ML24281A018
| ML24281A018 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 10/07/2024 |
| From: | Shaver M NuScale |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| RAIO-174796 | |
| Download: ML24281A018 (1) | |
Text
RAIO-174796 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com October 07, 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. 017 (RAI-10135 R1) on the NuScale Standard Design Approval Application
REFERENCE:
- 1. NRC Letter to NuScale, Request for Additional Information No. 017 (RAI-10135 R1), dated March 02, 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 the NuScale response to the following RAI question from NRC RAI-10177 R1:
3.6.2.7-2 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 October 07, 2024.
Sincerely, Mark W. Shaver Director, Regulatory Affairs NuScale Power, LLC
RAIO-174796 Page 2 of 2 10/07/2024 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-10135 R1, Question 3.6.2.7-2, nonproprietary
RAIO-174796 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-10135 R1, Question 3.6.2.7-2, nonproprietary
Response to Request for Additional Information Docket: 052000050 RAI No.: 10135 Date of RAI Issue: 03/02/2024 NRC Question No.: 3.6.2.7-2 Regulatory Basis
- GDC 4, Environmental and Dynamic Effects Design Bases, states that 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.
- GDC 14, Reactor coolant pressure boundary, states that the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.
- GDC 30, Quality of reactor coolant pressure boundary, states that the components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical.
- 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, describes, in part, that ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated.
Issue In NuScale response to Design Certification Application (DCA) RAI (E1902227t034905-RAIO-0219-64694, Supplemental RAI No 340, eRAI9358; ML19058A670), it was stated in part that containment leakage monitoring systems are sensitive to a leak rate as low as 0.01 lbm per minute (or ~0.001 gallon per minute). Whereas, in SDAA FSAR section 3.6.2.7, it was stated in NuScale Nonproprietary NuScale Nonproprietary
part that containment leakage monitoring systems are sensitive to a leak rate as low as 0.05 gallons per minute.
There is a significant change in the leakage system sensitivity, It is not clear whether the role of the leakage detection system is used for GDC 4 dynamic effects exclusion or LOCA break exclusion or both. It is also not clear whether each of the bolted connection to RV has its own dedicated leakage detection system and a limit for action.
Requested Information A. For bolted connections and piping inside CNV The sensitivity of the containment leak detection system is increased from 0.001 gallons per minute (gpm) in DCA to 0.05 gpm in SDAA. To support the staffs evaluation of the reduction in sensitivity, provide the following information:
(1) Clarification of the role of leakage detection system(s) for areas seeking exclusion of dynamic effects (i.e., GDC-4) and/or LOCA break exclusion, including the bolted connections of the RVV, RRV, CRDMs, I&C penetrations, etc.
(2) Explanation of the acceptable leakage limits and the ability of the leakage detection system to detect those leakage limits. Include a discussion of the leakage levels from the ECCS RRV and RVV, CRDM and other bolted flange connections, and leakage limits on piping systems, and whether there are different limits for a plant shutdown or additional inspection at the next opportunity.
(3) Explanation of how the allowed leakage limits ensure remedial actions are taken before unacceptable degradation occurs in the flanged joints or piping systems.
(4) Discussion of whether the bolted connections to the RV, including those for the RRVs, RVVs, and CRDMs, I&C penetrations, etc. will have dedicated leakage detection systems and associated actions to preempt significant degradation of these locations, or discuss other considerations (design, inspection) that address the potential for leakage to degrade the bolted connections.
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B. For connections and piping outside CNV NuScale is requesting approval for GDC break exclusion for areas outside containment such as within the containment isolation system (CNV nozzle safe-end to containment isolation valve connection). To support the staffs evaluation of this request, provide the following information:
(1) Clarification of the role of leakage detection system(s) for areas seeking exclusion of dynamic effects (i.e., GDC-4) outside containment.
(2) Explanation of the sensitivity of the leakage detection system and how this sensitivity supports detecting the leakage that ensures remedial actions are taken before unacceptable degradation occurs at CNV Nozzle safe end to containment isolation valve connections.
(3) Discussion of whether these connections to the CNV will have dedicated leakage detection systems and associated actions to preempt significant degradation of these locations, or if other considerations exist (design, inspection) that address the potential for leakage to degrade the connections.
In your response, include a markup of the relevant sections of the FSAR pertaining to leakage detection (including section A.2) of SDAA document TR-121507-P incorporating this information.
NuScale Response:
Response to Part A The design and capabilities of the leakage detection system have not changed between the US600 DCA and US460 SDAA. The leakage detection system is no longer credited for Leak Before Break (LBB) methodology, and therefore the sensitivity requirements relaxed from 0.001 gallons per minute (gpm) in the DCA to 0.05 gpm in the SDAA. This aligns with the Leakage-Monitoring-Related Positions specified in RG 1.45.
Leakage detection and monitoring is not provided for individual components inside containment.
Leakage into containment is monitored by the Containment Evacuation System and is considered unidentified reactor coolant leakage regardless of the source. The limit for this leakage is established in Technical Specification (TS) 3.4.5 (0.5 gpm). If this Limiting Condition for Operation (LCO) is exceeded, then required operator actions are performed as described in NuScale Nonproprietary NuScale Nonproprietary
TS 3.4.5. If leakage cannot be reduced to within limits in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, then operator action is required to shutdown the reactor to Mode 2 and subsequently Mode 3. TS 3.4.7 describes the Reactor Coolant System (RCS) leakage detection instruments that are required to be operable to support reactor coolant leakage monitoring (i.e. TS 3.4.5).
The capabilities of the leakage detection system described in the US460 SDAA meet the limits specified in RG 1.45 for the purpose of detecting degradation to the Reactor Coolant Pressure Boundary (RCPB). RCPB leakage is indicative of degradation of pressure retaining components that could ultimately result in a loss of component structural integrity. Therefore, leakage detection system requirements beyond the specifications of RG 1.45 do not result in a commensurate decrease in risk to limit the potential for gross failures of the pressure boundary, including the emergency core cooling system (ECCS) reactor vent valves and reactor recirculation valves, which are the bolted connections on the Reactor Pressure Vessel (RPV) that have applied break exclusion of dynamic effects (refer to the response for RAI 10134 Question 3.6.3-1).
The US460 SDAA applies break exclusion methodologies to the ECCS reactor vent valves and reactor recirculation valves as described in Pipe Rupture Hazards Analysis, TR-121507, Revision 0, Appendix A. This treatment includes the addition of inservice inspections as described in SDAA Chapter 3.6. In accordance with the requirements of BTP 3-4, the scope of postulated breaks is limited to piping. Control rod drive mechanism (CRDM) pressure housings, bolted access covers, instrument seal assemblies, reactor safety valves are not piping; and therefore, break exclusion criteria are not applied.
Response to Part B NuScales application of BTP 3-4 B.1.ii (i.e., break exclusion) is described in TR-121507, Revision 0, Appendix A. Section A.1 describes the BTP 3-4 B.1.ii criteria, which do not include leakage detection as a consideration. Section A.1.2 and Section A.1.3 describe NuScales adherence to BTP 3-4 B.1.ii criteria for systems outside containment for the Containment System piping and Decay Heat Removal System piping, respectively. As described in the referenced sections, required volumetric examinations are meant to preempt significant degradation of these locations. Additionally, more conservative piping stress criteria for these locations are satisfied. Leakage detection has no role in the application of break exclusion in these areas. These connections to the containment vessel (CNV) do not have dedicated (i.e.,
local) leakage detection.
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Impact on US460 SDAA:
There are no impacts to US460 SDAA as a result of this response.
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