ML24313A052

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LLC Response to NRC Request for Additional Information No. 14 (RAI-10131 R1) on the NuScale Standard Design Approval Application
ML24313A052
Person / Time
Site: 05200050, 99902078
Issue date: 11/08/2024
From: Shaver M
NuScale
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
RAIO-175396
Download: ML24313A052 (1)
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RAIO-175396 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com November 08, 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) on the NuScale Standard Design Approval Application

REFERENCE:

NRC Letter to NuScale, Request for Additional Information No. 14 (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:

3.9.4-9 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 November 08, 2024.

Sincerely, Mark W. Shaver Director, Regulatory Affairs NuScale Power, LLC

RAIO-175396 Page 2 of 2 11/08/2024 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, nonproprietary

RAIO-175396 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, nonproprietary

Response to Request for Additional Information Docket: 052000050 RAI No.: 10131 Date of RAI Issue: 03/15/2024 NRC Question No.: 3.9.4-9 Regulatory Bases

  • 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.
  • 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.

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  • 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.

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 be performed to ensure the assembly maintains integrity throughout the design life of the plant.

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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 NuScale Nonproprietary NuScale Nonproprietary

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 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).

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NuScale Response:

Executive Summary The control rod drive mechanism (CRDM) pressure housings are a Class 1 appurtenance per American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC),

Section III, NCA-1271. The CRDM seismic supports located on the reactor pressure vessel (RPV) and containment vessel (CNV) head are ASME Code Class 1, Seismic Category I component supports. The connections for the CRDM housings meet ASME Section III NB-3200 design criteria to assure their design function for reactor coolant pressure boundary (RCPB) integrity during normal conditions, including anticipated operational occurrences (AOOs).

In the US600 design and as described in the Design Certification Application (DCA), the connections for the CRDM to the reactor pressure vessel (RPV) head utilize a welded connection. For the US460 design and as described in the Standard Design Approval Application (SDAA), the same connections (CRDM to RPV) consist of a bolted flange. The change in connection (welded versus bolted) does not impact the design standards or stress analysis to be performed. The DCA Safety Evaluation Report (SER) reached a safety finding on CRDM housing integrity without requiring preliminary stress or cumulative usage factors (CUF).

The DCA SER also found that control rod drive mechanism (CRDM) housings, are ASME BPV Code Class 1... and therefore are not considered credible missile sources.

The SDAA follows the same design basis as the DCA without deviation from the underlying codes and safety standards established therein. A change to a bolted connection has no impact to the expected performance of the connections for the lifespan of the plant during operations, and assures the design will perform sufficiently in postulated AOOs and associated accidents.

Notwithstanding the foregoing, in a discussion with the staff in May 7 & 8 audit meetings, NuScale agreed to provide a representative analysis for connections demonstrating an example connection to the RPV head. NuScale understands this request as justified by the need to demonstrate design finality consistent with the Commissions standardization objectives. The reactor vent valve (RVV) and reactor recirculation valves (RRV) were selected in alignment with the staff as these connections represent an example connection for stress and fatigue characteristics. To demonstrate design sufficiency of an RPV bolted connection using a threaded insert, RAI 3.13-1 response provides analysis on the RVV and RRV threaded insert connections showing design margin and represents the CRDM pressure housing connections.

The completion of the analysis performed in RAI 3.13-1 accomplishes a standardization NuScale Nonproprietary NuScale Nonproprietary

objective of flange connections on the RPV as a sample of an ASME preliminary analysis that is to be completed for all connections for safety-related objectives.

US600 DCA Precedent The design basis of the US600 CRDM housings as Class 1 appurtenances was evaluated by the NRC first in an audit specific to the CRDMs and then in the SER for the designs capability to ensure RCPB integrity under normal operations and AOOs. The audit included a review of the failure modes and effects analysis. NuScale provided NuScale CRDM Failure Modes and Effects Analysis (FMEA) ER-A022-3459, Revision 1. The staff audit of the US600 CRDM was documented in the NRC Staff Audit Report on CRDM (NRC Agencywide Document Access and Management System (ADAMS) Accession No. ML17331A357). The audit report conclusions are as follows.

Staff Audit Report on CRDM (ADAMS Accession No. ML17331A357) 2.3 Failure Modes and Effects Analysis The NRC staff audited ER-A022-3459, Revision 1, NuScale CRDM Failure Modes and Effects Analysis (FMEA), that demonstrated that no single failure in the CRDS could prevent a reactor trip and the ability to rod drop on command was retained. Therefore, the NRC staff was satisfied that a failure modes and effects analyses was completed and determine that the CRDS is capable of performing its safety-related function following the loss of any active component.

The FMEA described the criticality and consequence of failure of various design functions, including the latch housing assembly to provide a pressure boundary for the RPV. The FMEA identified the fatigue failure for cumulative thermal and pressure transients as a high concern, and the preventative action to mitigate this consequence was the design analysis per NB-3200.

The NRC audit report verified the design change sufficiently addressed the FMEA. The change from a welded to bolted connection does not alter the design function of this connection and the same NB-3200 design code applies to the US460 CRDM housing connections. The SDAA FMEA describes the same conclusion.

The subsequent safety evaluation summarized the results of this audit and included potential missile generations associated with the housing, including rod ejection accidents, and their demonstrated suitability using the codes and standards associated with Class 1 appurtenances.

The safety evaluation statements that are relevant to the scope of this RAI are as follows.

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SER 3.5.1.1.4 Technical Evaluation The staff reviewed the potential for missiles generated from pressurized systems. DCA Part 2, Tier 2, Section 3.5.1.1.1, Pressurized Systems, considers the following potential missiles from pressurized systems as noncredible (P1 < 10-7):

moderate-and low-energy systems with operating pressures of less than 1.90 megapascals (MPa) (gauge) (275 pounds-force per square inch, gauge (psig)),

because of insufficient stored energy to generate a missile

piping and valves designed in accordance with ASME BPV Code,Section III, and maintained in accordance with the ASME BPV Code,Section XI, inspection program

threaded valve stems with back seats because they are designed to prevent ejection of the stems and valve stems with power actuators because they are effectively restrained by the actuator

nuts, bolts, and a combination of the two because of the small amount of stored energy The staff reviewed the reasons stated above to eliminate certain missile sources. These missile sources are either designed to a high level of quality in accordance with ASME BPV Code,Section III, thus demonstrating that missile generation is unlikely, or they do not have sufficient energy to generate a credible missile. Therefore, the staff finds the above list of noncredible missile sources acceptable.

The safety evaluation then continued with the following.

The staff also reviewed the potential for internally generated missiles from inside containment. DCA Part 2, Tier 2, Section 3.5.1.2, states that the NPMs use a steel containment that encapsulates the reactor pressure vessel (RPV) and that there is no rotating equipment inside containment. All pressurized components inside containment, including control rod drive mechanism (CRDM) housings, are ASME BPV Code Class 1 or 2 and therefore are not considered credible missile sources. The applicant does not consider these pressurized components a credible missile source because of the material characteristics, inspections, quality control during fabrication and erection, and prudent operation. The staff reviewed the applicants bases as described above and finds the applicants conclusion on the elimination of the above components as credible missile sources acceptable, because it is consistent with the guidance in SRP Section 3.5.1.1. DCA Part 2, Tier 2, Section 15.4.8, Spectrum of Rod Ejection Accidents, NuScale Nonproprietary NuScale Nonproprietary

presents the safety analyses of the rod ejection accident and documents the associated staff review.

Based on its review, the staff finds the applicants approach to identify potential missiles, determine the statistical significance of potential missiles, and provide measures for SSCs needing protection against the effects of missiles to be acceptable. Therefore, the staff concludes that the applicants evaluation of potential internally generated missiles resulting from equipment and component failures satisfies the applicable requirements related to GDC 4.

SER 3.9.4.4.2 Descriptive Information DCA Part 2, Tier 2, Section 4.6.2, discusses a failure modes and effects analysis (FMEA) that evaluated failures of the CRDM. The staff reviewed this analysis during an audit of the design and testing programs for the CRDS (ADAMS Accession No. ML17331A357). The staffs review determined that the CRDS is capable of performing its safety-related function following the loss of any active component, as documented in the audit report and SER Section 4.6.

SER 3.9.4.4.7 Control Rod Drive Housing Integrity NuScale DCA Part 2, Tier 2, Section 3.5.1.2, states that 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, is noncredible. Section 3.5.1.2 also notes that the CRDM housing is a Class 1 appurtenance per ASME BPV Code,Section III. In its letter dated July 15, 2019 (ADAMS Accession No. ML19196A368), in response to NRC staff questions on the potential for a control rod ejection accident, NuScale indicated that the CRDM nozzles are an integral part of the reactor pressure vessel (RPV) closure head forging. NuScale specifies that the CRDM nozzle to Alloy 690 safe-end welds are full penetration butt welds, using Alloy 52/152 weld filler materials for corrosion resistance.

As a result, NuScale states that the connection between the CRDM nozzles and the RPV closure head will be structurally robust.

Similar to DC reviews for other reactor designs, the NRC staff evaluated the specific aspects of the NuScale CRDM housing to evaluate the potential for a control rod to be ejected from the reactor core. Based on its review, the NRC staff does not consider a gross failure of a NuScale CRDM housing sufficient to create a missile from a piece of the housing, or to allow a control rod to be ejected rapidly from the core, to be credible.

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For example, control rod missile generation would be prohibited by a combination of (1) the design of the CRDM nozzles as an integral part of the RPV head, (2) the likelihood that a postulated failure of a CRDM nozzle would be preceded by leakage flow in a lateral orientation such that a catastrophic failure of the nozzle would not occur, (3) the length of each control rod, and (4) the ASME BPV Code Class 1 design requirements.

CRDM housings are ASME BPV Code Class 1 components, which are subject to stringent requirements for material characteristics, inspections, and quality control during fabrication, erection, and operation. These Class 1 components are also subject to preservice and inservice inspections upon installation in the nuclear power plant, per Section XI of the ASME BPV Code, as discussed in NuScale DCA Part 2, Tier 2, Section 5.2.4, Reactor Coolant Pressure Boundary Inservice Inspection and Testing. Therefore, the NRC staff finds that 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, is noncredible for the NuScale nuclear power plant. Section 15.4.8 of this SER documents the NRC staff review of NuScale DCA Part 2, Tier 2, Section 15.4.8, Spectrum of Rod Ejection Accidents, which presents the safety analyses for a postulated control rod ejection accident, even though this event is not expected to occur during the life of the plant. In summary, the potential for a control rod ejection accident at a NuScale nuclear power plant is a nonmechanistic assumption for the purpose of evaluating reactivity with respect to 10 CFR Part 50, Appendix A, GDC 28, Reactivity Limits. Sections 15.4.8 and 15.0.6.4.3 of this SER provide additional discussion of the NRC staff review.

Based on the safety evaluation by the staff, the CRDM design described in the DCA was sufficient for a safety evaluation finding with a level of detail that did not include stress or fatigue analysis, or the need to apply BTP 3-4 criteria. The design of the CRDM housings as ASME Class 1 appurtenances provides confidence that the components will function as described in the final safety analysis report. The US460 design change from welded to bolted connections maintains the same design codes and standards, along with the requisite margins of safety, to not materially impact the conclusions of the staff from the DCA.

NuScale Response to: Provide a list of bolted connections on the RPV head, noting which connections are designed to the criteria of BTP 3-4.

The list of bolted connections on the RPV are listed below in Table 1 with a comparison between the US600 and US460 designs. The US460 design reflects a bolted connection for the CRDM pressure housings and CRDM structural supports, which is a change from the US600 welded connection. The US600 and US460 designs both reflect bolted connections for the NuScale Nonproprietary NuScale Nonproprietary

remainder of components on the RPV. The RVV and reactor recirculation RRV apply the criteria of BTP 3-4.

NuScale Response to: 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.

Representative preliminary stress analysis of a CRDM welded connection to the RPV was not provided for the DCA review. The change to the attachment method of the CRDM pressure housings from a welded to a bolted connection does not alter the capability of the housings to perform their design function to maintain the reactor coolant pressure boundary and ensure that upon reactor trip the control rod drive shaft (CRDS) with connected control rod assembly (CRA) is able to pass freely through the pressure housing. The US460 CRDM housing connections are designed and analyzed to the same requisite codes as the welded connections used in the US600, specifically as pertaining to ASME Section III, Mandatory Appendix XIII requirements.

The ASME design specification requires that the purchaser (NuScale) identify the interface loading requirements and the N-Certificate Holder (NuScales component design supplier) analyze for the certified load inputs. The interface loading requirements are an input to the Section III Mandatory Appendix XIII evaluation.

DCA SER 3.9.4.4.7 includes a statement pertaining to RAI 9647, wherein the staff questioned the evaluation of a control rod housing failure with respect to generic design criteria (GDC) 28, Standard Review Plan 15.4.8, and Regulatory guide 1.77. Specifically, the staff had a concern with a rod ejection accident (REA). In response, NuScale provided a justification crediting the continuous complete joint penetration weld to Alloy 690 safe-ends, using alloy 52/152 weld filler.

Use of a flanged connection still performs the same functions for maintaining a reactor coolant pressure boundary and the connection maintains the requisite ASME Class 1 design, fabrication and inspection requirements. Thus the change from a welded nozzle does not impact the conclusions reached in SER 3.9.4.4.7.

Once the N-Certificate Holder stress analysis is completed, the flange bolting shall meet the design criteria specified in Section III Mandatory Appendix XIII, article XIII-4230(a) for fatigue limits, which establishes the evaluation method and fatigue reduction strength factors to be considered in the qualification of the flange bolting.

The change from the US600 to the US460 design has not significantly altered the expected pressures or thermal characteristics at the CRDM connections on the RPV upper head; there has not been a significant change to the stress and fatigue margins established in the DCA for NuScale Nonproprietary NuScale Nonproprietary

the RPV at these locations. Therefore, the US460 design is comparable to the US600 design basis and safety is maintained at these connections.

As the CRDM pressure housings are located on the top of the RPV upper head, they act as a dead end to the pressurizer steam space, therefore these locations do not experience the postulated dynamic loads represented by a fluid system during operation. At the top of the head, the RPV connection point for the CRDM housings undergoes lower stress intensities than CRDM pad to RPV Flat Head Surface Transition Region. Thus the connections are not the limiting location for the stresses experienced in the RPV head and provide confidence that these locations will perform their expected design functions with significant margin. These loads are described in SDAA Table 3.9-8 and include the operating loads, transients, and design basis accidents including loads unique to the CRDM housings, such as control rod drop and stepping.

The CRDM support structure is a structural member bolted to the top of the RPV upper head and provides lateral seismic support for the CRDMs. This structure is designed as a seismic category I structure to withstand a safe-shutdown earthquake as part of the design basis. The change from a welded to a bolted connection on the RPV does not impact the connections capability to fulfill this function. The CRDM support structure does not serve a pressure-retaining function and is not a credible failure point.

As discussed in a Chapter 3 audit meeting with staff on May 7 & 8, 2024 to provide reasonable assurance on the design adequacy of threaded inserts, and for the purposes of representative analyses, the RPV bolted connection threaded inserts of the emergency core cooling system (ECCS) reactor recirculation and vent valves were compared from the US600 design with parameters from the US460 in RAI 3.13-1. The evaluation described in the response to RAI 3.13-1 concludes that margins of stress and fatigue are maintained in the SDA design. The ECCS valve locations undergo higher stresses and fatigue than the CRDM pressure housing connections, so the evaluation of the ECCS valve bolted connections represent a bounding condition for the function to maintain the RCPB.

NuScale Response to: A comparison to the criteria outlined in Branch Technical Position 3-4 may be appropriate for this connection.

The bolted connections between the CRDM pressure housings and the RPV are not piping; therefore, the criteria of BTP 3-4 B.A(ii), which are intended for piping, are not directly applicable. NuScale evaluated extending the stress and fatigue criteria of BTP 3-4 B.A(ii)(1) to these bolted connections in order to demonstrate a lower likelihood of failure, as was done for the RVVs and RRVs. However, as demonstrated below: (1) for the other connections such as NuScale Nonproprietary NuScale Nonproprietary

the CRDM, the stress limits for bolting required by ASME BPVC Section III, Subsection NB (and Mandatory Appendix XIII) are already more conservative than the BTP 3-4 B.A(ii)(1) limits, and (2) more conservative cumulative usage factor (CUF) limits are not justified. Standard Design Approval Application Section 3.6.2.7 discusses the bolted flanged connections between the RVVs and RRVs, and the RPV (which are classified as break exclusion zones) for further clarification on those connections.

Stress Criteria The limits on primary and secondary stress given in BTP 3-4 ensure that, after "shake down",

the strains in the break exclusion area are within the elastic limit (excluding the effects of stress concentrations). This is accomplished by the requirement that the stress as calculated by Eq.

(10), or Eq. (12) and Eq. (13) in NB-3653 do not exceed 2.4 Sm (i.e., 80% of the Code allowable). The majority of NuScale ASME Section III piping is SA-312 TP304, which at room temperature has Sy=30.0 ksi and Sm=20.0 ksi. Therefore, imposing a limit on the primary plus secondary stress range of 2.4 Sm allows a limit of 48.0 ksi, or 1.6 Sy. Alternating loads that remain below 2.0 Sy allow "shake down" to elastic action (i.e., after repetitive loading, a state is reached where plastic action is no longer produced), therefore an imposed limit of 1.6 Sy provides ample margin to non-recoverable yielding. Note that NB-3653.6 allows this limit to be applied separately to thermal loads and non-thermal primary and secondary stresses.

The stress limits for Class 1 studs are contained in Section III Mandatory Appendix XIII-4000.

XIII-4200 includes a limit on the maximum service stress resulting from direct tension plus bending of 1.0 Sy. Because the bolt will never be in compression (i.e., the minimum stress will always be zero), stress ranges are also effectively limited to 1.0 Sy (less than 1.6 Sy in piping).

Further, the bolting criteria in XIII-4200 do not allow for qualifying the thermal and non-thermal stresses separately, as is allowed for piping in NB-3653.6. Therefore, the stress limits for bolting contained in XIII-4200 provide more margin against progressive yielding than do the rules of NB-3653 for typical piping system materials, even when considering the more restrictive limits of BTP 3-4 B.A(ii)(1).

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Fatigue Branch Technical Position 3-4 limits the CUF to 0.1 rather than the ASME Code limit of 1.0, which is comparatively much more restrictive than the 80% factor that BTP 3-4 applies to stress ranges. Electric Power Research Institute Report 1022873, Improved Basis and Requirements for Break Location Postulation, revisits the bases of the BTP 3-4 guidance, which has not changed significantly since the 1970s. Since then, industry experience has shown that thermal fatigue, which typically dominates code calculated CUF, is much less likely to result in piping failures than other degradation mechanisms. The report concludes that, fatigue usage based on design basis calculations is, at most, a minimal contributor to the potential for pipe failures, and that, consideration of fatigue usage by itself is not a reliable approach to predict crack initiation or leakage. Therefore, it is also an unreliable parameter for estimating rupture.

However, in all cases evaluated, the use of a CUF criterion of 1.0 resulted in a minimal impact on core damage frequency (CDF) within limits commonly found to be acceptable to the NRC.

Therefore, the application of more restrictive limits on CUF, similar to BTP 3-4 methodology, has not been found to result in a significant reduction of the probability of failure over that of the ASME Code limits.

NuScale Response to: 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.

The SDAA follows the same design basis as the DCA for considering the CRDMs as Class 1 appurtenances in NB-3200 design criteria. As evaluated in the DCA SER 3.9.4.4.7, the CRDMs function as an integral continuity of the RPV head. A failure of the CRDM pressure housing connections is not considered credible as the connections are designed to the same standards described in the DCA in accordance with NB-3200 criteria for bolted connections. This design basis ensures that the RCPB integrity is maintained during events analyzed in the design basis and transients described in SDAA Chapter 15.

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Similar to the RPV, CRDM pressure boundary materials and design function maintain the RCPB for design basis events. As a failure of the ASME Class 1 RPV is not considered credible, by extension neither is a failure of the ASME Class 1 CRDM housings and their connections. The DCA SER 3.9.4.4.7 concluded that (2) the likelihood that a postulated failure of a CRDM nozzle would be preceded by leakage flow in a lateral orientation such that a catastrophic failure of the nozzle would not occur. The bolted connection does not alter this conclusion as the design standards applied ensure the connection fulfills the same function.

Inservice inspections and reactor coolant leakage monitoring provide means for early detection of potential degradation of the CRDM pressure boundary. In the NuScale design, leakage into containment, including CRDM pressure housing leakage, is classified as unidentified reactor coolant system (RCS) leakage and operators are required to stay within the unidentified leakage limits described in Technical Specification 3.4.5. If limits are exceeded, the Technical Specifications require corrective actions, up to and including shutting down the reactor. Multiple methods of RCS leakage detection are required to be operable during plant operation as specified in Technical Specification 3.4.7 and these instruments can detect and monitor leakage rates as described in SDAA Section 5.2.5.1. The containment evacuation system (CES) detects leakage into the containment vessel from unidentified sources at rates greater than or equal to 0.05 gpm using either pressure or sample tank for flow rate. The response time for an unidentified source greater than 1 gpm is less than one hour. NRC review regarding the reactor coolant leakage detection sensitivity of the instrumentation and the compliance to RG 1.45 were raised and addressed in docketed audit response A-5.2.5-2.

For AOOs that could potentially impact the CRDM housings, SDAA Chapter 15 discusses the potential scenarios and impact of those transients. Standard Design Approval Application Section 15.6 describes the design basis events associated with a decrease in RCS inventory, which includes the inadvertent opening of the reactor safety valve. While not a credible event, the thermal-hydraulic response of a decrease in RCS inventory through a failed CRDM connection would be bounded by this AOO, which itself is bounded by the inadvertent operation of the ECCS. These connections and their potential failures are bounded by the transients analyzed in SDAA Chapter 15. Section 15.4.8 describes the spectrum of rod ejection accidents and the considerations for the subsequent thermal hydraulic loads. The design of the CRDM housings and their connections results in a mechanical failure being a non-credible event, aligning with the DCA SER 3.5.1.1.4.

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NuScale Response to: 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).

Given the criteria discussed above, the CRDM connection design anticipates the stresses upon the RPV in its design function. The CRDM connections and the housings themselves are classified as ASME Class 1 appurtenances and function as an extension of the vessel itself, as stated above and supported by DCA SER 3.9.4.4.7. With these considerations, a missile is not a credible threat as the design basis criteria include the integrity of the connection and housing for the expected loads and transients experienced during the plants operating lifetime.

This same justification was provided in the US600 DCA Section 3.5.1.2, which states A control rod drive mechanism (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, is non-credible. The CRDM housing is a Class 1 appurtenance per ASME Section III.

The NRC approved the US600 design basis. The design functions and code requirements for the pressure housings and their connections are unchanged. The configuration of the housings does not degrade or alter the design basis of these connections. Further, beyond the design basis events analyzed, the generation of a missile by a bolt or piece of the flange does not have significant mass or velocity to present a hazard for the purposes of risk-significance in a severe accident scenario. The critical equipment would not be affected by this small mass. This aligns with the evaluation statements in the DCA SER 3.5.1.1.4 eliminating nuts and bolts as missile concerns.

Conclusion NuScale summarizes the responses to the questions as follows:

1.

A summary list of RPV connections with their respective design functions and specifications is provided in Table 1.

2.

In a clarification meeting May 7 & 8, an analysis of the RVV and RRV valve connections was agreed upon by NuScale to represent a standardized connection applicable to the CRDM connections.

3.

For the request pertaining to the stress and CUF analysis, the selection of a bolted connection designed to NB-3200 code is consistent with the US600 DCA. This basis was accepted by the NRC for the DCA without needing additional calculations for stress and NuScale Nonproprietary NuScale Nonproprietary

CUF ahead of the ASME Section III NCA-1210 design report. The change from welded to bolted connections does not deviate from this basis.

4.

The stress limits for the CRDM connections under NB-3200 are more conservative than BTP 3-4 and more conservative fatigue limits are not justified.

5.

A failure at the connection is not considered credible since the connections are designed to NB-3200 code. The CES provides additional confidence for uncertainty as it detects unidentified leakage, and operators must take action, including the option for shutting down the reactor.

6.

The generation of missiles is not considered credible given the aforementioned design to NB-3200. To address additional uncertainty and consequence, a failure of these components does not provide sufficient energy to threaten the design functions of components located in this area of the NuScale Power Module.

The SDAA licensing basis provides sufficient justification for the designs capability to fulfill the requirements of 10 CFR 50.55a, GDC 1, 2,4,14, 26, 27, 28, and 29. The US460 design follows the same basis as the US600 without deviating from the same standards accepted during US600 DCA review. The change from welded to bolted connections does not alter the conclusions of the DCAA safety evaluation of the CRDMs not being considered credible missile sources.

The US460 design for the CRDM housings and their connections fulfills SRP 3.5.1.2 AC 1 as it follows the same design basis for the connection as the US600 design. The housings and their connections are ASME BPV Code Class 1 appurtenances and are thus not considered credible missiles, in alignment with the DCA SER 3.5.1.1.4.

The US460 design for the CRDM housings and their connections fulfills the acceptance criteria of SRP 3.9.4 AC 1.A as the CRDM is designed and will be constructed in accordance with the requirements of ASME Section III code for Class 1 appurtenances and the requisite NB-3200 criteria.

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Table 1: List of Connections on RPV US600 (see RAIs 9358, 9436, 9437, 9188)

US460 ID RPV Component RPV Connection Postulated Break Design Criteria Inspection Criteria (NOTE 1, NOTE 3)

RPV Connection Postulated Break Design Criteria Inspection Criteria (NOTE 1, NOTE 4)

A B

ECCS RRVs ECCS RVVs Bolted (w/

inserts)

LOCA DBEs (DCA Chap.15):

No break postulated. Inadvertent opening events are postulated and analyzed (see DCA Section 15.6).

Dynamic Effects (DCA Chap.3.6):

Break Exclusion applied. No break postulated.

Break Exclusion: Yes Break exclusion criteria are applied from BTP 3-4 B.A(ii) due to RAI resolution (DCA Section 3.6.2.7).

Bolting - Design criteria per NB-3200 (2013 Ed.), which provides more margin to yielding than rules for piping NB-3653 with BTP 3-4 B.A(ii) applied [see RAI 9358, 03.06.02-17S5, and DCA 3.6.2.7]

ASME Section XI Category B-G-2 (VT-1) plus the following augmented ISI requirements:

Bolting - UT inspection of bolts if the connection is disassembled during the interval, if not disassembled during the interval volumetric inspection performed in-place. Additionally, sampling is not permitted, all ECCS valve bolts are inspected during each inspection interval.

[Ref. DCA Section 3.6.2.7]

Insert - VT-1 for all insert seal welds (see RAI 9188, Ref. DCA Table 5.2-4)

Same Same Same(NOTE 2)

ASME Section XI B-G-2 (VT-1) plus the following augmented ISI requirements:

Bolting - Same (see SDAA 3.6.2.7)

Inserts - VT-1 of ECCS valve bolting threaded inserts and their seal welds in support of BTP 3-4 (Audit Item A-15.6.5-13)

C Reactor Safety Valve Bolted (w/

inserts)

LOCA DBEs (SDAA Chap.15):

No break postulated. Inadvertent opening events are postulated and analyzed (see DCA Section 15.6).

Dynamic Effects (DCA Chap.3.6):

No break postulated.

Break Exclusion: No. Not piping, BTP 3-4 not applied.

Bolting - Bolting design criteria are per NB-3200 Bolting - ASME Section XI Category B-G-2 (VT-1 of bolting) (Ref. DCA Section 5.2.4)

Insert - VT-1 for all insert seal welds (see RAI 9188, Ref. DCA Table 5.2-4)

Same Same Same(NOTE 2)

Bolting - Same. (Ref. SDAA Section 5.2.4)

Inserts - VT-1 of threaded insert and seal weld when disassembled during the interval.

D E

F G

SG Feedwater Plenum Access Cover SG Main Steam Plenum Access Cover PZR Heater Bundle Instrument Seal Assembly (ISA)

Bolted (w/

inserts)

LOCA DBEs (SDAA Chap.15):

No break postulated.

Dynamic Effects (DCA Chap.3.6):

No break postulated.

Break Exclusion: No. Not piping, BTP 3-4 not applied.

Bolting - Bolting design criteria are per NB-3200 Bolting - ASME Section XI Category B-G-2 (VT-1 of bolting) (Ref. DCA Section 5.2.4)

Insert - VT-1 for all insert seal welds (see RAI 9188, Ref. DCA Table 5.2-4)

Same Same Same(NOTE 2)

Bolting - Same. (Ref. SDAA Section 5.2.4)

Inserts - VT-1 of threaded insert and seal weld when disassembled during the interval.

H CRDM Pressure Housings Welded LOCA DBEs (SDAA Chap.15):

No break postulated.

Dynamic Effects (DCA Chap.3.6):

No break postulated.

Break Exclusion: No. Not piping, BTP 3-4 not applied.

Weld - Design criteria are per NB-3200 Welds - ASME Section XI Category B-O (Volumetric or Surface every interval) (Ref. DCA Table 5.2-6)

Bolted (w/

inserts)

Same Same(NOTE 2)

Bolting - ASME Section XI Category B-G-2 (VT-1 of bolting)

(Ref. SDAA Section 5.2.4)

Inserts - VT-1 of threaded insert and seal weld when disassembled during the interval.

I CRDM Support Structure Welded Not pressure retaining. No break postulated.

Break Exclusion: No. Not piping, BTP 3-4 not applied.

Designed to Section III, Subsection NF Not pressure retaining. ASME Section XI Category F-A (VT-3)

Bolted (w/

inserts)

Same Same Same Inserts - VT-1 of threaded insert and seal weld when disassembled during the interval.

Notes:

Notes:

1.

In addition to component specific ISI elements, there are inspection requirements for the RCPB (i.e. entire RPV head), ASME Section XI Category B-P Bare Metal Exam (VT-2).

2.

DCA used ASME BPVC Section III 2013 Edition where bolting design criteria are within NB-3200. SDAA uses ASME BPVC Section III 2017 Edition, NB-3200 refers to Mandatory Appendix XIII for bolting criteria.

3.

DCA has COL Item 3.13-1 for ISI program for Class 1/2/3 threaded fasteners, COL item 5.2-6 to develop site-specific PSI/ISI/IST programs, COL Item 6.6-2 to develop PSI/ISI program referencing DCA standard design 4.

SDA has COL Item 3.13-1 for ISI program for Class 1/2/3 threaded fasteners, COL Item 5.2-4 to develop site-specific PSI/ISI/IST programs, COL Item 6.6-1 to develop PSI/ISI program referencing SDAA standard design.

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