General Electric Company

Dale F Taylor

Louis S Sammler

Leonard W Niedrach

John S Sicko

William H Stoddard, Jr

William S Passman

A method for increasing the resistance of zirconium alloy tubing to nodular corrosion by applying a protective anneal at a temperature within a clearly defined temperature range. Also, a zirconium alloy tubing having such protective anneal is disclosed. The protective anneal comprises heating exposed surfaces of zirconium tubing to a temperature range bounded at its lower limit by the temperature T , T being the temperature which at equilibrium conditions a critical concentration of solute exists in &agr;-matrices of the zirconium alloy to resist nodular corrosion, and bounded at its upper limit by the maximum temperature at which precipitates exist in association with the &agr; and &bgr; matrices in the particular zirconium alloy. In respect of Zircalloy-2 containing zirconium and the following metals by weight, namely 1.2-1.7% tin, 0.13-0.20% iron, 0.06-0.15% chromium, and 0.05-0.08% nickel, the lower temperature limit T is approximately 840 C and the upper limit is approximately 855 C.

Protective coarsening anneal for zirconium alloys

A method for increasing the resistance of zirconium alloy tubing to nodular corrosion by applying a protective anneal at a temperature within a clearly defined temperature range. Also, a zirconium alloy tubing having such protective anneal is disclosed. The protective anneal comprises heating exposed surfaces of zirconium tubing to a temperature range bounded at its lower limit by the temperature T.sub.c, T.sub.c being the temperature which at equilibrium conditions a critical concentration of solute exists in .alpha.-matrices of the zirconium alloy to resist nodular corrosion, and bounded at its upper limit by the maximum temperature at which precipitates exist in association with the .alpha. and .beta. matrices in the particular zirconium alloy. In respect of Zircalloy-2 containing zirconium and the following metals by weight, namely 1.2-1.7% tin, 0.13-0.20% iron, 0.06-0.15% chromium, and 0.05-0.08% nickel, the lower temperature limit T.sub.c is approximately 840 C and the upper limit is approximately 855 C.

A method for determining the susceptibility of a sample of Zircaloy alloy to nodular corrosion. A specimen of such Zircaloy sample is annealed at a temperature within a temperature range bounded at its upper limit by the temperature T.sub.c being the temperature at equilibrium wherein sufficient solute would exist in the (.alpha.-matrix of the particular zircaloy to resist nodular corrosion, and bounded by a lower temperature, such temperature being the temperature of the (.alpha.+precipitate)/(.alpha.+.beta.+precipitate) transus for the particular Zircaloy sample. For Zircaloy-2, such temperature range is from approximately 825.degree. C. to 841.degree. C. The specimen is maintained at such selected temperature within such temperature range for a measured period, and subsequently exposed to steam at a fixed temperature and pressure for a fixed time. If nodular corrosion does not appear the above steps are repeated, but on each different specimen increasing the measured time period until a last measured time period is obtained when nodular corrosion first appears on a specimen. The length of the last measured time period is used as an indicator of the sample's resistance to nodular corrosion, and thus conversely its susceptibility to nodular corrosion.

Method for determining corrosion susceptibility of nuclear fuel cladding

A zirconium-based alloy with a reduced ahoy content is described that has resistance to both uniform and nodular corrosion comparable to present zirconium-based alloy compositions, such as Zircaloy-2. The alloy represents in essence a modified or diluted Zircaloy-2 or Zircaloy-4. The alloys of this invention are also expected to have improved uniform corrosion resistance at under high burn-up conditions The alloy comprises 0.05-0.09 weight percent of iron, 0.03-0.05 weight percent of chromium, 0.02-0.04 weight percent of nickel, 1.2-1.7 weight percent of tin and 0-0.15 weight percent oxygen, with a balance of zirconium. The iron chromium and nickel alloying elements form precipitates in the alloy matrix. The alloy is suitable for use as a cladding material for a fuel element housing fissionable nuclear materials in water cooled nuclear fission reactor. These alloys can also be formed with a reduced nitrogen content of no more than 20 parts per million nitrogen by weight in order to provide enhanced resistance to nodular corrosion.

Corrosion resistant zirconium alloy for extended-life fuel cladding

A cladding is provided for use in housing fissionable material in water cooled nuclear fission reactors. The cladding has inner and outer surfaces and includes (1) a cross-section of a Zirconium-based alloyed matrix, and (2) alloying elements in sufficient concentration to form precipitates disposed in the matrix. The cladding includes no more than 20 parts per million nitrogen by weight and is typically a modified Zircaloy-2 or Zircaloy-4. Metallurgically bonded to the inner region of the cladding is a zirconium barrier layer.

Zirconium alloy fuel cladding

A reference electrode is disclosed, comprised of a ceramic housing comprised of a bottom and a sidewall means extending from the bottom to a cylindrical housing top to define an inner channel. The channel contains a silver electrode extending therefrom and a deposit of a silver salt. The housing having a bore substantially radial to the axis of the housing extending through oppositely facing sidewall means, and a ceramic rod configured to fit in the bore and form a friction bond is positioned in the bore to form an annular junction that permits transfer of ions across the junction. Means for electrically isolating the silver electrode are mounted on the housing top.

Reference electrode with annular junction

A method for annealing a Zircaloy member having a cold worked or beta quenched crystal structure to mitigate the reduction in nodular corrosion resistance caused by the anneal comprises, annealing the member in an atmosphere comprising oxygen and the balance an inert atmosphere to form an adherent black oxide on the member.

Method for annealing zircaloy to improve nodular corrosion resistance

A container for electrochemical reactants for an electrode is provided, as well as electrodes employing the same. The electrodes can withstand the rigorous environment of a nuclear reactor core. The design of the container reduces stresses in the electrode and significantly improves its reliability. The container has a generally cylindrical member of an insulator. The member has a base region with a cut-out extending therethrough. Sidewall means extend from the base to define a cavity for containing electrochemical reactants, e.g., silver chloride. An insert of an insulator at one end mates with the cut-out and at another end mates with and is brazed to a cap for the container. The insert has a second sidewall means for holding the electrochemical reactants, and is shaped to fit within the first sidewall means. The base of the insert and the the cut-out are metallized for acceptance of a silver base. The design of the container and insert obviate the need for a wire to hold down the cap, thus overcoming problems of previous electrodes.

Integral cap for electrode and electrode employing same

A container for electrochemical reactants for an electrode is provided, as well as electrodes employing the same. The electrodes can withstand the rigorous environment of a nuclear reactor core. The design of the container reduces stresses in the electrode and significantly improves its reliability. The container has a generally cylindrical member of an insulator. The member has a base region whose outer surface is a surface attachment region. The base region has an annular channel for a lead. The base region also has a generally conical cut-out. Sidewall means extend from said base to define a cavity for containing electrochemical reactants, e.g., silver chloride. A generally conical insert of an insulator mates with the cut-out. The outer surface of the insert, cut-out, and surface attachment region is metallized, and the insert is set in place by brazing. In the electrode the lead is in electrical contact with the apex of the insert.

Container and reference electrode for use in radiated aqueous

Zirconium-based corrosion resistant alloys for use primarily as a cladding material for fuel rods in a boiling water nuclear reactor consist essentially of by weight percent about 0.5 to 2.0 percent tin, about 0.24 to 0.40 percent of a solute composed of copper, nickel and iron, wherein the copper is at least 0.05 percent, and the balance zirconium. Nuclear fuel elements for use in the core of a nuclear reactor have improved corrosion resistant cladding made from these zirconium alloys or composite claddings have a surface layer of the corrosion resistant zirconium alloys metallurgically bonded to the outside surface of a Zircaloy alloy tube. The claddings may contain an inner barrier layer of moderate purity zirconium metallurgically bonded on the inside surface of the cladding to provide protection from fission products and gaseous impurities generated by the enclosed nuclear fuel.

Corrosion resistant zirconium alloys containing copper, nickel and iron

A nuclear fuel element for use in the core of a nuclear reactor is disclosed having an improved corrosion resistant cladding. The cladding is comprised of zirconium alloys containing in weight percent 0.5 to 2.0 percent tin, or 0.5 to 2.5 percent bismuth, or 0.5 to 2.5 percent bismuth and tin, and about 0.5 to 1.0 percent of a solute composed of a member selected from the group consisting of molybdenum, niobium, tellurium and mixtures thereof, and the balance zirconium. Composite claddings are disclosed having a surface layer of one of the corrosion resistant zirconium alloys metallurgically bonded to a Zircaloy alloy tube. Claddings may contain an inner barrier layer of a moderate purity zirconium metallurgically bonded on the inside surface of the cladding to provide protection from fission products and gaseous impurities generated by the enclosed nuclear fuel.

Corrosion resistant cladding for nuclear fuel rods

Corrosion resistant composite claddings for nuclear fuel rods

A conductivity probe able to withstand high water temperatures and pressures, and intense nuclear radiation is suitable for use within the core of the nuclear reactor. The probe is made with a ceramic insulator, such as sapphire, and a central electrode that is mounted on the insulator. A sleeve that supports the insulator and the central electrode is made from metals having a coefficient of thermal expansion compatible with the coefficient of thermal expansion of the insulator, such as Kovar. The central electrode and sleeve are bonded to the insulator by a brazed compressive seal that protects an internal conductor in the probe from intrusion of the reactor water. A counter electrode made from a corrosion resistant metal, such as Kovar, is attached to the sleeve so that it extends over the central electrode a fixed, spaced distance. A positioning and signal transfer arrangement operatively supports the sleeve and conveys electrical signals to and from the two electrodes.

Conductivity probe for use in the presence of high intensity nuclear

Zirconium-based corrosion resistant alloys for use primarily as a cladding material for fuel rods in a boiling water nuclear reactor consist essentially of by weight percent about 0.5 to 2.0 percent thin, about 0.24 to 0.40 percent of a solute composed of copper, nickel and iron, wherein the copper is at least 0.05 percent, and the balance zirconium. Nuclear fuel elements for use in the core of a nuclear reactor have improved corrosion resistant cladding made from these zirconium alloys or composite claddings have a surface layer of the corrosion resistant zirconium alloys metallurgically bonded to the outside surface of a Zircaloy alloy tube. The claddings may contain an inner barrier layer of moderate purity zirconium metallurigcally bonded on the inside surface of the cladding to procide protection from fission products and gaseous impurities generated by the enclosed nuclear fuel.

Zirconium-based corrosion resistant alloys for use primarily as a cladding material for fuel rods in a boiling water nuclear reactor which consist essentially of 0.5 to 2.5 weight percent bismuth, or alternatively, 0.5 to 2.5 weight percent of a mixture of tin and bismuth, 0.5-1.0 weight percent of a solute composed of a member selected from the group consisting of niobium, molybdenum, tellurium, and mixture thereof, alternatively, the solute will be composed of tellurium and will be in the range of 0.3-1.0 weight percent, and the balance being zirconium.

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Schenectady, NY, United States of America

Dale F Taylor