ML23346A016
| ML23346A016 | |
| Person / Time | |
|---|---|
| Site: | Hermes File:Kairos Power icon.png |
| Issue date: | 12/11/2023 |
| From: | Michael Orenak NRC/NRR/DANU/UAL1 |
| To: | Peebles D, Gardner D Kairos Power |
| References | |
| EPID L-2023-CPS-0000 | |
| Download: ML23346A016 (1) | |
Text
From:
Michael Orenak To:
Darrell Gardner; Drew Peebles Cc:
Pravin Sawant; Matthew Hiser; Edward Helvenston; William Jessup; Candace de Messieres
Subject:
Hermes 2 Audit Questions 3.5-1, 4.3-1, 5.2-1, 7.4-1, 9.8-1, 11.1-1, 13.1-1, 13.1-2, 13.1-3, 13.1-4, 13.1-5, 13.1-6, 13.2-1, 14.1-1, and 14.1-2 Date:
Monday, December 11, 2023 4:20:00 PM Darrell and Drew,
Below is the fifth set of audit questions for the Hermes 2 General Audit.The NRC staff would like to have a discussion regarding these questions when Kairos is ready. Kairos is welcome to post answers to these questions on the Kairos electronic reading room, but no written answers are requested by the NRC staff at this time.
PSAR Section Request Number Request/Question 3.5, Plant Structures 3.5-1 KP-FHR PDC 4, as modified by the discussion in PSAR Section 3.1, Introduction, states that safety-related structures, system, and components (SSCs) shall be designed to be compatible with the environmental conditions associated with accidents. Since Hermes 2 has a power generation system that results in steam piping in the non-safety related portion of the reactor building, protecting the safety-related reactor building from adverse effect of steam line break is important. However, PSAR Section 3.5.2, Design Bases, does not include PDC 4 within the design bases for the safety-related portion of the reactor building. Please clarify the design basis for the protection of safety-related SSCs from environmental effects of a steam break in the non-safety related portion of the reactor building (e.g.,
distance or barriers). Additionally, address the effects of potential pressurization of the non-safety related portion of the reactor building on the conditions within the safety-related portion of the reactor building.
4.3, Reactor Vessel System and 4.7, Reactor Vessel Support System 4.3-1 Table 4.3-2, Load Combinations for the Reactor Vessel System, and Table 4.7-1, Load Combinations for the Reactor Vessel Support System, provide the load combinations for the reactor vessel (RV) system and reactor vessel support system (RVSS) and show that loads of service level A/B/C/D are considered. Additionally, the fatigue analyses for the RV system and RVSS might be impacted by the difference between Hermes 1 and Hermes 2 considering the longer operation life for Hermes 2:
- a. Provide calculations or documents on the
Kairos electronic reading room, if available, that discuss load development methodology for loads of different service levels.
- b. Discuss how the fatigue analyses will be updated for 11 years of operation.
5.2, Intermediate heat Transport System 5.2-1 PSAR Table 3.6-1, Structures, Systems, and Components, appears to show several subsystems of the intermediate heat transport system (IHTS) that are new to the proposed Hermes 2 facility, as compared to the proposed Hermes 1 facility. Clarify whether these subsystems (e.g., intermediate chemistry control, intermediate inert gas system) are unique systems that are separate from the auxiliary systems described in Chapter 9, Auxiliary Systems, of the preliminary safety analysis report (PSAR). For any subsystem that is separate and unique from those described in Chapter 9, provide the following information (which is discussed for all other facility systems listed in Table 3.6-1):
- a. Description of the system
- b. Design bases for the system
- c. A description for how the system meets its design bases
- d. Any testing and inspection requirements for the system
- e. Place the system requirements document on the portal for NRC staff audit 7.4.3.1, Main Control Room 7.4-1 Anhydrous hydrogen fluoride is used for tritium management (see PSAR Section 9.1.3, Tritium Management System). Section 2.7, Storage, Treatment, and Transportation of Radioactive and Nonradioactive Materials, of the Hermes 2 Environmental Report states that anhydrous hydrogen fluoride would be stored in the IHTS tritium management system. Section 4.8.1, Nonradiological Impacts, of the Hermes 2 Environmental Report states that quantities of hydrogen fluoride would be maintained at less than the 29 CFR 1910.119 threshold quantity of 1,000 pounds. No specific discussion is provided in PSAR Section 2.2.3.3, Evaluation of Airport Hazards and Helicopter Operations, regarding the potential hazards of anhydrous hydrogen fluoride toxicity, its ability to affect Main Control Room personnel, or protections against it. Please discuss how protection against the toxicity of anhydrous hydrogen fluoride is addressed with respect to control room habitability, including considerations for control room and ventilation system design.
9.8.4, Cranes and Rigging 9.8-1 Section 9.8.4.2, Design Bases, of the Hermes 2 PSAR states that, consistent with PDC 4, the crane and rigging are designed to protect against the dynamic effects potentially created by the failure of the crane and rigging equipment.
However, the discussion in Section 9.8.4.3, "System Evaluation," emphasizes administrative controls instead of discussing design features and states that "administrative controls and interlocks prevent the crane and rigging from moving heavy loads over safetyrelated SSCs except when the reactor is shut down..." Please clarify the extent to which the design of the crane and rigging protect against dynamic effects of their failure. Also, please clarify protection of SSCs performing safety-related functions during reactor shutdown from the dynamic effects of crane or rigging failure.
11.1, Radiation Protection 11.1-1 Please clarify whether activation products are produced in the IHTS salt or cover gas (in addition to those transported from the Primary Heat Transport System).
Chapter 13, Accident Analysis 13.1-1 The maximum hypothetical accident (MHA) source term, functional containment, and consequence analysis modeling effectively neglect transport or retention in the IHTS and Power Generation System (PGS). Discuss how the consequences of the MHA will be shown to be bounding for the events with expected releases from or failure of the IHTS or PGS.
Chapter 13, Accident Analysis 13.1-2 Table 3-2, Derived Figures of Merit and Acceptance Criteria for Postulated Events, of KP-TR-022, Hermes 2 Postulated Event Analysis Methodology, identifies an intermediate heat exchanger tube break as an applicable event, but Table 13.1-1, Acceptance Criteria for Figures of Merit, of the Hermes 2 PSAR does not. Clarify the reason for this discrepancy.
13.1.1, Maximum Hypothetical Accident 13.1-3 PSAR Section 9.9.1.2, System Evaluation, states that the steam system contains radiological contaminants, including tritium. It is not clear from Table 13.1-1 (nor from KP-TR-022) which figures of merit (FOM) would assure that events with releases due to steaming from the main steam power relief valves or safety valves would be shown to be bounded by the MHA.
13.1.1, Maximum Hypothetical Accident and 13.2.1, Maximum Hypothetical 13.1-4 It is not clear from Table 13.1-1 (nor from KP-TR-022) which FOM would assure that events with failure of the IHTS would be shown to be bounded by the MHA. Please discuss whether the FOM for the Flibe salt spill event are applicable to or shown
Accident to be bounding for the BeNaF salt in the IHTS (e.g., airborne release fraction of spilled/splashed Flibe).
13.1.1, Maximum Hypothetical Accident and 13.2.1, Maximum Hypothetical Accident 13.1-5 The discussion of the amount of radioactive material at risk for release (MAR) for the MHA in Hermes 2 PSAR Sections 13.1.1 and 13.2.1 does not include the IHTS or the PGS. Clarify how the MHA consequence analysis remains bounding when not modeling the potential MAR in these systems and transport through these systems.
13.1.6, Radioactive Release from a Subsystem or Component 13.1-6 Hermes 2 PSAR Section 13.1.6 states that there is a design requirement on the MAR available for release from the subsystems and components to remain below the MAR for release assumed in the MHA. Discuss whether these MAR limitations consider the concurrent failure of the relevant systems in both Hermes 2 units, considering that the limiting event for the category is assumed to be a seismically initiated event. Additionally, discuss if these limitations also include consideration of the concurrent failure of the relevant systems in the Hermes 1 test reactor located nearby.
PSAR Section 13.2.1, Maximum Hypothetical Accident 13.2-1 Section 13.2.1.1 of the Hermes 2 PSAR states that the quantity of retained tritium is conservatively bound within the graphite and structures over 10 years of operation. With the understanding that the Hermes 2 units would be licensed for 11 years of operating life (10 effective full power years) as described in the Hermes 2 Environmental Report, discuss whether this assumption is bounding for the tritium retention for the expected full licensed operating life and conditions.
Chapter 14, Technical Specifications 14.1-1 In Table 14.1-1, Proposed Variables and Conditions for Technical Specifications, one of the Limiting Conditions for Operation (LCOs) under Section 3.3 is proposed to limit the quantities of MAR in the primary heat transport system cover gas, the IHTS, and the PGS. Clarify whether this LCO would include MAR quantities for the IHTS cover gas and the coolant, and if there would be separate values for the IHTS cover gas and the coolant, or if it is a combined value for the entire system boundary. Also, it is not clear if the quantities of MAR for the PGS are activity concentrations in the water and steam.
Chapter 14, Technical Specifications 14.1-2 Section 13.1.10.11 of the PSAR states that superheater leaks are expected to be limited by technical specification. Please confirm that the LCO presented in Table 14.1-1 under Section 3.3
[t]he quantity of water in the intermediate coolant shall be maintained below an upper bound limit is specified to limit superheater leaks.
If you have any questions or need clarifications on the questions before the discussion, please do not hesitate to contact me. This email will be added to ADAMS and will be made public.
Mike
Michael D. Orenak, Project Manager Advanced Reactor Licensing Branch 1 (UAL1)
NRR - Division of Advanced Reactors and Non-Power Production and Utilization Facilities (DANU) 301-415-3229 Michael.Orenak@nrc.gov