual deformation observed during the tests to determine the input parameters for the DGE model, or (2) the deformation limits defined in ANP-10337P-A, that is, ((
))
Evaluation: The allowable GAIA grid deformation limits have been defined in accordance with items (1) and (2).
L&C 2: The applicability of the DGE model is limited to grids comprised of M5 alloy, Q12 alloy, or materials that maintain an irradiated ductility.
Evaluation: M5 material is used for the GAIA grids, which are the only components to use the DGE model.
Based on the above, all L&Cs are met for fuel transition at MPS3.
3.6 Design Evaluations
As discussed in section 2.0 of this SE, the fuel system design bases must reflect four objectives:
(1) the fuel system is not damaged as a result of normal operation and AOOs, (2) fuel system damage is never so severe as to prevent control rod insertion when it is required, (3) the number of fuel rod failures is not underestimated for postulated accidents, and (4) coolability is always maintained. To satisfy these objectives, acceptance criteria are needed for fuel system damage, fuel rod failure, and fuel coolability. The LAR evaluates mechanical design-related aspects of the fuel system design basis; a demonstration of the satisfaction of other SRP objectives, including analysis of fuel integrity for AOOs, analysis of the number of fuel rod failures in postulated accidents, and demonstration of fuel coolability requirements for a postulated LOCA event are under review by NRC staff as part of separate license amendment requests.
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3.6.1 Mechanical Evaluations
The GAIA fuel mechanical design criteria are met up to the licensed peak UO 2 fuel rod burnup of 62 GWd/MTU and peak GAD fuel rod burnup of 55 GWd/MTU for batch implementation at MPS3.
The design criteria relating to fuel system damage should not be exceeded during normal operation including AOOs. Fuel rod failure should be precluded and fuel damage criteria should ensure that fuel system dimensions remain within operational tolerances and that functional capabilities are not reduced below those presumed in the safety analysis. Each damage mechanism listed in SRP Section 4.2 will be reviewed to confirm that the design criteria are not exceeded during normal operation for the GAIA design.
3.6.1.1 Normal Operation and AOO Component Stress
The licensed method associated with the normal operation and AOO component stress analysis is in the GAIA Mechanical Design topical report (Reference 8). The design criteria for stress are that the stress intensities for GAIA fuel assembly components shall be less than the stress limits based on the American Society of Mechanical Engineers Code,Section III criteria (Reference 22). These design criteria are consistent with the acceptance criteria of SRP Section 4.2; therefore, the stress criteria are acceptable for application to the GAIA fuel design.
A deterministic method is used to obtain the most limiting stress value. The method provides the most conservative stress value for each fuel assembly component. Positive margin to the design criteria is shown for each of the fuel assembly components; therefore, the NRC staff concludes that the fuel assembly design satisfies the design criteria for design stress.
3.6.1.2 Shipping and Handling Component Stress
Licensing criterion for the shipping and handling component stress fuel damage mechanism is in the GAIA Mechanical Design topical report (Reference 8). Stresses and/or loads associated with shipping and handling shall be less than limits based on Section III of the ASME Code (Reference 22) for all components, unless otherwise specified. All axial loads for shipping and handling are evaluated for beginning of life (BOL) conditions and consider tension and compression. Shipping conditions are evaluated based on the maximum MAP-12 shipping container temperatures for normal conditions of transport. The applicable shipping and handling load limits for GAIA fuel include (1) 6g 7 lateral acceleration, (2) 4g maximum axial acceleration, and (3) 2.5g axial handling. The associated component stresses are within the specified Code limits, showing that the GAIA fuel assembly is structurally adequate for all shipping and handling loading conditions.
3.6.1.3 General Component Fatigue
For all components other than fuel rod cladding, the cumulative usage factor (CUF) shall be less than 1.0. The licensed method associated with the general component fatigue analysis is per Section 8.1.2.2 of the GAIA Mechanical Design topical report (Reference 8). Due to the significant load-bearing function of the GTs, in conjunction with their relatively thin-walled
7 g is acceleration of earth's gravity which is approximately 32.2 feet per second (ft/s2)
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construction, GT fatigue is considered the most limiting and bounds all other components of the fuel assembly with the exception of the holddown springs.
Fatigue calculations were performed in accordance with approved methods for the GAIA structural components, including fuel rods, guide tubes, and holddown springs. All component margins are positive, showing that the GAIA fuel assembly is structurally adequate for all normal operating conditions and AOOs.
3.6.1.4 Fretting Wear
The design criteria for fretting are that the GAIA fuel assembly design shall be shown to have no failure due to fretting (Reference 8). This criterion is consistent with the acceptance criteria of SRP Section 4.2; therefore, the fretting criteria are acceptable for application to the GAIA fuel design.
The licensee performed extensive autoclave testing using expected EOL condition for the GAIA fuel assemblies. Fretting wear and performance testing were performed at the HERMES P (Cadarache, France and Richland, Washington) flow test facilities. The 1000-hour endurance flow tests were performed and followed up by additional tests at the PETER loop (Erlangen Germany). Additionally, individual component pressure drop tests were performed at MAGALY test loop (Le Creusot, France). Evaluations of this extensive testing showed that the GAIA fuel assembly is predicted to meet all criteria through EOL. Therefore, the licensee has demonstrated that the GAIA fuel has the ability to meet this criterion.
3.6.1.5 Fuel Rod and Fuel Assembly Growth
Axial and lateral dimensional changes in the fuel rod and fuel assembly can occur due to irradiation growth, irradiation relaxation, creep, thermal expansion, etc. and can cause component-to-component or component-to-core interferences. These may lead to component failures and/or impacts on thermal-hydraulic limits, control rod insertion, and/or handling damage.
Fuel rod irradiation growth is addressed by pr oviding sufficient clearance between the fuel rod and assembly top and bottom nozzles at EOL. Fuel assembly irradiation growth is addressed by providing sufficient clearance between the fuel assembly and reactor core plates at EOL. To assess fuel rod and fuel assembly growth, empirical models are used to compute the irradiation growth of the applicable components and the resulting changes are compared with the specified dimensions. This is in accordance with cr iteria previously approved by the NRC.
The upper bound fuel rod growth and lower bound fuel assembly growth is used in conjunction with component manufacturing tolerances to determine the fuel rod shoulder gap margin. The upper bound fuel assembly growth is used in conjunction with component and core plate manufacturing tolerances to determine the fuel assembly gap margin. Limiting burnups and temperatures are considered.
The NRC-approved M5 fuel rod growth model in Reference 19 is used for the fuel rod growth bounds. The NRC-approved Q12 guide tube growth model in Reference 9 is used for the fuel assembly growth bounds.
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A limiting fuel rod shoulder gap clearance of (( )) inch and fuel assembly core plate clearance of (( )) inch was determined at EOL, showing that the GAIA fuel assembly is structurally adequate with respect to fuel rod and fuel assembly growth for analyzed normal operating conditions and AOOs.
3.6.1.6 Fuel Assembly Liftoff
Framatomes licensing criterion for the fuel assembly lift-off fuel damage mechanism is per Section 8.1.7.1 of the GAIA Mechanical Design topical report (Reference 8).
Section 5.1.5 of Attachment 3 to the proposed LAR summarizes the licensee's proposed approach for addressing fuel assembly lift-off. The design criteria for assembly liftoff are that during normal operating conditions and AOOs (with the exception of a pump over-speed transient), the GAIA fuel holddown springs maintain positive holddown margin (i.e., fuel assembly contact with the lower support plate). Assuming a pump over-speed transient, fuel assembly lift-off can occur, but the fuel assembly top and bottom nozzles maintain engagement with reactor internal pins, and the holddown springs maintain positive holddown margin after the event.
The fuel assembly lift-off methodology makes use of conventional open-literature equations to obtain a balance of forces on the fuel assembly in the vertical direction. The forces are due to fluid friction loss, buoyancy, momentum change, holddown spring force, and gravity. Forces due to friction losses are obtained through the use of loss coefficients derived from flow testing.
Holddown forces are obtained from testing. Other forces due to momentum and buoyancy are calculated based on the applicable fluid condition s. The evaluation includes the assessment of bounding operating conditions, component dimensional characteristics, and material characteristics. The Q12 GT growth model in Reference 9 is used to determine the fuel assembly growth bounds. Uncertainties are accounted for using a combination of deterministic and statistical methods.
In Section 3.3.1.9 of its SE on ANP-10342P-A, the NRC staff found that Framatome performs a combination of deterministic and statistically based analysis and can demonstrate that during all conditions considered, except for the pump over-s peed transient, the fuel assembly liftoff criteria are met. During the pump over-speed transient, any liftoff will not be capable of disengaging the fuel assembly from core support plate alignment pins, and the hold-down spring deflection is less than the worst-case normal operating cold-shutdown condition. Therefore, the NRC staff concluded in its SE on ANP-10342P-A that for the GAIA fuel assembly design, the fuel assembly liftoff criteria are met.
For the present review, the NRC staff reviewed the submitted information in the proposed LAR concerning the licensees methodology and finds that the methods and acceptance criteria proposed are consistent with those generically approved in ANP-10342P-A. Positive holddown margin is maintained for all events except the pump over-speed transient, for which the lift is small and would not be expected to lead to misalignments or other adverse impacts on the fuel assemblies or core internals. Therefore, consistent with the approved methodology in ANP-10342P-A, the staff finds the proposed approach to be acceptable.
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3.6.1.7 General Component Structural Deformation
Earthquakes and postulated pipe breaks in the reactor coolant system would result in external forces on the fuel assembly. The fuel assembly is designed to withstand the structural loads from the Operating Basis Earthquake (OBE), Safe Shutdown Earthquake (SSE), and LOCA events without loss of the capability to perform in a manner credited in its design basis for these events.
Licensing criteria for the general component structural deformation fuel damage mechanism are per Section 4 of the topical report (ANP-10337P-A, Revision 0, Reference 17). Specific to the GAIA spacer grid, additional criteria per Section 3.2 of the Deformable Grid Element (DGE) topical report (Reference 18) are applicable. Per Section 4 of Reference 17, OBE stress and load limits are set at the level A limits defined in the ASME Code, unless otherwise specified.
SSE and LOCA stress and load limits are set at t he Level D limits defined in the ASME Code, unless otherwise specified.
Due to their special functions (i.e. forming a path for control rod insertion, ensuring coolable geometry is maintained), IGM grids and GTs are subject to more stringent service limits including:
OBE Spacer Grid Acceptance Criteria:
IGM grid deformation experienced during an OBE event should not exceed the magnitude of the tolerance band to which the grid was designed. This acceptance criterion is established in the form of a grid impact load limit, which correspon ds to a small amount of plastic deformation in the spacer grid that is within the envelope tolerance and does not exceed the deformation at the buckling point of the grid.
SSE/LOCA Spacer Grid Acceptance Criteria:
IGM grid deformation experienced during an SSE/LOCA event ((
))
SSE/LOCA Guide Tube Acceptance Criteria:
Sudden and severe changes in the geometry of the GT (e.g. local collapse or plastic hinge) shall not occur. This acceptance criterion is further delineated by requiring that (1) stresses do not exceed a limit prohibiting local collapse of the GT, and (2) the structural stability of the GT must be maintained. The first criterion is met by limiting GT stresses to the Level C criteria in accordance with the ASME Code. The second criterion is satisfied by evaluating the critical buckling load margin.
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The licensed method associated with the general component structural deformation analysis is per Sections 7 and 8 of the topical report (ANP-10337P-A, Revision 0, Reference 17) and Sections 4 and 5 of the Deformable Grid Element topical report (Reference 18).
All component margins are positive, showing that the GAIA fuel assembly is structurally adequate for all postulated accident conditions and AOO's.
3.6.1.8 Fuel Assembly Drop Accident
The licensee stated that analysis of the fuel assembly drop accident is completed to ensure the number of failed fuel rods assumed in the fuel handling accident analysis of record remains bounding. This analysis is not explicitly associated with a SAFDL included in Reference 8, since SAFDLs are typically associated with the establishment of appropriate limits for operating conditions. Rather, analysis of the fuel assembly drop accident is performed to ensure that the public is appropriately protected from potential releases of radioactive materials.
From a regulatory perspective, the licensee's object ive of ensuring that the number of failed fuel rods allowed in the FSAR basis remains bounding would facilitate minimization of the potential for reanalysis and licensing basis changes that could require regulatory review. However, design limits for the number of failed GAIA fuel rods due to a fuel handling accident would ultimately be established by ensuring that regulatory dose limits for the event remain satisfied.
There are three (3) fuel handling accident cases summarized in the MPS3 FSAR, and the number of failed fuel rods currently predicted for each scenario are as follows:
i) 283 total failed fuel rods for the fuel handling accident in the fuel building spent fuel pool (SFP) - (Section 15.7.4.2.1 of the MPS3 FSAR);
ii) 283 total failed fuel rods for the fuel handling accident in containment (Section 15.7.4.2.2); and
iii) 30 total failed fuel rods for the fuel handling accident involving the drop of an insert component in the SFP (drop of RCCA -Section 15.7.4.2.3).
As discussed below, in all of these fuel handling accident scenarios, the estimated number of failed fuel rods for the GAIA fuel assembly is bounded by the number of failed rods defined above in the current licensing basis.
- For the case where a GAIA fuel assembly is dropped on another GAIA fuel assembly, a total of (( )) failed fuel rods are predicted in the SFP accident and (( )) failed fuel rods are predicted in the containment accident, both below the existing analyzed value of 283. In response to RAI-1-(b.) (Reference 2), the licensee stated that the difference in quantity of failed fuel rods predicted between the containment fuel handling accident and the spent fuel pool fuel handling accident is primarily attributed to the significant difference in drop height inputs for each scenario (drop height is approximately 8.5 times higher in containment than in the spent fuel pool). Thus, Framatomes prediction for number of failed fuel rods during a fuel handling accident is sensitive to drop height (RAIs, Reference 2).
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- For the case where a Westinghouse fuel assembly is dropped on a GAIA fuel assembly, a total of (( )) failed fuel rods are predicted in the SFP accident and (( )) failed fuel rods are predicted in the containment accident, both below the existing analyzed value of 283.
- For the case where a RCCA is dropped on a GAIA fuel assembly, a total of (( )) failed fuel rods are predicted, below the existing analyzed value of 30.
3.6.2 Thermal Mechanical Evaluations
Design criteria relating to the fuel rod failure are applied in two ways. When they are applied to normal operation including AOOs, they are used as limits (i.e., SAFDLs), since fuel failure should not occur under these conditions. When they are applied to postulated accidents, limited fuel failures are permitted and must be accounted for in the fission product releases. Fuel rod failure is defined as the loss of fuel rod hermeticity. Each fuel rod failure mechanism listed in SRP Section 4.2 will be reviewed to confirm that the design criteria are not exceeded during normal operation and are properly accounted for during postulated accidents for the GAIA design.
Fuel rod thermal-mechanical evaluations are dependent on the rod power and core operating parameters. Because these parameters may vary for each operating cycle, verification on a cycle-specific basis is needed to ensure the actual cycle design will not result in SAFDL non-compliance. All fuel rod analyses are based on inputs which either represent or bound expected operating conditions at MPS3.
3.6.2.1 Cladding Stress Normal Operation and AOO
Licensing criteria for the normal operation and AOO cladding stress fuel damage mechanism are per the M5 topical report (Reference 10, Section 3.3). The licensed method associated with the fuel rod cladding stress normal operation and AOO analysis is per Section 8.1.1.2 of the GAIA Mechanical Design topical report (Reference 8), with material properties per the M5 topical report (Reference 10) used as inputs.
((
)) The criteria for normal operation and AOO cladding stress have been met, with a limiting margin of (( ))
3.6.2.2 Cladding Buckling (( ))
Licensing criterion for the cladding buckling (( )) fuel damage mechanism is per the M5 topical report (Reference 10, Section 3.3).
(( ))
The licensed method associated with the cladding buckling (( )) analysis is per the M5 topical report (Reference 10, Section 3.3), including material properties used as input.
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((
))
The criterion for cladding buckling (( )) has been met, with a limiting margin of ((
))
3.6.2.3 Cladding Structural Deformation (Faulted Stress)
The general licensing criterion for the faulted cladding stress fuel damage mechanism is per topical report ANP-10337P-A (Reference 17). Explicit criteria for the faulted cladding stress fuel damage mechanism are per the M5 topical report (Reference 10, Section 3.3), as directed by Section 8.3.1.1 of Reference 8. M5 fuel rod cladding stress criteria are in accordance with those defined in Section 5.2.1.1 of Attachment 3 to the LAR.
The licensed method for calculating the overall faulted stresses and margins is consistent with Section 8.1.1.2 of the GAIA Mechanical Design topical report (Reference 8), with material properties per the M5 topical report (Reference 10) used as input.
The criteria for faulted cladding stress have been met, with a limiting margin of ((
)) respectively for BOL and EOL. Fuel rod mechanical fracturing does not occur and the strength of the M5 cladding is not exceeded under faulted condition loads.
3.6.2.4 Transient Cladding Strain
Licensing criterion for the transient cladding strain (TCS) fuel damage mechanism is per Section 8.1.1.1 of the GAIA Mechanical Design topical report (Reference 8).
Maximum uniform hoop strain (elastic plus plastic) shall not exceed 1 percent.
The licensed method associated with the TCS analysis is per the COPERNIC topical report (Reference 11, Section 12.4). COPERNIC predicts the linear heat rates at which the cladding uniform hoop strain equals 1%. Cases are run with ((
))
For the TCS fuel damage mechanism, linear heat rate limits have been predicted at which the cladding uniform hoop strain equals 1 percent. The TCS limits were provided to Dominion and need to be verified by Dominion on a cycle-specific basis.
3.6.2.5 Cladding Fatigue
The design criterion for M5 cladding fatigue is that the GAIA maximum fuel rod fatigue cumulative usage factor (CUF) shall not exceed 0.9. Licensing criterion for the cladding fatigue
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fuel damage mechanism is per Section 8.1.2.1 of the GAIA Mechanical Design topical report (Reference 8). This design criterion is consistent with the acceptance criteria of SRP Section 4.2; therefore, this cladding fatigue criterion is acceptable for application to the GAIA fuel design.
The licensed method associated with the cladding fatigue analysis is per the M5 topical report (Reference 10, Section 3.6), including material properties used as input.
Procedures for the fatigue analysis follow those outlined in the ASME Code. The analysis uses all the Condition I and II events and one Condition III event to determine the total cladding fatigue usage factor. The maximum fatigue usage factor was determined to be well below the design criteria limit. Since the methodology is consistent with the guidance in SRP Section 4.2 and the maximum fatigue is well below the design criteria limit, it is demonstrated that the cladding fatigue acceptance criterion has been met.
3.6.2.6 Cladding Oxidation
The criterion for the cladding oxidation fuel damage mechanism is per Section 8.1.4.1 of the GAIA Mechanical Design topical report (Reference 8). The design criterion for cladding oxidation is that the GAIA fuel rod cladding best-estimate corrosion shall not exceed 100 microns. While there is no established defined limit on hydrogen pickup for the cladding, hydrogen content in the cladding must be accounted for because it affects cladding material properties (e.g., ductility), such that margins to regulatory criteria can be affected. The M5 cladding material fuel rods are expected to have less than 50 microns of oxidation at the maximum allowable peak rod burnup of 62 GWd/MTU. Hydrogen pickup is a material-dependent property that is driven by cladding alloying elements and accelerated by thicker oxide layers. Initial hydrogen pickup is limited by vendor-controlled manufacturing processes that remain unchanged for GAIA. Additionally, the M5 cladding material has an improved resistance to hydrogen pickup relative to previous generations of zirconium-alloy claddings.
The licensed method associated with the cladding oxidation analysis is per the COPERNIC topical report (Reference 11, Section 12.6).
Criteria for cladding oxidation are intended to preclude potential fuel damage mechanisms. The SRP does not specify limits on cladding oxidation but does specify that its effects should be accounted for in the thermal and mechanical analyses performed for the fuel. The licensee accounts for the corrosion based on a database established for the M5 cladding material from its in-reactor performance over a number of years. The methodology and limits for cladding oxidation defined in Reference 8 are applicable and acceptable in the evaluation of the GAIA fuel assembly. This approach is acceptable because it uses realistic data that is representative of the material and burnup limits for the GAIA fuel assembly design.
Based on the data for M5 cladding material taken under prototypical irradiation conditions, there is reasonable assurance that oxidation will remain well below its acceptance criterion under the expected operating conditions at MPS3.
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3.6.2.7 Fuel Rod Internal Pressure
The design criterion for fuel rod internal pressure is that rod internal pressure is limited to that which would cause (1) the diametral gap to increase due to outward creep during steady-state operation or (2) reorientation of the hydrides in the radial direction in the cladding. Explicit criteria for the fuel rod internal pressure fuel damage mechanism are per the Fuel Rod Gas Pressure Criterion (FRGPC) topical report (Reference 14). The M5 topical report (Reference 10) extends the applicability of the FRGPC topical report to M5 material. These design criteria have been applied in previous fuel assembly designs and will continue to be valid since the parameters used in the methodology remain unchanged. Therefore, these criteria are acceptable for application to the GAIA fuel design.
The fuel rod internal pressure analysis uses the COPERNIC code with the methodology approved in Reference 11, Section 12.1. This analysis, performed on a plant-specific basis, includes the use of the most limiting manufacturing variations and a bounding power history for that plant. If the bounding analysis does not meet the fuel rod internal pressure criteria, then on a cycle-specific basis a rod-specific analysis using the actual power history and manufacturing data for that rod can be performed to demonstrate that the internal rod pressure criteria are satisfied. These dual analysis paths using the approved methodology are acceptable for use because either approach will demonstrate that the fuel rod internal pressure criterion is met.
3.6.2.8 Internal Hydriding
The design criterion for internal hydriding is that internal hydriding shall be precluded by appropriate manufacturing controls. The licensed method associated with fuel rod internal hydriding is per Section 8.2.1.2 of the GAIA Mechanical Design topical report (Reference 8). For the GAIA assembly design, hydriding is prevented by keeping the level of moisture and hydrogenous impurities within the fuel to very low levels. GAIA UO 2 and GAD fuel pellets have a total hydrogen content ((
)). This design criterion is consistent with the acceptance criteria of SRP Section 4.2 and is therefore acceptable.
Framatome maintains the low hydrogen levels in the fuel rod through manufacturing controls.
Because these controls will remain in place for the GAIA fuel assembly design and the limits are lower than the SRP Section 4.2 values, the design criteria will continue to be met with the GAIA fuel assembly design.
3.6.2.9 Cladding Creep Collapse
The design criterion for cladding collapse is that the predicted creep collapse life of the fuel rod must exceed the maximum expected in-core life. The SRP states that if axial gaps in the fuel pellet column occur due to densification, the cladding has the potential to collapse into such a gap. Because of the large local strains that accompany this process, any collapsed cladding is assumed to fail. Because the design criterion is consistent with the acceptance criteria of SRP Section 4.2, it is acceptable for application to the GAIA fuel assembly design.
Framatome uses their approved creep collapse methodology CROV topical report (Reference 16), to determine the potential for creep collapse of the GAIA fuel assembly design.
This methodology uses conservative values to determine the creep collapse life of the fuel rod.
When the ovality creep rate of the cladding exceeds (( )) the cladding pressure
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differential exceeds the bifurcation buckling limit, or the generalized stress exceeds the generalized yield strength, the cladding is deemed failed. Based on these definitions of creep collapse, the creep collapse lifetime was shown to be greater than the allowable rod-average burnup limit of 62 GWd/MTU. Therefore, the GAIA fuel assembly design is adequately designed to prevent creep collapse for a service life up to 62 GWd/MTU.
3.6.2.10 Overheating of Fuel Pellets (Centerline Fuel Melt, CFM)
The design criterion for overheating of the fuel pellets is that fuel pellet centerline melting shall not occur during normal operation and AOOs. This design criterion is consistent with the acceptance criteria of SRP Section 4.2; therefore, it is acceptable for application to the GAIA fuel assembly design.
SRP Section 4.2 states that this analysis should be performed for the maximum LHGR anywhere in the core, including all hot spots and hot channel factors, and should account for the effects of burnup and composition on the melting point. Framatome uses the COPERNIC computer code and fuel melt methodology (Reference 11) to determine the local LHGR throughout the fuel rod lifetime that could result in centerline temperature predictions exceeding the limit. As discussed in Section 3.3.2.4 of the NRC staff SE that is incorporated into ANP-10342P-A, for future analyses using the COPERNIC code and fuel melt methodology with the GAIA fuel assembly design, if the peak LHGR is not at the BOL, then the limiting time in life will be determined and thermal conductivity degradation will be accounted for in the centerline temperature predictions. Therefore, this analysis demonstrated that for the GAIA fuel assembly design the acceptance criteria are met.
3.6.3 Thermal-Hydraulic Evaluations
The proposed LAR addressed in this SE largely concerns technical assessments associated with fuel mechanical design. While the transition to a new fuel design also involves a broad set of thermal-hydraulic and other analyses, the licensee has largely addressed thermal-hydraulics and other analyses in separate LARs that the NRC staff reviewed separately (e.g. References 3 and 23). However, one thermal-hydraulic topic included in the proposed LAR and addressed in the present SE is fuel rod bow. Specifically, while fuel rod bow impacts downstream thermal-hydraulic analyses, the governing phenomena that drive fuel rod bow are mechanical in nature.
3.6.3.1 Fuel Rod Bow
The design criterion for fuel rod bow is that fuel rod bowing shall be evaluated with respect to the mechanical and thermal-hydraulic performance of the fuel assembly. There is not a specific limit for fuel rod bow specified in SRP Section 4.2; the SRP only calls for rod bow to be included in the design analysis to assure that it does not affect the satisfaction of other design criteria (e.g., DNB).
The methodology for fuel rod bow was approved in Reference 12. This database is representative of Zircaloy clad fuel. Because M5 cladding grows at a lower rate under irradiation conditions, the database for Zircaloy is conservative relative to the performance of M5. This approach remains unchanged and has been previously approved in Reference 19. Therefore, the NRC staff concludes that use of this database for predicting the rod bow of M5 clad fuel and continuing use of the penalty generated by the Zircaloy database for M5 fuel is conservative and acceptable for use.
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3.7 Technical Conclusion
The NRC staff reviewed the proposed LAR, in conjunction with additional information in supplemental submittals, related to the licensee's proposal to revise the MPS3 TS to incorporate information relevant to the mechanical design of Framatome GAIA fuel assemblies.
As discussed above in section 3.1 of this SE, the proposed TS changes included a revision to the safety limit for fuel centerline melt (TS 2.1.1.2), inclusion of a reference to M5 cladding in the reactor core design features (TS 5.3.1), and the addition of Framatome topical report ANP-10342-P-A in the core operating limits report references section (TS 6.9.1.6.b). The NRC staff notes that this safety evaluation and the subsequent conclusions presented herein are applicable to the LAR in support of full batch loading of Framatome GAIA fuel assemblies with M5 cladding to the MPS3 reactor as long as the licensee complies with the methodologies as described within the LAR.
Based on its review as described in detail above, the NRC staff concludes that the licensee has provided adequate technical basis to support the proposed TS changes. Specifically, the NRC staff finds the licensee has demonstrated that (1) the methods proposed by the licensee are applicable for the intended purpose, (2) the licensee complies with the staff-imposed limitations and conditions imposed for application of the topical reports where applicable, (3) the Framatome GAIA fuel assembly specific safety analyses results meet the applicable licensing criteria, and (4) the proposed TS changes are acceptable and satisfy the 10 CFR 50.36 requirements.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Connecticut State official was notified of the proposed issuance of the amendment on May 1, 2024. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, published in the Federal Register on August 1, 2023 (88 FR 50186), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),
no environmental impact statement or env ironmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the
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amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
- 1. Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station Unit 3, Proposed Amendment to Revise Technical Specifications for Reacto r Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report Related to Framatome GAIA Fuel, May 23, 2023, ADAMS (ADAMS) Accession No. ML23145A195.
- 2. Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station Unit 3, Response to Request for Additional Information Regarding Proposed License Amendment Request to Revise Technical Specifications for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report Related to Framatome GAIA Fuel, December 21, 2023, ML23361A094.
- 3. Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station Unit 3, License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits and Exemption Request for use of M5TM Cladding, May 2, 2023, ML23123A279.
- 4. Guzman, R., U.S. Nuclear Regulatory Commission, letter to Carr, E., Dominion Energy Nuclear Connecticut, Inc., Millstone Power Station, Unit No. 3 - Exemption from the Requirements of 10 CFR Part 50, Section 50.46 and Appendix K Regarding Use of M5 Cladding Material (EPID L-2023-LLE-0013), May 21, 2024, Package ML24110A058.
- 5. Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station, Units 1, 2, and 3 Updates to the Final Safety Analysis Reports, dated June 28, 2023, Package ML23193A862.
- 6. Standard Review Plan, Section 4.2, NUREG-0800 Revision 3, U.S. Nuclear Regulatory Commission, March 2007.
- 7. U.S. Nuclear Regulatory Commission, Removal of Cycle-Specific Parameter Limits from Technical Specifications, Generic Letter 1988-16, October 4, 1988, ML031200485.
- 8. ANP-10342NP-A, Revision 0, GAIA Fuel Assembly Mechanical Design, Topical Report, Framatome, Inc., September 2019, ML19309D916.
- 9. ANP-10334NP-A, Revision 0, Q12' Structural Material, Topical Report, Framatome, Inc., September 2017, ML17320A132.
- 10. BAW-10227P-A, Revision 1, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, June 2003 (non-publicly available).
- 11. BAW-10231NP-A, Revision 1, COPERNIC Fuel Rod Design Computer Code, January 2004, ML042930240.
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- 12. XN-75-32P-A, Supplements 1, 2, 3, & 4, Computational Procedure for Evaluating Fuel Rod Bowing, October 1983.
- 13. BAW-10227NP-A, Revision 2, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, January 2023, ML23037A888.
- 14. BAW-10183P-A, Revision 0, Fuel Rod Gas Pressure Criterion (FRGPC), July 1995.
(non-publicly available)
- 15. ANP-10311NP-A, Revision 1, COBRA-FLX: A Core Thermal-Hydraulic Analysis Code, Topical Report, AREVA NP Inc. October 2017, ML18103A141.
- 16. BAW-10084P-A, Revision 3, Program To Determine In-Reactor Performance of BWFC Fuel Cladding Creep Collapse, July 1995 (non-publicly available).
- 17. ANP-10337NP-A, Revision 0, PWR Fuel Assembly Structural Response to Externally Applied Dynamic Excitations, April 2018, ML18144A821.
- 18. ANP-10337NP-A, Revision 0, Supplement 1P -A, Revision 0, Deformable Spacer Grid Element, September 2020, ML20345A284.
- 19. BAW-10240NP-A, Revision 0, Incorporation of M5' Properties in Framatome ANP Approved Methods, May 2004, ML042800314.
- 20. ANP-3941P, Revision 0, GAIA Lead Test Assembly PIE Report, Technical Report, July 2021 (NRC Submittal Number ML21218A137, dated 7/30/2021, and ML21218A136).
- 21. ANP-10338NP-A, Revision 0, AREA' - ARCADIA Rod Ejection Accident, December 2017, ML18059A782.
- 22. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, Division 1, 2010, 2011a Addenda, 2013.
- 23. Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station Unit 3, Proposed Amendment to Support Implementation of Framatome GAIA Fuel, October 30, 2023, ML23304A047.
- 24. Peters, G., AREVA Inc., letter to U. S. Nuclear Regulatory Commission, Response to Request for Additional Information Regarding ANP-10334P, Q12 TM Structural Material, November 4, 2016, ML16313A235.
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- 25. Peters, G., AREVA Inc., letter to U. S. Nuclear Regulatory Commission, Response to RAI3 in Request for Additional Information Regarding ANP-10334P, Q12 TM Structural Material, December 16, 2016, ML16356A013.
Principal Contributors: R. Fu, NRR J. Lehning, NRR
Date: June 4, 2024
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