The invention relates to SiC ceramic matrix composite (CMC) claddings with metallic, ceramic and/or multilayer coatings applied on the outer surface for improved corrosion resistance and hermeticity protection. The coating includes one or more materials selected from FeCrAl, Y, Zr and Al—Cr alloys, Cr.sub.2O.sub.3, ZrO.sub.2 and other oxides, chromium carbides, CrN, Zr- and Y-silicates and silicides. The coatings are applied employing a variety of known surface treatment technologies including cold spray, thermal spray process, physical vapor deposition process (PVD), and slurry coating.

The invention relates to SiC ceramic matrix composite (CMC) claddings with metallic, ceramic and/or multilayer coatings applied on the outer surface for improved corrosion resistance and hermeticity protection. The coating includes one or more materials selected from FeCrAl, Y, Zr and Al—Cr alloys, Cr.sub.2O.sub.3, ZrO.sub.2 and other oxides, chromium carbides, CrN, Zr- and Y-silicates and silicides. The coatings are applied employing a variety of known surface treatment technologies including cold spray, thermal spray process, physical vapor deposition process (PVD), and slurry coating.

A manufacturing method provides a nuclear reactor component comprising a substrate and a coating covering a surface of the substrate. The manufacturing method comprises laser-marking a pattern on the surface of the substrate, the marking being carried out so as to form recessed reliefs outlining the pattern in the surface of the substrate, and then applying the coating to the surface of the substrate over the pattern.

A manufacturing method provides a nuclear reactor component comprising a substrate and a coating covering a surface of the substrate. The manufacturing method comprises laser-marking a pattern on the surface of the substrate, the marking being carried out so as to form recessed reliefs outlining the pattern in the surface of the substrate, and then applying the coating to the surface of the substrate over the pattern.

A nuclear fuel rod comprises nuclear fuel contained in a cladding, the cladding comprising a tube and two plugs, the tube extending along a central axis and having two ends, each plug being attached to a corresponding end of the tube by sealing the end of the tube. The tube is covered by a tube coating extending over the entire length of the tube from one end of the tube to the other.

A nuclear fuel rod comprises nuclear fuel contained in a cladding, the cladding comprising a tube and two plugs, the tube extending along a central axis and having two ends, each plug being attached to a corresponding end of the tube by sealing the end of the tube. The tube is covered by a tube coating extending over the entire length of the tube from one end of the tube to the other.

An assembly configured to seal an end plug on a plugged end of a fuel tube is disclosed. The assembly includes a seal weld chamber assembly and a helium flow stop assembly (HFSA) removably coupled to the seal weld chamber assembly. The seal weld chamber assembly includes a welding chamber and a plunger fluidically coupled to the welding chamber. A helium source is configured to supply helium to the welding chamber. The end plug of the fuel tube is positionable within the welding chamber via the plunger. The HFSA is configured to prevent helium from escaping the welding chamber through the plunger.

An assembly configured to seal an end plug on a plugged end of a fuel tube is disclosed. The assembly includes a seal weld chamber assembly and a helium flow stop assembly (HFSA) removably coupled to the seal weld chamber assembly. The seal weld chamber assembly includes a welding chamber and a plunger fluidically coupled to the welding chamber. A helium source is configured to supply helium to the welding chamber. The end plug of the fuel tube is positionable within the welding chamber via the plunger. The HFSA is configured to prevent helium from escaping the welding chamber through the plunger.

Embodiments of the invention are directed to a method for production of a nuclear fuel pellet by spark plasma sintering (SPS), wherein a fuel pellet with more than 80% TD or more than 90% TD is formed. The SPS can be performed with the imposition of a controlled uniaxial pressure applied at the maximum temperature of the processing to achieve a very high density, in excess of 95% TD, at temperatures of 850 to 1600° C. The formation of a fuel pellet can be carried out in one hour or less. In an embodiment of the invention, a nuclear fuel pellet comprises UO.sub.2 and a highly thermally conductive material, such as SiC or diamond.

Embodiments of the invention are directed to a method for production of a nuclear fuel pellet by spark plasma sintering (SPS), wherein a fuel pellet with more than 80% TD or more than 90% TD is formed. The SPS can be performed with the imposition of a controlled uniaxial pressure applied at the maximum temperature of the processing to achieve a very high density, in excess of 95% TD, at temperatures of 850 to 1600° C. The formation of a fuel pellet can be carried out in one hour or less. In an embodiment of the invention, a nuclear fuel pellet comprises UO.sub.2 and a highly thermally conductive material, such as SiC or diamond.