The present disclosure is generally related to methods, devices and systems for improving the performances of a Rod Cluster Control Assembly (RCCA) and/or a Control Element Assembly (CEA) to mitigate clad strain, especially in the high fluence region, during normal operation conditions and accident conditions. One method may include incorporating a device such as a powder collection and blockage device between the ceramic upper and ceramic lower absorber materials of the RCCA and/or CEA. Another method may include increasing the plenum volume by incorporating an axial hole into the top end plug extension. Another method may include increasing the plenum volume by incorporating an axial hole into the bottom end plug and optionally incorporating radial grooves in the bottom of the lower absorber material to provide a flow channel for gas expansion or generation to ensure that the lower absorber does not block the opening in the bottom end plug.

Radiation shielding and methods of manufacture are disclosed. A radiation shielding apparatus includes a matrix including matrix material; and a mixture positioned in the matrix, the mixture including: a neutron thermalizing material; and a neutron absorbing material mixed with the neutron thermalizing material. A reactivity control system includes a container rotatable around an axis; a divider positioned inside the container to define two or more compartments within the container; at least one neutron absorber positioned in at least one of the two or more compartments; and at least one neutron reflector positioned in another of the two or more compartments that is fluidly isolated from the at least one of the two or more compartments. A method of manufacturing radiation shielding material includes: fabricating a matrix; generating a mixture by mixing a neutron absorbing material, a neutron thermalizing material, and additive materials; and loading the mixture into the matrix.

Radiation shielding and methods of manufacture are disclosed. A radiation shielding apparatus includes a matrix including matrix material; and a mixture positioned in the matrix, the mixture including: a neutron thermalizing material; and a neutron absorbing material mixed with the neutron thermalizing material. A reactivity control system includes a container rotatable around an axis; a divider positioned inside the container to define two or more compartments within the container; at least one neutron absorber positioned in at least one of the two or more compartments; and at least one neutron reflector positioned in another of the two or more compartments that is fluidly isolated from the at least one of the two or more compartments. A method of manufacturing radiation shielding material includes: fabricating a matrix; generating a mixture by mixing a neutron absorbing material, a neutron thermalizing material, and additive materials; and loading the mixture into the matrix.

The invention relates generally to nuclear engineering and more particularly to controlled reactor start-up. The invention improves reliability of an operational neutron source by creating additional safety barriers between the coolant and the source active part materials. The operational neutron source is designed as a steel enclosure housing an ampule containing antimony and beryllium with separate antimony and beryllium cavities positioned coaxially. The antimony is contained in the central enclosure made of a niobium-based alloy unreactive with antimony. A beryllium powder bed is located between the antimony enclosure and the ampule enclosure. The ampule enclosure is made of martensite-ferrite steel poorly reacting with beryllium. An upper gas collector is located above the ampule, which serves as a compensation volume collecting gaseous fission products. At the bottom, the ampule is supported by a reflector and a bottom gas collector. The gas collectors, reflector and washers are made of martensite-ferrite grade steel.

The present disclosure relates to a pellet containing an oxide additive to improve a nuclear-fission-gas-adsorption ability of a uranium-dioxide pellet used as nuclear fuel and increase the grain size thereof, and to a method of manufacturing the same. A La.sub.2O.sub.3—Al.sub.2O.sub.3—SiO.sub.2 sintering additive is added to uranium dioxide so that mass movement is accelerated due to the liquid phase generated during sintering of the uranium-dioxide pellet, which promotes the growth of grains thereof. Further, since less volatilization occurs during sintering due to the low vapor pressure of the liquid phase, efficient additive performance is exhibited, so the liquid phase surrounding the grain boundary effectively adsorbs cesium, which is a nuclear fission gas.

The present disclosure relates to a pellet containing an oxide additive to improve a nuclear-fission-gas-adsorption ability of a uranium-dioxide pellet used as nuclear fuel and increase the grain size thereof, and to a method of manufacturing the same. A La.sub.2O.sub.3—Al.sub.2O.sub.3—SiO.sub.2 sintering additive is added to uranium dioxide so that mass movement is accelerated due to the liquid phase generated during sintering of the uranium-dioxide pellet, which promotes the growth of grains thereof. Further, since less volatilization occurs during sintering due to the low vapor pressure of the liquid phase, efficient additive performance is exhibited, so the liquid phase surrounding the grain boundary effectively adsorbs cesium, which is a nuclear fission gas.

An objective of the invention is to provide an austenitic stainless steel article having superior irradiation resistance and stress corrosion cracking resistance than before while maintaining mechanical properties equivalent to those of conventional ones. There is provided a dispersion strengthened austenitic stainless steel article, including: 16-26 mass % of Cr; 8-22 mass % of Ni; 0.005-0.08 mass % of C; 0.002-0.1 mass % of N; 0.02-0.4 mass % of O; at least one of 0.2-2.8 mass % of Zr, 0.4-5 mass % of Ta, and 0.2-2.6 mass % of Ti; and a balance consisting of Fe and inevitable impurities. The Zr, Ta and Ti components form inclusion particles in the stainless steel article by combining with the C, N and O components. The stainless steel article has an average grain size of 1 m or less and a maximum grain size of 5 m or less.

An objective of the invention is to provide an austenitic stainless steel article having superior irradiation resistance and stress corrosion cracking resistance than before while maintaining mechanical properties equivalent to those of conventional ones. There is provided a dispersion strengthened austenitic stainless steel article, including: 16-26 mass % of Cr; 8-22 mass % of Ni; 0.005-0.08 mass % of C; 0.002-0.1 mass % of N; 0.02-0.4 mass % of O; at least one of 0.2-2.8 mass % of Zr, 0.4-5 mass % of Ta, and 0.2-2.6 mass % of Ti; and a balance consisting of Fe and inevitable impurities. The Zr, Ta and Ti components form inclusion particles in the stainless steel article by combining with the C, N and O components. The stainless steel article has an average grain size of 1 m or less and a maximum grain size of 5 m or less.

A method for guaranteeing fast reactor core subcriticality under conditions of uncertainty involves, after assembling the reactor core, conducting physical measurements of reactor core subcriticality and comparing the obtained characteristics with design values; then, if there is a discrepancy between the values of the obtained characteristics and the design values, installing adjustable reactivity rods in the reactor at the level of a fuel portion of the reactor core, wherein the level of boron-B10 isotope enrichment of the adjustable reactivity rods is selected to be higher than the level of boron-B10 isotope enrichment of compensating rods of the reactor core. The technical result consists in improving the operating conditions of absorbing elements of a compensating group of rods, eliminating the need for increasing the movement thereof, simplifying monitoring technologies used during production, and simplifying the algorithm for safe reactor control.

A method for guaranteeing fast reactor core subcriticality under conditions of uncertainty involves, after assembling the reactor core, conducting physical measurements of reactor core subcriticality and comparing the obtained characteristics with design values; then, if there is a discrepancy between the values of the obtained characteristics and the design values, installing adjustable reactivity rods in the reactor at the level of a fuel portion of the reactor core, wherein the level of boron-B10 isotope enrichment of the adjustable reactivity rods is selected to be higher than the level of boron-B10 isotope enrichment of compensating rods of the reactor core. The technical result consists in improving the operating conditions of absorbing elements of a compensating group of rods, eliminating the need for increasing the movement thereof, simplifying monitoring technologies used during production, and simplifying the algorithm for safe reactor control.