Abstract: A bolt installation and removal (BIR) system is used for assembling and disassembling a nuclear vessel. The BIR system includes a platform with a stand for supporting the nuclear vessel. A track extends around an outside perimeter of the platform and multiple tool carts include wheels that roll on the track. Tool towers are located on the carts and include tool assemblies configured to install and remove bolts on the nuclear reactor vessel. Magazine towers also extend up from the tool carts next to the tool towers and include magazines that hold bolts for exchanging with the tool assemblies. Drive mechanisms move tool heads in the tool assemblies around a first vertical axis, vertically up and down, and laterally to more simply and reliably install and remove the bolts in a radioactive underwater environment.

Abstract: A packaging material comprised of a gas-impermeable base layer, an indicator layer, and a gas-impermeable top layer is disclosed. The top layer is selected to rupture upon experiencing an impingement or impact that exceeds a predetermined threshold. The indicator layer is configured to change visual appearance upon exposure to ambient air. The top layer further may be configured into a plurality of cells, with each cell filled with a gas that is inert with respect to the indicator layer. The cells may be of a size and shape configured to indicate the location and shape of an impingement or impact, to alert a recipient of an item protected by the packaging material of an area that may require further inspection.

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: The drive assembly includes annular drive magnets extending around a top end of a drive shaft and annular drive coils extending around the drive magnets, separated by a pressure boundary. A latch assembly is coupled to the drive magnets and engages with the drive shaft in response to actuation of the drive assembly. The drive coils also rotate the drive magnets and the engaged latch assembly to axially displace the drive shaft. Deactivating the drive coils disengages the latch assembly from the drive shaft, dropping a connected control rod assembly via gravity into a nuclear fuel assembly. A disconnect assembly includes a disconnect magnet coupled to the top end of a disconnect rod that extends through the drive shaft. Annular disconnect coils extend around the disconnect magnet, separated by a pressure boundary, to hold the disconnect magnet and the disconnect rod in a raised position to remotely disconnect from, or reconnect the drive shaft to, the control rod assembly.

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 through the wall of the reactor pressure vessel. The plurality of heat transfer tubes are connected to the plate.

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 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 system for measuring flow rate within a volume includes one or more transmission devices that transmit one or more signals through fluid contained within the volume. The volume may be bounded, at least in part, by an outer structure and by an object at least partially contained within the outer structure. A transmission device located at a first location of the outer structure transmits a first signal to a second location of the outer structure. A second signal is transmitted through the fluid from the second location to a third location of the outer structure. The flow rate of the fluid within the volume may be determined based, at least in part, on the time of flight of both the first signal and the second signal.

Abstract: A tube support assembly for a steam generator system is disclosed herein, including a sheet or support bar configured to support a plurality of heat transfer tubes of the steam generator system and a set of projections extending from a surface of the sheet or support bar. A distance that the set of projections extend from the surface of the sheet or support bar may be greater than, or equal to, an external diameter of a heat transfer tube. The adjacent tubes of the plurality of tubes may be separated from each other by one or more projections.

Abstract: A heat transfer system includes a plenum configured to provide a secondary side fluid such as feedwater to a plurality of heat transfer tubes from a primary side fluid, and a tube sheet coupled to the plurality of heat transfer tubes. An orifice plate is mounted within the plenum and located adjacent to the tube sheet, and one or more orifice devices are supported by the orifice plate and are configured for insertion into or sealing against the plurality of heat transfer tubes. The one or more orifice devices may include center flow orifices, and/or rectangular or helical shaped transition stepped annular flow orifices, and an insertion of a number of the transition steps into the plurality of heat transfer tubes may determine a corresponding pressure drop of the secondary side fluid in the heat transfer system.

Abstract: A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

Abstract: A system for draining a containment vessel may include a drain inlet located in a lower portion of the containment vessel. The containment vessel may be at least partially filled with a liquid, and the drain inlet may be located below a surface of the liquid. The system may further comprise an inlet located in an upper portion of the containment vessel. The inlet may be configured to insert pressurized gas into the containment vessel to form a pressurized region above the surface of the liquid, and the pressurized region may operate to apply a surface pressure that forces the liquid into the drain inlet. Additionally, a fluid separation device may be operatively connected to the drain inlet. The fluid separation device may be configured to separate the liquid from the pressurized gas that enters the drain inlet after the surface of the liquid falls below the drain inlet.

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: Fault-tolerant power distribution systems for modular power plants are discussed. The systems enable the transmission of a portion of the power generated by a modular power plant to remote consumers. The systems also enable the local distribution of another portion of the generated power within the power plant. The various fault-tolerant systems enable the plant to continuously, and without degradation or disruption, transmit power to remote consumers and distribute power within in the power plant in the event of one or more faults within the distribution system. The various embodiments include redundant power-transmission paths, power-distribution module feeds, switchgear, and other hardware components.

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 through the wall of the reactor pressure vessel. The plurality of heat transfer tubes are connected to the plate.

Abstract: A method of loading fuel in multiple reactor cores associated with a plurality of fuel cycles. The method includes, in a first fuel cycle, loading a first reactor core with a first fuel assembly selected from a first batch of fuel, loading the first reactor core with a first partially spent fuel assembly from a second batch of fuel, loading a second reactor core with a second fuel assembly from the first batch of fuel, and loading the second reactor core with a second partially spent fuel assembly from the second batch of fuel. In a second fuel cycle, which is performed after a completion of the first fuel cycle, the method includes loading the second reactor core with a fresh fuel assembly, and loading the second reactor core with the first fuel assembly from the first batch of fuel.

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 nuclear reactor control rod drive assembly includes a control rod drive mechanism coupled to a drive shaft and operable to bi-directionally urge the drive shaft through a portion of an inner volume of a reactor vessel at a first force; a control rod manifold coupled to the drive shaft; a plurality of control rods coupled to the control rod manifold, the plurality of control rods adjustable among a plurality of positions within the inner volume of the reactor vessel based on operation of the control rod drive mechanism; and at least one variable strength joint positioned between the control rod drive mechanism and the plurality of control rods.

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: Systems and methods for servicing a nuclear reactor module are provided. Such systems include a crane that couples to the reactor module. The crane includes a conduit for routing signals from sensors of the reactor module to sensor receivers. The crane includes a drive mechanism to move the reactor module from a reactor bay to a servicing area. The drive mechanism moves the reactor module in a first direction and a second direction that is orthogonal to the first direction. The crane includes a support bracket for mounting sensor signal receivers that receive the signals from the sensors within the nuclear reactor module. The system includes a display to display representations of the signals from the sensors of the nuclear reactor module in a servicing area. Another display displays representations of the signals from the sensors of the nuclear reactor module in an operator area.