ML24205A156
| ML24205A156 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 07/23/2024 |
| From: | Shaver M NuScale |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| Shared Package | |
| ML24205A155 | List: |
| References | |
| RAIO-172351 | |
| Download: ML24205A156 (1) | |
Text
RAIO-172351 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com July 23, 2024 Docket No.52-050 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Response to NRC Request for Additional Information No. 14 (RAI-10131 R1, Question 3.9.4-8) on the NuScale Standard Design Approval Application
REFERENCE:
- 1. NRC Letter to NuScale, Request for Additional Information No. 014 (RAI-10131 R1), dated March 15, 2024 The purpose of this letter is to provide the NuScale Power, LLC (NuScale) response to the referenced NRC Request for Additional Information (RAI).
The enclosure to this letter contains NuScale's response to the following RAI question from NRC RAI-10131 R1:
x 3.9.4-8 is the proprietary version of the NuScale Response to NRC RAI No. 014 (RAI-10131 R1, Question 3.9.4-8). NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 is the nonproprietary version of the NuScale response.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Elisa Fairbanks at 541-452-7872 or at efairbanks@nuscalepower.com.
I declare under penalty of perjury that the foregoing is true and correct. Executed on July 23, 2024.
Sincerely, Mark W. Shaver Director, Regulatory Affairs NuScale Power, LLC
RAIO-172351 Page 2 of 2 07/23/2024 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Distribution:
Mahmoud Jardaneh, Chief, New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Manager, NRC Prosanta Chowdhury, Senior Project Manager, NRC
- NuScale Response to NRC Request for Additional Information RAI-10131 R1, Question 3.9.4-8, proprietary : NuScale Response to NRC Request for Additional Information RAI-10131 R1, Question 3.9.4-8, nonproprietary : Affidavit of Mark W. Shaver, AF-172352
RAIO-172351 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com :
NuScale Response to NRC Request for Additional Information RAI-10131 R1, Question 3.9.4-8, proprietary
RAIO-172351 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com NuScale Response to NRC Request for Additional Information RAI-10131 R1, Question 3.9.4-8, nonproprietary
Response to Request for Additional Information Docket: 052000050 RAI No.: 10131 Date of RAI Issue:03/15/2024 NRC Question No.: 3.9.4-8 Regulatory Basis Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants, contains the following requirements applicable to the NuScale control rod drive system (CRDS):
- GDC 2, as it relates to the important-to-safety functions performed by the CRDS, requires that the CRDS be designed to withstand the effects of an earthquake without loss of capability to perform its safety functions.
- GDC 26, as it relates to the CRDS, requires that the CRDS be one of the independent reactivity control systems that is designed with appropriate margin to assure its reactivity control function under conditions of normal operation, including Anticipated Operational Occurrences (AOOs).
- GDC 27, as it relates to the CRDS, requires that the CRDS be designed with appropriate margin for stuck rods, and, in conjunction with the emergency core cooling system, be capable of controlling reactivity changes and cooling the core under postulated accident conditions.
- GDC 29, as it relates to the CRDS, requires that the CRDS, in conjunction with reactor protection systems, be designed to assure an extremely high probability of accomplishing its safety functions in the event of AOOs.
Issue During an in-person audit conducted on September 28, 2023, a new configuration was identified for the steam generator tube supports in the NuScale US460 design, which are located in the vicinity of the CRDS. Specifically, a large number of hex nuts are located inside the riser assemblies directly above the reactor core. The staff requested information on these features, NuScale Nonproprietary NuScale Nonproprietary
with an emphasis on any potential impact that they may have on the safety-related function of the nearby CRDS to insert control rods, particularly in a scenario where a nut becomes detached and drops into the path of travel for the control rod assemblies. In the initial response, the applicant provided information on the configuration (a threaded assembly with three welds to prevent rotation) and stated that the welded components are not postulated to impact the safety-related function of the control rods. A general statement was provided through an audit response that an augmented inspection program is being considered for the upper riser set screws. The FSAR does not contain any details of an augmented inspection program for the upper riser set screws.
Operating experience in similar configurations, where threaded fasteners have degraded into loose parts despite design elements included to ensure their longevity (e.g., baffle former bolts welded lock tabs), suggests that augmented inspection of these components to identify potential degradation may be needed. In addition to their function supporting the steam generator tubes, the subject components (particularly the hex nuts) are located above many potentially sensitive internals, such as the lower riser assembly guide tubes and fuel assemblies. Consequently, management of these components should be implemented to identify potential degradation before it becomes self-revealing or widespread.
Information Requested The staff requests the following information be provided:
- a description of the steam generator tube support assemblies with adequate detail of the elements comprising the assemblies, including features intended to prevent loose part generation,
- how the applicant will maintain assurance that the integrity of the welds and threaded elements (e.g., threaded insert, set screw, and hex nuts) remains adequate, Provide corresponding markups of the FSAR to reflect this information.
NuScale Nonproprietary NuScale Nonproprietary
NuScale Response:
Executive Summary:
The Staff's question pertains to the design of the steam generator (SG) tube support assemblies and the capability to perform their function, which includes preventing loose parts that may impact other structures, systems, and components (SSC). Staff noted GDCs 26, 27, and 29, however, these apply to the design and function of the reactivity control systems. Instead, NuScale understands the overall requirements of GDCs 1, 2, and 4 as considerations that are relevant to the design of the SG tube support assemblies. In particular, this response addresses the design, fabrication, and inspection of the SG tube support assemblies to quality standards commensurate with the importance of their functions, including SG support and the prevention of loose parts, pursuant to GDC 1.
The steam generator support assemblies consist of a T-shaped beam that is welded to the baffle plate and integral steam plenum and a beam that is cantilevered off of the inner reactor pressure vessel (RPV) surface. Set screw assemblies are used to close the clearance between the upper riser shell and backing strip to ensure that a firm contact is achieved between the upper riser assembly, the backing strips, each SG tube support, and the upper RPV. Mitigation strategies utilized to prevent loose part generation include welds to prevent rotation of the set screw assembly and assessment of operating experience. The operating experience review allowed NuScale to develop a design that precludes industry failure mechanisms.
NuScale Nonproprietary NuScale Nonproprietary
Description of the steam generator tube support assemblies:
The SG tube support assemblies are composed of a set of column support assemblies positioned between each column of helical tubes. The column support assemblies hang vertically down through the tube bundle from an upper SG support as depicted in Figure 1 and SDAA Figure 5.4-5. The upper SG support is a T-shaped beam that is welded to the baffle plate and integral steam plenum. The lower ends of the tube support assemblies are restrained in the circumferential direction by the lower SG support. The lower SG support is a beam that is cantilevered off of the inner RPV surface.
Figure 1: Steam Generator Supports Eight tube support assemblies are spaced around the circumference of the RPV (Figure 2). The outermost column support abuts with the RPV inner cladding surface. An SG tube support spacer provides support for the outermost SG column (i.e., Column 21) support in the region where the RPV inner diameter flares. There are socket head screw caps that affix the SG tube support spacers to the Column 21 SG tube support.
NuScale Nonproprietary NuScale Nonproprietary
Figure 2: Steam Generator Tube Support Assembly
((2(a),(c) Machined slots in the tube supports create a pocket that encompasses the tube outer diameter, providing support to the tube. The tube support design is such that adjacent tube supports are nested (i.e., interlocked) with each other to ensure radial and circumferential load transfer through the tube support assembly. A backing strip is positioned between the upper riser assembly and the first column of tubes that completes the enclosure of the Column 1 tubes and functions as the interface between the upper riser assembly and the tube support. NuScale Nonproprietary NuScale Nonproprietary
The set screw assemblies are installed in the upper riser shell to close the clearance between the upper riser shell and backing strip and ensure that firm contact is achieved between the upper riser assembly, the backing strips, each SG tube support, and the upper RPV, as depicted by Figure 3 and Figure 4 below. The set screw assemblies experience a compressive force to maintain firm contact, and do not provide a component support function as defined by ASME Section III NF. Figure 3: Support Assembly Configuration (( }}2(a),(c),ECI NuScale Nonproprietary NuScale Nonproprietary
Figure 4: Backing Strip and Set Screw Interface (( }}2(a),(c),ECI (( }}2(a),(c) The approximate dimensions of each part are as follows: (( }}2(a),(c) NuScale Nonproprietary NuScale Nonproprietary
Features intended to prevent loose part generation: }}2(a),(c),ECI The presence of weld filler metal and corresponding fusion at the threads in the root of the threads prevents components from unthreading through flow-induced or other means of vibration because of the torque required to shear the weld. Figure 5: Set Screw Assembly Drawing (( }}2(a),(c),ECI NuScale Nonproprietary NuScale Nonproprietary
Table 1: Set Screw Assembly Parts (( }}2(a),(c) Fabrication of these welds applies ASME BPVC Section III, Div. 1, Article NG-4000, including the use of qualified welding procedures. These procedures apply qualification requirements in accordance with ASME BPVC Section IX with representative testing of these welds. The fabricator inspects welds to verify proper fusion, in accordance with NG-5000. The verified fusion at the weld location assures a fracture during operation does not impact the function of the weld and inhibits unthreading of the set screws as the threads are destroyed in the fusion of the weld material, set screw, and nut. NuScale Nonproprietary NuScale Nonproprietary
Integrity of the welds and threaded elements: The staff incorrectly compares the configuration of the US460 set screw assemblies to components within currently operating reactors that have different environmental conditions and degradation mechanisms. The degradation and failure of baffle-former assembly bolts and associated welded lock tabs is an existing industry issue with older Westinghouse reactor vessels, as described in the NRC issued Information Notice No. 98-11, "Cracking of Reactor Vessel Internal Baffle-Former Bolts in Foreign Plants," and the NRC-approved Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) topical report, MRP-227-A, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The primary degradation mechanism for the failure of baffle-former bolts is irradiation-assisted stress-corrosion cracking (IASCC). Specifically, this age-related intergranular stress-corrosion cracking process is influenced by bolt material, fluence, stress, and temperature. The NuScale US460 set screw assembly components share none of these contributing factors. A summary of failure mechanisms and justification that the failure mechanism does not apply to the US460 set screws can be seen in Table 2. Table 2: Baffle-Former Bolt and Set Screw Comparison (( }}2(a),(c) Additionally, the design specification for reactor vessel internals NuScale Nonproprietary NuScale Nonproprietary
requires that cold work is removed by solution annealing, reducing risks for stress corrosion cracking. Baffle-former bolts are bolts, and remain in tension as they hold the baffle plates against the former. The US460 set screws are set screws, and remain in compression as they close the clearance between the upper riser shell and backing strip and ensure that firm contact is achieved between the upper riser assembly, the backing strips, each SG tube support, and the upper RPV. Baffle-former bolts are located directly around the core barrel region and experience higher levels of irradiation than the US460 set screw assemblies. Figure 6 represents the equivalent location of baffle-former bolts in comparison to the set screw assemblies. The set screw assemblies are located in the flow path of the primary coolant; therefore, it is not expected that impurities can build up around them that would contribute to stress corrosion cracking. NuScale Nonproprietary NuScale Nonproprietary
Figure 6: Set Screw Assembly Location Comparison Furthermore, there are no postulated mechanisms for wear relative to the set screw assemblies. The set screw assemblies are not susceptible to fluid elastic instability, acoustic resonance, gallop, or flutter due to the geometry. These components are designed with sufficient stiffness such that they do not develop significant motion due to excitation and are screened out of the detailed analysis for turbulent buffeting, ((
}}2(a),(c). The length of the set screw assembly protruding from the upper riser shell is less than (( }}2(a),(c) and the high natural frequency indicates that vortex shedding lock-in is not expected to occur.
NuScale Nonproprietary NuScale Nonproprietary
Summary of reviewed operating experience: NuScale reviewed the operating experience, listed in Table 3, much of which is not applicable to the NuScale design. The following design elements mitigate potential issues:
use of SA479 TP304 as the material for the set screws and threaded elements
low fluence levels compared to the core beltline region
fluid velocities lower than traditional pressure water reactors (PWR) Industry operating experience was reviewed from Institute of Nuclear Power Operations (INPO) SOER 84-5 and MRP-227 and is summarized below: Table 3: Operating Experience Operating Experience Description Application to NuScale Design SOER 84-5 Boric Acid Corrosion outside the RPV Not applicable: NuScale utilizes SB-637 Alloy 718 studs SOER 84-5 Inconel X-750 control rod guide tube split pins (special application bolts) in Westinghouse and Framatome reactors Not applicable: Not in NuScale Design SOER 84-5 Inconel X-750 control rod drive donut hold-down flexures (special application bolts) in Westinghouse reactors Not applicable: Not in NuScale Design SOER 84-5 Reactor coolant pump diffuser adapter-to-casing bolts in Westinghouse and combustion engineering reactor coolant pumps Not applicable: Not in NuScale Design SOER 84-5 Alloy A286 (ASME SA 453 Grade 660) thermal shield attaching bolts and core support barrel bolts in Babcock & Wilcox reactors Not applicable: Not in NuScale Design MRP-227 Intergranular stress corrosion cracking (IGSCC) - observed in Alloy A-286 ASTM A453 bolting. Not applicable: Utilize SA479 TP304 MRP-227 IGSCC - failures of core barrel bolts limited to Alloy X-750 Not applicable: Utilize SA479 TP304 MRP-227 Primary water stress-corrosion cracking (PWSCC) - Alloy X-750 failures in control rod guide tube support pins. Not applicable: Not in NuScale Design MRP-227 Irradiation-assisted stress-corrosion cracking (IASCTC) - Type 347 stainless baffle-former bolt failures. Not applicable: set screw assemblies located outside of the core beltline region and utilizes SA479 TP304 NuScale Nonproprietary NuScale Nonproprietary
In summary, there are no postulated failure mechanisms for the set screw assembly weld and threaded elements within the 60-year service life of the US460 design. The requirements of 10 CFR 50.55a are met with the development and implementation of the ASME Section XI Inservice Inspection Program. In order to provide reasonable assurance that the set screws do not degrade over time, augmented inservice inspection requirements are applied to the set screw assemblies. A visual (VT-1) examination is performed on a 10 percent sample of the set screw assemblies once per inservice inspection interval. The scope of the visual (VT-1) examination is limited to the welded nut and welded set screw and performed in accordance with ASME Section XI IWA-2211. The attached markup shows FSAR Section 3.9.5 changes for the steam generator tube support assembly augmented examination requirements. Impact on US460 SDAA: FSAR Section 3.9 has been revised as described in the response above and as shown in the markup provided in this response. NuScale Nonproprietary NuScale Nonproprietary
NuScale Final Safety Analysis Report Mechanical Systems and Components NuScale US460 SDAA 3.9-45 Draft Revision 2 The upper riser assembly is located immediately above the lower riser assembly and extends upward to the pressurizer (PZR) baffle plate. It channels the reactor coolant exiting the lower riser upwards and into the space above the top of the upper riser shell and below the PZR baffle plate. Reactor coolant flow then turns downward through the annular space outside of the upper riser and inside of the RPV where the SG helical tube bundles are located. Flow paths are located in the upper riser to permit a small amount of reactor coolant to bypass the top of the riser and flow into the SG tube bundle region. These flow paths ensure sufficient boron concentration remains in the reactor coolant during DHRS-driven riser uncovery conditions following non-LOCA transients, while not introducing structural integrity and fatigue concerns. RAI 3.9.4-8 The steam generator tube support assemblies include set screw assemblies installed in the URS. The set screw assemblies ensure that firm contact is achieved between the upper riser assembly, each SG tube support, and the upper RPV. The set screw assemblies do not provide a component support function. Augmented inservice inspection (ISI) requirements are applied to the set screw assemblies. A visual (VT-1) examination is performed on a 10 percent sample of the set screw assemblies once per ISI interval. The scope of the visual (VT-1) examination is limited to the welded nut and welded set screw. NuScale evaluated the potential for acoustic noise from vortices that may be formed at the upper riser holes, and other potential flow-induced vibration effects of flow through the holes onto the SG tubes. The riser flow holes are not expected to produce vortices due to the flow through the holes. If, however, vortices form, they do not coincide with a relevant acoustic mode of the riser. Additionally, the fluid passing through the riser holes produces minimal forces on the nearest SG tube column, and the frequency of the normal operation flow through the holes does not coincide with a predominant structural mode of the adjacent SG tubes. Therefore, the riser flow holes do not introduce structural integrity concerns due to FIV. Audit Question A-5.1.3.4-1 The RVIupper riser assembly is fastened to the pressurizer baffle plate at the hanger plate. The CRDS shaft sleeves connect the hanger plate to the rest of the upper riser assembly. During refueling and maintenance outages the upper riser assembly stays attached to the upper section of the NPM (upper CNV, upper RPV, and SG) while providing physical access for inspection to the RPV, SG, and upper riser assembly components. The set of upper CRDS support plates in the upper riser assembly, in conjunction with the CRA guide tube support plate, CRA guide tubes, and upper core plate in the lower riser assembly align and provide lateral support for the control rod drive shafts. These component geometries ensure adequate alignment of the CRDS with the fuel assemblies and permit full insertion of control rods under design-basis events (DBEs).
RAIO-172351 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Affidavit of Mark W. Shaver, AF-172352
AF-172352 Page 1 of 2
NuScale Power, LLC AFFIDAVIT of Mark W. Shaver I, Mark W. Shaver, state as follows: (1) I am the Director of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale. (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying Request for Additional Information response reveals distinguishing aspects about the response by which NuScale develops its NuScale Power, LLC Response to NRC Request for Additional Information (RAI No. 10131 R1, Question 3.9.4-8) on the NuScale Standard Design Approval Application. NuScale has performed significant research and evaluation to develop a basis for this response and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScales competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScales intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed response to NRC Request for Additional Information RAI 10131 R1, Question 3.9.4-8. The enclosure contains the designation Proprietary at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, (( }} in the document.
AF-172352 Page 2 of 2 (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale. (b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScales technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on July 23, 2024. Mark W. Shaver M k W Sh}}