ML22062B686

Enclosure 1: Kairos Power Response to NRC Question Q14

ML22062B686
Person / Time
Site:	Hermes
Issue date:	03/03/2022
From:	Kairos Power
To:	Office of Nuclear Reactor Regulation
Shared Package
ML22062B684	List: ML22062B685 ML22062B686
References
KP‐NRC‐2203‐002
Download: ML22062B686 (19)

Text

Enclosure 1 Kairos Power Response to NRC Question Q14 (NonProprietary)

Question # 2322Q14 Question Number: 2322Q14 PSAR Section 4.3 states The antisiphon feature also limits the loss of reactor coolant inventory from inside the reactor vessel in case of a [primary heat transport system (PHTS)] breach. Describe the antisiphon feature that limits the loss of vessel inventory in the event of a break in the PHTS.

Could a leak in the piping connected to the defueling chute or the pebble insertion line establish a siphon and result in a loss of coolant within the core?

Kairos Power Response:

Design features are found in fluid systems connected to the reactor vessel to limit loss of coolant inventory in the event of a break in the primary heat transport system (PHTS), the inventory management system (IMS) and the pebble handling and storage system (PHSS). PSAR Sections 5.1, 9.1.4, and 9.3 have been clarified as shown below. Conforming changes have been made to Chapter 3 and 13.

The casing design of the primary salt pump (PSP) sets the inlet elevation of the antisiphon surface for the hot leg should a leak occur in the external portion of the PHTS. This antisiphon feature limits the loss of reactor coolant inventory from inside the reactor vessel in the event of a PHTS break or in breaches in the IMS piping connected to the PHTS.

The IMS interfaces with the reactor vessel (RV) through the RV fill/drain transfer line and the RV level management line. During normal operation, when the reactor is fueled, the RV fill/drain transfer line is equipped with passive RV isolation features such as caps, flanges and/or a transfer line disconnect, designed to preclude inadvertent reactor coolant draining from the RV by siphoning.

During RV fill/drain operations, when the reactor vessel is defueled, and the RV fill/drain transfer line is connected, an isolation valve is used to interrupt the reactor coolant flow and a cover gas inlet is used to break the siphon in the transfer lines. The configuration of the RV level management lines short dip tube and overflow weir precludes inadvertent reactor coolant draining from the RV by design.

The PHSS interfaces with the reactor vessel at the PEM and the pebble insertion line. The elevation of the pebble extraction machine (PEM) relative to the Flibe free surface is such that coolant inventory loss from the reactor vessel in the event of breaks in the PEM would be limited. The pebble insertion line is designed with overflow protection cutouts to direct any coolant in the insert line back down into the reactor vessel. The pebble insertion line is designed to limit inventory loss such that reactor coolant level is no lower than the elevation of the primary salt pump intake, in the event of a break in the insertion line.

In preparation of this response, Kairos Power noted that Section 9.1.4 of the PSAR incorrectly cited PDC 15 as a design basis. Figure 9.1.41 also incorrectly referenced the PHX which was removed per Letter KPNRC2202002. The attached changes make these corrections.

References:

None Impact on Licensing Document:

This response impacts Tables 3.13, 3.61, and Sections 4.3, 5.1, 9.1.4, 9.3, and 13.1 of the Kairos Power Preliminary Safety Analysis Report. A markup of the affected sections is provided with this response.

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Preliminary Safety Analysis Report Design of Structures, Systems, and Components Table 3.13: Principal Design Criteria Principal Design Criteria SAR Section PDC 1, Quality Standards and Records 3.5, 4.3, 6.3, 7.3, 7.4, 7.5 PDC 2, Design bases for protection against natural phenomena 3.5, 4.2.2, 4.3, 4.7, 5.1, 6.3, 7.3, 7.4, 7.5, 8.2, 8.3, 9.1.1, 9.1.2, 9.1.3, 9.1.4, 9.1.5, 9.2, 9.3, 9.4, 9.7, 9.8.2, 9.8.4, 9.8.5, 11.2 PDC 3, Fire Protection 6.3, 7.3, 7.5, 9.3, 9.4 PDC 4, Environmental and dynamic effects design bases 4.2.2, 4.3, 4.7, 6.3, 7.3, 9.1.2, 9.1.4, 9.3, 9.7, 9.8.2 9.8.4 PDC 5, Sharing of structures, systems, and components 3.1 PDC 10, Reactor Design 4.2.1, 4.3, 4.5, 4.6, 5.1, 6.3, 7.3 PDC 11, Reactor Inherent Protection 4.5 PDC 12, Suppression of reactor power oscillations 4.5, 4.6, 5.1 PDC 13, Instrumentation and Control 7.2, 7.3, 7.5, 9.1.3 PDC 14, Reactor Coolant Boundary 4.3 PDC 15, Reactor coolant system design 7.3, 9.1.4 PDC 16, Containment design 4.2.1, 5.1 PDC 17, Electric Power systems 8.2, 8.3 PDC 18, Inspection and testing of electric power systems 8.2, 8.3 PDC 19, Control room 7.4 PDC 20, Protection system functions 7.3 PDC 21, Protection system reliability and testability 7.3, 7.5 PDC 22, Protection System Independence 7.5 PDC 23, Protection system failure modes 4.2.2, 7.3 PDC 24, Separation of protection and control systems 7.5 PDC 25, Protection system requirements for reactivity control 7.3 malfunctions Kairos Power Hermes Reactor 36 Revision 0

Preliminary Safety Analysis Report Design of Structures, Systems, and Components Principal Design Criteria SAR Section PDC 26, Reactivity control systems 4.2.2 4.5 PDC 28, Reactivity limits 4.2.2, 7.3 PDC 29, Protection against anticipated operation occurrences 4.2.2, 7.3, 7.5 PDC 30, Quality of reactor coolant boundary 4.3 PDC 31, Fracture prevention of reactor coolant boundary 4.3 PDC 32, Inspection of reactor coolant boundary 4.3 PDC 33, Reactor coolant inventory maintenance 5.1, 9.1.4, 9.3 PDC 34, Residual heat removal 4.6, 6.3 PDC 35, Passive residual heat removal 4.3, 4.6, 6.3 PDC 36, Inspection of passive residual heat removal system 6.3 PDC 37, Testing of passive residual heat removal system 6.3 PDC 44, Structural and equipment cooling 9.1.5, 9.7 PDC 45, Inspection of structural and equipment cooling systems 9.1.5, 9.7 PDC 46, Testing of structural and equipment cooling systems 9.1.5, 9.7 PDC 60, Control of releases of radioactive materials to the 5.1, 5.2, 9.1.3, 9.2, 11.2 environment PDC 61, Fuel storage and handling and radioactivity control 9.3 PDC 62, Prevention of criticality in fuel storage and handling 9.3 PDC 63, Monitoring fuel and waste storage 9.3, 11.2 PDC 64, Monitoring radioactivity releases 9.1.2, 9.1.3, 9.2 PDC 70, Reactor coolant purity control 9.1.1 PDC 71, Reactor coolant heating systems 9.1.5 PDC 73, Reactor coolant system interfaces 5.2 Kairos Power Hermes Reactor 37 Revision 0

Preliminary Safety Analysis Report Design of Structures, Systems, and Components SSC Name Safety Seismic Quality Program SAR Section Plant Area Classification Classification DHRS Makeup Water SSCs Nonsafety related SDC2 Not QualityRelated 6.3 SR and NSR areas Pebble Handling and Storage System (PHSS)

New Pebble Insertion SSCs Nonsafety related SDC2 Not QualityRelated 9.3 SR and NSR areas Pebble Extraction Machine Nonsafety related SDC2 Not QualityRelated 9.3 SR area Pebble Processing SSCs Nonsafety related SDC2 Not QualityRelated 9.3 SR area Pebble Inspection SSCs Nonsafety related SDC2 Not QualityRelated 9.3 SR area Debris Removal SSCs Nonsafety related SDC2 Not QualityRelated 9.3 SR and NSR areas Pebble Insertion Machine Nonsafety related SDC2 Not QualityRelated 9.3 SR area Full Core Offload and Spent Safetyrelated SDC3 QualityRelated 9.3 SR area Fuel Storage Rack Canister Transporter Nonsafety related SDC2 Not QualityRelated 9.3 SR area Spent Fuel Air Cooled Safetyrelated SDC3 QualityRelated 9.3 SR area Storage Rack Spent Fuel Storage Nonsafety related SDC2 Not QualityRelated 9.3 SR area Canisters Primary Heat Transport System (PHTS)

Primary Salt Pump Nonsafety related SDC2 Not QualityRelated 5.1.1 SR area Primary Heat Exchanger Nonsafety related SDC2 Not QualityRelated 5.1.1 SR area Primary Loop Piping Nonsafety related SDC2 Not QualityRelated 5.1.1 SR area System Primary Loop Auxiliary Nonsafety related SDC2 Not QualityRelated 5.1.1 SR area Heating Reactor Coolant Safetyrelated N/A QualityRelated 5.1.1 SR area AntiSiphon Feature Safetyrelated SDC3 Quality Related 5.1.1 SR area Primary Heat Rejection System (PHRS)

Intermediate Salt Pump(s) Nonsafety related SDC2 Not QualityRelated 5.2 NSR area Kairos Power Hermes Reactor 335 Revision 0

Preliminary Safety Analysis Report Reactor Description 4.3 REACTOR VESSEL SYSTEM 4.3.1 Description This section provides an overview of the reactor vessel system (see Figure 4.31) which includes the reactor vessel and the reactor vessel internals. The reactor vessel forms a major element of the reactor coolant boundary and the inert gas boundary. The reactor vessel and vessel internals define the flow path for reactor coolant and fuel into the core. The reactor vessel system contains the reactor core and provides for circulation of reactor coolant and pebbles as well as insertion of the reactivity control and shutdown elements through the reactor core.

The reactor vessel system provides a flow path for reactor coolant to transfer heat from the reactor core to the primary heat transport system (PHTS) during normal operations. The reactor coolant enters the reactor vessel through two side inlet nozzles and flows downward through a downcomer annulus formed between the metallic core barrel and the reactor vessel shell. Coolant flow moves through the reflector support structure and is distributed into the core by the design of the reflector blocks. Upon exiting the core, the coolant leaves the reactor vessel via the primary salt pump (PSP) (see Section 5.1.1) which draws suction directly from a pool of reactor coolant above the core and inside the vessel. An antisiphon feature is provided to limit loss of vessel inventory in the event of a break in the PHTS.

Design features are provided in fluid systems connected to the reactor vessel to limit loss of coolant inventory in the event of a break in those systems as described in Sections 5.1, 9.1.4, and 9.3.

The reactor vessel system also provides a flow path for pebbles to allow online refueling and defueling of the reactor core by the pebble handling and storage system (PHSS) (Section 9.3) during normal operation. The PHSS inserts pebbles into the reactor vessel and delivers them to the fueling chute below the reactor core by the pebble insertion line (Section 9.3.1). The buoyant pebbles float upward, and pebbles inserted via the insertion line will join the packed pebblebed in the reactor core. Upon circulating through the core, the pebbles accumulate in the defueling chute at the top of the reactor core. The pebble extraction machine (PEM) (Section 9.3.1) at the top of the reactor core removes pebbles from the reactor vessel (see Figure 4.32.)

During postulated events when the PHTS and the primary heat rejection system (PHRS) are not available, the reactor vessel provides an alternative flow path as discussed in Section 4.6.1 to allow natural circulation of the reactor coolant to remove heat from the reactor core. The reactor coolant leaving the core flows back into the downcomer annulus via fluidic diodes. The heat from the core is transferred to the reactor vessel shell which transfers the heat to the decay heat removal system (DHRS)

(Section 6.3).

The reactor vessel system interfaces with fuel (Section 4.2.1), primary heat transport system (PHTS)

(Section 5.1), reactivity control and shutdown system (RCSS) (Section 4.2.2), reactor vessel support system (RVSS) (Section 4.7), decay heat removal system (DHRS) (Section 6.3), pebble handling and storage system (PHSS) (Section 9.3), reactor thermal management system (RTMS) (Section 9.1.5), inert gas system (IGS) (Section 9.1.2), inventory management system (IMS) (Section 9.1.4), and instrumentation and controls (Chapter 7).

4.3.1.1 Reactor Vessel The reactor vessel is a vertical cylinder design with flat top and bottom heads. The vessel houses the reactor vessel internals. The reactor vessel shell and bottom head provide a major element of the reactor coolant boundary. The vessel is constructed of 316H stainless steel (SS) with ER1682 weld metal and is designed and fabricated per ASME BPVC Section III, Division 5 (Reference 1). It contains the inventory of reactor coolant such that the reactor core is covered by the coolant during normal Kairos Power Hermes Reactor 428 Revision 0

Preliminary Safety Analysis Report Reactor Description into the cover gas and precludes corrosion of the internals. The high temperature, high carbon grade 316H SS of the core barrel and reflector support structure have high creep strength and are resistant to radiation damage, corrosion mechanisms, thermal aging, yielding, and excessive neutron absorption.

Vessel fluence calculations, as described in Section 4.5, confirm adequate margin relative to the effects 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. ASME Boiler & Pressure Vessel Code,Section III, Division 5 (2019)

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.

Kairos Power Hermes Reactor 433 Revision 0

Preliminary Safety Analysis Report Heat Transport Systems functional containment and enhanced safety. The circulating activity of the reactor coolant is sampled (see Section 9.1.1) to remain within limits established in the technical specifications.

5.1.1.2 Primary Salt Pump The PSP is a variable speed, cartridge style pump located on the reactor vessel head that controls system flow rate and pressure in the PHTS under normal operation. The PSP circulates the reactor coolant between the reactor core, where the Flibe is heated as it contacts with the fuel, and the PHX where the heat is transferred to the PHRS. PHTS flow rates are varied based on maintaining a specified temperature change across the core as thermal output changes and maintaining positive pressure differential between the primary and intermediate coolants in the PHX. The design of the PSP operates continuously at full thermal power flow rates and temperatures, as well as at reduced power and flow rates.

The cantilever pump design extends the shaft down into the reactor coolant while keeping the bearings and seals in a lower temperature region above the coolant. The pump flow discharges horizontally above the reactor vessel head and has a highpoint vent that is used for vacuum fill. The pump has a positive pressure inert gas space with a purge gas flow which discharges into to the reactor vessel cover gas space. The pump motor rotor is directly mounted on the shaft and operates in the cover gas environment, eliminating the need for conventional shaft seals and providing a hermetic boundary for cover gas. The inert gas system is described in Section 9.1.2.

The design of the pump suction controls and prevents entrainment of cover gas at normal submergence levels. Residual gas in the PHTS at start up is removed by deentrainment locations in the upper reflector. The pump casing design sets the inlet elevation of the antisiphon surface for the hot leg should a leak occur in the external portion of the PHTS, and for when the external PHTS piping is drained. The antisiphoning function is described in Section 4.3.

5.1.1.3 Primary Heat Exchanger The PHX serves as the heat transfer interface and coolant boundary between the PHTS and PHRS. The PHX does not perform any safetyrelated functions. The reactor coolant is circulated from the PSP outlet nozzle through the primary piping before it enters the PHX, where the heat is transferred from the reactor coolant to intermediate coolant on the cooling side. The reactor coolant enters the PHX at approximately 600650°C and leaves the PHX at approximately 550°C during normal, steadystate operation at full power. After transferring its heat, the reactor coolant leaves the outlet nozzle of the PHX and is returned to the inlet nozzle of the reactor vessel.

The PHX design and location assures that a positive pressure differential is maintained between the reactor coolant and intermediate coolant under all normal operation and normal transient conditions, so that tube leakage is from the PHTS to the PHRS. The PHX is located at an intermediate elevation between the reactor vessel coolant free surface and the PHRS coolant free surface to assure positive pressure differential due to hydrostatic head under shut down conditions. PHTS and PHRS pump speeds are controlled to maintain positive pressure differential under all normal operating modes. A pressure differential measurement between the PHX intermediate coolant inlet pressure (highest intermediate coolant pressure) and reactor coolant outlet pressure (lowest reactor coolant pressure) monitors for positive pressure differential and initiates trips of the PHTS and PHRS pumps if the differential falls below a predetermined limit.

Kairos Power Hermes Reactor 52 Revision 0

Preliminary Safety Analysis Report Heat Transport Systems 5.1.1.4 Primary Loop Piping The primary loop piping consists of the interconnecting piping and small components not specifically allocated within the other architectural elements. This includes cross connection piping, valves, and interfaces with the inventory management system.

The primary loop piping does not perform any safetyrelated functions and is not credited to mitigate the consequences of postulated events.

The PHTS piping is designed to the ASME B31.3 Code and accommodates the reactor coolant temperature, pressure, and corrosion properties. The section of piping from the PSP discharge to the PHX inlets is termed the hot leg and the section of piping from the PHX outlets to the reactor vessel inlet is termed the cold leg. An antisiphon feature is implemented in the design that can break the siphon from the reactor vessel if a leak in the PHTS occurs. This feature is discussed in Section 4.3.

5.1.1.5 Primary Loop Auxiliary Heating The auxiliary heating function provides nonnuclear heating to the PHTS as needed for various operations including initial coolant melt, plant startup and shutdown, and supplemental heating during normal operation. Auxiliary heating maintains the PHTS piping and PHX above the trace heating setpoint temperature. The source of the heat depends on the subsystem or component requiring the heat. For example, electrical heating is used in some areas of the plant that would be susceptible to coolant freezing with the use of insulation alone. Sufficient heating is provided to maintain reactor coolant temperature in external piping and PHX above freezing throughout the filling, operation, and draining processes.

5.1.1.6 Normal Shutdown Cooling The PHTS provides normal shutdown cooling following plant trips. The transition from power operation to normal shutdown cooling involves a programed rundown of the PSP and intermediate salt pump speeds, to minimize the thermal transient experienced by the reactor vessel and PHTS. Normal shutdown cooling uses the PHRS as the heat sink.

5.1.2 Design Basis Consistent with PDC 2, the safetyrelated SSCs located near the PHTS are protected from the adverse effects of postulated PHTS failures during a design basis earthquake.

Consistent with PDC 10, the design of the reactor coolant supports the assurance that specified acceptable system radionuclide release design limits (SARRDLs) are not exceeded during any condition of normal operation, as well as during any unplanned transients.

Consistent with PDC 12, the design of the reactor coolant, in part, ensures that power oscillations cannot result in conditions exceeding specified acceptable SARRDLs.

Consistent with PDC 16, the design of the reactor coolant, in part, provides a means to control the release of radioactive materials to the environment during postulated events as part of the functional containment design.

Consistent with PDC 33, the design of the PHTS includes antisiphon features to maintain reactor coolant inventory in the event of breaks in the system.

Consistent with PDC 60, the design of the PHTS supports the control of radioactive materials during normal reactor operation.

Kairos Power Hermes Reactor 53 Revision 0

Preliminary Safety Analysis Report Heat Transport Systems Consistent with 10 CFR 20.1406, the design of the PHTS, to the extent practicable, minimizes contamination of the facility and the environment, and facilitate eventual decommissioning.

5.1.3 System Evaluation The design of the nonsafetyrelated PHTS is such that a failure of components of the PHTS does not affect the performance of safetyrelated SSCs due to a design basis earthquake. In addition to protective barriers, the PHTS pipe connections to the reactor vessel nozzles have sufficiently small wall thickness, such that if loaded beyond elastic limits, inelastic response occurs in the PHTS piping which is nonsafety related. These features, along with the seismic design described in Section 3.5, demonstrate conformance with the requirements in PDC 2 for the PHTS.

While the PHTS is a closed system, there are conceivable scenarios that may result in the release of radioactive effluents. The fuel design locates the fuel particles near the periphery of the fuel pebble, enhancing the ability of the fuel to transfer heat to the coolant. The thermal hydraulic analysis of the core (see Section 4.6) ensures that adequate coolant flow is maintained to ensure that SARRDLs, as discussed in Section 6.2, are not exceeded. These features demonstrate conformance with the requirements in PDC 10.

The design of the reactor coolant, in part, ensures that power oscillations cannot result in conditions exceeding SARRDLs. The reactor is kept near ambient pressure and the reactor coolant in the PHTS does not experience two phase flow. The coolant has a high thermal inertia making the reactor resilient to thermalhydraulic instability events. These features, in part, demonstrate conformance with the requirements in PDC 12.

The functional containment is described in Section 6.2. The design relies primarily on the multiple barriers within the TRISO fuel particles to ensure that the radiological dose at the exclusion area boundary as a consequence of postulated events meets regulatory limits. However, the reactor coolant also serves as a distinct physical barrier for fuel submerged in Flibe by providing retention of fission products that escape the fuel. The design of the reactor coolant composition provides, in part, a means to control the accidental release of radioactive materials during normal reactor operation and postulated events (PDC 60), and supports, in part, demonstration of the functional containment aspects.

The design aspects of the reactor coolant are discussed in Reference 5.1.51. The Flibe also accumulates radionuclides from fission products, and transmutation products from the Flibe and Flibe impurities. The retention properties of the Flibe are credited in the safety analysis as a barrier to release of radionuclides accumulated in the coolant, and radionuclide concentration is limited by technical specifications. The transport of radionuclides through Flibe is based on thermodynamic data that will be justified in the application for an Operating License. These features demonstrate conformance with the requirements in PDC 16.

The PSP casing design sets the inlet elevation of the antisiphon surface for the hot leg should a leak occur in the external portion of the PHTS. This antisiphon feature limits the loss of reactor coolant inventory from inside the reactor vessel in the event of a PHTS breach or in breaches of inventory management system piping connected to the PHTS (see Section 9.1.4.) These antisiphon features demonstrate compliance with PDC 33.

The design of the PHTS controls the release of radioactive materials in gaseous and liquid effluents in the event the PHTS working fluid is inadvertently released to the atmosphere via leaks in the piping system. The PHTS SSCs that are part of the reactor coolant boundary are designed to the ASME B31.3 Code (for the piping) and Section VIII (for the PHX) such that leaks are unlikely. Means are provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage in the PHTS SSCs. A postulated event in the PHTS would be a PHX tube failure. This event would cause Flibe Kairos Power Hermes Reactor 54 Revision 0

Preliminary Safety Analysis Report Auxiliary Systems 9.1.4.1.2 RV Fill/Drain Tank The RV fill/drain tank provides a means of filling and draining the RV through a transfer line and a dip tube. The transfer into the RV is pump driven and the transfer out of the RV is gravity driven. The RV fill/drain tank transfer line is equipped with a passive RV isolation system to prevent unintentional draining, which is discussed in Section 9.1.4.3. The RV fill/drain tank is sized to hold the RV coolant inventory.

9.1.4.1.3 PHTS Fill/Drain Tank The PHTS fill/drain tank provides a means of filling and draining the reactor coolant from the PHTS (see Section 5.1), including the primary heat exchanger (PHX), through a transfer line. The PHTS drain is gravity driven and the fill is pump driven between the PHTS fill/drain tank and the PHTS.

The PHTS fill/drain tank is sized to hold the PHTS and PHX reactor coolant inventory.

9.1.4.1.4 Solid IMS New and used reactor coolant is stored in transfer canisters used to transport reactor coolant to and from the site in solid state at ambient temperature. Within the IMS, the reactor coolant is transferred -

in liquid form - through transfer lines, driven by a cover gas pressure differential. The solid IMS function is to melt new reactor coolant in the canisters prior to a transfer into the IMS or to freeze the used reactor coolant in the canisters following a transfer from the IMS. The used reactor coolant presents a potential hazard due to radiological contamination.

The transfer canisters are constructed of stainlesssteel and are designed per ASME BPVC,Section VIII.

The transfer canisters are designed and fabricated to meet the pressure, mechanical loads, corrosion, and temperature requirements of the system.

9.1.4.2 Design Bases Consistent with PDC 2, safetyrelated SSCs located near the IMS are protected from the adverse effects of IMS failures during a design basis earthquake.

Consistent with PDC 4, safetyrelated SSCs located near the IMS are protected from the adverse effects of IMS failures during dynamic events.

Consistent with PDC 15, the IMS is designed to ensure the design conditions of the reactor coolant boundarys safetyrelated elements are maintained during normal and accident conditions.

Consistent with PDC 33, the design of the IMS includes design features to limit the sufficient reactor coolant inventory is provided to protect against a loss of reactor vessel coolant inventory in the safety related portions of the reactor coolant boundaryevent of breaks in the system.

Consistent with 10 CFR 20.1406, the IMS is designed, to the extent practicable, to minimize contamination of the facility and the environment, and facilitate eventual decommissioning.

9.1.4.3 System Evaluation The IMS does not perform safetyrelated functions and is not credited for the mitigation of postulated events. The system is also not credited for performing safe shutdown functions. The system is not credited to maintain the integrity of the reactor coolant pressure boundary.

Portions of the IMS may be located in proximity to SSCs with safetyrelated functions. Those safety related SSCs are protected from failure of the IMS during a design basis earthquake by either seismically Kairos Power Hermes Reactor 916 Revision 0

Preliminary Safety Analysis Report Auxiliary Systems mounting the applicable IMS components, physical separation, or barriers to preclude adverse interactions. The IMS is designed to preferentially fail in a way that does not impact the RV system. This satisfies PDC 2 for the IMS.

The IMS is designed such that safetyrelated systems in proximity to the IMS are protected against the dynamic effects potentially created by the failure of IMS equipment. The IMS is a low pressure system, as the reactor coolant pressures are bounded by the reactor coolant static head pressures, thus precluding pipe whip. This satisfies PDC 4 for the IMS.

The IMS is designed to preclude the inadvertent draining of the RV during normal operation and during RV fill/drain operations. During normal operation, when the reactor vessel is fueled, the RV fill/drain transfer line is equipped with a passive RV isolation system features such as caps, flanges and/or a transfer line disconnect, designed to preclude inadvertent reactor coolant draining from the RV by siphoning. In the event of a leak in the RV fill/drain transfer line, while connected to the reactor vessel during fueled operation, the reactor coolant leak is detected by the plant control system, the PSP is tripped, and the RV cover gas pressure is limited to an upper bound thus precluding the ejection of reactor coolant through the transfer line diptube. During RV fill/drain operations, the reactor vessel is defueled, and the fill/drain line is connected, an isolation valve is used to interrupt the reactor coolant flow and a cover gas inlet is used to break the siphon in the transfer lines. These design features satisfy the requirements of PDC 33.

The RV coolant level management tank line short dip tube and overflow weir designs preclude inadvertent reactor coolant draining from the RV into the RV level management tank. Additionally, the overflow line weir is designed in a way that precludes the uncovering of fuel due to thermal expansion of the reactor coolant. In the event of a leak in the RV level management tank or transfer line, the reactor coolant leak is detected by the plant control system, and the pump for the reactor level management is tripped to minimize the overflow of reactor coolant from the RV through the overflow weir. This design configuration satisfies the requirements of PDC 33.

The IMS encompasses a The PHTS drain line is, equipped with a PHTS drain valve which interfaces with the PHTS fill/drain tank. In the event of a leak in the PHTS fill/drain tank or drain line, the reactor coolant leak is detected by the control system, the PSP is tripped. While a PHTS leak cannot be precluded, Tthe PHTS design contains an RV antisiphon feature (see Section 4.35.1), thus precluding inadvertent reactor coolant drain from the RV,. precluding the IMS from draining the RV. These design features satisfy the requirements of PDC 33 15.

The safetyrelated portions of the reactor coolant boundary are limited to the RV (see Section 4.3).

Failures of other SSCs containing reactor coolant (e.g., salt spill), do not result in unacceptable consequences as described in Chapter 13. A failure of the RV is a prevented event. Thus, Tthe makeup inventory function of IMSreactor coolant to the reactor vessel is not relied on to mitigate the consequences of a postulated event. As described in Section 4.3, the safety related portions of the reactor coolant boundary are limited to the reactor vessel and a failure of the reactor vessel is precluded by design. Therefore, the makeup functional requirements of PDC 33 have been addressed by design.

The system is expected to handle reactor coolant with fission as well as activation products; therefore, the system will be designed to minimize contamination and support eventual decommissioning, consistent with the requirements of 10 CFR 20.1406.

9.1.4.4 Testing and Inspection The components of the IMS, including valves, tanks, pumps and other components, are located such that they are accessible for periodic inspection and testing.

Kairos Power Hermes Reactor 917 Revision 0

Preliminary Safety Analysis Report Auxiliary Systems Figure 9.1.41: Inventory Management System Figure 9.1.41: Inventory Management System Kairos Power Hermes Reactor 919 Revision 0

Preliminary Safety Analysis Report Auxiliary Systems 9.3 PEBBLE HANDLING AND STORAGE SYSTEM For fuel pebbles in the PHSS the TRISO fuel particles provide a functional containment such that radionuclides are contained within the particle. The pebbles are designed to prevent damage to the TRISO fuel particles during normal operation, storage, shipping and handling thus the fuel particle is credited for confining radioactive material rather than the pebble matrix material, the handling equipment and the storage system. The fuel pebbles can experience thermal and mechanical loads while being handled, inspected, operated, and stored but such loads are within the design basis of the fuel pebble design.

9.3.1 Description The PHSS provides for handling and storing fuel and other pebbles. The system encompasses receipt and inspection of new fuel upon delivery, core loading, sensing, inspection and sorting during downstream circulation, reinsertion, core unloading, and removal and transfer to storage.

Major components and features of the PHSS include the pebble extraction machine (PEM), debris removal, offhead conveyance line, pebble processing, pebble inspection, pebble insertion, PHSS inert gas boundary, pebble storage, and new pebble introduction. A process flow diagram is provided in Figure 9.31.

The PHSS interfaces with the reactor vessel (Section 4.3), IGS (Section 9.1.2), spent fuel cooling system (SFCS) (Section 9.8.2), and the CCWS (Section 9.7.3) as shown in Figure 9.32.

9.3.1.1 Pebble Extraction Machine The PEM removes buoyant pebbles which accumulate in the reactor defueling chute at the top of the reactor core and routes them towards the offhead conveyance. The PEM is comprised of a single screw shaft located at the top of the reactor vessel head. As the pebbles traverse the screw, they are removed from the molten Flibe and moved into the inert gas space. The PEM also acts as a pathway for debris removal from the vessel to the debris removal portion of the system. Components in the PEM are cooled by the reactor thermal management system (see Section 9.1.5) to preclude overheating. The elevation of the PEM relative to the coolant limits coolant leaks from the reactor vessel in the event of breaks in the PEM.

9.3.1.2 Debris Removal Pebble or graphite debris removal is accomplished by extracting Flibe primary coolant up the PEM via a pressure differential, transferring debriscarrying Flibe to a filtering tank through a debris pipe, filtering debris from the coolant in an offhead tank, and returning filtered Flibe back to the vessel through the PEM.

9.3.1.3 OffHead Conveyance An offhead conveyance line routes pebbles from the PEM to a buffer storage prior to the processing system, located off the reactor head as shown in Figure 9.32. The offhead conveyance mechanism is a downward angled chute with a diameter that is larger than the pebbles. The offhead conveyance line includes design features for removing debris or jams that could impede pebble movement. This design minimizes the risk of pebbles and debris from jamming the line, such that a geometrically safe configuration is maintained at all times.

9.3.1.4 Pebble Processing PHSS pebble processing directs pebbles to the correct insertion channel or to a storage canister for spent fuel, based on results from the inspection system via an automated mechanism. A rotating wheel Kairos Power Hermes Reactor 925 Revision 0

Preliminary Safety Analysis Report Auxiliary Systems in the processing system moves pebbles from the offhead conveyance to the inspection area. After inspection, the pebbles are directed for reinsertion into the core, or to pebble storage for removal from the circulating pebble inventory, based on inspection results.

9.3.1.5 Pebble Inspection An automated inspection system provides information to the processing portion of the PHSS for determining pebble health. This includes inspection of the physical condition of the pebble for unacceptable wear or damage, identifying moderator and fuel pebbles, as well as an evaluation of the burnup of the fuel relative to a maximum burnup limit using the burn up measurement sensor (BUMS).

The burnup measurement is done by means of a gamma spectrometer. Further details pertaining to inspections for wear and damage will be provided with the application for an Operating License.

9.3.1.6 Pebble Insertion Pebbles are received from the processing system and placed in a buffer storage until required for reinsertion. The pebble buffer storage is sized and orientated to prevent a critical configuration.

Individual pebbles are fed into the step feeder insertion machine from this pebble buffer storage as shown in Figure 9.32. The pebbles are inserted into the top of the reactor vessel head, then pushed through the insertion line and enter the reactor core via the invessel fueling chute at the bottom of the core (see Section 4.3). There is a single active insertion line into the vessel and is designed with overflow protection cutouts to limit coolant loss from the reactor vessel in the event the insertion line breaks.

9.3.1.7 PHSS Inert Gas Boundary The components of the PHSS are designed to maintain an inert gas boundary outside of the reactor vessel for pebble handling. The function of the inert gas environment is to prevent absorption of moisture and oxygen into pebbles for pebble handling during normal operations. The inert gas boundary within the PHSS (see Figure 9.32) is created by a mechanical structure that encloses the aforementioned components with penetrations for motor shafts, storage outlets, inspection viewport, data channels, electrical power, and pebbles from the offhead conveyance mechanism and for insertion. Portions of the inert gas boundary that are adjacent to personnel access areas have the appropriate radiation shielding.

9.3.1.8 Pebble Storage Pebble storage is provided for pebble debris, damaged pebbles, spent fuel, and end of life moderator pebbles. The storage portion of the system is composed of a stainless steel storage canister and transporter device. Individual storage canisters are sized to hold approximately 1,9002,100 pebbles.

The dimensions of the canister and quantity of pebbles are sized to maintain a noncritical configuration.

A transporter device is used to transfer canisters to either the spent fuel storage area during normal operation or the full core offload area in the event of a periodic maintenance full core offload or an emergent full core offload.

9.3.1.8.1 Spent Fuel Storage Spent fuel is discharged from service in the core under normal operating conditions, placed in sealed storage canisters, and moved to the spent fuel storage area as shown in Figure 9.32. The initial storage area is a cooling pool designed to hold spent fuel canisters while the decay heat of the pebbles drops.

The pool is designed to limit radiation exposure to personnel. After cooling in pool storage, the canisters are moved to a concrete storage bay with radiation shielding and forced air cooling. The pool is actively cooled by the CCWS using an inpool heat exchanger. Water is recirculated in the pool by the SFCS and makeup water is provided by the treated water system (see Section 9.7.2). The pool and concrete Kairos Power Hermes Reactor 926 Revision 0

Preliminary Safety Analysis Report Auxiliary Systems process is done via two sequential valves to prevent introduction of contaminants to the PHSS inert gas boundary or new pebbles. The interstitial space between the valves is purged prior to opening of either valve to limit the ingress of oxygen.

9.3.2 Design Bases Consistent with PDC 2, the PHSS is designed to withstand the effects of natural phenomena without exceeding the offsite dose consequences of the MHA, compromising decay heat removal, or criticality as a result of a system failure or breach.

Consistent with PDC 3, the PHSS is designed and located within the facility to minimize the probability and dose consequences of fires and explosions.

Consistent with PDC 4, the PHSS is designed to accommodate environmental conditions associated with normal operation, maintenance, testing and postulated events.

Consistent with PDC 33, the PHSS is designed to limit the loss of reactor coolant from the reactor vessel due to potential breaks in the system.

Consistent with PDC 61, the PHSS is designed to permit periodic inspection and testing and is suitably shielded for radiation protection. The PHSS design includes appropriate confinement and adequately accounts for decay heat and a reduction in fuel storage cooling under postulated events.

Consistent with PDC 62, the PHSS is designed to prevent criticality.

Consistent with PDC 63, the PHSS is designed to detect conditions that may result in excessive radiation levels and initiates appropriate safety actions.

Consistent with 10 CFR 70.24(a)(1), the PHSS design includes a monitoring system capable of detecting criticality.

Consistent with 10 CFR 20, the PHSS is designed to be shielded to support worker occupational dose limits and adhere to a radiation protection program.

Consistent with 10 CFR 20.1406, the PHSS is designed, to the extent practicable, to minimize contamination of the facility and the environment, and facilitate eventual decommissioning.

9.3.3 System Evaluation The concrete structures associated with the storage bay, pool, and support restraints in the pool are designed as SDC 3 structures to ensure the geometry of the storage area is maintained to preclude an inadvertent criticality during a design basis earthquake. The design of the support restraints and storage bay also ensures adequate spacing is maintained for air cooling between each canister. During a postulated earthquake, the fuel particles prevent radionuclide release. The particles are supported in their safety function during a postulated earthquake by the pool and by the canister transporter, both of which provide passive cooling and spacing to restrict pebble movement thereby preventing recriticality.

Other portions of the PHSS that do not perform a safety function are designed to be either seismically mounted or physically separated to preclude adverse interactions with other safetyrelated SSCs during a design basis earthquake. These design features satisfy the requirements of PDC 2.

The PHSS is designed to minimize the probability of a fire or explosion by limiting the accumulation of potentially combustible material such as graphite dust and debris within the system. Grinding of pebbles which contribute to graphite dust generation is precluded by system design. The small amount of graphite dust that might be generated is directed through pebble motion to the storage canisters for removal from the system. The PHSS is not located near nor interfaces with pneumatic systems with the Kairos Power Hermes Reactor 928 Revision 0

Preliminary Safety Analysis Report Auxiliary Systems potential for air inleakage. The system is filled with an inert gas operated at a slightly positive pressure to further prevent air ingress in the event of a PHSS breach. Locations where pebbles are not submerged in coolant, such as the PEM, will either not exceed temperatures that would induce oxidation of the graphite or are expected to cool quickly such that oxidation, if any, would be minimal and not affect the acceptability of the pebble for reuse. These design features satisfy the requirements of PDC 3 for the PHSS. Fire protection systems are further discussed in Section 9.4.

The pebble handling portion of the PHSS is protected from the effects of discharging fluids. There are no pressurized piping systems in or around the PHSS thus precluding the design from high energy line considerations. A hypothetical water line break in the area of the storage system does not pose a criticality risk as the analyses supporting the storage system assume complete submergence and internal flooding of the storage canisters in water. The PHSS is designed in consideration of the high radiation environment where equipment will be functioning. The PHSS design also considers and accounts for the temperature within the system to preclude oxidation of graphite pebbles. The stainless steel PHSS storage canisters are designed to accommodate pressure due to the accumulation of radionuclides and thermal loads associated with the amount of spent fuel loaded in each canister during normal and postulated event conditions. The canisters are also designed to accommodate the tensile stress exerted during transfer and are compatible with handling equipment. The interior of the stainless steel canisters is also designed to account for radiolysis products from spent nuclear fuel and ensures the integrity of the canister, seal, and weld thus precluding the potential release of radionuclides from the canister.

These design features demonstrate that the PHSS satisfies the environmental and dynamic effects in PDC 4.

The PHSS interfaces with the reactor vessel at the PEM and the pebble insertion line. The elevation of the PEM relative to the coolant free surface is such that coolant inventory loss from the reactor vessel is limited in the event the PEM breaks. The pebble insertion line is designed to limit inventory loss to an elevation no lower than the primary salt pump elevation, in the event of a break in the insertion line.

The pebble insertion line uses overflow protection cutouts to direct any coolant in the insertion line back down into the reactor vessel. These design features of the PHSS satisfy the requirements in PDC 33.

PDC 61 requires that the safetyrelated portions of the PHSS that contain radioactivity be designed to ensure (1) capability to permit appropriate periodic inspection and testing of components, (2) suitable shielding for radiation protection, (3) appropriate containment, confinement, and filtering, (4) residual heat removal capability, and (5) significant reduction in fuel storage cooling under postulated event conditions is precluded. The design features which address PDC 61 for the PHSS are discussed below:

The TRISO fuel particle provides a functional containment as described in Section 6.2. Radioactive material and fission products are contained within the particle unless the TRISO layers are compromised or defective (see Section 4.2.1). The fuel pebble, as described in Section 4.2.1, is designed to preclude physical damage or changes in geometry to the TRISO particle during anticipated loads from normal operation, storage, shipping and handling. Therefore, the TRISO particle is credited for the confinement of radioactive materials rather than the PHSS. The pebble can experience thermal and mechanical loads while being handled, inspected, operated, and stored; however, such loads do not introduce incremental failures of TRISO particles. Furthermore, the PHSS design precludes pebble damage from overheating and oxidation. Heat removal mechanisms within the system, such as thermal radiation and convection via natural circulation, are sufficient to remove the decay heat produced by individual pebbles during their transit through the PHSS. Also, oxidation associated with air or moisture ingress into the PHSS is negligible for pebbles at temperatures experienced in the system. The system also minimizes pebble wear. The limiting PHSS Kairos Power Hermes Reactor 929 Revision 0

Preliminary Safety Analysis Report Accident Analysis volatile products and oxidizes portions of the structural graphite above the surface of the Flibe and the carbon matrix for pebbles in transit above the surface of the Flibe. Radionuclides from the coolant circulating activity in the broken pipe are released into the facility air when aerosols are generated from the coolant that exits the pipe. All the floor surfaces where Flibe may be spilled will have design features such as steel liners to prevent Flibeconcrete interaction, as described in Section 3.5. The spilled Flibe spreads on top of the liner and forms a Flibe pool. Radionuclides in the spilled Flibe is released through evaporation until the top surface of the Flibe pool is solidified.

The limiting salt spill postulated event bounds other salt spill events, including:

Spurious draining and smaller leaks from the primary heat transport system Leaks from other Flibe containing systems and components (e.g., IMS fill/drain tank, IMS piping, chemistry control system piping)

Leaks up to the hypothetical doubleended guillotine primary salt piping break size Mechanical impact or collision events involving Flibe Containing SSCs (except the vessel)

Leaks from the primary heat rejection system that contains a nonFlibe coolant, which may contain nonzero amount of Flibe from heat exchanger leaks These following sections describe key assumptions associated with the limiting salt spill event. The quantitative values associated with these assumptions, as well as the methods used to evaluate the surrogate figures of merit that ensure the event consequences are bounded by the MHA are provided in Reference 2.

13.1.3.1 Initial Conditions Assumptions Normal operating parameters are provided in Section 4.1. Conservative initial values are assumed for each operating parameter to ensure a bounding result for the figures of merit that demonstrate the event is bounded by the MHA.

A hypothetical doubleended guillotine break in the PHTS hot leg piping is assumed as the event initiator. The initial Flibe conditions are discussed in Section 5.1.

13.1.3.2 Structures Systems and Components Mitigation Assumptions This section describes the SSCs performing a function to mitigate the consequences of the event.

The RPS is credited with detecting the break on low reactor coolant level and initiating a reactor trip, PSP trip, ISP trip, and the PHSS trip. The DHRS is operating when the reactor is above a threshold power, as discussed in Section 6.3, and remains in an always on mode. The RPS initiates a reactor trip to shut down the reactor and limits the addition of heat to the system. The RPS trips the PSP limit the amount of spilled Flibe. The ISP is tripped concurrently to ensure a positive pressure differential between the primary and intermediate loops. The PHSS trip stops pebble extraction and insertion following the reactor trip to preclude any damage to pebbles from faults during the event. The DHRS remains active to ensure that an adequate amount of decay heat is removed from the system. The design bases of the RPS are discussed in Section 7.3. The RPS detection and actuation capabilities are automatic and do not rely on manual operator action to perform these functions.

The RCSS is credited with shutting down the reactor upon receiving the reactor trip signal. The shutdown and control elements are assumed to have sufficient worth to shut down the reactor and maintain long term shutdown. The design bases of the RCSS shutdown function are provided in Section 4.2.2.

The antisiphon design features of the reactor vessel PHTS (see Section 5.1) discussed in Section 4.3 are credited with limiting the amount of Flibe available to spill out of the break. The design features below Kairos Power Hermes Reactor 137 Revision 0

Preliminary Safety Analysis Report Accident Analysis ensure there is no recriticality after the RCSS has initiated shutdown, as described in Section 4.5.

Additionally, the graphite reflector blocks are designed to maintain structural integrity and ensure misalignments do not prevent the insertion path of the shutdown elements, as discussed in Section 4.3.

13.1.10.2 Degraded Heat Removal or Uncooled Events In postulated events where the normal heat rejection is not available, natural circulation in the reactor vessel and the heat removal function of the DHRS are relied upon to remove heat from the reactor core.

Degraded heat removal or uncooled events are excluded from the design basis. The initiation of natural circulation is completely passive, and the design features, including the structural integrity of the reactor vessel internals, that ensure a continued natural circulation flow path are discussed in Section 4.6. The DHRS is aligned and operating when the reactor power is above a threshold power and remains in this state as described in Section 6.3, precluding the need for an actuation to occur for the DHRS to remove heat during a postulated event. The DHRS design includes sufficient redundancy to perform its safety function assuming the loss of a single train, as discussed in Section 6.3.

13.1.10.3 Flibe Spill Beyond Maximum Volume Assumed in Postulated Salt Spills In the salt spill postulated event category, an upper bound volume of Flibe is assumed to spill out of the PHTS onto the floor. A volume of Flibe spilling out of the system beyond the amount assumed in the bounding salt spill event is excluded from the design basis. There are several design features ensuring the amount of Flibe available to spill is limited to an upper bound value. The PHTS reactor vessel is designed with antisiphon features discussed in Section 4.3 5.1. These features are designed to passively break the siphon in the event of a break. The PSP also trips to allow the primary system to depressurize.

The reliability of the RPS, which trips the PSP and ISP in the event of a salt spill, is discussed in Section 7.3. The reactor vessel shell also maintains integrity in postulated events to ensure the fuel in the core remains covered with Flibe. The reactor vessel shell design features that prevent leakage are discussed in Section 4.3.

13.1.10.4 InService TRISO Failure Rates and Burnups Above Assumptions in Postulated Events The inservice fuel failure rates and the burnup of pebbles assumed in the postulated events are based on the fuel qualification specifications in Section 4.2.1. Inservice TRISO failure rates above the rate assumed in postulated events are excluded from the design basis. The insertion of pebbles with a burnup higher than the fuel qualification envelope is excluded from the design basis. As described in Section 7.3, the RPS includes a function to stop the pebble insertion and extraction functions to ensure pebbles are not damaged in faults occurring after an event initiation. The fuel qualification program includes testing, inspection, and surveillance to ensure the fuel operating envelope is within the fuel qualification envelope. Inspection and surveillance of the fuel in service is performed in the PHSS as discussed in Section 9.3.

13.1.10.5 Significant Intermediate Coolant Ingress Into PHTS The postulated events assume a positive pressure differential between the primary and intermediate coolant systems. Events where significant quantities of intermediate coolant enter the PHTS are excluded from the design basis. Chapter 5 discusses the design features of the PHTS and PHRS that maintain a positive pressure differential.

13.1.10.6 DHRS Reactor Cavity Flooding The DHRS is a waterbased system that removes heat from the reactor vessel shell. Events where the water from the DHRS leaks into the reactor cavity in quantities significant enough to wet the reactor Kairos Power Hermes Reactor 1315 Revision 0