Note: Descriptions are shown in the official language in which they were submitted.
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BU~IED ZIRCONI~M LAYER
Field of the Inventi'on
This invention relates broadly to an improvement
in nuclear fuel eIements for use in the core of nuclear
fission reactors and, more particularly, to an improved
nuclear fuel element having a composite cladding container
having a zirconium alloy substrate, a barrier of non-
alloyed zirconium metallurgically bonded to the inside
surface of the substrate and an inner zirconium alloy
layer metallurgically bonded to the zirconium barrier.
Background of the I'nvention
Nuclear reactors are presently being designed
constructed, and operated in which the nuclear fuel is
contained in fuel elements which can have various
geometric shapes, such as plates, tubes, or rods. The
fuel material is usually enclosed in a corrosion-
resistant, non-reactive, heat conductive container or
cladding. The fuel elements are assembled together in
a lattice at fixed distances from each other in a
2~ coolant flow channeI or region forming a fuel assembly
and sufficient fuel assemblies are combined to form
the nuclear fission chain reacting assembly or reactor
core capable of a self-sustained -fission reaction.
The core, in turn, is enclosed with a reactor
~essel through ~hich a coolant is passed.
The cladding serves several purposes and two
primary purposes are: first, to prevent contact and
chemical reactions between the nuclear fuel and the
-' coolant or the moderator if a moderator is present, or
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both if both the coolant and the moderator are present;
and second, to prevent the radioactive fission products,
some of which are gases, from being reIeased from the
fuel into the coolant or the moderator, or both if both
the coolant and the moderator are present. Common
cladding materials are stainless steel, aluminum and its
alloys, zirconium and its alloys, niobium (columbium),
certain magnesium alloys, and others. The failure of
the cladding, i.e, a loss of the leak tightness, can
contaminate the coolant or moderator and the associated
systems with radioactive long-lived products to a degree
which interferes with plant operation.
Problems have been encountered in the manufacture
and in the operation of nuclear fuel elements which
employ certain metals and alloys as the clad material
due to mechanical or chemical reactions of these cladding
materials under certain circumstancesO Zirconium and its
alloys, under normal circumstances~ are excellent nuclear
fuel claddings since they have low neutron absorption
cross-sections and at temperatures below about 750F
(about 398C~ are strong, ductile~ extremely stable and
non-reactive in the presence of demineralized water or
steam which are commonly used as reactor coolants and
moderators.
Howe~er, fuel element performance has re~ealed
a problem with the brittle splitting of the cladding due
to the combined interactions between the nuclear fuel,
the cladding and the fission products produced duriny
nuclear fission reactions. It has been discovered that
-this undesirable performance is promoted by localized
mechanical stresses due to fueI cladding differential
expansion (stresses in the cladding are localized at
fuel pellet interfaces and sometimes at cracks in the
nuclear fuel). The phenomenon is defined by the terms
pellet-cladding-interaction or PCI. Corrosive fission
products are released from the nuclear fuel and are
~ 7Z~ 24-NT-04471
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present at the intersection of the fuel pellet interfaces
with the cladding surface. Such fission products are
created in the nuclear ~uel during the fission chain
reaction during operation of a nuclear reactor.
~ithin the confines of a sealed fuel element,
hydrogen gas can be generated by the slow reaction
between the cladding and residual water inside the
cladding. This hydrogen gas may build up to levels
which, under certain conditions, can result in localized
hydriding of the cladding with concurrent local
deterioration in the mechanical properties of the
cladding. The cladding is also adversely affected by
such gases as oxygen, nitrogen, carbon monoxide, and
carbon dioxide over a wide range of temperatures. The
zirconium cladding of a nuclear fuel element is exposed
to one or more of the gases listed above and fission
products during irradiation in a nuclear reactor and
this occurs in spite of the fact that these gases may
not be present in the reactor coolant or moderator, and
further may have been excluded as far as possible from
the ambient atmosphere during manufacture of the cladding
and the fuel element. Sintered refractory and ceramic
compositions, such as uranium dioxide and other
compositions used as nuclear fuel, release measurable
quantities of the aforementioned gases upon heating, such
as during fuel element manufacture, and further release
fission products during irradiation. Particulate
refractory and ceramic compositions/ such as uranium
dioxide powder and other powders used as nuclear fuel,
have been known to release even larger quantities of the
aforementioned gases during irradiation. These released
gases are capable of reacting with the zirconium cladding
containing the~nuclear fuel.
Thus, in light of the foregoing, it has been
found desirable to minimize attack of the cladding from
water, water vapor and other gases, especially hydrogen,
12~7~:~
24-NT-04471
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reactive with the cladding from inside the fuel element
throughout the time the fuel element is used in the
operation of nuclear power plants. One such approach
has been to find materials which will chemically react
rapidly with the water, water vapor and other gases to
eliminate these from the interior of the cladding. Such
materials are called getters.
Another approach has been to coat the nuclear
fuel material with a ceramic to prevent moisture coming
in contact with the nuclear fuel material as disclosed
in U.S. Patent No, 3~108r936/ issued October 29, 1963
to Gale. U.S. Patent No. 3,085~059, issued April 9, 1963
to Burnham describes a fuel element including a metal
casing containing one or more pellets of fissionable
ceramic material and a layer of vitreous material
bonded to the ceramic pellets so that the layer is
between the casing and the nuclear fuel to assure
uniformly good heat conduction Erom the pellets to
the casing. U.S. Patent No. 2,873~238, issued
20 February 10, 1959 to OhIinger et al, describes
jacketed fissionable slugs of uranium canned in a
metal case in which the protective jackets or coverings
for the slugs are a zinc-aluminum bonding layer. U.S.
Patent No. 2,849,387, issued August 26, 1958 to
Brugmann~ discloses a jacketed fissionable body
comprising a plurality of open-ended jacketed body
sections of nuclear fuel which have been dipped into
a molten bath of a bonding material giving an
e~fective thermally conductive bond between the uranium
3Q body sections and the container (or cladding). The
coating is disclosed as any metal alloy having good
thermal conduction properties with examples including
aluminum-silicon and zinc-aluminum alloys. Japanese
Patent Publication No. SF~O 47-46559 dated November 24,
1972, disclosed consolidating discrete nuclear fuel
particles into a carbon-containing matrix fuel composite
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by coating the fuel particles with a high density, smooth
carbon-containing coating around the pellets. Still another
coating disclosure is Japanese Patent Publicatlon No.
SHO 14200 in which the coating of one of two groups of
pellets is with a layer of silicon carbide and the other group
is coated with a layer of pyrocarbon or metal carbide.
The coating of nuclear fuel material introduces
reliability problems in that achieving uniform coatings
free of faults is difficult. Further, the deterioration
of the coating can introduce problems with the long-lived
performance of the nuclear fuel material.
One known method for preventing corrosion of
nuclear fuel cladding consists of the addition of a metal
such as niobium to the fuel. The additive can be in the
form of a powder, provided the subsequent fuel processing
operation does not oxidize the metal, or the additive can
be incorporated into the fuel element as wires, sheets,
or other forms in, around or between fuel pellets.
General Electric Atomic Power Document 4555,
20 dated February 1964,at GE NEBO Library 175 Curtner Avenue,
San Jose, Calif. 95125, discloses a composite cladding of a
zirconium alloy with an inner lining of stainless steel
metallurgically bonded to the zirconium alloy, and the
composite cladding is fabricated by use of extrusion of a
hollow billet of the zirconium alloy having an inner lining
of stainless steel. This cladding has the disadvantage that
the stainless steel develops brittle phases, and the
stainless steel layer involves a neutron absorption
penalty of about ten to fifteen times the penalty for
a zirconium alloy layer of the same thickness.
U.S. Patent No. 3,502,549, issued March 24,
1970 to Charveriat, discloses a method fvr protecting
zirconium and its alloys by the electrolytic deposition
of chromium to provide a composite material useful for
nuclear reactors. A method for electrolytic deposition
of copper on Zircaloy-2 surfaces and subsequent heat
~ 7~7 2~-NT-0~471
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treatment for the purpose of obtaining surface diffusion
of the elctrolytieally deposited metal is presented in
Energia Nucleare, Volume 11, No. 9 (September, 1964) at
.
pages 505-508. In Stability and Compatibility of' Hydrogen
Barriers Applied to Zirconium All'oys, by ~. ~rossa et al
(European Atomic Energy Community, Joint Nuelear Research
Center, EUR 4098e~ 1969), methods of deposition of
different coatings and their efficiency as hydrogen
diffusion barriers are described along with an Al-Si
coating as the most promising barrier against hydrogen
diffusion. Methods for electroplating nickel on
zireonium and zirconium tin alloys and heat treating
these alloys to produce alloy-diffusion bonds are disclosed
in Electro;plating-on Zirconium and Zirconium-Tin, by W.C.
Schnickner et al (BMI-757, Technical Information Service,
1952). U.S. Patent No. 3,625,821, issued December 7, 1971
to Ricks, presents a fuel element for a nuclear reactor
having a duel cladding tube with the inner surface of the
tube being coated with a metal of low neutron capture
2Q cross-section such as nickel and having finely-dispersed
particles of a burnable poison disposed therein. Reactor
Development Program Proc'es's Report of August, 1973
(ANL-RDP-l9) discloses a chemical getter arrangement of
a sacrificial layer of chronium on the inner surface of
a stainless steel cladding.
Another approach has been to introduce a barrier
between the nuclear fuel material and the cladding holding
the nuclear fuel material as disclosed in U.S. Patent No.
3,320,150, issued January 18, 1966 to Martin et al
30 (copper foil), German Patent Publication DAS 1,238,115
(titanium layer), U.S. Patent No. 3,212,988, issued
October 19, 1965 to Ringott et al (sheath of zirconium~
aluminum or beryllium), U.S. Patent No. 3~018,238, issued
January 23, 1962 to Layer et al(barrier of crystalline
earbon between the UO2 and the zirconium cladding), and
U.S. Patent No. 3,088,893, issued May 7t 1963 to
Spalaris (stainless steel foil). While the barrier concept
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2~-NT-Q4471
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proves promising, some of the foregoing references invol~e
incompatible materials with~either the nuclear fuel (e.g.
carbon can combine with oxygen from the nuclear fuel), or
the cladding (e.g., copper and other metals can react with
the cladding, altering the properties of the cladding),
or the nuclear fission reaction ~e.g., by acting as
neutron absorbers). None of the listed references
disclose solutions -to the recently discovered problem
of localized chemical-mechanical interactions between
the nuclear fuel and the cladding.
Further approaches to the barrier concept are
disclosed in U.S. Pat~nt No. 3,969,186~ issued
July 13, 1976 to Thompson et al (refractory metal such
as molybdenum, tungsten, rhenium, niobium and alloys
thereof in the form of a tube of foil of single or
multiple layers or a coating on the internal surface of
the cladding), and U.S. Patent No. 3,925,151, issued
December 9, 1975 to Klepfer (Iiner of zirconium, niobium,
or alloys thereof between the nuclear fuel and the
cladding with a coating of a high lubricity material
between liner and the cladding).
U.S. Patent No. 4,045,288, issued
August 30, 1977 to Armijo discloses a compo~ite cladding
of a zirconium alloy substrate with a metal barrier
metallurgically bonded to the substrate and an inner layer
of zirconium alloy metallurgically bonded to the metal
barrier. The barrier is selected from a group of niobium,
aluminum, copper7 nickel~ stainless steel, and iron.
With the exception of the niobium barrier, all the other
materials will form low melting eutectic phases with the
zirconium alloy substrate ! making them undesirable in
postulated loss-of-coolant accidents.
U~S. Patent No. 4,200,492, issued April 29, 1980
to Armijo et al discloses a composite cladding of a
zirconium alloy substrate with an unalloyed zirconium liner.
The soft zirconium l~ner minimizes loaallzéd strain, and
24-NT-04471
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reduces stress corrosion cracking and liquid metal
embrittlement, but is subject to damage and losses due
to honing and the like during fabrication and to corxosion
in the event that the cladding is breached.
Accordingly, it has remained desirable to
develop nuclear fuel elements minimizing the problems
discussed above.
Summary of the Invention
A particularly effectlve nuclear fuel element
for use in the core of a nuclear reactor has a composite
cladding having a substrate, a non-alloyed zirconium
barrier metallurgically bonded to the inside surface
of the substrate, and an inner layer metallurgically
bonded to the inside surface of the zirconium barrier.
The substrate of the cladding is completely unchanged
in design and function from previous practice for a
nuclear reactor and is selected from conventional
cladding materials such as zirconium alloys.
The zirconium barrier and the inner layer form a
shield between the substrate and the nuclear fuel material
held in the cladding, as well as shielding the substrate
from fission products and gases. The inner layer in turn
shields the zirconium barrier from fission products
released from the fuel and other reactive elements present
in the fuel element. This sheilding allows the zirconium
barrier to retain a maximum amount of purity and ductility
by preventing hardening by recoiliny fission products or
hy reaction with chemical elements present in the Euel
element.
The zirconium barrier forms about 1 to about 30
percent of the thickness of the cladding. A zirconium
barrier forming less than about 1 percent of the thickness
of the cladding would be difficult to achieve in commercial
production, and a zirconium barrier forming more than 30
percent of the thickness of the cladding would provide no
additional benefit for the added thickness. Further, a
2~-NT-0~471
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barrier more than about 30 percent of the thickness of the
cladding would produce a concomitant reduction in thickness
of the substrate and weakening of the composite cladding.
The inner layer may be fabricated to constitute
from 1 percent to 10 percent of the total cladding
thickness. This thickness range has been specified to
provide an inner layer of minimum thickness fabricable
by tubing co-extrusion and co-reduction techniques. Because
of its purity and the shielding effect of the inner layer,
the barrier remains soft during irradiation and minimizes
localized strain inside the nuclear fuel element, thu~
serving to protect the substrate from stress corrosion
cracking or liquid metal embrittlement. The inner la~er
and the zirconium barrier provides a preferential reaction
site for reaction with volatile impurities or fission
products present inside the nuclear fuel element and, in
this manner, serve to protect the barrier and the cladding
from attack by the volatile impurities or fission products.
In addition, the inner layer is useful during
fabrication to prevent losses or damage to the soft barrier
and thus improves fabricability. Further, the inner layer
protects the barrier from aqueous corrosion in the event
of fuel element failure.
This invention has a striking advantage that the
substrate of the cladding and the barrier are protected
from s-tress corrosion cracking and liquid metal embrittlement,
in addition to contact with fission products, corrosive gases,
etc., by the inner layer which does not introduce an
appreciable neutron capture penalties, heat transfer
penalties, or materials incompatibility problems.
Objects of the Invention
-
It ls an object of this invention to provide a
nuclear uel element capable of operating in nuclear
reactors for extended periods of time without the
oCcurrence of splitting of the cladding/ corrosion of
; the cladding, or other fuel failure problems.
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It is another object o~ this invention to
provide a nuclear fuel element with a composite cladding
having a substrate, a zirconium barrier metallurgically
bonded to the inside surface of the substrate, and an
inner layer metallurgically bonded to the inside surface
of the zirconium barrier so that the metallurgical bonds
provide a long-lived connection between the substrate
and the zirconium barrier and between the zirconium
barrier and the inner layer.
The foregoing and other objects of this invention
will become apparent to persons skilled in the art from
reading the following specification and the drawings
described immediately hereinafter.
Description of the Drawing
Figure 1 is a partial cutaway sectional view of
a nuclear fuel assembly containing nuclear fuel elements
constructed according to the teaching of this invention
and
Figure 2 is an enlarged trans~erse cross-
sectional view of the nuclear fuel element in Figure 2illustrating the teaching of this invention.
Description of the Invention
Referring now more particularly to Figure 1, there
is shown a partially cutaway sectional view of a nuclear
fuel assembly 10. This fuel assembly 10 consists of a
tubular flow channel 11 of generally-square cross-section
provided at its upper end with a lifting bail 12 and at
its lower end with a nose piece ~not shown due to the
lower portion of the 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 the
channeI 11 and supported therein by means of an 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
t
24-NT-04471
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14, and discharges through the upper outlet 13 at an
elevated temperature in a partially vaporized condition
for boiling reactors in an unvaporized condition for
pressurized reac-tors.
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 assemhly. 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 gaes 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
f 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.
A nuclear fuel element or rod 14 constructed
according to the teachings of this invention is shown in
a partial section in Figure 1. The fuel element includes
a core or central cylindrical portion of nuclear fuel
material 16, here shown as a plurality of fuel pellets
of fissionable and/or fertile material positioned within
a structural cladding or container 17. In some cases,
the fuel pellets may be of various shapes such as
cylindrical pellets or spheres and, in other cases,
different fuel forms such as a particulate fuel may be
used. The physical form of the fuel is immaterial to
this invention. Various nuclear fuel materials may be
used including uranium compounds, plutonium compounds,
thorium compounds, and mixtures thereof. A preferred
fuel is uranium dioxide or a mixture comprising uranium
dioxide and plutonium dioxide.
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24-NT-04471
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Referring now to Figure 2, the nuelear ~uel
material 16 forming the central eore of the fuel element
14 is surrounded by a cladding 17 which, in this
invention, is also referred to as a composite cladding.
The composite cladding container encloses the fissile
core so as to leave a gap 24 between the eore and the
eladding container during use in a nuclear reactor. The
composite eladding has an external substrate 21 selected
for conventional fuel eladding materials and, in a
preferred embodiment of this invention, the substrate is
a zireonium alloy sueh as Zirealoy-2 or Zirealoy-4.
The substrate 21 has metallurgieally bonded
on the inside eircumference thereof an unalloyed
æirconium barrier 22 so that the zireonium barrier
forms a shield o~ the substrate from the nuelear fuel
material 16 inside the composite cladding. The
zirconium barrier preferably forms about 1 percent of
about 30 percent of the thiekness of the eomposite
eladding.
The zireonium barrier 22 has metallurgieally
bonded on the inside eireumference thereof an inner layer
23 so that the inner layer is the portion of the composite
cladding cosest to the nuelear fuel material 16. The
inner layer preferably forms about 1 percent to about
10 percent of the thickness of the cladding and is
comprised of conventional cladding materials, and in a
preferred embodiment of this invention, the substrate
is a zirconium alloy such as Zircaloy-2 or Zircaloy-4.
The zirconium barrier serves as a reaetion site for
gaseous impurities and fission products whieh have
penetrated through craeks or defects in the inner layer
23 and protects the substrate portion oF the cladding from
contact and reaction with such impurities and fission
products, and minimizes the oeeurrenee of localized stress
and cladding failure by pellet-eladding-interaetion.
~2~ 7Z~ 24-NT-04~71
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In an exemplary embodiment, the zirconium
barrier layer is about three mils thick and the inner
layer of Zircaloy-2 is about one mil thick. Each of
the inner layers and barrier layers should be continuous,
that is be free of perforations or seams.
The composite cladding o~ the nuclear fuel
element of this invention has an unalloyed zirconium
barrier metallurgically bonded to the substrate and an
inner layer metallurgically bonded to the zirconium
barrier. ~etallographic examination shows that there
is sufficient cross-diffusion between the substrate and
the zirconium barrier and between the zirconium barrier
and the inner layer to form metallurgical bonds, but
insufficient cross diffusion to significantly reduce the
purity of the zirconium barrier itself. Also, from
Figure 2, it is apparent that the zirconium barrier could
be termed a "buried" zirconium barrier since it is
sandwiched between the substrate and inner layer.
Non alloyed zirconium forming the barrier in the
composite cladding is highIy resistant to radiation
hardening and this enables the zirconium barrier, after
prolonged irradiationl to maintain desirable structural
properties such as yield strength and hardness at levels
considerably lower than those of conventional zirconium
alloys. In effect, the zirconium barrier does not harden
as much as conventional zirconium alloys when subjected
to irradiation and this, together with its initially low
yield strength, enables the zirconium barrier to deform
plastically and relieve pellet-induced stresses in the
fuel elemeIlt duxing power transients. Pellet-induced
stresses in the fuel element can be brought about, for
example, by thermal expansion and/or swelling of the
pellets of nuclear fuel at re~ctor operating temperatures
(300 C to 350 C) so that the pellet comes into contact
with the cladding.
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It has ~urther been discovered -that a zirconium
barrier of the order of preferably about 5 percent to
15 percent of the thickness of the cladding and a
particularly preferred thickness of 10 percent of the
cladding bonded to the external substrate of zirconium
alloy provides stress reduction and a barrier effect
sufficient to prevent failures in the composite cladding.
A preferred embodiment according to the
principles of this invention comprises "low oxygen sponge"
grade zirconium as the buried barrier layer, although
higher purity "crystal bar zirconiuml' grade and lower
purity "reactor grade sponge" zirconium are also possible.
The residual impurity content of the sponge zirconium
serves to impart special properties to the zirconium
barrier. Generally, there are at least about 1000 parts
ppm impurities in sponge zirconium and preferably less
than 4200 ppm. Oxygen is preferably kept within the
range of about 200 to about 1200 ppm. Other typical
impurity levels are listed as follows: aluminum -
75 ppm or less; boron - 0.4 ppm or less; cadmium -
0.4 ppm or less; carbon- 270 ppm or less; chromium -
200 ppm or less; cobalt - 20 ppm or less; copper -
50 ppm or less; hafnium - 100 ppm or less; hydrogen -
25 ppm or less; iron - 1500 ppm or less; magnesium -
20 ppm or less; manganese - 50 ppm or less; molybdenum -
50 ppm or less; nickel ~ 70 ppm or less; niobium -
100 ppm or less; nitrogen - 80 ppm or less; silicon -
120 ppm or less; tin 50 ppm or less; tungsten -
100 ppm or less; titanium - 50 ppm or less; and uranium -
3.5 ppm or less.
Sponge zirconium is typically prepared by
reduction with elemental magnesium at elevated temperatures
at atmospheric pressure. The reaction takes place in an
inert atmosphere such as helium or argon.
~nother preferred embodiment comprises a buried
barrier layer -formed of crystal bar zirconlum.
.
~ 24-NT-04471
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Crystal bar zirconium is produced by the vapor-phase
decomposition of zirconium tetraiodide~ Crystal bar
zirconium is more expensive, but has fewer impurities
and displays a greater resistance to radiation damage
than sponge zirconium.
The use of a buried layer of zirconium also
results in desirable fabricati~n benefits. Tube
reduction to finishing tends to remove some material
from the inside of the tube. sy burying the more
expensive unalloyed zirconium in the wall or the tube,
the manufacturing losses are of the less-costly
zirconium alloy, resulting in 100 percent utilization
of the unalloyed zirconiumr Further any manufacturing
defects on the inside of the tube are in the less-
critical inner layer, assuring continuity of thezirconium barrier which is typically only a few mils
thick. Further, an alloyed zirconium inner layer is
better than a container with an inner layer of unalloyed
zirconium, since zirconium alloys are more readily
machined, honed, etc., than the softer unalloyed zirconium.
However, if it is desired to have the buried
layer at the inside surface of the cladding container,
the inner layer of zirconium alloy can be removed by
etching after the tube is finished to its final
dimensions.
Among the zirconium alloys serving as suitable
alloy substrates are Zircaloy-2 and Zircaloy-4.
Zircaloy-2 has on a weight basis about 1.5 percent tin;
0.12 percent iron; 0.09 percent chromium and 0.005
percent nic~sel and is extensively employed in water-
cooled reactors. Zircaloy-4 has less nickel than
Zircaloy-2, but contains slightly more iron than
Zircaloy-2. The composite cladding used in the nuclear
fuel elements of this invention can be fabricated by
any of the following methods.
,,
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24~NT-04471
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~ n one method, a tube of unalloyed zirconium
barrier material is inserted into a hollow billet of
the material selected to be the substrate, a tube of the
material selected to be the inner layer is inserted into
the zirconium barrier tube, and then the assembly is
subjected to explosive bonding of the tubes to the
billet. The composite is extruded using conventional
tube shell extrusion at elevated temperatures of about
1000F to 1400F (about 538C to 760 C). Then the
extruded composite is subjected to a process involving
conventional tube reduction until the desired size of
cladding is achieved. The relative wall thickness o~ the
hollow billet, the zirconium barrier tube and the inner
layer tube are selected to give the desirable thickness
ratios in the finished cladding tube.
In another method, a tube of unalloyed zirconium
barrier material is inserted into a hollow billet of
the material selected to be the substrate, a tube of
the material selected to be the inner layer is inserted
into the tube of the zirconium barrier, and then the
assembly is subjected to a heating step ~such as at
750 C for 8 hours) under compressive stress to assure
good metal-to-metal contact and diffusion bonding between
the tubes and the billet. The diffusion bonded composite
is extruded using conventional tube shell extrusion such
as described above in the immediately preceding paragraph.
Then the extruded composite is subjected to a process
involving conventional tube reduction until the desired
size of cladding is achieved.
The foregoing process of fabricating the composite
cladding of this invention gives economies over other
processes used in fabricating cladding such as
electroplating or vapor deposition. The invention includes
a method of producing a nuclear fuel element comprising
making a composite cladding container which is open at one
end, the cladding container having a substrate, an unalloyed
zirconium barrier metallurgically bonded to the inside
37~ ~
24-NT-04471
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of the substrate, and an inner layer metallurgically
bonded to the inside surface of the zirconium barrier,
filling the composite cladding container with nuclear
fuel material and leaving a cavity at the open endl
inserting a nuclear fuel material retaining means into
the cavity, applying the cavity in communication with
the nuclear fuel, and then bonding the end of the
cladding container to said enclosure to form a tight
seal therebetween.
The present invention offers several advantages
promoting a lony operating life for a nuclear fuel
element, including the reduction of chemical interaction
of the cladding, the minimization of localized stress
on the zirconium alloy substrate portion of the cladding,
the minimization of stress corrosion on the zirconium
alloy substrate portion of the cladding, and the
reduction of the probability of a splitting failure
occurring in the zirconium alloy substrate as a result
of pellet-cladding-interaction. The invention further
prevents direct contact between the fission products and
the zirconium alloy substrate and the occurrence of
localized stress on the zirconium alloy substrate. The
invention thus prevents the initiation or propagation of
stress corrosion cracks in the alloy substrateO
There are particular advantages to using unalloyed
zirconium as a buried barrier. The unalloyed zirconium
is quite ductile and in the event stress corrosion
cracks initiate in the inner layer, -their propagation can
be effectively stopped in the zirconium barrier. It is
believed that the radius of curvature at the end of a
crack at the unalloyed zirconium is much larger than in
zirconium alloys; thereby re~uiring much higher stress
levels for propagation. The unalloyed zirconium is also
less susceptible to iodine stress corrosion~ further
~! ~ tending to inhibit crack propagation.
.,
~Z~ 7
24-NT-04~71
-18-
An importan-t property of the composite cladding
of this invention is that the foregoing improvements
are achieved with no additional neutron penalty. Such a
~ladding is readily accepted in nuclear reactors since
the cladding would have no eutectic formation during a
loss-of-coolant accident or a reactivity insertion
accident involving the dropping of a nuclear control rod.
Further, the composite cladding has no heat transfer
penalty in -that there is no thermal barrier to transfer
of heat such as results in the situation where a separate
foil or liner is inserted in a fuel element. Also, the
composite cladding of this invention is inspectable by
conventional, non-destructive testing methods during
various stages of fabrication and operation.
As will be apparent to those skilled in the art,
various modifications and changes may be made in the
invention described herein. It is accordingly the
intention that the invention be construed in the broadest
manner within the spirit and scope as set forth in the
accompanying claims.
.~
