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LLC, Submittal of Presentation Material Entitled ACRS Subcommittee Meeting (Open Session) Chapter 6, Section 17.4 and Chapter 19, PM-179462, Revision 0
	ML25043A431
Person / Time
Site:	05200050
Issue date:	02/12/2025
From:	Griffith T; NuScale
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	LO-179461, PM-179462
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LO-179461 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com February 12, 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 6, Section 17.4 and Chapter 19, PM-179462, 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 February 18th, 2025. The materials support NuScales presentation of the subject chapters and section for the US460 Standard Design Approval Application.

The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Meeting (Open Session) Chapter 6, Section 17.4 and Chapter 19, PM-179462, 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 Manager, NRC Michael Snodderly, Senior Staff Engineer, Advisory Committee on Reactor Safeguards, NRC Prosanta Chowdhury, Senior Project Manager, NRC

LO-179461 Page 2 of 2 02/12/2025 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) Chapter 6, Section 17.4 and Chapter 19, PM-179462, Revision 0, Nonproprietary

LO-179461 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) Chapter 6, Section 17.4 and Chapter 19, PM-179462, Revision 0, Nonproprietary

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

February 18, 2025 Chapter 6 Engineered Safety Features Presenter: Tyler Beck

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 Chapter 6 Overview

Section 6.1, Engineered Safety Feature Materials

Section 6.2, Containment Systems

Section 6.3, Emergency Core Cooling System

Section 6.4, Control Room Habitability

Section 6.5, Fission Product Removal and Control Systems

Section 6.6, Inservice Inspection and Testing of Class 2 and 3 Systems and Components

Note: The Chapter 6 presentation covers design of engineered safety features as discussed in FSAR Chapter 6 o The presentation does not cover specifics of accident sequences or evaluations (Ch. 15), Probabilistic Risk Assessment (Ch. 19), etc.

o The presentation focuses on differences from the US600 DCA to the US460 SDAA

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 6.1, Engineered Safety Feature Materials

Containment vessel (CNV) material changes:

o US600: CNV composed of FXM-19 (austenitic stainless steel) and SA-508 (low-alloy steel) o US460: CNV composed of FXM-19 and F6NM (martensitic stainless steel) o Addition of new Table 6.1-1, Dissimilar Metal Welds Addition of weld metals due to CNV materials changes Provisions for welding dissimilar metals

Implemented additional welding controls in response to NRC staff audits (e.g., post weld heat treatment)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Containment Vessel FSAR Figure 6.2-1: Containment System Figure: Lower Containment Vessel

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 6.2, Containment Systems

Containment system (CNTS) changes:

o Containment vessel:

Material changes (discussed in Section 6.1)

Number of CNV penetrations changed from 42 penetrations to 45 penetrations Design pressure rating changed from 1050 psia to 1200 psia Design temperature rating changed from 550°F to 600°F CVCS injection and discharge line include venturis integral to the CNV penetration

Mitigates line breaks outside the CNV o Combustible gas control:

Addition of safety-related passive autocatalytic recombiner (PAR) to maintain inert containment atmosphere

Removal of combustion loads as a result of maintaining an inert environment Removal of combustible gas monitoring and an exemption from monitoring requirements

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 6.2, Containment Systems (cont.)

CNTS changes:

o Containment isolation:

Addition of containment isolation test fixture (CITF) valve between the CNV nozzle and the containment isolation valve (CIV)

Improves ability to perform Appendix J testing

DCA design included first of a kind leak testing features integrated into the CIV assembly CIVs are welded directly to CITF, which are welded directly to the CNV nozzle safe-end CIV closure time changed from 7 to 10 seconds FSAR Figure 6.2-4: Primary System Containment Isolation Valves Dual Vale, Single Body Design

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 6.2, Containment Systems (cont.)

US600 DCA US460 SDAA Primary Events Peak Pressure (psia) 994 (IORV) 937 (DL break)

Primary Events Peak Temperature (°F) 526 (IL break) 533 (DL break)

Secondary Events Peak Pressure (psia) 449 (MSLB) 900 (MSLB)

Secondary Events Peak Temperature (°F) 433 (MSLB) 530 (MSLB)

CNTS changes:

o Containment response analysis:

Initial conditions align with US460 standard design

Similar stored energy to US600

US460 includes more design margin

Methodology included in the LOCA topical report o Removal of COL item related to containment leakage rate testing program o Addition of ITAAC verifying CNV free volume (and removal of previous COL item) 17 audit items and 4 RAIs resolved

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 6.3, Emergency Core Cooling System

ECCS changes:

o ECCS valves:

Changes related to safety analysis optimization:

ECCS includes two reactor vent valves (RVVs) from three in the DCA (change coincident with UHS pool level change)

RVVs do not include inadvertent actuation block (IAB) valve: RVVs open upon ECCS actuation

RRV IABs modified to 900 psid threshold (block) pressure and 450 psid release pressure

Addition of integral venturi to RRVs/RVVs to limit flow during high differential pressure conditions

- Decouples flow limiting function of valve internals Other operational enhancements:

Two in-series trip solenoid valves per RRV/RVV from a single trip solenoid valve per RRV/RVV in the DCA o ECCS actuation:

Removal of high CNV level and low RCS pressure ECCS actuation signals Addition of low and low-low RPV riser level actuation signal Addition of high-high RCS pressure and high-high RCS Tave ECCS actuation setpoints for BDBEs

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 6.3, Emergency Core Cooling System (cont.)

ECCS changes:

o ECCS includes an ECCS supplemental boron (ESB) feature:

Boron hoppers, condensate channels, dissolvers, mixing tubes o Addition of 8-hour ECCS actuation timer following reactor trip

14 audit items and 5 RAIs resolved FSAR Figure 6.3-2: Emergency Core Cooling System Operation

NuScale Nonproprietary Template #: 0000-21727-F01 R10 ECCS Supplemental Boron Detail from FSAR Figure 6.3-5: ECCS Emergency Supplemental Boron Feature Details

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 6.4, Control Room Habitability

CRHS changes:

o Ten minute delay added to actuation due to a loss of power to battery chargers o Toxic gas detection is within the scope of COL Item 6.4-1

Removed previous COL Item 6.4-5 that required testing and inspection requirements be specified for CRHS

Audit and RAI Results o One audit item concerning test method for test 16.02.03 (FSAR Table 14.2-16) and COL Item 6.4-1, resolved successfully

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 6.5, Fission Product Removal and Control Systems Unchanged from DCA

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 6.6, Inservice Inspection and Testing of Class 2 and 3 Components

No significant changes from DCA o Inservice Inspection of Class 2 and 3 components satisfies relevant 50.55a requirements and allows optional RG 1.147 code cases

Removed previous COL Item 6.6-1 o Inservice testing program is described in Section 3.9.6

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

February 18, 2025 Section 17.4 Reliability Assurance Program Presenter: Peter Shaw

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 17.4: Reliability Assurance Program As in the DCA, the Design Reliability Assurance Program (D-RAP) reviews and approves safety and risk classification NuScale re-evaluated the structures, systems, and components (SSC) classifications for the US460 standard plant design D-RAP expert panel insights resulted in changes to methodology for panel insights, without design changes

Steam generator tubes are safety-related, not risk-significant

Control rod drive mechanisms are safety-related, not risk-significant Audit Results o 10 items resolved in audit and resulted in updates to FSAR Section 8.2 and Figure 17.4-1 to clarify the SSC classification process and corresponding section references.

RAI Results o RAI 10199, Question 17.4-11 Resolved

Clarified the process does not assume risk significance based on safety-related classification

Resulted in clarifications to the default classification in FSAR Section 17.4.3.2 and role of backup diesel generators in Table 19.1-56 (Revision 2)

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

February 18, 2025 Chapter 19 Probabilistic Risk Assessment and Severe Accident Evaluation Presenters: Jim Schneider and Peter Shaw

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Chapter 19 Probabilistic Risk Assessment and Severe Accident Evaluation 19.1 Probabilistic Risk Assessment 19.2 Severe Accident Evaluation 19.3 Regulatory Treatment of Nonsafety Systems 19.4 Strategies and Guidance to Address Mitigation of Beyond-Design-Basis Events 19.5 Adequacy of Design Features and Functional Capabilities Identified and Described for Withstanding Aircraft Impacts Application review summary:

156 audit issues resolved in the audit, including 84 document requests 15 RAI questions resolved Note: an asterisk (*) indicates information that was added to Revision 2 of the SDAA

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 19.1: Probabilistic Risk Assessment General Overview 10 CFR 52.137(a)(25) requires a standard design applicant to develop a design-specific PRA.

When a site is chosen and a plant built, a licensee will develop and maintain a plant-specific PRA for the life of the plant (that is, each plant shall have a living PRA).

o The SDAA includes COL items that ensure the applicant has a PRA in the combined license, construction, and operational phases.

The purposes of the PRA at the design phase include:

o evaluate the overall safety of the plant design o provide insights for potential design improvements The safety goals of the Commission are a core damage frequency (CDF) of less than 1.0E-4 each reactor year, and a large release frequency (LRF) of less than 1.0E-6 each reactor year.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 US600 DCA US460 SDAA Full Power Hazard CDF (per mcyr)

LRF (per mcyr)

Internal Events 3.0E-10 2.3E-11 Internal Fires 9.7E-10 4.3E-11 Internal Floods 6.1E-11

<1E-15 External Floods 8.7E-10 7.9E-14 High Winds (Tornado) 9.9E-11

<1E-15 High Winds (Hurricane) 7.2E-10 6.4E-14 Seismic (SMA) 0.88g Low Power and Shutdown Hazard CDF (/mcyr)

LRF (/mcyr)

Internal Events 4.9E-13 2.0E-14 Module Drop 8.8E-08 N/A Multi-Module Hazard Conditional Probability of Core Damage Conditional Probability of Large Release Multi-Module 0.13 0.01 Composite CCFP < 0.1 Full Power Hazard CDF (per mcyr)

LRF (per mcyr)

Internal Events 6.0E-09 6.6E-13 Internal Fires 4.6E-09 1.3E-11 Internal Floods 1.6E-10 3.4E-14 External Floods 9.5E-09 1.4E-12*

High Winds (Tornado) 2.6E-09 1.6E-13 High Winds (Hurricane) 1.9E-08 1.3E-12 Seismic (SMA) 0.92g Low Power and Shutdown Hazard CDF (/mcyr)

LRF (/mcyr)

Internal Events 4.0E-11 3.5E-12 Module Drop 1.8E-08 N/A Multi-Module Hazard Conditional Probability of Core Damage Conditional Probability of Large Release Multi-Module 0.21 0.03 Composite CCFP < 0.1 Comparison of PRA Results (mean values) mcyr = module critical year CCFP = conditional containment failure probability SMA = seismic margin assessment

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 19.1: Probabilistic Risk Assessment Overview of PRA Results Internal events CDF increased, in part because of changes to ECCS, such as reducing the number of RVVs from three to two, the addition of an 8-hour actuation timer, and the addition of redundant trip valves on RRVs and RVVs.

o from 3.0E-10 per module critical year (mcyr) to 6.0E-09 per mcyr Internal events LRF decreased, primarily because of changes to ECCS that allow breaks outside of containment with failed containment isolation to be mitigated without the need for operator action or inventory makeup.

o from 2.3E-11 per mcyr to 6.6E-13 per mcyr

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 19.1: Probabilistic Risk Assessment Mitigation of unisolated breaks outside of containment Early ECCS actuation can limit coolant loss from the break by reducing system to atmospheric pressure.

o core stays covered and core damage is avoided without requiring addition of coolant to the module Relevant design changes:

o removal of inadvertent actuation blocks on the reactor vent valves o addition of low reactor pressure vessel riser level ECCS actuation signal o addition of venturi flow restrictors to CVCS injection and discharge lines to limit maximum break flow NuScale added an uncertainty to Table 19.1-28 addressing the low likelihood of weld failures between the CNV and the CIVs for CVCS*.

o The low likelihood of this weld failure, combined with leak identification and response requirements, minimize the impact of this event on the LRF.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 19.1: Probabilistic Risk Assessment Criteria for Risk Significance For determining component candidates for risk significance, NuScale uses both an absolute criterion and a sliding scale.

The sliding scale only applies to relative FV threshold; there is no change to the absolute conditional core damage frequency (CCDF) and conditional large release frequency (CLRF) thresholds.

At lower CDF and LRF, a higher Fussell-Vesely (FV) value is tolerated due to the low absolute risk.

The criteria are listed in FSAR Table 19.1-19, Criteria for Risk Significance:

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 19.2: Severe Accident Evaluation New COL Item 19.2-4 related to survivability*:

o An applicant that references the NuScale Power Plant US460 standard design will identify from Table 19.2-8 (Equipment Survivability List) the components and their severe accident doses for cases where the severe accident dose is greater than the environmental qualification dose.

o This COL item ensures that severe accident dose requirements are captured by the licensee in equipment specifications.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 19.3: Regulatory Treatment of Nonsafety Systems No change in methodology or results from the DCA: no SSC satisfy the criteria for Regulatory Treatment of Nonsafety Systems.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 19.4: Strategies and Guidance to Address Mitigation of Beyond-Design-Basis Events An applicant that references the NuScale Power Plant US460 standard design has the responsibility of addressing mitigation of beyond-design basis events in accordance with 10 CFR 50.155.

NuScale has presented its topical report on the NuScale Power Plant Design Capability to Mitigate Beyond-Design-Basis Events to the ACRS.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 19.5: Adequacy of Design Features and Functional Capabilities Identified and Described for Withstanding Aircraft Impacts High-level SDAA design changes reflected in the Aircraft Impact Analysis (AIA):

o The SDAA Reactor Building (RXB) reflects 6 modules (12 modules in the DCA) with updated building and site layout configuration.

o In the SDAA the RXB uses steel-plate composite (SC) walls along with reinforced concrete (RC) members.

Additional AIA differences in the SDAA:

o No other buildings are credited as intervening structures in the analysis (DCA credited the Radioactive Waste Building) o FSAR Section 19.5.1 updates how the assessment was performed, including models for concrete and steel o FSAR Section 19.5.4.1 Physical Damage updates reflect key design changes with the updated analysis for SC construction and site layout o Reactor Building equipment door design changed (with the SC construction) and details updated for the key design feature including reinforcement and connection details o Emergency core cooling system (ECCS) identified as a key design feature to ensure adequate core cooling

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 19.5: Adequacy of Design Features and Functional Capabilities Identified and Described for Withstanding Aircraft Impacts Audit Responses o 12 audit questions, 4 resolved with no changes to the SDAA, 8 transitioned to RAI RAI Results o 8 RAIs: Resulted in additional design details additions in FSAR Section 19.5 to support the RAI responses

Clarification on the basis of steel composite wall efficacy for resisting aircraft impact

Clarified details of certain structural features credited as key design features for aircraft impact analysis

Reactor building equipment door details were discussed for equivalence to SC walls

Key design features added to the SDAA consistent with NEI 07-13 guidance SDAA Revision 2 updates to include AIA key design feature updates in FSAR Section 19.5 with supporting Figure 1.2 updates, conclusions remain the same:

o Consistency with NEI 07-13 Revision 8 o Meets 10 CFR 50.150(a) with containment intact, core cooling capability, and spent fuel pool integrity

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms ACRS Advisory Committee on Reactor Safeguards AIA Aircraft Impact Analysis BDBE beyond-design-basis event CCDF conditional core damage frequency CCFP conditional containment failure probability CDF core damage frequency CFR Code of Federal Regulations CITF containment isolation test fixture CIV containment isolation valve CLRF conditional large release frequency CNTS containment system CNV containment vessel COL combined license CRHS control room habitability system CVCS chemical and volume control system DCA Design Certification Application DL discharge line D-RAP Design Reliability Assurance Program ECCS emergency core cooling system ESB ECCS supplemental boron ESF engineered safety feature FSAR Final Safety Analysis Report FV Fussell-Vesely IAB inadvertent actuation block IL injection line IORV inadvertent operation of a relief valve ISI inservice inspection IST Inservice Testing ITAAC Inspections, Tests, Analyses, and Acceptance Criteria LOCA loss-of-coolant accident LRF large release frequency mcyr module critical year MSLB main steam line break NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission PAR passive autocatalytic recombiner PRA Probabilistic Risk Assessment PZR pressurizer RAI Request for Additional Information RC reinforced concrete RCS reactor coolant system RG Regulatory Guide RRV reactor recirculation valve RVV reactor vent valve RXB Reactor Building SC steel-plate composite SDAA Standard Design Approval Application SER Safety Evaluation Report SG steam generator SMA seismic margin assessment SSC structures, systems, and components

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