Zirconium alloy and components for the core of light water-cooled nuclear reactors

Abstract

A zirconium alloy has the following composition in percent by mass: Sn: 0.2-0.5%; Nb: 0.2-0.8%; Fe: 0.05-0.40%; V: 0-0.20%; 0:0.12-0.20%; Si: 80-120 ppm; C:=120 ppm; and a remainder of reactor-pure zirconium and acceptable impurities. The alloy is particularly suitable for components for the core of light water reactors, in particular, for pressurized water reactors.

Description

CROSS-REFERENCE TO RELATED APPLICATION

This is a continuing application, under 35 U.S.C. § 120, of copending international application PCT/EP2004/007822, filed Jul. 15, 2004, which designated the United States; this application also claims the priority, under 35 U.S.C. § 119, of German patent application No. 103 32 239.6, filed Jul. 16, 2003; the prior applications are herewith incorporated by reference in their entirety.

BACKGROUND OF THE INVENTION

Field of the Invention

The invention lies in the nuclear reactor technology field. More specifically, the invention relates to a zirconium alloy, or zirconium-based alloy, and to structural parts made from an alloy of this type for the core of light-water-cooled nuclear reactors, in particular of pressurized-water reactors. Structural parts of this type are in particular fuel cladding tubes, spacers and control rod guide tubes.

For physical reasons, zirconium, which has a low neutron absorption, is used as base metal for the structural parts in reactor cores. On account of the separation of the neutron absorber hafnium, it is customary to use reactor-pure zirconium sponge, the composition of which is governed by standards.

Zircaloy-2 (for boiling-water reactors) and Zircaloy-4 (for pressurized-water reactors) or other zirconium-based alloys, for example those known from U.S. Pat. Nos. 5,940,464 and 4,938,920 (corresp. German published patent application DE 38 05 124 A1); U.S. Pat. Nos. 5,230,758 and 5,112,573 (cf. German DE 690 10 115 T2), and from international PCT publication WO 01/24193 A1, are nowadays generally used for the above-mentioned purpose. Binary Zr—Nb alloys are also used to a lesser extent.

The following table gives the compositions of zirconium sponge and the standardized alloys which have hitherto been customary in Western engineering. In this context, it should be mentioned that nowadays some of the permitted impurities can be set in a particularly controlled way or even, by using suitable additions, set to specific values. By way of example, on account of its hardening action on zirconium, oxygen was originally controlled only to levels corresponding to manufacturing requirements, but nowadays it is actually used deliberately as a hardening addition.

Table

Reactor-pure zirconium (maximum contents in ppm):

Al	B	Cd	C	Cl	H	Hf	Fe	O	Si
75	0.5	0.5	250	1300	25	100	1500	1600	20

Composition of Zircaloy and ZrNb alloys (in % by mass):

					Other
	Sn	Fe	Cr	Ni	stipulations
Zircaloy-2:	1.2-1.7	0.07-0.20	0.05-0.15	0.03-0.08	0.18-0.36
					FeCrNi
Zircaloy-4:	1.2-1.7	0.18-0.24	0.07-0.13	≦0.007	0.28-0.37
					FeCr
Zr-2,	≦0.05	≦0.150	≦0.02	≦0.007	2.40-2.80%
5% Nb:					Nb

SUMMARY OF THE INVENTION

It is accordingly an object of the invention to provide a zirconium alloy, which is further improved relative to the above-mentioned materials and structural parts of the general type and which, in particular, provides for further improved zircaloy structural parts for light water reactor cores.

With the foregoing and other objects in view there is provided, in accordance with the invention, a zirconium alloy, comprising, in percent by mass:

• • Sn: 0.2-0.5%
  • Nb: 0.2-0.5%
  • Fe: 0.05-0.40%
  • V: 0-0.20%
  • O: 0.12-0.20%
  • Si: 80-120 ppm
  • C: ≦120 ppm
    and a remainder of reactor-pure zirconium together with standard impurities.

In accordance with an added feature of the invention, there is provided an amount of vanadium with a content of at least 0.07%.

In accordance with an additional feature of the invention, a sum of the contents of Sn, Nb, Fe, and V is at most 1.3%.

In other words, the novel alloy is composed of a matrix of reactor-pure zirconium together with 0.2 to 0.5% of Sn, 0.2 to 0.5% of Nb, 0.05 to 0.40% of Fe and 0 to 0.20% of V, with the carbon content being restricted to at most 120 ppm, and a range of from 80 to 120 ppm for Si and from 0.12 to 0.20% for O being maintained. It has been found that alloys of this type can be used to produce components, such as cladding tubes, spacers, guide tubes and further structural elements of a fuel assembly, for the core of light water reactors, in particular of pressurized water reactors. The components have an improved resistance to corrosion compared to components made from Zircaloy-4, while maintaining substantially the same production and the same heat treatment. This property is particularly pronounced if the sum of the alloying constituents Sn, Nb, Fe and V must drop if the total amount of Sn and Nb increases.

Values higher than 0.5% of Sn, for example up to 0.75%, have an adverse effect on the corrosion resistance, increase the radiation-induced growth, while the mechanical properties are significantly improved, which means that the proposed value of at most 0.5% represents a good compromise. The minimum Sn content at which components with good mechanical properties can still be produced is 0.2%.

Vanadium is an addition which is not absolutely imperative with a view to improving the corrosion properties. For example, it is possible to increase the corrosion resistance with a high burn-up using Sn contents of 0.4% to 0.5%. However, if some of the iron is replaced by V or if V is added to the alloy in small quantities (0.02 to 0.20%), the hydrogen pickup factor (HPUF) and therefore the formation of hydrides, which in addition to embrittling the material also cause material growth, is reduced.

To achieve an optimum creep rupture strength and at the same time a yield strength with a high value, it is possible to add Nb to the alloy in an amount of up to 0.8%, preferably up to its solubility limit, i.e. up to 0.5%. If this limit is not significantly exceeded, there is no risk of uncontrolled phase transitions, which result at relatively high temperatures, e.g. when welding spacers or cladding tubes to their end stoppers, on account of the complicated phase diagrams of ZrNb. Consequently, it should not be necessary for the alloy according to the invention to be subjected to a further heat treatment following welding.

Furthermore, the alloys are relatively insensitive to the effects of high heating surface stresses and local boiling processes at the interface with water. In this context, in particular a low uptake of lithium and a low level of nodular corrosion—as is found with cladding tubes made from Zircaloy-4 under standard pressurized-water conditions—are observed. Moreover, they have a low radiation-induced growth.

With the above and other objects in view there is also provided, in accordance with the invention, a component for the core of a light water reactor, in particular, a pressurized-water reactor, the is formed of the above-outlined zirconium-based alloy.

In a preferred embodiment of the invention, the component is produced while maintaining a cumulative annealing parameter of (10-40) E-18 h.

Other features which are considered as characteristic for the invention are set forth in the appended claims.

Although the invention is illustrated and described herein as embodied in a zirconium alloy and components for the core of light-water-cooled nuclear reactors, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims.

The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments of the invention.

BRIEF DESCRIPTION OF THE DRAWING

The sole FIGURE of the drawing is a chart in which the thickness of a resultant oxide layer is plotted over burn-up.

DESCRIPTION OF THE PREFERRED EMBODIMENTS

The following table illustrates four exemplary embodiments of the invention:

	Sn(%)	Nb(%)	Fe(%)	V(%)	O(%)	Si(ppm)	C(ppm)
A	0.30	0.25	0.35	0.16	0.14	110	100
B	0.30	0.45	0.15	0.10	0.14	110	100
C	0.40	0.45	0.10	0.07	0.14	110	100
D	0.30	0.75	0.13	0.07	0.14	110	100

Remainder: reactor-pure unalloyed zirconium in each, with permitted foreign substances or impurities.

To produce cladding tubes, ingots of the alloys A to D are melted in vacuo in a number of melting steps and then forged in the β-range of the alloys below the melting point. The forgings are heated again to a temperature in the β-range and then quenched in a water bath with a cooling rate of at least 30 K/s. The forgings are then forged to form rods.

The forged rods are machined and cut into pieces which are used to extrude tubes. To obtain a fully recrystalized microstructure, an anneal is carried out after the extrusion. The tubes treated in this way are pilgered in a number of steps by cold-forming to form cladding tubes. Prior to each deformation operation, an intermediate anneal is carried out in vacuo at temperatures of approximately 700° C., which brings about recovery and recrystalization. The final deformation, which leads to the definitive cross section of the cladding tube, is followed by a final anneal at approximately 600° C. In this way, a low creep deformation with a high yield strength is set for the intended reactor use. A cumulative annealing parameter in the range A=10-40 E-18 h is maintained during production. It is in addition optionally possible to carry out an anneal in the alpha range following production of the forged rods.

The cladding tubes which have been produced in the manner outlined are finally filled with fuel pellets and welded to end stoppers in a gastight manner at both ends. This concludes the production of the fuel rods. Control rod guide tubes are also produced by the same process.

In another exemplary embodiment, following corresponding heating and quenching of an ingot of the same composition, the forging is hot-rolled (once or in a number of steps with anneals between them) to form plates. For the hot-forming and intermediate annealing steps, the temperatures are selected in such a way that they are in the α-range of the alloys. Then, the plate is cold-rolled in a number of steps to form a metal sheet of the desired thickness. Between the deformation steps and following the final deformation, a vacuum anneal is carried out, which can also take place as a continuous process and brings about complete recrystalization. These metal sheets are processed further to form spacers.

If spacers, guide tubes and fuel rods produced in this way are used in a pressurized-water reactor, these components have better corrosion properties, in particular after a prolonged operating period, compared to components made from conventional Zircaloy-4 with a low tin content (low tin Zirc-4), as can be established from empirical calculations. The results of these calculations are disclosed in the diagram of the FIGURE. The burn-up is plotted on the abscissa and the oxide layer thickness on the ordinate. It can be seen that the alloys according to the invention can remain in the reactor for approximately twice as long (6 cycles) as the conventional alloys (3 cycles) before they have to be replaced for corrosion reasons. As noted above, all percentages cited herein are in percent by mass.

Claims

1. A zirconium alloy, comprising, in percent by mass:

Sn: 0.2-0.5%

Nb: 0.2-0.5%

Fe: 0.05-0.40%

V: 0-0.20%

O: 0.12-0.20%

Si: 80-120 ppm

C: ≦120 ppm

and a remainder of reactor-pure zirconium together with standard impurities.

2. The alloy according to claim 1, wherein a content of vanadium is at least 0.07%.

3. The alloy according to claim 1, wherein a sum of Sn, Nb, Fe, and V contents is at most 1.3%.

4. A component for the core of a light water reactor, comprising the alloy according to claim 1.

5. The component according to claim 4 configured for use in a pressurized-water reactor.

6. The component according to claim 4, produced maintaining a cumulative annealing parameter of (10-40) E-18 h.