Note: Descriptions are shown in the official language in which they were submitted.
~2`~ 72~
NT- 0 4 4 81
Z IRCONIUM ALLOY BARRIER HAVING
.
IMPROVED CORROSION RESISTANCE
EIELD OF TXE 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 having a composite cladding
container having a metal liner of dilute zirconium
alloy consisting of zirconium and a metal selected
from the group consisting of iron, chromium, iron plus
chromiuml and copper bonded to the inside surface of
the zirconium alloy cladding substrate.
~ACKGROUND OF T~IE INVENTION
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 corroslon-
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
coolant flow channel or region forming a fuel assembly,
and sufficient fuel assemblies are combined to orm
the nuclear fission chain reacting assembly or reactor
core capabIe of a sel-f-sustained fission reaction.
The corel in turn, is enclosed within a reactor vessel
through which a coolant ls passed.
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The cladding serves several purposes and two
primary purposes are: first, to prevent contact and
chemical reactions between the nuelear fuel and the
coolant or the moderator if a moderator is present,
or both if both the coolant and the moderator are
present; and second, to prevent the radioactive
fission products, some o:E which are gases, from being
released 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
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 claddings since they have
low neutron absorption cross-sections and at temperatures
below about 750 F (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.
However, fuel element performance has
revealed a problem with the brittle splitting of the
cladding due to the eombined interactions between the
nuelear fuel, the cladding and the fission products
produced during nuclear fission reac-tions. It has
been discovered that this undesirable performance is
promoted by localized mechanical stress due to
differential expansion of fuel and cladding (stresses
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24--NT-044 ~1
--3--
in the cladding are concentrated at cracks in the nuclear
fuel). Corrosive fission products are released from the
nuclear fuel and are present at the intersection of the
fuel 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 ~y 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 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 heatingl such as during
fuel element manufacture and urther release ~ission
products during irradiation. Particulate refractory
and ceramic compositions~ such as uranium dioxide
powder and other powders used as nuclear fuel, have
been known to reIease even larger quantities of the
a-forementioned gases during irradiation. These
~2~
- 4 - 24-NF-044~1
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,
which are 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 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 difficult. 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 NEBO 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 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 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 of chromium to provide a composite material
useful for nuclear reactors. A method for electrolytic
~2~97~ 24-NT-0~481
deposition of copper on Zircaloy-2 surfaces and
subsequent heat treatment for the purpose of obtaining
surface diffusion of the electrolytically deposited
metal is presented in Energia Nuclearer ~olume 11~
No. 9 ~September, 1964) at pages 505-508. In Stability
and Compatibility of Hydrogen Barriers Applied to
zirconium Alloys, by F. Brossa et al (~uropean Atomic
Energy Community, Joint Nuclear 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 zirconium and
zirconium-tin alloys and heat treating these alloys to
produce alloy-diffusion bonds are disclosed in
Electroplating on Zirconium and Zirconium-Tln, by
W.C. Schickner et al (BMI-757/ Technical Information
Service, 1952~.
U.S. Patent No. 3,625,821, issued
December 7l 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. Reactor Development Program
- ~ 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
barrie~ between the nuclear fuel material and the
cladding holding the nuclear fuel material as disclosed
in U.S. Patent No. 3,230,150, issued ~anuary 18, 1966
to Martin et al (copper foil); German Patent Publication
DAS lr238,115 (titanium layer); U.S. Patent No. 3,212,988,
issued October 19, 1965 to Ringot et al (sheath of
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24-NT-04481
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zirconium, aluminum, aluminum or beryllium); U.S. Patent
No. 3,018,238, issued January ?3~ 1962 to Layer et al
(barrier of crystalline carbon between the U02 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 promising, some of
the foregoing references involve incompatible ma-terials
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 andthe cladding.
Further approaches to the barrier concept are
disclosed in U~S. Patent No. 3,969,186t issued
July 13, 1976 to Thompson, (refractory metal such as
molybdenum~ tungsten, rhenium, niobium and alloys
thereof in the form of a tube or foil of single or
multiple layers or a coa~ing on the internal surface
of the cladding), and U.5. 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. 4rO45,288, 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. The barrier i5 selected
from a group of niobiumr aluminuml copper, nickel,
stainless steel, and iron. The buried metal barrier
reduces corrosion due to fission products and corrosive
gases, but is subject to stress corrosion cracking
,
~ ~ 24-NT-04~81
-7-
and liquid metal embrittlement.
U.S. Patent No. 4,200,492~ issued April 29, 1980,
to Armijo, discloses a composite cladding of a zirconium
alloy substrate with a sponge zirconium liner. The soft
zirconium liner minimizes localized strain, 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.
SU~ARY OF THE INVENTION
A particularly effective nuclear fuel elemen-t
for use in -the core of a nuclear reactor has a composite
cladding having a metal liner of dilute zirconium alloy
metallurgically bonded to the inside surface of the
substrate. The dilute zirconium alloy comprises
zirconium and a metal selected from the group consisting
of iron, chromium, iron plus chromium and copper, wherein
the amount of iron alloyed with the zirconium is from about
0.2% to about 0.3% by weight, the amount of chromium is
from about 0.05% to about 0.3% by weight; the total
amount of iron plus chromium is from about 0.15% to
about 0.3% by weight and wherein the ratio of the weights
of iron to chromium is in the range of from about 1:1 to
about 4:1; and wherein the amount of copper is from about
0.002% to about 0.2% by weight.
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
~ 7~ 24-NT-04481
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dilute zirconium alloy liner forms a continuous shield
between the substrate 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 zirconium alloy liner forms from
about 1 to about 20 percent of the thickness of the
cladding. The liner remains soft, relative to the
substrate, during irradiation and minimizes localized
stress inside the nuclear fuel element, thus serving
to protect the cladding from stress corrosion cracking
or liquid metal embrittlement. The dilute zirconium
alloy liner shields the substrate from reaction with
volatile impurities or fission products present inside
the nuclear fuel element and, in this mannerJ serves
to protect the cladding substrate from attack by the
volatile impurities or fission products~
This invention has a striking advantage
that the substrate 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 liner does not introduce any appreciable neutron
capture penalties~ heat transfer penalties, or fuel/
liner incompatibility problems. In addition, the
liner provides superior resistance to steam or hot
water oxidation as compared to unalloyed zirconium
in the event of a breach in the cladding.
DESCRIPTION OF THE DRA~INGS
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 described hereinafter.
FIG. 1 is a partial cutaway sectional view
of a nuclear fuel assembly containing nuclear fuel
. .
37;~6 24-NT-04481
g
elements constructed according 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 of this invention.
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 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 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
suppor-ted -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 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 longitudinal expansion of the fuel
material and accumulation of gases released from the
fuel material~ ~ nuclear fuel material retainer means
24 in the form of a helical member is positioned
within space 20 to provide restraint against the axial
movement of the pellet column, especially during
handling and transportation of the fuel element.
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24-NT-04481
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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 FIG~ 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 thereo. A preferred fuel is uranium
dioxide or a mixture 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 23 be-tween 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.
35The substrate 2I has metallurgically bonded
on the inside circumference thereof a dilute zirconium
24-NT-0~81
--11--
alloy liner 22 so that the dilute zirconium alloy liner
forms a shield of the substrate from the nuclear fuel
material 16 inside the composite cladding The dilute
zirconium alloy liner preferably forms about 1 to about
20% of the thickness of the cladding. A dilute
zirconium alloy liner forming less than about 1% of
the thickness of the cladding would be difficult to
achieve in commercial production, and a dilute
zirconium alloy liner forming more than 20% 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 is comprised of
zirconium and an alloy addition selected from the
group consisting of: iron, chromium, iron pIus
chromium and copper. As used herein, dilute zirconium
alloy means a zirconium alloy with an alloy content
sufficiently low to display greater ductility and
higher strain rate than does the substrate material
under equivalent conditions of stress.
The amount of iron alloyed with zirconium is
from about 0.2% to about 0.3% by weight, and preferably
from about 0.2% to about 0.25% by weight.
Chromium is in the range of from about 0.05%
to about 0.3% by weight and preferably from about 0.15%
to about 0.25~ by weight.
Iron plus chomium may be included so that
the total amount of both components is from about
0.15% to about 0.3% by weight and preferably from about
0.2% to about 0.25% by weight and wherein the ratio
of the weights of iron to chromium is from about 1:1
to about ~:1 and preferably about 2:1.
Copper is used in the amount of from about
0.02% to about 0.2% by weight and preferably from
about 0.05~ to about 0.15% by weight.
..
7~
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The dilute zirconium alloy liner shields
the substrate from gaseous impurities and fission
products and protects the substrate portion of the
cladding from contact and reaction with such impurities
and fission products and prevents the occurrence of
localized stress.
The addition to zirconium of small amounts of
a metal selected from the group of iron, chromium,
iron plus chromium and copper improves corrosion
resistance, especially resistance to oxidation by
hot water or steam if the addition is within the
stated range for that metal. The lower limit of the
amount of each metal alloyed with zirconium provides
su~ficient quantity of that metal to signiEicantly
improve the corrosion resistance as compared to unalloyed
zirconium.
The upper limit of the amount of each metal
alloyed with zirconium is generally set at the maximum
amount of the metal which significantly improves the
corrosion resistance as compared with sponge zirconium.
Additions of the metal exceeding the upper limit fail
to significantly enhance the corrosion resistance
properties of zirconium and may have a detrimental effect
in reducing the softness and ductility of the liner.
The additions of each metal to zirconium that
impart the greatest improvement in corrosion resistance
are stated as the preferred ranges.
Iron, chromium and copper are sparingly soluble
in zirconium. Dilute zirconium alloys involving one
or more of these metals can be heat treated to provide
a material with a fine dispersion of intermetallic
particles which ar~ noble with respect to the zirconium
matrix. Because the alloy constituents are sparingly
soluble, little solid solution strengthening of the
zirconium occurs. The strengthening effect is
sufficiently low to maintain the softness required
7;~j
2~-NT-04~81
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for the dilute zirconium alloy liner to preven-t or
mitigate fuel failure by pellet-cladding interaction.
The dilute zirconium alloy liner in the
composite cladding resists irradiation hardening relative
to Zircoloy or other conventional zirconium alloys,
and this enables the dilute zirconium alloy liner,
after prolonged irradiation, to .naintain desirable
structural properties such as yield strength and
hardness at levels considerably lower than those of
conventional zirconium alloys. In effect, the dilute
zirconium alloy ]iner does not harden as much as
conventional zirconium alloys when subjected to
irradiation and this, thogether with its initially
low yield strength, enables the dilute zirconium
alloy liner to deform plastically and relieve pellet-
induced stresses in the fuel element that can be
broughtabout, for example, by swelling of the pellets
of nuclear fuel a-t reactor operating -temperatures
(300 C to 350C) so that the pellet comes into contact
with the cladding.
A dilute zirconium alloy liner comprising
zirconium and a metal selected from the group
comprising iron, chromium, iron plus chromium, and
copper and preferably about 5 to 15 percent of the
thickness of the cladding bonded to a conventional
z.irconium alloy substrate provides stress reduction
sufficient to prevent or mitigate failures in the
composite claddiny.
The purity of the zirconium metal that is
alloyed with iron, chromium, iron plus chromium or
copper is important and serves to impart special
properties to the dilute zirconium alloy liner.
Generally, there is less than 5000 ppm impurities in
the zirconium metal. Of these oxygen should be as
low as practical but may vary up to about 1000 ppm
~,
7Z~i
24-NT-04481
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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 dilute zirconium alloy liner to form metallurgical
bonds, but insufficient cross-diffusion to alloy
siynificantly with the dilute zirconium alloy liner
itself.
Among the conventional zirconium alloys serviny
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 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~
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 explGsive bonding
of the tube to the billet. The composite is extruded
using conventional tube shell extrusion at elevated
temperatures of about 1000 F to 1400 F ~about 538 C
to 760C~. Then the extruded composite is subjected
to a process involving conventional tube reduction
until the desired size o~ 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
,
~2~ 6 2~-NT-04~81
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The present invention offers several advantages
promoting a long operating life for a nuclear fuel
element, lncluding 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 corroslon on the zirconium
alloy substrate portion of the cladding, and the
reduction o~ the probability of a splitting failure
occurring in the zirconium alloy substrate.
l~ In addition to minimizing stress and stress
corrosion on the substrate~ the dilute ~irconium alloy
liner is resistant to oxidation by steam and hot
water in the event that the cladding is breached,
whereas unalloyed zirconium oxidizes rapidly under
-these conditions. The dilute zirconium alloy exhibi-ts
plasticity similar to unalloyed zirconium and provides
the benefits thereof while also providing increased
resistance to corrosion~ especially to oxidation by
hot water and steam.
2~ An important property of the composite cladding
of this invention is that the foregoing improvements
are achieved with no substantial additional neutron
penalty. Such a cladding is readily accepted in nuclear
reactors since the cladding would have no eutectic
formation during a loss-of-coolant accident or an
accident involving the dropping of a nuclear control
rod. Further, 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 inser-ted
in a fuel element. Also, the composite cladding of
this invention is inspectable by conventional non-
destructive testing methods during various stage of
fabrication and operation.
~s will be apparent to those skilled in the art,
various modifications and changes may be made in the
' ;
12~972~ 24-NT-04481
-17-
invention described herei.n. It is accordingly the
intenti~n that the invention be construed in the
broadestmanner within the spirit and scope as set
forth in the accompanying claims.
~2~72~ 2~-NT-04481
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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.
