Thorium-based fuel bundles according to one or more embodiments of the present invention are used in existing PHWR reactors (e.g., Indian 220 MWe PHWR, Indian 540 MWe PHWR, Indian 700 MWe PHWR, CANDU 300/600/900) in place of conventional uranium-based fuel bundles, with little or no modifications to the reactor. The fuel composition of such bundles is 60+ wt % thorium, with the balance of fuel provided by low-enriched uranium (LEU), which has been enriched to 13-19.95% 235U. According to various embodiments, the use of such thorium-based fuel bundles provides (1) 100% of the nominal power over the entire life cycle of the core, (2) high burnup, and (3) non-proliferative spent fuel bundles having a total isotopic uranium concentration of less than 12 wt %. Reprocessing of spent fuel bundles is also avoided.

The invention relates to the nuclear industry and can be used for producing fuel pellets from uranium-molybdenum metal powders enriched to 7% uranium 235 for nuclear reactor fuel elements. The pellets are sintered in an inert atmosphere of argon at a temperature ranging from 1100° C. to 1155° C., and the initial powder is a uranium-molybdenum powder having a fraction size of 160 .Math.m and a molybdenum con¬tent of 9.0 to 10.5 wt%. The powder is pre-heated at a temperature of 500° C. for 10-20 hours (in an atmosphere of argon) and is subsequently cold pressed into pellets in a die under a force of up to 950 MPa. In an alternative emb¬odiment for producing uranium-molybdenum pellets with a binder (plasticizer), the step of sintering is preceded by heating the pellets in an atmosphere of argon at 300° C. to 450° C. for 2-4 hours to remove the binder. The invention makes it possible to increase the uranium intensity of the fuel, reduce the amount of heat buildup in a reactor core, and lower the amount of energy released in the event of abnormalities in the operation of a nuclear reactor, thus providing increased reactor safety and resilience to accidents.

Thorium-based fuel bundles are used in existing PHWR reactors (e.g., Indian 220 MWe PHWR, Indian 540 MWe PHWR, Indian 700 MWe PHWR, CANDU 300/600/900) in place of conventional uranium-based fuel bundles, with little or no modifications to the reactor. The fuel composition of such bundles is 60+ wt % thorium, with the balance of fuel provided by low-enriched uranium (LEU), which has been enriched to 13-19.95% .sup.235U. According to various embodiments, the use of such thorium-based fuel bundles provides (1) 100% of the nominal power over the entire life cycle of the core, (2) high burnup, and (3) non-proliferative spent fuel bundles having a total isotopic uranium concentration of less than 12 wt %. Reprocessing of spent fuel bundles is also avoided.

Disclosed are a nuclear fuel pellet having enhanced thermal conductivity and a method of manufacturing the same, the method including (a) a step of manufacturing a mixture including a nuclear fuel oxide powder and a thermally conductive plate-shaped metal powder; and (b) a step of molding and then heat-treating the thermally conductive plate-shaped metal powder to have an orientation in a horizontal direction in the mixture, thereby forming a pellet.

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.

A method for manufacturing a nuclear fuel compact is provided. The method includes forming an additive structure, consolidating a fuel matrix around the additive structure, and thermally processing the fuel matrix to form a fuel compact in which the additive structure is encapsulated therein. The additive structure optionally includes a vertical segment and a plurality of arm segments that extend generally radially from the vertical segment for conducting heat outwardly toward an exterior of the fuel compact. In addition to improving heat transfer, the additive structure may function as burnable absorbers, and may provide fission product trapping.

Disclosed herein is a method for manufacturing oxide fuel pellets. The method for manufacturing the oxide fuel pellets includes (step 1) preparing nuclear fuel powder containing uranium dioxide (UO2+x, x=0 to 0.20), (step 2) compacting the nuclear fuel powder prepared in step 1 to manufacture green pellets, sintering the green pellets manufactured in step 2 at a temperature of about 1,200° C. to about 1,400° C. by using an atmosphere gas, and reducing the green pellets sintered in step 3 at a temperature of about 800° C. to about 1,000° C. by using a reducing atmosphere gas. The method for manufacturing the oxide fuel pellets according to the present invention performs the sintering at a low temperature of about 1,200° C. to 1,400° C. to manufacture economical and safe oxide fuel pellets that are adequate for the nuclear fuel specification.

A device for transferring a given powder or a mixture of given powders contained in a container including a side wall and at least one discharge opening, the container with axisymmetric shape having an axis of rotation being arranged in the transfer device such that the discharge opening thereof is located in a lower portion of the container, the transfer device including rotating the container about the axis thereof, on which the discharge opening is located and control for controlling the rotation such that the rotation impose on at least one portion of the side wall of the container, referred to as movable portion, a first moving phase wherein an acceleration no lower than a minimum acceleration is capable of causing the powder to slide relative to the movable portion.

Device and method for checking fuel rods with IFBA, their zirconium diboride coating. The device includes a variable magnetic field generator and a magnetic field pickup device, arranged in the vicinity of the rod, as well as a control system for comparing both fields in order to measure the electric conductivity of the rod. The method includes the steps of: arranging the rod to be measured between the generator and the pickup device; generation of a variable magnetic field in the generator; picking-up of the magnetic field; comparison between the generated magnetic field and the picked-up one in order to quantify the electric conductivity of the rod; if the electric conductivity differs from a reference value, consider the rod for checking or recycling.

The present invention relates to a process for sintering a compacted powder of at least one oxide of a metal selected from an actinide and a lanthanide, this process comprising the following successive steps, carried out in a furnace and under an atmosphere comprising an inert gas, dihydrogen and water: (a) a temperature increase from an initial temperature T.sub.I up to a hold temperature T.sub.P, (b) maintaining the temperature at the hold temperature T.sub.P, and (c) a temperature decrease from the hold temperature T.sub.P down to a final temperature T.sub.F, in which the P(H.sub.2)/P(H.sub.2O) ratio is such that: 500<P(H.sub.2)/P(H.sub.2O)≦50 000, during step (a), from T.sub.I until a first intermediate temperature T.sub.i1 between 1000° C. and T.sub.P is reached, and P(H.sub.2)/P(H.sub.2O)≦500, at least during step (c), from a second intermediate temperature T.sub.i2 between T.sub.P and 1000° C., until T.sub.F is reached.