Note: Descriptions are shown in the official language in which they were submitted.
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ZIRCONIUM ALLOY BARRIER HAVING
IMPROVED_CORROSI'ON RESIST~NCE
Field o'f the'Invention
This invention relates broadly to an improvement
in nuclear fuel elements for use in the core of nuclear
fission reactors and, more particularly, to an improved
nuclear fuel element havlng a composite cladding
container having a metal liner of dilute ~irconium
alloy comprising zirconium and niobium bonded to the
inside surface of a zirconium alloy cladding substrate.
Background'of''the- Inve'nti'on
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 coolan-t
flow channel or region forming a fuel assem~ly, and
sufficient fuel assemblies are combined to form -the
nuclear fission chain reacting assembly or reac-tor
core capable of a self-sustained fission reaction.
The core, in -turn, is enclosed within a reac-tor vessel
-through which a coolant is passed.
The cladding serves several purposes and -two
primary purpose, are: first, to prevent contact and
chemical reactions between the nuclear fuel and the
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coolant or the moderator i~ a moderator is present, or
both i~ both the coolant and the moderator are present;
and second; to prevent the radioactive fission products,
some of which are gases, from being released from the
fuel into the coolant or the moderator or both iE 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 ~ the leak
tightness, can con~aminate the coolant or moderator
and the associated systems with long-lived radioactive
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 cladding
material due to mechanical or chemical reactions of these
cladding materials under certain circumstances.
Zirconium and its alloys, under normal circumstances,
make excellent nuclear fuel c]addings since they have
low neutron absorption cross-sections and at temperatures
below about 750F (about 398 C) are strong, ductile,
extremely stable and relatively non-reactive in the
presence of demineralized water or steam which are
commonly used as reactor coolants and moderators.
~ owever, ~uel element performance has revealed
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 during
nuclear fission reactions. It has been discovered that
this ~Indesirable perEormance is promoted by localized
mechanical stresses due to fuel cladding differential
expansion o:E fuel and cladding (stresses in the cladding
are concentrated at cracks in the nuclear fuel).
Corrosive fission products are released ~rom the nuclear
24-NT-04493
fuel and are present at the intersection of the uel
cracks with the cladding surface. Such fission products
are created in the nuclear fuel during the fission
chain reaction during operation of a nuclear reactor.
The localized stress is exaggerated by high friction
between the fuel and the cladding.
Within 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 mechanieal 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 zireonium 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 eeramic compositions, such
as uranium dioxide and other compositions used as
nuelear fuel, release measurable quantities of the
aforementioned gases upon heating, such as during
fuel element manufacture and further release fission
produets during irradiation. Partieulate refractory
and eeramic 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
zireonium eladding con-taining the nuele~r fuel.
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Thus, in liyht of the foregoing, i-t has been found
desirable to minimize attack of the cladding from water,
water vapor and other yases, especially hydrogen, which are
reactive with the cladding from inside the fuel element
throughou-t 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 cal]ed getters.
Another approach has been to coat the nuclear
fuel material with any of a variety of materials to prevent
moisture coming in contact with the nuclear fuel material.
The coating of nuclear fuel material introduces reliability
problems in that achieving uniform coatings free of faults
is difficulto Further, the deterioration of the coating
can introduce problems with the long-lived performance of
the nuclear fuel material.
General Electric Atomic Power Document 4555
of February 1964, at GE NEB0 Library, 175 Curtner Ave.,
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 extrusion of a hollow
billet of the zirconium alloy having an inner lining of
stainless steel. This cladding has ihe disadvantages
that the stainless steel develops brittle phases, and
-the s-tainless steel layer involves a neutron absorption
penalty of about ten to fifteen ti.mes the penal-ty for a
zirconium alloy of the same thickness.
U.S~ Patent No. 3,502,5~9, issued
March 24, 1970 to Charveriat, discloses a method for
protecting zirconium and its alloys by the electrolytic
deposition oE chronium to provide a composite material
useful Eor nucJear reactors. A method for electrolytic
deposition of copper on Zircaloy-2 surfaces and
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24-NT-04493
subsequent heat treatment for the purpose of obtaining
surface diffusion of -the electrolytically deposited
metal is presented in Ene-rgia Nucleare, Volume 11,
No~ 9 (September, 196~) at pages 505-508. In Stability
and Compatibility of Hydrogen Barrie~ plied to
_ . _
Zirconium Alloys, by ~. ~rossa et al (European Atomic
Energy Communi-ty~ ~oint Nuclear Research Center,
EUR 4098e, 1969), methods of deposition of different
coatings and their efficiency as hydrogen diffusion
barriers are described along wi-th an Al-Si coating as
the most promising barrier against hydrogen diffusion.
Methods for electroplating nickel on zirconium and
zirconium-tin alloys and heat treating these alloys to
produce alloy-diffusion bonds are disclosed in
Electropl~a'ting o'n Zirconium 'and Zirconlum=Tin, by
W.C. Schickner et al (BMI-757, Technical Information
Service, 19521.
U.S. Patent No. 3,625,821, issued
December 7, 1971, to Ricks, presents a fuel element
for a nuclear reactor having a fuel cladding tube with
the inner surface of the tube being coated with a metal
of low neutron capture cross-section such as nickel and
having finely dispersed particles of a burnable poison
disposed therein. Reac'tor Devel'opment Program Progress
Report of August, 1973 (ANL-RDP-l9) discloses a chemical
getter arrangement of a sacrificial layer of chromium
on the inner surface of a stainless steel cladding.
Another approach has been to introduce a barr:ier
between the nuclear fuel material and the cladding
holding the nuclear fuel material as disclosed in
U.S. Patent No. 3,230,150, :issued January 18, 1966
to Martin et al, tcopper foil); German Patent Publication
DAS 1,238,115 (titanium layer); U.S. Patent No.
3,212,988,issuecl October 19, 1965 to Ringot et al
(sheath of zirconium, aluminum or beryllium); U.S.
Patent No. 3,018,238, issued ~anuary 23, 1962 to
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layer et al (barrier of crystalline carbon between the
U2 and the zirconium alloy cladding); and U.S. Patent
No. 3,088,893~ issued May 7, 1963 to Spalaris
(stainless steel foil). While the barrier concept
proves pro~ising~ some of the foregoing references
involve 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
problem of localized chemical-mechanical interactions
between the nuclear fuel and the cladding.
Further approaches to the barrier concept are
disclosed in U.S. Patent No. 3,969,186, issued
July 13, 1976 to Thompson, (refractory metal such as
molybdenuml tungsten, rhenium, niobium and alloys
thereof in the ~orm of a tube or foil of ~ingle 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 (liner 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 composite 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. ~he barrier is selected from a group of
niobium, aluminum, copper, nickel, stainless steel, and
iron. The buri.ed metal barrier reduces corrosion due to
fission product:s and corrosive gases, but is subject to
stress corrosion crackin~ and liquid metal embrit-tlement.
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U.S. Patent No, 4,200,~9~, issued ~pril 29, 1980
-to Armijo et al discloses a composite cladding of a
zirconium alloy substrate with a sponge zirconium liner.
The soft zirconium liner minimizes localized stress,
and reduces stress corrosion cracking and liquid metal
embrittlement, but is subject to loss due to honing
and the like during fabrication and due to oxidation.
Furthermore, if a breach in the cladding should occur,
allowing water and/or steam to enter the fuel rod,
the zirconium liner tends to oxidize rapidly.
Accordingly~ it has remained desirable to
develop nuclear fuel elements minimizing the problems
discussed above.
Summary of the Invention
. . . _
A particularly effective nuclear fuel element
for use in the core of a nuclear reactor has a composite
cladding having an inner liner of dilute zirconium
alloy metallurgically bonded to the inside surface
of the substrate. The dilute zirconium alloy comprises
from about 0.1% to about 0.5~ by weight niobium
and preferably from about 0.2~ to about 0.4~ by weight
niobium, the balance being zirconium.
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. A zirconium alloy cladding substrate has
a higher alloy content than the dilute zirconium
alloy liner.
The clilute zirconium alloy liner forms a
continuous shield be-tween the suhstrate and the
nuclear fuel material held in the cladding, as well
as shielding the zirconium alloy or other substrate
cladding from fission products and gases. The dilute
3S zirconium a]loy liner forms Erom aboutl to about 20
percent of the thickness of the cladding.
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24~NT-04493
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The liner remains soft, relative to the
substrate, during irradiation and minimizes localized
strain inside the nuclear fuel element~ thus serving
to protect the cladding from stress corrosion
cracking or liquid metal embrittlement. The dllute
zirconium alloy liner gives a preferential reaction
site for reaction with volatile impurities or fission
products present inside the nuclear fuel element and~
in this manner, serves to protect the cladding substrate
from attack by the volatile impurities or fission
products.
This invention has a striking advantage that
the substxate of the cladding is protected from stress
corrosion cracking and liquid metal embrittlement, in
addition to contact with fission products, corrosive
gases, etc., by the dilute zirconium alloy liner and
the lin~r does not introduce any appreciable neutron
capture penalties, heat transfer penalties, or fuel/
liner incompatibility problems. In addition, the
liner provides superior resistance to hot water or
steam o~idation as compared to unalloyed zirconium
in the event of a breach in the cladding.
Description of the; Drawings
_ ... . . .... _ _ _
The foregoing and other objects of this
invention will become apparent to persons skilled in
the art from reading the following specification and
the appended claims with reference to the accompanying
drawings, wherein:
FIG. 1 is a partial cukaway sectional view of
a nuclear fuel assembly containing nuclear fuel elements
constructed acc:ording to the teaching of this invention;
and
FIG. 2 is an enlarged transverse cross-sectional
view of the nuclear fuel element in FIG. 2 illustrating
the teaching oE this invention.
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_.9_
Description of the Invention
Referring now more particularly to FIG. 1, there
is shown a partially cutaway sectional view of a
nuclear fuel assembly 10. This ~uel 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 assembly
10 being omitted). The upper end of channel 11 is
open at outlet 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
channel 11 and supported therein by means of an
upper end plate 15 and a lower end plate (not. shown
due to the lower portion beinc3 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 through the upper
outlet 13 at an elevated temperature in a partially
vaporized condition for boiling reactors or in an
unvaporized condition for pressurized reactors.
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 longit.udinal expansion of the fuel
material and accumulat.ion of gases released from the
~uel material. A nuclear fuel material retainer means
24 in the form of a helical member i.s positioned within
space 20 to provide restrain-t against the axial
movement of the pellet colllmn, espec.ially duri.ng
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
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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 sectior) in FIG. 1. The fuel
element includes a core or central cylindrical
portion of nuclear fuel material 16, here shown as a
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 mi~ture comprising uranium
dioxide and plutonium dioxide.
Referring now to FIG. 2~ the nuclear fuel
material 16 forming the central core of the fuel element
14 is surrounded by a cladding 17 which, in this invention,
is also referred to as a composite cladding container.
The composite cladding container encloses the fissile
core so as to leave a gap 2~ between the core and
the cladding during use in a nuclear reactor. The
composite cladding container has an external substrate
21 selected from conventional cladding materials
such as a stainless steel and zirconium alloys and,
in a preferred embodiment of this invention, the
substrate is a zirconium alloy such as Zircaloy-2.
The substrate 21 has metallurgically bonded on
the inside circumference thereof a dilute zirconium
alloy liner 22 so that the zirconium alloy liner
forms a continuous shield of the substrate from the
nuclear fuel material 16 inside the composite cladding.
24-NT-04493
The dilute zirconium alloy liner preferably forms from
about 1 to about 20 percent of the thickness of the
cladding. As used herein, dilute zirconium alloy means
a zirconium alloy with an alloy content sufficiently
low to display greater ductility than the substrate
material.
A dilute zirconium alloy liner forminy less
than about 1 percent of the thickness of the cladding
would be difficult to achieve in commercial production,
and a dilute zirconium alloy liner forming more than
20 percent of the thickness of the cladding provides
no additional benefit for the added thickness. Further
a liner more than about 20% of the thickness of the
cladding means a concomitant reduction in thickness
of the substrate and possible weakening of the cladding.
The dilute zirconium alloy liner serves as a
prefarential reaction site for gaseous impurities
and fission products and protects the substrate portion
of the claddin~ from contact and reaction with such
impurities and fission products and ameliorates
the occurence of localized stresses.
The dilute zirconium alloy liner comprises
from about 0.1% to about 0.5~ by weight niobium, and
preferably from about 0.2% to 0.4~ by weight niobium,
the balance being zirconium.
A dilute zirconium alloy comprising from
about 0.1~ to about 0.5% by weight niobium exhibits
increased resistance to corrosion or oxidation by
contact with hot water and steam as compared with
unalloyed zirconium. Dilute zirconium alloys
comprising less than about 0.1~ by weight niobium
exhibit no significant increase in corrosion
resistance and are difficult to achieve in commercial
production.
Niobium is soluble in zirconium in the range
from about 0.1% to above 0.5~ by we:ight. While some
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24-NT-0~493
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solid solution strengthening does occu:r due to the
solubility of niobium, the amount ls sufficiently
low to enable the dilute zirconium alloy to resist
fuel rod failure from pellet-cladding interaction.
Above about 0.5% by weight, niobium forms
precipitates which increase the strength of the
zirconium alloy and significantly decreases i-ts
ductility or plasticity. An upper limit of about
0.5% by weight is therefore preferred to insure that
the dilute zirconium alloy remains highly ductile
and resistant to radiation hardening which enables
the dilute zirconium alloy liner, after prolonged
irradiation, to maintain desirable structural
properties such as yield strength and hardness at
levels considerably lower than those of conventional
zirconium alloys. In effect the dil.ute zirconium
alloy liner does not harden as much as conventional
zirconium alloys when subjected to irradiation and
thisl together with its initial low yield strength,
20. enables the dilute zirconium alloy liner to deform
plastically and relieve pellet-induced stresses in
the cladding. Such stresses can be brought about,
for example, by swelling of the pellets of nuclear
fuel at reactor operating temperature (300 C to
350C) so that the pellet comes into contact with
the cladding.
A dilute zirconium alloy liner comprising
about 0.2% to about 0.~% by weight niobium is
particularly preferred because a dilute zirconium
alloy in this range exhibits the preferred combination
of corrosion resistance and ductility. ~elow about
0.2% nic~bium .in the zirconium, the corrosion resis-
tance beyins to approach that of sponge zirconium.
It is particularly preferred to h~e a ma.ximum
niobium content of about 0.4% to assure t.hat the liner
does not lie outside the solid soluhility range
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24-NT-04493
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regardless of thermal exposure of the fuel rod over
long periods of time~ thereby assuring continued
ductility.
The dilute zirconium alloy liner comprising
about 0.1% to about 0.5~ by weight niobium, and
preferably from about 0~2~ to about 0.4% by weight
niohium and forming from about 1 to 20 percent of the
thickness of the claddîng and preferably Erom about 5
to 15 percent of the cladding bonded to a conventional
zirconium alloy substrate provides stress reduction
while improving corrosion resistance, especially
resistance to oxidation by hot water and steam in
the event of a cladding breach.
The purity of the zirconium metal that is
alloyed with niobium is important and serves to impart
particular properties to the dilute zirconium alloy
liner. Generally, there is at least 1000 parts per
million (ppm~ by weight and less than 5000 ppm
impurities in the zirconium metal and preferably less
than 4200 ppm. Of these oxygen may vary up l:o about
1000 ppm.
The composite cladding of the nuclear fuel
element of this invention has a dilute zirconium
alloy liner metallurgically bonded to the substrate.
Metallographic examination shows that there is
sufficient cross-diffusion between the substrate and
the zirconium liner to form metallurgical bonds, but
insuEficient cross-difEusion to significantly alloy
with the dilute ~irconium alloy liner itselE.
Among the conventional zirconium alloys
serving as suitable 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 nickel and is extemsively
employed in water--cooled reactors. Zircaloy-4 has
less nickel than Zircaloy-2l but contains slightly
more iron than Zircaloy-2. The composite cladding used
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in the nuclear fuel elements oE this invention can be
fabricated by any of t~e following methods.
In one method/ a tube of -the dilute zirconium
alloy liner material is inserted into a hollow billet
of the material selected to be the substrate, and
then the assembly is subjected to explosive bonding
of the tube to the billet~ The composite is ex-truded
using conventional tube shell extrusion at elevated
temperatures of about 1000F to lA00F (about 538 C
to 760C). 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 of the hollow billet and the
dilute zirconium alloy liner tube are selected to
give the desired thickness ratios in the finished
cladding tube.
In another method, a tube of the dilute
zirconium alloy liner material is inserted into a
hollow billet of the material selected to be the
substrate, and then the assernbly is subjected to a
heating step (such as at 750C for 8 hours) under
compressive stress to assure good metal-to~metal
contact and diffusion bonding between the tube and
the billet. The diffusion bonded composite is
extruded using conventional tube shell extrusion such
as described above in the immedia-tely preceding
paragraph. Then the extruded cornposite is subje!cted
to a process involving conventional tube reduction
until the desired size of cladding is achieved.
In still another method, a tube of the!
dilute zirconium alloy liner material is inserted into
a hollow bille-t of the material selected to be the
substrate~ and the assembly is extruded using
conventional tube shell extrusion as described above.
Then the extrud,ed composite is subjected to a process
involvingconvention~l tube reduction until the desired
size of cladding is achieved.
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The foregoing processes of fabricating t:he
composite cladding of this invention give economies
over other processes used in fabricating cladding such
as electroplating or vapor c1eposition. A nuclear
fuel element can be forged by making a composite
cladding container which is open at one end, the
cladding container having a substrate and an inner
dilute zirconium alloy liner metallurgically bonded
to the inside surface of the substrate. The element
is completed by fil.ling the composite cladding container
with nuclear fuel material~ leaving a cavity at the
open end~ inserting a nuclear fuel material retaining
means into the cavity, applying an enclosure to the
open end of the container leaving 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 long operating life ~or a rluclear fuel
element, including the reduction of chemical interaction
o~ the cladding, the minimization of localized stress
on the æirconium alloy substrate portion of the cladding,
the minimization of stress corrosion on the zirconi.um
alloy substrate portion of the cladding, and the
reduction of the probability of a splitting failure
occurring in the zirconium alloy substrate.
In addi-tion to minimizing stress and stress
corrosion on the substrate, the dilute zirconium alloy
liner is resistant to oxidation by steam and hot
water in theevent that the cladding is breached,
whereas unalloyed zirconium oxidizes very rapidly
under these conditions. The dilute zirconium alloy
exh.ibits plasticity similar to unalloyed zirconium
and provides the benefits thereof while also provid:ing
increased resistance to corrosion, especially to
oxidation by hot water and steam.
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24-NT-0~493
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An important property ~ the composite cladding
of this invention is that the foregoing improvements
are achieved with no substan-tial additional neutron
penalty. Such a cladding is readily accepted in
nuclear reactors since the cladding would have no
eutectic foxmation during a loss-of-coolant accident
or an accident involving the dropping of a nuclear
control rodA Furtherl the composite cladding has a
very small 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 inspectahle by conventional
non-destructive testing methods during various stages
of fabrication.
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
broadestmanner within the spirit and scope as set
forth in the accompanying claims.
