ML19207A257

LLC Response to NRC Request for Additional Information No. 427 (Erai No. 9408) on the NuScale Design Certification Application

ML19207A257
Person / Time
Site:	05200046
Issue date:	07/26/2019
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
References
RAIO-0719-66436
Download: ML19207A257 (5)

Text

RAIO-0719-66436 July 26, 2019 Docket No.52-048 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.

427 (eRAI No. 9408) on the NuScale Design Certification Application

REFERENCES:

1. U.S. Nuclear Regulatory Commission, "Request for Additional Information No. 427 (eRAI No. 9408)," dated April 17, 2018

2. NuScale Power, LLC Response to NRC "Request for Additional Information No. 427 (eRAI No. 9408)," dated July 25, 2018

3. NuScale Power, LLC Response to NRC "Request for Additional Information No. 427 (eRAI No. 9408)," dated February 8, 2019 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 eRAI No. 9408:

03.09.02-77 The response to questions 03.09.02-75 and 03.09.02-76 were provided in Reference 2 and 3.

The response to questions 03.09.02-73 and 03.09.02-74 will be provided by July 31, 2019.

This letter and the enclosed response make no new regulatory commitments and no revisions to any existing regulatory commitments.

If you have any questions on this response, please contact Marty Bryan at 541-452-7172 or at mbryan@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8H12 Samuel Lee, NRC, OWFN-8H12 Marieliz Vera, NRC, OWFN-8H12 : NuScale Response to NRC Request for Additional Information eRAI No. 9408 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0719-66436 :

NuScale Response to NRC Request for Additional Information eRAI No. 9408 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9408 Date of RAI Issue: 04/17/2018 NRC Question No.: 03.09.02-77 10 CFR 50, Appendix A, GDC 4 requires structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents. Given the uncertainties and potential non- conservatisms regarding the CRAGT FIV analyses and potential significant degradation, low margin against VS lock- in for ICIGT, CRDS, and SG tubes, low margin against FEI of SG tubes, and lack of LFI analyses or testing of RVI (with the exception of the SG inlet flow restrictors), provide an updated initial startup testing plan which includes instrumentation and test conditions to confirm the long-term integrity of those components. Also provide pre-test predictions of responses at instrumentation locations and acceptance criteria.

DCD Tier 2, Table 5.2-7 provides the long-term in-service inspection (ISI) method for RVI components such as the ICIGTs, and CRAGTs. Typically the components in the ISI program are inspected at the end of the 10-year interval. The components identified in the CVAP as susceptible to FIV may need more frequent inspections to identify abnormal wear early. It is unclear to the staff whether these components will have additional inspections prior to the end of the 10-year ISI program interval. Finally, summarize or provide the DCD sections that discusses the long term inspection plan for the CRD shafts and SG inlet flow restrictors.

Alternatively, NuScale may propose other options to resolve the staff's concerns.

Update the comprehensive vibration assessment program report TR-0716-50439 or the measurement and inspection plan to include the requested information.

NuScale Nonproprietary

NuScale Response:

The initial startup vibration monitoring plan is outlined in Section 6.0 of the NuScale Comprehensive Vibration Assessment Program (CVAP) Measurement and Inspection Plan Technical Report, TR-0918-60894. Vibration monitoring is performed on the lead NuScale Power Module (NPM) unit in order to detect unexpected, large amplitude flow-induced vibration (FIV) during the initial startup test phase. These measurements supplement the scope of the detailed validation measurement program, discussed in other sections of this technical report, by confirming the lack of large amplitude vibration in the prototype NPM. Optimal sensor type and installation details are provided for each NPM region of interest in Section 6.3 of TR-0918-60894. These regions of interest include the steam generator assembly, in-core instrument guide tubes (ICIGT), control rod drive (CRD) shaft support and shaft sleeve, and the connection of the upper and lower riser assemblies. Pre-test predictions for the purpose of CVAP analysis validation, per Regulatory Guide 1.20, are not performed for this vibration monitoring, because there are no bias errors for tests performed at normal operating conditions in the installed NPM.

Anticipated vibration amplitudes for the FIV mechanisms being tested are defined per Section 6.0 of TR-0918-60894.

The technical issue related to control rod assembly guide tube (CRAGT) wear was addressed in the NuScale response to RAI 9408 Question 03.09.02-75, with a status of resolved-closed. The low margin against vortex shedding lock-in for the ICIGT and CRD shaft is no longer a concern due to design changes to these components. Refer to the response to RAI 8884, Question 03.09.02-9 for a description of these changes and their improvements to the vortex shedding safety margins. Quantitative screening of reactor module components for leakage flow instability (LFI) has been added to Section 2.3.8 of the NuScale CVAP Technical Report, TR-0716-50439.

See also the improved margins to vortex shedding and fluid elastic instability for the steam generator tubes discussed in Sections 3.2.1 and 3.2.2 of TR-0716-50439.

The Inservice Inspection (ISI) program is outlined in the FSAR Section 5.2.4, Reactor Coolant Pressure Boundary Inservice Inspection and Testing. This program detects degradation in components over the life of the plant, before the ability of the components to perform their functions is compromised. The requirements and guidelines for the reactor vessel internals ISI program are based on ASME BPVC,Section XI and MRP 227. The MRP 227 addresses the screening of components for various degradation mechanisms, including wear, based on operating experience with similar components. The NuScale ISI program follows the MRP approach without exception.

The CVAP results show that the components screened as being susceptible to FIV have positive safety margin and that FIV corresponds to small or zero fatigue usage. Components NuScale Nonproprietary

with less than 100% margin are tested to provide a precise characterization of the safety margin.

The CVAP, as allowed in Regulatory Guide 1.20, specifically excludes ISI from its scope.

The SG inlet flow restrictors are not part of the NuScale ISI program. As described in FSAR Section 5.4.1.5, the inlet flow restrictors are non-structural attachments to the RPV, for which inspections are not required per ASME Section XI rules. Combined License (COL) Item 5.4-1 requires that the COL applicant develop and implement a steam generator program for periodic monitoring of the degradation of the steam generator components. As the inlet flow restrictors interface with the SG tubes, the specific component inspections therefore are controlled by the COL applicants steam generator program.

As described in FSAR Section 3.9.5, the upper CRD shaft supports are classified as internal structures per ASME Subsection NG. In accordance with ASME Section XI, Table IWB-2500-1 (B-N-1, B-N-2, B-N-3), the types of internal structures that require inspection are core support structures and interior attachments. The upper CRD shaft supports are neither of these types, thus inspections are not required per ASME rules. However, based on the guidance of MRP-227, inspections of the internal structures have been specified. Therefore, the upper CRD shaft supports receive a VT-3 examination every 10 years.

The CRD shaft support is made from dual certified 304/304L stainless steel and the CRD shaft is made from 410 stainless steel. The 410 stainless is a harder material than dual certified 304/304L. Additionally, some of the CRD shafts when operating move vertically such that the same section of the shaft is not always adjacent to a support. The CDR shaft support is at a fixed position and any wear would be localized. For these reasons the CRD shaft supports are more susceptible to wear than the CRD shafts. Accordingly, VT-3 examinations of the CRD shaft supports are planned.

Additionally, per FSAR Section 3.9.4.4 and COL Item 3.9-11, the CRD System design is subject to an Operability Assurance Program (OAP). The OAP is a series of tests designed to demonstrate acceptable performance with respect to wear and other design requirements.

Impact on DCA:

The CVAP Analysis Technical Report TR-0716-50439 and CVAP Measurement and Inspection Plan Technical Report TR-0918-60894 will be revised as described in the response above. The technical report revisions will be submitted separately.

NuScale Nonproprietary