ML20176A157

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Enclosure 1 - Chapter 15 Audit Summary Public Version
ML20176A157
Person / Time
Site: NuScale
Issue date: 07/10/2020
From: Bavol B
NRC/NRR/DNRL/NRLB
To: Michael Dudek
NRC/NRR/DNRL/NRLB
Bavol B
Shared Package
ML20176A155 List:
References
Download: ML20176A157 (27)
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AUDIT REPORT FOR THE REGULATORY AUDIT OF NUSCALE POWER, LLC DESIGN CERTIFICATION APPLICATION, CHAPTER 15, TRANSIENT AND ACCIDENT ANALYSES

1.0 BACKGROUND

By letter dated December 31, 2016, NuScale Power, LLC (hereinafter referred to as NuScale or the applicant) submitted to the U.S. Nuclear Regulatory Commission (NRC) its design certification application (DCA) for the NuScale design (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17013A229). Over the course of the technical review, the NRC staff issued several requests for additional information (RAIs) that resulted in NuScale revising existing analyses or performing additional analyses. In addition, NuScale updated its systems analysis code, NRELAP5, from Version 1.3 to Version 1.4.

One purpose of this regulatory audit was to clarify the NRC staffs understanding of the RAI responses and the associated analyses. Another objective of this audit was for the NRC staff to gain a better understanding of changes to calculations or information that supports the DCA resulting from the transition to NRELAP5 Version 1.4. The NRC staff needed to fully comprehend the new and changed calculations and information to ensure it supported the applicants docketed conclusions.

To facilitate the NRC staffs understanding of the calculations and rationale underlying RAI responses and the updates to DCA Part 2, Tier 2, Chapter 15, Transient and Accident Analyses, as a result of implementing a new version of NRELAP5, the NRC staff conducted the following:

  • An audit entrance meeting was held January 10, 2019, via conference call.
  • Audit Phase 3 focused on certain calculations or information supporting RAI responses.
  • Audit Phase 4 focused on updates to DCA Part 2, Tier 2, Chapter 15, as a result of reanalysis using NRELAP5, Version 1.4.
  • The audit was performed via the NuScale electronic reading room (ERR), at NuScales Rockville office, and at NuScales Corvallis office.
  • An audit exit meeting was conducted on December 18, 2019, via conference call.

2.0 REGULATORY AUDIT BASIS This regulatory audit was based on the following:

  • NuScale DCA Part 2, Tier 2, Chapter 15.
  • Information related to the planned revision to DCA Part 2, Tier 2, Chapter 15, to update NRELAP5 calculations and results.
  • RAI responses provided by NuScale identified in the table below:

1 Enclosure 1

DCA Part 2, Tier 2, Section RAI/Question No. ADAMS Accession No.

15.0.6 8930/15-27 ML18257A308 15.2.1-15.2.9 9407/15.02.01-9 ML18141A880 15.2.4 9407/15.02.01-11 ML18141A880 15.4.3 9512/15.04.03-2, 15.04.03-4 ML18163A421 15.4.7 9504/15.04.07-3 ML18179A522 3.0 AUDIT LOCATION AND DATES The audit was conducted from the NRC headquarters via NuScales ERR, telephone, and at NuScales Rockville, Maryland, and Corvallis, Oregon, offices.

Dates: January 10, 2019, through December 18, 2019 Locations: NRC Headquarters Two White Flint North 11545 Rockville Pike Rockville, MD 20852-2738 NuScale 11333 Woodglen Drive, Suite 205 Rockville, MD 20852 NuScale 1100 NE Circle Blvd., Suite 200 Corvallis, OR 97330 4.0 NRC AUDIT TEAM MEMBERS Jeffrey Schmidt Sr. Reactor Systems Engineer, Lead (NRR/DANU/ARTB)

Carl Thurston Nuclear Engineer (NRR/DSS/SNRB)

Shanlai Lu Sr. Nuclear Engineer (NRR/DSS/SNRB)

Antonio Barrett Reactor Systems Engineer (NRR/DANU/ARTB)

Alex Siwy Reactor Systems Engineer (NRR/DSS/SNRB)

Ryan Nolan Nuclear Engineer (NRR/DSS/SNRB)

Andrew Bielen Nuclear Engineer (RES/DSA/FSCB)

Timothy Drzewiecki Reactor Systems Engineer (NRR/DANU/ARTB)

Andrew Ireland Reactor Systems Engineer (RES/DSA/CRAB)

Matt Thomas Reactor Systems Engineer (RII)

Ray Skarda Reactor Systems Engineer (RES/DSA/CRAB)

Grant Garrett Nuclear Engineering Intern (NRR/DSS/SNRB)

Rebecca Patton Branch Chief (NRR/DSS/SNRB) 5.0 APPLICANT PARTICIPANTS Adam Brigantic John Fields Matthew Presson Morris Byram 2

Ben Bristol Paul Infanger Ken Rooks Allyson Callaway Meghan McCloskey Emil Weil Taylor Coddington Rebecca Norris Andy Lingenfelter 6.0 AUDIT DOCUMENTS The NRC staff audited the following documents provided by NuScale:

Document Number Document Title Documents Generically Applicable to Chapter 15 ER-A030-3027, Revision 0 RCS Chemistry Report EC-A021-2066, Revision 1 Moderator Density Coefficient Calculation EC-A021-6822, Revision 0 Core Bypass Hydraulic Analysis EC-T050-3638, Revision 1 Assessment of NRELAP5 Using SIET Fluid Heated Test Facility (TF-2) Data EC-T050-3234, Revision 0 NRELAP5 Model for the SIET Fluid Heated Test Facility ECN-T050-6039, Revision 0 Changes to SIET TF-2 NRELAP5 Base Model EC-T050-3234 Revision 0 ER-0000-2486, Revision 7 Safety Analysis Limits Report DCA Part 2, Tier 2, Section 15.0.5 EC-A010-4270, Revision 2 Long Term Cooling Analysis EC-T080-4505, Revision 0 NRELAP5 Assessment of Spurious RVV Opening Test NIST-1 HP-19a EC-T080-4506, Revision 0 NRELAP5 Assessment of Spurious RVV Opening Test NIST-1 HP-19b DCA Part 2, Tier 2, Section 15.0.6 EC-0000-4820, Revision 2 Overcooling Return to Power Analysis EC-0000-7127, Revision 0 Boron Dilution Analysis for IORV (RRV) Followed by ECCS Actuation EC-0000-7144, Revision 0 Boron Dilution Analysis for IORV (RVV) Followed by ECCS Actuation EC-0000-6622, Revision 0 Boron Dilution for RCCW Line Break Followed by ECCS Actuation EC-0000-2044, Revision 1 Boron Dilution Transient Analysis ER-0000-6621, Revision 1 Boron Transport and Distribution Methodology EC-0000-7110, Revision 0 Extended ECCS Transient for Boron Dilution Analysis EC-0000-4848, Revision 0 ECCS Overcooling Reactivity Coping Analysis EC-0000-7532, Revision 0 Nominal Seven Day Boron Transport Evaluation N/A REWET-II Volatility Talking Points CP-1856 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.0.6 DCA Part 2, Tier 2, Section 15.1.1 EC-0000-2017, Revisions 0 Decrease in Feedwater Temperature Analysis and 1 ECN-0000-7369, Revision 0 Decrease in Feedwater Temperature Additional Plots and 6507 Update Disposition EC-0000-2898, Revision 1 Subchannel Analysis of Decrease in Feedwater Temperature 3

Document Number Document Title ECN-0000-6005, Revision 0 NSP4 Update - Subchannel Analysis of Decrease in Feedwater Temperature CP-1730 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.1.1 DCA Part 2, Tier 2, Section 15.1.2 EC-0000-2016, Revisions 0 Increase in Feedwater Flow Analysis and 1 ECN-0000-6486, Revision 0 RAI response to Increase in Feedwater Flow Analysis (EC-0000-2016, Revision 0)

EC-0000-2016, Revision 1 SG Level Overfill, Most Limiting Cases EC-0000-3077, Revision 1 Subchannel Analysis of Increase in Feedwater Flow ECN-0000-5705, Revision 0 Subchannel Analysis of Increase in Feedwater Flow ECN-0000-5705, Revision 0, NSP4 Update - Subchannel Analysis of an Increase in Feedwater Flow CP-1731 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.1.2 CP-1977 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.1.2 DCA Part 2, Tier 2, Section 15.1.3 EC-0000-2906, Revision 2 Increase in Steam Flow/Inadvertent Opening of Steam Generator Relief or Safety Valve Analysis CP-1732 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.1.3 DCA Part 2, Tier 2, Section 15.1.5 EC-0000-2714, Revision 1 Steam System Piping Failure Analysis EC-0000-2900, Revision 2 Subchannel Analysis of Steam System Piping Failures CP-1733 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.1.5 DCA Part 2, Tier 2, Section 15.1.6 CP-1734 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.1.6 DCA Part 2, Tier 2, Section 15.2.1-15.2.3 EC-0000-1997, Revisions 1 Turbine Trip, Loss of External Load, and Loss of Condenser and 2 Vacuum Transient Analysis EC-0000-3073, Revision 1 Subchannel Analysis of Loss of External Load, Turbine Trip, Condenser Vacuum ECN-0000-6007, Revision 0 NSP4 Update - Subchannel Analysis of Loss of External Load, Turbine Trip, Condenser Vacuum CP-1630 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Sections 15.2.1, 15.2.2, 15.2.3 DCA Part 2, Tier 2, Section 15.2.4 EC-0000-2995, Revision 2 Closure of Main Steam Isolation Valve Transient Analysis ECN-0000-7390, Revision 0 Update to MSIVC transient using latest non-LOCA model, EC00006507, Revision 2 EC-0000-2907, Revision 1 Subchannel Analysis of Closure of Main Steam Isolation Valve 4

Document Number Document Title ECN-0000-6012, Revision 0 NSP4 Update - Subchannel Analysis of Closure of Main Steam Isolation Valve CP-1610 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.2.4 DCA Part 2, Tier 2, Section 15.2.6 EC-0000-2908, Revision 2 Loss of Non-Emergency AC Power to the Station Auxiliaries Analysis EC-0000-3081, Revision 1 Subchannel Analysis of Loss of Nonemergency AC Power ECN-0000-6009, Revision 0 NSP4 Update - Subchannel Analysis of Loss of Nonemergency AC Power CP-1726 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.2.6 DCA Part 2, Tier 2, Section 15.2.7 EC-0000-1998, Revision 1 Loss of Normal Feedwater Transient Analysis EC-0000-3001, Revision 2 Subchannel Analysis of Loss of Normal Feedwater Flow CP-1735 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.2.7 DCA Part 2, Tier 2, Section 15.2.8 EC-0000-2250, Revision 1 Feedwater Piping Failure Analysis CP-1736 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.2.8 DCA Part 2, Tier 2, Section 15.2.9 EC-0000-2735, Revision 2 Inadvertent Operation of the Decay Heat Removal System EC-0000-3004, Revision 2 Subchannel Analysis of Inadvertent Operation of Decay Heat Removal System CP-1737 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.2.9 DCA Part 2, Tier 2, Section 15.4.1 EC-0000-7292, Revision 0 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition Analysis EC-0000-3080, Revision 2 Subchannel Analysis of an Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power State CP-1738 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.4.1 DCA Part 2, Tier 2, Section 15.4.2 EC-0000-1999, Revision 4 Uncontrolled Control Rod Assembly Withdrawal at Power Transient Analysis EC-0000-2899, Revision 3 Subchannel Analysis of Uncontrolled Rod Assembly Withdrawal at Power CP-1739 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.4.2 DCA Part 2, Tier 2, Section 15.4.3 EC-A021-7235, Revision 0 Cycle-Specific Nuclear Analysis EC-A021-1922, Revision 3 Single Rod Withdrawal Analysis ECN-A021-4989, Revision 0 Corrected RPF Ratios 5

Document Number Document Title EC-A021-5277, Revision 0 Assembly Power and Rod Worth Relationship for the Single Rod Withdrawal Event EC-0000-2139, Revision 2 Control Rod Misoperation Transient Analysis ECN-0000-7435, Revision 0 Extra Plots for EC-0000-2139 EC-0000-2897, Revision 2 Subchannel Analysis of Control Rod Misoperation EC-0000-4309, Revisions 2 Subchannel Analysis of a Control Rod Misalignment and 3 CP-1740 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.4.3 DCA Part 2, Tier 2, Section 15.4.6 CP-1741 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.4.6 DCA Part 2, Tier 2, Section 15.4.7 EC-A021-7235, Revision 0 Cycle-Specific Nuclear Analysis EC-A021-1858, Revision 1 Inadvertent Assembly Misloading Analysis EC-0000-2646, Revisions 2 Subchannel Analysis of Inadvertent Loading and Operation and 3 of a Fuel Assembly in an Improper Position CP-1742 DCA Revision 3 ,Change Package for DCA Part 2, Tier 2, Section 15.4.7 DCA Part 2, Tier 2, Section 15.4.8 CP-1743 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.4.8 DCA Part 2, Tier 2, Section 15.5.1 EC-0000-2726, Revision 1 Increase in RCS Inventory from CVCS Malfunction CP-1744 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.5.1 DCA Part 2, Tier 2, Section 15.6.2 EC-0000-2786, Revision 3 Failure of Small Lines Carrying Primary Coolant Outside Containment CP-1746 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.6.2 DCA Part 2, Tier 2, Section 15.6.3 EC-0000-1735, Revisions 1 Steam Generator Tube Failure Transient Analysis and 2 CP-1747 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.6.3 DCA Part 2, Tier 2, Section 15.6.5*

CP-1748 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Section 15.6.5 DCA Part 2, Tier 2, Section 15.6.6*

CP-1745 DCA Revision 3, Change Package for DCA Part 2, Tier 2, Sections 15.6.1, 15.6.6

  • Note: Additional audit documents and activities related to these DCA sections are documented in a separate Audit Report (ML20054A060).

6

7.0 DESCRIPTION

OF AUDIT ACTIVITIES AND

SUMMARY

OF OBSERVATIONS The NRC staff audited information related to NuScales Chapter 15 analyses. The NRC staff examined calculations and other documents supporting RAI responses, as well as those supporting design changes made during the audit, including revisions to Chapter 15 analyses made due to the change to NRELAP5 Version 1.4. The specific observations are documented below.

7.1 Documents Generically Applicable to Chapter 15 The NRC staff reviewed several documents that provided overarching support to the transient and accident analyses, focusing on documents that supported input assumptions used in Chapter 15. Areas of review addressed during this audit included: (1) mixing evaluation for chemistry, (2) moderator density coefficients (MDCs), (3) core bypass hydraulic analysis, (4)

NRELAP5 assessment using SIET TF-2 experimental data, and (5) safety analysis limits. The NRC staffs observations associated with these areas of review are discussed in the subsections that follow.

In addition, the NRC staff reviewed the change packages listed in Section 6 of this audit summary to ensure that the DCA sections were appropriately updated based on the results of the revised calculations.

Mixing Evaluation for Chemistry Adjustments The NRC staff examined Appendix B of ER-A030-3027, RCS Chemistry Report, which describes an evaluation of the time estimate for chemical and volume control system (CVCS) chemistry adjustments to completely mix into the reactor coolant system (RCS). During this audit, the NRC staff noted the analysis method and key inputs, including:

  • The methodology for evaluating circumferential mixing in turbulent pipes is from the following reference: ((

)).

  • The methodology does not credit additional sources of turbulence associated with flow obstructions (e.g., steam generator tubes).
  • Chemical addition is assumed to be conducted over the span of the loop transport time to eliminate the need for axial diffusion.
  • For startup conditions, the average temperature ((

)).

  • The analysis assumes a turbulent Schmidt number of (( )).
  • The analysis determined that it takes approximately ((

)) for complete mixing.

Moderator Density Coefficients 7

The NRC staff examined EC-A021-2066, Moderator Density Coefficient Calculation, Revision 1. During this audit, the NRC staff noted the purpose of the calculation, the analysis method, key inputs, and results, including:

  • The purpose of this calculation is to generate MDC values against a range of densities at specific NuScale reactor core (RXC) conditions.
  • This calculation applies to a range of densities for specific RXC conditions. The generated MDCs documented in this calculation are used in downstream calculations and to support RXC design.
  • No uncertainties are applied to the results of the calculation.
  • The analyses are performed using SIMULATE5.
  • The analyses use the best-estimate thermal-hydraulic conditions provided in the table below.

Power RCS Average Primary Flow Primary Flow (kg/s)

(percent rated) Temperature (percent rated)

(( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( ))

  • The analyses use the power-dependent insertion limits (PDILs) and analyzed rod positions provided in the table below.

PDIL Analyzed Rod Position Power (steps withdrawn) (steps withdrawn)

(percent rated)

Reg. Bank 1 Reg. Bank 2 Reg. Bank 1 Reg. Bank 2

(( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( ))

  • To bound all allowed PDILs, an additional series of analyses were performed ((

)). The resulting rod positions are provided in the table below.

High PDIL Analyzed Rod Position Power (steps withdrawn) (steps withdrawn)

(percent rated)

Reg. Bank 1 Reg. Bank 2 Reg. Bank 1 Reg. Bank 2

(( )) (( )) (( )) (( )) (( ))

  • MDC calculations are performed at:

o (( )):

(( )).

8

(( )).

o (( )):

(( )).

(( )).

  • ((

))

  • ((

))

(( )) (( )) (( )) (( ))

Density MDC Density MDC Density MDC Density MDC (kg/m3) (pcm/kg/m3) (kg/m3) (pcm/kg/m3) (kg/m3) (pcm/kg/m3) (kg/m3) (pcm/kg/m3)

(( )) (( )) (( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( )) (( )) (( ))

  • ((

))

9

(( )) (( )) (( ))

Density MDC Density MDC Density MDC (kg/m3) (pcm/kg/m3) (kg/m3) (pcm/kg/m3) (kg/m3) (pcm/kg/m3)

(( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( )) (( )) (( ))

  • ((

))

(( )) (( ))

Condition CBC (parts per Condition CBC (ppm) million (ppm))

(( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( ))

(( )) (( )) (( )) (( ))

(( )) (( ))

Core Bypass Hydraulic Analysis The NRC staff examined EC-A021-6822, Core Bypass Hydraulic Analysis, Revision 0. During the audit of this document, the NRC staff noted the purpose, analysis method, key inputs, and results, including:

  • The purpose of this calculation is to estimate the bypass flow fraction of the RCS.
  • The applicant performed sensitivity studies to estimate the bounding bypass flow fraction limit.
  • Analysis inputs and assumptions include:

o ((

)).

o ((

)).

10

o ((

)).

o ((

)).

  • The calculation is performed ((

)).

o ((

)).

  • Calculations are run at ((

)).

  • ((

)).

11

  • The cases with the highest calculated bypass flows are presented in the table below:

Percent Bypass Gap Total bypass Case between (does not Guide Instrument Reflector core and include tubes tubes reflector core/reflector wall gap)

(( )) (( )) (( )) (( ))

((

(( ))

))

(( )) (( )) (( )) (( ))

((

(( ))

))

  • The NRC staff notes that the ((

)).

NRELAP5 Assessment using SIET TF-2 Data The NRC staff audited the assessment report EC-T050-3638, Revision 1, Assessment of NRELAP5 Using SIET Fluid Heated Test Facility (TF-2) Data, and the associated NRELAP5 model results and modeling inputs document, EC-T050-3234, Revision 0, NRELAP5 Model for the SIET Fluid Heated Test Facility. ((

)) The NRC staff also noted that the loss coefficient value appears comparable to NPM, but the flow characteristics are likely very different. ((

))

Safety Analysis Limits The NRC staff audited ER-0000-2486, Revision 7, Safety Analysis Limits Report, and noted that the analytical limits, initial conditions, and assumptions used in DCA Part 2, Tier 2, Chapter 15, were consistent with those listed in ER-0000-2486.

7.2 DCA Part 2, Tier 2, Section 15.0.5, Long-Term Decay and Residual Heat Removal 12

The NRC staff audited EC-A010-4270, Revision 2, Long Term Cooling Analysis, in support of its review of Technical Report TR-0916-51299, Revision 1, Long-Term Cooling Methodology.

The NRC staff noted that the NRELAP5 long-term cooling (LTC) model was developed from noding modeling studies for loss-of-coolant accident (LOCA) modeling reported in EC-0000-4888, NuScale LOCA TR Evaluation Model Supporting Calculations, which the NRC staff audited as part of the LOCA methodology review (ML20010D133 and ML20042E024), with the addition of (( )).

Additionally, the NRC staff noted that NuScale only ran LTC events to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and then used a state point method to extrapolate conditions at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> based on the expected decay heat.

However, 1 or 2 cases were run with NRELAP5 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to confirm that the state point method was sufficiently accurate.

In addition, the NRC staff examined NIST-1 facility tests HP-19a and HP-19b in EC-T080-4505, Revision 0, NRELAP5 Assessment of Spurious RVV Opening Test NIST-1 HP-19a, and EC-T080-4506, Revision 0, NRELAP5 Assessment of Spurious RVV Opening Test NIST-1 HP-19b, which were used to validate the long-term cooling evaluation model. The NRC staff reviewed these two NIST LTC benchmarks to confirm the capability of NRELAP5 to compute key phenomena for LTC transients. The NRC staff observed that benchmarks showed ((

)) of pressures, temperatures, and collapsed liquid level in the reactor vessel and containment vessel. The(( )) for the cooling pool vessel, but those are of less importance.

7.3 DCA Part 2, Tier 2, Section 15.0.6, Evaluation of a Return to Power EOC Return to Power The NRC staff audited the end of cycle return to power analysis calculations supporting DCA Part 2, Tier 2, Section 15.0.6. The applicant employed a new analysis methodology in the DCA using a combination of NRELAP5 and SIMULATE5. The NRC staff audited how the two codes were used to determine the equilibrium return to power value and associated MCHFR.

NRELAP5 is used to determine the average moderator temperature for a series of constant powers for each cooldown scenario, referred to as statepoint cases. The use of statepoint cases to predict conservative thermal-hydraulic conditions was reviewed as part to the long-term cooling analysis Technical Report, TR-0916-51299, Revision 1. The NRELAP5 cases consisted of the three long term cooling modes, uncovered riser, covered riser and ECCS. The NRC staff reviewed the NRELAP5 statepoint cases, including comparisons to confirmatory transient cases, and found the statepoint cases yielded conservative thermal-hydraulic conditions for the reactivity evaluation. The NRC staff also reviewed that conservative input assumptions such as ultimate heat sink level and the minimum TS reactor pool building temperature were used to yield the lowest expected moderator temperature. For the DHRS cooldown cases, the NRC staff also confirmed that normal and adiabatic riser heat transfer models were applied to bound the riser wall heat transfer. For the NRELAP5 statepoint cases, the NRC staff noted that the following conservative inputs were used:

  • Reactor pool level is maximized, and temperature was minimized.

13

  • Conservative low decay heat was assumed.

A SIMULATE5 power search was used to determine critical power for a variety of input conditions. The NRELAP5 core power versus RCS flow rate curve was used to determine a single SIMULATE5 criticality line versus average moderator temperature. The use of NRELAP5 core power verse flow rate curve is conservative as it assumes a flow rate consistent with the higher, single phase natural circulation (riser covered) flow which is conservative for the uncovered riser and ECCS cooling modes. The intersection of the NRELAP5 constant power versus average moderator temperature curve and the SIMULATE5 line of criticality determines the equilibrium return to power magnitude. The NRC staff noted the following conservatisms in the applicants calculation:

  • Reactivity is biased to address SIMULATE5 uncertainties (((

)) consistent with Nuclear Analysis Topical Report methodology).

  • Conservative coolant density is assumed (( )).
  • Conservative peaking factors are applied for MCHFR determination (local peaking factor of (( )) accounting for the worst rod stuck out).
  • Zero xenon concentration is assumed consistent with 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a reactor trip.

The NRC staff examined the applicants hand calculation of the minimum critical heat flux ratio (MCHFR) calculation using a pool boiling CHF correlation with an input power a factor of two higher than the SIMULATE5 determined equilibrium power. The NRC staff considers the factor of two times the SIMULATE5 equilibrium power adequate to cover any reasonable transient power overshoot. The NRC staff noted that significant MCHFR margin exists for all three passive cooling modes.

Non-EOC Boron Redistribution In its response to RAI 8930, the applicant provided several new calculations analyzing boron dilution for various LOCA scenarios for the NRC staffs audit in the ERR, as listed in Section 6 of this audit summary.

The NRC staff had requested that the applicant specify and describe, in sufficient detail, a methodology to calculate boron transport during long-term cooling following ECCS actuation after any Chapter 15 event, to include LOCA and non-LOCAs. The methodology is primarily given in ER-0000-6621, and depends on NRELAP5 to provide thermal-hydraulic conditions (mass flow rate, temperature, and pressure) provided in EC-0000-7110 for each lumped segment. The boron transport analyses were essentially performed in an Excel spreadsheet used to post-process the NRELAP5 output data. The NRC staff examined the spreadsheet in an audit conference at NuScales Rockville, Maryland office in early 2019. The spreadsheet and the documentation of the methodology in ER-0000-6621, were complex and difficult to 14

understand due to use of multiple donor cell transport factors. The NRC staff determined that boron volatility needed to be addressed and subsequently examined the applicants volatility calculations, as described below. The NRC staff noted that the overall goal of the analyses was to show that the final boron concentration at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> was not less than the initial boron concentration at the start of the transient. The limiting case was observed to be the RRV IORV documented in EC-0000-7127.

The applicant used test data from the VEERA facility to justify that the boron distribution will be axially well mixed. The applicants rationale for the VEERA mixing justification was well documented in the submitted information (ML19332A120), as well as publicly available literature. Therefore, the NRC staffs evaluation is detailed in the staffs SER.

The applicant calculated the NPM core elevation where the onset of saturated boiling occurs for the cold ECCS conditions at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Above this point, the boiling is sufficient to ensure a uniform boron concentration, as demonstrated by the test data (i.e., middle and top measurement points). The NRC staff audited the applicants calculation of this core elevation (designated as Zcrit), which was determined using a set of equations involving decay heat, core inlet and exit enthalpies, to determine the core axial position at which saturated boiling starts.

Below Zcrit, the applicant assumed (( )) was present, and above it, a conservative uniform boron concentration was assumed. The NRC staff audited the applicants NPM analysis, which showed that the core remains subcritical assuming a (( ))

concentration up to the point where uniform mixing would occur, and the core reactivity was below that calculated assuming the homogenized core/riser boron concentration. The NPM analysis utilized a MCNP model simulating a ((

)). The resulting case demonstrated the core remained subcritical. The NRC staff performed a similar analysis using PARCS which likewise showed the core remained subcritical.

To further examine the core/riser mixing, the NRC staff audited NIST-1 long-term cooling test data from tests HP-19a, HP-19b, and HP-43 to determine if sufficient voiding exists such that two-phase mixing would promote riser and core fluid exchange. The NRC staffs examination of the NIST-1 data, indicated that the amount of two-phase mixing in the riser would promote sufficient core and riser fluid exchange to prevent a significant boron gradient. The NRC staff also noted that riser/core and within-core temperature differences will exist (e.g., across the riser/downcomer and between high- and low-power assemblies) to further aid flow recirculation and lessen any potential boron concentration gradient.

The NRC staff evaluated the applicants use of the Bhlke correlation1 to determine the volatility of the boric acid in the NuScale design and audited calculations supporting the submitted information (ML19332A120). The NRC staffs review of the use and applicability of the Bhlke correlation is detailed in the staffs SER, as the underlying information is provided in the applicants submittal. The NRC staff audited the applicants spreadsheet that implemented the 1 Steffen Bhlke, Christoph Schuster, and Antonio Hurtado, About the Volatility of Boron in Aqueous Solutions of Borates with Vapour in Relevance to BWR-Reactors, International Conference on the Physics of Reactors, Nuclear Power: A Sustainable Resource, Interlaken, Switzerland, September 14-19, 2008.

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correlation and noted that the correlation was implemented in the calculations and are consistent with the submitted information.

As part of the audit, the NRC staff reviewed the applicants comparison of the Bhlke predictions of volatility to existing test data from a variety of other volatility studies (ML19332A120). The comparison shows good agreement between the Bhlke predictions and the measured data, except for the low temperature range of two specific sets of data, which are dispositioned as representing theoretical maximum volatility rates at fully equilibrium conditions.

In contrast, at low temperatures, the NPM is not expected to approach equilibrium conditions.

The NRC staff also considered information from the REWET-II tests and audited engineering documents that evaluated the calculation of volatility from measured data from those tests, as well as reviewed the applicants justification (ML19332A120) that the Bhlke correlation is appropriate considering the data from these tests. Based on the applicants calculation of the REWET-II BOR008 test, the NRC staff noted that the predicted value using the Bhlke correlation appeared reasonably similar to the measured data, even though the REWET-II BOR008 test included concentrations well above the range of applicability of the Bhlke correlation and the conditions expected in the NPM. The NRC staff also examined the other REWET-II tests and the applicants underlying documentation in the audited document REWET-II Volatility Talking Points and noted that the difference in the volatility rates for the tests could be easily explained by the difference in concentrations between the REWET-II tests and the predictions for the NPM and the impact of the concentration differences on calculated volatility. The NRC staff also noted that some tests, such as REWET-II BOR004, did not appear well controlled, showing widely varying volatility rates from one test to the next. As such, the NRC staff noted that an assessment of the volatility should not take a single test point as representative.

7.4 DCA Part 2, Tier 2, Section 15.1.1, Decrease in Feedwater Temperature The NRC staff audited the underlying transient calculations for DCA Part 2, Tier 2, Section 15.1.1, including sensitivity studies that confirmed the applicants choice of input parameters and initial conditions for the limiting analysis presented in DCA Revisions 3 and 4. The NRC staff also confirmed that the calculation was consistent with the event-specific methodology and strategy of using suitably conservative inputs described in the applicants non-loss-of-coolant accident (non-LOCA) methodology topical report.

In addition, the NRC staff examined the differences in the transient calculation resulting from design and methodology changes subsequent to DCA Revision 2. Some of the differences include:

  • The limiting case changed from a temperature decrease over 160 seconds to a decrease over 86 seconds, which may be due to some or all of the below factors.
  • The calculation was updated to use NRELAP5 Version 1.4, and the model was derived from NRELAP5 Basemodel, Revision 2.
  • SG and DHRS heat transfer uncertainties were not applied.

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  • The Doppler coefficient applied was changed from the least negative (-1.40 pcm/°F) to most negative (-2.5 pcm/°F).
  • ((

))

  • ((

))

  • ((

))

The subchannel calculation engineering change notice (ECN) that the NRC staff audited was consistent with the applicants subchannel analysis methodology as well as the results presented in DCA Revisions 1 and 2. The NRC staff did not request to audit the subchannel calculation revision corresponding to later DCA revisions because of the larger MCHFR margin for this event compared to the at-power reactivity-initiated events (i.e., DCA Part 2, Tier 2, Section 15.4).

7.5 DCA Part 2, Tier 2, Section 15.1.2, Increase in Feedwater Flow The NRC staff audited many of the listed documents associated with DCA Part 2, Tier 2, Section 15.1.2, to better understand the basis for the applicants response to RAI 9483, Question 15.01.01-7 (ML18191B326), about potential overfill of the SGs and potential resulting degradation of DHRS performance. This concern was discussed in a public teleconference with the applicant on January 29, 2019 (ML19240A389), and subsequent audit calls to discuss the relevant audit documents. Ultimately, EC-0000-2016, Revision 1, and the attachment; support the applicants position that the DHRS is still capable of performing its safety-related function under conditions that maximize SG level and present the greatest challenge to DHRS heat removal, as evidenced by reactor coolant system (RCS) temperatures and pressures decreasing over time. These calculations also supported the related information the applicant added to DCA Revision 4, Part 2, Tier 2, Section 15.1.2.

The NRC staff also audited the underlying transient calculations for DCA Part 2, Tier 2, Section 15.1.2, to examine the sensitivity studies, which confirmed the applicants choice of input parameters and initial conditions for the limiting analysis presented in DCA Revisions 3 and 4.

In addition, the NRC staff confirmed that the calculation was consistent with the event-specific methodology and strategy of using suitably conservative inputs described in the applicants non-LOCA methodology topical report.

The NRC staff also examined the differences in the transient calculation resulting from design and methodology changes subsequent to DCA Revision 2. Some of the differences include:

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  • The maximum feedwater flow increase considered was changed from a 100 percent increase from nominal in EC-0000-2016, Revision 0, to the flow rate determined by the maximum feedwater pump curve in EC-0000-2016, Revision 1. This presents a more realistic, yet bounding, feedwater flow increase.
  • The limiting case changed from a 100-percent feedwater flow increase to a 15-percent increase.
  • The calculation was updated to use NRELAP5 Version 1.4, and the model was derived from NRELAP5 Basemodel, Revision 2.
  • SG and DHRS heat transfer uncertainty were investigated but were not applied to the limiting DCA case due to their small impact.
  • ((

))

The subchannel calculation ECN that the NRC staff audited, was consistent with the applicants subchannel analysis methodology as well as the results presented in DCA Revisions 1 and 2.

The NRC staff did not request to audit the subchannel calculation revision corresponding to later DCA revisions because of the larger MCHFR margin for this event compared to the at-power reactivity-initiated events.

7.6 DCA Part 2, Tier 2, Section 15.1.3, Increase in Steam Flow The NRC staff audited the underlying transient calculations for DCA Part 2, Tier 2, Section 15.1.3, including sensitivity studies that confirmed the applicants choice of input parameters and initial conditions for the limiting analysis presented in DCA Revisions 3 and 4. The NRC staff also confirmed that the calculation was consistent with the event-specific methodology and strategy of using suitably conservative inputs described in the applicants non-LOCA methodology topical report. In addition, the calculation used NRELAP5 Version 1.4, and the model was derived from NRELAP5 Basemodel, Revision 2.

The NRC staff noted that the DCA reported maximum RCS and SG pressure values that corresponded to the limiting MCHFR case. While the applicant did not perform sensitivity studies to maximize system pressures, the maximum RCS and SG pressures from the spectrum of MCHFR sensitivity cases were 2,022 and 1,270 psia, respectively.

7.7 DCA Part 2, Tier 2, Section 15.1.5, Steam Piping Failures Inside and Outside of Containment The NRC staff audited calculation files associated with the applicants main steam line break analysis to support the review of DCA Part 2, Tier 2, Section 15.1.5.

The NRC staff examined the initial conditions, boundary conditions, assumptions, and input for the applicants method of biasing input parameters in the conservative direction. The NRC 18

staffs audit included examining outcomes of certain parameter biasing and break size/location sensitivity calculations for their impact on calculation results. The NRC staff also examined the audit calculation files for the applicants single failure assumptions presented in DCA Part 2, Tier 2, Section 15.1.5.

The NRC staff examined the audited documents with respect to following the non-LOCA methodology referenced in this section of the DCA. In addition, the NRC staff examined the subchannel analysis calculation with respect to following the subchannel analysis methodology for calculating the MCHFR associated with this Chapter 15 event.

The NRC staff also examined the audited calculations to understand the differences between how the calculations were performed for DCA Revision 2, compared to the updated analyses that included design changes and other updates.

7.8 DCA Part 2, Tier 2, Sections 15.2.1-15.2.3, Loss of External Load, Turbine Trip, and Loss of Condenser Vacuum The NRC staff audited the underlying transient calculations for DCA Part 2, Tier 2, Sections 15.2.1-15.2.3, including sensitivity studies that confirmed the applicants choice of input parameters and initial conditions for the limiting analyses presented in DCA Revisions 3 and 4.

The NRC staff also confirmed that the calculation, particularly the latest revision, was consistent with the event-specific methodology and strategy of using suitably conservative inputs described in the applicants non-LOCA methodology topical report. In addition, the NRC staff observed that:

  • The transient calculation was updated to use NRELAP5 Version 1.4, and the model was derived from NRELAP5 Basemodel, Revision 2.
  • SG and DHRS heat transfer uncertainties were investigated and were applied to the limiting DCA cases, as appropriate, although the impacts are not significant.

The NRC staff audited the subchannel calculation and ECN and engaged the applicant in audit calls to understand differences in the time-dependent trend of MCHFR between DCA Revisions 0 and 1, for heatup events, a topic also discussed with the applicant in a public teleconference (ML19086A041). The NRC staff came to understand that the minor differences were due to the use of the NSP2 and NSP4 correlations, respectively, and a corresponding difference in prediction of local conditions. The NRC staff also noted that the ECN was consistent with the applicants subchannel analysis methodology as well as the results presented in DCA Revisions 1 and 2. The NRC staff did not request to audit later revisions of the subchannel calculation for this event because of the significant MCHFR margin for this event compared to the at-power reactivity-initiated events.

7.9 DCA Part 2, Tier 2, Section 15.2.4, Closure of Main Steam Isolation Valve(s)

The NRC staff audited the underlying transient calculations for DCA Part 2, Tier 2, Section 15.2.4, including sensitivity studies that confirmed the applicants choice of input parameters and initial conditions for the limiting analyses presented in DCA Revisions 3 and 4. The NRC staff also confirmed that the calculations were consistent with the event-specific methodology 19

and strategy of using suitably conservative inputs described in the applicants non-LOCA methodology topical report. In addition, the NRC staff observed that:

  • The transient calculation was updated to use NRELAP5 Version 1.4, and the model was derived from NRELAP5 Basemodel, Revision 2.
  • SG heat transfer uncertainties were investigated and were applied to the limiting DCA cases, as appropriate, although their impacts are not significant.

The NRC staff audited the subchannel calculation and engineering change notice (ECN) in a similar manner to what was described under DCA Part 2, Tier 2, Sections 15.2.1-15.2.3, and had similar observations.

7.10 DCA Part 2, Tier 2, Section 15.2.6, Loss of Non-Emergency AC Power to the Station Auxiliaries The NRC staff audited the underlying transient calculations for DCA Part 2, Tier 2, Section 15.2.6, including sensitivity studies that confirmed the applicants choice of input parameters and initial conditions for the limiting analyses presented in DCA Revisions 3 and 4. The NRC staff also confirmed that the calculations were consistent with the event-specific methodology and strategy of using suitably conservative inputs described in the applicants non-LOCA methodology topical report. In addition, the NRC staff observed that:

  • The transient calculation was updated to use NRELAP5 Version 1.4, and the model was derived from NRELAP5 Basemodel, Revision 2.
  • SG and DHRS heat transfer uncertainty were investigated but were not applied to the limiting DCA case due to their small impact.

The NRC staff audited the subchannel calculation and ECN in a similar manner to what was described under DCA Part 2, Tier 2, Sections 15.2.1-15.2.3, and had similar observations.

7.11 DCA Part 2, Tier 2, Section 15.2.7, Loss of Normal Feedwater Flow The NRC staff audited calculation files associated with the applicants loss of normal feedwater flow analysis to support the review of DCA Part 2, Tier 2, Section 15.2.7.

The NRC staff examined the initial conditions, boundary conditions, assumptions, and input for the applicants method of biasing input parameters in the conservative direction. The NRC staffs audit included examining outcomes of certain parameter biasing for their impact on calculation results. The NRC staff also examined the audit calculation files for the applicants single failure assumptions presented in DCA Part 2, Tier 2, Section 15.2.7.

The NRC staff examined the audited documents with respect to following the non-LOCA methodology referenced in this section of the DCA. In addition, the NRC staff examined the subchannel analysis calculation with respect to following the subchannel analysis methodology for calculating the MCHFR associated with this Chapter 15 event.

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The NRC staff also examined the audited calculations to understand the differences between how the calculations were performed for DCA Revision 2, compared to the updated analyses that included design changes and other updates.

7.12 DCA Part 2, Tier 2, Section 15.2.8, Feedwater System Pipe Breaks Inside and Outside of Containment The NRC staff audited calculation files associated with the applicants feedwater line break analyses to support the review of DCA Part 2, Tier 2, Section 15.2.8.

The NRC staff examined the initial conditions, boundary conditions, assumptions, and input for the applicants method of biasing input parameters in the conservative direction. The NRC staffs audit included examining outcomes of certain parameter biasing and break size/location sensitivity calculations for their impact on calculation results. The NRC staff also examined the audit calculation files for the applicants single failure assumptions presented in DCA Part 2, Tier 2, Section 15.2.8.

The NRC staff examined the audited documents with respect to following the non-LOCA methodology referenced in this section of the DCA. In addition, the NRC staff examined the subchannel analysis calculation with respect to following the subchannel analysis methodology for calculating the MCHFR associated with this Chapter 15 event.

The NRC staff also examined the audited calculations to understand the differences between how the calculations were performed for DCA Revision 2, compared to the updated analyses that included design changes and other updates. During the audit, the applicant identified an error in the analyses. The applicant provided updated, corrected analyses for the NRC staff to examine. The NRC staff examined the updated, corrected analyses and the changes are consistent with previous audit activities described above.

7.13 DCA Part 2, Tier 2, Section 15.2.9, Inadvertent Operation of the Decay Heat Removal System The NRC staff audited calculation files associated with the applicants inadvertent DHRS operation analyses to support the review of DCA Part 2, Tier 2, Section 15.2.9.

The NRC staff examined the initial conditions, boundary conditions, assumptions, and input for the applicants method of biasing input parameters in the conservative direction. The NRC staffs audit included examining outcomes of certain parameter biasing as well as actuation scenario sensitivity calculations for their impact on calculation results. The NRC staff also examined the audit calculation files for the applicants single failure assumptions presented in DCA Part 2, Tier 2, Section 15.2.9.

The NRC staff examined the audited documents with respect to following the non-LOCA methodology referenced in this section of the DCA. In addition, the NRC staff examined the subchannel analysis calculation with respect to following the subchannel analysis methodology for calculating the MCHFR associated with this Chapter 15 event.

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The NRC staff also examined the audited calculations to understand the differences between how the calculations were performed for DCA Revision 2, compared to the updated analyses that included design changes and other updates.

7.14 DCA Part 2, Tier 2, Section 15.4.1, Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power or Startup Condition The NRC staff audited the underlying transient calculation for DCA Part 2, Tier 2, Section 15.4.1, including sensitivity studies that confirmed the applicants choice of input parameters and initial conditions for the limiting analysis presented in DCA Revisions 3 and 4.

The NRC staff also confirmed that the calculation was consistent with the event-specific methodology and strategy of using suitably conservative inputs described in the applicants non-LOCA methodology topical report.

In addition, the NRC staff examined the differences in the transient calculation resulting from design and methodology changes subsequent to DCA Revision 2. Some of the differences include:

  • The initiating power level for the limiting case changed from 1 MW to 24 MW, and the limiting reactivity insertion rate changed from 13.37 to 0.014 pcm/s, which may be due to some of the following factors. Despite these changes, there is still ample margin to acceptance criteria for this event.
  • The calculation was updated to use NRELAP5 Version 1.4, and the model was derived from NRELAP5 Basemodel, Revision 2.
  • The applicant updated the kinetics parameters to be consistent with the most recent nuclear analysis.
  • The applicant performed a more robust set of sensitivity studies that considered additional reactivity insertion rates at each power level for cases initiating from intermediate range conditions.
  • The applicant did not examine sensitivities of primary or secondary pressures because the pressure response for this event is bounded by that of other events.

The NRC staff also confirmed that the subchannel calculation followed the applicants subchannel analysis methodology, including the use of limiting axial and radial power shapes.

In addition, the NRC staff confirmed that the applicant analyzed the three potentially limiting cases identified from the NRELAP5 transient calculation.

7.15 DCA Part 2, Tier 2, Section 15.4.2, Uncontrolled Control Rod Assembly Withdrawal at Power The NRC staff audited the underlying transient calculation for DCA Part 2, Tier 2, Section 15.4.2, including sensitivity studies that confirmed the input parameters, initial conditions, and loss of power assumption for the limiting analysis presented in DCA Revisions 3 and 4. The NRC staff also confirmed that the calculation was consistent with the event-specific 22

methodology and strategy of using suitably conservative inputs described in the applicants non-LOCA methodology topical report.

In addition, the NRC staff examined the differences in the transient calculation resulting from design and methodology changes subsequent to DCA Revision 2. Some of the differences include:

  • The limiting reactivity insertion rate for MCHFR changed slightly, from 0.9 to 0.92 pcm/s.
  • The applicant updated the kinetics parameters to be consistent with the most recent nuclear analysis.
  • The Doppler temperature coefficient was updated from -1.4 pcm/°F to -1.377 pcm/°F since -1.4 pcm/°F may not be bounding above 130 percent power, and such power levels may be encountered in the calculation. Negative 1.377 pcm/°F applies up to 150 percent power.
  • The calculation was updated to use NRELAP5 Version 1.4, and the model was derived from NRELAP5 Basemodel Revision 2.
  • The applicant did not examine sensitivities of primary or secondary pressures in depth because the pressure response for this event is no more challenging than that of other events.

The NRC staff also confirmed that the subchannel calculation followed the applicants subchannel analysis methodology, including the use of limiting axial and radial power shapes.

In addition, the NRC staff noted that the applicant performed a subchannel analysis of all cases from the NRELAP5 transient calculation except the best-estimate (e.g., unbiased) case.

7.16 DCA Part 2, Tier 2, Section 15.4.3, Control Rod Misoperation (System Malfunction or Operator Error)

The NRC staff audited several documents related to DCA Part 2, Tier 2, Section 15.4.3, to confirm that the applicant identified the most limiting cases and used appropriately conservative inputs and initial conditions.

The NRC staff audited the calculation EC-0000-4309, Revision 3, to understand aspects of the applicants response to RAI 9512, Question 15.04.03-2 (ML18163A421) related to the control rod misalignment event, as well as differences in calculation results between DCA Revisions 2 and 3. Most notably, the calculated control rod misalignment MCHFR dropped from 2.509 to 1.437. Revisions 2 and 3, of EC-0000-4309 were generally similar and used limiting axial and radial power shapes from the time of the calculation. The major differences between the revisions, which explain the MCHFR change in the DCA, include:

  • Revision 3 was updated to use the most current limiting power shapes.

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  • Revision 3 used a minimum system flow of 535 kg/s simultaneously with a maximum core inlet temperature of 510 °F. This is more conservative than Revision 2, which used the minimum core inlet temperature corresponding to the minimum system flow, then biased high by 10 °F (essentially translating to a nominal core inlet temperature relative to Chapter 15 initial conditions).
  • Revision 3 applied a radial peaking augmentation factor of (( )). As described in document ER-0000-2486, Revision 7, Safety Analysis Limits Report (further discussed in Section 7.1 of this summary), this is a cycle-specific limit intended to be confirmed for each reload. EC-A021-7235, Revision 0, shows that the bounding radial peaking augmentation factor calculated for a control rod misalignment is (( )), which is within the limit of (( )).

The NRC staff audited calculations EC-A021-1922, ECN-A021-4989, and EC-A021-5277 to understand the steady-state nuclear calculations that provided input to the single rod withdrawal transient calculation. EC-A021-1922 (as modified by ECN-A021-4989) calculated static rod worth, radial peaking augmentation factors, power shapes, and changes in relative power fraction for the peripheral fuel assemblies considering several different scenarios, including:

various power levels (100, 75, 50, and 25 percent); three times in life (beginning, middle, and end of cycle); both regulating banks at the power-dependent insertion limits and regulating bank 1 at the regulating bank 2 limits; minimum and maximum axial offsets; two different withdrawn rod locations; and peak and equilibrium xenon. The NRC staff confirmed that the radial peaking augmentation factors for a single rod withdrawal event used in EC-0000-2897 bound the ones calculated in EC-A021-1922. However, results presented in EC-A021-7235 supersede the EC-A021-1922 results. EC-A021-5277 defined the relationship between the static rod worth and relative power fraction ratios.

EC-0000-2139 documented the transient calculations for the single rod withdrawal and control rod drop events, including sensitivity studies that confirmed the input parameters and initial conditions for the limiting analyses presented in DCA Revisions 3 and 4. The NRC staff also confirmed that the calculation was consistent with the event-specific methodology and strategy of using suitably conservative inputs described in the applicants non-LOCA methodology topical report. The NRC staff also noted new or different information compared to the last revision of the document, which was part of the previous Chapter 15 audit (ML19270G302). The major differences included:

  • The limiting reactivity insertion rate for MCHFR for the single rod withdrawal changed from 2.5 to 1.32 pcm/s. The NRC staff noted that DCA Revisions 3 and 4, still said the limiting rate was 2.5 pcm/s, and the applicant has submitted errata to the DCA (ML20092L899) with markups that correct this error.
  • The applicant updated the applicable nuclear parameters to be consistent with those in EC-A021-7235.
  • The moderator temperature coefficient applied for the single rod withdrawal was updated from a power-dependent value to a more conservative value of 0 pcm/°F.

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  • The most negative Doppler temperature coefficient applied for the control rod drop was updated from -2.25 pcm/°F to 2.5 pcm/°F, consistent with ER-0000-2486.
  • The calculation was updated to use NRELAP5 Version 1.4, and the model was derived from NRELAP5 Basemodel, Revision 2.
  • The applicant did not examine sensitivities of primary or secondary pressures in depth because the pressure response for this event is no more challenging than that of other events.

The NRC staff also confirmed that the subchannel calculation for the control rod drop and single rod withdrawal events in EC-0000-2897 used limiting axial and radial power shapes and generally followed the applicants subchannel analysis methodology. This included the radial peaking augmentation factors from EC-A021-7235 and ER-0000-2486. In addition, the axial power shape used for the single rod withdrawal event, the pre-event limiting axial power profile, was a departure from the subchannel methodology, which states that the post-event axial power distribution should be used. However, the applicant noted this was conservative since the power peaking was higher for the pre-event profile than the post-event profile. The NRC staff also noted that the applicant performed a subchannel analysis of all cases from the NRELAP5 transient calculation except the best-estimate (e.g., unbiased) case.

In addition, based on its audit of EC-0000-2897, the NRC staff noted that the applicant did not apply a radial peaking augmentation factor in the calculation of linear heat generation rate (LHGR) for the control rod drop and single rod withdrawal events, which resulted in a small non-conservative decrease in the LHGR values reported in DCA Revisions 3 and 4. The applicant has committed to include a correction in future errata to be submitted to the NRC.

EC-A021-7235 Revision 0, shows that the bounding radial peaking augmentation factors calculated for control rod drop at 100, 75, 50, and 25 percent power are ((

)), respectively, which are within the limits of (( )),

respectively. For the single rod withdrawal event, the bounding value of (( )) was within the limit of (( )); however, this did not account for power-specific values, as was done for the control rod drop.

7.17 DCA Part 2, Tier 2, Section 15.4.7, Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position The NRC staff audited the listed calculations underlying DCA Part 2, Tier 2, Section 15.4.7, to understand aspects of the applicants response to RAI 9504, Question 15.04.07-3 (ML18179A522), as well as differences in calculation results between DCA Revisions 2 and 3.

Most notably, the calculated MCHFR dropped from 1.916 to 1.437.

The NRC staff confirmed that EC-A021-1858, Revision 1, corrected the deficiency in calculating the limiting radial peaking augmentation factor mentioned in the response to the above RAI and further discussed during a public teleconference (ML19240A389). The applicant calculated augmentation factors for all possible fuel misload cases, and the limiting one was (( )).

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Revisions 2 and 3 of the subchannel analysis were generally similar and used limiting axial and radial power shapes at the time of the calculation. In addition, both used an RCS pressure bias of 70 psia. The major differences between the revisions include:

  • Revision 3 was updated to use the most current limiting power shapes.
  • Revision 3 used a minimum system flow of 535 kg/s simultaneously with a maximum core inlet temperature of 510 °F. This is more conservative than Revision 2, which used the minimum core inlet temperature corresponding to the minimum system flow, then biased high by 10 °F (essentially translating to a nominal core inlet temperature relative to Chapter 15 initial conditions).
  • Revision 3 applied a radial peaking augmentation factor of 1.25, corresponding to the cycle-specific limit in ER-0000-2486. EC-A021-7235, Revision 0, shows that the bounding radial peaking augmentation factor calculated for a fuel misload is ((

)), which is within the limit of (( )).

7.18 DCA Part 2, Tier 2, Section 15.5.1, Chemical and Volume Control System Malfunction The NRC staff examined EC-0000-2726, Revision 1, to understand the differences between how the calculations were performed for DCA Revision 2, compared to the updated analyses that were revised to address base model changes and the change to NRELAP5, Version 1.4.

7.19 DCA Part 2, Tier 2, Section 15.6.2, Failure of Small Lines Carrying Primary Coolant Outside Containment The NRC staff audited calculation files associated with the applicants small line break analyses to support the review of DCA Part 2, Tier 2, Section 15.6.2.

The NRC staff examined the initial conditions, boundary conditions, assumptions, and input for the applicants method of biasing input parameters in the conservative direction. The NRC staffs audit included examining outcomes of certain parameter biasing and break type sensitivity calculations for their impact on calculation results. The NRC staff also examined the audit calculation files for the applicants single failure assumptions presented in DCA Part 2, Tier 2, Section 15.6.2.

The NRC staff examined the audited documents with respect to following the non-LOCA methodology referenced in this section of the DCA.

The NRC staff also examined the audited calculations to understand the differences between how the calculations were performed for DCA Revision 2, compared to the updated analyses that included design changes and other updates.

7.20 DCA Part 2, Tier 2, Section 15.6.3, Steam Generator Tube Failure (Thermal Hydraulic)

The NRC staff examined EC-0000-1735, Revisions 1 and 2, to understand the differences between how the calculations were performed for DCA Revision 2, compared to the updated 26

analyses that were revised to address base model changes and the change to NRELAP5, Version 1.4.

8.0 EXIT BRIEFING The NRC staff conducted an audit exit meeting via teleconference on December 18, 2019.

During the meeting, the NRC staff reiterated the purpose of the audit and discussed the audit activities and outcome.

In summary, the NRC staff examined numerous calculation notes and supporting documents on paper and in the ERR as well as held multiple teleconferences to discuss various topics and issues as they emerged and developed during the audit. In general, the audit greatly enhanced the NRC staffs understanding of the applicants RAI responses and updated analyses.

9.0 REQUESTS FOR ADDITIONAL INFORMATION RESULTING FROM AUDIT The NRC staff did not issue any RAIs or supplemental RAIs as a result of the audit.

10.0 OPEN ITEMS AND PROPOSED CLOSURE PATHS Not applicable.

11.0 DEVIATIONS FROM THE AUDIT PLAN Not applicable.

12.0 REFERENCES

1. Audit Plan for the Regulatory Audit of NuScale Power, LLC Design Certification Application, Chapter 15, Transient and Accident Analysis, December 20, 2018, ML19004A098.
2. NuScale Power, LLC Submittal of the NuScale Standard Plant Design Certification Application, NuScale, December 31, 2016, ML17013A229.

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