Text

LLC, Submittal of Presentation Material Entitled ACRS Subcommittee Meeting (Open Session) Chapters 1, 4, and 15, PM-180495, Revision 0
	ML25084A368
Person / Time
Site:	05200050
Issue date:	03/25/2025
From:	Shaver M; NuScale
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	LO-180546
	Download: ML25084A368 (1)
	v • d • e

LO-180546 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com Docket No. 052-050 March 25, 2025 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) Chapters 1, 4, and 15, PM-180495, 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 April 1st, 2025. The materials support NuScales presentation of Chapters 1, 4 and 15 for the US460 Standard Design Approval Application.

The enclosure to this letter is the nonproprietary presentation entitled ACRS Subcommittee Meeting (Open Session) Chapters 1, 4, and 15, PM-180495, 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:

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

ACRS Subcommittee Meeting (Open Session) Chapters 1, 4, and 15, PM-180495, Revision 0, Nonproprietary

LO-180546 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 1, 4, and 15, PM-180495, Revision 0, Nonproprietary

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

April 1, 2025 Chapters 1, 4, and 15

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 ACRS Subcommittee Meeting (Open Session)

April 1, 2025 Presenter: Tyler Beck Chapter 1 - Introduction and General Information

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 1.1 through Section 1.4 Section 1.1, Introduction o Optimized to reduce redundant content from other sections o Multi-module considerations Previously included in US600 Design Certification Application (DCA) Chapter 21 Section 1.2, General Plant Description o Includes changes (e.g., figures of plant overview) reflecting the US460 standard design Section 1.3, Comparison with Other Facilities o Reflects US460 design features (e.g., thermal power output)

Section 1.4, Identification of Agents and Contractors o Unchanged from US600 DCA

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 1.5 through Section 1.8 Section 1.5, Requirements for Additional Technical Information o Verification and confirmation tests of unique design features (e.g., emergency core cooling system (ECCS) supplemental boron) o Boron dissolution testing performed at NuScale Integral System Test (NIST) facility o Additional ECCS valve functional testing performed with fully prototypic valve assemblies Section 1.6, Material Referenced o Incorporation by Reference was an issue resolved during audit o NuScale incorporates by reference most technical and topical reports Section 1.7, Drawings and Other Detailed Information o No significant changes from the US600 DCA Section 1.8, Interfaces with Standard Design o Removal of Conceptual Design Information list from the US600 DCA (e.g., potable water system)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 1.9 and Section 1.10

Section 1.9, Conformance with Regulatory Criteria o Includes comprehensive list of conformance with NUREG-0800 criteria, Design Specific Review Standard (DSRS) criteria, regulatory guides (RGs), generic communications, etc.

o Changes in conformance reflect US460 standard design o Examples of changed conformance from US600 DCA to US460 SDAA:

RG 1.7, Control of Combustible Gas Concentrations in Containment: NuScale utilizes a passive autocatalytic recombiner in the SDAA, as opposed to no specific control system in the DCA DSRS 5.3.1, Reactor Vessel Materials: Criteria pertaining material surveillance are no longer applicable because the design supports an exemption from 10 CFR 50.61 and 10 CFR 50.60 due to using austenitic stainless steel in the reactor pressure vessel (RPV) beltline

Section 1.10, Sites with Multiple Nuclear Power Plants o No significant changes from US600 DCA

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms DCA Design Certification Application DSRS Design Specific Review Standard ECCS Emergency Core Cooling System NIST NuScale Integral System Test NPM NuScale Power Module NRC Nuclear Regulatory Commission RAI Request for Additional Information RG Regulatory Guide RPV Reactor Pressure Vessel SDAA Standard Design Approval Application

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

April 1, 2025 Presenter: Sarah Turmero Chapter 4 - Reactor

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Agenda for Chapter 4: Reactor 4.1 Summary Description 4.2 Fuel System Design 4.3 Nuclear Design 4.4 Thermal and Hydraulic Design 4.5 Reactor Materials 4.6 Functional Design of Control Rod Drive System

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 4.1 Summary Description Information from DCA FSAR Section 4.1 was separated and incorporated into other sections of Chapter 4 -

reduced redundancy in Chapter 4 DCA FSAR Table 4.1-1 NuScale Reactor Design Parameters o SDAA FSAR Table 4.4-1 and Table 4.2-2 DCA FSAR Table 4.1-2 NuScale Core Design Parameters o SDAA FSAR Table 4.3-1 DCA FSAR Table 4.1-3 NuScale Reactor Control Rod Assembly Parameters o SDAA FSAR Table 4.2-3 DCA FSAR Table 4.1-4 NuScale Core Design Analytical Tools o Provided in the text of SDAA FSAR Section 4.3.3 for Nuclear Analysis No audit questions or RAIs

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 4.2 Fuel System Design Majority of the fuel design remains the same as the DCA design o Fuel rod array, rod per assembly, spacer grids, active fuel length, cladding material Control rod design remains the same as the DCA design Changes from DCA o Administrative - Incorporation of classification of SSC table, removal of redundant information o Fuel rod length increased by ~1 inch in the upper portion of the fuel pin where the plenum spring is o Core loading changed from 9,213 kgU to 9,269 kgU o TR-117605-P, NuFuel-HTP2'Fuel and Control Rod Assembly Designs, Revision 1

Faulted limits applied to the fuel rod cladding are derived from ASME BPVC,Section III, Table XIII-3110-1 (2019) o TR-108553-P-A Framatome Fuel and Structural Response Methodologies Applicability to NuScale, Revision 0, for applicability of previously approved codes and methods to the SDAA design.

21 audit questions resolved and no RAIs o 11 questions on TR-117605-P

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 4.3 Nuclear Design Changes from DCA o

Fuel pellet density change from 96 to 96.5%

o Changes related to power uprate

Linear heat rate

Peaking factors

Cycle length Addition of emergency core cooling system (ECCS) supplemental boron (ESB) 29 audit questions and 1 RAI resolved o

RAI requested a limiting condition for operation (LCO) on the heat flux hot channel factor (FQ) or justification for not having an LCO o

FQ does not require an LCO per 10 CFR 50.36(c)(2)(ii)(B), Criterion 2 because it is not a direct input or initial condition for safety analysis calculations

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Nuclear Design Parameter Comparison Parameter NPM-160 NPM-20 Core Average Linear Power (kw/ft) 2.5 3.9 Heat Flux Hot Channel Factor 1.860 2.196 Maximum Enthalpy Rise Hot Channel Factor 1.386 1.400 Fuel pellet density (% theoretical density) 96 96.5 Doppler (least negative) ($/F)

-8.4E-03

-2.1E-03 Doppler (most negative) ($/F)

-1.4E-02

-4.7E-03 Shutdown Margin Available (pcm - EOC) 2696 2436 Cycle Length (months) 24 18

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 4.4 Thermal and Hydraulic Design Changes from DCA o Treatment of CHF uncertainties implementing TR-108601-P-A, Revision 4, Statistical Subchannel Analysis Methodology, Supplement 1 to TR-0915-17564-P-A, Revision 2

New technical report - TR-169856-P, Revision 0, NuScale US460 Statistical Subchannel Critical Heat Flux Analysis Probabilistic Uncertainties o NSPN-1 critical heat flux correlation for rapid depressurization events

NSPN-1 analysis limit - 1.20 o NSP4 analysis limit - 1.43 o Flow reduction of 20 percent applied to the limiting fuel assembly in the subchannel analysis o Changes related to power uprate

Flow rate

Average temperature

System pressure 3 audit questions resolved and no RAIs

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Reactor Design Parameter Comparison Parameter NPM-160 NPM-20 Core thermal output 160 250 System pressure (psia) 1850 2000 Inlet temperature - best estimate flow (°F) 497 481 Core average temperature - best estimate flow (°F) 543 540 Core bypass flow (%)

8.5 7.5

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Analytical Design Operating Limits US600 US460

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Thermal Margin Limit Map US600 US460

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 4.5 Reactor Materials Control Rod Drive System Structural Materials o Change from DCA

Control Rod Drive Mechanism (CRDM) cooling water pressure boundary components and water connections outside of the reactor coolant pressure boundary (RCPB) designed to ASME BPVC, 2018 Edition, B31.1.

Removed applicability of Paragraph NC-2160 and Subarticle NC-3120 for materials exposed to borated water Materials selected for the SDAA comply with NB-2160 and NB-3120

Added additional alloy options such as Alloy 625, Alloy 718, and Type 440C to improve strength Reactor Internals and Core Support Structure Materials o No significant material changes from DCA to SDAA o RVI materials are austenitic stainless steel of various grade, class, or type 9 audit questions resolved and no RAIs

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Section 4.6 Functional Design of Control Rod Drive System Changes from DCA o Mechanical design changes are described in SDAA FSAR Section 3.9.4

Pressure housing is bolted instead of welded to reactor pressure vessel (RPV) head

Addition of rod hold out device Safety function of the CRDM remains the same between the DCA and SDAA o Release the control rod assemblies (CRAs) during a reactor trip o Maintain the pressure boundary for the RPV 3 audit questions resolved and no RAIs

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms AO Axial Offset ASME American Society of Mechanical Engineers BPVC Boiler and Pressure Vessel Code CHF Critical Heat Flux CRA Control Rod Assembly CRDM Control Rod Drive Mechanism DCA Design Certification Application ECCS Emergency Core Cooling System EOC End of Cycle ESB ECCS Supplemental Boron FSAR Final Safety Analysis Report GDC General Design Criterion HFP Hot Full Power HZP Hot Zero Power LCO Limiting Condition for Operation RAI Request for Additional Information RCPB Reactor Coolant Pressure Boundary RPV Reactor Pressure Vessel RVI Reactor Vessel Internals SSC Systems, Structures, and Components SDAA Standard Design Approval Application

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

April 1, 2025 Presenters: Kevin Lynn, Meghan McCloskey, Ben Bristol Chapter 15 - Transient and Accident Analyses

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Agenda for Chapter 15 Summary of review and current status Overview of analysis results o Primary and secondary pressure o Minimum critical heat flux ratio (MCHFR) o Loss-of-coolant accident (LOCA) and inadvertent opening of a reactor valve (IORV) event results o Radiological consequences Key differences from prior review o Long-term cooling without return to power o LOCA break spectrum high impact technical issues (HITIs) o Secondary side oscillation analysis Additional topic - augmented direct current (DC) power system (EDAS) considerations

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Chapter 15 Review Summary Total of 105 audit questions received by NuScale o 96 audit questions resolved during the audit o 9 audit questions sent to request for additional information (RAI) process Total of 10 RAI questions received by NuScale o 8 RAI questions resolved o 2 draft RAI questions on LOCA break spectrum HITI resolved by supplemental audit responses

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Primary and Secondary Pressure Results vs. Acceptance Criteria (Non-LOCA) 500 1000 1500 2000 2500 3000 15.1.1 15.1.2 15.1.3 15.1.5 15.1.6 15.2.1 15.2.4 15.2.6 15.2.7 15.2.8 15.2.9 15.4.1 15.4.2 15.4.3.b 15.4.8 15.5.1 15.6.2 15.6.3 Pressure (psia) vs. Chapter 15 Event RCS Pressure SG Pressure RSV Lift AOO Limit PA Limit

NuScale Nonproprietary Template #: 0000-21727-F01 R10 MCHFR Results vs. Acceptance Criteria (Non-LOCA) 1 1.5 2

2.5 3

3.5 4

15.1.1 15.1.2 15.1.3 15.1.5 15.1.6 15.2.1 15.2.4 15.2.6 15.2.7 15.2.8 15.2.9 15.4.1 15.4.2 15.4.3.a15.4.3.b 15.4.3.c 15.4.7 15.4.8 15.5.1 MCHFR vs. Chapter 15 Event MCHFR MCHFR Limit

NuScale Nonproprietary Template #: 0000-21727-F01 R10 LOCA and IORV Events Results vs. Acceptance Criteria Design-basis LOCA break spectrum is for breaks inside containment o Chemical and volume control system (CVCS) discharge and injection lines (liquid-space breaks) o Pressurizer spray and reactor pressure vessel (RPV) high point vent (HPV) lines (vapor-space breaks)

Design-basis IORV spectrum is for valve opening events o Single valve opens: reactor vent valve (RVV), reactor recirculation valve (RRV), reactor safety valve (RSV) o Two valves open: emergency core cooling system (ECCS) actuation (i.e., both RVVs open) o Multiple valves open: single valve opens (RRV or RSV) plus loss of EDAS causes RVVs to open Parameter Acceptance Criteria LOCA Results IORV Results MCHFR

> 1.20 1.35 1.41 Minimum collapsed liquid level

> 0 ft above top of core

> 8 ft above top of core Containment pressure

< 1200 psia

< 920 psia (from Chapter 6)

Containment temperature

< 600°F

< 535°F (from Chapter 6)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Dose Results vs. Acceptance Criteria Event Offsite Exclusion Area Boundary (EAB) and Low Population Zone (LPZ)

Main Control Room Results Acceptance Criteria Results Acceptance Criteria Failure of small lines carrying primary coolant outside containment Steam generator tube failure Main steam line break Iodine spike design-basis source term 0.83 (maximum of EAB and LPZ for listed events for either spiking)

< 2.5 (event with coincident spike)

< 25 (event with pre-incident spike) 0.25 (maximum of listed events)

< 5 Fuel handling accident 1.60 (EAB) 1.60 (LPZ)

< 6.3 0.55

< 5 Core damage event 2.39 (EAB) 4.95 (LPZ)

< 25 1.31

< 5

All values are in rem total effective dose equivalent (TEDE)

NuScale Nonproprietary Template #: 0000-21727-F01 R10 US600 (certified design) evaluated return to power and took exemption from general design criterion (GDC) 27 US460 prevents return to power and meets GDC 27 o ECCS supplemental boron provides additional negative reactivity to maintain subcriticality, assuming highest worth control rod stuck out Conservative analysis scope and method per the extended passive cooling (XPC) evaluation model (EM) o Analysis bounds anticipated operational occurrence (AOO), infrequent event, postulated accident (PA) initiating events o Analysis bounds wide range of off-normal power operating histories o High-biased critical boron concentration (CBC) calculation and boron transport method results in low-biased core concentration to conservatively minimize margin Results:

o Non-LOCA event analyses more limiting than LOCA due to later ECCS actuation o Minimum margin in non-LOCA cases occurs 28-40 hours after event initiation due to xenon decay; then margin increases as core boron concentration continues to increase o Lower riser holes assures fluid in the downcomer remains near the core boron concentration Long-Term Cooling without Return to Power Event Minimum Margin to CBC (ppm)

[Time]

Approximate Margin to CBC (ppm) at 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months LOCA Injection line break 134

[at 4.2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s]

> 200 Non-LOCA Reactor component cooling water (RCCW) break Slow-biased ESB 30

[at 42.2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s]

~ 50 Non-LOCA RCCW break Fast-biased ESB 28

[at 29.4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s]

~ 150

NuScale Nonproprietary Template #: 0000-21727-F01 R10 LOCA Break Spectrum HITIs Connections between ECCS valves and RPV Four valves total per NuScale power module (NPM)

In design-basis valve opening events, flow is restricted by venturi (figure on left below)

Hypothetical break at flange (figure on right below) would allow flow path without venturi Larger flow area has potential to be more limiting for MCHFR and containment (CNV) response (but non-limiting for liquid level above top of fuel)

Connections between CNV and CVCS piping Four CVCS lines total per NPM Hypothetical break would not be isolated by containment isolation valves and not all inventory would be retained within CNV Breaks in these locations have potential to be limiting for liquid level above top of fuel (but not for MCHFR and CNV response)

Based on FSAR Figure 6.3-3 Based on FSAR Figure 6.2-4

NuScale Nonproprietary Template #: 0000-21727-F01 R10 LOCA Break Spectrum HITIs (continued)

Failures at these locations are unlikely due to design of the connections, design stress and fatigue limits applied, inspections, and detection capabilities Exemption from 10 CFR 50.46 and GDC 35 requested to classify these postulated failures as beyond-design-basis events o Analyses are performed for these postulated failures with alternate acceptance criteria o Analyses are performed with alternative assumptions compared to design-basis events Results show that event-specific acceptance criteria for core cooling, CNV response, and dose are met o Met with credit only for passive, safety-related design features o Consideration of active makeup systems provides additional defense-in-depth

==

Conclusion:==

these failures are very unlikely, but US460 NPM design can passively mitigate these failures

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Stability and Oscillation Analysis Primary coolant and power stability analyzed with PIM code in same manner as US600 (certified design) using previously approved methodology o Stability to small perturbations during normal operation o Stability during operational occurrences o Analyses confirm acceptance criteria are met (decay ratio < 0.8 or reactor trip prior to loss of riser subcooling)

New scope of stability evaluation: consideration of continuous secondary side oscillations o Addresses potential control system issues - was Combined License (COL) Item 7.0-1 for US600 o Analyzed in NRELAP5 with secondary side oscillation imposed on steam pressure or feedwater flow o Spectrum of cases with varied oscillation amplitudes, oscillation periods, initial reactor power levels, and times in cycle o Variety of module protection system (MPS) signals provide protection to terminate oscillations prior to challenging specified acceptable fuel design limits (SAFDLs) o Limiting cases for SAFDLs look similar to existing Chapter 15 events

Example: oscillation induced cooldown causes control system rod withdrawal that behaves like other rod withdrawal events

==

Conclusion:==

operational events do not result in unstable behavior or are terminated by MPS prior to challenging SAFDLs

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Bottom Line Up Front - Augmented DC Power System (EDAS)

Safety: US460 exceeds Commission Safety Goals by orders of magnitude o The design includes nonsafety-related EDAS EDAS: NuScale went beyond DCA requirements and included additional OCRM requirements to address failure modes, reliability, and test and maintenance unavailability ECCS: The fundamental function of ECCS is the same for the US600 and US460 designs o ECCS actuation establishes continual, passive recirculation, requires no operator action, and requires no electrical power

Removal of RVV IABs allows earlier ECCS valve opening and improves ECCS effectiveness o Both designs include nonsafety-related electrical power to ECCS valves RCPB Integrity :

o ECCS valve actuation as it pertains RCPB integrity was raised and resolved by NRC staff during the DCA review of the Safety Classification of Passive Nuclear Power Plant Electrical Systems topical report SRM-SECY-19-0036:

o In any licensing review or other regulatory decision, the staff should apply risk-informed principles when strict, prescriptive application of deterministic criteria such as the single failure criterion is unnecessary to provide for reasonable assurance of adequate protection of public health and safety.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Loss of Power Considerations - General Background The GDCs require safety functions to be performed with onsite or offsite electric power available o GDC 17 addresses electric power systems, generally: safety functions to be performed assuming the other system is not functioning o GDCs 34, 35, 38, 41, and 44 require system-specific performance for either onsite or offsite power operation Typical operating plant implements GDC 17 in safety analyses by assuming:

o Offsite power available throughout event, or o Loss of offsite power (prompting safety-related onsite power to take over)

Coincident with event initiation

After reactor trip as consequence of the reactor trip and turbine trip with delay times crediting grid stability

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Loss of Power Approach for NuScale NuScale design goes further: performs safety functions with or without electric power o Supports exemption from GDC 17 o Intent of GDC 17 is met as described in FSAR Section 3.1: With electric power unavailable, safety-related SSC have sufficient capacity and capability to ensure (1) specified acceptable fuel design limits and design conditions of the RCPB are not exceeded as a result of AOOs and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.

o Conforming PDCs require systems perform their safety functions without electric power US460 implements design-specific principal design criteria (PDC) in safety analyses by assuming electric power is unavailable o Chapter 15 event analyses evaluate availability of alternating current (AC) power and EDAS

Loss of AC power at time of event initiation or time of reactor/turbine trip

EDAS power supply available or unavailable coincident with event intiation o Conservative, nonmechanistic assumption o Demonstrates electric power is not credited to mitigate design-basis events, and therefore AC and DC power supply systems are nonsafety-related

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Loss of Power Considerations - Maintaining Reactor Coolant Pressure Boundary (RCPB) Integrity Background & US600 History The ECCS valves are designed to open if electric power (EDSS in the US600 design) is lost o

Ensures the key safety function of ECCS is fulfilled by establishing passive core cooling o

Fundamental safety feature of the US460 design, as with the US600 design In the review of US600 DCA, the Commission determined inadvertent ECCS operation was not a loss of RCPB integrity o

Staff considered during review of TR-0815-16497-P-A, Safety Classification of Passive Nuclear Power Plant Electrical Systems o

Staff questioned whether nonsafety-related was sufficient to maintain RCPB integrity

On loss of EDSS, ECCS valves opened when IAB set pressure reached (~950 psid)

GDC 15 requires that the design conditions of the RCPB are not exceeded during normal operation or AOOs

NuScale understands GDC 15 to concern gross failure of the RCPB due to over-pressurization

ECCS valve opening does not challenge the design conditions of the RCPB o

Staff concluded that ECCS opening during AOO may not be consistent with the underlying defense-in-depth purpose of GDC 15

Resolved by limiting the expected frequency of occurrence via limitation and condition (L&C) 4.4, requiring a probabilistic determination that the expected frequency of an AOO and an actuation of the ECCS is not expected to occur in the lifetime of the module o

With L&C 4.4 satisfied, NRC concluded RCPB integrity was consistent with requirements - no exemption required for nonsafety-related EDSS o

Commissions Statements of Consideration for US600 Design Certification Rulemaking confirmed:

The NRC reviewed topical report TR-0815-16497 and concluded that NuScale Power demonstrated that the safety-related systems do not rely on Class 1E electrical power.

Because no safety-related functions of NuScale rely on electrical power, NuScale does not need any safety-related electrical power systems.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Loss of Power Considerations - Maintaining RCPB Integrity (continued)

US460 Approach The ECCS valves are designed to open if electric power (EDAS) to the ECCS valves is lost US460 licensing basis follows approach approved in TR-0815-16497-P-A o Similar augmented requirements to ensure reliability of EDAS o Applies and meets L&C 4.4 to ensure frequency of an AOO and an actuation of the ECCS is less than once in the lifetime of a module US460 design does not include IABs on ECCS RVVs o Improves overall plant safety by enhancing ECCS mitigative capabilities for some events o As a result, on loss of EDAS the ECCS would open at a higher RCS pressure than would occur for the US600 design o Not a material difference with respect to RCPB integrity:

Underlying defense-in-depth purpose of GDC 15 still met by limiting frequency

Inadvertent ECCS on loss of EDAS is an analyzed event (assumed AOO) with substantial safety margins for core cooling and containment integrity EDAS is not relied upon to ensure RCPB integrity

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Loss of Power Considerations - US460 Safety Analyses US460 design change to remove RVV IABs improves plant safety overall in context of public health and safety Removal of IABs addressed in safety analysis event sequences o Multiple pre-application engagements with NRC discussing EDAS failure treatment in Chapter 15 analysis, and whether evaluating unrelated EDAS random failure was required to demonstrate that the system was not relied upon to remain functional to assure RCPB integrity, in context of the 10 CFR 50.2 definition of safety-related NuScale submitted a new technical report, referenced in FSAR Chapter 15, describing:

o Augmented requirements on EDAS o Evaluation of how the augmented requirements protect EDAS from effects of design-basis initiating events, to demonstrate that other initiating events are not expected to cause EDAS failure during the event progression o How Chapter 15 evaluates EDAS failure to demonstrate the system is not relied upon in the design-basis safety analysis o Quantification of frequency of an AOO and actuation of the ECCS as less than once in the lifetime of a module - providing assurance that the underlying purpose of GDC 15 is met, consistent with the L&C 4.4 on the previously approved topical report o Quantification of frequency of random EDAS failure and ECCS valve opening during a separate event: ~1E-8/year o Evaluation of consequences of assuming random EDAS failure and ECCS valve opening during a separate event under worst conditions

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Loss of Power Considerations - US460 Safety Analyses (continued)

NPM safety systems are designed to actuate to their safe position when power supply is removed from components Loss of EDAS power supply actuates the safety systems:

o ECCS actuation - RVVs open (valves opening timing ~1 sec), RRVs remain closed initially due to IABs o Reactor trip (rod insertion timing ~2 sec) o Containment isolation, secondary system isolation, DHRS actuation (valve repositioning timing ~10-30 sec)

Depressurization from RVVs opening reduces coolant temperature and causes flow reduction as power decreases due to rod insertion - very short duration (i.e., less than 2 sec) reduction in MCHFR

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Loss of Power Treatment in US460 Chapter 15

AC power loss timing consistent with regulatory requirements and guidance o Event initiation - Deterministic assumption In some scenarios, initiating event may disrupt normal AC power supply (e.g., seismic event).

o Reactor/turbine trip - Consequential failure Normal AC power supply is disrupted after turbine trip because grid disruption is identified as a causal failure in the event progression (assumed consistent with traditional practice even though single NPM is small).

EDAS loss timing consistent with regulatory requirements and guidance o

Event initiation - Deterministic assumption Demonstrates not relied upon for safety functions.

o Unlike loss of offsite power, there is no failure mode where the initiating event progression would cause the EDAS power supply to fail o

Treatment is consistent with EDAS design and augmented requirements

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Loss of Power Treatment in US460 Chapter 15 (continued)

Random loss of EDAS at time of worst conditions in the event progression - not considered to be a design-basis event, but submitted technical report included bounding assessment of consequences o Regardless of initial condition, no core damage occurs o Conservative MCHFR limits met for a subset of power conditions exceeding 102%

o Significant margin to peak cladding temperature (PCT) criteria of 10 CFR 50.46 even if MCHFR limit not met Technical report was originally referenced in Chapter 15, but was later removed at NRC request NRC requested consideration of Technical Specifications (TS) for EDAS o NuScale provided justification for no need for TS

Power operation is not possible if EDAS is not functional

Loss of EDAS during power operation ensures safety functions of reactor trip, containment isolation, secondary system isolation, DHRS actuation, and ECCS actuation occur as designed

On loss of EDAS plant is placed in safe, stable condition with no need for further actions NuScale committed to control EDAS under Owner Controlled Requirements Manual (OCRM) and maintenance rule program (10 CFR 50.65) o Ensures system reliability and availability is maintained throughout plant lifetime

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Loss of Power Considerations - Risk-Informed Review In SRM-SECY-19-0036, Commission directed NRC staff that the inadvertent actuation block (IAB) feature of ECCS valves for NPM did not need to be assumed as a single active failure o US600 has IABs on RRVs and RVVs, NRC staff believed it necessary to treat IABs as an active single failure o Commission directed that treating IAB failure as a passive failure was consistent with risk-informed review principles o SRM-SECY-19-0036 went further by providing more general direction to NRC staff: In any licensing review or other regulatory decision, the staff should apply risk-informed principles when strict, prescriptive application of deterministic criteria such as the single failure criterion is unnecessary to provide for reasonable assurance of adequate protection of public health and safety.

Strict, prescriptive application of RCPB integrity criterion is unnecessary to provide for reasonable assurance of adequate protection o US600 review established that ECCS opening on loss of power is an issue of underlying purpose, not compliance o A conflicting, stricter interpretation here does not advance public health and safety

As with IAB single failure, loss of EDAS is a low frequency event with insignificant consequences

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Loss of Power Considerations - Conclusions Nonsafety-related classification of EDAS is appropriate Control of EDAS in OCRM and under maintenance rule program combined with augment requirements is appropriate to ensure reliability and availability is maintained during operation Safety analyses considering EDAS available or unavailable at event initiation is sufficient to demonstrate that EDAS is not relied upon to mitigate design-basis events, consistent with nonsafety-related classification Design-basis event progressions do not require consideration of random loss of EDAS during unrelated event at time of worst conditions Even if random loss of EDAS during unrelated event at time of worst conditions is considered, consequences are minimal (core cooling maintained)

The removal of IABs was a design change made to improve overall plant safety Commission direction in SRM-SECY-19-0036 emphasizes that strict, prescriptive application of deterministic criteria are unnecessary when risk informed principles provide for reasonable assurance of safety

NuScale Nonproprietary Template #: 0000-21727-F01 R10 EDAS Related Topic - ACRS Question on ECCS Solenoid Valves ECCS valves have two in series safety-related trip solenoid valves o Both must actuate to actuate ECCS - prevents single failure from causing inadvertent ECCS actuation o Valves fail in safe (i.e., actuated) position - ensures single failure does not prevent ECCS actuation Previous ACRS meetings identified question regarding one solenoid valve failed o For RVVs, subsequent failure of other solenoid valve would cause that RVV to open o For RRVs, IAB would prevent that RRV from opening even if other solenoid valve subsequently failed Known failure of a solenoid valve during operation would require operability determination for the supported ECCS valve under TS 3.5.1 o If supported ECCS valve is inoperable, TS 3.5.1 requires restoration of operability within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or else shut down o If supported ECCS valve is operable, TS 3.5.1 has no time-limiting restrictions, so continued operation may be possible. However, licensee remains responsible for compliance with licensing basis, including Section 15.0.0.6.3:

An analysis is conducted to quantify the frequency for which a combination of an AOO and an actuation of the ECCS is expected to occur, and the analysis concludes that ECCS actuation in response to an AOO or IE is not expected to occur in the lifetime of an NPM.

NuScale Nonproprietary Template #: 0000-21727-F01 R10 EDAS Related Topic - ACRS Question on ECCS Solenoid Valves (continued)

In Chapter 15 safety analyses, single failures are applied to mitigating systems o For events where ECCS is needed - single failure of other solenoid valve opens the RVV (i.e., safety function met) o For events where ECCS is not needed - random single failure of other solenoid valve does not need to be considered Initiating events are analyzed separately o Event consequences are analyzed (e.g., if the initiating event results in failure of some other system or component) o Random component failures are not assumed to occur during the event Evaluating a reactivity insertion event or cooldown event with random failure of a solenoid causing the ECCS valves to open combines two initiating events and is not required in the deterministic design basis event scope o For example: Operating plants are not required to evaluate a reactivity insertion event with a random failure of feedwater flow, or reactor coolant flow, that could otherwise be postulated due to failure of the nonsafety-related pump or failure of the nonsafety-related normal AC power supply.

NRC review focus on EDAS (not solenoid valve) failure due to interest in system safety classification per 10 CFR 50.2 Consequences of a random solenoid valve failure (with one already failed) causing an ECCS valve to open would be similar to previous analyses

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Chapter 15 Conclusions All review questions resolved All acceptance criteria met US460 NPM design passively mitigates Chapter 15 events with reasonable assurance of adequate protection of the public health and safety

NuScale Nonproprietary Template #: 0000-21727-F01 R10 Acronyms AC alternating current AOO anticipated operational occurrence CBC critical boron concentration CNV containment vessel COL combined license CVCS chemical and volume control system DC direct current EAB exclusion area boundary ECCS emergency core cooling system EDAS augmented DC power system (US460)

EDSS highly reliable DC power system (US600)

EM evaluation model ESB ECCS supplemental boron FSAR final safety analysis report GDC general design criterion HITI high impact technical item HPV high point vent IAB inadvertent actuation block IORV inadvertent opening of a reactor valve L&C limitation and condition LOCA loss-of-coolant accident LPZ low population zone MCHFR minimum critical heat flux ratio MCR main control room MPS module protection system NPM NuScale power module OCRM owner controlled requirements manual PA postulated accident PCT peak cladding temperature PDC principal design criteria RAI request for additional information RCCW reactor component cooling water RCPB reactor coolant pressure boundary RCS reactor coolant system RPV reactor pressure vessel RRV reactor recirculation valve RSV reactor safety valve RVV reactor vent valve SAFDL specified acceptable fuel design limit SDAA standard design approval application SE safety evaluation SG steam generator TEDE total effective dose equivalent TS technical specification XPC extended passive cooling

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