Abstract: A reactor core (104) includes fuel assemblies (202) arranged in a loading pattern (1100) in response to corresponding power levels of the fuel assemblies (202). A crud deposition model (412) is used to predict crud deposition on the fuel assemblies (202) and fuel pins (300) within the fuel assemblies (202). The prediction of crud deposition enables generation of the loading pattern (1100) and design of lattice structures (1600) for the fuel assemblies (202) that results in the reduction of total crud deposition in the reactor core (104) and causes a substantially uniform crud deposition on the fuel assemblies (202) of the reactor core (104).

Abstract: The present invention relates to a method for inhibiting stress corrosion cracking occurring on steam generator tubes in the secondary side of nuclear power plants and an inhibitor therefor. The method comprises supplying lanthanum boride as an inhibitor for stress corrosion cracking into the secondary feed water of the nuclear power plants. The method according to the present invention improves the resistance of stress corrosion cracking ten times or higher than no inhibitor, and five times or higher than the conventional inhibitor. In particular, the method according to the present invention shows a superior inhibiting effect in the lead-contaminated environment such that the complete inhibition of stress corrosion cracking can be achieved. As thus, the method can reduce the unexpected stop of operating nuclear power plants caused by the stress corrosion cracking on steam generator tubes, thereby reducing the cost required for maintaining and repairing the tubes.

Abstract: The method involves measuring hydrogen permeation in the tubes by mass spectrometry, wherein the tube is inserted into a high or ultrahigh vacuum device in which a mass spectrometer and a total pressure gauge are located. H2 or H2 and inert gas mixtures are circulated inside the tube at the required partial pressure. The tube is then heated and the appearance of H2 outside the tube is observed. The flow thereof inside the tube and emergence time, called permeation time, are determined based on permeation curves. The emergence time of the first microcrack is also determined.

Abstract: A method for controlling water quality in a nuclear reactor comprises a first and second steps. The first step is to make an amount of iron, which is carried into the nuclear reactor and corrosively eluted from structural material within the nuclear reactor into reactor water, at least twice as much as any one of an amount of nickel, which is carried into the nuclear reactor, and an amount of nickel, which is generated in the nuclear reactor. The second step is to limit an upper limit of concentration value of iron in system water supplied into the nuclear reactor to up to 0.10 ppb.

Abstract: A process for mitigating stress corrosion cracking of steam turbine components in a steam environment, includes coating the metal components of the steam turbine with a noble metal. The noble metal is preferably a platinum group metal selected from the group consisting of platinum, palladium, osmium, rhodium, ruthenium, iridium, and combinations comprising at least one of the foregoing platinum group metals. In another embodiment, the process comprises coating the metal components with a platinum group metal and introducing a reductant into the steam to mitigate the stress corrosion cracking. Also disclosed herein is a steam turbine comprising a metal component having a surface coated with a platinum group metal.

Abstract: Oxygen is mixed with heater drain water of power plant which contains fine particles of iron oxides such that the concentration of the dissolved oxygen in the heater drain water after the mixing of oxygen is in a range from 1 &mgr;g/liter to 20 &mgr;g/liter. The obtained heater drain water after oxygen mixing is filtered through a filter such as a hollow fiber filter to remove fine particles of iron oxides.

Abstract: A method and system for reducing stress corrosion cracking in a hot water system, such as a nuclear reactor, by reducing the electrochemical corrosion potential of components exposed to high temperature water within the structure. The method includes the steps of: providing a reducing species to the high temperature water; and providing a plurality of noble metal nanoparticles having a mean particle size of up to about 100 nm to the high temperature water during operation of the hot water system. The catalytic nanoparticles, which may contain at least one noble metal, form a colloidal suspension in the high temperature water and provide a catalytic surface on which a reducing species reacts with least one oxidizing species present in the high temperature water. The concentration of the oxidizing species is reduced by reaction with the reducing species on the catalytic surface, thereby reducing the electrochemical corrosion potential of the component.

Abstract: A support plate for retaining tube array spacing within a heat exchanger tube and shell structure. The support plate having a plurality of individual tube receiving apertures formed therein. Each aperture has at least three inwardly protruding members and bights are formed therebetween when the tube associated therewith is lodged in place to establish secondary fluid flow through the support plate. The inwardly protruding members terminate in flat lands that restrain but do not all contact the outer surface of the respective tube. These flat lands minimize fretting wear and eliminate potential gouging of the outer wall of the tube. The plate wall forming each aperture has an hourglass configuration which, inter alia, reduces pressure drop, turbulence and local deposition of magnetite and other particulates on the support plates.

Abstract: A filter for separating particles from cooling water in a nuclear plant of light water model. The filter has an inlet end (2) and an outlet end (3) and allow flowing through of the cooling water in a flow direction % from the inlet end to the outlet end. The filter comprises a plurality of channels (5) extending mainly in the flow from the inlet end to the outlet end. Each of the channels comprise a first (6) channel section provided closer to the inlet end, a second (8) channel section arranged closer to the outlet end, and an intermediate section (7) provided between the first and the second channel section. The intermediate section (7) has an extension d2 in a direction x transversely to the flow direction, which is essentially larger than the extension d2.

Abstract: A method for mitigating crack initiation and propagation on a surface of a metal component due to susceptibility to corrosion comprises depositing a metallic material on the surface of the component to form a coating, and then converting at least an outer layer of the coating to an electrically insulating material. The deposition of the metallic material is carried out by a method selected from the group consisting of wire-arc spraying, physical vapor deposition, and chemical vapor deposition. Electrochemical corrosion potential less than −0.23 VSHE based on the standard hydrogen electrode can be achieved with the method of coating of the present invention. This method is applied to produce coated structural components of water-cooled nuclear reactor.

Abstract: Method for controlling the level of iron in feedwater of a BWR wherein an iron-containing sacrificial electrode in placed contact with the feed water dissolution of the electrode is induced at a rate which provides a desired level of iron in the feedwater

Abstract: Method for controlling the amount of metal atoms deposited into an oxide layer present on a metal surface, which metal atoms increase the corrosion resistance of metal when present in the oxide film, wherein the metal surface is submerged in water at a selected temperature within the range of about to 200° to 550° F.; and a solution of a compound containing the metal which increases the corrosion resistance of the metal surface when present in the oxide film is injected into the water. The compound decomposes at the selected temperature to release atoms of the metal which incorporate in the oxide film at a desired loading.

Abstract: In a demineralization apparatus, a mixed bed of a gel type cation exchange resin having a moisture holding capacity of 41% or less or a degree of crosslinkage of 12% or greater is employed along with a porous type anion exchange resin. As a result, oxidation degradation of the cation exchange resin due to hydrogen peroxide can be inhibited and the performance of the ion exchange resins and of the condensate water demineralizer can be stabilized and maintained for a long period of time.

Abstract: Method for controlling the amount of metal atoms deposited into an oxide layer present on a metal surface, which metal atoms increase the corrosion resistance of metal when present in the oxide film, wherein the metal surface is submerged in water at a selected temperature within the range of about to 200° to 550° F.; and a solution of a compound containing the metal which increases the corrosion resistance of the metal surface when present in the oxide film is injected into the water. The compound decomposes at the selected temperature to release atoms of the metal which incorporate in the oxide film at a desired loading.

Abstract: A method for mitigating crack initiation and propagation on a surface of a metal component due to susceptibility to corrosion comprises depositing a metallic material on the surface of the component to form a coating, and then converting at least an outer layer of the coating to an electrically insulating material. The deposition of the metallic material is carried out by a method selected from the group consisting of wire-arc spraying, physical vapor deposition, and chemical vapor deposition. Electrochemical corrosion potential less than −0.23 VSHE based on the standard hydrogen electrode can be achieved with the method of coating of the present invention. This method is applied to produce coated structural components of water-cooled nuclear reactor.

Abstract: A stud enclosure for protecting a stud extending upwardly from a nuclear reactor pressure vessel (RPV) flange has a cylindrical can with a capped end and an open end. The capped end has an axially extending hole with a screw extending therein for fastening the stud enclosure to the stud. A seal ring is disposed adjacent the open end of the cylindrical can for sealing the ring on the RPV flange. A gas valve is disposed in the capped end of the cylindrical can for pressurizing the interior portion of the can with air. The stud enclosure is used to protect the RPV studs after the RPV head has been removed in order to permit access to the interior portions of the RPV.

Abstract: A preventive maintenance method and apparatus for a structural member in a reactor pressure vessel according to the present invention reduce a tensile residual stress on a surface of the structural member by impinging a water jet from a nozzle onto a plane surface of a deflector to thereby change direction of flow of the water jet, and impinging the water jet after being deflected onto the surface of the structural member. This method and apparatus are applicable to a narrow space portion, and can improve a residual stress on the surface of the structural member and can also prevent damage such as stress corrosion cracking.

Abstract: An opening portion 3 is produced by cutting a recirculating piping 72 connected to a recirculating nozzle 2 of a reactor pressure vessel 1. A repairing device is brought into an annulus 10 of the reactor pressure vessel 1 from the opening 3, thereby repairing the annulus 10.

Abstract: A jet pump for a nuclear reactor includes a riser and an inlet mixer having a set of nozzles and a mixing section for receiving coolant flow from the nozzles and suction flow from an annular space between the reactor vessel and the shroud core. To minimize or eliminate electrostatic deposition of charged particulates carried by the coolant on interior wall surface of the inlet-mixer of the jet pump, and also to inhibit stress corrosion cracking, the interior wall surfaces of the nozzles and mixing section are coated with a ceramic oxide such as TiO2 and Ta2O5 to thicknesses of about 0.5-1.5 microns.

Abstract: A sensor for measuring electrochemical corrosion potential, and a method for manufacturing a sensor, the sensor comprising a tubular ceramic probe having a closed tip at one end, the probe at least partially filled with a powder comprising metal and metal oxide; a metal support tube having one end receiving an opposite end of the probe, and joined thereto by a braze joint therewith; an electrical conductor extending through the support tube and into the probe, and having an end buried in the powder for electrical contact therewith; and a protective band bridging the probe and tube at the joint for sealing thereof, the protective band consisting essentially of a metallic coating.

Abstract: In a nuclear power plant, by injecting an amount of hydrogen small enough not to increase a radiation dose rate of the main steam system, ECP of metallic component materials composing a nuclear reactor can be decreased to suppress the potential of occurrence of IGSCC, and the control can be easily performed, and the operating cost can be suppressed to increase. Occurrence of intergranular stress corrosion cracking in metallic component materials in contact with reactor cooling water is suppressed by injecting zirconium hydroxide and hydrogen into the reactor cooling water to decrease the electrochemical corrosion potential of the metallic component materials.

Abstract: Oxygen is mixed with heater drain water of power plant which contains fine particles of iron oxides such that the concentration of the dissolved oxygen in the heater drain water after the mixing of oxygen is in a range from 1 &mgr;g/liter to 20 &mgr;g/liter. The obtained heater drain water after oxygen mixing is filtered through a filter such as a hollow fiber filter to remove fine particles of iron oxides.

Abstract: The present invention relates to a method for inhibiting stress corrosion cracking occurring on steam generator tubes in the secondary side of nuclear power plants and an inhibitor therefor. The method comprises supplying cerium boride as an inhibitor for stress corrosion cracking into the secondary feed water of the nuclear power plants. The method according to the present invention improves resistance of stress corrosion cracking three times or higher than no inhibitor, and two times or higher than the conventional inhibitor. In particular, the method according to the present invention shows a superior inhibiting effect in the lead-contaminated environment such that the complete inhibition of stress corrosion cracking can be achieved. As thus, the method can reduce the unexpected stop of operating nuclear power plants caused by the stress corrosion cracking on steam generator tubes, thereby reducing the cost required for maintaining and repairing the tubes.

Abstract: A method for reducing in situ the electrochemical corrosion potential and susceptibility to stress corrosion cracking of a nickel-base alloy and boiling water nuclear reactor components formed therefrom when in contact with high temperature water. The method comprises the steps of: adding a metal hydride to the high temperature water; dissociating the metal hydride in the high temperature water to form a metal and at least one hydrogen ion; and reducing the concentration of the oxidizing species by reacting the hydrogen ions with an oxidizing species, thereby reducing in situ the electrochemical corrosion potential of the nickel-base alloy. The method may further include the steps of reacting the metal with oxygen present in the high temperature water to form an insoluble oxide and incorporating the metal into the surface of the nickel-base alloy, thereby reducing the electrical conductivity of the surface of the nickel-base alloy.

Abstract: To suppress a decrease of thickness due to corrosion of structural members and to achieve a removal of radionuclides with good efficiency in a nuclear power plant, oxidation decontamination is first conducted. An aqueous potassium permanganate solution is supplied from a circulation line to a reactor pressure vessel, which is a stainless steel structural member, and a reactor water cleanup system piping and a drain piping, which are carbon steel structural members. These structural members are oxidation-decontaminated by the action of potassium permanganate. Then the above-mentioned structural members are reduction-decontaminated by using an aqueous oxalic acid solution. The aqueous oxalic acid solution contains hydrazine.

Abstract: An incore piping section maintenance system of a reactor comprises a maintenance system main body which is fixed to a maintenance target portion in a reactor pressure vessel or in the vicinity thereof to which a preventive-maintenance operation is executed, a support mechanism provided for the maintenance system main body so as to be movable in a reciprocal manner towards the maintenance target portion, a laser generator for generating a laser beam, a laser de-sensitization treatment apparatus which is rotatably supported around an axis of the support mechanism and which includes a laser irradiation section for irradiating the laser beam to the maintenance target portion, and an optical transmission element which guides the laser beam outputted from the laser generator to the laser de-sensitization treatment mechanism.

Abstract: An improved fuel element for use in a nuclear reactor comprised of a central core of nuclear material, which is surrounded by a composite cladding. The cladding has an outer metallic tubular portion comprised of well-known cladding alloys used for such purposes. Metallurgically bonded to the outer metallic tubular portion is a commercially pure zirconium microalloyed with a controlled quantity of iron. The zirconium microalloyed with iron produced an inner metallic barrier having a beneficial balance between stress corrosion crack resistance and corrosion resistance while retaining other beneficial properties of pure zirconium, such as ductility.

Abstract: A component (1) designed for use in a light water reactor and at least partly comprised by a metal and/or a metal alloy presents a coatings (4, 7) at its outer surface (3) and its inner surface(5). The coating (4 and 7 respectively) has as its task to protect the surface (3 and 5 respectively), against oxidation, corrosion, wear and hydration. The coating (4 and 7 respectively) suitably comprises at least one of zirconium dioxide (ZrO2) and zirconium nitride (ZrN).

Abstract: The present invention provides an improved scale conditioning composition and method that results in improved dissolution and disruption of tube scale, hardened sludge and other deposits composed primarily of highly densified magnetite such as those found in heat exchange vessels. After treatment with the advanced scale conditioning composition, these magnetite rich deposits are more easily removed using known and commercially available high pressure hydro-mechanical cleaning techniques. The present invention further provides effective cleaning in a short period of time and at relatively low temperatures, while reducing the amount of waste produced and reducing the resulting corrosion of carbon and low alloy steel components within the steam generator during the cleaning process.

Abstract: Surfaces of the structural components of a nuclear power plant exposed to reactor water are wetted with an electroless plating solution containing an electrical insulating substance. The electrical insulating substance has a high resistivity about 105 (100000) times those of the structural components of the nuclear power plant or above. A metal film containing the electrical insulating substance is formed on the surfaces of the structural components exposed to the reactor water by wetting the surfaces of the structural components with the electroless plating solution. Thus, the electrochemical corrosion potential of the structural components is reduced regardless of whether hydrogen is injected into the reactor water.

Abstract: A preventive maintenance method and apparatus for a structural member in a reactor pressure vessel according to the present invention reduce a tensile residual stress on a surface of the structural member by impinging a water jet from a nozzle onto a plane surface of a deflector to thereby change direction of flow of the water jet, and impinging the water jet after being deflected onto the surface of the structural member. This method and apparatus are applicable to a narrow space portion, and can improve a residual stress on the surface of the structural member and can also prevent damage such as stress corrosion cracking.

Abstract: A non-steady state computer model of water in a Boiling Water Reactor (BWR) primary water flow circuit is used to represent the water chemistry and noble metal loading during, for example, an in situ noble metal application process. The modeling software is provided on a laptop or portable computer for real-time use in the field at different reactor sites. After inputting data representing the initial state of reactor water chemistry and operating conditions of the reactor, the model determines the water chemistry, pH, conductivity and noble metal loading throughout the BWR primary water flow circuit, including selected sample locations, as a function of time. Results are used to determine whether technical specifications on conductivity or other chemistry-related parameters will be exceeded during the noble metal application process. Values of rate constants used for modeling noble metal reactions may be changed on site at the reactor during the application process.

Abstract: Surfaces of the structural components of a nuclear power plant exposed to reactor water are wetted with an electroless plating solution containing an electrical insulating substance. The electrical insulating substance has a high resistivity about 105 (100000) times those of the structural components of the nuclear power plant or above. A metal film containing the electrical insulating substance is formed on the surfaces of the structural components exposed to the reactor water by wetting the surfaces of the structural components with the electroless plating solution. Thus, the electrochemical corrosion potential of the structural components is reduced regardless of whether hydrogen is injected into the reactor water.

Abstract: A novel high-purity polymer dispersant and sludge conditioner added to the feedwater entering the secondary side of a nuclear steam generator for minimizing the accumulation of metal-oxide deposits within the nuclear steam generator during the continuing operation of the generator is disclosed. The high-purity polymer is selected from a group consisting of acrylic acid polymer, methacrylic acid polymer, acrylate polymer, methacrylate polymer, copolymers, and terpolymers, acrylate/acrylamide copolymer, acrylate/methacrylate copolymer, terpolymers, and mixtures thereof. Methods of making and applying the polymer dispersant and sludge conditioner are described. Means for removing the metal oxides and polymer from the blowdown stream are disclosed.

Abstract: A metal cooling tube of a water-cooled nuclear reactor, having an inner surface thereof exposed to an aqueous cooling medium containing hydrogen peroxide. The cooling tube has its inner surface coated with matter selected from the group consisting of the element manganese, molybdenum, zinc, copper, cadmium for absorbing such hydrogen peroxide and then affecting decomposition of the hydrogen peroxide in the aqueous medium. In preferred embodiment such coating is manganese and oxides thereof. A method for lowering the electrochemical corrosion potential of a metal allow cooling tube exposed to an aqueous medium in a water-cooled nuclear reactor is also disclosed. Such method comprises the step of coating an inner surface of such tube with matter selected from the group of elements comprising manganese, molybdenum, zinc, copper, cadmium, so as to permit absorption and hydrogen peroxide in such aqueous medium and effect decomposition of hydrogen peroxide in such aqueous medium.

Abstract: Method for controlling erosion and cracking in a metal component of a nuclear reactor, particularly in the highly concentrated primary and secondary systems of a PWR, comprising creating a catalytic surface on the component; and generating a stoichiometric excess of reductant the water of the reactor to reduce the oxidant concentration at the surface to substantially zero.

Abstract: An ECP sensor includes a tubular ceramic probe having a closed tip at one end packed with a metal and metal oxide powder. A metal support tube receives an opposite end of the probe, and is joined thereto by a braze joint therewith. An electrical conductor extends through the support tube and probe, and has an end buried in the powder for electrical contact therewith. A ceramic band bridges the probe and tube at the joint for sealing thereof.

Abstract: An electrochemical corrosion potential sensor includes a ceramic tip insulating member, and a sensor tip joined to the ceramic tip insulating member, the sensor tip comprising an alloy. Further, a coating is provided on an outer surface of the sensor tip, the coating including a noble metal, and a conductor electrically connected to said sensor tip.

Abstract: An ECP sensor includes a tubular ceramic probe having a closed tip at one end packed with a metal and metal oxide powder. A metal support tube receives an opposite end of the probe, and is joined thereto by a braze joint therewith. An electrical conductor extends through the support tube and probe, and has an end buried in the powder for electrical contact therewith. A ceramic band bridges the probe and tube at the joint for sealing thereof.

Abstract: In a nuclear power plant, by injecting an amount of hydrogen small enough not to increase a radiation dose rate of the main steam system, ECP of metallic component materials composing a nuclear reactor can be decreased to suppress the potential of occurrence of IGSCC, and the control can be easily performed, and the operating cost can be suppressed to increase. Occurrence of intergranular stress corrosion cracking in metallic component materials in contact with reactor cooling water is suppressed by injecting zirconium hydroxide and hydrogen into the reactor cooling water to decrease the electrochemical corrosion potential of the metallic component materials.

Abstract: A fuel assembly for a boiling water reactor comprising a plurality of fuel units, stacked on top of each other, each of which comprising a plurality of fuel rods extending vertically between a top tie plate and a bottom tie plate, and means for keeping the fuel elements together. The fuel elements are surrounded by a fuel channel with a substantially square cross section. At least two of the fuel units differ from each other in regard to fuel distribution or free flow area.

Abstract: An ECP sensor includes a tubular ceramic probe having a closed tip at one end packed with a metal and metal oxide powder. A metal support tube receives an opposite end of the probe, and is joined thereto by a braze joint therewith. An electrical conductor extends through the support tube and probe, and has an end buried in the powder for electrical contact therewith. A ceramic band bridges the probe and tube at the joint for sealing thereof.

Abstract: A preventive maintenance method and apparatus for a structural member in a reactor pressure vessel according to the present invention reduce a tensile residual stress on a surface of the structural member by impinging a water jet from a nozzle onto a plane surface of a deflector to thereby change direction of flow of the water jet, and impinging the water jet after being deflected onto the surface of the structural member. This method and apparatus are applicable to a narrow space portion, and can improve a residual stress on the surface of the structural member and can also prevent damage such as stress corrosion cracking.

Abstract: A preventive maintenance method and apparatus for a structural member in a reactor pressure vessel according to the present invention reduce a tensile residual stress on a surface of the structural member by impinging a water jet from a nozzle onto a plane surface of a deflector to thereby change direction of flow of the water jet, and impinging the water jet after being deflected onto the surface of the structural member. This method and apparatus are applicable to a narrow space portion, and can improve a residual stress on the surface of the structural member and can also prevent damage such as stress corrosion cracking.

Abstract: Disclosed is a process for introducing zinc into a water-containing component of the primary system of a nuclear power plant, in particular for preventing the deposition of radioactive cobalt and/or for reducing the possible susceptibility of materials in the component to corrosion. An effective level of zinc is introduced into the component as a dissolved zinc compound which is a zinc complex and/or metal zincate.

Abstract: An incore piping section maintenance system of a reactor comprises a maintenance system main body which is fixed to a maintenance target portion in a reactor pressure vessel or in the vicinity thereof to which a preventive-maintenance operation is executed, a support mechanism provided for the maintenance system main body so as to be movable in a reciprocal manner towards the maintenance target portion, a laser generator for generating a laser beam, a laser de-sensitization treatment apparatus which is rotatably supported around an axis of the support mechanism and which includes a laser irradiation section for irradiating the laser beam to the maintenance target portion, and an optical transmission element which guides the laser beam outputted from the laser generator to the laser de-sensitization treatment mechanism.

Abstract: The invention relates to a sensor for a measuring an electrochemical corrosion potential comprising a sensor tip, a conductor electrically connected to the sensor tip, an insulating member which surrounds the conductor, a connecting member which surrounds the conductor; and a sleeve which fits over the sensor tip, the insulating member, and the connecting member, the sleeve having inner threads which engage with corresponding outer threads on at least one of the sensor tip and the connecting member.

Abstract: An incore piping section maintenance system of a reactor comprises a maintenance system main body which is fixed to a maintenance target portion in a reactor pressure vessel or in the vicinity thereof to which a preventive-maintenance operation is executed, a support mechanism provided for the maintenance system main body so as to be movable in a reciprocal manner towards the maintenance target portion, a laser generator for generating a laser beam, a laser de-sensitization treatment apparatus which is rotatably supported around an axis of the support mechanism and which includes a laser irradiation section for irradiating the laser beam to the maintenance target portion, and an optical transmission element which guides the laser beam outputted from the laser generator to the laser de-sensitization treatment mechanism.

Abstract: A metal cooling tube of a water-cooled nuclear reactor, having an inner surface thereof exposed to an aqueous cooling medium containing hydrogen peroxide. The cooling tube has its inner surface coated with matter selected from the group consisting of the element manganese, molybdenum, zinc, copper, cadmium for absorbing such hydrogen peroxide and then affecting decomposition of the hydrogen peroxide in the aqueous medium. In preferred embodiment such coating is manganese and oxides thereof. A method for lowering the electrochemical corrosion potential of a metal allow cooling tube exposed to an aqueous medium in a water-cooled nuclear reactor is also disclosed. Such method comprises the step of coating an inner surface of such tube with matter selected from the group of elements comprising manganese, molybdenum, zinc, copper, cadmium, so as to permit absorption and hydrogen peroxide in such aqueous medium and effect decomposition of hydrogen peroxide in such aqueous medium.

Abstract: An improved fuel element for use in a nuclear reactor comprised of a central core of nuclear material, which is surrounded by a composite cladding. The cladding has an outer metallic tubular portion comprised of well-known cladding alloys used for such purposes. Metallurgically bonded to the outer metallic tubular portion is a commercially pure zirconium microalloyed with a controlled quantity of iron. The zirconium microalloyed with iron produced an inner metallic barrier having a beneficial balance between stress corrosion crack resistance and corrosion resistance while retaining other beneficial properties of pure zirconium, such as ductility.