JPH10104378A - Nuclear fuel clad - Google Patents

Abstract

PROBLEM TO BE SOLVED: To improve reliability of nuclear fuel element by making the average grain diameter in the region forming innermost surface layer of clad be over a specific value. SOLUTION: The average grain diameter of zirconium liner layer 2 on a zirconium liner pipe lined inside with zirconium on the base 3 of zircalloy-2 is made over ca. 10μm. By controlling the iron concentration in the zirconium liner layer at high purity, for example, below ca. 100ppm, growth of grain in final heat treatment process (ca. 500 deg.C, 3h) in clad production is promoted. In this case, the clad is heated simultaneously cooing the outer surface with water, oil, inert gas, etc., so that the precipitation distribution of the base 3 and grain diameter do not change. Thus, hydrogen absorption of the clad is suppressed even under oxygen depleted atmosphere at the position apart from a primary break hole in the case where the fuel rod fails owing to fretting and the like and the occurrence probability of secondary hydrogenation and secondary failure can be lowered.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a nuclear fuel cladding tube used in a nuclear reactor, and more particularly to a nuclear fuel cladding tube for improving reliability in the event of fuel damage.

[0002]

2. Description of the Related Art A so-called zirconium liner in which pure zirconium having a thickness of, for example, 80 to 100 .mu.m is stretched as a barrier on the inner peripheral surface of a zircaloy-2 cladding substrate for the purpose of preventing stress corrosion cracking of a nuclear fuel cladding tube. Tubes are known (see JP-A-55-164396). The zirconium liner tube is expected to prevent the cladding tube from coming into contact with the corrosive fission products and to alleviate local stress generated in the cladding tube to prevent stress corrosion cracking of the cladding tube. With good operating results.

[0003]

However, if the cladding tube breaks due to an unexpected event such as fretting, the water vapor that has entered from the damaged portion immediately reacts with the pellets and the inner surface of the cladding tube near the damaged portion. There is a concern that the cladding tube may rapidly absorb hydrogen by generating hydrogen and forming a so-called oxygen-deficient atmosphere having a low water vapor pressure ratio at a position distant from the damaged portion, thereby causing secondary hydrogenation.

[0004] In order to prevent such secondary hydrogenation, a zirconium alloy having good corrosion resistance, for example, an alloy such as zircaloy-2 or zircaloy-4 is disposed inside zirconium, which is easily oxidized, or a conventional zirconium liner. There has been proposed a cladding tube in which the corrosion resistance is improved by adding a gas.

[0005] However, although these cladding tubes have improved corrosion resistance, their ductility near the inner surface of the tube is reduced due to a decrease in P resistance.
It is expected that the CI characteristics will deteriorate. Recently, it was reported that the oxidation rate of the pellet was much faster than that of the inner surface of the cladding tube. That is, it occurs under an oxygen-deficient atmosphere. Therefore, in the process up to the occurrence of secondary hydrogenation, the hydrogen absorption resistance under an oxygen-deficient atmosphere is more important than the oxidation of the inner surface of the cladding tube as a hydrogen generation source. In other words, it can be said that a cladding tube that hardly absorbs hydrogen from the inside has excellent secondary hydrogenation resistance.

The present invention has been made in view of the above circumstances, and suppresses the absorption of hydrogen into a cladding tube in an oxygen-deficient region away from a primary damage hole when a fuel cladding tube is damaged, thereby achieving secondary hydrogenation. It is an object of the present invention to provide a highly reliable nuclear fuel element by reducing the probability of occurrence of the nuclear fuel element.

[0007]

In order to achieve the above object, the present invention provides a zirconium alloy nuclear fuel cladding tube in which a plurality of nuclear fuel pellets are stacked and accommodated inside a cladding tube constituting an innermost surface layer of the cladding tube. Average grain size of 10μ
It is characterized by being larger than m.

In the present invention, the region constituting the innermost surface layer of the cladding tube is not particularly limited, and examples thereof include zirconium or a barrier layer containing zirconium as a main component disposed inside a zirconium alloy base material.
Alternatively, a zirconium alloy layer disposed further inside the zirconium barrier layer may be used.

In order to investigate the effect of the crystal grain size on the hydrogen absorption behavior of zirconium in an oxygen-deficient atmosphere, the inventors heat-treated high-purity zirconium (iron concentration: 80 ppm) that had been cold-worked, and subjected to heat treatment from 5 μm to 40 μm. A zirconium sample having an average crystal grain size of 400 ° C.
Were compared with each other. The result is shown in FIG.

As apparent from FIG. 2, the hydrogen absorption rate sharply decreases when the crystal grain size is in the range of 5 to 10 μm.
Thereafter, it gradually decreases as the crystal grain size increases. In particular, when the crystal grain size is larger than 10 μm, the hydrogen absorption rate is low, and the hydrogen absorption resistance is excellent. Therefore, by increasing the crystal grain size even on the inner surface of the fuel cladding for a nuclear reactor, In addition, it is possible to suppress hydrogen absorption from the inner surface of the tube under an oxygen-deficient atmosphere.

On the other hand, it is known that if the crystal grain size on the inner surface of the fuel cladding tube is increased, it is generally not preferable in view of the crack propagation phenomenon caused by corrosive fission products. The present inventors considered the stress intensity factor at the tip of the crack when the crack propagates and the very small unpenetrated crack observed in the actual fuel cladding tube, and found that the crystal grain size was 19% or less of the barrier thickness. In the case of, it was found that the crack did not grow. Therefore, it is desirable that the crystal grain size exceeds 10 μm and is 19% or less of the barrier thickness.

[0012]

DESCRIPTION OF THE PREFERRED EMBODIMENTS Hereinafter, the present invention will be described with reference to embodiments. FIG. 1 is a cross-sectional view of a cladding tube according to one embodiment of the present invention, wherein 1 is an inner surface of the tube, 2 is a zirconium liner portion,
3 is Zircaloy-2 part.

Zircaloy currently used in BWRs
In a zirconium liner tube having zirconium lined inside using 2 as a base material, the average value of the crystal grain size of the zirconium liner layer was set to be larger than 10 μm.

There are the following methods for increasing the value of the crystal grain size as described above. For example, there is a method in which the concentration of iron in the zirconium liner layer is controlled to a high purity of 100 ppm or less, and crystal grains are easily grown by the final heat treatment (577 ° C. × 3 hours) in the cladding tube manufacturing process.

According to experiments by the present inventors, the crystal grain size of zirconium having an iron concentration of 80 ppm is 650 ° C. × 2.
h, the final heat treatment temperature is raised to a range of 600 to 650 ° C., and the crystal grains of the zirconium liner layer are coarsened. Alternatively, there is a method in which a heating element is inserted inside the cladding tube in the final step of manufacturing the cladding tube, and the inner surface of the tube is heated to 600 to 800 ° C. to grow crystal grains. In this case, it is necessary to heat the outer surface side while cooling it with a coolant, for example, water, oil, an inert gas or the like, so that the distribution of precipitates and the crystal grain size of the Zircaloy-2 tube of the base material do not change.

The above embodiment is an example, and is not limited to a zirconium liner tube. For example, the base material is Zircaloy-2, and the thickness is 40 to 150 μm inside thereof.
A zirconium barrier layer of
Even in a triple tube in which a zirconium alloy layer of 0 μm or less is arranged, only the inner surface side is heated to increase the crystal grain size on the inner surface, and the nuclear fuel cladding tube of the present invention can be obtained.

In addition, the nuclear fuel cladding tube of the present invention can be obtained in any cladding tube having a zirconium-based alloy property by increasing the crystal grain size by heating the inner surface side. In the case of a zirconium liner material, the present invention is not limited to high purity, and the present invention is applicable to a zirconium liner to which iron is added and a liner to which an alloy element is added.

[0018]

As described above, according to the present invention,
If the fuel rod breaks for any reason, such as fretting, the hydrogen absorption of the cladding can be suppressed even in an oxygen-deficient atmosphere at a position away from the primary damage hole, reducing the probability of secondary hydrogenation and secondary damage. Can be reduced.

[Brief description of the drawings]

FIG. 1 is a cross-sectional view of a cladding tube according to an embodiment of the present invention.

FIG. 2 is a graph showing the relationship between the hydrogen absorption rate and the crystal grain size, which is the basis of the present invention.

[Explanation of symbols]

1 ... inner surface of cladding tube 2 ... zirconium liner part 3 ... zircaloy-2 part

Continued on the front page (72) Inventor Masafumi Nakaji 2163 Narita-cho, Oarai-machi, Higashiibaraki-gun, Ibaraki Pref. Japan Nuclear Fuel Development Co., Ltd.

Claims (4)

[Claims] In a zirconium alloy nuclear fuel cladding tube in which a plurality of nuclear fuel pellets are stacked and housed inside, an average crystal grain size of a region constituting an innermost surface layer of the cladding tube is 10%.
A nuclear fuel cladding characterized by being larger than μm. 2. The nuclear fuel cladding tube according to claim 1, wherein the region constituting the innermost surface layer of the cladding tube is zirconium or a barrier layer containing zirconium as a main component. 3. The nuclear fuel cladding tube according to claim 1, wherein the region constituting the innermost surface layer of the cladding tube is a zirconium alloy layer disposed further inside the zirconium barrier layer. 4. An average crystal grain size of a region constituting the innermost surface layer of the cladding tube exceeds 10 μm and a barrier layer thickness of 19
%.