GB2065955A - Production of Tritium in a Nuclear Reactor - Google Patents

Abstract

Tritium and electricity are co- generated in a high temperature gas- cooled nuclear reactor. Nuclear fuel material and minute target particles are disposed in chambers formed in the fuel elements of the reactor. The target particles comprise cores of a lithium compound, e.g. LiAlO2, an inner coating of porous material and an outer gas-tight coating of relatively dense material. Irradiation of the target particles by thermal neutrons results in transmutation of Li<6> nuclides to create helium and tritium that is retained within said outer gas-tight coating. Tritium is recovered from the irradiated target particles.

Description

SPECIFICATION Production of Tritium in a Nuclear Reactor This invention relates to a method of making tritium in a 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 an alpha particle (i.e., helium nucleus), which is referred to as an (n a) reaction. Lithium naturally contains about 7.49% of the isotope Lie and the remainder of the isotope Li7. Commercial production has been carried out using lithium in its natural or enriched form within a sealed, double-wall container that is hydrogenimpermeable which is located in the core of a plutonium production reactor. Lithium has been used in the form of an alloy with magnesium or aluminium and has also been used in oxide form.

These large water-cooled reactors, which have been either graphite-moderated or heavy water moderated, produce tritium essentially as a byproduct of plutonium production, and as a result of the ever-increasing demand for tritium (which will be magnified substantially when fusion technology becomes economically feasible), additional economic methods for the production of tritium are desirable.

It has been found that high temperature gascooled graphite-moderated reactors (HTGRs) offer a unique opportunity for the dual production of tritium and usable electric power, using to-day's technology. These reactors inherently have a relatively high conversion ratio which makes excess neutrons available for breeding fertile material from fissile material and have heretofore been employed to breed uranium-233 from thorium. By using minute coated particles containing lithium, 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 production of tritium can be carried out without disrupting the powergenerating function of such a nuclear reactor and without the creation of significant safety hazards.

Substantially all of the tritium created in the lithium target particles is retained therewithin; however, should minor amounts of tritium escape outside the coatings, it is recoverable from the primary gas-coolant stream. Following the fuel lifetime of the fuel elements, the tritium is recovered. Usually the fuel elements are first removed from the reactor, and the tritium is then recovered by heating in a facility designed for that purpose. The recovery of 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.

It is an object of the invention to provide an improved method for the production of tritium in a nuclear reactor. In accordance with the invention, a method of making tritium from a lithium compound in a nuclear reactor is characterized by forming minute coated particles including cores which contain a lithium compound, an inner coating of porous material and an outer gas-tight coating of relatively dense material, irradiating the coated particles with thermal neutrons in a gascooled nuclear reactor to cause the transmutation of a major proportion of the Lie nuclides to form helium and tritium, and recovering the tritium from the particles.

The invention will be explained in more detail by way of example with reference to the accompanying drawings, wherein: Figure 1 is a perspective view of a fuel element designed for use in the prismatic core of a gascooled nuclear reactor which may be employed for the production of tritium by the method of the invention: Figure 2 is a diagrammatic view showing a nuclear reactor including a reactor core formed of fuel elements of the type shown in Figure 1; Figure 3 is a fragmentary sectional view through one of 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 Figure 5 is a view, enlarged in size, of a target particle of the type which may be employed in the fuel element of Figure 1.

Gas-cooled high temperature graphitemoderated nuclear reactors have been developed which utilize heat produced by nuclear fission to produce steam in steam generators which is then used to drive electricity-producing turbines and also to provide heat for other applications. Such reactors have utilized cores made up of prismatic block-type fuel elements which contain coolant holes that extend axially therethrough and that are spaced between parallel, elongated fuel chambers, and have also utilized cores made up of a bed of spherical graphite 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 have been provided in the form of minute particles having coatings which function as fission-productretentive 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 material. Examples of such nuclear fuel particles are described in detail in U.S. Patent Specification No. 3,649,452, the disclosure of which is incorporated herein by reference. Such nuclear fuel particle designs have been found to be particularly wellsuited for gascooled nuclear reactors.

Illustrated in Figure 2 is a gas-cooled nuclear reactor of the prismatic core type which utilizes fuel elements 11 (Figure 1) in the form of graphite blocks 13 having hexagonal top and bottom surfaces 1 5 interconnected by perpendicular side faces 19. The nuclear reactor core 21 is formed from a plurality of vertical columns of such fuel elements 11 stacked one atop another, arranged within a pressure vessel 23. The fuel element block 13 is formed with a plurality of coolant holes 25 (Figures 3 and 4) located on a constant triangular 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 from the nuclear fission reactions.Fuel chambers 27 of slightly lesser diameter than the coolant holes 25 are located in a triangular array of lesser pitch so that each coolant hole is surrounded by a number of fuel chambers.

To facilitate alignment of the individual fuel elements 11 is stacked columns, the 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 fuel elements 11 is located within a pressure vessel 23 of a pre- or post-stressed concrete or the like, which may have a steel inner liner 33.

A surrounding reflector 35 is provided by unfuelled prismatic blocks of graphite, as is well known in the nuclear reactor art. The primary helium coolant is circulated downward through the core 21 by blowers 37 which are preferably disposed in suitable cavities located in the pressure vessel 23. To confine the primary coolant stream to the pressure vessel 23, steam generators 39 are also provided in cavities within the pressure vessel.

The primary helium coolant is accordingly circulated by the blowers 37 through the reactor core 21 where 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 HTGR design are needed when target particles containing Lie are included along with nuclear fuel in the individual fuel elements 11. The target particles are neutronically compatible with the nuclear fission reaction and can be employed in the place of burnable poison (e.g., boron) that would normally be included in such fuel elements and also in place of most of the fertile thorium. In fuel elements 11 of prismatic design, the fuel chambers 27 generally define the volume available for disposition of nuclear fuel material.

When an HTGR is 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 of the available volume holds nuclear fuel material.

Lis has an extremely large cross section, equal to about 953 barns, for the absorption of neutrons in the thermal energy range and the consequent transmutation to produce tritium and helium. As a result, lithium is inherently self-shielding, and in order to induce efficient conversion of Li6 to tritium, it is important to disperse the lithium throughout the reactor core. Excellent dispersal is achieved by forming small cores or kernels of a Li compound, having a size on the order of about 300 to 1000,um, and spacing these cores from one another by means of exterior coatings which totally surround the cores. By locating some of the target material in each of the fuel chambers 27, overall dispersion is still further enhanced.

In the fuel element design depicted in Figure 3, short rods 41 of target particles joined together by suitable bonding material such as carbonized pitch, are located adjacent to the top and bottom ends ot each of the fuel chambers 27, with similar rods 43 of nuclear fuel material filling up the major portion of the fuel chambers 27 in the regions between the top and bottom target material rods. Graphite plugs 45 close the upper end, and heat-decomposable plastic spacers 47 are provided beneath the plugs. This arrangement facilitates the selective recovery of tritium by physically slicing the upper and lower ends of the fuel elemente 11 from the remainder and then reprocessing these sliced portions 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 formed into fairly flat wafers 41' which are then disposed alternately between adjacent nuclear fuel rods 43 throughout the entire length of each fuel chamber 27. Such an arrangement facilitates even better dispersal throughout the nuclear reactor core; however, it does not facilitate selective tritium recovery by processing only portions of the individual fuel elements.

Examples of representative target particles 51 are depicted in Figure 5 and include minute cores 53, preferably spheroidal in shape, which are between about 300 and about 1000,um in diameter. The cores 53 are surrounded by an inner coating 57 of a generally porous nature which accommodates the build-up of helium and tritium from the transmutation of the lithium and which is in turn surrounded by an outer coating 59 which forms the gas-tight barrier that prevents escape of the helium and tritium.

The cores 53 are formed from a solid compound of lithium which is preferably stable at the temperatures employed for the vapourdeposition of the surrounding coatings. Lithium in oxide form, ether by itself or in a combination with another refractory-like element, may be employed as the core material. Examples are lithium oxide, lithium aluminate (LiAlO2), and lithium silicates (Li2SiO3)(Li4SiO4). The lithium compound should havea 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 vapour-deposition coater.For example, small kernels can be formed by a powder agglomeration process or by coldpressing in steel dies and then sintered to provide strength and higher density. For example, lithium aluminate powder can be cold-pressed in a die at about 20,000 kPa and then sintered in a vacuum at about 1 2000C for eight hours. Kernels made by powder agglomeration can also be sintered to provide mechanical strength. If high density is desired, the sintered kernels can be made spheroidal by being dropped through a hot zone at between 1 8000C and 22000C to cause them to melt and densify into spheroidal shapes in accordance with known technology.

Generally, the cores 53 are densified to at least about 70% of theoretical density. By theoretical density is meant the maximum density for that particular stoichiometric compound. The preferred lithium compound is lithium aluminate which has a theoretical density of about 2.55 grams per cm3.

Although densification to a density approaching theoretical density is possible, it may be preferred to employ cores 53 in the range of about 70% to 80% of theoretical density from the standpoint both of spacial dispersion and ultimate accommodation of the gaseous products of the lithium transmutation.

To prevent the reaction of lithium aluminate with subsequent vaporous materials to which there will be exposure during the coating operations, an impervious carbon seal layer 55 is applied at a temperature below 1 1000C and preferably at a temperature not higher than about 10000C. Such a seal coating may be deposited in a particulate bed fluidized by gas flow or in a rotating drum or other type of agitated bed coater.

Pyrocarbon seal layers should have a density of about 1.8 to 2.0 grams/cm3 and are preferably oriented. A thickness of about 10 to 30,,lm of such pyrocarbon provides an adequate seal coating and can be deposited from a mixture of acetylene plus an inert gas such as argon.

The porous layer 57 that is provided for the accommodation of the helium and tritium within the miute pressure vessels is preferably pyrocarbon having a density between about 0.9 and 1.2 gram/cm3. The thickness of the porous pyrocarbon layer 57 is dependent upon the amount of Lie included within the cores 53 and the pressure which the outer gas-tight coating is designed to withstand.

If there are no constraints on the amount of space occupied by the target particles in the nuclear reactor core, larger amounts of porous material can be included so as to prevent the build-up of high gas pressures within the gastight outer barrier. On the other hand, if particular constraints limit the amount of space, a lesser thickness of the porous pyrocarbon may be employed along with a slightly thicker outer coating, which will withstand the higher gas pressure build-up: In general, for cores in the 300 to 500 um range, it is likely that at least about 75 um of prous pyrocarbon would be used.

In the outer coating which provides the gastight barrier to prevent the escape of tritium, the outer coating 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 silicon carbide and zirconium carbide provide an effective barrier to the passage of tritium. As in any such barrier material, the thicker the material, the more effective the barrier, and it is expected that at least about 50 ,'m of SiC or ZrC would be used. For examlple, a silicon carbide layer having a thickness of 90 ym or even greater might be employed. The carbide barrier layer should have a density of at least 95% of theoretical density.Deposition of silicon carbide from a vaporous atmosphere can be fairly 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 outer coating 59 are layers 61, 63 of isotropic polycarbon having a density between about 1.7 and 2.0 grams/cm3 and having individual thicknesses of between about 35 and 45 ym.

Such isotropic coatings are deposited from a mixture of acetylene, propylene and inert gas at a temperature of about 1 3500C which have a BAF (Bacon Anisotrophy Factor) of less than about 1.05. The interior pyrocarbon layer 61 retards the outward diffusion of materials from the core to the silicon carbide during irradiation in the reactor core and, during the process when the silicon carbide is being deposited, serves as a further barrier to prevent chlorine (which is present in the coating atmosphere) from reaching the core where undesirable chemical reactions may occur.

The exterior pyrocarbon layer 63 has a larger strain to fracture than the relatively brittle silicon and thus provides mechanical handling strength for the target particles following completion of the coating operation, particuiarly during bonding of the particles with pitch or the like to form short rods before loading into the fuel chambers. During operation in the reactor core, the isotropic pyrolytic carbon layer 63 undergoes a controlled shrinkage as a result of exposure to high temperature and fast neutrons, and it shrinks onto the silicon carbide layer 59 placing it in compression and increasing its strength as a minute pressure vessel.

Although the employment of an outer coating which includes a layer of silicon carbide or zirconium carbide sandwiched between isotropic pyrocarbon layers is preferred, other suitable gastight 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 grams/cm3 may be employed.

Such a layer should have a thickness of at least about 70,us, and depending upon the size of the lithium cores and the amount of porous pyrocarbon employed, a layer up to about 200 m thick might be used. Such oriented pyrocarbon layer can be deposited from an atmosphere of acetylene, propylene and argon at a temperature of about 1 3500C in a fluidized bed coater using a lower coating rate than is used to deposit isotropic pyrocarbon.

It is contemplated that the core of a 1000 MWt HTGR could be designed with approximately 1000 kg of U-235 and 140 kg of lithium (measured as lithium metal). Operation of such a HTGR for about two years would result in fissioning of about 80% to 90% of the fissile isotopes of uranium, in the transmutation of over 90% of the Liss nuclides and in the recovery of about 5 to 6 kg of tritium. Although such a batchloaded cycle wherein all the fuel elements 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 of the fuel elements are reloaded. The batch scheme has the advantage of having a flatter power across the core and therefore lower fuel temperatures.

The HTGR has the unique advantage to allow various such schemes through balancing the reactivity effect of the uranium depletion with the shielding 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 competitive on the basis of electricity production alone. Of course the fuel cost would be somewhat higher because it would not be offset by breeding fairly large amounts of U-233 from the fertile thorium; however, this additional fuel cost would be more than offset by the economic value of the tritium produced. Accordingly, the dual operation of the HTGR both as an electricity producer and as a tritium producer appears to be extremely attractive economically.

As earlier indicated, the lithium target particles 51 could be restricted to zones adjacent to the top and bottom of prismatic fuel elements, which could then be severed and reprocessed separately for the recovery of tritium, or the target material could be distributed throughout the fuel chambers. In either instance, heating of such target particles to a temperature of about 1300- 1 4000C or above effects the relatively prompt diffusion of tritium through a silicon carbide barrier layer, which was very effective in restraining passage of tritium at lower temperatures. For example, autoclaving of the fuel elements, or the remains thereof, at temperatures of 1 5000C can be employed to release the tritium into a 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 8000C to destroy the graphite blocks and the isotropic pyrolytic carbon exterior layers. Thereafter, crushing the carbide barrier layers in a controlled environment, coupled with heating to a temperature of at least about 5000C, would quickly effect release of the tritium being held therewithin, and this procedure is preferred for ZrC coatings which are more resistant to hydrogen diffusion at higher temperatures.

Ultimate recovery of tritium (T) from a gaseous atmosphere is preferably effected by conversion of the tritium to T20 by oxidation using a suitable oxygen source, such as copper oxide. T20 has physical characteristics quite similar to ordinary water and is then removed from the gas stream by a molecular sieve or by freezing in a suitable cold trap, such as liquid nitrogen. Alternatively, tritium can be recovered as a hydride, instead of being oxidized, by exposure to zirconium or titanium sponge metal.

Although the foregoing containment system is preferred wherein the target particles are provided with gas-tight barriers which retain therewithin the tritium which is bred, if it is desired to continuously recover tritium from a producing HTGR, a less effective coating system may be employed so as to permit the controlled release of tritium throughout the life of the reactor core. In such an instance, a helium purification system 71 of a substantially greater capacity is employed to continuously treat a side stream of the circulating primary 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 hall either controlled release characteristics or excellent retention characteristics based upon SiC barrier layers, it might be possible to recover the tritium while the fuel elements remain in the reactor. In such an instance, the temperature of the reactor might be raised to heat the target particles to about 1 3000C for a sufficient period, e.g., a week, to assure the diffusion of substantially all of the tritium into the helium atmosphere, whence it is removed by the recovery system.

The following Example illustrates a presently preferred embodiment of target particles for the production and retention therewith in of tritium; however, it should not be understood to in any way limited the scope of the invention.

Example Lilo2 powder is cold-pressed into small cylindrical pellets using a steel die and about 20,000 kPA. After sintering in a vacuum for about 1 2000C for about an hour, the sintered pellets are dropped through a zone heated to about 22000C in an inert atmosphere to cause them to spheroidize. The resultant particles are found to have a density equal to about 99% of theoretical density. An impervious layer of oriented pyrocarbon of about 10 pm thick and having a density of 1.9 gram/cm3 is applied in a rotating drum coater using a mixture of acetylene and argon at a temperature of about 0000C. The particles are then transferred to a fluidized bed coater and, at a temperature of about 1 1000C, are coated using a mixture which is about 90% by vol. acetylene and 10% by vol. helium.Spongy pyrocarbon having a density of about 1.1 gram/cm3 is deposited, and a layer about 80 zim thick is applied to the cores which average about 500u.m in diameter.

Following deposition of the porous coating, the temperature is raised to about 1 3500C, and a mixture of propylene, acetylene and argon is employed to deposit about 35,um of isotropic pyrocarbon having a density of about 1.9 gram/cm3 and a BAF of about 1.02.

The temperature of the coater is then raised to about 15000C, and hydrogren is employed as the fluidizing gas. Approximately 10% of the hydrogen stream is bubbled through a bath of methyltrichlorosilane. Under these conditions, silicon carbide having a density of about 3.20 grams/cm3, which is beta-phase SiC, is deposited to create a layer about 90 ym thick.

Thereafter, argon is again used as the fluidizing gas, and the temperature is lowered to about 1 3700C. A mixture of acetylene, propylene and argon is then employed to deposit about 45,um of isotropic pyrolytic carbon having a density of about 1.85 gram/cm3 onto the silicon carbide layers. Thereafter, the particles are slowly cooled in a stream of inert gas until they approach room temperature and are removed from the coater.

The particles are mixed with pitch to form a paste which is injected into moulds to form short rods in the same manner as heretofore employed to produce nuclear fuel rods. Baking these green rods at a temperature of about 16000C for about one hour drives off 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 carbonized bonding agent.

These bonded target particle rods are then irradiated in a suitable capsule subjecting them to a thermal neutron bombardment at about 1 0000 C. Irradiation is continued until a sufficient dosage of neutrons has been encountered so that more than 95% of the Lie isotopes should have been transmuted to tritium and helium.

Monitoring of the capsule atmosphere shows that barely measurable amounts of tritium are present during irradiation.

After removal of the capsule from the reactor, the target rods are removed from the capsule and disposed in an autoclave which is supplied with a controlled recirculating helium gas atmosphere.

The autoclave is heated to about 1 5000C and held at this temperature for 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 Lie isotopes present in the target material. Accordingly, such target particles are capable of producing and retaining tritium when exposed to thermal neutrons, which tritium can be released therefrom by heating to about 1 5000C. These target particles are considered to be well-suited for use in an HTGR designed for the co-production of tritium and electrical energy.

Claims (25)

Claims

1. A method of making tritium from a lithium compound in a nuclear reactor, characterized by forming minute coated particles including cores which contain a lithium compound, an inner coating of porous material and an outer gas-tight coating of relatively dense material, irradiating the coated particles with thermal neutrons in a gas-cooled nuclear reactor to cause the transmutation of a major proportion of the Lie nuclides to form helium and tritium, and recovering the tritium from the particles.

2. A method in accordance with claim 1, characterized in that the lithium compound is an oxide of lithium.

3. A method in accordance with claim 1, characterized in that the compound is lithium aluminate.

4. A method in accordance with any one of claims 1 to 3, characterized in that the cores are initially coated with a layer of pyrocarbon having a density of at least 1.8 g/cm3 at a temperature not higher than 10000C.

5. A method in accordance with any one of claims 1 to 4, characterized in that the inner coating includes pyrocarbon having a density of not greater than about 1.2 g/cm3 and a thickness of at least 75,cm.

6. A method in accordance with any one of claims 1 to 5, characterized in that the cores are spheroids between 300,um and 1000 Mm in diameter.

7. A method in accordance with any one of claims 1 to 6, characterized in that the outer coating includes layer of dense SiC or ZrC.

8. A method in accordance with claim 7, characterized in that the outer coating includes SiC and said recovering of tritium is effected by heating the particles to a temperature of at least about 1 3000C to cause diffusion through the SiC coating.

9. A method in accordance with claim 5, characterized in that the pressure of tritium within said outer coating is between 500 and 2000 kPa at the time when said irradiation is completed.

10. A method in accordance with claim 7, characterized in that the irradiated particles are removed from the reactor and said carbide layers are broken to recover the tritium.

11. A fuel element for use in a gas-cooled nuclear reactor, characterized in that the element comprises a body of refractory material having relatively good thermal conductivity and neutron moderating characteristics, and nuclear fuel material and minute coated target particles contained in one or more chambers formed in said body, said target particles including cores which contain a lithium compound, an inner coating of porous material and an outer gas-tight coating of relatively dense material, whereby irradiation of said target particles with thermal neutrons in a gas-cooled nuclear reactor causes the transmutation of a major proportion of the Lie nuclides to form helium and tritium.

12. The element in accordance with claim 11, characterized in that the lithium compound is an oxide of lithium.

13. The element in accordance with claim 11, characterized in that the lithium compound is lithium aluminate.

14. The element in accordance with any one of claims 11 to 13, characterized in that the cores are spheroids between 300 and 1000 ym in diameter.

1 5. The element in accordance with any one of claims 11 to 14, characterized in that the cores are surrounded by an adjacent seal layer of pyrocarbon at least 10 fly thick having a density of at least 1.8 g/cm3.

1 6. The element in accordance with claim 15, characterized in that the seal layer is surrounded buy a porouslayer of pyrocarbon having a density of not greater than 1.2 g/cm3 and a thickness of at least 75 ym.

1 7. The element in accordance with any one of claims 11 to 16, characterized in that the outer coating contains a layer of dense SiC or ZrC at least 50 ,um thick.

18. The element in accordance with claim 17, characterized in that the SiC or ZrC layer is disposed immediately between layers of isotropic pyrocarbon.

1 9. The element in accordance with any one of claims 11 to 18, characterized in that the outer coating contains a layer of pyrocarbon having a BAF (Bacon Anisotrophy Factor) of 1.1 or above, a density of at least 1.85 g/cm3 and a thickness of at least 70 ym.

20. The element in accordance with any one of claims 11 to 19, characterized in that said body is a block which has a pair of parallel flat top and bottom end faces and a plurality of sides which are substantially perpendicular to said end faces, and said block also contains a plurality of coolant holes which extend axially completely therethrough from end face to end face and a plurality of said chambers which are elongated and extend parallel to said coolant holes.

21. The element in accordance with claim 20, characterized in that the target particles are disposed in zones at the top and bottom of said chambers and nuclear fuel material is located therebetween.

22. The element in accordance with claim 20, characterized in that said refractory material is graphite, said block has a cross section shape of a regular hexagon, and said fuel chambers are disposed in a triangular array.

23. A nuclear reactor core, characterized by comprising a plurality of vertical columns of fuel elements in accordance with claim 20 stacked one atop another.

24. An nuclear reactor, characterized by including a reactor core in accordance with claim 23 and means for circulating a helium coolant therethrough.

25. A nuclear reactor in accordance with claim 24, characterized by the provision of means for continuously removing tritium from said circulating helium stream.