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The current status and the progress of R&D activities for a Fusion DEMO reactor in the QST Rokkasho Fusion Institute is reported. In order to advance the JA DEMO activity, not only Japanese domestic activity but also international collaborations of Broader Approach Activity and ITER related activities are conducted in the QST Rokkasho Fusion Institute. Activities for DEMO design and relevant R&D, design of a fusion neutron source and development of an accelerator, ITER-TBS, tritium handling technology and information technology infrastructures including a supercomputer system and a remote experimentation system are carried out for a Fusion DEMO reactor.

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...

核融合炉ブランケットはプラズマからの熱および中性子負荷を繰り返し受け、その第一壁はクリープ温度以上になることが想定されるため、クリープ疲労損傷の評価は重要な検討課題の一つである。本研究では、ブランケットの構造材料である低放射化フェライト鋼F82Hを用いて、ASTM規格に沿って、大気中かつ550℃の全ひずみ範囲1%の両振りひずみ制御条件において、圧縮保持時間180 sから1800 sまでのクリープ疲労試験を実施した。純疲労に比べて、圧縮保持の導入に伴い、寿命の低下が見られた。保時時間700 s以上のとき、寿命の低下は僅かであり、応力緩和の挙動や非弾性ひずみの増加は飽和する傾向が示唆された。, 第38回プラズマ・核融合年会

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.

Several key components, such as superconducting coils, remote handling equipment, heating systems, have been designed and manufactured by JADA (Japan Domestic Agency). These activities have been carried out in accordance with the agreed schedule; in collaboration with the ITER organization and other domestic agencies. As a significant technical program using ITER, to design and to manufacture the TBS (Test blanket system), some R&D and design activities have also been conducted in Japan. Under the IFERC (the International Fusion Energy Research Center) projects of BA (Broader Approach) activities, design and R&D activities on fusion DEMO reactor have been carried out. For the DEMO R&D activity, five basic R&D subjects for a DEMO blanket system have been selected, and been studies under close collaborations between EU and JA: structure materials (RAFM steels and SiC/SiC composites), functional materials (tritium breeders and neutron multipliers), and tritium technology. From 2007, the above project...

A steep hill slope composed of bedrock often deforms slowly but steadily by the action of gravity due to its heavy weight of bedrock slope. This is called “gravitational deformation”. On the other hand, in the Himalayas and other hilly regions in India, road administrators have suffered from sinking of hill roads for long years at many sections on steep hill slopes. According to the steep gradient of the sinking section, normal landslides with landslide clay on slip surface are not responsible for the sinking. Through a technical cooperation project in India funded by JICA, we confirmed relation between sinking of hill roads and gravitational deformation.

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.

Rabbits are valuable experimental animals used to evaluate the teratogenicity of chemical entities. If teratogenicity is recognized, it is very important to investigate the pathogenesis to assess the potential risk in humans. For this investigation, whole-mount in situ hybridization (WISH) is highly useful for visualizing spatial and temporal changes in the expression patterns of the targeted genes and or proteins. In this chapter, the WISH method in rabbit embryos, using a digoxigenin-labeled RNA probe, and the morphometric analysis of the expression pattern is described.

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).

The authors have applied the isotherm equations derived from the reactive vacancy solution theory (RVST) to correlation of experimental and highly non-ideal adsorption isotherms of hydrogen and deuterium on a mordenite adsorbent, and have examined the ability of the isotherm equations to match this correlation. Several isotherm equations such as Langmuir, Freundlich, Toth, Vacancy Solution Theory and so forth were also tested, but they did not work. For the Langmuir-Freundlich equation tests have indicated that its 'ability to correlate' of the adsorption isotherms is not satisfactory. For the multi-site Langmuir-Freundlich (MSLF) equation the correlation of the isotherms appears to be somewhat improved but remains unsatisfactory. The results show that the isotherm equations derived from RVST can better correlate the experimental isotherms.

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...

Rabbits are valuable experimental animals used to evaluate the teratogenicity of chemical entities. If teratogenicity is recognized, it is very important to investigate the pathogenesis to assess the potential risk in humans. For this investigation, whole-mount in situ hybridization (WISH) is highly useful for visualizing spatial and temporal changes in the expression patterns of the targeted genes and or proteins. In this chapter, the WISH method in rabbit embryos, using a digoxigenin-labeled RNA probe, and the morphometric analysis of the expression pattern is described.

It is revealed that binding position of fatty acid in triacylglycerol (TAG) deeply relates to the expression of its function. Therefore, we investigated the binding positions of individual trans-octadecenoic acid (trans-C18:1) positional isomers, known as unhealthy fatty acids, on TAG in partially hydrogenated canola oil (PHCO), milk fat (MF), and beef tallow (BT). The analysis was carried out by the sn-1(3)-selective transesterification of Candida antarctica Lipase B and by using a highly polar ionic liquid capillary column for gas chromatography-flame ionization detection. Trans-9-C18:1, the major trans-C18:1 positional isomer, was selectively located at the sn-2 position of TAG in PHCO, although considerable amounts of trans-9-C18:1 were also esterified at the sn-1(3) position. Meanwhile, trans-11-C18:1, the major isomer in MF and BT, was preferentially located at the sn-1(3) position. These results revealed that the binding position of trans-C18:1 positional isomer varies between various fats and oils.

Trans fatty acids (TFA) are considered risk factors for cardiovascular disease (CVD), while the details of distribution and metabolism of the individual isomers are not clear. Here we investigated the accumulation and catabolic rate of TFA positional isomers of octadecenoic acid (18:1) in mice. ICR mice were fed deuterium- and [1-(13)C] stable isotope-labeled trans-9-18:1 (9t-18:1*), trans-10-18:1 (10t-18:1*), or trans-11-18:1 (11t-18:1*) for 2 or 4 weeks, or a TFA mixture (9t-18:1*, 10t-18:1*, and 11t-18:1*) for 3 weeks. Analysis of whole-body tissues by gas chromatography-chemical ionization mass spectrometry revealed the highest 9t-18:1* levels in the heart. Significant differences in the accumulation of the respective trans-18:1 were observed in the heart and erythrocytes, where 9t- > 11t- > 10t-18:1*, but no significant difference was observed in the liver or white adipose tissue (WAT). Mice fed on 11t-18:1 demonstrated accumulation of endogenously synthesized conjugated linoleic acid in the liver, WAT, and heart, but any other metabolites were not found in other groups. Furthermore, we analyzed catabolic rates of single-dose-administered trans-18:1* isomers into [(13)C]-labeled CO2 using isotope-ratio mass spectrometry, and the 10t-18:1*catabolic rate was significantly higher than those of 9t- and 11t-18:1*. We found that the accumulation and catabolism of trans-18:1 positional isomers varied in these mice. Differential accumulation in tissues suggests that individual TFA positional isomers may play different roles in human health.

Since the current financial reporting uses various measurement attributes for measuring assets and liabilities, it is often referred to as the “mixed measurement model.” This paper addresses the issues related to why and how different measurement attributes, notably fair value and historical cost, are used in financial statements. The cost-benefit analysis is adopted as a research methodology throughout this paper by formalizing as mathematical functions the preparer’s measurement cost and the users’ benefits of accounting information that the preparer produces. The cost-benefit analysis suggests that the mixed measurement model can be justified through the preparer’s economic incentives for minimizing cost. It also suggests that the cost-minimization solution in the marketplace would represent a consensus among the interests of the preparer and the users within a constraint in which only one set of accounting information is provided. This paper also identifies the situations in which providing two sets of accounting information is preferable.

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.

Thalidomide (TM) induces limb defects in humans and some animal species including rabbits. Although the mechanism of TM-induced limb defects has been investigated for a long period, the limb development-related genes expressions have not been vigorously characterized in rabbits. In this study, we investigated the Fgf8, Bmp4 and Hoxa11 expressions in TM-treated JW rabbit embryos on gestation days (GDs) 10, 11 and 12 by whole mount in situ hybridization. On GDs 10 and 11, growth retardation of the embryo was induced by TM treatment. The Fgf8 expression lengths on GDs 10 and 11 in the forelimb bud were significantly or tended to be decreased in the TM-treated embryos, which was correlated to the growth retardation and was not considered to be directly relevant to the teratogenic effect of TM in the forelimb. The TM-induced characteristic changes in the expression pattern of Hoxa11 and Bmp4 on GDs 10 and/or 11 were not noted. On GD 12, TM-induced growth retardation was not noted and the Fgf8 and Bmp4 expressions were not changed. On the contrary, Hoxa11 expression was narrowed at the anterior region, which was located on the radial side, and was not changed at the middle and posterior regions in the forelimb bud and in all regions in the hindlimb bud. Because the radius malformations were induced by TM treatment, we concluded the decrease in the Hoxa11 expression was related to the TM-induced limb defects and can be a good marker for early prediction of the teratogenic effect of TM.