Development of an iron(II) complex exhibiting thermal- and photoinduced double proton-transfer-coupled spin transition in a short hydrogen bond

Nakanishi, Takumi*; Hori, Yuta*; Shigeta, Yasuteru*; Sato, Hiroyasu*; Kiyanagi, Ryoji; Munakata, Koji*; Ohara, Takashi; Okazawa, Atsushi*; Shimada, Rintaro*; Sakamoto, Akira*; et al.

Journal of the American Chemical Society, 145(35), p.19177 - 19181, 2023/08

https://doi.org/10.1021/jacs.3c06323

R&D status on water cooled ceramic breeder blanket technology

Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.

Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10

https://doi.org/10.1016/j.fusengdes.2014.01.035

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. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed 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 simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.

High-resolution Compton cameras based on Si/CdTe double-sided strip detectors

Odaka, Hirokazu*; Ichinohe, Yuto*; Takeda, Shinichiro*; Fukuyama, Taro*; Hagino, Koichi*; Saito, Shinya*; Sato, Tamotsu*; Sato, Goro*; Watanabe, Shin*; Kokubun, Motohide*; et al.

Nuclear Instruments and Methods in Physics Research A, 695, p.179 - 183, 2012/12

https://doi.org/10.1016/j.nima.2011.12.061

We have developed a new Si/CdTe semiconductor double-sided strip detector (DSD) Compton camera. The camera consists of a 500-m-thick Si-DSD and four layers of 750-m-thick CdTe-DSDs all of which have common electrode configuration segmented into 128 strips on each side with pitches of 250m. In order to realize high angular resolution and to reduce size of the detector system, a stack of DSDs with short stack pitches of 4 mm is utilized to make the camera. Taking advantage of the excellent energy and position resolutions of the semiconductor devices, the camera achieves high angular resolutions of 4.5 degrees at 356 keV and 3.5 degrees at 662 keV. To obtain such high resolutions together with an acceptable detection efficiency, we demonstrate data reduction methods including energy calibration using Compton scattering continuum and depth sensing in the CdTe-DSD. We also discuss imaging capability of the camera and show simultaneous multi-energy imaging.

Development of a two-dimensional nuclear-thermal-coupled analysis code for conceptual blanket design of fusion reactors

Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Sato, Satoshi; Seki, Yohji; Takase, Haruhiko

Fusion Engineering and Design, 86(9-11), p.2378 - 2381, 2011/10

https://doi.org/10.1016/j.fusengdes.2011.03.069

For DEMO reactor blanket design, a two-dimensional (2-D) nuclear-thermal-coupled analysis code, DOHEAT, has been developed. In DOHEAT, the neutron flux is calculated by a 2-D transport code, DOT3.5, with the nuclear data library, FUSION-40, and the nuclear heating rate and the local TBR profile of blanket are calculated using the 2-D neutronics calculation code, APPLE-3. Use of the code has showed outstanding usefulness in the blanket design where detailed evaluation of neutron flux, nuclear heating rate, tritium breeding ratio (TBR) and the temperature of materials is required for various blanket concepts and trial-and-error-basis iteration is sometimes necessary. DOHEAT can replace the actual blanket structure by a more realistic model including cooling tubes, multipliers and breeders. A validation calculation indicates that DOHEAT provides reasonable results on the temperature profile.

Compact DEMO, SlimCS; Design progress and issues

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Kawashima, Hisato; Kurita, Genichi; Tanigawa, Hiroyasu; Nakamura, Hirofumi; Honda, Mitsuru; Saito, Ai*; Sato, Satoshi; et al.

Nuclear Fusion, 49(7), p.075029_1 - 075029_10, 2009/07

https://doi.org/10.1088/0029-5515/49/7/075029

Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with the previous advanced concept with supercritical water, the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m or lower, which can be a critical constraint for determining a handling power of DEMO (a combination of alpha heating power and external input power for current drive).

SlimCS; Compact low aspect ratio DEMO reactor with reduced-size central solenoid

Tobita, Kenji; Nishio, Satoshi; Sato, Masayasu; Sakurai, Shinji; Hayashi, Takao; Shibama, Yusuke; Isono, Takaaki; Enoeda, Mikio; Nakamura, Hirofumi; Sato, Satoshi; et al.

Nuclear Fusion, 47(8), p.892 - 899, 2007/08

https://doi.org/10.1088/0029-5515/47/8/022

The concept for a compact DEMO reactor named "SlimCS" is presented. Distinctive features of the concept is low aspect ratio ( = 2.6) and use of a reduced-size center solenoid (CS) which has a function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field (TF) coil system which contributes to reducing the weight and construction cost of the reactor. SlimCS is as compact as advanced commercial reactor designs such as ARIES-RS and produces 1 GWe in spite of moderate requirements for plasma parameters. Merits of low-, i.e. vertical stability for high elongation and high beta limit are responsible for such reasonable physics requirements.

Design study of fusion DEMO plant at JAERI

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Sato, Masayasu; Isono, Takaaki; Sakurai, Shinji; Nakamura, Hirofumi; Sato, Satoshi; Suzuki, Satoshi; Ando, Masami; et al.

Fusion Engineering and Design, 81(8-14), p.1151 - 1158, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.08.058

no abstracts in English

Development of risk management approach for underground facility construction in granite

Shimono, Masato*; Motoshima, Takayuki*; Ijiri, Yuji*; Otsu, Hiroyasu*; Sakai, Ryosuke*; Sakai, Kazuo*; Sato, Toshinori; Mikake, Shinichiro

no journal, , 

It is considered that the uncertainty of construction cost due to geological heterogeneity is quite large. Therefore, it is crucial to make a long-term consistent budget plan for a construction period of tens years and to have engineering strategies for reducing a total budget. In this study, a risk management approach for construction in a heterogeneous fractured rock is developed. In this approach, methodologies of calculating construction costs for excavating drifts and taking countermeasures at high-hydrostatic pressured fracture zones in a heterogeneous rock are established. The cost for excavating drift is calculated from the heterogeneity of rock strength estimated from a geostatistical modeling approach. The cost for taking countermeasures is calculated from the heterogeneity of large fracture zone estimated from a probabilistic discrete fracture network modeling approach. In addition, a risk for a construction is assessed by a risk cost curve commonly used in the financial business.

Development of core damage evaluation technology (level 2 psa) for fast reactors, 3; Development of analytical code for ex-vessel phenomena

Seino, Hiroshi; Ohno, Shuji; Ishikawa, Hiroyasu; Nishimura, Masahiro; Sato, Isamu; Miyahara, Shinya

no journal, , 

no abstracts in English

Development of core damage evaluation technology (level 2 PSA) for fast reactors, 5; Progress of R&D in FY2007

Nakai, Ryodai; Kurisaka, Kenichi; Sato, Ikken; Tobita, Yoshiharu; Kamiyama, Kenji; Yamano, Hidemasa; Miyahara, Shinya; Ohno, Shuji; Seino, Hiroshi; Ishikawa, Hiroyasu; et al.

no journal, , 

To develop a core damage evaluation technology (level-2 PSA) in sodium-cooled fast reactors, a new analysis method is developed for core-material relocation phase and internal containment vessel event. This study also develop technical basis necessary for the level-2 PSA.