JPS6362716B2 - - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION BACKGROUND OF THE INVENTION The present invention relates generally to nuclear fuel elements for use in nuclear fission reactor cores, and more particularly to an improvement comprising a composite cladding having a metallic barrier of sponge zirconium bonded to the inner surface of a zirconium alloy tube. This invention relates to a method for producing type nuclear fuel elements.

Nuclear reactors as currently designed, manufactured, and operated contain nuclear fuel in fuel elements that can be of various geometries, such as plates, tubes, or rods. Nuclear fuel material is typically enclosed within a corrosion-resistant, non-reactive and thermally conductive container or cladding. Fuel elements are assembled in a lattice at regular intervals within coolant flow channels or regions to form fuel assemblies, which can be assembled in appropriate numbers to produce a self-sustaining nuclear fission reaction. form a nuclear fission chain reaction type assembly or core. This core is placed in a reactor vessel through which coolant flows.

Cladding (Cladding)
Vessels (also called vessels, cladding pipes, cladding tubes) are used for several purposes, the first of which is to
The aim is to prevent contact and chemical reactions between the nuclear fuel and the coolant or moderator (or the moderator if present, or both coolant and moderator if both are present). The second purpose is to prevent radioactive fission products, which are partially gaseous, from escaping from the fuel into the coolant or moderator, or both, if both are present. It is in. Common coating materials include stainless steel, aluminum and its alloys, zirconium and its alloys, niobium (columbium), and certain magnesium alloys. If coating failure or loss of leak sealing occurs, the coolant or moderator and its associated system may become contaminated with radioactive long-lived products to the extent that operation of the blunt is impaired.

Various problems arise in the manufacture and operation of nuclear fuel elements using certain metals and alloys as cladding materials due to mechanical or chemical reactions that occur in these cladding materials under certain conditions. Zirconium and its alloys are excellent nuclear fuel cladding materials under normal conditions. The reason is,
Zirconium and its alloys have a small neutron absorption cross section and, at temperatures below about 750°C, can be used in the presence of demineralized water or steam, which is commonly used as a reactor coolant and moderator. This is because it is strong, ductile, extremely stable, and non-reactive.

However, operation of fuel elements has revealed the problem of brittle cracks in the cladding due to complex interactions between the nuclear fuel, the cladding, and the fission products produced during the fission reaction. It has been determined that this undesirable behavior is facilitated by localized mechanical stresses in the cladding due to differential fuel-cladding expansion and friction. During operation of a nuclear reactor, fission products are generated within the nuclear fuel by a fission chain reaction and these fission products are released from the nuclear fuel and present on the cladding surface. In the presence of certain fission products, such as iodine and cadmium, these localized stresses and strains can cause failure of the double cladding through a phenomenon known as stress corrosion cracking or liquid metal embrittlement.

Within the sealed fuel element, hydrogen gas can be generated by the slow reaction of the cladding with residual water inside it;
This hydrogen gas builds up to a certain level, so that under certain conditions it can occur that the coating is locally hydrogenated and at the same time there is a local deterioration of the mechanical properties of the coating. Coatings are also adversely affected by gases such as oxygen, nitrogen, carbon monoxide and carbon dioxide over a wide range of temperatures.

The zirconium cladding of nuclear fuel elements is exposed to one or more of the gases and fission products mentioned above during irradiation in a nuclear reactor, which means that these gases are not present in the reactor coolant or moderator. This occurs despite the fact that they are removed as much as possible from the ambient atmosphere during the manufacture of the cladding and fuel elements. Sintered refractory ceramic compositions, such as uranium dioxide and other compositions used as nuclear fuels, release measurable amounts of the above gases when heated, e.g. during the manufacture of fuel elements; It also releases fission products during the fission chain reaction. Granular refractory ceramic compositions used as nuclear fuel, such as uranium dioxide powder and other powders, are known to release greater amounts of the gases during irradiation. These released gases can react with the zirconium cladding containing the nuclear fuel.

As is clear from the above, during the entire period of use of a fuel element in the operation of a nuclear power plant, water, water vapor and other gases, especially water, from inside the fuel element and reactive with the cladding. It has been found that it is desirable to minimize damage to the coating by hydrogen. One approach to achieving this goal has been to find materials that rapidly chemically react with water, water vapor, and other gases to remove them from the interior of the coating. Such materials are called getters.

Other strategies include coating the nuclear fuel material with various materials, such as coating the nuclear fuel material with ceramics to prevent moisture from coming into contact with the nuclear fuel material, as disclosed in U.S. Pat. No. 3,108,396. things are being done. The fuel element disclosed in U.S. Pat. No. 3,085,059 includes a metal casing containing one or more pellets of fissionable ceramic material and a layer of vitreous material bonded to the ceramic pellets, the layer being connected to the casing and the nuclear fuel. between the pellets and the casing to ensure uniform and good heat transfer from the pellets to the casing. In a jacketed fissionable uranium slug enclosed in a metal case described in US Pat. No. 2,873,238, the protective jacket is a zinc-aluminum bond layer. U.S. Patent No. 2,849,387 includes
A jacketed fissionable body is disclosed comprising a plurality of end-opening jacketed body portions of nuclear fuel, the body portions being bonded to a bonding material forming an effective thermally conductive bond between the uranium body portion and the cladding. immersed in a molten bath. The coating is described as any metal alloy with good thermal conductivity properties, examples include aluminum-silicon and zinc-aluminum alloys. Special Public Service 1977
No. 46559 (November 24, 1972), each individual nuclear fuel particle is combined into a carbon-containing matrix fuel composition by coating the nuclear fuel particles with a dense smooth carbon-containing coating around the pellet. It is disclosed that it hardens. Another coating is described in Japanese Patent Publication No. 47-14200, in which one of the two groups of pellets is coated with a layer of silicon carbide and the other group is coated with a coating of pyrolytic carbon or metal carbide. do.

Coatings of nuclear fuel materials have reliability problems in that it is difficult to obtain uniform coatings that are free of defects. Additionally, coating deterioration creates problems regarding the long-life operation of nuclear fuel materials.

A method for preventing corrosion of nuclear fuel cladding, disclosed in U.S. Patent Application No. 330,152 (filed February 6, 1973), consists of adding metals such as niobium to the fuel. The additive can be added in powder form, provided that the metal is not oxidized by subsequent fuel treatment operations, or it can be placed within the fuel element as a wire, sheet or other form, or around the fuel pellet. or between them.

Document GEAP-4555, dated February 1974, describes a composite coating of zirconium alloy with an inner lining of stainless steel metallurgically bonded to the zirconium alloy. This composite coating is manufactured by extruding a hollow billet of zirconium alloy with an inner lining of stainless steel. In this coating, the stainless steel develops a brittle phase, and the neutron absorption penalty of the stainless steel layer is the same as that of the zirconium alloy layer of the same thickness.
It has the disadvantage of being 10 to 15 times more expensive.

In a method for protecting zirconium and its alloys disclosed in US Pat. No. 3,502,549, chromium is electrodeposited to form a composite material useful in nuclear reactors.
There is a method of electrodepositing copper on the surface of Zircaloy-2 and then heat-treating it to cause the electrodeposited metal to diffuse.
“Energia New Clear”
Nucleare) Volume 11, No. 9 (September 1964), No.
It is described on pages 505-508. “Stability and compatibility of hydrogen barriers applied to zirconium alloys” by F. Brossa et al.
and Compatibility of Hydrogen Barriers
Applied to Zirconium Alloys, European Atomic Energy Community, Joint Nuclear Research Center
Atomic Energy Community Joint Nuclear
Research Center) EUR4098e (1969),
Deposition methods of various coatings and their efficiency as hydrogen diffusion barriers are described, and Al-Si
Coatings have been shown to be a very promising barrier to hydrogen diffusion. Electroplating on Zirconium and Zirconium by WCSchickner et al.
―Tin)” Technical Information
Service (Technical Information Service)
BMI-757 (1952), electroplated with nickel on zirconium and zirconium-tin alloy,
Methods of heat treating these alloys to form alloy diffusion bonds are described. U.S. Patent No. 3,625,821 describes a nuclear reactor fuel element having a fuel cladding tube, in which the inner surface of the fuel cladding tube is coated with a metal having a small neutron capture cross section, such as nickel, and a burnable poisonous material is coated on the inner surface of the fuel cladding tube. Place finely dispersed particles. “Reactor Development Program Progress Report”
Report) ANL-RDP-19 (August 1973) discloses a chemical getter arrangement in which a sacrificial layer of chromium is provided on the inner surface of a stainless steel coating.

There are several other patents that disclose coatings on the inner surface of the cladding of a fuel element. No. 3,145,150 claims a fuel element having a hollow sealed pressure vessel of metal hydride loosely supporting a core of fissile material and a thin corrosion-resistant jacket surrounding the pressure vessel. US Patent No.
The fuel element disclosed in No. 3053743 has a metal clad tube whose inner wall is coated with metallic nickel or a nickel-iron-chromium alloy, and which surrounds a core of nuclear fuel pellets and which There are spacers here and there in between. In the modified cross-section nuclear fuel element described in British Patent Specification No. 933500, individual fuel particles are coated on their surface with one or more substances, encapsulated in a covering member and reducing the cross-section of the element. has been put into a transformation process.

Other methods involve interposing a self-supporting barrier between the nuclear fuel material and the cladding that retains it. Such methods are described in US Pat. No. 3,230,151 (copper foil), German Patent Publication No. DAS 1238115 (titanium barrier), US Pat.
No. 3,018,238 (crystalline carbon barrier between UO ₂ and zirconium coating), and US Pat. No. 3,088,893 (stainless steel foil). Although the idea of providing a barrier is promising, some of the documents mentioned above include nuclear fuel or cladding or materials that are incompatible with nuclear fission reactions. For example, with respect to nuclear fuel, carbon may combine with oxygen from the nuclear fuel, and with respect to cladding, certain metals may react with the cladding, changing its properties, and, with respect to nuclear fission reactions, may act as a neutron absorber.

Other solutions along the barrier philosophy include U.S. Pat. No. 3,969,186, in which refractory metals such as molybdenum, tungsten, rhenium, niobium and their alloys are used in single or multi-layer tubes or foils between the cladding and the fuel. and U.S. Pat. No. 3,925,151 describes a method in which a foil of zirconium, niobium, or an alloy thereof is located between the nuclear fuel and the cladding, and a coating of highly lubricious material is provided on the liner or cladding. Each is listed.

OBJECTS OF THE INVENTION It is an object of the present invention to provide a nuclear fuel element that can be operated in a nuclear reactor for extended periods of time without cladding cracking or corrosion or other fuel defect problems.

Another object of the present invention is to use composite cladding (composite cladding, composite cladding, composite cladding, composite cladding, composite cladding, composite cladding, coated tube)
The purpose of the present invention is to provide a nuclear fuel element provided with a.

Another object of the present invention is to provide a nuclear fuel element having a composite cladding consisting of a metal barrier bonded to the inner surface of a zirconium alloy tube, said bond providing a long-life bond between the alloy tube and the metal barrier. It is to provide.

Yet another object of the present invention is to provide a nuclear fuel element having a composite cladding consisting of a metal barrier bonded to the inner surface of a zirconium alloy tube, the metal barrier comprising sponge zirconium.

Structure of the Invention A particularly effective nuclear fuel element for use in the core of a nuclear reactor has a composite cladding, which is a sponge metallurgically bonded to the inner surface of a zirconium alloy tube containing more than 5000 ppm by weight of components other than zirconium. Zirconium has a metal barrier of moderate purity, such as zirconium. The composite cladding surrounds the nuclear fuel material, leaving a gap between the nuclear fuel material and the composite cladding.
The metal barrier protects the alloy tube from nuclear fuel material trapped within the composite cladding and protects the alloy tube from fission products and gases. The metal barrier forms 5-30% of the thickness of the composite coating. Metal barriers thinner than 5% of the composite cladding thickness are insufficient to prevent microcracks from propagating into the zirconium alloy tube, and metal barriers thicker than 30% of the composite cladding thickness Adding weight to the composite coating does not add weight to the benefits and further reduces the overall strength of the composite coating. A liner barrier is
Contains more than 1000 ppm and less than 5000 ppm impurities by weight, but due to its purity it remains soft during irradiation and minimizes localized strains within the nuclear fuel element, thus preventing stress corrosion cracking or liquid metal embrittlement. Helps protect the alloy tube from. In design and function, the alloy tube section of the composite cladding is no different from previous nuclear reactor practice and is selected from conventional cladding materials such as zirconium alloys.

High purity, i.e. the amount of impurities by weight
Metallurgical bonding of a zirconium barrier of 1000 ppm or less to a zirconium alloy tube has been described, for example, in
It is known from No. 69795. However, high purity zirconium is expensive, difficult to obtain, and is sensitive to water attack. In contrast, sponge zirconium contains a moderate amount of impurities, is easily available at a low price, and is
zirconium is only slightly inferior to high-purity zirconium in terms of resistance to zirconium.

Furthermore, sponge zirconium has an advantage over high purity zirconium in that the resistance of sponge zirconium to water erosion is significantly greater, but this is neither suggested nor obvious from JP-A-51-69795.

The method of manufacturing a composite coating according to the present invention includes (1) inserting a hollow collar of a metal barrier into a hollow zirconium alloy billet, explosively bonding the hollow collar to the billet, and then extruding the composite; (2) inserting a hollow collar of the metal barrier into the hollow zirconium alloy billet of the substrate and heating under collar and billet compressive loads to diffusion bond the tube to the billet; Extrude the composite and then process the tube reduction, or
or (3) inserting a metal barrier collar into a hollow zirconium alloy billet and then extruding the composite before tube reduction.

The present invention provides several advantages that provide nuclear fuel elements with a long operating life. These benefits include reducing chemical interactions between the composite cladding and the nuclear fuel, minimizing localized stresses on the zirconium alloy tubing section of the composite cladding, and reducing stress corrosion on the zirconium alloy tubing section of the composite cladding. and to minimize strain corrosion, reducing the possibility of cracking failure in zirconium alloy tubes.

An important property of the composite coating of the present invention is that the various improvements described above can be achieved with little increase in neutron penalty. Such composite coatings can easily be used in nuclear reactors. This is because the composite coating does not form a eutectic mixture in the event of an accident such as loss of coolant or nuclear control rod fall. Additionally, the composite cladding has a very low heat transfer penalty since there is no thermal barrier to heat transfer as would occur with a separate foil or liner inserted into the fuel element.

The metal barrier does not introduce significant neutron capture penalties, heat transfer penalties, or material incompatibility problems to the fuel elements of the present invention. The composite coatings of the present invention can also be tested by conventional non-destructive measurement methods at various stages of manufacture and operation.

DESCRIPTION OF THE EMBODIMENTS In order that the invention may be better understood, it will be described below with reference to the drawings.

FIG. 1 shows a partially broken cross-sectional view of a nuclear fuel assembly 10. As shown in FIG. This fuel assembly 10 has a lifting handle 12 at the upper end and a nose member at the lower end.
(not shown since the lower part of the assembly 10 has been omitted) are each provided with tubular flow channels 11 of generally square cross section. The channel 11 has an opening 13 at its upper end and a coolant flow opening at its lower tip. An array of fuel elements or rods 14 is contained within the channel 11 and is supported therein by an upper end plate 15 and a lower end plate (not shown since the lower part of the assembly 10 has been omitted). Coolant typically enters through an opening in the lower end of the leading member and enters the fuel element 1
4 and from the upper outlet 13 in a partially evaporated state in the case of a boiling reactor.
Additionally, in the case of a pressurized nuclear reactor, the water exits without being evaporated and at a high temperature.

The ends of the nuclear fuel elements or rods 14 are sealed by end plugs 18 welded to the composite cladding 17, the plugs 18 being provided with studs 19 to facilitate attachment of the fuel elements to the assembly. I have to. A void space or cavity 20 is provided at one end of the fuel element to allow longitudinal expansion of the fuel material and to allow gases released from the fuel material to accumulate. A nuclear fuel material retention member 24 in the form of a spiral member is disposed within the cavity 20 to inhibit axial movement of the pellet column, particularly during handling and transportation of the fuel elements.

The fuel element is designed so that the composite cladding and fuel material have good thermal contact, minimize parasitic neutron absorption, and are resistant to warping and vibrations sometimes caused by high velocity coolant flow. .

A nuclear fuel element or rod 14 constructed in accordance with the present invention is shown in partial cutaway in FIG. The fuel element is a core or central cylindrical portion 16 of nuclear fuel material, shown in this example as a plurality of fuel pellets of fissionable and/or fuel parent material.
are arranged within the structural composite coating 17. In some cases, the fuel pellets can be of various shapes, such as cylindrical pellets or spheres, and in other cases, different fuel shapes can be used, such as granular fuel. The physical form of the fuel is not important to the invention. A variety of nuclear fuel materials can be used for nuclear fuel, including uranium compounds, plutonium compounds, thorium compounds, and mixtures thereof. A preferred fuel is uranium dioxide or a mixture of uranium dioxide and plutonium dioxide.

In FIG. 2, nuclear fuel material 16 forming the central core of fuel element 14 is surrounded by a composite cladding 17. In FIG. The composite cladding surrounds the core during use in a nuclear reactor, leaving a gap 23 between the core and the composite cladding. Composite coated zirconium alloy tube 2
1 is made of Zircaloy-2 in the preferred embodiment of the present invention. The zirconium alloy tube is metallurgically bonded on its inner surface to a metal barrier 22, which thus forms a shield between the zirconium alloy tube 21 and the nuclear fuel held within the composite cladding. The metal barrier constitutes 5-30% of the thickness of the composite coating and is formed from a low neutron absorbing metal, ie, zirconium of moderate purity (e.g. sponge zirconium). The metal barrier 22 protects the zirconium alloy portion of the composite cladding from contact and reaction with gases and fission products and prevents the development of localized stresses and strains.

The content of the metal barrier with adequate purity is an important factor and serves to impart special properties to the metal barrier. Generally, the impurities in the metal barrier material are at least greater than about 1000 ppm and about 5000 ppm by weight.
The amount should be less, preferably about 4200 ppm or less.
Among these impurities, oxygen is maintained in the range of about 200-1200 ppm. All other impurities are within the normal range of commercially available products. Nuclear reactor grade sponge zirconium is shown below.

Aluminum - 75ppm or less, Boron - 0.4ppm
Below, cadmium - 0.4ppm or less, carbon - 270ppm
Below, chromium - 200ppm or less, cobalt - 20ppm
Below, copper - 50ppm or less, hafnium - 100ppm or less, hydrogen - 25ppm or less, iron - 1500ppm or less, magnesium - 20ppm or less, manganese - 50ppm or less,
Molybdenum - 50ppm or less, Nickel - 70ppm or less, Niobium - 100ppm or less, Chitsune - 80ppm or less,
Silicon - 120ppm or less, Tin - 50ppm or less, Tungsten - 100ppm or less, Titanium - 50ppm or less,
Uranium - 3.5ppm or less.

The composite cladding of the nuclear fuel element of the present invention has a metal barrier bonded to the substrate, ie, the zirconium alloy tube, with a rigid bond. According to metallographic examination,
There is sufficient cross diffusion of the substrate material and metal barrier to form a bond, but this cross diffusion is not away from the bond area.

The inventors found that the sponge zirconium metal forming the metal barrier wall of the composite coating has high radiation hardening resistance, which allows the metal barrier to maintain its structure, such as yield strength and hardness, after long-term irradiation. It has been discovered that the optical properties can be suitably maintained at a significantly lower level than similarly irradiated common zirconium alloys.

Data obtained by the Applicant regarding the room temperature hardness of actual composite coatings with zirconium barriers shows that two purity levels of zirconium remain in the zircaloy, despite the hardening effect of high neutron irradiation, as shown in Figure 3. - It maintained its softness much more than 2. The irradiation temperatures in these experiments were within the range applied to nuclear fuel cladding.
The highest purity zirconium remains softer than less pure "low oxygen sponge" zirconium. However, low-oxygen sponge zirconium retains its softness much better than zircaloy.

In fact, metal barriers do not harden as much as common zirconium alloys when subjected to irradiation, and this, along with their low initial yield strength, results in plastic deformation and power transients.
during which pellet-induced stresses in the fuel element can be relieved. Pellet-induced stresses within the fuel element are affected by e.g. reactor operating temperatures (300-350
This can occur due to swelling of the fuel pellets at temperatures (°C) and contact of the pellets with the coating.

Additionally, we have determined that the sponge zirconium metal barrier preferably has a thickness of about 5 to 100% of the thickness of the composite coating.
around 15%, particularly preferably 10% of the thickness of the composite coating
They found that when bonded to a zirconium alloy tube as thick as , the stress is reduced and a barrier effect sufficient to prevent failure of the composite cladding is obtained. Zirconium alloys that serve as suitable alloy tubes include Zircaloy-2 and Zircaloy-4. Zircaloy
2 is about 1.5% Sn, 0.12% Fe, 0.09% Cr by weight
It contains 0.005% Ni and is widely used in water-cooled nuclear reactors. Zircaloy-4 is Zircaloy-
It has less nickel content and slightly more iron content than No.2.

The composite cladding used in the nuclear fuel element of the present invention can be manufactured by any of the methods described below.

In one method, a hollow collar of sponge zirconium selected as the metal barrier is inserted into a hollow billet of zirconium alloy selected as the alloy tube, and the assembly is then explosively bonded to join the collar to the billet. . Using conventional tube and shell extrusion techniques, this composite material is manufactured to approximately 1000-1400〓
Extrusion at high temperature (538~750℃). The extruded composite is then subjected to conventional tube reduction processing to obtain the desired dimensions of the composite sheath.

In another method, a hollow collar of sponge zirconium selected as the metal barrier is inserted into a hollow billet of zirconium alloy selected as the alloy tube, and the assembly is then subjected to a heat treatment [e.g. approximately 8 hours] to effect diffusion bonding between the collar and billet. The composite is then extruded using conventional tube shell extrusion techniques, and the extruded composite is subjected to conventional tube reduction processing to obtain the composite sheathing of the desired dimensions.

In yet another method, a hollow collar of sponge zirconium selected as the metal barrier is inserted into a hollow billet of zirconium alloy selected as the alloy tube, and the assembly is then pressed using conventional tube shell extrusion techniques. put out. The extruded composite is then subjected to conventional tube reduction processing to obtain the desired dimensions of the composite sheath.

The above-described method of producing the composite coating of the present invention is more economical than other methods used to produce the coating, such as electroplating or vapor deposition.

The method of manufacturing a nuclear fuel element of the present invention includes creating a composite cladding having a sponge zirconium metal barrier bonded to the inner surface of a zirconium alloy tube;
The composite cladding is open at one end, and a core of nuclear fuel material is filled into the composite cladding, leaving a void at the open end and a gap between the core and the composite cladding, and the nuclear fuel material is filled into the void. A closure member is provided at the open end of the composite cladding with the material retention device inserted and the cavity communicated with the nuclear fuel, and then the end of the composite cladding is bonded to the closure member to form a tight seal between them. form.

Next, examples of nuclear fuel elements manufactured and utilized for nuclear power plants according to the present invention will be shown.

EXAMPLE A plurality of nuclear fuel elements each having a composite coating of moderately pure (sponge) zirconium metallurgically bonded to the inner surface of a Zircaloy-2 alloy tube as a metal barrier were manufactured according to the following steps.

A hollow collar of low oxygen (less than 600 ppm) sponge zirconium was inserted into a hollow Zircaloy-2 billet and the assembly was extruded using conventional extrusion techniques to a 2.50 inch (6.35 cm) outer diameter and total wall thickness.
A 0.480 inch (1.22 cm) composite tube was manufactured.
Among them, the thickness of the sponge zirconium layer is 0.05
It was an inch (0.13cm).

Next, the composite tube was subjected to tube reduction treatment using the Pilger mill method, with an outer diameter of 0.483 inches (1.23 cm) and a total wall thickness of 0.483 inches (1.23 cm).
A 0.032 inch (0.08 cm) composite coating was produced.
The thickness of the composite coating sponge zirconium barrier is
It was 0.003 inch (0.008 cm).

Composite cladding has approximately 0.410 inch (1.04 cm) diameter
It was filled with sintered pellets of 95% theoretical density uranium dioxide (UO ₂ ). The composite cladding was then evacuated, filled with helium, and sealed using conventional fuel element manufacturing techniques.

The nuclear fuel elements thus produced were mechanically assembled into a nuclear fuel assembly suitable for boiling water reactor (BWR) applications. The characteristics of this nuclear fuel assembly are as follows.

Table Dimensions Shape 8 x 8 Nuclear Fuel Elements Pitch of Nuclear Fuel Elements 0.640 inches (1.63 cm) Water-to-Fuel Volume Ratio 2.55 Heat Transfer Area 94.33 square feet (8.76 m ² ) _UO2 Weight 199.3Kg U Weight 175.7Kg U-235 Enrichment 2.82% by weight (average) These nuclear fuel assemblies were loaded into a nuclear power plant (direct cycle BWR type, output 810 MWe). The operation progressed completely smoothly.

Next, since the nuclear fuel element according to the present invention has remarkable features and effects compared to those of the prior art, these will be explained.

PCI Characteristics A power ramp test designed to test resistance to PCI damage.
test), the performance of the nuclear fuel according to the present invention was tested. Data from this test are shown in FIG. The figure directly compares the performance of a nuclear fuel element with a zirconium barrier to that of a conventional Zircaloy-coated nuclear fuel element. Several lamp sequences were used in these tests. That is, the data indicated by C-lamp or C'-lamp is the ultra fast lamp rate (328 KW/m/min or more), and the other data is the lamp at 18 KW/m/min or less.

This data shows that nuclear fuel elements with high purity zirconium or low oxygen sponge zirconium metal barriers have superior PCI resistance compared to conventional ones. Conventional nuclear fuel elements with burn maps larger than 14MWd/KgU are 50KW/KgU.
Destroyed whenever ramped to a linear power level greater than m. However, it turns out that nuclear fuel elements with zirconium barriers (high purity or low oxygen sponges) are indeed robust to lamps up to this power level. However, purity also has an effect on PCI resistance. Nuclear fuel elements with a metal barrier of high-purity zirconium (at highest purity, less than 200 ppm oxygen) can withstand the highest speed power ramps to approximately 59 KW/m at high burn-ups. However, the hypoxic sponge zirconium barrier nuclear fuel element
16.4MWd/Kg, U burn-up, 56KW/
If you play Ultrafast Trump on m, it will be destroyed. Still, it is well above the 95% failure rate curve for lamp tests in conventional (barrier-free) nuclear fuel elements. Additionally, it has a high oxygen content (approximately
The performance of nuclear fuel elements with a sponge zirconium metal barrier (1000 ppm) was also very good;
It wasn't as bad as the hypoxic one. Thus, a practical range of purity was established.

Deterioration of composite cladding after water (or steam) incorporation Nuclear fuel elements with zirconium barriers are designed such that the zircaloy is in contact with the coolant and moderator water and the zirconium barrier is in contact with the nuclear fuel body. Zircaloy's oxidation resistance in hot water (or steam) is greater than zirconium. Although the composite cladding is designed to prevent direct contact between the zirconium barrier and water, defects in the composite cladding can allow water (or steam) to enter the nuclear fuel element and react with the metal barrier. There is a possibility that this may occur. The applicant has conducted a simulation test under this condition, and FIG. 5 shows data regarding the unirradiated composite coating. The oxidation rate at 350°C (a reasonable estimate of the temperature of the zirconium barrier during operation) is significantly lower for hypoxic sponge zirconium than for high-purity zirconium (83.3 mg/d for the former).
m ² /day, whereas the latter is 62.5 mg/dm ² /day). According to these data, if a nuclear fuel element with a hypoxic sponge zirconium barrier is defective,
It is assumed that the rate of degradation will be slower than if a nuclear fuel element with a barrier of high purity zirconium were similarly defective.

The above was confirmed in a nuclear fuel irradiation experiment as follows. Defects (holes) were created in the composite cladding of a nuclear fuel element having a composite cladding with a zirconium barrier to allow water (steam) to enter. Such nuclear fuel elements are used in test reactors to produce approximately 3 MWd/
KgU and then tested. All of the defective nuclear fuel elements (conventional and with zirconium barriers) survived these tests. However, it has been found that the zirconium barrier of nuclear fuel elements with high purity zirconium barrier is severely oxidized and hydride (zirconium hydride) is formed within the composite coating. The hydrogen in this hydride is a reaction product when zirconium is oxidized by water. The reaction is as follows.

At pellet surface: 600℃ UO ₂ +H ₂ O→UO _2+x +XH ₂ +(1-X)H ₂ O Within barrier: 350℃ Zr+2H ₂ O→ZrO ₂ +2H ₂ Zr+H ₂ →Zr+ 2H Zr+ 2H →ZrH ₂ However, nuclear fuel elements with low-oxygen sponge barriers had significantly less oxidation of the barrier and significantly less hydride within the barrier. At the peak power of the nuclear fuel element, the high purity zirconium barrier was actually completely oxidized. The resulting hydrogen was deposited as a layer of zirconium hydride on the outer surface of the composite coating (as shown by the dark area at the top of the photograph in Figure 6). Similar results for a nuclear fuel element with a hypoxic sponge zirconium barrier show that the barrier was only partially oxidized, with no concentration of hydrogenation on the outer surface of the composite cladding, as shown in Figure 7. was not observed, and the accumulation of hydrides within the composite coating was extremely small.

Pricing and Availability In the United States, high purity zirconium is available from only one supplier (Teledyne Wah Chang), while low oxygen sponge zirconium is available from several suppliers. According to recent comparative quotes, the price of high-purity zirconium is lower than that of low-oxygen sponge zirconium.
It will be more than 3.5 times. This price quote is for a zirconium collar prior to insertion into a hollow zircaloy biret, which is then extruded to form a composite tube with a zirconium barrier. In practice, this composite tube is subjected to a tube reduction process to produce a composite cladding with a zirconium barrier.
This has already been said.

According to the applicant's review, current global production capacity for high purity zirconium is not sufficient to meet the current demands of the nuclear energy market. of course,
It is ultimately possible to increase production capacity sufficiently through investment. However, under current supply and market conditions, the supply capacity of the zirconium barrier market will depend on the present invention utilizing low oxygen sponge zirconium.

In Japan, sponge zirconium is available on the market, but its market price is highly volatile. On the other hand, high purity zirconium is not available on the market.

As will be apparent to those skilled in the art, various modifications and improvements may be made within the scope of the present inventors.

[Brief explanation of drawings]

FIG. 1 is a partially cutaway sectional view showing a nuclear fuel assembly having a nuclear fuel element constructed according to the present invention, FIG. 2 is a cross-sectional view showing an enlarged view of the nuclear fuel element of FIG. 1, and FIG. and Zircaloy-2
Figure 4 is a diagram showing the relationship between neutron irradiation dose and hardness, Figure 4 is a diagram showing the lamp test results of a nuclear fuel element with a zirconium barrier, Figure 5 is a diagram showing the relationship between weight increase due to corrosion of a composite coating and number of days, Figures 6a and 6b are micrographs showing the structure of the cladding, metal barrier and part of the nuclear fuel pellet, showing the oxidation of the high purity zirconium barrier at the peak power position (Figure 6a, as polished, 50X) and the hydrogen compounds in the cladding. Distribution (Figure 6b after corrosion, 100X), Figure 7a, 7b
The figures are micrographs showing the structure of the cladding, metal barrier and part of the nuclear fuel pellet, including the oxidation of the sponge dillium barrier at the pea power position (Figure 6a, as polished, 50X) and the hydride distribution of the cladding (Figure 7).
Figure b shows 100X after corrosion. 10... Nuclear fuel assembly, 11... Channel, 1
4... Nuclear fuel element, 16... Nuclear fuel, 17... Coating, 18... End plug (plug member), 20... Cavity, 21... Tube, 22... Metal barrier, 24... Holding device, a... _ZnO2 , b... _UO2 , c...zirconium hydride.

Claims (1)

[Claims] 1 Hollow zirconium collar containing impurities of more than 1000 ppm and less than 5000 ppm by weight
A nuclear fuel characterized by being inserted into a hollow billet of zirconium alloy containing impurities of 5000 ppm or more, extruding this assembly together by tube extrusion to make a composite tube, and performing tube reduction processing on this composite tube. Manufacturing method of zirconium composite coated container for elements.