ML23285A150

Official Exhibit - NRC-004-MA-CM01 - NRC Staff Responses to Commission Pre-Hearing Questions (September 28, 2023)

ML23285A150
Person / Time
Site:	Hermes
Issue date:	10/12/2023
From:	NRC/OGC
To:	NRC/OCM
SECY RAS
References
Construction Permit Mndtry Hrg, RAS 56823, 50-7513-CP
Download: ML23285A150 (0)

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: NRC-004 Exhibit

Title:

NRC Staff Responses to Commission Pre-Hearing Questions (September 28, 2023) - ML23271A250

NRC Staff Exhibit NRC-004 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION In the Matter of KAIROS POWER LLC Docket No. 50-7513-CP (Hermes Test Reactor)

NRC STAFF RESPONSES TO COMMISSION PRE-HEARING QUESTIONS Pursuant to the Commissions Order (Transmitting Pre-Hearing Questions) of September 15, 2023, the staff of the U.S. Nuclear Regulatory Commission hereby responds to the questions posed in that Order. These questions generally pertain to subjects discussed in the staffs final safety evaluation (SE) 1 or final environmental impact statement (FEIS). 2 The Commissions Order directed some questions only to the staff, some only to the applicant, and some to both the staff and the applicant. The attachment to this filing presents the staffs responses.

1 Safety Evaluation Related to the Kairos Power LLC Construction Permit Application for the Hermes Test Reactor (June 13, 2023) (Agencywide Documents Access and Management System (ADAMS) No. ML23158A268).

2 Environmental Impact Statement for the Construction Permit Application for the Kairos Hermes Test Reactor (August 17, 2023) (ADAMS No. ML23214A269).

Respectfully submitted,

/Signed (electronically) by/

Megan Wright Counsel for NRC Staff Mail Stop: O-14-A44 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Telephone: (516) 765-6523 E-mail: megan.wright@nrc.gov Executed in accord with 10 CFR 2.304(d)

Anita Ghosh Naber Counsel for NRC Staff Mail Stop: O-14-A44 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Telephone: (301) 415-0764 E-mail: anita.ghoshnaber@nrc.gov Blake Vaisey Counsel for NRC Staff Mail Stop: O-14-A44 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Telephone: (301) 415-0995 E-mail: blake.vaisey@nrc.gov Dated at Lewes, Delaware this 28th day of September, 2023

NRC STAFF RESPONSES TO COMMISSION PRE-HEARING QUESTIONS

1. On page ii of the SE, the Staff states:

On the basis of its review of the construction permit application, the staff has determined that the preliminary design and analysis of the Hermes test reactor, including the principal design criteria; design bases; information relative to materials of construction and general arrangement; and preliminary analysis and evaluation of the design and performance of structures, systems, and components of the facility: (1) provides reasonable assurance that the final design will conform to the design basis; (2) includes an adequate margin of safety; (3) describes the structures, systems, and components which will provide for the prevention of accidents and the mitigation of consequences of accidents; and (4) meets applicable regulatory requirements and satisfies applicable NRC guidance.

These statements appear to go beyond the criteria specified in 10 C.F.R. §§ 50.34 and 50.35. As noted in SE section 13, Kairos is not requesting Commission approval of the safety of any design feature or specification in the construction permit (CP) application, as permitted by 10 CFR 50.35(b).

What is the regulatory basis for the conclusions stated on page ii of the SE and does this conclusion impose any limitations or constraints on the Staff at the operating licensing review stage?

Staff Response: The conclusions stated on page ii of the SE (the abstract) summarize findings from the Staffs SE based on information Kairos provided in accordance with the requirements of 10 C.F.R. §§ 50.34(a)(3), specifically 50.34(a)(3)(iii), and 50.34(a)(4).

The Staff made these determinations to support its conclusions that regulatory requirements, including 10 C.F.R. §§ 50.35 and 50.40, are met, and to support its recommendation that the Commission make the necessary findings for the issuance of a CP in accordance with 10 C.F.R.

§§ 50.35, 50.40, and 50.50.

In SE section 1.2, the Staff notes that its review and findings are based on the preliminary design of the facility and further technical information or design information required to complete the safety analysis in support of operation can reasonably be left for later consideration in the final safety analysis report (FSAR). The Staff further emphasizes the limitations of its approval at the CP stage in SE section 1.1.2:

The CP, if issued, would constitute an authorization for Kairos to proceed with construction. The staffs evaluation of the preliminary design and analysis of the Hermes facility does not constitute approval of the safety of any design feature or specification. Such approval will be made following the evaluation of the final design of the facility, as described in the FSAR as part of Kaiross operating license (OL) application for Hermes.

Based on the discussion in SE Sections 1.1.2 and 1.2, the Staff SE, including the abstract, does not impose any limitations or constraints on the Staff at the operating license review stage and 1

appropriately reflects the provisions and criteria from 10 C.F.R. §§ 50.34 and 50.35.

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?

Staff Response: For applicant only.

3. Are any notifications to agencies or bodies under 10 C.F.R. § 50.50 required prior to issuance of the construction permit? If so, have they been completed?

Staff Response: The NRC is required to make certain notifications to agencies or bodies under 10 C.F.R. § 50.50 prior to issuance of the Hermes CP. For example, 10 C.F.R. Part 2 specifies notifications that are required in conjunction with the NRCs receipt and review of a testing facility construction permit application, such as notices of hearing required by 10 C.F.R. § 2.104.

10 C.F.R. Part 51 also requires various notifications to be made in relation to the environmental review and environmental impact statement (EIS) for a testing facility construction permit application. The Staff has determined that all notifications that are required prior to the issuance of a Hermes CP have been duly made.

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-15.162015 (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-15.162015 (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 23),

states, The staff verified that the Hermes EPZ size is appropriate and consistent with guidance in ANSI/ANS-15.162016 [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. 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.

Staff Response: In response to this question, the Staff performed independent calculations to confirm that accident doses at the emergency planning zone (EPZ) boundary would not exceed 2

EPA protective action guides of 1 rem total effective dose equivalent (TEDE) or 5 rem projected child thyroid dose. Specifically, the Staff adjusted the MHA dose results to roughly estimate the child thyroid dose and TEDE. Notably, the exclusion area boundary (EAB) dose results in the PSAR are reported for exposure to the initial 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the release, consistent with the requirements in 10 C.F.R. § 100.11, whereas the EPA PAGs would be compared to a 4-day projected dose at the time of the decision. The MHA dose results at the outer boundary of the low population zone (LPZ) are calculated for 30 days, but at 800 meters instead of the site boundary. While information was not provided in the PSAR, a more precise result for the 4-day period can be calculated using the actual release rates assumed in the applicants MHA considering changes over time for atmospheric dispersion factors. Adjusting the MHA thyroid dose at the LPZ to account for the difference in child thyroid dose conversion factors for inhalation from International Commission on Radiological Protection (ICRP) Publication 60 (ICRP-60) 1, child breathing rate from ICRP-66 2, and the EAB atmospheric dispersion factors, the estimated adjusted EAB 30-day dose results would range from 0.8 to 2.0 rem child thyroidadj,30. Taking the MHA LPZ whole-body dose and combining it with the organ-weighted thyroid dose to give an approximate 30-day TEDE at the LPZ and adjusting for the EAB atmospheric dispersion factors gives a range of 0.26 to 0.64 rem TEDEadj,30 for the approximate EAB 30-day TEDE. These adjusted 30-day dose results are bounding for the 4-day dose at the EAB and confirm that the MHA doses at the boundary of the EPZ, which is coincident with the EAB, would not exceed the early phase PAG values in the 2017 PAG Manual.

The Staff maintains its findings in SE section 2.1.3 that the PSAR contains sufficient information to assess the preliminary description of the EPZ to support the issuance of the Hermes CP. The Staffs additional calculations also 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.

However, the question raises an important point regarding how the Hermes plume exposure pathway EPZ accounted for sensitive populations. The Staff has considered sensitive populations through its guidance on non-power reactor EPZs. The guidance on non-power reactor emergency planning given in ANSI/ANS-15.162015 (R2020), which is endorsed in NRC Regulatory Guide (RG) 2.6 3, as well as the 2017 EPA PAG Manual address considerations for sensitive populations in emergency planning. The 1975 EPA PAG manual PAGs for the general population were 1 to 5 rem whole-body dose and 5 to 25 rem thyroid dose. Consistent with these PAGs, the ANSI/ANS-15.162015 (R2020) guidance for non-power reactor emergency planning zones uses the lower end of this range (i.e., 1 rem whole-body or 5 rem thyroid dose).

The 1975 PAG Manual, section 2.3, Interpretation of PAGs, states that Consideration has been made of the higher sensitivity of children and pregnant women and the need to protect all members of the public. That is, even though the PAG is defined as the dose to the reference man or designated individual, the PAGs are applicable to the general population, including sensitive populations, by use of the lower end of the range. Although the PAGs have since been updated, as analyzed in Appendix C to the 1992 PAG Manual, the old PAGs provide the same level of protection as the new PAGs. The PAG basis for nuclear power plants was carried forward to the 2017 PAG Manual.

1 ICRP, 1991. 1990 Recommendations of the International Commission on Radiological Protection. ICRP Publication 60. Ann. ICRP 21 (13).

2 ICRP, 1994. Human Respiratory Tract Model for Radiological Protection. ICRP Publication 66.

Ann. ICRP 24 (13).

3 RG 2.6, Emergency Planning for Research and Test Reactors and Other Non-Power Production and Utilization Facilities, Revision 2, 2017 (ML17263A472).

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The 2017 PAG Manual, section 2.2.3, Considerations for Potassium Iodide (KI), states that KI is a supplemental action, secondary to evacuation or sheltering. It should not be used as a substitute for evacuation or sheltering-in-place. Many communities do not use KI. The prior 1992 PAG Manual did include, through a footnote to the early phase PAG table, a PAG for thyroid dose-based evacuation. As stated in 2017 PAG Manual, section 2.2.1, Thyroid Based Evacuation, with regard to sensitive subpopulations and thyroid:

The former range recommended for thyroid dose-based evacuation (5 to 25 rem adult thyroid dose) is well covered by projections of whole-body dose, with evacuation recommended at 1 to 5 rem (10 to 50 mSv) adult TED [total effective dose]. The conservatism built into the PAG levels when they were set results in an appropriate level of dose avoidance for the whole community, including all age groups, for an emergency.

ANSI/ANS-15.162015 (R2020) doses for determining the EPZ are adult doses, including thyroid, and equate to the lower end of the PAG range which provides consideration for sensitive populations. Although KI is a possible protective measure in the event of a radiological emergency at a nuclear power plant, the dose criterion for administration is not a decisive factor in determining plume exposure pathway EPZ size. The NRC and EPA in NUREG-0396 4 concluded that thyroid dose, other organ dose, and external whole-body dose showed similar characteristics when evaluated over the same distances. The NRC does not require the use of KI. The use and dosage of KI by the public in the unlikely event of a severe nuclear reactor accident is the responsibility of offsite response organizations. Therefore, the Staff would not use child thyroid dose for EPZ size determination.

With respect to the Hermes MHA radiological consequence analyses, SE section 13.2 states that the Staff will confirm that the final design conforms to the design basis during the evaluation of the FSAR as part of the operating license application review. Similarly, the Staff will review the final justification of the Hermes EPZ size based on more detailed plant design information in its review of the operating license application.

5. In PSAR section 2.1.2, the Applicant states that population data from the 2010 census was used. However, in SE section 2.1.3, the Staff used the more recent 2020 census data in assessing the population center distance. Did the Staff also use the more recent 2020 census data when assessing other demographic information provided by Kairos in PSAR section 2.1.2 (which was based on the 2010 census)? For example, are the Staffs conclusions based on demographic information documented in SE section 2.1.3 (such as Kaiross data indicate that current and future projected populations within 1 mi (1.6 km) of the proposed site are very small) and the population information provided in PSAR figures 2.14 and 2.15 consistent with 2020 census data?

Staff Response: In assessing demographic information provided by Kairos in PSAR section 2.1.2, the Staff used the more recent 2020 Census data by considering the differences between the 2010 and 2020 Census data. The Staffs findings in SE section 2.1.3, which are based on information provided in the PSAR and documented in that SE section, are consistent with 2020 4 NUREG-0396 (EPA 520/178-016), Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light-Water Nuclear Power Plants, December 1978 (ML051390356).

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Census data in that the updated data in the 2020 Census do not affect the Staffs findings in SE section 2.1.3.

The proposed Hermes site is located in the City of Oak Ridge in Roane County, Tennessee, within the East Tennessee Technology Park. A small portion of Morgan County, Tennessee, is also within 5 miles of the proposed site. Only nine residents lived in that portion of Morgan County (based on 2010 Census data), and the population is not expected to change significantly over time (see PSAR Table 2.12). Significantly more people (7,487 based on 2010 Census data) reside in Roane County within 5 mi of the site; however, the population of this portion of Roane County is expected to continue to gradually decrease over time (see PSAR Table 2.11).

The Staff notes that, based on a comparison of the 2020 Census data with the 2010 Census data, the population in Roane County decreased from 54,173 in 2010 to 53,841 in 2020.

Because Census data indicate only small differences in populations within 5 miles of the proposed site between 2010 and 2020, and also based on the gradual downward trends in population within 5 miles of the site, the Staff determined that the information in PSAR section 2.1.2 (including PSAR figures 2.14 and 2.15), which is based on 2010 Census data, is sufficient to adequately characterize the population distribution near the proposed Hermes site for the purposes of assessments of the potential radiological impact on the public, as discussed in SE section 2.1.3.

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?

Staff Response: For applicant only.

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?

Staff Response: In PSAR section 3.5.3.4, Kairos states that it analyzed the impact response on the safety-related portion of the Hermes reactor building using aircraft models representative of the projected types of aircraft that will be used at the Oak Ridge Airport that is proposed to be constructed near the Hermes facility. The Staff reviewed the Environmental Assessment (EA) developed by the Department of Energy for the future Oak Ridge Airport, which Kairos references in the PSAR as providing operation forecasts for this proposed airport. As discussed in this EA and in SE section 2.2.3.1.1, the Beechcraft King Air 350i would be the most common aircraft type to use the proposed Oak Ridge Airport. Based on the characteristics (for example, typical weight and airspeed) of this aircraft compared to other aircraft types expected to use the airport, as well as the expected number of airport operations involving this aircraft relative to other types, the Staff expects that the Beechcraft King Air 350i likely bounds the other general aviation aircraft that would be expected to use the Oak Ridge Airport. Therefore, the Staff expects that aircraft characteristics similar to those of the Beechcraft King Air 350i may be used in establishing the aircraft impact design basis for the reactor building.

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During its review of a Hermes operating license application, the Staff will confirm that the design of the safety-related portion of the reactor building is sufficient to withstand potential impacts from any aircraft types likely to use the Oak Ridge Airport and/or transit nearby airways for which the likelihood of impact could exceed applicable screening values (see response to Question 8 below), as appropriate.

8. When assessing aircraft impact and flooding hazards, the Staff notes that risk during the four-year operating life of the facility is acceptable. For example, in SE section 2.4.3.2, the Staff states, extreme flood events causing inundation of the site are unlikely during the planned 4-year Hermes operational lifetime.

What basis did the Staff use to determine that the lifetime risk over a 4-year period is acceptable for these hazards? Was consideration of a limited operating life necessary to approve the siting? If so, how will any operating lifetime limit be imposed on the Applicant to ensure that the basis for these conclusions remains valid?

Staff Response: In SE section 2.2.3.1.5, the Staff notes that the lifetime frequency of potential aircraft crashes into the proposed Hermes facility would be reduced due to the relatively short operating lifetime (4 years) for Hermes. The aircraft crash risk is primarily from general aviation aircraft; the total crash probabilities for aircraft types other than general aviation are very low and significantly below the DOE-STD-30142006 screening value of 1.00E-6 per year (see PSAR Table 2.29). As discussed in PSAR section 3.5.3.4 and SE section 3.5.3.3, Kairos will design the safety-related portions of the reactor building to withstand a crash of a general aviation aircraft. Although the Staff noted that the 4-year operating lifetime would reduce the crash risk, the Staffs determination that lifetime aircraft crash risks are acceptable was not based on the operating time because Hermes will be constructed to withstand a general aviation aircraft impact, and because the crash risk for other aircraft types is very low, even if a longer operating life were assumed. Therefore, with regard to aircraft impact risks, consideration of a limited operating life was not necessary for the Staff to find the Hermes siting acceptable.

In SE section 2.4.3.2, in reference to floods resulting from a probable maximum precipitation event or a hypothetical sunny-day dam failure that could exceed the Hermes grade elevation, the Staff finds that extreme flood events causing inundation of the site are unlikely during the planned 4-year Hermes operational lifetime. For flooding hazards from potential dam failures, consideration of a limited operating life was necessary to approve the siting because the Staff did rely on the 4-year operating period, in part, to assess the suitability of the site. The Staffs basis for this determination with respect to potential dam failures is discussed in the Staffs response to Question 9 below. With respect to extreme floods due to precipitation events that could inundate the site, the Staff notes that although the lifetime probability of such events is reduced by the short operating lifetime, the Staffs finding that such events are unlikely and the lifetime risk of such events is acceptable was not based on the 4-year operating period. The Staff considers such events to be unlikely (even over a longer operating period) because they would be beyond-design basis events occurring even less frequently than the design basis flood, which has a 25,000-year return period as discussed in PSAR section 2.4.3. The Staff considers a 25,000-year return period for a design basis flood to be acceptable for a testing 6

facility such as Hermes based on guidance in Department of Energy Standard 10202016 5, which was referenced by Kairos in the PSAR. The Staff notes that this return period is vastly more conservative than the 500-year flood used by the Federal Emergency Management Agency for critical facilities such as hospitals.

A Hermes operating license, if issued, would be issued with a license term that would be limited to ensure the validity of assumptions and conclusions in the safety analysis for the final design of the Hermes facility, as appropriate.

9. SE section 2.4.3.3 states that the staff notes that potential flood elevations for the hypothetical sunny-day dam failure could exceed the Hermes grade elevation but finds that such extreme flood events causing inundation of the site are unlikely during the planned 4-year period of Hermes operation. However, neither the Staff nor the Applicant appear to provide a probability or likelihood estimation for the hypothetical sunny-day dam failure. Please provide the likelihood of this event and the basis the Staff used to determine that this dam failure event was unlikely.

Staff Response: Based on information from the literature on generic dam failure probabilities, such as Fell, et al. 6, the Staff determined that sunny-day failure of the dams discussed in SE section 2.4.3.3 (including the Norris Dam which the Staff determined would be the critical case for potential effects on the Hermes site) would be unlikely (i.e., not a credible hazard), and that a more detailed site-specific analysis is not warranted. This determination was also based on the limited (4-year) period of Hermes operation combined with the lower risks associated with management and operations programs of the responsible federal agency for relevant dams in the vicinity of the proposed Hermes site. Tennessee Valley Authority owns, operates, and regularly inspects the dams following the requirements of its dam safety program. This oversight further supports Staffs determination that sunny-day failure of the dams would be unlikely.

10. SE section 3.4.3.1 states:

The staff reviewed the information provided in the PSAR and finds the approach acceptable because the development of the DRS [design response spectrum] follows the guidance of ASCE 4319 and the PSAR states that the SSCs will be designed in accordance with SDC 3 of ASCE 4319. In addition, the site seismic hazard characterization is used to develop the DRS. Following this approach aligns with the applicable guidance and acceptance criteria in NUREG-1537, Part 1 and 2, Section 3.4, and provides reasonable assurance that the reactor can be shut down and maintained in a safe condition following a seismic event.

NUREG-1537, part 1, section 3.4, Seismic Damage, references ANSI/ANS 157, IAEA-TECDOC403, and IAEA-TECDOC-348. How did the Staff determine that ASCE 5 DOE-STD-20102016, Natural Phenomena Hazards Analysis and Design Criteria for DOE Facilities, U.S. Department of Energy, 2016, https://www.standards.doe.gov/standards-documents/1000/1020-astd-2016.

6 Fell, R., Bowles, D., Anderson, L., Bell, G. 2000. The Status of Methods for Estimation of the Probability of Failure of Dams for Use in Quantitative Risk Assessment. International Commission on Large Dams, September 1422, 2000, Beijing, China.

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4319 was an acceptable approach for the development of the design response spectrum and that it aligned with the guidance and acceptance criteria in NUREG-1537?

Staff Response: The portion of guidance in Section 3.4 of NUREG-1537, Part 1, which recommends the use of ANSI/ANS 15.7, along with IAEA-TECDOC-403 and IAEA-TECDOC, to determine the seismic design, is applicable to research reactors. For the licensing of test reactors such as the Hermes test reactor, NUREG-1537 states that regulations in 10 CFR Part 100 must be met (i.e., 10 C.F.R. § 100.10(c)). Both ANSI/ANS-15.71977 (now withdrawn) and 10 C.F.R. § 100.10, along with section 2.5.5 of NUREG-1537, Part 1, specify the use of deterministic seismic hazard analyses to establish the vibratory ground motion design spectrum.

The applicant selected a more robust approach for the development of design response spectra (DRS) offered by ASCE 4319, which is based on inputs obtained from a probabilistic seismic hazard analysis (PSHA). An earlier version of this standard, ASCE 4305, has been endorsed by the Staff in RG 1.208 for the development of DRS for commercial power plants. The Staff considers the use of the most recent standard, ASCE 4319, for the Hermes test reactor to be appropriate as the technical approach for developing a DRS does not change between the subsequent versions of the standard. The main difference is that ASCE 4319 applies to multiple seismic design categories (SDCs) based on the risk profile of the nuclear facility that range from small nuclear facilities (SDC-2) to large facilities (SDC-5), while ASCE 4305 focuses on nuclear facilities that are categorized as SDC-3 and above.

In summary, although the deterministic approach is suitable for developing the seismic DRS for test reactors, the applicant went beyond this requirement by performing a PSHA to develop seismic hazard curves and using ASCE 4319 to determine the DRS for the Hermes test reactor.

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?

Staff Response: For applicant only.

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 3.61 as safety related and are identified to be SDC-3 in accordance with ASCE 4319. 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?

Staff Response: Section 1.2.3 of the Hermes PSAR outlines the approach that Kairos proposed for defining safety-related structures, systems, and components, considering the design of the Hermes test reactor in the context of 10 C.F.R. § 50.2. Further discussion of the Staffs determination regarding the acceptability of this definition is provided in response to Question 14. Table 3.61 of the Hermes PSAR summarizes key design characteristics of the 8

structures, systems, and components (SSCs) that make up the Hermes test reactor facility, including safety classification for the various SSCs. The various components that make up the reactivity control and shutdown system (RCSS) are included in this table. As noted in the table, the shutdown elements, including the latching/release mechanisms, are safety related. The RCSS drive mechanisms (except the shutdown element latching/release mechanisms) and control elements are not safety related because they are not relied upon to remain functional during or following design basis events. The release function of the electric clutch (i.e., the latching/release mechanism) and the shutdown elements themselves are relied upon to shut down the reactor and maintain it in a safe condition and to prevent and mitigate the consequences of accidents. This function is described in Hermes PSAR § 4.2.2.1, Description, as follows:

On a reactor trip, the electric clutch opens, which allows the sheave to rotate freely. With the sheave rotating freely, the shutdown and control elements are released from their drives and drop into the core and reflector, respectively, as a result of gravity.

The drive mechanism is not intended to insert the shutdown or control elements into the core upon a reactor trip signal.

While overpower transients may involve a malfunction of the drive mechanism, the drive motor is electrically isolated, and the opening of the electric clutch is responsible for inserting the shutdown elements into the core upon reactor trip signal. The control elements receive a trip signal but serve the purpose of defense-in-depth. Only the combined worth of two of the three available shutdown elements is needed to shut down the reactor and maintain it in a safe condition. The Staff will confirm the adequacy of the final RCSS design, including the acceptability of the RCSS component safety classifications, during the review of the operating license application.

b. Please identify what individual components of the RCSS are considered or are expected to be safety related.

Staff Response: The Staff considers or expects the following RCSS components to be safety related (or to perform safety-related functions): the shutdown element latching/release mechanisms (i.e., electric clutch), motor electrical isolation equipment, and shutdown elements.

This list is based on the Staffs review of the RCSS, as it is described in Hermes PSAR

§ 4.2.2. According to Hermes PSAR Table 4.24, and as discussed in Hermes PSAR § 4.2.2.1, the drive mechanisms include a counter-weighted winch (i.e., the sheave, wire-rope, and counter-weight), whereas the release mechanisms include an electric clutch, which is located between the motor and sheave, and electrical isolation for the motors. As discussed in response to Question 12.a, the drive mechanisms (except the shutdown element latching mechanisms) and control elements are non-safety related.

The RCSS design provided in the PSAR is preliminary. The Staff will confirm the adequacy of the final RCSS design, including the acceptability of the RCSS component safety classifications, during the review of the operating license application.

13. The Staff states in SE § 3.6.3.1.3 that the safety and seismic classification of these SSCs conforms with the guidance in RG 1.29, because safety related SSCs needed to control reactivity in the core are assigned correctly to the seismic classification, SDC-3 in accordance with ASCE 4319. This explanation seems to 9

address whether appropriately identified safety related SSCs have been assigned the appropriate seismic classification, but it does not appear to explain why the safety classification (of the components in the RCSS) itself is acceptable. Please discuss why this safety classification is acceptable.

Staff Response: As discussed in the response to Question 12, the safety classification of the components in the RCSS is acceptable because the RCSS components that are relied upon to shut down the reactor, to maintain it in a safe condition, and to prevent and mitigate the consequences of accidents are classified as safety related. Specifically, the release function of the electric clutch (i.e., the shutdown element latching/release mechanism) and the shutdown elements are relied upon to shut down the reactor and maintain it in a safe condition and to prevent and mitigate the consequences of accidents.

Conversely, the RCSS drive mechanisms (except the shutdown element latching/release mechanisms) and control elements are not safety related because they are not relied upon to remain functional during or following design basis events. The control elements also receive a trip signal but serve the purpose of defense-in-depth. Only the combined worth of two of the three available shutdown elements is needed to shut down the reactor and maintain it in a safe condition. The reflector blocks, which are also safety related, are expected to maintain the element insertion pathway.

The Staff will confirm the adequacy of the final RCSS design, including the acceptability of the RCSS component safety classifications, during the review of the operating license application.

14. In section 3.6.3.2.1 of the SE, the Staff states:

The Hermes reactor uses the definition of 10 CFR 50.2 for safety related SSCs to establish those SSCs that are classified as safety related, with the exception of integrity of the reactor coolant pressure boundary which has been modified to integrity of the portions of the reactor coolant boundary relied upon to maintain coolant level above the active core. This modification was made because the Hermes reactor does not rely on the functional capability of the primary heat transport system (PHTS) to remove decay heat from the reactor core. PSAR Table 3.61 identifies the safety related classification of SSC in accordance with the 10 CFR 50.2 definition with the modification. The staff finds that the modification of integrity of the portions of the reactor coolant boundary relied upon to maintain coolant level above the active core, is acceptable based on NRCs staff approval of KP-TR-004-NP-A Regulatory Analysis for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor.

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.

a. For the Staff: Are the 10 C.F.R. § 50.2 definitions of Safety related structures, systems, and components and Reactor coolant pressure boundary applicable to the Hermes facility? Is an exemption necessary for 10

the use of Kaiross modified definition of Safety-related structures, systems, and components for this application?

Staff Response: The 10 C.F.R. § 50.2 definition of safety-related structures, systems, and components is not applicable to the Hermes facility. This definition is only applicable to nuclear power reactors. This interpretation is consistent with the NRCs approach to licensing existing non-power reactors, as well as medical isotope production facilities such as the SHINE Medical Radioisotope Production Facility. In particular, the Staff has not applied the 10 C.F.R. § 50.2 definition of safety-related structures, systems, and components when reviewing licensing submittals for non-power reactors or production facilities, and exemptions from the 10 C.F.R.

§ 50.2 definition of safety-related structures, systems, and components have not been granted in licensing such facilities. In addition, as discussed in the Staffs response to Commission Pre-Hearing Question 2.d for the construction permit application hearing for the SHINE Medical Radioisotope Production Facility (ML15342A392), the history of this term and its usage in the regulations imply that the formal 10 C.F.R. § 50.2 definition of safety-related structures, systems, and components only applies to nuclear power plants.

Additionally, the 10 C.F.R. § 50.2 definition of reactor coolant pressure boundary is only applicable to nuclear power reactors, and not non-power reactors such as Hermes. This limited applicability is based on the definition of the term reactor coolant pressure boundary in 10 C.F.R. § 50.2 which reads, in part, all those pressure-containing components of boiling and pressurized water-cooled nuclear power reactors The Hermes test reactor is not a boiling or pressurized water-cooled nuclear power reactor.

Further, the Staff notes that Table A-3 from the referenced topical report is specific to power reactors. Conversely, Table D-3, Design Regulatory Requirements in 10 CFR That Require an Exemption for a KPFHR Test Reactor, of the referenced topical report applies to the Hermes test reactor facility and contains a set of expected exemptions, including exemptions from the requirements of 10 C.F.R. § 50.2. However, the Staffs SE for this topical report concluded, in part, that no exemption related to the use of the 10 C.F.R. § 50.2 definition of safety-related structures, systems, and components definition would be required for a KP-FHR testing facility.

Based on these considerations, Kairos proposed to replace the first portion of the 10 C.F.R

§ 50.2 definition of safety-related structures, systems, and components and an exemption was not necessary for them to do so. Kaiross alternate definition is described in section 1.2.3 of the Hermes PSAR. The Staff found this replacement acceptable, as discussed in SE section 3.1 and 3.6.

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?

Staff Response: For applicant only.

c. For the Staff: In light of this topical report, please explain why no exemptions are necessary at the construction permit stage.

Staff Response: In Table D-3 of the referenced topical report, Kairos cites six regulations as requiring exemptions for a KP-FHR test reactor. The first two of the six regulations involve the definitions of safety-related structures, systems, and components and reactor coolant pressure boundary from 10 C.F.R. § 50.2. The Staffs rationale for not requiring an exemption for these two items is addressed in the response to item (a) of this question.

11

The third regulation listed in Table D-3 of the referenced topical report involves 10 C.F.R.

§ 50.34(a)(4), which includes two parts. The first part of this regulation refers to the analysis and evaluation of the design and performance of SSCs. This is applicable to the Hermes test reactor and was analyzed by Kairos and evaluated by the Staff in the Hermes review. The second part of this regulation requires an analysis and evaluation of the emergency core cooling system performance to be provided in accordance with 10 C.F.R. §§ 50.46 and 50.46a. The Staff addressed this requirement specifically in SE section 5.1.3.2.8:

The regulation in 10 C.F.R. § 50.34(a)(4) requires an analysis in accordance with 50.46 and 50.46a of emergency core cooling system performance and the need for high point vents following postulated loss-of-coolant accidents. However, the regulation in 50.46 is specifically for light-water reactors, and thus is not applicable to the molten salt-cooled Hermes reactor. Similarly, the regulation in 50.46a states that it is for power reactors, so this regulation also does not apply to the Hermes reactor.

Therefore, the portion of 10 C.F.R. § 50.34(a)(4) that references 10 C.F.R. §§ 50.46 and 50.46a does not apply to the Hermes test reactor and does not require an exemption.

In addition to the discussion above, the Staff further notes the following portion of § 3.2.1 from the Staffs SE for the referenced topical report (underline and italics added for emphasis):

In addition, the NRC staff also reviewed the regulations that Kairos designated as exemption, which reflect Kairos position that certain requirements should not apply to the KP-FHR design or that Kairos plans to seek relief premised on an alternate solution. The NRC staff finds that, with the exception of the conclusions that the 10 C.F.R. § 50.2 safety-related SSC definition and that 10 C.F.R. § 50.34(a)(4) requires a test reactor applicant to submit analyses related to 10 C.F.R. § 50.46 and 10 C.F.R. § 50.46a, the regulations designated as exemption may apply (in whole or in part) to the licensing of the KP-FHR. The 10 C.F.R. § 50.2 safety-related SSC definition and the portion of 10 C.F.R. § 50.34(a)(4) related to 10 C.F.R. § 50.46 are not required when applied to a testing facility.

Therefore, the Staffs conclusion that no exemptions from the requirements of 10 C.F.R. §§ 50.2 and 50.34(a)(4) are required at the CP stage is consistent with the Staffs SE for the referenced topical report.

The final three regulations cited in Table D-3 in KP-TR-004-NP-A are 10 C.F.R. §§ 50.34(b)(4),

50.34(b)(9), and 50.36(c)(2)(ii)(A). These regulations involve information that must be provided, as applicable, with operating license applications, including final safety analysis reports, and therefore do not require an exemption at the CP stage for the Hermes test reactor.

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.

12

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?

Staff Response: Section 1.2.3 of the Hermes PSAR outlines the approach that Kairos proposed for defining safety-related structures, systems, and components, considering the design of the Hermes test reactor in the context of 10 C.F.R. § 50.2. Further discussion regarding the Staffs determination regarding the acceptability of this definition is provided in response to Question 14.

The moderator pebbles are not considered safety related because they are not relied upon to remain functional during and following design basis events. Specifically, they are not credited for shutting down the reactor or maintaining safe shutdown in the safety analyses.

As described in section 4.2.1.1 of the PSAR, the moderator pebble physical characteristics (e.g., size, shape, buoyancy) are the same as the pyrolytic carbon layers of fuel pebbles.

Moderator pebbles are tested using the same processes as are used for the fuel pebbles. Two deviations from the design bases for the moderator pebbles that could adversely impact the capability to shut down the reactor and maintain it in a safe shutdown condition are: (1) excessive wear causing carbon dust accumulation and (2) manufacturing flaws or mis-load resulting in additional carbon in the reactor. However, these deviations would not impact the ability to shut down the reactor and maintain safe shutdown as described below.

1. Qualification tribology tests will quantify the wear rates ensuring that the amount of carbon dust introduced into the coolant is analyzed and managed. PSAR section 4.3.3 states that dust could be postulated to accumulate in the natural circulation pathway.

The functional capability of the coolant pathway can be periodically checked during operation which would provide an early indication of the issue. Section 4.3.3.12 of the Staffs SE documents the Staffs evaluation of the inspection and monitoring capabilities of the components needed to ensure integrity of the natural circulation flow path.

2. The quality assurance process should ensure anticipated moderator pebble density and appropriate pebble loading. However, if manufacturing flaws are not identified and the density (i.e., carbon content) is incorrect, it could affect the neutron multiplication factor.

Alternatively, inadvertent moderator pebble mis-load causing significant deviations away from the target moderator pebble fraction would also affect core reactivity. However, either scenario would be identified during reactor start-up activities (e.g., operational confirmation of anticipated reactivity) and would not impact the ability to shut down the reactor and maintain 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)?

Staff Response: Safety-related functions would not be adversely impacted by failure of one or 13

more moderator pebbles. For example, if one or more moderator pebbles were to fail, Flibe does not significantly infiltrate into the carbon matrix material manufactured to the Kairos Hermes specifications and the carbon matrix itself does not appreciably form any compounds with Flibe. Therefore, it is anticipated that the thermophysical properties of Flibe will not be impacted by chemical interactions caused by failed moderator pebbles. Additionally, since the shutdown elements are inserted directly into the pebbles without the use of guide tubes, there is no potential for a moderator pebble fragment from a failed pebble impeding shutdown element insertion by wedging between the control element and a surrounding tube.

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?

Staff Response: The Staff expects that fouling or plugging of the DHRS can be distinguished and monitored separately from fouling or plugging of the PHTS but has not made a finding related to these capabilities. Consistent with 10 C.F.R. § 50.34(a)(4), the information required to be provided by an applicant consists of a preliminary analysis of the performance of structures, systems, and components. Kairos provided preliminary information related to the function of the DHRS in the PSAR and associated audits, but not the final system design or performance monitoring measures.

Based on the preliminary information, the Staff has reasonable assurance that performance monitoring measures can be shown to be sufficiently independent because separate testing will be conducted to assess the PHTS and DHRS, and Kairos stated that separate instrumentation will be used to monitor the performance of the DHRS. The Staff will confirm during the operating license application review that the final design will enable DHRS performance monitoring to be sufficiently independent. Additional detail is provided below.

In response to the Staffs DHRS audit (ML23115A480), Kairos provided additional detail regarding DHRS qualification testing (ML22244A235). SE section 6.3.3.2, Staff Evaluation of DHRS Design, notes the testing will cover factors that could challenge DHRS performance.

The Treated Water System, described in PSAR section 9.7.2, contains components such as filters and demineralizers and supplies water to the DHRS. The Staff expects that these chemistry controls will help mitigate or prevent fouling in the DHRS.

The Staff SE, Appendix A, section A.2, Additional Items for an Operating License Application, states that testing and inspection of the DHRS will be reviewed during the operating license application review. This will allow the Staff to ensure DHRS fouling or plugging can be mitigated and monitored.

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?

Staff Response: Kairos stated that there will be monitoring instrumentation specific to the DHRS. Based on this, the Staff expects that it will be possible for Kairos to demonstrate that fouling and plugging can be monitored separately, so fouling or plugging of the DHRS would not be masked by PHTS operation. The Staff will confirm during the operating license application 14

review that the final design will enable DHRS performance monitoring to be sufficiently independent.

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 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.

Staff Response: The tritium mitigation system uses instrumentation to monitor tritium accumulation in the tritium capture beds. The instrumentation monitors tritium accumulation by measuring the difference in the tritium activity in the process streams entering and exiting the tritium capture beds. Based on information provided by Kairos in the PSAR, the term proper instrumentation for tritium monitoring is defined as a system which meets PDC 13. Accordingly, proper instrumentation for tritium monitoring provides tritium measurements over the range of anticipated tritium activity levels during normal operation to ensure that the amount of tritium accumulation in the capture beds is sufficiently small to meet the assumptions for the contribution of tritium for the MHA dose assessment described in PSAR Chapter 13. PSAR section 11.1.4 states that additional details of tritium monitoring will be provided in the operating license application.

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.

Staff Response: Section 1.2.3 of the Hermes PSAR outlines the approach that Kairos proposed for defining safety-related structures, systems, and components, considering the design of the Hermes test reactor in the context of 10 C.F.R. § 50.2. Further discussion 15

regarding the Staffs determination regarding the acceptability of this definition is provided in response to Question 14.

TMS instrumentation is designated as non-safety related because it is not relied upon to remain functional during and after design basis events. The instrumentation is used during normal operation to measure tritium accumulation in the tritium capture beds. It is used by the operators to ensure that the amount of tritium accumulation in the capture beds is sufficiently small to meet the assumptions for the contribution of tritium for the MHA dose assessment described in PSAR Chapter 13.

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?

Staff Response: 10 C.F.R. § 50.55(f)(3) does not apply to the Hermes reactor because the Hermes reactor is a non-power reactor, and therefore, is not a nuclear power plant or fuel reprocessing plant subject to the quality assurance criteria in Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants to 10 C.F.R. Part 50.

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.

Staff Response: Because Hermes is neither a nuclear power plant nor a fuel reprocessing plant, the requirements of 10 C.F.R. § 50.55(f), which require, in part, that nuclear power plant or fuel reprocessing plant construction permit holders implement the quality assurance program described in their Safety Analysis Report, would not apply to Hermes. Therefore, the Staff recommended including the permit condition related to Kaiross quality assurance program to (1) ensure consistency in expectations of construction permit holders implementation of quality assurance programs developed pursuant to 10 C.F.R. § 50.34(a)(7); (2) establish criteria for changes to the quality assurance program and Kaiross notifying the NRC of such changes; and (3) facilitate the correction of any identified deficiencies in the implementation of the quality assurance program through the NRCs enforcement process during construction inspection.

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 G9917, 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 16

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?

Staff Response: The fuel pebbles in the Hermes reactor design will experience both Flibe and argon environments during their operating lifetimes. For example, the reactor vessel environment is Flibe while the pebble handling and storage system environment is argon.

Pebble wear rates are a function of several parameters, including Flibe properties (e.g., density, viscosity, wetting behavior) and environmental conditions (e.g., temperature). While parameter details are not available in the PSAR that would allow the Staff to evaluate whether wear is higher in argon or Flibe, the Staff maintains its understanding that wear rates will be measured in Filbe and in argon as stated in SE section 13.1.5.3. Results from this testing will allow the Staff to evaluate wear rates during the review of the Hermes operating license application.

Discussion of tribology tests described in section 3.6.2 of Fuel Qualification Methodology for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor, Rev. 2 indicate that wear and friction tests would occur in both Flibe and argon environments. For example, the first paragraph of section 3.6.2 begins with:

Tribology testing will be performed to quantify coefficients of friction and wear rates during pebble-to-pebble, pebble-to-graphite, and pebble-to-stainless steel contact. These measurements will be made in Flibe and in argon gas.

While it is mentioned multiple times that the tests are performed in both Flibe and argon environments, the word or is also used in the topical report two times when discussing the environment used in the tests. First, the third paragraph of section 3.6.2 of the topical report states, [t]he contact surface will be immersed in argon gas or in molten Flibe Based on the context in and around this paragraph, the Staff interprets this phrase to mean that an individual test would use either environment such that the test matrix as a whole would include both argon and Flibe tests.

Second, the Staff acknowledges that the example provided by the bullets at the end of section 3.6.2 could be interpreted to indicate that wear tests will be performed in one environment or the other, but not both. However, based on the overall discussion provided in section 3.6.2 of the topical report, the Staff concluded that the planned tribology tests would encompass both Flibe and argon environments over a range of temperatures. Further, Kairos stated that wear rates would be assessed in different environments during the October 17, 2022, ACRS subcommittee meeting regarding the Kairos fuel qualification topical report (ML22299A013).

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.

17

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.

Staff Response: The source term and modeling of radionuclide transport, retention, and release, including cesium (Cs) and iodine (I) radioisotopes that support the MHA analysis is consistent with the approved mechanistic source term (MST) topical report (TR), KP-TR-012-NP-A, KP-FHR Mechanistic Source Term Methodology (ML22136A288).

The following summarizes how Cs and I are accounted for in the in the MHA analysis. PSAR section 13.1.1.3 and 13.2.1, Transient Assumptions and Maximum Hypothetical Accident, respectively, describe the overall Hermes system as a set of sources of radionuclide material at risk for release (MAR), identifies the succession of barriers to release, and applies a release fraction for each barrier that contains the MAR. The sources of MAR and associated barriers are listed in PSAR section 13.2.1.1, Methodology and Inputs; the pertinent sources for Cs and I radioisotopes are the TRISO fuel in the reactor core and the Flibe circulating activity. The applicant developed an integral release fraction for each barrier on a radionuclide class or isotopic basis, based on transport and retention phenomena such as diffusion or evaporation driven by temperature. These transport and retention phenomena are described in more detail in the PSAR and the MST TR. Cs and I are in separate radionuclide transport groups for each barrier, but each Cs or I radioisotope is accounted for as it moves from the fuel into the Flibe, then from the Flibe into the gas space, then into the environment.

Releases from the fuel are modeled for fuel with manufacturing defects, heavy metal contamination, in-service failure, and diffusive releases from intact TRISO particles. Diffusion from the fuel kernel and through TRISO layers is modeled by fuel radionuclide transport group, with the element members of that group assumed to be released at the same release fraction as the representative element for that class. Cs diffusivity information is representative for Cs isotopes, while Krypton (Kr) diffusivity information is representative for I isotopes. The release from the fuel is modeled as a set of release fractions for a fuel radionuclide transport group inventory, over time intervals. Fuel is not uncovered for the MHA conditions; therefore, the releases from the fuel are only into the Flibe.

Radionuclides are grouped by chemical behavior in Flibe, with retention and releases modeled from the Flibe into the cover gas. Radionuclide release from the Flibe to the gas space occurs by bubble burst from entrained cover gas and evaporation from the Flibe driven by the MHA time-at-temperature curve (PSAR Figure 13.21). Bubble burst on the surface of the Flibe releases produces Flibe aerosols, which includes Cs and I from the Flibe circulating activity.

Evaporation allows for release of the Flibe circulating activity, as well as radionuclides (including Cs and I) that have diffused from the fuel into the Flibe. The release from the Flibe is modeled as a set of release fractions for a Flibe radionuclide transport group inventory, over time intervals. Cs and I are not assumed to be retained in graphite structures or the fuel pebble matrix.

Transport through gas spaces includes the cover gas in the reactor vessel and the air space in the reactor building. The reactor building and reactor vessel headspace are not credited as part of the functional containment. Cs and I radioisotopes included in the Flibe aerosols are available for aerosol deposition or release to the environment with a leakage rate that models release of all the radionuclides in the building within a two-hour window, consistent with the approved MST 18

TR. Although the TRISO particles do provide retention of Cs and I, transport of the remaining fractions of Cs and I through the remaining barriers are modeled in the applicants MHA radiological consequence analysis. Therefore, the MHA offsite dose results include the effect of the estimated postulated releases of MAR, including Cs and I radioisotopes.

21. Section 4.4 of the Environmental Protection Plan (EPP) includes a requirement that the permit holder shall request a license amendment to incorporate Terms and Conditions set forth in Incidental Take Statements of Biological Opinions issued subsequent to the effective date of the EPP. This provision was not included in other recent EPPs (e.g. the Shine Medical Technologies, Inc., EPP (ML16041A473) and Northwest Medical Isotopes, LLC, EPP (ML18037A468)). Why is it necessary to request a license amendment to incorporate these terms and conditions?

Staff Response: The provision requiring a license amendment to incorporate Terms and Conditions issued subsequent to the effective date of the license was included in EPPs prepared as part of several recent NRC licenses for new reactors. Examples include the Clinch River Nuclear Site ESP (ML19352D868) and the Turkey Point Units 6 and 7 COLs (ML17088A301 and ML17088A319). However, the NRC Staff reconsidered including this provision as a result of the Commissions question and due to the circumstances of this licensing action and decided to remove it from the EPP for the Kairos Hermes CP. For example, unlike for recently licensed large light-water reactors but consistent with the SHINE and Northwest Medical Isotopes (NWMI) licensing actions, the Staff did not perform formal consultation under Section 7 of the Endangered Species Act (ESA) for Kairos Hermes and did not receive an Incidental Take Statement containing Terms and Conditions needing to be addressed in the EPP. Instead, the Staff received through informal consultation a concurrence letter from the U.S. Fish & Wildlife Service agreeing with a determination that the Kairos Hermes CP may affect but would not likely adversely affect resources protected under the ESA. As such, the NRC Staff agrees that the Kairos Hermes CP EPP should be consistent with the EPPs for SHINE and NWMI rather than with the EPPs for recent new reactors. Furthermore, the Staff must comply with Section 7 again upon receipt of an operating license application from Kairos for the Hermes test reactor. At that time, Staff can assess whether any additional Terms and Conditions have been issued to Kairos for Hermes since issuance of the CP as well as the likelihood of future impacts the operating life of Hermes will have on resources protected under the ESA. The Staff will consider at that time whether the EPP provision requesting Kairos submit a license amendment is appropriate.

22. 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?

Staff Response: Prior to issuing the FEIS, the Staff reviewed the Fiscal Responsibility Act (FRA) and the amendments to NEPA and found that the FEIS was consistent with the requirements of the FRA, given our current understanding. The related footnote in the draft Record of Decision (ROD) will be revised to read NEPA was amended in June 2023 by the Fiscal Responsibility Act of 2023 (FRA). The Staff determined that the FEIS is consistent with the requirements of the FRA. No information in the EIS for the Kairos Hermes CP or in this 19

Record of Decision was altered as a result of the amendments.

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?

Staff Response: The Staff has made all the findings necessary in the FEIS to support the Commissions issuance of a construction permit. However, to complete the record, the following information is also available to the Commission.

1) The new and significant evaluation of the proposed Hermes 2 facility (ML23220A164) determined the new information is not significant for cumulative impact assessment in the FEIS (see response to Question 23).

2) The Staff is in the process of closing National Historic Preservation Act (NHPA) section 106 consultation. The Staff plans to provide updated information on the status of consultation to the Commission when that information is available (see response to part C of Question 25).

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?

Staff Response: For applicant only.

23. On July 14, 2023, the Applicant submitted a construction permit application for the Hermes 2 facility, a two-unit fluoride salt-cooled, high temperature test reactor that would be situated within the 185- acre Hermes project boundary. Does the cumulative impacts analysis in the FEIS take into account the proposed Hermes 2 facility?

Staff Response: The application for the Hermes 2 facility was submitted after the Hermes CP EIS was in final stages of publication; therefore, the cumulative impacts analysis in the FEIS does not take into account this proposed facility. Upon receipt of the Hermes 2 application, however, the Staff performed an evaluation in accordance with 10 C.F.R. § 51.92(a) to determine whether the new information provided by Kairos regarding the proposed Hermes 2 facility application was significant for cumulative impacts as evaluated in the Hermes CP EIS.

The Staff followed guidance provided in Staff Process for Determining if a Supplement to an Environmental Impact Statement is Required in Accordance with 10 C.F.R. § 51.92(a) or 51.72(a) (ML120950050) to evaluate and document the significance of this new information.

The Staff concluded that the new information did not meet either of the criteria in 10 C.F.R.

§ 51.92(a) that would require the Staff to prepare a supplement to the Hermes CP EIS.

The Staffs evaluation of the new information provided by Kairos regarding the proposed 20

Hermes 2 facility application and its significance on cumulative impacts as evaluated in the Hermes CP EIS is located in Agencywide Documents Access and Management System (ADAMS) Package No. ML23220A164.

24. In describing the Applicants proposed Atlas Fuel Fabrication Facility as a reasonably foreseeable new project with the potential to affect land use and visual quality in the area around the Hermes site, the FEIS states that [t]he NRC staff would ensure the compatibility of the Atlas facility with the land uses and visual quality of the Hermes site and the Heritage Center in general when reviewing a future licensing application for the Atlas facility project.

How would the Staff ensure such compatibility? What is the basis of the Staffs legal authority to ensure such compatibility?

Staff Response: In this context, the Staff used the term ensure to indicate Staffs intent to evaluate fully the compatibility of the Atlas facility with the land use and visual quality of the Hermes site and the Heritage Center at the time the NRC receives an application for the Atlas facility. The Staff did not intend to suggest any authority to directly regulate the visual compatibility of the future project. However, if the NRC receives an application for a fuel fabrication facility on the Hermes site, the Staff does intend to ensure our review addresses the compatibility of the facility with other facilities on the site and would provide an opportunity for stakeholders to comment on the visual compatibility and for Kairos to hear their concerns.

25. The Staffs analysis in the FEIS concludes that the potential environmental impacts on cultural and historic resources from constructing, operating, and decommissioning the Hermes project would be small. The Staff also concludes, for the purposes of satisfying section 106 of the NHPA, that there would be no adverse effects to historic properties from the proposed project. The bases for both conclusions rely on the Staffs determination that there are no known historic and cultural resources (NEPA) or historic properties (NHPA) within the proposed Hermes reactor site. However, the Staff also indicates that the Applicant is working with a consulting Tribe to develop and carry out an additional reconnaissance field investigation of the site.

a. What is the purpose of the additional reconnaissance field investigation?

Staff Response: After issuance of the draft EIS, the Staff received a comment from a Federally Recognized Indian Tribe requesting that the Tribes Historic Preservation Officer (THPO) be included in the development of the Archaeological Resource Monitoring and Unanticipated Discovery Plan (monitoring plan) and that a cultural resource survey be conducted for the project. The Staff contacted the Department of Energy (DOE) to obtain additional information related to site geology, geomorphology, and the scope and extent of prior disturbance. Based on new information obtained from DOE, the Staff determined there was a potential for deeply buried paleosols (old soil surfaces where archaeological artifacts are a potential). The Staff concurred with the consulting Tribes request for field investigation for the purpose of informing the tribal participation in the development of the monitoring plan and shared this information with Kairos in a series of closed meetings. Kairos has completed the field investigation and has used the information from the survey to update its monitoring plan accordingly. The Staff and the THPO are currently reviewing draft versions of the field investigation report and updated monitoring plan.

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b. How did the Staff reach its conclusions regarding impacts under NEPA and the NHPA when it appears that additional work to determine the presence of cultural resources or historic properties at the Hermes site is yet to be carried out?

Staff Response: The Staff reached its conclusions regarding impacts under NEPA and the NHPA with the available information submitted through the application as well as the Staffs independent environmental review, including information the Staff received from its NHPA Section 106 consulting parties throughout the environmental review process. The information Staff gathered during the review process was varied and detailed enough for the Staff to draw conclusions under NEPA and the NHPA regarding impacts to historic and cultural resources.

The Staffs NEPA and NHPA findings did not change as a result of new information regarding potential for the presence of cultural resources as the Staffs findings in the draft EIS had acknowledged the potential for historic and cultural resources to exist.

The Staffs analysis included basic information regarding the characteristics of the site that Staff needed to draw reasonable conclusions as to potential impacts. The field investigation was meant to provide additional information regarding the extent and characteristics of the fill and to better understand which areas could have the potential for deeply buried historic and cultural resources (areas which would require enhanced monitoring and reporting and at what depths).

The low likelihood of historic and cultural resources being present in areas where project activities will occur, and the addition of monitoring in areas where the depths of previous disturbance are not certain, provides additional assurance for the Staffs finding that there are no adverse effects to historic properties or historic and cultural resources.

c. How does the Staff plan to inform the Commission of the status of consultations with the Tribe during the pendency of the mandatory hearing process?

Staff Response: The Staff will continue to meet with the consulting parties until NHPA Section 106 consultation closure is complete. Once all parties agree that the information in the field investigation report is complete for the purposes of informing the monitoring plan, and the monitoring plan has been updated accordingly, the Staff will update the draft ROD to describe the completion of consultations and the plans for monitoring and mitigation. The draft CP will also be updated to document Kaiross responsibility to follow the monitoring plan.

The Staff will issue formal NHPA Section 106 correspondence to the consulting Tribe that documents the steps taken to resolve their comments and request concurrence. This formal exchange of NHPA Section 106 correspondence will conclude consultation. There are several mechanisms available to the Staff to inform the Commission when consultations have concluded (e.g., CA Note, CA Briefing, or Daily Note), depending on the complexity, nature of any emergent issues, or level of tribal engagement necessary to complete consultations. Staff would seek to use the most efficient and expeditious means to inform the Commission.

26. Section 3.7.1.3 of the FEIS states:

Nonradioactive gaseous wastes from operating areas would be passed through a high-efficiency particulate air filtration system before being vented to the atmosphere, and additional controls may be implemented as required by local permit conditions.

How was effluent that is vented directly to atmosphere from the Decay Heat 22

Removal System (DHRS) considered in the environmental impact of operations?

For example, PSAR section 6.3.1, states:

The DHRS is an ex-vessel system that continuously operates when the reactor is operating above a threshold power by removing energy from the vessel wall via thermal radiation and convective heat transfer to water-based thermosyphons.

Inventory in the thermosyphons is boiled off and vents directly to the atmosphere outside of the reactor building.

Staff Response: The statement cited in the question above that appears in the nonradioactive health section of the FEIS does not include the emissions from the DHRS. Emissions from the DHRS would consist of water vapor formed by exposure of water to decay heat generated in the reactor core. The feedwater used to trap the heat delivered to the DHRS would be contained in annular thermosyphons (called thimbles) that are positioned circumferentially around the outside of the reactor vessel. Thus, the DHRS is physically separated from the reactor systems and relies on demineralized water from the Treated Water System in four separate water storage tanks (one for each DHRS train). The Treated Water System is comprised of supply piping, pumps, filters, storage tanks, demineralization exchangers and tanks, demineralization support components, and distribution piping. Because the effluent from the DHRS is water vapor, the effluent would chemically consist of only water since solutes contained in the water would be captured as residue in the thimbles rather than being emitted to the atmosphere.

Emission to the atmosphere of water vapor would not have a significant effect on human health.

27. Section 3.9.1 of the FEIS states: While the Hermes reactor is not a light-water-cooled nuclear power reactor, Kairos will rely upon the same uranium fuel cycle addressed by Table S-3.

How was the impact of use of higher enriched High-Assay Low Enriched Uranium (HALEU) considered (e.g., the impacts associated with achieving an enrichment level up to 20 weight percent) when addressing the applicability of Table S-3?

Staff Response: While the exact source of the 0.93 MTU of HALEU for the Hermes test reactor is not known, the Staff did evaluate the environmental effects of the uranium fuel cycle with HALEU enrichment levels in Section 3.14.2.3 of NUREG-2249, Generic Environmental Impact Statement for Advanced Nuclear Reactors: Draft Report for Comment (ML21222A055) 7. In that environmental evaluation, the Staff determined that the enrichment process efficiencies of gaseous centrifuge enrichment technology were such that Table S-3 would bound the environmental impacts for a gaseous centrifuge enrichment facility to produce HALEU. This evaluation in Draft NUREG-2249 is also in line with the expected outcome of the DOEs efforts to spur demand for additional HALEU production and private investment in the nations nuclear fuel supply infrastructureultimately removing the federal governments initial role as a supplier.

To this end, DOE has been working through various activities to establish a HALEU gaseous centrifuge-based supply chain over the time-period of the Kairos Hermes test reactor environmental review. Thus, while the exact HALEU source for the Hermes test reactor cannot be identified at this time and some of the potential sources could differ from what was assessed for Table S-3, the related environmental impacts for the small quantity of HALEU needed by Kairos (0.93 MTU) should be bounded by Table S-3 and would need to be confirmed in the operating license application review.

7 NUREG-2249 is an enclosure of SECY-210098, which is currently before the Commission and is not yet available for public comment.

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28. NEPA section 102(2)(C)(v), as amended by the Fiscal Responsibility Act of 2023, requires federal agencies to describe any irreversible and irretrievable commitment of federal resources which would be involved in the proposed agency action. FEIS section 5.3.3 contains an analysis of the irreversible and irretrievable commitment of resources noted in the environmental review of the Hermes project, but it does not specify whether any of these resources are federal resources.

Please clarify whether the FEIS analysis accounts for the irreversible and irretrievable commitment of federal resources.

Staff Response: The FRA changed the requirement from considering the irreversible and irretrievable commitment of resources to the irreversible and irretrievable commitment of federal resources. While Section 5.3.3 of the FEIS discusses the irreversible and irretrievable commitment of resources such as land, water, raw materials, and other natural resources, this section also states, In general, the commitment of capital, energy, labor, and material resources for a project such as Hermes are also irreversible. As some of these types of resources are expended by the NRC during its review of the Hermes application, the Staff considers that these could be considered federal resources under the FRA. In this regard, the FEIS does account for the irreversible and irretrievable commitment of federal resources. For future environmental reviews, the Staff will consider any new or changing guidance issued by the Council on Environmental Quality regarding the irreversible and irretrievable commitment of federal resources.

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

CERTIFICATE OF SERVICE Pursuant to 10 C.F.R. § 2.305, I hereby certify that copies of the foregoing NRC STAFF RESPONSES TO COMMISSION PRE-HEARING QUESTIONS, dated September 28, 2023, have been served upon the Electronic Information Exchange, the NRCs E-Filing System, in the above-captioned proceeding, this 28th day of September, 2023.

/Signed (electronically) by/

Megan Wright Counsel for NRC Staff Mail Stop: O-14-A44 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Telephone: (516) 765-6523 E-mail: megan.wright@nrc.gov Dated at Lewes, Delaware this 28th day of September, 2023