Abstract: A system includes a nuclear reactor pressure vessel comprising a heat exchange wall. An interior of the vessel contains nuclear fuel which heats a primary fluid. A primary fluid circuit extends vertically within the wall. The primary fluid circulates in a loop that includes the interior and the primary fluid circuit in the wall. A secondary fluid circuit also extends vertically within the wall, and is fluidly isolated from the primary fluid circuit. A secondary fluid circulates in a loop that includes a power generating system and the secondary fluid circuit in the wall. The wall acts as a heat exchanger in which thermal energy is transferred to the wall from primary fluid flowing through the primary fluid circuit, and then is transferred from the wall to secondary fluid flowing through the secondary fluid circuit.

Abstract: The present invention relates to a method for dealing with anticipated transient without scram in a nuclear power plant, and a system therefor. The method for dealing with anticipated transient without scream in a nuclear power plant, according to the present invention, comprises the steps of: identifying the pressure of a pressurizer; determining an accident level on the basis of the determined pressure of the pressurizer; and activating safeguards related to the pressure, according to the accident level.

Abstract: A system and method for employing centrifugal mirror confinement in a nuclear fusion reactor for in-space propulsion and power generation. Centrifugal confinement represents an advancement beyond traditional mirror devices by further confining the fusion fuel and inhibiting instability growth, while still offering a simple magnet geometry and allowing a controllable fraction of the charged fusion products to leave each end of the mirror.

Abstract: An optical counter is used in a method and system for producing a nuclear fuel element having a known volume of homogeneously distributed nuclear material. The method includes feeding nuclear fuel kernels along a channel having a conveyer configured to transmit the kernels to an exit; driving the conveyer until a target number of kernels exits the channel through the exit; and counting a number of kernels which pass through the exit of the channel with an optical counter. The conveyer is stopped after the target number of nuclear fuel kernels exits the channel. The target number of nuclear fuel kernels are fed into a mold, and void space remaining in the mold is filled with a particulate matrix material so as to homogeneously distribute the target number of nuclear fuel kernels within the matrix material. The matrix material is then converted into a solid matrix material.

Abstract: A nuclear component is a nuclear fuel cladding, a spacer grid, a guide tube a plate fuel or an absorber rod. The nuclear component includes a support containing a substrate based on a metal which may be zirconium or titanium, the support being supported by a protective layer which is composed of a protective material comprising chromium. The nuclear component has improved resistance to oxidation, hydriding and/or migration of undesired material.

Abstract: A system comprising an emergency core cooling system (ECCS) valve, a nuclear reactor vessel, and a containment vessel surrounding the nuclear reactor vessel, is provided. The ECCS valve comprises an outer shell connected to the nuclear reactor vessel and protruding toward the containment vessel and having a connector formed therein; an inner shell provided inside the outer shell at a preset distance from an inner wall of the outer shell; a piston movably constrained and inserted through a piston opening formed in the inner shell at a position facing the connector of the outer shell to open and close the connector; a spring provided on an outer circumferential surface of the piston to provide a restoring force for moving the piston in a direction toward the nuclear reactor vessel; a fluid pump; a swing check valve; a reset valve; and a trip valve.

Abstract: The present disclosure relates to a device for use in a fuel assembly of a nuclear power plant. The device includes at least one rod, each rod including a plurality of containers having a space being filled with material to be activated. The device further includes a flow restrictor including a plurality of fingers adapted to extend respectively into a control rod guide tube of the fuel assembly. The at least one rod is connected to a finger of the flow restrictor. The containers are arranged subsequently in a direction of the longitudinal axis of the respective rod. The containers are connected to each other at a connecting portion. Each rod has a marking at a connecting portion. Each marking identifies the end of a space in a container and/or the end of a container.

Abstract: A reactor comprises a reactor vessel having therein a riser tube and a below-ground core. An outer annulus surrounds the vessel, and forms a downcomer. The reactor can produce a primary coolant flow rate above 10,000 kg/s with a pressure within a range of 6-15 bars. A steam drum is connected to the riser tube, and located at ground level with a pressure of 1-10 bars. A borated water storage tank is connected to the vessel. An injection system can cause the borated water to be injected into the vessel. The riser tube converts core-heated water to steam, which can be delivered to steam turbines that can generate power at a pressure of 1-10 bars. The conversion creates a difference in a primary coolant density that achieves a density-driven natural circulation of the primary coolant in the riser tube, the downcomer, the steam drum, and the core.

Abstract: In a sodium fast reactor, a bypass pipe is fluidly coupled to the primary sodium pump discharge and diverts a portion of the primary sodium coolant to an instrument assembly. The bypass pipe routes flowing sodium upward toward the reactor head where it fluidly couples to the instrument assembly. The instrument assembly includes an instrument tank and selectively swappable instrument modules. The instrument modules can be configured to measure flow, pressure, temperature, and fluid level, among other things. The instrument assembly is located relatively close to the reactor head and close to the sodium level in the sodium pool and is accessible from above the reactor head for quick and efficient removal and replacement of the entire instrument assembly or individual instruments.

Abstract: An inner sleeve for an integrated monopile system may be contained within an outer monopile that is installed in an earthen substrate. The inner sleeve includes at least one side wall, an open top end, a closed bottom end, and a cavity formed within the side wall, the open top end, and the closed bottom end. The cavity configured to house a power generator positioned inside the inner sleeve on or adjacent to the closed end, where the power generator being accessible through the open top end.

Abstract: A plasma processing method includes driving electric current through a Z-pinch plasma column within an atmospheric isolation device to produce a Lorentz force on the Z-pinch plasma column. Radiofrequency (RF) energy is supplied to the Z-Pinch plasma column with one or more RF generators. One or more RF applicators direct RF energy from the one or more RF generators toward the Z-pinch plasma column in a direction generally perpendicular to the direction of the electric current through the Z-pinch plasma column. One or more RF distributors distribute the RF energy to the one or more RF applicators.

Abstract: A method for measuring drop time of a control rod cluster integrated with a rod position measurement device is provided, wherein the method is used to measure the drop time of the control rod cluster, and includes following steps. A voltage Ua of coils in Group A is monitored to capture a rod cluster drop signal. A point (tmax, Vmax) with a maximum drop speed is searched. An end of a time period when the control rod cluster starts to drop is retroactively calculated from tmax. A start of the time period when the drop reference signal DROPref drops from a maximum value to 33% thereof is retroactively searched from a minimum value point of a drop reference signal DROPref. A time point t6 when a drop speed of the control rod cluster is lower than 0 is determined from tmax forward.

Abstract: A method of repairing a divertor or first wall surface in a tokamak plasma vessel. The divertor or first wall surface comprises a refractory metal having a melting point of at least 2000° C. A pressure of less than 25 mbar is maintained within the plasma vessel following the end of operation of the plasma vessel. The refractory metal is deposited onto the divertor or first wall surface within the plasma vessel via a deposition process which is one of: additive manufacturing; physical vapour deposition; thermal spray; arc ion plating; diode laser cladding; and chemical vapour deposition.

Abstract: A reactor unit cell is disclosed including a graphite moderator structure, a heat pipe positioned in the graphite moderator structure, and a fuel assembly positioned in the graphite moderator structure. The fuel assembly comprises a beryllium-oxide sleeve and nuclear fuel positioned in the beryllium-oxide sleeve.

Abstract: A nuclear fuel rod comprises nuclear fuel contained in a cladding, the cladding comprising a tube and two plugs, the tube extending along a central axis and having two ends, each plug being attached to a corresponding end of the tube by sealing the end of the tube. The tube is covered by a tube coating extending over the entire length of the tube from one end of the tube to the other.

Abstract: A system for mitigating and/or preventing potential fire hazards in a nuclear reactor including a heat pipe reactor core is provided. The system further comprises a plurality of heat pipes, the heat pipe reactor core engaging the plurality of heat pipes between first and second ends of each respective heat pipe. A heat exchanger device defining an enclosed gas chamber annularly surrounds at least a portion of at least one heat pipe. The system further comprises a first valve and a second valve positioned proximate the first end of the at least one heat pipe, the first valve fluidly coupling the at least one heat pipe to a first suppressant chamber and the second valve fluidly coupling the enclosed gas chamber to a second suppressant chamber. Each of the first and second valves is movable between open and closed positions to regulate the flow of fire suppressant material from the first and second suppressant chambers, respectively.

Abstract: Disclosed herein is a nuclear reactor system comprising a vessel encompassing a reactor core. An external loop line is connected to top and bottom portions of the vessel. A pump can circulate molten fuel through the reactor core and the external loop line. An external neutron reflector encompasses the vessel. External heaters heat the loop line and the reflector. The loop line, the reactor, or both, may be covered with insulation.

Abstract: An integrated control logic device and an operating method thereof adjust a position of a main control valve and a position of an auxiliary control valve. In particular, the position of the auxiliary control valve is adjusted by determining whether a change in the position of the main control valve is in a preset deadband range, thereby preventing a periodic water level fluctuation of a steam generator of a nuclear power plant.

Abstract: A method for determining real-time thermal power of a nuclear reactor based on number of gamma rays counted comprising measuring a number of gamma-ray counts by a gamma-ray detector that is placed outside a biological shield at a primary cooling circuit of a fission nuclear power plant so that maintenance of the detector is possible during normal operation of the fission nuclear power plant, and determining, by a computing device in real-time, a thermal power of the fission nuclear power plant based on the number of gamma-ray counts measured by the gamma-ray detector.

Abstract: A BWR channel box comprises a substrate, a first layer, and a second layer. The substrate comprises silicon carbide fibers. The first layer is deposited on a first surface of the substrate. The second layer is deposited on an opposite second surface of the substrate. Each layer comprises a corrosion resistant metallic composition. The combined weight of both layers is in a range from greater than 0% to 10% of the total weight of the channel box.