ML19183A488

LLC Submittal of Changes to Final Safety Analysis Report, Section 4.3, Nuclear Design

ML19183A488
Person / Time
Site:	NuScale
Issue date:	07/02/2019
From:	Rad Z NuScale
To:	Document Control Desk, Office of New Reactors
References
LO-0719-66116
Download: ML19183A488 (6)

Text

LO-0719-66116 July 2, 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 Submittal of Changes to Final Safety Analysis Report, Section 4.3, Nuclear Design

REFERENCES:

1. Letter from NuScale Power, LLC to Nuclear Regulatory Commission, NuScale Power, LLC Submittal of the NuScale Standard Plant Design Certification Application, Revision 2, dated October 30, 2018 (ML18311A006)

2. Letter from NuScale Power, LLC to Nuclear Regulatory Commission, NuScale Power, LLC Submittal of Fluence Claculation Methodology and Results, TR-0116-20781, Revision 1, dated July 2019 (ML19183A485)

NuScale Power, LLC (NuScale) is updating Final Safety Analysis Report (FSAR), Section 4.3, Nuclear Design. This is a corresponding change associated with Reference 2. The Enclosure to this letter provides a mark-up of the FSAR pages incorporating revisions in redline/strikeout format.

NuScale will include this change as part of a future revision to the NuScale Design Certification Application.

This letter makes no regulatory commitments or revisions to any existing regulatory commitments.

If you have any questions, please feel free to contact Matthew Presson at 541-452-7531 or at mpresson@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Bruce Bavol, NRC, OWFN-8H12

Enclosure:

Changes to NuScale Final Safety Analysis Report Section 4.3, Nuclear Design NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0719-66116

Enclosure:

Changes to NuScale Final Safety Analysis Report Section 4.3, Nuclear Design NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Final Safety Analysis Report Nuclear Design The Monte Carlo N-Particle transport code (MCNP6) version 1.0 is used to perform the vessel fluence calculations. The specific cross-section set used is based on ENDF/B-VII.

The calculations are performed for 60 years of operation at a 95 percent capacity factor.

The values are based on the reference equilibrium cycle. A cross-section view of the MCNP6 model used to calculate vessel flux is provided in Figure 4.3-25.

Neutron flux distribution values (in units of n/cm2-sec) are summarized in Table 4.3-12.

in, above, below the core, at the inside diameter of the pressure vessel, and at the core outer radius at mid-height. These calculations are based on exposure averaged axial and radial power profiles with soluble boron concentration assumed to be zero ppm.

The core average flux values in Table 4.3-12 areis calculated using SIMULATE5 instead of MCNP6. SIMULATE5 is used because it is provides a more accurate calculation of neutron flux in the core. This SIMULATE5 calculation is based on exposure averaged axial and radial power profiles with soluble boron concentrations that correspond to the nominal middle of cycle concentration. The other Table 4.3-12 values are calculated with MCNP6 using assumptions consistent with Reference 4.3-6. The flux and radiation damage estimates are verified through the analysis of actual surveillance test samples from the irradiation surveillance program as described in Section 5.3.1. The methodology used by NuScale to calculate neutron fluence on the NPM pressure vessel and containment vessel is provided in Reference 4.3-6.

4.3.3 Analytical Methods The NuScale nuclear analysis is performed with the Studsvik Scandpower Core Management Software simulation tools. These simulation tools include the lattice physics code CASMO5, the linkage code CMSLINK5 for nuclear data library generation, and the core simulator code SIMULATE5 for power distribution and stability calculations. These codes and the modelling methodology are described in detail in Reference 4.3-1. The SIMULATE-3K code is used for transient core physics calculations and is described in detail in Reference 4.3-5. In addition, the MCNP6 code is used to perform fluence calculations.

These codes are used to perform both steady-state and transient neutronic analyses of light water reactors for core design and input to safety analysis.

RAI 29739 As described in Reference 4.3-1, the methodology for the design and analysis of a single core is independent of the presence of other NPMs. A conservative neutron flux attenuation analysis, which considers the barriers between modules (several feet of both borated water and concrete wall) confirms that the neutron flux contributed by the next closest NPM operating at full power has an insignificant neutronic impact on the reactor core of a neighboring NPM.

CASMO5 CASMO5 is a multi-group transport theory physics code which uses two-dimensional methods of characteristics transport theory for fuel assembly analysis and isotopic depletion. Cross sections and group constants are generated based on a wide range of potential conditions and cover several hundred energy groups and isotopes. CASMO5 uses the ENDF/B-VII cross section library. CASMO5 calculates eigenvalue results, power and flux Tier 2 4.3-24 Draft Revision 3

NuScale Final Safety Analysis Report Nuclear Design 4.3-3 NuScale Power LLC, "Applicability of AREVA Fuel Methodology for the NuScale Design," TR-0116-20825-P-A, Rev. 1, June 2016.

4.3-4 Krimer, M., G. Grandi, and M. Carlsson, "PWR Transient XENON Modeling and Analysis Using Studsvik CMS," Proceedings of 2010 LWR Fuel Performance/Top Fuel/WRFPM, Orlando, Florida, September 26-29, 2010.

4.3-5 NuScale Power, LLC, "Rod Ejection Accident Methodology," TR-0716-50350, Rev. 0, December 2016.

4.3-6 NuScale Power, LLC, Fluence Calculation Methodology and Results, TR-0116-20781, Rev. 10, December 2016July 2019.

Tier 2 4.3-26 Draft Revision 3

NuScale Final Safety Analysis Report Nuclear Design Table 4.3-12: Typical Fast Neutron Flux Levels (n/cm2-sec) in the Reactor Core and Reactor Pressure Vessel at Full Power Location E 1.0 MeV 1.00 MeV > E E < 0.625 eV 0.625 eV Core average 3.410E+13 9.56E+13 1.93E+13 Core outer radiusReflector block at mid-height 2.812.16E+13 8.67E+13 8.29E+12 Core top, on axisUpper core plate 4.30E+101.34E+11 0.999E+11 3.65E+11 Core bottom, on axisLower core plate 1.48E+111.12E+12 4.34E+11 1.49E+12 Pressure vessel IDinside diameter azimuthal peak 1.02E+109.57E+09 5.40E+10 2.39E+11 Tier 2 4.3-38 Draft Revision 3

NuScale Final Safety Analysis Report Nuclear Design Figure 4.3-25: Cross-section view of MCNP6 Model for Vessel Irradiation Flux Calculation Tier 2 4.3-63 Draft Revision 3