ML24291A085

NPM-20 - Nonproprietary - NuScale SDAA Sections 15.0 and 15.1 - Request for Additional Information No. 037 (RAI-10357-R1)

ML24291A085
Person / Time
Site:	99902078
Issue date:	10/17/2024
From:	NRC
To:	NRC/NRR/DNRL/NRLB
References
Download: ML24291A085 (9)

Text

From:

Getachew Tesfaye Sent:

Thursday, October 17, 2024 9:12 AM To:

Request for Additional Information Cc:

Stacy Joseph; Mahmoud -MJ-Jardaneh; Griffith, Thomas; Bode, Amanda; NuScale-SDA-720RAIsPEm Resource

Subject:

Nonproprietary - NuScale SDAA Sections 15.0 and 15.1 - Request for Additional Information No. 037 (RAI-10357-R1)

Attachments:

SECTION 15.05 and 15.1 - RAI-10357-R1-FINAL NON-PROPRIETARY.pdf Attached please find NRC staffs nonproprietary 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). The encrypted proprietary version will be submitted in a separate email.

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:

40 Mail Envelope Properties (DM6PR09MB56863E7D1AF4B7E159FC6C8C8C472)

Subject:

Nonproprietary - NuScale SDAA Sections 15.0 and 15.1 - Request for Additional Information No. 037 (RAI-10357-R1)

Sent Date:

10/17/2024 9:12:11 AM Received Date:

10/17/2024 9:12:15 AM From:

Getachew Tesfaye Created By:

Getachew.Tesfaye@nrc.gov Recipients:

"Stacy Joseph" <stacy.joseph@nrc.gov>

Tracking Status: None "Mahmoud -MJ-Jardaneh" <Mahmoud.Jardaneh@nrc.gov>

Tracking Status: None "Griffith, Thomas" <tgriffith@nuscalepower.com>

Tracking Status: None "Bode, Amanda" <abode@nuscalepower.com>

Tracking Status: None "NuScale-SDA-720RAIsPEm Resource" <NuScale-SDA-720RAIsPEm.Resource@nrc.gov>

Tracking Status: None "Request for Additional Information" <RAI@nuscalepower.com>

Tracking Status: None Post Office:

DM6PR09MB5686.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 673 10/17/2024 9:12:15 AM SECTION 15.05 and 15.1 - RAI-10357-R1-FINAL NON-PROPRIETARY.pdf 165341 Options Priority:

Normal Return Notification:

No Reply Requested:

No Sensitivity:

Normal Expiration Date:

1 REQUEST FOR ADDITIONAL INFORMATION No. 037 (RAI-10357-R1)

BY THE OFFICE OF NUCLEAR REACTOR REGULATION NUSCALE STANDARD DESIGN APPROVAL APPLICATION DOCKET NO. 05200050 CHAPTER 15, TRANSIENT AND ACCIDENT ANALYSES SECTION 15.0.5, EXTENDED PASSIVE COOLING FOR DECAY AND RESIDUAL HEAT SECTION 15.1.1, DECREASE IN FEEDWATER TEMPERATURE ISSUE DATE: 10/17/2024

=

Background===

By letter dated October 31, 2023, NuScale Power, LLC (NuScale or the applicant) submitted Revision 1 of its US460 standard plant design approval application (SDAA) (Agencywide Documents Access and Management System Accession No. ML23306A033). Supporting documents TR-0516-49416-P, Non-Loss-of-Coolant Accident Analysis Methodology, Revision 4 (ML23005A306) (Non-LOCA LTR) and TR-124587-P, "Extended Passive Cooling and Reactivity Control Methodology," Revision 0 (ML23005A308) (XPC LTR), were submitted on January 5, 2023. The applicant submitted the US460 standard plant (NPM-20) 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 the Final Safety Analysis Report (FSAR) provided in SDAA Part 2, specifically information in Chapter 15, Transient and Accident Analysis, (ML23304A365) and other FSAR Chapters as necessary. The NRC staff has determined that additional information is required to complete its review.

Regulatory Basis General Design Criterion (GDC) 5, Sharing of structures, systems and components, in 10 CFR Part 50, Appendix A, requires that any sharing among nuclear power units of structures, systems, and components (SSCs) important to safety will not significantly impair their safety function.

GDC 10, Reactor Design, requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLS) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs).

GDC 13, Instrumentation and control, requires that instrumentation be provided to monitor variables and systems over anticipated ranges for normal operations, for AOOs, and for accident conditions and that controls be provided to maintain these variables and systems within prescribed operating ranges.

GDC 15, Reactor coolant system design, requires that the reactor coolant system (RCS) and its associated auxiliaries shall be designed with sufficient margin to assure that the design conditions of the pressure boundary are not exceeded during normal operations, including AOOs.

GDC 17, Electric power systems, requires that an onsite and offsite electric power system shall be provided to permit the functioning of SSCs important to safety. The safety function for each system (assuming the other system is not working) shall be to provide sufficient capacity and capability to ensure that the acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary (RCPB) are not exceeded as a result of

2 an AOO and that core cooling, containment integrity, and other vital functions are maintained in the event of an accident.

GDC 20, Protection system functions, requires that the reactor protection system shall be designed to initiate automatically the operation of appropriate systems, including the reactivity control systems, to ensure that the plant does not exceed SAFDLs during any condition of normal operation, including AOOs.

GDC 25, Protection system requirements for reactivity control malfunctions, requires that the reactor protection system shall be designed to ensure that SAFDLs are not exceeded for any single malfunction of the reactivity control system, such as accidental withdrawal of control rods.

GDC 26, Reactivity control system redundancy and capability, requires that a reactivity control system shall be provided that is capable of reliably controlling reactivity changes to assure that SAFDLs are not exceeded even during AOOs. This is accomplished by ensuring that the applicant has allowed an appropriate margin for malfunctions such as stuck rods.

GDC 27, Combined reactivity control systems capability, requires that reactivity control systems shall be designed with the combined capability of reliably controlling reactivity changes to assure that, under postulated accident conditions and with appropriate margin for stuck rods, the capability to cool the core is maintained.

GDC 28, Reactivity limits, requires that reactivity control systems shall 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 RCPB greater than limited local yielding nor (2) sufficiently disturb the core, its support structures, or other reactor pressure vessel (RPV) internals to impair significantly the capability to cool the core.

GDC 31, Fracture prevention of reactor coolant pressure boundary, requires that the RCS shall be designed with sufficient margin to ensure that the boundary behaves in a nonbrittle manner, and that the probability of propagating fracture is minimized.

GDC 34, Residual heat removal, requires that a system shall be provided with the capability to transfer decay heat and other residual heat from the reactor so that SAFDLs and pressure boundary design limits are not exceeded.

10 CFR 50.46 provides the acceptance criteria for emergency core cooling systems (ECCS) for light-water nuclear power reactors.

10 CFR 52.137(a)(2) which states a standard design application must include [a]

description and analysis of the SSCs of the facility, with emphasis upon performance requirements, the bases, with technical justification, upon which the requirements have been established, and the evaluations required to show that safety functions will be accomplished. Additionally, 10 CFR 52.137(a)(4) which states a standard design application must include [a]n analysis and evaluation of the design and performance of SSC with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of SSCs provided for the prevention of accidents and the mitigation of the consequences of accidents.

3 Question 15.0.5-1 Issue FSAR Section 15.0.5 is missing details of the design that are used in the SDA analyses when applying the Extended Passive Cooling and Reactivity Control topical report (XPC LTR). The design of the ECCS Supplemental Boron (ESB) system needs to be sufficiently complete to resolve all safety issues, and the application needs to include system descriptions sufficient to permit understanding of the system designs and their relationship to the safety evaluations. It is noted that during the audit, NuScale documentation stated that ((. NuScale also stated that (( }}. Information Requested The following information is missing from the FSAR and needs to be provided in FSAR markups:

1. The size, location, orientation, key dimensions, and resistances used in the analyses for the boron dissolver basket, containment mixing pipes and respective collector rail(s).

To address this request during the audit, NuScale provided FSAR markups that include the ESB dissolver basket and mixing tube minimum collection area, nominal dissolver basket diameter, nominal mixing tube size, and dissolver basket and mixing tube minimum condensate channel flow capacity. In an effort to make a risk informed safety-finding, as an alternative to providing the specific information requested above, the ESB FSAR markups provided during the audit are sufficient provided NuScale includes a test in their initial test program to confirm the as-built ESB design adequately reflects the analysis response, as requested in RAI Question 6.3-7.

2. The flow area and corresponding losses used in the analyses for the riser holes.

3. The description of the application of the XPC methodology (Section 5.2.3 in the XPC LTR) regarding the assessment of lower riser hole flow to justify the use of the rates in SDA boron transport analysis needs to be included in the SDA. In addition, the results as requested in Question 15.0.5-6 below, need to be incorporated into this description.

Question 15.0.5-4 Issue Section 15.0.5.3.2 describes the long-term cooling minimum collapsed liquid level analyses. The Steam Generator Tube Failure (SGTF) event is identified as the most limiting event for minimum collapsed liquid level above the top of active fuel. The collapsed liquid level for the SGTF case is 0.23 ft (2.76 inches) above the top of active fuel. This is lower than the other limiting events, breaks outside of containment, which have a calculated collapsed liquid level of at least 1.57 ft (18.8 inches) above the top of active fuel. During its audit, the staff observed that the SGTF event appears to be modeled differently with NRELAP5 than how it is modeled for the SGTF event presented in SDA Section 15.6.3 using the methodology from the non-LOCA LTR. Given the minimal margin in the collapsed liquid level for the SGTF event presented in SDA Section 15.0.5, the apparent difference in modeling from

4 the SGTF event presented in 15.6.3 and the notable difference between the results from the other limiting long term cooling cases the staff is seeking information on the modeling of the event and clarification of the modeling in the SDA. Additionally, the staff is seeking information and clarification of assumptions, including conservatisms, used. Information Requested In addition to the expected general plant response that is already provided in the FSAR, revise the FSAR to describe how the SGTF case is modeled, relative to the break outside containment cases, and an explanation of why the SGTF case has a lower collapsed liquid level. Revise the FSAR to describe assumptions and conservatism in the modeling of the SGTF case. Question 15.0.5-5 Issue FSAR Section 15.0.5 is missing a summary and results of evaluations for extended DHRS cooling up to 8 hours (prior to ECCS actuation) that demonstrate subcriticality is maintained (1) at EOC conditions, (2) and over a range of powers. In addition, the evaluations need to consider boron transport when the upper riser holes are uncovered. The moderator temperature during extended DHRS cooling (without ECCS actuation) can fall below the temperature used for shutdown margin, meaning that the SDM calculations do not ensure subcriticality with the highest worth CRA fully withdrawn A key aspect of demonstrating subcriticality before ECCS actuation is that the boron concentration in the downcomer remains above the critical boron concentration when boron redistribution effects are maximized. In addition, subcriticality needs to be demonstrated for conditions where boron redistribution effects are less important and moderator temperature effects are maximized. Information Requested Revise FSAR Section 15.0.5 to include a summary of evaluations for extended DHRS cooling up to 8 hours (prior to ECCS actuation) sufficient to demonstrate subcriticality in the scenarios described above. Question 15.0.5-6 Issue FSAR Section 15.0.5 is missing the results from the lower riser hole flow evaluation calculations, as required by the XPC topical report methodology, that shows that the flow rates used in the boron transport analysis are justified. The results are needed to show that the applicable regulations identified by NuScale in the XPC topical report are met when the methodology is applied to the NPM-20 design. The calculation results from the application of the topical report methodology to a particular design is within the scope of an FSAR and is not considered proprietary similar to the results of the application of other topical report methodologies. The results from the lower riser hole flow evaluation calculations are used in the boron transport either directly as stated in the XPC topical report methodology or as an input based on its results. The XPC topical report methodology states This section provides an acceptable approach for evaluation of riser hole flow rates for use in boron transport analyses, or for justification of use of rates determined from a different source. During the audit, the staff observed that the NuScale calculation that appears to contain the lower riser hole flow evaluation calculations states that ((

5 }}. The results from the lower riser hole flow evaluation calculations are required to perform the boron transport safety analysis and therefore must be verified per 10 CFR 50 Appendix B. Information Requested Revise FSAR Section 15.0.5 to present the results of the lower riser hole flow evaluation calculations. These results need to be calculated following the XPC topical report methodology and show that the flow rates used in the boron transport analysis are appropriate. The lower riser hole flow evaluation calculation results need to be verified by NuScale in accordance with 10 CFR 50 Appendix B given the calculations are used in the boron transport analyses either directly as stated in the XPC topical report methodology or as an input based on its results. Question 15.1.1-7 Issue As described in FSAR Section 15.1, malfunctions resulting in an increase in primary heat removal by the secondary system cause the moderator density to increase. Increased moderator density in the downcomer region affects the excore detector signal. NuScale models an increased high power level trip setpoint during overcooling events to account for this factor. NRC staff audited (ML23067A300) calculations supporting this modification to the high power level trip setpoint and found that the modification applied to the safety analysis is larger than the calculated value. Changes in downcomer temperature will also affect reactor trip on the high power rate signal. However, the FSAR and methodology documentation, including information provided during the audit, does not contain sufficient information concerning modeling of the high power rate trip setpoint during overcooling events. During the audit, NuScale clarified that the high power rate trip is assumed to operate in the sensitivity studies on the decrease in feedwater temperature event discussed in FSAR Section 15.1.1 and the increase in steam flow event discussed in FSAR Section 15.1.3. It is unclear whether inferences made from sensitivity studies in which the high power rate signal results in a reactor trip are applicable to limiting cases that do not model operation of this trip, since terminating the transient at an earlier time may obscure trends resulting from variation of other parameters (e.g., initial power level) in these sensitivity studies. Additionally, the staff noted during its audit that these sensitivity studies do not account for the effect of changes in downcomer moderator temperature and density on the high power rate trip. It is unclear to NRC staff whether this approach will result in an accurate or conservative assessment of reactor trip time. Information Requested Revise calculations supporting the decrease in feedwater temperature and increase in steam flow events such that modeled reactor trip signals are consistent between the limiting case and sensitivity studies, unless the sensitivity study is performed to examine modeling of a trip signal. If the high power rate trip will be credited in limiting cases, confirm that the modeled change in excore detector signal from a change in downcomer temperature is conservative or accurate by providing, for limiting cases, the magnitude of the modification to the excore detector signal or trip setpoint, trip time, and by showing progressions of downcomer temperature, core average

6 temperature, actual reactor power, and indicated reactor power, and other parameters as necessary such that the phenomena leading to the high power rate trip are clear. Additionally, update portions of the licensing basis discussing the methodology for evaluation of overcooling transients (i.e., the non-LOCA LTR or a subsection of FSAR 15.0) to specify how the evaluation model accounts for changes in excore detector signal caused by changes in downcomer temperature when assessing high power rate trips. Provide justification that this methodology results in a conservative or accurate trip time. If the magnitude of this effect is biased in a specific direction, provide justification that trip time will be accurate or conservative in circumstances in which the effect encourages a high power rate trip (e.g., downcomer temperature is increasing when a positive rate trip is assessed) and discourages a high power rate trip (e.g., downcomer temperature is decreasing when a positive rate trip is assessed), as information provided during the audit does not address all of these circumstances. If justification for this biasing is based on a specific event progression or timing of the high power rate trip, state this basis such that conservatism or appropriateness of the approach for overcooling events is clear. Specify any modeling details (e.g., method of modeling this effect such as a scalar multiplication on core power, formulas used to assess the high power rate trip) needed to support justification that the specific method is conservative, as information provided during the audit does not sufficiently specify this information. Question 15.1.1-8, NonLOCA.LTR-53 Issue Table 7-7 of the Non-LOCA LTR states that an applicant implementing the methodology will perform sensitivity studies on initial feedwater (FW) temperature and initial steam generator (SG) pressure in order to identify the limiting MCHFR for the decrease in feedwater temperature event. The applicant provided for NRC staff audit (ML23067A300) the sensitivity studies to support the standard design approval application for the NPM-20. However, the staff noted that the initial feedwater temperature and initial steam generator pressure were varied simultaneously in these studies. The staff also noted that these studies varied bypass of the high power rate trip and operation of the CRA regulating bank. Because these four parameters are varied simultaneously, the impact of any individual change on MCHFR cannot be determined. As such, the NRC staff cannot conclude that initial SG pressure and FW temperature were varied to identify the limiting MCHFR consistent with the Non-LOCA LTR evaluation model. During the staffs audit, the applicant stated that examining biased-high initial feedwater temperature (( }} that is evaluated in the decrease in feedwater temperature event. The staff considers evaluation of cases with biased-high initial feedwater temperature to be a necessary step to ensure that the limiting MCHFR is identified, but it is not clear whether ((

}} will minimize margin to acceptance criteria.

During the audit discussion, the applicant also stated that maintaining consistent steam pressure allows comparison of the cooldown effect driven by the initiating event (i.e., decreasing feedwater temperature). However, NRC staff does not consider that studying the cooldown effect with consistent initial SG pressure precludes variation of initial SG pressure in a separate set of cases in order to assure that the limiting MCHFR is identified.

7 The staff also noted additional calculations that deviate from the methodologies defined in the Non-LOCA LTR. The audited calculation that supports the single CRA withdrawal evaluation, uses a (( }}. However, Table 7-68 of the Non-LOCA LTR specifies that a biased-low initial fuel temperature should be used. The audited calculation does not specify that assuming a ((

}} deviates from the LTR methodology nor does it provide an acceptable justification for deviation from the proposed NPM-20 licensing basis. The audited calculation also states that using a biased-low initial fuel temperature is more conservative.

Information Requested

1. Provide results of sensitivity studies that vary initial feedwater temperature independently of other parameters to demonstrate that the limiting MCHFR is identified for the decrease in feedwater temperature event.

2. Provide results of sensitivity studies that vary initial steam generator pressure independently of other parameters to demonstrate that the limiting MCHFR is identified for the decrease in feedwater temperature event.

3. Provide revised calculations for the single CRA withdrawal event with initial fuel temperature assumptions that bias initial fuel temperature consistent with the Non-LOCA LTR methodology, or provide appropriate justification for use of the (( }} initial fuel temperature and modify the FSAR to describe the deviation from the LTR methodology.}}