12 results about "Neutron moderator" patented technology

The invention relates to a combined cold and thermal neutron moderator (10) having a longitudinal axis (t) and at least one neutron exit window (50) at a first end (11, 12), and a cold moderator (20) arranged along the longitudinal axis (t), a thermal pre-moderator (30) and a vacuum space (40), said thermal pre-moderator (30) having a pre-moderator mantle (32) surrounding the cold moderator (20) along the longitudinal axis (t) and separated from the cold moderator (20) by the vacuum space (40), characterized in that the pre-moderator (30) comprises a pre¬moderator collar (34) extending at least at the first end (11, 12) from the pre¬moderator jacket (32) towards the longitudinal axis (t), which substantially covers the portion of the vacuum space (40) between the cold moderator (20) and the pre¬moderator mantle (32) from the direction of the neutron exit window (50), leaving substantially free at least one end (21, 22) of the cold moderator (20) at the end (11, 12).WO

A radioisotope production capsule is described. Each capsule includes generally an inner container for housing one of a target material and a neutron moderator, an outer container surrounding the inner container for housing the one of the target material and the neutron moderator not housed by the inner container, and cladding for isolating the target material from the neutron moderator. One or more modular capsules are placed in each of a plurality of plug fingers. Each single plug finger loaded with one or more capsules is inserted into a guide thimble of an array of guide thimbles in a fuel assembly.

The application discloses a neutron gamma detector with high dynamic range and time resolution, comprising: a plurality of layers of detection units, a plurality of neutron moderators and a detector shell; each detection unit is arranged along the axis of the detector shell, and a neutron moderator is arranged between each detection unit; the detection unit comprises a neutron / gamma sensitive scintillator array, a photomultiplier tube array and a preamplifier array, and each layer of neutron / gamma sensitive scintillator array is matched with one layer of photomultiplier tube array and one layer of preamplifier array. The detector utilizes the multiple layers of neutron moderators to form the gradient of the neutron and gamma ray detection efficiency of each layer of scintillator, the neutron / gamma sensitive scintillator array matched with the photomultiplier tube array and the preamplifier array can measure and obtain a pulse count matrix, and then the neutron and gamma count rate at each layer of detector can be extracted. Meanwhile, the time resolution and dynamic range are improved based on the structural design of the detector.

The invention discloses a composite neutron slowing-down material, and relates to the technical field of composite neutron slowing-down materials, the composite neutron slowing-down material comprises a composite hydrogen-containing material and fluoride.The preparation process comprises the steps that the hydrogen-containing material is smashed and screened to obtain particles with specific granularity, and the particles are mixed and dispersed with dried fluoride particles; adsorbing the fluoride on the surface of the hydrogen-containing material through mechanical stirring; carrying out melting treatment on the mixed material, melting the hydrogen-containing material to form a continuous phase, and uniformly dispersing fluoride particles under the action of shearing force and combining with a macromolecular chain through weak interaction; after the molten material is injected into a mold, heat preservation forming is conducted at the specific temperature, and annealing treatment is conducted to eliminate internal stress after cooling demolding. By regulating and controlling the particle size of the hydrogen-containing material, the hydrogen / fluorine mass ratio and the forming heat preservation temperature, uniform distribution of hydrogen atoms in the material, interface bonding strengthening and inhibition of fluoride on thermal degradation are achieved, so that the neutron moderation efficiency, the tensile strength and the thermal stability of the material are improved, and the material is suitable for neutron protection scenes in the nuclear industry.

A fuel rack for wet storage of nuclear fuel in a fuel pool comprises, in one embodiment, a base plate and a cellular body formed from slotted plates bonded to each other and stacked, coupled to the base plate. This cellular body comprises densely packed, upward-opening cells, each holding a nuclear fuel assembly. When the rack is submerged in water, open flax straps are formed between at least some of the water-filled cells, acting as neutron moderators to control the reactivity within the rack. In one embodiment, flax straps are interspersed between adjacent cells in a first direction along a first horizontal axis of the rack. However, flax straps are not interspersed between adjacent cells in a second direction perpendicular to a second horizontal axis. Other flax strap and cell arrangements are also provided. Tension elements penetrate some of the flax straps, and these tension elements compress the stack of plates to ensure stability.

The application belongs to the technical field of boron neutron capture therapy of tumors, and relates to an interventional boron neutron capture therapy device. The interventional boron neutron capture therapy device comprises a neutron generator, a neutron moderator and a neutron export injector. The neutron generator is used for generating neutrons by bombarding a target material with an incident high-energy particle beam. The neutron moderator is connected to the neutron generator and is used for slowing down the neutrons generated by the neutron generator into thermal neutrons and shielding gamma rays generated in the slowing down process. The neutron export injector is connected to the neutron moderator and is used for guiding and transporting the slowed-down thermal neutrons to a treatment site in a patient's body. Thus, the problem that the neutron transport scheme in the prior art cannot effectively balance neutron utilization rate, treatment depth and additional radiation damage borne by a patient can be solved.

A core of a fast nuclear reactor has at least one gas expansion module, at least one neutron absorber, and at least one neutron moderator. The gas expansion module has a hollow tubular structure with one end closed and the other end opened. The at least one neutron absorber adjoins an outer face, in a radial direction of the core, of each gas expansion module. The at least one neutron moderator adjoins an outer face, in a radial direction of the core, of each gas expansion module. The at least one neutron absorber also adjoins the at least one neutron moderator.

This invention provides a yttrium hydride material with a hydrogen permeation barrier coating and a method for preparing the hydrogen permeation barrier coating. The method includes the following steps: adding aluminum salt and chromium salt to a mixed solvent containing alcohol, reacting with stirring, and aging to prepare an alumina and chromium oxide composite sol-gel system; immersing a cleaned yttrium hydride substrate into the alumina and chromium oxide composite sol-gel system, and forming a composite sol-gel film on the surface of the yttrium hydride substrate using an dip-coating method; and calcining the yttrium hydride substrate containing the composite sol-gel film to transform the composite sol-gel film into a dense hydrogen permeation barrier coating. This invention prepares the raw materials constituting the coating into a sol system, and uses an dip-coating + calcination method to form a hydrogen permeation barrier coating on the surface of the yttrium hydride substrate. The yttrium hydride material with the hydrogen permeation barrier coating can be used in small nuclear reactors as a neutron moderator, which can effectively improve the service life of the yttrium hydride material.

The application relates to a high-temperature-resistant composite neutron moderator material and a manufacturing method thereof, and belongs to the nuclear power field. The manufacturing method of the high-temperature-resistant composite neutron moderator material comprises the following steps: determining the hydrogen partial pressure of a metal hydride at a working temperature, establishing a moderator material particle model comprising a metal hydride core, a buffer layer and a SiC sealing layer and calculating the stress borne by the SiC sealing layer at the working temperature; calculating the allowable stress of the SiC sealing layer at the working temperature and adjusting the moderator material particle model according to the allowable stress to obtain a corrected moderator material particle model; and manufacturing a high-temperature-resistant composite neutron moderator material finished product based on the corrected moderator material particle model. The method can be used for manufacturing a neutron moderator material with good stability and reliability at a high temperature, and improves the safety and reliability of a commercial nuclear power plant reactor.

A nuclear fission reactor comprising a passive thermal shut-down device, and a reactor core containing fissile fuel. The passive thermal shutdown device comprises a reservoir and a Pythagorean syphon. The reservoir is in thermal communication with the reactor core, and containing a neutron poison material which has a free surface. The Pythagorean syphon has an inlet within the reservoir, an outlet in fluid communication with the reactor core. When the neutron poison material overtops the syphon due to thermal expansion, the neutron poison material flows through the Pythagorean syphon and into the reactor core. Also provided are nuclear fission reactors having passive thermal shutdown devices using a Pythagorean syphon to control the flow of fissile fuel, neutron moderators, or neutron reflectors.

This application provides a spherical bed fuel molten salt breeder reactor system and its operation method. The system includes: a reactor body containing a solid spherical bed, the solid spherical bed comprising solid fuel spheres and neutron moderator spheres. The solid fuel spheres are used to initiate fission reactions to produce neutrons and completely confine fission products inside the spheres. The neutron moderator spheres are used to moderate the neutrons. A breeder cooling circuit connected to the reactor body circulates a liquid molten salt containing thorium but not fissile nuclides. When the liquid molten salt flows through the solid spherical bed, thorium-232 in the liquid molten salt absorbs neutrons and is then multiplied into uranium-233 or its precursors. An online chemical processing subsystem connected to the breeder cooling circuit is used to extract and enrich uranium-233 or its precursors from the liquid molten salt. A fuel loading and unloading system connected to the reactor body is used to adjust the composition ratio of solid fuel spheres and neutron moderator spheres in the solid spherical bed.

The application relates to a neutron source azimuth measurement device and method. The method comprises a neutron source to be measured, a polyhedral neutron detector, a neutron moderator structure and an electronic and data acquisition system; the neutron moderator structure is wrapped outside the polyhedral neutron detector; the polyhedral neutron detector is connected by a plurality of planar neutron detectors, each planar neutron detector is connected with the electronic and data acquisition system; the electronic and data acquisition system comprises a ratio calculation module and an azimuth acquisition module; the ratio calculation module obtains a neutron count rate according to the number of electric pulse signals output by the planar neutron detector per unit time, and obtains a relative ratio of the neutron count rate between each planar neutron detector, and the azimuth acquisition module is used for inputting the relative ratio into a pre-trained azimuth acquisition model to obtain the azimuth of the neutron source to be measured. The method can shorten the neutron source azimuth measurement time and improve the neutron source azimuth measurement precision.