ML23285A130

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Official Exhibit - KRS-006-MA-CM01 - Kairos Power Llc'S Responses to Commission'S Prehearing Questions
ML23285A130
Person / Time
Site: Hermes
Issue date: 10/12/2023
From:
Kairos Power
To:
NRC/OCM
SECY RAS
References
Construction Permit Mndtry Hrg, RAS 56821, 50-7513-CP
Download: ML23285A130 (0)
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Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of KAIROS POWER LLC Docket No. 50-7513-CP (Hermes Test Reactor)

Hearing Exhibit Exhibit Number: KRS-006 Exhibit

Title:

Kairos Power LLCS Responses to Commissions Pre-hearing Questions

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION

)

In the Matter of ) Docket No. 50-7513

)

KAIROS POWER LLC )

)

(Kairos Power Hermes Reactor) ) October 1, 2023

)

KAIROS POWER LLCS RESPONSES TO COMMISSIONS PRE-HEARING QUESTIONS Kairos Power LLC (Kairos Power) provides the following responses to the questions in the Commissions September 15, 2023 Order (Transmitting Pre-Hearing Questions) regarding the mandatory hearing for the Construction Permit Application for the Hermes Test Reactor. Kairos Powers responses are limited to those questions directed to it.

Responses to Commission Questions Question 2 10 C.F.R. § 50.35(a) states, in part, that the Commission may issue a construction permit if the Commission finds that (1) the applicant has described the proposed design of the facility, 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 . . . .

Did you identify the major features or components for your design and, if so, what criteria did you use?

Response to Question 2 Yes. Kairos Powers PSAR for the Hermes facility describes the major features and components of the overall plant design. This includes those major features and components that are required to:

  • prevent or mitigate the consequences of a postulated event,
  • meet the principal design criteria (PDC), or

Table 3.6-1 of the PSAR provides a summary listing of the major structures, systems, and components (SSCs) and identifies the safety-related SSCs that are credited with mitigating the effects of postulated events described in PSAR Chapter 13. The maximum hypothetical accident, which bounds the postulated events demonstrates that the dose consequences meet the site dose criteria in 10 CFR Part 100 with significant margin.

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Table 3.6-1 of the PSAR also identifies non-safety related SSCs, in addition to the safety-related SSCs, that are necessary to demonstrate that the Hermes facility meets the PDC described in Section 3.1.1 of the PSAR or to meet 10 CFR § 50.34(a). Additionally, the design bases of these non-safety related SSCs provide assurance that the safety-related SSCs will not be adversely affected by the failure of a non-safety related SSC.

10 CFR § 50.34(a)(1)(i) states:

A description and safety assessment of the site on which the facility is to be located, with appropriate attention to features affecting facility design. Special attention should be directed to the site evaluation factors identified in part 100 of this chapter. The assessment must contain an analysis and evaluation of the major structures, systems and components of the facility which bear significantly on the acceptability of the site under the site evaluation factors identified in part 100 of this chapter...

The SSCs in Table 3.6-1 of the PSAR meet the above regulation by providing a design basis that satisfies the siting dose criteria in 10 CFR Part 100.

By identifying the components that prevent or mitigate the consequences of a postulated event, meet the PDC, or meet 10 CFR § 50.34(a)(1)(i), Kairos Power has met the requirement in 10 CFR

§ 50.35(a) to identify the major features or components in the Hermes facility.

Question 4 PSAR section 2.1.1.2 Boundary and Zone Area Maps, states that, The doses at the EPZ are below the Environmental Protection Agency (EPA) Protective Action Guide (PAG) Manual guidelines for protective action, as recommended by ANSI/ANS-15.16-2015 (R2020), Emergency Planning for Research Reactors and pursuant to Regulatory Guide 2.6, Emergency Planning for Research and Test Reactors and Other Non-Power Production and Utilization Facilities.

ANSI/ANS-15.16-2015 (R2020) specifies in Table 1 - Emergency classes, an actual or projected dose in the plume exposure pathway of 10 mSv (1 rem) TEDE or 50 mSv (5 rem) committed dose equivalent to the thyroid.

The SE, section 2.1.3 (page 2-3), states, The staff verified that the Hermes EPZ size is appropriate and consistent with guidance in ANSI/ANS-15.16-2016 [sic], based on the Hermes preliminary maximum hypothetical accident (MHA) dose calculations in PSAR Chapter 13, which indicate that accident doses at the EPZ boundary (based on the assumption that the EAB and EPZ boundary is 250 m (820 ft) from the reactor) would not exceed EPA protective action guides of 1 rem whole body or 5 rem thyroid.

The EPAs PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents, incorporates the Food and Drug Administrations 2001 guidance to lower the PAG for administration of potassium iodide (KI) to 5 rem (50 millisieverts (mSv)) projected child thyroid dose.

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Given that there is a substantial difference in the projected dose to a childs thyroid versus an adults thyroid, please confirm that accident doses at the EPZ boundary would not exceed EPA protective action guides of 1 rem TEDE or 5 rem projected child thyroid dose.

Response to Question 4 The accident doses at the EPZ boundary will not exceed EPA guidelines for protective action of 1 rem TEDE for sheltering in place and evacuating or 5 rem projected child thyroid dose from exposure to radioiodine for administration of prophylactic drugs (i.e., potassium iodine) (Reference 1).

The Hermes Maximum Hypothetical Accident (MHA) analysis demonstrates that the functional containment provided by the combination of TRISO fuel and Flibe coolant is effective at containing fission products such as iodine. As a result, the MHA analysis dose-driving elements are mobile activation products (e.g., argon, tritium) instead of fission products (e.g., iodine). In the MHA analysis, iodine results in calculated thyroid doses less than 0.001 rem (1 mrem) over the duration of the plume using the dose conversion factors from Federal Guidance Report (FGR) 11 (Reference 2). The ICRP-71 I-131 thyroid dose conversion factor for a one-year-old (Reference 3)

(which the EPA PAGs and the FDA use to define child doses) is approximately a factor of 11 higher than the FGR 11 thyroid dose conversion factor. Thus, the projected child thyroid dose from exposure to radioiodine is under the 5 rem PAG limit for the Hermes MHA.

References:

1. U.S. Environmental Protection Agency, PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents, EPA-400/R-17/001 (Jan. 2017),

https://www.epa.gov/radiation/protective-action-guides-pags

2. U.S. Environmental Protection Agency, Federal Guidance Report No. 11: Limiting Values Of Radionuclide Intake And Air Concentration And Dose Conversion Factors For Inhalation, Submersion, And Ingestion, EPA-520/1-88-020, 1988.
3. International Commission on Radiological Protection, Age Dependent Doses to Members of the Public from Intake of Radionuclides: Part 4 Inhalation Dose Coefficients, ICRP Publication 71, Vol. 25, Nos 3, 1995.

Question 6 According to SE section 2.2.3.1.1, PSAR Section 2.2.2.1 states that the proposed 5,000 ft (1,524 m) long runway of [the Oak Ridge Airport] would be oriented in such a way that the aircraft would not land or take off from this airport on a trajectory over the proposed site.

a. How was the hazard associated with the proposed site lying within an airport traffic pattern zone considered?
b. Because the airport has not yet been built, how did Kairos account for any potential future changes to flight patterns?

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Response to Question 6

a. Given the approximate location of the proposed airport and the proposed number and type of operations described in PSAR Section 2.2.2.1, the Hermes site aircraft crash impact frequency from flight operations at the proposed nearby airport is calculated in accordance with PSAR Equation 2.21 using methodologies outlined in DOE Standard DOESTD 30142006 (Reference 1) for airport operations.
b. An evaluation of crash frequencies related to the proposed Oak Ridge Airport is presented in PSAR Section 2.2.2.3. This evaluation uses the methodologies outlined in DOE Standard DOESTD30142006 for airport operations (Reference 1) in combination with available site-specific data provided in the 2016 DOE environmental assessment (EA) for the proposed location of the Oak Ridge Airport (Reference 2). The calculated aircraft impact frequencies for the proposed airport are provided in PSAR Table 2.2-8 and exceed the DOE-STD-3014-2006 screening criteria of 1.00E-06. As a result, the design bases in the PSAR commits that the safety-related portion of the Reactor Building will be designed to withstand the impact of general aviation aircraft associated with flight operations at the proposed nearby airport. Any configuration of the airport runways is not anticipated to change the need to design for aircraft impact.

References:

1. U.S. Department of Energy, Accident Analysis for Aircraft Crash Into Hazardous Facilities.

DOE Standard DOESTD30142006, Reaffirmation, Washington DC. May 2006.

2. U.S. Department of Energy, Office of Energy, Oak Ridge Office of Environmental Management, Environmental Assessment, Property Transfer to Develop a General Aviation Airport at the East Tennessee Technology Park Heritage Center, Oak Ridge, Tennessee.

DOE/EA2000, FINAL. Oak Ridge, Tennessee. February 2016.

Question 7 The Applicant states that the safety related portion of the reactor building will be designed to withstand the impact of small nonmilitary general aviation aircraft or light general aviation aircraft. What are the characteristics of the aircraft that will establish the aircraft impact design basis for the reactor building?

Response to Question 7 The summary of hazards from aircraft impacts in PSAR Section 2.2.2.4 indicates that a small risk from small non-military general aviation aircraft exist from the potential future nearby airport and two federal airways (J46 and V16).

As stated in PSAR Section 2.2.2.4, The risk from large commercial aviation aircraft is well below the screening criterion and, therefore, the characteristics of the aircraft that will establish the aircraft impact design basis for the reactor building are limited to smaller aircraft.

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The DOE Environmental Assessment (EA) for the Property Transfer to Develop a General Aviation Airport at the East Tennessee Technology Park Heritage Center, Oak Ridge, Tennessee (Reference 1) describes representative aircraft for the proposed nearby Oak Ridge airport. As stated in EA Section 2.1.5.2, The operational fleet mix forecast for the proposed Oak Ridge airport was conducted for three aircraft types: (1) turbine-powered fixed-wing aircraft, (2) helicopters, and (3) piston-powered, fixed-wing aircraft. Table 2.5 of the EA report then indicates that over 90% of projected flight operations for the proposed airport are the lighter piston-powered, fixed-wing, general aviation aircraft (for example, a Cessna 172R (single engine) or Beechcraft Baron 58P (multi-engine).

References:

1. U.S. Department of Energy, Office of Energy, Oak Ridge Office of Environmental Management, Environmental Assessment, Property Transfer to Develop a General Aviation Airport at the East Tennessee Technology Park Heritage Center, Oak Ridge, Tennessee.

DOE/EA2000, FINAL. Oak Ridge, Tennessee. February 2016.

Question 11 What design standards were used to design features protecting safety related SSCs from the seismic failure of nearby non-safety related equipment? Further, were potential adverse seismic impacts from the failure of any non-safety related SSCs assessed to ensure all safety related SSCs are adequately protected during a design basis earthquake?

Response to Question 11 PSAR § 3.5.3, System Evaluation, states that Consistent with PDC 2, the safety-related portion of the reactor building is designed so that it will be able to perform its physical protection safety functions described in Section 3.5.1, even if the non-safety related portion of the reactor building is damaged due to the design basis wind, water or earthquake events described in Sections 3.2, 3.3, and 3.4.

As described in PSAR § 3.4.2, non-safety related SSCs will be designed according to the local building code, the 2012 International Building Code (IBC). For the seismic input, the design basis ground motion is defined in accordance with the deterministic processes of local building code, the 2012 IBC, which refers to ASCE/SEI 7-10. ASCE/SEI 7-10 defines the requirements for general structural design to address earthquake loads. No specific design codes were cited in the PSAR to design non-safety SSCs from affecting safety-related SSCs during a seismic event. The PSAR does not reflect a detailed design for either safety-related or non-safety SSCs.

Kairos Power expects that the adverse impacts resulting from seismically induced failures of non-safety SSCs on safety-related SSCs can be mitigated in the design by either: 1) demonstrating insufficient proximity to cause damage, 2) designing the safety-related SSC to withstand the consequences of an adverse interaction, or 3) designing the non-safety SSC to preclude failure under design basis earthquake loadings.

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The specific design features or mitigation features, and the SSCs to which they are applied, will be provided in the Operating License application as well as an evaluation describing how PDC 2 is satisfied by the selected design solution.

References:

1. PSAR § 3.5.3.

Question 12 The reactivity control and shutdown system (RCSS) appears to operate similarly to pressurized water reactors in that a loss of power de-energizes the rod mechanism, allowing the rods to drop into the core via gravity. SE § 3.6.3.1.3 states SSCs in the RCSS (shutdown elements only) . . . are identified in PSAR Table 3.6-1 as safety related and are identified to be SDC-3 in accordance with ASCE 43-19. Additionally, the most significant overpower transient involves a malfunction of the rod control mechanisms.

a. Why are the rod control mechanisms not also safety related?
b. Please identify what individual components of the RCSS are considered or are expected to be safety related.

Response to Question 12

a. The rod control mechanisms are not safety-related because they do not meet the specific definition of safety-related as defined in Section 1.2.3 of the Hermes PSAR (Reference 1) which states that safety-related structures systems and components (SSCs) are those that are relied upon to remain functional during normal and operating events to assure:
  • The integrity of the portions of the reactor coolant boundary relied upon to maintain coolant level above the active core;
  • The capability to shut down the reactor and maintain it in a safe shutdown condition; or
  • The capability to prevent or mitigate the consequences of accidents which could result in potential exposures exceeding the limits set forth in 10 CFR § 100.11.

The rod control mechanisms are not part of the reactor coolant boundary and are not relied on to maintain coolant level above the active core. The ability to shut down the reactor and maintain it in a safe shutdown condition is met by the shutdown elements alone (assuming one element does not insert), and the rod control mechanisms are not credited for this function. The rod control mechanisms are not credited in the PSAR Chapter 13 safety analysis to prevent or mitigate consequences of accidents which could result in potential exposures exceeding the limits in 10 CFR § 100.11. Even though they are not credited, the control rods would receive a reactor trip signal and would be expected to insert into the core.

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The most significant postulated overpower transient is a malfunction of the rod control mechanisms and is an initiating event. However, the consequences of this event are mitigated without reliance on the rod control mechanisms or elements.

b. The safety-related function of the shutdown elements is to rapidly insert negative reactivity into the reactor core to shut down the reactor. The portions of the RCSS that are safety-related include the reactivity shutdown elements and associated drive mechanism release.

The specific portions of the shutdown elements that are safety related are the metallic structure of the shutdown elements themselves, the boron carbide neutron absorber pellets within the elements, and the metallic capsules that encapsulate the boron carbide pellets within the shutdown elements. The specific portions of the shutdown drive mechanism release that are safety related are the clutch and the sheave that release the shutdown element into the core. The safety-related function of the shutdown element drive mechanism is to release the shutdown elements in response to a reactor trip signal from the reactor protection system to allow rapid insertion of the shutdown elements.

Question 14b In approved topical report KP-TR-004-NP-A, it is noted that use of the modified definition of safety related would require an exemption (e.g., see Table A-3, All Regulatory Requirements in 10 CFR That Require an Exemption for a KP-FHR Power Reactors, which discusses exemptions needed for the 10 C.F.R. § 50.2 definitions as well as other regulatory requirements). However, the SE does not discuss any exemptions for the Hermes construction permit. Kairos identifies additional items in KP-TR-004-NP-A as requiring exemptions from NRC regulations.

b. For the Applicant: Does Kairos intend to request any exemptions from the regulations for the Hermes reactor? If so, are the exemptions applicable at the construction permit or operating license stage?

Response to Question 14b Kairos Power does not intend to request any of the exemptions listed in the Regulatory Analysis for the Kairos Power Fluoride Salt-Cooled, High Temperature Reactor topical report, KP-TR-004-NP-A (Reference 1) for the Hermes reactor. However, Kairos Power is still evaluating whether any exemptions will be required at the Operating License Application (OLA) phase.

Hermes is a non-power test reactor, so Table A-3 (All Regulatory Requirements in 10 CFR That Require an Exemption for a KP-FHR Power Reactors) in Regulatory Analysis for the Kairos Power Fluoride Salt-Cooled, High Temperature Reactor topical report, KP-TR-004-NP-A (Reference 1) does not apply. Table D-3 Design Regulatory Requirements in 10 CFR That Require an Exemption for a KP-FHR Test Reactor" could apply to Hermes. However, the Hermes Preliminary Safety Analysis Report (PSAR) Revision 3 (Reference 2) departs from the exemptions identified for a test reactor in Table D-3.

There are 6 regulations listed in Table D-3 of the topical report (Reference 1) as requiring exemption for a KP-FHR test reactor:

  • 10 CFR 50.2 - Definition of Safety-Related Structures, Systems, and Components
  • 10 CFR 50.36(c)(2)(ii)(A) - Reactor Coolant Pressure Boundary After further review and conversations with the NRC staff during the CPA review, Kairos Power determined that the definitions of reactor coolant pressure boundary and safety-related structures, systems, and components did not require an exemption because the statements of consideration for these definitions in 10 CFR 50.2 made it clear that they were specific to light water reactor technology. Therefore, no exemptions are necessary for the two 10 CFR § 50.2 definitions listed in Table D-3. PSAR Section 1.2.3 identifies the definition of safety-related that is used for the Hermes test reactor. This also obviates the need for an exemption to the reactor coolant pressure boundary wording in 10 CFR 50.36(c)(2)(ii)(A).

Kairos Power submitted Enclosure 3 to Reference 3 which included a request for exemptions to 10 CFR § 50.34(a)(4) for Hermes at the CPA stage. Since the same regulation is repeated for the FSAR stage, 10 CFR § 50.34(b)(4) was also included in the exemption request. However, after further review and discussion with NRC staff during the CPA review, the exemptions identified within Table D-3 of Reference 1 were determined to be not technically relevant and therefore, do not require an exemption. This discussion and justification was provided in Enclosure 1 to Reference 4. As described in the Reference 4 submittal, the requirements in the last sentence of both 10 CFR § 50.34(a)(4) and 10 CFR § 50.34(b)(4) were concluded to be not technically relevant to the KPFHR design and the requested evaluations are not provided in the PSAR nor planned for the FSAR.

The remaining regulation in Table D-3 is 10 CFR § 50.34(b)(9), which was not discussed in the PSAR because the regulations in Section 50.34(b) are applicable to the Final Safety Analysis Report, which Kairos Power will provide with the Operating License Application. However, after further review, Kairos Power determined that the regulation is not technically relevant to KP-FHR technology. KP-FHRs, and Hermes specifically, operate at near atmospheric pressures, and therefore cannot experience pressurized thermal shock phenomena. Therefore, Kairos Power does not intend to request an exemption to this regulation at the Operating License Application phase.

Kairos Power will evaluate whether any exemptions are needed during the OLA phase. If any exemptions are identified, they will be included with the OLA submittal.

References:

1. Kairos Power LLC, Regulatory Analysis for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor (Topical Report), KP-TR-004- NP-A, rev. 4, Jan. 2022 (ML22159A358)
2. Kairos Power LLC, Submittal of the Preliminary Safety Analysis Report for the Kairos Power Fluoride Salt-Cooled, High Temperature Non-Power Reactor (Hermes), Revision 3, May 31, 2023 (ML23151A744) 8
3. Kairos Power LLC, Submittal of the Preliminary Safety Analysis Report for the Kairos Power Fluoride Salt-Cooled, High Temperature Non-Power Reactor (Hermes), September 29, 2021 (ML21272A376)
4. Kairos Power LLC, Transmittal of Changes to Construction Permit Application -

Exemptions Enclosure, PSAR Chapter 1, PSAR Chapter 3, and PSAR Chapter 14, September 19, 2022 (ML22263A035)

Question 15 SE section 4.2.1.1 states:

PSAR Section 4.2.1.1 states that, in addition to the fuel pebbles, the reactor also contains moderator pebbles. The moderator pebbles have the same diameter as the fuel pebbles but contain no uranium and are made of graphite material. The graphite pebbles are non-safety related and serve to provide sufficient moderation for the thermal spectrum Hermes reactor. Similar to the fuel pebbles, the moderator pebbles are designed to maintain positive buoyancy under normal operation and postulated events.

a. Why are the graphite moderator pebbles not considered safety related? In particular, do the design bases for the Hermes reactor include any assumptions regarding the physical characteristics or reactivity influence of the moderator pellets? If so, could a deviation from these design basis assumptions have an adverse impact on the capability to shut down the reactor and maintain it in a safe shutdown condition?
b. Could failure of one or more moderator pellets adversely impact a safety related function (e.g., the thermophysical properties of the Flibe coolant needed for natural circulation heat transfer chemistry)?

Response to Question 15

a. The moderator pebbles are not safety-related because they do not meet the specific definition of safety-related as defined in PSAR Section 1.2.3 (Reference 1) which states:

The SSCs in the facility are assigned a nuclear safety classification, as follows:

Safety-related SSCs: Those SSCs that are relied upon to remain functional during normal operating conditions and during and following design basis events to assure:

The integrity of the portions of the reactor coolant boundary relied upon to maintain coolant level above the active core.

The capability to shut down the reactor and maintain it in a safe shutdown condition; or The capability to prevent or mitigate the consequences of accidents which could result in potential exposures exceeding the limits set forth in 10 CFR § 100.11.

Non-safety related: Those SSCs that are not in the above safety classification.

No assumptions are made regarding the physical characteristics or reactivity influence of the moderator pebbles. Structural failure of moderator pebbles, like fuel pebbles, is precluded by pebble design and qualification testing. Qualification of the moderator pebble 9

is performed in accordance with the methodology described in KP-TR-011-NP-A (Reference 2). Regardless, a hypothetical loss of moderating properties by the moderator pebbles would cause negative reactivity and would not have an adverse impact on the ability of the reactor to shut down or maintain a safe shutdown state.

b. A failure of moderator pebbles would not adversely impact a safety related function. A statistically significant number of failed moderator pebbles could hypothetically result in debris that reduces the effectiveness of the natural circulation flow path and could increase forces on the shutdown elements in the pebble bed during insertion. For this reason, the moderator pebbles, like the fuel pebbles, are designed to maintain integrity and are subject to testing using the same methodology described in the Fuel Qualification Methodology for the Kairos Power Fluoride Salt-cooled High Temperature Reactor (KP-FHR) Topical Report, KP-TR-011-NP-A (Reference 2). Moderator pebble material is also not chemically reactive with the Flibe coolant and does not chemically impact the thermophysical properties of the Flibe.

All pebbles including the moderator pebbles will be inspected by the pebble handling and storage system (PHSS) as described in Section 4.2.1 of the PSAR (Reference 1). Moderator pebbles with indications of wear, cracking, or missing surfaces will be removed from service. Both the fuel and moderator pebbles are therefore not expected to produce debris or dust in the reactor coolant that could inhibit the removal of heat from the core.

References:

1. Kairos Power LLC, Submittal of the Preliminary Safety Analysis Report for the Kairos Power Fluoride Salt-Cooled, High Temperature Non-Power Reactor (Hermes), Revision 3 May 31, 2023 (ML23151A744)
2. Kairos Power LLC, Fuel Qualification Methodology for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor (KP-FHR), Revision 2 (Non-Proprietary)

(ML23089A398).

Question 16 Potential fouling or plugging of the heat exchanger in the non-safety related primary heat transport system (PHTS) is to be monitored via observing downcomer and core temperatures (see SE § 5.1.3.2.6 at 5-6).

a. Will potential fouling or plugging of the safety related decay heat removal system (DHRS) be monitored in a separate, distinguishable way?
b. If the DHRS and PHTS are monitored via the same or similar parameters, how is plugging or fouling of the DHRS distinguishable such that it would not be masked by strong or efficient performance of the PHTS?

Response to Question 16 10

a. Kairos Power provided the NRC staff information regarding qualification testing of the DHRS during the audit of PSAR Section 6.3, which included testing the potential for evaporator tube fouling (Reference 1). Specific parameters that are monitored to ensure system integrity and capability of the DHRS to perform its safety function will be addressed in the application for an Operating License. Kairos Power has not requested the NRC staff to make a final determination regarding the acceptability of the DHRS design as part of the construction permit application. There is a commitment in the PSAR to ensure DHRS operability will be addressed by the technical specifications (PSAR Section 6.3.4). The surveillance requirements for the DHRS in the technical specifications are expected to account for degraded performance such as fouling or plugging.
b. Parameters used to monitor DHRS performance may be influenced by the performance of the PHTS. For example, degraded performance of the DHRS could be detected by means such as long-term monitoring of makeup water consumption rates, which could be affected by PHTS performance. However, other indicators of PHTS performance, such as vessel temperature and heat rejection radiator removal rate, would not be significantly impacted by DHRS performance during normal operations, and can therefore be used to decouple PHTS and DHRS performance. This level of design detail was not provided in the PSAR since Kairos Power did not request the NRC staff to make a final safety determination of the DHRS design. Specific parameters that are monitored for degraded performance will be provided with the Operating License application.

References:

1. Kairos Power, LLC, "Transmittal of Response to NRC Question on DHRS Testing from PSAR Section 6.3 Audit on Hermes Preliminary Safety Analysis Report," KP-NRC-2209-003, September 2022. ML22244A236.
2. Kairos Power, LLC, "Hermes Non-Power Reactor Preliminary Safety Analysis Report,"

HER-PSAR-001, Revision 3. May 2023. ML23151A744.

Question 17 PSAR section 9.1.3 states that the Tritium Management System (TMS) does the following: (1) provides tritium separation from argon in the inert gas system (IGS), (2) provides tritium separation from air in the reactor building cells, and (3) provides final collection and disposal of tritium.

PSAR section 9.1.3.2, Design Bases, states: Consistent with PDC 13, proper instrumentation is provided to measure tritium inventories in the TMS and demonstrate compliance with imposed inventory limits. (Emphasis added).

PSAR Table 3.6.-1, Structures, Systems, and Components, indicates that the TMS and the Inventory Management System are classified as non-safety related.

Section 9.1.3.3 of the SE states that:

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 11

releases from the TMS are bounded by the MHA and would therefore meet the accident dose criteria in 10 CFR 100.11.

The SE then states that [t]he 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.

a. Please provide additional information describing the term proper instrumentation in PSAR section 9.1.3.2.
b. In addition, please provide additional information describing the basis for the designation of the instrumentation provided to measure tritium inventories in the TMS as non-safety related.

Response to Question 17

a. In the context of PSAR Section 9.1.3, proper instrumentation means that the instrumentation provided in the TMS will be designed to ensure the capability of measuring tritium inventory levels for comparison with limits derived from the MHA. The summary description of this instrumentation in the PSAR is consistent with the high-level requirements in 10 CFR § 50.34(a) for a preliminary safety analysis report. The specific instrumentation of the TMS and its capabilities to measure tritium will be discussed in further detail, consistent with the requirements in 10 CFR § 50.34(b) for a final safety analysis report, in the application for an Operating License.
b. The instrumentation used to measure the tritium activity for conformance with the technical specification limits is not safety-related because it does not meet the specific definition of safety-related as defined in PSAR Section 1.2.3 which states:

The SSCs in the facility are assigned a nuclear safety classification, as follows:

o Safety-related SSCs: Those SSCs that are relied upon to remain functional during normal operating conditions and during and following design basis events to assure:

o The integrity of the portions of the reactor coolant boundary relied upon to maintain coolant level above the active core.

o The capability to shut down the reactor and maintain it in a safe shutdown condition; or o The capability to prevent or mitigate the consequences of accidents which could result in potential exposures exceeding the limits set forth in 10 CFR 100.11.

o Non-safety related: Those SSCs that are not in the above safety classification.

The TMS instrumentation is not relied upon to maintain any portion of the reactor coolant boundary nor to maintain coolant level above the active core. The TMS instrumentation has no shut down capabilities nor functions and is not relied on to maintain a safe shutdown condition. The TMS instrumentation is not relied upon to prevent or mitigate the 12

consequences of accidents which could result in potential exposures exceeding the limits set forth in 10 CFR 100.11.

The TMS instrumentation is the means by which the tritium inventory is measured and compared with limits derived from the MHA to provide assurance that the plant is operated consistent with the assumptions in the safety analysis. A loss of this capability would be controlled by technical specification action statements provided with the Operating License application. Action statements for measured tritium levels that are above allowable limits derived from the MHA could include reducing tritium inventory by replacing active capture beds that are at or near capacity with fresh capture beds and packaging and dispositioning the used capture beds for shipment and offsite disposal as described in PSAR Section 9.1.3.

Question 18 The Staff proposes to include a condition related to Kaiross quality assurance program.

a. Does 10 C.F.R. § 50.55(f)(3) apply to the Hermes reactor?
b. Section 12.9.4 of the SE states that the staff concludes that the information in PSAR Section 12.9 and PSAR Appendix 12B is sufficient and meets the applicable guidance and regulatory requirements identified in this section for the issuance of a construction permit in accordance with 10 CFR 50.35 and 50.40, and, as such, the Hermes QAPD is acceptable for implementation during the design and construction of the Hermes facility. Given this conclusion, please discuss why the license condition is necessary to provide reasonable assurance that regulatory requirements and license commitments for QA are adequately included in the design, procurement, and construction of the Hermes facility.

Response to Question 18

a. As described in Table E-2 (NonDesign Regulatory Requirements in 10 CFR That Do Not Apply to a KPFHR Test Reactor) of KP-TR-004-NP-A, (Regulatory Analysis for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor) (Reference 6), 10 CFR § 50.55(f) does not apply to this Hermes Reactor. 10 CFR § 50.55(f)(3) applies to construction permit holders described in 10 CFR § 50.55(f)(1). 10 CFR § 50.55(f)(1) describes nuclear power plant and fuel reprocessing plant construction permit holders subject to the quality assurance criteria in 10 CFR 50 § Appendix B. This Hermes Reactor is a non-power reactor and is not subject to 10 CFR 50 § Appendix B. Therefore 10 CFR § 50.55(f)(3) is not applicable.
b. The license condition proposed by the staff and described in Section 12.9.4 of the SE is necessary to provide reasonable assurance that regulatory requirements and license commitments for QA are adequately included in the design, procurement, and construction of the Hermes facility because the conditions for construction permits described in 10 CFR

§ 50.55(f)(1) and (f)(3) apply only to nuclear power reactors and fuel reprocessing plants.

10 CFR § 50.55(f)(1) requires nuclear power reactors and fuel reprocessing plants to implement, pursuant to 10 CFR § 50.34(a)(7), the quality assurance program described or referenced in the Safety Analysis Report, including changes to that report. 10 CFR § 13

50.55(f)(3) provides an allowance for construction permit holders for nuclear power reactors and fuel reprocessing plants to make changes to the approved QAPD that do not reduce the commitments in the program description previously accepted by the NRC and to submit those changes to the NRC within 90 days.

By including this proposed license condition, the subject Construction Permit will have an explicit requirement comparable to 10 CFR § 50.55(f)(1) to implement the QA program found acceptable by the NRC in Section 12.9.4 of the SE, including revisions, and an allowance to make changes to the program that do not reduce commitments without prior NRC approval consistent with the allowance of 10 CFR § 50.55(f)(3). Absent the license condition, any changes to the quality assurance plan would require the submittal and approval of a license amendment irrespective of whether the change reduces a commitment or not.

References:

1. 10 CFR § 50.34(a)(7)
2. 10 CFR § 50.55(f)(1)
3. 10 CFR § 50.55(f)(3)
4. 10 CFR 50, Appendix B
5. SE § 12.9.4 (p. 12-21 to 12-22)
6. Kairos Power LLC, Regulatory Analysis for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor (Topical Report), KP-TR-004- NP-A, rev. 4, Jan. 2022 (ML22159A358)

Question 19 Section 3.6.2, Tribology, in Fuel Qualification Methodology for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor, Rev. 2, states that tribological testing will be informed by the corresponding American Society for Testing and Materials (ASTM) Standard ASTM G99-17, Standard Test Method for Wear Testing with a Pin-on-Disk Apparatus, where the contact surface will be immersed in argon gas or in molten Flibe with a controlled argon atmosphere representative of KP-FHR.

SE section 13.1.5.3, Technical Evaluation, indicates tribology testing will be conducted in Flibe and argon. Section 3.6.2 of the fuel qualification program indicates the coefficient of friction will be measured in both Flibe and argon, but the wear rates will be measured in Flibe or argon.

Is erosion of the fuel pebbles expected to be higher in Flibe than in argon? If so, why will wear rates be measured in Flibe or in argon and not both?

Response to Question 19 Tribology testing will be conducted on fuel pebbles in two separate environments: 1) Flibe with an argon cover gas, and 2) an argon only environment. The coefficient of friction and wear rates will be determined during tribology tests in both of these environments. This was the intent of the second sentence of Section 3.6.2 of KP-TR-011-NP-A (Reference 1). The erosion of the fuel pebbles is expected to be slightly higher in Flibe than in argon however, testing will show the 14

behavior of fuel pebbles in both environments.

References:

1. Kairos Power LLC, Fuel Qualification Methodology for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor (KP-FHR), KP-TR-011-NP-A, Revision 2 (June 2022),

Non-Proprietary, (ML23089A398).

Question 20 The Staff SE, section 13.2.1.3, Technical Evaluation, states:

Transport of the radionuclides within a fuel transport group is based on the transport of a representative element which has a complete set of diffusion information (i.e., diffusivities in the kernel, coating layers, and matrix). Diffusivity information is currently available for four elements (Cs, Sr, Ag, Kr). The radionuclides are assumed to be retained completely in the TRISO particle, or completely released depending on the element class, as described in the MST TR.

Given the importance of the isotopes of iodine and cesium for short term and long term dose consequences respectively, please provide additional information describing how these radionuclides are accounted for in the maximum hypothetical accident (MHA) dose consequence analysis.

Response to Question 20 The Hermes MHA demonstrates that the functional containment provided by the combination of TRISO fuel and Flibe coolant is effective at containing fission products such as cesium and iodine.

As a result, the MHAs dose driving elements are mobile activation products (e.g., argon, tritium) that both originate from and are held up outside of the TRISO fuel, instead of fission products.

While cesium and iodine are not consequence-driving isotopes in the Hermes MHA, their transport and release from the functional containment is modeled per the NRC staff-approved Kairos Power Mechanistic Source Term Topical Report (KP-TR-012-P-A) (Reference 1). KP-TR-012-P defines the radionuclide grouping structures and representative elements/species which are then used to model the transport of a wider set of elements/species.

  • Transport out of the kernel, PyC, and SiC layers is modeled using diffusion relationships with cesium transporting with the cesium diffusion group and iodine transporting with the krypton diffusion group. The rate of diffusion out of the fuel is determined from the MHA time-temperature curve. Radionuclides that leave the TRISO join the appropriate radionuclide groups in the Flibe.
  • Transport out of the Flibe is modeled through evaporation, with both cesium and iodine evaporating with the cesium fluoride transport group. The rate of evaporation from the Flibe is primarily determined from the MHA time-temperature curve. Radionuclides that leave the Flibe join the appropriate radionuclide groups in the cover gas and reactor building.
  • Transport out of the cover gas and reactor building is modeled as either gas or aerosol 15

transport with both cesium and iodine transporting with the aerosol transport group using no-holdup unfiltered release rates from the cover gas and two-hour holdup unfiltered release rates from the reactor building. Aerosols are subject to settling using the Henry correlation while inside the reactor building. Radionuclides that leave the reactor building joins the appropriate radionuclide groups for off-site transport to the Exclusion Area Boundary and Low Population Zone.

  • Off-site transport to the Exclusion Area Boundary and Low Population Zone is modeled as non-depositing and non-decaying gases. Site-specific weather data is used to calculate dispersion over source values (i.e., /Q values) to evaluate the radionuclide concentrations and associated doses of the maximally exposed individual at the Exclusion Area Boundary and Low Population Zone.

References:

1. Kairos Power, LLC, KP-FHR Mechanistic Source Term Methodology, KP-TR-012-P-A, May 2022.

Question 22c On June 3, 2023, President Biden signed into law the Fiscal Responsibility Act of 2023 (the Act).

Section 321 of the Act included amendments to NEPA. Congress did not include any delay in the effective date of these new amendments to NEPA; accordingly, these amendments became applicable to the NRC upon enactment.

a. For the Staff: Do the FEIS and the Staffs initial testimony account for these amendments to NEPA?

Section 321(a) of the Act amended NEPA section 102(2), which outlines an agencys NEPA responsibilities and imposes requirements for the preparation of EISs. As part of the findings necessary to support issuance of this construction permit, we must determine whether the requirements of NEPA sections 102(2)(A), (C), and (E) have been met.

b. For the Staff: Does the Commission need to consider any other information outside of what was presented in the FEIS to make required findings under NEPA?
c. For the Applicant: Does the Applicant have any views it wishes us to consider regarding the environmental findings we must make in light of these recent amendments to NEPA?

Response to Question 22c Kairos Power believes the cited amendments to NEPA from Section 321(a) of the Fiscal Responsibility Act of 2023 are adequately addressed in the NRC staffs FEIS.

16

CERTIFICATION AND DECLARATION OF WITNESS I certify that Kairos Power LLCs responses to the Commissions questions were prepared by me or under my direction. I declare under penalty of perjury that the foregoing testimony is true and correct to the best of my information, knowledge, and belief.

Executed on September 28, 2023.

Respectfully submitted, Executed in Accord with 10 C.F.R. § 2.304(d)

Signed by Peter Hastings Kairos Power LLC - Vice President, Regulatory Affairs & Quality 2115 Rexford Road, Suite 325 Charlotte, NC 28210 Phone: 704.336.9596 Email: hastings@kairospower.com

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