High strength zirconium alloys containing bismuth

High strength zirconium alloys with improved strength and creep resistance having 1.5 to 6 weight percent bismuth and an element or mixtures of elements selected from the group of molybdenum, tin and niobium.

FIELD OF THE INVENTION 
The present invention relates to zirconium based alloys suitable for use in 
nuclear reactors, and more particularly for use in the cladding of nuclear 
fuel elements used in nuclear fuel assemblies for pressurized water 
reactors. 
BACKGROUND OF THE INVENTION 
Cladding for use in nuclear fuel rods for light water reactors functions to 
prevent fission products from being released from the fuel into the 
coolant/moderator and to prevent contact and chemical reactions between 
the fuel and the coolant/moderator. The cladding is required to have 
excellent mechanical properties and high corrosion resistance in the 
environment and for the conditions expected during reactor operations. 
Cladding is therefore required to have adequate corrosion resistance for 
the lifetime of the fuel rod for operation in steam and water at 
temperatures up to approximately 345.degree. C., adequate strength and 
creep behavior over the lifetime of the fuel rod, and typically have low 
parasitic neutron absorption for economic use of the fissionable fuel 
material. 
Common cladding materials include zirconium, zirconium alloys, and 
stainless steel. Zirconium based alloys in which the major component is 
zirconium have been used in the cladding of nuclear fuel rods or elements 
for several decades. Two of the most commonly used zirconium alloys that 
have given satisfactory performance are Zircaloy 2 and Zircaloy 4 and are 
described in American Society for Testing and Materials standard 
B350-93(1993), Standard Specification For Zirconium and Zirconium Alloy 
Ingots For Nuclear Application, compositions R60802 and R60804, 
respectively. Zircaloy 2 (composition R60802) is composed of from 1.20 to 
1.70 weight percent tin, 0.07 to 0.20 weight percent iron, 0.05 to 0.15 
weight percent chromium, 0.03 to 0.08 weight percent nickel, where the 
iron plus chromium plus nickel content is from 0.18 to 0.38 weight 
percent, and the balance is zirconium plus impurities. Zircaloy 4 
(composition R60804) is composed of from 1.20 to 1.70 weight percent tin, 
0.18 to 0.24 weight percent iron, 0.07 to 0.13 weight percent chromium, 
where the iron plus chromium content is 0.28 to 0.37 weight percent, and 
the balance is zirconium plus impurities. The maximum impurities for 
Zircaloy 2 and Zircaloy 4 are given in the following table which is from 
Table 1 of the ASTM B350-93 Standard. 
TABLE I 
______________________________________ 
MAXIMUM IMPURITIES, WEIGHT % 
R 60802 R 60804 
______________________________________ 
Aluminum 0.0075 0.0075 
Boron 0.00005 0.00005 
Cadmium 0.00005 0.00005 
Carbon 0.027 0.027 
Cobalt 0.0020 0.0020 
Copper 0.0050 0.0050 
Hafnium 0.010 0.010 
Hydrogen 0.0025 0.0025 
Oxygen * * 
Magnesium 0.0020 0.0020 
Manganese 0.0050 0.0050 
Molybdenum 0.0050 0.0050 
Nickel -- 0.0070 
Niobium 0.010 0.010 
Nitrogen 0.0065 0.0065 
Silicon 0.012 0.0120 
Tin -- -- 
Titanium 0.0050 0.0050 
Tungsten 0.010 0.010 
Uranium (Total) 0.00035 0.00035 
______________________________________ 
Although these and other alloys have provided generally adequate 
performance, they possess some deficiencies that have prompted further 
analysis and research to find alternative materials for and alternative 
constructions of nuclear fuel rod cladding to single walled cladding 
comprised of a single metal or alloy (sometimes referred to as "through" 
wall cladding) which does not possess both optimum strength and resistance 
to corrosion. Alternative constructions to single or through wall cladding 
for use as nuclear fuel rod cladding includes two layer or multiple layer 
tubing. These types of cladding have (a) an outer layer of a highly 
corrosion resistant alloy and (b) an inner layer that provides the bulk of 
the mechanical strength of the cladding. Cladding of this type, sometimes 
referred to as duplex cladding, with an extra low tin Zircaloy-type outer 
layer (nominally 0.8 wt. % tin) and a Zircaloy-4 inner layer is currently 
in use for nuclear fuel rod cladding. Zircaloy-4 inner layer cladding with 
a thin outer layer (3 to 5 mil) of various other corrosion resistant 
alloys has been produced and tested in-reactor. An outer layer alloy 
containing 0.5 wt. % tin, 0.5 wt. % iron, balance zirconium, and another 
outer layer alloy containing 0.5 wt. % tin, 0.5 wt. % iron, 0.2 wt. % 
chromium, balance zirconium have each shown exceptional corrosion 
performance in a high temperature pressurized water reactor. Examples of 
multiple layered tubing constructions and alloys for nuclear fuel rods are 
discussed in U.S. Pat. Nos. 5,493,592; 4,963,316; 4,735,768, which are 
each hereby incorporated by reference. 
With the higher burnups and longer in-reactor residence times that are 
being pursued and which, for largely economic reasons, continue to be 
increased, performance limits of commonly used alloys for nuclear fuel rod 
cladding are being reached. The corrosion resistance of the Zircaloys has 
been a major concern, especially in modern high coolant temperature 
pressurized water reactors that employ low leakage core loadings where the 
corrosion film on Zircaloy can build up to unacceptable levels for burnups 
around 50 to 60 MWd/kgU. In efforts to optimize the corrosion performance 
of the Zircaloys, through a reduction in the tin level, the strength and 
creep properties of the cladding material have thereby been diminished. 
For example, over the last decade the tin level of the Zircaloys used as 
cladding materials in nuclear fuel rods which was nominally held at 
approximately 1.55 wt. % has been lowered to a nominal level of 
approximately 1.30 wt. %. This reduction in the level of tin has resulted 
in substantially better corrosion performance specifically at higher 
burnups, but the reduction in tin has negatively impacted the mechanical 
properties of the cladding. Tin is a solute solution strengthening alloy 
element in Zircaloy and improves the strength and creep resistance of the 
alloy. However, lowering the tin level in Zircaloy reduces the resistance 
of the cladding to creepdown as well as the strength of the cladding. 
In attempts to overcome the limitations in the higher burnup performance of 
the zirconium alloys and the Zircaloys, alloy development programs have 
been initiated and research and development continue to this date for 
zirconium alloys for use as a nuclear fuel rod cladding that would have a 
more favorable combination of corrosion resistance, high strength and 
creep resistance as well as a low neutron cross section. 
An object of the present invention is to improve upon the nuclear fuel rod 
claddings produced to date by using (I) an alloy for the outer layer of a 
multiple layered cladding tube with exceptional in-reactor corrosion 
characteristics and in accordance with the present invention to utilize 
(II) a new alloy for the inner part of the cladding that is of 
substantially higher strength than Zircaloy-2 or Zircaloy-4, while 
maintaining low parasitic neutron absorption characteristics of the latter 
alloys. 
By using such a higher strength alloy for an inner layer of a multiple 
layer cladding tube, the overall cladding tube wall thickness can be 
reduced while still meeting the mechanical design and performance criteria 
of the fuel rod. By being able to reduce the cladding wall thickness, the 
cladding weight per unit length of cladding can be reduced and the cost of 
a cladding tube of a given length is reduced since less material is needed 
for the production of the cladding. Furthermore, by being able to reduce 
the cladding wall thickness, improvements in fuel cycle costs resulting 
from a reduction in the parasitic thermal neutron absorption can be 
obtained since parasitic neutron absorption for cladding of a given 
composition is directly proportional to the cladding wall thickness. 
Alloying elements with a smaller thermal neutron cross section than 
currently employed tin or niobium additions can reduce the parasitic 
neutron absorption of the alloy even further and gain additional 
improvements in fuel cycle costs. 
By using such a higher strength alloy for an inner layer of the multiple 
layer cladding, significant energy production cost savings can also be 
obtained by reducing cladding wall thickness and increasing fuel rod 
fissionable material weight which is achieved by being able to use larger 
diameter fuel pellets while maintaining a constant fuel rod outer 
diameter. For a given fuel rod design, the outer diameter of the cladding 
is primarily determined by thermal hydraulic considerations and therefore 
cannot readily be changed. Thin wall cladding can accommodate larger 
diameter fuel pellets than a thicker wall cladding of the same outside 
diameter. A larger diameter fuel pellet can have a lower uranium 
enrichment than a smaller diameter pellet to produce the same amount of 
energy. For slightly enriched uranium dioxide nuclear fuel, the lifetime 
energy production of a unit length of fuel rod is proportional to the 
total number of U.sup.235 atoms per unit length. Thus, for example, by 
using cladding with a 0.005 inch thinner wall than a thick wall design 
fuel rod containing 0.300 inch diameter pellets enriched to 4.00 wt. % 
U.sup.235, fuel pellets of 0.310 inch diameter may be used. The reduced 
U.sup.235 enrichment of these pellets would be 
##EQU1## 
(where L is a unit length of fuel) to maintain approximately the same 
number of U.sup.235 atoms per unit length of fuel. Alternatively, by 
maintaining the same U.sup.235 enrichment and increasing the pellet 
diameter, the number of U.sup.235 atoms per unit length of fuel rod is 
increased and the lifetime energy production of a unit length of fuel 
would be increased as well. Either alternative would lead to reactor fuel 
cycle cost reductions by using relatively higher cost, but thin wall, 
multiple layer cladding compared to using thicker through wall Zircaloy 
cladding. 
SUMMARY OF THE INVENTION 
The present invention relates to high strength zirconium based alloys which 
contain bismuth and which, in one embodiment consists essentially of 
molybdenum and 3 to 6 weight percent bismuth, balance zirconium. 
In another preferred embodiment, the high strength zirconium based alloy 
consists essentially of 1.5 to 6 weight percent bismuth and about 1 to 4 
weight percent tin, the balance zirconium. 
In another preferred embodiment, the high strength zirconium based alloy 
consists essentially of 1.5 to 3 weight percent bismuth, 0.5 to 3 weight 
percent niobium, 0.5 to 1.5 weight percent molybdenum, the balance 
zirconium, and where the sum of the amounts of niobium and molybdenum is 
greater than 1.5 weight percent.

DETAILED DESCRIPTION OF THE INVENTION 
Zircaloy-4 and Zircaloy-2 are much stronger and have much better creep 
resistance than unalloyed commercially pure zirconium. Zirconium alloys 
can typically be strengthened by two mechanisms; solid solution hardening 
and precipitation hardening. A combination of these strengthening 
mechanisms is employed in many high strength zirconium alloys. The most 
prominent precipitation strengthener is niobium. It is among others used 
in the Russian developed zirconium alloys having 1% niobium, 1.2% tin, 
0.4% iron, and in the zirconium 2.5%-2.8% niobium alloy used in Canada for 
CANDU pressure tubes. Other precipitation strengtheners are molybdenum and 
silicon. The strength of Zircaloy-4 and Zircaloy-2 derive mainly from the 
addition of tin which, because of its solubility in the zirconium matrix, 
is a solid solution strengthener. The atomic radius of tin, 0.1584 nm, is 
nearly the same as that of zirconium, 0.1602 nm, and tin atoms take the 
place of or substitute for zirconium atoms in the crystallographic lattice 
of the alloy. Tin, therefore, is also called a substitutional alloying 
element when used in zirconium base alloys. The addition of iron and 
chromium to Zircaloy-4 and iron, chromium and nickel to Zircaloy-2 does 
not substantially affect the mechanical properties of these alloys since 
these elements are nearly insoluble in the zirconium alpha phase and are 
added in small amounts only. These alloy elements are added mainly to 
improve the corrosion behavior of the Zircaloys. At reactor operating 
temperatures and below, these transitional elements are present in the 
form of small intermetallic particles with the approximate compositions 
Zr(CrFe).sub.2 or Zr.sub.2 (NiFe). 
It has been determined in the present invention that the addition of 
certain alloying elements to zirconium produces alloys possessing improved 
strength and creep resistance. More particularly, the addition of bismuth 
making up about 1.5 to 6 weight percent and an element or mixtures of 
elements selected from the group consisting of molybdenum, tin and 
niobium, making up about 1 to 4 weight percent tin, 0.5 to 3 weight 
percent niobium and/or 0.5 to 1.5 weight percent molybdenum, the balance 
being zirconium, produces alloys which possess substantial improvement in 
strength and creep resistance. 
In accordance with the present invention, alloys for use as the inner layer 
of two layered cladding tube or an inner layer of a three or more layered 
cladding tube having high strength and improved creep behavior as well as 
reduced parasitic neutron absorption comprise zirconium with an addition 
of from 1.5 to 6 weight percent bismuth (Bi). Similar to tin, bismuth is a 
solid solution strengthener. The atomic radius of bismuth is 0.1700 nm 
compared to the atomic radius of zirconium which is 0.1602 nm. This makes 
bismuth a substitutional alloying element similar to tin. The added 
advantage of using bismuth as an alloying element is its very low thermal 
neutron cross section; 0.034 barns compared to zirconium with a cross 
section of 0.184 barns. The thermal neutron cross section of tin is 0.610 
barns. Whereas the addition of tin to zirconium increases the parasitic 
neutron absorption of the alloy over that of pure zirconium metal, the 
addition of bismuth lowers the parasitic neutron absorption by the alloy 
compared to either zirconium metal or to Zircaloy. 
The following zirconium alloys with concentration levels of alloying 
elements have higher yield strength and creep resistance than Zircaloy and 
zirconium. 
I. Ternary Alloys 
a. Zirconium-Bismuth-Molybdenum alloys with 3-6 weight percent Bismuth, and 
Molybdenum, balance Zirconium, preferably 0.5 to 1.5 weight percent 
Molybdenum 
b. Zirconium-Bismuth-Tin alloys with 1-4 weight percent Tin and 1.5-6 
weight percent Bismuth, balance Zirconium 
II. Quarternary Alloys 
a. Zirconium-Bismuth-Molybdenum-Niobium alloys with 
(A) 3-6 weight percent Bismuth, and Molybdenum and Niobium, balance 
Zirconium, preferably 0.5-1.5 weight percent Molybdenum and 0.5-3 weight 
percent Niobium; and 
(B) 1.5-3 weight percent Bismuth, 0.5-3.0 weight percent Niobium and 
0.5-1.5 weight percent Molybdenum, balance Zirconium where the sum of 
Molybdenum and Niobium is greater than 1.5 weight percent 
b. Zirconium-Bismuth-Molybdenum-Tin alloys with 1-4 weight percent Tin, 
1.5-6 weight percent Bismuth, and Molybdenum, balance Zirconium preferably 
0.5-1.5 weight percent Molybdenum 
III. Quinary Alloys 
a. Zirconium-Bismuth-Molybdenum-Tin-Niobium alloys with 1-4 weight percent 
Tin, 1.5-6 weight percent Bismuth, and Molybdenum and Niobium, balance 
Zirconium preferably 0.5-1.5 weight percent Molybdenum and 0.5-3 weight 
percent Niobium 
All the above alloys could furthermore contain up to approximately 0.1 
weight percent silicon for added strength and grain refinement purposes. 
In a preferred embodiment, the minimum silicon content should be 0.008 
weight percent (80 ppm). These alloys could also contain between 
approximately 0.008 and 0.02 weight percent (80 and 200 ppm) of carbon for 
grain size control. The oxygen level in the above alloys could be adjusted 
to fall in the range of 0.06 to 0.018 weight percent (600-1800 ppm) and 
preferably in the range of 0.06 to 0.09 weight percent (600-900 ppm) in 
order to impart low temperature strength to the alloys. 
Referring to the drawings, FIG. 1 represents a nuclear fuel assembly 10 for 
a pressurized water reactor (PWR) comprising a lower tie plate 12, guide 
tubes 14, nuclear fuel rods 18 which are spaced radially and supported by 
spacer grids 16 spaced along the guide tubes, an instrumentation tube 28, 
and an upper tie plate 26 attached to the upper ends of the guide tubes. 
Control rods which are used to assist in controlling the fission reaction 
are disposed in the guide tubes during reactor operation, but are not 
shown. Each fuel rod 18 generally includes a metallic tubular fuel rod 
cladding 110 (120) within which are nuclear fuel pellets 80 composed of 
fissionable material and an upper end plug 22 and a lower end plug 24 
which hermetically seal the nuclear fuel pellets within the metallic 
tubular fuel rod cladding. A helical spring member can be positioned 
within the fuel rod between upper end plug 22 and fuel pellet 80 to 
maintain the position of the fuel pellets in a stacked relationship. 
Referring to FIG. 2A which is a schematic representation of a 
cross-sectional view of a nuclear fuel rod for a PWR such as shown in FIG. 
1 constructed according to the teachings of the present invention having a 
composite cladding 110 comprising an outer layer 111 composed of a 
corrosion resistant zirconium and/or zirconium alloy metal and an inner 
layer 114 bonded metallurgically to inner wall 113 of outer layer 111 and 
composed a zirconium alloy consisting essentially of molybdenum and 3 to 6 
weight percent bismuth and the balance zirconium, and preferably where the 
amount of molybdenum is in the range of 0.5 to 1.5 weight percent. 
In another embodiment, the inner layer 114 is made from another zirconium 
alloy consisting essentially of molybdenum, niobium, and 3 to 6 weight 
percent bismuth and the balance zirconium, and preferably where the amount 
of molybdenum is in the range of 0.5 to 1.5 weight percent and the amount 
of niobium is in the range of 0.5 to 3 weight percent. 
In another embodiment, inner layer 114 is composed of a zirconium alloy 
consisting essentially of 1.5 to 6 weight percent bismuth and 1 to 4 
weight percent tin, the balance zirconium. 
In another embodiment, inner layer 114 is composed of a zirconium alloy 
consisting essentially of molybdenum and 1.5 to 6 weight percent bismuth 
and 1 to 4 weight percent tin, the balance zirconium. 
In another embodiment, inner layer 114 is composed of a zirconium alloy 
consisting essentially of molybdenum, niobium, and 1.5 to 6 weight percent 
bismuth and 1 to 4 weight percent tin, the balance zirconium, and 
preferably where the amount of molybdenum ranges from 0.5 to 1.5 weight 
percent and the amount of niobium ranges from 0.5 to 3 weight percent. 
In another embodiment, inner layer is composed of a zirconium alloy 
consisting essentially of 1.5 to 3 weight percent bismuth, 0.5 to 3 weight 
percent niobium, 0.5 to 1.5 weight percent molybdenum, the balance 
zirconium, where the sum of the amounts of niobium and molybdenum is 
greater than 1.5 weight percent. 
Referring to FIG. 2B which is a schematic representation of a 
cross-sectional view of another nuclear fuel rod for a PWR such as shown 
in FIG. 1 constructed according to the teachings of the present invention 
having a composite cladding 120 comprising an outer layer 121 composed of 
a corrosion resistant zirconium and/or zirconium alloy metal, an inner 
layer 124 composed a high strength zirconium alloy and an innermost layer 
127 or liner bonded metallurgically on the inside surface 126 of the inner 
layer 124. 
In accordance with the present invention, inner layer 124 of composite 
cladding 120 is composed of a high strength zirconium alloy consisting 
essentially of molybdenum and 3 to 6 weight percent bismuth and the 
balance zirconium, and preferable where the amount of molybdenum is in the 
range of 0.5 to 1.5 weight percent. 
In another embodiment, the inner layer 124 is made from another zirconium 
alloy consisting essentially of molybdenum, niobium, and 3 to 6 weight 
percent bismuth and the balance zirconium, and preferably where the amount 
of molybdenum is in the range of 0.5 to 1.5 weight percent and the amount 
of niobium is in the range of 0.5 to 3 weight percent. 
In another embodiment, inner layer 124 is composed of a zirconium alloy 
consisting essentially of 1.5 to 6 weight percent bismuth and 1 to 4 
weight percent tin, the balance zirconium. 
In another embodiment, inner layer 124 is composed of a zirconium alloy 
consisting essentially of molybdenum and 1.5 to 6 weight percent bismuth 
and 1 to 4 weight percent tin, the balance zirconium. 
In another embodiment, inner layer 124 is composed of a zirconium alloy 
consisting essentially of molybdenum, niobium, and 1.5 to 6 weight percent 
bismuth and 1 to 4 weight percent tin, the balance zirconium, and 
preferably where the amount of molybdenum ranges from 0.5 to 1.5 weight 
percent and the amount of niobium ranges from 0.5 to 3 weight percent. 
In another embodiment, inner layer 124 is composed of a zirconium alloy 
consisting essentially of 1.5 to 3 weight percent bismuth, 0.5 to 3 weight 
percent niobium, 0.5 to 1.5 weight percent molybdenum, the balance 
zirconium, where the sum of the amounts of niobium and molybdenum is 
greater than 1.5 weight percent. 
To provide additional protection against pellet cladding interactive (PCI) 
induced failures, innermost layer 127 can be zirconium or a zirconium 
alloy, or another metal and preferably is made of pure or sponge zirconium 
or a dilute zirconium iron alloy of about 0.4 weight percent iron. 
While the foregoing description and drawings represent the preferred 
embodiments of the present invention, it will be apparent to those skilled 
in the art that various changes and modifications may be made therein 
without departing from the true spirit and scope of the present invention.