ML24075A188
| ML24075A188 | |
| Person / Time | |
|---|---|
| Site: | 99902078 |
| Issue date: | 03/15/2024 |
| From: | NRC |
| To: | NRC/NRR/DNRL/NRLB |
| References | |
| Download: ML24075A188 (7) | |
Text
From:
Getachew Tesfaye Sent:
Friday, March 15, 2024 9:22 AM To:
Request for Additional Information Cc:
Prosanta Chowdhury; Mahmoud -MJ-Jardaneh; Griffith, Thomas; Osborn, Jim; Fairbanks, Elisa; NuScale-SDA-720RAIsPEm Resource
Subject:
NuScale SDAA Section 3.9.4 - Request for Additional Information No. 014 (RAI-10131-R1)
Attachments:
SECTION 3.9.4 - RAI-10131-R1-FINAL.pdf Attached please find NRC staffs request for additional information (RAI) concerning the review of NuScale Standard Design Approval Application for its US460 standard plant design (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23306A033).
Please submit your technically correct and complete response by the agreed upon date to the NRC Document Control Desk.
If you have any questions, please do not hesitate to contact me.
Thank you, Getachew Tesfaye (He/Him)
Senior Project Manager NRC/NRR/DNRL/NRLB 301-415-8013
Hearing Identifier:
NuScale_SDA720_RAI_Public Email Number:
27 Mail Envelope Properties (BY5PR09MB56822607E968F8A3354CAB2C8C282)
Subject:
NuScale SDAA Section 3.9.4 - Request for Additional Information No. 014 (RAI-10131-R1)
Sent Date:
3/15/2024 9:22:23 AM Received Date:
3/15/2024 9:22:28 AM From:
Getachew Tesfaye Created By:
Getachew.Tesfaye@nrc.gov Recipients:
"Prosanta Chowdhury" <Prosanta.Chowdhury@nrc.gov>
Tracking Status: None "Mahmoud -MJ-Jardaneh" <Mahmoud.Jardaneh@nrc.gov>
Tracking Status: None "Griffith, Thomas" <tgriffith@nuscalepower.com>
Tracking Status: None "Osborn, Jim" <josborn@nuscalepower.com>
Tracking Status: None "Fairbanks, Elisa" <EFairbanks@nuscalepower.com>
Tracking Status: None "NuScale-SDA-720RAIsPEm Resource"
<NuScale-SDA-720RAIsPEm.Resource@usnrc.onmicrosoft.com>
Tracking Status: None "Request for Additional Information" <RAI@nuscalepower.com>
Tracking Status: None Post Office:
BY5PR09MB5682.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 584 3/15/2024 9:22:28 AM SECTION 3.9.4 - RAI-10131-R1-FINAL.pdf 160092 Options Priority:
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1 REQUEST FOR ADDITIONAL INFORMATION No. 014 (RAI-10131-R1)
BY THE OFFICE OF NUCLEAR REACTOR REGULATION NUSCALE STANDARD DESIGN APPROVAL APPLICATION DOCKET NO. 05200050 CHAPTER 3, DESIGN OF STRUCTURES, SYSTEMS, COMPONENTS AND EQUIPMENT SECTION 3.9.4, CONTROL ROD DRIVE SYSTEM ISSUE DATE: 03/15/2024
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Background===
By letter dated October 31, 2023, NuScale Power, LLC (NuScale or the applicant) submitted Part 2, Final Safety Analysis Report (FSAR), Chapter 3, Design of Structures, Systems, Components and Equipment, Revision 1 (Agencywide Documents Access and Management System Accession No. ML23304A321), of the NuScale Standard Design Approval Application (SDAA) for its US460 standard plant design. The applicant submitted the US460 standard plant SDAA in accordance with the requirements of Title 10 Code of Federal Regulations (10 CFR)
Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, Subpart E, Standard Design Approvals. The NRC staff has reviewed the information in FSAR Chapter 3 of the SDAA and determined that additional information is required to complete its review.
Question 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
2 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, 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.
Question 3.9.4-9 Regulatory Bases 10 CFR 50.55a:The pressure housings for the control rod drive mechanisms (CRDMs) constitute part of the reactor coolant pressure boundary (RCPB) and are American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code)Section III Class 1 components. The connection between these Class 1 components and the Reactor Pressure Vessel head (also a Class 1 component) must meet the ASME BPV Code Section III Subsection NB requirements in accordance with 10 CFR 50.55a.
GDC 1, Quality standards and records, and 10 CFR 50.55a, as they relate to the CRDS, require that the CRDS be designed to quality standards commensurate with the importance of the safety functions to be performed.
GDC 2, Design bases for protection against natural phenomena, as it relates to the CRDS, requires that the CRDS be designed to withstand the effects of an earthquake without loss of capability to perform its safety functions.
3 GDC 4, Environmental and dynamic effects design bases, as it relates to the CRDS, requires that structures, systems, and components important to safety (including the CRDS) be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.
GDC 14, "Reactor coolant pressure boundary, as it relates to the CRDS, requires that the RCPB portion of the CRDS be designed, constructed, and tested for the extremely low probability of leakage or gross rupture.
GDC 26, Reactivity control system redundancy and capability, 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.
GDC 27, Combined reactivity control systems capability, as it relates to the CRDS, requires that the CRDS be designed with appropriate margin, and in conjunction with the emergency core cooling system, be capable of controlling reactivity and cooling the core under postulated accident conditions.
GDC 28, Reactivity limits, as it relates to the CRDS, requires that the reactivity control systems be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.
GDC 29, Protection against anticipated operational occurrences, 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 anticipated operational occurrences.
Issue Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition (NUREG-0800), Section 3.9.4, Control Rod Drive Systems, addresses review and acceptance criteria for Control Rod Drive System (CRDS). Part of this review includes a review of applicable design loads (static and alternating) and their appropriate combinations, the corresponding design stress and fatigue limits, and the corresponding allowable deformations. If applicable, fatigue assessments, primarily of connections between CRDS components, should
4 be performed to ensure the assembly maintains integrity throughout the design life of the plant.
This topic is also addressed in Section 3.9.3, ASME Code Class 1, 2, and 3 Components, and Component Supports, and Core Support Structures.
Section 3.13, Threaded Fasteners - ASME Code Class 1, 2, and 3, provides guidance for reviewing and evaluating the adequacy of an applicants criteria in regard to selection of materials, design, inspection and testing of its threaded fasteners (i.e., threaded bolts, studs, etc.) prior to initial service and during service in ASMEBPV Code Class 1, 2 or 3 systems.
Section 15.4.8, Spectrum of Rod Ejection Accidents (PWR) and Regulatory Guide 1.236, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents, describe the initiating event of the rod ejection accident as the rapid rejection of a control rod caused by an assumed control rod mechanism housing failure. General Design Criteria (GDC) 28 requires the evaluation of the rod ejection accident. Consistent with GDC 28, SRP 15.0 classifies control rod ejection as a postulated accident. NuScale SDAA Section 15.4.8.1 also classifies the rod ejection accident as a postulated accident.
Branch Technical Position (BTP) 3-4, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment, provides criteria that the NRC staff finds acceptable to preclude the need to postulate breaks and cracks in certain break locations of particular piping systems.
As part of its review of the SDAA, the NRC staff compared what had been accepted under the Design Certification Application (DCA) review and noted a significant change from the DCA design related to the connection between the control rod drive mechanism (CRDM) pressure housing and the Reactor Pressure Vessel (RPV) head. The DCA design utilized welded connections while the SDAA design utilized bolted connections with threaded inserts. This is a significant change from the previously reviewed design that warrants additional review. This change impacts safety-related ASME BPV Code Section III Class 1 components that constitute the reactor coolant pressure boundary. This change also deviates from the basis used to deem a CRDM housing failure, sufficient to create a missile from a piece of the housing or to allow a control rod to be ejected rapidly from the core, as non-credible for the DCA. A significant number of bolted connections exist on the RPV head for this design, which has resulted in NRC staff questions regarding the design of these connections. The NRC staff has attempted to consolidate questions on these connections (with a linkage to Section 3.13 of the SER),
intending to sample from those connections designed to the criteria of BTP 3-4, but it is unclear to the staff, with the information currently available, which bolted connections on top of the RPV head (including the CRDM connection) are designed to the criteria of BTP 3-4. Should a sample of bolted connections designed to BTP 3-4 be selected, it is uncertain if the CRDM connection would be considered as part of the sample. Due to the safety significant nature of these CRDM connections and the change in the SDAA design from that reviewed in the DCA, the NRC requests the following information to support its safety findings for the review of the structural integrity of the CRDM connections. The staff does not consider regulatory commitments to complete the design evaluation and meet the ASME BPV Code acceptance criteria through ITAAC 02.01.01as an appropriate resolution for this matter.
Due to the nature of construction and the potential for deviations between as-designed and as-built configurations, the NRC realizes that minor changes may occur through the fabrication process for these bolted connections. Therefore, the final as-built evaluation or a reconciliation
5 for the CRDM pressure housing bolted connection analyses would be appropriate to be resolved through an ITAAC.
Information Requested Provide a list of bolted connections on the RPV head, noting which connections are designed to the criteria of BTP 3-4.
Provide a summary of preliminary analysis results based on current design for 1 representative CRDM Bolted connection to RPV and 1 representative CRDM support structure bolted connection to RPV that use threaded inserts.. Include ASME BPV Code Section III Class 1 stress intensities and cumulative usage factors (CUFs) for all the key components of the bolted connections (bolting, threaded inserts, flanges, and others), based on the current design of the CRDM bolted connections. This evaluation should be made available for staff audit. The discussion should address the inputs, boundary conditions, and margins in stresses and CUF values for threaded inserts, bolts, flanges and other key components. A comparison to the criteria outlined in Branch Technical Position 3-4 may be appropriate for this connection.
Further, discuss the credibility, and underlying basis, of a mechanistic failure of a CRDM pressure housing, considering the design change of the connection between the CRDM pressure housing and the RPV head. Include a discussion of any added features for missile protection (such as those deployed in the existing fleet of PWRs) in light of this change).