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Implantation driven permeation (IDP) behaviour was investigated for deuterium implanted at low energy (100–1800 eV) into pure aluminium to simulate the behaviour of tritium. The experimental results showed that steady state IDP fluxes φP of deuterium depended on the incident deuterium ion (D+) fluxes φi according to the relation φp = α(φ)12, under the following conditions: φi ≈ 7 × 1013 ∼ 1 × 1015 D+ cm−2 s−1, sample temperature approximately 550 ∼ 825 K, and incident D+ ion energy approximately 300 ∼ 1500 eV. This fact suggests that the IDP process through pure aluminium was controlled by recombination of deuterium at the incident surface and by deuterium diffusion at the permeation side. The activation energy for implantation driven permeation of deuterium through pure aluminium was determined to be 59.9 ± 5.0 kJ mol−1 under the above experimental conditions.

Hydrogen isotope exchange reaction rate in tritium and methane mixed gas, as induced by tritium decay and beta radiation, has been experimentally measured. Initially T2 gas was filled to 40 kPa and 20 kPa of CH4 gas was added. The mixed gas spectrum was analyzed periodically by laser Raman spectrometry. The first order HT and H2 formation rates and T2 and CH4 decay rates by hydrogen isotope exchange reaction were observed between 2.9·10−3 h−1 and 4.8·10−3 h−1. Although the estimated hydrogen isotope exchange reaction rate was 1/20–1/10 slower than the rate of H2+T2 mixed gases, it was nearly equivalent to the ion formation rate by tritium beta radiation. This suggested that isotopic hydrogen radicals formed via ionization would disappear in the presence of methane.

This paper reports that to study application o laser Raman spectroscopy for fusion fuel gas analysis by an in situ method, methane (CH{sub 4}) and tritium (T{sub 2}) mixed gases were measured. In the mixed gases, hydrogen isotope exchange reactions were induced by beta decay, and various isotopic hydrogens and methanes were produced. Spectral peaks of {nu}{sub 1} and {nu}{sub 3} bands were detected individually for CH{sub 4} and four tritiated methanes. The {nu}{sub 1} bands between 1700-1900 cm{sup {minus}1} were selected as suitable ones for quantitative analysis. After mixing T{sub 2} and CH{sub 4} gases, while large amounts of tritiated methanes were produced as time lapsed, the equilibrium state was not reached by the time 1000 h had passed. It was presumed that the isotope exchange reactions were very slow compared to mixed gases of just hydrogen isotopes.

The R and D for tritium technology and studies of tritium science in Japan has been carried out in the national institutions (JAERI, PNC and others), in the universities, and in industry. In fusion fuel technology, the technological R and D of tritium processing and handling has been carried out by TPL-JAERI, including collaboration with TSTA (United States) and basic and innovative research conducted by the universities. PNC has conducted R and D and engineering studies in the heavy water quality control system for FUGEN reactor, a heavy water cooled reactor ATR, and in the tritium recovery system for the heavy water coolant. Since the last meeting in Toronto, there are many developments in the above researches which will be summarized in this paper with some results.

A polyimide membrane has a selective permeability for hydrogen gas and water vapor. A small scale experimental apparatus for reducing the tritium concentration of air was designed and fabricated at...

Laser Raman spectroscopy has been applied to the on-line analysis of the operation of the cryogenic Isotope Separation System (ISS) at the Tritium Systems Test Assembly (TSTA). A flow-through cell was employed to permit near real-time observation of the dynamic response of the 3-column ISS. Accurate analysis of hydrogen isotopic mixtures may be made in less than 2 minutes. Full response to a change in the sampling point is achieved in approximately one minute. In this paper, response measurements are shown as well as static column profiles and dynamic response to induced parameter changes. Cross check of analysis was performed with radio-gas chromatography.

The recent results of the Isotope Separation System (ISS) operations at the Tritium Systems Test Assembly (TSTA) with 100 g of tritium indicate that the system generally satisfies design goals, while system stability problems remain to be solved. In this paper, the authors configure the ISS system for the three column mode. which is one of the promising cascade configurations in a fusion fuel cycle, to eliminate such instability and operated it for six days. Fluctuations in flows and liquid levels were improved. Column separation characteristics obtained were satisfactory and agreed with the numerical analysis. The amount of discharged tritium was an acceptable effluent level. This means that the existing ISS system can be used as a three column system and possibly be applied to numerous fuel concepts. Presently, a new laser Raman spectroscopic gas analyzer has been installed at the ISS. This on-line system enables studies of the ISS dynamic behavior for further stability and performance data.

This paper reports the preliminary results of studies of the kinetics of the reactions of T2 with H2, D2, and CH4. The gas phase reactions were studied using laser Raman spectroscopy for the real-t...

In order to discuss the long-term reliability and safety of zirconium-cobalt alloy for tritium use, the release behavior of decay helium from ZrCo tritide has been investigated for one and a half years with a radio-gaschromatograph. In this paper, the results show that the release fractions of the total decay {sup 3}He in ZrCo tritide are less than 3% and has been almost constant for 18 months under the following conditions: the operating temperatures = 293 {approximately} 523 K, the atom ratios (T/ZrCo) = 0.3 {approximately} 1.4, and the number of hydrogenation-dehydrogenation cycles before tritiation = 1 {approximately} 10. Moreover, residual decay {sup 3}He was not released even if ZrCo was heated to 873 K, though most of the tritium was released. It became clear that the decay {sup 3}He was quite immovable in ZrCo tritide under these experimental conditions.

A brief survey on recent advances in the field of tritium technology concerning the most important systems of the fuel cycle of a fusion reactor, i.e. the plasma exhaust pumping system, the exhaust gas clean up system, the isotope separation system, the tritium storage facility and the tritium extraction from a blanket is provided. Experimental results, single component developments, and technical tests including those with relevant amounts of tritium that constitute the basis of proposed integral process concepts are described.

The Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI) has developed a Fuel Cleanup System (FCU) which accepts simulated fusion reactor exhaust and produces pure hydrogen isotopes and tritium-free waste. Under the collaboration program of USDOE and JAERI, an integrated process, the “JAERI Fuel Cleanup System” (JFCU) was designed and fabricated for testing at the Tritium Systems Test Assembly (TSTA) as a major subsystem of the simulated fusion fuel cycle. The JFCU has the capability to continuously process a mixture of hydrogen isotopes and impurities of 15 mol/h. The major components are: a palladium diffuser for hydrogen isotopes purification, a catalytic reactor for oxidation of tritium in impurities, cold traps for separation of tritiated water to produce exhaust containing a very low level of tritium, a ceramic electrolysis cell for decomposition of tritiated water, and zirconium-cobalt beds for hydrogen isotopes storage. In early 1990, the JFCU was installed in the TSTA and preliminary testing without tritium was initiated. Components such as the palladium diffuser, cold traps, catalytic reactor, scroll pump and electrolysis cell were tested independently as well as in the integrated process tests.

The Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI) has developed a Fuel Cleanup System (FCU) which accepts simulated fusion reactor exhaust and produces pure hydrogen isotopes and tritium-free waste. The major components are; a palladium diffuser, a catalytic reactor, cold traps, a ceramic electrolysis cell, and zirconium-cobalt beds. In 1988, an integrated FCU process loop was installed in the TPL. A number of “hot” runs were performed to study the system characteristics and improve performance. Under the US-Japan collaboration program, the “JAERI Fuel Cleanup System” (JFCU) was designed and fabricated by JAERI/TPL for testing at the Tritium Systems Test Assembly (TSTA) in Los Alamos National Laboratory as a major subsystem of the simulated fusion fuel cycle. The JFCU was installed in the TSTA in early 1990.

Under a collaborative agreement between US and Japan, two tritium processing components, a palladium diffuser and a ceramic electrolysis cell have been tested with tritium for application to a Fuel Cleanup System (FCU) for plasma exhaust processing at the Los Alamos National Laboratory. The fundamental characteristics, compatibility with tritium, impurities effects with tritium, and long-term behavior of the components, were studied over a three year period. Based on these studies, an integrated process loop, JAERI Fuel Cleanup System'' equipped with above components was installed at the TSTA for full scale demonstration of the plasma exhaust reprocessing.

An overview is given of the design, construction and initiation of the operation of the facility, Tritium Process Laboratory, JAERI, to study and develop tritium processes for fusion fuel cycle and safe handling technology.

Compound cryopumps have been added to the Tritium Systems Test Assembly (TSTA) integrated fusion fuel loop. Operations have been performed which closely simulate an actual fusion reactor pumping scenario. In addition, performance data have been taken that support the concept of using coconut charcoal as a sorbent at 4 K for pumping helium. Later tests, reported in detail in Batzer et al, Proc 13th Symp Fusion Engng , Knoxville, TN, 2–6 Oct (1989), show that coconut charcoal may be used to co-pump D, T and He mixtures on a single 4 K panel. Rotary spiral pumps have been used successfully in several applications at TSTA and have required more than 9000 h of maintenance-free operation. Metal bellows pumps have been used to back the spiral pumps and have been relatively trouble-free in loop operations. Bellows pumps also have more than 9000 h of maintenance-free operation.

Future experimental fusion devices and power reactors will continuously pump large amounts of helium, and deuterium and tritium (D-T). D-T to helium ratios greater than 10:1 are projected. Cryopumps are the logical choice to pump gases in the 10/sup 3/-m/sup 3//s range. A test program undertaken at Los Alamos that evaluated the capability of copumping the fusion exhaust gases on a liquid-helium cooled surface of activated carbon (coconut charcoal) is described. A compound cryopump installed in the VAC system of the Tritium Systems Test Assembly (TSTA) was used as the test article. In the test, copumping of helium and deuterium, and helium-3 and tritium was undertaken. Test results show that copumping is practical and that compound pumps may not be required for fusion applications. >

An all-metal apparatus has been constructed and installed in the main cell of the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. The apparatus is being used in an experiment that is part of a study to determine the feasibility of using metal tritides for storing tritium in fusion fuel processing. The structure will handle about 2600 Ci of tritium in the experiment. It allows measurement of equilibrium pressure on metal-hydrogen systems and was made to assure safe handling of tritium. The pressure-composition isotherms for the LaNi/sub 3/Mn/sub 2/-protium (H), deuterium (D), and tritium (T) systems were measured to study isotopic effects in the temperature range of 60 degrees C to 250 degrees C, and the pressure range below 120 kPa. >

The Tritium Systems Test Assembly (TSTA) project at Los Alamos National Laboratory is charged with developing and demonstrating the fusion fuel processing requirement for the magnetic fusion energy program. The key fusion fuel streams are the plasma exhaust and the blanket stream. At present, the technology under development at TSTA includes only the plasma exhaust. The blanket is a logical upgrade to TSTA and the upgrade is the subject of an ongoing study. The program, called the Breeding Blanket Interface (BBI), is being investigated with the intention of installing such a processing system at TSTA in the 1992-1993 time frame. The BBI program was initiated in July, 1987. The conditions of the tritium product streams of various blankets were determined. Pre-conceptual designs of the BBI systems were developed for two blanket concepts: aqueous lithium salt solution blanket and solid breeder blanket. Preliminary estimates of costs have been performed and the schedule of the TSTA/BBI program has been developed. The schedule and planning of the TSTA/BBI program are discussed. The goal of the program is to establish the engineering data basis of the fuel cycle for an International Thermonuclear Engineering Reactor. >

Research and development of a fusion reactor fuel-processing system are being performed with a large amount of tritium at the Tritium Process Laboratory (TPL). Major experiments involve the following: the fuel cleanup system (FCU) for processing simulated plasma exhaust; the isotope separation system (ISS) with cryogenic distillation; the tritium permeation study with triton ion beam; isotope separation with the thermal diffusion columns; and analysis and measurement of tritium. Experiments have been performed at TPL since 1988. In recent experiments, some results are obtained with and without a large amount of tritium. >

The Tritium Systems Test Assembly (TSTA) at Los Alamos, New Mexico, is a full-scale facility dedicated to testing tritium processing for fusion reactors. Adding a breeder blanket interface (BBI) to the TSTA is being studied. The BBI is to test the processing required for the tritium output streams for the various fusion-reactor breeder blankets. In the current phase of the study, the characteristics of the output from the various blanket types are being evaluated. Defining the output stream with respect to H/T ratio, impurity content, and radionuclide content is emphasized. Assessments of the solid breeder blanket (ceramic, Li/sub 2/O) and the aqueous salt solution blanket are reported. For the aqueous lithium salt solution blanket, the chemical systems are very complex, involving a number of phenomena, including radiolysis, corrosion, and electrolysis. A system of about 0.01 ITER scale may be added to the TSTA. For the solid breeder blanket, no critical issues are identified. A system of about half-ITER scale, possibly even full scale, may be added to the TSTA. >

The Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory has conducted its longest continuous integrated loop operation to data. This provided an opportunity to test some hitherto unproven capabilities of the TSTA fuel cleanup system (FCU). The purpose of the FCU is to remove impurities from a stream of hydrogen isotopes (Q/sub 2/) representative of torus exhaust gas. During this run, impurities loadings ranging from 60 to 179 sccm of 90% N/sub 2/ and 10% CH/sub 4/ were fed to the FCU. Each of the two FCU main flow molecular sieve beds (MSBs) were filled to breakthrough three times. The MSBs were regenerated during loop operations. The experiment yielded two principle findings. The first is that using the cryogenic molecular sieve is very robust method for removing impurities from fusion fuels. While loop flow rates, pressures, and impurities concentrations varied, the MSBs continued unaffected, completely removing impurities. The second is that the first step in the MSB regeneration process proved to be very effective. >

Tritium Systems Test Assembly (TSTA) in the Los Alamos National Laboratory is a reactor-scale fusion fuel processing loop and is operated by the Japan Atomic Energy Research Institute (JAERI) and the USDOE under the collaboration. An extended experimental campaign of the TSTA loop was performed with 100 grams of tritium in April-May 1992 in order to simulate a steady-state fusion fuel processing for 25 days. Continuous processing of simulated plasma exhaust (DT and H mixture, He, CH/sub 4/ and N/sub 2/) was successfully demonstrated. Impurity processing was successfully tested by two different techniques. US developed Fuel Cleanup System (FCU) purifies DT by cryogenic molecular sieve beds and decomposes tritiated water by Magnesium beds. Japanese developed J-FCU is based on purification by permeation through palladium alloy membrane and vapor electrolysis with a ceramic electrolyte cell. Isotope Separation System was stably operated with four interlinked columns and continuously produced pure tritium and deuterium stream while exhausting protium impurity. Real-time analysis in the ISS process was demonstrated with the laser Raman Spectroscopy. The result demonstrated the safe and stable operation of reactor-scale fusion fuel loop in a steady state.

This paper discusses the possibility that tritium might exchange with water trapped in aluminum anodize cryopanels in JET which prompted a test program at the Tritium systems Test Assembly, TSTA. Los Alamos, New Mexico. JET furnished two test pieces of cryopanel which were exposed to tritium at approximately liquid nitrogen temperature and 25 torr pressure for nearly two weeks. One specimen was removed and the retained tritium was measured. The second specimen was subjected to several increasing temperature vacuum bakeouts and the effectiveness of the bakeouts were inferred from the pressure history of the chamber. When the retained tritium in the second specimen was measured, it was found that nearly 96% of the tritium, as measured in the first specimen, had been removed during the vacuum bakeouts. If the tritium retained in the cryopanel without bakeout were scaled to JET conditions according to a linear pressure-time relationship, the tritium expected to become trapped in the JET cryopanels would be approximately 0.6 gram.

In 1988, a prototypical vacuum system was added to the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. Since then various pumping scenarios, which might be expected in a fus...

In 1987 the US Department of Energy (DOE) and th Japan Atomic Energy Research Institute (JAERI) signed collaborative aggrement for the joint operation and join funding of the Tritium Systems Test Assembly (TSTA) for a five year period ending June 1992. The tritium research and development accomplished during this five year period has provided a very valuable data base for the fusion energy program. During the past year a number of significant experiments were completed and major programmatic milestones were met. This paper summarizes some of the major accomplishments during the initial five year collaboration. The integrated TSTA gas processing loop was operated continuously for 25 days. Both the JAERI fuel cleanup system and the original TSTA fuel cleanup system were operated under similar conditions of flow, pressure, and impurity content of the DT gas. The isotope separation system was operated continuously, using the recently installed laser Raman spectroscopy system for in situ analysis of the gas in the loop. Additional studies include studies on the interaction of tritium with sulfur hexaflouride, studies on the compatibility of cryosorption pump materials with tritium, and studies of tritium interaction with gaseous materials such as methane and deuterated methane.

The Tritium Process Laboratory (TPL) has been constructed to carry out research and development on tritium processing and tritium safety handling technologies for fusion reactors. Since 1988, many tritium experiments including simulated plasma exhaust gas cleanup, hydrogen isotope separation, tritium accountability, tritium removal and containment, and so on have been carried out using gram level tritium at TPL. The experiments have produced remarkable accomplishments on tritium technology for fusion reactors.

The Tritium Process Laboratory (TPL) was constructed at the Japan Atomic Energy Research Institute (JAERI) in June, 1985. The objectives of this laboratory are the research and development of the fusion reactor fuel processing system and safe handling technology of large amounts of tritium. Total tritium inventory in the facility is 10 g and each experiment will be conducted with up to 1 g. Safety systems have been in operation since 1985 and have been tested without tritium. Tritium removal systems, gloveboxes, vacuum pumping system and tritium storage systems etc. have been proven to work as designed so that tritium can be handled safely in this facility. The various items of experimental apparatus were designed, fabricated, installed and checked early in 1988. First tritium was shipped from Oak Ridge National Laboratory in February, and then the “hot” operation was initiated in March, 1988.

A mathematical simulation procedure is developed for a single cryogenic distillation column with a feedback stream and a catalytic equilibrator used for protium-tritium (HT) separation. Large error...

At the Tritium Process Laboratory (TPL) in the Japan Atomic Energy Research Institute, the Thermal Diffusion process is applied to enrich and recovery tritium diluted by the operation and experiments of the other apparatus installed in the gloveboxes. The apparatus is required to process 200-liter of hydrogen isotope mixture containing up to lg tritium by batch operation and enrich tritium to 90%. Preliminary studies were performed numerically as well as experimentally, and good agreement was obtained. Based on the results, the apparatus was designed. Major components are four 2.5m-high diffusion columns, reservoirs, pumps, active metal beds and a mass spectrometer.

A ZrCo intermetallic compound is proposed as the material for recovery, storage and supply of gaseous tritium. It is a substitute for the uranium that has been extensively used for this purpose in spite of its restricted handling as a nuclear fuel material. Pressure-composition isotherms of a ZrCoH2 system were investigated and this material proved to be applicable and advantageous. Some ZrCo beds were designed, fabricated and tested for actual use. A bed for the handling of large amounts of hydrogen isotopes was developed. The test establishes ZrCo beds as practical components for tritium service.

The first loop operation tests of the Tritium Systems Test Assembly (TSTA) with 100 grams-level of tritium were performed at the Los Alamos National Laboratory (LANL) in June and July, 1987. The July run was resumption of the June run, which was halted because of a loss of cryogenic refrigerant in the hydrogen isotope separation system.

In March of 1988 full operation of the 4-column isotope separation system (ISS) was realized in runs that approximated the design load of tritium. Previous operations had been fraught with operating difficulties principally due to external systems. This report will examine the recent highly successful 6-day period of operation. During this time the system was cooled from room temperature, loaded with hydrogen isotopes including 109 grams of tritium, integrated with the transfer pumping, impurity injection, and impurity removal systems, as well as the remote computer control system. At the end of the operation 12 grams of tritium having a measured purity of 99.987% (remainder deuterium) were offloaded from the system. Observed profiles in the columns in general agree with computer models. A Height Equivalent to a Theoretical Plate (HETP) of 5.0 cm is confirmed.

An experimental apparatus has been developed to carry out tritium permeation experiments for candidate first-wall materials subjected to a high flux of low energy tritium ions, and installed in a g...

At the Tritium Process Laboratory (TPL) in the Japan Atomic Energy Research Institute, an apparatus for the Fuel Cleanup Process was designed, fabricated and installed for the experiments with up to 1g of tritium. The function of the system is continuous processing of a simulated plasma exhaust and separation of hydrogen isotopes and impurity elements in it. Main components are, palladium diffusers, catalytic reactors, cold traps, an electrolysis cell and zirconium-cobalt beds. The apparatus was installed in a glovebox and tested with hydrogen by early 1988.

An experimental apparatus for the development of cryogenic hydrogen isotope separation system (ISS) was designed and fabricated at the Tritium Process Laboratory (TPL) in Japan Atomic Energy Research Institute (JAERI). This apparatus will serve to study the separation characteristics of hydrogen isotope mixture with 10/sup 4/ Ci of tritium. It consists of two distillation columns with isotopic equilibrators for H-D-T mixtures and can be operated with both single column distillation mode and two interlinked cascade modes. Installation of this system in a glovebox was completed by late January, 1988. Tests with tritium will start late 1988.

The requirements of tritium technology are centered in three main areas, i.e., (1) fuel processing, (2) breeder tritium extraction, and (3) tritium containment. The gaseous tritium stream from the breeder tritium extraction system is significantly different from the plasma exhaust stream and, therefore, may have an important impact on the operation of the fuel processing system. For some blankets, such an aqueous solution blanket, the blanket tritium stream may dominate the fuel processing system in terms of component size and power consumption. The importance of the blanket interface to a fuel processing experiment, such as TSTA, has been identified. The initial work to define the blanket processing system, which is proposed to be added as part of TSTA, will be discussed here.

An in-situ tritium recovery experiment was conducted in the JRR-2 reactor on Li 2 O and LiAlO 2 spheres. A technique for obtaining tritium diffusion coefficients from the time dependency of tritium release after incremental temperature changes was developed. The tritium retention in LiA1O 2 was higher than in Li 2 O for the same temperature. The apparent diffusion coefficients for LiAlO 2 at temperatures up to 900°C was one order magnitude higher than the previous, lower temperature TRIO data.

Nuclear Fusion Council, Atomic Energy Commission of Japan, has started to review the nuclear fusion R and D plan for the next stage, post JT-60. The council launched a subcommittee on fundamental issues in the nuclear fusion development in 1985, for review of the basic strategy of a development plan. The subcommittee presented an interim report in Feb. 1986 after 6 months discussion and the report was approved by the Nuclear Fusion Council. Two major R and D programs described in the interim report are the development of a Tokamak type large facility and the comprehensive development of the fusion reactor technology. The latter means to promote the reactor technologies which will be essential in the future to construct not only a D/T burning but also a DEMO reactor. The Nuclear Fusion Development Program in Japan is shown. The interim report recommended to organize two subcommittees to establish an integrated national R and D plan; one was for the design of the next step large facility and the other was for the R and program of the fusion technology. The subcommittee for the latter consisted of 7 working groups; one of them was organized for the tritium technology.

This paper reports that the Japan Atomic Energy Research Institute (JAERI) has developed a full scale Fuel Cleanup System (JFCU) that processes a simulated plasma exhaust at the Tritium Systems Test Assembly (TSTA) in the Los Alamos National Laboratory (LANL). The JFCU was designed by the Tritium Process Laboratory (TPL) of the JAERI based on the component studies performed under previous TSTA-TPL collaboration and the pilot scale experiment using grams of tritium at the TPL. The JFCU accepts a simulated fusion reactor exhaust, a mixture of hydrogen isotopes with an impurity level of up to 15% at a throughput of 4.2 {times} 10{sup {minus}3} mol/s continuously and produces pure hydrogen isotopes while exhausting tritium-free waste gas. Some newly developed components, such as the Ceramic Electrolysis Cell and the large Zirconium-Cobalt bed, required special attention during fabrication and assembly.

The pressure-composition isotherms for the LaNi3Mn2 -H and D systems were obtained in the temperature range 25–300 C. The absorption plateau pressure increases from 4 Pa at 25 C to 200 kPa at 300 C...

In situ measurement of tritium release during irradiation provides a more realistic understanding of the tritium release mechanism than that obtained by the post-irradiation annealing method. The tritium release curves reached saturated values after irradiation of a few days at the reactor operation of 10 MW. The saturated values were in good agreement with the predicted values of the tritium generation rate obtained from neutronics calculation. Temperature dependence of tritium release under the temperature of 550°C at the pellet center was interpreted by the decomposition of LiOH(T) at the pellet surface. The ratio of tritium gas and tritiated water (HT/HTO) was strongly affected by the moisture generated from the piping system and the pellet. The retained tritium in Li2O was 0.5–0.9 wppm after the final stage of the irradiation of 760°C and the burnup of 0.24% of total Li.

A ceramic electrolysis cell and a palladium diffuser have been developed in Japan and tested with tritium at the Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory, to test their feasibility as possible upgrades for the fuel cleanup system (FCU). The ceramic electrolysis cell, made of stabilized zirconia, was operated at 630/sup 0/C for an extended period with a mixture of 3% T/sub 2/O in He carrier gas in a circulation system with an oxidizing catalyst bed. The palladium diffuser was tested with pure tritium gas, circulated at 280/sup 0/C, to verify the compatibility of the alloy with tritium, since the /sup 3/He produced in the metal could cause degradation. The isotopic effects were also measured for both devices.