ML24215A108
| ML24215A108 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 08/02/2024 |
| From: | NuScale |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24215A000 | List:
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| References | |
| LO-169995 | |
| Download: ML24215A108 (1) | |
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Response to SDAA Audit Question Question Number: A-5.3.1.5-1 Receipt Date: 10/20/2023 Question:
SDAA Section 5.3.1.5 provides some discussion and reference to associated exemption requests in the application. These exemptions present how certain aspects of regulations do not apply to the subject design due to specific design elements. Normally compliance with the subject regulations provides a substantial portion of the justification that a design is compliant with the associated general design criteria (GDCs).
SDAA Section 5.3.1.5 does not provide a clear description of how the associated GDCs are met by the proposed design. The staff cannot make a finding without this information in the SDAA.
Discussion is necessary regarding how the RV materials were assessed to support compliance with GDCs 14, 15, and 31. In particular, discussion of how the upper RV was analyzed under the spirit of 10 CFR Part 50, Appendix G is missing, though this is documented to some extent in the associated P-T Limits methodology. The 10 CFR Part 50 Appendix G applies to all ferritic materials for pressure-retaining components of the reactor coolant pressure boundary regardless of whether the subject materials are in the beltline or subject to neutron radiation effects of consequence to the function of the components.
Similarly, address the lack of description regarding meeting the associated GDCs as it relates to the associated Part 7 Exemption.
Follow-up question received on January 19, 2024:
The staff stated that the SDAA Section 5.3.1.5 does not provide a verifiable description of how the GDCs are met and specifically that the SDAA needs a discussion of how the upper RV was analyzed under Part 50 App G. Other than stating in the response that the upper RPV complies with the requirements of Part 50 App G and that App G requires qualification by methods equivalent to ASME Section XI, Appendix G, the response did not provide any revisions to the SDAA to describe how the GDCs are met with respect to the upper RPV. Neither Appendix G, nor the GDCs are limited to the lower head, they apply to the entire module. Please provide a NuScale Nonproprietary NuScale Nonproprietary
response that revises the SDAA to addresses the acceptability of the entire module. The following GDCs apply:
GDC 1 and 30, as they relate to quality standards for design, fabrication, erection, and testing of structures, systems, and components.
GDC 4, as it relates to the compatibility of components with environmental conditions.
GDC 14, as it relates to prevention of rapidly propagating fractures of the reactor coolant pressure boundary (RCPB).
GDC 31, as it relates to material fracture toughness.
Follow-up question received on January 19, 2024:
Section 5.2.3, third bullet should be updated to be consistent with proposed SDAA changes in the audit response (e.g. not solely referencing 5.2-10 which is focused on the lower RPV).
The referenced paragraph is below:
Appendix G to 10 CFR 50. The RPV ferritic pressure retaining and integrally attached materials meet applicable fracture toughness acceptance criteria (Reference 5.2-10). The design supports an exemption from the requirements of 10 CFR 50.60 which invokes compliance with 10 CFR 50, Appendix G. Section 5.3.1.6 provides further details.
Response
The Standard Design Approval Application (SDAA) Section 5.3.1.5 states that the reactor pressure vessel (RPV) design prevents non-ductile fracture in accordance with General Design Criterion (GDC) 14, GDC 15, and GDC 31.
Appendix G of 10 CFR Part 50 is met by upper RPV compliance with its requirements. As stated in Appendix G, the ASME Code forms the basis for the requirements of this appendix.
Regulation 10 CFR 50, Appendix G, requires qualification by methods equivalent to those described in American Society of Mechanical Engineers (ASME) Section Section XI, Appendix G, for ferritic materials in the reactor coolant pressure boundary (RCPB). In conformance with 10 CFR 50 Appendix G, the RPV is designed to meet the requirements of ASME Section III, Subsection NB, paragraph NB-3210. The qualification requirements in ASME NuScale Nonproprietary NuScale Nonproprietary
Section III, Appendix G, are equivalent to ASME Section XI, Appendix G (as noted in the audit response to A-6.2.7-1). The qualification of the upper RPV is verified during inspections, tests, analyses, and acceptance criteria (ITAAC) 02.01.01 (SDAA Part 8, Section 2.1), which ensures the ASME design reports meet the requirements of ASME Section III. This qualification plan is unchanged from the US600 design.
Section 5.3.1.1 of SDAA addresses material compliance with GDC 1 and 30, GDC 4, GDC 14 and 31, GDC 15, and GDC 32. Section 5.3.1.5 of SDAA addresses compliance with GDC 14, GDC 15, and GDC 31 for prevention of non-ductile fracture. There is an update to SDAA Section 5.3.1.5 attached to this response.
Response to the follow-up question received on January 19, 2024:
The attached markup adds Technical report TR-130877-P, Revision 0, Pressure and Temperature Limits Methodology, to Section 5.2.3 as Reference 5.2-11. This change ensures consistency with other changes in Chapter 5 related to this audit question and audit questions A-5.3.2.3-1 and A-5.3.2.4-1. In addition, this markup corrects the section number cited at the end of the quoted paragraph.
Markups of the affected changes, as described in the response, are provided below:
NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Material Referenced NuScale US460 SDAA 1.6-3 Draft Revision 2 Audit Question 5.3.1.5-1, Audit Question A-5.3.2.3-1, Audit Question A-5.3.2.4-1 Table 1.6-2: NuScale Referenced Technical Reports Report Number Title FSAR Section TR-118318, Revision 1 NuScale Design of Physical Security Systems 9.5, 13.6, 14.2 TR-122844-P, Revision 0 NuScale Instrument Setpoint Methodology Technical Report 7.0, 7.2 TR-121353-P, Revision 0 NuScale Comprehensive Vibration Assessment Program Analysis Technical Report 3.9, 14.2 TR-121515-P, Revision 0 NuScale Power Module Seismic Analysis 3.7, 3.12 TR-130877-P, Revision 0 Pressure and Temperature Limits Methodology 5.2, 5.3 TR-121517-P, Revision 0 NuScale Power Module Short-Term Transient Analysis 3.9 TR-130721-P, Revision 0 Use of Austenitic Stainless Steel for US460 Standard Design Reactor Pressure Vessel 5.2, 5.3 TR-121354-P, Revision 0 NuScale Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report 3.9, 14.2 TR-121507-P, Revision 0 Pipe Rupture Hazards Analysis 3.6
NuScale Final Safety Analysis Report Integrity of Reactor Coolant Boundary NuScale US460 SDAA 5.2-12 Draft Revision 2
GDC 14 and 31. Design and fabrication of the RPV and pressure retaining components associated with the RCPB ensure sufficient margin such that the RCPB behaves in a non-brittle manner and minimizes the probability of rapidly propagating fracture and gross rupture of the RCPB (Reference 5.2-10).
Criterion XIII of 10 CFR 50, Appendix B. The Quality Assurance Program requires procedures for the control of the on-site cleaning of RPV and the RCPB during construction.
Audit Question 5.3.1.5-1, Audit Question A-5.3.2.3-1, Audit Question A-5.3.2.4-1 Appendix G to 10 CFR 50. The RPV ferritic pressure retaining and integrally attached materials meet applicable fracture toughness acceptance criteria (Reference 5.2-10 and Reference 5.2-11). The design supports an exemption from the requirements of 10 CFR 50.60 which invokes compliance with 10 CFR 50, Appendix G. Section 5.3.1.6Section 5.3.1.5 provides further details.
5.2.3.1 Material Specifications The materials for the Class 1 components and supports that comprise the RCPB, including the RPV and SGs, are in Table 5.2-3. Table 5.2-3 also includes materials and specifications associated with the RPV attachments and appurtenances. The table lists the grade or type, as applicable, of the ferritic low alloy steels, austenitic stainless steels, and nickel-based alloys specified for the RCPB. Except where noted in Table 5.2-3, the associated ASME BPVC material specification provides the final metallurgical condition. Further discussion of the materials associated with the RPV is in Section 5.3, Reactor Vessel.
The RCPB surface materials in contact with reactor coolant or in contact with pool water during refueling, including welds, are corrosion resistant alloys or clad with austenitic stainless steel or nickel-based alloy. The SG tubesheet bores are the exception, the SA-508 tubesheet bores do not have corrosion resistant cladded surface. The SG tubes expand into the tubesheet bore to provide corrosion protection in the crevice between the SG tube and tubesheet.
Processing and welding of unstablized American Iron and Steel Institute Type 3XX series austenitic stainless steels for pressure retaining components comply with RG 1.44 to prevent sensitization caused by chromium depletion at the grain boundaries during welding and heat treatment operations. For unstablized American Iron and Steel Institute Type 3XX series austenitic stainless steel subjected to sensitizing temperatures subsequent to solution heat treatment, the carbon content is no more than 0.03 weight percent.
Processing and welding of American Iron and Steel Institute Type 2XX series austenitic stainless steels for pressure retaining components comply with ASME BPVC paragraph NB-2433 and RG 1.31 for delta ferrite composition. The ferrite number are in the range of 5 FN to 16 FN. The carbon content of the weld filler materials is restricted to 0.04 percent maximum.
Nickel-based Alloy 690 is a base metal in RCPB components and structures along with Alloy 52/152 cladding and weld metals and similar alloys developed for improved weldability. Alloy 690 and 52/152 have a high resistance to general
NuScale Final Safety Analysis Report Integrity of Reactor Coolant Boundary NuScale US460 SDAA 5.2-28 Draft Revision 2 5.2.6 References 5.2-1 NuScale Power, LLC, Non-Loss-of-Coolant Accident Analysis Methodology, TR-0516-49416-P, Revision 4.
5.2-2 NuScale Power, LLC, Loss-of-Coolant Accident Evaluation Model, TR-0516-49422-P, Revision 3.
5.2-3 Electric Power Research Institute, Pressurized Water Reactor Primary Water Chemistry Guidelines, EPRI #3002000505, Revision 7, Palo Alto, CA, 2014.
5.2-4 American Society of Mechanical Engineers, Quality Assurance Requirements for Nuclear Facility Applications, ASME NQA-1-2015, New York, NY.
5.2-5 Electric Power Research Institute, Materials Reliability Program:
Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111), EPRI #1009801, Palo Alto, CA, 2004.
5.2-6 Electric Power Research Institute, Materials Reliability Program:
Resistance to Primary Water Stress Corrosion Cracking of Alloy 690 in Pressurized Water Reactors (MRP-258), EPRI #1019086, Palo Alto, CA, 2009.
5.2-7 Electric Power Research Institute, Steam Generator Management Program: Pressurized Water Reactor Steam Generator Examination Guidelines, EPRI #1013706, Revision 7, Palo Alto, CA, 2007.
5.2-8 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2017 Edition,Section III, Division 1, Rules for Construction of Nuclear Facility Components, New York, NY.
5.2-9 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2017 Edition,Section XI, Division 1, Rules for Inservice Inspection of Nuclear Power Plant Components, New York, NY.
5.2-10 NuScale Power, LLC, Use of Austenitic Stainless Steel for NPM Reactor Pressure Vessel, TR-130721-P, Revision 0.
Audit Question 5.3.1.5-1, Audit Question A-5.3.2.3-1, Audit Question A-5.3.2.4-1 5.2-11 NuScale Power, LLC, Pressure and Temperature Limits Methodology, TR-130877-P, Revision 0.
NuScale Final Safety Analysis Report Reactor Vessel NuScale US460 SDAA 5.3-4 Draft Revision 2 Section 4.5.1, Control Rod Drive - Materials Specifications, addresses the use of cold-worked austenitic stainless steel.
5.3.1.5 Fracture Toughness Audit Question 5.3.1.5-1, Audit Question A-5.3.2.3-1, Audit Question A-5.3.2.4-1 The RPV design prevents non-ductile fracture in accordance with GDC 14, GDC 15, and GDC 31. However, the design supports an exemption from the requirements of 10 CFR 50.60, which includes an exemption from the requirements of 10 CFR 50, Appendix G for the lower RPV. The materials used for the lower RPV are not applicablesubject to the fracture toughness analyses required by 10 CFR 50, Appendix G, and the upper RPV does not meet the neutron fluence levels to be assessed for the effects of neutron embrittlement.
Audit Question 5.3.1.5-1, Audit Question A-5.3.2.3-1, Audit Question A-5.3.2.4-1 10 CFR 50, Appendix G, requirements apply to ferritic materials of pressure-retaining components of the RCPB of light water nuclear power reactors.
The NPM uses austenitic stainless steel materials (Table 5.2-3) in the lower RPV shell. The requirements of 10 CFR 50, Appendix G, rely on impact testing data performed in accordance with ASME BPVC Section III, Paragraph NB-2331.
However,Paragraph NB-2331 follows NB-2311, which does not require impact testing of austenitic stainless steel.
Audit Question 5.3.1.5-1, Audit Question A-5.3.2.3-1, Audit Question A-5.3.2.4-1 The fluence values for the upper RPV shell do not exceed 1.0E+17 n/cm2 (E > 1 MeV), which is the peak neutron fluence at the end of the design life of a reactor vessel that requires an assessment of the effects of neutron embrittlement as specified in 10 CFR 50, Appendix H. Therefore, the upper RPV, which is made of ferritic steel, does not require an assessment for the effects of neutron embrittlement and, there are no considerations of adjustments for embrittlement necessary using the RG 1.99 methodology.
Audit Question 5.3.1.5-1, Audit Question A-5.3.2.3-1, Audit Question A-5.3.2.4-1 The neutron flux and fluence calculation methods follow the guidance of RG 1.190 with exceptions as described in the NuScale Technical Report "Fluence Calculation Methodology and Results" (Reference 5.3-7). Reference 5.3-7 provides further details regarding the fracture toughness capabilities of the austenitic stainless steel material used in the lower RPV. Reference 5.3-6 provides the methodology used for derivation of the pressure-temperature limits for the RPV, which ensures that the upper RPV meets GDC 14, GDC 15, GDC 31, and 10 CFR 50 Appendix G.
5.3.1.6 Material Surveillance 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, applies to ferritic materials in the reactor vessel beltline region of light water nuclear power reactors. The RPV design prevents non-ductile fracture in accordance with GDC 14, GDC 15, and GDC 31. However, the design supports
NuScale Final Safety Analysis Report Reactor Vessel NuScale US460 SDAA 5.3-6 Draft Revision 2 The welding of the stud to the cladding requires a cladding preservice liquid penetrant exam, per ASME BPVC Section III, paragraph NB-5272, Weld Metal Cladding. The welding of the stud to the cladding also complies with ASME BPVC Section III, paragraph NB-4435, Welding of Nonstructural Attachments.
There are no inservice exam requirements for the lock plate stud welds or the lock plates.
5.3.2 Pressure-Temperature Limits, Pressurized Thermal Shock, and Charpy Upper-Shelf Energy Data and Analyses Audit Question 5.3.1.5-1, Audit Question A-5.3.2.3-1, Audit Question A-5.3.2.4-1 The information in this section describes the bases for setting operational limits on pressure and temperature for the RCPB. The RPV design prevents non-ductile fracture in accordance with GDC 14, GDC 15, and GDC 31. The design supports an exemption from the requirements of 10 CFR 50.60, which includes an exemption from the requirements of 10 CFR 50, Appendix G, and 10 CFR 50, Appendix H, for the lower RPV. The design supports an exemption from the requirements of 10 CFR 50.61. Reference 5.3-9 provides further details regarding austenitic stainless steel used in the lower RPV, which is resistant to the effects of neutron and thermal embrittlement.
5.3.2.1 Limit Curves The calculation of a generic set of pressure-temperature limits at 57 EFPY uses the methodology provided in ASME BPVC Section XI, Appendix G, and the applicable limits provided in 10 CFR 50, Appendix G, as described below.
Consideration of only the initial RTNDT temperature is necessary because the lower portion of the RPV is not a ferritic material, and the peak fluence for the upper portion of the RPV shell is less than the 10 CFR 50, Appendix H, criteria (1.0E+17 n/cm2(E > 1 MeV)). Therefore, no adjustment is necessary to account for fluence embrittlement effects (Reference 5.3-5). For conservatism, the 10 CFR 50, Appendix G, Table 1, limits have been applied to the final pressure-temperature limits.
The pressure-temperature limits for normal heatup and criticality conditions, normal cooldown, and inservice leak and hydrostatic (ISLH) tests including transient conditions are in Figure 5.3-2, Figure 5.3-3, and Figure 5.3-4, respectively. The corresponding numerical values are in Table 5.3-2 and Table 5.3-3. RCS pressure maintained below the limit of the pressure-temperature limit curves ensures protection against non-ductile failure.
Acceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the applicable pressure-temperature curves. These pressure-temperature curves include neither location correction nor instrument uncertainty. For the purpose of location correction, the allowable pressure in the pressure-temperature curves is the pressure at the RPV bottom. The reactor is not permitted to be critical until the pressure-temperature combinations are to the right of the criticality curve shown in Figure 5.3-2.
NuScale Final Safety Analysis Report Reactor Vessel NuScale US460 SDAA 5.3-7 Draft Revision 2 Further information on the methodology used to develop the limits is in the NuScale Technical Report, "Pressure and Temperature Limits Methodology" (Reference 5.3-6).
5.3.2.2 Operating Procedures Section 13.5, Plant Procedures, addresses development of plant operating procedures that ensure pressure-temperature limit compliance. These procedures ensure compliance with the technical specifications during normal power operating conditions and anticipated transients.
COL Item 5.3-1:
An applicant that references the NuScale Power Plant US460 standard design will develop operating procedures to ensure that transients will not be more severe than those for which the reactor design adequacy had been demonstrated. These procedures will be based on material properties of the as-built reactor vessels.
5.3.2.3 Pressurized Thermal Shock Audit Question 5.3.1.5-1, Audit Question A-5.3.2.3-1, Audit Question A-5.3.2.4-1 The RPV design prevents non-ductile fracture in accordance with GDC 14, GDC 15, and GDC 31. The design supports an exemption from the requirements of 10 CFR 50.61 due to the use of austenitic stainless steel in the lower RPV. The methodology described in 10 CFR 50.61 determines RTPTS, which is the RTNDT evaluated for the end of design life peak fluence for each beltline material.
Because the lower RPV material is austenitic stainless steel, this material is exempt from impact test requirements per ASME BPVC Section III, NB-2311. As a result, the PTS screening methodology in 10 CFR 50.61 is not applicable to RPV beltline materials. Further, 10 CFR 50.61 is not applicable to the upper RPV shell. The end of design life fluence value for the upper RPV shell does not exceed 1.0E+17 n/cm2 (E > 1 MeV). Therefore, 10 CFR 50.61 is not applicable to this area of the RPV.This fluence means that the entire upper RPV shell is outside the RPV beltline region per 10 CFR 50.61. Therefore, 10 CFR 50.61 PTS screening is not required for the upper RPV shell. Reference 5.3-7 provides further details regarding the effects of neutron and thermal embrittlement on the austenitic stainless steel material used in the lower RPV. The requirements of 10 CFR 50.61 are addressed in the NuScale Technical Report, Pressure and Temperature Limits Methodology (Reference 5.3-6).
5.3.2.4 Upper-Shelf Energy Audit Question 5.3.1.5-1, Audit Question A-5.3.2.3-1, Audit Question A-5.3.2.4-1 The evaluation of effects of neutron embrittlement on RPV materials uses Charpy Uupper-Sshelf Eenergy. A decrease in Charpy Uupper-Sshelf Eenergy level as defined in ASTM E 185-82 occurs based on fluence levels and copper content in the material. The design does not require this evaluation because the lower RPV shell is not a ferritic material. The upper RPV is ferritic and meets the requirements of ASME BPVC Section III, Subsection NV, Paragraph NB-3210,
NuScale Final Safety Analysis Report Reactor Vessel NuScale US460 SDAA 5.3-8 Draft Revision 2 which requires a minimum Charpy upper-shelf energy of 50 ft-lb., and the fluence levels for the upper RPV shell are less than the peak neutron fluence at the end of the design life of 1.0E+17 n/cm2 (E>1 MeV) (Reference 5.3-5). Reference 5.3-7 provides further details regarding the resistance to the effects of neutron and thermal embrittlement of the austenitic stainless steel material used in the lower RPV.
5.3.3 Reactor Vessel Integrity 5.3.3.1 Design Section 5.3.1, Reactor Vessel Materials describes the compatibility of the RPV design with established standards. Section 5.2.4, RCPB Inservice Inspection and Testing, and Section 5.3.1, Reactor Vessel Materials describes how the basic design of the RPV establishes compatibility with required inspections.
5.3.3.2 Materials of Construction Section 5.2.3, RCPB Materials, and Section 5.3.1, Reactor Vessel Materials describe the reactor vessel materials of construction.
5.3.3.3 Fabrication Methods Section 5.2.3, RCPB Materials, and Section 5.3.1, Reactor Vessel Materials describe the fabrication methods used in the construction of the reactor vessel.
5.3.3.4 Inspection Requirements Section 5.3.1, Reactor Vessel Materials describes the nondestructive examinations performed.
5.3.3.5 Shipment and Installation Section 5.2.3.4.2 describes the packaging, shipment, handling, and storage of the RPV.
A dry environment is maintained for RPV surfaces, both primary and secondary sides, by an installed non-chloride, non-corrosive desiccant. Humidity indicators covering a suitable range of moisture content are shipped with the RPV. Both the primary and secondary sides of the RPV ship under positive pressure. The internal atmosphere on both sides of the SG tubes are evacuated to eliminate residual moisture and filled with nitrogen having a dew point less than
-20 degrees F.
In preparation for shipping the RPV, the fabricator takes appropriate foreign material exclusion measures.
There are cleanliness and contamination controls in place during handling, storage, shipping, and during installation of the RPV. Section 5.2.3.4.2, Cleaning