Zirconium-base alloy nuclear fuel container and method

A fast neutron-irradiated zirconium-base alloy body having load-carrying capacity substantially greater than similar conventional zirconium-base alloy bodies likewise irradiated is produced by subjecting a body heat treated at 930.degree. C. and then water-quenched and containing 0.2 weight percent beryllium and at least 95 weight percent zirconium to integrated neutron flux approximating 1.2.times.10.sup.21 nvt while maintaining the body at about 330.degree. C.

The present invention relates generally to materials of construction of 
nuclear reactors and is more particularly concerned with a novel 
zirconium-base alloy nuclear reactor structural member or body having 
unique corrosion resistance, ductility and load-carrying capacity 
(resistance to stress corrosion) and possibly corrosion resistance in the 
irradiated condition. 
CROSS REFERENCE 
This invention is related to that disclosed and claimed in copending patent 
application Ser. No. 535,419, filed Dec. 23, 1974, abandoned in favor of 
Ser. No. 934,948 filed Aug. 18, 1978, allowed, in the names of Rodney E. 
Hanneman, Daeyong Lee and Craig S. Tedmon, Jr., which is based on the 
concept that a small amount of lanthanum or praseodymium will 
substantially improve the slow strain rate ductility of certain 
zirconium-base alloys and on the additional concept that these new 
zirconium alloys and certain others in the irradiated condition can under 
certain circumstances have surprising load-carrying capacity. 
BACKGROUND OF THE INVENTION 
Important requirements for materials used in boiling water nuclear reactor 
construction include low absorption for thermal neutrons, corrosion and 
stress corrosion resistance and mechanical strength. Zirconium-base alloys 
sufficiently satisfy these requirements that they are widely used for such 
purposes, "Zircaloy-2" (containing about 1.5 percent tin, 0.15 percent 
iron, 0.1 percent chromium, 0.05 percent nickel and 0.1 percent oxygen) 
and "Zircaloy-4" (containing substantially no nickel but otherwise similar 
to Zircaloy-2) being two of the important commercial alloys commonly 
finding such use. These alloys, however, are not nearly all that one would 
desire, particularly in respect to useful service life, despite many 
efforts of others during the past two decades to improve them. 
Mainly, these efforts have been aimed at improving corrosion resistance and 
usually this has involved changes in composition. Thus, in U.S. Pat. No. 
3,005,706, it is proposed that from 0.03 to 1.0 percent of beryllium be 
added to zirconium alloys intended for use in conventional boilers, 
boiling water reactors and similar apparatus. Similarly, in U.S. Pat. Nos. 
3,261,682 and 3,150,972, cerium and/or yttrium additions and a calcium 
addition, respectively, are proposed as zirconium alloy additions in like 
proportions for the same purpose. Accounts and reports of the results of 
such compositional changes are sparse, however, and the present commercial 
alloys do not include any of these additional constituents. 
The literature in this field, however, contains little concerning efforts 
to improve upon the mechanical strength of zirconium-base alloys and 
particularly the load-carrying capacity of fuel cladding and other reactor 
parts subjected to prolonged exposure to typical boiling water reactor 
conditions. This is in spite of the fact that it has long been general 
knowledge that slow strain rate ductility of these alloys is lost to a 
great extent as a result of radiation exposure over periods of a year or 
more. The problem of premature termination of service life because of fast 
neutron radiation-induced embrittlement is particularly aggravated in the 
case of nuclear fuel containment channels and tubes or cladding. The 
natural swelling of the fuel as it is burned produces high localized 
stresses leading to stress-corrosion cracking of the cladding at a time 
before corrosion of the type described in the above patents might normally 
necessitate cladding replacement. 
THE INVENTION 
This invention, which is predicated on my discovery and new concept to be 
described, provides an answer to both the iodine stress-corrosion problem 
and the embrittlement problem in the form of a process which can result in 
doubling the length of the service life of zirconium-base alloy nuclear 
fuel cladding. Moreover, this result is obtained without incurring any 
significant offsetting cost or performance disadvantage. 
My discovery is that a zirconium-base alloy of the kind presently used in 
nuclear reactors will have a much greater load-carrying capacity after 
being subjected to fast neutron radiation for a period of a year or so if 
it contains from 0.5 to 0.25 percent beryllium. More specifically, such an 
alloy will characteristically exhibit 500 to 600 percent greater 
load-carrying capacity (i.e., uniform strain to maximum load) than 
conventional beryllium-free cladding and can therefore be expected to 
serve in that use and environment much longer and possibly twice as long 
as the zirconium-base alloys in general use in nuclear reactors. 
My new concept is to prepare a zirconium-base alloy containing 0.05 to 0.25 
percent beryllium for use as nuclear fuel cladding by heating to a 
temperature of the order of 900.degree. C. and then water-quenching it. 
Although such heat treatment results in a phase transformation, the 
zirconium alloy transforming in part or totally from the alpha to the beta 
phase, and the prior art warns that detrimental effects on mechanical 
properties will result, I have found that there are substantial advantages 
to be gained by effecting such transformation. For one thing, ductility is 
enhanced materially, as will subsequently be described in more detail. For 
another, resistance to corrosion under boiling water nuclear reactor 
conditions resulting in heavy oxide coating formation on fuel cladding may 
thereby be substantially reduced or limited, as set forth more fully and 
generically claimed in copending patent application Ser. No. 552,794, 
filed Feb. 25, 1975, abandoned in favor of Ser. No. 852,906, filed Nov. 
18, 1977. 
In its method aspect, this invention in brief description includes the 
steps of forming a zirconium-base alloy body containing 0.05 to 0.25 
percent beryllium, heating the body to a temperature above 900.degree. C. 
and then quenching, and finally subjecting the body to boiling water 
reactor conditions for a long period of time such as a year or more. More 
specifically, the alloy body will be of at least 95percent zirconium, the 
quenching will be done with water and the nuclear reactor conditions will 
be a temperature of about 325.degree. C. and a fast-neutron flux of 1.0 to 
10.0.times.10.sup.21 nvt. 
In its product or article aspect, this invention takes the form of a 
zirconium-base alloy body of substantially greater load-carrying capacity 
than similar conventional bodies of the alloy irradiated in the same way 
and to the same extent and having at 325.degree. C. an unique combination 
of physical properties including 2.5 percent uniform elongation, 8.2 
percent total elongation and 35 percent area reduction, yield strength 
greater than 76,000 psi, and tensile strength greater than 80,000 psi. In 
more specific terms, the body is a nuclear fuel container for use in a 
nuclear reactor, and is in the form of a tube having microstructure in 
which the intermetallic phase is to some extent segregated at the grain 
boundaries as a consequence of the heat treatment and quenching steps 
stated above. Additionally, the fuel container or cladding is irradiated 
as a result of having been subjected for a long period to fast-neutron 
flux and has greater load-carrying capacity than a counterpart fuel 
container similarly irradiated but containing no berrylium.

DETAILED DESCRIPTION OF THE INVENTION 
As indicated by FIG. 1, a primary application of the present invention is 
for the fabrication of nuclear fuel assemblies such as that illustrated at 
10 consisting of a tubular flow channel 11 of generally square cross 
section provided at its upper end with lifting bale 12 and at its lower 
end with a nose piece (not shown due to the lower portion of assembly 10 
being omitted). The upper end of channel 11 is open at 13 and the lower 
end of the nose piece is provided with coolant flow openings. An array of 
fuel elements or rods 14 is enclosed in channel 11 and supported therein 
by means of upper end plate 15 and a lower end plate (not shown due to the 
lower portion being omitted). The liquid coolant ordinarily enters through 
the openings in the lower end of the nose piece, passes upwardly around 
fuel elements 14, and discharges at upper outlet 13 in a partially 
vaporized condition for boiling water reactors or in an unvaporized 
condition for pressurized reactors at an elevated temperature. 
The nuclear fuel elements or rods 14 are sealed at their ends by means of 
end plugs 18 welded to the cladding 17, which may include studs 19 to 
facilitate the mounting of the fuel rod in the assembly. A void space or 
plenum 20 is provided at one end of the element to permit longitudinal 
expansion of the fuel material and accumulation of gases released from the 
fuel material. A nuclear fuel material retainer means 24 in the form of a 
helical member is positioned within space 20 to provide restraint against 
the axial movement of the pellet column, especially during handling and 
transportation of the fuel element. 
The fuel element is designed to provide an excellent thermal contact 
between the cladding and the fuel material, a minimum of parasitic neutron 
absorption, and resistance to bowing and vibration which is occasionally 
caused by flow of the coolant at high velocity. 
Cladding 17 is produced in accordance with this invention by a process 
which includes in addition to the usual tube-forming operations a heat 
treatment in argon or other inert atmosphere above the alpha-alpha plus 
beta transformation temperature followed by a water quench. The rate at 
which the work piece is heated up to the transformation temperature range 
is a matter of choice, but the time it is maintained in that range is 
preferably about 30 seconds and the cooling rate down to 700.degree. to 
750.degree. C. may be as low as 50.degree. C. per second. As so treated, 
the zirconium alloy body is made more easily workable and forming 
operations are facilitated through the warm-working stage. It also 
appears, as indicated above, that the physical properties and particularly 
the ductility of the ultimate cladding product may be considerably 
enhanced in this manner. As a further advantage, depending upon the nature 
of the finishing operations involved in producing the cladding, the 
tendency toward corrosion may be to a large extent suppressed as a 
consequence of the heat treatment above the alpha-alpha plus beta 
transformation temperature of about 810.degree. C. This latter effect 
would be attributable, possibly, to the segregation of the intermetallic 
phase at the grain boundaries, as set out in the aforesaid copending 
application, Ser. No. 552,794. In any event, the zirconium alloy employed 
in this process is one which contains beryllium in amount from 0.05 to 
0.25 weight percent, and preferably also contains about 1.5 weight percent 
tin and 0.05 weight percent nickel, and at least 95 weight percent 
zirconium. In other words, it is preferably either Zircaloy-2 or 
Zircaloy-4. 
The method and products of this invention are set forth in more detail 
together with actual test results in the following illustrative example in 
which Zircaloy-2 was used, being melted in an electric arc furnace under 
vacuum to provide control specimens as well as test specimens meeting the 
special compositional requirements of this invention. 
EXAMPLE 
Of the total of seven test specimens, four were of commercial Zircaloy-2 
composition and the others differed therefrom only in that they each 
contained 0.2 weight percent beryllium. These specimens in the form of 
cast buttons about 2.5 inches in diameter and about one-half inch thick 
were machined to provide a smooth surface and then wrapped in zirconium 
foil, offset-forged approximately 30 percent, heated to 930.degree. C. in 
argon and then again offset-forged. They were sandblasted and wrapped 
again in zirconium foil and reheated to 930.degree. C. for 20 minutes and 
water-quenched. The four specimens (Nos. 1, 3, 4 and 5 in Table I below) 
containing no beryllium were then rolled to ultimate thickness of 
one-sixteenth inch by a multiple pass method, the final passes being 
cold-rolling operations. These sheets were sandblasted, pickled in aqueous 
2.0 percent HF and 6.0 percent HNO.sub.3, and then Specimens 1 and 3 were 
finally annealed at 650.degree. C. for one hour while Specimens 4 and 5 
were annealed at 580.degree. C. for four hours. Beryllium-containing 
Specimens 2, 6 and 7 were processed in the same manner as Specimens 4 and 
5 through the final annealing stage. Specimens 1 and 2 were maintained at 
327.degree. C. (620.degree. F.) in a neutral atmosphere for one year, and 
Specimens 3, 4, 5, 6 and 7 were exposed to fast-neutron radiation at 
temperatures of either 250.degree. C. or 327.degree. C. for the same 
twelve-month period, being located within standard-size fuel cladding 
dummy fuel rods installed in fuel bundles in a working boiling water 
reactor core. Flux wires of nickel and iron indicated that these specimens 
were subjected to radiation exposure, peaking at 3.1.times.10.sup.21 nvt 
for corresponding peak fast flux values of 7.times.10.sup.13 n/cm.sup.2 
-sec. Thus, the typical specimen of this series was exposed to the fast 
flux over a period of one to 11/2 years in a neutral or an inert helium 
atmosphere at the temperature indicated in Table I. 
The results of all of the tests made on these irradiated and unirradiated 
specimens are set out in Table I: 
TABLE I 
__________________________________________________________________________ 
Neutron 
Test Yield Tensile 
Fluence 
Temp, 
Oxygen 
Strength 
Strength, 
Uniform 
Total 
Reduction 
Specimen 
10.sup.21 
.degree.C. 
ppm wt 
(0.2%) psi 
psi Elong, % 
Elong, % 
of area, % 
__________________________________________________________________________ 
1 -- 327 1500 22,400 
28,400 
16.4 30.6 79.9 
2 -- 327 1200 27,100 
37,600 
16.0 27.8 56.8 
3 1.5 327 1500 73,800 
75,200 
0.35 5.9 42 
4 1.5 250 920 56,500 
56,700 
0.45 8.0 44 
5 1.4 250 1600 74,600 
76,000 
0.80 9.3 53 
6 1.2 250 1200 78,600 
84,800 
3.30 8.6 37 
7 1.2 327 1200 76,800 
81,200 
2.50 8.2 35 
__________________________________________________________________________ 
The test temperature stated in Table I is the temperature at which the 
mechanical properties of the specimen were tested in each instance, all 
these specimens being subjected to the same 327.degree. C. temperature 
environment over the twelve-month period under the conditions as set forth 
above. 
The effect of the beryllium addition is demonstrated in FIG. 2 where the 
dramatic difference in load-carrying capacity between Specimens 3 and 7 is 
indicated by Curves A and B, respectively. Also, it will be noted in this 
connection that in Table I the same inherent characteristic is reflected 
in the uniform strain-to-maximum-load which increased from 0.35 percent in 
Specimen 3 to 2.5 percent in Specimen 7, a net increase over 600 percent. 
The tests yielding these data were conducted at a strain rate of 
8.3.times.10.sup.-4 cm/cm/sec. Metallographic examination of these two 
specimens revealed that deformation was noticeably more diffuse in 
Specimen 7 than in Specimen 3. 
The effect of the beryllium addition is further illustrated in FIG. 3 
where, again, there is a dramatic difference in strength between Specimens 
3 and 7 as indicated by Curves C and D, respectively. As previously noted, 
the tests resulting in the data represented by these curves were conducted 
in an aggressive environment (i.e., under an iodine atmosphere) at a 
strain rate of 2.83.times.10.sup.-6 cm/cm/sec. 
The iodine atmosphere tests were conducted by subjecting the work piece at 
325.degree. C. in each case to an atmosphere of helium gas containing 
iodine in amount approximating the room temperature iodine partial 
pressure. Thus, helium gas is flowed continuously through an iodine 
crystal bed from which it entered the test chamber. Helium gas flow 
through the chamber was continuous as the pressure within the chamber was 
maintained slightly greater than atmospheric pressure.