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LLC Submittal of Presentation Materials Entitled Statistical Subchannel Analysis Methodology and Rod Ejection Accident Methodology Topical Reports, ACRS Open Session, PM-154736, Revision 0 (Open Session)
	ML24025A144
Person / Time
Site:	99902078, 05200050
Issue date:	01/25/2024
From:	Griffith T; NuScale
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	LO-156239
	Download: ML24025A144 (1)
	v • d • e

LO-156239 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com January 25, 2024 Docket No.52-050 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 Presentation Materials Entitled Statistical Subchannel Analysis Methodology and Rod Ejection Accident Methodology Topical Reports, ACRS Open Session, PM-154736, Revision 0 (Open Session)

The purpose of this submittal is to provide presentation materials to the NRC for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on February 6, 2024. The materials support NuScales Statistical Subchannel Analysis Methodology and Rod Ejection Accident Methodology topical reports of the NuScale Standard Design Approval Application.

The enclosure to this letter is the nonproprietary version of the presentation entitled Statistical Subchannel Analysis Methodology and Rod Ejection Accident Methodology Topical Reports, ACRS Open Session.

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

If you have any questions, please contact Wren Fowler at 541-452-7183 or sfowler@nuscalepower.com.

Sincerely, Tom Griffith Manager, Licensing NuScale Power, LLC Distribution:

Mahmoud Jardaneh, NRC Getachew Tesfaye, NRC Mike Snodderly, NRC

Enclosure:

Statistical Subchannel Analysis Methodology and Rod Ejection Accident Methodology Topical Reports, ACRS Open Session, PM-154736, Revision 0 (Open Session)

Tom Grifffif th

LO-156239 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

Enclosure:

Statistical Subchannel Analysis Methodology and Rod Ejection Accident Methodology Topical Reports, ACRS Open Session, PM-154736, Revision 0 (Open Session)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Statistical Subchannel Analysis Methodology and Rod Ejection Accident Methodology Topical Reports February 6, 2024 ACRS Open Session

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0008928.

This presentation was prepared as an account of work sponsored by an agency of the United States (U.S.)

Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Statistical Subchannel Analysis Methodology, TR-108601-P Supplement 1 to Subchannel Analysis Methodology, TR-0915-17564-P-A, Rev. 2

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Statistical Subchannel Analysis Methodology - History

Subchannel Analysis Methodology, TR-0915-17564-P-A, Rev. 2 o Approved by NRC in December 2018 and previously presented to ACRS August 24, 2018 subcommittee meeting September 6, 2018 full committee meeting o Used for the NuScale US600 design codified in 10 CFR 52 Appendix G

Statistical Subchannel Analysis Methodology, TR-108601-P, Rev. 0 submitted in December 2021 o Serves as a supplement to TR-0915-17564-P-A, Rev. 2

NRC staff performed review and audit of TR-108601-P o One request for supplemental information (RSI) o No requests for additional information (RAIs) o Multiple audit questions

TR-108601-P was revised during the review to address NRC staff feedback

Current revision is TR-108601-P, Rev. 4 - submitted November 2023

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Overview of Previous Subchannel Methodology in TR-0915-17564-P-A, Rev. 2

VIPRE-01 used for steady-state and transient analysis

Methodology fulfills requirements of VIPRE-01 generic safety evaluation report (SER) limitations

Methodology application and treatment of uncertainties

Objective: critical heat flux (CHF) and fuel centerline melt

General methodology approach:

o Input uncertainties treated deterministically; no credit for statistical randomness o Conservative basemodel development o Generic cycle-independent radial power distribution o Bounding axial power shapes o Detailed radial and axial nodalization evaluations o Detailed checklist to ensure compliance with method

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Subchannel Methodology Changes in TR-108601-P

Changes from TR-0915-17564-P-A, Rev 2:

o Treatment of uncertainties - statistical for a set of parameters instead of deterministic approach o Radial nodalization o Axial domain o Axial nodalization

Unchanged:

o Fuel conduction o Grid and frictional losses o Cross-flow and mixing o Qualification (validation and applicability)

The Statistical Subchannel Analysis Methodology utilizes a statistical approach in defining the CHF analysis limit; but, many aspects of the methodology continue to employ a conservative deterministic approach (e.g., axial and radial power profiles)

The intent of introducing a statistical treatment of uncertainties in certain areas was to reduce some of the overly conservative treatments with a defendable basis and to provide a better representation of the physical response

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Subchannel Methodology: Statistical vs. Deterministic

Deterministic: Event analysis input uncertainties (power distributions, boundary conditions, tolerances, etc.) are biased independently in the limiting direction o Range of axial and radial power distributions allowed by operations not treated statistically o Variations possible from: exposure, power, boron concentration, control rod insertion, axial offset, etc.

o As in existing approved methodology:

Radial power distribution: Artificially created to preserve measured Technical Specification allowed radial peaking and minimize beneficial cross-flow in analysis Axial power distribution: Search performed for limiting shape allowed by axial offset

Statistical: All uncertainties associated with both CHF correlation and event analysis inputs are statistically treated in order to determine the CHF analysis limit o Statistical approach accounts for all uncertainties with a 95% probability at the 95% confidence level o Statistical approach continues to require use of an approved CHF correlation with a 95/95 design limit

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Rod Ejection Accident Methodology, TR-0716-50350-P

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Rod Ejection Accident Methodology - History

Rod Ejection Accident Methodology, TR-0716-50350-P-A, Rev. 1 o Approved by NRC in June 2020 o Previously presented to ACRS February 19, 2020 subcommittee meeting March 5, 2020 full committee meeting o Used for the NuScale US600 design codified in 10 CFR 52 Appendix G

TR-0716-50350-P, Rev. 2 submitted in December 2021

NRC staff performed review and audit of TR-0716-50350-P o No RSIs o One RAI with two questions o Multiple audit questions

TR-0716-50350-P was revised during the review to address NRC staff feedback

Current revision is TR-0716-50350-P, Rev. 3 - submitted October 2023

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Overview of Previous Rod Ejection Methodology in TR-0716-50350-P-A, Rev. 1

Methodology for modeling rod ejection accident (REA)

Bounding reactivity initiated accident (RIA) from General Design Criteria (GDC) 28

REA is unique in comparison to other Chapter 15 events o Phenomena, time-scales, acceptance criteria, methods

Combination of codes and methods:

o SIMULATE-3K: Transient nuclear physics simulations o NRELAP5: Transient systems thermal-hydraulics o VIPRE-01: Transient detailed core thermal-hydraulics o Adiabatic Fuel Model: Conservative analytical model of fuel enthalpy and temperature

Unique acceptance criteria from Regulatory Guide (RG) 1.77, NUREG-0800

Justification for software, acceptance criteria, applicability, and treatment of uncertainties

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Rod Ejection Accident Methodology Changes

Changes from TR-0716-50350-P-A, Rev. 1:

o Replacement of RG 1.77 with RG 1.236 (issued in June 2020) o Change to pellet clad mechanical interaction (PCMI) fuel failure acceptance criteria from RG 1.236 o Calculation of fuel enthalpy and temperature via VIPRE-01 instead of adiabatic fuel model o Subchannel statistical analysis limit o Other minor changes to accommodate updated statistical subchannel method o Incorporate content from previous RAIs and add new detail, justification, and explanation to address NRC staff questions during review

Unchanged:

o SIMULATE-3K analysis and uncertainty treatment o Qualification (validation and applicability)

Primary driver of the revision was the new RG 1.236

REA method effectively not impacted by design changes o Increase in power was not a driver of the changes

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Summary and Conclusions

Subchannel:

o Statistical treatment of uncertainties allows for improved results while maintaining overall robust analysis approach

Rod ejection:

o Incorporate changes from RG 1.236 issuance o Simplify analysis structure to use VIPRE-01 for fuel calculations o Interface with updated subchannel method

Improvements in methods while maintaining conservative results

Results from calculations utilizing these methodologies are contained in Chapters 4 and 15 of the NuScale standard design approval application (SDAA) for the US460 design

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms ACRS Advisory Committee on Reactor Safeguards CFR Code of Federal Regulation CHF Critical Heat Flux GDC General Design Criteria NRC Nuclear Regulatory Commission PCMI Pellet Clad Mechanical Interaction RAI Request for Additional Information REA Rod Ejection Accident RG Regulatory Guide RIA Reactivity Initiated Accident RSI Request for Supplemental Information SDAA Standard Design Approval Application SER Safety Evaluation Report

ML24025A144

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