ML18110A359

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LLC Response to NRC Request for Additional Information No. 372 (Erai No. 9364) on the NuScale Design Certification Application
ML18110A359
Person / Time
Site: NuScale
Issue date: 04/20/2018
From: Rad Z
NuScale
To:
Document Control Desk, Office of New Reactors
References
RAIO-0418-59636
Download: ML18110A359 (17)
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RAIO-0418-59636 April 20, 2018 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.

372 (eRAI No. 9364) on the NuScale Design Certification Application

REFERENCE:

U.S. Nuclear Regulatory Commission, "Request for Additional Information No.

372 (eRAI No. 9364)," dated February 27, 2018 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 Questions from NRC eRAI No. 9364:

14.03.03-8 14.03.03-9 14.03.03-10 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 Steven Mirsky at 240-833-3001 or at smirsky@nuscalepower.com.

Sincerely, Zackary W. Rad Director Regulatory Affairs

Director, NuScale Power, LLC Distribution: Samuel Lee, NRC, OWFN-8G9A Prosanta Chowdhury NRC, OWFN-8G9A Demetrius Murray, NRC, OWFN-8G9A : NuScale Response to NRC Request for Additional Information eRAI No. 9364 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0418-59636 :

NuScale Response to NRC Request for Additional Information eRAI No. 9364 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.: 9364 Date of RAI Issue: 02/27/2018 NRC Question No.: 14.03.03-8 10 CFR 52.47(b)(1) requires The proposed inspections, tests, analyses, and acceptance criteria that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, a facility that incorporates the design certification has been constructed and will be operated in conformity with the design certification, the provisions of the [Atomic Energy] Act, and the Commission's rules and regulations. In supporting this requirement, discrepancies have been identified in Tier 1 material. Furthermore, as the Tier 1 material becomes a part of the design certification rule, it is of the utmost importance that this information be free of errors. Below are some specific instances that should be addressed:

a) DCD Tier 1, Table 2.1-4, ITAAC #6 is inconsistent between Tier 1 and Tier 2.

Specifically, Tier 2 states that the initial RPV beltline Charpy upper-shelf energy is no less than 75 ft-lb but Tier 1 states greater than 75 ft-lb. The inconsistency is impossible to reconcile in the event that the test results are exactly 75 ft-lb. Correct this inconsistency in the DCD.

b) Tier 1, Page 2.1-1 contains a typographical error: The SG supports the RCS by suppling part of the RCPB (should be supplying). Correct the typographical error.

c) ASME Piping ITAAC (Table 2.1-4, ITAAC #1, for instance) needs to have an Acceptance Criteria that relates back to the Design Commitment, namely that the Report exists and concludes that the system meets the requirements of ASME Code Section III. This is also consistent with the Tier 2 discussion in Table 14.3-1. The current Acceptance Criteria wording for Table 2.1-4, ITAAC #1 specifies that a Report meets the Section III requirements for a Report - this has no direct tie to what the Design Commitment entails, namely that the piping system complies with ASME Code Section III requirements. As an example, please see Table 2.2-3 ITAAC #1. Correct the affected ITAAC.

d) The definition of ASME Code presented in Tier 1 Section 1.1 does not contain the provisions for conditions and alternatives contained in 10 CFR 50.55a. A verbatim interpretation of the definition would not allow the phrase ASME Code to account for the conditions and alternatives provided in 10 CFR 50.55a. Clarification should be added to indicate that the phrase ASME Code, as used in the DCD, means ASME Code, as endorsed in 10 CFR 50.55a.

NuScale Nonproprietary

e) The narrative in Tier 2 Table 14.3-1 for DCD Tier 1 Table 2.8-2 ITAAC #2 is inconsistent with the Tier 1 material, as it discusses seismic Category I equipment rather than Class 1E equipment.

NuScale Response:

Part a) response Tier 1, Section 2.1.1 Design Commitments; Tier 1, Table 2.1-4, ITAAC #6; and Tier 2 Table 14.3-1, ITAAC 02.01.06 are revised to agree with the minimum RPV beltline Charpy upper-shelf energy stated in Tier 2 Section 5.3.1.5.

Part b) response The typographical error was corrected in Revision 1 to the DCA. The sentence reads "The SG supports the RCS by supplying part of the RCPB."

Part c) response Tier 1, Table 2.1-4, ITAAC #1 Acceptance Criteria is revised to correlate to the Design Commitment.

Part d) response The definition of ASME Code in Tier 1, Section 1.1 is revised to include "as endorsed in 10 CFR 50.55a".

Part e) response The discussion of ITAAC 02.08.02 in Tier 2, Table 14.3-1 was revised in Revision 1 to the DCA and is consistent with Tier 1, Table 2.8-2, ITAAC #2 in that it discusses Class 1E equipment rather than seismic Category I equipment.

Impact on DCA:

Tier 1, Section 1.1 and Section 2.1.1, and Tier 1, Table 2.1-4 and Tier 2, Table 14.3-1 have been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

NuScale Tier 1 Definitions 1.1 Definitions The definitions below apply to terms that may be used in the design descriptions and associated Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC).

Acceptance Criteria refers to the performance, physical condition, or analysis result for structures, systems, and components (SSC), or program that demonstrates that the design commitment is met.

Analysis means a calculation, mathematical computation, or engineering or technical evaluation. Engineering or technical evaluations could include, but are not limited to, comparisons with operating experience or design of similar SSC.

As-built means the physical properties of an SSC following the completion of its installation or construction activities at its final location at the plant site. In cases where it is technically justifiable, determination of physical properties of the as-built SSC may be based on measurements, inspections, or tests that occur prior to installation, provided that subsequent fabrication, handling, installation, and testing do not alter the properties.

RAI 14.03.03-8 ASME Code meansSection III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, as endorsed in 10 CFR 50.55a, unless a different section of the ASME Code is specifically referenced.

ASME Code Data Report means a document that certifies that a component or system is constructed in accordance with the requirements of the ASME Code. This data is recorded on a form approved by the ASME.

Component, as used for reference to ASME Code components, means a vessel, concrete containment, pump, pressure relief valve, line valve, storage tank, piping system, or core support structure that is designed, constructed, and stamped in accordance with the rules of the ASME Code. ASME Code Section III classifies a metal containment as a vessel.

Design Commitment means that portion of the design description that is verified by ITAAC.

Design Description means that portion of the design that is certified. Design descriptions consist of a system description, system description tables, system description figures, and design commitments. System description tables and system description figures are only used when appropriate. The system description is not verified by ITAAC; only the design commitments are verified by ITAAC. System description tables and system description figures are only verified by ITAAC if they are referenced in the ITAAC table.

Inspect or Inspection means visual observations, physical examinations, or reviews of records based on visual observation or physical examination that compare (a) the SSC condition to one or more design commitments or (b) the program implementation elements to one or more program commitments, as applicable. Examples include walkdowns, configuration checks, measurements of dimensions, or nondestructive examinations. The terms, inspect and inspection, also apply to the review of Emergency Planning ITAAC requirements to determine whether ITAAC are met.

Tier 1 1.1-1 Draft Revision 2

NuScale Tier 1 NuScale Power Module

  • The CNTS supports the DHRS by closing CIVs for main steam valves and feedwater valves when actuated by MPS for DHRS operation.
  • The ECCS supports the RCS by opening the ECCS reactor vent valves and RRVs when their respective trip valve is actuated by MPS.
  • The DHRS supports the RCS by opening the DHRS actuation valves on a DHRS actuation signal.
  • The CNTS supports the MPS by providing electrical penetration assemblies to route instrument cables for MPS actuation through the CNV.

The NPM performs the following nonsafety-related, risk-significant function that is verified by Inspections, Tests, Analyses, and Acceptance Criteria:

  • The CNTS supports the RXB crane by providing lifting attachment points that the RXB crane can connect to so that the NPM can be lifted.

The NPM performs the following nonsafety-related functions that are verified by Inspections, Tests, Analyses, and Acceptance Criteria:

  • The CNTS supports the SG by providing structural support for the SG piping.
  • The CNTS supports the CRDS by providing structural support for the CRDS piping.
  • The CNTS supports the RCS by providing structural support for the RCS piping.
  • The CNTS supports the feedwater system by providing structural support for the feedwater system piping.

Design Commitments

  • The NPM American Society of Mechanical Engineers (ASME) Code Class 1, 2 and 3 piping systems listed in Table 2.1-1 comply with ASME Code Section III requirements.
  • The Nuscale Power Module ASME Code Class 1 and 2 components conform to the rules of construction of ASME Code Section III.
  • The Nuscale Power Module ASME Code Class CS components conform to the rules of construction of ASME Code Section III.
  • Safety-related structures, systems, and components (SSC) are protected against the dynamic and environmental effects associated with postulated failures in high- and moderate-energy piping systems.
  • The Nuscale Power Module ASME Code Class 2 piping systems and interconnected equipment nozzles are evaluated for leak-before-break (LBB).

RAI 14.03.03-8

  • The RPV beltline material has a Charpy upper-shelf energy of greater than 75 ft-lb minimum.
  • The CNV serves as an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment.
  • The CIV closure times limit potential releases of radioactivity.
  • The length of piping shall be minimized between the containment penetration and the associated outboard CIVs.

Tier 1 2.1-3 Draft Revision 2

NuScale Tier 1 NuScale Power Module RAI 08.01-1, RAI 08.01-1S1, RAI 08.01-2, RAI 14.03.03-8 Table 2.1-4: NuScale Power Module Inspections, Tests, Analyses, and Acceptance Criteria No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

1. The NuScale Power Module ASME An inspection will be performed of the The ASME Code Section III Design Code Class 1, 2 and 3 piping systems NuScale Power Module ASME Code Reports (NCA-3550) exist and listed in Table 2.1-1 comply with ASME Class 1, 2 and 3 as-built piping system conclude that the for the NuScale Code Section III requirements. Design Reports required by ASME Power Module ASME Code Class 1, 2 Code Section III. and 3 as-built piping systems listed in Table 2.1-1 meet the requirements of ASME Code Section III, NCA-3550.
2. The NuScale Power Module ASME An inspection will be performed of the ASME Code Section III Data Reports for Code Class 1 and 2 components NuScale Power Module ASME Code the NuScale Power Module ASME conform to the rules of construction of Class 1 and 2 as-built component Data Code Class 1 and 2 components listed ASME Code Section III. Reports required by ASME Code in Table 2.1-2 and interconnecting Section III. piping exist and conclude that the requirements of ASME Code Section III are met.
3. The NuScale Power Module ASME An inspection will be performed of the ASME Code Section III Data Reports for Code Class CS components conform to NuScale Power Module ASME Code the NuScale Power Module ASME the rules of construction of ASME Class CS as-built component Data Code Class CS components listed in Code Section III. Reports required by ASME Code Table 2.1-2 exist and conclude that the Section III. requirements of ASME Code Section III are met.
4. Safety-related SSC are protected An inspection will be performed of the Protective features are installed in against the dynamic and as-built high- and moderate-energy accordance with the as-built Pipe environmental effects associated with piping systems and protective features Break Hazard Analysis Report and postulated failures in high- and for the safety-related SSC. safety-related SSC are protected moderate-energy piping systems. against or qualified to withstand the dynamic and environmental effects associated with postulated failures in high- and moderate-energy piping systems.
5. The NuScale Power Module ASME An analysis will be performed of the The as-built LBB analysis for the ASME Code Class 2 piping systems and ASME Code Class 2 as-built piping Code Class 2 piping systems listed in interconnected equipment nozzles are systems and interconnected Table 2.1-1 and interconnected evaluated for LBB. equipment nozzles. equipment nozzles is bounded by the as-designed LBB analysis.
6. The RPV beltline material has a Charpy A vendor test will be performed of the An ASME Code Certified Material Test upper-shelf energy of greater than 75 Charpy V-Notch specimen of the RPV Report exists and concludes that the ft-lb minimum. beltline material. initial RPV beltline material Charpy upper-shelf energy is greater than 75 ft-lb minimum.
7. The CNV serves as an essentially leak- A leakage test will be performed of the The leakage rate for local leak rate tight barrier against the uncontrolled pressure containing or leakage- tests (Type B and Type C) for pressure release of radioactivity to the limiting boundaries, and CIVs. containing or leakage-limiting environment. boundaries and CIVs meets the requirements of 10 CFR Part 50, Appendix J.
8. Containment isolation valve closure A test will be performed of the Each CIV listed in Table 2.1-3 travels times limit potential releases of automatic CIVs. from the full open to full closed radioactivity. position in less than or equal to the time listed in Table 2.1-3 after receipt of a containment isolation signal.

Tier 1 2.1-12 Draft Revision 2

NuScale Tier 1 NuScale Power Module Table 2.1-4: NuScale Power Module Inspections, Tests, Analyses, and Acceptance Criteria (Continued)

No. Design Commitment Inspections, Tests, Analyses Acceptance Criteria

15. The DHRS safety-related valves change A test will be performed of the DHRS Each DHRS safety-related valve listed position under design differential safety-related valves. in Table 2.1-2 strokes fully open and pressure. fully closed by remote operation.
16. The RCS safety-related check valves A test will be performed of the RCS Each RCS safety-related check valve change position under design safety-related check valves. listed in Table 2.1-2 strokes fully open differential pressure and flow. and closed under forward and reverse flow conditions, respectively.
17. The RCS safety-related excess flow A test will be performed of the RCS Each RCS safety-related excess flow check valves change position under safety-related excess flow check check valve listed in Table 2.1-2 excess flow conditions. valves. strokes fully closed under excess flow conditions.
18. The CNTS safety-related hydraulic- A test will be performed of the CNTS Each CNTS safety-related hydraulic-operated valves fail to their safety- safety-related hydraulic-operated operated valve listed in Table 2.1-2 related position on loss of electrical valves. fails to its safety-related position on power under design differential loss of motive power.

pressure.

19. The ECCS safety-related RRVs and RVVs A test will be performed of the ECCS Each ECCS safety-related RRV and RVV fail to their safety-related position on safety-related RRVs and RVVs. listed in Table 2.1-2 fails open on loss loss of electrical power to their of electrical power to its corresponding trip valves under corresponding trip valve.

design differential pressure.

20. The DHRS safety-related hydraulic- A test will be performed of the DHRS Each DHRS safety-related hydraulic-operated valves fail to their safety- safety-related hydraulic-operated operated valve listed in Table 2.1-2 related position on loss of electrical valves. fails open on loss of motive power.

power under design differential pressure.

21. The CNTS safety-related check valves A test will be performed of the CNTS Each CNTS safety-related check valve change position under design safety-related check valves. listed in Table 2.1-2 strokes fully open differential pressure and flow. and closed under forward and reverse flow conditions, respectively.
22. i. TheA CNTS containment electrical i. An analysis will be performed of the i. A circuit interrupting device penetration assemblyies isare rated to CNTS as-built containment electrical coordination analysis exists and withstand fault currents for the time penetration assemblyies. concludes that the current carrying required to clear the fault from its capability for eachthe CNTS power source. containment electrical penetration OR assemblyies listed in Table 2.1-3 is ii. A CNTS containment electrical greater than the analyzed fault penetration assembly is rated to currents for the time required to clear withstand the maximum fault current the fault from its power source.

for its circuits without a circuit OR interrupting device. ii. An analysis of the CNTS containment penetration maximum fault current exists and concludes the fault current is less than the current carrying capability of the CNTS containment electrical penetration Tier 1 2.1-14 Draft Revision 2

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.05 NPM Section 3.6.3, Leak-Before-Break Evaluation Procedures, describes X the application of the mechanistic pipe break criteria, commonly referred to as leak-before-break (LBB), to the evaluation of pipe ruptures. The LBB analysis eliminates the need to consider the dynamic effects of postulated pipe breaks for high-energy piping that qualify for LBB.

An analysis, which includes material properties of piping and welds, stress analyses, leakage detection capability, and degradation mechanisms, confirms that the as-designed LBB analysis is bounding for the ASME Code Class 2 as-built piping listed in Tier 1 Table 2.1-1 and interconnected equipment nozzles.

A summary of the results of the plant specific LBB analysis, including material properties of piping and welds, stress analyses, leakage detection capability, and degradation mechanisms is provided in the as-built LBB analysis report.

14.3-17 02.01.06 NPM Section 5.3.1.5, Fracture Toughness, discusses the fracture X Certified Design Material and Inspections, Tests, Analyses, and toughness properties of the reactor pressure vessel (RPV) beltline material and the Material Surveillance Program. A Charpy V-Notch test of the RPV beltline material specimen is performed by the vendor to ensure that the initial RPV beltline Charpy upper-shelf energy is no less than 75 ft-lb minimum.

02.01.07 NPM Section 6.2.6, Containment Leakage Testing, provides a discussion X of the leakage testing requirements of the containment vessel (CNV), which serves as an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment. As discussed in Section 6.2.6, the NuScale CNV is exempted from the integrated leak rate testing specified in the General Design Criterion (GDC) 52.

Acceptance Criteria In accordance with Table 14.2-43, a preoperational test demonstrates that the leakage rate for local leak rate tests (Type B Draft Revision 2 and Type C) for pressure containing or leakage-limiting boundaries and containment isolation valves (CIVs) meet the leakage acceptance criterion of 10 CFR Part 50, Appendix J.

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9364 Date of RAI Issue: 02/27/2018 NRC Question No.: 14.03.03-9 10 CFR 52.47(b)(1) requires The proposed inspections, tests, analyses, and acceptance criteria that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, a facility that incorporates the design certification has been constructed and will be operated in conformity with the design certification, the provisions of the [Atomic Energy] Act, and the Commission's rules and regulations. In supporting this requirement, the Tier 2 material provides important clarifications to the Tier 1 material and should therefore be as clear as possible with respect to referenced information. The staff notes that references to Tables in the narrative discussion for DCD Tier 2, Table 14.3-1 do not specify Tier 1 or Tier 2. This may provide confusion to a user of this document. For instance, In accordance with Table 14.2-63, a preoperational test demonstrates that the ECCS safety-related valves listed in Table 2.1-2 stroke fully open, refers to both a Table in Tier 2 and Tier 1 without differentiation. Please provide clarification to the language in the DCD.

NuScale Response:

Note 1 is added to Tier 2, Table 14.3-1 to clarify that any references in Table 14.3-1 to sections, figures, and tables refer to Tier 2 unless the reference specifically states Tier 1 sections, figures, or tables.

Note 1 is also added to Tier 2, Table 14.3-2 to clarify that any references in Table 14.3-2 to sections, figures, and tables refer to Tier 2 unless the reference specifically states Tier 1 sections, figures, or tables.

Impact on DCA:

Tier 2, Tables 14.3-1 and 14.3-2 have been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

Tier 2 NuScale Final Safety Analysis Report RAI 08.01-1S1, RAI 08.01-2, RAI 14.03.03-6, RAI 14.03.03-7, RAI 14.03.03-8, RAI 14.03.03-9 Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.01 NPM As required by ASME Code Section III NCA-1210, each ASME Code X Class 1, 2 and 3 component (including piping systems) of a nuclear power plant requires a Design Report in accordance with NCA-3550. NCA-3551.1 requires that the drawings used for construction be in agreement with the Design Report before it is certified and be identified and described in the Design Report. It is the responsibility of the N Certificate Holder to furnish a Design Report for each component and support, except as provided in NCA-3551.2 and NCA-3551.3. NCA-3551.1 also requires that the Design Report be certified by a registered professional engineer when it is for Class 1 components and supports, Class CS core support structures, Class MC vessels and supports, Class 2 vessels designed to NC-3200 (NC-3131.1), or Class 2 or Class 3 components designed 14.3-14 to Service Loadings greater than Design Loadings. A Class 2 Design Report shall be prepared for Class 1 piping NPS 1 or smaller that is Certified Design Material and Inspections, Tests, Analyses, and designed in accordance with the rules of Subsection NC. NCA-3554 requires that any modification of any document used for construction, from the corresponding document used for design analysis, shall be reconciled with the Design Report.

An ITAAC inspection is performed of the NuScale Power Module ASME Code Class 1, 2 and 3 as-built piping system Design Report to verify that the requirements of ASME Code Section III are met.

Acceptance Criteria Draft Revision 2

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.09 EQ Section 3.11, Environmental Qualification of Mechanical and X Electrical Equipment, presents information to demonstrate that the CNTS electrical penetration assemblies, including its connection assemblies, located in a harsh environment are qualified by type test or a combination of type test and analysis to perform its safety-related function under design basis harsh environmental conditions, experienced during normal operations, anticipated operational occurrences, DBAs, and post-accident conditions in accordance with 10 CFR 50.49. As defined in IEEE-Std-572-2006, IEEE Standard for Qualification of Class 1E Connection Assemblies for Nuclear Power Generating Stations, a connection assembly is any connector or termination combined with related cables or wires as an assembly. The qualification method employed for the equipment is the same as the qualification method described for that type of equipment in Section 3.11.

14.3-57 The ITAAC verifies that: (1) an equipment qualification record form Certified Design Material and Inspections, Tests, Analyses, and exists for the CNTS electrical penetration assemblies listed in Tier 1 Table 2.8-1and addresses connection assemblies; (2) the equipment qualification record form concludes that the CNTS electrical penetration assemblies, including its connection assemblies, performs its safety-related function under the environmental conditions specified in Section 3.11 and the equipment qualification record form; and (3) the required post-accident operability time for the CNTS electrical penetration assemblies in the equipment qualification record form is in agreement with Section 3.11.

After installation in the plant, an ITAAC inspection is performed to verify that the CNTS electrical penetration assemblies listed in Tier Acceptance Criteria 1 Table 2.8-1, including its connection assemblies, is installed in its design location in a configuration bounded by the equipment qualification record form.

Draft Revision 2 Note:

1. References to Sections, Figures and Tables in Table 14.3-1 refer to Tier 2 unless the reference specifically states Tier 1 Sections, Figures or Tables

Tier 2 NuScale Final Safety Analysis Report RAI 09.01.04-1, RAI 09.05.01-6, RAI 14.03.02-1, RAI 14.03.02-2, RAI 14.03.03-1, RAI 14.03.03-6, RAI 14.03.03-7, RAI 14.03.03-8, RAI 14.03.09-1, RAI 14.03.09-2, RAI 14.03.09-3, RAI 14.03.12-2, RAI 14.03.12-3 Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.01.01 CRH Testing is performed on the CRE in accordance with RG 1.197, X Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, Revision 0, to demonstrate that air exfiltration from the CRE is controlled. RG 1.197 allows two options for CRE testing; either integrated testing (tracer gas testing) or component testing. Section 6.4 Control Room Habitability, describes the testing requirements for the CRE habitability program. Section 6.4 provides the maximum air exfiltration allowed from the CRE.

In accordance with Table 14.2-18, a preoperational test using the tracer gas test method demonstrates that the air exfiltration from the CRE does not exceed the assumed unfiltered leakage rate 14.3-58 provided in Table 6.4-1: Control Room Habitability System Design Certified Design Material and Inspections, Tests, Analyses, and Parameters for the dose analysis. Tracer gas testing in accordance with ASTM E741 will be performed to measure the unfiltered in-leakage into the CRE with the control room habitability system (CRHS) operating.

03.01.02 CRH The CRHS valves are tested by remote operation to demonstrate X the capability to perform their function to transfer open and transfer closed under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-18, a preoperational test demonstrates that each CRHS valve listed in Tier 1 Table 3.1-1 strokes fully open and fully closed by remote operation under preoperational test conditions.

Acceptance Criteria Preoperational test conditions are established that approximate design-basis temperature, differential pressure, and flow Draft Revision 2 conditions to the extent practicable, consistent with preoperational test limitations.

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference(1) (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.18.01 RM Section 11.5.2.2.9, Containment Flooding and Drain System, X discusses the operation of the containment flooding and drain system (CFDS). For each high radiation signal listed in Tier 1 Table 3.18-1, the CFDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table.

In accordance with the information presented in Table 14.2-42, a preoperational test demonstrates the CFDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table upon initiation of a real or simulated CFDS high radiation signal from 6B-CFD-RT-1007.

03.18.02 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the BPDS. For each high radiation signal listed in Tier 1 Table 3.18-1, the BPDS automatically aligns the components 14.3-102 identified in Tier 1 Table 3.18-1 to the required positions identified in the table.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with the information presented in Table 14.2-24, a preoperational test demonstrates the BPDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table upon initiation of a real or simulated BPDS high radiation signal from 6B-BPD-RIT-0552.

03.18.03 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the BPDS. For each high radiation signal listed in Tier 1 Table 3.18-1, the BPDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table.

In accordance with the information presented in Table 14.2-24, a Acceptance Criteria preoperational test demonstrates the BPDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table upon initiation of a real or Draft Revision 2 simulated BPDS high radiation signal from 6B-BPD-RIT-0529.

Note:

1. References to Sections, Figures and Tables in Table 14.3-2 refer to Tier 2 unless the reference specifically states Tier 1 Sections, Figures or Tables.

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9364 Date of RAI Issue: 02/27/2018 NRC Question No.: 14.03.03-10 10 CFR 52.6 requires, in part, that information provided to the Commission by an applicant for a standard design certification be complete and accurate in all material respects. Guidance in SRP 14.3.3 suggests that the reviewer ensure that all Tier 1 information is consistent with Tier 2 information and that ASME code classification, safety classification, and seismic classification of the piping systems should be indicated clearly on the figures or described in the design descriptions and consistent with DCD Tier 2, Section 3.2. The reviewer should also ensure that system boundaries and interfaces are indicated clearly in Tier 1 and that the figures are in accordance with the legends.

DCD Tier 2, Figure 3.6-1, Piping Systems Associated with the NuScale Power Module, appears inconsistent with both DCD Tier 1, Figure 2.1-1 and DCD Tier 1, Table 2.1-1, NuScale Power Module Piping Systems. Specifically, the classification of the piping systems between the containment isolation valves and the NPM flange connection are depicted as ASME B31.1 in Tier 2, but appear to be ASME Code Section III Class 3 in Tier 1. Additionally, the DHRS penetrations are depicted as penetrations in the CNV head in Tier 2, but are depicted as penetrations in the CNV shell in Tier 1. Correct these inconsistencies.

NuScale Response:

Tier 2, Figure 3.6-1 was deleted as shown in Revision 1 to the DCA. Tier 2, Figure 6.6-1 shows the lines that interface with the CNV.

The classification of piping systems between the containment isolation valves and the NPM flange connection depicted in Tier 2, Figure 6.6-1 agree with the classification of the piping contained in Tier 1, Table 2.1-1. A previously self-identified change to Tier 1, Table 2.1-1 is attached showing appropriate classifications of the piping systems.

Tier 1, Figure 2.1-1 shows the containment system boundaries, and is not intended to depict a physical location of CNV penetrations. No change to Tier 1, Figure 2.1-1 is needed.

NuScale Nonproprietary

Impact on DCA:

A change to Tier 1, Table 2.1-1 was approved and has been revised as described in the response above and as shown in the markup provided in this response.

NuScale Nonproprietary

NuScale Tier 1 NuScale Power Module RAI 10.03-1 Table 2.1-1: NuScale Power Module Piping Systems Piping System Description ASME Code High/ Evaluated for Length of Section III Moderate LBB Containment Class Energy Piping (ft)

Outside CNV CNTS reactor coolant system injection line from valves 3 High No 0 CVC-ISV-0331 & CVC-ISV-0329 at CNV nozzle CNV6 to NPM (see Note 1) disconnect flange CNTS reactor coolant system pressurizer spray line from valves 3 High No 0 CVC-ISV-0325 & CVC-ISV-0323 at CNV nozzle CNV7 to NPM (see Note 1) disconnect flange CNTS reactor coolant system discharge line from valves 3 High No 0 CVC-ISV-0334 & CVC-ISV-0336 at CNV nozzle CNV13 to NPM (see Note 1) disconnect flange CNTS reactor coolant system RPV high point degasification line 3 High No 0 from valves CVC-ISV-0401 & CVC-ISV-0403 at CNV nozzle (see Note 1)

CNV14 to NPM disconnect flange CNTS containment evacuation line from valves CE-ISV-0101 & N/A3 No No 0 CE-ISV-0102 at CNV nozzle CNV10 to NPM disconnect flange (see Note 1)

CNTS flood and drain line from valves CFD-ISV-0130 & N/A3 No No 0 CFD-ISV-0129 at CNV nozzle CNV11 to NPM disconnect flange (see Note 1)

CNTS control rod drive mechanism cooling water supply line N/A3 No No 0 from valves RCCW-ISV-0185 & RCCW-ISV-0184 at CNV nozzle (see Note 1)

CNV12 to NPM disconnect flange CNTS control rod drive mechanism cooling water return line N/A3 No No 0 from valves RCCW-ISV-0190 & RCCW-ISV-0191 at CNV nozzle (see Note 1)

CNV05 to NPM disconnect flange CNTS steam generator #1 feedwater line from valves N/A2 High No 0 FW-ISV-1003 & FW-CKV-1002 at CNV nozzle CNV1 to NPM (see Note 1) disconnect flange CNTS steam generator #2 feedwater line from valves N/A2 High No 0 FW-ISV-2003 & FW-CKV-2002 at CNV nozzle CNV2 to NPM (see Note 1) disconnect flange CNTS steam generator #1 steam line from CNV nozzle CNV3 to 2 High No 4 NPM disconnect flange including to and including valves MS-ISV-1005 & MS-ISV-1006 CNTS steam generator #2 steam line from CNV nozzle CNV4 to 2 High No 4 NPM disconnect flange including to and including valves MS-ISV-2005 & MS-ISV-2006 DHRS #1 lines from steam generator #1 steam line to DHRS 2 High No N/A Passive Condenser A including valves DHR-HOV-1002A and DHR-HOV-1002B DHRS #1 condensate line from DHRS Passive Condenser A to 2 High No N/A CNV nozzle CNV22 DHRS #2 lines from steam generator #2 steam line to DHRS 2 High No N/A Passive Condenser B including valves DHR-HOV-2002A and DHR-HOV-2002B DHRS #2 condensate line from DHRS Passive Condenser B to 2 High No N/A CNV nozzle CNV23 Tier 1 2.1-5 Draft Revision 2

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