Development of in-vessel components of the microfission chamber for ITER

Ishikawa, Masao; Kondoh, Takashi; Okawa, Kiyofumi*; Fujita, Kyoichi*; Yamauchi, Michinori*; Hayakawa, Atsuro*; Nishitani, Takeo; Kusama, Yoshinori

Review of Scientific Instruments, 81(10), p.10D308_1 - 10D308_3, 2010/10

https://doi.org/10.1063/1.3485080

Microfission chambers (MFCs) will provide total neutron source strength in ITER. The MFC is a pencil-sized gas counter containing the fissile material, U. The MFCs will be installed behind blanket modules in the vacuum vessel (VV). Double coaxial mineral insulated (MI) cables will carry signals from the MFCs to the upper port. Though the MI cables will be installed at a factory of the vacuum vessel or ITER assembly hall, detectors with U will be installed to the vacuum vessel at the tokamak pit. Then, the MI cable should be connected in the vacuum vessel. In this work, the connection of the MI cable with the MFC was conceptually designed. The MI cable should be also installed with small curvature radius (R) of 100 200 mm to avoid the VV structure and other diagnostics. So, the vending test of the MI cable was conducted. As a result, damages leak, electrical disconnection and the change in insulation resistance have not been observed at R = 100 mm.

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

https://doi.org/10.1016/j.fusengdes.2008.08.002

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.

Monte Carlo analyses of blanket neutronics experiments at FNS with latest nuclear data libraries

Sato, Satoshi; Ochiai, Kentaro; Wada, Masayuki*; Yamauchi, Michinori; Iida, Hiromasa; Nishitani, Takeo; Konno, Chikara

Proceedings of International Conference on Nuclear Data for Science and Technology (ND 2007), Vol.2, p.995 - 998, 2008/05

https://doi.org/10.1051/ndata:07403

no abstracts in English

Thin slit streaming experiment for ITER by using D-T neutron source

Ochiai, Kentaro; Sato, Satoshi; Wada, Masayuki*; Kubota, Naoyoshi; Kondo, Keitaro; Yamauchi, Michinori; Abe, Yuichi; Nishitani, Takeo; Konno, Chikara

Fusion Engineering and Design, 82(15-24), p.2794 - 2798, 2007/10

https://doi.org/10.1016/j.fusengdes.2007.05.076

Neutron streaming experiments have been conducted by using the FNS D-T neutron source at Japan Atomic Energy Agency under the ITER/ITA Task 73-10 in order to evaluate effects of the slit on nuclear properties and validate prediction accuracies on numerical simulations. The experimental assembly with a slit of 2 cm in width and 55 cm in depth was prepared with two iron blocks of 30 cm in height, 100 cm in width and 55cm in thickness as first campaign. The slit was located in the 12-cm upper part from the D-T neutron source point. In order to evaluate distributions of the neutron fluxes along the slit as a function of the depth from the assembly surface, fission reaction rates were measured by U-238 and U-235 micro-fission chambers. The experimental accuracies of these fission reaction rates are within 5%. Monte-Carlo calculation code, MCNP-4c, was used to calculate the U-238 and U-235 reaction rates and neutron energy spectra due to each measured position. From our first experiment, the following facts were found: (1) At d = 20 and 40 cm, reaction rates on U-238, which represent fast neutron flux, decreased by about three orders of magnitude along slits with 50 cm in depth. Monte Carlo calculation results agree well with measured values within 6 %. (2) Reaction rates on U-235, which represent thermal neutron flux, decrease by about one order of magnitude along slits with 50 cm in depth. Values of C/E of U-238 and U-235 reaction rates were 1.10-1.22 and 1.10-1.23 respectively and the calculated values overestimated slightly.

ACT-XN; Revised version of an activation calculation code for fusion reactor analysis; Supplement of the function for the sequential reaction activation by charged particles

Yamauchi, Michinori; Hori, Junichi*; Sato, Satoshi; Nishitani, Takeo; Konno, Chikara; Kawasaki, Hiromitsu*

JAEA-Data/Code 2007-016, 58 Pages, 2007/09

JAEA-Data-Code-2007-016.pdf:7.33MB

The ACT-XN is a revised version of the ACT4 code, which was developed in the Japan Atomic Energy Agency (JAEA) to calculate the transmutation, induced activity, decay heat, delayed -ray source etc. for fusion devices. The ACT4 code has not dealt with the sequential reactions of charged particles generated by primary neutron reactions. However, the reactions cannot be disregarded in the design employing low activation material, and the code was newly supplemented with the function to calculate the activation for sequential reactions and renamed the ACT-XN. The FISPACT data were adopted for (x,n) reaction cross sections, charged particles emission spectra and stopping powers. An application of the code to the analysis of FNS experiment for LiF activation confirmed that the function is enough reliable, and a notice was presented through the design calculation of the Demo-reactor with FLiBe that the dose rate may be enhanced by sequential reactions.

Measurement of tritium production rate in water cooled pebble bed multi-layered blanket mockup by DT neutron irradiation experiment

Sato, Satoshi; Ochiai, Kentaro; Verzilov, Y.*; Wada, Masayuki*; Kubota, Naoyoshi; Kondo, Keitaro; Yamauchi, Michinori; Nishitani, Takeo; Konno, Chikara

Nuclear Fusion, 47(7), p.517 - 521, 2007/07

https://doi.org/10.1088/0029-5515/47/7/003

no abstracts in English

Experimental study on nuclear properties of water cooled pebble bed blanket

Sato, Satoshi; Verzilov, Y.*; Ochiai, Kentaro; Wada, Masayuki*; Kubota, Naoyoshi; Kondo, Keitaro; Yamauchi, Michinori; Nishitani, Takeo; Konno, Chikara

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03

no abstracts in English

Possibility of tritium self-sufficiency in low aspect ratio tokamak reactor with the outboard blanket only

Hayashi, Takao; Tobita, Kenji; Nishio, Satoshi; Sato, Satoshi; Nishitani, Takeo; Yamauchi, Michinori

Fusion Engineering and Design, 81(23-24), p.2779 - 2784, 2006/11

https://doi.org/10.1016/j.fusengdes.2006.07.045

We have studied the possibility of tritium self-sufficiency with the outboard blanket only in low aspect ratio tokamak reactor to simplify the inboard structure in high magnetic field. The tritium breeding ratio (TBR) of the outboard blanket increases by applying the inboard reflector working as a neutron multiplier such as lead and beryllium. Assuming that the coverage of the breeder zone to plasma facing components is 0.78, the requirement of local TBR is above 1.35. The local TBR calculated with both the inboard reflector of lead and the outboard blanket using beryllium multiplier and LiO breeder is larger than 1.35 for aspect ratios smaller than 2.9. These results indicate there is a solution for tritium self-sufficiency using the outboard breeding blanket only in low-A tokamak reactor when the appropriate inboard reflector and outboard blanket are adopted. In case both inboard and outboard blankets are applied, the total TBR with the combination of LiO and Be is more than 1.5. However, BeTi and LiTiO are recommended from the viewpoint of safety, and the TBR with the combination is slightly larger than 1.35 for aspect ratios from 2 to 4.

Effect of photobleaching on radiation-induced transmission loss of fused-silica-core optical fibres under -ray and 14 MeV neutron irradiation

To, Kentaro*; Shikama, Tatsuo*; Nagata, Shinji*; Tsuchiya, Bun*; Yamauchi, Michinori; Nishitani, Takeo

Measurement Science and Technology, 17(5), p.955 - 959, 2006/05

https://doi.org/10.1088/0957-0233/17/5/S03

The effect of photobleaching on radiation-induced transmission loss of fused-silica-core optical fibres was examined under Co -ray and 14 MeV neutron irradiation. In the visible wavelength range, the radiation-induced transmission loss could be reduced by photobleaching under both types of irradiation. It is considered that the number of radiation-induced defects such as E' centre and NBOHC that cause optical absorption was reduced by photobleaching. For using optical fibres in the visible wavelength range as light guides or image guides, photobleaching is an effective technique to improve their radiation resistance. However, a so-called microbending loss is induced by radiation in the entire wavelength range. The loss response to photobleaching was different under -ray and fast-neutron irradiation. The loss under the -ray irradiation increased due to photobleaching, while that under the fast-neutron irradiation decreased. It is considered that there are multiple reasons for the radiation-enhanced microbending loss. In general, it is concluded that the photobleaching technique is effective in improving the radiation resistance of fused-silica-core optical fibres under neutron and moderate flux radiation environments.

Measurements of deuteron-induced activation cross sections for IFMIF accelerator structural materials

Nakao, Makoto; Hori, Junichi*; Ochiai, Kentaro; Kubota, Naoyoshi; Sato, Satoshi; Yamauchi, Michinori; Ishioka, Noriko; Nishitani, Takeo

FAPIG, (172), p.3 - 7, 2006/03

The IFMIF (International Fusion Materials Irradiation Facility) is in the design planning stage as neutron source to test fusion reactor materials for high fluence neutron. In the design of it long-term operation that total facility is 70 % at least is required. However, deuteron beam activates the structural materials and these activation limits maintenance and makes long-term operation difficult. Thus the accurate estimation of deuteron-induced activity and the selection of structural materials are important. In this work, measurements of deuteron-induced activation cross sections for aluminum, iron, copper, tantalum, and tungsten were performed.