JPS6029915B2 - Composite cladding, fuel elements and manufacturing methods - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION The present invention relates generally to fuel elements for use within the core of a nuclear reactor.

More particularly, the present invention relates to a zirconium-containing composite coating for nuclear fuel having a copper coating and an intermediate boundary layer of zirconium oxide on the inner surface adjacent to the nuclear fuel. Nuclear reactors currently designed, constructed, and operated contain nuclear fuel in fuel elements having a variety of geometric shapes, such as plates, tubes, and rods.

Typically, nuclear fuel materials are enclosed within a corrosion-resistant, non-reactive, and flame-conducting container or cladding that has low neutron absorption. Fuel assemblies are formed by assembling the fuel elements thus obtained in a lattice pattern at regular intervals within coolant channels, and then by combining a sufficient number of fuel assemblies, nuclear fission is possible for sustained nuclear fission. A chain reaction system or core is formed. Such a core is housed within a reactor vessel, into which coolant is flowed. Coatings serve several purposes, but the two main ones are:

The first is nuclear fuel and coolant or (if moderator is present)
If there is a moderator, the goal is to prevent contact and chemical reaction with the coolant and moderator (if they coexist). Second, it prevents radioactive fission products, including gases, from being released from the nuclear fuel into the coolant or moderator (if the coolant and moderator coexist). be. Common coating materials include copper and its alloys, zirconium and its alloys, niobium (columbium) and its alloys, and the like. If the cladding is breached, or the seal is lost, the coolant or moderator and associated systems may become contaminated with radioactive fission products, thereby interfering with plant operation. However, in the manufacture and use of fuel elements coated with certain metals and metal alloys, problems have been encountered due to the mechanical behavior and chemical reactions of the coating under certain circumstances.

For example, zirconium and its alloys are excellent nuclear fuel cladding materials under normal circumstances. Because they have small neutron absorption cross sections, are strong, ductile, and extremely stable, they are commonly used as coolants or moderators in nuclear reactors at temperatures below about 75 qo. It shows no reactivity in the presence of deionized water and/or water vapor. However, when examining the performance of the fuel element, it was discovered that in the presence of certain fission products produced by nuclear fission reactions, cladding defects could occur due to mechanical interaction between the nuclear fuel and the cladding.

Furthermore, it has been found that such undesirable performance is facilitated by the localization of mechanical stresses (due to differences in thermal expansion between the nuclear fuel and the cladding) at the cracks and inter-pellet interfaces in the nuclear fuel. . That is, corrosive fission products are released from the nuclear fuel and are present at the inter-bellet interface and at the interface between the nuclear fuel cracks and the coating surface. Such fission products are produced in nuclear fuel during nuclear fission chain reactions during nuclear reactor operation. In that case, if the friction between the nuclear fuel and the cladding is large, the localized stress will be further magnified. During irradiation in a nuclear reactor, the zirconium alloy cladding of the fuel element is exposed to fission products.

Refractory bran-silk ceramic compositions (eg, uranium dioxide and other compositions) used as nuclear fuels release large amounts of fission products upon irradiation. Some of these fission products can react with the zirconium or zirconium alloy cladding that encapsulates the nuclear fuel. Another approach taken in nuclear reactor design is to coat the nuclear fuel materials with ceramics to protect them from exposure to moisture, as described in U.S. Pat. No. 3,108,936.

The fuel element described in U.S. Pat. No. 3,085,059 is comprised of a metal container enclosing one or more pellets of fissile ceramic fuel and a layer of vitreous material bonded thereto. . As a result, there is a layer of glassy material between the container and the nuclear fuel, which improves uniform heat transfer from the pellet to the container. No. 2,873,238 describes a skinned fissile uranium slug enclosed within a metal container, the protective skin for the slag consisting of a zinc-aluminum bonding layer. US Patent No. 2849
No. 387 discloses a skinned nuclear fuel fragment that forms an effective thermally conductive bonding layer between the uranium fragment and the vessel (or cladding) by immersion into the molten layer of the bonding material. has been done. The skin in this case is defined as being made of any alloy with good thermal conductivity, examples of which include aluminum-silicon alloys and zinc-aluminum alloys. Hakama Kosho No. 47-114200 describes a method in which a silicon carbide coating is applied to one of the two pellets and a pyrolytic carbon or metal carbide coating is applied to the other. Placing a coating on nuclear fuel material is unreliable because it is difficult to obtain a uniform coating free of defects.

Moreover, degradation of such coatings can also cause problems in terms of the performance life of the nuclear fuel material. It has also been proposed to add metals such as niobium to nuclear fuel as a way to prevent the nuclear fuel cladding from developing defects.

Such additives can be incorporated as powders as long as subsequent nuclear fuel processing operations do not cause oxidation of the metal. Alternatively, twilled, lamellar or other shaped additives may be introduced within, around or between the fuel pellets. Document GEAP-4555, dated February 196, describes a composite cladding made of a zirconium alloy with a lining of stainless copper bonded to the zirconium alloy.

Such a composite cladding can be produced by extruding a hollow billet of zirconium alloy with a stainless steel lining. However, in such a composite coating, the stainless steel layer has a thickness of about 70% compared to a zirconium alloy layer of the same thickness.
It has the disadvantage of resulting in neutron absorption losses of ~15. US Pat. No. 3,502,54 describes a method for producing composite materials for nuclear reactors by depositing chromium to protect zirconium and its alloys.

Energia-Nucleale (Energia-Nucleale)
eare) Volume 11, No. 9 (September 1964), 505
Pages 508 to 508 introduce a method in which copper is exposed on the surface of Zircaloy 2 and then heat treated to achieve surface diffusion of the deposited copper. Fehu Hlotsa (F.B.
"Stability and compatibility of hydrogen barriers used for zirconium alloys" (European Atomic Energy Community, Joint Nuclear Research Center, EUR 409$)
196 Gyakuten) describes methods for depositing various coatings and the efficiency of such coatings as hydrogen diffusion barriers;
And the most promising barrier to hydrogen diffusion is A.
A Si film is mentioned. “Zirconium and Zirconium-Tin Electroplating” (BMI) by W. C. Schiekner et al.
757, Technical Information Bureau, 1952) discloses a method of forming an alloy diffusion bonding layer by plating nickel on zirconium and a zirconium-tin alloy with open air and then subjecting it to heat treatment. US Patent No. 362
The specification of No. 5821 describes a fuel for a nuclear reactor that includes a fuel cladding tube having a coating of a retaining metal (for example, nickel) with a small neutron capture cross section on the inner surface, and in which fine particles of a burnable poison are arranged. elements are listed. 1973
The Nuclear Reactor Development Program Progress Report (ANL-RDP-19), dated August 2013, discloses a chemical getter mechanism that places an ant-like chromium layer on the inner surface of the stainless steel cladding. has been done.

Yet another option is to introduce a barrier between the nuclear fuel material and the cladding.

Examples are US Pat. No. 3,230,150 (copper foil), German Patent DAS 1,238,115 (titanium layer), US Pat.
crystalline carbon barrier between uranium dioxide and zirconium coating) and US Pat. No. 3,088,893 (stainless steel foil). Although the idea of using such a barrier is controversial, the materials mentioned in some of the above references cannot coexist with nuclear fuel (for example, because carbon reacts with oxygen from nuclear fuel). Sometimes it is impossible to coexist with the coating (e.g. because knots or other metals react with the coating and change its properties), and sometimes it is impossible (e.g. because it acts as a neutron absorber). coexistence with fission products may be impossible. Also, no solution to the recently discovered problem of localized chemical and mechanical interactions between the nuclear fuel and the cladding is described in any of the above references. Other solutions based on the idea of using a barrier are disclosed in U.S. Pat. method of installing) and U.S. Patent No. 3925
No. 151 (Method of disposing a zirconium, niobium, or alloy filter liner between the nuclear fuel and the coating and at the same time disposing a coating of a material with high anti-friction properties between the liner and the coating) disclosed in the specification has been done.

Other efforts to solve the problem of protecting zirconium or zirconium alloy coated containers are shown in U.S. Pat. No. 4,029,545 to Wrdon et al., assigned to the same assignee as the present invention. There is.

According to this, chromium is hazy-plated onto the zirconium or zirconium alloy coating, and then copper is hazy-plated onto the chromium layer. However, it has been discovered that electroplating chromium with a zirconium or zirconium alloy coating (hereinafter sometimes referred to simply as a zirconium coating) is not economical, and the overall process therefore turns out to be less promising than originally anticipated. U.S. Pat. No. 402,266 to Gordon et al.
An alternative plan is presented in the specification No. 2, according to which a metal liner (for example, a steel liner) is placed between the nuclear fuel and the cladding.
A fuel element is provided having a diffusion barrier (eg, a chrome coating) between the liner and the cladding. The fuel element in this case is also uneconomical due to the requirement for debonding and the production of live copper. Therefore, our goal is to find an economical solution to the problem of preventing perforation and failure of the cladding substrate due to metal warping and stress corrosion cracking caused by the interaction of fuel pellets with the cladding. Research efforts continued. Now, in the present invention, if the surface of the zirconium-coated substrate attached to the nuclear fuel material is covered with a copper coating and an intermediate boundary layer of zirconium oxide is disposed between the copper coating and the zirconium-coated substrate, the fuel pellet and the coating can be separated. It is based on the discovery that metal arming and stress corrosion cracking caused by interactions can be substantially reduced.

In that case, the intermediate boundary layer of zirconium oxide was found to provide an excellent diffusion barrier for copper. Furthermore, although copper cannot be electroplated directly onto non-conductive zirconium oxide, it is possible to install copper coatings by electroless plating if the surface of the zirconium-coated substrate is modified prior to oxidation. It also became clear that this was the case. According to one aspect of the invention, a core material comprising a nuclear fuel material; and b) a zirconium or zirconium alloy substrate, copper, nickel, iron or the like covering the inner surface of the zirconium or zirconium alloy substrate in close proximity to the nuclear fuel material. A fuel element consisting of both an elongated and composite cladding vessel for nuclear fuel material consisting of a metal coating of a zirconium or zirconium alloy substrate and an oxide film of a zirconium or zirconium alloy substrate disposed between the zirconium or zirconium alloy substrate and the metal coating. provided.

According to another aspect of the present invention, 'a] the surface of the zirconium or zirconium alloy container is roughened or corroded, and 'b) the surface of the zirconium or zirconium alloy container thus treated is oxidized. forming an oxide film, {c-activating the oxide film surface of the zirconium or zirconium alloy container so that a metal coating can be installed by electroless plating, and then {d-activating the zirconium alloy or zirconium alloy container An improved fuel element that prevents stress corrosion cracking or metal embrittlement in zirconium or zirconium alloy containers for containing nuclear fuel materials by installing a metal film of copper, nickel, iron, or their alloys on the surface of the oxidized film. A method is provided that allows the production of.

The invention will now be described in more detail with reference to the accompanying drawings.

Turning first to FIG. 1, a partially cutaway cross-sectional view of fuel assembly 10 is shown.

Such a fuel assembly 10' consists of a tubular coolant channel 11 with a substantially square cross section, the upper end of which is provided with a lifting handle 12, and the lower end of which is provided with a tip member ( However, since the lower part of the fuel assembly 10 is omitted, it is not shown). The upper end of the channel 11 is open at 13, and the lower end of the tip member is provided with an opening for coolant flow. Within the channel 11 is a group of fuel elements (or fuel rods) 14.
are arranged and supported by an upper end plate 15 and a lower end plate (not shown because the lower part is omitted). Typically, the liquid coolant enters through an opening at the lower end of the tip, passes upwardly around the fuel element 14, and then exits at an elevated temperature through an outlet 13 at the top. In that case,
In a boiling water reactor, the coolant is partially vaporized, and in a pressurized water reactor, the coolant is not vaporized. Both ends of the fuel element 14 have plugs 18 welded to the cladding 17.
sealed by.

The bung 18 may also be provided with a strut 19 to facilitate installation of the fuel element into the fuel assembly.
A cavity (or plenum) 20 is provided at one end of the fuel element, which allows longitudinal expansion of the nuclear fuel material and accumulation of gas released from the nuclear fuel material. A nuclear fuel material holding means 24 in the form of a helical member is arranged inside, which serves to prevent the pellet from moving in the vertical direction, especially during handling and transport of the fuel elements. Such fuel elements are designed to provide excellent heat transfer between the nuclear fuel material and the cladding while avoiding bowing and vibrations that can sometimes occur due to high velocity coolant flow. Fuel element 14, shown in partial section in FIG. 1, is constructed in accordance with the present invention.

Such fuel element 14 includes a core of nuclear fuel material 16, illustrated as a number of fuel pellets of fissile material and/or fuel parent material disposed within a cladding 17. ing. In some cases,
Fuel pellets may have various shapes, such as cylindrical or spherical, and may also contain other forms of nuclear fuel (e.g. granular,
Nuclear fuel) may also be used. It should be noted that the physical form of the nuclear fuel is not important to the present invention. A variety of nuclear fuel materials can be used, including uranium compounds, plutonium compounds, thorium compounds, and mixtures thereof, but preferred is uranium dioxide or a mixture of uranium dioxide and rutonium dioxide. Referring now to FIG. 2, the nuclear fuel material 16 forming the core material of the fuel element 14 is surrounded by a cladding (hereinafter also referred to as a composite cladding) 17. Composite coating 17 includes a substrate 21 of zirconium or a zirconium alloy (eg Zircaloy-2). An oxide film 22 of the substrate as a diffusion barrier is bonded to the inner surface of the substrate. As a result, the diffusion barrier 2
2 forms a shield that prevents other chemical species from diffusing into the substrate 21 through the diffusion barrier 22 . The diffusion barrier 22 is approximately 1×
10-5 inches (about 2.54 x 10-5 to about 12.7 x
Preferably, it consists of a zirconium dioxide layer having a thickness of 10-4 mm. Such a diffusion barrier may also be applied to the substrate 21
It also prevents the metal coating 23 from contacting and reacting.
A metal coating 23 is bonded to the diffusion barrier 22 .

As a result, the metal coating 23 covers the diffusion barrier 22 and at the same time provides a shield that protects the substrate 21 against fission products and gaseous impurities released from the nuclear fuel material encapsulated within the coating. The metal coating 23 has a thickness of about 2×10-4 to about 4×1
0-4 inches (approx. 5. female x 10-4 to approx. 1.02 x 10
It consists of a metal layer with a thickness of -3 arc) for neutron loss. Such metals include copper, nickel, iron, and alloys thereof, with copper being preferred among them. Such a steel coating serves as a primary or preferential reaction site for the fission products, and also acts as a barrier to prevent the substrate 21 from contacting and reacting with harmful fission products. The purity of the copper coating is important from the perspective of neutron loss. The total amount of impurities in the two layers is limited to less than 4 pm in terms of boron.

Moreover, in order to maintain high ductility and good thermal conductivity, it is necessary to keep impurities at a level of less than 1% (by weight), preferably less than 100 pm. In the composite coating of the fuel element of the present invention, the diffusion barrier is tightly bonded to the substrate and the metal coating is tightly bonded to the diffusion barrier. When we conducted tests to examine the strength of the bond between the diffusion barrier and the substrate, we found that the diffusion barrier remained firmly attached even when the elastic region was bent or a permanent strain of approximately 2% was applied. It has been found. Conditions found in commercial nuclear reactors such as 500-75 capes (260-39900)
At temperatures of , copper coatings are more resistant to deleterious effects such as radiation hardening and damage than zirconium and zirconium alloys. Therefore, under nuclear reactor operating conditions, copper has a greater ability to withstand plastic deformation without mechanical failure than zirconium and zirconium alloys. Moreover, the copper plastically changes in accordance with the stresses that the pellet experiences during power transients, thereby relieving such stresses. Furthermore, such metals will not fracture mechanically, thus protecting the zirconium or zirconium alloy substrate from the deleterious effects of fission products. A diffusion barrier bonded to a zirconium or zirconium alloy substrate has a thickness of about 0.0001 to about 0.001 inches (about 0.0
Bonded metal coatings with thicknesses on the order of 0.00254 to about 0.00254 C) provide sufficient stress relief and chemical protection to prevent failure centers from forming in the composite coating substrate. It was found that resistance was obtained.

Such metal coatings provide significant chemical resistance to fission products and gases that may be present in the fuel element, thereby preventing such fission products and gases from contacting the substrate of the composite coating. For example, it has been found that the copper coating undergoes little oxidation, thus stabilizing the quantitative relationship of the components of the uranium dioxide fuel.

If the copper coating were not present, the zirconium or zirconium alloy would react with the oxide nuclear fuel to form zirconium dioxide, thus changing the quantitative relationship of the components of the oxide nuclear fuel. The chemical states of the various fission products are highly dependent on the quantitative relationship of the components of the oxide nuclear fuel. For example, at high oxygen/uranium ratios, cesium forms compounds with uranium dioxide. However, as the ratio decreases, the compound becomes unstable and cesium can therefore migrate to cold regions of the fuel element (e.g. to the inner surface of the cladding), in which case the cesium (alone or together with other fission products) This will prevent stress corrosion of the cladding.For fuel elements with uncoated cladding, even if the initial oxygen/uranium ratio of the oxide nuclear fuel is large,
Oxidation of the zirconium or zirconium alloy results in a lower ratio of oxygen as a result of which cesium can be released and migrate to the inner surface of the coating. In the case of the present invention using metal coatings and diffusion barriers, the above ratio remains approximately constant or changes at a slow rate. Thus, oxide nuclear fuels with any desired quantitative relationships can be used within such composite claddings, and the quantitative relationships can remain constant or vary at a much slower rate. It is expected. According to the practice of the invention, the zirconium or zirconium alloy tube (hereinafter simply referred to as zirconium tube or zirconium substrate) comprises a zirconium tube with a copper coating on its inner surface and a zirconium tube oxide layer forming an intermediate boundary layer. Converted to composite cladding.

For this purpose, the inner surface of the zirconium tube is first modified. Refurbishment of the internal surface of the zirconium tube can be accomplished by grit blasting or Roramil treatment or by the use of certain caustic agents. After modifying the inner surface of the zirconium tube, its oxidation is carried out. The oxidized inner surface of the zirconium tube is then activated to allow electroless plating of metal (eg, copper) onto the oxide film of the zirconium tube. If the inner surface of the zirconium tube is modified by roughening techniques, grit blasting is carried out using aluminum oxide grit or a weight having an outer diameter of about 8 to IQ and an inner diameter of about 5 to 7 feet. The inner surface may be roller milled using a coated aluminum oxide tube. In the latter case, the zirconium tube may be rotated at a speed of 12 to 20 RPM for 24 to 7 hours after filling the moist aluminum oxide powder and plugging the ends of the zirconium tube. When modifying the inner surface of a zirconium tube by the use of corrosion techniques, it is preferred to clean the inner surface of the tube with a detergent, expose it to a brightening solution, and then rinse with water. Suitable caustic agents are shown in US Pat. No. 4,017,368. To describe the curved corrosion procedure, about 10 to 2 degrees of ammonium hydrogen fluoride and about 0.
75-2. A zirconium tube is brought into contact with an aged activated aqueous solution containing sulfuric acid. To age such a solution, a piece of zirconium with an area of 100 square meters coated with Solution 1 may be immersed for 10 hours. If desired, the loose coating on the corroded interior surface of the zirconium tube may then be stripped away. In order to oxidize a zirconium tube that has been subjected to scattering or corrosion and film removal as described above, it is sufficient to expose it to an oxygen atmosphere for a period of 1 to 10 hours at a temperature of 300 to 50,000 °C. Alternatively, the inner surface of the zirconium tube after modification may be subjected to steam treatment at a temperature of 350 to 450 qo for 5 to 5 days. Experience has shown that tin salts and various noble metal salts can be used to activate the oxidized interior surfaces of zirconium tubes.

A preferred combination consists of an alkaline solution of stannous (e.g. sodium stannite) and palladium chloride. However, other noble metal salts, such as silver nitrate, platinum chloride, gold chloride, etc., as well as alkaline solutions of noble metals, such as sodium goldate, sodium palladate, sodium platinate, etc., may also be used. Curved activation mixtures are described in C.R. Shipley, U.S. Pat.
It is shown in the 5-mono technical bulletin. In fact, the oxidized inner surface of the zirconium tube is manufactured by Cuposit Catalyst, a product of the Shipley Company of Newton, Massachusetts, USA.
atalyst) 9F. Next, after washing the oxidized inner surface with water, the Cuposit Accelera, also a product of Shipley, was applied.
tor) 19.

Electroless plating of the activated oxidized inner surface of the zirconium tube can be performed according to standard procedures. For example, a plating solution may be uniformly flowed through the zirconium tube to uniformly deposit the metal on the inner surface of the tube. Although the preferred metal is copper, other metals (such as nickel or iron) can be plated onto the oxidized interior surface of the zirconium tube with satisfactory results. For electroless plating, an aqueous plating grade having the following composition can be used.

i.e. 600% of water, potassium sodium tartrate (
KNaC4 was 141.5 g on 2006-4, 41.5 g of sodium hydroxide (NaCH) and copper sulfate (CuS).
04・Spot 20) After mixing the two groups, add water and leave it overnight. Also, 73% formaldehyde solution (2 CO per day) 16.7% may be added immediately before use. This is a modification of the known Fehling steel plating. Besides that, MacDermid 90
Commercially available electroless steel plating grades such as No. 38, Sibley CP74 and Sel-Rex Cu510 can be used. In practice, it may be uniformly flowed through the zirconium tube while maintaining a temperature of about 25 to about 75 degrees centigrade. According to this procedure, a plated layer containing almost no pores and exhibiting extremely good adhesion while remaining stuck is obtained. To ensure that pre-plated zirconium tubes can be used at elevated temperatures without substantial loss of adhesion, it is necessary to
Degassing takes place at a temperature of 0.00 to about 40 Cape (149 to 20400). During such degassing, approximately 5
The temperature may be raised from room temperature to the final temperature at a rate of 0 to 125 degrees Fahrenheit (10 to 520 degrees Fahrenheit). When electroless copper is plated on zirconium tubes, a considerable amount of hydrogen gas may be generated.

Since hydrogen gas tends to adhere to the tube walls and thus may interfere with the electroless plating process, it is preferred to facilitate the removal of hydrogen gas by flowing the plating layer through the tube. Moreover, it is also preferable to perform electroless plating with the tube in an upright position. In order to apply nickel plating to zirconium tubes, a water-based plating grade having the following composition is used.

That is, 3 females/its nickel chloride (Njc〆・fine 20
), log/its sodium hypophosphite (Na ratio P02
・QO), 12.6g/sodium citrate (Na
After preparing an aqueous solution containing 2C6day507, Pan20) and 5g of its sodium acetate (NaC2ma0), p may be adjusted within the range of 4 to 6 by adding sodium hydroxide (NaOH). Alternatively, commercially available electroless nickel plating products such as Enpla 410 and 416 can be used. Actually, about 19
It may be uniformly poured into a zirconium tube while maintaining a temperature of 4 to about 212 degrees Fahrenheit (90 to 100 degrees Celsius). In addition, the suitable target temperature in that case is 95±
It is 2℃. According to this procedure, a plated layer that does not contain pores and exhibits extremely good adhesion even when stuck is obtained. To ensure that plated zirconium tubes can be used at elevated temperatures without substantial loss of adhesion, degassing is carried out following the same procedure as for copper. Zirconium tubes treated by the method of the invention are
It may be as obtained from the cutting operation, or it may have been previously subjected to mechanical cleaning operations (e.g. grit blasting) or chemical cleaning operations (e.g. acid or alkaline corrosion). It may be.

In order to enable those skilled in the art to more easily practice the invention, the following examples are provided.

These examples are merely illustrative of the practice of the invention and are not intended to limit the scope of the invention.
EXAMPLE 1 Two 5-inch long (12.7 arc) Zircairo tubes with an outside diameter of 10.490 inches (1.24 skins) and an inside diameter of 0.425 inches (1.08 skins) were subjected to 50 watts ultrasound. It was washed for 10 minutes with detergent liquid in a washer.

Next, it was washed with distilled water for 10 minutes. Then, 500 g of water, 500 g of concentrated nitric acid, and 1 l of ammonium hydrogen fluoride.
A brightening solution consisting of 0.0 g was flowed into the tube for 2 minutes at a flow rate of about 1000'/min. The tube was then washed with water and neutralized with an aqueous sodium hydroxide solution. After rinsing with distilled water for 5 minutes, the tubes were eroded for 1 minute in an ultrasonic cleaner using a solution consisting of 1,000 parts water, 1 part ammonium hydrogen fluoride, and 0.5 parts sulfuric acid. The corrosive solution was premixed and then aged by contacting it for 10 minutes with a piece of two Zircaloy tubes with an area of approximately 100 cm of fin.An ultrasonic cleaner was used to remove the scale-like material produced during corrosion. It was useful. After corrosion, the tubes were flushed with distilled water for approximately 1 minute and then dried using dry nitrogen. The tube was then placed in an oxidation furnace and oxidized at 400 DEG C. for two periods with a flow of about 0.2 cubic feet of oxygen per hour. After cooling, the tube was removed from the furnace and cleaned in an ultrasonic cleaner containing an aqueous sodium hydroxide solution for 5 minutes.
Then, it was washed with distilled water for 1 minute. To activate the tubes, a solution of Cuposite Catalyst Group, manufactured by Sibley, Inc., Newton, Mass., was run into the tubes for 3 minutes at a flow rate of 1000' per minute, followed by a 3 minute water rinse. Next, the cuposit
Accelerator 19 solution at a flow rate of 1000 m/min 6
The tube was flushed for 10 minutes and then rinsed with distilled water for 10 minutes. After that, Metex, a product of Magdermid Company of Waterford, Connecticut, USA, was introduced.
The tubes were plated at 60° C. for 2 hours in a +9038 plating Yunaka. Note that the metal was poured into the tube at a flow rate of 1,000 minutes per minute from a temperature-controlled container. The result is a copper coating approximately 3.8 x 10-4 inches thick on the inner surface and a copper coating approximately 4 inches thick.
A composite coating consisting of two Zircaloy tubes with an intermediate boundary layer of oxide film was obtained. Then 0.4 inch x 1.5 inch (1.02 skin x
A fuel element usable in the core of a nuclear reactor was produced by loading uranium dioxide pellets of 3.81g) into the composite cladding described above. An intermediate boundary layer of an oxide film on the zirconium substrate disposed between the copper coating and the zirconium substrate is exposed to reactor conditions (i.e., conditions in which it is exposed to temperatures exceeding 290 qo while in contact with cadmium dissolved in cesium, etc.). In order to prove that it has excellent ability as a means to reduce metal arm formation and breakage at the bottom,
1/8 inch (0.318 C) gauge thickness and 1'
A series of Zircaloy 2 tensile specimens having a length of 2 inches (1.27 mm) were prepared. The tensile test specimen was immersed in liquid cesium saturated with cadmium while an Instron tensile tester was used to
Evaluation was performed at a temperature of oo. Some tensile test specimens include
Prior to the above tensile test in liquid cesium, a heat treatment was performed in argon or vacuum at about 58,000 °C for 21/2 hours. The tensile test specimens evaluated were Zircaloy-2 without wind coating, Zircaloy-2 with curved steel coating, and Zircaloy-2 with copper coating and Zircaloy-2 between the copper coating and the Zircaloy-2 substrate. It consisted of Zircaloy 2 containing an intermediate boundary layer of an oxide film on the substrate.

The results obtained are shown in the table below. In the table, the "Y" in the "Heat Treatment" column indicates that the tensile specimen was exposed to a temperature of 580 oo for 21'4 in argon or vacuum prior to the Instron tensile test. According to the authors, it was confirmed that a Zircaloy-2 tensile test piece with a copper coating and an intermediate boundary layer of an oxide film on a Zircaloy-2 substrate exhibited the maximum plastic strain at break.Unexpectedly, is also the plastic strain at break of a tensile test piece placed in an environment of liquid cesium saturated with cadmium after heat treatment.It is 3.8% compared to the plastic strain at break of a tensile test piece that was not heat treated. It was even bigger.

These facts may suggest that the fuel element of the present invention exhibits superior failure resistance when exposed to actual operating conditions for extended periods of time. Such a zirconium coating resists verticalization better because it is protected by the copper coating and at the same time the diffusion of copper into the zirconium substrate is also prevented by the oxide layer of the zirconium substrate as a diffusion barrier. As is obvious to those skilled in the art, 1%
This indicates that it can withstand cracking quite well even though it has plastic strain at break. It is also important to note that the tensile test specimen [B] having only a copper coating was damaged after the heat treatment. Since there is no oxide film on the zirconium substrate as a diffusion barrier, copper diffusion into the zirconium substrate occurs when heated to 580°C, resulting in arm formation and Damage had occurred. Example 2 The procedure of Example 1 was repeated, except that instead of etching the zirconium substrate prior to oxidation, a 1 x 1.5 arc flat coupon was roughened by grit blasting.

That is, the surface was roughened by applying mechanical abrasion with a 90-mesh aluminum oxide grout over one inkstone sand interval. Then, according to the procedure of Example 1, 400 qo
The coupons were oxidized for 224 hours. Example 3 The procedure of Example 2 was repeated except that two tubes of Zircaloy were used instead of flat coupons.

In this case, the surface is roughened by roller milling using an aluminum oxide tube with an outer diameter of 0.31 inch (0.787 inch) and an inner diameter of 0.28 inch (0.711 inch) as a roller. I did it. That is, the above aluminum oxide tube was filled with mercury to add weight, and then sealed inside a Zircaloy 2 tube together with 90 mesh wet aluminum oxide grit as described above. After plugging both ends to prevent leakage of grit and water, the tube was rotated at a speed of 12 PM for M hours. Afterwards, rinse the tube with distilled water, and then
Surface oxidation was carried out for 24 hours at °C. The tube oxidized as described above was activated according to the procedure of Example 1 and then subjected to electroless plating of steel.

The Zircaloy 2 tubes thus obtained were not only similar in appearance to each other, but also similar to the Zircaloy 2 tubes in which the copper coating and the oxide film of the Zircaloy 2 tube were provided as described in Example 1. The foregoing examples are merely illustrative of a very large number of materials and procedures that may be used in the method of the present invention to provide a variety of useful fuel elements and nuclear fuel containment claddings. It goes without saying that a much wider range of materials and procedures may be used in the method of the present invention, as described prior to the examples above.

[Brief explanation of drawings]

FIG. 1 is a cut-away cross-sectional view of a fuel assembly including a fuel element made in accordance with the present invention, and FIG. 2 is an enlarged cross-sectional view of a fuel element made in accordance with the present invention. In the figure, 10 is a fuel assembly, 11 is a coolant channel, 1
4 is a fuel element, 16 is a nuclear fuel material, 17 is a composite cladding, 2
0 represents a void, 21 represents a zirconium substrate, 22 represents an oxidized layer of the zirconium substrate, 23 represents a metal coating, and 24 represents a nuclear fuel material holding means. Mitsuru Z. Sparrow. I

Claims (1)

[Scope of Claims] 1. A fuel element having a core material made of a nuclear fuel material and an elongated composite cladding container for the nuclear fuel material, in which the container includes a base body, a metal coating covering the surface of the base body in proximity to the nuclear fuel material, and a base material. It consists of a base oxide film interposed between a metal coating, and the base is zirconium or a zirconium alloy.
A fuel element characterized in that the metal coating is copper, nickel, iron or an alloy thereof. 2. The fuel element according to claim 1, wherein the oxide film consists essentially of zirconium dioxide. 3. A fuel element having a core material made of nuclear fuel material, a base, a metal coating covering the surface of the base close to the nuclear fuel material, and an elongated composite coated container consisting of an oxide film of the base interposed between the base and the metal coating. A manufacturing method comprising: (a) roughening or corroding the surface of a zirconium or zirconium alloy substrate 21, (b) oxidizing the thus treated substrate to form an oxide film, and (c) electroless The surface of the oxide film obtained in step (b) is activated so that a metal coating can be installed by plating, and then (d) the zirconium or zirconium alloy substrate obtained in step (c) is activated. A method that includes the steps of placing a metal film of steel, nickel, iron, or an alloy thereof on the surface of an oxide film. 4. The method according to claim 3, characterized in that a copper coating is provided on the surface of the activated oxide film by electroless plating. 5. The method according to claim 3 or 4, characterized in that the surface of the zirconium alloy is roughened by subjecting it to mechanical abrasion prior to oxidation. 6. Claims 3 to 5, characterized in that the surface of the zirconium alloy is corroded by treatment with a solution of ammonium hydrogen fluoride and sulfuric acid, and then the loosely attached skin is removed by sealing. The method according to any one of the above. 7. A method according to any one of claims 3 to 6, characterized in that the oxidation is achieved by heating the zirconium alloy to a high temperature in steam. 8. A method according to any one of claims 3 to 6, characterized in that the oxidation is achieved by heating the zirconium alloy to a high temperature in an oxygen atmosphere.