Provided is a nuclear fuel pellet having excellent compression resistance and a reduced missing pellet surface, the pellet being a cylindrical shape and including, on each of a top surface and a bottom surface thereof: a dish configured as a spherical groove shape having a predetermined curved surface at a center; a land configured as a horizontal annular shape along a rim of the dish; and a chamfer configured as a shape in which a corner is chamfered at a predetermined angle along a rim of the land, wherein the chamfer includes: a primary chamfer adjacent to the rim of the land; and a secondary chamfer configured as a shape additionally chamfered along a rim of the primary chamfer.

Provided is a nuclear fuel pellet having excellent impact resistance, the pellet being prepared with UO.sub.2 powder and having a cylindrical shape with a height of 9 to 13 mm and a horizontal cross-section diameter of 8 to 8.5 mm, and including: at each of a top surface and a bottom surface thereof, a dish configured as a spherical groove shape having a predetermined curvature and a groove diameter of 4.8 to 5.2 mm on a center; a shoulder configured as an annular plane along a rim of the dish; and a chamfer configured as a shape in which a corner is chamfered along a rim of the shoulder, wherein a width of the shoulder is 0.20 mm to 0.80 mm, and an angle between the chamfer and a horizontal plane is a 14-degree angle to 18-degree angle.

The present disclosure provides systems and methods for fast molten salt reactor fuel-salt preparation. In one implementation, the method may comprise providing fuel assemblies having fuel pellets, removing the fuel pellets and spent fuel constituents from the fuel assemblies, granulating the removed fuel pellets or process feed to a chlorination process, processing the granular spent fuel salt into chloride salt by ultimate reduction and chlorination of the uranium and associated fuel constituents chloride salt solution, enriching the granular spent fuel salt, chlorinating the enriched granular spent fuel salt to yield molten chloride salt fuel, analyzing, adjusting, and certifying the molten chloride salt fuel for end use in a molten salt reactor, pumping the molten chloride salt fuel and cooling the molten chloride salt fuel, and milling the solidified molten chloride salt fuel to predetermined specifications.

A nuclear power plant spent fuel negative pressure unloading system comprises a fuel element transport pipe and a gas transport pipe. The fuel element transport pipe comprises a fuel element output pipe, a fuel element lifting pipe, and a fuel element unloading pipe connected in series. The fuel element unloading pipe is arranged obliquely downward in the direction of fuel element movement. The distal end of the fuel element unloading pipe is connected sequentially to fuel loading apparatus and a transfer apparatus. Two nozzles of the gas transport pipe are connected to set positions on the fuel element output pipe and the fuel element unloading pipe respectively. A gas driving mechanism is connected to the gas transport pipe. An inlet of the gas driving mechanism is arranged at one end in proximity to the fuel element unloading pipe.

Disclosed herein is a method comprising coating a fissile, uranium-containing ceramic material with a water-resistant layer, the layer being non-reactive with the fissile, uranium-containing ceramic material. The coating is applied to a surface of the fissile, uranium-containing ceramic material. Also disclosed is a fuel for use in a nuclear reactor.

The known fully ceramic microencapsulated fuel (FCM) entrains fission products within a primary encapsulation that is the consolidated within a secondary ultra-high-temperature-ceramic of Silicon Carbide (SiC). In this way the potential for fission product release to the environment is significantly limited. In order to extend the performance of this fuel to higher temperature and more aggressive coolant environments, such as the hot-hydrogen of proposed nuclear rockets, a zirconium carbide matrix version of the FCM fuel has been invented. In addition to the novel nature to this very high temperature fuel, the ability to form these fragile TRISO microencapsulations within fully dense ZrC represent a significant achievement.

The invention is related to nuclear technologies, in particular, to the technology of producing nuclear oxide fuel for fuel elements, this oxide fuel can be used for manufacturing palletized nuclear fuel from uranium dioxide to be consumed by NPPs. The essence of the invention: this method of producing palletized nuclear fuel from uranium dioxide involves preparation of uranium dioxide moulding powder with/without uranium oxide, at this point powdered uranium dioxide is used as a raw material for preparation of moulding powder. Powdered uranium dioxide should be in the following proportion: O/U=2.370.04, it is obtained using a renowned methodby air heating of powdered uranium dioxide (ceramic grade) with the following proportion O/U=2.012.15. The technical result of the invention is increased mechanical strength of sintered pellets and a larger grain size of sintered pellets.

The present disclosure provides systems and methods for fast molten salt reactor fuel-salt preparation. In one implementation, the method may comprise providing fuel assemblies having fuel pellets, removing the fuel pellets and spent fuel constituents from the fuel assemblies, granulating the removed fuel pellets or process feed to a chlorination process, processing the granular spent fuel salt into chloride salt by ultimate reduction and chlorination of the uranium and associated fuel constituents chloride salt solution, enriching the granular spent fuel salt, chlorinating the enriched granular spent fuel salt to yield molten chloride salt fuel, analyzing, adjusting, and certifying the molten chloride salt fuel for end use in a molten salt reactor, pumping the molten chloride salt fuel and cooling the molten chloride salt fuel, and milling the solidified molten chloride salt fuel to predetermined specifications.

A method of forming a water resistant boundary on a fissile material for use in a water cooled nuclear reactor is described. The method comprises coating the fissile material, such as a pellet of U.sub.3Si.sub.2 and/or the grain boundaries, to a desired thickness with a suitable coating material, such as atomic layer deposition or a thermal spray process. The coating material may be any non-reactive material with a solubility at least as low as that of UO.sub.2. Exemplary coating materials include ZrSiO.sub.4, FeCrAl, Cr, Zr, AlCr, CrAl, ZrO.sub.2, CeO.sub.2, TiO.sub.2, SiO.sub.2, UO.sub.2, ZrB.sub.2, Na.sub.2OB.sub.2O.sub.3SiO.sub.2Al.sub.2O.sub.3 glass, Al.sub.2O.sub.3, Cr.sub.2O.sub.3, carbon, and SiC, and combinations thereof. The water resistant layer may be overlayed with a burnable absorber layer, such as ZrB.sub.2 or B.sub.2O.sub.3SiO.sub.2 glass.

A nuclear fuel element for a nuclear reactor comprises a body having a first region and a second region surrounded by the first region. The first segment comprises a poison material, and the second region comprises a nuclear fuel material and is substantially free of the poison material. A nuclear fuel element for use in a nuclear reactor comprises the body and a cladding material at least partially surrounding the body. Related methods of forming the nuclear fuel pellet include additive manufacturing processes to form first and second segments.