ML26043A152
| ML26043A152 | |
| Person / Time | |
|---|---|
| Site: | Westinghouse |
| Issue date: | 03/09/2026 |
| From: | Jeremy Dean, Fu R Licensing Processes Branch |
| To: | |
| Lenning E, NRR/DORL/LLPB | |
| References | |
| EPID L-2024-TOP-0007, WCAP-18869-P, Rev. 0, WCAP-18869-NP, Rev. 0 | |
| Download: ML26043A152 (0) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION FINAL SAFETY EVALUATION TOPICAL REPORT WCAP-18869-P/NP, REVISION 0 HIGH PERFORMANCE CLADDING FOR USE IN BOILING WATER REACTOR FUEL WESTINGHOUSE ELECTRIC COMPANY DOCKET NO. 99902038; EPID L-2024-TOP-0007
1.0 INTRODUCTION
By letter dated March 8, 2024 (Ref. 1), Westinghouse Electric Company (Westinghouse) submitted Topical Report (TR) WCAP-18869-P/NP, Revision 0, High Performance Cladding for Use in Boiling Water Reactor Fuel (Proprietary/Non-proprietary), for the U.S. Nuclear Regulatory Commission (NRC) staff review and approval.
This cladding material is a zirconium (Zr)-based alloy designed to maximize the safety margins for boiling water reactor (BWR) fuel, amid increasing demands for higher fuel duties and burnup, by reducing the hydrogen uptake. Westinghouse stated that it [
].
The improved performance of HiFi' cladding is the result of optimizing the Zr-based alloy dopants of iron (Fe) and chromium (Cr).
The NRC staff conducted a regulatory audit of the WCAP-18869-P/NP, Revision 0, on December 12-13, 2024 (Refs. 2 and 3), to ensure that the NRC staff had proper understanding of the TR, had access to all the information it needed to perform the review, and fully understood the scope of the TR. After the NRC staffs preliminary review and results of the audit, the NRC staff issued its request for additional information (RAI) (Ref. 4).
2.0 REGULATORY EVALUATION
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, contains the general design criteria (GDC) described in Appendix A to Part 50, including, GDC 10, Reactor design, GDC 25, Protection system requirements for reactivity control malfunctions, GDC 26, Reactivity control system redundancy and capability, GDC 27, Combined reactivity control systems capability, GDC 28, Reactivity limits, and GDC 35, Emergency core cooling.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION GDC 10 states:
The reactor core and associated coolant, control, and protection systems shall be designed with the appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
GDC 10 establishes specified acceptable fuel design limits to ensure that the fuel is not damaged. That means that fuel rods do not fail, fuel system dimensions remain within operational tolerances, and functional capabilities are not reduced below those assumed in the safety analysis. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - LWR Edition (SRP), Section 4.2, Fuel System Design, acceptance criteria are based on meeting the requirements of GDC 10.
Requirements for analyzing the design-basis loss-of-coolant accident (LOCA) are provided in 10 CFR 50.46, Appendix K to 10 CFR Part 50, and GDC 35. The most relevant regulations to this review are:
Per 10 CFR 50.46(a)(1)(i), each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated LOCAs conforms to the criteria set forth in Section 50.46(b). ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated LOCAs are calculated.
10 CFR Part 50, Appendix K, sets forth the documentation requirements for each evaluation model and establishes the required and acceptable features of evaluation models for heat removal by the ECCS.
GDC 35 requires abundant emergency core cooling sufficient to (1) prevent fuel and cladding damage that could interfere with effective core cooling and (2) limit the metal-water reaction on the fuel cladding to negligible amounts.
Regulatory guidance for the NRC staff review of fuel system designs that shows conformance to these GDC is provided in SRP, specifically, Section 4.2 (Ref. 5). Additionally, SRP Section 4.3, Nuclear Design (Ref. 6), and Section 4.4, Thermal and Hydraulic Design (Ref. 7), are pertinent to the review of fuel systems.
Consistent with SRP Section 4.2, the objectives of the fuel system safety review are to provide assurance that:
The fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs),
Fuel system damage is never so severe as to prevent control rod insertion when it is
- required,
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The number of fuel rod failures is not underestimated for postulated accidents, and Coolability is always maintained.
SRP Chapter 15, Transient and Accident Analyses (Ref. 9), including acceptance criteria for AOOs and postulated accidents and their impact on HiFi cladding, is addressed in the TR. The NRC staffs review of this TR is based on the acceptance criteria for each of the events described in SRP Chapter 15.
3.0 TECHNICAL EVALUATION
3.1 Alloy Definition The HiFi alloy was developed to reduce the hydrogen uptake to meet evolving requirements imposed on fuel cladding materials for BWR nuclear fuel, in particular, increasing demands for higher fuel duties and burnup. [
]. HiFi is a Zr alloy with higher Fe content and [
] than that of Zircaloy-2 cladding, the material currently in use in BWR fuel produced by Westinghouse. Table 1-1 in Section 1.3 of the TR WCAP-18869-P/NP, Revision 0, compares the chemical composition of HiFi and Zircaloy-2. HiFi cladding is defined to have a nominal [
] specified for Zircaloy-2 cladding in Reference 13. [
].
Westinghouse determined the final chemical composition of the HiFi cladding following extensive out-of-core testing, aiming to optimize manufacturability, mechanical properties, corrosion and hydrogen pickup.
3.2 Phase Transformation and Microstructure For temperatures below about 750 degrees Celsius (°C), Zr is in the so called phase, and for temperatures above 1000°C, Zr is in the phase. There is a mixed phase region, which varies depending on the alloy. In the Zircaloy-2, the mixed phase region covers the temperature range from 750 to 1000°C, and the precipitates are dissolved at about 850°C.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
[
].
3.3 Properties and Experience The data for the HiFi alloy has been collected in two separate development programs: (1) a high Fe cladding that was developed by NFI and (2) Westinghouse acquired relevant experience with
[
].
3.3.1 Manufacturing Process WCAP-18869-P/NP, Revision 0, states that HiFi cladding for Westinghouse nuclear fuel designs is manufactured following the same steps as the current Zircaloy-2 LK3 cladding. Processing of cladding tubes is commonly tailored to the chemical composition and to the equipment capabilities of each manufacturer to optimize the microstructure for robust performance.
In the response to RAI 7 (Ref. 4), Westinghouse states that with the exception of [
].
In summary, [
]:
[
].
[
].
[
].
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
[
].
[
].
[
].
[
].
[
].
Table 3-3 and Table A-1 in TR WCAP-18869-P/NP, Revision 0, demonstrated that the most obvious trend is that [
]. Typical average SPP sizes vary [
].
The NRC staff reviewed the information and determined that a very small difference in SPP size between Zircaloy-2 and HiFi is not expected to result in any detrimental effects on performance, as demonstrated by testing results presented in later sections of the TR.
After reviewing the manufacturing process information provided by Westinghouse, the NRC staff concluded that the manufacturing processes used for HiFi cladding show only minimal differences between manufacturers, driven mainly by the capabilities of different manufacturers.
These small differences do not result in significant differences in the resulting annealing parameters and associated microstructures, which demonstrate the equivalence of the
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.3.2.3.1 Thermal Creep WCAP-18869-P/NP, Revision 0, states that samples of HiFi cladding fabricated for different fuel designs were tested [
]. The results are shown in Section 3.2.3.1, Table 3-6 of the TR, and are compared with the similar tests performed for Zircaloy-2 LK3 cladding. The NRC staff concludes that the results demonstrate that the thermal creep of HiFi cladding is [
].
Irradiation creep is discussed in Section 3.4.2.3 of WCAP-18869-P/NP, Revision 0, which states that [
] regarding irradiation creep.
Westinghouse stated in the response to RAI 4 (Ref. 4) that since the measurements demonstrate that the creep behavior of HiFi cladding [
].
Figure 10 in the Westinghouses response to RAI 4 presents results from calculation of the irradiation cladding creep for a TRITON11 fuel rod with the fuel performance code STAV7.2.
The irradiation creep is a function of fluence and no pellet cladding mechanical interaction (PCMI) or differential pressure effects are considered.
Figure 11 in Westinghouses RAI 4 response contains calculations of lift-off pressure for TRITON11 fuel with the BWR fuel performance code described in WCAP-15836-P-A (STAV7.2). Since [
].
The accuracy of the predicted lift-off pressure will depend on the ability of the fuel performance code and methods to correctly capture the behavior of the specific fuel/cladding materials being used. The NRC staff reviewed Westinghouses response to RAI 4 and concluded that [
] applicable to the HiFi 3.3.2.3.2 Texture and Contractile Strain Ratio The texture of HiFi cladding tubes fabricated in multiple designs and reference Zircaloy-2 tubes produced with the same manufacturing parameters has been evaluated using laboratory X-ray diffraction by Westinghouse. Samples [
], were also tested for CSR, a bulk The NRC staff has reviewed the differences in the [
].
Therefore, the NRC staff concluded that [
] for HiFi cladding acceptable for use in Westinghouse fuel rod safety analysis evaluation models.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
[
].
Westinghouse demonstrated in Section 3.3.2 of TR WCAP-18869-P/NP, Revision 0, that the findings on the relationship between the annealing parameter A and nodular corrosion resistance aligned with the ex-core nodular corrosion tests. Specifically, [
].
3.3.4.2 Kashiwazaki-Kariwa 5 As described in TR WCAP-18869-P/NP, Revision 0, the irradiation program at Kashiwazaki-Kariwa Unit 5 (K5) was developed to [
]. This extensive program involved
[
].
The coupons inside the capsules underwent six irradiation cycles over roughly 2,500 days, achieving an equivalent burnup of 72 GWd/MTU and a fast fluence of 15x1025 n/m². Although coolant circulated within the capsules, the coupon surfaces were estimated to be in a boiling state due to -heating. TRAC code simulations indicated a void fraction of about 20 percent and a coupon temperature of 285°C. The interior of the capsules remained stable and consistent throughout the six cycles. TEM discs were kept in an inert atmosphere to prevent oxidation.
[
]. Measurements of oxide thickness during irradiation of coupons in K5, along with
[
], showed close proximity of data points, including some overlap, from side-by-side samples of HiFi and Zircaloy-2 tested under the same conditions. This indicates [
].
Figure 3-13 in TR WCAP-18869-P/NP, Revision 0, illustrates the hydrogen pickup fraction of the K5 coupons. In alignment with the out-of-pile tests, the uniform corrosion between Zircaloy-2 and HiFi cladding is comparable. However, the hydrogen pickup fraction of HiFi cladding is significantly lower than that of Zircaloy-2 cladding. For both materials, the hydrogen pickup fraction remains [
]
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.3.4.2.1 Tensile Test Tensile tests at 343°C and room temperature (Ref. 23) were performed on both HiFi cladding and reference Zircaloy-2 cladding coupon samples in K5. The results of these tensile tests, which measured yield strength and elongation for both materials as a function of fluence, along with additional results presented in the following sections, are shown alongside [
].
3.3.4.2.2 Fatigue Tests Westinghouse described in WCAP-18869-P/NP, Revision 0, fatigue tests that were carried out on both unirradiated and irradiated K5 coupons at room temperature show that HiFi material has comparable or superior performance, in terms of cycles to failure, compared to the reference Zircaloy-2 at a given stress amplitude. The results for irradiated HiFi material surpass the ODonnell-Langer best fit for unirradiated Zircaloy-2 (Ref. 32).
3.3.4.2.3 Irradiation Creep Closed specimens pressurized with noble gas were irradiated in K5 for 1,666 days. As shown in Figure 3-18 of TR WCAP-18869-P/NP, Revision 0, outer diameter (OD) measurements taken on these samples indicate that [
].
3.3.4.2.4 Characterization of Second Phase Precipitates TEM images of HiFi and Zircaloy-2 cladding that were taken after six irradiation cycles revealed that most of the precipitates were amorphous. Irradiation caused a reduction in the Fe content of the precipitates. This is demonstrated in Figure 3-20 of TR WCAP-18869-P/NP, Revision 0, which shows a decrease in the Fe-to-(Fe+Cr) ratio with increasing neutron fluence. The ratio eventually stabilizes, with HiFi material maintaining a slightly higher level than Zircaloy-2, owing to its higher Fe content.
3.3.4.3 Halden BWR Westinghouse indicated in TR WCAP-18869-P/NP, Revision 0, that fuel rods with HiFi and Zircaloy-2 cladding underwent irradiation testing in a BWR corrosion test loop within the Halden BWR (HBWR). The fuel rods were irradiated successfully, without any failures, achieving a rod average burnup of 60 GWd/MTU.
Intermediate inspections of the fuel rods were carried out after irradiation to fuel rod averages of 20 GWd/MTU and 40 GWd/MTU. Oxide thickness measurements, taken using eddy current at 40 GWd/MTU and 60 GWd/MTU, showed total oxide thicknesses ranging from 20 to 60 m, with thicker oxides observed on the surface nearer to the stainless-steel flask. HiFi cladding had a thinner oxide layer on the side facing the stainless-steel flask compared to Zircaloy-2 cladding.
Additionally, visual inspections revealed that the oxide film on the Zircaloy-2 cladding surface facing the flask had flaked off, whereas HiFi cladding showed minimal flaking. Significant oxide flaking was noted in the non-boiling region upstream of the Zircaloy-2 material. The thicker
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION oxide and the flaking can be attributed to the phenomenon known as shadow corrosion. These findings suggest that HiFi cladding may be less prone to shadow corrosion than Zircaloy-2 cladding.
3.3.4.4 Plant B - [
]
[
] fuel assemblies with HiFi cladding tubes were irradiated for [
], achieving [
] without any performance issues. Poolside inspections of these assemblies involved visual examinations, fuel rod diameter measurements, oxide thickness measurements using eddy current, and fuel rod growth measurements at various stages. Hot cell examinations included metallography, hydrogen content analysis, hardness tests, and burst tests.
[
].
The oxide thickness measurements for the rods shown in Figure 3-26 in the TR WCAP-18869-P/NP, Revision 0, confirm that HiFi and Zircaloy-2 cladding [
]. The hydrogen content of the fuel rod, obtained in hot cell, is shown in Figure 3-27 of the TR. [
]. Additionally, the hydrogen content of HiFi is [
].
3.3.4.4.1 Burst Test Westinghouse stated in TR WCAP-18869-P/NP, Revision 0, that burst tests were performed to assess the mechanical properties of irradiated HiFi cladding tubes as shown in Figures 3-29 and 3-30 of the TR. [
].
3.3.4.5 Plant C - [
]
In conjunction with the experience gained by NFI with HiFi cladding in Japan, Westinghouse initiated a development program focused on three different high Fe alloys. [
].
During the irradiation program, the fuel rods were inspected multiple times. Results from poolside inspections, as shown in Table 3-11 and Figure 3-31 to Figure 3-34 of TR
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[
].
The Westinghouse response to RAI 1 (Ref. 4) reported that [
].
Figure 3 in Westinghouses response to RAI 1 illustrates that [
].
[
].
[
].
The NRC staff reviewed the results and determined that [
].
3.4 Fuel Design and Accident Analysis 3.4.1 Fuel Assembly Mechanical Design The fuel assembly designs can be impacted by changes in unirradiated yield strength and ultimate strength. The mechanical strength for both irradiated and un-irradiated HiFi cladding is [
]. Therefore, the NRC staff concluded that the HiFi cladding meets the fuel assembly mechanical design criteria.
3.4.2 Fuel Rod Design Westinghouse BWR fuel designs are analyzed employing the following fuel rod design criteria by using the NRC-approved BWR methods and methodologies for [
]. Each criterion is specified along with the evaluation of the use of HiFi cladding on the specific criterion.
Section 4.2 of TR WCAP-18869-P/NP, Revision 0, contains Westinghouses evaluation of how
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.4.3 Nuclear Design The NRC staff reviewed the statements in Section 4.3 of the TR WCAP-18869-P/NP, Revision 0, indicating that [
]. The NRC staff has determined that [
] and no other nuclear design inputs result from the HiFi material changes. Based on this determination, the NRC staff concluded that Westinghouses disposition of Nuclear Design [
]
acceptable.
3.4.4 Thermal and Hydraulic Design Westinghouse states in Section 4.4 of the TR, WCAP-18869-P/NP, Revision 0, that the thermal-hydraulic analysis depends on the fuel assembly geometric conditions, the cladding surface finish, and the heat transferred to the surface of the cladding. The NRC staff has determined that [
].
Based on this determination, the NRC staff concluded that Westinghouses disposition of [
] is acceptable.
3.4.5 Non-LOCA Accident Design Westinghouse states in TR WCAP-18869-P/NP, Revision 0, Section 4.5, that in non-LOCA events the cladding temperature remains below the to + phase transition temperature, precluding any significant differences in specific heat. The specific heat, and in general all the thermo-mechanical properties, of Zircaloy-2 and HiFi cladding are [
]. The NRC staff reviewed the statement in the TR and determined that it was acceptable based on the engineering judgement of non-LOCA transient analyses and which parameters affect the results. This is typical of previous fuel rod cladding material changes across the industry.
3.4.6 LOCA Design Westinghouse states in TR WCAP-18869-P/NP, Revision 0, Section 4.6, that thermal conductivity, specific heat, density, thermal expansion, and emissivity of HiFi material and Zircaloy-2 material are [
]. The NRC staff reviewed the statement in the TR and determined that it was acceptable based on the engineering judgement of LOCA analyses and which parameters affect the results. This is typical of previous fuel rod cladding material changes across the industry.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.4.7 Pellet-Cladding Interaction and Reactivity-Initiated Accident The susceptibility to pellet-cladding interaction (PCI) is determined by the properties of the inner wall (i.e., the liner). Westinghouse states in TR WCAP-18869-P/NP, Revision 0, Section 4.7, that
[
].
The susceptibility of cladding to reactivity-initiated accident (RIA) is determined by a combination of materials microstructure and the hydrogen content. [
].
Therefore, the NRC staff reviewed HiFi material behaviors described in the TR and determined that the PCI and RIA methods and evaluations are acceptable, as discussed above.
3.4.8 Hydrogen Pickup Figure 4-1 in TR WCAP-18869-P/NP, Revision 0, summarizes hydrogen data from both unirradiated and irradiated programs. The TR states that the benefit of a lower hydrogen pickup with HiFi cladding [
]. Based on the NRC staffs review of the hydrogen complete data set from both unirradiated and irradiated programs presented in Figure 4-1 of the TR, the NRC staff concluded [
].
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 4.0 LIMITATIONS AND CONDITIONS HiFi cladding as described in TR WCAP-18869-P/NP, Revision 0, is intended for use within the following limitations and conditions:
- 1. The fuel rod burnup limit for this approval shall remain at currently established limits as specified in the SEs for the following TRs:
CENPD-287-P-A, Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors (Ref. 11),
WCAP-15942-P-A, Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors Supplement 1 to CENP-287 (Ref. 12), and WCAP-15836-P-A, Revision 0, Fuel Rod Design Methods for Boiling Water Reactors - Supplement 1, Volumes I & II (Ref. 10).
- 2. All the Limitations & Conditions listed in NRC final SEs for the NRC-approved TRs in References 10, 11, and 12 for methodologies used for Zircaloy-2 fuel analysis shall continue to be met, except that the use of HiFi cladding in addition to or in place of Zircaloy-2 cladding is now approved. Those include Oxide Thickness < 100 m, Hydrogen Pickup [
].
5.0 CONCLUSION
The NRC staff reviewed TR WCAP-18869-P/NP, Revision 0 (Ref. 1), to ensure that all applicable NRC regulations and requirements for fuel rod cladding material would be able to be met. Based on its evaluation discussed above, the NRC staff has determined that Westinghouse demonstrated that it has up to date evaluation models and methods to evaluate HiFi cladding performance in plant specific license amendment requests to support establishing compliance with all applicable NRC regulations and requirements for plant licensees.
The NRC staff reviewed TR WCAP-18869-P/NP, Revision 0, and determined that Westinghouse has demonstrated acceptable performance of HiFi cladding material using its currently approved methodology for evaluation when done in accordance with the limitations and conditions in Section 4.0 of this SE. Westinghouse has demonstrated that [
] to HiFi cladding The NRC staff reviewed all the information and data provided in Reference 1 and the RAI responses regarding HiFi performance beyond the current fuel burnup limits and has determined that Westinghouse can acceptably analyze HiFi cladding operation up to [
] with respect to cladding material figures of merit only. This acceptability determination does not raise the current burnup operating limit in References 10 and 12.
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6.0 REFERENCES
- 1. Letter from Z. Harper, Westinghouse, to NRC Document Control Desk (DCD), Submittal of Westinghouse Topical Report WCAP-18869-P/NP, Revision 0, High Performance Cladding for Use in Boiling Water Reactor Fuel (Proprietary/Non-Proprietary), March 8, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24072A267).
- 2. NRC, U. S. NRC Regulatory Audit Plan of Westinghouse Topical Report WCAP-18869-P/NP, Revision 0, High Performance Cladding for Use in Boiling Water Reactor Fuel, November 25, 2024 (ML24324A166).
- 3. NRC, U. S. NRC Report for Regulatory Audit of Westinghouse Topical Report WCAP 18869-P/NP, Revision 0, High Performance Cladding for Use in Boiling Water Reactor Fuel, April 30, 2025 (ML25108A038).
- 4. NRC, U. S. NRC Report for Regulatory Audit of Westinghouse Topical Report WCAP18869-P/NP, Revision 0, High Performance Cladding for Use in Boiling Water Reactor Fuel, April 30, 2025 (ML25108A038).
- 5. Letter from J. Ewing, Westinghouse, to NRC DCD, LTRNRC2523, Submittal of Responses to Requests for Additional Information on Westinghouse Topical Report WCAP-18869-P/NP, High Performance Cladding for Use in Boiling Water Reactor Fuel, April 21, 2025 (ML25112A272).
- 6. NRC, SRP, NUREG-0800, Section 4.2, Fuel System Design, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition - Reactor (NUREG-0800, Chapter 4), March 2007 (ML070740002).
- 7. NRC, SRP, NUREG-0800, Section 4.3, Nuclear Design, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition -
Reactor (NUREG-0800, Chapter 4), March 2007 (ML070740003).
- 8. NRC, SRP, NUREG-0800, Section 4.4, Thermal and Hydraulic Design, Revision 2, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition - Reactor (NUREG-0800, Chapter 4), March 2007 (ML070550060).
- 9. NRC, SRP, NUREG-0800, Chapter 15, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition - Transient and Accident Analysis (NUREG-0800, Chapter 15).
- 10. WCAP-15836-P-A, Revision 0, Fuel Rod Design Methods for Boiling Water Reactors -
Supplement 1, Volumes I & II, April 2006 (ML061220485 (Proprietary/Non-publicly available)).
- 11. CENPD-287-P-A, Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, July 1996 (ML19353D872 (Proprietary/Non-publicly available)).
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- 12. WCAP-15942-P-A, Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors Supplement 1 to CENP-287, March 31, 2006 (ML061110351 (Proprietary/Non-publicly available)).
- 13. Letter from K. Hosack, Westinghouse, to NRC DCD, Submittal of WCAP17769PA and WCAP-17769-NP-A, Revision 0, Reference Fuel Design SVEA-96 Optima3 (Proprietary/Non-Proprietary), LTR-NRC-20-38, May 27, 2020 (ML20150A239).
- 14. Standard Specification for Wrought Zirconium Alloy Seamless Tubes for Nuclear Reactor Fuel Cladding, ASTM B811, 2013.
- 15. C. Lemaignan, and A.T. Motta, Zirconium Alloys in Nuclear Applications, in Materials Science and Technology - A Comprehensive Treatment, Volume 10B, Nuclear Materials, Part II, Edited by R.W. Cahn, P. Haasen & E.J. Kramer, Volume Editor Brian R.T. Frost, VCH Verlagsgesellschaft mbH, Weinheim (Germany), VCH Publisher Inc.,
- 16. Waterside corrosion of zirconium alloys in nuclear power plants, IAEA Report, IAEATECDOC-996, ISSN 1011-4289, Vienna, Austria, January 1998.
- 17. F. Garzarolli, H. Stehle, and E. Steinberg, Behavior and Properties of Zircaloys in Power Reactors: A Short Review of Pertinent Aspects in LWR Fuel, Zirconium in the Nuclear Industry: Eleventh International Symposium, ASTM STP 1295, 1996, pp. 1232.
- 18. P. Tgtstrm, M. Limbck, M. Dahlbck, T. Andersson, and H. Pettersson, Effects of Hydrogen Pickup and Second-Phase Particle Dissolution on the In-Reactor Corrosion Performance of BWR Claddings, Zirconium in the Nuclear Industry: Thirteenth International Symposium, ASTM STP 1423, 2002, pp. 96118.
- 19. K. Kakiuchi, N. Itagaki, T. Furuya, A. Miyazaki, Y. Ishii, S. Suziki, T. Terai, and M. Yamawaki, Role of Iron for Hydrogen Absorption Mechanism in Zirconium Alloys, Zirconium in the Nuclear Industry: Fourteenth International Symposium, ASTM STP 1467, 2005, pp. 349366.
- 20. A.T. Motta, R.J. Comstock, A. Couet, Corrosion of Zirconium Alloys used for Nuclear Fuel Cladding, Annual Review of Materials Research, July 2015, pp. 311343.
- 21. Y. Takagawa, S. Ishimoto, Y. Etoh, T. Kubo, K. Ogata, and O. Kuboto, The Correlation Between Microstructures and in-BWR Corrosion Behavior of Highly Irradiated Zr-based Alloys, Zirconium in the Nuclear Industry: Fourteenth International Symposium, ASTM STP 1467, 2005, pp. 386403.
- 22. N. Itagak, et al., Development of New High Corrosion Resistance Zr Alloy HiFi for High Burnup Fuel, Proceedings European Nuclear Society TopFuel, Wüstzburg, 2003.
- 23. K. Ohira, et al., Recent Experience and Development of BWR Fuel at NFI, Proceedings of the 2005 Water Reactor Fuel Performance Meeting, Kyoto, 2005.
- 24. K. Kakiuchi, et al., Irradiated Behavior for BWR Advanced Zr Alloy (HiFi Alloy),
Proceedings of the 2005 Water Reactor Fuel Performance Meeting, Kyoto, 2005.
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- 25. K. Kakiuchi, et al., Irradiated Behavior at High Burnup for HiFi Alloy, Journal of Nuclear Science and Technology, Vol. 43, No. 9, 2006.
- 26. K. Kataoka, et al., The Irradiation Performance and Experience of Advanced Zirconium Alloy HiFi in a Commercial BWR, Proceedings European Nuclear Society TopFuel, Manchester, 2012.
- 27. M. Dahlbck, L. Hallstadius, M. Limbck, G. Vesterlund, T. Andersson, P. Witt, J. Izquierdo, B. Remartinez, M. Díaz, J. L. Sacedon, A.-M. Alvarez, U. Engman, R. Jakobsson, and A. R. Massih, The Effect of Liner Component Iron Content on Cladding Corrosion, Hydriding and PCI Resistance, Zirconium in the Nuclear Industry:
Fourteenth International Symposium, ASTM STP1467, 2004, pp. 873895.
- 28. F. Garzarolli, E. Steinberg, and H. Weidinger, Microstructure and Corrosion Studies for Optimized PWR and BWR Zircaloy Cladding, Zirconium in the Nuclear Industry: Eighth International Symposium, ASTM STP 1023, 1988, pp. 202212.
- 29. IAEA-TECDOC-1496, Thermophysical properties database of materials for light water reactors and heavy water reactors, IAEA, Wien, 2006, ISBN 92-0-104706-1.
- 30. NUREG/CR-0497: MATPRO v11, A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior, February 1979.
- 31. W.J. ODonnell and B.F. Langer, Fatigue Design Basis for Zircaloy Components, Nuclear Science and Engineering, Vol. 20, 1964, pp. 112.
Principal Contributors: R. Fu, NRR J. Dean, NRR Date: March 9, 2026 NRC Resolution to Westinghouse Proprietary Markings and Voluntary Comments By letter dated February 2, 2026 (ML26034C024), Westinghouse submitted the proprietary markup and voluntary comments. The NRC staff reviewed and incorporated Westinghouses proprietary markings requests (including Westinghouses requests for removal of some proprietary markings) and voluntary minor editorial comments into the final SE.