ML25077A004
| ML25077A004 | |
| Person / Time | |
|---|---|
| Site: | 05200050 |
| Issue date: | 04/01/2025 |
| From: | Joseph S NRC/NRR/DNRL/NRLB |
| To: | |
| References | |
| Download: ML25077A004 (1) | |
Text
Non-Proprietary Presentation to the Advisory Committee on Reactor Safeguards Subcommittee Staff Review of NuScales US460 Standard Design Approval Application (SDAA)
Final Safety Analysis Report (FSAR), Revision 1 April 1, 2025 (Open Session) 1 Chapters 1, 4 and 15
Non-Proprietary Presentation to the ACRS Subcommittee Staff Review of NuScale SDAA FSAR, Revision 1 Chapter 1, Introduction and General Description of the Plant April 1, 2025 (Open Session) 2
Non-Proprietary NuScale SDAA FSAR Chapter 1 Review Technical Reviewer
- Getachew Tesfaye, Lead PM, NRR/DNRL/NRLB Project Manager
- Getachew Tesfaye, Lead PM, NRR/DNRL/NRLB Contributors 3
Non-Proprietary US460 SDAA Review Overview US460 pre-application activities begun in 2019 with the submittal of a regulatory engagement plan followed by a public meeting Eight topical reports submitted during the preapplication phase SDAA staged submittal was completed in January 2023, including four new topical reports The NRC staff issued the results of its acceptance review with a request for supplemental information (RSI) on March 17, 2023 The staff began detailed safety evaluation of portions of the application not impacted by the RSI on March 20, 2023 Following the receipt of the supplemental information on July 14 and 17, 2023, a docketing letter was issued on July 31, 2023, that included a four phase, 24-month review schedule 4
Non-Proprietary Staff Review Approach for SDAA Four Phase Review for SDAA vs Six phase review for DCA Use of extended audit process via NuScales electronic reading room (eRR) for efficient review of the application Facilitated easy access to calculations and other supporting documents Minimized the number of RAI 5
Non-Proprietary NuScale SDAA FSAR Chapter 1 Review NuScale submitted Chapter 1, Introduction and General Description of the Plant Revision 0 of the SDAA FSAR on December 31, 2022, and Revision 1 on October 31, 2023 NRC regulatory audit of Chapter 1 was performed from March 2023 to August 2023, generating one audit issue that was resolved in the audit No RAI resulted from chapter 1 review Staff completed Chapter 1 review and issued an advanced safety evaluation to support todays ACRS Subcommittee meeting The draft SE provided to ACRS on March 3/4/25 was updated to include supplemental information submitted by NuScale on March 17, 2025, and is reflected in the SE submitted on 3/25/25.
Overview 6
Non-Proprietary NuScale SDAA FSAR Chapter 1 Review Notable differences between NuScale DCA FSAR and SDAA FSAR with Impact to Chapter 1 SE Elimination of Chapter 20, Mitigation of Beyond-Design-Basis Events, and Chapter 21, Multi-Module Design Considerations from SDAA SDAA does not use Topical Report TR-0815-16497-P-A, Safety Classification of Passive Nuclear Power Plant Electrical Systems Two exemption requested in the DCA were not requested for the SDAA.
7
Non-Proprietary NuScale SDAA FSAR Chapter 1 Review Notable differences between NuScale DCA FSAR and SDAA FSAR with Impact to Chapter 1 SE (Continued)
Three new exemptions requests were added in the SDAA Staff evaluation of exemption request for GDC 19 is in Chapter 6 SE. It was in Chapter 1 SE for DCA.
For the SDAA, only applicable sections of topical reports and technical reports are incorporated by reference (IBR). For the DCA all sections of topical and technical reports were IBRed.
8
Non-Proprietary NuScale SDAA FSAR Chapter 1 Review 9
Conclusions Information from topical and technical reports incorporated by reference (IBR) in Section 1.8 adequately address applicable regulatory requirements Chapter 1 SE does not include a safety finding. SDAA safety findings are in chapters 2 through 19.
Non-Proprietary Presentation to the ACRS Subcommittee Staff Review of NuScale SDAA FSAR, Revision 1 Chapter 4, Reactor April 1, 2025 (Open Session) 10
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review NuScale submitted Chapter 4, Reactor Revision 0 of the SDAA FSAR on December 31, 2022, and Revision 1 on October 31, 2023 NRC regulatory audit of Chapter 4 was performed from March 2023 to August 2024, generating 76 audit issues Questions raised during the audit were resolved within the audit. One RAI was issued, and the response was acceptable Staff completed Chapter 4 review and issued an advanced safety evaluation to support todays ACRS Subcommittee meeting One significant change between draft SE provided to ACRS on 3/4/25 and SE submitted on 3/25/25 Overview 11
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review Significant differences between previously submitted SER 12 One significant difference in Section 4.3.4 Technical Evaluation following closure of RAI question 4.3-28:
Section 4.3.4.1, Power Distributions, and Section 4.3.4.9, Technical Specifications -
revised evaluation of the TS to include assessment of why a limiting condition for operation (LCO) is not needed for the heat flux hot channel factor (FQ)
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review Technical Reviewers
- Dr Zhian Li, NRR/DSS/SNRB
- Dr Adam Rau, NRR/DSS/SNRB*
- Antonio Barrett, NRR/DSS/SNRB
- Dr Rosie Sugrue, NRR/DSS/SNRB*
- Dr Cory Parker, NRR/DNRL/NVIB
- John Honcharik, NRR/DNRL/NPHP*
- Ryan Nolan, NRR/DSS/SNRB
- Hiruy Hadgu, NRR/DSS/SNRB
- Dr Andrew Bielen, RES/DSA/FSCB
- Dr Kevin Kadooka, PNNL(Contractor)
- Nicholas Klymyshyn, PNNL (Contractor)
- Kenneth Geelhood, PNNL (Contractor)
- Presenters Contributors 13 Project Management
- Stacy Joseph, NRR/DNRL/NRLB
- Getachew Tesfaye, Lead PM, NRR/DNRL/NRLB
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review Section 4.1 - Summary Description Section 4.2 - Fuel System Design Section 4.3 - Nuclear Design Section 4.4 - Thermal-Hydraulic Design Section 4.5 - Reactor Materials Section 4.6 - Functional Design of the Control Rod Drive System Sections 14
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review Significant differences between NuScale DCA FSAR and NuScale SDAA FSAR include:
Implementation of TR-108553-P-A: Applicability of Framatome methodologies for the new NPM-20 design
- Approved in 2022 for NPM-20 operating parameters (power, pressure, flow)
- NuScale Performance Calculation
- FAST confirmatory analyses Cladding stress intensity limits Fuel Seismic Analysis with new core plate input motions
- Changed building footprint, UHS dimensions and pool level, construction materials, hydrodynamic loads Section 4.2 Fuel System Design 15
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review Significant differences between NuScale DCA FSAR and NuScale SDAA FSAR include:
New equilibrium core design for higher power level
- Increased power, power density, linear power generation rate Fuel does not include axial blankets (i.e., reduced U-235 enrichment or natural uranium)
Added emergency supplemental boron (ESB) system Section 4.3 Nuclear Design 16
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review Staff reviewed & audited updated calculations for:
Normalized power distributions:* assembly, pin-wise, axial Control rod worth and lifetime limit
- Integral control rod worth*
- Differential control rod worth
- Loss of control rod worth is limited through exposure limits Shutdown margin
- Short term* (min 2436 pcm, most reactive rod stuck out)
- Long term (Extended Passive Cooling (XPC) methodology) - discussed in 15.0.5 Doppler*, moderator temperature, and power defect coefficients Updated RPV fluence calculation
- Indicates the staff performed confirmatory analyses with POLARIS/PARCS Section 4.3 Nuclear Design 17
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review US460 Generic Technical Specifications (GTS) include two power distribution LCOs:
Enthalpy rise hot channel factor (FH)
Axial Offset (AO)
Staff issued RAI 10269, Question 4.3-28 on the need for an LCO restricting peak linear heat generation rate (e.g., FQ(z), LHR)
Staff findings:
Local peaking may exceed that considered in the AO window analysis Higher peak LHGR may reduce MCHFR Staff is not requiring a US460 FQ LCO because:
NPM-20 LHGR remains lower than operating PWRs Safety analysis shows that fuel thermal limits would not likely be challenged Section 4.3 Nuclear Design 18
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review Significant differences between NuScale DCA FSAR and NuScale SDAA FSAR include:
Statistical critical heat flux analysis limit (SCHFAL)
New critical heat flux correlation NSPN-1: used for rapid depressurization portions of applicable events. The correlation description and development is provided in the LOCA TR.
Section 4.4 Thermal-Hydraulic Design 19
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review Subchannel analysis Statistical CHFR analytical limit NSPN-1 CHF correlation Bypass flow calculations Core bypass flow methodology and analysis was provided during audit Effects of Crud Conservative heat transfer inputs for fuel rod conduction are used in COPERNIC to account for Crud Section 4.4 Thermal-Hydraulic Design: Review Items 20
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review Significant differences between NuScale DCA FSAR and NuScale SDAA FSAR include:
Use of bolted connection for control rod drive mechanism (CRDM) in lieu of welded connection Use of threaded inserts as part of bolted connection for the CRDM CRDM not routinely disassembled for inspection Degradation of the bolted connection (including stainless steel threaded inserts and alloy steel vessel head) could lead to shifting of the CRDM and could affect the safety function of the CRDM.
Section 4.5 Reactor Materials 21
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review Augmented VT-1 examination on threaded inserts and its seal welds whenever an ASME Class 1 component is disassembled (routinely, such as):
SG Feedwater Plenum Access Covers, the SG Main Steam Plenum Access Covers, the Pressurizer Heater Bundles and the Instrument Seal Assemblies.
Detection of defects in these areas requires sample expansion to include threaded inserts and seal welds for the CRDM connections.
Staff finds this provides adequate assurance of the integrity of the threaded inserts and seal welds based on statistically significant number of threaded inserts being inspected Section 4.5 Reactor Materials 22
Non-Proprietary NuScale SDAA FSAR Chapter 4 Review While there are some differences between the DCA and the SDAA, the staff found that the applicant provided sufficient information to support the staffs safety finding.
The staff found that all applicable regulatory requirements were adequately addressed.
Conclusion 23
Non-Proprietary Presentation to the ACRS Subcommittee Staff Review of NuScale SDAA FSAR, Revision 1 Chapter 15, Transient and Accident Analysis April 1, 2025 (Open Session) 24
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review NuScale submitted Chapter 15, Transient and Accident Analysis Revision 0 of the SDAA FSAR on December 31, 2022, and Revision 1 on October 31, 2023 NRC regulatory audit of Chapter 15 was performed from March 2023 to August 2024, generating 105 audit issues Questions raised during the audit were resolved within the audit.
Eight RAIs were issued, and the responses were acceptable Staff completed Chapter 15 review and issued an advanced safety evaluation to support todays ACRS Subcommittee meeting Two significant changes between draft SE provided to ACRS on 3/4/25 and SE submitted on 3/25/25 Overview 25
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Two significant differences Section 15.0.5, Extended Passive Cooling for Decay and Residual Heat Removal, revised to include evaluation of XPC TR RAIs Section 15.6.5.3, Beyond Design Basis Event Breaks, revised to reflect closure and evaluation of LOCA break spectrum open item.
Significant differences between previously submitted SER 26
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Technical Reviewers
- Antonio Barrett, NRR/DSS/SNRB*
- John Parillo, NRR/DRA/ARCB
- Ed Stutzcage, NRR/DRA/ARCB
- Adam Rau, NRR/DSS/SNRB*
- Zhian Li, NRR/DSS/SNRB*
- Joshua Miller, NRR/DSS/SNRB
- Rosie Sugrue, NRR/DSS/SNRB*
- Ryan Nolan, NRR/DSS/SNRB*
- Shanlai Lu, NRR/DSS/SNRB
- Carl Thurston, NRR/DSS/SNRB
- Sean Piela, NRR/DSS/SNRB*
- Dong Zheng, NRR/DSS/SNRB*
- Hiruy Hadgu, NRR/DSS/SNRB
- Peter Lien, RES/DSA/CRAB II
- Presenters Contributors 27 Project Management Stacy Joseph, NRR/DNRL/NRLB Getachew Tesfaye, Lead PM, NRR/DNRL/NRLB
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Section 15.0 - Introduction - Transient and Accident Analysis Section 15.1 - Increase in Heat Removal by the Secondary System Section 15.2 - Decrease in Heat Removal by the Secondary System Section 15.3 - Decrease in Reactor Coolant System Flow Rate Section 15.4 - Reactivity and Power Distribution Anomalies Section 15.5 - Increase in Reactor Coolant Inventory Section 15.6 - Decrease in Reactor Coolant Inventory Section 15.7 - Radioactive Release from Subsystem or Component Section 15.8 - Anticipated Transients without a Scram Section 15.9 - Stability Section 15.10 - Core Damage Event Sections 28
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Design and methodology changes that impact Chapter 15 include:
Power uprate and NRELAP version/numerous basemodel changes Emergency core cooling system (ECCS) valve design and number
- Removal of inadvertent actuation block (IAB) valves on RVVs
- Addition of flow restricting venturis ECCS actuation on riser level vs CNV level, new riser level instrumentation Credit for decay heat removal system (DHRS) for LOCA and LOCA-like (IORV) events No return to power during extended passive cooling Addition of ECCS supplemental boron feature and additional riser flow holes Change to dc power availability assumptions and reliance on augmented dc power system (EDAS)
Significant Changes Between DC and SDA 29
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Extended Passive Cooling Analyses 15.0.5 Analysis of Key Chapter 15 Events & Key Issues 15.4.8 - Rod Ejection
- Operator Actions Cooldown & Reactivity Events (15.4.3 - CRA Misoperation & 15.1.3 -
Increase in Steam Flow)
- EDAS HITI 15.2.8 - Feedwater Line Break 15.6.3 - Steam Generator Tube Rupture 15.6.6 - Inadvertent Operation of a Reactor Valve 15.6.5 - LOCA
- Thermal Dispersion Sensor
- LOCA Break Spectrum HITI Focus Areas for Review 30
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Limiting Minimum Level Event - Steam Generator Tube Failure Staff performed independent confirmatory analysis Xc value for RVV compressible flow expansion factor is part of the ASME QME-1 qualification program Collapsed Liquid level above TAF - 1.8 ft Boron Transport Precipitation Analysis - Inadvertent RVV Opening Conservative assumptions for thermal hydraulic conditions Staff confirmatory/sensitivity studies show fair amount of mixing Assumed initial RCS boron concentration at maximum operational limit Margin to precipitation limit - 6250 ppm Core peak concentration - 8490 ppm Section 15.0.5-Extended Passive Cooling Analyses 31
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Boron Transport Subcriticality Analysis - RCCW Line Break Staff sensitivity calculations performed for NRELAP and MATLAB script Nuclear Reliability Factor implementation review Minimal non-condensable gas in the CNV Mixing delay due to liquid density differences accounted for Margin to critical boron concentration - 28 ppm Section 15.0.5-Extended Passive Cooling Analyses 32
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Initial startup test (first module) for CNV boron dissolution and transport (RAI-10350 R1, 6.3-7) FSAR Table 14.2-40, Test #40 Emergency Core Cooling System Section 15.0.5-Extended Passive Cooling Analyses 33
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Consideration of operating history of reduced-power impacts on short term xenon changes and potential for low decay heat Technical specification LCO 3.5.4 - The ESB shall be OPERABLE LCO 3.5.4 Condition A - ESB operational limits specified in the COLR not met SR 3.5.4.2 - Verify RCS boron concentration is within the ESB operational limits specified in the COLR Technical Specification Bases 3.5.4 - Initial RCS boron concentrations greater than the ESB operational boron limit specified in the COLR, combined with other limitations associated with the boron limit, ensure core boron concentration remains above the critical boron concentration for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after event initiation.
Section 15.0.5-Extended Passive Cooling Analyses 34
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Section 15.0.5-Implementation of XPC TR in Chapter 15 35 SDAA Figure 15.0 RCS minimum boron concentration limit requirements considering integral downpower (example COLR Limit)
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Key Assumptions Most reactive rod CRA stuck out MPS actuation, pressurizer spray on CRA ejection event with five initial power levels (0, 20, 50, 75, and 100%)
Delay in core trip, most positive MTC 15.4.8 - Limiting Rod Ejection Analysis Results MCHFR = 3.13 (Limit 1.43)
Peak RCS pressure = 2231 psia (Limit 2640 psia)
Peak radial enthalpy = 65 cal/g (Limit = 100 cal/g, RG 1.236)
PCMI failure threshold limit = 21 cal/g (Limit = 33 cal/g, RG 1.236)
Peak fuel temperature 2417 °F (Limit = 4791 °F)
Section 15.4.8 -Rod Ejection Analysis 36
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Implementation of Rod Ejection Methodology TR-0716-50350-P, Rev. 3 New peak radial enthalpy & PCMI failure thresholds per RG 1.236 All Limitations and Conditions are met
- Demonstrate the applicability of the rod ejection methodology to the specific NPM design. NPM-20 was used in TR development
- The rod ejection methodology is limited to evaluation of rod ejection accidents for fuel that has not experienced significant depletion with control rods inserted, such as from non-baseload operation. SDAA only addresses baseload operation
- Rod ejection methodology must use TR-0616-48793-P-A, Revision 1, Nuclear Analysis Codes and Methods Qualification, and TR-108601-P-A, Revision 3, Statistical Subchannel Analysis Methodology. These codes and methods are used in NPM-20 analyses Section 15.4.8 -Rod Ejection (Cont.)
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Non-Proprietary Evaluates remaining shutdown margin before automatic isolation of dilution source Considers Modes 1 through 5, HZP to HFP During the review, earlier calculations credited operator action to secure the dilution source for Modes 1 and 5 Staff issued questions to NuScale on crediting of operator actions for boron dilution and other events NuScale revised necessary calculations to ensure operator actions were not credited NuScale SDAA FSAR Chapter 15 Review Section 15.4.6 - Boron Dilution 38
Non-Proprietary Mode 1 analysis response dependent on time-in-cycle BOC: faster response, higher initial boron concentration, smaller MTC Mode 1 EOC uses alternate method:
Isolation based on high pressurizer level Automatic letdown prohibited when DWS unisolated Assumes high initial boron concentration (bounds later times-in-cycle)
Results:
47 pcm SDM remaining at DWS isolation No operator action required to terminate the dilution NuScale SDAA FSAR Chapter 15 Review Section 15.4.6 - Boron Dilution (Cont.)
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Non-Proprietary NuScale SDAA FSAR Chapter 15 Review 15.4.3: CRA misalignment, single CRA withdrawal, CRA drop (bank and single)
Staff audited NuScales detailed calculations and confirmed the non-LOCA EM TR was followed Limiting cases:
MCHFR: 1.71 (Limit 1.43) - Static CRA Misalignment
- 102% RTP
- One regulating CRA inserted to the 20% PDIL + 6 steps of rod position uncertainty, other CRAs fully withdrawn LHGR: 14.0 kW/ft (Limit 15.0 kW/ft) - Single CRA Withdrawal
- Initial power level: 45% RTP
- Reactivity insertion rate: 0.0101 $/s
- Reactor trip, SSI, and DHRS actuation on high PZR pressure Cooldown & Reactivity Events 40
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review 15.1.3: Instantaneous opening of TBV Staff audited the applicants detailed calculations and confirmed they followed the Non-LOCA EM TR Analysis Results MCHFR = 1.55 (Limit 1.43)
No trip in limiting case Key Assumptions:
EDAS is relied on to remain functional during cooldown &
reactivity events Cooldown & Reactivity Events (Cont.)
41 EDAS loss during the event would cause blowdown from higher power, pressure, and temperature
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Due to removal of the IAB valves from the RVVs, EDAS is now directly supporting the ECCS valve function to remain closed when a valid actuation signal is not present This raised concerns regarding the design and safety classification of the system resulting in the identification of a High Impact Technical Issue Based on its review of the FSAR and audited documentation the staff determined EDAS is relied on in the safety analysis to perform, at a minimum, the following safety functions:
Relied on to assure the integrity of the reactor coolant pressure boundary during power operation Relied on to ensure the SAFDLs are not exceeded during certain AOOs EDAS has augmented quality and was evaluated in Chapter 8 of the SER A staff differing view raised during the review will be discussed in the following slides A staff-initiated exemption to safety-related requirements in Chapter 8 is a potential option under consideration to address the differing view Cooldown & Reactivity Events - EDAS HITI 42
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review On December 13, 2024, the following staff submitted a non-concurrence on the NuScale SDAA Chapter 15 safety evaluation report:
Antonio Barrett, Senior Nuclear Engineer Craig Harbuck, Senior Safety and Plant Systems Engineer John Lehning, Senior Nuclear Engineer Zhian Li, Senior Nuclear Engineer Joshua Miller, Nuclear Engineer Ryan Nolan, Senior Nuclear Engineer Rebecca Patton, Branch Chief Marie Pohida, Senior Reliability and Risk Analyst Adam Rau, Nuclear Engineer Sheila Ray, Senior Electrical Engineer Thomas Scarbrough, Senior Mechanical Engineer Staff raised concerns regarding insufficient technical or regulatory basis for the acceptability of the EDAS classification and regulatory controls EDAS HITI - Staff Differing Opinion 43
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review The specific issues raised include:
EDAS meets the definition of a safety-related structure, system, or component prescribed in 10 CFR 50.2 EDAS meets 10 CFR 50.36 criteria for establishing limiting conditions for operation in the technical specifications Management decision made early in the SDAA review on the acceptability of EDAS did not provide defensible technical or regulatory bases, and was not conducted in accordance with applicable policies, procedures, and regulations The differing view also provided acceptable risk-informed approaches to resolve the concerns, including:
Use of regulatory exemptions to applicable requirements and application of the RTNSS process Use of the risk-informed classification process provided in 10 CFR 50.69 EDAS HITI - Staff Differing Opinion (Cont.)
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Non-Proprietary NuScale SDAA FSAR Chapter 15 Review As an outcome of the differing views process, NRR management is evaluating whether an exemption is needed to treat EDAS as non-safety-related Information pertaining to the EDAS design and its reliability and availability controls would be sufficient to support the exemptions Classifying EDAS as safety-related is not necessary for adequate protection A staff-initiated exemption could be documented in SER Chapter 8 Exemption from safety-related requirements of 10 CFR 50.55a(h)
Exemption from safety-related requirements of 10 CFR 50 Appendix B, Criterion III through XVIII This approach would clarify that EDAS is exempted from safety-related classification and is therefore non-safety-related EDAS HITI - Staff Differing Opinion - Path Forward 45
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Most limiting case in group 2: Decrease in Heat Removal by the Secondary System Analysis Results MCHFR = 2.4 Maximum RCS pressure = 2,316 psia Maximum peak secondary pressure = 1,446 psia Key Assumptions:
Initial power level is assumed to be 102% of nominal to account for measurement uncertainty Conservative reactor trip characteristics: maximum time delay, holding the most reactive rod out of the core, and bounding control rod drop rate Limiting BOC reactivity feedback for limiting power response analyses AC power lost at the time of the break, immediate turbine and FW pump trip Section 15.2.8 - Feedwater System Pipe Breaks Inside and Outside Containment 46
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Key Assumptions (cont.):
FWIV is assumed to fail close on the faulted FW line SSI valves are assumed to close and DHRS valves are assumed to open at their maximum times System biases: high RCS temperature, high fuel temperature, low PZR pressure, low PZR level, minimum RCS flow Limiting cases: double ended guillotine break:
- DHRS cooling case: FW line inside containment Section 15.2.8 - Feedwater System Pipe Breaks Inside and Outside Containment (Cont.)
47
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review 15.6.3 - Analysis Results MCHFR is not limiting for SGTF (screened out)
Limiting RPV pressure scenario: 20% partial tube failure at top of SG with coincident loss of normal AC power Limiting SG pressure scenario: 100% split break tube failure at top of SG with loss of normal AC power Maximum radiological consequences confirmed to be bounded by FSAR 15.0.3 assumption Key Assumptions:
Core power at 102%; highest worth rod stuck out Assuming no single failure is conservative Tube failure at the top of the SG results in higher RCS and SG pressure Section 15.6.3 - Steam Generator Tube Failure 48
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review There are few valves in the design, and the ECCS valves are the ones that IO that cause the biggest challenge to FoMs This means the limiting IORV event is an inadvertent ECCS operation A loss of dc power to MPS causes both RVVs, which do not have IABs, to open without delay This means results will be insensitive to ECCS actuation signal timing Note that ECCS valves now have venturi internal to the valve body Section 15.6.6 - Inadvertent Operation of a Reactor Valve 49
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review IORVs are MCHFR-challenge events LOCA EM has special sub-methodology for "phase 0" MCHFR analysis
- Hot assembly inlet flow blockage
- 102% initial thermal power
- Distributed primary loop losses
- Special, new NSPN-1 CHF correlation Section 15.6.6 - Inadvertent Operation of a Reactor Valve (Cont.)
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Non-Proprietary NuScale SDAA FSAR Chapter 15 Review The worst IORV event is found to be an IO of one RRV with a loss of ac and EDAS dc power IABs on the RRVs make inadvertent opening of more than one RRV improbable Using the LOCA LTR methodology for "phase 0"
- The limiting MCHR is 1.41
- This is not the limiting Chapter 15 MCHFR (unlike for US600)
- Acceptance criterion for MCHFR is 1.2 or greater for NSPN-1 The IORV events are also not the design's limiting transients for:
Containment response RCS pressure Steam generator pressure CLL (is about 10' for this worst IORV event)
DHRS is not a factor in the limiting IORV event Section 15.6.6 - Inadvertent Operation of a Reactor Valve (Cont.)
51
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Section 15.6.5 - Loss of Coolant Accidents 52 LOCA for the NPM-20 design characterized by:
Small break sizes <2, and limited RCS pipe break locations ECCS actuation logic changes - triggered by riser level Credit DHRS during LOCA for passive cooling of the RCS (important for SBLOCAs)
LOCA scenario and limiting analysis results:
Limiting case: 100% CVCS discharge line break w/o AC/DC MCHFR > 1.35; CLL > 9.7 above TAF Staff performed confirmatory analysis using TRACE NRELAP5s LOCA FoMs are more conservative
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Section 15.6.5 - Use of Thermal Dispersion Switch for ECCS Actuation 53 LOCA TR SE L/C - ECCS RPV Riser Level Instrument Setpoint Modeling Method follows LOCA EM TR modeling setpoint based on mixture level Level Detection by heat transfer differences between liquid and vapor phase ECCS Actuation Trip Implementation Low level signal trigger: 90% void near the riser outlet (CLL 540-552")
Low-low level signal trigger: 95% void (CLL 460-472")
ECCS Timing Evaluation LOCA not sensitive to ECCS actuation timing delay Staffs Finding the level sensor responses corresponding to the specific setpoints and analytical limits results in acceptable collapsed water level above the core
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Staff determined certain locations are subject to the requirements of 10 CFR 50.46 and GDC 35 and were not considered within the design-basis LOCA break spectrum. This resulted in two High Impact Technical Issues:
- Applying the LOCA EM at these locations result in more severe consequences than IORV events HITI #10: CVCS piping systems between the CNV and CIVs
- Breaks at these locations result in the loss of coolant outside the CNV with more severe consequences than LOCAs analyzed inside containment Staff was open and supportive of a risk-informed alternative approach for the analyses of losses of coolant from these locations NuScale submitted an exemption request, with supporting analysis, to treat these locations as beyond-design-basis Section 15.6.5 - LOCA Break Spectrum Exemption 54
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review Framework used to evaluate a risk-informed exemption to 10 CFR 50.46:
The design implements a holistic safety approach that reduces LOCA risk through both prevention and mitigation
- Reduced penetrations, large volume of water above the core, slower accident progression that provides more time for operators to respond, etc.
Enhanced design and operational programs provide assurance that failures at the location of interest are highly unlikely
- Limits on stresses at the locations beyond those specified in the ASME BPV Code, leakage detection, enhanced inservice inspections, etc.
Realistic, best-estimate analyses of LOCAs at the location of interest as beyond-design-basis accidents demonstrate that the consequences are acceptable
- Analysis demonstrates the core remains cooled, consideration of uncertainties to avoid cliff edge effects Section 15.6.5 - LOCA Break Spectrum Exemption (Cont.)
55
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review NuScale Analysis and Acceptance Criteria:
Developed acceptance criteria for core cooling, containment, and radiological figures of merit Thermal-hydraulic analysis was performed using the LOCA EM with modification to represent best-estimate initial conditions.
Demonstrates the results meet the acceptance criteria Staff Review:
Audited NuScale calculations to understand modifications to the LOCA EM and verified the results Performed independent confirmatory and sensitivity analyses to confirm NuScales assumptions and inputs do not result in cliff edge effects Concludes the analysis is acceptable for a BDB event and supports the exemption to 10 CFR 50.46 and GDC 35 Section 15.6.5 - LOCA Break Spectrum Exemption-Cont.
56
Non-Proprietary NuScale SDAA FSAR Chapter 15 Review While there are some differences between the DCA and the SDAA, the staff found that the applicant provided sufficient information to support the staffs safety finding.
The staff found that all applicable regulatory requirements were adequately addressed.
Staff does not expect the decision on the EDAS exemption to change the analysis or design. As an outcome of the NCP review, the staff will modify the relevant SERs to clarify the regulatory basis and document the justification that EDAS is non-safety related.
Conclusion 57