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The current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokka...

A passive catalytic reactor without heating is required to enhance the safety of a fusion facility. A precious metal catalyst without heating is not suitable to oxidize tritium under conditions of ...

A detritiation system (DS) is required to remove tritium from the atmosphere of a nuclear containment in any extraordinary situations. Realization of a DS that does not require heating of a catalys...

To design a detritiation system (DS) applicable to a fire event, it is important to investigate influence of moisture and hydrocarbons produced by combustion of cables in fire on tritium oxidation efficiency of a catalytic reactor. In our experiments, when we assumed that tritium release in a simulated room in the following scenarios as fire situation, (I) tritium only; (II) tritium and excess moisture; (III) tritium and excess hydrocarbons, we conducted demonstration test of tritium oxidation using by DS consist of a catalytic reactor packed with platinum-alumina catalyst and a dryer with molecular sieves. We investigated the detritiation behavior and determined the volume of catalyst to satisfy requirement of tritium oxidation efficiency of a catalytic reactor packed with the catalyst. The tritium concentration theoretically decreased by DS since the tritium oxidation efficiency kept high under the conditions of three scenarios. Our results showed that the concentration of hydrogen in air and temperature of catalyst had an influence on the tritium oxidation efficiency of the catalytic reactor. We determined the worst scenario for tritium oxidation and evaluated the amount of catalyst to meet the requirement of the tritium oxidation efficiency.

To enhance public acceptance of a nuclear fusion reactor, the realization of passive detritiation system is a technical challenge taking hypothetic extraordinary situations occurred in a nuclear fusion facility into consideration. To achieve passive tritium oxidation with a room temperature recombiner (RTR), the hydrophobic platinum catalyst has been realized and data on kinetics have been accumulated. Since the efficiency of hydrogen oxidation at room temperature decreases with a decrease in hydrogen concentration, the worst condition of the process gas for RTR is the lowest hydrogen in the process gas. To overcome this condition, we have proposed to add hydrogen intentionally to be 100 ppm at the inlet of RTR. In this study, the dynamic behavior of a tritium leakage in a 12 m3 simulated room and the detritiation with a RTR was demonstrated to confirm the effectiveness of our proposal considering the worst condition for RTR. Initial tritium concentration in the containment was 1.0 GBq/m3. The flow rate was set to 1.8 Nm3/h. The volume of the hydrophobic platinum catalyst was 1000 cm3. Tritium concentration in the containment smoothly decreased one order magnitude after processing for 900 min. We obtained a prospect that a sufficient hydrogen conversion rate can be secured by adding hydrogen of about 100 ppm. From the kinetics point of view, the required amount of catalyst for a RTR is estimated to be three times that for a heated reactor at 473 K in order to obtain the same conversion efficiency of tritium.

The Tokamak exhaust processing system (TEP) in the deuterium-tritium fuel cycle for DEMO is the key system where hydrogen isotopes and impurity gases are separated. Then the impurity gases are decomposed into hydrogen isotopes and other chemical species. The major impurity gas to be processed in TEP is tritiated hydrocarbons. Among tritiated hydrocarbons, tritiated methane is hard to be decomposed directly since the temperature of catalyst for thermal decomposition is 500 °C or above with conventional precious metal supported catalysts such as Pt, Pd, and Rh. The effect of constitutive elements of catalysts on methane decomposition was investigated for a catalyst that could decompose methane efficiently at low temperatures. The investigation showed that the activity of the catalyst for methane decomposition depends on some constitutive elements such as precious metal species, particle sizes of metals, pore sizes of catalyst supports. The experimental results indicated that methane could be decomposed effectively by optimizing precious metal species and the particle sizes of precious metals. Especially the particle size of precious metals had a great influence on methane decomposition efficiency. It was notable that the decomposition efficiency rose 5 times by enlarging a particle size from 12 nm to 23 nm.

In the fuel cycle of ITER and the future DEMO fusion reactors, which has many processes of tritium provision/ consumption/recovery/loss, such as, fueling, DT burning, vacuum pumping, purification, isotope separation and storage, detritiation, etc., tritium accountancy control will be required in its operation and safety management. In contrast to ITER, which will have frequent intervals between operations, a DEMO fusion reactor generating electricity must be operated continuously for long periods. This requires a dynamic tritium accountancy control, especially at the mass balance area boundary of the Tokamak including the vacuum vessel. To make a dynamic tritium accountancy system for a DEMO fusion reactor, based on our past work carried out during the ITER EDA, we are developing a new compact hydrogen isotope gas analysis system by the application of laser Raman spectroscopy advanced by technological progress in laser and detection systems. A prototype system has been made at TPL using a semiconductor laser with a Peltier-cooled CCD detector, with a simple optical system. Results of some preliminary tests being conducted to confirm sensitivity and accuracy under static and dynamic conditions are presented.

The potential halogen effects on Detritiation System (DS) was investigated. DS is required to remove tritium from the atmosphere of a nuclear containment even in an event of fire. Basically several precautions should be taken to avoid the formation of halogenated gases by fire since there is concern that halogen can cause loss of both catalyst and water absorber activities. However, halogens will be present in many facility components of a fusion facility such as cables and electrical cubicles which are in direct contact with room atmosphere. Combustion products from halogenated hydrocarbons by fire will be a mixture of many compounds such as CO, CO2, NOx, HTO, HCl. We have already presented the effect of HTO, H2O, hydrocarbons, CO, CO2 on DS. In this study we assessed the impact of halogen on DS and proposed a mitigation measure. HCl and Cl2 caused decrease in activity of a platinum catalyst. We make clear a mitigation measure by poisoning with halogenated gases is the application of a noble catalyst alloyed with platinum and palladium. The outstanding proof of the alloyed catalyst was measured with molecular iodine which is known as the worst halogen to poison catalyst. Chlorine gas caused decrease in water absorbing capacity of molecular sieve absorbent. The data in this paper affords the perspective on the limited use of halogenated materials in a fusion facility.

A method combined oxidation catalytic reactor and water bubbler is the most quantitative analysis method for gaseous tritium. We developed a tritium measurement system with high accuracy using an organic-based hydrophobic platinum catalyst previously. However, the previous measurement system had a shortcoming that the catalyst was combustible over 100 °C. To overcome the shortcoming, we developed an inorganic-based hydrophobic platinum catalyst (Pt/SiO 2 ) which was noncombustible. The developed Pt/SiO 2 catalyst was specially enhanced the hydrophobicity to prevent adsorption of water vapor on the surface of catalyst. To oxidize tritium efficiently, the particle diameter of 1mmϕ was selected for the catalyst and hydrogen was intentionally added at the inlet of the catalytic reactor to be the optimal concentration for oxidation. The performance tests showed that a modified quantitative tritium analysis system using by the hydrophobic Pt/SiO 2 catalyst provides tritium measurement with high accuracy more than 99.9%. We have confirmed the modified quantitative tritium analysis system is available in various conditions.

Technical reliability of tritium confinement shall be elevated taking hypothetic extraordinary situations occurred in a nuclear fusion facility such as events of fire and loss of electric power fully into consideration in order to enhance public acceptance of a nuclear fusion reactor. Considerable attention has been paid to the research of passive tritium conversion in the research field of detritiation system. Demonstration of detritiation to grasp the dynamic behavior is practically important to enhance the tritium confinement. In this study, passive detritiation of a 12 [m3] container was demonstrated with hydrophobic catalyst packed in a catalytic reactor. Initial tritium concentration in the container was 1.0 [GBq/m3]. The volume of hydrophobic catalyst packed in the passive catalytic reactor was 1000 [cm3]. The flow rate was set to 2.4 [Nm3/h] which is equivalent to atmosphere exchange rate of 5 times per day. The tritium concentration in the container successfully decreased two order magnit...

Thoughtful consideration of abnormal events such as fire is required to design and qualify a detritiation system (DS) of a nuclear fusion facility. Since conversion of tritium to tritiated vapor over catalyst is the key process of the DS, it is indispensable to evaluate the effect of excess moisture and hydrocarbons produced by combustion of cables on tritium conversion rate considering fire events. We conducted demonstration tests on tritium conversion under the following representative conditions: (I) leakage of tritium, (II) leakage of tritium plus moisture, and (III) leakage of tritium plus hydrocarbons. Detritiation behavior in the simulated room was assessed, and the amount of catalyst to fulfill the requirement on tritium conversion rate was evaluated. The dominant parameters for detritiation are the concentration of hydrogen in air and catalyst temperature. The tritium in the simulated room was decreased for condition (I) following ventilation theory. An initial reduction in conversion rat...

To design a recombiner packed with hydrophobic catalyst for passive tritium oxidation, the development of manufacturing technology for hydrophobic platinum catalyst and the following analysis on reaction rate are essential. This presentation deals with the reaction rate for tritium oxidation over a hydrophobic platinum catalyst at temperature between room temperature and 473 K. Two kinds of hydrophobic platinum catalysts; supported with styrene-divinylbenzene and with silica were used for this test. The results indicate that 1) the particle size of platinum affects the reaction rate, 2) the order of reaction is 0.5 of hydrogen concentration at room temperature in case of a hydrogen content below 100 ppm. The rate-determining step approximation method is applied for the discussion on the reaction mechanism of tritium oxidation over platinum catalyst.

Various methods of tritium measurement have been applied depending on a chemical form of tritium. A method combined oxidation catalyst and water bubblers has been used as one of the most quantitative analysis methods for gaseous tritium. The major technical shortcoming with the conventional methods is the lowering of quantitative accuracy caused by adsorption of produced tritiated vapor on hydrophilic oxidation catalyst. Then, we developed an organic-based hydrophobic platinum catalyst. The catalyst enhanced its hydrophobicity to prevent absorption of water vapor on catalyst surface. In order to ensure to oxidize tritium, we additionally devised 1) the particle size of the catalyst made small to be around 1 mm ϕ and 2) intentional addition of hydrogen. As the result of experimental verification of the proposed method, we succeeded to achieve the high accuracy of more than 99.9% in the quantitative measurement of tritium. The developed tritium measurement system is applicable widely for various tritium measurements, estimation of tritium generation rate at TBM, tritium behavior in various materials and so on.

In the JA DEMO, the deuterium-tritium (DT) fuel will be supplied by a combination of gas puff, pellet injection and neutral beam injection. Since the combustion rate of fuel in the JA DEMO is approximately 1.7 %, it is necessary to establish the closed deuterium-tritium fuel cycle in the facility for fuel recycling. The fuel cycle for DEMO inevitably differs from that for ITER. Compared to the ITER fuel cycle, which requires high flexibility for fuel supply based on the requirements of plasma experiments, the flexibility required for the fuel cycle decreases in the DEMO reactor. The flow in the fuel cycle becomes steadier. Therefore, the fuel cycle of the DEMO shifts to the continuous processing system. In addition, considering the tritium breeding in the blanket, the fuel cycle must be established so that tritium in the DEMO facility can be balanced. The process components of the fuel cycle that process a large amount of tritium-containing gas are provided with multiple barriers that confine tritium, and the detritiation system processes the confined tritium. The confinement of tritium in the DEMO facility according to nuclear safety regulations is an important safety issue. As a result, the fuel cycle becomes a complex chemical plant. Assuming safety requirements, it is necessary to consider the reduction of the tritium inventory as much as possible at the design stage of the DEMO fuel cycle. The tritium inventory in the fuel cycle is closely related to the requirements to pellet manufacturing and hydrogen isotope separation, and the reduction of the requirements to the hydrogen isotope separation is the major issue in the fuel cycle. For this purpose, it is necessary to minimize the D/T separation based on the concept of direct fuel cycle with the D/T mixture. For the possible shortage in the preparation of the initial loading tritium for the DEMO reactor, it is necessary to design the fuel cycle through dynamic numerical simulation so that the fuel cycle can be started with the minimum initial loading tritium. In the system design of the fuel cycle, it is necessary to support various operation modes from the startup of DT operation to steady operation. The requirements to design the JA DEMO fuel cycle are being investigated in the JA DEMO design activity.

The development of a water cooled ceramic breeder (WCCB) test blanket module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

We have performed the tritium release experiment on the fusion reactor blanket at JAEA/FNS since 2009, and then clarified the ratio of tritium release and the recovered tritium chemical form. In order to acquire the detailed tritium recovery performances, we have started a new blanket tritium recovery experiment with ionization chamber (IC) at JAEA/FNS. For the appropriate tritium measurement with IC, we improved the experimental container and carried out with an intense DT neutron source at JAEA/FNS. From our new experiment, the tritium recovery radioactivity from the LSC measurement corresponds with the calculation within 6%. However, it was pointed out that further improvement in the quantitative tritium measurement by IC method was needed.

Li2TiO3 with excess Li (Li2+xTiO3+y) has been developed as an advanced tritium breeder. With respect to the tritium release characteristics of the blanket, the optimum grain size after sintering was less than 5 μm. Therefore, an emulsion method was developed to fabricate pebbles with this target grain size. The predominant factor affecting grain growth was assumed to be the presence of binder in the gel particles; this remaining binder was hypothesized to react with the excess Li, thereby generating Li2CO3, which promotes grain growth. To inhibit the generation of Li2CO3, calcined Li2+xTiO3+y pebbles were sintered under vacuum and subsequently under a 1% H2–He atmosphere. The average grain size of the sintered Li2+xTiO3+y pebbles was less than 5 μm. Furthermore, the tritium release properties of Li2+xTiO3+y pebbles were evaluated, and deuterium–tritium (DT) neutron irradiation experiments were performed at the Fusion Neutronics Source facility in the Japan Atomic Energy Agency. To remove the tritium produced by neutron irradiation, 1% H2–He purge gas was passed through the Li2+xTiO3+y pebbles. The Li2+xTiO3+y pebbles exhibited good tritium release properties, similar to those of Li2TiO3 pebbles. In particular, the released amount of tritiated hydrogen gas for easier tritium handling was greater than the released amount of tritiated water.

The accurate measurement of behavior of bred tritium released from a tritium breeder is indispensable to understand the behavior for a design of a tritium extraction system. The tritium collection method combined a CuO bed and water bubbles was not suitable to measure transient behavior of tritium released from Li2TiO3 during neutron irradiation because tritium released behavior was changed to be delayed due to adsorption of oxidized tritium on the CuO. Hence, the tritium collection method with hydrophobic catalyst instead of the CuO was demonstrated and succeeded the accurate release measurement of tritium from Li2TiO3. With the method, we assessed the behavior of tritium release under the various conditions since tritium should be released from Li2TiO3 as the form of HT as much as possible from the view point of the fuel cycle. Our results indicated; promotion of isotopic exchange reaction on the surface of Li2TiO3 by addition of hydrogen in sweep gas is mandatory in order to release tritium smoothly from Li2TiO3 irradiated with neutrons; the favorable sweep gas to release as the form of HT was hydrogen added inert gas; and the temperature of Li2TiO3 was the dominant parameter to control the chemical form of tritium released from the Li2TiO3.

Detritiation system of a nuclear fusion plant is mandatory to be designed and qualified taking carefully into consideration all the possible extraordinary situations in addition to that in a normal condition. We focused on the change in the efficiency of tritium oxidation of a catalytic reactor in an event of fire where the air accompanied with hydrocarbons, water vapor, and tritium is fed into a catalytic reactor at the same time. Our test results on the effect of these gases on the efficiency of tritium oxidation of the catalytic reactor indicated; (1) tritiated hydrocarbon produces significantly by reaction between tritium and hydrocarbons in a catalytic reactor; (2) there is little possibility of degradation in the detritiation performance because the tritiated hydrocarbons produced in the catalyst reactor are combusted; (3) there is no possibility of uncontrollable rise in the temperature of the catalytic reactor by heat of reactions; and (4) saturated water vapor could temporarily poison the...

Absract Air detritiation system of a fusion facility consists of catalyst reactors for tritium oxidation and a system to remove tritiated vapor from air. In the event of fire in the facility, the impact of gaseous impurities produced by polymeric materials on catalytic oxidation of tritium is one of the evaluation items. The point is the impact of reaction heat needs to be detected since gaseous impurities such as hydrocarbons are likely to combust in the catalyst reactors as well as heat generation by the combustion reaction is large. The distributions of temperature and concentration in a catalytic reactor are numerically evaluated using the thermophysical and chemical properties. The numerical analysis showed that the temperature in a catalytic reactor rose to 1173 K by the combustion of ethylene and hydrogen when air containing 1% concentration of hydrogen and 1% concentration of ethylene was fed into the reactor heated initially at 473 K. Since the increase in temperature is large, further technical consideration will emerge for designing such as the selection of material for a reactor, the increase in tritium permeation, and the impact of excessively heated air on equipment located at the downstream. The heat removal mechanism needs to be designed for catalytic reactors.

We have developed two types of hydrophobic platinum honeycomb catalyst to be used for tritium oxidation reactors. One is the hydrophobic platinum catalyst on a metal honeycomb. The other is the hyd...

The temperature of Li 2 TiO 3 pebble breeder in a fusion DEMO blanket is assumed to be more than 1000 K. For the investigation of tritium release from a Li 2 TiO 3 pebble breeder blanket at such a high temperature, we have carried out a tritium release experiment with the DT neutron source at the JAEA-FNS. The Li 2 TiO 3 pebble (1.0–1.2 mm in diameter) of 70 g was put into a stainless steel container and installed into an assembly stratified with beryllium and Li 2 TiO 3 layers. During the DT neutron irradiation, the temperature was kept at 1073 K with wire heaters in the blanket container. Helium gas including 1% hydrogen gas (H 2 /He) mainly flowed inside the container as the purge gas. Two chemical forms, HT and HTO, of extracted tritium were separately collected during the DT neutron irradiation by using water bubblers and CuO bed. The tritium activity in the water bubbler was measured by a liquid scintillation counter. To investigate the effect of moisture in the purge gas, we also performed the same experiments with H 2 O/He gas (H 2 O content: 1%) or pure helium gas. From our experiment at 1073 K, in the case of the purge gas includes H 2 , it is indicated that the increasing tendency of HT release is similar to that of the dry H 2 /He.

Activities on Broader Approach (BA) were started in 2007. In Phase 2-2, many R&Ds, development of tritium accountancy technology, development of basic tritium safety research and tritium durability test, were implemented successfully by JAEA and Japanese Universities. In Phase 2-3, new collaborative study for tritium measurement and new R&D activities for JET ILW are started. R&D activities on BA have continued in Phase 2-3 (2014–2016).

Development of an efficient tritium recovery method is indispensable in order to compose a liquid blanket system of a D-T fusion reactor in the near future. Here, tritium recovery using a bubbling ...

Hydrogen transfer under a fluidized condition of Li-Pb is investigated experimentally to design a Li-Pb blanket system. Li-Pb eutectic alloy flows through inside of a Ni tube in the experimental sy...

JAEA has two tritium handling laboratories, such as Tritium Processing Laboratory (TPL) in Tokai and DEMO R&D building in Rokkasho. Specially, TPL has been accumulated a gram level tritium safety h...

What affects tritium migration through porous concrete walls coated with a hydrophobic paint is reviewed from the viewpoint of tritium safety. Being taken into consideration of multi-structural con...

Li2TiO3 advanced tritium breeders with excess Li (Li2+xTiO3+y) are stable in a reducing atmosphere at high temperatures. Although the tritium release properties of tritium breeders are documented in databases for DEMO blanket design, no in situ examination under fusion neutron (DT neutron) irradiation has been performed. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed, and DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. Considering the tritium release characteristics, the optimum grain size after sintering is >5 μm. From the results of the optimization of granulation conditions, prototype Li2+xTiO3+y pebbles with optimum grain size (>5 μm) were successfully fabricated. The Li2+xTiO3+y pebbles exhibited good tritium release properties similar to the Li2TiO3 pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water.

The present study aims at obtaining fundamental data on tritium migration in porous materials, which include soaking effect, interaction between tritium and cement paste coated with paints and tran...

Behavior of tritium transfer through hydrophobic paints of epoxy and acrylic-silicon resin was investigated experimentally. The amounts of tritium permeating through their paint membranes were measured under the HTO concentration condition of 2–96 Bq/cm 3 . Most of tritium permeated through the paints as a molecular form of HTO at room temperature. The rate of tritium permeating through the acrylic-silicon paint was correlated in terms of a linear sorption/release model, and that through the epoxy paint was controlled by a diffusion model. Although effective diffusivity estimated by a diffusion model was obtained 1.1 × 10 −13 –1.8 × 10 −13 m 2 /s for epoxy membranes at the temperature of 21–26 °C, its value was found to be hundreds times larger than that for cement-paste coated with epoxy paint. Hence, resistance of tritium diffusion through interface between cement-paste and the epoxy paint was considered to be the most effective in the overall tritium transfer process. Clarification of tritium transfer behavior at the interface is important to understand the mechanism of tritium transfer in concrete walls coated with various paints.

To investigate the behavior of hydrogen on a water–metal boundary, a series of experiments have been performed that studied tritium permeation into a pressurized water jacket through pure iron piping, which contained approximately 1 kPa of pure tritium gas at 423 K, while monitoring the chemical forms of tritium. Additionally, the behavior of deuterium, which was generated on the heavy water–metal boundary and transferred to the metal, was also investigated using a pressure-proof heavy water vessel. Actual deuterium transfer through various metal piping, such as Fe, Ni, SS304, etc., was detected clearly by QMS at 423–573 K. Moreover, using the above heavy water system, we have succeeded in detecting hydrogen isotopes simultaneously transferring from and to the metal surface by introducing hydrogen gas to the Ni piping after deuterium permeation from the heavy water side had stabilized.

The Li2TiO3 is one of candidate breeding materials of a water cooled ceramic breeding (WCCB) blanket. In order to clarify the tritium recovery property of the WCCB blanket with the Li2TiO3 breeding material, we have performed the tritium recovery online experiment with the DT neutron source at the Fusion Neutronics Source facility in Japan Atomic Energy Agency (JAEA-FNS). We irradiated an experimental assembly simulating the WCCB blanket and recovered the tritium recovered from the Li2TiO3 pebbles put into the assembly with a heater system, sweep gases and bubblers. The activity of the recovered tritium was measured with a liquid scintillation counter. From our tritium recovery online experiment and calculation, the followings were found out: (1) the recovered tritium corresponded to the calculated tritium production within the experimental error in the range of 573–1073 K and (2) the recovered HTO tended to be easily recovered at lower temperature and high water moisture. The recovered HT increases at higher temperature and dry hydrogen circumstance. However, the maximum level of the tritium gas recovery is around 90% even at higher temperature and 1% H2 circumstance.

Tritium recovery system using adsorption or catalytic isotope exchange has already been proposed for a solid breeding blanket system of a nuclear fusion reactor. Synthetic zeolite is often used as an adsorbent or a substrate of chemical exchange catalyst. And, it is well known that its property is changed easily by exchanging its cations. Synthetic mordenite is one of zeolites having fairly large silica/alumina ratio. There are many reports about hydrogen adsorption properties of cation-exchanged mordenite so far. And, the present authors also have reported that cation-exchanged mordenite with Ca ion (Ca–MOR) indicated fairly large hydrogen adsorption capacity at 77 K in comparison with Molecular Sieves 5A (MS5A). So, in this work, hydrogen adsorption properties of cation-exchanged mordenite with transition metal ion were investigated mainly. The cation-exchanged mordenite with Ag ion (Ag–MOR) has indicated considerably large hydrogen adsorption capacity in lower pressure range at 77 K in comparison with Ca–MOR. The discussion from the viewpoint of adsorption rate is still remaining, but more compact cryosorption column for tritium recovery system is possible to design if Ag–MOR is adopted.

Recovery of D dissolved in a liquid Li flow at low D concentration is experimentally investigated using a Y metal absorber under the two fluidized conditions: (a) in a vertical cylindrical tube and (b) in an agitated vessel. The target concentration is 1 appm in Li around at 300 degrees C. The two concentrations of D remaining in Li and recovered by Y are detected by a dissolution method using H2O with depleted-D and HNO3. The main released species is HD. A small amount of HDO released is reduced to HD by a Mg particle bed. It is found that HF-treated Y can absorb H isotopes at the target temperature and concentration. The chemical dissolution technique is found to be useful to specify the two absolute concentrations of D recovered by Y and D remaining in Li. (C) 2014 Elsevier B.V. All rights reserved.

Hydrogen isotope separation is important process in fusion reactor. Cryosorption is one of the methods for hydrogen isotope separation, and is often employed to gas chromatograph (GC) for hydrogen isotope analysis. Synthetic zeolite is used as the packing material of separation column for GC. And, it is well known that the retention time becomes long at 77 K. A zeolite having good separation capability at higher temperature may bring short retention time with good sensitivity. Mordenite type zeolite (MOR) in which cation was exchanged has been reported having comparatively good isotope separation capability. The present authors have investigated adsorption capacity of hydrogen isotope on cation-exchanged MORs at various temperatures. And, it has been shown that MOR with Ca 2+ as exchanged cation (Ca-MOR) has fairly large adsorption capacity. In this work, the breakthrough curves of hydrogen adsorption on Ca-MOR at 194, 175 K have been observed, and mass transfer coefficients or effective diffusivity in Ca-MOR particle have been estimated. The rate-controlling step is the diffusion process in the adsorbent particle. The isotope effect on hydrogen diffusion in Ca-MOR particle is larger than raw material (Na-MOR). Therefore, Ca-MOR is more suitable for the separation column than Na-MOR.

Hydrogen permeation is investigated in a conjugated system of Li15.8Pb84.2 eutectic alloy or Ar under natural convection inside a cylindrical vertical sus316 tube. Two different temperature conditi...

IFMIF-EVEDA progresses in Japan as one of the EU-Japan Broader Approach Activities. The research is performed to decide whether or not IFMIF is constructed after some uncertainties included in the ...

In order to clarify the tritium behavior in Pb­Li blanket system, the solubility of hydrogen isotopes in the Pb­Li eutectic alloy was experimentally investigated. The effects of (i) Li preferential evaporation from Pb­Li, (ii) thermal convection in a quartz-glass reaction tube and (iii) difference in crucible materials between Al2O3 and W on the solubilities were discussed. Another important issue of the Pb­Li blanket system is to develop a new ceramic coating to decrease tritium permeation rate. Tritium permeation test was performed for Er2O3 coating on F82H. The permeation reduction factor achieved here was higher than 10 3 . Precise experiments were carried out to determine the solubility, diffusivity and permeability of hydrogen isotopes in Pb­Li eutectic alloy by means of a transient permeation method. The isotope effects in solubility and diffusivity between H and D were clarified. No isotope effect in solubility was observed between H and D, and the H diffusivity was 1.4 times larger than the D one. The ratio was near to a value predicted from the classical theory. Experiment using tritium is designed under the collaboration work with INL. [doi:10.2320/matertrans.MG201203]

A passive reactor for tritium oxidation at room temperature has been widely studied in nuclear engineering especially for a detritiation system (DS) of a tritium process facility taking possible extraordinary situation severely into consideration. We have focused on bacterial oxidation of tritium by hydrogen-oxidizing bacteria in natural soil to realize the passive oxidation reactor. The purpose of this study was to examine the feasibility of a bioreactor with hydrogen-oxidizing bacteria in soil from a point of view of engineering. The efficiency of the bioreactor was evaluated by kinetics. The bioreactor packed with natural soil shows a relative high conversion rate of tritium under the saturated moisture condition at room temperature, which is obviously superior to that of a Pt/Al2O3 catalyst generally used for tritium oxidation in the existing tritium handling facilities. The order of reaction for tritium oxidation with soil was the pseudo-first order as assessed with Michaelis-Menten kinetics model. Our engineering suggestion to increase the reaction rate is the intentional addition of hydrogen at a small concentration in the feed gas on condition that the oxidation of tritium with soil is expressed by the Michaelis-Menten kinetics model.

The use of D–Li stripping reaction to generate multiple neutron is expected in the International Fusion Materials Irradiated Facility (IFMIF). Tritium generated by its side reaction needs to be recovered from liquid Li for safety. Y hot trap is expected as an effective purification system to recover tritium from the liquid Li loop. In IFMIF, since liquid Li circulates in a loop, it is important to trace tritium behavior in Y and liquid Li under dynamic conditions. In this study, the authors investigate the effectiveness of tritium recovery from liquid Li under dynamic conditions of Y hot trap. Hydrogen absorption in Li instead of tritium is investigated under dynamically fluidized conditions by stirring liquid Li with an arbitrary rotational rate. The experimental results prove that the hydrogen absorption rate from liquid Li to Y is independent of the rotational rate. This means that the rate-determining step is hydrogen diffusion in Y.

A lead–lithium eutectic alloy (Pb–Li) is one of the most promising candidate materials for the liquid blanket of an advanced fusion reactor. We have experimentally determined mass-transfer properties by an unsteady permeation method, which data are necessary to design a system to recover tritium (T) from a Pb–Li blanket. An experiment of simultaneous H and D permeation through Li17Pb83 is performed to clarify interactions between atoms in the two-component permeation process. The experimental results are analyzed by a model of one-dimensional or two-dimensional permeation through Li17Pb83. The major permeation proceeds in the longitudinal direction of the present system, and the ratio of hydrogen leak in the radial direction is evaluated using the simulation. As a result, it was found that H and D atoms permeate independently regardless of the H/D component ratio within the present experimental conditions. The permeability and diffusivity of H are 1.4 times higher than that of D. The solubility of H is close to that of D. The isotope effect in diffusivity is in proportion to the square root of the mass ratio of D to H. When these data can be extended to the case of T, T permeability and diffusivity is predicted as 1/1.7 times lower than that of H in the temperature range from 773 K to 973 K.

Tritium release behavior for thermal neutron-irradiated Li0.17Pb0.83 eutectic alloy was studied. Main tritium release peak was observed in the temperature just a little higher than melting point in...

An experimental study on tritium (T) transfer in porous concrete for the tertiary T safety containment is performed to investigate (i) how fast HTO penetrates through concrete walls, (ii) how well concrete walls contaminated with water-soluble T are decontaminated by a solution-in-water technique, and (iii) how well hydrophobic paint coating works as a protecting film against HTO migrating through concrete walls. The experiment is comparatively carried out using disks of cement paste which W(water)/C(cement) weight ratio is 0.6:1 with or without hydrophobic paints, and mortar disks which W/C/sand ratio is 0.6:1:2 with or without the paints. The hydrophobic paints tested in the present study are an epoxy polymer resin paint and an acrylic-silicon polymer resin one. After T exposure during specified time under a constant HTO vapor pressure in an acrylic box, the amount of water-soluble HTO on/in the disks is determined using a technique of H2O dissolution during specified time. The results obtained here are summarized as follows: (1) HTO penetration in porous concrete can be correlated in terms of the effective diffusivity. (2) Its value in porous cement without coating is 1.2 × 10−11 m2/s at 25 °C. (3) HTO penetrates only through pores in cement, and there is no path for HTO transfer in non-porous sand. (4) Rates of sorption and dissolution of HTO in disks of cement and mortar coated with the epoxy resin paint are correlated in terms of the effective diffusivity through the paint film which value is DT = 1.0 × 10−16 m2/s. The rate-determining step is diffusion through the paint. (5) The epoxy resin paint works more effectively as an anti-HTO diffusion coating. (6) Another acrylic-silicon resin paint does not work well as anti-HTO diffusion coating. This may be because the hydrophobic property of the silicon resin paint is deteriorated with elongating the contact time with H2O vapor or liquid. (7) The HTO uptake inside the epoxy paint is greater than that of the silicon one. (8) The permeation reduction factor (PRF) of HTO for the epoxy paint at steady-state is expected large, if HTO vapor only contributes to diffusion. However, when concrete surfaces coated with the epoxy paint are under wet conditions, the PRF value becomes smaller. All those results can be used to estimate the effect of HTO soaking in concrete walls in case of accidental T release in a fusion reactor room and to decontaminate wastes of tritiated concrete.

Systems to recover tritium (T) from fusion liquid blankets of LiPb (Li17Pb83) and Flibe (Li2BeF4) under safety conditions are discussed based on available data of T transport properties. We recentl...

In IFMIF/EVEDA, tasks for lithium target system consist of 5 validation tasks (LF1-5) and a design task (LF6), and are shared by Japan and Europe. Japan is covering LF1, LF2, LF4, LF5 and LF6, and JAEA as an Implementing Agency is organizing these tasks. Japanese universities contribute these tasks. The purpose of LF1 task is to construct and operate EVEDA Li test loop (ELTL) with diagnostics and purification systems for the validation of Li target facility. ELTL will be constructed at the end of February in 2011 and the operation will be started at the summer in 2011. LF2 and LF4 are basic research for the diagnostics and purification system, respectively. And LF2 and LF4 have also contributed the construction of the ELTL (LF1). LF5 is focusing on the remote handling for the replacement of the target assembly. Major deliverable of LF6 is finally engineering design report, which is assembled with all validation tasks, and this activity will be gradually enhanced to the end of IFMIF/EVEDA.

In order to estimate mass-transfer properties of tritium (T) in Li + Pb binary alloys as a blanket material for fusion reactors, variations of the H solubility with different Li and Pb compositions are presented. Deviation of the Li + Pb binary liquid mixture from an ideal solution model is discussed, and a correlation is proposed for predicting properties of various Li + Pb alloys. Isotope effects on solubility and diffusivity of Li 17 Pb 83 between H and D are correlated based on an isotropic harmonic oscillation model. The zero-point energy of H atoms dissolved in Li 17 Pb 83 is determined from the comparison with the model. Wettability at the interface between liquid Li 17 Pb 83 and metallic walls and effects of impurities on H solubility are discussed.

We performed an experiment of H and D permeation through Li17Pb83 in the two-component (H+D) system by an unsteady permeation method to clarify interactions between H and D atoms. It was found that...

The permeability, diffusivity and solubility of hydrogen isotopes (H, D) in a Li–Pb eutectic alloy are measured in the temperature range from 573 to 973 K using a permeation method. Their isotope effects between H and D are obtained. It is shown that the solubility of hydrogen isotopes in Li–Pb obeys the Sieverts’ law. The solubility of D is around 1.4 times larger than that of H. The isotope effect of the diffusivity between H and D is very small.