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Nuscale Power, LLC - Submittal of Presentation Material Entitled, ACRS Subcommittee Meeting (Open Session) Non-Less-of-Coolant Accident Topical Report and Extended Passive Cooling and Reactivity Control Methodology Topical Report, PM-179845
	ML25057A410
Person / Time
Site:	05200050
Issue date:	02/26/2025
From:	Shaver M; NuScale
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	LO-179859
	Download: ML25057A410 (1)
	v • d • e

LO-179859 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com February 26, 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) Non-Loss-of-Coolant Accident Topical Report and Extended Passive Cooling and Reactivity Control Methodology Topical Report, PM-179845, 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 March 4th, 2025. The materials support NuScales presentation of the subject topical reports for the US460 Standard Design Approval Application.

The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Meeting (Open Session) Non-Loss-of-Coolant Accident Topical Report and Extended Passive Cooling and Reactivity Control Methodology Topical Report, PM-179845, Revision 0.

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

If you have any questions, please contact Amanda Bode at 541-452-7971 or at abode@nuscalepower.com.

Sincerely, Mark W. Shaver Director, Regulatory Affairs NuScale Power, LLC Distribution:

Mahmoud Jardaneh, Chief New Reactor Licensing Branch, NRC Getachew Tesfaye, Senior Project Manager, NRC Michael Snodderly, Senior Staff Engineer, Advisory Committee on Reactor Safeguards, NRC Thomas Hayden, Project Manager, NRC David Drucker, Senior Project Manager, NRC

LO-179859 Page 2 of 2 02/26/2025 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) Non-Loss-of-Coolant Accident Topical Report and Extended Passive Cooling and Reactivity Control Methodology Topical Report, PM-179845, Revision 0, Nonproprietary

LO-179859 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) Non-Loss-of-Coolant Accident Topical Report and Extended Passive Cooling and Reactivity Control Methodology Topical Report, PM-179845, Revision 0, Nonproprietary

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

March 4, 2025 Non-Loss-of-Coolant Accident Topical Report and Extended Passive Cooling and Reactivity Control Methodology Topical Report

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

March 4, 2025 Non-Loss-of-Coolant Accident Topical Report Presenters: Kevin Lynn, Meghan McCloskey, Ben Bristol

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 Agenda Non-loss-of-coolant accident (non-LOCA) topical report history Non-LOCA evaluation model (EM) analysis purpose, transient class, acceptance criteria Relevant power uprate design and operating changes Summary of EM applicability assessment and updates

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Non-LOCA Topical Report History Non-LOCA topical report Revision 3 was approved by NRC in 2020 Approved Revision 3 was used in Final Safety Analysis Report (FSAR) analyses for US600 (with NPM-160) design that has been certified Approved Revision 3 contained limitations and conditions (L&Cs) restricting use to NPM-160 design Revision 4 was submitted in January 2023 along with FSAR for US460 (with NPM-20)

Updates to Revision 4 have been made since January 2023 in response to NRC questions Revision 5 will incorporate these updates, but has not been submitted at this time Focus of discussion today is changes since prior NRC approval of Revision 3

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Non-LOCA EM: Analysis Purpose, Transient Class, Acceptance Criteria Scope consistent with the NRC-approved Revision 3 Plant design, core design, fuel rod design, plant initial conditions, structures, systems, and components (SSC) performance Primary

pressure, secondary

pressure, safe stabilized condition Fuel cladding integrity Radiological dose acceptance criteria Mass & energy release input Non-LOCA topical report TR-0516-49416-P Subchannel topical reports TR-0915-17564-P-A TR-108601-P-A Accident source term topical report TR-0915-17565-P-A

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 (Tavg changed 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 Module protection system (MPS) actuations optimized for US460 design o Adjusted to accommodate modified operating conditions o Added reactor trip on high Tavg to terminate slower reactivity transients earlier (e.g., reactivity transient initiated from lower power) o Additional decay heat removal system (DHRS) actuations - for any containment vessel (CNV) isolation signal during power operation o Pressurizer line isolation on low pressurizer pressure

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Pressure/Temperature Operation and Limit Changes US600 (Certified Design)

US460 (Design currently under review)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Analytical Assumptions for Non-LOCA Analysis Approach from NRC-approved Revision 3 methodology maintained:

Scope of event progression Safety analyses of design-basis events are performed from event initiation until a safe, stabilized condition is reached Operator action No operator actions required to achieve safety functions for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after an initiating event occurs Loss of power Evaluate whether power available, loss of alternating current (AC) power, or loss of AC and direct current (DC) power is more limiting Nonsafety-related module or plant control systems Operation of nonsafety-related control system that leads to a less severe plant response is not credited Operation of nonsafety-related control system that leads to a more severe plant response is assumed Nonsafety-related SSC credited Nonsafety-related secondary main steam isolation valves (MSIVs) and feedwater (FW) regulating valves serve as backup for safety-related valve single failure Nonsafety-related check valves in FW piping serve as backup for safety-related check valve failure

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Non-LOCA EM Updates Design changes have no substantial change in non-LOCA event progressions or important phenomena o Reactor pressure vessel (RPV) pressure protected by reactor safety valve (RSV) lift o Secondary pressure protected by design pressure equal to RPV design pressure, physically limited to saturation pressure at maximum primary hot side temperature o Minimum critical heat flux ratio (MCHFR) limited under high power, high temperature conditions (e.g., reactivity insertion events)

Non-LOCA phenomena identification and ranking table (PIRT) from NPM-160 remains applicable Current EM employs NRELAP5 v1.7 (NRC-approved Revision 3 used NRELAP5 v1.4)

NRELAP5 assessment basis expanded with NIST-2 steam generator (SG)-DHRS tests Methodology changes for event-specific analyses o Provided additional detail on when more extensive sensitivity calculations performed

Dependent on margin to acceptance criteria - more sensitivity studies needed where margin is smaller o Generally expanded scope to vary parameters rather than bias in only one direction o Option for radiological analyses to use bounding input rather than transient-specific input o Option to demonstrate control rod drop analyses bounded by single rod withdrawal or steady-state conditions o Option to use increase in level during boron dilution events to determine shutdown margin at event termination

==

Conclusion:==

EM remains adequate to evaluate an NPM design

NuScale Nonproprietary Template #: 0000-21727-F01 R10 11 Questions?

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms AC Alternating current CNV Containment vessel DC Direct current DHRS Decay heat removal system EM Evaluation model FSAR Final safety analysis report FW Feedwater L&C Limitation and condition LOCA Loss-of-coolant accident MCHFR Minimum critical heat flux ratio MPS Module protection system MSIV Main steam isolation valve NIST NuScale Integral System Test Facility Non-LOCA Non-loss-of-coolant accident NPM NuScale Power Module PIRT Phenomena identification and ranking table RPV Reactor pressure vessel RSV Reactor safety valve SG Steam generator SSC Structures, systems, and components Tavg Average temperature Thot Hot temperature

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

March 4, 2025 Extended Passive Cooling and Reactivity Control Methodology Topical Report Presenters: Thomas Case, Meghan McCloskey, Ben Bristol

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Agenda Evaluation model (EM) scope, regulations, acceptance criteria NuScale Power Module (NPM) design features Phenomena identification and ranking table (PIRT) evolution EM structure EM validation basis EM adequacy assessment and conclusions

NuScale Nonproprietary Template #: 0000-21727-F01 R10 NPM Extended Passive Cooling and Reactivity Control Scope InitiatingEvent NonLOCAevent DHRSactuation Inadv.ECCSor Valveopening event PipebreakLOCA insideCNV NonLOCAevent Rxtriponly NoDHRSactuation Controlrod ejection ShorttermDHRS operation ShorttermECCS operation ShorttermECCS operation Normalshutdown ExtendedDHRS cooling EDAScapacity ECCSactuation at24hrs ECCSvalve opening ECCSvalvepassive opening InsufficientSDM ECCSactuation ext.DHRStimer Longterm ECCScooling ARIorWRSO anypoweravailability ARIorWRSO anypoweravailability ARIorWRSO AC,DCpoweravailable ACpoweravailable ARI ACpowerlost EDASlost ARIorWRSO ACpoweravailable WRSO Veryeffective DHRScooling ACpowerlost EDASavailable ARI Shortterm toLongterm ECCS Transition ExtendedDHRS toECCS Shortterm toLongterm ECCS Extended passive cooling Shortterm nonLOCA Shortterm nonLOCA Shortterm LOCA Rodejection Leakage DHRStoECCS RTNSS7days ELAP14days 72hours Postevent returntoservice designcapability LeakageECCS actutation New topical report to support 250 MWt NPM design and US460 submittal Regulations:

o 10 CFR 50.46(b)(4) and (5) o Principal design criterion (PDC) 35 -

emergency core cooling o PDC 34 - residual heat removal o General design criterion (GDC) 26, GDC 27 - reactivity control and subcriticality, normal operation or following anticipated operation occurrences (AOOs) or accidents o Supports application exemptions to GDC 33 for system with safety function to provide makeup in response to reactor coolant pressure boundary leakage

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Extended Passive Cooling (XPC) Figures of Merit Safety Objective Acceptance Criteria Provide decay and residual heat removal Collapsed liquid level remains above top of core Reactivity control Core remains subcritical Maintain coolable geometry Boron concentration remains below precipitation limits Key assumptions/requirements:

Demonstrate subcriticality (keff<1) with highest worth control rod withdrawn from core Demonstrate acceptance criteria met for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

NuScale Nonproprietary Template #: 0000-21727-F01 R10 NPM Design - Long-term ECCS Collapsed Level After emergency core cooling system (ECCS) actuation, decay and residual heat generate vapor and energy transferred to reactor pool ultimate heat sink:

o ECCS recirculation and condensation on containment wall, heat transfer through vessel wall o Steam generator (SG)-decay heat removal system (DHRS) operation with condensation on outside of SG tubes During ECCS long-term cooling, reactor coolant distributes between reactor pressure vessel (RPV) and containment vessel (CNV)

Distribution of reactor coolant depends on o ECCS venting capacity and demand o Containment heat transfer capacity reactor pool reactor recirculation valves containment vessel reactor vent valves

NuScale Nonproprietary Template #: 0000-21727-F01 R10 NPM Design - Long-term ECCS Collapsed Level Minimum Level Conditions High CNV wall heat transfer Maximum Temperature Conditions Low CNV wall heat transfer 4-24 hours: module pressure

~ 100 psia - ~ 50 psia 2-24 hours: module pressure

~ 100 psia - ~ 5 psia

NuScale Nonproprietary Template #: 0000-21727-F01 R10 NPM Design Features - Boron Transport Method Applicability ECCS actuation designed for core cooling and reactivity control Upper riser flow paths between riser and downcomer o Sustain liquid flow over the SG for decay heat removal after riser uncovery o Maintain boron transport during DHRS operation Lower riser flow paths between riser and downcomer o Maintain boron transport during ECCS operation Low RCS level Low-low RCS level Upper riser flow paths Lower riser flow paths

NuScale Nonproprietary Template #: 0000-21727-F01 R10 NPM Design Features - Boron Transport Method Applicability (continued)

ECCS supplemental boron (ESB) feature o Passive design feature to maintain subcriticality during design basis extended passive cooling o Boron oxide (B2O3) pellets in dissolver basket(s) o Mixing tube(s) in containment o Condensate collection channels to dissolver basket(s) and mixing tube(s)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 PIRT Evolution for XPC Design Certification Application (DCA) NPM-160 Design Standard Design Approval Application (SDAA) CORE250B/NPM-20 Design Phase Figure of Merit (FOM)

Phase FOM Loss-of-coolant accident (LOCA) long-term cooling (LTC) Phase 2 Period beginning after reactor recirculation valve (RRV) flow direction reverses and flows from CNV to RPV Critical heat flux ratio (CHFR);

Collapsed liquid level (CLL);

Subcriticality ECCS Phase 2 Period beginning after RRV flow direction reverses and flows from CNV to RPV CLL; Subcriticality; Coolable geometry Non-LOCA Phase 3 Stable natural circulation CHFR; Mixture level (phase 3, 4);

Subcriticality DHRS Phase 3 Stable natural circulation Non-LOCA Phase 4 Intermittent natural circulation n/a n/a Non-LOCA Phase 5 Interrupted natural circulation n/a n/a Previously developed PIRTs for NPM-160 long-term ECCS or DHRS cooling were re-assessed holistically, expanded as needed due to ESB design changes Requirement to maintain subcriticality

NuScale Nonproprietary Template #: 0000-21727-F01 R10 EM Structure NRELAP5 thermal-hydraulic analysis o Evaluate collapsed liquid level above top of fuel and containment response

Minimum level conditions

Maximum temperature conditions o Provide boundary conditions for boron transport SIMULATE5 core reactivity analysis o Provide critical boron concentrations o Evaluate range of operating cycle exposures, operating histories, thermal-hydraulic conditions Boron transport analysis o Implemented in MATLAB scripts or other appropriate computational script o Map NRELAP5 conditions to critical boron concentration from SIMULATE5 to demonstrate subcriticality o Evaluate maximum concentration to demonstrate margin to precipitation concentrations

NuScale Nonproprietary Template #: 0000-21727-F01 R10 EM Validation Basis NRELAP5 validation o Builds on validation basis for LOCA and non-LOCA EMs o Additional validation against NIST-2 LTC and LOCA tests Boron dissolution validation o Separate effects tests performed o Methods for slow or fast-biased dissolution assessed against test data SIMULATE5 o Extensive validation basis developed for other applications o Nuclear reliability factor (NRF) for XPC conditions evaluated and included in critical boron concentration Boron transport o Relies on thermal-hydraulic input o Conservative treatment of phenomena specific to boron transport

NuScale Nonproprietary Template #: 0000-21727-F01 R10 EM Adequacy Assessment and Conclusions Adequacy assessment evaluated from bottom-up and top-down perspectives o Models and correlations in NRELAP5 or phenomena treatment in boron transport considered o Top-down assessments considered NIST-2 integral tests and overall approach/conservatism in the EM Adequacy assessment discusses limitations in the models and correlations EM requires conservative or bounding approaches to address limitations in models and correlations

==

Conclusion:==

EM provides conservative method to demonstrate that an NPM, with specified design features, provides adequate core cooling and decay heat removal, remains subcritical following design basis events, and maintains coolable geometry.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 25 Questions?

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms AOO Anticipated operational occurrence CHFR Critical heat flux ratio CLL Collapsed liquid level CNV Containment vessel 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 GDC General design criterion/criteria LOCA Loss-of-coolant accident LTC Long-term cooling NIST NuScale Integral System Test Facility Non-LOCA Non-loss-of-coolant accident NPM NuScale Power Module NRF Nuclear reliability factor PDC Principal design criterion/criteria PIRT Phenomena identification and ranking table RCS Reactor coolant system RPV Reactor pressure vessel RRV Reactor recirculation valve RVV Reactor vent valve SDAA Standard design approval application SG Steam generator XPC Extended passive cooling

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