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In this study, creep-fatigue (CF) tests on F82H at 823 K were conducted according to the standard specimen and requirement of ASTM. Tension holding (TH) showed higher lifetime than compression holding (CH). Compared with 9Cr-1Mo-V steels, CF lifetime degradation was not observed at the holding time less than 0.05 h in case of CH and 0.17 h in case of TH. CF damages were summarized from the CF test results by Campbell method. The obtained damages distributed around CF damage envelope of Grade 91 in ASME NH. CF damage of cylindrical water-cooled ceramic breeder (WCCB) test blanket module (TBM) container at the zenith of hemisphere body under the ITER thermo-mechanical load was calculated. The calculated damage located within of the damage envelope of Grade 91. It was indicated that the present design of cylindrical WCCB TBM might resist CF interaction under the ITER thermo-mechanical load conditions.

Water cooled ceramic breeder (WCCB) blankets have been regarded as a primary concept in Japan. We in-vestigated the tritium breeding capability of a WCCB blanket with a cylindrical structure, an advantageous shapefor withstanding high coolant pressure, in this study. The breeding area in the blanket was modeled as ahomogeneous mixture or heterogeneous structures. Nuclear analyses with Monte Carlo method were conductedfor evaluating the tritium breeding ratio (TBR) of the blanket. The neutron balance of the blanket was analyzedto elucidate the mechanism related to the increase in TBR when an additional thin breeding layer (envelope) wasapplied to the blanket. The neutron multiplication in (n, 2n) reaction increased concomitantly with increase ofthe beryllium volume ratio. However, numerous neutrons were not used efficiently for tritium production, butwere captured by the container, which was made of the reduced activation ferritic martensitic steel, F82H.Capture by the container can be reduced by introducing an envelope. The effect of the envelope was considerablewhen modeling the internal structure such as U-shaped pipes in the breeding area. The envelope was also appliedto a blanket with different container designs. An increase in TBR was shown irrespective of blanket design. Theenvelope effect was remarkable when it was difficult to achieve the internal configuration, which was equivalentto the homogeneous mixture in the breeding area

A water-cooled ceramic breeder blanket has been developed. Heat flux and coolant pressure are applied to the first wall of the blanket in a fusion reactor. A square cooling channel is adopted in the first wall in Quantum and Radiological Science and Technology (QST). This study specifically addressed the relation between the thermal mechanical responses with different cooling channel shapes and the allowable load of the first wall. Thermal mechanical responses of the first wall were evaluated using finite element method (FEM). Based on those results, the allowable loads for the structural integrity of the first wall were assessed in accordance with the requirements of ASME Sec. III NH. Among the failure modes the creep-fatigue damage was the most dominant on the allowable loads of the first wall. The allowable loads were determined by characteristics of the thermal mechanical responses with the cooling channel shape. The major determinants were the surface heat flux and the coolant pressure, respectively, for the circular and the square channels.

In this study, creep-fatigue (CF) tests on F82H at 823 K were conducted according to the standard specimen and requirement of ASTM. Tension holding (TH) showed higher lifetime than compression holding (CH). Compared with 9Cr-1Mo-V steels, CF lifetime degradation was not observed at the holding time less than 0.05 h in case of CH and 0.17 h in case of TH. CF damages were summarized from the CF test results by Campbell method. The obtained damages distributed around CF damage envelope of Grade 91 in ASME NH. CF damage of cylindrical water-cooled ceramic breeder (WCCB) test blanket module (TBM) container at the zenith of hemisphere body under the ITER thermo-mechanical load was calculated. The calculated damage located within of the damage envelope of Grade 91. It was indicated that the present design of cylindrical WCCB TBM might resist CF interaction under the ITER thermo-mechanical load conditions.

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.

It is known that the chemical forms of tritium released from a ceramic breeder blanket are hydrogen form and water form. To recover tritiated water vapor, adoption of dryer that is packed column of synthetic zeolite has been proposed. On the other hand, synthetic zeolite is often used as a support of precious metal catalyst. Such catalysts usually have a capability of hydrogen isotope exchange between gas and water vapor. If this catalyst is used to dryer, the dryer may obtain a preferable function for tritium recovery by isotopic exchange reaction. To assess such functions, reaction rate should be estimated. The results of water adsorption experiment on cation-exchanged mordenite-type zeolite suggested the possibility that state of adsorbed water varied by exchanged cation. So, in this work, precious metal catalyst based on cation-exchanged mordenite was prepared, and the reaction rate of chemical exchange between hydrogen and tritiated water was investigated under temperature range between 30 °C and 80 °C by the steady-state approximation. In the case of platinum on Na-mordenite, the reaction between gaseous hydrogen and tritiated water vapor was almost expressed as first-order reaction concerning tritiated water vapor concentration.

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.

Test Blanket Systems (TBS) will be operated in ITER in order to prepare the next steps towards fusion power generation. After the initial operation in H/He plasmas, the introduction of D and T in ITER will mark the transition to nuclear operation. The significant fusion neutron production will give rise to nuclear heating and tritium breeding in the in-vessel part of the TBS. The management of the activated and tritiated structures of the TBS from operation in ITER is described. The TBS specific features like tritium breeding and power conversion at elevated temperatures, and the use of novel materials require a dedicated approach, which could be different to that needed for the other ITER equipment.

The Water Cooled Ceramic Breeder Test Blanket module (TBM) has been developed for a demonstration of the major functions under the fusion reactor environment in ITER. Testing of the TBM shall neither hinder the ITER operation, nor threaten safety of operation even at water ingress into a box structure of the TBM (In-Box Loss of Coolant Accident, In-Box LOCA). In present study, the structural responses on the TBM under In-Box LOCA were evaluated by using Finite Element Method. Based on the analysis results, the TBM design was modified. The filled region and Tritium Breeding Ratio of the modified TBM design adding the fillet with 35 mm of radius increased by up to 20%, 12%, respectively, compared to the original TBM model without fillet.

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.

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.

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.

Damage cause and mechanism of the tungsten specimen under ELM-like heat pulse were examined by using a Nd:YAG laser experiment assuming the ITER divertor allowable heat load, 0.5 MJ/m2 and finite-element analysis in terms of thermal mechanical response. The thermal stress distribution with the heat load area was compared by FEM and the stress distribution by a Nd:YAG laser corresponded to that expected in the actual divertor target. The crack growth direction was changed from a perpendicular to parallel and this was caused by the thermal stress distribution in isotropic microstructure under ELM-like heat pulse. In addition the change of crack growth direction caused the deterioration of thermal conduction near the surface and it led to the local surface melting near cracks even at 0.51 MJ/m2 with 3000 heat load cycles.

A Water-Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) System is being developed as one of the most important steps toward DEMO blanket in Japan. As for the ITER-TBM program, the conceptual design of WCCB-TBM system has been approved by the ITER organization (IO). And two years have already passed after the start of the preliminary design activity. WCCB TBM Team had a concern about TBM box structure withstanding over 15 MPa of the coolant pressure, and decided to change the design as the result of study on structural integrity. Recently, the design change of the TBM structure has been approved domestically. So, WCCB TBM Team has applied the design change to the IO, and agreed with the IO to receive the conceptual design review which is limited the scope to the TBM structure. This paper provides an overview of the recent achievements of the development of the WCCB Test Blanket Module System in Japan.

In test blanket module (TBM) program blanket major functions, which are tritium production, heat extraction for electric power production, and neutron shielding, will be demonstrated under fusion reactor environments in ITER. National Institutes for Quantum and Radiological Science and Technology (QST) has performed developments of water cooled ceramic breeder (WCCB) TBM as a primary concept in Japan. The container design of WCCB TBM was changed from the Box-shaped to a cylindrical structure to increase tritium breeding capability while maintaining pressure resistance under In-Box LOCA caused by water ingress into the container. The nuclear responses related to the blanket major functions would be changed with the container design change. The nuclear responses of WCCB TBM with the container design change were evaluated and the effects of the changed nuclear responses were described in this study. Tritium production rate increased by about 2 times with the container design changes. In contrast neutron shielding performance decreased due to the container design change. Degradation of neutron shielding in the cylindrical TBM led to an increase in nuclear heating of the TBM frame as well as an average of neuron flux behind TBM-set.

In this study, electromagnetic analyses of cylindrical water-cooled ceramic breeder test blanket module under the three types of ITER plasma disruption scenarios were conducted. Highest Lorentz force (LF) along radial direction was observed under plasma disruption scenario of MD-II. Symmetrical LF along two-eight o'clock position was observed in side wall (SW) of container and could be canceled in summation. Integrated LF of a container was mainly contributed by first wall and rear part of SW. Compared with previous design of box-shaped container, lower magnitude of LF was shown in current design of cylindrical container, which was mainly understood by difference in the volume of the solid structure. Maxwell force (MF) was additionally evaluated considering the ferromagnetic characteristic of F82H. Magnitude and distribution of MF was much higher than LF. Besides connection structure, the structural integrity of container should be taken into account when ferromagnetic characteristics were considered.

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).

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.

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li2TiO3 pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.

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.

The multi-purpose RI facility has been constructed at Rokkasho site in DEMO R&D building until 2011. The facility is the first and quite unique facility in Japan, where tritium, beta and gamma RI species, and beryllium (Be) can simultaneously be used. The amounts of tritium used and stored are 3.7 TBq per glove box and 7.4 TBq, respectively. Some tritium water samples of 38 GBq has been stored at the equipment on March 2012. The material of the column of the micro gas chromatograph has been studied to develop a real time analysis tool for the hydrogen isotope composition in gas phase. The calorimeter has also been studied as a possible tritium measurement method in solid waste. A set of basic data on the interaction between materials and tritium has been measured by various methods. The behavior of tritium in Fe and W has been studied as a typical subject. As a study for the tritium durability, the endurance of ion exchange membrane has been tested by using high concentration tritium water. The curves of strength vs. dose for the Nafion membranes in tritium water were well consistent with those by gamma rays and electron beams irradiations.

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.

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 evaluation toward DEMO blanket, the module fabrication technology development by a candidate structural material, reduced activation martensitic/ferritic steel, F82H, is one of the most critical items from the viewpoint of realization of TBM testing in ITER. In Japan, fabrication of a real scale first wall, side walls, a breeder pebble bed box and assembling of the first wall and side walls have succeeded. Recently, the real scale partial mockup of the back wall was fabricated. The fabrication procedure of the back wall, whose thickness is up to 90 mm, was confirmed toward the fabrication of the real scale back wall by F82H. Important key technologies are almost clarified for the fabrication of the real scale TBM module mockup. From the view point of testing and evaluation, development of the technology of the blanket tritium recovery, development of advanced breeder and multiplier pebbles and the development of the blanket neutronics measurement technology are also performed. Also, tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been started as the verification test of tritium production performance. This paper overviews the recent achievements of the development of the WCCB TBM in Japan.

In a fusion reactor, the prediction of tritium release behavior from breeder blanket is important to design the tritium recovery system, but the amount of tritium generated is necessary information to do that. Hence, tritium generation and recovery studies on lithium ceramics packed bed have been started by using fusion neutron source (FNS) in Japan Atomic Energy Agency (JAEA). Lithium titanate (Li 2 TiO 3 ) was selected as tritium breeding material, and its packed bed was enclosed by the beryllium blocks, and was kept at certain temperature during fusion neutron irradiation. During irradiation, the packed bed was purged with the sweep gas continuously, and tritium released was trapped in each gas absorber selectively by chemical form. In this work, the effect of sweep gas species on tritium release behavior was investigated. In the case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in the case of sweep by helium without water vapor, tritium in gaseous form was released first, and release of tritium in water form was delayed from gaseous tritium and was gradually increased.

In JAEA, the tritium processing and handling technologies have been studied at TPL (Tritium Process Laboratory). The main RD the basic tritium behavior in confinement materials; and detritiation and decontamination. The R&D activities on tritium processing and handling technologies for a demonstration reactor (DEMO) are also planned to be carried out in the broader approach (BA) program by JAEA with Japanese universities. The ceramic proton conductor has been studied as a possible tritium processing method for the blanket system. The BIXS method has also been studied as a monitoring of tritium in the blanket system. The hydrogen transfer behavior from water to metal has been studied as a function of temperature. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). A new hydrophobic catalyst has been developed for the conversion of tritium to water. The catalyst could convert tritium to water at room temperature. A new Nafion membrane has also been developed by gamma ray irradiation to get the strong durability for tritium.

Electrochemical hydrogen pump can extract hydrogen isotopes selectively from a blanket sweep gas, because one of its driving forces of hydrogen transportation is electric potential difference. In this work, transportation properties of multi-component hydrogen isotopes including tritium were investigated experimentally. Ceramic proton conductor membrane used in this work was SrCe 0.95 Yb 0.05 O 3− α made by TYK Co. H 2 –D 2 mixture balanced with He was supplied to the hydrogen pump, that was kept at 873 K and applied the voltage between 0 and 1000 mV, and the permeated gas was analyzed with a gas chromatograph. The rate-determining step of hydrogen transfer via ceramic membrane seems to be close to the diffusion step in the bulk of ceramic with increase of the applied voltage. HT–H 2 mixture balanced with He was also supplied to the hydrogen pump, and recovery of tritium via ceramic membrane was done. Diffusivity of hydrogen in the proton conductor ceramic is larger than that of tritium. Hence, hydrogen is enriched in the gas of the permeation side. In this experimental condition, about 40% of tritium supplied was recovered and the decontamination factor was 1.5.

The RD (2) basic tritium safety research; and (3) tritium durability test. The EU joins the discussions and assessment of the R&D results. As the recent activities on the tritium technologies of the BA program, a multi-purpose RI facility, where the above R&D subjects will be carried out, has been designed in detail. A preliminary safety study has also been carried out for the amount of tritium released to the environment and for the radiation dose of workers. The facility is now under construction at Rokkasho in Aomori. The main subjects of the R&D of tritium analysis are the technologies for real-time analysis for hydrogen isotopes, gas, liquid and solid (such as microGC and calorimeter). The materials of interest include F82H, SiC, ZrCo, solid and liquid advanced breeder and multipliers. In the tritium durability tests, organic materials and metals planned to be used in a DEMO plant are studied for the radiation and the corrosion damage. A series of preliminary studies for the above subjects has been started.

At Tritium Process Laboratory (TPL) of Japan Atomic Energy Agency (JAEA), tritium technologies for a fusion reactor have been carried out up to date. The design studies of Air Detrtiation and Vent Detritiation System (ADS/VDS) of ITER have been carried out in JAEA as a contribution of Japan to ITER. For the tritium processing technologies, our efforts have been focused on the R&D of the tritium recovery system of ITER test blanket system. A ceramic proton conductor has been studied as an advanced blanket system. A series of fundamental studies on tritium safety technologies for ITER and for fusion DEMO plants has also been carried out. The main R&D activities in this field are the tritium behavior in a confinement and its barrier materials, monitoring, accountancy, detritiation and decontamination, etc. Especially, for the fundamental studies for a DEMO plants, a part of the studies will be carries out a new facility at Rokkasho in Aomori in the Broader Approach (BA) program in coming 10 years. For this purpose, a design study of the facility at Rokkasho has first been started.

A gas chromatography using a cryogenic separation column is one of the methods for hydrogen isotope analysis. However, the use of the refrigerant such as liquid nitrogen is the cause of long analysis time and is not suitable for easy installation. The development of the column material having separation capability at comparatively high temperature region is one of the solutions for these weak points. Mordenite (MOR) is a kind of a zeolite, and it has been reported that the separation column using MOR has possibility to separate hydrogen isotope mixture at comparatively high temperature. In this work, the separation columns using MOR were made and tested. The peaks of H2 and D2 were mostly separated at 144 K, but they were not separated at 195 K. MOR column adjusted in this work was still not for the practical use. However, this result suggests the possibility of the existence of the synthetic zeolite that can separate hydrogen isotope mixture at comparatively high temperature.

Electrochemical hydrogen pump with ceramic proton conductor membrane has been proposed to apply for a blanket tritium recovery system (BTR) of a fusion reactor. The advantage of this system is that it can apply to the system of low hydrogen pressure, because the driving force of hydrogen permeation is the electric potential difference. Perovskite-type ceramic such as SrCe0.95Yb0.05O3−α (SCO) is one of the candidates of membrane. To apply hydrogen pump to BTR, the enhancement of the hydrogen transportation capability is necessary. Modification of electrode is one of the methods to enhance the hydrogen transportation capability. In this work, the electrodes of platinum (Pt) and palladium (Pd) were attached to the SCO sample by the sputtering method (sputtering electrode), and its electric conductivity and proton conductivity were measured. Then, they were compared with that of the usual Pt paste electrode. Hydrogen transportation capability was enhanced when the sputtering electrode was applied. Especially, in case of the Pd sputtering electrode, the current density which was about 4 or 5 times larger than the usual Pt paste electrode was observed at 0.1% of H2 concentration.

In a fusion reactor system, a monitoring of hydrogen isotopes including tritium is necessary for the safety of system control and operation. A gas chromatography using a cryogenic separation column is one of the methods for hydrogen isotope analysis. Synthesis zeolite such as molecular sieve 5A (CaA) is a candidate material of the separation column, and its property varies by the ratio of silica to alumina, the kinds of cation and so on. If the factor affected the hydrogen adsorption property of the synthesis zeolite is clarified, it may lead to the development of the new zeolite optimized to the separation column. So, in this work, adsorption capacity of hydrogen (H2) and deuterium (D2) for mordenite (MOR) and NaY type zeolite (NaY) were investigated at various temperatures, and were compared with CaA. The amount of adsorption per unit weight of MOR was larger than that of CaA, and that of NaY was smaller than that of CaA. The adsorption isotherms were expressed by sum of two Langmuir equations, and the Langmuir coefficients of H2 and D2 were proposed. Furthermore, the Langmuir coefficients of HD, HT, DT and T2 were estimated by the reduced mass. The correlation between the adsorption properties and the physical parameters of the zeolite were not confirmed.

The blanket tritium recovery system using the electrochemical hydrogen pump with proton conductor membrane has been proposed. The feature of the electrochemical hydrogen pump is that the driving force of hydrogen transportation is a potential difference. Therefore, it might be effective to apply the hydrogen pump to the blanket sweep gas (the low hydrogen and water vapor pressure). Perovskite-type ceramic such as SrCe 0.95 Yb 0.05 O 3− α , is one of the candidate proton conductor for hydrogen pump and its ionic hydrogen transportation properties have been investigated. In this work, the basic mass transfer equation for hydrogen, in which the apparent proton conductivity is used as the over-all mass transfer coefficient, is proposed. And then, the apparent proton conductivities were estimated from experimental data using these equations, and mass transfer of hydrogen via proton conductor membrane was discussed by using the apparent conductivity.

Our idea on the DEMO plant is that it must demonstrate (1) an electric power generation of one GW level, (2) self-sufficiency of tritium fuel (TBR is more than 1.05), (3) year-long continuous operation, etc. At the same time, DEMO is expected to use technologies to be proven by 2020 and present an economical prospect of fusion energy in the operational time of the reactor. The design guidelines for the blanket are defined in order to meet the mission of the DEMO plant as mentioned above. Major design conditions are surface heat flux of 0.5 MW/m2 with peaking factor of 2, a neutron wall load of 3.5 MW/m2 with peaking factor of 1.5 and a neutron fluence of about 10 MW/m2. To moderate the technological extrapolation, reduced activation ferritic steel (F82H) structural material, Li2TiO3 and Be neutron multiplier are considered. To improve the economical aspect, supercritical water with inlet/outlet temperatures of 280/510 °C is chosen as coolant material, with coolant pressure of 25 MPa. As a result, a thermal efficiency of 41% is achieved. To obtain higher plasma performance, MHD instabilities suppressing shell structure is adopted with structural compatibility to the blanket structure. To meet higher plant availability requirements (more than 75%), the hot cell maintenance approach is selected for the replaceable power core components.

Several systems for the tritium recovery from cooling water of a blanket of a fusion power plant have been designed. The tritium concentration in the cooling water and the tritium permeation rate to the coolant is assumed to be 370 GBq(10 Ci)/kg, 13 and 130 g/day, respectively. For the case of 13 g/day, the system can be composed of a water distillation (WD: 2.6 m inner diameter and ∼50 m height) and a catalytic exchange column with an electrolysis cell (0.7 m inner diameter and 22 m in height). The WD column can be replaced by a system of catalytic exchange columns in vapor and liquid phases. For the case of 130 g/day, no solution can be found as long as quite a large flow rate to a main fuel cycle for its final treatment is permitted. Otherwise, an electrolysis cell that can be used under a high concentration of tritium water will need to be developed. The tritium inventory of the WD column is appreciably large, so that it is desirable to develop a system having a high separation factor and with no liquid phase.

Tritium bred in the solid breeder blanket of a fusion reactor is extracted by passing H 2 added helium sweep gas through the blanket itself. In the blanket tritium recovery system (BTR), tritium is separated from sweep gas. Palladium (Pd) membrane diffuser is one of the applicable processes for BTR. Recently, the conceptual design study on the demonstration reactor with supercritical water cooling blanket, that is named ‘DEMO2001’, has been carried out in JAERI. In this report, the application of the Pd diffuser to the blanket sweep gas condition is discussed based on DEMO2001 conditions. The counter flow may be the most opportune flow type for the Pd diffuser. However, Pd diffuser is suitable for the secondary process, like the purification process, after the tritium in the sweep gas has been concentrated by another method.

Various analytical techniques exist to determine the compositions of gases handled in the fuel cycle of future fusion machines. Gas chromatography was found to be the most appropriate method. The main disadvantages of conventional gas chromatography were the long retention times for the heavy hydrogen species of >30 min. Recent progress in the development of micro-gas chromatography has reduced these retention times to ≈3 min. The usefulness of micro-gas chromatography for the analysis of hydrogen and impurity gas mixtures in the fuel cycle of future fusion machines is presented and the advantages and drawbacks are discussed.

Analysis of hydrogen isotopes is very important in the fuel cycle system of fusion reactors. Gas chromatography with a cryogenic separation column is one of the most extensively used methods for the analysis of hydrogen isotopes. Micro gas chromatograph (micro GC) with a cryogenic column is expected to improve the analysis time, that is a major disadvantage of conventional GC. The present authors have modified the micro GC to use its separation column at cryogenic temperature. In previous work a micro packed column has been tested and indicated fairly good performance. In this work a capillary column was tested at cryogenic temperature for more improvement of analysis time. Obtained retention time of H 2 , HD and D 2 were about 43, 47 and 54 s, respectively, for the column with 0.53 mm of I.D., 4.0 m of length and 0.08 mm of film thickness. Peak resolution between H 2 and HD was about 1.12. These results suggest that the column developed in this work attained the practical level for the separation and short analysis time. The estimation of the retention time was carried out using the dispersion model. The retention times of HT, DT and T 2 were also estimated using the reduced mass.

In the fuel cycle system of fusion reactors, analysis of hydrogen isotopes is very important from the view point of system control. The gas chromatograph (GC) with cryogenic separation column (cryogenic GC) is one of the most extensively used methods for the analysis of hydrogen isotopes. The micro GC with cryogenic column is expected to improve analysis time, that is a major disadvantage of conventional GC. The present authors have modified the micro GC to use its separation column at cryogenic temperature for H2, HD and D2 mixture analysis. Obtained retention time of H2, HD and D2 was about 85, 100 and 130 s, respectively. Peak resolution between H2 and HD, these are nearest each other, was about 1.0. These result suggests that the column developed in this work attained the practical level for the separation of hydrogen isotopes without tritium. Present detection limit of hydrogen isotopes was about 100–200 p.p.m., and it can be improved further by adjustment of separation column.

In order to confirm tritium confinement ability in the deuterium–tritium (DT) fusion reactor, intentional tritium release experiments have been started in a specially fabricated test stand called ‘Caisson’, at Tritium Process Laboratory in Japan Atomic Energy Research Institute. The Caisson is a stainless steel leak–tight vessel of 12 m3, simulating a reactor room or a tritium handling room. In the first stage experiments, about 260 MBq of pure tritium was put into the Caisson under simulated constant ventilation of four times air exchanges per h. The tritium mixing and migration in the Caisson was investigated with tritium contamination measurement and detritiation behavior measurement. The experimental tritium migration and removal behavior was almost perfectly reproduced and could almost be simulated by a three-dimensional flow analysis code.

The Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute has been working on the development of fuel processing technology for fusion reactors as a major activity. A fusion fuel processing loop was installed and is being tested with tritium under reactor relevant conditions. The loop at the TPL consists of ZrCo based tritium storage beds, a plasma exhaust processing system using a palladium diffuser and an electrolytic reactor, cryogenic distillation columns for isotope separation, and analytical systems based on newly developed micro gas chromatographs and Raman Spectroscopy. Several extended demonstration campaigns were performed under realistic reactor conditions to test tritiated impurity processing. A sophisticated control technique of distillation column was performed at the same time, and integrated fuel circulation was successfully demonstrated. Major recent design work on the International Thermonuclear Experimental Reactor (ITER) tritium plant at the TPL is devoted to water detritiation based on liquid phase catalytic exchange for improved tritium removal from waste water.

Although lithium ceramic materials such as Li 2 O, LiAlO 2 , Li 2 ZrO 3 and Li 4 SiO 4 are considered as breeding materials in the blanket of a D-T fusion reactor, the release behavior of the bred tritium in these solid breeder materials is not yet fully understood. Most results of in situ tritium release experiments are analyzed assuming that the overall release process of tritium is mainly controlled with tritium diffusion in the crystal grain of a solid breeder material. However, the diffusivities by various authors for solid breeder materials do not agree with each other. An insufficient estimation of the surface reactions, the effect due to irradiation defects or the system effect may be the reason. Effects of diffusion of tritium in the grain, absorption of water in the bulk of grain, and adsorption of water on the surface of grain, together with two types of isotope exchange reaction on the tritium inventory in the steady-state condition, are discussed in this study. Comparisons of tritium inventory estimated in this study with data obtained in several in-situ experiments are also performed, and good agreement is obtained when the existence of some water vapor is assumed in the purge gas.

To understand the release behavior of bred tritium in a solid breeder blanker, the tritium transfer rate and tritium inventory for various mass transfer processes should be investigated. The contribution of the surface reactions (adsorption, desorption and two kinds of isotope exchange reactions) to the release process cannot be ignored. It is believed that two kinds of isotope exchange reactions (gaseous hydrogen-tritiated water and water vapor-tritiated water) occur on the surface of the solid breeder materials when hydrogen is added to the sweep gas to enhance the tritium release rate. The isotope exchange reaction study in H–D systems was carried out using a Li 2 ZrO 3 packed bed. The exchange reaction between gaseous hydrogen and water was the rate controlling step among the two kinds of exchange reactions. The reaction rate constants were quantified, and experimental equations were proposed. The equilibrium constant of the isotope exchange reaction in the H–D system was obtained from experimental data and was found to be 1.17.