Status of development of functional materials with perspective on beyond-ITER

Shikama, Tatsuo*; Knitter, R.*; Konys, J.*; Muroga, Takeo*; Tsuchiya, Kunihiko; Mslang, A.*; Kawamura, Hiroshi; Nagata, Shinji*

Fusion Engineering and Design, 83(7-9), p.976 - 982, 2008/12

https://doi.org/10.1016/j.fusengdes.2008.07.034

Functional materials should play an important role not only in ITER but also in fusion machines beyond ITER, though it is occasionally claimed that future fusion plants should have much more simple structures and they should be free from auxiliary systems such as diagnostics. Studies on test blanket modules (TBM) clearly show the importance of functional materials there. The paper will review the present status of studies and developments of functional ceramics for nuclear fusion applications, with a perspective on their application in electric-power generating fusion power plant, namely DEMO.

Safe handling experience of a tritium storage bed

Hayashi, Takumi; Suzuki, Takumi; Yamada, Masayuki; Shu, Wataru; Yamanishi, Toshihiko

Fusion Engineering and Design, 83(10-12), p.1429 - 1432, 2008/12

https://doi.org/10.1016/j.fusengdes.2008.07.010

In ITER facility, about 3 kg of tritium will be stored in more than 30 ZrCo hydride beds, as a reference design. The safe design and operation of tritium storage beds will be one of the most important points to enhance total safety of the facility. In the Tritium Process Lab. in Japan Atomic Energy Agency, many tritium storage beds with ZrCo have been used with/without self-accountancy measure, and the safe handling experiences have been accumulated for almost 20 years. From these experiences, the key issues to be considered for the safety design are the effect of tritium decay, such as decay heat transfer and He behavior with the normal protection of over temperature, over pressure and leak for a metal-hydride bed. Concerning the safety operation, the key issues are the procedure of hydrogenation-dehydrogenation cycle under the requirements of the storage system and the emergency performances, such as a rapid hydrogen recovery and loss of normal cooling function.

Development of a virtual reality simulator for the ITER blanket remote handling system

Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Shibanuma, Kiyoshi; Tesini, A.*

Fusion Engineering and Design, 83(10-12), p.1837 - 1840, 2008/12

https://doi.org/10.1016/j.fusengdes.2008.05.042

The maintenance activity in the ITER has to be performed remotely because 14 MeV neutron caused by fusion reaction induces activation of structural material and emission of ray. In general, it is one of the most critical issues to avoid collision between the remote maintenance system and in-vessel components. Therefore, the visual information in the vacuum vessel is required strongly to understand arrangement of these devices and components. However, there is a limitation of arrangement of viewing cameras in the vessel because of high intensity of ray. Furthermore, visibility of the interested area such as the contacting part is frequently disturbed by the devices and components, thus it is difficult to recognize relative position between the devices and components only by visual information even if enough cameras and lights are equipped. From these reasons, the simulator to recognize the positions of each devices and components is indispensable for remote handling systems in fusion reactors. The authors have been developed a simulator for the remote maintenance system of the ITER blanket using a general 3D robot simulation software "ENVISION". The simulator is connected to the control system of the manipulator which was developed as a part of the blanket maintenance system in the EDA and can reconstruct the positions of the manipulator and the blanket module using the position data of the motors through the LAN. In addition, it can provide virtual visual information, such as the connecting operation behind the blanket module with making the module transparent on the screen. It can be used also for checking the maintenance sequence before the actual operation.

Recent results of R&D activities on tritium technologies for ITER and fusion reactors at TPL of JAEA

Yamanishi, Toshihiko; Hayashi, Takumi; Shu, Wataru; Kawamura, Yoshinori; Nakamura, Hirofumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Isobe, Kanetsugu; Arita, Tadaaki; Hoshi, Shuichi; et al.

Fusion Engineering and Design, 83(10-12), p.1359 - 1363, 2008/12

https://doi.org/10.1016/j.fusengdes.2008.08.010

At TPL (Tritium Process Laboratory) of JAEA, ITER relevant tritium technologies have been studied. The design studies of Air Detritiation System 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 research 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 not only for ITER but also for fusion DEMO plants has also been carried out at TPL of JAEA. The main research activities in this field are the tritium behavior in a confinement and its barrier materials; monitoring; accountancy; detritiation and decontamination etc. In this paper, the results of above recent activities at TPL of JAEA are summarized from viewpoint of ITER relevant and future fusion DEMO reactors.

Progress of R&D and design of blanket remote handling equipment for ITER

Kakudate, Satoshi; Takeda, Nobukazu; Nakahira, Masataka; Matsumoto, Yasuhiro; Shibanuma, Kiyoshi; Tesini, A.*

Fusion Engineering and Design, 83(10-12), p.1850 - 1855, 2008/12

https://doi.org/10.1016/j.fusengdes.2008.08.029

The design of in-vessel transporter (IVT) including vehicle manipulator has been updated according to the design changes such as blanket segmentation and structure, taking account of the interface between modules and vehicle manipulator. In particular, the updated design of the vehicle manipulator and rail has been carried out in order to avoid the interference between modules and vehicle manipulator. According to the updated design, the vehicle manipulator has been reduced by about 30%, compared with the reference design. In parallel with design activities, the R&D to clarify the specifications of the IVT design in detail is also performed, i.e., simulation system to provide the visual information during maintenance, dry lubricant to prevent the lubricant oil from spreading in the VV. The rail connection and cable handling in the transfer cask, which are critical issues for IVT system, are under preparation of the demonstration tests to finalize the design of the IVT system. Connection of the rail joint and cable handling test facilities are planned and under fabrication now. These test facility will be installed by the end of March 2008, and the performance tests will be carried out from April 2008.

Evolution of ITER tritium confinement strategy and adaptation to Cadarache site conditions and French regulatory requirements

Murdoch, D.*; Glugla, M.*; Hayashi, Takumi; Perevesentsev, A.*; Stephan, Y.*; Taylor, C.*

Fusion Engineering and Design, 83(10-12), p.1355 - 1358, 2008/12

https://doi.org/10.1016/j.fusengdes.2008.04.006

Latest design of liquid lithium target in IFMIF

Nakamura, Hiroo; Agostini, P.*; Ara, Kuniaki; Cevolani, S.*; Chida, Teruo*; Ciotti, M.*; Fukada, Satoshi*; Furuya, Kazuyuki*; Garin, P.*; Gessii, A.*; et al.

Fusion Engineering and Design, 83(7-9), p.1007 - 1014, 2008/12

https://doi.org/10.1016/j.fusengdes.2008.06.014

This paper describes the latest design of liquid lithium target system in IFMIF. Design requirement of the Li target is to provide a stable Li jet with a speed of 20 m/s to handle an averaged heat flux of 1 GW/m. A double reducer nozzle and a concaved flow are applied to the target design. On Li purification, a cold trap and two kinds of hot trap are applied to control impurities below permissible levels. Nitrogen concentration shall be controlled below 10 wppm by one of the hot trap. Tritium concentration shall be controlled below 1 wppm by an yttrium hot trap. To maintain reliable continuous operation, various diagnostics are attached to the target assembly. Among the target assembly, a back-plate made of RAFM is located in the most severe region of neutron irradiation (50 dpa/y). Therefore, two design options of replaceable back wall and their remote handling systems are under investigation.

Analyses of fusion integral benchmark experiments at JAEA/FNS with FENDL-2.1 and other recent nuclear data libraries

Konno, Chikara; Ochiai, Kentaro; Wada, Masayuki*; Sato, Satoshi

Fusion Engineering and Design, 83(10-12), p.1774 - 1781, 2008/12

https://doi.org/10.1016/j.fusengdes.2008.05.010

Many integral benchmark experiments with DT neutrons have been carried out for nuclear data verification for fusion nuclear design at JAEA FNS; simple benchmark experiments, time-of-flight, breeding blanket experiments. For a few years several nuclear data libraries have been newly released; JENDL-3.3, FENDL-2.1, JEFF-3.1 and ENDF/B-VII.0. It is essential to verify these libraries through analyses of integral benchmark experiments. Thus we carried out a series of analyses for the benchmark experiments at JAEA FNS with FENDL-2.1, JENDL-3.3, JEFF-3.1 and ENDF/B-VII.0. The Monte Carlo code MCNP-4C was used for this analysis. Calculated results were compared with measured ones. They were also compared each other. The calculated results with FENDL-2.1, JENDL-3.3, JEFF-3.1 and ENDF/B-VII.0 were almost the same and represented the measured ones except for some experiments.

Heatup event analyses of the water cooled solid breeder test blanket module

Tsuru, Daigo; Enoeda, Mikio; Akiba, Masato

Fusion Engineering and Design, 83(7-9), p.1238 - 1243, 2008/12

https://doi.org/10.1016/j.fusengdes.2008.07.008

One-dimensional analyses are carried out for the three representative event sequences about heatup of WCSB TBM. For the event sequence of loss of cooling of the TBM during plasma operation, the requirement to the TBM design becomes apparent that no cooling pipe rupture can be guaranteed under the predicted highest temperature condition to prevent the activation of the Be-HO chemical reaction under the high temperature condition. For the event sequence of ingress of coolant into the TBM during plasma operation, the requirement for the cooling system of TBM becomes apparent that the cooling system for the TBM should be designed to continue its operation against partial loss of coolant due to the ingress of coolant. For the loss of off-site power after the ingress of coolant, it is confirmed that the event converges.

Verification of KERMA factor for beryllium at neutron energy of 14.2 MeV based on charged-particle measurement

Kondo, Keitaro; Ochiai, Kentaro; Murata, Isao*; Konno, Chikara

Fusion Engineering and Design, 83(10-12), p.1674 - 1677, 2008/12

https://doi.org/10.1016/j.fusengdes.2008.06.008

In previous direct measurements of nuclear heating for beryllium induced with DT-neutrons, it was pointed out that the calculation with JENDL-3.2 underestimated the measured one by 25 %. However, reasons of this large discrepancy have not been understood clearly. In order to reveal the reason of this discrepancy, we examined KERMA factors for beryllium deduced with three latest nuclear data libraries: JENDL-3.3, ENDF/B-VII.0 and JEFF-3.1. As a result, the partial KERMA factors for Be(n,2n+2) reaction channel at incident neutron energy of 14.2 MeV deduced from JENDL-3.3 was significantly smaller than that deduced from the other libraries. These partial KERMA factors were compared with a new partial KERMA factor calculated based on our experimental model from our recent measurement of the -particle emission double-differential cross-section for beryllium. The partial KERMA factor from JENDL-3.3 was smaller by 20 % than our experiment-based one. The reason of the discrepancy in the previous nuclear heating measurement comes from smaller partial KERMA factor for beryllium in JENDL-3.3, which is caused by significant underestimation of higher energy part of -particle emission DDX at forward emission angles.