Abstract: A nuclear reactor protection system includes a plurality of functionally independent modules, each of the modules configured to receive a plurality of inputs from a nuclear reactor safety system, and logically determine a safety action based at least in part on the plurality of inputs, each of the functionally independent modules comprising a digital module or a combination digital and analog module, an analog module electrically coupled to one or more of the functionally independent modules, and one or more nuclear reactor safety actuators communicably coupled to the plurality of functionally independent modules to receive the safety action determination based at least in part on the plurality of inputs.

Abstract: A power module assembly may include a reactor vessel containing a primary coolant and one or more inlets configured to draw a secondary coolant from the containment cooling pool in response to a loss of power and/or a loss of coolant. One or more outlets may be submerged in the containment cooling pool and may be configured to vent the secondary coolant into the containment cooling pool. A heat exchanger may be configured to remove heat from the primary coolant, wherein the heat may be removed by circulating the secondary coolant from the containment cooling pool through the heat exchanger via natural circulation.

Abstract: A nuclear reactor protection system includes a plurality of functionally independent modules, each of the modules configured to receive a plurality of inputs from a nuclear reactor safety system, and logically determine a safety action based at least in part on the plurality of inputs; and one or more nuclear reactor safety actuators communicably coupled to the plurality of functionally independent modules to receive the safety action determination based at least in part on the plurality of inputs.

Abstract: Nuclear reactor systems and associated devices and methods are described herein. A representative nuclear reactor system includes a heat pipe network having an evaporator region, an adiabatic region, and a condenser region. The heat pipe network can define a plurality of flow paths having an increasing cross-sectional flow area in a direction from the evaporator region toward the condenser region. The system can further include nuclear fuel thermally coupled to at least a portion of the evaporator region. The heat pipe network is positioned to transfer heat received from the fuel at the evaporator region, to the condenser region. The system can further include one or more heat exchangers thermally coupled to the evaporator region for transporting the heat out of the system for use in one or more processes, such as generating electricity.

Abstract: Nuclear reactor systems and associated devices and methods are described herein. A representative nuclear reactor system includes a reactor vessel having a barrier separating a core region from a shield region. A plurality of fuel rods containing a liquid nuclear fuel are positioned in the core region. A liquid moderator material is also positioned in the core region at least partially around the fuel rods. A plurality of heat exchangers can be positioned in the shield region, and a plurality of heat pipes can extend through the barrier. The moderator material is positioned to transfer heat received from the liquid nuclear fuel to the heat pipes, and the heat pipes are positioned to transfer heat received from the moderator material to the heat exchangers. The heat exchangers can transport the heat out of the system for use in one or more processes, such as generating electricity.

Abstract: A system for servicing a nuclear reactor module comprises a crane operable to attach to the nuclear reactor module, wherein the crane includes provisions for routing signals from one or more sensors of the nuclear reactor module to one or more sensor receivers.

Abstract: A damping area or “dash pot” on the upper ends of control rods absorb energy from dropped control rod assemblies without narrowing the diameter of guide tubes. As a result, coolant can freely flow through the guide tubes reducing boiling water issues. The dampening area reduces a separation distance between an outside surface of the control rod and an inside surface of the guide tubes decelerating the control rods when entering a top end of the guide tubes. In another example, the dampening area may be located on a drive shaft. The dampening area may have a larger diameter than an opening in a drive shaft support member that decelerates the drive shaft when dropped by a drive mechanism.

Abstract: A system includes a containment vessel configured to prohibit a release of a coolant, and a reactor vessel mounted inside the containment vessel. An outer surface of the reactor vessel is exposed to below atmospheric pressure, wherein substantially all gases are evacuated from within the containment vessel.

Abstract: A nuclear power system includes a reactor vessel that includes a reactor core mounted within a volume of the reactor vessel. The reactor core includes one or more nuclear fuel assemblies configured to generate a nuclear fission reaction. The nuclear power system further includes a containment vessel sized to enclose the reactor vessel such that an open volume is defined between the containment vessel and the reactor vessel. A boron injection system is positioned in the open volume of the containment vessel and includes an amount of boron sufficient to stop the nuclear fission reaction or maintain the nuclear fission reaction at a sub-critical state. The boron injection system is positioned to deliver the amount of boron into the open volume.

Abstract: An inadvertent actuation block valve includes inlet and outlet orifices being in selective fluid communication via a chamber. A disc is disposed within the chamber and a bellows is configured to contract at a predetermined pressure differential between reactor fluid entering a reference pressure orifice and control fluid entering the inlet orifice. When the bellows contracts, the disc engages the outlet orifice and isolates fluid communication between the inlet and outlet orifices. The inadvertent actuation block valve prevents inadvertent opening of an emergency core cooling valve when a reactor is at operating pressure that is above the predetermined set pressure range. The inadvertent actuation block valve permits the emergency cooling valves to open and to remain open when reactor pressure is below the predetermined set pressure range. The inadvertent actuation block valve does not impede long term emergency cooling that occurs when the reactor is at low pressure.

Abstract: Embodiments are directed to providing a user interface (UI) that streamlines and simplifies the process of monitoring critical power-generation module (PGM) parameters after a PGM assembly is shutdown. The UI displays, in real-time, indicators corresponding to one or more post-shutdown PGM parameters. The UI provides indications of whether the post-shutdown PGM parameters meet post-shutdown criteria of the PGM assembly. When a post-shutdown PGM parameter does not meet the post-shutdown criteria, a user alert is provided to the user. A protocol may additionally be provided to the user. In some embodiments, the protocol may enable the user to return the PGM assembly to a condition that satisfies the post-shutdown criteria. The protocol may be a safety protocol and/or an asset protection protocol.

Abstract: A nuclear power system includes a reactor vessel that includes a reactor core mounted therein. The reactor core includes nuclear fuel assemblies configured to generate a nuclear fission reaction. The reaction vessel does not include any control rod assemblies therein. The nuclear power system further includes a riser positioned above the reactor core, a primary coolant flow path, a primary coolant that circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the received heat to generate electric power in a power generation, and a control system communicably coupled to the power generation system and configured to control a power output of the nuclear fission reaction independent of any control rod assemblies.

Abstract: A representative cooling system for a nuclear reactor control rod drive mechanism (CRDM) includes an evaporation section located within or next to the CRDM and a condensation section fluidly coupled to the evaporation section. The cooling system includes a set of heat fins coupled to drive coils in the CRDM and heat pipes that extend through the drive coils and heat fins. A fluid evaporates while in the evaporation section of the heat pipes from heat generated by the CRDM and moves out of the evaporation section into the condensation section in the heat fins. The fluid cools and condensates while in the condensation section, recirculating back into the evaporation section. This passive natural circulation cooling system reduces or eliminates the number of water hoses, piping, and other water pumping equipment typically used for cooling a CRDM thereby increasing nuclear reactor reliability and simplifying nuclear reactor operation and maintenance.

Abstract: A nuclear power system includes a reactor vessel that includes a reactor core that includes nuclear fuel assemblies configured to generate a nuclear fission reaction. A representative nuclear power system further includes a riser positioned above the reactor core and a primary coolant flow path that extends from a bottom portion of the reactor vessel, through the reactor core, and through an annulus between the riser and the reactor vessel. A primary coolant circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the heat to a power generation system configured to generate electric power. The nuclear power system further includes a control rod assembly system positioned in the reactor vessel and configured to position control rods in only two discrete positions.

Abstract: In an example, a raised face flange assembly, comprises an upper flange to couple to a lower flange using one or more bolts: wherein the upper flange or the lower flange comprises: a bolting face defining one or more openings for the one or more bolts, respectively; a pair of raised faces including a first raised face and a second raised face to make contact with a mating surface of the other of the upper flange or the lower flange; wherein a distance between an area of the second raised face and a plane corresponding to the bolting face is greater than a distance between an area of the first raised face and the plane to distribute contact force with a mating surface over the area of the second raised face to maintain a seal.

Abstract: An in-core instrumentation system for a reactor module includes a plurality of in-core instruments connected to a containment vessel and a reactor pressure vessel at least partially located within the containment vessel. A reactor core is housed within a lower head that is removably attached to the reactor pressure vessel, and lower ends of the in-core instruments are located within the reactor core. The in-core instruments are configured such that the lower ends are concurrently removed from the reactor core as a result of removing the lower head from the reactor pressure vessel.

Abstract: In some embodiments, a nuclear reactor vessel comprises a containment vessel for a reactor pressure vessel (RPV); a control rod drive mechanism (CRDM) located in the containment vessel, the CRDM including drive motors configured to move control rods into and out of a nuclear reactor core located in the RPV; and a partition extending across a portion of the containment vessel configured to retain the drive motors in a separate fluid-tight barrier region within the containment vessel. Other embodiments may be disclosed and/or claimed.

Abstract: A nuclear reactor protection system includes a plurality of functionally independent modules, each of the modules configured to receive a plurality of inputs from a nuclear reactor safety system, and logically determine a safety action based at least in part on the plurality of inputs, each of the functionally independent modules comprising a digital module or a combination digital and analog module, an analog module electrically coupled to one or more of the functionally independent modules, and one or more nuclear reactor safety actuators communicably coupled to the plurality of functionally independent modules to receive the safety action determination based at least in part on the plurality of inputs.

Abstract: A thermal control system for a reactor pressure vessel comprises a plate having a substantially circular shape that is attached to a wall of the reactor pressure vessel. The plate divides the reactor pressure vessel into an upper reactor pressure vessel region and a lower reactor pressure vessel region. Additionally, the plate is configured to provide a thermal barrier between a pressurized volume located within the upper reactor pressure vessel region and primary coolant located within the lower reactor pressure vessel region. One or more plenums provide a passageway for a plurality of heat transfer tubes to pass fluid through the wall of the reactor pressure vessel. The plurality of heat transfer tubes are connected to the plate.

Abstract: A damping area or “dash pot” on the upper ends of control rods absorb energy from dropped control rod assemblies without narrowing the diameter of guide tubes. As a result, coolant can freely flow through the guide tubes reducing boiling water issues. The dampening area reduces a separation distance between an outside surface of the control rod and an inside surface of the guide tubes decelerating the control rods when entering a top end of the guide tubes. In another example, the dampening area may be located on a drive shaft. The dampening area may have a larger diameter than an opening in a drive shaft support member that decelerates the drive shaft when dropped by a drive mechanism.