Note: Descriptions are shown in the official language in which they were submitted.
lO9iO33
BACRGROUND OF THE INVENTION
Field of the Invention
This invention relates to a waste control process for
the treatment of an actinide-containing organic waste. It
particularly relates to a molten salt process for reducing
the volume of an organic waste material contaminated with
radioactive actinide elements and further provides for the
separation and recovery of such radioactive actinide elements.
Prior~Art
In the processing of fuel for nuclear reactors, and in
the operation of such reactors, a considerable amount of waste
material is generated, which is contaminated with radioactive
actinide elements. It is repor$ed that the various U.S. Energy
Research and Development Administration (ERDA) facilities
lS generate approximately 350,000 cu.ft. of solid transuranic
waste per year. Since the current costs of storing such waste
are high and are likely to increase, there i8 an urgent economic
incentive to reduce the volume of such waste. In addition, the
high toxicity, high specific activity and long-retention in
the body (half-life 300 years~ of plutonium requires s,pecial
handling. Since about 1970, plutonium wastes-have been stored
on a temporary basis until ultimate disposal techniques could
~ be developed. The wastes are contained in polyethylene-lined
! 55 gallon drums so as to be retrievable without external con-
tamination. As much as 1600 kgms of plutonium are associated
with the waste currently in interim storage.
lV91033
The majority of the waste is made up of organic
combustible materials such as rags, paper, plastic and rubber.
A summary of the characteristics of the solid waste is given
below.
CHARACTERISTICS OF TYPICAL SOLI~ TRANSURANIC WASTE
(Source: ERDA3
Composition (wt%)
Paper 55
Rags 5
Plastic 30
(50% polyvinyl chloride
and 50% polyethelene)
Rubber 10
Bul~ Density (lb/cf) 7
Ash Content (%) 8
Plutonium Content tg/lb)0.014
Heating Value (Btu/lb)9,000 to 12,000
Because of the high proportion of halogenated (usually
chlorinated) plastic and the danger of plutonium carryover
in the particulate, a conventional incinerator is not wholly
satisfactory for the combustion of this material as a means
of volume reduction. Thus, more complex incinerators or
special combustion methods are reguired. More particularly,
in processing such waste, ideally as much volume reduction
as possible is obtained with a minimum of pollution.
-3-
1~91033
Various processes have been suggested for treating
different radioactive waste. None of these processes, however,
have proven altogether satisfactory. British Patent No.
1,035,330 discloses a process and apparatus for treating solid
radioactive wastes. The patent suggests that low level
radioactive wastes be incinerated in a furnace to reduce their
volume, and a multi-step filtering technique for combustion
gases is proposed. The disadvantage of this process is that
it requires ela~orate filters for the offgases, and further,
makes no provision for recovery of radioactive actinide elements.
More particularly, since normal incineration temperatures are
around 800C, it would be extremely difficult to recover the
actinide elements as most of such elements form refractory
oxides at temperatures in excess of around 750C. The
refractory transuranic oxides are not readily amenable to
conventional recovery techniques.
U. S. Pat. No. 3,479,295 suggests a method of reducinq
a radioactive waste solution, obtained in the processing of
nuclear fuel elements, to dryness. Broadly, the process
comprises blowing an oxygen-containing gas upwardly through a
bed of particles, formed by calcination of the salts in the
waste solution, to fluidize the bed and feeding additional
waste solution into the fluidized bed so formed. A hydrocarbon
fuel also is introduced into the fluidized bed in the presence
Z5 of nitrate ions at a temperature above the ignition temperature
of the fuel to burn the fuel and provide the heat necessary to
evaporate the solution and calcine the salts contained therein.
--4--
. . .
1~19103;~
A disadvantage of this process is that it requires a source of
nitrate ions and does not provide for the recovery of any
actinide elements contained in the waste solution.
In U. S. Pat. No. 3,716,490 there is disclosed another
method for the treatment of radioactive liquids. The method
comprises providing a solid, fusible, partly sulfonated
bituminous substance and contacting that substance with a
liquid waste containing radioactive ions to ion exchange the
radioactive ions with the suifonated portion of the bituminous
substance. Thereafter, the bituminous substance is melted to
reduce its volume and encapsulate the radioactive ions. Thus,
while this patent provides a method for reducing the volume of
radioactive waste, it does not provide a means for isolation
and recovery of the radioactive elements.
U. S. Pat. No. 3,764,552 discloses a method for storing
radioactive combustible waste material. The method comprises
the steps of placing the waste material in a container provided
with oxide getters selected from the group consisting of
magnesium oxide, calcium oxide, barium oxide, and strontium
oxide in an amount sufficient to react with sorbed water and
combustion products formed by oxidation and pyrolysis of the
waste material. The container then is sealed and heated to
pyrolyze the waste.
, _ . ,, _ .... . .
31(~3;~
In other processes, an actinide-containing waste material
is combusted and encapsulated. Examples of patents relating
to such processes are U. S. Pat. Nos. 3,00~,904; 3,262,885 and
3,332,8~4.
The combustion per se of carbonaceous fuels and
carbon-containing wastes in a molten alkali metal salt for
various purposes is kno~n. U. S. Pat. No. 3,710,737 shows
the generation of heat for external use employing a variety
of carbonaceous materials. U. S. Pat. Nos. 3,567,412,
3,708,270 and 3,916,617 show the use of such techniques
for the production of pyrolysis gases. In U. S. Pat. Nos.
3,77~,320 and 3,845,190, such techniques are involved,
respectively, in the non-polluting disposal of explosives
and of organic pesticides. In U. S. Pat. No. 3,8g9,322,
valuable metals are recovered from organic scrap in a
molten salt bath. None of these patents are concerned with
the treatment of radioactive wastes for the isolation and
retention of volatile radioactive elements.
SUMMARY OF THE INVENTION
In accordance with the present invention, there is
provided a method of treating organic wastes containing at
least one actinide element to reduce the volume of said waste
and recover the actinide element therefrom. Broadly, the
method comprises introducing an actinide-containing organic
waste and a source of gaseous oxygen, such as air, into a
molten salt comprising an alkali metal carbonate. The bath
is maintained at a temperature of about 400C to 1000C and
a pressure of from about 0.5 to 10 atmospheres to thermally
3~.0~
decompose and at least partially oxidize the actinide-
containing waste. Under such conditions the volume of
organic waste is substantially reduced, and combustion
products are formed, which include a gaseous effluent
consisting essentially of carbon dioxide and water vapor.
The gaseous effluent is vented to the atmosphere preferably
through a series of filters to remove any trace amount of
the actinide element which may be entrained therein, or
particulate alkali metal salts which may be entrained in the
gaseous effluent. The remaining combustion products of the
organic waste are retained in the molten salt as will be
explained more fully later. At least ~ portion of the
molten salt containing the combustion products is withdrawn
and mixed with an aqueous medium. The aqueous medium then is
treated to remove the insoluble combustion products forming a
substantially solids-free solution. The insoluble combustion
products formed contain the actinide element and are leached
with an inorganic acid to solubilize the actinide elements and
recover them from the combustion products. The actinide
elements are readily recoverable from the acid solution
utilizing conventional solvent extraction or anion exchange
resin techni~ues known to those versed in the art.
Preferably the alkali metal carbonate is sodium carbonate
which may optionally contain from about 1 to 25 wt.% of
sodium sulfate. An advantage of using sodium carbonate is
that it is lower in cost, and further, when substantial
portions of the carbonate are reacted with other consti-
tuents of the waste, such as halides and sulfur, to form
--7--
10~ 1033
alkali metal halides and sulfates, the sodium form is more
readily amenable to processing to recover the sulfur and
reform the carbonate constituents for recycle in the process.
When sodium carbonate is used the preferred range of temper-
ature and pressure is from about 800C to 900C and 0.8 to
1.0 atmosphere, respectively.
When the organic waste also contains sulfur and halogen
constituents, the sulfur and halogens react with the alkali
metal carbonate to form alkali metal sulfates and halides
which are retained in the molten salt. Such wastes are typical
of those which result from processing nuclear reactor fuel,
for example. When such wastes are used as a feed material,
there also is provided a method wherein the carbonate and
sulfate component of the molten salt is recoverable for recycle.
A method also is provided for the separation of the alkali
metal halide component for disposal.
In its preferred aspects, the combustion is a complete
one, whether or not air, oxygen-enriched air, or fine oxygen
is used as the oxygen-containing gas. Although air is gener-
ally preferred, pure oxygen can be used where it is desiredto reduce the volume of gaseo~s products.
BRIEF DESCRIPTION OF THE DRAWING
The sole figure is a block diagram illustrating the
method of the present invention.
DESCRIPTION OF PREFERRED EMBODIMENTS
The present invention relates to the treatment of
combustible organic waste containing one or more elements
of the actinide series. A significant source of such waste
--8--
arises from the operation and utilization of atomic energy
such as nuclear power plants. The waste resulting from the
operation of such plants can be solid, liquid or gaseous.
In addition to the radioactive waste produced in the reactors
involved, other wastes are developed in operation and maintenance
of those installations. For example, tools, equipment and
clothing that are used in close proximity to these installations
may themselves become radioactive. Similarly, paper, rags
and organic solvents used in such installations may also be
contaminated as a consequence of such use. Further, in
processing or reprocessing of the used fuel elements, waste
materials are developed. The present invention is applicable
to all combustible organic wastes, including such items as
plastic, paper and rags. In view of the importance of
having a convenient, rapid, safe, effective method of economic
interest for complete treatment and reduction of volume of
radioactive actinide-contaminated organic waste, the present
invention will be particularly described with reference to
the treatment of such waste, and particularly to the recovery
of uranium and plutonium of the actinide series from such
waste.
In accordance with the present method, the molten salt
may consist of a single alkali metal carbonate or a mixture of
alkali metal carbonates. Preferably, the molten salt also
will contain a minor amount of alkali metal sulfate, since it
has been found the presence of sulfate enhances the combustion
; rate of organic materials. The organic waste is combusted;
_g_
~ O 3 ~
some of the combustion products being retained in the molten
salt, while others are further oxidized to CO2 and H2O.
When it is desired to effect the combustion of tne waste at
a relatively low temperature, a low melting binary or a
ternary mixture of alkali metal carbonates may be utilized.
For example, the ternary alkali metal carbonate eutectic
consisting of 43.5, 31.5 and 25.0 mole % of carbonates of
lithium, sodium and potassium, respectively, melts at 397C.
A preferred binary mixture is the sodium carbonate-potassium
L0 carbonate eutectic which melts at 710C. The alkali metal
sulfate utilized may consist of any of the sulfates of the
foregoing alkali metals. In general, sodium sulfate is
preferred because of its ready availability and low cost.
Referring to the drawing, an overall flow diagram for
the method is shown. The waste, after manual sorting, is
shredded in a shredder 1 and fed with air into a molten salt-
containing furnace 2. Under steady state conditions, a
typical composition of salt in the furnace is approximately
50 wt.% sodium carbonate, 10 wt.% sodium sulfate, 20 wt.%
sodium chloride and 20 wt.% ash. The sodium chloride results
from, for example, the combustion of plastics such as poly-
vinyl chloride ~its chloride content is converted to sodium
chloride in the molten salt). The ash principally comprises
the inorganic constituents of the waste. From processing
and viscosity considerations, an upper limit for the inorganic
ash concentration (insolubles) of approximately 25 wt% has
been set assuming an operating temperature of about ~00C.
-10-
l(J'~iO3;~
The desired steady state concentrations are maintained by
periodically or continuously draining a portion of the
molten salt for treatment to remove the insolubles and the
alkali metal halide content and further provide for recycle
of the alkali metal carbonate.
In operation, the alkali metal carbonate serves as a
heat transfer medium for the combustible material and also as
a neutralizing agent for any acidic gaseous combustion products
from the waste material such as hydrochloric acid and sulfur
dioxide. The actinide-contaminated organic waste and air are
fed continuously into the molten salt furnace 2 below the surface
of the salt, so that all gases formed during combustion are
forced to pass through the molten salt before being released
into the atmosphere. Thus, in accordance with the present
method, the only gaseous combustion products involved are CO2
and water vapor. The effluent gas also will contain some
unreacted oxygen and nitrogen when air is used as a source of
; gaseous oxygen. The inorganic content of the waste, such as
metal, is oxidized and retained in the salt principally as the
ash. Another advantage of the present invention is that the
temperatures of combustion are sufficiently low so that no
significant amount of nitrogen oxides are formed during combustion.
In addition, the actinide elements do not form refractory
oxides in the molten salt even at high temperatures, i.e.,
750 to 1000C.
- The melt-ash mixture withdrawn from the furnace by
way of a conduit 3 preferably is mixed with recycled solution
by way of a conduit 4 to dissolve the soluble sodium salts in
a vessel 5. The solution then is treated using any conven-
3 tional solids liquid separation technique, such as a centrifuge
--11 -
1~9 103~
or the like, to separate and remove the insoluble ash by way
of a conduit 6. Advantageously, the ash is leached with an
inorganic acid at 7 to solubilize the actinide constituents
for recovery and for use, for examp e, as a source of nuclear
reactor fuel. Suitable inorganic acids include HCl, HN03,
HF and mixtures thereof.
The substantially solids-free filtrate then is treated
in a series of evaporative crystallizer zones 8 and 9 at
succeedingly lower temperatures to remove first a sodium
carbonate or mixed sodium carbonate-sodium sulfate fraction
and then a sodium chloride fraction. This phase of the
method is somewhat analogous to the treatment of brines
commonly practiced, for example, at Trona, California, in
the treatment of the Searles Lake brine. More particularly,
it has been found that it is possible to achieve a selective
precipitation of salts in the mixed brine, whereby it is
possible to remove a first crop of carbonate or mixed sodium
æulfate-sodium carbonate crystals frequently referred to as
burkeite, and subsequently, at a lower temperature to recover
a second crop of crystals consisting essentially of alkali
metal chlorides substantially free of carbonate or burkeite.
Examples of patents relating to the separation of
sodium salts from mixtures thereof are U. S. Pat. Nos. 1,836,426
and U. S. 2,347,053. Generally, the first crop of crystals is
obtained by boiling the filtrate in an evaporator crystallizer
operated at a temperature of between about 60 and 170C.
After separation of this crop of crystals, the substantially
crystal-free liquor is then introduced into a second evaporator
crystallizer, which preferably is maintained under a vacuum,
-12-
0;~
and the liquor is cooled to a temperature of less than about43 C. Generally, a temperature of 20 to 35C is used to
produce a second crop of crystals, comprising the alkali metal
chlorides, substantially free of alkali metal carbonates and
sulfates.
It is seen, therefore, that the present invention
provides a closed cycle and produces a substantially pollutant-
free offgas containing carbon dioxide and steam. In addition,
there are no liquid wastes in the process, but rather two
solid products for disposal, namely the ash and sodium chloride.
If desired, these can be consolidated by melting the sodium
chloride, adding the ash and solidifying the mass. Alternatively,
they may, of course, be kept separate for different disposal
options.
The following examples are set forth to further illustrate
the practice of the present invention, and are not intended to
be construed as limited in scope.
EXAMPLE 1
The following example demonstrates the application of
the present invention to a waste which simulates the character-
istics of radioactive transuranic waste materials. Two
different ash-melt mixtures were used in these combustion
tests. The first mixture consisting of 16 wt.~ ash and
16 wt.% NaCl, lQ wt.% ~a2SO4, and 58 wt.% Na2CO3; the second
mixture consisting of 20 wt.% ash, 20 wt.% NaCl, 10 wt.~
; Na2S04 and 58 wt.~ Na2CO3. These two mixtures were selected
as representing the concentration extremes which would be
anticipated in the actual treatment of radioactive actinide-
containing waste. The composition of a typical waste was
obtained from the Los Alamos Scientific Laboratory and
comprised paper and plastic mixtures having essentially the
same heating value, ash content, ash composition and halide
-13-
10~103;~
content found in actual radioactive-contaminated waste. The
waste and air were introduced below the surface of the
molten salt-ash mixture in a furnace. The offgas from the
combustion was monitored for N2, 2~ C02, CO, HC and NOx to
determine the completeness of combustion. Particulate
samples of the cffgas also were collected to determine the
particular loading and composition of the particulates. The
offgas was scrubbed with aqueous sodium hydroxides to
determine if any gaseous chloride escaped from the melt and
passed through the particulate filter. None was found. The
results of these tests are shown in Table 1.
-14-
l~t'3103;~
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* I ~ ~ ~ I l, o o C~ o ,~ o V
~ t~ ~ b,O 1 o o o o o o o o o
~ ~ l ~.
V I . .
~ ~ lo~ o o~ou~ ~
U~ I O ~ I ....... ...... Ti
~ C ~ c~ _ ~o r\ u~ L~ U~ 1' ~
~ ~ ;o~ u~ ~D ~ ~ 0~ ;t
~31 ~
i ~ ~ ~ ~ ~ i ~
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H O V
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lV~ 33
From the foregoing table, it is seen that the nitrogen
oxide emissions are less than 20 ppm, and CO and HC emissions
are below the detection limits of the instruments. Further,
it is seen that below about 800C the particulate emissions
are within the EPA standards for particulate emissions from
incinerators; i.e~, 0.1 grains/scf (0.2 gm/m3). It was found
that the amount of particulate is directly related to the
sodium chloride vapor pressure and mole fraction of sodium
chloride in the melt. Specifically, analysis of the particulates
indicates that at 800C they are essentially sodium chloride
crystals.
EXAMPLE 2
The following example demonstrates the treatment of
the melt-ash mixture to recover the sodium carbonate and sulfate
for recycle to the combustor and to separate an ash and sodium
chloride fraction for disposal. The melt-ash mixture was quenched
in an aqueous mediu~ comprising sodium carbonate, sulfate and
chloride. The aqueous medium then was filtered to remove the
insoluble ash constituents. The filtrate was subjected to a
first fractional crystallation at a temperature of about 106C
to precipitate a first mixed crystal crop of sodium carbonate
and sodium sulfate, which was removed by filtration. The filtrate
from this step was introduced into a second crystallization
zone wherein sodium chloride was precipitated at a temperature
of about 35C. Samples of the crystals were obtained and
analyzed and the results are listed in Table ~.
-16-
iO3;~
TABLE 2
COMPOSITION OP` CRYSTALS OBTAINED IN
AQUEOUS PROCESSING TESTS
_ ..
Composition
Crystallization
Flow Sheet Experimental
(Calculations~ Results
:
1st 84% Na2CO3 80~ Na2CO3
16% Na2SO4 12% Na2SO4
8% NaCl
2nd 100% NaCl 96% NaCl
4% Na2CO3
0.2% Na2SO4
For comparison, the theoretical composition of the crystals,
based on solubility calculations, also is given.
From the foregoing table it is seen that the carbonate-
sulfate crystals were substantially free of sodium chloride;
i.e., less than 10%. Further, in the second crystallization
it is seen that sodium chloride was obtained substantially
free of carbonate and sulfate. Since the crystals were not
washed, the difference between the experimental results and
the theoretical is due primarily to the retention of mother
liquor in the crystals. Based on weight changes, when the
crystals were dried, and the composition of the solution from
which the crystals were separated, it is found that nearly
80 wt% of the undried crystal weight was mother liquor in the
first crystallization and about 40 wt% in the second
crystallization. Thus, the use of a centrifuge instead of a
filter, for example, to separate the crystals would reduce the
liquor retention in the crystals, and thus even further improve
the crystal purity.
-17-
10'~11~33
EXAMPLE 3
The following example will demonstrate the disposition
of the actinide elements when treated in accordance with the
present method. Four different plutonium and uranium compounds
were added as solutions to separate ash-melt mixtures prepared
from combustion tests. One hundred grams of crushed-ash melt
mixture was placed in a 1/2 inch alumina tube, and a plutonium
compound (approximately 2 ml) and corresponding uranium compound
(approximately 2 ml) were placed in the center of the ash-melt
mixture. The mixture was melted, and gas was bubbled through
the melt to simulate the agitation present during combustion.
The test conditions are given in Table 3.
After each test the alumina tube was transferred to an
analytical glove box. The alumina tube was cracked away from
the solidified ash-melt mixture which was subsequently crushed
to less than about 400 mesh. About 12.2 grams of the ash-melt
mixture were placed in a reflux flask and 25 grams of a recycle
solution (dilute sodium chloride-carbonate-sulfate solution)
were added. The ash-melt mixture was refluxed with the recycled
solution for two hours. The mixture then was filtered to
separate the ash from the solution~
Aliquots of the reflux filtrate, the water used to wash
any particulates out of the condenser and sections of the
filter were analyzed to determine if any plutonium or uranium
was carried out of the melt during simulated combustion. The
uxanium in the various fractions was determined by a fluorometric
procedure.
-18-
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1033
The disposition of plutonium and uranium in the various
processing fractions and components is shown in Table 4. The
percent of the original plutonium and uranium in the ash
fraction was obtained by difference. In addition, one test
was made to determine if plutonium dissolved from the ash
fraction. The ash was leached with a mixture of HF-HN03 and
plutonium so dissolved was determined coulometrically, the
result confirming the percent calculated by difference.
From the table it is seen that g9.9% of plutonium was
present in the ash fraction. Only a small percentage of the
plutonium (generally less than 0.1%) was present in the reflux
solution. Only a negligible amount of plutonium was carried
away with the product gases, generally less than about 0.01%.
Further, it is seen that essentially no difference in the
particulate evolution is observed with gas-melt superficial
velocities of from about 1 to 2 ft/sec. In the actual combustion
of radioactive-actinide-containing waste it is anticipated
that superficial velocities of from 0.5 to about 2 ft/sec will
be satisfactory.
The uranium results are similar to that of plutonium.
As with plutonium, only trace amounts of uranium were found in
the heat exchanger and filter used. However, unlike plutonium,
it appears as if a small fraction of the uranium may dissolve
in the reflux solution. Nonetheless, by far, the majority of
uranium is present in the ash.
-20-
10'~1033
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EXAMP~E 4
The following example is set forth to demonstrate the
combustion of an actinide-containing organic waste to reduce its
volume followed by the recovery of the actinide from the
combustion products. Three combustion tests were conducted with
plutonium-contaminated waste. The waste was a mixture of paper,
paper products, plastic (a mixture of polyethylene and polyvinyl
chloride), and rubber. The waste was contaminated with plutonium
by adding a known volume of plutonium nitrate, sulfate, and
chloride solutions to the waste. After contamination, the
waste was mixed thoroughly. In two of the tests, the plutonium
concentration in the waste was 9 x lO 5 grams per gram of waste,
and in one test the plutonium concentration was l.l x 10 3 grams
per gram, which corresponds roughly to the level of plutonium
expected in low level and intermediate level actinide-containing
~ wastes, respectively.
! The waste was introduced into a molten sodium carbonate
bath maintained at a temperature within the range from about
850 to 905C. The off-gas from each test was monitored and
analyzed for plutonium content. By difference it was determined
that about 99.9% of the plutonium was retained in the melt.
To demonstrate that the actinide ~plutonium) could be
recovered from the spent salt, solidified carbonate from the
foregoing test was dissolved in water and filtered. The insoluble
ash (containing the plutonium) was leached with various inorganic
acids. The results are shown in Table 5.
-22-
,~
10~31033
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-~3-
1.0~ 10:33
From the foregoing table it i~ seen th~t greater than 90%
;, of the plutonium is recovered. Further, as a result of the
foregoing examples, it was determined that a substantial
reduction in the volume of organic waste was obtained. More
particularly, even when the entire melt-ash mixture is
withdrawn for disposal, the weight of the organic waste is
reduced by a factor of 2.5, and the volume of waste is
reduced by a factor of 25. However, in accordance with the
particularly preferred embodiment, wherein the melt-ash
mixture is treated to recover and recycle the carbonate-
sulfate fraction, the weight is reduced by a factor of 5.7,
and the volume of waste is reduced by a factor of 57. Thus,
, the foregoing examples clearly demonstrate the efficacy and
advantage of the present method for the treatment of actinide-
containing waste.
While the present process has been described with
regard to certain particular sources of waste, actinide
elements, process conditions, temperatures, concentrations
and the like, and has been illustrated, in part, with various
synthetically prepared radioactive wastes, it will be readily
apparent to those versed in the art that many variations
thereof may be used. Accordingly, this invention is not to
be limited by the illustrative and specific embodiments
thereof. Rather, its scope should be determined in accordance
with the following claims.
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