Cladding tube for nuclear fuel and nuclear fuel element having this cladding tube

Disclosed are a cladding tube for a nuclear fuel and a nuclear fuel element incorporating the cladding tube. The cladding tube consists of an inner zirconium liner layer and an outer zirconium alloy layer. The cladding tube has at least one of the following features: (I) the ratio a/b of the oxygen content a to iron content b in the zirconium liner layer is greater than 1.0, (II) the zirconium liner layer is made of a zirconium into the matrix of which impurities are dissolved, and (III) the second phase particles having microscopic sizes and dispersed in the inner surface of the zirconium liner layer and/or the outer surface of the zirconium alloy layer have been removed substantially. Owing to these features, undesirable stress corrosion cracking and local corrosion are remarkably suppressed in the cladding tube and the nuclear fuel element of the invention.

BACKGROUND OF THE INVENTION 
The present invention relates to a cladding tube for nuclear fuel used in a 
nuclear reactor and, more particularly, to a cladding tube made of a 
zirconium alloy lined on its inner surface with a zirconium liner. The 
invention is also concerned with a nuclear fuel element having the 
above-mentioned cladding tube. 
A nuclear fuel element comprises a cladding tube accommodating therein a 
stack of a plurality of fuel pellets, which are formed by sintering an 
uranium oxide, thorium oxide, plutonium oxide or a compound thereof, and 
end plugs sealing both open ends of the cladding tube. Further, in an 
upper portion of the nuclear fuel element, there are provided a gas 
storage plenum and a spring for stably holding the fuel pellets. 
In the nuclear fuel element having a construction as mentioned above, for 
the cladding tube such functions are required as to prevent mutual contact 
and chemical reaction of the fuel pellets with the coolant or moderator 
and to prevent contamination of the coolant by radioactive fission 
products emitted from the fuel pellets. Accordingly, the material of the 
cladding tube is required to have excellent mechanical properties and high 
corrosion resistance under the operating conditions in the nuclear 
reactor, as well as a small neutron absorption. From these points of view, 
zirconium alloys containing zirconium as the major component, such as 
Zircaloy material, are widely used as the material of the cladding tube. 
The zirconium alloys have a small neutron absorption cross section, and 
exhibit only a small reactivity with pure water or steam and suitable 
strength and ductility at temperatures below 400.degree. C., and so they 
have excellent properties as a cladding material used under normal 
condition. However, from the operating experiences up to now it has become 
clear that at high degree of burn-up the cladding tube causes a stress 
corrosion cracking because of a synergetic effect of chemical reaction 
with corrosive fission products (iodine, etc.) and stress caused in the 
cladding tube by thermal expansion of the fuel pellets. In addition, the 
outer surface of the cladding tube is locally oxidized by the coolant or 
moderator of the nuclear reactor. 
In order to prevent the aforesaid stress corrosion cracking of cladding 
tube, a so-called zirconium-lined cladding tube has been developed, in 
which the cladding tube is lined on its inner surface with a zirconium 
liner. The zirconium liner is expected to prevent stress corrosion 
cracking by eliminating contact between the cladding tube and the 
corrosive fission products and by relieving local stress caused by thermal 
expansion of the fuel pellets. Thus, the zirconium liner is required to 
have a high resistance to corrosive fission products and a high ductility 
to relieve effectively any local stress. 
In general, zirconium is less sensitive to the stress corrosion cracking in 
comparison with zirconium alloys and has high ductility and small neutron 
absorption cross section, and so it has excellent properties as the liner 
material. 
SUMMARY OF THE INVENTION 
Accordingly, a first object of the invention is to enhance the stress 
corrosion cracking resistance of a zirconium liner which lines the inner 
surface of the cladding tube for a nuclear fuel. 
A second object of the invention is to reduce the unfavorable effect of 
impurities contained in the zirconium liner on the stress corrosion 
cracking resistance. 
A third object of the invention is to prevent local corrosion of the outer 
surface of a cladding tube due to reaction of the cladding tube material 
with the coolant or moderator. 
A fourth object of the invention is to provide a nuclear fuel element 
having a cladding tube which is improved to achieve the above-mentioned 
first to third objects and thus has an enhanced reliability over a long 
period of use. 
The present inventors have made an intensive study and carried out various 
experiments to achieve these objects, and found that these objects can be 
attained by the following measures I to III and, as a result, attained the 
invention: 
I. To reduce the stress corrosion cracking susceptibility of a zirconium 
liner by limiting the total content of impurities to a level not greater 
than 5,000 ppm and maintaining the ratio a/b of the oxygen content a (ppm) 
to iron content b (ppm) greater than 1.0, paying specific attention to 
oxygen and iron among all impurities contained in the zirconium liner. 
II. To reduce the stress corrosion cracking susceptibility of a zirconium 
liner by beforehand dissolving the impurities contained in the zirconium 
liner into the zirconium matrix, especially the impurities liable to form 
precipitates (referred to as "second phase particles", hereinunder) which 
are intermetallic compounds with zirconium. 
III. To remove the second phase particles of compounds containing Fe, Cr, 
Ni, Sn and so forth, which are dispersed on the inner and/or the outer 
surface of the zirconium-lined cladding tube. 
By combining the measure I with the measure II and by combining the measure 
I with the measure III, it is possible to achieve the objects of the 
invention more effectively.

DETAILED DESCRIPTION OF THE INVENTION 
The invention is described in detail hereinunder with reference to the 
accompanying drawings. 
Firstly, the description is made as to the aforementioned measure I. As 
explained already, it has become clear that zirconium is less liable to 
cause the stress corrosion cracking in comparison with zirconium alloys. 
However, it becomes liable to cause the stress corrosion cracking with 
increase in contents of impurities such as oxygen, iron, etc. Hitherto, it 
has been believed that the oxygen content is an important factor because 
among the impurities the oxygen enhances the mechanical strength of 
zirconium, and a technical thought of suppressing the oxygen content below 
a certain level is disclosed in Japanese Patent Publication No. 33037/80 
[corresponding to U.S. patent application Ser. No. 522,856 (Nov. 11, 
1974)] and Japanese Patent Application Laid-Open Publication No. 59600/79 
(corresponding to U.S. Pat. No. 4,200,492). However, according to the 
results of experiments recently carried out by the present inventors, it 
became clear that the iron content was a more important factor than the 
oxygen content. FIG. 1 shows, from the results of the experiments recently 
carried out, the effects of oxygen content and iron content on the stress 
corrosion cracking susceptibility (that is, the liability to cause the 
stress corrosion cracking) of zirconium. The Figure shows that, while the 
oxygen content causes no remarkable effect, the stress corrosion cracking 
becomes liable to occur with increase in the iron content. By the way, as 
shown in FIG. 2, in the case of the zirconium liner in cladding tube 
commercially produced for nuclear fuel elements, the ratio a/b of the 
oxygen content a (ppm) to iron content b (ppm) was about 1.0. It became 
clear that the zirconium liner, whose ratio a/b was greater than 1.0, had 
excellent properties in both of stress corrosion cracking susceptibility 
and mechanical strength. FIG. 4 shows the results of experiments which 
concretely represent the content of FIG. 2. In the region wherein the 
ratio a/b is greater than 1.0, it is seen that the stress corrosion 
cracking susceptibility decreases with decrease in the iron content and 
the mechanical strength increases with increase in the oxygen content. 
Accordingly, even in the region wherein the ratio a/b is greater than 1.0, 
the zirconium liner with a smaller iron content and a larger oxygen 
content has the most preferable properties. 
Presently, highly pure crystal bar zirconium with oxygen and iron contents 
less than 200 ppm, and sponge zirconium with oxygen and iron contents of 
about 500 to 1000 ppm, are considered as the material for the zirconium 
liners. 
On the basis of the aforementioned knowledges, it became clear that the 
zirconium had the advantages and disadvantages mentioned below. Since the 
crystal bar zirconium has a low iron content, it has a low stress 
corrosion cracking susceptibility, but since it has a too small oxygen 
content, the solid solution strengthening owing to oxygen can not be 
obtained and so it has a small mechanical strength, and in addition it has 
a disadvantage that its price is too high. On the other hand, although the 
sponge zirconium is inferior in its stress corrosion cracking 
susceptibility to the crystal bar zirconium, it has advantages that owing 
to the high oxygen content it has higher mechanical strength than the 
crystal bar zirconium. In addition the sponge zirconium is more economical 
than the crystal zirconium. 
On the basis of the results of experiments shown in FIG. 1, the present 
invention provides an economical zirconium liner material which has both 
of the stress corrosion cracking susceptibility equivalent to that of 
crystal bar zirconium and the mechanical strength equivalent to that of 
sponge zirconium, by making the ratio a/b of the oxygen content a to iron 
content b in the zirconium greater than 1.0. This can be achieved by 
reducing iron content in the sponge zirconium, or by adding oxygen to the 
crystal bar zirconium. That is, by lowering the iron content the stress 
corrosion cracking susceptibility is suppressed to a low level, and by 
containing oxygen of a suitable content the mechanical strength is 
maintained at a high level. 
Preferable embodiments of the invention are described hereinunder. 
Embodiment 1 
The nuclear fuel element of this embodiment comprises, as shown in FIG. 3, 
a cladding tube 1 accommodating therein a plurality of fuel pellets 2 and 
end plugs 3a and 3b sealing both ends of the cladding tube 1. In an upper 
portion of the nuclear fuel element, a gas plenum 4 is formed. A spring 5 
for holding the fuel pellets 2 is provided in the gas plenum 4. A 
zirconium liner 6 is lined on an inner surface of the cladding tube 1 over 
its entire length. The zirconium liner 6 interposes between the cladding 
tube 1 and the fuel pellets 2. The cladding tube 1 is made of Zircaloy-2, 
whose composition is shown in Table 6 submitted later. The zirconium liner 
6 is made of zirconium containing impurities, whose composition is shown 
at No. 1 in Table 1. 
TABLE 1 
______________________________________ 
(Unit: ppm) 
No. O Fe Cr Hf Si W C H N 
______________________________________ 
1 670 420 123 80 30 &lt;10 &lt;50 12 11 
2 1040 610 117 70 &lt;30 &lt;10 &lt;50 13 22 
______________________________________ 
Using the zirconium-lined cladding tube (having an inner zirconium liner 6 
and an outer zirconium alloy layer 1) as a sample, stress corrosion 
cracking test was carried out by both of the uni-axial tensioning method 
and the expanding mandrel method in order to evaluate the performance of 
the liner. The test result of the zirconium-lined cladding tube in the 
nuclear fuel element of this embodiment is shown at A.sub.l in FIG. 4. The 
testing conditions are as follows: 
iodine concentration: 0 to 20 torr. 
strain rate: 10.sup.-6 to 10.sup.-3 sec.sup.-1. 
testing temperature: 350.degree. C. 
A cladding tube having a zirconium liner made of zirconium containing 
impurities, whose composition is shown at No. 2 in Table 1, was also 
tested under the same conditions as mentioned above. The result thereof is 
shown at A.sub.2 in FIG. 4. 
The mechanical properties of the zirconium liners made of zirconium 
containing impurities, whose compositions are shown at Nos. 1 and 2 in 
Table 1, are shown in Table 2. 
TABLE 2 
______________________________________ 
tensile strength 
No. (at room temp.) 
0.2% proof stress 
elongation 
______________________________________ 
1 206 MPa 207 MPa 10% 
2 359 MPa 295 MPa 9% 
______________________________________ 
The same tests as mentioned above were made on 19(nineteen) kinds of 
zirconium liners having different iron and oxygen contents, besides two 
kinds shown in Table 1. The results thereof are also plotted in FIG. 4. 
As shown in FIG. 4, the zirconium-lined cladding tube, wherein the ratio 
a/b of oxygen content a to iron content b in the zirconium liner 6 is 
greater than 1.0, has superior property. 
The zirconium liner 6 having a reduced iron content is obtainable by means 
of classifying ingots produced by melting a sponge zirconium and using 
only the portions having small iron content. The iron content can be 
reduced also by removing iron by heating the ingot in a high vacuum 
atmosphere nearly to the melting point of iron, by making use of the fact 
that the melting point of iron is lower than that of zirconium. The use of 
sponge zirconium affords a greater economy as compared with the crystal 
bar zirconium. 
Next, a description will be made hereinunder as to the aforementioned 
measure II. 
The present inventors have found that the effectiveness against stress 
corrosion cracking produced by the zirconium liner is impaired as the 
amount of aforementioned second phase particles (precipitates of 
intermetallic compounds with zirconium) in the zirconium matrix is 
increased. Namely, when second phase particles of sizes above 0.1 .mu.m 
are irregularly dispersed in the grain boundaries and within the grains 
under corrosive environment formed by iodine which is a fission product, 
the stress is concentrated around the second phase particles to promote 
the initiation and propagation of the cracks. Therefore, a higher stress 
relieving effect, which is one of the purposes of the provision of the 
zirconium liner, can be attained by reducing the amount of the second 
phase particles. This can be achieved by dissolving insoluble impurities 
already mentioned into the zirconium matrix. 
Table 3 shows the maximum solubility limits and maximum dissolving 
temperature of the elements which form the second phase particles, as well 
as the compositions of the second phase particles. These elements are more 
or less contained in the zirconium which is a material used in nuclear 
reactors. 
TABLE 3 
______________________________________ 
maximum solubility 
limit to Zr secondary 
ele- dissolving temp. phase 
ment amount (wt %) 
(.degree.C.) 
particles 
remarks 
______________________________________ 
Fe 0.02 800 Fe.sub.2 Zr 
Cr trace 835 ZrCr.sub.2 
Si 0.1 860 Zr.sub.4 Si 
W 0.5 860 ZrW.sub.2 
Ni trace 808 Zr.sub.2 Ni 
Al 0.35 700 Zr.sub.3 Al 
at 940.degree. C., dissolving 
amount of Al is 3.5% 
C trace ZrC 
______________________________________ 
From Table 3, it will be seen that, among the impurity elements, Fe, Cr, Ni 
and C have very small solubility limits and, hence, are liable to form 
second phase particles. The elements Si, W and Al are generally contained 
in zirconium in small amounts, but they have a possibility of forming 
second phase particles when they are segregated in zirconium. 
According to the invention, therefore, in order to re-dissolve the impurity 
elements into the zirconium matrix, the zirconium to be used as the liner 
is heat treated at temperatures around 800.degree. C. or 860.degree. C. 
near the phase transformation temperature (862.degree. C.), and then 
rapidly cooled to room temperature, thereby uniformly re-dissolving the 
second phase particles. Further, in order to remove the strain caused by 
the solution heat treatment of the zirconium, a strain relief annealing is 
performed at such a low temperature range that the second phase particles 
are prevented from being re-precipitated. 
It is preferable that the amount of second phase particles is as small as 
possible. However, a sufficient effectiveness is attained even when the 
impurities are not perfectly dissolved into the zirconium matrix, since 
the initiation probability of the stress corrosion decreases with a 
decrease of an amount of second phase particles. In fact, a satisfactory 
improving effect is obtained when the number of particles greater than 
about 0.1 .mu.m among the materially observable second phase particles is 
reduced to less than 50% of that in the zirconium liner to which the 
present invention has not been applied. 
Preferable embodiments of the invention will be described hereinunder. 
Embodiment 2 
The nuclear fuel element of this embodiment is shown in FIG. 5, and is 
almost identical in its construction with that of the embodiment 1, except 
the zirconium liner 34. This zirconium liner 34 is made of zirconium 
containing the impurities, whose composition is shown at No. 1 in Table 4, 
which impurities have been dissolved into the zirconium matrix. The 
zirconium liner 34 shown in FIG. 5 may be made of zirconium containing the 
impurities, whose composition is shown at No. 2 or 3 in Table 4, which 
impurities also have been dissolved into the zirconium matrix. 
FIG. 6 illustrates, by way of example, a flow chart for producing a 
cladding tube 1 having a zirconium liner 6. In a step 11, zirconium is 
melted and formed into an ingot. The ingot is then forged and formed in a 
step 12 to become a hollow billet. The hollow billet is inserted into 
another hollow billet which has been formed beforehand from a zirconium 
alloy such as Zircaloy-2 (point D). Two hollow billets thus assembled 
monolithically are subjected to a hot extrusion (step 13) and become a 
blank tube of zirconium-lined cladding tube. Thus, this blank tube is 
composed of an inner zirconium layer and an outer zirconium alloy layer. 
The blank tube is then subjected to repetitional cycles of treatment in 
which cold rolling (step 14) and annealing (step 15) are effected 
alternatingly for a predetermined number of cycles, and becomes the 
zirconium-lined cladding tube 1 (step 16). According to this embodiment, 
the following steps (A), (B) and (C) are added in the flow chart shown in 
FIG. 6. 
(A) Solution heat treatment after formation of hollow zirconium billet. 
(B) Solution heat treatment on the blank tube after the hot extrusion. 
(C) Solution heat treatment after final annealing. 
In the production of the nuclear fuel element according to this embodiment, 
the best result is obtained when all of the steps (A), (B) and (C) are 
applied. However, any one of the steps (A), (B) and (C) has sufficient 
effectiveness. In particular, it is effective to add the step (C). 
Impurities contained in the zirconium and shown in Table 3 can be almost 
dissolved into the zirconium matrix by improving any one of the steps (A), 
(B) and (C), or any combination of them. 
The impurities contained in the zirconium are subjected to the solution 
heat treatment at the aforesaid step (C). That is, after the final 
annealing, the cladding tube thus obtained is heated in a vacuum 
atmosphere at 850.degree. C. for 2 hours and then quenched to the room 
temperature. Further, the cladding tube is annealed in a vacuum atmosphere 
at 530.degree. C. for 2 hours. In this way, a zirconium-lined cladding 
tube having the zirconium liner 6 into the matrix of which the impurities 
have been dissolved is obtained. 
Using zirconium containing the impurities shown at Nos. 1, 2 and 3 in Table 
4, three kinds of zirconium-lined cladding tubes, the impurities of which 
have been dissolved in compliance with the aforesaid flow chart, were 
produced, and they were subjected to the stress corrosion cracking tests 
by the expanding mandrel method. The testing conditions are as follows: 
iodine concentration: 1 mg/cm.sup.2. 
strain rate: 1.times.10.sup.-3 min.sup.-1. 
testing temperature: 350.degree. C. 
TABLE 4 
______________________________________ 
(Unit: ppm) 
No. Fe Cr Ni Si W Al C O N 
______________________________________ 
1 110 &lt;50 &lt;35 &lt;70 &lt;40 &lt;40 50 115 11 
2 455 84 &lt;10 &lt;70 &lt;40 54 50 525 14 
3 930 135 18 &lt;70 &lt;40 61 75 1060 17 
______________________________________ 
FIG. 7 shows the relationship between the mean circumferential stress at 
rupture of the cladding tube and the total contents of impurities, i.e., 
(Fe+Cr+Ni+Si+W+Al+C) content, contained in the zirconium. In this Figure, 
a curve 7 shows the property of the cladding tube of this embodiment, 
while a curve 8 shows the property of a conventional cladding tube, as 
obtained in the tests mentioned above. In both curves, although the strain 
at rupture is lowered as the total impurity content is increased, it is 
generally higher in all samples of this embodiment than in the samples of 
the conventional cladding tubes. This means that the cladding tube of this 
embodiment has lower susceptibility to stress corrosion cracking than the 
conventional cladding tubes. 
Embodiment 3 
The nuclear fuel element of this embodiment is identical in its 
construction with that shown in FIG. 5, except that the composition of 
impurities contained originally (i.e., before being dissolved into the 
zirconium matrix) in a blank material used for the zirconium liner 34 is 
different. A zirconium, which becomes a blank material of the zirconium 
liner in the nuclear fuel element of this embodiment, contained the 
impurities of the composition shown at No. 5 in Table 5, before being 
dissolved into the zirconium matrix. The zirconium liner in the nuclear 
fuel element of this embodiment is made of the zirconium into the matrix 
of which the impurities shown at above No. 5 have been dissolved. Such 
zirconium-lined cladding tube is produced in compliance with the 
aforementioned flow chart shown in FIG. 6, and the solution heat treatment 
is performed at the aforesaid step (C) similarly to the embodiment 2. This 
solution heat treatment is carried out by heating in a vacuum atmosphere 
the cladding tube obtained in the final annealing at 800.degree. C. for 3 
hours, and then quenching it to room temperature. Further, the cladding 
tube is annealed by being heated at 550.degree. C. for 2 hours. 
Incidentally, the zirconium liner in the nuclear fuel element of this 
embodiment may be made of zirconium containing the impurities, whose 
composition is shown at No. 6 in Table 5, instead of the zirconium 
containing the impurities, whose composition is shown at No. 5 in Table 5. 
In this case, the solution heat treatment is performed at the same 
conditions as in the case where the zirconium containing the impurities, 
whose composition is shown at No. 5 in Table 5, is used. 
TABLE 5 
______________________________________ 
(Unit: ppm) 
No. Fe Cr Ni Si W Al C O N H 
______________________________________ 
5 800 117 &lt;10 &lt;30 &lt;10 25 &lt;50 945 17 12 
6 610 117 &lt;10 &lt;30 &lt;10 25 &lt;50 1140 22 13 
______________________________________ 
In order to confirm the effect of the liner in the cladding tube of this 
embodiment, a stress corrosion cracking test was carried out by a 
uni-axial tensioning method in an iodine atmosphere, using zirconium 
plates having impurity contents shown at No. 5 in Table 5, under the 
following testing conditions: 
(i) iodine concentration in testing atmosphere: 10 torr. 
(ii) strain rate: about 10.sup.-5 /sec. 
(iii) testing temperature: 350.degree. C. 
From the above zirconium plates (having impurity contents shown at No. 5 in 
Table 5), three kinds of samples were prepared: namely, in FIG. 6, (1) 
sample obtained by being subjected to the steps till cold rolling, (2) 
sample obtained by being subjected to the steps till final annealing, and 
(3) sample obtained by being subjected to solution heat treatment 
following the final annealing. The solution heat treatment was made, 
similarly to that mentioned above, by heating in a vacuum atmosphere the 
zirconium plate obtained in the final annealing step at 800.degree. C. for 
3 hours and water-cooling it to the room temperature, and then annealing 
it at 550.degree. C. for 2 hours. 
FIG. 8 shows the test results of the above-mentioned samples (1), (2) and 
(3), in terms of strain ratio at rupture (i.e., the ratio between the 
strain at rupture in iodine and the strain at rupture in argon). It will 
be seen that the material of this embodiment exhibits the highest strain 
ratio at rupture. That is, the material of this embodiment has an 
extremely low susceptibility to stress corrosion cracking. 
FIG. 9 shows the photomicrographs by scanning electron microscope, of the 
above-mentioned samples' surfaces; FIG. 9a shows the structural state of 
the sample [the above-mentioned sample (2)] before being subjected to the 
solution heat treatment, and FIG. 9b shows the structural state of the 
sample [the above-mentioned sample (3)] after being subjected to the 
solution heat treatment. From these photomicrographs, it becomes clear 
that almost no second phase particles are observable in the sample (3), 
i.e., the cladding tube of this embodiment. 
Additionally, the zirconium plates, which would become the samples, were 
produced from the zirconium containing impurities, whose composition were 
shown respectively at Nos. 5 and 6 in Table 5. From respective zirconium 
plates, the samples subjected to the same solution heat treatment as in 
the case of the above-mentioned No. 5 and the samples not subjected to 
such solution heat treatment were prepared. For these four kinds of 
zirconium plates, the stress corrosion cracking test was carried out on 
the aforesaid conditions (i), (ii) and (iii). FIG. 10 shows, after 
arranging the data, the result of this test. It may be said that the 
smaller the reduction of area is, the higher the stress corrosion cracking 
susceptibility becomes. From FIG. 10, it will be seen that the stress 
corrosion cracking susceptibility becomes high with increase in the iron 
content. This is attributable to the fact that a greater amount of second 
phase particles, which are intermetallic compounds of iron and zirconium, 
are formed. However, in FIG. 10, the sample 21 subjected to the solution 
heat treatment as in this embodiment exhibits a higher reduction of area 
in comparison with the sample 20 not subjected to such solution heat 
treatment, and so it has an extremely low stress corrosion cracking 
susceptibility. Incidentally, the curve 19 represents the reduction of 
area of the sample 21 tested in an argon atmosphere. 
With the nuclear fuel element of this embodiment, the probability of 
causing the stress corrosion cracking becomes very small. 
Next, a description will be made as to the aforementioned measure III. 
The present inventors have found that the stress corrosion cracking 
susceptibility and the local corrosion can be lowered by reducing the 
second phase particles containing F, Cr, Ni, Sn, etc., which are dispersed 
in inner surface and/or outer surface of the cladding tube made of a 
zirconium alloy, the inner surface of which is lined with a zirconium 
liner. 
In general, the portions of the inner and outer surfaces of a cladding tube 
where the second phase particles appear exhibit high chemical reactivity 
and stresses tend to be concentrated to such portions when a load is 
applied to the cladding tube. The studies made by the present inventors 
proved a fact that the stress corrosion cracking susceptibility is 
increased as the amount of the second phase particles in the inner surface 
of the zirconium liner is increased. This may be attributable to the fact 
that, when the cladding tube is loaded under a corrosive environment, the 
initiation and propagation of cracks are promoted due to high chemical 
reactivity and stress concentration around the second phase particles. 
On the other hand, it is considered that the second phase particles also 
trigger the local corrosion (such as nodular corrosion) of the outer 
surface of the cladding tube due to reaction with water or steam. Namely, 
it is considered that the local corrosion is attributable to the fact that 
the second phase particles, which act as electron conductors, promote the 
corrosive reaction in the areas in the vicinity of the second phase 
particles. 
Since both of the stress corrosion cracking and the local corrosion occur 
exceedingly from a surface layer of the inner or outer surface of the 
cladding tube and proceed inwardly, it is necessary that the surface 
layer's susceptibility to stress corrosion cracking or local corrosion be 
maintained to a low level at an initial stage. Accordingly, it is also an 
aim of the invention to lower the surface's susceptibility to stress 
corrosion cracking or local corrosion by reducing the second phase 
particles precipitated in the surface portion of the cladding tube. 
Preferable embodiments of the invention will be described hereinunder. 
Embodiment 4 
The nuclear fuel element of this embodiment is shown in FIG. 11. The 
nuclear fuel element of this embodiment is almost identical in its 
construction with that shown in FIG. 3, except that the former's zirconium 
liner is somewhat different in its construction from the latter's 
zirconium liner. As shown in FIG. 11, in the nuclear fuel element of this 
embodiment, the region 35 wherein an amount of the second phase particles 
is reduced (hereinafter referred to as the particles reduced region 35) is 
formed in an inner surface side of the zirconium liner 6. The mean 
distribution density of the second phase particles existing in the 
precipitate particles reduced region 35 is extremely lower than that of 
the zirconium liner 6. Namely, the zirconium liner 6 is made of the 
zirconium containing the impurities, whose composition is shown at No. 2 
in Table 4, and the zirconium liner 6 includes the second phase particles 
containing these impurities. The particles reduced region 35 is a region 
wherein the amount of the second phase particles is smaller in comparison 
with the zirconium liner 6. The particles reduced region 35 is formed over 
the entire inner surface of the zirconium liner 6 and faces the outer 
surfaces of a plurality of fuel pellets 2. The cladding tube 1 in this 
embodiment is made of a zirconium alloy having a composition shown in 
Table 6, i.e., Zircaloy-2. The particles reduced region 35 is obtained by 
means of electropolishing the inner surface of the zirconium liner 6 
provided inside the cladding tube 1 by flowing an ethyl alcohol solution 
containing aluminum chloride and zinc chloride through the cladding tube 1 
in such a manner as mentioned later in relation to the embodiment 7 with 
reference to FIG. 18. 
TABLE 6 
__________________________________________________________________________ 
main components 
(wt. %) impurities (ppm) 
Sn Fe Cr Ni Zr Al B C Ca Cd Cl 
__________________________________________________________________________ 
1.53 
0.14 
0.10 
0.05 
Bal. 
&lt;35 &lt;0.25 
170 
&lt;10 &lt;0.25 
5 
__________________________________________________________________________ 
impurities (ppm) 
__________________________________________________________________________ 
Co Cu H Hf Mg Mn N O Pb Si Ti U W 
__________________________________________________________________________ 
&lt;10 
12 8 &lt;50 
&lt;10 
&lt;25 
40 1380 
&lt;25 
59 &lt;25 
&lt;1.0 
&lt;25 
__________________________________________________________________________ 
FIG. 12 shows a photomicrograph (magnification: about 10,000) by scanning 
electron microscope, of the inner surface of the zirconium liner 6, and 
from this photomicrograph a multiplicity of the second phase particles of 
the size smaller than about 0.5 .mu.m are observable. FIG. 13 shows a 
photomicrograph (magnification: about 10,000) by scanning electron 
microscope, of the same portion after being electropolished at a current 
density of about 10 mA/mm.sup.2 for 10 seconds in an ethyl alcohol 
solution containing aluminum chloride and zinc chloride, and from this 
photomicrograph it became clear that the second phase particles had been 
removed by the electropolishing in such a degree that almost no particles 
could be confirmed. 
Embodiment 5 
The nuclear fuel element of this embodiment is shown in FIG. 14. In this 
embodiment, the region 36 where amount of the second phase particles is 
reduced (hereinafter referred to as the particles reduced region 36) is 
formed in an outer surface side of the cladding tube 1, reversely to the 
embodiment 4. The number of the second phase particles in the particles 
reduced region 36 is smaller than that in the cladding tube 1. In the 
inner surface side of the zirconium liner 6, there exists no particles. 
The zirconium liner 6 is made of zirconium having the impurity composition 
shown at No. 2 in Table 4, and the cladding tube 1 is made of Zircaloy-2 
having the composition shown in Table 6. 
The particles reduced region 36 in the cladding tube's outer surface is 
obtained by pickling the cladding tube's outer surface in such a manner as 
mentioned later in relation to the embodiment 7 with reference to FIG. 19. 
As a pickling liquid, an aqueous solution containing 5 vol. % of 
hydrofluoric acid and 45 vol. % of nitric acid was used. 
FIG. 15 shows a photomicrograph by scanning electron microscope, of the 
outer surface of the cladding tube 1 made of the Zircaloy-2 (the 
composition of which is as shown in Table 6), after having been pickled 
for 20 seconds in the aqueous solution containing 5 vol. % of hydrofluoric 
acid and 45 vol. % of nitric acid. Although the mean number of the second 
phase particles per a surface area of 1 mm.sup.2 is about 2.times.10.sup.5 
in this case, when the outer surface is pickled for 2 minutes in the 
aqueous solution containing the above-mentioned acids the number of the 
second phase particles existing therein is very small. Namely, the mean 
number of the second phase particles in the outer surface portion (i.e., 
the particles reduced region 36) of the cladding tube 1 after having been 
pickled was reduced to about 1/10 of that in a portion of the cladding 
tube 1 inside the region 36. 
Embodiment 6 
In the nuclear fuel element of this embodiment, zirconium containing 800 
ppm of iron (i.e., the zirconium having impurity composition shown at No. 
5 in Table 5) is used as a blank material of the zirconium liner 6 in the 
nuclear fuel element of the embodiment 4. In the inner surface side of the 
zirconium liner 6 of this embodiment, there is formed the particles 
reduced region 35 similarly to the embodiment 4. This particles reduced 
region 35 is obtained by electropolishing with an ethyl alcohol solution 
mentioned later. 
When the inner surface of such zirconium liner 6 containing 800 ppm of iron 
as an impurity was microscopically observed at a magnification of about 
10,000, a multiplicity of the second phase particles of size smaller than 
about 2 .mu.m were detected; and after these particles were analyzed by an 
X-ray microanalyzer (XMA) and electron beam diffraction, it was identified 
that they were intermetallic compounds of zirconium and iron. By 
electropolishing the inner surface of this zirconium liner 6 in the ethyl 
alcohol solution containing aluminum chloride and zinc chloride at a 
current density of 20 mA/mm.sup.2 for 10 seconds similarly to the 
embodiment 4, it became that no particles were confirmed when observed at 
a magnification of about 10,000. Namely, it was possible to remove the 
impurity iron existing as the second phase particles in the zirconium 
liner 6 by the electropolishing. 
Embodiment 7 
The nuclear fuel element of this embodiment has, as shown in FIG. 17, the 
particles reduced regions 35 and 36 mentioned in the embodiments 5 and 6. 
The materials of the cladding tube 1 and the zirconium liner 6 are 
identical with those in the embodiment 6. The particles reduced regions 35 
and 36 are respectively positioned in the inner surface side of the 
zirconium liner 6 and in the outer surface side of the cladding tube 1. 
The zirconium-lined cladding tube used in the nuclear fuel assembly of this 
embodiment is produced in such a manner as mentioned hereinunder. Namely, 
the zirconium-lined cladding tube 1 made of Zircaloy-2 (the composition of 
which is as shown in Table 6) such as shown in FIG. 3 and having the 
zirconium liner 6 containing 800 ppm of iron as an impurity was prepared, 
and both ends of the cladding tube 1 were connected, as shown in FIG. 18, 
to polyethylene tubes 30A and 30B; and by flowing an aqueous solution 
containing 5 vol. % of hydrofluoric acid and 45 vol. % of nitric acid 
through the cladding tube for 2 minutes as indicated with arrow marks, the 
inner surface of zirconium liner 6 was firstly pickled. Then, as shown in 
FIG. 19, after closing both ends of the cladding tube with polyethylene 
plugs 31, the cladding tube was immersed in the above-mentioned aqueous 
solution 33 in a vessel 32 for 2 minutes, thus pickling the outer surface 
of the cladding tube. Further, after each pickling, the cladding tube was 
rinsed with water to completely get rid of the aqueous solution. By the 
picklings mentioned above, it was made possible to reduce the number of 
the second phase particles in the inner surface portion (i.e., the 
particles reduced region 35) of the zirconium liner 6 to less than about 
1/10 of that in a portion of the zirconium liner 6 other than the region 
35 and, further, to reduce the number of the second phase particles in the 
outer surface portion (i.e., the particles reduced region 36) of the 
cladding tube 1 to less than about 1/10 of that in a portion of the 
cladding tube 1 other than the region 36. 
Here, a concept "ratio of reduction of area" was assumed as an index for 
the evaluation of stress corrosion cracking susceptibility. More 
specifically, the samples of cladding tubes were subjected to uniaxial 
tensile tests in an iodine atmosphere and an argon atmosphere. 
Representing the reduction of area at ductile rupture in argon atmosphere 
by E.sub.1 and the reduction of area at rupture in iodine atmosphere by 
E.sub.2, the ratio expressed by (E.sub.1 -E.sub.2)/E.sub.1 was determined 
as the stress corrosion cracking susceptibility index. Thus, the smaller 
value of this index represents a lower stress corrosion cracking 
susceptibility, i.e., a higher resistance to stress corrosion cracking. 
FIG. 20 shows the stress corrosion cracking susceptibility indices as 
observed with zirconium having impurity contents of 2000 ppm and 1500 ppm, 
respectively before and after the removal of precipitates by the 
electropolishing explained in the embodiment 4. More specifically, in FIG. 
20, symbols E and F show the indices of the zirconium having the impurity 
content of 2000 ppm, before and after the removal of precipitates, 
respectively, while G and H show the indices of the zirconium having the 
impurity content of 500 ppm, before and after the removal of the 
precipitates, respectively. In the case of the zirconium containing about 
2000 ppm of impurity, the stress corrosion cracking susceptibility index, 
which was 0.25 in the state before the removal of precipitates, was 
reduced to 0.10 as a result of the removal of the precipitates. Similarly, 
in the case of the zirconium having the impurity content of about 1500 
ppm, the stress corrosion cracking susceptibility was reduced from 0.15 to 
0.05, as a result of removal of the precipitates. It is thus possible to 
remarkably reduce the stress corrosion cracking susceptibility by removing 
the second phase particles existing in the surface layer. 
Further, a zirconium containing 800 ppm of iron as an impurity (the 
zirconium shown at No. 5 in Table 5) and a zirconium containing 610 ppm of 
iron (the zirconium shown at No. 6 in Table 5) were subjected to a 
uni-axial tensile test in an iodine gas atmosphere. The test proved a fact 
that the zirconium having greater iron content causes the stress corrosion 
cracking with smaller elongation. This means that the iron as an impurity 
increases the stress corrosion cracking susceptibility. The zirconium 
having the iron content of 800 ppm was subjected to an electropolishing in 
the same way as explained before in the embodiment 6 to remove the 
iron-containing particles on the zirconium surface. The zirconium thus got 
rid of particles was subjected to a uni-axial tensile test in an iodine 
gas atmosphere, and it was confirmed that an elongation till rupture of 
this zirconium was greater than that of the zirconium having a lower iron 
content of 610 ppm. Namely, the susceptibility to stress corrosion 
cracking could be extremely lowered by removing the iron-containing 
particles precipitated in the zirconium surface. 
As mentioned above, it becomes possible to extremely lower the 
susceptibility to the stress corrosion cracking by removing or decreasing 
the second phase particles in the inner surface of the cladding tube, and 
so a sufficiently high effect of preventing the stress corrosion cracking 
can be obtained even when a low-priced zirconium having a somewhat low 
purity is used for the zirconium liner 6. Further, the local corrosion of 
the outer surface of the cladding tube during the long use can also be 
remarkably suppressed and so the reliability of the nuclear fuel element 
can be improved, by removing the second phase particles which are 
considered to have a strong influence on the local corrosion. 
Incidentally, the invention aims at removing the second phase particles in 
the surface of the cladding tube, but the methods therefor are not limited 
to those explained in the aforementioned embodiments, and it is possible 
to execute the invention by suitably conducting a chemical or 
electrochemical surface treatment. Further, even when the particles are 
removed not perfectly it is possible, by reducing the amount thereof, to 
suppress the stress corrosion cracking at initial stage which occurs in 
probability relationship in the surface layer or to lower the probability 
of occurrence for the local corrosion, thereby bringing about an effect of 
improving the cladding tube. 
Further, since the stress corrosion cracking and the local corrosion are 
phenomena which occur in probability relationship from the portions 
wherein the second phase particles exist, it is effective to reduce the 
amount of the second phase particles even though they are not perfectly 
removed, and so a sufficient improving effect can be obtained by reducing 
the materially observable particles to less than 50% in terms of 
volumetric ratio or less than 1/10 in terms of number of particles. 
As has been described, according to the invention, it is possible to 
suppress the occurrence of stress corrosion cracking in the cladding tube 
and, therefore, to obtain a highly durable and reliable nuclear fuel 
element by using this cladding tube.