ML24346A165
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Response to SDAA Audit Question Question Number: A-4.2-5 Receipt Date: 11/20/2023 Question:
The stress intensity limits of Table 4-6 (TR-117605) Cladding Stress Intensity Limits do not appear to be consistent with the cladding acceptance criteria defined in ANP-10337 (the modeling and analysis methodology) or BAW-10227 (the M5 material topical report), but both of these reports are referenced as the source of the acceptance criteria. Instead of using the established acceptance criteria, the intent appears to be to define new acceptance criteria in Table 4-6, which appears to be inconsistent with previously approved and referenced documents. Provide the basis for NuScales new acceptance criteria and demonstrate that the new criteria is appropriate and the methodology for determining the criteria. Additionally, provide the nexus for the new criteria with respect to the existing documents and references. In addition to the request in the sentences above, provide the following related information:
a.
Table 4-6 is proposing to use ASME BPVC Service Level C limits for cladding in the faulted condition. What is the justification for this given that ASME BPVC limits do not assure operability? The previously defined acceptance criteria (in ANP-10337 and BAW-10227) prohibit plastic deformation of the cladding, which has broader implications beyond the safety topic of structural analysis.
b.
Table 4-6 is proposing acceptance criteria that allow the cladding to experience plastic (permanent) deformation.
i.
ANP-10337 limits the cladding to 90% of yield, which eliminates any concern of permanent deformation in the analysis methodology. What is the justification for using acceptance criteria that are higher than those defined in ANP-10337?
ii.
What other safety basis analyses are affected by allowing cladding to experience permanent deformation? It is common to assume that fuel rods will remain elastic during an SSE. Which NuScale safety basis analyses account for permanently deformed fuel rods?
c.
BAW-10337 defines stress acceptance criteria for M5 cladding that are lower than the values listed in Table 4-9. Where do the fuel rod stress limits (stress values in parentheses) in NuScale Nonproprietary NuScale Nonproprietary
Table 4-9 come from? Provide the source of the stress limit values for M5 and any derivation performed to calculate the values.
Response
The original response to this audit question was posted on December 20, 2023. The audit response was supplemented on May 9, 2024 and the content in this response starts with the section titled Response Supplement - May 2024. The most recent supplement begins with the section titled Response Supplement - July 2024.
The stress intensity limits of Table 4-6 of TR-117605 are consistent with Framatome Topical Report BAW-10227P-A, Revision 2.0, and were reviewed and approved by the NRC for use on PWR fuel. Section 7 of FS1-0064723 shows results consistent with Table 4-6 of TR-117605, but also shows that positive margins were obtained with the previous M5 cladding stress intensity limits from BAW-10227P-A, Revision 1.0. FS1-0064723 is provided in the eRR as part of the response to A-4.2-4. The additional questions from A-4.2-5 are included below.
a.
The fuel rod related criterion from NUREG-0800, SRP 4.2 Appendix A is that fuel rod fracturing shall not occur. The method for ensuring this criterion is met is to apply the ASME Code Service Level C limits based on the classifications in NUREG-0800, Section 3.9.3, Revision 3, Table 1. The more restrictive limits of Service Level C are applied to Faulted conditions. Functional capability is also addressed by applying the limits of Service Level C as described below.
The ASME Code Service Level C limits are discussed in Section E.2 of BAW-10337PA Appendix E where it is stated that The Level C service limits are intended to protect a component against loss of control over deformation, while allowing for a limited amount of plastic deformation. Appendix E goes on to describe the concept of limit stress when based on elastic analysis. In the case described, the limit stress exceeds the yield stress by a factor which depends on the section shape. For the fuel rod cladding annular cross-section, the shape factor is a function of the inner and outer diameters as given by equation E-1. Typically, the shape factor for fuel rod cladding results in a multiplier on Sm greater than 1.8. The more restrictive Service Level C Sm multiplier of 1.8 was chosen for conservatism and simplicity.
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b.
Framatome Topical Report BAW-10337PA does not provide the acceptance criteria for fuel rods, see the SER limitation and condition #6 for BAW-10337PA. See above for discussion of the new limits from BAW-10227PA, Revision 2.
The new limits are applied only to LOCA/SSE. Note that the new limits are based on the more restrictive limits of Service Level C (Emergency) while SSE and LOCA are classified as Faulted (Level D) in NUREG-0800, SRP 3.9.3, Rev. 1.
c.
The values in TR-117605 Table 4-9 are based on the Faulted limit definitions in Table 4-
- 6. The values are determined using the M5 cladding material equation for bi-axial hoop yield strength in Table K-1.1 from BAW-10227PA, Revision 1 at the operating temperature of interest (752 degF).
Response Supplement - May 2024 The original supplemented response was provided on May 9, 2024. The cladding stress intensity limits adopted in TR-117605-P are derived from ASME BPVC Section III Table XIII-3110-1 and Article XIII-3420 as described in Section 4.3.5 of the technical report. These same limits were approved in BAW-10227P-A Revision 2.
NUREG-800 Standard Review Plan (SRP) for Section 4.2 acceptance criteria, specifically criterion 1.A.i., the states the following with emphasis in bold:
Stress, strain, or loading limits for spacer grids, guide tubes, thimbles, fuel rods, control rods, channel boxes, and other fuel system structural members should be provided.
Stress limits that are obtained by methods similar to those given in Section III of the Boiler and Pressure Vessel Code of the American Society of Mechanical Engineers (ASME) are acceptable. Other proposed limits must be justified.
The cladding stress intensity limits that are evaluated against in the technical report were obtained from ASME Section III as noted in Section 4.3.5 and satisfies the statement above.
For seismic and LOCA response loads, SRP Section 4.2 Appendix A.III.2. states the following regarding the determination of strength for components other than grids with emphasis in bold:
Strengths of fuel assembly components other than spacer grids may be deduced from fundamental material properties or experimentation. Supporting evidence for strength values should be supplied. Since structural failure of these components (e.g., fracturing NuScale Nonproprietary NuScale Nonproprietary
of guide tubes or fragmentation of fuel rods) could be more serious than grid deformation, allowable values should bound a large percentage (about 95 percent) of the distribution of component strengths. Therefore, ASME Code values and procedures may be used when appropriate for determining yield and ultimate strengths. Specification of allowable values may follow the ASME Code requirements and should consider buckling and fatigue effects.
Further, in SRP Section 4.2 Appendix A.IV.1., the following is stated as acceptance criteria for externally applied forces with emphasis added:
Two principal criteria apply for the LOCA(1) fuel rod fragmentation must not occur as a direct result of the blowdown loads and (2) the 10 CFR 50.46 temperature and oxidation limits must not be exceeded. The first criterion is satisfied if the combined loads on the fuel rods and components other than grids remain below the allowable values defined above. The second criterion is satisfied by an ECCS analysis. If combined loads on the grids remain below P(crit), as defined above, then no significant distortion of the fuel assembly would occur and the usual ECCS analysis is sufficient. If combined grid loads exceed P(crit), then grid deformation must be assumed and the ECCS analysis must include the effects of distorted fuel assemblies.
An assumption of maximum credible deformation (i.e., fully collapsed grids) may be made unless other assumptions are justified.
The paragraph above states that in order to satisfy prevention of fuel rod fracture, evaluation of the combined loads on the fuel rods must remain below the allowable values defined in Appendix A.III.2.
Finally, in SRP Section 4.2 Appendix A.IV.1., the following is stated with emphasis in bold:
Control rod insertability is a third criterion that must be satisfied. Loads from the worst-case LOCA that requires control rod insertion must be combined with the SSE loads, and control rod insertability must be demonstrated for that combined load. For a PWR, if combined loads on the grids remain below P(crit), as defined above, then significant deformation of the fuel assembly would not occur and lateral displacement of the guide tubes would not interfere with control rod insertion. If combined loads on the grids exceed P(crit), then additional analysis is needed to show that the deformation is not severe enough to prevent control rod insertion.
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Thus, while the faulted limits employed in TR-117605-P Table 4-6 allow for small plastic deformation of the fuel cladding, deformations do not need to be quantified or accounted for in ECCS evaluations because the combined loads on the spacer grids remain below the approved limit as reported in TR-117605-P Table 4-8 and therefore significant deformation of the fuel assembly does not occur.
The fuel in the NuScale design maintains core coolability by precluding cladding rupture and grid deformation as shown in the results presented in TR-117605-P. Quantification of plastic deformation of the fuel rods is not required to assure core coolability is maintained. The margins to grid load limits in TR-117605-P Table 4-8 indicate significant control over grid deformation.
The grid load limits in TR-117605-P Table 4-8 are defined [
] As noted in Section 15 of ANP-10337P-A, Rev. 0, [
] NuScales safety analyses are not impacted by the use of Level C Service limits for cladding stress intensity limits for faulted conditions.
However, separately, NuScales safety analyses account for fuel rod bowing effects on core neutronics and thermal-hydraulics in all critical heat flux and peak linear heat rate calculations.
Additional feedback was received on April 26, 2024. The feedback states that the third set of acceptance criteria, the criteria that is applied to the NuScale US460 design, is new. However, the applied stress criteria in Table 4-6 of TR-117605-P is equivalent to the ASME BPVC stress criteria from Table XIII-3110-1 and Article XIII-3420 noted in column 3.
Response Supplement - July 2024 Based on feedback received on June 3, 2024, TR-117605-P is revised to adopt BAW-10227P-A Revision 2 for fuel faulted fuel stress intensity limits and to add a requirement in TR-117605-P Section 4.3.5.2.3 to perform analyses of the fuel if these limits are reached.
This response references a proprietary version of a topical report that is marked as containing export controlled information (ECI). However, the extracted pages of the topical report that is attached to this response does not contain ECI as submitted herein. Notwithstanding, any proprietary information included in the response and the attachment hereto shall be withheld per 10 CFR 2.390.
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NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Draft Revision 1
© Copyright 2024 by NuScale Power, LLC 4
1.0 Introduction NuScale Power, LLC (NuScale) has developed a 17x17 fuel assembly, NuFuel-HTP2',
and control rod assembly (CRA) design for use in the NuScale Power Module (NPM) based on existing Framatome methods and technology.
1.1 Purpose This report describes the NuFuel-HTP2' fuel assembly and CRA designs, summarizes the key analysis results, and demonstrates the designs meet regulatory requirements.
1.2 Scope This report uses Framatome methodologies to evaluate the NuFuel-HTP2' fuel design against the requirements of NUREG-0800 Section 4.2 (Reference 9.1.1). The evaluations are focused on the mechanical aspects of the design. Neutronic and thermal-hydraulic performance of the fuel assembly are addressed in the NuScale US460 Final Safety Analysis Report.
This report also summarizes the CRA design and analyses.
Chapter 2 provides the regulatory framework that identifies the requirements against which the fuel design is evaluated.
Chapter 3 describes the mechanical design and function of the fuel assembly and provides a brief description of applicable operating experience.
Chapter 4 summarizes the results of the fuel assembly design evaluations using NRC-approved methods.
Audit Question A-4.2-5
The NuFuel-HTP2' fuel mechanical design is evaluated using generic mechanical design criteria for pressurized water reactor (PWR) fuel designs (Reference 9.1.2, and Reference 9.1.3, and Reference 9.1.14).
The COPERNIC Fuel Rod Design Computer Code (Reference 9.1.4) is used in the fuel performance analysis of the NuFuel-HTP2' fuel design.
The CROV code (Reference 9.1.5) is used to evaluate the cladding creep performance.
The computational procedure for evaluating fuel rod bowing (Reference 9.1.6) is used to evaluate the impacts of fuel rod bow for the NuFuel-HTP2' design.
The PWR fuel assembly structural response to externally applied dynamic excitations (Reference 9.1.7) is used to perform seismic structural analysis on the fuel design.
The applicability of the fuel analysis methods is demonstrated in Applicability of AREVA Fuel Methodology for the NuScale Design, TR-0116-20825-P-A Revision 1 (Reference 9.1.8), NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces, TR-0716-50351-P-A
NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Draft Revision 1
© Copyright 2024 by NuScale Power, LLC 51 4.3.3 Generalized Cladding Melting As stated in NUREG-0800 Section 4.2, criteria for cladding embrittlement in Section 4.3.1 are more stringent than generalized cladding melting criteria. Therefore, additional specific criteria are not used.
4.3.4 Fuel Rod Ballooning Burst strain and flow blockage caused by ballooning of the cladding relates to the ECCS performance evaluation and is not addressed in this report.
4.3.5 Fuel Assembly Structural Damage from External Forces Design Criterion The fuel assembly shall withstand the loads from an SSE and a LOCA. Specific acceptance criteria for fuel assembly components are identified in Reference 9.1.7.
NuFuel-HTP2' Design Evaluation The external load analysis is performed using the methodology of Reference 9.1.7 as modified by Reference 9.1.10. The analysis demonstrates positive margins to all applicable criteria.
Cladding Faulted Audit Question A-4.2-5 The faulted limits applied to the fuel rod cladding are shown in the far right column of Table 4-6. The table also shows the limits used in the previous faulted analysis that were based on Table 3-3 of the Framatome M5 Topical report (Reference 9.1.3).
The middle column of values is derived from Table XIII-3110-1 and Article XIII-3420 of Reference 9.1.13. They are simplified for application to the cladding allowable stresses. The applied M5 cladding stress intensity limits are adopted from Revision 2 of the Framatome M5 Topical report (Reference 9.1.14).
NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Draft Revision 1
© Copyright 2024 by NuScale Power, LLC 52 Audit Question A-4.2-5 Audit Question A-4.2-5
[
]
Audit Question A-4.2-5 Strengths of fuel assembly components other than spacer grids may be deduced from fundamental material properties or experimentation. Supporting evidence for strength values should be supplied. Since structural failure of these components (e.g.,
fracturing of guide tubes or fragmentation of fuel rods) could be more serious than grid deformation, allowable values should bound a large percentage (about 95 percent) of the distribution of component strengths. Therefore, ASME Code values and procedures may be used when appropriate for determining yield and ultimate strengths. Specification of allowable values may follow the ASME Code requirements and should consider buckling and fatigue effects.
4.3.5.1 Analysis Inputs 4.3.5.1.1 Lateral Model The lateral fuel assembly model is developed using the model parameters from Section 6.1 of Reference 9.1.7. Damping coefficients specific to the NuFuel-HTP2' design are summarized in Table 4-7 and defined in Reference 9.1.9. The analysis uses values that are more conservative than those listed in Table 4-7.
Table 4-6 M5 Cladding Stress Intensity Limits
[
]
NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Draft Revision 1
© Copyright 2024 by NuScale Power, LLC 57 loads are combined, along with steady-state normal operating loads, for each component for comparison to its respective acceptance criteria.
As defined in Section 8.1.2 of Reference 9.1.7, [
Audit Question A-4.2-5 Audit Question A-4.2-5
]
The results of the component evaluations are presented in Table 4-9.
Fuel rods The accident loads on the fuel rod consist of the lateral bending stress, the axial stress from vertical loads, and an additional loading corresponding to impacts to the fuel assembly. The impact induced stress is a localized bending stress resulting from the portion of the grid impacts that are passed through the fuel rods (i.e., internal impact loads). The overall combined stress is the SRSS of the individual components combined with steady-state stresses to determine the maximum stress intensity.
The faulted stress intensity limits in Table 4-6 are based on Service Level C limits defined in Reference 9.1.13. As such, [
]
If the calculated maximum stress intensity is [
]
The fuel rods are also evaluated for buckling under compressive loads.
Guide tubes The accident loads on the guide tube consist of [
] The overall combined stress is calculated as the SRSS of the individual components combined with steady-state stresses
NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Draft Revision 1
© Copyright 2024 by NuScale Power, LLC 80 7.0 Design Change Process The following defines the mechanical design change process for the NuFuel-HTP2' fuel design. The design change process is modeled after the NRC-approved change process defined in Reference 9.1.2 and is in conformance to the NuScale design control requirements and 10 CFR 50 Appendix B Quality Assurance Program.
The fuel design criteria for the NuFuel-HTP2' design are distributed across several sources. Reference 9.1.8, Table 7-1 identifies the mechanical design criteria and source of applicability to the NuFuel-HTP2' design. The design criteria sources are Audit Question A-4.2-5
Framatome methodologies that are applicable to the NuFuel-HTP2' design provided in Reference 9.1.4, Reference 9.1.5, Reference 9.1.6, Reference 9.1.7, Reference 9.1.2, and Reference 9.1.3, and Reference 9.1.14.
NuScale methodologies provided in Reference 9.1.8, Reference 9.1.9, and Reference 9.1.10.
the remaining criteria that are evaluated by the methodologies as approved in the FSAR.
Applicable design criteria are summarized in Section 4.0. Compliance to these criteria is demonstrated by
documenting the fuel system and fuel assembly design drawings.
performing analyses with the NRC-approved models and methods described in Section 4.0.
using lead test assemblies, prototypic testing and engineering analyses, where appropriate, to demonstrate that analytical methods and acceptance criteria remain applicable and to demonstrate in-reactor performance.
continuing irradiation surveillance programs, including post irradiation examinations, to confirm fuel assembly performance.
using the quality assurance procedures, quality control inspection program, and design control requirements set forth in the NRC-approved NuScale Quality Assurance Program.
Design changes can be made without NRC review and approval if the following conditions are met:
demonstrating compliance with the approved criteria as defined above
changes in plant technical specifications are not required
the applicability of NRC-approved methodologies as described in Section 4.0 is demonstrated to be valid
burnup limits are within those approved by the NRC
NuFuel-HTP2' Fuel and Control Rod Assembly Designs TR-117605-NP Draft Revision 1
© Copyright 2024 by NuScale Power, LLC 83 9.0 References 9.1 Source Documents 9.1.1 NUREG-0800, U.S. NRC Standard Review Plan Section 4.2, Revision 3, Fuel System Design, March 2007.
9.1.2 EMF-92-116(P)(A), Rev. 0, Generic Mechanical Design Criteria for PWR Fuel Designs, February 2015.
9.1.3 BAW-10227P-A, Rev. 1, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, June 2003.
9.1.4 ANP-10231PA-01, COPERNIC Fuel Rod Design Computer Code, January 2004.
9.1.5 ANP-10084PA-03, Program to Determine In-Reactor Performance of B&W Fuels
- Cladding Creep Collapse (CROV computer code), October 1980.
9.1.6 XN-75-32 (P) (A), Supplements 1-4, Computational Procedure for Evaluating Fuel Rod Bowing, February 1983.
9.1.7 ANP-10337P-A, Rev. 0, PWR Fuel Assembly Structural Response to Externally Applied Dynamic Excitations, April 2018.
9.1.8 TR-0116-20825-P-A, Applicability of AREVA Fuel Methodology for the NuScale Design, June 2016, Revision 1.
9.1.9 TR-0716-50351-P, NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to Externally Applied Forces, December 2019, Revision 1.
9.1.10 TR-108553-P-A, Framatome Fuel and Structural Response Methodologies Applicability to NuScale, Supplement 1 to TR-0116-20825-P-A, Revision 1, Supplement 1 to TR-0716-50351-P-A, Revision 1, October 2022, Revision 0.
9.1.11 ASME Boiler and Pressure Vessel Code,Section III, Division 1 - Subsection NG, Core Support Structures, 2010 Edition with 2011a Addenda, July 1, 2011.
9.1.12 ODonnell, W.J. and B.F. Langer, Fatigue Design Basis for Zircaloy Components, Nuclear Science and Engineering, Volume 20, pp. 1-12, September 1964.
9.1.13 ASME Boiler and Pressure Vessel Code,Section III, Division 1, 2019 Edition.
Audit Question A-4.2-5 9.1.14 BAW-10227P-A, Rev. 2, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, January 2023.