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LLC, Submittal of Presentation Material Entitled ACRS Subcommittee Meeting (Open Session) Chapter 16, Part 4, LOCA LTR, and Hiti Status, PM-177830, Revision 0
	ML25009A012
Person / Time
Site:	05200050, 99902078
Issue date:	01/09/2025
From:	Griffith T; NuScale
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	LO-177832
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LO-177832 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com January 09, 2025 Docket No. 052-050 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 Material Entitled ACRS Subcommittee Meeting (Open Session) Chapter 16, Part 4, LOCA LTR, and HITI Status, PM-177830, Revision 0 The purpose of this submittal is to provide presentation materials for use during the upcoming Advisory Committee on Reactor Safeguards (ACRS) NuScale Subcommittee Meeting on January 15, 2025. The materials support NuScales presentation of the subject chapter, topical report and status of the US460 Standard Design Approval Application.

The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Meeting (Open Session) Chapters 16, Part 4, LOCA LTR, and HITI Status, PM-177830, Revision 0.

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

If you have any questions, please contact Jim Osborn at 541-360-0693 or at josborn@nuscalepower.com.

Sincerely, Thomas Griffith Director, Regulatory Affairs NuScale Power, LLC Distribution:

Mahmoud Jardaneh, Chief New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Engineer, NRC Michael Snodderly, Senior Staff Engineer, Advisory Committee on Reactor Safeguards, NRC : ACRS Subcommittee Meeting (Open Session) Chapters 16, Part 4, LOCA LTR, and HITI Status, PM-177830, Revision 0

LO-177832 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com ACRS Subcommittee Meeting (Open Session) Chapters 16, Part 4, LOCA LTR, and HITI Status, PM-177830, Revision 0

NuScale Nonproprietary Template #: 0000-21727-F01 R10 ACRS Subcommittee Meeting (Open Session)

January 15, 2025 Chapter 16, Part 4, LOCA LTR and HITI Status

NuScale Nonproprietary Template #: 0000-21727-F01 R10 ACRS Subcommittee Meeting (Open Session)

January 15, 2025 Presenter Gene Eckholt Chapter 16 Technical Specifications

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 NuScale SDAA Part 4, US460 Generic Technical Specifications (GTS)

Subpart E of 10 CFR 52, Standard Design Approvals, does not require submittal of Technical Specifications for consideration.

In the Statements of Consideration for the 2007 rule change to 10 CFR Part 52, the commission expressed its expectation that the contents of applications for design approvals should contain essentially the same technical information that is required of design certification applications.

NuScale included Part 4, Generic Technical Specifications, in the SDAA.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Noteworthy US460 Design Changes Affecting GTS Rated thermal power increase from 160 MWt to 250 MWt Modification of ECCS design from three to two reactor vent valves Addition of ECCS supplemental boron system Addition of passive autocatalytic recombiner in containment

NuScale Nonproprietary Template #: 0000-21727-F01 R10 US460 GTS Development Started with NuScale US600 Certified Design Technical Specifications as model Addressed US460 design changes Applied 10 CFR 50.36 criteria to plant design, operations, and safety analyses Used industry STS Writers Guide format and guidance Incorporated recent industry STS changes as appropriate Technical Report TR-101310 Rev 0 describes the differences between US600 and US460 GTS at the time of SDAA submittal

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Noteworthy GTS Changes Described in TR-101310 Revision 0 The MODE definition revised to better align with the plant response behavior The reactor core critical heat flux correlations and limits, and the RCS pressure safety limits revised to reflect the increased reactor power and changes to the plant design New Surveillance Requirement to ensure isolation of Module Heatup System between modules Module Protection System requirements modified to align with design changes Remote Shutdown Station LCO removed RCS Operational Leakage LCO and definition modified to align with industry standards to the extent appropriate for the NuScale Design

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Noteworthy GTS Changes Described in TR-101310 Revision 0 (continued)

LTOP and ECCS LCOs modified to reflect reduced number of reactor vent valves UHS LCO modified to reflect design changes New LCO to ensure OPERABILITY of ECCS Supplemental Boron System New LCO to ensure containment closure during module movement between operating location and containment closure tool LCO 3.7.3 removed due to change from leak-before-break to break exclusion Chapter 5 Administrative Controls modified to reflect approved control room staffing plan

NuScale Nonproprietary Template #: 0000-21727-F01 R10 US460 GTS Review Audit Results 68 audit items resolved Most changes were editorial or clarifications Noteworthy changes included o

Core reactivity balance surveillance frequency clarified o

Module Heatup System flow paths added to Boron Dilution Control specification o

ECCS Supplemental Boron specification revised to include requirements for the geometric form of boron pellets RAI Results No RAI questions on Chapter 16 or GTS GTS change resulting from RAI associated with another FSAR Chapter o

Steam Generator Program revised to update the SG tube integrity discussion

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Noteworthy GTS Change Not Associated with SDAA Review Added LCO to ensure OPERABILITY of passive autocatalytic recombiner (PAR) o NuScale determined the PAR mitigates design-basis events, making the component safety-related and appropriate for inclusion in the GTS

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms CFR Code of Federal Regulations ECCS Emergency Core Cooling System FSAR Final Safety Analysis Report HVAC Heating, Ventilation, and Air Conditioning LCO Limiting Condition for Operation LTOP Low Temperature Overpressure Protection MWt Megawatts Thermal PAR Passive Autocatalytic Recombiner RAI Request for Additional Information RCS Reactor Coolant System SDAA Standard Design Approval Application STS Standard Technical Specifications TS Technical Specifications UHS Ultimate Heat Sink

NuScale Nonproprietary Template #: 0000-21727-F01 R10 ACRS Subcommittee Meeting (Open Session Session)

January 15, 2025 Presenter: Sarah Turmero Loss-of-Coolant Accident Topical Report

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Agenda Summary of significant changes since TR-0516-49422-P Revision 2 approval o Analysis scope addressed by topical report o Design changes from 160 MWt NPM-160 design to 250 MWt NPM-20 design o Summarize effects on PIRT o EM structure and assessment basis updates o Adequacy assessment process and conclusions Conclusions

NuScale Nonproprietary Template #: 0000-21727-F01 R10 LOCA Topical Report: Analysis Purpose and Transient Class Scope maintained from DCA Topical Report:

o Loss-of-coolant accident (LOCA) pipe break inside containment analysis method o Event classification: Postulated accident o Figures of merit: Phase 1a, 1b collapsed liquid level (CLL) over top of fuel, minimum critical heat flux ratio (MCHFR) o Key Regulations: 10 CFR 50.46, 10 CFR 50 Appendix K, GDC 35 Scope modified from DCA Topical Report:

o Inadvertent opening of a reactor valve (IORV) analysis method o Event classification: conservatively classified as Anticipated Operational Occurrence (AOO)

Realistically not expected to occur during a module lifetime o Figure of merit: MCHFR during Phase 0 initial blowdown o Key Regulations: GDC 10 Scope added from DCA Topical Report:

o Containment vessel (CNV) pressure/temperature response analysis method

Similar to method used in DCA technical report o Response to LOCA pipe break, secondary line breaks, IORV events, or inadvertent ECCS actuation o Figures of merit: Maximum CNV pressure, maximum CNV wall temperature, CNV pressure reduction over time o Key Regulations: GDC 16, PDC 38, GDC 50 Scope addressed elsewhere:

o Extended passive cooling and reactivity control (XPC) topical report addresses long-term core cooling and subcriticality EM Capability Requirements

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Power Uprate and Design Changes Summary NPM-160 to NPM-20 Power uprate from 160 MWt to 250 MWt Module SSC design essentially maintained Operating conditions o Increased primary pressure from 1850 psia to 2000 psia o Primary and secondary side design pressures increased from 2100 psia to 2200 psia o Use Tavg control instead of Thot control (RCS Tavg change from ~545°F to 540°F) o Decreased secondary side feedwater temperature at 100% power from 300°F to 250°F o Reduced minimum temperature for criticality from 420°F to 345°F Containment vessel o Design pressure increased from 1050 psia to 1200 psia o Design temperature increased from 550°F to 600°F o Upper containment material change from SA-508 to SA-336 F6NM EM Capability Requirements

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Power Uprate and Design Changes Summary NPM-160 to NPM-20, contd ECCS actuation signals modified o Low RCS level (top of riser), Tcold interlock o Low-low RCS level (mid-riser) always active (no interlocks) o Timers:

8-hour timer on all reactor trips; operators can block the actuation if subcriticality at cold conditions is confirmed and combustible gas mixture in RPV is precluded

24-hour timer after loss of AC power supply (unchanged from DCA)

ECCS valve design changes o Removed IAB from vent valves to enhance depressurization capability in DBE and BDBE o Modified IAB threshold/release pressures on recirculation valves o Added second trip valve to each ECCS valve to prevent inadvertent opening on solenoid failure o Added flow venturi to RVVs and reactor recirculation valves (RRVs) o Removed third RVV Long-term passive cooling enhancements for collapsed liquid level and subcriticality FOM o Addressed by Extended Passive Cooling and Reactivity Control methodology o Design changes reduce but maintain ample CNV cooling capacity:

Lowered reactor pool level from ~68 ft to ~53 ft

Reduced conductivity in upper CNV due to material change o Mitigation of boron redistribution during DHRS and ECCS cooling with riser hole flow paths o Supplemental ECCS Boron (ESB) to maintain subcriticality during extended passive cooling EM Capability Requirements

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Phenomena Identification and Ranking NuScale reviewed the NPM-160 LOCA PIRT for applicability to the 250 MWt NPM-20 design based on:

o Plant break spectrum comparisons o Scaling analyses o Other NPM-160 work subsequent to the PIRT development NuScale convened a PIRT panel for focused evaluation of IORV phenomena during rapid initial depressurization Overall conclusions:

o Existing phenomena identification and rankings applicable for 250 MWt NPM-20 design o Clarified phenomena importance based on work for NPM-160 and NPM-20 designs performed after development of NPM-160 LOCA PIRT Limited scope of methodology changes identified, and supported by additional NRELAP5 validation and sensitivity analyses EM Capability Requirements

NuScale Nonproprietary Template #: 0000-21727-F01 R10 EM Structure and Assessment Basis NRELAP5 v1.4 previously approved in DCA Current EM employs NRELAP5 v1.7 Upgraded the NIST-1 integral effects test facility to NIST-2 Expanded NRELAP5 assessment basis:

o NIST-2 LOCA test series o NIST-2 IORV test series Additional benchmark calculations, sensitivity cases as needed to support EM changes

NuScale Nonproprietary Template #: 0000-21727-F01 R10 EM Adequacy Assessment Bottom-up and top-down evaluations performed o Builds on LOCA EM adequacy assessment performed for DCA o Builds on the non-LOCA EM development for SG/DHRS heat transfer phenomena Compared NPM-160 and NPM-20 geometry, operating conditions, range of conditions for LOCA spectrum Evaluated scope of NRELAP5 code changes since v1.4 Top-down scaling analyses demonstrated important PI group similarity between NPM-160, NPM-20, NIST-2 No significant changes to NRELAP5 field equations or numerical solution NIST-2 LOCA and IORV tests expand the NRELAP5 assessment basis for NPM integral response

==

Conclusion:==

NRELAP and updated EM are applicable and adequate for the defined scope.

EM Adequacy Assessment

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Conclusions Updated topical report describes evaluation model for use of NRELAP5 to analyze:

o NPM-20 LOCA or valve opening events, to assess Phase 0 MCHFR, Phase 1a/1b MCHFR, collapsed liquid level, o NPM-20 LOCA, valve opening, and secondary pipe break containment pressure response Design changes from NPM-160 to NPM-20 evaluated for effect on LOCA or valve opening transient and important phenomena NRELAP5 code changes incorporated as necessary to support the NPM-20 EMs NRELAP5 validation basis expanded with NIST-2 tests Topical report provides robust methodology to analyze NPM response to LOCA and valve opening events, and for containment pressure/temperature response analysis.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Questions?

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms AC Alternating Current AOO Anticipated Operational Occurrence BDBE Beyond Design Basis Event CLL Collapsed Liquid Level CNV Containment Vessel DBE Design Basis Event DCA Design Certification Application DHRS Decay Heat Removal System ECCS Emergency Core Cooling System EM Evaluation Model ESB ECCS Supplemental Boron FOM Figure of Merit IAB Inadvertent Actuation Block IORV Inadvertent Opening of an RPV Valve GDC General Design Criteria LOCA Loss-of-Coolant Accident MCHFR Minimum Critical Heat Flux Ratio NIST NuScale Integral NPM NuScale Power Module PDC Principal Design Criteria PIRT Phenomena Identification and Ranking Table RAI Request for Additional Information SDAA Standard Design Approval Application SDA Standard Design Approval SSC Systems, Structures, and Components RCS Reactor Coolant System RPV Reactor Pressure Vessel RRV Reactor Recirculation Valve RVV Reactor Vent Valve XPC Extended Passive Cooling

NuScale Nonproprietary Template #: 0000-21727-F01 R10 ACRS Subcommittee Meeting (Open Session)

January 15, 2025 Presenter: Thomas Griffith Update - High Impact Technical Issues

NuScale Nonproprietary Template #: 0000-21727-F01 R10 High Impact Technical Issues (HITIs)

1. Design and classification of the augmented DC power system (EDAS)

2. Loss-of-Coolant (LOCA) break spectrum

3. Incorporated by reference (IBR)

4. Containment Vessel (CNV) material change

5. Lower reactor pressure vessel (RPV) material change

6. Secondary side controller design for density wave oscillation (DWO) events

7. DWO and steam generator inlet flow restrictor (IFR) design changes

8. ASME qualification of the helical coil steam generator for the onset of DWO-induced loads

9. Upper-to-lower RPV flange bolted joint shear loading that results from differential thermal expansion 10.LOCA break at CVCS/CIV connection Note: Green indicates issues that have been considered resolved by NuScale and NRC Management

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Density Wave Oscillation Safety Case

Analyses - DWO transient, which is used to assess SG structural integrity. SG structural integrity ensured for longer than NPM lifetime limit for time in DWO.

Real-Time Monitoring - Defined operational space where DWO is precluded and where time in DWO is conservatively accounted for against the NPM lifetime limit.

Physical Inspections - Examinations of SG tubes and IFRs ensure RCPB integrity is maintained. Degradation assessment will ensure that any damage to the tubes will inform future examination locations and frequencies.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 DWO Safety Case (Continued)

Analyses o DWO transient defined in SDAA FSAR Section 3.9.1 and lifetime limit specified in Table 3.9-1 o SG structural integrity is evaluated beyond the DWO lifetime limit for the NPM 60-year design life.

Real-Time Monitoring o SG approach temperature Comparison between RCS hot temperature and main steam temperature o Time is counted in DWO against FSAR Table 3.9-1 Summary of Design Transients 60-year US460 design life limit of 2840 days in DWO.

Technical Specifications 5.5.3 cyclic limits

NuScale Nonproprietary Template #: 0000-21727-F01 R10 DWO Safety Case (Continued)

Real-Time Monitoring o DWO is precluded during normal operations by maintaining an adequate SG approach temperature.

DWO is precluded in Region 2 Margin between normal operation and the Region 1/Region 2 boundary; Margin between the Region 1/Region 2 boundary and DWO onset.

Operation with DWO is avoidable for most of the NPM operating life.

SDAA FSAR Figure 5.4-16

NuScale Nonproprietary Template #: 0000-21727-F01 R10 DWO Safety Case (Continued)

Physical Examinations o SG tube examination requirements in Technical Specifications 5.5.4 100 percent SG tube examination at first refueling outage 100 percent SG tube examination over 72 EFPM (~ 6 years) after first refueling outage:

Maximum time below approach temperature boundary (2840 days or >7 years) is greater than maximum time between SG tube examinations.

Additional requirement to inspect at least 20 percent of tubes per outage for the first NPM to undergo a refueling outage Degradation assessment program will ensure that examination results factor into future examination frequency and location.

VT-3 examination of IFRs

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms ASME American Society of Mechanical Engineers CIV Containment Isolation Valve CNV Containment Vessel CVCS Chemical and Volume Control System DWO Density Wave Oscillation EDAS Augmented DC Power System EFPM Effective Full Power Months FSAR Final Safety Analysis Report HITI High Impact Technical Issue IBR Incorporate by Reference IFR Inlet Flow Restrictor LOCA Loss-of-Coolant Accident NPM NuScale Power Module RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RPV Reactor Pressure Vessel SDAA Standard Design Approval Application SG Steam Generator

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