A control rod assembly having a connecting rod, a spider, a plurality of control rods, a coupling sleeve secured to the connecting rod and including a body portion defining at least one cam pin, and a barrel cam defining a cam track, the barrel cam being rotatably secured to the spider. The cam track includes a first camming surface and a second camming surface that are configured so that the barrel cam rotates in a first direction as the at least one cam pin slides along the first and second camming surfaces. The barrel cam rod is rotatable between a first position in which the spider is connected to the connecting rod and a second position in which the spider is disconnected from the connecting rod.

A nuclear reactor system includes a nuclear reactor core disposed in a pressure vessel. Nuclear reactor system further includes control drums disposed longitudinally within the pressure vessel and laterally surrounding fuel elements and at least one moderator element of the nuclear reactor core to control reactivity. Each of the control drums includes a reflector material and an absorber material. Nuclear reactor system further includes an automatic shutdown controller and an electrical drive mechanism coupled to rotatably control the control drum. Automatic shutdown controller includes a counterweight to impart a bias and an actuator. To automatically shut down the nuclear reactor core during a loss or interruption of electrical power from a power source to the electrical drive mechanism, the actuator is coupled to the counterweight and responsive to the bias to align the absorber material of one or more control drums to face inwards towards the nuclear reactor core.

An enhanced architecture for a nuclear reactor core includes several technologies: (1) nuclear fuel tiles (S-Block); and (2) a high-temperature thermal insulator and tube liners with a low-temperature solid-phase moderator (U-Mod) to improve safety, reliability, heat transfer, efficiency, and compactness. In S-Block, nuclear fuel tiles include a fuel shape designed with an interlocking geometry pattern to optimize heat transfer between nuclear fuel tiles and into a fuel coolant and bring the fuel coolant in direct contact with the nuclear fuel tiles. Nuclear fuel tiles can be shaped with discontinuous nuclear fuel lateral facets and have fuel coolant passages formed therein to provide direct contact between the fuel coolant and the nuclear fuel tiles. In U-Mod, tube liners with low hydrogen diffusivity retain hydrogen in the low-temperature solid-phase moderator even at elevated temperatures and the high-temperature thermal insulator insulates the solid-phase moderator from the nuclear fuel tiles.

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

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

A method for replacing a damaged sleeve lining a tube passing through a nuclear reactor pressure vessel. The damaged sleeve has an end including a radially enlarged end portion for resting on a support section of the tube for retaining the damaged sleeve in the tube. The method includes removing the damaged sleeve from a tube; providing a sleeve assembly including a first sleeve with a radially variable end and a retainer; installing the sleeve assembly in the tube so the radially variable end of the first sleeve is received by the support section, the radially variable end being in a radially contracted configuration during installation and being in a radially expanded configuration after the sleeve assembly is installed in the tube; and deforming the retainer from an installation configuration to a retention configuration to retain the radially variable end of the first sleeve in the radially expanded configuration.

A method for replacing a damaged sleeve lining a tube passing through a nuclear reactor pressure vessel. The damaged sleeve has an end including a radially enlarged end portion for resting on a support section of the tube for retaining the damaged sleeve in the tube. The method includes removing the damaged sleeve from a tube; providing a sleeve assembly including a first sleeve with a radially variable end and a retainer; installing the sleeve assembly in the tube so the radially variable end of the first sleeve is received by the support section, the radially variable end being in a radially contracted configuration during installation and being in a radially expanded configuration after the sleeve assembly is installed in the tube; and deforming the retainer from an installation configuration to a retention configuration to retain the radially variable end of the first sleeve in the radially expanded configuration.

A method to determine a global core reactivity bias and the corresponding estimated critical conditions of a nuclear reactor core prior to achieving reactor criticality. The method first requires collection and evaluation of the inverse count rate ratio (ICRR) data; specifically, fitting measured ICRR vs. predicted ICRR data. The global core reactivity bias is then determined as the amount of uniform reactivity adjustment to the prediction that produces an ideal comparison between the measurement and the prediction.

A method to determine a global core reactivity bias and the corresponding estimated critical conditions of a nuclear reactor core prior to achieving reactor criticality. The method first requires collection and evaluation of the inverse count rate ratio (ICRR) data; specifically, fitting measured ICRR vs. predicted ICRR data. The global core reactivity bias is then determined as the amount of uniform reactivity adjustment to the prediction that produces an ideal comparison between the measurement and the prediction.

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.