Abstract: A nuclear reactor trip apparatus includes a remote circuit breaker trip device operatively connected to a reactor trip breaker to release a control rod into a nuclear reactor core, an active power source, a passive power source, and a local circuit breaker trip device operatively connected to the reactor trip breaker including a sensor to trigger the local circuit breaker trip device upon sensing a predefined condition. The active power source is electrically coupled to energize the remote circuit breaker trip device under normal operating conditions. The passive power source is electrically coupled to energize the remote circuit breaker trip device based on a loss of the active power source.

Abstract: A rod position indication system includes a drive rod operably coupled to a control rod that is configured to be both withdrawn from and inserted into a reactor core. A number of sensing devices are arranged along a path of the drive rod, and an end of the drive rod passes by or through one or more of the sensing devices in response to movement of the control rod relative to the reactor core. The sensing devices are arranged into a plurality of groups, and each group includes two or more of the sensing devices electrically coupled together. The rod position indication system further includes a control rod monitoring device electrically coupled to each group of sensing devices by a routing wire.

Abstract: A system for managing notifications for a modular power plant that includes one or more power-generation module (PGM) assemblies and one or more common-systems (CS) includes a transceiver that receives data over a network; a display device that displays a graphical display; a memory that stores at least instructions; and a processor device that executes instructions that perform actions. The actions include receiving power plant data over the network that includes signals generated by one or more sensors included in at least one of the PGM assemblies or one of the CS; detecting an event based on the power plant data and one or more notification thresholds; in response to the event, determining a notification type based on the data and the one or more notification thresholds; providing a notification to a user via the graphical display based on the notification type and the event; and enabling the user to acknowledge the notification.

Abstract: A power module assembly includes a reactor core immersed in a coolant and a reactor vessel housing the coolant and the reactor core. An internal dry containment vessel submerged in liquid substantially surrounds the reactor vessel in a gaseous environment. During an over-pressurization event the reactor vessel is configured to release the coolant into the containment vessel and remove a decay heat of the reactor core through condensation of the coolant on an inner surface of the containment vessel.

Abstract: A steam generator for a nuclear reactor comprises plenums proximate with a first plane, wherein the first plane intersects a bottom portion of a riser column of a reactor vessel. The steam generator may further comprise plenums proximate with a second plane, approximately parallel with the first plane, wherein the second plane intersects a top portion of the riser column of the reactor vessel. The steam generator may further include a plurality of steam generator tubes that convey coolant from a plenum located proximate with the first plane to one of the plenums proximate with the second plane.

Abstract: A nuclear reactor seismic isolation assembly includes an enclosure that defines a volume; a plastically-deformable member mounted, at least in part, within the volume; and a stretching member moveable within the enclosure to plastically-deform the plastically-deformable member in response to a dynamic force exerted on the enclosure.

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 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 containment seal seals a cable or tube to an opening accessing a containment vessel. The containment seal includes a lower body that attaches over the opening into the containment vessel. The cable or tube is inserted through a hole that extends axially through an upper body of the containment seal. Compression fittings are attached to the top and bottom ends of the upper body sealing the cable inside of the upper body. The cable sealed inside of the upper body in inserted through the lower body and into the opening accessing the containment vessel and a lower portion of the upper body is seated into the opening formed in the lower body. A retaining device compresses the upper body down against the lower body forming a seal between the upper body and lower body.

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 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 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 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.