ML23048A326
| ML23048A326 | |
| Person / Time | |
|---|---|
| Site: | 99902069 |
| Issue date: | 03/02/2023 |
| From: | Jessup W NRC/NRR/DANU/UAL1 |
| To: | Hastings P Kairos Power |
| Beasley B | |
| References | |
| EPID L-2020?TOP?0046, CAC 000431 | |
| Download: ML23048A326 (14) | |
Text
Mr. Peter Hastings Vice President, Regulatory Affairs and Quality Kairos Power LLC 707 W Tower Ave Alameda, CA 94501
SUBJECT:
KAIROS POWER LLC - FINAL SAFETY EVALUATION FOR TOPICAL REPORT FUEL QUALIFICATION METHODOLOGY FOR THE KAIROS POWER FLUORIDE SALT-COOLED HIGH TEMPERATURE REACTOR (KP-FHR) (REVISION 2) (EPID: L-2020TOP0046/CAC NO. 000431)
Dear Mr. Hastings:
This letter provides the final safety evaluation for the Kairos Power LLC (Kairos) topical report Fuel Qualification Methodology for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor (KP-FHR). By letter dated June 30, 2020, Kairos submitted the topical report for the U.S. Nuclear Regulatory Commission (NRC) staffs review (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML20182A830). The NRC staff provided questions on the topical report to Kairos on November 20, 2020 (ML20328A147), and August 4, 2021 (ML21216A398). On October 13, 2021, February 8, 2022, May 2, 2022, May 16, 2022, June 2, 2022, and June 23, 2022, the NRC staff and Kairos held closed meetings to discuss additional clarification questions. Kairos submitted Revision 1 of the topical report by letter dated April 5, 2022, and Revision 2 by letter dated July 5, 2022 (ML22095A277 and ML22186A213, respectively).
The NRC staffs final safety evaluation for Fuel Qualification Methodology for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor (KP-FHR), Revision 2, is enclosed. The NRC staff provided a draft of the safety evaluation to Kairos for the purpose of identifying proprietary information and Kairos confirmed that the draft safety evaluation did not contain proprietary information on October 5, 2022 (ML22278B094).
The Advisory Committee on Reactor Safeguards (ACRS) was briefed on this topical report and the NRC staffs draft safety evaluation on October 17, 2022, and November 30, 2022. The ACRS provided its recommendations for the publication of this safety evaluation in a letter dated December 20, 2022 (ML22340A628). By letter dated February 8, 2023 (ML23024A263), the NRC staff provided its response to the ACRS recommendations. The enclosed safety evaluation is final and is publicly available.
March 2, 2023
P. Hastings The NRC staff requests that Kairos publish an accepted version of this topical report within 3 months of receipt of this letter. The accepted version shall incorporate this letter and the enclosed safety evaluation after the title page. The accepted version shall include an -A (designating accepted) following the topical report identification symbol.
If you have any questions, please contact Ben Beasley at (301) 415-2062 or via email at Benjamin.Beasley@nrc.gov.
Sincerely, William Jessup, Chief Advanced Reactor Licensing Branch 1 Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation Docket No.: 99902069
Enclosure:
As stated Cc via ListServ: Distribution via Kairos Power FHR Signed by Jessup, William on 03/02/23
ML23048A326 NRR-043 OFFICE NRR/DANU/UAL1:PM NRR/DANU/UAL1:LA OGC/NLO NRR/DANU/UAL1:BC NAME BBeasley CSmith AGhosh WJessup DATE 2/21/ 2023 2/27/2023 3/1/2023 3/2/2023
Enclosure FINAL SAFETY EVALUATION FOR TOPICAL REPORT KPTR011, REVISION 2, FUEL QUALIFICATION METHODOLOGY FOR THE KAIROS POWER FLOURIDE SALT-COOLED HIGH TEMPERATURE REACTOR (KP-FHR) (EPID L2020TOP0046)
APPLICANT INFORMATION Applicant:
Kairos Power LLC Applicant Address:
707 W. Tower Ave.
Alameda, CA 94501 Docket /Project No(s).:
99902069 APPLICATION INFORMATION Submittal Date: June 30, 2020 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML20182A830 Revision letter Dates and ADAMS Accession Nos: Revision 1, April 5, 2022 (ML22095A276),
Revision 2, July 5, 2022 (ML22186A212)
Brief Description of the Topical Report: This topical report (TR) describes the fuel qualification methodology for the Kairos Power Fluoride Salt Cooled High Temperature Reactor (KPFHR) power and test reactors. Kairos Power LLC (Kairos, the vendor) relies on the Electric Power Research Institute (EPRI) TR on Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO)-Coated Particle Fuel Performance (Reference 1) for qualification of the TRISO fuel particles, but KPTR011P TR details the qualification of the fuel pebble form which is used in the KPFHR power and test reactor designs. The qualification methodology relies on an evaluation of a KPFHR Phenomena Identification and Ranking Table (PIRT) to provide the scope of the qualification needs. Kairos presents seven limitations in Section 4.2 of the TR, which serve to provide boundaries to the applicability of the TR.
The methodology considers mechanical and structural performance as well as chemical, thermal, and irradiation effects. Mechanical and chemical performance are established or confirmed by laboratory testing. Thermal and irradiation performance is established by the Advanced Gas Reactor (AGR)-2 testing program or additional irradiation testing, if outside the AGR2 performance envelope. AGR2 is the second of the planned irradiations for the AGR Fuel Development and Qualification Program, a program at the Idaho National Laboratory funded by the Department of Energy for purposes of determining irradiated TRISO particle behavior.
The KPFHR power and test reactor fuel qualification methodology is intended to provide reasonable assurance that the KPFHR power and test reactor fuel pebbles can operate with a low failure fraction similar to what has been previously demonstrated for the AGR TRISO fuel particles. To that end, this TR presents a series of tests, available data, and surveillance plans to support the qualification of KPFHR power and test reactor fuel pebbles.
For additional details on the submittal, please refer to the documents located at the ADAMS Accession numbers identified above.
EVALUATION CRITERIA Regulations For construction permit applications, Title 10 of the Code of Federal Regulations (10 CFR) 50.34(a) provides that each application for a construction permit shall include a preliminary safety analysis report (PSAR) and describe the minimum information to the be included in the PSAR.
The regulation at 10 CFR 50.34(a)(1)(ii)(C) provides that the application shall include the extent to which the reactor incorporates unique, unusual or enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials.
The regulation at 10 CFR 50.34(a)(1)(ii)(D) provides, in part, that the application shall include the safety features that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur.
The regulation at 10 CFR 50.34(a)(2) provides that the application shall include a summary description and discussion of the facility, with special attention to design and operating characteristics, unusual or novel design features, and principal safety considerations.
For operating license applications, 10 CFR 50.34(b)(4) requires that each application, for an operating license, include a final safety analysis report that includes an analysis and evaluation of the design and performance of structures, systems, and components with the objective of evaluating the risk to the public. The analysis and evaluation shall take into account any pertinent information developed since the submittal of the PSAR. Similar regulations exist for Part 52 applications.
The requirements in 10 CFR 50.34(a) and 10 CFR 50.34(b), specified above, apply to this TR because the TRISO particle is the primary means of fission product retention.
The regulation at 10 CFR 50.41(b) provides additional standards for class 104 test reactor licenses, including that in addition to applying the standards set forth in 10 CFR 50.40, the Commission will permit the conduct of widespread and diverse research and development.
The regulation at 10 CFR 50.43(e) provides requirements regarding analysis, appropriate test programs, and experience for designs that differ significantly from light water reactors designs licensed before 1997.
The regulation at 10 CFR Part 100, Reactor Siting Criteria, establishes approval requirements for proposed sites for stationary power and testing reactors subject to 10 CFR Part 50 or Part
- 52. The regulation in 10 CFR 100.11, Determination of exclusion area, low population zone, and population center distance requires, in part, the determination of an exclusion area, a low population zone, and a population center distance. The vendor plans to credit the retention of the TRISO particle to inform its dose analysis, and this TR will, therefore, play a role in the NRC staffs review of the siting criteria.
Principal Design Criteria The regulation in 10 CFR 50.34(a)(3)(i) requires, in part, that an applicant for a construction permit to build a reactor is to provide principal design criteria (PDC) for the facility. The NRC staff approved PDC for the KPFHR in KPTR003NP-A, Revision 1 (ML20167A174).
KPFHR PDC 10, Reactor design states [t]he reactor core and associated heat removal, control, and protection systems shall be designed with appropriate margin to ensure that specified acceptable system radionuclide release design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
Establishing fuel design limits and ensuring these limits are not exceeded represent a fundamental underpinning of the safety assessment of a nuclear power plant required by 10 CFR 50.34(a)(1).
KPFHR PDC 16, Containment Design states that the [r]eactor functional containment, consisting of multiple barriers internal and/or external to the reactor and its cooling system, shall be provided to control the release of radioactivity to the environment and to ensure that the functional containment design conditions which are safety significant are not exceeded for as long as postulated accident conditions require. Preventing uncontrolled release of radioactivity represents a fundamental underpinning of the safety assessment of a nuclear power plant required by 10 CFR 50.34(a)(1).
TECHNICAL EVALUATION In the following technical evaluation, a section, table, or figure number without additional description refers to the vendors TR. If section, table, or figure numbers are referenced from other reports or sections for this safety evaluation, additional descriptive information will be provided.
Applicability of Existing Data Section 2 of TR KPTR011P, Fuel Qualification Methodology for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor (KP-FHR) provides a description of the KPFHR fuel design. The NRC staff compared the KPFHR particle and pebble design with existing designs to understand the applicability of existing data. Specifically, the NRC staff compared Table 22 with Tables 52 and 53 of TR EPRI-AR1(NP)-A, Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO)-Coated Particle Fuel Performance (Reference 1). Based on its review, the NRC staff determined that the KPFHR particle design is acceptable because it is in good agreement with the kernel dimensions for the AGR2 experiments.
Table 317 lists the desired fuel performance parameters for the planned KPFHR power and test reactor equilibrium cores in comparison with the AGR1 and AGR2 experiments. Since the Kairos fuel particles are designed with the AGR2 kernel specifications, the AGR2 fuel performance envelope is used to determine the operability bounds. The NRC staff notes that the desired KPFHR power reactor equilibrium core fuel performance parameters exceed the AGR2 program fuel performance parameters with the exception of time-averaged temperature.
As stated in TR Limitation #3, Kairos will develop additional supporting data per Reference 1 before extending the operability bounds. Section 3.7 details the process by which additional data will be collected to support extension of the operability bounds.
Based on the findings above regarding KPFHR fuel particle specifications and Limitation #3 as presented in Section 4.2, the NRC staff finds that the use of existing data from the AGR program (Reference 1) is acceptable for operations within the AGR2 performance envelope. As stated in Limitation #3 of Section 4.2, additional data will be necessary to support reactor operations outside of the AGR2 envelope.
Fresh Fuel Pebble Laboratory Testing As discussed in Section 2.2.2, the Kairos fuel pebble design is slightly smaller than traditional TRISO fuel pebbles and has a non-fueled low-density carbon matrix center. The next layer of the Kairos fuel pebbles is a carbon matrix with embedded TRISO fuel particles. The final (outside) layer of the Kairos fuel pebble is fuel-free matrix material. Although the Kairos pebble design is different, the TRISO fuel particle specifications are consistent with the AGR2 particle specifications.
Kairos developed a PIRT to identify fuel pebble phenomena with a low to medium knowledge base and develop a testing program to increase the knowledge base for these phenomena.
Section 3.6 summarizes the laboratory testing used to support fuel qualification of the KPFHR fuel pebbles based on the PIRT. The NRC staff reviewed mechanical, tribology, buoyancy, and material compatibility testing to determine if these tests would support a fuel qualification determination and reasonable assurance finding for the fuel pebbles to be used in the KPFHR power and test reactors.
The mechanical testing consists of compression tests and impact tests. Irradiated mechanical testing of the fuel pebbles is not planned due to the increase in strength that is expected based on publicly available experimental results from the German test program on TRISO pebble fuel (Reference 2). Similarly, the tests will be performed at room temperature due to publicly available information demonstrating that strength increases with temperature. The compression tests involve compressing fresh fuel pebbles between two steel plates until failure at room temperature.
The impact tests will be performed either by: (1) dropping a pebble onto a non-yielding surface from a height bounding all possible drop heights that could occur in the core, pebble handling systems, or during storage/transportation, or (2) accelerating a pebble to a velocity bounding all possible drop heights and striking a non-yielding target. The NRC staff reviewed References 2, 3, and open literature regarding temperature and irradiation effects on carbon and agrees that for the temperatures of interest, fresh fuel mechanical testing at room temperature would provide bounding results compared with irradiated fuel and operating temperatures. Based on the NRC staffs review of the mechanical testing methods in comparison with previous TRISO pebble testing methods, the NRC staff finds the Kairos mechanical testing methods acceptable for determining crush and impact strength for the Kairos fuel pebbles.
Kairos presented the acceptance criteria for both the crush strength and impact testing in Section 3.6.1. The NRC staff reviewed the acceptance criteria and confirmed that they would reasonably cover any expected loads and that they were consistent with historical TRISO fuel pebble acceptance criteria. Therefore, the NRC staff finds that the crush strength and impact load acceptance criteria are conservative, and they provide reasonable assurance that the pebbles will not fail as a result of anticipated mechanical loads.
Kairos states that a subset of the failed pebbles from mechanical testing will be investigated to determine the failure fraction of TRISO particles. The deconsolidation leach burn leach (DLBL) method will be used to determine the particle failure fraction, which the NRC staff notes is consistent with the method used to investigate TRISO particle failed fuel fractions in EPRI-AR1(NP)-A (Reference 1). Therefore, the NRC staff concludes that the method for determining mechanical limits of the pebbles and failed fuel fractions of the particles is acceptable.
Kairos states in Section 3.6.2, that tribology tests will be conducted on fresh pebbles to obtain coefficient of friction and bounding wear rate data for comparison with Discrete Element Modeling (DEM). This DEM will be used to confirm the pebble lifetime, single point contact wear rate. The tribology tests are designed to bound pebble degradation as a result of wear against other pebbles, stainless steel, and graphite. The NRC staff reviewed the temperature ranges, method of testing, and applied force. The procedure used is based on American Society of Testing and Materials standard G9917, Standard Test Method for Wear Testing with a Pin-on-Disk Apparatus, but the test conditions were chosen to reflect the Kairos pebble design and reactor conditions. While the maximum temperature used during testing does not bound all possible accident conditions, the NRC staff agrees with Kaiross justification in that the limited time expected above the maximum tested conditions for any accident is insufficient for excessive wear to occur. Additionally, the NRC staff reviewed publicly available data on the graphite coefficient of friction as a function of temperature and noted that the coefficient of friction at the maximum accident conditions for KPFHR is bounded by the coefficient of friction at the lower test conditions. The NRC staff reviewed the assumptions used to develop the tribology testing and wear calculations and finds that they are conservative. The NRC staff finds that: (1) the tribology testing proposed for the KPFHR power and KPFHR test reactor is acceptable, and (2) the assumptions used to model lifetime wear of pebbles (e.g., the use of a single contact point) are conservative, and therefore, acceptable.
In Section 3.6.3, Kairos presents the analytical data and experimental methods which will be used to ensure the ability of the Kairos pebble to maintain net positive buoyancy under normal operations and licensing basis event conditions, as well as identify salt intrusion characteristics for the Kairos fuel pebbles in a molten salt environment. Three phenomena affecting buoyancy of the Kairos pebbles are: (1) salt infiltration, (2) densification of graphite with irradiation, and (3) densification of graphite with increasing temperature. Densification due to temperature and irradiation has been publicly investigated and can be addressed in fuel performance modeling.
Salt infiltration into a graphite matrix has also been studied, but Kairos will perform additional testing to collect data based on Kairos materials over the expected temperature and pressure ranges.
The NRC staff reviewed the salt intrusion test parameters and notes that the temperature ranges bound the expected pebble temperatures as presented in Table 317. Additionally, the pressure used to investigate salt intrusion conservatively bounds the expected maximum pressure which is at the vessel inlet. The salt intrusion tests are performed on an unirradiated pebble which maximizes salt ingress as the carbon matrix pore size decreases with irradiation due to graphite densification. This is conservative since the maximum pebble fluence for the non-power and power KPFHRs is less than the turnaround fluence where continued densification stops, and the minimum pore size is achieved.
Based on the NRC staffs review of the buoyancy testing methods and parameters described in the TR relative to the planned reactor operating conditions, and a review of the references associated with salt intrusion and graphite density changes as a function of temperature and irradiation (References 4, 5, and 6), the NRC staff finds the method of determining buoyancy based on densification and salt intrusion characteristics for the Kairos fuel pebbles to be acceptable.
Section 3.6.4 covers material compatibility between the Kairos pebble carbon matrix and Flibe.
Flibe is a lithium fluoride, beryllium fluoride salt mixture which serves as the coolant in KPFHR reactors. Kairos presents a test plan to investigate material compatibility behavior between the Kairos pebble carbon matrix with Flibe and air. The NRC staff confirmed that the Kairos test plan parameters bound pebble surface temperatures expected during normal operation and cycle times for a pebble to pass through the core. Kairos plans to also investigate the test specimens from the salt intrusion test for material compatibility at accident conditions.
In addition to the planned testing for Kairos fuel pebble graphitic material, Kairos cites available information from the Molten Salt Reactor Experiment (MSRE), University of Wisconsin-Madison (UW), the Chinese Thorium Molten Salt Reactor Energy System-Solid Fuel (TMSR-SF), and Massachusetts Institute of Technology (MIT). The MSRE operated for 3 years in which a graphitic core was submerged in Flibe. The UW, TMSR-SF, and MIT experiments investigated salt intrusion into TRISO particles which contained a graphitic overpack. The NRC staff notes that while the MSRE, UW, TMSR-SF, and MIT data was not collected from specimens that were identical to the Kairos design, they are similar and still serve to support the data which will be collected by Kairos in its material compatibility test program.
Kairos test plans also include material compatibility with air which will be used to develop oxidation models for the Kairos fuel pebble. The NRC staff reviewed the material presented in Section 3.6.4 and finds that the test plan for the Kairos fuel pebble in combination with Kairos use of available data to investigate the interactions of Flibe and the graphitic carbon matrix of the Kairos fuel pebble are acceptable for developing appropriate correlations to model Flibe and air material compatibility with pebble matrix graphite. It should be noted that the NRC staff makes no finding as to the acceptability of the future correlations based on the Kairos tests and supporting information. The acceptability of any future correlations will be addressed in future licensing actions.
Irradiation Testing In Section 3.7, Kairos describes planned irradiation testing to support Kairos fuel pebble qualification if the planned KPFHR operations exceed the operating ranges investigated in the AGR2 program (Reference 1). The test plan covers the KPFHR equilibrium core conditions, which is the most limiting of the planned operating conditions. The NRC staff reviewed the test plan, post-irradiation examination (PIE) description, and the acceptance criteria as presented by Kairos.
The test plan involves Kairos pebble irradiation in a non-KPFHR facility within a gas environment. The NRC staff reviewed the test plan based on the information provided in Section 3.7.1 and Table 319, and finds that the test plan can successfully be used to simulate pebble exposure in equilibrium core conditions if these conditions exceed the AGR2 testing envelope because the test conditions will bound operating conditions as addressed in Limitation
However, this is addressed in Limitation #5 in Section 4.2 which states, [f]uture license applications for the commercial electric power KP-FHRs will include justification (testing or analysis based on an approved methodology) of the applicability of this methodology during rapid reactor transient events. The NRC staff notes that Limitation #5 does not apply to the non-power test reactor. The need for reactivity transient testing in a non-power reactor will be addressed in the NRC staffs safety evaluation for that facility and is identified as Limitation and Condition #1 of this safety evaluation.
In Section 3.2.2, Kairos states that while it is expected that the Kairos fuel pebble design can safely operate within the desired fuel performance parameters, KPFHR will be limited to AGR2 parameter values until the results of testing (as outlined in Section 3.7) can be completed and used to support operation within the desired fuel performance envelope. This is further discussed in TR Table 34, Limitations and Conditions from SER on TRISO Topical Report in relation to Condition 1 of the referenced EPRI TR (Reference 1).
Section 3.7 presents the testing and statistical methods which will be used to support operation of KPFHR beyond the fuel performance parameters used in the AGR2 program. The NRC staff reviewed this irradiation test plan and notes that it generally aligns with the AGR2 test methods but would operate at the desired more limiting conditions. Based on the foregoing, the NRC staff finds this test plan to be acceptable for developing a database to support future operation at the desired KPFHR equilibrium core conditions. Based on the NRC staffs review of the statistical methods described in Section 3.7.4, the NRC staff finds that the statistical methods use a well-known statistical analysis methodology, are appropriate for calculating the upper confidence bounds, and are acceptable for determining the minimum number of test fuel particles needed to support operation at the desired KPFHR equilibrium conditions.
Kairos plans to use Non-Destructive Examination (NDE) and Destructive Examination (DE) techniques as part of the PIE of the TRISO particles. The NRC staff reviewed the NDE techniques described in Section 3.7.2 and finds them to be reasonable for confirming the burnup and identifying any gross external damage of the fuel pebbles. Fuel pebble burnup is determined by gamma spectrometry and is a standard industry method to determine burnup.
Follow-on pebble destructive testing will also be used to confirm the gamma spectrometry determined burnup. Kairos also states that KPFHR reactors will be capable of evaluating two surface criteria, both of which evaluate pebble surface degradation that could expose TRISO particles to Flibe coolant. The NRC staff agrees that these are reasonable pebble surface criteria to determine if significant pebble surface degradation exists leading to potential TRISO particle/Flibe interactions.
Kairos will use DLBL techniques for DE of the TRISO particles from the irradiation experiment in order to determine the particle failure fraction. The NRC staff notes that this technique is the same technique used by the AGR program to determine the condition of the TRISO particle layers and was used to support determination of the particle failure fraction in the approved EPRI TR (Reference 1).
The NRC staff notes that the irradiation test plan presented in Section 3.7 of the TR does not include testing prior to reactor startup of fuel pebbles which have been irradiated in a molten salt environment and were manufactured using the planned production process. Experiments with TRISO compacts (as opposed to pebbles) in a gas environment have shown chemical attack of the TRISO particle SiC layer from contaminants (Fe, Cr, or Ni). For the KP-FHR design, the NRC staff observes that:
1.
KPFHR designs will control salt coolant chemistry, including minimizing oxidation potential; 2.
KPFHR designs will perform NDE of the pebbles as they transit the pebble handling systems; 3.
KPFHR designs will monitor cover gas and coolant activity for indications of TRISO particle failures; 4.
there was a lack of non-gaseous precursor fission product migration in the MSRE graphite core; and 5.
a reduction in pebble graphite matrix pore size with irradiation will reduce the migration of salt impurities which could damage the TRISO particle.
The NRC staff finds that these observations provide reasonable assurance that the fuel pebbles will have an acceptably low failure fraction. The NRC staff notes, however, that it is appropriate to seek confirmation prior to commercial reactor operation that corrosion products in the KP-FHR design do not attack the SiC layer. Therefore, the NRC staff has included Limitation and Condition #2 in this safety evaluation to address the need for confirmatory data on Kairos fuel pebble irradiation in a molten salt environment to support fuel qualification for a commercial power KPFHR.
The NRC staff notes that Limitation and Condition #2 does not apply to a non-power test reactor. A commercial power KP-FHR contains significantly more pebbles, operates at a higher power density, and will result in higher fuel burnup compared to test reactors. Further, as noted above, the NRC staff finds that there is reasonable assurance that the fuel pebbles will have an acceptably low failure fraction. Therefore, the NRC staff concludes that no additional fuel pebble testing beyond that which has already been addressed in the TR and this safety evaluation is necessary for a non-power test version of the KPFHR. The NRC staff notes that data gained from the destructive examination of fuel pebbles from a non-power test version of a KPFHR may be used to support the qualification of fuel pebbles in a commercial power KPFHR, thereby satisfying Limitation and Condition #2. This is consistent with 10 CFR 50.41(b) which permits the conduct of widespread and diverse research and development for class 104 test reactor licenses.
Based on the findings in the preceding paragraphs, the NRC staff concludes that:
given the limitations on use presented in Limitations #3 and 5 of the TR, the Kairos irradiation test plan and PIE can be used to acquire data to support the fuel qualification of the Kairos fuel pebbles for operating conditions which exceed the AGR2 test envelope;
the statistical analysis presented in the TR is acceptable for identifying the number of pebbles and particles necessary to provide the desired confidence level and to calculate the failure fraction;
the PIE methods and irradiation test acceptance criteria are acceptable for identifying failed particles;
evaluation of irradiated test reactor fuel as described in Limitation and Condition
- 2 of this safety evaluation will provide data to support full qualification of commercial power reactor fuel; and
irradiation testing described in the TR for test reactor fuel is sufficient for referencing in future test reactor operating license applications.
Fuel Performance Modeling Section 3.8 addresses fuel performance modeling of the Kairos fuel pebbles. The report addresses both TRISO particle and fuel pebble modeling. The TRISO modeling is based on KPBison code and the fuel pebble modeling utilizes DEM and Finite Element Method (FEM).
While the NRC staff is aware of these tools, the correlations built from testing have not been developed yet; therefore, the NRC staff can make no findings regarding the use of KPBison, DEM, or FEM beyond the use of any approval already obtained. Once the data have been collected from the tests as outlined in the fuel qualification TR, correlations can be either validated or modified, as necessary. The NRC staff makes no findings regarding fuel performance modeling as part of this safety evaluation; any information derived from the fuel qualification testing program which impacts the fuel performance modeling will be incorporated in a future licensing submittal.
Fuel Surveillance Program Section 3.9 presents the KPFHR fuel surveillance program, which utilizes different techniques to monitor the fuel and ensure it performs as designed. Specifically, the surveillance program includes the following:
1.
use of online cover gas and coolant activity monitoring to detect failed fuel of sufficient quantity to result in a detectible release of reactivity into the coolant and cover gas; 2.
gamma spectroscopy measurements of pebbles as they transit the pebble handling systems to detect burnup and ensure that burnup limits are not exceeded by reintroduction into the core; 3.
pebble inspections for damage which could lead to particle failure if the pebbles were reintroduced into the core; and 4.
destructive examinations in both the power and non-power KPFHR reactors of a select number of initial core pebbles to confirm other aspects of the surveillance program and to provide validation data for particle failure fractions.
Based on the NRC staffs review of the above elements of the surveillance program, the NRC staff concludes that the surveillance program is robust and can detect failed particles, pebbles at the end of life, and damaged pebbles which could lead to particle failure. Therefore, the NRC staff finds the surveillance program as described in Section 3.9, to be acceptable.
LIMITATIONS AND CONDITIONS Kairos included seven limitations on the use of this TR in Section 4.2. The NRC staff considered these limitations in the technical review presented in this safety evaluation. The NRC staff applies the following additional limitations and conditions on the acceptance of this TR:
1.
Future license applications for non-power KPFHRs shall include justification of the applicability of this methodology during rapid reactor transient events.
2.
Future operating license applications for a commercial power KPFHR shall provide information demonstrating that TRISO particle failures due to chemical attack from impurities (e.g., Fe, Cr, or Ni) in the molten salt coolant during expected reactor conditions are precluded or can be demonstrated to be insignificant. This information shall include test data from representative fuel pebbles in salt under irradiated conditions (e.g., in a test reactor).
CONCLUSION As discussed above, the NRC staff has determined that Kairos TR Fuel Qualification Methodology for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor provides an acceptable methodology to qualify fuel for the KPFHR power reactor and KPFHR test reactor based on: (1) identification of knowledge gaps pertaining to the Kairos fuel pebble, (2) a test plan to build the necessary pebble knowledge database, (3) a surveillance program to validate fuel performance, and (4) destructive examinations designed to support reactor operations which exceed the AGR2 parameters. Accordingly, the NRC staff concludes that Kairos TR Fuel Qualification Methodology for the Kairos Power Fluoride Salt-Cooled High Temperature Reactor, complies with the applicable NRC regulations and can be used to support fuel qualification for the KPFHR power and KPFHR test reactor designs, subject to the limitations and conditions discussed above.
Based on the technical evaluations presented above, the NRC staff concludes that the TR is consistent with KPFHR PDC 10 because the testing plan outlined in the TR will increase fuel behavior knowledge which supports the ability to demonstrate that specified acceptable radionuclide release design limits (SARRDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
Based on the NRC staffs technical evaluations of the Kairos fuel design, the NRC staff concludes that the TR is consistent with KPFHR PDC 16 because the use of TRISO particles, within the operating conditions allowed by the TR, provides multiple barriers internal to the reactor and its cooling system to control the release of radioactivity to the environment and to ensure that the functional containment design conditions which are safety significant are not exceeded for as long as postulated accident conditions require.
The NRC staff finds that the qualification plan outlined in the TR can be used to support a finding that 10 CFR 50.43(e) is met because: (1) the safety aspects of the Kairos fuel design are demonstrated through a combination of analysis, test programs, and experience, (2) interdependent effects among the safety features of the Kairos fuel design are acceptable, as demonstrated by analysis, appropriate test programs, experience, and (3) sufficient data exist on the safety features of the Kairos fuel design to assess the analytical tools used for safety analyses over a sufficient range of normal operating conditions, transient conditions, and specified accident sequences, including equilibrium core conditions.
The NRC staff finds that the qualification plan outlined in the TR can be used to support a finding that 10 CFR 100.11 is met by supporting the determination of the exclusion area, low population zone, and population center distance.
REFERENCES 1.
Electric Power Research Institute, Uranium Oxycarbide (UCO) Tristructural Isotropic (TRISO)-Coated Particle Fuel Performance, Topical Report EPRI-AR(NP)-A, issued November 2020 (ADAMS Accession Number ML20336A052).
2.
K. Rollig, W. Theyman, Operational Requirements of Spherical HTR Fuel Elements and their Performance. Hochtemperatur-Reactorbau GmbH. Mannheim, Federal Republic of Germany. IAEA. INIS Volume 31, Issue 41, Reference Number: 31049614.
3.
A. Schmidt, et.al., Strength Test on Spherical Elements, issued 1973.
4.
H. X. Xu et al. Characterization of Molten 2LiF-BeF2 Salt Impregnated into Graphite Matrix of Fuel Elements for Thorium Molten Salt Reactor, Nuclear Science and Techniques, Volume 30, Article 74, 2019.
5.
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Principal Contributor(s):
Christopher Van Wert Jeffrey Schmidt Antonio Barrett Dated: March 2, 2023