JPS6313160B2 - - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION The present invention relates to thermal and mechanical treatment of composite cladding for fuel elements.

Nuclear reactors using water as a coolant and moderator are known and are available from e.g.
MMEl-Wakil's ``Nuclear Power Engineering''
Power Engineering)” McGraw-Hill, 1962
Discussed in 2010).

The fuel element for such reactors typically consists of uranium oxide and/or plutonium oxide pellets contained within an elongated protective cladding made of a suitable metal such as a zirconium alloy (e.g. Zircaloy-2). It is customary that
Such fuel elements are described, for example, in U.S. Pat.
No. 3365371.

In order to prevent premature failure of the fuel element cladding and extend its mechanical useful life, it has been proposed to provide various protective barriers between the columnar array of fuel pellets and the inner surface of the cladding. There is. Some of these partition walls have a layer of metallic zirconium bonded to the inner surface of a zirconium alloy cladding tube.

Belgian Patent No. 835,481 describes the case where the partition, which is joined to the inner surface of the cladding tube, consists of a layer of almost pure metallic zirconium.

Belgian Patent No. 870,342 also describes the case where such a partition consists of a layer of metallic zirconium of low purity, such as sponge zirconium.

According to the conventional method of manufacturing cladding tubes with internally bonded partitions, a hollow billet of zirconium alloy is fitted with a metal zirconium sleeve for the partition, followed by coextrusion of the composite. The diameter of the composite is then reduced to the desired final value by cold working in multiple passes in equipment such as a Pilger tube squeezer.

After each pass, the composite is typically annealed by heat treatment using a temperature and time sufficient to substantially completely recrystallize the zirconium alloy.

However, the annealing temperatures and times required for complete recrystallization of the zirconium alloy have been found to be undesirable due to grain growth in the metallic zirconium partitions.

According to one feature of the present invention, the heat treatment of the composite material after the final drawing step results in substantially complete recrystallization of the metallic zirconium layer and imparts a fine microscopic structure therein. At the same time, temperatures and times are used that allow stress relief of the zirconium alloy but do not result in its complete recrystallization.

According to another feature of the invention, the crystal structure of the metallic zirconium layer can be improved without deforming the zirconium alloy of the composite tube by applying a compressive deformation to the surface of the zirconium layer of the composite, for example using shot peening techniques. can be improved at will.

The invention will be explained in more detail below with reference to the accompanying drawings.

A fuel element 11 for a nuclear reactor as shown in FIGS. 1 and 2 consists of fuel pellets 13 arranged in columns.
and an elongated composite cladding tube 12 housing it, the ends of which are sealed by a lower end plug 14 and an upper end plug 16.

A plenum space 17 is provided to allow longitudinal expansion of the fuel and to provide room for gases released from the fuel during operation within the reactor. Top fuel pellet and top end plug 16
A spring 18 is disposed between the fuel pellets and the fuel pellets to hold the fuel pellets in place. As best seen in FIG. 2, the dimensions of the composite cladding tube 12 relative to the diameter of the fuel pellets are:
It is determined that an annular gap 19 is obtained between the fuel pellets and the inner surface of the cladding tube.

In accordance with one preferred embodiment of the present invention, composite cladding 12 includes a zirconium alloy cladding 21 and a metallic zirconium partition 22 metallurgically bonded to its inner surface.

Zirconium alloys suitable for the cladding tube 21 include Zircaloy-2 and Zircaloy-4.
Zircaloy-2 is approximately 1.5% (by weight) tin, 0.12
(wt)% iron, 0.09 (wt)% chromium, 0.005
(by weight)% nickel and the balance zirconium. Zircaloy-4 has a lower nickel content than Zircaloy-2, but a slightly higher iron content. In either case, such alloys contain more than 5000 ppm of components other than zirconium.

The partition wall 22, which accounts for about 1% to about 30% of the thickness of the composite cladding 12, is comprised of metallic zirconium with a limited impurity content. That is, such materials can range from substantially pure (ie, highly pure) zirconium metal having an impurity content of less than 500 ppm up to zirconium metal having an impurity content of up to 5000 ppm and preferably less than about 4200 ppm.

Among impurities, oxygen should be kept as low as possible, with its content ranging from less than 200 ppm to a maximum of about
Must be kept within a range of up to 1200ppm.
Other impurities may be within the standard range for commercially available nuclear reactor sponge zirconium. If we list them, aluminum is 75ppm or less, boron is
0.4ppm or less, cadmium 0.5ppm or less, carbon
270ppm or less, chromium 200ppm or less, cobalt
20ppm or less, copper 50ppm or less, hafnium
100ppm or less, hydrogen 25ppm or less, iron 1500ppm
Below, magnesium is 20ppm or less, manganese is
50ppm or less, molybdenum 50ppm or less, nickel 70ppm or less, niobium 100ppm or less, nitrogen
80ppm or less, silicon 120ppm or less, tin
less than 50ppm, and uranium less than 3.5ppm.

The metallic zirconium partition wall 22 is metallurgically joined to the zirconium alloy cladding tube 22. Specifically, although there is sufficient interdiffusion between the two to create a strong bond, it is unlikely that the barrier 22 that is outside approximately 1/2 to 1 mil from the interface will be contaminated by diffusion. Make it not exist.

If there is a partition wall 22 that occupies about 5 to 15% (particularly preferably 10%) of the thickness of the composite cladding tube 12,
It has been found that exposure of the zirconium alloy cladding 21 to corrosive fission products is prevented.

The partition 22 also prevents the zirconium alloy cladding 21 from having direct mechanical interaction with the fuel pellets, thus reducing the stresses that may result therefrom. It has been found that the desired structural properties (eg, yield strength and hardness) of the septum 22 are maintained at significantly lower levels than with conventional zirconium alloys. In fact, upon irradiation, the metallic zirconium partition walls 22 do not harden as much as normal zirconium alloys. Therefore, considering that the initial load-bearing strength is small, the partition wall 22
exhibits plastic deformation during power transients, thus making it possible to remove pellet-induced stresses in the fuel element. Pellet-induced stresses in the fuel element include those caused by the expansion of the fuel pellets into contact with the cladding under reactor operating temperatures, for example.

The composite cladding tube of the present invention can be manufactured by any of the following methods.

According to the first method, inside a hollow billet made of zirconium alloy selected for the cladding tube,
A hollow tube made of metallic zirconium, selected for the bulkhead, is inserted. The hollow tubes are joined to the hollow billet by subjecting the assembly to an explosive joining process. The composite material thus obtained is processed according to conventional tube extrusion techniques to a thickness of about 1000 to about 1400
Extruded at high temperatures (538 to about 760°C). Then,
A composite cladding tube of a desired size is obtained by subjecting the extruded composite material to processing operations including a normal tube drawing step.

According to the second method, a hollow billet made of metallic zirconium selected for the partition wall is inserted into a hollow billet made of zirconium alloy selected for the cladding tube. For such a collection, for example
Diffusion bonding between the hollow tube and the hollow billet is achieved by carrying out a heating step at 1400° C. (760° C.) for about 8 hours. The composite thus obtained is extruded according to conventional tube extrusion techniques. Next, by subjecting the extruded composite material to processing operations including a normal tube drawing step, a composite cladding tube of desired dimensions is obtained.

According to the third method, a hollow billet made of metallic zirconium selected for the partition wall is inserted into a hollow billet made of zirconium alloy selected for the cladding tube. Such a mass is extruded according to conventional tube extrusion techniques. Next, by subjecting the extruded composite material to processing operations including a normal tube drawing step, a composite cladding tube of desired dimensions is obtained.

The dimensions of the starting material are determined by the ratio of the cross-sectional areas of the septum and cladding body in the desired composite cladding product. For example, the total cross-sectional area of the final product is given by:

A _TF = π/4 (OD _TF ² − ID _TF ² ) where A _TF is the total cross-sectional area of the final product, OD _TF is the outer diameter of the final product, and ID _TF is the inner diameter of the final product. The desired septum cross-sectional area is given by:

A _BF =π/4 (OD _BF ² − ID _BF ² ) where A _BF is the cross-sectional area of the partition wall, OD _BF is the outer diameter of the partition wall, and ID _BF is the inner diameter of the partition wall. The total cross-sectional area of the original hollow billet including the bulkhead is given by:

A _TI = π/4 (OD _TI ² − ID _TI ² ) where A _TI is the total cross-sectional area of the original hollow billet including the partition wall, OD _TI is the outer diameter of the original hollow billet, and ID _TI is the outer diameter of the hollow billet. It is the inner diameter. In that case, the required initial cross-sectional area of the bulkhead A _BI is given by:

A _BI =A _TI (A _BF /A _TF ) Example An example of the method for manufacturing the composite cladding tube 12 based on the present invention is shown below.

The billet for the zirconium alloy cladding and the insert for the metallic zirconium septum are machined, cleaned and assembled in a conventional manner. It should be noted that these dimensions are selected to be compatible with extrusion of the assembly in a hot extrusion press. The billet for the cladding is standard Zircaloy-2 that meets ASTM B 3533, grade RA-1.
and the insert for the septum is made of metallic zirconium with an impurity content within the aforementioned range. The bores of the billet and insert are molded with a taper of 8 mils per inch (203 microns) so that their mating under pressure ensures good contact between the mating surfaces.

Examples of dimensions of such machined parts are as follows. In other words, the billet for cladding is 9.0 inches (22.9 cm) long and the outside diameter is 9.0 inches (22.9 cm).
5.74 inches (14.6 cm) and has an inner diameter of 2.44 inches (6.20 cm). The bulkhead insert has an outer diameter of 2.44 inches and an inner diameter of 1.66 inches (4.22 cm).

Prior to assembly, trace impurities are removed by lightly etching the billet and insert contact surfaces. A suitable caustic agent for such purpose is a solution consisting of 70 ml H ₂ O, 30 ml HNO ₃ (70% aqueous solution) and 5 ml HF (48% aqueous solution).

To ensure a satisfactory bond during extrusion, the assembly may be pre-bonded by pressing a tapered insert into the tapered bore of the billet in a vacuum below 20 μm of mercury. In that case, 30 to 45,000 pounds (13.7
Approximately 1400〓 while applying an initial pressure of ~20400Kg)
(760°C) for 8 hours. It has been found that the result is a bond over 20-25% of the contact area.

To reduce end loss during extrusion, a 2-inch long piece of Zircaloy-2 tubing was welded to each end of the pre-bonded assembly and machined to a flat surface. good.

When extruding such a pre-bonded aggregate into a composite material, the following parameters may be used. That is, the extrusion speed is 6 inches/
The working ratio is 6:1, the temperature is 1100㎜, and the extrusion load is 3500 tons.

All billet surfaces, except the bore and the floating mandrel, may be lubricated with a water-soluble lubricant that has been heat treated at 1300°C (704°C) for 1 hour. After extrusion, additional tubing is cut from each end of the composite and the interior surface is honed to remove blemishes and improve the finish.

In order to obtain a cladding tube of suitable dimensions for a fuel element by drawing such a composite material, cold working is performed in three passes in a known Pilger tube drawing machine, and heat treatment and cleaning operations are performed between successive passes. That's fine. The steps of a typical operating procedure are shown in FIG.

The operating procedure is conventional except for the changes made in accordance with the present invention. The rationale for such changes and the beneficial results obtained therefrom will be discussed below.

The intense cold working performed after the tube drawing process distorts the shape of the metal microcrystals and creates a large number of lattice defects inside the microcrystals. That is, the metal after cold working is in a relatively high energy state, which is thermally unstable. The annealing process uses heat to impart mobility to the atoms of the metal, thereby serving to rearrange such atoms into a lower energy state. The parameters that affect such annealing are temperature and time, with temperature being the more sensitive. in general,
The annealing temperature and time are selected to be sufficient to provide substantially complete recrystallization but insufficient to cause excessive grain growth.

That is, with respect to annealing steps 5 and 8 in the operating sequence of FIG. 3, temperatures and times are selected to provide substantially complete recrystallization of the zirconium alloy of cladding tube 21.

However, because the relatively pure metal that makes up partition wall 22 recrystallizes at lower temperatures, typical annealing temperatures and times suitable for zirconium alloys, such as those used in steps 5 and 8, do not result in a finished product. A particularly undesirable degree of grain growth will be observed in the partition walls 22.

Therefore, according to one feature of the invention, step 1 is applied to the cladding obtained after the final tube drawing step.
As shown in 2, heat treatment is performed at a low temperature.

In that case, the temperature and time of the heat treatment in step 12 are selected so that the metallic zirconium constituting the partition wall 22 undergoes substantially complete recrystallization, but no crystal grain growth occurs. The partition walls thus obtained have a fine equiaxed crystal structure and therefore exhibit improved strength and ductility, increased resistance to stress corrosion cracking, and a high degree of plastic stabilization.

The temperature and time of the heat treatment step 12 are also selected to allow complete stress relief of the zirconium alloy of the cladding tube 21 but not complete recrystallization thereof. As a result, zirconium alloys retain the elongated grain structure created by the tube drawing operation and have the advantage of exhibiting increased strength under high strain rates along with internal stress relief.

Suitable conditions for annealing steps 2, 5 and 8 include temperatures of about 1000-1300°C (538-704°C) for about 1-15 hours, preferably about 1-4 hours.

In addition, conditions suitable for heat treatment step 12 include approximately
Examples include conditions for about 1 to 4 hours at 825 to 950°C (440 to 510°C).

According to another feature of the invention, the crystal structure (ie the degree of preferred crystal orientation) of the metallic zirconium partition walls can optionally be improved by applying mechanical compressive deformation to the surface thereof. For example, by subjecting the partition wall to shot peening from the inside of the composite material prior to the heat treatment step 12, compressive deformation of the partition wall can be achieved without substantially deforming the zirconium alloy cladding tube.

Such mechanical treatment, shown as step 10 in FIG. 3, results in an improved crystal structure in which the bottom poles {0002} are highly aligned along the radial direction of the composite cladding. It turns out.

[Brief explanation of the drawing]

FIG. 1 is a partially sectional elevational view of a fuel element, FIG. 2 is a cross-sectional view of the fuel element of FIG. 1, and FIG. 3 is a typical operating procedure for tube throttling and treatment in accordance with the present invention. This is a diagram. In the figure, 11 is a fuel element, 12 is a composite cladding tube, 1
3 is a fuel pellet, 14 is a lower end plug, 16 is an upper end plug, 17 is a plenum space, 18 is a spring,
21 represents a cladding tube, and 22 represents a partition wall.

Claims (1)

[Scope of Claims] 1. Made of a zirconium alloy tube containing more than about 5000 ppm of components other than zirconium, and a cladding tube material metallurgically bonded to the inner surface thereof and consisting of a commercially available nuclear reactor sponge zirconium layer, In a method of manufacturing an elongated composite cladding tube useful for containing nuclear fuel and forming a fuel element for a nuclear reactor,
(1) Through cold working in a series of tube drawing processes,
reducing the dimensions of the cladding to a desired inner diameter and wall thickness; (2) between successive tube drawing steps, a temperature and time sufficient to effect substantially complete recrystallization of the zirconium alloy; (3) After the final tube drawing step, substantially complete recrystallization of the cancellous zirconium layer is carried out to impart a fine microscopic structure to the interior thereof, and at the same time, A method comprising heat treating the cladding using a temperature and time so low as to allow stress relief of the zirconium alloy but not complete recrystallization thereof. 2. The method of claim 1, wherein the sponge zirconium contains up to 5000 ppm of impurities. 3. A method according to claim 1 or 2, wherein the sponge zirconium contains up to 1200 ppm oxygen. 4 The temperature and time of step (2) are approximately
538 to about 704°C and about 1 to about 15 hours, and the temperature and time of step (3) are about 440 to about 440°C, respectively.
5. The method of claim 1, wherein the temperature is about 510<0>C and about 1 to about 4 hours. 5. Claim 1 or 4, further comprising the step of applying substantially uniform compressive deformation to the surface of the sponge zirconium layer prior to step (3).
The method described in section. 6. The method of claim 5, wherein the compressive deformation is applied by shot peening. 7 Composed of a zirconium alloy tube containing more than about 5000 ppm of components other than zirconium and a commercially available nuclear reactor sponge zirconium layer metallurgically bonded to its inner surface, the tube contains nuclear fuel and is used as a fuel element for a nuclear reactor. The zirconium alloy tube undergoes substantially complete recrystallization of the cancellous zirconium layer to exhibit a fine microscopic structure in the elongated composite cladding tube that serves to construct the zirconium alloy tube, and the zirconium alloy tube exhibits substantially complete stress relief. Composite cladding shown but not completely recrystallized. 8. The method of claim 7, wherein the sponge zirconium contains up to 5000 ppm of impurities. 9. A method according to claim 7 or 8, wherein the sponge zirconium contains up to 1200 ppm oxygen. 10. The method according to claim 7, wherein compressive deformation is applied to the surface of the metal zirconium layer. 11 Claim 10, wherein the compressive deformation is applied by shot peening.
The method described in section.