Patent Number: 061920983
Section: description

DETAILED DESCRIPTION OF THE INVENTION With the foregoing and other objectives in view there is provided, in accordance with the invention, a nuclear fuel rod comprising a cladding tube of a zirconium alloy and having a highly corrosion resistant outer portion in which zirconium hydride precipitation is inhibited and an inner portion in which zirconium hydride precipitation is promoted. In accordance with the present invention, there is also provided a nuclear fuel rod comprising a multiple-layer composite cladding having an inner zirconium or zirconium alloy layer and a highly corrosion resistant zirconium or zirconium alloy outer layer having an outer portion in which hydride precipitation is inhibited and an inner portion of the outer layer in which zirconium hydride precipitation is promoted. In accordance with the present invention, there is further provided a nuclear fuel rod comprising a multi-layered composite cladding having a highly corrosion resistant zirconium or zirconium alloy outer layer in which hydride precipitation is inhibited, and a zirconium or zirconium alloy inner layer where hydride precipitation is promoted. In a recent research program involving corrosion studies of a composite two-layer cladding for a nuclear fuel rod for a light water reactor having an outer layer of Zircaloy 4 and a zirconium inner layer, it was observed by the present inventor that zirconium hydrides precipitated in the lower oxygen content zirconium inner layer and not in the outer higher oxygen content Zircaloy 4 layer whereas the temperature gradient across the wall of the composite cladding was such that hydride precipitation would have been expected in the Zircaloy 4 outer layer. The present inventor discovered that the amount of hydride precipitation is in part a function of the amount of oxygen in the zirconium alloy and zirconium metal and by limiting the oxygen content to a low level in an inner portion or layer of a zirconium metal and/or zirconium alloy fuel rod cladding and increasing the oxygen content to a higher level in an outer portion or layer of the cladding, enhanced resistance to hydride formation and corrosion is obtained in the outer portion or layer compared to conventional single layer or multiple layered claddings made of zirconium and/or zirconium alloys with a non-varying or uniform oxygen content across the cladding wall. During continued corrosion of the cladding when exposed in a nuclear reactor, the absence of hydride formation near the cladding outside surface limits the corrosion reaction to that described by Equation (1), whereas if hydrides precipitate in large quantities near the outside surface of the cladding, the corrosion reaction is more appropriately described by Equation (2) which, as explained above, leads to an acceleration in the corrosion rate. In accordance with the present invention, by inhibiting the formation of hydride precipitates in the outer portion or layer of the cladding, accelerated corrosion is inhibited. In accordance with the present invention, a nuclear fuel rod for water moderated or cooled reactors is provided having a metallic tubular cladding comprising a zirconium alloy and having a decreasing oxygen concentration gradient from the outer wall to the inner wall where despite being subject to an increasing temperature gradient from the outer wall to the inner wall of the cladding during reactor operations, zirconium hydrides preferentially precipitate in the inner portions of the cladding away from the cladding outside wall and are inhibited from forming on the outer wall. In a preferred embodiment, the oxygen content in the cladding tube decreases from an amount greater than approximately 1600 ppm at the outer wall of the cladding to less than approximately 1200 ppm at the inner wall of the cladding. In an alternative embodiment, a nuclear fuel rod for water moderated or cooled reactors is provided having a cladding tube comprising a composite of two or more layers of zirconium and/or zirconium alloy metals, the outermost layer having a higher oxygen content than an inner layer of the cladding where despite being subject to an increasing temperature gradient across the cladding wall during reactor operations, zirconium hydrides will preferentially precipitate in the inner layer of the cladding and will be preferentially inhibited from precipitating in the outermost layer of the cladding. In a preferred embodiment, the oxygen content of the outermost layer is greater than about 1600 ppm and the oxygen content of the inner layer is less than about 1200 ppm. In another alternative embodiment, a nuclear fuel rod for water moderated or cooled reactors is provided having a cladding tube comprising a composite of two or more layers, the outermost layer comprising a zirconium or zirconium alloy metal having a decreasing oxygen concentration gradient from the outer wall to the inner portion of the outermost layer, where despite a temperature gradient across the cladding wall, zirconium hydrides will preferentially precipitate in the inner portion of the outermost layer of the cladding and will be preferentially inhibited from precipitating near the outer wall of the outermost layer of the cladding. In a preferred embodiment, the oxygen content in the outermost layer of the composite cladding decreases from an amount greater than about 1600 ppm at the outer wall to less than about 1200 ppm in the inner portion of the outer layer of the composite cladding. In the outer portion or layer of the cladding comprised of the zirconium metal or alloy having the higher oxygen concentration, the hydrogen which was formed as a result of the reaction between the reactor coolant water and the zirconium and which was picked up in the outer portion or layer, continues to diffuse through the outer portion or layer and into the inner portion or layer of the cladding. The hydrogen does not precipitate to form hydrides in the oxygen-enriched outer portion or layer because the hydrogen concentration does not reach the raised solubility limit and diffuses into the low oxygen inner portion or layer where it preferentially precipitates as a result of the hydrogen concentration exceeding the lowered solubility limit. Opposing the preferential precipitation of hydrogen in the inner portion or layer low oxygen alloy is the temperature gradient across the wall of the cladding tube. Since the portions of the cladding closer to the fuel pellet are at a higher temperature than the portions of the cladding closer to or in contact with the reactor coolant water, the inner portion or layer low oxygen concentration alloy is at a higher temperature which raises the solubility limit of hydrogen in the metal alloy. Thus, the temperature gradient across the wall of the cladding tube affects the solubility limits of the inner layer and the outer layer and tends to encourage the precipitation of hydrogen as hydrides in the inner layer low oxygen metal alloy near the interface with the outer layer high oxygen metal alloy. Therefore, even in the presence of a temperature gradient which would promote hydride precipitation in the outermost layer of conventional cladding, such precipitation near the outside surface is prevented when the outer layer of the cladding comprises a high oxygen alloy and the inner portion of the cladding has a lower oxygen content. FIG. 1 represents a nuclear fuel assembly 10 for a pressurized water reactor (PWR) comprising a lower tie plate 12, guide tubes 14, nuclear fuel rods 18 which are spaced radially and supported by spacer grids 16 spaced along the guide tubes, an instrumentation tube 28, and an upper tie plate 26 attached to the upper ends of the guide tubes. Each fuel rod 18 generally includes a metallic tubular fuel rod cladding 100 within which are nuclear fuel pellets 80 composed of fissionable and/or fertile material and an upper end plug 22 and a lower end plug 24 which hermetically seal the nuclear fuel pellets within the metallic tubular fuel rod cladding as shown in FIG. 2. A helical spring member 21 can be positioned within the fuel rod to maintain the position of the fuel pellets in a stacked relationship. Control rods which are used to assist in controlling the fission reaction are disposed in the guide tubes, but are not shown. Referring to FIG. 3 which is a schematic representation of cross-sectional view of the nuclear fuel rod shown in FIG. 2, cladding 100 is a metallic tube having a single metal layer 101 of a zirconium alloy with a decreasing oxygen concentration gradient (depicted as a decreasing density in stippling) from outer wall 102 to inner wall 103. During reactor operations, fuel pellets 80 which are positioned within the cladding generate heat which is transferred through the cladding to outer wall 102 to the reactor coolant which results in a decreasing temperature gradient from inner wall 103 to outer wall 102 of the cladding. In a preferred embodiment, the oxygen concentration gradient from outer wall 102 to inner wall 103 decreases from at least about 1600 ppm to less than about 1200 ppm. Referring to FIG. 4A which is a schematic representation of a cross-sectional view of a nuclear fuel rod for a PWR such as shown in FIGS. 1 and 2 but instead has a composite cladding 110 comprising an outer layer 111 and an inner layer 114 each of which is composed of a zirconium and/or zirconium alloy metal. Outer layer 111 has a higher oxygen content than inner layer 114 (which is depicted as different stippling) so that hydrides will preferentially precipitate in inner layer 114 and be inhibited from precipitating in outer layer 111. In a preferred embodiment, outer layer 111 has an oxygen content greater than about 1600 ppm, and inner layer 25 has an oxygen content less than about 1200 ppm. Referring to FIG. 4B which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a PWR such as shown in FIGS. 1 and 2 but instead has a composite cladding 120 comprising an outer layer 121, an inner layer 124 and an innermost layer 127. Outer layer 121 and inner layer 124 are composed of a zirconium and/or zirconium alloy metal. Outer layer 121 has a higher oxygen content than inner layer 124 (which is depicted as different stippling) so that hydrides will preferentially precipitate in inner layer 124 and be inhibited from precipitating in outer layer 121. Innermost layer 127 can be zirconium or a zirconium alloy, or another metal. In a preferred embodiment, outer layer 121 has an oxygen content greater than about 1600 ppm, and inner layer 124 has an oxygen content less than about 1200 ppm. Referring to FIG. 4C which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a PWR such as shown in FIGS. 1 and 2 but instead has a composite cladding 130 comprising an outer layer 131, an inner layer 134 and an innermost layer 137 each of which is composed of a zirconium and/or zirconium alloy metal. Outer layer 131 has a higher oxygen content than inner layer 134 (which is depicted as different stippling) so that hydrides will preferentially precipitate in inner layer 134 and be inhibited from precipitating in outer layer 131. In a preferred embodiment, outer layer 131 has an oxygen content greater than about 1600 ppm, and inner layer 134 as an oxygen content less than about 1200 ppm. In another preferred embodiment, innermost layer 137 has an oxygen concentration which is at least that of inner layer 134 but less than or equal to outer layer 131. Referring to FIG. 5 which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a PWR such as shown in FIGS. 1 and 2 but instead has a composite cladding 140 which comprises at least two layers of zirconium and/or zirconium alloy metals, including an outermost layer 141 and an inner layer 144. Outermost layer 141 has a decreasing oxygen concentration gradient depicted as a decreasing variation in stippling from outer wall 142 to an inner portion 141a, where despite a temperature gradient across the cladding wall, zirconium hydrides will preferentially precipitate in inner portion 141a of outermost layer 141 of cladding 140 and will be preferentially inhibited from precipitating near outer wall 142 of outermost layer 141 of cladding 140. In an alternative embodiment, composite cladding can include an innermost layer formed from a zirconium metal or alloy, or another metal or alloy. In a preferred embodiment, the innermost layer is a zirconium metal or alloy, and the oxygen content in outermost layer 141 of cladding 140 decreases from an amount greater than about 1600 ppm at outer wall 142 to less than about 1200 ppm in the inner portion 141a of outer layer 141. Referring to FIG. 6, a nuclear fuel assembly for a boiling water reactor (BWR) in the U.S. is generally shown at 30 having nuclear fuel rods 32 which are supported between a lower tie plate 34A and upper tie plate 36. Each fuel rod generally includes a metallic tubular fuel rod cladding 150 within which are nuclear fuel pellets 80 which are hermetically sealed within the tubular cladding by end sealing means such as end plugs. Lower tie plate 34A and upper tie plate 36 are connected structurally by tie rods 40 positioned within the array of fuel rods or by other means such as an inner water channel. Spacer grids 38 provide intermediate support of the fuel rods 32 over the length of the fuel assembly and maintain them in a spaced relationship while restraining them from lateral vibration. Outer channel 42 surrounds the fuel assembly and extends from the lower tie plate to the upper tie plate. An example of nuclear fuel assembly for use in boiling water reactors outside the U.S. and typically in Europe is generally shown at 30 in FIG. 7 and similarly has tie rods 40, spacer grids 38, outer channel 42, and fuel rods 32 each generally including a metallic tubular fuel cladding 150 within which are nuclear fuel pellets 80. The fuel rods 32 are supported between a lower tie plate 34B and upper tie plate 36. Referring to FIG. 8, nuclear fuel rod 32 shown in FIGS. 6 and 7 includes nuclear fuel shown as a plurality of fuel pellets 80 of fissionable and/or fertile material positioned within a metallic tubular fuel rod cladding 150. The metallic tubular fuel rod cladding is sealed at its ends by means of end plugs 54 which may include alignment pins 33 to facilitate the mounting of the fuel rod in the assembly. A void space or plenum 56 is provided at one end of the fuel rod to permit longitudinal expansion of the nuclear fuel and accumulation of gases released from the nuclear fuel. A helical spring member 58 is positioned within space 56 and is capable of maintaining the position of the fuel pellets during handling and transportation of the fuel rods. Cladding 150 is secured to end plugs 54 by means of circumferential welds 62. Referring to FIG. 9 which is a schematic representation of a cross-sectional view of the nuclear fuel rod shown in FIG. 8, cladding 150 is a metallic tube having a single metal layer 151 of a zirconium alloy with a decreasing oxygen concentration gradient (depicted as a decreasing density in stippling) from outer wall 152 to inner wall 153. In a preferred embodiment, the oxygen concentration gradient from outer wall 152 to inner wall 153 decreases from at least about 1600 ppm to less than about 1200 ppm. Referring to FIG. 10A which is a schematic representation of a cross-sectional view of a nuclear fuel rod for a BWR such as shown in FIGS. 6-8 but instead has a composite cladding 160 comprising an outer layer 161 and an inner layer 164 each of which is composed of a zirconium and or zirconium alloy metal. Outer layer 161 has a higher oxygen content than inner layer 164 (which is depicted as different stippling) so that hydrides will preferentially precipitate in the inner layer 164 and be inhibited from precipitating in outer layer 161. In a preferred embodiment, outer layer 161 has an oxygen content greater than about 1600 ppm, and inner layer 164 has an oxygen content less than about 1200 ppm. Referring to FIG. 10B which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a BWR such as shown in FIGS. 6-8 but instead has a composite cladding 170 comprising an outer layer 171, an inner layer 174 and an innermost layer 177. Outer layer 171 and inner layer 174 are composed of a zirconium and or zirconium alloy metal. Outer layer 171 has a higher oxygen content than inner layer 174 (which is depicted as different stippling) so that hydrides will preferentially precipitate in the inner layer 174 and be inhibited from precipitating in outer layer 171. Innermost layer 177 can be zirconium or a zirconium alloy, or another metal. In a preferred embodiment, outer layer 171 has an oxygen content greater than about 1600 ppm, and inner layer 174 has an oxygen content less than about 1200 ppm. Referring to FIG. 10C which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a BWR such as shown in FIGS. 6-8 but instead has a composite cladding 180 comprising an outer layer 181, an inner layer 184, and an innermost layer 187 each of which is composed of a zirconium and or zirconium alloy metal. Outer layer 181 has a higher oxygen content than inner layer 184 (which is depicted as different stippling) so that hydrides will preferentially precipitate in the inner layer 184 and be inhibited from precipitating in outer layer 181. In a preferred embodiment, outer layer 181 has an oxygen content greater than about 1600 ppm, and inner layer 184 has an oxygen content less than about 1200 ppm. In another preferred embodiment, innermost layer 187 has an oxygen concentration which is at least that of inner layer 184 but less than or equal to outer layer 181. Referring to FIG. 11 which is a schematic representation of a cross-sectional view of another nuclear fuel rod for a BWR such as shown in FIGS. 6-8 but instead has a composite cladding 190 which comprises at least two layers of zirconium and/or zirconium alloy metals, including an outermost layer 191 and an innermost layer 197. Outermost layer 191 has a decreasing oxygen concentration gradient depicted as a decreasing variation in stippling from outer wall 192 to an inner portion 191a where despite a temperature gradient across the cladding wall, zirconium hydrides will preferentially precipitate in inner portion 191a of outermost layer 191 and will be preferentially inhibited from precipitating near outer wall 192 of outermost layer 191. Although innermost layer 197 can be zirconium metal or alloy, or another metal or alloy, in a preferred embodiment, innermost layer 197 is a zirconium metal or alloy, and the oxygen content in outermost layer 191 of cladding 190 decreases from an amount greater than about 1600 ppm at the outer wall 192 to less than about 1200 ppm in the inner portion 191. While the present invention has been particularly shown and described with reference to preferred embodiments thereof, it will be understood by those skilled in the art that various changes in form and details may be made therein without departing from the spirit and scope of the invention.