Zirconium alloy fuel cladding

A cladding is provided for use in housing fissionable material in water cooled nuclear fission reactors. The cladding has inner and outer surfaces and includes (1) a cross-section of a Zirconium-based alloyed matrix, and (2) alloying elements in sufficient concentration to form precipitates disposed in the matrix. The cladding includes no more than 20 parts per million nitrogen by weight and is typically a modified Zircaloy-2 or Zircaloy-4. Metallurgically bonded to the inner region of the cladding is a zirconium barrier layer.

This invention relates to Zircaloy cladding for use in nuclear fuel 
elements. More particularly, the invention relates to cladding having 
improved nodular corrosion resistance while maintaining uniform corrosion 
resistance and axial crack propagation resistance. 
BACKGROUND OF THE INVENTION 
Nuclear reactors have their fuel contained in sealed cladding for the 
isolation of the nuclear fuel from the moderator/coolant system. The term 
cladding, as used herein, refers to a zirconium-based alloy tube composed 
of at least one metal in addition to the zirconium base. The cladding may 
be composed of more than one layer including a zirconium alloy substrate 
and an unalloyed zirconium barrier. Typically, the cladding is formed in 
the shape of a tube with the nuclear fuel contained in pellet form 
therein. These pellets are stacked in contact with one another for almost 
the entire length of each cladding tube, which cladding tube is in the 
order of 160 inches in length. Typically, the cladding tube is provided 
with springs for centering the fuel pellets and so-called "getters" for 
absorbing fission gases. Thereafter, the internal portions of the fuel rod 
are pressurized with various gases for optimum dissipation of gases 
produced from the fission reaction, and sealed at both ends. 
Zirconium and its alloys, under normal circumstances, are excellent nuclear 
fuel cladding since they have low neutron absorption cross sections and at 
temperatures below about 398.degree. C. (at or below the core temperature 
of the operating reactor) are strong, ductile, stable, and nonreactive in 
the presence of demineralized water or steam. "Zircaloys" are a widely 
used family of corrosion-resistant zirconium alloy cladding materials. The 
Zircaloys are composed of 97-99% by weight zirconium, with the balance 
being tin, iron, chromium, and nickel. "Zircaloy-2" and "Zircaloy-4" are 
two widely-used zirconium-based alloys for cladding. Zircaloy-2 has on a 
weight basis about 1.2 to 1.7 percent tin; 0.13-0.20 percent iron; 
0.06-0.15 percent chromium and 0.05 to 0.08 percent nickel. Zircaloy-4 has 
essentially no nickel and about 0.2% iron but is otherwise substantially 
similar to Zircaloy-2. 
The presence of alloying elements which are relatively insoluble in 
zirconium under normal conditions, will generally result in "precipitates" 
forming within a zirconium matrix. Under equilibrium conditions, the 
matrix--which is a single phase--will contain the alloying elements at 
concentrations no higher than their solubility limit. The 
precipitates--which form a second phase--contain higher concentrations of 
the alloying elements. For example, the precipitates found in the 
Zircaloys are represented by chemical formulas such as Zr(Fe,Cr).sub.2 and 
Zr.sub.2 (Fe,Ni). 
Cladding corrosion is a potential problem both in boiling water reactors 
(BWRs) and pressurized water reactors (PWRs). Corrosion in a BWR occurs in 
nodular or uniform formats on the zirconium cladding. Nodular corrosion is 
usually a porous, stoichiometric ZrO.sub.2 oxide forming on the surface of 
the cladding. It can rapidly cover the entire surface of pure zirconium, 
but it tends to form as small patches ("nodules" or "pustules") on the 
surface of the Zircaloys. Uniform corrosion is also a ZrO.sub.2 oxide 
forming on the surface of the cladding, but it usually contains a small 
excess of zirconium. As such, it contains excess electrons giving it a 
black or gray color and semiconductor properties. 
Nodular or pustule corrosion is not inherently bad. However, where fuel in 
the reactor has longer life, nodular corrosion may concentrate. And where 
such concentrated nodular corrosion acts in conjunction with certain 
contaminants--such as copper ions--localized spalling and ultimately 
penetration of the cladding wall can occur. 
Various approaches have been taken to minimize or eliminate nodular 
corrosion and the damage it can cause to cladding. In one widely used 
approach, the concentration of alloying elements (particularly iron, 
nickel, and chromium) in Zircaloy alloy is increased. This has been found 
to actually reduce the severity of nodular corrosion under reactor 
conditions. Unfortunately, increased concentrations of alloying elements 
also leads to increased rates of corrosion due to uniform corrosion. Even 
at such elevated rates, uniform corrosion has not been a significant 
problem in reactors operated under conditions common in the past. Today, 
however, it is increasingly common to operate reactors at high "burn-up" 
(i.e., to nearly complete consumption of the nuclear fuel). Under these 
conditions, the cladding is exposed to a neutron flux for long periods, a 
condition which tends to increase the degree of uniform corrosion. Thus, 
uniform corrosion can become a significant problem in modern reactor 
operation. 
In another approach to nodular corrosion containment, the precipitates in 
the Zircaloy matrix are purposely made small (e.g., less than about 0.1 
micrometer in diameter). They may be made small throughout the entire 
cross-section of the cladding or only in certain regions. For example, it 
is known to externally treat the outer water exposed surface of cladding 
with heating from a coil to produce a fine precipitate exterior surface. 
See Eddens et al. U.S. Pat. No. 4,576,654. Unfortunately, some research 
has suggested that small precipitates in the Zircaloy metal matrix can 
increase the danger of crack propagation in the cladding axial direction. 
See, for example, U.S. patent application No. 08/052,793, entitled 
ZIRCALOY TUBING HAVING HIGH RESISTANCE TO CRACK PROPAGATION and U.S. 
patent application Ser. No. 08/052,791, entitled METHOD OF FABRICATING 
ZIRCALOY TUBING HAVING HIGH RESISTANCE TO CRACK PROPAGATION, both of which 
were filed on Apr. 23, 1993, naming Adamson et al. as inventors, assigned 
to the assignee hereof, and are incorporated herein by reference for all 
purposes. These applications describe a cladding having a microstructure 
in which coarse precipitates predominate in the inner regions of the 
cladding and fine precipitates predominate in the outer regions of the 
cladding, the regions where corrosion is a problem. 
Corrosion and cracking can both damage cladding, but they are fundamentally 
different phenomena. Cracking is a mechanical breaking or splitting of the 
cladding wall, while corrosion is an electrochemical conversion of the 
cladding metal into an oxide or other non-metallic compound. Cracks may be 
initiated by a variety of causes including mechanical stresses as well as 
corrosion. Once a crack is initiated, it may pose little problem, so long 
as it remains confined to a small area. However, if the crack propagates, 
the cladding can be breached and the fission material eventually contacts 
the coolant or moderator. Ultimately, this can lead to an expensive 
reactor outage. 
The mechanical initiation of cracks can be attributed to various stresses 
in a conventional reactor. Cracks can start when debris such as wires or 
metallic shavings or particles find their way into reactor water that 
flows within the fuel bundles between the fuel rods. The debris may lodge 
at a fuel rod spacer adjacent the cladding wall. As a result, the debris 
vibrates or frets against the cladding wall under the influence of the 
passing steam/water mixture. Corrosion can be the source of initial crack 
propagation. Moreover, manufacturing defects can be the points of crack 
origin. Still further, crack propagation can start on the inside of the 
fuel rods in the corrosive high pressure environment present during 
in-service reactor life. 
U.S. Pat. Nos. 4,200,492 and 4,372,817, to Armijo et al as well as Adamson 
U.S. Pat. No. 4,894,203--each of which is incorporated herein by reference 
for all purposes--suggest solutions to preventing crack initiation by 
including a barrier on the inside of the cladding. Cladding containing 
barriers are sometimes referred to as "composite" cladding or cladding 
having two distinct metallurgical layers. 
Although it is highly desirable to prevent nodular corrosion of zirconium 
alloy cladding, it is also desirable to prevent uniform corrosion at high 
bum-up and to prevent axial crack propagation. There exists a need for 
cladding which is resistant to nodular corrosion while retaining 
resistance to uniform corrosion at high bum-up and axial crack 
propagation. 
SUMMARY OF THE INVENTION 
The present invention is directed to low-nitrogen zirconium alloy cladding 
resisting nodular corrosion while used to house fissionable material in 
water cooled nuclear fission reactors. The invention also provides methods 
of making such cladding. Preferably, the low nitrogen zirconium alloy 
cladding is a Zircaloy tube having less than 20 parts per million (ppm) 
nitrogen by weight. 
The low-nitrogen alloys of this invention show surprising resistance to 
nodular corrosion. In addition to their obvious direct benefit (resistance 
to nodular corrosion), the alloys of this invention have other secondary 
advantages. First, because they show improved resistance to nodular 
corrosion, they may employ lower concentrations of alloying elements 
(e.g., iron, nickel, and chromium). Such low concentration alloys (e.g., a 
modified Zircaloy having low nickel, iron, and chromium, in addition to 
low-nitrogen) will resist uniform corrosion at high bum-up. Second, the 
claddings of this invention need not rely on a microstructure having fine 
precipitates for nodular corrosion resistance. In fact, preferred cladding 
will have course precipitates (e.g., greater than about 0.2 micrometers in 
diameter) throughout the Zirconium alloy matrix. Such cladding resists 
crack propagation in the axial direction as well as nodular corrosion. 
Further, such cladding is relatively easy to produce because the late beta 
quenches and localized induction anneals normally employed to generate a 
fine precipitate distribution are unnecessary in this invention. 
In a preferred embodiment, the cladding will include no more than about 20 
ppm nitrogen by weight in a Zirconium-based alloy matrix having alloying 
elements in concentrations sufficient to form precipitates disposed 
throughout matrix. Preferably, the Zirconium-based alloy will be a 
modified Zircaloy-2 or Zircaloy-4 having alloying elements present in the 
concentration ranges of about 0.05 to 0.09 weight percent iron, about 0.03 
to 0.05 weight percent chromium, and about 0.02 to 0.04 weight percent 
nickel. In some embodiments, the cladding will have a controlled 
microstructure in which the precipitates proximate to an inner surface of 
a cladding tube will have an average size distribution of at least about a 
first predefined diameter and the precipitates proximate to an outer 
surface will have an average size distribution of at most about a second 
predefined diameter, such that the first predefined diameter is greater 
than the second predefined diameter. In especially preferred embodiments, 
the first predefined diameter will be about 0.2 micrometers, and the 
second predefined diameter 0.1 micrometers. 
In another aspect, the present invention provides a method of preparing a 
cladding for use in housing fissionable material in a water cooled nuclear 
fission reactor. The cladding will be prepared by converting a Zircaloy 
tube containing at most about 20 ppm nitrogen to a cladding containing at 
most 20 ppm nitrogen through a plurality of steps including cold working 
and annealing. Each step in this process that is conducted at a 
temperature above about 500.degree. C. is conducted in a low-nitrogen 
environment (i.e., an environment sufficiently low in nitrogen that the 
zirconium can not pick up significant additional nitrogen, e.g. 1-2 ppm). 
Preferably the low-nitrogen environment is a vacuum or argon atmosphere. 
In some embodiments, the plurality of processing steps will include at 
least one step of conditioning the surface of the cladding such that an 
outer layer of material is removed. Such processing steps are well known 
to those of skill in art and include, for example, honing, machining with 
a lathe, chemical etching and mechanical polishing. Such surface 
conditioning steps, serve to remove the material that is most likely to 
have taken up nitrogen during processing. Thus, the underlying 
nitrogen-free regions form the outer surface of the cladding. 
These and other features of the present invention will be presented in more 
detail in the following specification of the invention and the figures.

DESCRIPTION OF THE PREFERRED EMBODIMENTS 
I. THE TUBING STRUCTURE 
As used herein, the term "tubing" refers to a metal tube having various 
uses, and the term "fuel rod container" or simply "container" refers to 
tubing used in fuel rods to enclose fuel pellets. Sometimes the fuel rod 
container is referred to as "cladding" or "cladding tube". The container 
will have an associated thickness or cross-section formed from a 
zirconium-based alloy of this invention. 
Referring to FIG. 1, a fuel element 14 (commonly referred to as a fuel rod) 
includes a fuel rod container 17 surrounding a fuel material core 16. The 
fuel element 14 is designed to provide excellent thermal contact between 
the fuel rod container 17 and the fuel material core 16, a minimum of 
parasitic neutron absorption, and resistance to bowing and vibration which 
are occasionally caused by flow of coolant at high velocity. The fuel 
material core is typically a plurality of fuel pellets of fissionable 
and/or fertile material. The fuel core may have various shapes, such as 
cylindrical pellets, spheres, or small particles. Various nuclear fuels 
may be used, including uranium compounds, thorium compounds and mixtures 
thereof. A preferred fuel is uranium dioxide or a mixture comprising 
uranium dioxide and plutonium dioxide. 
The container 17 is a composite cladding having a structure including a 
zirconium alloy substrate 21 and a zirconium barrier 22. In alternative 
embodiments, the cladding also includes an inner layer or liner (not 
shown) metallurgically bonded to the inner surface of the zirconium 
barrier. In other alternative embodiments, the container will contain only 
substrate 21 and not the zirconium barrier layer. The substrate will have 
an outer circumferential region and an inner circumferential region, with 
the zirconium barrier metallurgically bonded to the inner circumferential 
region. 
The substrate may be made from a low-nitrogen version of a zirconium alloy 
employed in conventional cladding. Most generally, any zirconium alloy may 
be employed that contains alloying elements in sufficient concentration to 
form precipitates while retaining the strength and ductility necessary in 
fuel cladding tubes. Suitable zirconium alloys for the substrate 
preferably include at least about 98% zirconium, up to about 0.25% iron, 
up to about 0.1% nickel, up to about 0.15% chromium, and up to about 1.7% 
tin (all percents by weight). In a preferred embodiment of this invention, 
the substrate is a low-nitrogen version of Zircaloy-2 or Zircaloy-4. As 
will be explained below, it is often desirable that the cladding have 
relatively low concentrations (in comparison to the Zircaloys) of some 
alloying elements, most notably iron, nickel, and chromium. 
Preferably, the zirconium-alloys of this invention will have no more than 
about 50 ppm nitrogen, more preferably no more than about 30 ppm nitrogen, 
and most preferably no more than about 20 ppm nitrogen. While not wishing 
to be bound by theory, it is believed that the presence of nitrogen raises 
the activity coefficient of alloying elements such as iron and nickel 
thereby reducing their concentration in the matrix. The activity of an 
alloying element (or any chemical species) is the product of its 
concentration and activity coefficient. Thus, when nitrogen is present in 
a significant concentration (e.g., greater than about 20 ppm), it is 
believed that the concentration of alloying elements dissolved in the 
matrix decreases and the alloy becomes more susceptible to nodular 
corrosion. 
Unfortunately, when the concentration of alloying elements (nickel and iron 
in particular) becomes too great, uniform corrosion can become a 
significant problem at high bum-up. Because the low-nitrogen cladding of 
this invention exhibits increased resistance to nodular corrosion, it is 
now possible to employ cladding having lower concentrations of the 
alloying elements. Thus, resistance to both nodular and uniform corrosion 
is improved. In preferred embodiments, the alloying element concentrations 
are provided in a dilution-factor range of 0.3 to 0.5 (of the 
concentrations employed in conventional Zircaloys). Thus, preferred alloys 
will have the following concentrations (on a per weight basis): 0.05-0.09 
percent iron, 0.03-0.05 percent chromium, and 0.02-0.04 percent nickel. 
Although any alloy having alloying elements within these ranges is 
suitable, especially preferred alloys will have iron: chromium: nickel in 
the ratio of 3:2:1. This should provide an alloy having the precipitates 
Zr(Fe,Cr).sub.2 and Zr.sub.2 (Fe,Ni) in approximately equal 
concentrations. 
In some preferred embodiments, the substrate will have a microstructure 
(i.e. precipitate size distribution) that resists corrosion and/or crack 
propagation. It is known that the microstructure of Zircaloys and other 
alloys can be controlled by the anneal temperature and time as well as 
other fabrication parameters. It is also known that in boiling water 
reactors (BWRs), smaller precipitates generally provide superior 
resistance to corrosion while in pressurized water reactors (PWRs), larger 
precipitates generally provide superior resistance to corrosion. In either 
environment, coarse precipitates provide improved resistance to axial 
crack propagation. In a preferred embodiment, the substrate will have a 
distribution of coarse precipitates (e.g., greater than about 0.2 
micrometers in diameter and preferably between about 0.2 and 1 micrometers 
in diameter) in the substrate. This will provide significant resistance to 
crack propagation in the axial direction. 
In an alternative embodiment, a dense distribution of fine precipitate 
(e.g., between about 0.01 and 0.15 micrometers in diameter) is provided in 
the outer region (radially) of the substrate and a less dense distribution 
of coarse precipitates (e.g., between about 0.2 and 1 micrometers in 
diameter) in the inner region of the substrate. This embodiment will be 
especially preferred in BWRs. In PWRs, preferred substrates will have 
coarse precipitates distributed throughout. Detailed discussions of 
Zircaloy microstructure and methods of fabricating cladding having a 
desired microstructure are found in U.S. patent application Ser. Nos. 
08/052,793 and 08/052,791, both of which were previously incorporated by 
reference. 
Metallurgically bonded on the inside surface of substrate 21 is the 
zirconium barrier 22. See U.S. Pat. Nos. 4,200,492, and 4,372,817, to 
Armijo and Coffin, U.S. Pat. No. 4,610,842, to Vannesjo, and U.S. Pat. No. 
4,894,203, to Adamson, each of which is incorporated herein by reference 
for all purposes. The barrier shields the substrate from the nuclear fuel 
material inside the composite cladding. Fuel pellet-induced stress may be 
introduced by, for example, swelling of the pellets at reactor operating 
temperatures so that the pellet presses against the cladding. In effect, 
the zirconium barrier deforms plastically to relieve pellet-induced 
stresses in the fuel element during swelling. The barrier also serves to 
inhibit stress corrosion cracking and protects the cladding from contact 
and reaction with impurities and fission products. The zirconium barrier 
maintains low yield strength, low hardness, and other desirable structural 
properties even after prolonged use because it is resistant to radiation 
hardening. 
In preferred embodiments, the thickness of the barrier layer is between 
about 50 and 130 micrometers (approximately 2 to 5 mils) and more 
preferably between about 75 and 115 micrometers (approximately 3.2 to 4.7 
mils). In a typical cladding, the zirconium barrier forms between about 5% 
to about 30% of the thickness or cross-section of the cladding. 
Generally, the zirconium barrier layer may be made from unalloyed zirconium 
possessing the desired structural properties. Suitable barrier layers are 
made from "low oxygen sponge" grade zirconium, "reactor grade sponge" 
zirconium, and higher purity "crystal bar zirconium". In alternative 
embodiments, the barrier layer is alloyed with low concentrations of 
alloying elements such as the chromium, nickel, and iron used in the 
substrate. The alloying elements and the concentrations at which they 
appear should be chosen to impart additional corrosion resistance to the 
barrier layer while maintaining compliance sufficient to prevent damage 
from pellet-cladding interaction. 
Referring now to FIG. 2, a cutaway sectional view of a nuclear fuel bundle 
or assembly 10 is shown. The fuel bundle is a discrete unit of fuel 
containing many individual sealed fuel elements or rods R each containing 
a cladding tube of this invention. In addition, the fuel bundle consists 
of a flow channel C provided at its upper end with an upper lifting bale 
12 and at its lower end with a nose piece L and lower lifting bale 11. The 
upper end of channel C is open at 13 and the lower end of the nose piece 
is provided with coolant flow openings. The array of fuel elements or rods 
R is enclosed in channel C and supported therein by means of upper tie 
plate U and lower tie plate (not shown). Certain of the fuel rods serving 
to "tie" the tie plates together--thus frequently being called "tie rods" 
(not shown). In addition, one or more spacers S may be disposed within the 
flow channel to hold the fuel elements in alignment with one another and 
the flow channel. During the in service life of the fuel bundle, the 
liquid coolant ordinarily enters through the openings in the lower end of 
the nose piece, passes upwardly around fuel elements R, and discharges at 
upper outlet 13 in partially vaporized condition. 
Referring now to FIG. 3, the fuel elements or rods R are sealed at their 
ends by end plugs 18 welded to the fuel rod container 17, which may 
include studs 19 to facilitate the mounting of the fuel element in the 
fuel assembly. A void space or plenum 20 is provided at one end of the 
element to permit longitudinal expansion of the fuel material 16 and 
accumulation of gases released by the fuel material. A getter (not shown) 
is typically employed to remove various deleterious gases and other 
products of the fission reaction. A nuclear fuel material retainer 24 in 
the form of a helical member is positioned within space 20 to provide 
restraint against axial movement of the pellet column during handling and 
transportation of the fuel element. 
II. MANUFACTURE OF THE TUBING 
The low-nitrogen cladding tubes of this invention may be formed by various 
conventional processes with only minor modification. Most importantly, the 
process steps should be conducted so that exposure to nitrogen is limited, 
particularly in those steps where cladding is most susceptible to nitrogen 
diffusion. First, in preferred methods, the zirconium alloy starting 
material used to form the cladding will have a low bulk concentration of 
nitrogen, preferably less than about 20 ppm. Zircaloy ingots having this 
concentration of nitrogen are available from Teledyne Wahchang (Albany, 
Oreg.), Western Zirconium (Ogden, Utah) and Cezus (France). 
Next, the process steps--particularly those conducted at temperatures at or 
above about 500.degree. C.--are performed in a low-nitrogen environment. 
Typically, an inert atmosphere such as a vacuum or an argon atmosphere is 
used for this purpose. Suitable vacuum annealing furnaces are available 
from Centorr Vacuum Industries of Nashua, N.H. 
Finally, if it appears that some nitrogen may have penetrated a short 
distance through the surface, a chemical or mechanical surface 
conditioning step such as etching can be employed to remove any nitrogen 
that might have entered the Zircaloy. Chemical and mechanical surface 
conditioning steps are currently employed in cladding fabrication. These 
include honing, grinding, sanding, machining with a lathe, buffing, 
chemical etching, and chemical mechanical polishing. 
As noted, in preferred embodiments, the zirconium alloy matrix will contain 
a distribution of relatively coarse precipitates. The size of the 
precipitate can be controlled by various manufacturing processes. 
Initially, the precipitate size is governed essentially by the cooling or 
quenching rate from the beta phase. The beta phase refers to the 
body-centered cubic crystal lattice structure of crystalline zirconium and 
Zircaloy that is stable at higher temperatures (i.e., above about 
960.degree. C. for Zircaloy-2). A different phase, the alpha phase, is a 
close-packed hexagonal crystal lattice structure of zirconium and Zircaloy 
that is stable at lower temperatures. Between about 825.degree. C. and 
960.degree. C., the alpha and beta phases can coexist in Zircaloys. Rapid 
quenching rates from the beta phase (e.g. faster than about 50.degree. C. 
per second) give smaller precipitates, while slower cooling rates give 
larger precipitates. The initial precipitate sizes (obtained by quenching 
from the beta phase) can be altered somewhat by later heat treatments such 
as annealing at a high temperature within the alpha phase field (e.g. 
between about 600.degree.-825.degree. C.). This allows the smaller 
precipitates to dissolve and some of the nickel, iron and chromium 
components of the Zircaloy matrix phase to diffuse to larger precipitates, 
causing the precipitates to coarsen. A guideline widely applicable to 
various processes is provided by the "accumulated normalized annealing 
time" defined in F. Garzarolli, et at., "Progress in the Knowledge of 
Nodular Corrosion", Zirconium in the Nuclear Industry, ASTM STP939, pp. 
417-430 (1987), which is incorporated herein by reference for all 
purposes. The accumulated normalized annealing time includes contributions 
from the duration, temperature, and number of annealing steps employed in 
the complete process. Generally, longer and higher temperature anneals 
will produce coarser precipitates. Preferably, to ensure sufficiently 
coarse precipitates, the accumulated normalized annealing time should be 
greater than about 10.sup.-17 hours. 
To obtain the final tubing of the necessary dimensions, various other 
manufacturing steps such as cold-working, extruding, heat treating, and 
annealing may be employed. The equipment and operating conditions 
necessary to carry out these various steps will be readily apparent to 
those of skill in the art, and are described in U.S. patent application 
Ser. No. 08/052,791, entitled METHOD OF FABRICATING ZIRCALOY TUBING HAVING 
HIGH RESISTANCE TO CRACK PROPAGATION and previously incorporated herein by 
reference. 
In an exemplary embodiment, a hollow billet of zirconium alloy having a 
nitrogen concentration at or below about 20 ppm is beta quenched from 
1000.degree. C. to about 700.degree. C. by immersion in a tank of water. 
Next, the tube is extruded with the tube temperature being at about 
570.degree. C. by putting the tube through a set of tapered dies under 
high pressure. The extruded product is referred to as a "tubeshell" which 
is available in specified dimensions (and nitrogen levels of 20 ppm) from 
various vendors such as Teledyne Wahchang (Albany, Oreg. USA), Western 
Zirconium (A Westinghouse company of Ogden, Utah), and Cezus (France). 
Next, a first pass cold work to 70% is performed as in known processes 
followed by an anneal at a relatively high temperature (e.g. 650.degree. 
C. for four hours in a vacuum). Next, a second pass cold work to 70% is 
performed followed by annealing at 650.degree. C. for 2 hours (also in a 
vacuum). A third pass cold work and a recrystallization or stress relief 
anneal are performed under the same conditions as the known processes. At 
this point the tubing is suitable (with only minor modifications and 
testing) for use in a fuel rod. 
If the cladding is to have a gradient in precipitate size (with smaller 
precipitates near the outer regions and larger precipitates for inner 
regions), a heat treatment is performed to regain the smaller precipitates 
on the outside of the tube. This treatment is performed at about 
1045.degree. C. in the pure beta phase (although it could be performed in 
the alpha plus beta phase). An induction coil rapidly heats the outer 15% 
of the tube to the desired temperature and then shuts off while water (or 
other cooling fluid) is flowing through the tube interior. This allows the 
tube to rapidly cool (sometimes within a matter of 2 seconds). The 
penetration of the induction coil energy can be tuned by adjusting the 
induction coil frequency, the induction coil energy, the speed at which 
the tube moves through the induction coil, and the water temperature (flow 
rate). One of skill in the art will appreciate how to adjust these 
conditions to achieve the type of heat treatment that will form small 
precipitates at the outer 15% of the tube. Further details can be found in 
U.S. Pat. No. 4,576,654, to Eddens. The resulting tube will have good 
nodular corrosion resistance while retaining coarse precipitates in the 
inner regions. 
If the tubing of this invention is to contain a barrier layer, the 
fabrication process will have at least one additional step. Details of 
such a step are known in the art and are provided in, for example, U.S. 
Pat. No. 4,894,203. Usually, the barrier layer is bonded to the tubing as 
liner during an extrusion step. Other steps in the process are performed 
as described above. 
III. EXAMPLE 
FIG. 4 is a graph showing that the severity of nodular corrosion increases 
as nitrogen concentration increases in Zircaloy-2. More specifically, the 
number of nodules on a radial cladding surface increases dramatically as 
the nitrogen content in the Zircaloy increases. In this example, 0.40 inch 
length sections of Zircaloy-2 tubing were annealed for 9 to 81 hours at 
750.degree. C. The sections were exposed initially to a mixture containing 
equal concentrations of high purity nitrogen and argon gases for 0.25 to 
12 hours. Only argon flowed through the annealing furnace for the balance 
of the time at 750.degree. C. The total annealing time for most of the 
specimens was 24 hours. 
This treatment allowed the nitrogen to partially--but not 
completely--diffuse through the Zircaloy sections. Thus, the nitrogen 
concentration most likely was greatest near the surfaces of the sections 
and dropped off progressively toward the centers. The total amount of 
nitrogen picked up by each section was measured gravimetrically. As would 
be expected, those sections exposed to the argon-nitrogen mixture for the 
longest time picked up the most nitrogen. Before exposure to corrosive 
conditions, the cladding sections were etched briefly to remove the 
exposed regions of the sections. The specimens were etched for 60 seconds 
in a solution containing 10:9:1 parts by volume of water, 70% nitric acid, 
and 50% hydrofluoric acid. This procedure removed about 0.6 mils (or 15 
micrometers) from the exposed surface, and certainly removed some of the 
nitrogen containing material. Obviously, it did not remove all the 
nitrogen, as evidenced by the data showing progressively less resistance 
to corrosion as exposure to nitrogen increased. 
To study the effect of nitrogen concentration on nodular corrosion, the 
sections were all subjected to the same corrosion conditions: exposure to 
steam at 510.degree. C. and 1500 PSIG for 24 hours. The resulting sections 
were then examined and the number of nodules counted. As can be seen from 
FIG. 4, the number of nodules ranged from zero for low nitrogen alloys to 
near 100 for high nitrogen alloys. Also, the number of nodules on the 
sections remained at or near zero from a nitrogen pick-up of 20 ppm to a 
nitrogen pick-up of 60 ppm. 
IV. CONCLUSION 
Although the foregoing invention has been described in some detail for 
purposes of clarity of understanding, it will be apparent that certain 
changes and modifications may be practiced within the scope of the 
appended claims. For instance, although the specification has described 
preferred zirconium alloy tubes, other shapes may be used as well. For 
example, plates and metal sections of other shapes may also be used. In 
addition, the reader will understand that the metallurgy herein can be 
used in reactor pans other than cladding. For example, the Zirconium alloy 
composition here taught may be used with water rods, spacers, channels and 
other Zirconium alloy structures and their equivalent within the reactor.