ML24270A019
| ML24270A019 | |
| Person / Time | |
|---|---|
| Site: | 99902069 |
| Issue date: | 09/25/2024 |
| From: | Bettes B NRC/NRR/DANU/UAL1 |
| To: | |
| References | |
| EPID L-2024-TOP-0022 | |
| Download: ML24270A019 (1) | |
Text
APPENDIX A: KAI ROS POWER, LLC - SAFETY ANALYSIS METHODOLOGY FOR THE KAIROS POWER FLUORIDE SALT-COOLED HIGH-TEMPERATURE TEST REACTOR TOPICAL REPORT AUDIT QUESTIONS
[Question GEN-1] The descriptions of test reactor design features (TR section 1.1.3, "KP FHR Structures, Systems, and Components"), postulated events (TR section 3.2, "Postulated Event Descriptions"), phenomena identification and ranking tables (PIRTs) (TR section 3.5, "PIRT Summary"), and KP-SAM input model (TR section 4.3, "KP-SAM Base Model") are specific to the Hermes 1 design. The NRC staff acknowledges that the significant portions of these methodology elements presented in the TR are also applicable to the Hermes 2 design. However, as identified in the safety evaluation of the Hermes 2 CP application (ML24200A114 ), there are many unique design and safety features that need to be considered for the safety analysis of the Hermes 2 design. Without the Hermes 2 design specific information, the NRC staff's review and approval of the methodology as presented in the TR will be limited to only the Hermes 1 design. Any extension of the safety analysis methodology in the TR to the Hermes 2 design would require review and approval of the design specific information. Please confirm that the scope of the methodology presented in the TR is limited to only the Hermes 1 design.
[Question GEN-2) Please describe the findings of any commercial grade dedication (CGD) studies or early gap evaluations performed for the selection of the System Analysis Module (SAM) code developed by Argonne National Laboratory. Please discuss changes implemented in the SAM code to develop KP-SAM. Please discuss the process for the implementation and verification of these updates to the SAM code. State the specific version of the KP-SAM code that is being used.
[Question GEN-4) SAM relies on multiphysics object-oriented simulation environment (MOOSE) for some of the thermal fluids modules and fluid properties it uses, but once SAM is compiled these modules and properties would have already been imported from MOOSE and would be included within the SAM executable. However, the TR states that the fluid properties are hard coded in KP-SAM. Does KP-SAM need to import any information from MOOSE for code compilation or is it independent? Also, does SAM need to interact with any of the MOOSE modules (e.g., tensor mechanics module) during execution?
[Question GEN-5] The reactor core is modeled using KP-SAM 2-D [111111111111111)) and the coolant loops are modeled using KP-SAM 1-D. Please discuss how bomains are connected while ensuring consistency, continuity, and preserving energy and momentum. Please explain if the "domain overlapping" technique developed by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) or a similar method is used to achieve the domain overlapping and any validation approaches used to validate the method.
[Question 1.1-1] Anti-Siphon features play a critical role in limiting the loss of coolant from the reactor vessel in the event of a primary heat transport system (PHTS) break. The description of this important function and the methodology for modeling its performance are
not provided in the TR. Please discuss the design and the modeling approach for the cold and hot leg anti-siphon design features. Please identify the phenomena from the PIRT tables (tables 3-1, Thermal Fluids PIRT Summary (High and Medium Importance Phenomena), through 3-3, Source Term PIRT Summary (High and Medium Importance Phenomena)) that address anti-siphon features.
[Question 2.2-1] In section 2.2.1.1, Quantification of MAR [material at risk for release]
Sources, of the TR, the provided methodology describes activation of argon (Ar) and a calculation of Ar-41 inventory based on an evaluation for a given region. Discuss what is meant by a region and how are they determined? How many regions are there? What is the sensitivity of the calculation to the region definition and use of region-averaged values?
[Question 2.2-2] What is the basis for TR equation (Eq) 2.35 regarding the natural convection mass transfer relationship between LiF-BeF2 (Flibe) and cover gas?
[Question 2.2-3] Discuss an example of the Ar-41 solubility-limited pore release model (TR sections 2.2.1.1, Argon-41, and 2.2.4.2, Argon-41 Release) calculation to give Ar-41 release fractions from the reflector and fuel graphite pores.
[Question 2.2-4] It is unclear from TR section 2.2.6, Gas Space, if the control room radiological habitability analyses will model post-accident isolation of the control room, unfiltered in leakage, and ventilation system operation. In addition, how will any potential radiation shine dose contribution be evaluated for the control room? See discussion in NRC regulatory guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 1, section 4.2, Control Room Does Consequences, for further guidance (ML23082A305).
[Question 2.2-5] The discussion of control room breathing rates and occupancy factors in TR section 2.2.6 refers to TR reference 21, which is NRC RG 1.194, Atmospheric relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants (ML031530505). RG 1.194 does not provide guidance on the control room habitability analysis assumptions on breathing rates and occupancy factors. The staff notes that TR reference 33, RG 1.183, does give such guidance. Discuss why is RG 1.194 referenced instead of RG 1.183 in TR section 2.2.6.
[Question 3.2-1] In section 3.2 of the TR, the methodology addresses seven postulated event groups. It states that the methodology for other postulated event groups will be included in a future licensing submittal. Discuss potential event categories left out of the current methodology.
[Question 3.2-2] Section 3.2.6, Pebble Handling and Storage Malfunction, of the TR states that the mechanical damage and loss of pebble handling and storage system (PHSS) cooling scenarios that could result in a release of radioactive MAR are assumed to be mitigated or precluded by design. These precluded scenarios are not discussed in the Hermes 1 CP application (ML23151A743); please explain the discrepancy.
[Question 3.2-3] Section 3.2.6 of the TR indicates that a break in the PHSS is detected by the reactor protection system (RPS) through the low pressure in the PHSS actuation signal.
Is this a safety-related signal? Is the low pressure signal used to detect other PHSS malfunctions? If not, how are other PHSS malfunctions detected?
states that the detailed discretization is managed in MOOSE and the user can control numerical method orders.
o Please discuss any limitations or assumptions for the use of this specific numerical method and discretization, as well as the spatial stabilization method.
o Please discuss justifications for the selected numerical method order.
o Please describe the approach or guidelines established to control or manage the numerical errors.
[Question 4.2-1] Table 4-1, Parameter Ranges to Address Heat Transfer in KP-FHR Test Reactor Pebble Bed, of the TR provides ranges for Reynolds number and Prandtl numbers covered in the Pebble Bed Heat Transfer SET. If available, please discuss the simulant fluid, pebble sizes, and bed porosities covered in the experiment. Please confirm that SET instrumentation will be adequate to calculate contributions to the heat transfer from the near-wall heat transfer and pebble-to-fluid heat transfer mechanisms.
[Question 4.2-2] Impact of asymmetric flow, temperature and power (neutron flux) distributions in different regions of the reactor is accounted for by using several factors such as HPF, peak vessel temperature factor, peak temperature factor, and peak reflector temperature. The following phenomena are addressed using this approach:
o Volumetric heating of structures and coolant by gamma and neutron radiation o Core flow 3-D effects; o Conjugate heat transfer from fuel to coolant and the resulting 3-D fuel temperature distribution; o Conjugate heat transfer from graphite structures to coolant and the resulting 3-D temperature distribution; and o Conjugate heat transfer from coolant and cavity to reactor vessel structure and the resulting 3-D temperature distribution.
Please discuss how these peaking factors are applied and biased. Please discuss the methodology for validating these factors. Please explain what, if any, consideration was given to the use of 3-D calculations to validate 1-D or 2-D modeling approaches.
[Question 5.6-1] Discuss how the number of tri-structural isotropic (TRISO) particles with intact silicon carbide layers that are dislodged from the oxidized fuel pebbles is determined.
[Question 5.6-2] Discuss how the TR methodology is implemented and the choices made by the analyst in the example analysis in appendix A.4, Pebble Handling and Storage System Malfunction (Transfer Line Break), including the modeling of damaged fuel pebbles and graphite dust.