ML18156A578

LLC Response to NRC Request for Additional Information No. 411 (Erai No. 9467) on the NuScale Design Certification Application

ML18156A578
Person / Time
Site:	NuScale
Issue date:	06/05/2018
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
References
RAIO-0618-6308
Download: ML18156A578 (8)

Text

RAIO-0618-6308 June 05, 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.

411 (eRAI No. 9467) on the NuScale Design Certification Application

REFERENCE:

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

411 (eRAI No. 9467)," dated April 09, 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. 9467:

06.02.01.01.A-13 06.02.01.01.A-14 06.02.01.01.A-15 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 Di t Regulatory Director, R l t Aff i Affairs NuScale Power, LLC Distribution: Gregory Cranston, NRC, OWFN-8G9A Omid Tabatabai, NRC, OWFN-8G9A Samuel Lee, NRC, OWFN-8G9A : NuScale Response to NRC Request for Additional Information eRAI No. 9467 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvalis, Oregon 97330, Office: 541.360.0500, Fax: 541.207.3928 www.nuscalepower.com

RAIO-0618-6308 :

NuScale Response to NRC Request for Additional Information eRAI No. 9467 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.: 9467 Date of RAI Issue: 04/09/2018 NRC Question No.: 06.02.01.01.A-13 ITAAC for Containment Safety Analyses The regulatory requirement 10 CFR 52.47(b)(1) listed in DSRS Section 6.2.1.1.A states that a DC application contain the proposed inspections, tests, analyses, and acceptance criteria (ITAAC) that are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and analyses are performed and the acceptance criteria met, a plant that incorporates the DC is built and will operate in accordance with the DC, the provisions of the Atomic Energy Act, and the U.S. Nuclear Regulatory Commission's (NRC's) regulations. DSRS Section 6.2.1.1.A, paragraph III.4 requires the staff to review assumptions used in the containment response analysis to determine if the analyses are acceptably conservative.

Section C.II.1.2.11 of Regulatory Guide 1.206 (ITAAC for Containment Systems (SRP Section 14.3.11)) guides the applicant to develop ITAAC to verify key input parameters used in the containment safety analyses for the design, such as about containment heat removal during LOCA, main steam line break, and main feedline break. Following 10 CFR 52.47, "Contents of applications; technical information," Part 2 of the NuScale FSAR presents the Tier 1 information developed for the NuScale Power Plant. Staff's SRP Section 14.3.11 review showed that the NuScale containment ITAAC provided in FSAR Tier 1 do not address some of the key physical parameters relied on in the containment safety analyses. Substantial changes to these parameters in the as-built NuScale containment structure could adversely impact the FSAR Section 6.2.1 safety analyses. The applicant needs to include containment ITAAC in the FSAR to validate the as-built safety-related functions of the NuScale containment structure and the reactor pool. In this regard, staff requests the applicant to address the following questions and update the FSAR, accordingly. The regulatory bases identified above are applicable to all questions in this RAI.

Containment net free volume is one of the principal containment design parameters. NuScale FSAR Tier 1 provided no ITAAC to confirm that the as-built containment net free volume is conservative with respect to the value assumed in the containment peak pressure and temperature analyses in FSAR Tier 2, Section 6.2.1. As the containment free volume is a key input parameter in the containment pressure analyses (both for calculating peak pressure in Section 6.2.1.1.A and the minimum pressure in Section 6.2.1.5), the applicant is requested to provide an ITAAC to verify its as-built value to conservatively bound the value assumed in the NuScale Nonproprietary

design-basis containment analyses.

NuScale Response:

Tier 2 FSAR Section 14.3.2 identifies that the top-level design features for the NuScale Power Plant are contained in Tier 1 design descriptions. Section 14.3.2 describes the first principles approach as the basis for design descriptions and ITAAC in DCAs. First principles are applied to determine top level design descriptions in Tier 1 and whether an ITAAC is appropriate. Tier 2 FSAR Section 14.3.2.1 describes the aforementioned first principles as limiting the design description (and ITAAC) to top-level features, which are described in detail in all the subsections of 14.3.2.1. As specified in Tier 2 FSAR Section 14.3.2.1.1, the top-level design feature contained in Tier 1 design descriptions with respect to containment is "containment pressure boundary." There is no top-level design feature of "containment net free volume". Since "containment net free volume" is not a top level design feature, there is no ITAAC for this parameter. The calculation of peak containment pressure includes many input parameters beyond that of "containment net free volume". These inputs are delineated in the Containment Response Analysis Methodology Technical Report (TR-0516-49084-P). It is the aggregate of all these inputs that constitute the calculation of peak containment pressure.

Further, there is no 10 CFR Part 52 DCA precedent for a containment net free volume ITAAC.

This ITAAC was not required for the AP 1000 design, which has a very small margin of 1.1 percent to containment design pressure in its DCA as compared to the NuScale DCA containment peak pressure margin of 4.9 percent to containment design pressure.

Compliance with 10 CFR Part 50, Appendix A, GDC 16 and GDC 50 constitutes the regulatory basis which ensures that the containment peak pressure does not exceed containment design pressure.

Impact on DCA:

There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9467 Date of RAI Issue: 04/09/2018 NRC Question No.: 06.02.01.01.A-14 Following a high energy line break within containment, containment internal heat structures (heat sinks) are important for condensing steam from the containment atmosphere and storing energy. As condensation influences the containment pressure and temperature following an accident, the heat sinks credited in the containment pressure analyses make a key input assumption (both for calculating peak pressure in Section 6.2.1.1.A and the minimum pressure in Section 6.2.1.5). However, NuScale FSAR Tier 1 provided no ITAAC to verify that the passive heat sinks credited to the design-basis analyses conform to the information provided in Section 8.2.1 through 8.2.4 of the Containment Response Analysis Methodology technical report. The applicant is requested to provide an ITAAC to verify the as-built heat sink parameters and compositions to conservatively bound the heat sink input design assumptions made in the design-basis containment analyses, e.g., the containment heat sinks dimensions and materials.

NuScale Response:

TR-0516-49084 Rev. 0, "Containment Response Analysis Methodology Technical Report",

Section 8.2.1 states that "The containment vessel shell is the only passive heat sink credited in the containment response analysis methodology." The modeling of the passive heat sink (i.e.,

the containment vessel (CNV) shell) is identified in technical report Table 8-3, and includes the CNV materials, thicknesses, and "group". Technical report Table 8-4 identifies the thickness ranges for each thickness "group" and is linked to Table 8-3.

Since containment response only models passive heat transfer through the CNV shell to the reactor pool, ITAAC 02.01.02, FSAR Tier 1, Table 2.1-4, item 2, verifies that the passive heat sink that is credited in the design basis containment response analyses conforms to Tables 8-3 and 8-4 of the technical report. Data Reports for the ASME Code Class 1 CNV shell will be inspected to ensure that the as-built CNV shell characteristics (i.e., its materials, thicknesses, and dimensions) are within the parameters identified in Tables 8-3 and 8-4 of the technical report.

NuScale Nonproprietary

Impact on DCA:

There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary

Response to Request for Additional Information Docket No.52-048 eRAI No.: 9467 Date of RAI Issue: 04/09/2018 NRC Question No.: 06.02.01.01.A-15 FSAR Tier 1 Section 3.6 states that the ultimate heat sinks (UHS) safety-related system function of heat removal via direct water contact with the containment vessel is verified by inspections, tests, analyses, and acceptance criteria, but the staff found that no such ITAAC was provided. The applicant is requested to clarify the statement and justify why it is not supported by an ITAAC, or include an appropriate ITAAC, accordingly.

NuScale Response:

Inspections, tests, analyses, and acceptance criteria (ITAAC) is already provided for the

ultimate heat sink (UHS) safety-related function of heat removal via direct water contact with the

containment vessel. Tier 1, Section 3.6 describes the UHS for the NuScale design, and states

The configuration of the UHS includes the combined volume of water in the reactor pool,

refueling pool (RFP), and spent fuel pool (SFP). The pool areas are open to each other with a

weir wall partially separating the SFP from the RFP. The dry dock area is not considered part of

the UHS volume.

Tier 1, Section 3.6.1 also states that The structural components of the reactor pool, RFP, and

SFP (i.e., structural walls, weir wall, and floor) and associated pool liners are a component of

the Reactor Building (RXB) structure.

When the NuScale design refers to the UHS, it is describing the water in the reactor pool,

RFP, and SFP that provides heat removal and radiation shielding functions.

The UHS verifies the following safety-related system functions by ITAAC 03.06.02

The UHS supports the containment system by providing the removal of heat via direct water contact with the containment vessel.

The UHS supports the decay heat removal system by accepting the heat from the decay heat removal heat exchanger.

NuScale Nonproprietary

The UHS supports the spent fuel system by providing the removal of decay heat from the spent fuel via direct water contact with the spent fuel assemblies.

The UHS verifies the following nonsafety-related system functions by ITAAC 03.06.02:

The UHS supports the containment system by providing the radiation shielding for the NPMs via the water surrounding the components.

The UHS supports the spent fuel system by providing radiation shielding for spent fuel via the water surrounding the components.

ITAAC 03.06.02 is contained in Tier 1, Table 3.6-2 as shown below.

Inspections, Tests, No. Design Commitment Acceptance Criteria Analyses The spent fuel pool, refueling There are no gates, pool, reactor pool, and dry An inspection will be openings, drains, or piping dock piping and connections performed of the as- within the SFP, RFP, reactor are located to prevent the built SFP, RFP, pool, and dry dock that are 2

drain down of the SFP and reactor pool and dry below 80 ft building elevation reactor pool water level dock piping and (55 ft pool level) as below the minimum safety connections. measured from the bottom of water level. the SFP and reactor pool.

Note the design commitment for ITAAC 03.06.02 references the spent fuel pool, refueling pool and reactor pool water, which is the UHS. Thus, ITAAC 03.06.02 verifies that the UHS capability to support heat removal and radiation shielding cannot be affected by a drain down through penetrations below the 80 ft. Reactor Building elevation.

ITAAC 03.06.02 is further discussed in Tier 2, 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.

Impact on DCA:

There are no impacts to the DCA as a result of this response.

NuScale Nonproprietary