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Journal Articles

Irradiation tests of a small-sized motor with radiation resistance

Nakamichi, Masaru; Ishitsuka, Etsuo; Shimakawa, Satoshi; Kan, Satoshi*

Fusion Engineering and Design, 83(7-9), p.1321 - 1325, 2008/12

 Times Cited Count:1 Percentile:88.61(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Engineering design and R&D of impurity influx monitor (divertor) for ITER

Ogawa, Hiroaki; Sugie, Tatsuo; Kasai, Satoshi*; Katsunuma, Atsushi*; Hara, Hirotsugu*; Takeyama, Norihide*; Kusama, Yoshinori

Fusion Engineering and Design, 83(10-12), p.1405 - 1409, 2008/12

 Times Cited Count:9 Percentile:42.28(Nuclear Science & Technology)

no abstracts in English

Journal Articles

A Proposal of ITER vacuum vessel fabrication specification and results of the full-scale partial mock-up test

Nakahira, Masataka; Takeda, Nobukazu; Kakudate, Satoshi; Onozuka, Masanori*

Fusion Engineering and Design, 83(10-12), p.1578 - 1582, 2008/12

 Times Cited Count:4 Percentile:67.02(Nuclear Science & Technology)

The structure and fabrication methods of the ITER vacuum vessel have been investigated and defined by the ITER international team. However, some of the current specifications are very difficult to be achieved from the manufacturing point of view and will lead to cost increase. This report summarizes the Japanese proposed specification of the VV mock-up describing differences between the ITER supplied design. A series of the fabrication and assembly procedures for the mock-up are presented in this report, together with candidates of welding configurations. Finally, the report summarizes the results of mock-up fabrication, including results of non-destructive examination of weld lines, obtained welding deformation and issues revealed from the fabrication experience.

Journal Articles

Experimental durability studies of electrolysis cell materials for a water detritiation system

Iwai, Yasunori; Hiroki, Akihiro; Yagi, Toshiaki*; Tamada, Masao; Yamanishi, Toshihiko

Fusion Engineering and Design, 83(10-12), p.1410 - 1413, 2008/12

 Times Cited Count:5 Percentile:60.95(Nuclear Science & Technology)

The radiation durability of the solid-polymer-electrolyte (SPE) water electrolyzer composed of the Water Detritiation System (WDS) was investigated. A series of $$gamma$$-ray and electron beam irradiation tests of Nafion N117 ion exchange membrane, a key polymer in a SPE electrolyzer, beyond ITER-WDS requirement (530kGy) indicated Nafion N117 has enough radiation durability up to 1600 kGy from the view points of mechanical strength and ion exchange capacity. To keep electrolysis function of a SPE cell up to 530 kGy, we suggest replacing Teflon with polyimide. A series of $$gamma$$-ray irradiation test of Kapton polyimide observed no serious damage in strength up to 1500 kGy. Concerning rubber material for O-ring seal, we observed that soaking VITON rubber keeps the constant value of tensile strength up to 1500 kGy. Moreover organic elution was not observed from a soak of VITON. From the viewpoint of stable strength and organic elution, VITON is a first candidate for rubber material.

Journal Articles

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

 Times Cited Count:8 Percentile:46.21(Nuclear Science & Technology)

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 $$gamma$$ 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 $$gamma$$ 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.

Journal Articles

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

 Times Cited Count:14 Percentile:27.98(Nuclear Science & Technology)

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$$^{2}$$. 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.

Journal Articles

Multi-scattering time-of-flight neutron spectrometer for deuterium to tritium fuel ratio measurement in fusion experimental reactors

Asai, Keisuke*; Yukawa, Kyohei*; Iguchi, Tetsuo*; Naoi, Norihiro*; Watanabe, Kenichi*; Kawarabayashi, Jun*; Yamauchi, Michinori*; Konno, Chikara

Fusion Engineering and Design, 83(10-12), p.1818 - 1821, 2008/12

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

The fuel ratio in a DT burning plasma can be derived from the intensity ratio of DD/DT neutrons, and detecting a trace of DD neutrons in the DT burning plasma is a key issue. A new type of neutron spectrometer is proposed to monitor the fuel ratio in the core of the ITER plasma. The system based on a conventional time-of-flight method consists of a water cell as a neutron scattering material and tens of scintillator pairs arranged around the first scintillator in a corn shape. We call it a multi-scattering time-of-flight neutron spectrometer (MS-TOF). A trial experiment was conducted for the prototype MS-TOF system with a DT neutron beam (20-mm diameter) at the Fusion Neutronics Source (FNS), Japan Atomic Energy Agency. The experimental results show that the DD and DT neutron peaks are clearly observed, and the experiment has successfully demonstrated the feasibility of the MS-TOF concept for detecting trace-DD neutrons within a DT neutron beam extracted from a DT burn plasma.

Journal Articles

Temperature dependence of blistering and deuterium retention in tungsten exposed to high-flux and low-energy deuterium plasma

Shu, Wataru; Isobe, Kanetsugu; Yamanishi, Toshihiko

Fusion Engineering and Design, 83(7-9), p.1044 - 1048, 2008/12

 Times Cited Count:37 Percentile:6.91(Nuclear Science & Technology)

At 315 K, only sparse low-dome blisters appeared even the fluence was increased to 10$$^{27}$$ D/m$$^{2}$$. At around 400 K, the blisters became much denser and the dome of blisters became a little higher. Peculiar change occurred around 500 K, where two kinds of blisters appeared. One is the large blisters with sizes of a few tens of microns and varying ratios of height against chord (up to 0.6), and the other is the small blisters with chords of less than a few microns and large ratio of height against chord (about 0.7). In high temperature region (higher than 600 K), the blisters became much sparser with the increasing temperature and disappeared at 1000 K. Deuterium retention showed the maximum around 500 K, corresponding to the appearance of two kinds of high-dome blisters.

Journal Articles

ITER vacuum vessel, in-vessel components and plasma facing materials

Ioki, Kimihiro*; Barabash, V.*; Cordier, J.*; Enoeda, Mikio; Federici, G.*; Kim, B. C.*; Mazul, I.*; Merola, M.*; Morimoto, Masaaki*; Nakahira, Masataka*; et al.

Fusion Engineering and Design, 83(7-9), p.787 - 794, 2008/12

 Times Cited Count:17 Percentile:22.34(Nuclear Science & Technology)

This paper presents recent results of ITER activities on Vacuum Vessel (VV), blanket, limiter, and divertor. Major results can be summarized as follows. (1) The VV design is being developed in more details considering manufacturing and assembly methods, and cost. Incorporating manufacturing studies being performed in cooperation with parties, the regular VV sector design has been nearly finalized. (2) The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. (3) The divertor activities have progressed with the aim of launching the procurement according to the ITER project schedule.

Journal Articles

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

 Times Cited Count:11 Percentile:35.75(Nuclear Science & Technology)

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.

Journal Articles

Structural material properties and dimensional stability of components in first wall components of a breeding blanket module

Hirose, Takanori; Enoeda, Mikio; Ogiwara, Hiroyuki; Tanigawa, Hiroyasu; Akiba, Masato

Fusion Engineering and Design, 83(7-9), p.1176 - 1180, 2008/12

 Times Cited Count:16 Percentile:23.83(Nuclear Science & Technology)

This paper summarize the fabrication process of the first wall structure and provides the material properties of the structural material F82H and dimensional stability of the components through whole of the process. A cold-rolling process introduced typical stretched rolling structure and ferrite/martensite dual phase structure, which lead reduction in strength. These anisotropic microstructural features were successfully recovered by optimized HIP process at 1373 K. As for dimensional stability of the components, a full-scale mockup has been developed with F82H tubes and plates. The HIPped mockup demonstrated good accordance with a design drawing. The dimensions of wall thickness and cooling channels were to size even after HIP. According to these results, the fabrication process does not degrade the material properties and demonstrates good dimensional accuracy and stability of the FW structure.

Journal Articles

D-T neutron streaming experiment simulating narrow gaps in ITER equatorial port

Ochiai, Kentaro; Sato, Satoshi; Wada, Masayuki*; Iida, Hiromasa; Takakura, Kosuke; Kutsukake, Chuzo; Tanaka, Shigeru; Abe, Yuichi; Konno, Chikara

Fusion Engineering and Design, 83(10-12), p.1725 - 1728, 2008/12

 Times Cited Count:1 Percentile:88.61(Nuclear Science & Technology)

Under the ITER/ITA task, we have conducted the neutron streaming experiment simulating narrow and deep gaps at boundaries between ITER vacuum vessel and equatorial port plugs. Micro fission chambers and some activation foils were utilized to measure fission rates and reaction rates to evaluate the relative fast and slow neutron fluences along the gap in the experimental assembly. The MCNP4C, TORT and Attila codes were used for the experimental analysis. From comparing our measurements and calculations, the following facts were found: (1) In case of a such narrow and deep gap structure, the calculation with MCNP, TORT and ATTILA codes and FENDL-2.1 is sufficient to predict fast neutron field inside the gap.: (2) Angular quadrature set of upward biased U315 and last collided source calculation on TORT and Attila were very important technique for accurate estimation of neutron transport.

Journal Articles

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

 Times Cited Count:6 Percentile:55.23(Nuclear Science & Technology)

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-H$$_{2}$$O 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.

Journal Articles

Recent progress in safety assessments of Japanese water-cooled solid breeder test blanket module

Tsuru, Daigo; Enoeda, Mikio; Akiba, Masato

Fusion Engineering and Design, 83(10-12), p.1747 - 1752, 2008/12

 Times Cited Count:12 Percentile:32.74(Nuclear Science & Technology)

This reports presents summary of safety assessment activities of the Japanese WCSB TBM. Nuclear analyses have been carried out to calculate neutron flux, tritium breeding ratio, nuclear heat, decay heat and induced activity of radioactive waste are calculated. For the purpose of evaluation of occupational radiolysis exposure, RI inventories in each component are estimated. FMEA has been carried out to identify the PIEs, i.e., the representative events that need safety evaluation. The PIEs are summarized into three groups, i.e., release of RI, pressurization and heatup. With respect to PIEs about release of RI, the maximum released RI is evaluated for three RI inventories, i.e., RI in VV (tritium and radio-activated dust), RI in purge gas (tritium) and RI in coolant (tritium and Active Corrosion Products (ACP)). With respect to PIEs about pressurization, the PIEs of pressurization of the compartment nearby the pipes of cooling system are evaluated by numerical analyses.

Journal Articles

Impact of reflected neutrons on accuracy of tritium production rate prediction in blanket mock-ups for fusion reactors

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

Fusion Engineering and Design, 83(7-9), p.1304 - 1308, 2008/12

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

In the experiment with a reflector and TPR around the boundary between the rear parts of the breeder layer and the beryllium, TPRs calculated by MCNP overestimated the experimental results by more than 10% for our breeding blanket neutronics experiments with DT neutrons carried out at JAEA FNS. We have pointed out that the nuclear data have some problems on back scattered neutrons. Thus we modified the backward part of angular distributions in FENDL-2.1 in order to investigate the possibility of improvement of the overestimation in this study. By decreasing backward part of angular distributions in the elastic scattering of $$^{56}$$Fe by uniformly 50% from the original data for the incident neutron less than 0.11 MeV, the overestimation was improved for the enriched breeding blanket experiment with a reflector. Also by decreasing backward part of $$^{9}$$Be by uniformly 20% for the incident neutron of 0.62-14.94 MeV, the overestimation was improved on the TPR around the boundary.

Journal Articles

Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, Kiyoyuki; Kasada, Ryuta*; Wakai, Eiichi; Serizawa, Hisashi*; Kawahito, Yosuke*; Jitsukawa, Shiro; Kimura, Akihiko*; Kono, Yutaka*; et al.

Fusion Engineering and Design, 83(10-12), p.1471 - 1476, 2008/12

 Times Cited Count:73 Percentile:1.61(Nuclear Science & Technology)

Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. F82H, which were developed and studied in Japan, was designed with an emphasis on high temperature properties and weldability. The database on F82H properties is currently the most extensive available among the existing RAFMs. The objective of this paper is to review the R&D status of F82H and to identify the key technical issues for the fabrication of an ITER Test Blanket Module (TBM) suggested by recent achievements in Japan.

Journal Articles

Critical heat flux experiments using a screw tube under DEMO divertor-relevant cooling conditions

Ezato, Koichiro; Suzuki, Satoshi; Dairaku, Masayuki; Akiba, Masato

Fusion Engineering and Design, 83(7-9), p.1097 - 1101, 2008/12

 Times Cited Count:6 Percentile:55.23(Nuclear Science & Technology)

As part of development of Plasma-Facing Components (PFCs) for fusion machines, JAEA has been developing high performance cooling tubes with pressurized water flow. Along this line, a cooling tube with a helical triangular fin on its inner surface has been proposed recently for application to a fusion DEMO to enhance heat removal. Since the fin can be machined by a simple mechanical threading, this tube is called as a screw tube. Divertor cooling conditions in the DEMO design in JAEA are envisaged at the pressure of 4 MPa and the outlet temperature of 200$$^{circ}$$C to improve thermal efficiency of power generation. In the this study, effect of subcooling on critical heat flux (CHF) of the screw tube has been investigated under DEMO-relevant condition with the local pressure of 4 MPa and the inlet coolant temperature up to 180$$^{circ}$$C. A test sample is the screw tube made of pure Cu instead of F82H, a candidate material of the DEMO divertor. The results show that the ICHF values of the screw tube remains more than double values of the smooth at the inlet coolant temperature of 180$$^{circ}$$C, although temperature rise of the cooling water with 140 K leads to reduction of ICHF by almost half compared with those values at room temperature.

Journal Articles

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

 Times Cited Count:23 Percentile:15.37(Nuclear Science & Technology)

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 $$^{3}$$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.

Journal Articles

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

 Times Cited Count:6 Percentile:55.23(Nuclear Science & Technology)

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 $$^{9}$$Be(n,2n+2$$alpha$$) 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 $$alpha$$-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 $$alpha$$-particle emission DDX at forward emission angles.

Journal Articles

Hydrogen isotope separation capability of low temperature mordenite column for gas chromatograph

Kawamura, Yoshinori; Onishi, Yoshihiro*; Okuno, Kenji*; Yamanishi, Toshihiko

Fusion Engineering and Design, 83(10-12), p.1384 - 1387, 2008/12

 Times Cited Count:11 Percentile:35.75(Nuclear Science & Technology)

A gas chromatograph using a cryogenic separation column is one of the methods for hydrogen isotope analysis. However, use of liquid nitrogen is a 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 is one of the solutions for these weak points. Mordenite (MOR) is a kind of the synthesis 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 H$$_{2}$$ and D$$_{2}$$ 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 synthesis zeolite that can separate hydrogen isotope mixture at comparatively high temperature.

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