Feeder components and instrumentation for the JT-60SA magnet system

Yoshida, Kiyoshi; Kizu, Kaname; Murakami, Haruyuki; Kamiya, Koji; Honda, Atsushi; Onishi, Yoshihiro; Furukawa, Masato; Asakawa, Shuji; Kuramochi, Masaya; Kurihara, Kenichi

Fusion Engineering and Design, 88(9-10), p.1499 - 1504, 2013/10

https://doi.org/10.1016/j.fusengdes.2012.12.010

The modifying of the JT-60U magnet system to the superconducting coils (JT-60SA) is progressing as a satellite facility for ITER by both parties of Japanese government and European commission (EU) in the Broader Approach agreement. The magnet system for JT-60SA consists of 18 Toroidal Field (TF) coils, a Central Solenoid (CS) with 4 modules, and 6 Equilibrium Field (EF) coils. The manufacturing of the JT-60SA magnet system is in progress in EU and Japan. The JT-60SA superconducting magnet system generates an average heat load of 3.2 kW at 4 K to the cryoplant, from nuclear and thermal radiation, conduction and electromagnetic heating, and requires current supplies 20 kA for 4 CS modules and 6 EF coils, 25.7 kA to 18 TF coils. The helium flow to remove this heat, consisting of supercritical helium at pressures up to 0.5 MPa and temperature between 4.4-4.8 K, is distributed to the coils and structures through the valve box (VB) from the cryoline connecting to the auxiliary cold box located outside the torus hall. The feeders also contain the electrical supplies from the current lead transitions to room temperature to the coil. The feeder components consist of the in-cryostat feeders with flexible parts to allow coil operational displacements from the connection pipes out of the cryostat, including S-bend conductor to allow differential thermal contraction and the coil terminal boxes (CTBs) with HIS current leads. A measurement and control system is required to monitor and control these coils and feeders for safety and optimal operational availability. For each coil, both current and supercritical helium are supplied from external systems and are controlled from a central system as part of the regular operation with plasma pulses. Quench detection instruments for superconducting coils, feeders and HTS current leads are provided as a separate, stand alone system.

Corrosion of austenitic steel in leakage lithium

Furukawa, Tomohiro; Hirakawa, Yasushi; Kato, Shoichi

Fusion Engineering and Design, 88(9-10), p.2502 - 2505, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.04.036

Lithium, which is used as the neutron source of the IFMIF, reacts with oxygen, nitrogen and moisture in atmosphere in case of the leakage accident. In this study, fundamental corrosion test was performed in order to obtain the corrosion behavior of austenitic stainless steel in the estimated lithium compounds. In the experiment, the lithium compounds were filled with the steel into the Tammann pipe made from AlO, and heated up to 850C. After the test, the specimen was cleaned by alcohol, and then the weight loss measurement and metallurgical examination were performed. Intense corrosion was observed at the environmental conditions containing lithium peroxide. No corrosion was observed in LiO environment. According to the consideration based on thermodynamics, LiO cannot oxidize iron and lithium is reducing agent. Slight corrosion was observed in LiOH and LiN environments.

Oxidation property and homogenization treatment of plasma sintered beryllides

Kim, Jae-Hwan; Nakamichi, Masaru

Fusion Engineering and Design, 88(9-10), p.2215 - 2218, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.03.052

Recent results on tritium technology in JAEA under BA program

Yamanishi, Toshihiko; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu

Fusion Engineering and Design, 88(9-10), p.2272 - 2275, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.04.046

The multi-purpose RI equipment has been constructed at Rokkasho site in DEMO R&D building until 2011. The equipment is the first and unique facility in Japan, where tritium, RI species, and beryllium can simultaneously be used. The amounts of tritium used and stored are 3.7 TBq per day and 7.4 TBq, respectively. The material of the column of the micro gas chromatograph has been studied. The calorimeter has also been studied as a possible tritium measurement method. A set of basic data on the interaction between materials and tritium has been measured especially for pure Fe. As for the tritium behavior in the blanket materials, the tritium release after neutron irradiation was studied. As a study for the tritium durability, the endurance of ion exchange membrane has been tested by using high concentration tritium water. The data of tritium water were well consistent with those obtained by irradiation.

Benchmark test of TRIPOLI-4 code through simple model calculation and analysis of fusion neutronics experiments at JAEA/FNS

Ota, Masayuki; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Konno, Chikara

Fusion Engineering and Design, 88(9-10), p.2160 - 2163, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.02.090

Three-dimensional Monte Carlo transport code is widely used in analyses of radiation protection and shielding, nuclear criticality safety, fission and fusion reactor design and nuclear instrumentation. The MCNP code is worldwide used as the standard code. The TRIPOLI code is adopted as one of ITER neutronics codes, but it is not always used worldwide, particularly in Japan. In this study we examined the basic performance of TRIPOLI through simple model calculations and analysis of iron neutronics experiments with DT neutrons at the Fusion Neutronics Source facility in Japan Atomic Energy Agency. Almost results by TRIPOLI are consistent with ones by MCNP. However, there is a difference between TRIPOLI and MCNP results in some calculations caused mainly by treatment with inelastic scattering data.

Progress of the ITER NBI acceleration grid power supply reference design

Toigo, V.*; Zanotto, L.*; Bigi, M.*; Decamps, H.*; Ferro, A.*; Gaio, E.*; Gutirrez, D.*; Tsuchida, Kazuki; Watanabe, Kazuhiro

Fusion Engineering and Design, 88(6-8), p.956 - 959, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.01.013

This paper reports the progress in the reference design of the Acceleration Grid Power Supply (AGPS) of the ITER Neutral Beam Injector (NBI). The design of the AGPS is very challenging, as it shall be rated to provide about 55 MW at 1 MV dc in quasi steady-state conditions; moreover, the procurement of the system is shared between the European Domestic Agency (F4E) and the Japanese Domestic Agency (JADA), resulting in additional design complication due to the need of a common definition of the interface parameters. A critical revision of the main design choices is presented also in light of the definition of some key interface parameters between the two AGPS subsystems. Moreover, the verification of the fulfillment of the requirements in any operational conditions taking into account the tolerance of the different parameters is also reported and discussed.

Welding technology R&D on port joint of JT-60SA vacuum vessel

Shibama, Yusuke; Masaki, Kei; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira; Onawa, Toshio*; Araki, Takao*; Asano, Shiro*

Fusion Engineering and Design, 88(9-10), p.1916 - 1919, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.01.100

This presentation focuses on the welding technology R&D between the JT-60SA vacuum vessel and the ports. The vacuum vessel is designed to allow port bore penetration to access the vessel inside for plasma diagnostics, and so on. There are various types of 73 ports and these are categorized by their locations; the upper/lower vertical, the upper/lower oblique, and the horizontal. Ports are onsite-welded onto the VV port stub after the assembly of the VV. This assembly sequence involves the out-vessel components such as VV thermal shield and toroidal field magnets, so that these ports welding are accessed from the inside of the vessel and limited by the internal port wall. The one of the most difficult ports are the upper vertical port with corner radius of 50 mm under narrow space, and it is necessary to clarify mobility of the weld torch head. The port weldability is discussed with the mock-up trial, which consists of the partial test pieces of the product size. The TIG welding manipulator, optimized for this R&D, is prepared by its operational simulation and examined not to interfere with the internal port wall.

Rail deployment operation test for ITER blanket handling system with positioning misalignment

Takeda, Nobukazu; Aburadani, Atsushi; Tanigawa, Hisashi; Shigematsu, Soichiro; Kozaka, Hiroshi; Murakami, Shin; Kakudate, Satoshi; Nakahira, Masataka; Tesini, A.*

Fusion Engineering and Design, 88(9-10), p.2186 - 2189, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.02.086

R&D for rail deployment equipment was performed for the ITER blanket remote handling system. The target torque for the automatic operation was investigated. The result shows that the 20% of the rated torque is adequate as the torque limitation for the automatic operation. A schedule for the procurement of the blanket remote handling system, which will be delivered to the ITER in 2020, was also shown.

Novel granulation process of beryllide as advanced neutron multipliers

Nakamichi, Masaru; Kim, Jae-Hwan; Yonehara, Kazuo

Fusion Engineering and Design, 88(6-8), p.611 - 615, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.02.077

Advanced neutron multipliers with lower swelling and higher stability at high temperature are desired in pebble-bed blankets, which would have a big impact on the design of a DEMO reactor, especially on the blanket operating temperature. Development of advanced neutron multipliers has been started by Japan and EU in the DEMO R&D as a part of the Broader Approach (BA) activities. Beryllium intermetallic compounds (beryllides) such as BeTi are one of the most promising advanced neutron multipliers. In order to fabricate the beryllide pebbles, beryllide with shapes of block and/or rod is necessary when a melting granulation process is applied such as a rotating electrode method. However, beryllide is too brittle for the fabrication of blocks or rods by these methods. A plasma sintering method has been proposed as new technique which uses a non conventional consolidation process, because this method is simple, and is easy to control. It was clarified that the beryllide could be simultaneously synthesized and jointed by the plasma sintering method in the insert material region between two beryllide blocks, with no variation of the phase and hardness. Beryllide rod of BeTi with 10 mm in diameter and 60 mm in length has been successfully fabricated by the plasma sintering method. Using this plasma-sintered beryllide rod, prototype pebble of beryllide was performed by a rotating electrode method. The prototype pebbles of BeTi with 1 mm in average diameter were successfully fabricated. The present paper describes novel granulation process of beryllide using these methods including fabrication and granulation techniques.

Structural analyses of HV bushing for ITER heating NB system

Tobari, Hiroyuki; Inoue, Takashi; Taniguchi, Masaki; Kashiwagi, Mieko; Umeda, Naotaka; Dairaku, Masayuki; Yamanaka, Haruhiko; Watanabe, Kazuhiro; Sakamoto, Keishi; Kuriyama, Masaaki*; et al.

Fusion Engineering and Design, 88(6-8), p.975 - 979, 2013/10

https://doi.org/10.1016/j.fusengdes.2013.02.020

The HV bushing, one of the ITER NB components, which is to be procured by JADA, is a multi-conductor feed through composed of five-stage double-layered insulator columns with large brazed ceramic ring and fiber reinforced plastic (FRP) ring. The HV bushing is a bulk head between insulation gas at 0.6 MPa and vacuum. The FRP ring is required to sustain the pressure load, seismic load and dead weight. Brazing area of the ceramic ring with Kovar is required to maintain vacuum leak tightness and pressure tightness against the air filled at 0.6 MPa. To design the HV bushing satisfying the safety factor of 3.5, mechanical analyses were carried out. As for the FRP ring, it was confirmed that isotropic fiber cloth FRP rings should be used for sufficient strength against shear stress. Also, shape and fixation area of the Kovar sleeve were modified to lower the stress at the joint area. As a result, a design of the insulator for the HV bushing was established satisfying the requirement.