According to embodiments, a light water reactor uranium fuel assembly is capable of reducing heating values of both Am-241 and Cm-244, to reduce the amount of generated vitrified waste without using fast reactors. The light water reactor uranium fuel assembly is a light water reactor uranium fuel assembly to be used in a nuclear fuel cycle that extracts. An americium isotope is extracted at the time of reprocessing of spent fuel to be added to a fuel, in which a weight fraction W (unit: wt %) of americium 241 to be added to a fuel heavy metal is in ranges of W<−0.006e.sup.2+0.12e−0.43 (enrichment: 5 wt % or more), W<−0.000356e+0.00357 (enrichment: 4.2 wt % or more and less than 5.0 wt %) with respect to an average enrichment of uranium 235 e (unit: wt %) of the fuel assembly.

A reactor control system for a nuclear reactor, the reactor control system comprising: one or more hollow tubes comprising neutron absorbing material, each having a first end and a second end; a pump connected to the first end of each hollow tube and operable to control the amount of a first fluid within the hollow tube, the first fluid comprising a neutron moderator, wherein: the pump is controlled based on a level of reactivity in the nuclear reactor, and the second end of the hollow tubes is in fluid communication with a second fluid, the second fluid having a neutron moderating capacity lower than 10% of that of the first fluid.

A computer-implemented simulation method of predicting local concentrations of constituents in coolant water anywhere along fuel rods within any fuel assembly mechanical design of a Boiling Water Reactor (BWR) potentially resulting in crud deposits on said fuel rods. The method is based on a sub-channel approach of predicting local mass fluxes of vapor and liquid in coolant water anywhere along fuel rods within any fuel assembly mechanical design of a Boiling Water Reactor (BWR) for given steady-state or transient boundary conditions. The sub-channel approach is based on the solution of mass, momentum and energy conservation equations for the vapor phase and the liquid phase, the liquid phase is represented by more than one field variable, and is specifically represented by three fields, with the vapor phase as a fourth field, consisting of droplets, a liquid base film, and disturbance waves. The method comprises:

simulating steady-state or transient boundary conditions, such as inlet coolant water flow into said sub-channels, the coolant water flow may have a predetermined flow velocity variation,

analyzing predefined parameters of said disturbance waves and base film, including wave velocity, wave frequency and base film thickness, and

analyzing liquid base film thickness between consecutive passing disturbance waves, to calculate local instantaneous impurity concentrations based on said simulated boundary conditions, the calculation is made for each fuel rod of the fuel assembly, wherein, for each fuel rod, the method further comprises comparing said calculated local instantaneous impurity concentration to a crud compound precipitation limit, and during the time said concentration is higher than said precipitation limit, crud is considered to have occurred. In a related simulation method also base film dryout, clad temperature increase, and drop entrainment from waves, may be determined.

A reactor core includes an inner core region that extends in a vertical direction, and has a plurality of first fuel pins accommodating an inner core fuel; an outer core region that extends in the vertical direction, is arranged to surround the inner core region from an outer peripheral side, and has a plurality of second fuel pins accommodating an outer core fuel; and a sodium plenum provided above the inner core region and the outer core region, in which a dimension of the outer core fuel in the vertical direction is larger than a dimension of the inner core fuel in the vertical direction, and the position of a center of the outer core fuel in the vertical direction is higher than the position of a center of the inner core fuel in the vertical direction.

A system and method for performing active scanning on a nuclear fuel rod are provided. The system includes an electrically-driven neutron generator including an ion source, an accelerator, and a target; a moderator surrounding the neutron generator and configured to moderate neutrons generated by the neutron generator; a fuel rod channel disposed within the moderator, the fuel rod channel configured to receive a nuclear fuel rod and subject the nuclear fuel rod to a predetermined neutron flux; and a plurality of radiation detectors. When the nuclear fuel rod is subjected to the predetermined neutron flux, neutrons induce a secondary radiation of prompt and delayed gamma emissions, neutron emission, or a combination thereof that are detected by the plurality of radiation detectors to determine an amount of fissile material in the nuclear fuel rod and a spatial distribution of the fissile material along a length of the nuclear fuel rod.

A system and method for performing active scanning on a nuclear fuel rod are provided. The system includes an electrically-driven neutron generator including an ion source, an accelerator, and a target; a moderator surrounding the neutron generator and configured to moderate neutrons generated by the neutron generator; a fuel rod channel disposed within the moderator, the fuel rod channel configured to receive a nuclear fuel rod and subject the nuclear fuel rod to a predetermined neutron flux; and a plurality of radiation detectors. When the nuclear fuel rod is subjected to the predetermined neutron flux, neutrons induce a secondary radiation of prompt and delayed gamma emissions, neutron emission, or a combination thereof that are detected by the plurality of radiation detectors to determine an amount of fissile material in the nuclear fuel rod and a spatial distribution of the fissile material along a length of the nuclear fuel rod.

One embodiment provides a multi-segment rod that includes a plurality of rod segments. The rod segments are removably mated to each other via mating structures in an axial direction. An irradiation target is disposed within at least one of the rod segments, and at least a portion of at least one mating structure includes one and/or more combinations of neutron absorbing materials.

A fuel assembly for a nuclear power boiling water reactor, including: a fuel channel defining a central fuel channel axis, fuel rods, each having a central fuel rod axis, at least 3 water channels for non-boiling water, each water channel having a central water channel axis and each water channel having a larger cross-sectional area than the cross-sectional area of (the average) fuel rod. The fuel rods comprise a first group of full length fuel rods and a second group of shorter fuel rods. The fuel assembly comprises at least 5 fuel rods which belong to said second group and which are positioned such that the central fuel rod axis of each of these at least 5 fuel rods is closer to the central fuel channel axis than any of the water channel axes of the water channels.

A fuel assembly for a nuclear power boiling water reactor, including: a fuel channel defining a central fuel channel axis, fuel rods, each having a central fuel rod axis, at least 3 water channels for non-boiling water, each water channel having a central water channel axis and each water channel having a larger cross-sectional area than the cross-sectional area of (the average) fuel rod. The fuel rods comprise a first group of full length fuel rods and a second group of shorter fuel rods. The fuel assembly comprises at least 5 fuel rods which belong to said second group and which are positioned such that the central fuel rod axis of each of these at least 5 fuel rods is closer to the central fuel channel axis than any of the water channel axes of the water channels.

A nuclear reactor fuel rod is a fuel rod for a light-water reactor. The nuclear reactor fuel rod includes a fuel cladding tube and an end plug, both of which are formed of a silicon carbide material. A bonding portion between the fuel cladding tube and the end plug is formed by brazing with a predetermined metal bonding material interposed, and/or by diffusion bonding. The predetermined metal bonding material has a solidus temperature of 1200° C. or higher. An outer surface of the bonding portion, and a portion of an outer surface of the fuel cladding tube and the end plug, which is adjacent to the outer surface of the bonding portion are covered by bonding-portion coating formed of a predetermined coating metal. The predetermined metal bonding material and the predetermined coating metal have an average linear expansion coefficient which is less than 10 ppm/K.