ML19092A361
| ML19092A361 | |
| Person / Time | |
|---|---|
| Site: | PROJ0769 |
| Issue date: | 04/02/2019 |
| From: | Bavol B NRC/NRO/DLSE/LB1 |
| To: | Samson Lee NRC/NRO/DLSE/LB1 |
| Bavol B | |
| Shared Package | |
| ML19092A359 | List: |
| References | |
| Download: ML19092A361 (14) | |
Text
AUDIT
SUMMARY
FOR THE REGULATORY AUDIT OF NUSCALE POWER, LLC DESIGN CERTIFICATION APPLICATION FINAL SAFETY ANALYSIS REPORT SECTION 4.2, FUEL SYSTEMS DESIGN, SECTION 4.3, NUCLEAR DESIGN, AND SECTION 4.4, THERMAL AND HYDRAULIC DESIGN I. BACKGROUND NuScale Power, LLC (NuScale) submitted by letter dated December 31, 2016, to the U.S.
Nuclear Regulatory Commission (NRC), a Design Certification Application (DCA) for the NuScale design (Reference 1). The NRC staff started its detailed technical review of NuScales DC application on March 27, 2017.
The purpose of this regulatory audit was to: (1) gain a better understanding of the evaluations supporting the fuel, nuclear, and thermal-hydraulic design presented in the NuScale Final Safety Analysis Report (FSAR), Tier 2, Chapter 4; (2) verify analyses inputs and methodologies; and (3) identify information that would require docketing to support the basis of the licensing or regulatory decision. Additional background in available in the audit plan associated with this audit summary (Reference 2).
II. REGULATORY AUDIT BASES Title 10 of the Code of Federal Regulations (CFR), Section 52.47(a)(3)(i) states:
A DC application must contain a final safety analysis report (FSAR) that includes a description of principle design criteria for the facility.
This regulatory audit is based on the following regulations:
- 10 CFR 52.47, Contents of applications; technical information in final safety analysis report.
- General Design Criteria (GDC) 2 Design Bases for Protection Against Natural Phenomena, of Appendix A to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, as it relates to the structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena without loss of capability to perform their safety functions.
- GDC 10, Reactor Design, which requires that reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
III. AUDIT LOCATION AND DATES The audit was conducted from the NRC headquarters via NuScales electronic reading room.
Enclosure 1
Date: March 19, 2018 - April 30, 2018 Location: NRC Headquarters Two White Flint North 11545 Rockville Pike Rockville, MD 20852-2738 IV. AUDIT TEAM MEMBERS Chris Van Wert (NRO, Audit Lead)
Tim Drzewiecki (NRO)
Bruce Bavol (NRO, Project Manager)
NRC staff was augmented with contract support. These additional participants include the following:
Pacific Northwest National Laboratory Ken Geelhood Nick Klymyshyn V. APPLICANT AND INDUSTRY STAFF PARTICIPANTS NuScale Power Steve Mirsky Jennie Wike VI. DOCUMENTS AUDITED FSAR Section 4.2:
- FS1-0023557, Rev. 2, NuScale Damping Estimation for Lateral Accident Condition Analysis
- NuScale Rod Bow Penalty Calculation and Application FSAR Section 4.3:
- EC-A021-1859, Rev. 1, Xenon Stability Analysis, 11/27/2016.
- ER-A021-4818, Rev. 0, Assessment of Control Rod Assembly Absorber Depletion, 10/27/2016.
- ER-A025-4219, Rev. 2, Final Summary Report for NuScale CRA, 11/17/2016.
FSAR Section 4.4:
- EC-A025-3562, Rev. 0, NuScale Hot Channel Factors Evaluation, 9/18/2015.
- ER-0000-2486, Rev. 5, Safety Analysis Analytical Limits Report, 9/13/2017.
- EC-0000-2347, Rev. 3, Steady-State Subchannel Analysis, 11/30/2017.
- EC-A010-3204, Rev. 1, RCS Loop CFD, 7/11/2017.
- EC-0000-2337, Rev. 5, Subchannel Analysis Methodology, 11/29/2017.
VII. DESCRIPTION OF AUDIT ACTIVITIES AND
SUMMARY
OF OBSERVATIONS FSAR Section 4.2:
NRC staff examined NuScales documentation regarding the rod bow penalty methodology and analysis. Information from this documentation was used by the staff to inform confirmatory analyses. The results from the staffs confirmatory analyses did not lead to any new RAIs.
NRC staff examined the NuScale fuel assembly damping document, FS1-0023557, Revision 2.
The information from this document was used for confirmatory analyses. No new RAIs were generated as a result of the confirmatory analysis.
FSAR Section 4.3:
Xenon Stability NRC staff examined engineering calculation EC-A021-1859, Rev. 1, Xenon Stability Analysis.
During this examination NRC staff noted the purpose of the calculation, identified key inputs and their sources, and noted the key results. In particular, NRC staff noted:
- The purpose of this calculation is to determine whether the NuScale reactor core (RXC) is stable with respect to axial and radial Xenon-135 induced power oscillations.
- There are no unverified input assumptions used in EC-A0221-1859, Rev. 1.
- The analysis uses design moderator temperatures [
].
- The analysis uses reactor coolant design flows [
].
- The analysis uses power dependent insertion limits (PDILs) [
]. NRC staff compared the PDILs used this EC-A021-1859, Rev. 1 against the PDILs provided in the FSAR, Tier 2, Figure 4.3-2, and found them to be consistent.
- The analysis uses an operating pressure that is consistent with the value provided in FSAR, Tier 2, Table 4.4-2.
- The analysis uses a control rod assembly (CRA) position uncertainty that is consistent with NuScale Generic Technical Specification (GTS) 3.1.4.
- The analyses is performed using the Studsvik Scandpower Core Management Software (CMS5).
- The analysis was performed for equilibrium Cycle 8 at beginning of cycle (BOC and end of cycle (EOC) in order to bound the reactivity coefficients.
- Xenon stability analysis was performed at multiple power levels of 100 percent, 75 percent, 50 percent, and 25 percent.
- Axial and radial xenon oscillations are induced by perturbing the RXC with the insertion and removal of the regulating CRA groups. The analysis introduces a bounding perturbation by inserting regulating CRA groups to a position of 105 steps withdrawn, which is about halfway into the core and below the PDIL. The xenon transient is performed [
] The results [
] stability indices.
o For axial xenon stability analysis [
].
o For radial xenon stability analysis [
]
- NRC staff compared the calculated axial and radial xenon stability indices against the values provided in FSAR, Tier 2, Table 4.3-9, Table 4.3-10, and Table 4.3-11 and found the values reported in the FSAR to be consistent with the results documented in EC-A021-1859, Rev. 1.
CRA Design NRC staff examined engineering report ER-A025-4219, Rev. 2, Final Summary Report for NuScale CRA. During this examination NRC staff noted the purpose of the report, identified key inputs and their sources, and noted the key results. In particular, NRC staff noted:
- The purpose of this report is to provide a summary of the CRA design work completed by AREVA.
- An evaluation of CRA clad strain and fluence lifetime was performed:
o CRA clad strain is primarily driven by the swelling of the control rod absorbers.
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- An evaluation of CRA clad stress was performed:
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Stress Category Temperature Allowable Stress
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- An evaluation of cladding fatigue was performed:
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- An evaluation of cladding wear was performed:
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- An evaluation of rod internal pressure was performed:
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o Over the lifetime of the CRA, there is very low depletion due to rod movements or reactor trips.
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- An evaluation of fluence accumulation was performed:
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CRA Depletion NRC staff examined engineering report ER-A021-4818, Rev. 0, Assessment of Control Rod Assembly Absorber Depletion. During this examination NRC staff noted the purpose of the report, identified key inputs and their sources, and noted the key results. In particular, NRC staff noted:
- The purpose of the report is to assess the uncertainty associated with the NuScale reactor control rods with respect to depletion of the CRA absorber.
- In a separate report, ER-A025-4219, Rev. 1, Final Summary Report for NuScale CRA, NuScale documented that, [
]
- [
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- The report concludes that, [
]:
- 1. [
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[
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FSAR Section 4.4:
Critical Heat Flux Ratio (CHFR) Penalties NRC staff examined engineering calculation EC-A025-3562, Rev. 0, NuScale Hot Channel Factors Evaluation. During this examination NRC staff noted the purpose of the calculation and identified key inputs and their sources. In particular, NRC staff noted:
- This calculation verifies that the [
] is appropriate.
- The calculation uses the following assumptions:
- 1. [
]
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- The analysis results in a local hot channel factor for heat flux (FQE) [ ] which is bounded by the assumed value [ ].
NRC staff examined engineering calculation EC-0000-2337, Rev. 5, Subchannel Analysis Methodology, with a focus on examining the CHFR penalties and core bypass flow through the guide tube and instrument tubes. During this examination NRC staff noted the purpose of the calculation and identified key inputs and their sources. In particular, NRC staff noted:
- The purpose of this calculation is to describe the methodology used for steady-state and transient subchannel analyses.
- The calculation states that, [
].
- The applicant provided a sample calculation, NuScale Rod Bow Penalty and Application, which shows that the rod bow penalty used for licensing calculations is conservative.
- [
].
Thermal Margin Limits NRC staff examined engineering calculation EC-0000-2347, Rev. 3, Steady-State Subchannel Analysis. During this examination NRC staff noted the purpose of the calculation and identified key inputs and their sources. In particular, NRC staff noted:
- The purpose of this calculation is to perform steady-state subchannel analysis of the NuScale Power Module (NPM).
o In order to support the thermal margin analysis for Anticipated Operational Occurrences (AOOs), [ ].
o A thermal margin assessment [
].
o An axial power shape analysis is performed [
].
- The analysis is performed in accordance with TR-0915-17564, Rev. 1, Subchannel Analysis Methodology.
- The acceptance criteria for the analysis [
].
- The analysis uses [
]
- The thermal margin limit analyses are performed [
].
Core Bypass Flow NRC staff examined engineering calculation EC-A010-3204, Rev. 1, RCS Loop CFD, with a focus on the evaluation of core bypass flow. During this examination NRC staff noted the purpose of the report and identified key inputs and their sources. In particular, NRC staff noted:
- The purpose of the calculation is to provide detailed results for many different aspects of the RCS loop flow, including bypass flow through the reflector cooling channels.
- The analyses were performed [
].
- [
].
- [
].
- [
]. NRC staff notes that the combined values for bypass flow (Reflector Cooling, Leakage Between Reflector Blocks and Core Barrel, and Guide Tube Bypass) is bounded by the design basis core bypass flow fraction from FSAR, Tier 2, Section 4.4.1.3 of the NuScale DCA.
Safety Analysis Analytical Limits NRC staff examined engineering calculation ER-0000-2486, Rev. 5, Safety Analysis Analytical Limits Report. During this examination NRC staff noted the purpose of the report and identified key inputs and their sources. In particular, NRC staff noted:
- The purpose of the report is to document the selection of a set of high level design parameters and analytical limits to use in the safety analysis of the NPM. [
].
- This report consolidates and documents [
]:
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- The following list comprises the safety-related process measurements with a description of their function and how they are credited in the safety analyses.
Measurement Safety Related Detection Purpose
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- NPM initial conditions are the associated basis are provided. This information is consistent with the information provided in FSAR, Tier 2, Table 15.0-6 of the NuScale DCA.
- Analytical limits and the associated basis are provided. This information is consistent with the information provided in FSAR, Tier 2, Table 15.0-7 of the NuScale DCA.
- The power dependent insertion limits (PDILs) are provided. This information is consistent with the information provided in FSAR, Tier 2, Figure 4.3-2 of the NuScale DCA.
- The axial offset window is provided. This information is consistent with the information provided in FSAR, Tier 2, Figure 4.3-3 of the NuScale DCA.
- Core design limits used assumed in the safety analyses are provided in the table below:
Parameter Conditions Design Limit
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- Safety related valve analytical characteristics are provided in the table below:
Description Value
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VIII. EXIT BRIEFING The NRC staff conducted an audit closeout meeting on June 14, 2018. At the exit briefing the NRC staff reiterated the purpose of the audit and discussed their activities. Additionally, the NRC staff stated that they did not identify areas where additional information would be necessary to support the review.
IX. REQUESTS FOR ADDITIONAL INFORMATION RESULTING FROM AUDIT No RAIs were generated as a result of this audit.
X. OPEN ITEMS AND PROPOSED CLOSURE PATHS Not applicable.
XI. DEVIATIONS FROM THE AUDIT PLAN The audit was extended to 06/14/2018.
XII. REFERENCES
- 1. NuScale Power, LLC Submittal of the NuScale Standard Plant Design Certification Application (NRC Project No. 0769), December 31, 2016. (ADAMS Accession No. ML17013A229)
- 2. Supplement 1 to the Audit Plan for the Regulatory Audit of NuScale Power, LLC Design Certification Application Chapter 4, Reactor; Chapter 5, Reactor Coolant and Connecting Systems; and Chapter 9, Auxiliary Systems, May 9, 2017. (ADAMS Accession No. ML17124A339)