Note: Descriptions are shown in the official language in which they were submitted.
5 ~ ~
U
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PRoc)uc~rI ON OF TRI TI UM
BACKGRC)UND OF THE INVENTION
This invention relates to the production of
trltium and more particularly to the economical
production of tritium in a power-generating nuclear
reactor.
Tritium has generally been produced from the
lithium isotope of the mass 6 by the absorption of slow
neutrons and the resultant transmutation to produce an
atom of tritium and an alpha particle (i.e., helium
: nucleus~, which is referrPd to as an (n~cx~ ~ raaction.
~ithium naturally contai~s about 7.49~ o the isotope
Li and the remainder of the isotope Li .
Commercial production has been carried out using lithium
in ~ts natural or enriched form within a sealed,
doub~le-wall container 1hat is hydrogen impermeable which
~s lo~ated in the core of a plut~nium production
~ reactor. Lithium has been used in the form of an alloy
: with magnesium or aluminum a~d has also been used in
oxide form. These large wa~er~cooled reactors, which
have been either graphite-moderated or heavy
~ater-moderated, produce tri~ium essentially as a
by~product of plutonium production, and as a result of
the e~er-increasing demand for tritium ~which will be
magnified substantially when fusion technology ~ecomes
economically feasible), additional economic me~hods for
the production of tritium are deslrable.
RIEF SUMMP~RY OF THE INVENTION
i I~ has been found that high temperature
gas cooled graphite-moderated reactors ~HTGRs~ offer a
unlque opportunity for the dual production of tritium
and usable electric power, uslng today's technology.
These reactors inherently have a relatively high
converslon ratlo which makes excess neutron~ available
3S ~or breedin~ fertile material from ~is~i}e material and
~.~
1 1555~5
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have hereto~ore been employed to breed uranlum-233 from
~horium. By using minute coated particles containing
lit~ium, which serve as individual pressure vessels that
retain tritium, as target material within the fuel
elements that constitute the core of a gas-cooled
graphite-moderated nuclear reactor, the p~roduction o
txitium can be carried out without disrupting the
power-generating function of such a nuclear reactor and
without the creation of significant safety hazards.
Substantially all of the tritium created in the lithium
t~rget particles is retained therewithin; however,
~hould minor amounts of tritium escape outside the
coatings, it is recoverable from the primary gas coolant
stream. Following the fuel lifetime of the ~uel
~lements, the tri~ium is recovered. Usually the fuel
elements are first removed from th0 reactor, and the
tritium is then recovered by heating in a facility
designed for that purpose. The recovery o~ the tritium
may also be carried out as a part of the reprocessing of
the overall fuel element and the reclaiming of the
remaining fissile uranium.
BRI F DESCRIPTION OF_THE DRAWINGS
FIGURE 1 is a perspective view of a fuel
element designed for use in the prismatic core of a
gas-cooled nuclear reactor which may be employed for the
production o~ tritium;
FIGURE 2 is a diagrammatic view showing a
nuclear reactor including a reactor core formed of fuel
elements of the type shown ln FIG~RE l;
FIGURE 3 is a ~ragmentary sectional view
throuqh one o~ the fuel elements shown in FIGURE 1
FIGURE 4 is a fragmentary sectional view of an
alternative embodiment of a fuel element similar to
FIGURE 3; and
- FX~URB 5 is a view, enlarged in s~ze, of a
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target particle of the type whlch may be employed in the
~uel element of FIGURE 1.
DETAI~,ED DESCRIPTION OF T~E PREFERRED EMBODIMENTS
.. . . . , . .. ~
Gas-cooled high temperature graphite moderated
nuclear reactors have been developed which utilize heat
produced by nuclear fission to produce steam in steam
- ~enerators which is then used to drive
electricity-producing turbines and also to provide heat
or other applications. Such reactors have utilized
cores made up of prismatic block-type fuel elements
which contain coolant hvles that extend axially
therethrough and that are spaced between parallel,
elongated fuel chambers, and have also utilized'cores
made up of a ~ed of sp~erical graphi~e balls which are
located randomly within a pressure vessel and provide
passageways for coolant gas through the interstices of
the bed (which is generally referred to as a Pebble-Bed
reactor core). Particulate nuclear fuel for such fuel
elements has been provided in the form of minute
particles having coa~ings which ~unction as
fission-product-retentive pressure vessels. Such fuel
particles include a core of fissile material, e.g.,
uranium enriched in isotope-235 in carbide form or in
oxide form or as a mixture of uranium carbide and
uranium oxide. Such cores often also include thorium as
a diluent and fertile materia}. Examples of such
nuclear fuel particles are described in detail in U.S.
Patent No. 3,649,~52 issued March 14, 1972, to Jack
Chin, et al., the disclosure o~ which is incorporated
herein by reference. Such nuclear fuel particle designs
have been found to be particularly well-suited for
gas-cooled nuclear reactorsO
Illustrated in the drawings is a gas-cooled
nuclear reactor of the prismatic core type which
utillzes ~uel elements 11 in the form of graphite blocks
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1 ~555~
13 havlng hexagonal top and bottom suraces lS
interconnected by perpendicular side faces 19 as shown
~n FIGURE 1. The nuclear reactor core 21 is formed from
a plurality of vertical columns o such fuel elements ~1
s~acked one atop another, arranged within a pressure
vessel 23 as illustrated in FIGURE 2. The fuel element
block 13 ls formed with a plurality of coolant holes 25
located on a constant ~rian~ular pitch. The coolant
holes 25 extend from top to bottom axially through the
blocks and provide the passageways for the gas coolant,
preferably helium, to extract the heat ~rom the nuclear
~ission reactions~ ~uel chambers 27 of slightly lesser
diameter than the coolant holes 25 are located in a
triangular array of lesser pitch so ~hat each coolant
hole is surrounded by a number of fuel chambers.
To ~acilitate alignment of the indivi~ual fuel
~- elements 11 in stacked columns, ~he blocks 13 are each
provided with short pins 31 which protrude from the top
end surface of each fuel element which are received in
corresponding cavities provided in the bottom end
surface. The pins 31 and cavities are aligned with
individual coolant holes.
The reactor core 21 made up of these columns of
fue~ elements 11 is located within a pressure vessel 23
o~ a pre- or post-stressed concrete or the like, which
may llave a steel inner liner 33. A surrounding
reflector 35 is prov~ded by un~fueled prismatic blocks
of qraphite, as is well known in the nuclear reactor
art. The primary helium coolant is circulated downward
through the core 21 by blowers or circulators 37 which
are preferably disposed in suitable cavities located ln
the pressure vessel ~3. To conine the primary coolant
stream to the pressure vessel 23, steam generators 39
are also provlded in cavit~es within the pressure
vessel. The primary hellum coolant is accordingly
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1 1S~5~5
circulated by the blowers 37 ~hrough the reactor core 21
~here it picks up heat and then through the steam
generators 39 where it exchanges its heat with a
secondary coolant stream in the form of water which is
being turned to steam. The blowers 37 take suction from
! the steam generators 39 and circulate the primary
coolant back downward through the core for another pass~
It has been found that only very minor
modifications to an existing ~TGR design are needed when
target particles containing Li6 are included along
with nuclear fuel in the individual fuel elements ll.
The target particles are neutronically compatible with
the nuclear ~ission reac~ion and can be employed in the
place of burnable poison (e.g. boron) that would
normally be included in such ~uel elements and also in
place o~ most o~ the ~ertile thorium. In fuel elements
ll of pr~smatic design, the fuel chambers 27 generally
define the volume available for disposi~ion of nuclear
fuel material. When an HTGR ~s used for the production
of tritium, only approximately 3% to about 5~ of the
volume of the fuel chambers is occupied by the target
particle material, and the remainder o the available
volume holds nuclear fuel material.
Li6 has an extremely large cross section,
equal to about 953 barns, for the absorption o~ neu~rons
in the thermal energy range and the consequent
transmutation to produce tritium and helim. As a
result, lithium i~ inherently self-shielding, and in
order to induce efficient conversion of Li6 to
tritium, it is important to dlsperse the lithium
throughout the reactor core. Excellent dispersal is
achieved by forming small cores or kernels of a Li
~ompound, having a size on the order of about 300 to
lO00 microns, an~ spaclng these cores from one another
by means of exterior coatings which totally surround the
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cores. By locatin~ some of the target material in each
o~ the fuel chambers 27, overall dispersion 19 still
further enhanced.
. In the fuel element design depicted in FIGURE
3, short sticks or rods 41 of target particles joined
together by sui~able bonding material, such as
carbonized pitch, are located adjacent to .the top and
bottom ends of each of the fuel chambers 27, with
~imllar rods 43 of nuclear fuel material filling up the
lQ major portion of the fuel chambers 27 in the regions
between the top and bottom target material rods. .
Graphlte plugs 45 close the upper end, and
. heat-decomposable plastic spacers 47 are provided. This
arrangement facilitates the selective recovery of
tr~tium by physically slicing the upper and lower e~ds
o~ ~he fuel elements 11 from the remainder and then
reprocessing these sllced por~ions separately for the
recovery of tritium, as discussed hereinafter. An
: . alternative version of a fuel element 11' is illustrated
in FIGURE 4 wherein the target material is ~ormed into
: ~airl~ flat wafers 41' which are.then disposed
alternately between adjacent nuclear fuel rods 43
t~roughout the entire length of each fuel chamber 27.
Such an arrangement facilitates even better dispersal
25 throughout the nuclear reactor.core; however, it.doe~
not facilitate selective tritium recovery by processing
only portions of the indiv.ldual fuel elementsO
- .~. Examples of representa~ive target particles 51
. ~are. depicted in ~IGVRE S and ~nclude minute cores or
Xernels 53, pre~erably spheroidal in shape, which are
between about 300 and about 1000 ~m in diameter. The
kernels 53 are surrounded by an inner coating region of
a generally p~rous nature which acommodates the build-up
o~ helium and tritium from the transmutation of the
lithium and wbich is in turn ~urrounded by an outer
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coatlng reglon which forms the gas-tlght barrier that
prevents escape of the hellum and tritium.
The cores 53 are formed ~rom a solid compound
of lithium which is preferably stable at the
temperatures employed for the vapor-deposition of the
surrounding coatings. Lithium in oxide form, elther by
ltself or in a combination with another refractory-like
element, may be employed as the kernel materials.
~xamples are lithium oxide, lithium aluminate
ILiA102), and lithium silicates (Li2Sio3)
~Li4Sio4). The l~thium compound should have a
melting point and other characteristics which render it
compatible with the coating processes. It can be
employed in any form in which the cores have sufficient
mechanical stability to render them physically suitable
to treatment in a vapor-deposition coater. For example,
small kernels can be formed by a powder agglomeration
process or by cold-pressing in steel dies and then
` ~intered to provide strength and higher density. For
example, lithium aluminate powder can be cold-pressed in
a 2ie at about 3000 psi and then sintered in a cacuum at
~bout 1200 C. for eight hours. Kernels made by powder
agglomeration can also be sintered to provide mechanical
~trength. I~ high density is desired, the sintered
kernels can be made spheroidal by be~ng dr~pped through
a hot zone at between 180~ C. and 2200 C. to cause
them to melt and densify into spheroidal shapes in
accordance with known technology.
Generally, the kernels 53 are densified to at
least about 70% of theoretical density. By theoretical
den~ity i~ meant the maximum density for that part~cular
stoichlometric compound. ~he preferred iithium compound
is lithium aluminate ~hich has a theoretical density of
about 2.55 grams per cm3~ Although densification to a
35 ~ density approaching theoretical density is possible, it
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may be preferred to employ kernels 53 in the range of
about 70% to ~0~ of theoretical density from the
standpoint both of spa~ial dispersion and ultimate
~ccommodation of the gaseous products vf the lithium
transmutation.
To prevent the reaction of lithium aluminate
with subsequent vaporous materials to which there will
be exposure durinq the coa~ing operations, an impervious
carbon seal layer 55 is applied at a temperature below
1100 C. and preferably at a temperature not higher than
about 1000 C. Such a seal coating may be depositecl in
a particulate bed 1uidized by gas flow or in a rotating
drum or other ~ype o~ agita~ed bed coater. Pyrocarbon
seal layers should have a density of about 1.8 to 2.0
gram/cm3 and are preferably oriented. A thickness of
~bout 10 to 30 microns o such pyrocarbon prov'~es an
adequate seal coating and can be deposited from a
mixture of acetylene plus an inert gas such as argon~
~he porous layer 57 that is provided for the
accommodation of the helium and tritium within the
minute pressure vessels is preferably pyrocarbon having
a density between about 0.9 and 1.2 gram/cm3. The
thickness o the poro~s pyrocarbon layer 57 is dependent
upon the amount of Li6 included within the kernels 53
and the pressure which the ou~er gas-tight coating is
designed to wlthstand.
If there are no constraints on the amount of
~pace occupled by the target particles in the nuclear
reactor core, larger amounts o~ porous material can be
included so as to prevent the build-up of high gas
pressures wi~.hin the gas-tight outer barrier. On the
other handl if particular constraints llmit the amount
o space, a lesser thickness of the porous pyrocarbon
may be employed along with a slightly thicker outer
coating, which will withstand the higher ga~ pres~ure
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bu~ld-up. In general, for cores in the 300 to 500
micron range, it is likely tha~ at least about 75
m~ crons of porous pyrocarbon would be used.
In the outer coating which provides the
gas-tight barrier to prevent the escape o tritium, the
key layer 59 is one of dense silicon carbide or
zirconium carbide. The reactor may be operated so that
the temperature of the target particles may be in the
range of about 900 to 1000 C.~ at which both dense
10 silicon carbide and zirconium carbide provide an
effective barrier to the passage of tritium. As ln any
~uch barrier material, the thicker ~he material, the
more efective the barrier, and it is expected that at
least about 50 microns of SiC or ZrC would be used~ For
axample, a silicon carbide layer having a thickness of
9~ microns or even greater might be employed. The
carbide barrier layer should ha~e a density of at least
95% ~ theoretical densityO Deposition of silicon
~arbide from a vaporous atmosphere can be ~airly readily
carried out to produce deposits having greater than 99~
of theoretical density. For example, for SiC, which has
a theoretical density of 3.22 grams/cm3, densities
greater than 3.20 grams/cm3 can be achieved.
Disposed immediately interior and exterior of
the carbide layer 59 ara layers 61, 63 of isotroPic
pyrocarbon having a density between about 1.7 and about
2.0 grams/cm3 and havin~ individua} thicknesses of
between about 35 and 45 microns. Such isotropic
~oatings are deposited from a mixture of acetylene,
propylene and inert qas at a temperature of about 1350
C. which have a BAF IBacon Anlsotrophy Factor) of less
than about 1.05. The interlor pyrocarbon layer 61
retards the outward diffusion of mater$als from the core
to the silicon carbide during irradia~ion in the reac~or
core and, during the process when the sllicon carbide is
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being deposited~ serves a~ a further barrier to prevent
chlorine (which is present ~n the coating atmosphere)
~rom reaching the kernel where undesirable chemical
reactions may occur. The exterior pyrocarbon layer 63
has a larger strain to ~racture than the relatively
brittle silicon and thus provides mechanical handling
strength for the target particles following completion
of the coating operation, particularly dur~ng bonding of
- the particles with pitch or the like to form short rods
before loading into the fuel chambers. During operation
ln the reactor core, the isotropic pyrolytic carbon
layer 63 undergoes a con~rolled shrinkage as a result of
exposure to hi~h temperature and ~ast neutrons, and it
shrinks onto the silicon carbide layer 59 placing it ln
compression and increasing its strength as a minuta
pressure vessel.
- - Although the employment of an outer coating
which includes a layer of silicon carbide or zirconium
carblde sandwiched between isotropic pyrocarbon layers
is preferred, other su;table ~as-tight coatings can also
be employed. For example, oriented pyrocarbon has
proved to be extremely effective in retaining tritium,
and it is contemplated that a single layer of such
pyrocarbon might be disposed exterior of the porous
layer 57. For example, pyrocarbon having a BAF of about
1.1 to 1.4 and a density between about 1.85 and 2.0
gram/cm3 may be employed. Such a layer should have a
thickness of at least about 70 microns, and depending
upon the size of the llthium kernels and the amount of
porous pyrocarbon employed, a layer up to about 200
microns thick might be used. Such oriented pyrocarbon
la~er can be deposited from an atmosphere of acetylene,
propylene and argon at a temperature o about 1350 C.
in a fluidi~ed bed coater uslng a lower coating rate
than is used to deposit lsotropic pyrocarbon.
5 ~ ~
It is contemplated ~hat the core of a 1000 ~Wt
HTGR could be designed w1~h approximately 10V0 kg of
U-235 and 140 kg of lithlum ~measured as lithium
metal). Operatlon of such an HTGR for about two years
would result in fissioning of about 80% and 90% of the
fissile isotopes of uranlum, in the transmutation of
over 90~ of the Li6 nuclides and in the recovery of
about 5 to 6 kq o tritium. Although such a
batch-loaded cycle wherein all the fuel el~ments are
loaded and discharged at the same time appears the most
promising, several fuel management schemes seem
feasible, including a 3-year graded cycle in which every
year 1/3 o the fuel elements are reloaded~ The batch
~cheme has the advantage of having a flatter power
across the core and therefore lower fuel temperatures.
The HTGR has the uni~ue advantage to allow
var~ous such schemes through balancing the reactivity
effect of the uranium depletion with th~ shieldin~ of
the lithium. During operation in all such fuel
management schemes, the HTGR produces electricity in the
same manner as the existing HTGRs, which are
economically compe~titive on the basis of electricity
production aloneO Of course the fuel cost would be
somewhat higher because it would not be offset by
breeding ~airly large amounts of U-233 from the fertile
thorium; however, this additional fuel cost would be
more than of~set by the economic value of the tritium
produced. Accordingly, the dual operation of the ~TGR
~oth as an electricity producer and as a tritium
p~oducer appears to be extremely attractlve economically.
As earlier indicatedt the lithium target
particles 51 could be restrlcted to zones adjacent to
the top and bottom o~ prlsmatic fuel elements, which
¢ould then be severed and reprocessed separately for the
recovery o~ tritium, or the target material could be
5 ~ 5
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distributed throughout the ~uel chambers. In either
lnstance, heating of ~uch target particles to a
temperature o about 1300 - 1400 C. or above effects
the relatlvely prompt diffusion of tritlum througb a
~ilicon carbide barrier layer, which was very efective
in restraining passage of tritium at lower
temperatures9 For example, autoclaving o~ the fuel
elements, or the remains thereof/ at temperatures of
1500 C~ can be employed to release the tritium into a
lQ controlled atmosphere. Alternatively, if the tritium
recovery is to be effected as a part of the overall
recovery treatment of the spent fuel elements, the hex
block fuel elements can be crushed and burned in a
controlled atmosphere at temperatures below about 800
C~ to destroy the graphite blocks and the isotropic
- pyrolytic carbon exterior layers. Tbereafter, crushing
the carbide barrier layers in a controlled environment,
coupled with heatinq to a temperature of at least about
500 C., woula quickly e~fect release of the tritium
2Q being held therewithin, and thi~ procedure is preferred
~or ZrC coatings which are more resistant to hydrogen
dif~usion at hlgher temperatures. ~ltimate reaovery of
tritium ~T) from a gaseous atmosphere is pre~erably
effected by conversion of the tritium to T2O by
oxidation using a suitable oxygen source, such as copper
oxide. T20 has physical characteristics quite
~milar to ordinary water and is then removed from the
gas stream by a molecular sieve or by freezing in a
~uitable cold trap, such as liquid nitrogen.
Alternatively, tritium can be recovered as a hydride,
instead of being oxidized, by exposure to zircanium or
t~tanium sponge metal~
Although the foregoing containment syYtem is
preferred wherein the target particles are provided with
~as-tight barriers which retain therewithin the tritium
~55~5
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which is bred, i~ it is desired to continuously recover
trit~um from a producing HTGR, a less efective coating
system may be employed so as to permit the controlled
release o~ ~ritium throughout the life of the reactor
core. In such an instance, a helium purification system
71 of a substantially greater capaci~y is employed to
continuously treat a side stream of thle circulating
pr~mary gas coolant stream to remove tritium therefrom,
employing one of the tritium recovery schemes just
described~ Moreover, with the employment of such a
tritium recovery system as a part of the overall reactor
design, and using target particles having either
controlled release characteristics or excellent
retention characteristics based upon SiC barrier layers,
lt migh~ be pos~ible to recover the trltium wh~le the
fuel elements remain in the reactor. In such an
lnstance, the temperature o the reactor might be raised
; to heat the target particles to about 1300 C. for a
~u~ficient period~ e.g., a week, to assure the dif~usion
of substantially all of the tritlum into the helium
a~tmosphere, whence it is removed by ~he recovery system.
The following example illustrates a presently
preferred embodiment of target particles ~or the
production and retention therewithin of tritium;
however, it should not be understood to in any way limit
the scope of the invention which is deflned solely by
claims at the end of this specification.
EXA~PLE
LlA102 powder is cold-pressed into small
3a cylindrical pellets using a steel die and about 3000
psi. After ~intering in a vacuum for about 1200 C. for
a~out an hour, the 3intered pellets are dropped through
a zone heated to about 2200 C. in an inert atmosphere
to cause them to spheroldize. The resultant particles
are found to have a dens~ty equal to about 99% o~
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theoretical denslty~ ~n impervious layer of orlented
pyrocarbon of about 10 microns thick and having a
density of 1.9 yrams/cm3 is applied in a rotating drum
coater using a mixture o~ acetylene ancl argon at a
S tempera~ur~ of about 1000 C. The particles are then
transferred to a ~luidized bed coater and, at a
temperature of about 1100 C., are coated using a
mixture which is about 90 volume percent acetylene and
10 volume percent helium. Spongy pyrocarbon having a
L0 dens~ty oE about 1.1 grams~cm is deposited, and a
layer about 80 microns thick is applied to the kernels
whioh averag~ about 500 microns in diameter.
Following depositivn o the porous coating, the
tempera~ure is raised to about 1350 C., an~ a mixture
O~ propylene, acetylene and argon is employed to deposit
about 35 microns o~ isotropic pyrocarbon having a
~ensity o~ about 1.9 grams~cm3 and a BAF of about 1.02.
The temperature of the coater is then raised to
about 150Q C., and hydrogen is employed as tile
~luidizing gas. Approximately 10% of the hydrog n
stream i5 bubbled ~hrough a bath of
m~thyltrichlorosilane. Under these conditions, silicon
carbide having a density of about 3.20 grams/cm3,
wh~ch is beta-phase SiC, is deposited to create a layer
about 90 microns thick.
Thereafter, argon is again used as the
~luidizing gas, and the temperature is lowered to about
1370 C. A mixture of acetylene, propylene and argon is
then employed to deposit about 45 microns of isotropic
pyrolytic carbon having a density o~ about 1.85
grams/cm onto the silicon carbide layers.
Therea~ter, the particle~ are slowly cooled in a stream
o~ inert ga3 until they approach room ~emperature and
are removed from the coater.
35 . The particle~ are mixed with p~ tch to form a
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paste which is lnjected into molds to form short sticks
or rods in the same ~anner as heretofore employed to
produce nuclear fuel sticks~ Baking these green ~ticks
at a temperature o~ about 1600 C. for about one hour
drives o~f the volatiles from the pitch and leaves short
cylindrical rods having a diameter of about 1.57 cm.
wherein the coated target particles are securely bonded
to one another by the car~onized bonding agent.
These bonded target particle rods are then
irradiated in a suitable capsule subjecting them to a
thermal neutron bombardment at about 1000 C.
Irradiation is continued until a sufficient dosage of
neutrons has been encountered so that more than 95~ of
the ~i6 isotopes should have been transmuted to
tritium and helium. Monitoriny of the aapsule
atmosphere shows that barely measurable amounts of
trit~um are present during irradiation.
After removal of the capsule from the reactor,
the targe~ rods are removed from the capsule and
disposed in an autoclave which is supplied with a
controIled recirculating helium gas atmosphere. The
autoclave is heated to about 1500 C. and held at this
temperature ~or about 10 hours. The circulating helium
atmosphere is passed over zirconium sponge material, and
the tritium which is released from the target particles
in the autoclave is absorbed on the metal zirconium as
zirconium hydride. Following completion of the
absorption, examination of the zirconium sponge shows
tritium has been recovered in an amount equivalent to
about 90% of the Li6 isotopes present in the target
material. Accordingly, such target particles are
capable o~ producing and retaining tritium when exposed
to thermal neutronsr which tritium can be released
ther~from by heating to about 1500 C. These target
particles are considered to be well-suited for use in an
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. HTGR designed for the co-production of tritium and
: electrical energy.
Although the ~nvention has been described with
regard to certain preferred embodiments, which
. 5 constitute the best mode presently known to the
; applicants, it should be understood that various changes
and modifications as would be obvious to one having the
ordinary skill in this art may be made without departing
~rom the scope o~ the invention which is defined in the
la claims appended hereto. Various features of the
lnvenLlon are emphasized ln che olaims whloh follow.
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