ML23017A152

Preliminary Hermes CP SE Chapter 9, Auxiliary Systems for ACRS

ML23017A152
Person / Time
Site:	Hermes File:Kairos Power icon.png
Issue date:	01/10/2023
From:	Benjamin Beasley NRC/NRR/DANU/UAL1
To:	Advisory Committee on Reactor Safeguards
References
Download: ML23017A152 (1)

Text

9 AUXILIARY SYSTEMS THIS NRC STAFF DRAFT SE HAS BEEN PREPARED AND IS BEING RELEASED TO SUPPORT INTERACTIONS WITH THE ACRS. THIS DRAFT SE HAS NOT BEEN SUBJECT TO FULL NRC MANAGEMENT AND LEGAL REVIEWS AND APPROVALS, AND ITS CONTENTS SHOULD NOT BE INTERPRETED AS OFFICIAL AGENCY POSITIONS.

The auxiliary systems of the Hermes non-power test reactor consist of the reactor coolant auxiliary systems (the chemistry control system, the inert gas system, the tritium management system, the inventory management system, and the reactor thermal management system); the reactor building heating, ventilation and air conditioning system (RBHVAC); the pebble handling and storage system (PHSS); the fire protection systems; the communication system; facilities for possession and use of byproduct, source, and special nuclear material; the plant water systems (service water system, Treated Water System, Component Cooling Water System, Chilled Water System); and other auxiliary systems (remote maintenance and inspection system, Spent Fuel Cooling System, Compressed Air System, cranes and rigging, and auxiliary site services).

The principal design criteria (PDC) for the facility structures, systems, and components (SSCs) are described in Section 3.1 of the Preliminary Safety Analysis Report (PSAR). The PDC are based on the NRC approved Kairos Power Topical Report, KP-TR-003-NP-A, Principal Design Criteria for the Kairos Power Fluoride Salt-Cooled, High Temperature Reactor, Revision 1, (Agencywide Documents Access and Management System ADAMS Accession No. ML20111A118,) dated May 22, 2020. Each of the sections below identify the applicable PDC for the system being evaluated.

9.1 Reactor Coolant Auxiliary Systems The RCAS consist of five systems providing support for the functionality and performance of the Flibe coolant. Each of the five systems is evaluated in a separate section below.

9.1.1 Chemistry Control System Introduction The PSAR describes the chemistry control system (CCS) as being used to monitor coolant chemistry in the reactor vessel system and primary heat transport system (PHTS) for compliance with Flibe specifications found in KP-TR-005, Reactor Coolant for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor Topical Report, Revision 1. The CCS allows for offline analysis of Flibe chemistry and can remove and replace a sufficient amount of coolant to restore conformance with the Flibe specification via the inventory management system (IMS). The CCS does not perform any safety-related functions, nor is it credited to mitigate postulated events.

Regulatory Evaluation The applicable regulatory requirements for the evaluation of the Hermes CCS design criteria are as follows:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, Including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), 50.34(a)(4), and 50.34(a)(5) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards As provided in § 20.1002, Scope, the regulations in 10 CFR Part 20, Standards for Protection Against Radiation, apply to persons licensed by the Commission to receive, possess, use, transfer, or dispose

of byproduct, source, or special nuclear material or to operate a production or utilization facility. Kairos has applied for a construction permit and has not specifically requested approval of any design information. A construction permit does not provide a license to operate the facility. In its Hermes construction permit application, Kairos also has not applied for licenses to receive, possess, use, transfer, or dispose of byproduct, source, or special nuclear material at the facility. Therefore, the NRC staff did not evaluate whether requirements in 10 CFR Part 20 would be met for the construction of the Hermes reactor. Instead, the NRC staff assessed whether Kairos had identified the relevant requirements for an operating facility and provided descriptions of the preliminary facility design to determine whether the PSAR provides an acceptable basis for the development of the auxiliary systems and whether there is reasonable assurance that Kairos will comply with the regulations in 10 CFR Part 20 during Hermes facility operation. This is consistent with 10 CFR 50.40(a), which provides that in determining whether a construction permit may be issued, the Commission will be guided by consideration of reasonable assurance that Kairos will comply with the regulations, including the regulations in 10 CFR Part 20, and that the health and safety of the public will not be endangered.

The applicable guidance for the evaluation of the CCS is NUREG-1537, Part 2, Section 5.4, Primary Coolant Cleanup System. This guidance was developed for water chemistry and does not contain specific guidance for molten salts. However, the rationale for maintaining coolant purity (e.g., limit corrosion, minimize radioactivity release) remains the same. The NRC staffs review was based on the guidance in NUREG-1537 Part 2 but did not use portions of the guidance only applicable to water chemistry (e.g., pH ranges).

Technical Evaluation The PDC for Hermes are described in PSAR Chapter 3, Design of Structures, Systems and Components, and are based on those specified in the NRC approved Kairos Power Topical Report, KP-TR-003-NP-A, Principal Design Criteria for the Kairos Power Fluoride Salt-Cooled, High Temperature Reactor, ML20167A174. The PDC applicable to the Hermes CCS are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases PDC 70, Reactor Coolant Purity Control.

The NRC staff evaluated the preliminary design of the CCS to determine whether it can meet its design functions stated above. Section 9.1.1.3, System Evaluation, of the PSAR states that the CCS may be located near safety-related SSCs but that the safety-related SSCs will be protected from failure of the CCS. The NRC staffs evaluation of seismic damage and effects of natural phenomena are contained in Sections 3.4 and 3.5, respectively, of this safety evaluation. Therefore, consistent with the evaluations in Safety Evaluation Report (SER) Sections 3.4 and 3.5, the staff has reasonable assurance that the preliminary design of the CCS is consistent with PDC 2 because the PSAR describes potential methods which will be employed to ensure that a postulated failure of the CCS does not impact safety-related SSCs.

The PSAR states that the CCS is consistent with PDC 4 because safety-related systems in the area of the CCS are protected against failures of the CCS equipment by being physically separate, the use of barriers, or by the design of the safety-related components. The staff finds that the preliminary design is sufficient to provide reasonable assurance that the CCS will not adversely impact safety-related SSCs.

Final design details in the Final Safety Analysis Report (FSAR) will allow the staff to verify any potential impacts of failures of the CCS.

PDC 70 addresses the ability to monitor and correct coolant chemistry. Section 9.1.1.3, System Evaluation, of the PSAR states that the CCS can periodically monitor the reactor coolant chemistry.

It describes some of the impurities that can be examined and states that if specifications are only

marginally met then the IMS can remove and replace enough coolant to increase the margin of conformance to specifications. The staff has reasonable assurance that the preliminary design is consistent with PDC 70 because the CCS can monitor coolant chemistry and because removing coolant and replacing it with cleaner coolant provides a way to maintain coolant purity.

The guidance in NUREG-1537, Part 2, Section 5.4, states that coolant purity should be maintained consistent with limits established in Chapters 4 and 11 of the safety analysis report. The NRC staff will review specific coolant purity limits in the FSAR. The management of coolant purity limits is part of Operational Programs and is appropriate to review at the operating license (OL) stage. The NRC staff reviewed the description of CCS capabilities in the PSAR and finds that the description provides the staff assurance that the CCS will be able to achieve the necessary functions. Details of specific purity limits required actions to be taken if the purity limits are not met, and time limits to complete required actions can be provided in the FSAR. Additionally, as confirmed in the response to Request for Confirmation of Information (RCI) 349 ML22231B228, the FSAR will describe how the CCS measures a well-mixed representative sample of the reactor coolant. This will allow the staff to ensure that purity specifications can be met throughout the primary system. The provisions for representative sampling support proposed technical specifications to measure various aspects of coolant purity and composition. The NRC staff will evaluate the specific methods to monitor and maintain coolant specifications during the FSAR review.

The PSAR states that the CCS is designed to meet 10 CFR 20.1406. As a sampling system, the CCS is not credited to correct coolant chemistry and therefore should not experience a build-up of radionuclides within the system relative to the bulk reactor coolant. Additionally, because the CCS is not credited to remove radionuclides from the coolant, the generation of radioactive waste should be minimized. The NRC staff finds the CCS will be consistent with the requirements of 10 CFR 20.1406 because it is designed, in conjunction with the IMS, to contain radionuclides from the reactor coolant.

The NRC staff will review the final details of how the CCS minimizes contamination in the FSAR.

Chapter 14, Technical Specifications, of the PSAR describes proposed technical specifications to be contained in the Hermes OL. Some of the proposed technical specifications relate to reactor coolant purity requirements (e.g., LiF to BeF2 ratio, circulating activity). The NRC staff has reasonable assurance that the preliminary design of the CCS will be able to support the proposed technical specifications related to the reactor coolant chemistry because the CCS has the ability to analyze and correct coolant chemistry (via the IMS). The NRC staff will evaluate the required samples and frequencies needed to meet the technical specification as part of its review of the OL application Conclusion Based on the discussion above, the NRC staff concludes that the information in Hermes PSAR Section 9.1.1 is sufficient and meets applicable guidance and regulatory requirements for issuance of a construction permit identified in the Regulatory Evaluation. The information provided gives the NRC staff reasonable assurance that the CCS will be built to maintain and monitor required coolant purity, will not lead to radiation exposure or releases that exceed limits in 10 CFR Part 20, and that it can support proposed technical specifications related to coolant chemistry. The information also gives the NRC staff reasonable assurance that the CCS design will conform with guidance in NUREG-1537 to maintain coolant purity in order to limit degradation of essential components in the primary system. Further information will be required before approving operation of the Hermes CCS (e.g., basis of sampling location to provide a well-mixed and representative chemistry sample, ability of CCS to correct coolant chemistry within specified timeframes) and will be reviewed at the OL stage.

9.1.2 Inert Gas System Introduction Section 9.1.2 of the PSAR describes the inert gas system (IGS). The design functions of the IGS are to:

Maintain an inert environment for components using argon Provide inert gas purge flow Remove impurities from cover gas Provide transport for tritium for treatment Provide reactor coolant motive force during filling and draining The PSAR also states that the IGS does not perform any safety-related functions.

Regulatory Evaluation The applicable regulatory requirements for the evaluation of the Hermes IGS design criteria are as follows:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), 50.34(a)(4), and 50.34(a)(5). 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards As described in subsection 9.1.1 above, the staff did not evaluate whether requirements in 10 CFR Part 20 would be met for the construction of the Hermes reactor. Instead, the staff assessed whether Kairos identified the relevant requirements for an operating facility and provided descriptions of the preliminary facility design. The NRC staff assessed this to determine whether the PSAR provides an acceptable basis for the development of systems and whether there is reasonable assurance that Kairos will comply with the regulations in 10 CFR Part 20 during Hermes facility operation.

Section 9.6, Cover Gas Control in Closed Primary Coolant Systems, of NUREG-1537, Part 2, provides the review scope and acceptance criteria for evaluating cover gas systems. Although the staff used the NUREG-1537 guidance in its review, this guidance was developed for reactors that use water as a coolant and does not contain specific guidance for cover gas systems for reactors with a molten salt coolant. However, the rationale for cover gas control (e.g., providing required pressure differential, assessing purity, processing and storing gases) remains the same. The staff used concepts from the guidance and did not use portions of the guidance only applicable to water chemistry (e.g.,

recombination of hydrogen and oxygen from radiolysis of the coolant).

Technical Evaluation The PDC applicable to the Hermes IGS are as follows:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases PDC 64, Monitoring Radioactivity Releases.

The NRC staff evaluated the preliminary design of the IGS to determine whether it can meet its design functions stated above. The NRC staff has reasonable assurance that the design bases and functional descriptions of the IGS provided in the PSAR will allow the NRC staff to verify that the primary coolant system remains a closed system, helps to prevent the uncontrolled release of radionuclides, maintains cover gas purity, allows for tritium to move downstream for treatment, and can provide the reactor coolant motive force during filling and draining operations. The preliminary design information in the PSAR provides reasonable assurance that the IGS will be able to ensure the required type of gas, concentrations of constituents, and design basis pressure. Additionally, the staff has reasonable assurance that the IGS will be able to provide an appropriate purging flow to system components because gas flow rates, temperature, and pressures will be individually regulated to meet design requirements for each component. The NRC staff also has reasonable assurance the IGS will be able to

provide Flibe vapor/aerosol control because it will allow for the forced flow of gas to different components and systems to prevent Flibe deposits. The IGS only provides a pathway for tritium to be

transported to the tritium management system and the IGS does not perform any tritium management functions. Tritium capture is discussed and evaluated in Section 9.1.3, Tritium Management System, of this SER.

The NRC staff reviewed the preliminary design of the IGS to ensure the purity of the gas can be assessed, and the gas can be processed, decontaminated and stored. The NRC staff has reasonable assurance that the preliminary design of the IGS will allow for management of the inert gas because Figure 9.1.2-1, Process Flow Diagram for the Inert Gas System, shows the appropriate subsystems needed to achieve these functions. Additionally, in the response to RCI 349, Kairos confirmed that the IGS will be capable of measuring both air and moisture in the cover gas. Therefore, based on the preliminary design of the IGS, the NRC staff has reasonable assurance that the IGS can ensure purity of the argon cover gas consistent with operational limits and proposed technical specifications. This is consistent with the guidance in NUREG-1537 which states the staff should review systems for assessing the required purity of contained gases as well as systems for processing and storing the contained gases.

Section 9.1.2.3 of the PSAR states that although the IGS is not safety-related, it may be in proximity or connect to safety-related SSCs. These safety-related SSCs will be protected from seismic failures of the IGS such that postulated failures of the IGS do not prevent safety-related SSCs from performing safety functions. The NRC staffs evaluation of seismic damage and effects of natural phenomena are contained in Sections 3.4 and 3.5, respectively, of this SER. Therefore, consistent with the evaluations in Sections 3.4 and 3.5 of this SER, the NRC staff has reasonable assurance that the IGS will be designed to meet PDC 2 because the PSAR describes potential methods which will be employed to ensure that a postulated failure of the IGS does not impact safety-related SSCs.

The PSAR states that the IGS is consistent with PDC 4 because pipe whip is precluded by design of the IGS and because escape of inert argon gas will not affect safety-related SSCs. The staff finds that this preliminary information is sufficient to provide reasonable assurance that the IGS will be designed consistent with PDC 4. The NRC staff will review the final design details in the FSAR to assess any potential impacts of pipe whip on safety-related SSCs. The NRC staff agrees that argon is an inert gas and its release will not cause damage to other SSCs.

The PSAR states that the IGS includes sampling systems that can evaluate gas radioactivity levels.

Additionally, Figure 9.1.2-1 of the PSAR shows subsystems that will be able to measure radioactivity in, and remove radioactive particles from, the cover gas. Based on this information, the NRC staff has reasonable assurance that the IGS will be designed consistent with PDC 64 because the IGS will be able to monitor for radioactivity that can potentially be released. Additionally, the NRC staff finds it acceptable to include a technical specification on circulating activity to support the determination of the specified acceptable radiological release design limit. This is consistent with PDC 64 which requires a means to monitor for radioactivity that may be released and guidance in NUREG-1537 which states systems should assess purity of gases.

Section 9.1.2.4, Testing and Inspection, of the PSAR states that the IGS backup argon system will be periodically checked for quantity leakage, argon volumes and gas purity will be included in the technical specifications. Additionally, the Kairos response to RCI 349 confirmed that the entirety of the IGS can be periodically checked for leakage.

The NRC staff has reasonable assurance that these capabilities will allow the IGS to meet its design functions to maintain an inert environment. The NRC staff also finds the proposed technical specifications for argon volume and purity acceptable as this is consistent with the guidance in NUREG-1537 that gas purity should be assessed. Based on the preliminary design, the NRC staff has reasonable assurance that leaks from the system that could release radionuclides, or introduce

contaminants (e.g., air or moisture), can be detected and appropriate action taken prior to exceeding operational limits. This is consistent with NUREG-1537 guidance which states that systems should be designed to ensure control and detection of leaks so there is no uncontrolled release of radioactive material.

The NRC staff has reasonable assurance that the preliminary design of the IGS will meet 10 CFR 20.1406 as there are systems provided to remove radioactive material from the IGS if necessary, and because the IGS can be monitored for leaks which would allow for action to limit release of radioactivity.

The NRC staff will verify the capability to monitor activity and releases with the design details provided in the FSAR and the OL application.

Conclusion Based on the review described above, the staff concludes that the information in Hermes PSAR Section 9.1.2 is sufficient and meets applicable guidance and regulatory requirements for issuance of a construction permit identified in the Regulatory Evaluation. The information provided gives the NRC staff reasonable assurance that the IGS can meet its design functions and will not lead to radiation exposure or releases that exceed limits in 10 CFR Part 20. The information also gives the NRC staff reasonable assurance that the IGS will be designed consistent with the guidance in NUREG-1537 to assess and maintain cover gas purity and meet any technical specification applicable to the IGS. Further information will be required to approve operation of the Hermes IGS (e.g., limit on circulating activity in the cover gas, details for allowable impurities, leakage detection) and will be reviewed at the OL stage.

9.1.3 Tritium Management System Introduction PSAR Section 9.1.3, Tritium Management System, describes a preliminary design for the tritium management system (TMS), which monitors and removes tritium from the vapor spaces of the reactor coolant system and the reactor building during normal operation. PSAR Section 9.1.3 states that the system does not perform safety-related functions.

Regulatory Evaluation The applicable regulatory requirements are as follows:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), 50.34(a)(4), and 50.34(a)(5) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards 10 CFR 100.11, Determination of exclusion area, low population zone, and population center distance As described in subsection 9.1.1 above, the NRC staff did not evaluate whether requirements in 10 CFR Part 20 would be met for the construction of the Hermes reactor. Instead, the NRC staff assessed whether Kairos identified the relevant requirements for an operating facility and provided descriptions of the preliminary facility design. The NRC staff assessed this to determine whether the PSAR provides an acceptable basis for the development of systems and whether there is reasonable assurance that Kairos will comply with the regulations in 10 CFR Part 20 during Hermes facility operation.

The applicable guidance for the NRC staffs evaluation is as follows:

NUREG-1537, Parts 1 and 2, Section 9.7, Other Auxiliary Systems. The guidance notes that the design, functions, and potential malfunctions of the auxiliary systems should not result in

reactor accidents or uncontrolled release of radioactivity and no function or malfunction of the auxiliary systems should interfere with or prevent safe shutdown of the reactor.

Technical Evaluation Principal design criteria applicable to the TMS are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 60, Control of Releases of Radioactive Materials to the Environment PDC 64, Monitoring Radioactivity Releases PSAR Section 9.1.3 states that the TMS does the following: (1) provides tritium separation from argon in the IGS, (2) provides tritium separation from air in the reactor building cells, and (3) provides final collection and disposal of tritium. The system captures tritium to reduce environmental releases.

The system is integrated into other Hermes systems based on expected tritium distribution among possible transport pathways and the feasibility of tritium capture in each environment. To accomplish these functions, getter alloy capture beds are used to separate tritium from argon, and molecular sieve capture beds are used to separate tritium from air. Following their in-service duty cycles, the tritium capture materials are stored in sealed canisters.

PSAR Section 9.1.3 states that the TMSs functions are not safety related. However, portions of the TMS may be near safety-related SSCs. Therefore, PSAR Section 9.1.3 states that TMS components will be designed and positioned to preclude adverse interactions with safety-related SSCs, consistent with PDC

2. Consistent with the evaluations in Sections 3.4 and 3.5 of this SER, the staff has reasonable assurance that the TMS will be designed to meet PDC 2 because the PSAR describes potential methods which will be employed to ensure that a postulated failure of the TMS does not impact safety-related SSCs.

PSAR Section 9.1.3 states that, consistent with PDC 13, Instrumentation and control, tritium inventories will be monitored to comply with the inventory limits set by Maximum Hypothetical Accident (MHA) assumptions. This will ensure that the dose due to accidental releases from the TMS are bounded by the MHA and would therefore meet the accident dose criteria in 10 CFR 100.11. PSAR Section 9.1.3 states that the TMS maintains a minimum level of overall tritium capture capacity to minimize tritium releases from the plant and satisfy PDC 60. PSAR Section 9.1.3 states that radiation monitoring is provided in the TMS for the evaluation of tritium levels in TMS subsystems which satisfies, in part, PDC 64. PSAR 9.1.3 states that, because the system contains radiological contaminants, the system is designed to minimize contamination and support eventual decommissioning consistent with 10 CFR 20.1406. The NRC staff finds the TMS will be consistent with the requirements of 10 CFR 20.1406 because it is designed, in conjunction with other system, to collect and contain tritium generated from reactor operation.

For the TMS, the NRC staff used the guidance and acceptance criteria in NUREG-1537, Parts 1 and 2, Section 9.7, Other Auxiliary Systems, to review the TMS preliminary design description in PSAR Section 9.1.3, Tritium Mitigation System. As part of its review, the NRC staff assessed whether PSAR Section 9.1.3 identifies the appropriate PDC for the TMS. Based on the staffs evaluation of the preliminary design features described above, the staff finds that Kairos has described the design bases for the TMS and that the preliminary information on the design and functional description of the TMS provides reasonable assurance that the TMS will conform to the design bases. The NRC staff finds that the TMS is a non-safety-related system that will be designed such that it will (1) not result in reactor accidents, (2) not prevent safe shutdown of the reactor, and (3) not result in unacceptable radioactivity releases or exposures. Therefore, based on its review, the NRC staff determined that the level of detail provided on the TMS, including the design bases, and identification of relevant PDC, is adequate for a preliminary design and is consistent with the applicable acceptance criteria of NUREG-1537, Part 2,

Section 9.7. Accordingly, the NRC staff finds that the PSARs preliminary information regarding the Hermes TMS provides reasonable assurance that it will comply with applicable requirements.

Conclusion Based on the review described above, the staff concludes that the information on the TMS in PSAR Section 9.1.3 is sufficient and meets applicable guidance and regulatory requirements for issuance of a construction permit identified in the Regulatory Evaluation. The NRC staff concludes that the preliminary design features intended to minimize contamination, support eventual decommissioning, and control releases to the environment will comply with the requirements of 10 CFR 20.1406 and 10 CFR 100.11.

Based on the information provided, the NRC staff concludes that the TMS will be designed to collect and contain tritium, consistent with the guidance in NUREG-1537 to prevent uncontrolled release of radioactivity, to limit potential radiation exposures to within 10 CFR Part 20 requirements, and for no function or malfunction of the TMS to interfere with or prevent safe shutdown of the reactor.

9.1.4 Inventory Management System Introduction PSAR Section 9.1.4, Inventory Management System, describes a preliminary design for the IMS, which adds and removes reactor coolant to maintain the desired level and volume within reactor-coolant-containing systems and components (e.g., reactor vessel, CCS). PSAR Section 9.1.4 states that the system does not perform safety-related functions.

Regulatory Evaluation The applicable regulatory requirements are as follows:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), 50.34(a)(4), and 50.34(a)(5) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards As described in subsection 9.1.1 above, the staff did not evaluate whether requirements in 10 CFR Part 20 would be met for the construction of the Hermes reactor. Instead, the staff assessed whether Kairos identified the relevant requirements for an operating facility and provided descriptions of the preliminary facility design. The staff assessed this to determine whether the PSAR provides an acceptable basis for the development of systems and whether there is reasonable assurance that Kairos will comply with the regulations in 10 CFR Part 20 during Hermes facility operation.

The applicable guidance for the staffs evaluation is as follows:

NUREG-1537, Parts 1 and 2, Section 9.7, Other Auxiliary Systems. The guidance notes that the design, functions, and potential malfunctions of the auxiliary systems should not result in reactor accidents or uncontrolled release of radioactivity and no function or malfunction of the auxiliary systems should interfere with or prevent safe shutdown of the reactor.

Technical Evaluation Principal design criteria applicable to the IMS are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases PDC 15, Reactor Coolant System Design PDC 33, Reactor Coolant Inventory Maintenance

PDC 70, Reactor Coolant Purity Control PSAR Section 9.1.4 states that the IMS does the following: (1) adds and removes reactor coolant to maintain the desired level and volume within the reactor vessel and the CCS and (2) provides an interface for new reactor coolant delivery and used reactor coolant removal. The free volume in the system is filled with inert gas from the IGS. The IGS can apply inert gas pressure or vacuum on the IMS tanks to circulate cover gas through the TMS or initiate reactor coolant transfers between tanks.

Electrical heating and thermal insulation of the tanks, pumps, and piping is provided to maintain the reactor coolant in a liquid phase for system operations. Process sensors are included for use by the plant control system to monitor the reactor coolant inventory (e.g., load cells and coolant level sensors) and temperature in the system tanks.

The IMS reactor coolant transfer lines will be constructed of stainless steel and designed per ASME B31.3 2016. The IMS tanks will be constructed of stainless steel and designed per ASME BPVC,Section VIII 2015. The tanks and reactor coolant transfer lines are being designed and fabricated to meet the pressure, mechanical loads, corrosion, and the temperature requirements of the system.

The reactor vessel coolant level management tank provides a means to maintain the level of reactor coolant in the reactor vessel through a transfer line, a dip tube, and an overflow weir. The transfer into the reactor vessel is pump driven through the dip tube. The return flow is collected by an overflow weir and the transfer is gravity driven back into the reactor vessel coolant level management tank.

The reactor vessel fill/drain tank provides a means of filling and draining the reactor vessel through a transfer line and a dip tube. The transfer to the reactor vessel is pump driven, and the transfer out of the reactor vessel is gravity driven. The reactor vessel fill/drain tank transfer line is equipped with a passive reactor vessel isolation system to prevent unintentional draining. The reactor vessel fill/drain tank is sized to hold the reactor vessel coolant inventory.

The system is being designed to preclude the inadvertent draining of the reactor vessel via siphoning.

PSAR Section 4.3 describes the anti-siphoning function.

PSAR Section 9.1.4 states that the IMSs functions are not safety related. However, portions of the IMS may be near safety-related SSCs. Therefore, PSAR Section 9.1.4 states that IMS components will be designed and positioned to preclude adverse interactions with safety-related SSCs, consistent with PDC

2. Consistent with the evaluations in Sections 3.4 and 3.5 of this SER, the staff has reasonable assurance that the IMS will be designed to meet PDC 2 because the PSAR describes potential methods which will be employed to ensure that a postulated failure of the IMS does not impact safety-related SSCs.

PSAR Section 9.1.4 states that the IMS will be designed such that safety-related systems near the IMS are protected against the dynamic effects potentially created by the failure of IMS equipment, consistent with PDC 4. PSAR Section 9.1.4 states that the system will be designed to prevent inadvertent draining of the reactor vessel, consistent with PDC 15. Safety-related portions of the reactor coolant boundary are limited to the reactor vessel, such that failures of other SSCs containing reactor coolant do not result in unacceptable consequences, consistent with PDC 33. The CCS will be used to monitor the coolant chemistry, and the IMS may be used to replace coolant to restore conformance to the Flibe specification, consistent with PDC 70. PSAR Section 9.1.4 states that the system will be designed to minimize contamination and support eventual decommissioning, consistent with 10 CFR 20.1406, Minimization of Contamination.

For the IMS, the NRC staff used the guidance and acceptance criteria in NUREG-1537, Parts 1 and 2, Section 9.7, Other Auxiliary Systems, to review the IMS preliminary design description in PSAR

Section 9.1.4, Inventory Management System. As part of its review, the staff assessed whether PSAR Section 9.1.4 identifies the appropriate PDCs for the IMS. Based on the staffs evaluation of the preliminary design features described above, the staff finds that Kairos has provided the design bases for the IMS and that the preliminary information on the design and functional description of the IMS provides reasonable assurance that the IMS will conform to the design bases. The staff finds that the IMS is a non-safety-related system that will be designed such that it will 1) not result in reactor accidents,

2) not prevent safe shutdown of the reactor, and 3) not result in unacceptable radioactivity releases or exposures. Therefore, based on its review, the staff determined that the level of detail provided on the IMS, including the design bases and identification of relevant PDC, is adequate for a preliminary design and is consistent with the applicable acceptance criteria of NUREG-1537, Part 2, Section 9.7.

Accordingly, the staff finds that the PSARs preliminary information regarding the Hermes IMS provides reasonable assurance that it will comply with applicable requirements.

The NRC staffs technical evaluation of the anti-siphoning function is in Section 4.3 of this SER.

Conclusion Because of the review of IMS features to maintain coolant level, volume, and purity described above, the NRC staff concludes that the information on the IMS in PSAR Section 9.1.4 is sufficient and meets applicable guidance and regulatory requirements for issuance of a construction permit identified in the Regulatory Evaluation. The NRC staff concludes that the preliminary design features intended to minimize contamination and support eventual decommissioning will comply with the requirements of 10 CFR 20.1406. Based on the information provided, the NRC staff concludes that the preliminary design of the IMS satisfies the applicable acceptance criteria in NUREG-1537 to prevent uncontrolled release of radioactivity and for no function or malfunction of the IMS to interfere with or prevent safe shutdown of the reactor.

9.1.5 Reactor Thermal Management System Introduction PSAR Section 9.1.5, Reactor Thermal Management System, describes a preliminary design for the reactor thermal management system (RTMS) which consists of two subsystems - the equipment and structural cooling subsystem (ESCS) and the reactor auxiliary heating system (RAHS). The purpose of the ESCS is to remove heat from SSCs in the reactor cavity to maintain the temperatures within operational limits. The purpose of the RAHS is to preheat the reactor vessel and to ensure Flibe in the vessel is maintained above a minimum operating temperature. PSAR Section 9.1.5 states that the RTMS does not perform safety-related functions.

Regulatory Evaluation The applicable regulatory requirements are as follows:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), 50.34(a)(4), and 50.34(a)(5) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards As described in subsection 9.1.1 above, the staff did not evaluate whether requirements in 10 CFR Part 20 would be met for the construction of the Hermes reactor. Instead, the NRC staff assessed whether Kairos identified the relevant requirements for an operating facility and provided descriptions of the preliminary facility design. The NRC staff assessed this to determine whether the PSAR provides an acceptable basis for the development of systems and whether there is reasonable assurance that Kairos will comply

with the regulations in 10 CFR Part 20 during Hermes facility operation. The applicable guidance for the staffs evaluation is as follows:

NUREG-1537, Parts 1 and 2, Section 9.7, Other Auxiliary Systems. The guidance notes that the design, functions, and potential malfunctions of the auxiliary systems should not result in reactor accidents or uncontrolled release of radioactivity and no function or malfunction of the auxiliary systems should interfere with or prevent safe shutdown of the reactor.

Technical Evaluation Principal design criteria applicable to the RTMS are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 44, Structural and Equipment Cooling PDC 45, Inspection of Structural and Equipment Cooling Systems PDC 46, Testing of Structural and Equipment Cooling Systems PDC 71, Reactor Coolant Heating Systems PSAR Section 9.1.5 states the ESCS removes heat from SSCs in the reactor cavity area to maintain the operational temperature limits of those structures and components. This is accomplished by active heat removal features (see below) as well as high temperature, load-bearing, irradiation-hardened insulation on SSCs. Heat removed by the subsystem is transferred to the Component Cooling Water System.

The systems that are actively cooled by the ESCS during normal operation are the steel liner of the concrete structure, the reactivity control and shutdown system, and the primary salt pump. The steel liner of the concrete structure is water cooled. The primary salt pump is gas cooled by active components (fans, blowers, pumps) in a closed loop system.

PSAR Section 9.1.5 states the RAHS pre-heats the reactor vessel and ensures coolant is maintained above a minimum operating temperature. This is accomplished by electric heaters adjacent to the heated surface. The system provides the initial startup heating required to achieve and maintain operational temperature for the reactor vessel, primary salt pump, reactor vessel internals, IMS, IGS, and PHSS reactor vessel head penetrations prior to the availability of nuclear heating. For initial commissioning, the system will be used to evaporate residual moisture out of the reactor vessel, internal graphite structures, and the reactor vessel, to preclude corrosion upon contact with the introduction of the molten salt coolant. Because the reactor coolant melting point is high, preheating of the vessel and internals prevents thermal shock damage to the reactor.

PSAR 9.1.5 states that the RTMSs functions are not safety related. However, portions of the RTMS may be near safety-related SSCs. Therefore, PSAR Section 9.8.2 states that RTMS components will be designed and positioned to preclude adverse interactions with safety-related SSCs, consistent with PDC

2. Consistent with the evaluations in Sections 3.4 and 3.5 of this SER, the staff has reasonable assurance that the RTMS will be designed to meet PDC 2 because the PSAR describes potential methods which will be employed to ensure that a postulated failure of the RTMS does not impact safety-related SSCs.

PSAR 9.1.5 states that the ESCS will be designed to detect gas and water leaks and isolate breaches in the system via the plant control system, and that the ESCS will be designed to permit inspection and testing to ensure the capability of the ESCS to cool SSCs and interface with the Component Cooling Water System (CCWS) to transfer heat to the ultimate heat sink, consistent with PDC 44, PDC 45, and PDC 46.

PSAR Section 9.1.5 states that the RAHS will ensure sufficient heat is added to components to compensate for parasitic heat loss during periods when the reactor is not supplying fission heat and to

heat the vessel and vessel head components uniformly to prevent thermal shock, consistent with PDC

71.

PSAR Section 9.1.5. states that the RTMS will be designed to minimize contamination and support eventual decommissioning, consistent with 10 CFR 20.1406.

For the RTMS, the NRC staff used the guidance and acceptance criteria in NUREG-1537, Parts 1 and 2, Section 9.7, Other Auxiliary Systems, to review the RTMS preliminary design description in PSAR Section 9.1.5, Reactor Thermal Management System. As part of its review, the NRC staff assessed whether PSAR Section 9.1.5 identifies the appropriate PDC for the RTMS. Based on staffs evaluation of the preliminary design features described above, the NRC staff finds that Kairos has described the design bases for the RTMS and that the preliminary information on the design and functional description of the RTMS provides reasonable assurance that the RTMS will conform to the design bases. The NRC staff finds that the RTMS is a non-safety-related system that will be designed such that it will 1) not result in reactor accidents, 2) not prevent safe shutdown of the reactor, and 3) not result in unacceptable radioactivity releases or exposures. Therefore, based on its review, the NRC staff determined that the level of detail provided on the RTMS, including the design bases, and identification of relevant PDC, is adequate for a preliminary design and is consistent with the applicable acceptance criteria of NUREG-1537, Part 2, Section 9.7. Accordingly, the NRC staff finds that the PSARs preliminary information regarding the Hermes RTMS provides reasonable assurance that it will comply with applicable requirements.

Conclusion Based on the review of RTMS features described above to maintain Flibe above a minimum operating temperature and cool nearby structures and equipment, the staff concludes that the information on the RTMS in PSAR Section 9.1.5 is sufficient and meets applicable guidance and regulatory requirements for issuance of a construction permit identified in the Regulatory Evaluation. The NRC staff concludes that the preliminary design features intended to minimize contamination and support eventual decommissioning will comply with the requirements of 10 CFR 20.1406. Based on the information provided, the NRC staff concludes that the preliminary design of the RTMS satisfies the applicable acceptance criteria in NUREG-1537 to not result in accidents or unacceptable radioactivity releases and for no function or malfunction of the RTMS to interfere with or prevent safe shutdown of the reactor.

9.2 Reactor Building Heating, Ventilation, and Air Conditioning System Introduction Section 9.2 of the Hermes PSAR describes the RBHVAC. The RBHVAC is not proposed to provide any safety-related function or support any safety-related SSCs. Although radiation monitoring and filtration will be provided, the RBHVAC is not needed to mitigate any postulated event. No technical specifications are proposed for the RBHVAC. The RBHVAC performs the following non-safety-related functions:

Maintain environmental conditions (air quality, temperature, humidity, pressure, and noise levels) for personnel health, habitability, and for SSC operability Provide a means to control and monitor tritium, beryllium, and other controlled effluents Monitor exhaust air vented from the reactor building for controlled effluents Ensure ventilation flow from areas of low hazard to areas of higher hazard potential Minimize contamination of facility areas

Regulatory Evaluation The applicable regulatory requirements for the evaluation of the Hermes non-power test reactor RBHVAC are as follows:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), and 50.34(a)(4) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards As described in subsection 9.1.1 above, the staff did not evaluate whether requirements in 10 CFR Part 20 would be met for the construction of the Hermes reactor. Instead, the NRC staff assessed whether Kairos identified the relevant requirements for an operating facility and provided descriptions of the preliminary facility design. The NRC staff assessed this to determine whether the PSAR provides an acceptable basis for the development of systems and whether there is reasonable assurance that Kairos will comply with the regulations in 10 CFR Part 20 during Hermes facility operation.

Guidance for this review includes:

NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Parts 1 and 2. Specifically, Section 9.1 Heating, Ventilation, and Air Conditioning Systems provides review criteria and procedures. The guidance notes that HVAC systems will support functions like the ALARA program, radiation monitoring, and contamination control which are described in Chapter 11, Radiation Protection Program and Waste Management, of the PSAR and this safety evaluation.

Regulatory Guide 4.20, Constraint on Releases of Airborne Radioactive Materials to the Environment for Licensees Other than Power Reactors, Revision 1. The constraint of 10 CFR 20.11.1(d) may be interpreted as that fraction of the public dose limit allocated to airborne emissions to ensure that doses are ALARA through this particular release pathway. Licensees are required to design their facilities and structure operations such that airborne emissions of radioactive materials generated from operations result in doses to the public that are below the constraint. The constraint serves as a starting point, or upper level, for ALARA assessments. If licensees exceed the constraint on airborne emissions, they are required to report the radiation dose to the NRC and to take corrective actions to lower the dose below the constraint value.

Technical Evaluation Principal design criteria applicable to the RBHVAC are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 60, Control of Releases of Radioactive Materials to the Environment PDC 64, Monitoring Radioactivity Releases The NRC staff reviewed material presented in the PSAR against acceptance criteria of NUREG-1537, Part 2 Section 9.1. The PSAR describes the design of the RBHVAC but does not provide details of design features. The NRC staff expects that the FSAR submitted with the OL application will provide details on design features.

The RBHVAC supplies filtered fresh air to the reactor building and monitors and filters air that is exhausted to the environment. The RBHVAC contains return air flow paths throughout the reactor building, including from the reactor and PHSS cells. The TMS (described in Section 9.1.3 of the PSAR and evaluated in Section 9.1.3 of this safety evaluation) has a subsystem within the RBHVAC to collect airborne tritium and convert it to water. The tritiated water is collected and handled as radioactive waste (described in Section 11.2.2 of the PSAR and evaluated in Section 11.2 of this safety evaluation).

This process assures that normal sources of airborne radioactive material are diverted and filtered to maintain occupational doses and doses to the public are ALARA. The NRC staff finds that Kairos has shown that the preliminary Hermes design will acceptably meet the requirements of 10 CFR 20.1101(b) because of RBHVAC design features, including the supply of fresh air to the reactor building, the filtering of exhaust air, and the integration of the TMS with the RBHVAC.

The RBHVAC is designed to direct flow and leakage toward areas of higher radioactivity to minimize diffusion or other uncontrolled release of airborne radioactive material from the reactor cell and the PHSS cell. The NRC staff finds that Kairos has shown that the description and level of detail on the RBHVAC will acceptably meet the requirements of 10 CFR 20.1101(b) and 10 CFR 20.1406(a) based on the design of the system to move contaminated air to areas of higher radioactivity.

The design of the reactor building and the RBHVAC minimize the uncontrolled release of airborne radioactive material to the environment. All exhaust paths will be monitored, filtered with high efficiency particulate air filters, and can be isolated. Final details of the design and isolation capability will be provided in the operating license application. Higher radiation areas will be sealed to minimize leakage and will be fully contained within lower radiation areas.

Analyses of the operation of the RBHVAC described in PSAR Section 11.1.5 demonstrate that planned releases of airborne radioactive material to the environment will be within regulatory limits. The analyses calculate a peak dose of 0.10 mrem/year at a nearby location accessible to the public. This dose is well within the constraint of 10 mrem/year required by 10 CFR 20.1101(d). Because Kairos demonstrated a peak dose of 0.10 mrem/year to the public, the staff finds that Kairos has shown that the Hermes design will acceptably meet the requirements of 10 CFR 20.1101(d).

Conclusion In the PSAR Sections 9.1.3, 9.2, and 11.2.2, Kairos has identified design features for the RBHVAC and the TMS. Based on the review described above, the NRC staff concludes that the information in Hermes PSAR Section 9.2 is sufficient and meets applicable guidance and regulatory requirements for issuance of a construction permit identified in the Regulatory Evaluation. On the basis of its review, the NRC staff has determined that the level of detail provided regarding the RBHVAC demonstrates an adequate basis for a preliminary design and satisfies the applicable acceptance criteria of NUREG-1537 to support the ALARA program, radiation monitoring and contamination control functions. Because of these descriptions and the findings discussed above, the NRC staff concludes that Kairos will meet the requirements of 10 CFR 50.34(a) for the RBHVAC. Based on the NRC staff review and findings above, the NRC staff concludes that the design features and analyses described in the PSAR demonstrate that Kairos will meet 10 CFR 20.1101(b), 10 CFR 20.1101(d), and 10 CFR 20.1406(a).

9.3 Pebble Handling and Storage System Introduction PSAR Section 9.3, Pebble Handling and Storage System, describes a preliminary design for the PHSS, which provides for handling and storing fuel and other pebbles. The system encompasses receipt and inspection of new fuel upon delivery, core loading, sensing, inspection and sorting during downstream circulation, re-insertion, core unloading, and removal and transfer to storage.

Regulatory Evaluation The applicable regulatory requirements are as follows:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary

safety analysis report. including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), 50.34(a)(4), and 50.34(a)(5) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards 10 CFR 70.24, Criticality accident requirements, paragraph (a)(1). Paragraph (a)(1) requires monitoring in areas of the plant where nuclear material is handled or stored As described in subsection 9.1.1 above, the NRC staff did not evaluate whether requirements in 10 CFR Part 20 would be met for the construction of the Hermes reactor. Instead, the staff assessed whether Kairos identified the relevant requirements for an operating facility and provided descriptions of the preliminary facility design. The NRC staff assessed this to determine whether the PSAR provides an acceptable basis for the development of systems and whether there is reasonable assurance that Kairos will comply with the regulations in 10 CFR Part 20 during Hermes facility operation.

The applicable guidance for the NRC staffs evaluation is NUREG-1537, Parts 1 and 2, Section 9.2, Handling and Storage of Reactor Fuel.

Technical Evaluation Principal design criteria applicable to the RBHVAC are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 3, Fire Protection PDC 4, Environmental and Dynamic Effects Design Bases PDC 33, Reactor Coolant Inventory Maintenance PDC 61, Fuel Storage and Handling and Radioactivity Control PDC 62, Prevention of Criticality in Fuel Storage and Handling PDC 63, Monitoring Fuel and Waste Storage PSAR Section 9.3 states the PHSS provides for receipt and inspection of new fuel upon delivery; performs core loading, sensing, inspection and sorting during downstream circulation; re-insertion; core unloading; and removal and transfer to storage. Major components, features, and functions include the pebble extraction machine, debris removal, off-head conveyance line, pebble processing, pebble inspection, pebble insertion, inert gas boundary, pebble storage, and new pebble introduction.

The spent fuel storage portion of the system is composed of stainless steel storage canisters and a transporter device to move canisters. Individual storage canisters will be designed to hold 1900 to 2100 pebbles. After the canister is filled with pebbles, the canisters fill valve is closed, and the canister is moved via an automated transfer system for sealing. After sealing, the canister is moved with a canister transporter to a water-filled cooling pool for initial spent fuel storage. After cooling in pool storage, the canister is moved to an air-cooled storage bay.

Consistent with PDC 2, the concrete structures associated with the storage bay, pool, and support restraints in the pool are being designed as Seismic Design Category 3 structures to ensure the geometry of the storage area is maintained to prevent inadvertent criticality during a design basis earthquake. For a postulated earthquake, the TRISO fuel is passively cooled by the transporter and the pool which provides passive cooling and spacing to restrict pebble movement to prevent criticality.

Other portions of the PHSS that do not perform a safety function will be either seismically mounted or physically separated to preclude adverse interactions with other safety-related SSCs during a design basis earthquake.

Consistent with PDC 3, the PHSS is being designed to minimize the probability of a fire or explosion by limiting the accumulation of potentially combustible material such as graphite dust within the system.

The PHSS is not located near nor interfaces with pneumatic systems with the potential for air leakage into the PHSS. The PHSS is filled with inert gas at a positive pressure to help prevent air ingress.

Locations within the PHSS where pebbles are not submerged in coolant, such as the pebble extraction machine, will either not exceed temperatures that would induce pebble oxidation or are expected to quickly cool such that oxidation, if any, would be minimal and not affect the acceptability of the pebble for reuse.

Consistent with PDC 4, the pebble handling portion of the PHSS is protected from the effects of discharging fluids. There are no pressurized piping systems in or around the PHSS thus precluding the need to consider high energy lines and related dynamic effects in the design. A hypothetical water line break does not pose a criticality risk, because the storage system criticality analysis assumes complete submergence and internal flooding of canisters in water. The PHSS is designed to handle the high radiation environment. The PHSS accounts for the temperature within the system to preclude oxidation of pebbles. The canisters are designed for the internal pressure due to accumulation of radionuclides and thermal loads associated with the spent fuel loaded in each canister. The canisters are designed for the tensile stress exerted during transfer. The canister interior is designed to handle radiolysis products from spent fuel to ensure the integrity of the canister, seal, and weld precluding release of radionuclides from the canister.

Consistent with PDC 33, the design of the PHSS minimizes the coolant drained from the reactor system in the event of a PHSS failure. The pebble extraction machine elevation is near the free surface of the coolant so that only a small amount of coolant is lost if the extraction machine breaks. Similarly, the pebble insertion line elevation is no lower than the salt pump elevation, limiting the amount of coolant lost with an insertion line break.

Consistent with PDC 61, the safety-related portions of the PHSS that contain reactivity will be designed to ensure (1) capability to permit appropriate inspection and testing of components, (2) suitable shielding for radiation protection, (3) appropriate containment, confinement, and filtering, (4) residual heat removal capability, and (5) precluding significant reduction in fuel storage cooling under postulated event conditions. Regarding the inspection and testing of components, pebbles exiting the core are inspected to identify abnormal wear, cracking, and missing surfaces due to pebble chipping, as well as to determine the burnup. Regarding shielding for radiation protection, the PHSS will be adequately shielded to limit worker dose, in accordance with the radiation protection program described in PSAR Chapter 11.

Containment and confinement of radioactivity within the fuel is maintained for the limiting PHSS malfunction event as discussed in Section 13.1.5. Regarding residual heat removal capability, heat removal mechanisms within the system, such as thermal radiation and convection via natural circulation are sufficient to remove decay heat produced by pebbles moving through the system. Regarding precluding significant reduction in fuel storage cooling under postulated event conditions, the PHSS is designed to ensure decay heat loads from pebbles in the spent fuel pool are passively cooled by water in the pool in the event of a loss of power. The canisters in the storage bay are cooled during postulated events by natural convection due to the spacing which allows sufficient air flow.

Consistent with PDC 62, the system design will incorporate design features needed to ensure criticality is prevented. Examples include (1) removing pebbles at a rate that prohibits formation of a critical configuration in the event a PHSS line breach spills pebbles outside the PHSS, (2) maintaining an inert gas environment precluding moisture intrusion in fuel handling areas, (3) including physical constraints and system interlocks in the fuel handling equipment, and (4) making conservative assumptions in determining spacing requirements for each spent fuel canister such as assuming the canisters are submerged under water but are not flooded internally. A summary of the criticality analyses confirming the system design maintains a geometrically safe configuration will be provided in the FSAR.

Consistent with PDC 63, the PHSS is designed to ensure thermal and mechanical loads to the pebble and oxidation during handling, inspection, and loading into canisters do not exceed pebble design limits.

Pebble inspection and sorting performed by the PHSS ensures that damaged pebbles extracted from the reactor are removed from use. Monitoring of the cover gas and reactor coolant radioactivity provides early indication of potential TRISO failure.

The PHSS includes a monitoring system capable of detecting criticality which satisfies the requirement of paragraph (a)(1) of 10 CFR 70.24, Criticality Accident Requirements..

PSAR Section 9.3 states that the PHSS contains radiological contaminants, and therefore, the PHSS will be designed to minimize such contamination and support eventual decommissioning, consistent with the requirements of 10 CFR 20.1406, Minimization of contamination.

For the PHSS, the staff used the guidance and acceptance criteria in NUREG-1537, Parts 1 and 2, Section 9.2, Handling and Storage of Reactor Fuel, to review the PHSS preliminary design description in PSAR Section 9.3, Pebble Handling and Storage System. As part of its review, the staff assessed whether PSAR Section 9.3 identifies the appropriate PDC for the PHSS. The staff finds that Kairos has described the design bases for the PHSS and that the preliminary information on the design and functional description of the PHSS provides reasonable assurance that the PHSS will conform to the design bases. The staff finds that the PHSS will be designed such that 1) criticality will not occur, 2) fuel handling tools and procedures will be designed to avoid damaging fuel, 3) methods for assessing fuel radioactivity and potential exposure rates will be adequate to avoid personnel overexposure, and 4) shielding methods will ensure doses are below occupational exposure limits and as low as reasonably achievable. Therefore, based on its review, the staff determined that the level of detail provided on the PHSS, including the design bases, and identification of relevant PDC, is adequate for a preliminary design and is consistent with the applicable acceptance criteria of NUREG-1537, Part 2, Section 9.2.

Accordingly, the staff finds that the PSARs preliminary information regarding the Hermes PHSS provides reasonable assurance that it will comply with applicable requirements.

Conclusion On the basis of its review, the staff has determined that the level of detail provided regarding the Hermes PHSS demonstrates an adequate basis for a preliminary design and satisfies the applicable acceptance criteria of NUREG-1537 to support functions like maintaining subcriticality, preventing damage to pebbles, limiting radiation exposure, and material control and accounting. Based on the review described above, the NRC staff concludes that information provided in the PSAR meets the requirements of 50.34(a). Based on the staff findings above, the staff concludes that the preliminary design of the PHSS, as described in the PSAR, is sufficient and meets the applicable regulatory requirements and guidance for the issuance of a construction permit identified in the Regulatory Evaluation. Based on the monitoring capability described above, the NRC staff concludes that the preliminary design will meet the requirements of 10 CFR 70.24. The NRC staff concludes that the preliminary design features intended to minimize contamination and support eventual decommissioning will acceptably meet the requirements of 10 CFR 20.1406. Further technical or design information required to approve operation of the PHSS will be evaluated in the review of the FSAR at the OL stage.

9.4 Fire Protection Systems and Programs Introduction PSAR Section 9.4 Fire Protection Systems and Programs, describes a preliminary design for fire protection systems and related programs. The fire protection program integrates components, procedures, analysis, and personnel used to define and carry out all activities of fire protection. The fire

protection system is designed to detect, control, and extinguish fires so that a continuing fire will not prevent safe shutdown or result in an uncontrolled release of radioactive material that exceeds acceptance criteria. Kairos stated that a description of the fire protection program and a fire hazards analysis will be provided with the application for an OL.

Regulatory Evaluation The applicable regulatory requirements are as follows:

10 CFR 50.34(a) Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), 50.34(a)(4), and 50.34(a)(5) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards The NRC staff used guidance from NUREG 1537 and Interim Staff Guidance Augmenting NUREG 1537, (ADAMS Accession Nos. ML12156A069 and ML12156A075). Specifically, criteria from NUREG-1537 Section 9.3, Fire Protection Systems and Programs were applied in the review.

Technical Evaluation Kairos states in PSAR Section 9.4.2.3 that the fire protection systems conform to local building and fire codes. Additionally, PSAR Section 9.4.2.3 states that the fire protection systems will be designed to ANSI/ANS 15.17, Fire Protection Program Criteria for Research Reactors, and National Fire Protection Association 801, Standard for Fire Protection for Facilities Handling Radioactive Materials, 2020 edition. Life safety provisions are included in the facility design in accordance with the National Fire Protection Association 101, Life Safety Code, 2021 edition.

Kairos states in PSAR Section 9.2.2.2 that noncombustible and fire-resistant materials are used whenever practical, particularly in locations with SSCs that are safety-related or required for safe shutdown. Safety-related SSCs and equipment required for safe shutdown of the reactor are located within the reactor building. The floors, walls, and ceilings of the reactor building are constructed almost entirely of reinforced concrete.

According to PSAR Section 9.4.2.1, the fire protection system is designed to detect, control, and extinguish fires so that a continuing fire will not prevent safe shutdown or result in an uncontrolled release of radioactive material that exceeds acceptance criteria. Design features such as fire barriers and fire area penetration protection are also included to provide for life safety provisions and to minimize the spread of fire. Fire protection system elements and design features will be identified and installed in defined fire areas based on the results of the fire hazards analysis. Kairos states in PSAR Section 9.4.2.4 that functional tests of the fire protection system will be performed prior to startup.

Kairos states in PSAR Section 9.4.2.1 that the fire protection system includes fire detection and alarm systems as well as automatic and manual fire suppression systems. These fire detection and firefighting systems will be of appropriate capacity and capability and are designed to minimize the adverse effects of fires on SSCs that are safety-related or required for safe shutdown. The PDC applicable to fire protection systems are evaluated below based on the information provided in PSAR Section 9.4.

PDC 2 Design Bases for Protection Against Natural Phenomena Kairos states in PSAR Section 9.4.2.3 that fire water piping is routed such that a rupture or inadvertent operation of the fire protection system does not significantly impair safety-related or safe shutdown functions. Kairos stated that the fire protection systems are not safety-related but portions of these systems may cross the isolation moat discussed in Section 3.5 of the PSAR, and that SSCs that cross a base-isolation moat may experience differential displacements as a result of seismic events.

Kairos stated that the fire protection systems are designed so that postulated failures of SSCs in these systems from differential displacements will not preclude a safety-related SSC from performing its safety function. Based on these features of the fire protections systems, staff finds that the preliminary design is consistent with PDC 2.

PDC 3 Fire Protection The NRC staff reviewed Kaiross submittal against the four criteria described in PDC 3. The fire protection features described in the paragraphs above demonstrate that the preliminary design will 1) minimize the effect of fires, 2) will minimize the use of combustible materials, 3) will provide fire detection and fighting systems, and 4) will not impair the function of safety-related SSC and is thus consistent with PDC 3.

PDC 19 Control Room The NRC staff reviewed the PSAR as it applies to providing the capability both inside and outside the control room to operate plant systems necessary to achieve and maintain safe shutdown conditions.

In Section 7.4 of the PSAR Main Control Room and Remote Onsite Shutdown Panel, Kairos stated that the main control room provides the means for operators to monitor the behavior of the plant, control performance of the plant, and manage the response to postulated event conditions in the plant.

The MCR also contains a central alarm panel for the fire protection system so that operators can monitor the status of fire protection equipment inside the reactor building. The central alarm panel includes controls for the ventilation and extinguishing systems related to the response to fires.

The remote onsite shutdown panel (ROSP) provides separate means to shut down the plant and monitor plant parameters in response to postulated event conditions. The ROSP is located in the safety-related portion of the reactor building and is used in the event that the MCR becomes uninhabitable. The NRC staff finds that capabilities to monitor and control the plant in the event of a fire demonstrate consistency with PDC 19.

PDC 23 Protection System Failure Modes The NRC staff reviewed the PSAR as it applies to the protection system being designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if the plant experiences adverse environments such as from a fire. In Section 4.2.2 of the PSAR, Reactivity Control and Shutdown System, (RCSS) Kairos stated that the RCSS fails into a safe state in the event of adverse conditions or environments. In Section 7.3 of the PSAR, Reactor Protection System, (RPS) Kairos stated that the RPS fails to a safe state upon detection of adverse environmental conditions, such as a fire. The NRC staff finds that the preliminary design for the protection system to fail in a safe state in the event of a fire are consistent with PDC 23.

Conclusion On the basis of its review, the NRC staff has determined that the level of detail provided regarding the fire protection program and fire protection systems demonstrates an adequate basis for a preliminary design and satisfies the applicable acceptance criteria of NUREG-1537: prevention of fires and limitation of combustible materials, methods to detect, control, and extinguish fires, and ensuring safe shutdown in the event of a fire. Therefore, the NRC staff concludes that information provided in the PSAR meets the regulatory requirements identified above. Based on the review described above, the NRC staff concludes that the preliminary description of the fire protection program and the fire protection systems, as

described in the PSAR, provides reasonable assurance that the applicable regulatory requirements and guidance for the issuance of a construction permit are met. Further programmatic, technical, or design information required to approve operation of the test reactor will be evaluated in the review of FSAR at the OL stage.

9.5 Communication Systems Introduction PSAR Section 9.5 Communication, describes a preliminary design for the Hermes communication systems. Hermes PSAR, Section 9.5.1, states that the communication system provides communications during normal and emergency conditions between essential areas of the facility and between locations remote to the facility. The communication system is not safety-related; it is not credited for mitigation of design basis events and has no safe shutdown function.

Regulatory Evaluation The applicable regulatory requirements are as follows:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), and 50.34(a)(3)(iii) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards 10 CFR 50.47(b)(6) and 10 CFR 50.47(d)(3) for capability of communication to emergency personnel and the public 10 CFR 50, Appendix E, Section IV.E.9.

The applicable guidance for the staffs evaluation is NUREG-1537, Parts 1 and 2, Section 9.4, Communication Systems. The guidance notes the need for communication capabilities among plant operating staff.

Technical Evaluation As stated in Hermes PSAR, Section 9.5.2, the communication system provides the capabilities for both normal and emergency communications and has the following subsystems: plant radio, public address and general alarm, communication capability in the event of a loss of normal power, distributed antenna, and security communications.

Two-way communication is provided between the main control room and other locations in the reactor facility. The communication system uses diverse voice over internet protocol, commercial land, and cellular phone lines in combination in the control room and several other locations within and outside the reactor building to provide communications between key areas of the facility. The communication system is designed so that a failure of any one station does not impact the other stations. In an emergency, the public address system is used to alert personnel.

The NRC staff evaluated the sufficiency of the preliminary information on Hermes communication systems, as described in PSAR Section 9.5, using the guidance and acceptance criteria from NUREG-1537, Parts 1 and 2. Based on its review, the NRC staff determined that the level of detail provided on Hermes communication systems is adequate and meets the applicable acceptance criteria of NUREG-1537, Part 2, Section 9.4 (Ref. 9) because:

The facility communication systems are designed to provide two-way communication between the main control room and all other locations necessary for safe reactor operation.

The communication systems allow the reactor operator on duty to communicate with the supervisor on duty and with health physics personnel.

The communication systems allow a facility-wide announcement of an emergency.

The communication systems have provisions for summoning emergency assistance from designated personnel, as discussed in the physical security and emergency plans. Physical security and emergency plans will be evaluated as part of the operating license application review.

Conclusion On the basis of its review, the NRC staff has determined that the level of detail provided regarding the Hermes communication systems demonstrates an adequate basis for a preliminary design and satisfies the applicable acceptance criteria of NUREG-1537. Based on its findings above, the NRC staff concludes the design of the Hermes communication systems, as described in Hermes PSAR Section 9.5, is sufficient and meets the applicable regulatory requirements and guidance for the issuance of a construction permit identified in the Regulatory Evaluation and demonstrates an adequate design basis for a preliminary design. Based on the review and the findings discussed above, the NRC staff concludes that the preliminary design of communication systems meet the requirements of 10 CFR 50.47(b)(6), 50.47(d)(3), and Part 50 Appendix E Section IV.E.9. A more detailed evaluation of information will occur during the review of the Hermes OL application, at which time the NRC staff will confirm that the final design conforms to this design basis.

9.6 Possession and Use of Byproduct, Source, and Special Nuclear Material Introduction PSAR Section 9.6, Possession and Use of Byproduct, Source, and Special Nuclear Material, discusses radioactive materials, including byproduct material, source material, and special nuclear material (SNM) that will be present at the Hermes facility. PSAR Section 9.6 also discusses locations where these materials will be stored or used at the facility, systems that interact with these materials and controls for handling these materials. PSAR Section 9.6 states that the design bases for systems interacting with byproduct, source, or SNM are to prevent uncontrolled release of radioactive materials and to maintain any Kairos personnel exposures within 10 CFR Part 20 dose limits and ALARA objectives.

PSAR Section 9.6 states that Kairos construction permit application does not request authorization to possess any radioactive material, and that amendments or applications for license(s) allowing such possession of such material would be submitted at later date(s). During the General Audit, Kairos stated that it planned to possess byproduct, source and SNM associated with Hermes operation under a 10 CFR Part 50 OL for Hermes, but that it might also request authorization to possess such materials prior to the issuance of an OL, for example, through a CP amendment request. Possession of radioactive material by Kairos would be evaluated when an application is submitted to the NRC.

Regulatory Evaluation The applicable regulatory requirements for the evaluation of Kairos possession and use of byproduct, source, and SNM are as follows:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary

safety analysis report. including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), 50.34(a)(4), and 50.34(a)(5) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards 10 CFR Part 30, Rules of General Applicability to Domestic Licensing of Byproduct Material 10 CFR Part 40, Domestic Licensing of Source Material 10 CFR Part 70, Domestic Licensing of Special Nuclear Material As described in subsection 9.1.1 above, the NRC staff did not evaluate whether requirements in 10 CFR Part 20 would be met for the construction of the Hermes reactor. Instead, the NRC staff assessed whether Kairos identified the relevant requirements for an operating facility and provided descriptions of the preliminary facility design. The NRC staff assessed this to determine whether the PSAR provides an acceptable basis for the development of systems and whether there is reasonable assurance that Kairos will comply with the regulations in 10 CFR Part 20 during Hermes facility operation.

The applicable guidance for the evaluation of Kairos possession and use of byproduct, source, and SNM is as follows:

NUREG-1537, Parts 1 and 2, Section 9.5, Possession and Use of Byproduct, Source, and Special Nuclear Material.

Technical Evaluation The Hermes core will contain two types of pebbles: fuel pebbles that contain enriched fuel particles and moderator pebbles that contain unenriched uranium particles. In PSAR Section 9.6.1, Kairos states that SNM at the Hermes facility during operation would consist of the fuel particles contained in fuel pebbles.

Hermes SNM fuel pebbles will use uranium enriched to various levels less than 20 percent enrichment.

PSAR Section 9.6.2 states Hermes will also utilize moderator pebbles with unenriched uranium particles, and therefore source material will also be present at the Hermes facility during operation. Handling, use, and storage of SNM and source material will occur in fresh fuel handling areas, the reactor vessel, and the PHSS.

In PSAR Section 9.6.3, Kairos states that byproduct material will be used to support Hermes operation, and byproduct material will also be generated by Hermes operation. Byproduct material at Hermes will include tritium, which is generated as a result of the nuclear reaction in the core, and which will be present in the irradiated fuel pebbles (in the reactor vessel and PHSS) and the primary cooling system.

PSAR Section 9.6.4 states that byproduct material, source material, and SNM may also be handled in laboratories at the Hermes facility for research and testing purposes. Further detail regarding the laboratory use of radioactive materials would be provided as part of an OL application and reviewed by the staff at the OL stage.

PSAR Sections 9.6.1 and 9.6.3 state that the designs of the fuel pebbles, reactor vessel, PHSS, and TMS prevent uncontrolled releases of radioactive material. Shielding is also used to reduce any direct doses from irradiated fuel pebbles, consistent with ALARA practices. PSAR Section 9.6.4 states that any material used in laboratories under the reactor license will be handled appropriately (e.g., using glove boxes and air exhaust systems, as applicable) to ensure any doses are ALARA and within 10 CFR Part 20 limits. In addition, as discussed in PSAR Section 9.6, spaces in which radioactive material is used and equipment used to handle radioactive material will be subject to administrative controls to minimize contamination, prevent radiological sabotage, theft, or diversion, and prevent uncontrolled release of the materials. Kairos states that byproduct material, source material, and SNM at the Hermes facility will be managed by compliance with the applicable provisions of 10 CFR Parts 30, 40, and 70, respectively.

The preliminary designs of the Hermes fuel, reactor vessel, PHSS, and TMS are discussed in further detail in PSAR Sections 4.2, 4.3, 9.3, and 9.1.3 respectively. The NRC staff reviewed the information on the preliminary design of these items and found it acceptable for the issuance of a CP, as discussed in Sections 4.2.1, 4.3, 9.3, and 9.1.3 of this SER.

PSAR Section 11.2.2 describes the preliminary design of the Hermes radioactive waste handling systems, and states that the systems will be designed to control releases of radioactive materials to the environment such that 10 CFR Part 20 limits will not be exceeded. The NRC staff reviewed the information on the preliminary design of the waste handling systems and found it acceptable for the issuance of a CP, as discussed in Section 11.2 of this SER.

PSAR Section 9.6 states that further information on administrative procedures related to the use of byproduct material, source material, and SNM during Hermes operation would be provided in an application for an OL. These administrative procedures would cover several technical and programmatic areas. PSAR Chapter 11 provides preliminary information on administrative controls for radioactive material at the Hermes facility and the Hermes ALARA program. The NRC staff reviewed the preliminary information provided in PSAR Chapter 11 and found it acceptable for the issuance of a CP as discussed in Chapter 11 of this SER. PSAR Sections 11.2.1 and 12.8 state that a description of the radioactive waste management program for the Hermes facility, and a description of the security plan for Hermes, respectively, will also be provided in an OL application. PSAR Section 9.4 states that the Hermes fire protection system will be designed to prevent a continuing fire from resulting in an unacceptable release of radioactive material, and that details of a Hermes fire hazards analysis, and fire protection program plan will be provided in an OL application. PSAR Chapter 12, Appendix A, provides a preliminary plan for addressing emergencies (including emergencies involving a potential release of radioactive material) at the Hermes facility, which the NRC staff reviewed and found acceptable for the issuance of a CP as discussed in Section 12.7 of this SER. PSAR Chapter 14 includes the potential items or variables that are expected topics of technical specifications for the Hermes facility. Administrative controls, including those for managing SNM, source, and byproduct material, will be provided in the application for an OL.

The NRC staff documented its review of Kairos probable subjects of technical specifications for the Hermes facility in Chapter 14 of this SER.

The NRC staff reviewed the information in PSAR Section 9.6 and other PSAR sections related to the Hermes facility preliminary design with respect to the byproduct material, source material, and SNM that will be used in the production facility. PSAR Section 9.6 and other PSAR sections describe a preliminary design with appropriate systems and controls to help ensure that doses from radioactive material at Hermes are within 10 CFR Part 20 dose limits and ALARA. In combination with this and based on its reviews of information in other PSAR sections as summarized above, the NRC staff finds that the information in the PSAR demonstrates an adequate design basis for a preliminary design and satisfies the applicable acceptance criteria of NUREG-1537, Part 2, Section 9.5. Based on this review, the NRC staff finds that: (1) the auxiliary facilities and systems are designed for the possession and use of source material, SNM and byproduct material produced by the Hermes reactor and (2) the Hermes design provides reasonable assurance that uncontrolled release of radioactive material to the unrestricted environment and public will not occur. Because the design bases include limits on potential personnel exposures, the NRC staff has reasonable assurance that Kairos will comply with the regulations in 10 CFR Part 20 and the ALARA program during Hermes facility operation.

Conclusion Based on the review and findings described above, the NRC staff concludes that the information provided in the PSAR and the preliminary design of the Kairos program and auxiliary facilities for the possession and use of byproduct material, source material, and SNM at Hermes, as described in the

PSAR, is sufficient and meets the applicable regulatory requirements and guidance for the issuance of a construction permit identified above. Based on the NRC staff review and findings above, the NRC staff concludes that the preliminary design and programs described in the PSAR demonstrate that Kairos will comply with 10 CFR 20.1101, 10 CFR 20.1201, and 10 CFR 20.1301. Further technical or design information required to approve operation of the test reactor will be evaluated in the review of the OL application.

9.7 Plant Water Systems Section 9.7 of the Hermes PSAR describes four water systems:

Service Water Treated Water Component Cooling Water Chilled Water The auxiliary water systems do not provide any safety-related function or support any safety-related SSCs, are not needed to mitigate any postulated event, and are not credited with performing safe shutdown functions. No technical specifications are proposed in the PSAR for these water systems.

These water systems perform the following non-safety-related functions:

Supply, treat, and store water Distribute water for cooling and maintenance Remove heat from non-essential loads Remove heat from essential loads Discharge heat to the environment The introduction to PSAR Section 9.7 states that water systems which directly interface with systems containing radioactive material will be designed to meet the requirements of 10 CFR 20.1406.

As indicated by PSAR Figure 9.7-1, only the CCWS interfaces with systems containing radioactive material in the current design. Kairos confirmed in the General Audit that, in the final design, any auxiliary water systems which connect to a system containing radioactive material will be designed to meet the requirements of 10 CFR 20.1406.

Common Regulatory Evaluation for Auxiliary Water Systems Common regulatory requirements and guidance for auxiliary water systems are identified here.

Any additional requirements or guidance specific to a system are identified in the subsection for that system. The applicable regulatory requirements for the evaluation of the Hermes non-power test reactor auxiliary water systems are:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report. including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), and 50.34(a)(4) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards Guidance for this review includes:

NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Parts 1 and 2. Specifically, Section 9.7, Other Auxiliary Systems provides review criteria and procedures. The guidance notes that the design, functions, and potential

malfunctions of the auxiliary systems should not result in reactor accidents or uncontrolled release of radioactivity and no function or malfunction of the auxiliary systems should interfere with or prevent safe shutdown of the reactor.

9.7.1 Service Water System Introduction The Hermes Service Water System draws water from municipal sources and provides the water to other water systems and supports general facility services (e.g., potable water). The Service Water System is not safety related and is not credited for the mitigation of postulated events. The Service Water System is designed in accordance with local building codes.

Regulatory Evaluation The requirements and guidance in the common regulatory evaluation for auxiliary water systems apply to the Service Water System. There are no additional regulatory requirements or guidance applicable to the Service Water System.

Technical Evaluation Principal design criteria applicable to the Service Water System are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases Section 9.7.1 of the PSAR describes the Service Water System as a supply to other water systems and for general facility use. PSAR Section 9.7.1 identifies traits of the system that will enable it to meet regulatory requirements. The detailed design of the Service Water System and its relation to safety evaluations will be reviewed as part of the OL application.

No portion of the Service Water System will be located in the proximity of safety-related SSCs.

Thus, staff finds that the preliminary design of the Service Water System is consistent with the objective of PDC 2 to withstand the effects of natural phenomena and is consistent with the guidance of NUREG-1537 Part 2, Section 9.7 for auxiliary systems to not cause accidents to the reactor, uncontrolled release of radioactivity, or interfere with safe shutdown of the reactor.

The Service Water System is a low-pressure system and thus is consistent with the objective of PDC 4 to accommodate the effects of environmental and dynamic effects during operation, maintenance, testing, and postulated accidents.

Conclusion Kairos has identified traits of the Service Water System in PSAR Section 9.7.1. The NRC staff finds that the traits described in the PSAR demonstrate that the preliminary design is consistent with the objectives of PDC 2, PDC 4, and the guidance of NUREG-1537 Part 2, Section 9.7. Based on the staffs review and the findings above, the NRC staff concludes that Kairos has met the requirements of 10 CFR 50.34(a) for the Service Water System because the PSAR provides a preliminary design and a preliminary evaluation of the design and performance of the Service Water System.

9.7.2 Treated Water System Introduction PSAR Section 9.7.2 states that the Treated Water System provides chemistry control for water supplied to the CCWS, the Chilled Water System, and the Decay Heat Removal System. Portions of the Treated Water System may be located in proximity to SSCs with safety-related functions and portions of the

system may cross the base-isolation moat that provides seismic protection for the reactor cell and PHSS cell. The Treated Water System is not safety related and is not credited for the mitigation of postulated events. The Treated Water System is designed in accordance with local building codes.

Regulatory Evaluation The requirements and guidance in the common regulatory evaluation for auxiliary water systems apply to the Treated Water System. There are no additional regulatory requirements or guidance for the Treated Water System.

Technical Evaluation Principal design criteria applicable to Treated Water System are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases The PSAR identifies traits of the Treated Water System that will enable it to meet regulatory requirements. Implementation of these traits and associated specific design features are not needed for the preliminary design required by a CP. The detailed design of the Treated Water System and its relation to safety evaluations will be reviewed as part of the operating license application.

Since portions of the Treated Water System may be located in proximity to SSCs with safety-related functions, the system will be designed through various means identified in the PSAR to protect those safety-related SSCs. Thus, staff finds that the preliminary design of the Treated Water System is consistent with the objective of PDC 2 to withstand the effects of natural phenomena and is consistent with the guidance of NUREG-1537 Part 2, Section 9.7 for auxiliary systems to not cause accidents to the reactor, uncontrolled release of radioactivity, or interfere with safe shutdown of the reactor.

The Treated Water System is a low-pressure system. Since portions of the Treated Water System cross the base-isolation moat, design features are provided to protect safety-related SSCs. Together, the low-pressure design and protective features is consistent with the objective of PDC 4 to accommodate the effects of environmental and dynamic effects during operation, maintenance, testing, and postulated accidents.

Conclusion Kairos has identified traits of the Treated Water System in PSAR Section 9.7.2. The NRC staff finds that the traits of the system, as described in the PSAR, demonstrate that the preliminary design is consistent with the objectives of PDC 2, PDC 4, and the guidance of NUREG-1537 Part 2, Section 9.7. Based on the NRC staffs review and the findings above, the staff concludes that Kairos has met the requirements of 10 CFR 50.34(a) for the Treated Water System because the PSAR provides a preliminary design and a preliminary evaluation of the design and performance of the Treated Water System.

9.7.3 Component Cooling Water System Introduction PSAR Section 9.7.3 states that the CCWS provides water cooling for the RBHVAC, the ESCS, the Spent Fuel Cooling System (SFCS), and the IGS. The CCWS is managed by the plant control system to maintain desired operational temperature limits. Heat from the CCWS is rejected to the environment.

The CCWS does not perform safety-related functions and is not credited for the mitigation of postulated events. Portions of the CCWS may be located in proximity to SSCs with safety-related functions and portions of the system may cross the base-isolation moat that provides seismic protection for the reactor cell and PHSS cell.

Regulatory Evaluation The requirements and guidance in the common regulatory evaluation for auxiliary water systems apply to the CCWS. As described in subsection 9.1.1 above, the NRC staff did not evaluate whether requirements in 10 CFR Part 20 would be met for the construction of the Hermes reactor. Instead, the NRC staff assessed whether Kairos identified the relevant requirements for an operating facility and provided descriptions of the preliminary facility design. The NRC staff assessed this to determine whether the PSAR provides an acceptable basis for the development of systems and whether there is reasonable assurance that Kairos will comply with the regulations in 10 CFR Part 20 during Hermes facility operation.

Technical Evaluation Principal design criteria applicable to the CCWS are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases PDC 44, Structural and Equipment Cooling PDC 45, Inspection of Structural and Equipment Cooling Systems PDC 46, Testing of Structural and Equipment Cooling Systems Section 9.7.3 of the PSAR provides a summary description of the CCWS and identifies traits of the CCWS that will enable it to meet regulatory requirements. Implementation of these traits and specific design features of the system are not needed for the preliminary design required by a CP. The detailed design of the CCWS and its relation to safety evaluations will be reviewed as part of the OL application.

The CCWS will interface with water systems that may contain radioactive material. The CCWS will be designed to minimize the contamination of the facility and the environment and minimize the generation or radioactive waste. Based on these design traits, staff finds that the preliminary design of the CCWS will acceptably meet the requirements of 10 CFR 20.1406.

Since portions of the CCWS may be located in proximity to SSCs with safety-related functions, the system is designed, through various means identified in the PSAR, to protect those safety-related SSCs.

Because of these design features, staff finds that the preliminary design of the CCWS is consistent with the objective of PDC 2 to withstand the effects of natural phenomena and meets the guidance of NUREG-1537 Part 2, Section 9.7 for auxiliary systems to not cause accidents to the reactor, uncontrolled release of radioactivity, or interfere with safe shutdown of the reactor.

The CCWS is a low-pressure system. Since portions of the CCWS cross the base-isolation moat, design features are provided to protect safety-related SSCs. Because of the low-pressure design and protective features, Kairos has shown that the CCWS is consistent with the objective of PDC 4 to accommodate the effects of environmental and dynamic effects during operation, maintenance, testing and postulated accidents.

The CCWS is designed to permit periodic inspection and testing to ensure the integrity and capability of the system to cool SSCs and to adequately transfer heat to the ultimate heat sink. Based on this capability, the NRC staff finds that the CCWS is consistent with the objectives of PDC 44, PDC 45, and PDC 46.

Conclusion Kairos has identified traits of the CCWS in PSAR Sections 9.7 and 9.7.3. The staff finds that the traits described in the PSAR demonstrate that the preliminary design is consistent with the objectives of PDC 2, PDC 4, PDC 44, PDC 45, PDC 46, and the guidance of NUREG-1537 Part 2, Section 9.7. Based on

the staffs review and the findings above, the NRC staff concludes that Kairos has met the requirements of 10 CFR 50.34(a) and 10 CFR 20.1406 for the CCWS because the PSAR provides a preliminary design and a preliminary evaluation of the design and performance of the CCWS.

9.7.4 Chilled Water System Introduction PSAR Section 9.7.4 states that the Chilled Water System provides cooling water to the RBHVAC system and other facility SSCs that are not safety-related. The Chilled Water System is not safety related and is not credited for the mitigation of postulated events. The Chilled Water System is designed in accordance with local building codes.

Regulatory Evaluation The requirements and guidance in the common regulatory evaluation for auxiliary water systems apply to the Chilled Water System. There are no additional regulatory requirements or guidance applicable to the Chilled Water System.

Technical Evaluation Principal design criteria applicable to Treated Water System are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases The PSAR identifies traits of the Chilled Water System that will enable it to meet regulatory requirements. Implementation of these traits and associated design features specific to the system are not needed for the preliminary design required by a CP. The detailed design of the Chilled Water System and its relation to safety evaluations will be reviewed as part of the OL application.

No portion of the Chilled Water System will be located in the proximity of safety-related SSCs.

Thus, staff finds that the preliminary design of the Chilled Water System is consistent with the objective of PDC 2 to withstand the effects of natural phenomena and meets the guidance of NUREG-1537 Part 2, Section 9.7 for auxiliary systems to not cause accidents to the reactor, uncontrolled release of radioactivity, or interfere with safe shutdown of the reactor.

The Chilled Water System is a low-pressure system and thus is consistent with the objective of PDC 4 to accommodate the effects of environmental and dynamic effects during operation, maintenance, testing, and postulated accidents.

Conclusion Kairos has identified traits of the Chilled Water System in PSAR Section 9.7.4. The NRC staff finds that the traits described in the PSAR demonstrate that the preliminary design is consistent with the objectives of PDC 2, PDC 4, and the guidance of NUREG-1537 Part 2, Section 9.7. Based on the NRC staffs review and the findings above, the NRC staff concludes that Kairos has met the requirements of 10 CFR 50.34(a) for the Chilled Water System because the PSAR provides a preliminary design and a preliminary evaluation of the design and performance of the Chilled Water System.

9.8 Other Auxiliary Systems 9.8.1 Remote Maintenance and Inspection System Introduction The Hermes test reactor will have a Remote Maintenance and Inspection System, as described in PSAR Section 9.8.1. The Remote Maintenance and Inspection System will provide the ability to remotely access, inspect, and handle components in the reactor system, the PHTS, and the PHSS. The Remote Maintenance and Inspection System is located in the reactor building and includes manipulators, tooling, cameras, monitors, cranes and rigging. The Remote Maintenance and Inspection System is not safety related and does not perform safety related functions. Portions of the system may cross the base-isolation moat that provides seismic protection for the reactor cell and PHSS cell.

Regulatory Evaluation The applicable regulatory requirements for the evaluation of the Hermes Remote Maintenance and Inspection System are as follows:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), and 50.34(a)(4) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards As described in subsection 9.1.1 above, the staff did not evaluate whether requirements in 10 CFR Part 20 would be met for the construction of the Hermes reactor. Instead, the NRC staff assessed whether Kairos identified the relevant requirements for an operating facility and provided descriptions of the preliminary facility design. The NRC staff assessed this to determine whether the PSAR provides an acceptable basis for the development of systems and whether there is reasonable assurance that Kairos will comply with the regulations in 10 CFR Part 20 during Hermes facility operation.

Guidance for this review includes:

NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Parts 1 and 2, Section 9.7, Other Auxiliary Systems.

Technical Evaluation Principal design criteria applicable to the Remote Maintenance and Inspection System are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases Section 9.8.1 of the PSAR provides a summary description of the Remote Maintenance and Inspection System and identifies traits of the system that will enable it to meet applicable regulatory requirements.

Implementation of these traits and associated specific design features are not needed for the preliminary design required by a CP. The detailed design of the system and its relation to safety evaluations will be reviewed as part of the OL application.

The purpose of the Remote Maintenance and Inspection System is to reduce personnel radiation exposure by providing tools for remote work on components in the reactor system, the PHTS, and the PHSS. The NRC staff finds that Kairos has shown that the Remote Maintenance and Inspection System will meet the requirements of 10 CFR 20.1101(b) through the use of tooling to remotely handle potentially radioactive material.

The Remote Maintenance and Inspection System will interface with components that may contain radioactive material. PSAR Section 9.8.1 states that the system will be designed to minimize the contamination of the facility and facilitate decommissioning. Based on these design traits, staff finds that the preliminary design of the system will acceptably meet the requirements of 10 CFR 20.1406.

Since portions of the Remote Maintenance and Inspection System may be located in proximity to SSCs with safety-related functions, the system will be designed so that it cannot interfere with a safety systems ability to perform a safety function. Because of these design features, staff finds that the preliminary design of the system is consistent with the objective of PDC 2 to withstand the effects of natural phenomena and will meet the guidance of NUREG-1537 Part 2, Section 9.7 for auxiliary systems to not cause accidents to the reactor, uncontrolled release of radioactivity, or interfere with safe shutdown of the reactor.

Since portions of the Remote Maintenance and Inspection System cross the base-isolation moat, design features are provided to protect safety-related SSCs. The protective features are consistent with the objective of PDC 4 to accommodate the effects of environmental and dynamic effects during operation, maintenance, testing, and postulated accidents.

Conclusion In PSAR Section 9.8.1, Kairos has identified design features for the Remote Maintenance and Inspection System. The NRC staff finds that the traits described in the PSAR demonstrate that the preliminary design is consistent with the objectives of PDC 2, PDC 4, and the guidance of NUREG-1537 Part 2, Section 9.7. Based on the NRC staffs review and the findings above, the NRC staff concludes that Kairos has met the requirements of 10 CFR 50.34(a) for the Remote Maintenance and Inspection System because the PSAR provides a preliminary design and a preliminary evaluation of the design and performance of the system. The NRC staff also concludes that the design features described in the PSAR demonstrate that Kairos will meet 10 CFR 20.1101(b) and 10 CFR 20.1406(a).

9.8.2 Spent Fuel Cooling System Introduction PSAR Section 9.8.2, Spent Fuel Cooling System, describes a preliminary design for the SFCS, which cools spent fuel canisters in the spent fuel storage pool and storage bay. PSAR Section 9.8.2 states that the system does not perform safety-related functions.

Regulatory Evaluation The applicable regulatory requirements are as follows:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), 50.34(a)(4), and 50.34(a)(5) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards As described in subsection 9.1.1 above, the staff did not evaluate whether requirements in 10 CFR Part 20 would be met for the construction of the Hermes reactor. Instead, the NRC staff assessed whether Kairos identified the relevant requirements for an operating facility and provided descriptions of the preliminary facility design. The NRC staff assessed this to determine whether the PSAR provides an acceptable basis for the development of systems and whether there is reasonable assurance that Kairos will comply with the regulations in 10 CFR Part 20 during Hermes facility operation.

The applicable guidance for the staffs evaluation is as follows:

NUREG-1537, Parts 1 and 2, Section 9.2, Handling and Storage of Reactor Fuel.

Technical Evaluation Principal design criteria applicable to Treated Water System are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases PSAR Section 9.8.2 states that the SFCS performs the following: (1) provides forced air cooling for the spent fuel storage canisters located in the storage bay of the PHSS and (2) recirculates water in the spent fuel storage pool to help cool spent fuel storage canisters in the pool. The SFCS transfers heat from the recirculating water to the CCWS. Temperatures in and around the storage canisters and other SSCs served by the SFCS will be monitored and controlled by the plant control system such that the SFCS fans and piping maintain the temperatures within desired limits. In addition, the SFCS will be capable of passively cooling the spent fuel storage canisters during postulated events, for example during a loss of power. PSAR Section 9.3.3 states that the storage part of PHSS will be designed with spacing between canisters in the spent fuel storage pool and the storage bay to ensure passive cooling (e.g., by natural convection) during postulated events.

PSAR Section 9.8.2 states that the SFCSs functions are not safety related. However, portions of the SFCS may be near safety-related SSCs. Therefore, PSAR Section 9.8.2 states that SFCS components will be designed and positioned to preclude adverse interactions with safety-related SSCs, consistent with PDC 2. PSAR Section 9.8.2 also states that because there will be no pressurized piping systems in or around the SFCS, the SFCS design is precluded from high energy line considerations, consistent with PDC 4.

PSAR Section 9.8.2 also states that the SFCS has the potential to become contaminated based on its location and system interfaces, and therefore, the SFCS will be designed to minimize such contamination consistent with the requirements of 10 CFR 20.1406, Minimization of contamination.

For the SFCS, the NRC staff used the guidance and acceptance criteria in NUREG-1537, Parts 1 and 2, Section 9.2, Handling and Storage of Reactor Fuel, to review the SFCS preliminary design description in PSAR Section 9.8.2, Spent Fuel Cooling System. As part of its review, the staff assessed whether PSAR Section 9.8.2 identifies the appropriate PDC for the SFCS. The NRC staff finds that Kairos has described the design bases for the SFCS and that the preliminary information on the design and functional description of the SFCS provides reasonable assurance that the SFCS will conform to the design bases. The NRC staff finds that the SFCS is a non-safety-related system that will be designed such that irradiated fuel can be cooled as necessary to avoid radionuclide release from the fuel during moving and storage within the facility. Therefore, based on its review, the NRC staff determined that the level of detail provided on the SFCS, including the design base, and identification of relevant PDC is adequate for a preliminary design and is consistent with the applicable acceptance criteria of NUREG-1537, Part 2, Section 9.2. Accordingly, the NRC staff finds that the PSARs preliminary information regarding the Hermes SFCS provides reasonable assurance that it will comply with applicable requirements.

Conclusion On the basis of its review, the staff has determined that the level of detail provided regarding the SFCS demonstrates an adequate basis for a preliminary design and satisfies the applicable acceptance criteria of NUREG-1537 to support functions like preventing thermal failure and limiting radiation exposure.

Based on the review described above, the NRC staff concludes that information provided in the PSAR

meets the requirements of 50.34(a). Based on the NRC staff findings above, the staff concludes that the preliminary design of the SCFS, as described in the PSAR, is sufficient and meets the applicable regulatory requirements and guidance for the issuance of a construction permit identified in the Regulatory Evaluation. Based on the review discussed above, the NRC staff concludes that the preliminary design features intended to minimize contamination and support eventual decommissioning will comply with the requirements of 10 CFR 20.1406. Further technical or design information required to approve operation of the test reactor will be evaluated in the review of the OL application.

9.8.3 Compressed Air System Introduction Section 9.8.3 of the PSAR states that the Compressed Air System provides compressed air for general facility services and for use in valve operation. The Compressed Air System is not safety related and is not credited with performing safe shutdown functions.

Regulatory Evaluation Regulatory requirements for the Compressed Air System are:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), and 50.34(a)(4) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards Guidance for this review includes:

NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Parts 1 and 2, Section 9.7, Other Auxiliary Systems.

Technical Evaluation Principal design criteria applicable to the Compressed Air System are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases PSAR Section 9.8.3 identifies traits of the Compressed Air System that will enable it to meet applicable regulatory requirements. Implementation of these traits and associated design features specific to the system are not needed for the preliminary design required by a CP. The detailed design of the Compressed Air System and its relation to safety evaluations will be reviewed as part of the OL application.

The Compressed Air System will be built so that failure of the system will not interfere with the ability of a safety-related system to perform its safety function. Thus, staff finds that the preliminary design of the Compressed Air System is consistent with the objective of PDC 2 to withstand the effects of natural phenomena, achieve the objective of PDC 4 to accommodate the effects of environmental and dynamic effects, and meets the guidance of NUREG-1537 Part 2, Section 9.7 for auxiliary systems to not cause accidents to the reactor, uncontrolled release of radioactivity, or interfere with safe shutdown of the reactor.

Conclusion Kairos has identified traits of the Compressed Air System in PSAR Section 9.8.3. The NRC staff concludes that the traits described in the PSAR demonstrate that the preliminary design is consistent with the objectives of PDC 2, PDC 4, and the guidance of NUREG-1537 Part 2, Section 9.7. Based on

the NRC staffs review and the findings above, the NRC staff concludes that Kairos has met the requirements of 10 CFR 50.34(a) for the Compressed Air System because the PSAR provides a preliminary design and a preliminary evaluation of the design and performance of the system.

9.8.4 Cranes and Rigging Introduction PSAR Section 9.8.4 describes a reactor building gantry crane that will be used to move equipment and support material receiving and shipping. Because of the heavy loads that would be lifted by the crane, failure or mis-operation of the crane could damage safety-related SSCs if there were a load drop.

The crane is not safety related and does not perform any safety related function.

Regulatory Evaluation Regulatory requirements for installed cranes are:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), and 50.34(a)(4) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards Guidance for this review includes:

NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Parts 1 and 2. Section 9.7, Other Auxiliary Systems.

Technical Evaluation Principal design criteria applicable to the crane are:

PDC 2, Design Bases for Protection Against Natural Phenomena PDC 4, Environmental and Dynamic Effects Design Bases Section 9.8.4 of the PSAR provides a summary description and evaluation of the reactor building crane and identifies traits of the crane that will enable it to meet regulatory requirements. Implementation of these traits and associated design features specific to the crane are not needed for the preliminary design required by a CP. The detailed design of the crane and its relation to safety evaluations will be reviewed as part of the OL application.

In PSAR Section 9.8.4.3, Kairos committed to implement the design standards in ASME B30.2-2016, Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist), for the reactor building crane. As indicated in the General Audit, a description of the use of ASME B30.2-2016 for testing, inspection, operator training, operation, and maintenance of the crane and rigging will be provided in the OL application.

Since portions of the crane and its rigging may be located in proximity to SSCs with safety-related functions, the crane is designed, through various means, to protect those safety-related SSCs, including seismic mounting of certain components, physical separation, and barriers. Because of these design features, staff finds that the preliminary design of the crane is consistent with the objective of PDC 2 to withstand the effects of natural phenomena and meets the guidance of NUREG-1537 Part 2, Section 9.7 for auxiliary systems to not cause accidents to the reactor, uncontrolled release of radioactivity, or interfere with safe shutdown of the reactor.

The crane will be designed in accordance with ASME B30.2-2016, Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist). Administrative controls will be

provided to ensure that a dropped load does not interfere with the ability of safety-related SSCs to perform their function during operation. The crane superstructure will be built to remain standing during and after a fire so that failure of the superstructure does not interfere with the ability of safety-related SSCs to perform their safety function(s). Because of the design, protective features, and administrative controls, Kairos has shown that the crane is consistent with the objective of PDC 4 to accommodate the effects of environmental and dynamic effects during operation, maintenance, testing, and postulated accidents.

Conclusion Kairos has identified traits of the reactor building crane in PSAR Section 9.8.4. The NRC staff concludes that the traits described in the PSAR demonstrate that the preliminary design is consistent with the objectives of PDC 2, PDC 4, and the guidance of NUREG-1537 Part 2, Section 9.7. Based on the NRC staffs review and the findings above, the NRC staff concludes that Kairos has met the requirements of 10 CFR 50.34(a) for the reactor building crane because the PSAR provides a preliminary design and a preliminary evaluation of the design and performance of the crane.

9.8.5 Auxiliary Site Services Introduction Section 9.8.5 of the PSAR states that the Hermes auxiliary site services include non-safety-related systems and equipment that support operation of the plant, such as machine shops, chemistry laboratory, sewers, lighting, warehousing, and storage. The auxiliary services are not credited for the mitigation of postulated events and will be built so that they will not interfere with the ability of safety-related SSCs to perform their safety function(s).

Regulatory Evaluation Regulatory requirements for the site auxiliary services are:

10 CFR 50.34, Contents of applications; technical information, paragraph (a), Preliminary safety analysis report, including subparagraphs 50.34(a)(3)(ii), 50.34(a)(3)(iii), and 50.34(a)(4) 10 CFR 50.35, Issuance of construction permits 10 CFR 50.40, Common standards Guidance for this review includes:

NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Parts 1 and 2, Section 9.7, Other Auxiliary Systems.

Technical Evaluation The principal design criterion applicable to the Compressed Air System is:

PDC 2, Design Bases for Protection Against Natural Phenomena The PSAR identifies traits of the auxiliary site services that will enable the SSCs that make up these services to meet applicable regulatory requirements. Implementation of these traits and associated specific design features are not needed for the preliminary design required by a CP. The detailed designs of the site services will be reviewed as part of the OL application.

Kairos states in the PSAR that safety-related SSCs located in proximity to auxiliary site service SSCs are protected from failure of the auxiliary site services during a design basis earthquake by either seismically mounting the applicable auxiliary site service components, physical separation, or barriers

to preclude adverse interactions. The NRC staff finds that this design requirement is consistent with the objective of PDC 2.

Kairos states that the services will be built so that failure of the SSCs making up these services will not interfere with the ability of safety-related SSCs to perform their safety function(s). Because auxiliary systems will be designed to not cause accidents to the reactor, uncontrolled release of radioactivity, or interfere with safe shutdown of the reactor, the NRC staff finds that the preliminary design of the auxiliary site services meet the guidance of NUREG-1537 Part 2, Section 9.7.

Conclusion Kairos has identified traits of the auxiliary site services in PSAR Section 9.8.5. The NRC staff concludes that the traits described in the PSAR demonstrate that the preliminary design is consistent with the objectives of PDC 2 and the guidance of NUREG-1537 Part 2, Section 9.7. Based on the NRC staffs review and the findings above, the NRC staff concludes that Kairos has met the requirements of 10 CFR 50.34(a) for the auxiliary site services because the PSAR provides a preliminary design and a preliminary evaluation of the design and performance of the services.

9.9 Summary and Conclusions for Auxiliary Systems The NRC staff evaluated the descriptions and discussions of the Hermes auxiliary systems as described in PSAR Chapter 9 and finds that the preliminary designs and information on the Hermes auxiliary systems meet the applicable guidelines of NUREG-1537, Part 2, allowing the staff to make findings that:

Kairos preliminary information and commitments to design the reactor coolant auxiliary systems, HVAC systems, pebble handling and storage systems, communication systems, water systems, and other auxiliary systems are sufficient and meet the applicable regulatory requirements and guidance for the issuance of a CP. Further information on these items can reasonably be left for later consideration in the FSAR.

The preliminary information on fire protection systems and programs is sufficient and meets the applicable regulatory requirements and guidance for the issuance of a CP. Further information can reasonably be left for later consideration in the FSAR, fire protection program, and fire hazards analysis submitted with an OL application.

The preliminary design of the Kairos program and auxiliary facilities for the possession and use of byproduct material, source material, and SNM at Hermes is sufficient and meets the applicable regulatory requirements and guidance for the issuance of a CP. Further information related to possession and use of byproduct material, source material, and SNM during operations and decommissioning can reasonably be left for later consideration during future reviews of a Hermes OL application and proposed decommissioning plan, respectively.

Based on these findings referenced above, the staff concludes the following regarding the issuance of a CP in accordance with 10 CFR Part 50:

Kairos has described the proposed facility design criteria for auxiliary systems, including, but not limited to, the principal architectural and engineering criteria for the design, and has identified the major features or components incorporated therein for the protection of the health and safety of the public.

There is reasonable assurance that, taking into consideration the site criteria contained in 10 CFR Part 100, the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public.

There is reasonable assurance: (i) that the construction of the Hermes facility will not endanger the health and safety of the public, and (ii) that construction activities can be conducted in compliance with the Commissions regulations.

Kairos is technically qualified to engage in the construction of its proposed Hermes facility in accordance with the Commissions regulations.

The issuance of a permit for the construction of the Hermes facility would not be inimical to the common defense and security or to the health and safety of the public.

9.10 References for Auxiliary Systems American Society of Mechanical Engineers (ASME) B30.2-2016, Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Hoist), May 2017 Kairos Power LLC, Reactor Coolant for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor Topical Report, KP-TR-005-P-A, Revision 1, July 2020, ADAMS No. ML20219A591

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. Principal Design Criteria for the Kairos Power Fluoride-Salt Cooled, High Temperature Reactor, KP-TR-003-NP-A, Revision 1, June 2020, ADAMS No. ML20167A174

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. Transmittal of Kairos Power CPA Changes, February 2022, ADAMS No. ML22049B555

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. Transmittal of Changes to Hermes Construction Permit Application, February 2022, ADAMS No.

ML22042A095

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. Transmittal of Changes to PSAR Sections 9.1.1 and 9.1.4, June 2022, ADAMS No.

ML22160A689

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. Transmittal of Responses to NRC Requests for Confirmation of Information Hermes Preliminary Safety Analysis Report, Section 9.1, August 2022, ADAMS No. ML22231B228

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. Submittal of the Preliminary Safety Analysis Report for the Kairos Power Fluoride Salt-Cooled, High Temperature Non-Power Reactor (Hermes), Revision 1, September 2022, ADAMS No.

ML22272A593.

U.S. Nuclear Regulatory Commission. NUREG-1537 Part 2, Guidelines for Preparing and Reviewing Application for the Licensing of Non-Power Reactors, Parts 1 and 2. NRC: Washington, D.C. February 1996, ADAMS Accession Nos. ML042430055 and ML042430048.

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. Regulatory Guide 4.20, Constraint on Releases of Airborne Radioactive Materials to the Environment for Licensees Other Than Power Reactors, Revision 1, April 2012

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. General Audit report