ML22181B159
| ML22181B159 | |
| Person / Time | |
|---|---|
| Site: | 99902069, Hermes File:Kairos Power icon.png |
| Issue date: | 06/30/2022 |
| From: | Kairos Power |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML22181B157 | List: |
| References | |
| KP-NRC-2206-014 | |
| Download: ML22181B159 (6) | |
Text
KP-NRC-2206-014 Enclosure 1 Changes to PSAR Chapters 3, 4, and 6 (Non-Proprietary)
Preliminary Safety Analysis Report Design of Structures, Systems, and Components SSCs that are nonsafety related are classified as SDC2. SDC2 SSCs are subject to the seismic design requirements of the local building code, ASCE/SEI 710 (Reference 3).
3.6.2.2.1 Seismic Qualification by Analysis Seismic qualification by analysis follows Section 8.2 of ASCE 4319. Depending on the characteristics and complexities of the subsystem or equipment, qualification by analysis is accomplished by either equivalent static analysis methods or dynamic analysis methods.
There are limitations to qualification by analysis. Per ASCE 4319:
Qualification of active electrical equipment by analysis is not performed.
Qualification of active mechanical equipment by analysis may be permitted if the component is such that the functionality during an earthquake can be established and a margin of loss of functionality during an earthquake can be quantified.
Qualification of active mechanical components by analysis shall be justified.
Seismic qualification by analysis is typically implemented for subsystems and equipment structural integrity related capacities (e.g. anchorage, pressure boundary / rupture, serviceability deformations, etc.).
3.6.2.2.2 Seismic Qualification by Testing Seismic qualification by testing follows Section 8.3 of ASCE 4319. Qualification by test is typically used for SSCs for which qualification by analysis is not permitted and for SSCs where dynamic behaviors are not sufficiently understood to support qualification by analysis.
3.6.2.3 Quality Classification The quality classification for SSCs conforms with the requirements of Kairos Powers Quality Assurance Program for the Hermes Reactor, which is discussed in Section 12.9. Safetyrelated SSCs are classified as QualityRelated, while nonsafety related SSCs are classified as Not QualityRelated. These classifications are shown in Table 3.61.
3.6.3 References
- 1. Kairos Power, LLC, Regulatory Analysis for the Kairos Power SaltCooled, High Temperature Reactor, KPTR004P, Revision 2. July 2020.
- 2. American Society of Civil Engineers, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, ASCE 4319. 2019.
- 3. American Society of Civil Engineers, Seismic Engineering Institute, Minimum Design Loads for Buildings and Other Structures, ASCE/SEI 710. 2011.
- 4. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section III, Division 5, High Temperature Reactors. 20179.
- 5. ASME, Boiler and Pressure Vessel Code,Section VIII, Divisions 1 and 2, Rules for Construction of Pressure Vessels, New York, NY. July 2017.
- 8. American Petroleum Institute, 610, Centrifugal Pumps for Petroleum, Heavy Duty Chemical, and Gas Industry Services, 1995.
- 9. American Petroleum Institute, 674, Positive Displacement PumpsReciprocating. 1995.
- 10. American Petroleum Institute, 675, Positive Displacement PumpsControlled Volume. 1994.
- 11. American Petroleum Institute, 650, Welded Steel Tanks for Oil Storage. 1998.
Kairos Power Hermes Reactor 332 Revision 0
Preliminary Safety Analysis Report Reactor Description of irradiation. The fast neutron fluence received by the reactor vessel from the reactor core and pebble insertion and extraction lines is attenuated by the core barrel, the reflector, and the reactor coolant.
Coolant purity design limits are also established in consideration of the effects of chemical attack and fouling of the reactor vessel. These features demonstrate conformance with PDC 31.
The MSS utilizes coupons and component monitoring to confirm that irradiationaffected corrosion is nonexistent or manageable. The 316H SS reactor vessel and ER1682 weld material, as a part of the reactor coolant boundary, will be inspected for structural integrity and leaktightness. As detailed in Reference 3, fracture toughness is sufficiently high in 316H SS under reactor operating conditions that additional tensile or fracture toughness monitoring and testing programs are unnecessary. These features demonstrate conformance to PDC 32.
Fluidic diodes are used to establish a flow path for continuous natural circulation of coolant in the core during postulated events to remove residual heat from the reactor core to the vessel wall. During and following a postulated event, the hot coolant from the core flows from the upper plenum through the low flow resistance direction of the fluidic diode to the cooler downcomer via natural circulation, thereby cooling the core passively. Continuous coolant flow through the reactor core prevents potential damage to the vessel internals due to overheating thereby ensuring the coolable geometry of the core is maintained. The antisiphon feature also limits the loss of reactor coolant inventory from inside the reactor vessel in the event of a PHTS breach. These features demonstrate compliance with PDC 35.
The reactor vessel reflector blocks permit insertion of the reactivity control and shutdown elements. The ETU10 grade graphite of the reflector blocks is compatible with the reactor coolant chemistry and will not degrade due to mechanical wear, thermal stresses and irradiation impacts during the reflector block lifetime. The graphite reflector material is qualified as described in the Kairos Power topical report Graphite Material Qualification for the Kairos Power Fluoride SaltCooled HighTemperature Reactor, KPTR014 (Reference 4). To preclude damage to the reflector due to entrained moisture in the graphite, the reflector blocks are baked (i.e., heated uniformly) prior to coming into contact with coolant and the reactor vessel is design to preclude air ingress. The reflectors, which act as a heat sink in the core, are spaced to accommodate thermal expansion and hydraulic forces during normal operation and postulated events. The gaps between the graphite blocks also allow for coolant to provide cooling to the reflector blocks. The reactor vessel permits the insertion of the reactivity control and shutdown elements as well. The vessel is classified as SDC3 per ASCE 4319 and will maintain its geometry to ensure the RCSS elements can be inserted during postulated events including a design basis earthquake.
These features demonstrate compliance with PDC 74.
4.3.4 Testing and Inspection The reactor vessel and internals will be included in an inservice inspection program which will be submitted at the time of the Operating License Application.
4.3.5 References
- 1. American Society of Mechanical Engineers, ASME Boiler & Pressure Vessel Code,Section III, Division 5 (2019), High Temperature Reactors. 2017.
- 2. ASCE 4319, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities.
- 3. Kairos Power, LLC, Metallic Materials Qualification for the Kairos Power Fluoride SaltCooled High Temperature Reactor, KPTR013P, Revision 1.
- 4. Kairos Power, LLC, Graphite Material Qualification for the Kairos Power Fluoride SaltCooled High Temperature Reactor, KPTR014P, Revision 1.
Kairos Power Hermes Reactor 433 Revision 0
Preliminary Safety Analysis Report Reactor Description 4.7 REACTOR VESSEL SUPPORT SYSTEM 4.7.1 Description The reactor vessel support system (RVSS) provides structural support to the reactor vessel support the full weight of the reactor vessel with fuel and coolant, vessel internals, and all headmounted components. The system transmits pressure, seismic, and thermal loads to the cavity structures during normal operation and design basis earthquakes. The RVSS provides adequate thermal management to support the vessels thermal expansion while transitioning from room temperature at assembly to nominal operating temperature for primary coolant fill. The RVSS also supports the vessels thermal expansion during postulated events.
The RVSS interfaces with the reactor vessel (see Section 4.3), the reactor thermal management system (RTMS) (See Section 9.1.5), and the safetyrelated portion of the Reactor Building (see Section 3.5). The safetyrelated portion of the Reactor Building is seismically isolated to reduce seismic loads (see Section 3.5.3).
The bottom support consists of a support tray, ledge, support columns, support pads, base plate, vessel connector, and anchoring connector as shown in Figure 4.71. All the components are made of 316H stainless steel. The reactor vessel bottom head sits directly on top of the tray and is connected to the tray by the vessel connector to prevent uplift and shear. The ledge around the edge of the tray contains spilled Flibe in case of leakage. The tray is reinforced by 316H SS support columns which are sized and spaced appropriately to provide structural support for the total weight of the vessel, vessel internals, head components, coolant, and fuel. The support columns are welded onto the support pad which allows relative sliding with the underlying base plate to accommodate thermal expansion. The support pads have slotted holes to allow relative sliding with the anchoring connectors. The anchoring connectors prevent the reactor vessel and RVSS from uplift and shear. The RVSS is designed and fabricated per ASME BPVC Section III, Division 5 (20179) (Reference 1).
The RTMS provides thermal management for the bottom support with a load bearing metallic insulation material which acts as a thermal break that reduces heat loss and cooling load for the RVSS support columns. The bottom insulation of the RTMS, as shown in Figure 4.71, protects the reactor building cavity concrete from thermal effects. The RVSS is also vertically anchored to the foundation through the bottom insulation. The bottom support insulation interface accommodates relative thermal expansion between the support columns and the insulation material.
There are no lateral seismic restraints for the reactor vessel and the headmounted components. The RVSS is designed to keep the reactor vessel from uplift and shear during seismic events. The design also leverages seismic isolation of the Reactor Building to reduce seismic effects on the reactor vessel, RVSS, and the headmounted components (see Section 4.3).
4.7.2 Design Basis Consistent with PDC 2, the RVSS can withstand the effects of natural phenomena and to perform its safety function in the event of a design basis earthquake.
Consistent with PDC 4, the RVSS accommodates the environmental conditions associated with normal operation, maintenance, testing, and postulated events.
Consistent with PDC 74, the design of the reactor structural support system ensures the integrity of the reactor vessel during postulated events to support the geometry for passive removal of residual heat from the core and to permit sufficient insertion of the control and shutdown elements providing for reactor shutdown.
Kairos Power Hermes Reactor 457 Revision 0
Preliminary Safety Analysis Report Reactor Description 4.7.3 System Evaluation The RVSS supports the reactor vessel in the event of an earthquake or other natural phenomenon thus ensuring the integrity of the reactor vessel and its ability to retain reactor coolant. The bottom support meets ASCE 4319 (2019) (Reference 2) and precludes linear buckling in the vessel support columns under static and design basis earthquake loads. The bottom support is also vertically anchored to the cavity to prevent the vessel from uplift during a design basis earthquake. The vessel connectors meet Reference 2 and provide sufficient lateral and uplift support to the vessel and the vessel top head components. The reactor cavity is also seismically isolated to reduce seismic loads. These design features demonstrate compliance with PDC 2 for the RVSS.
The RVSS is protected from discharging fluids by catch basins. Sensors and probes installed on catch basins including the bottom support tray can be used as a means of leak detection to preclude damage to the RVSS. There are no pressurized piping systems in proximity to the RVSS thus precluding by design any impacts from high energy line considerations. The RVSS accommodates the reactor vessel temperature loading cycles in combination with relevant mechanical loading cycles to ensure creep fatigue damages are precluded. The RVSS can also accommodate the growth of the reactor vessel due to thermal expansion between startup and equilibrium conditions. These design features satisfy PDC 4 for the RVSS.
PDC 74 states requires the design of the reactor vessel and reactor system shall be such that their integrity is maintained during postulated events (1) to ensure the geometry for passive removal of residual heat from the reactor core to the ultimate heat sink and (2) to permit sufficient insertion of the neutron absorbers to provide for reactor shutdown. The RVSS maintains the integrity of the reactor vessel by removing heat via the RTMS, actively during normal operation and passively during postulated events. Fission product decay heat and other residual heat from the reactor core is transferred to the reactor vessel; then to the anchored surface by the RVSS. The support columns of the RVSS are sized and spaced to maximize heat transfer between the bottom support and the environment. The thermal break between the RVSS and the reactor building provided by the bottom support insulation ensures the concrete integrity meets ACI 34913 to support maintenance and inspection of the vessel bottom head/vessel shell weld and to ensure conditions in the surrounding cavity do not exceed maximum allowable parameters. This demonstrates compliance with PDC 74 for the RVSS.
4.7.4 Testing and Inspection The RVSS temperature will be monitored during operation for conformance with design limits. The RVSS will be included in an inservice inspection program which will be submitted at the time of the Operating License Application.
4.7.5 References
- 1. American Society of Mechanical Engineers, ASME Boiler & Pressure Vessel Code,Section III, Division 5, (2019)High Temperature Reactors. 2017.
- 2. ASCE 4319, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities.
- 3. ACI 34913, Code Requirements for Nuclear SafetyRelated Concrete Structures and Commentary Kairos Power Hermes Reactor 458 Revision 0
Preliminary Safety Analysis Report Engineered Safety Features 6.3.5 References
- 1. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, Sec. III Div. 5, BPVC Section IIIRules for Construction of Nuclear Facility ComponentsDivision 5High Temperature Reactors, 20179.
- 2. American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, Sec. XI Div. 1 and 2, BPVS Section XIRules for Inservice Inspection of Nuclear Power Plant Components, 2019.
- 3. American Society of Civil Engineers, ASCE/SEI 4319, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, 2020.
- 4. American Society of Civil Engineers, ASCE/SEI 416, Seismic Analysis of SafetyRelated Nuclear Structures, 2017.
- 5. American Concrete Institute, ACI 34913, Code Requirements for Nuclear SafetyRelated Concrete Structures and Commentary, 2014.
Kairos Power Hermes Reactor 69 Revision 0