Hirose, Takanori; Seki, Yohji; Tanigawa, Hisashi; Tanigawa, Hiroyasu; Tsuru, Daigo; Enoeda, Mikio; Serizawa, Hisashi*; Yamaoka, Hiroto*
Fusion Engineering and Design, 85(7-9), p.1426 - 1429, 2010/12
This paper describes packing experiment of tritium breeder pebbles into a full-scale Tritium-Breeder-Container (TBC) mockup. A full scale mockup of the TBC for Water Cooled Solid Breeder - Test Blanket Module has been successfully developed using a reduced activation ferritic steel, F82H. A full-scale TBC mock-up was successfully fabricated with the fiber laser welding, and its dimensions are 74 112 990 mm. It was confirmed to be gastight under pressurized helium up to 0.5 MPa. By using the fabricated mockup, packing tests were performed with LiTiO pebbles of 1mm diameter. The pebbles were packed into the TBC through sweep gas lines penetrating the tube plates. X-ray tomography revealed that dense packing was uniformly achieved in the whole TBC.
Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Enoeda, Mikio; Mori, Seiji
Fusion Engineering and Design, 85(7-9), p.1451 - 1454, 2010/12
Japan Atomic Energy Agency (JAEA) is willing to procure the outer Vertical Targets (VT) in cooperation with ITER Organization. In advance of the start of the procurement, the JAEA has to first demonstrate its technical capability to carry out the procurement. This is achieved via the successful manufactures and quality tests of VT Qualification Prototypes. Non-Destructive Examination (NDE) with the infrared thermography is required as one of the quality tests to detect the defect in the CFC monoblock, and between the CFC/OFCu. In this research and development, the Facility of Infrared NDE for Divertor (FIND) has been built by the JAEA. The FIND successfully detects the position and the magnitude of the integrated defect in the CFC and in the bonding of CFC/OFCu. The infrared NDE system established in the JAEA contributes to keeping the quality of the ITER-divertor.
Iwai, Yasunori; Sato, Katsumi; Hiroki, Akihiro; Tamada, Masao; Hayashi, Takumi; Yamanishi, Toshihiko
Fusion Engineering and Design, 85(7-9), p.1421 - 1425, 2010/12
The deteriorations of polymeric materials for a SPE-type high-level tritiated water electrolyzer composed of the Water Detritiation System (WDS) against sulfonic acid environment and radiation environment were discussed. A long-term durability of VITON, AFLAS, denaturated polyphenylene ether, and Kapton polyimide immersed in a sulfonic acid was demonstrated. Negligible degradation in percent elongation at break of these polymeric materials was observed up to the immersing period of 2 years. The detectable radiation deterioration in ionic conductivity of Nafion N117CS ion exchange membrane irradiated with electron beams up to the integrated dose of 1500 kGy was measured. The ionic conductivity of Nafion N117CS ion exchange membrane irradiated at more than 1000 kGy was slightly deteriorated. As for the elastomers for its use as a seal, the radiation degradation in hardness of VITON, AFLAS was investigated. Negligible degradation in hardness of these rubbers was observed up to the integrated dose of 1500 kGy. The water uptake of rubbers was generally increased as the integrated dose was increased. However, irradiated VITON rubbers had constant water uptake up to the integrated dose of 1500 kGy.
Masaki, Kei; Miyo, Yasuhiko; Sakurai, Shinji; Ezato, Koichiro; Suzuki, Satoshi; Sakasai, Akira
Fusion Engineering and Design, 85(10-12), p.1732 - 1735, 2010/12
Steady-state research is indispensable to establish scientific and technological basis for the next fusion devices. In JT-60, long pulse operation of up to 65s (OH) with a neutral beam heating power of 12 MW (30s) was conducted to investigate the plasma behavior in several tens of seconds. However, the structure of the JT-60U first wall, which was composed of bolted graphite tiles and backings, restricted the flexibility of the plasma operation, because the first wall was not actively cooled. To improve the heat transfer characteristics of the first wall taking into account the cost, a candidate is to insert a graphite sheet between the graphite tile and the backing plate. Aiming at a design study for next fusion devices, the heat transfer characteristics of the first wall structure were investigated with a variety of graphite sheets and fixing-bolt torque conditions. The first wall mockup used for the experiment was composed of three CFC tiles (125(L) 110(W)24(T) mm for each tile) and a cupper-alloy heat sink (377(L)100(W)20(T) mm) with two cooling channels of 10 mm diameter. Four types of the graphite sheets, 0.1-mm thickness PGS (Pyrolytic Graphite Sheet; Panasoic Co., Ltd), 0.2-mm PF (Perma Foil; Toyo Tanso Co., Ltd) 0.38-mm PF, 0.6-mm PF, were examined in the experiment. The heat load tests of the mockup were performed with the heat fluxes of 1 and 3 MW/m on the JAERI electron beam irradiation stand. The experimental results showed that the structure with 0.1-mm thickness 3 PGSs had the highest heat transfer performance in the experiment. The first wall structure with the PGS sheets withstood the heat flux of 1 MW/m100s. The maximum surface temperature of the CFC tile was 500C. Furthermore, the results indicated that the structure could be used at the steady-state condition with the heat flux of 1 MW/m. In the paper, detail of the results will be presented and discussed.
Nakamura, Hirofumi; Isobe, Kanetsugu; Nakamichi, Masaru; Yamanishi, Toshihiko
Fusion Engineering and Design, 85(7-9), p.1531 - 1536, 2010/12
Tritium permeation reduction performance of various coatings was evaluated by means of deuterium permeation experiment and thermal stress analysis. Several coatings such as amorphous CrPO, plasma sprayed zirconia (ZrO) and gold plating. As the results, ZrO showed relatively good permeation reduction performance about 1/10, and deuterium seems permeate through the grain boundaries or cracks of the ZrO. Improvement such as filling the diffusion path will be effective for further permeation reduction performance. On the other hand, gold plating showed excellent permeation reduction performance around 550 K. Permeation reduction by gold plating would be attributed to low solubility of deuterium in gold. Results of thermal stress analysis were consistent with the results of permeation experiment. Therefore, thermal stress analysis could be an effective method for the selecting permeation reduction coatings prior to the permeation experiment.
Kondo, Keitaro; Tatebe, Yosuke; Ochiai, Kentaro; Sato, Satoshi; Takakura, Kosuke; Onishi, Seiki; Konno, Chikara
Fusion Engineering and Design, 85(7-9), p.1229 - 1233, 2010/12
In the previous blanket neutronics experiments conducted at the FNS facility of Japan Atomic Energy Agency, the following disagreements between experiments and analyses have been pointed out: (1) In the experiment with a Li-enriched LiTiO layer and a beryllium layer, approximately 10% overestimation was found for the tritium production rate (TPR) when a neutron reflector composed of SS316 was attached. (2) In the experiment with natural LiO pebbles sandwiched by beryllium layers, TPR was overestimated near the rear beryllium layer by up to 10%. In order to confirm the above problems clearly, a new blanket neutronics experiment using a natural LiTiO layer and beryllium layers with DT neutrons was conducted at FNS. TPR distributions inside the LiTiO layer were measured with LiCO pellets with and without the source reflector. The measured TPR well agreed with the calculation within an estimated experimental error of 6% in the both experiments. The influence of the reflector was not remarkable in the present experiment. Contrary to our expectation, no remarkable difference was observed in the TPR distribution around the rear beryllium layer.
Araki, Masanori; Sakamoto, Yoshiteru; Hayashi, Kimio; Nishitani, Takeo; Ouchi, Rei*
Fusion Engineering and Design, 85(10-12), p.2196 - 2202, 2010/12
For contributing to the ITER project and promoting a possible early realization of DEMO, the IFERC project shall perform the activities on (1) DEMO Design and R&D Coordination, (2) Computational Simulation Centre, and (3) ITER Remote Experimentation Centre in the framework under the BA agreement. The DEMO design activity aims at establishing a common basis for DEMO design, including design features of DEMO, a possible common concept of DEMO, a roadmap for DEMO, and so on. Based on the common interest toward DEMO, the DEMO R&D activities have been planned and carried out for the following five areas which are relevant to blanket development: (1) SiC/SiC composites, (2) tritium technology, (3) materials engineering for DEMO blanket, (4) advanced neutron multiplier for DEMO blanket, and (5) advanced tritium breeders for DEMO blanket. In the activity of the Computational Simulation Centre, the objective is to provide and exploit a supercomputer for large scale simulation activities to analyse experimental data on fusion plasmas, prepare scenarios for ITER operation, predict the performance of the ITER facilities and contribute to the DEMO design. At the initial phase, high-level benchmark codes in the fusion research field have been selected through Special Working Group.
Sato, Satoshi; Nashif, H.*; Masuda, Fukuzo*; Morota, Hidetsugu*; Iida, Hiromasa*; Konno, Chikara
Fusion Engineering and Design, 85(7-9), p.1546 - 1550, 2010/12
Automatic conversion systems from CAD data to MCNP geometry input data have been developed to convert the CAD data of the fusion reactor with very complicated structure. So far, three conversion systems (GEOMIT-1, ARCNCP and GEOMIT-2) have been developed. The void data can be created in these systems. GEOMIT-1 was developed in 2007, and a lot of manual shape splitting works for the CAD data were required to successfully convert the complicated geometry. ARCNCP was developed in 2008. The algorithm has been drastically improved on automatic creation of ambiguous surface in ARCNCP, and manual shape splitting works can be drastically reduced. The latest system, GEOMIT-2, does not require additional commercial software packages, though the previous systems require them. It has also functions of the CAD data healing and the automatic shape splitting. The geometrical errors of the CAD data can be automatically revised by the healing function, and the complicated geometries can be automatically split into the simple geometries by the shape splitting function. Any manual works are not required in GEOMIT-2. The latest system is very useful for nuclear analyses of fusion reactors.
Kondo, Hiroo; Kanemura, Takuji*; Sugiura, Hirokazu*; Yamaoka, Nobuo*; Ida, Mizuho; Nakamura, Hiroo; Matsushita, Izuru*; Muroga, Takeo*; Horiike, Hiroshi*
Fusion Engineering and Design, 85(7-9), p.1102 - 1105, 2010/12
This paper reports a measurement technique for surface waves on a liquid lithium jet for a Li target of the International Fusion Materials Irradiation Facility. The characteristic of the waves was successfully clarified by a contact-type liquid level detector. As a result, it was found that the wave distributions in the all jet velocity range up to 15 m/s were conformed each other in normalized form and Rayleigh distribution which is one of popular model to show irregular water wave.
Yamanishi, Toshihiko; Hayashi, Takumi; Kawamura, Yoshinori; Nakamura, Hirofumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Isobe, Kanetsugu; Suzuki, Takumi; Yamada, Masayuki
Fusion Engineering and Design, 85(7-9), p.1002 - 1006, 2010/12
The R&D for tritium technologies to a demonstration reactor (DEMO) is planned to be carried out in the Broader Approach (BA) program in Japan by JAEA with Japanese universities: (1) tritium analysis technology; (2) basic tritium safety research; and (3) tritium durability test. A multi-purpose RI facility is under construction at Rokkasho in Aomori to carry out the above R&D subjects. A preliminary safety study has been carried out for the amount of tritium released to the environment and for the radiation dose of workers. The main subjects of the R&D of tritium analysis are the technologies for real-time analysis for hydrogen isotopes, gas, liquid and solid. The materials of interest include F82H, SiC, ZrCo, solid and liquid advanced breeder and multipliers. In the tritium durability tests, organic materials and metals are studied for the radiation and the corrosion damage. A series of preliminary studies for the above subjects has been started.
Ishida, Shinichi; Barabaschi, P.*; Kamada, Yutaka; JT-60SA Team
Fusion Engineering and Design, 85(10-12), p.2070 - 2079, 2010/12
The mission of the JT-60SA project is to contribute to the early realization of fusion energy by supporting the exploitation of ITER and research towards DEMO by addressing key physics issues associated with these machines. The JT-60SA will be capable of confining break-even equivalent class high-temperature deuterium plasmas at a plasma current I of 5.5 MA and a major radius of 3 m lasting for a duration longer than the timescales characteristic of plasma processes, pursue full non-inductive steady-state operation with high plasma beta close to and exceeding no-wall ideal stability limits, and establish ITER-relevant high density plasma regimes well above the H-mode power threshold. Re-baselining of the project was completed in late 2008 which has been worked on since late 2007, where all the scientific missions are preserved with the newly designed machine to meet the cost objectives. The JT-60SA project made a large step forward towards its construction, which now foresees the first plasma in 2016. Construction of JT-60SA begins at Naka in Japan with launching the procurement of PF magnet, vacuum vessel and in-vessel components by Japan. In this year, the procurement of TF magnet, cryostat and power supply will be launched by Europe.
Batistoni, P.*; Angelone, M.*; Carconi, P.*; Fischer, U.*; Fleischer, K.*; Kondo, Keitaro; Klix, A.*; Kodeli, I.*; Leichtle, D.*; Petrizzi, L.*; et al.
Fusion Engineering and Design, 85(7-9), p.1675 - 1680, 2010/12
The EU is developing two test blanket modules (TBM), the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL), which will be tested in ITER. Here neutronics experiments with a mockup for HCLL TBM were carried out. Detail distributions of the tritium production rate inside the mockup were measured with various methods. A lithium diamond detector developed as a neutron monitor for fusion devices has also been used as a tritium detector. Activation reaction rates inside the mockup were also measured. These measured data agreed with calculation results buy using MCNP and FENDL-2.1 within 10%, which demonstrated that the prediction accuracy was high. Sensitivity and uncertainty analyses suggested that the uncertainty of the tritium production rate from the nuclear data uncertainty was small, usually below 2%.
Liu, C.; Tobita, Kenji
Fusion Engineering and Design, 85(7-9), p.979 - 982, 2010/12
Konno, Chikara; Ochiai, Kentaro; Takakura, Kosuke; Onishi, Seiki; Kondo, Keitaro; Wada, Masayuki*; Sato, Satoshi
Fusion Engineering and Design, 85(10-12), p.2054 - 2058, 2010/12
In the last ISFNT, we presented re-analyses of fusion neutronics benchmark experiments on beryllium at JAEA/FNS and reported that all the calculations with JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0 overestimated experimental data on low energy neutrons and that the calculation with JEFF-3.1 had a strange peak around 12 MeV. Here we investigate reasons for these problems. As a result, It was found out that the official ACE file MCJEFF3.1 of JEFF-3.1 had an inconsistency with the original JEFF-3.1, which caused the strange larger neutron peak around 12 MeV. We also find out that the calculated thermal neutron peak is probably too large. It is indicated that the coherent elastic scattering cross section data in the thermal neutron flux law data of beryllium metal are too large.
Hayashi, Takumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Nakamura, Hirofumi; Yamanishi, Toshihiko; Perevezentsev, A.*
Fusion Engineering and Design, 85(7-9), p.1386 - 1390, 2010/12
In order to establish effective ITER atmosphere detritiation system (DS), JAEA has investigated the performance and the durability at various incident/accident conditions, and supported to finalize the DS conceptual design through the ITER design review. The current DS at the safety important component has been discussed and mainly consists of catalytic reactors, wet scrubber column (SC) and blowers. The functional failure of the DS design with SC was evaluated using database of failure experiences of valves, controllers and components. Even in the tritium release into the biggest confinement sector of Tokamak gallery, it improved more than tow orders of magnitude comparing with that of original DS design using Molecular Sieve (MS) dryer beds in the 2001 design report. This improvement is achieved mainly by the minimization of valve operation like MS dryers and by the standardized module arrangement of DS with SC.
Isobe, Kanetsugu; Yamanishi, Toshihiko; Konishi, Satoshi*
Fusion Engineering and Design, 85(7-9), p.1012 - 1015, 2010/12
The measurement of tritium permeation behavior of NITE-SiC/SiC composites in both low tritium partial pressure and tritium diluted with hydrogen was carried out. Steady-state tritium permeation fluxes of NITE-SiC/SiC that has low permeability of hydrogen could be measured successfully by using tritium, even though low partial pressure (0.6 Pa). Steady-state tritium permeation fluxes were estimated to be 9.510[mol/msec]. In the experiment of tritium diluted with hydrogen, it was found that steady-states permeation fluxes was decreased with the increase of distillation rate, although the partial pressure of tritium in all condition was same (0.6 Pa).
Takeda, Nobukazu; Kakudate, Satoshi; Matsumoto, Yasuhiro; Kozaka, Hiroshi; Aburadani, Atsushi; Negishi, Yusuke; Nakahira, Masataka*; Tesini, A.*
Fusion Engineering and Design, 85(7-9), p.1190 - 1195, 2010/12
Several R&Ds for the ITER blanket remote handling system had been performed from the Engineering Design Activity phase until now and only several technical issues regarding the control system remained such as noise caused by slip ring, control of cable handling system, signal transmission through very long cable and radiation-hard amplifier. This study concentrates on these issues. As a conclusion, major issues for the control system have been solved and the ITER blanket remote handling system becomes further feasible.
Akiba, Masato; Enoeda, Mikio; Tanaka, Satoru*
Fusion Engineering and Design, 85(10-12), p.1766 - 1771, 2010/12
As the primary candidate of ITER Test Blanket Module (TBM) for the first day of ITER operation, development of Water Cooled Solid Breeder (WCSB) TBM has been performed toward the TBM milestones, which are necessary for acceptance of the TBM in ITER for testing from the first day of plasma operation. Regarding the liquid breeder blanket development, universities and NIFS are conducting the development. This paper overviews the recent achievements of the TBMs and DEMO blankets.
Ezato, Koichiro; Seki, Yohji; Tanigawa, Hisashi; Hirose, Takanori; Tsuru, Daigo; Nishi, Hiroshi; Dairaku, Masayuki; Yokoyama, Kenji; Suzuki, Satoshi; Enoeda, Mikio
Fusion Engineering and Design, 85(7-9), p.1255 - 1260, 2010/12
no abstracts in English
Tobita, Kenji; Uto, Hiroyasu; Liu, C.; Tanigawa, Hisashi; Tsuru, Daigo; Enoeda, Mikio; Yoshida, Toru; Asakura, Nobuyuki
Fusion Engineering and Design, 85(7-9), p.1342 - 1347, 2010/12
For a tokamak fusion DEMO reactor with the fusion output of 2.95 GW, neutronics and thermal design was carried out to find a blanket concept with reality. For the continuity with the Japanese ITER-TBM options, this study considered water-cooled blanket with solid breeding materials of Li ceramics and Be multipliers. A neutronics-heat coupled analysis determined an optimal arrangement of blanket interior under the constraints of the operating temperature of breeding materials and multipliers. When the cooling water is used under 23 MPa and 290-360 C, the overall tritium sufficiency is marginally satisfied although blankets with high neutron wall load ( = 5 MW/m) around the mid-plane do not meet the required local TBR. Based on the results, possible directions for further research are presented.
Sakurai, Shinji; Higashijima, Satoru; Hayashi, Takao; Shibama, Yusuke; Masuo, Hiroshige*; Ozaki, Hidetsugu; Sakasai, Akira; Shibanuma, Kiyoshi
Fusion Engineering and Design, 85(10-12), p.2187 - 2191, 2010/08
JT-60SA tokamak project has just started construction phase under both the Japanese domestic program and the Japan-EU international program "ITER Broader Approach". All of plasma facing components (PFC) shall be actively cooled due to high power long pulse plasma heating. Lower single null closed divertor with vertical target (VT) will be installed at the start of experiment phase. Each divertor module covers a 10-degree sector in toroidal direction. PFCs such as VTs, baffles and dome shall be assembled on a divertor cassette, which provides integrated coolant pipe connection to coolant headers in the VV. Static structural analysis for dead weight, coolant pressure and EM loads shows that displacement and stress of the divertor module are generally small but a part of support structure of PFC requires improvement.