ML25014A019
| ML25014A019 | |
| Person / Time | |
|---|---|
| Site: | 99902103 |
| Issue date: | 03/11/2025 |
| From: | Stephanie Devlin-Gill NRC/NRR/DANU/UAL1 |
| To: | Joshua Borromeo NRC/NRR/DANU/UAL1 |
| References | |
| EPID L-2023-LRM-0022 | |
| Download: ML25014A019 (1) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION On November 12, 2024, ARC Clean Technology (ARC) presented slides entitled Seismic [Soil-Structure Interaction] Analysis of ARC-100 [Small Modular Reactor] Plant: Summary of Seismic Isolation Results, Dynamic Responses, Building Design Evaluations, and Probabilistic Risk Assessment (ML24335A001). The following bullets describe the main points of discussion during ARCs presentation:
ARC stated that it intends to submit an application for the ARC-100 design using Title 10 of the Code of Federal Regulations (10 CFR) Part 53, Risk-Informed, Technology-Inclusive Regulatory Framework for Commercial Nuclear Plants, when the final rule is issued.
The United States (U.S.) Nuclear Regulatory Commission (NRC) staff asked ARC what type of feedback it is requesting during this meeting. ARC responded that it is requesting that the NRC provide high level feedback regarding whether ARC is proceeding in the correct or incorrect direction.
Isolated Building Description and Regulatory/Code Applicability:
On slide 6, the NRC staff requested clarification on the location of the seismic isolators. ARC responded by showing the green items labeled in the model as Isolator / Pedestal.
ARC stated that the modular construction would use steel-concrete composite construction for the vertical walls while the horizontal slabs (floors) would be partially steel-concrete composite construction.
ARC stated on slide 8 that the 28 identical seismic isolators intended to be used for the reactor building have been tested by Earthquake Protection Systems, Inc., in California.
Generic Site Seismic Design Parameters Seismic Modeling:
On slide 12, ARC stated that the two ground motion response spectra are bounding for most sites in the continental United States (U.S.).
On slide 12, the NRC staff asked how ARC concluded that the two ground motion response spectra were bounding for most continental U.S. operating reactor sites.
ARC stated that these were compared to many sites in the U.S. but excluded sites such as Diablo Canyon and the Rocky Mountain area.
The NRC staff stated that new ground motion models have been published since the completion of the post-Fukushima seismic hazard re-evaluation work. The NRC staff stated that ARC should be aware of these new ground motion models to ensure that the ground motion response spectra are still bounding.
Seismic Soil-Structure Interaction (SSI) Modeling and Analysis: ARC presented the slides and the NRC staff had no comments on the presented information.
Detrimental Dynamic Coupling Between [Reactor Vessel] System and Isolated Structure SSI Responses for Soil Site and Remedial Actions:
On slide 48, the NRC staff asked if there was consideration of the variable properties in the isolators (sensitivity studies). ARC stated that sensitivity studies were performed considering various effects.
The NRC staff asked if ARC has a plan for how to incorporate the sensitivity studies results into the design. ARC stated that they have a plan to incorporate the results.
The NRC staff requested further clarification on where the seismic isolation devices are placed. ARC stated that the isolators would support the reactor vessel and would be placed under the head of the reactor vessel.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION The NRC staff asked how the connection to the isolators was modeled. ARC stated that this was a rigid connection.
The NRC staff asked for clarification on slide 45, paragraph 2. ARC stated that the global response is dominant, thus the local vertical isolation does not improve the global vertical results.
The NRC staff asked if the isolators are accessible to ensure that they are performing as expected and whether they could be replaced. ARC stated that likely they will be accessible and that they may be replaceable. ARC stated that cases of maintenance and replacement of isolators using a jacking system, for example, exist elsewhere in industry.
The NRC asked what tests would be performed to ensure that the isolators are operating as expected. ARC listed some tests that could be performed; however, at this point, this will be a future determination on how exactly the isolators performance would be monitored.
On slide 40, the NRC staff asked if the isolators will be used in parallel or series.
ARC stated that the isolators will be used in parallel.
The NRC staff asked about using the isolators in series. ARC stated that the standard is to use in parallel.
Reactor Building Structural Design:
On slide 50, ARC stated that Step 2 of the seismic response analysis for seismically isolated structures would follow the methodology in Chapter 12, Seismically Isolated Structures, of the American Society of Civil Engineers standard (ASCE) 4-16, Seismic Analysis of Safety-Related Nuclear Structures, using a SAP200 model for the reactor vessel with only minor deviations from the model used in the SSI analysis (Step 1) and they will be equivalent models when final analysis is completed.
On slide 50, ARC stated that there are future plans to evaluate all soil site design basis event cases beyond the 15 currently evaluated.
On slide 52, ARC stated that the results shown would be smoothed and broadened to inform reactor vessel design.
ARC stated that the results presented on slides 58 through 60 are conservative due to minimal manipulation of finite element analysis results at this stage.
The NRC staff pointed out that ASCE 7, Minimum Design Loads and Associated Criteria for Buildings and Other Structures, is for standard (commercial) building design and was not developed for nuclear safety-related buildings which may not be applicable here and its applicability would have to be evaluated by the NRC.
The NRC staff pointed out the importance of analyzing the size of the gaps in the reactor building to accommodate displacements of the isolated system.
Seismic [Probabilistic Risk Assessment (PRA)] Results:
ARC stated on slide 91 that because the ratio for the Z axis is approximately 1, this would allow for scaling.
The NRC staff asked if the PRA is assuming that the isolators worked. ARC stated that this PRA assumes operability of the seismic isolators.
The NRC staff asked if the PRA models the seismic isolators randomness capability.
ARC stated that the 80th percentile results were used for the current PRA, and no modeling of a failed seismic isolators has been performed. ARC stated that the design process would direct select acceptance criteria of isolators.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION The NRC staff asked what guidance is being used to perform the seismic PRA. ARC stated that the Electric Power Research Institute guidance is being used.
Conclusions, Open Items, and Future Investigations:
The NRC staff commented that the NRC Office of Nuclear Regulatory Research (RES) will likely have guidance on seismic isolation at some point in the future.
The NRC staff asked about qualification of the isolators. ARC stated that the method for testing is not determined; however, ARC has recommended to the U.S.
Department of Energy (DOE) that testing of isolators be performed. The NRC staff stated that the upcoming guidance will discuss the testing of these devices. In addition, the NRC staff commented that there are testing provisions for prototype and production testing in, for example, ASCE 4-16, Seismic Analysis of Safety-Related Nuclear Structures, Section 12.7, Testing of Prototype and Production Isolators.
The NRC staff commented that the usage of seismic isolation will likely reduce the significance of seismic flooding which will be a future discussion.
The meeting on November 12, 2024, was adjourned at 4:35 PM eastern time.
On November 13, 2024, ARC presented slides entitled Safeguards & Physical Security, Fire Protection, [Heating, Ventilation, and Air Conditioning (HVAC)], and Plant Water, Alternate Shutdown Systems, Power Reactivity Decrement (PRD) [Research & Design (R&D)],
Program Plan for Future Testing, Simulation of a Postulated Assembly Blockage Accident in the ARC-100 Sodium Fast Reactor with the SAS-RES Safety Analysis Code, and Operating ARC-100 at 200 [megawatt electrical (MWe)] (ML24335A002). The following bullets describe the main points of discussion during ARCs presentation:
Safeguards and Physical Security:
The NRC staff commented that 10 CFR Part 37, Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material, is used for byproduct material, however this facility has special nuclear materials (SNM). The NRC staff commented that per NRC Regulatory Issue Summary (RIS) 2015-15, Information Regarding a Specific Exemption in the Requirements for the Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material, (ML15092A432), if the 10 CFR Part 73, Physical Protection of Plants and Materials, requirements for SNM are met, then no special security plan is required for the byproduct material. The NRC staff noted that during the week of November 17, 2024, there would be a public meeting on physical security requirements for 10 CFR Part 53.
The NRC staff commented that the DOE and NRC requirements on physical security are not always equivalent.
On slide 6, ARC stated that storing spent fuel outside is an issue that will be resolved later because of the high plutonium content of the spent fuel.
The NRC staff asked ((
))
ARC stated that ((
))
The NRC staff asked where reactor protection system (RPS) is in the hierarchy. ARC stated that RPS is ((
))
The NRC staff asked about if procurement will be in accordance with the supply chain requirements compared with retrofitting purchased equipment. ARC stated that the intent would be to meet the requirements when purchasing equipment.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION The NRC staff asked about two-way communication between the central alarm station (CAS) / secondary alarm system (SAS) and the outside world, operating plants are typically air-gapped. ARC stated that the intent is to follow a similar process to the operating plant.
The NRC staff asked what is the estimated size of the plant. ARC stated that the owner controlled area is approximately ((
)) The control room will be approximately ((
))
The NRC staff asked if the control room would be a digital or analog. ARC stated that control room would be mostly digital, and that this will result in a smaller control room than most operating plants.
The NRC staff asked to verify the symbols on slide 7 in the diagram. ARC explained the symbols being used.
The NRC staff asked if there are any novel technologies being considered in the system. ARC stated that currently it is unclear if any novel technologies will be considered. Specifically, ARC commented that fleet analytics could benefit from novel technologies.
The NRC staff stated that both RG 5.71, Cybersecurity Programs for Nuclear Facilities (ML22258A204) and Nuclear Energy Institute (NEI) 08-09, Cyber Security Plan for Nuclear Power Reactors (ML101180437) are listed, and that typically a plant follows a single guide. ARC said NEI 08-09 is the more complete document.
The NRC staff followed up by asking about the sister documents to NEI 08-09; such as NEI 10-04, Identifying Systems and Assets Subject to the Cyber Security Rule (ML12180A081) and NEI 13-10, Cyber Security Control Assessments (ML17046A658). ARC stated that this is likely to be followed.
The NRC staff asked about transmitting design information to plant operators in the future. ARC stated that the method of data transmission is not yet determined.
Fire Protection, HVAC and Water Systems:
ARC stated that water is not used in any areas that have sodium/sodium-potassium (Na/NaK).
ARC stated that fixed automatic fire suppression systems for Na fires are included in areas which they would be effective (small rooms) and catch pans and portable fire extinguishers would be used for large plant areas.
The NRC staff asked if the fire barriers will be used to separate fire areas. ARC stated that fire barriers will divide fire areas.
The NRC staff asked about secondary fires in Na areas that are not Na fires. ARC stated that water would not be used in these areas; however, this has not been considered and that ARC will provide an answer to the NRC on what systems will be used to stop these fires.
The NRC staff asked about automatic detection of Na fires. ARC responded that heat and smoke detectors will be used with a dry chemical extinguishing system, the exact chemical to be used will be provided to the NRC.
The NRC staff asked about the coping period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and what the basis of this time period was. ARC stated that the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> was based on a preliminary safety analysis that was performed.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION On slide 7, the NRC staff asked about the Category B designation. ARC stated that the Category B systems would be used for temperature control to ensure operability of safety-related systems.
On slide 7, the NRC asked about the Category D designation, particularly the containment performance goals. ARC stated that the reactor building is part of the functional containment, which would have the capability to clean up the air within that volume. The NRC staff asked about relying on these systems to meet dose requirements. ARC responded that a report was submitted on release under a severe accident without crediting HVAC cleanup.
The NRC staff asked how they would use the service water pump stations. ARC stated that this is a maintenance function throughout the plant.
Alternate Shutdown Systems:
ARC stated that during an anticipated transient without scram, the reactor inherent feedback brings the reactor to a stable state that is not subcritical, allowing for time to scram the reactor.
The NRC staff asked about ((
)) ARC clarified that the control rod systems would have ((
)) and asked for the NRC staffs feedback.
The NRC staff explained that 10 CFR Part 50, Appendix A, GDC-26, requires reactivity control systems to be both redundant (multiple independent systems) and diverse (utilizing different mechanisms or methods). The NRC staff noted that having
((
)) would not satisfy this regulatory requirement. The NRC staff further clarified that the backup system must not be vulnerable to the same failure modes as the primary system and should effectively address any shortcomings of the primary system in case it fails to perform as intended. The NRC staff discussed how this requirement could be met through various approaches, such as employing a different design, geometry, or alternative shutdown mechanisms. ARC stated they will consider using ((
)).
ARC stated that another alternate shutdown system using boron balls has been considered but is not the preferred option. The NRC staff stated that the boron ball injection system would align with the requirements and could be considered acceptable as an alternative shutdown system.
The NRC staff asked if ARC is planning to send the white paper on Alternative Shutdown System. ARC clarified that, based on the NRC's feedback during the presentation, they will withdraw the white paper and will submit it at a later time.
PRD R&D ARC presented the slides, and the NRC staff had no comments.
Program Plan for Future Testing:
ARC stated that a testing plan and characterization program is being developed to address data gaps and support fuel qualification activities.
The NRC staff commented that establishing a testing and characterization program aligns with regulatory requirements, as it supports the fuel qualification guidance outlined in NUREG-2246, Fuel Qualification for Advanced Reactors (ML22063A131) and addresses some of the data gaps discussed in NUREG-7305,
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - EXPORT CONTROLLED INFORMATION Metal Fuel Qualification: Fuel Assessment Using NRC NUREG-2246, Fuel Qualification for Advanced Reactors (ML23214A065). ARC clarified that while the program plan for future testing would not be submitted, the outcomes and work products developed from the program would be provided.
Simulation of a Postulated Assembly Blockage Accident in the ARC-100 Sodium Fast Reactor with the SAS-RES Safety Analysis Code:
ARC presented the slides, and the NRC staff had no comments on the presented material.
Operating ARC-100 at 200 MWe:
The NRC staff stated that the most significant consideration for operating at a higher power is fuel qualification as operating the reactor at higher power necessitates an increase in the fission rate. The NRC staff commented that metallic fuels are sensitive to fission rates, therefore, an increase in fission rate is likely to impact fundamental properties such as swelling, creep, and other factors critical to thermo-mechanical performance of the fuel, likely requiring filling additional data gaps Concluding Discussion:
ARC stated that currently electromagnetic pumps are classified as non-safety; however, the power bank used to provide coast down is safety-related.
ARC stated that 10 CFR Part 53 is the licensing path and will likely not submit until 2027.
The NRC staff commented on the importance of quality assurance when designing the ARC-100. In addition, the staff commented that addressing the data gap will be important to address for fuel qualification.
The NRC staff stated that Idaho National Laboratory may be developing further guidance on modeling and simulation for fuel qualification. This guidance would describe how a surveillance program could be used for fuel qualification.
The NRC staff asked how many white papers or topical reports are to be expected from ARC. ARC stated that there are no plans to submit topical reports at this time; however, there may be a couple more white papers.
The meeting on November 13, 2024, was adjourned at 3:15 PM eastern time.