Design for detecting recycling muon after muon-catalyzed fusion reaction in solid hydrogen isotope target

Okutsu, Kenichi*; Yamashita, Takuma*; Kino, Yasushi*; Nakashima, Ryota*; Miyashita, Konan*; Yasuda, Kazuhiro*; Okada, Shinji*; Sato, Motoyasu*; Oka, Toshitaka; Kawamura, Naritoshi*; et al.

Fusion Engineering and Design, 170, p.112712_1 - 112712_4, 2021/09

https://doi.org/10.1016/j.fusengdes.2021.112712

A muonic molecule which consists of two hydrogen isotope nuclei (deuteron (d) or tritium (t)) and a muon decays immediately via nuclear fusion and the muon will be released as a recycling muon, and start to find another hydrogen isotope nucleus. The reaction cycle continues until the muon ends up its lifetime of 2.2 s. Since the muon does not participate in the nuclear reaction, the reaction is so called a muon catalyzed fusion (CF). The recycling muon has a particular kinetic energy (KE) of the muon molecular orbital when the nuclear reaction occurs. Since the KE is based on the unified atom limit where distance between two nuclei is zero. A precise few-body calculation estimating KE distribution (KED) is also in progress, which could be compared with the experimental results. In the present work, we observed recycling muons after CF reaction.

Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

Kasahara, Shigeki; Kitsunai, Yuji*; Chimi, Yasuhiro; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka

Journal of Nuclear Materials, 480, p.386 - 392, 2016/11

https://doi.org/10.1016/j.jnucmat.2016.06.017

This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. Tensile tests at 290C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Influence of difference in the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. The influence was also found certainly in loss of strain hardening capacity and ductility, although the influence on the yield strength and the Vickers hardness was not clearly observed. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, were considered to contribute to deformation of the austenitic stainless steel.

Recent progress in the chopper spectrometer 4SEASONS at J-PARC

Kajimoto, Ryoichi; Nakamura, Mitsutaka; Inamura, Yasuhiro; Kamazawa, Kazuya*; Ikeuchi, Kazuhiko*; Iida, Kazuki*; Ishikado, Motoyuki*; Nakajima, Kenji; Kawamura, Seiko; Nakatani, Takeshi; et al.

JAEA-Conf 2015-002, p.319 - 329, 2016/02

https://doi.org/10.11484/jaea-conf-2015-002

Investigation on iodine release behavior during the operation of high temperature engineering test reactor (HTTR)

Ueta, Shohei; Inoi, Hiroyuki; Mizutani, Yoshitaka; Ohashi, Hirofumi; Iwatsuki, Jin; Sakaba, Nariaki; Sawa, Kazuhiro

Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 4 Pages, 2013/07

https://doi.org/10.1115/ICONE21-16128

Japan Atomic Energy Agency (JAEA) has planned to investigate on iodine release behavior from fuel through the testing operation of High Temperature Engineering Test Reactor (HTTR) in order to contribute to the reasonable estimation of the radiation exposure necessary for the realization of HTGR in the future. In this test, the fractional release of iodine will be measured and evaluated by measuring xenon isotopes, the daughter nuclides of iodine isotopes, in the primary coolant sampling under the loss-of-forced cooling (LOFC) test by which the primary coolant circulator is shut down and/or the manual scram test of HTTR. In parallel, the local area of primary coolant circuit where iodine is plated-out will be evaluated. This paper describes the testing plan and the preliminary analytical study on the release behavior of iodine and xenon isotopes through the operation of HTTR.

IASCC evaluation method for irradiated core internal structures in BWR power plants

Takakura, Kenichi*; Tanaka, Shigeaki*; Nakamura, Tomomi*; Chatani, Kazuhiro*; Kaji, Yoshiyuki

Proceedings of 2010 ASME Pressure Vessels and Piping Conference (PVP 2010) (CD-ROM), 10 Pages, 2010/07

https://doi.org/10.1115/PVP2010-25293

Irradiation Assisted Stress Corrosion Cracking (IASCC) is a matter of great concern as a degradation of core internal components in light water nuclear reactor. Japan Nuclear Energy Safety organization (JNES) had been conducting a project related to IASCC as a part of safety research and development study for the aging management and maintenance of the nuclear power plants. Based on the JNES project results, JNES proposed "IASCC evaluation guide for BWR core internals". The purpose of this paper is to describe the background of the guide, especially crack growth rate (CGR) tests for irradiated stainless steels.

Development of optical fiber detector for measurement of fast neutron

Yagi, Takahiro*; Misawa, Tsuyoshi*; Pyeon, C. H.*; Unesaki, Hironobu*; Shiroya, Seiji*; Kawaguchi, Shinichi*; Okajima, Shigeaki; Tani, Kazuhiro*

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8 Pages, 2008/09

In order to insert a neutron detector in a narrow space such as a gap of between fuel plates and measure the fast neutrons in real time, a neutron detector with an optical fiber has been developed. This detector consists of an optical fiber whose tip is covered with mixture of neutron converter material and scintillator such as ZnS(Ag). The detector for fast neutrons uses ThO as converter material because Th makes fission reaction with fast neutrons. The place where Th can be uses is limited by regulations because Th is nuclear fuel material. The purpose of this research is to develop a new optical fiber detector to measure fast neutrons without Th and to investigate the characteristic of the detector. These detectors were used to measure a D-T neutron generator and fast neutron flux distribution at Fast Critical Assembly. The results showed that the fast neutron flux distribution of the new optical fiber detector with ZnS(Ag) was the same as it of the activation method, and the detector are effective for measurement of fast neutrons.

IASCC crack growth rate of neutron irradiated low carbon austenitic stainless steels in simulated BWR condition

Chatani, Kazuhiro*; Takakura, Kenichi*; Ando, Masami*; Nakata, Kiyotomo*; Tanaka, Shigeaki*; Ishiyama, Yoshihide*; Hishida, Mamoru*; Kaji, Yoshiyuki

Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems (CD-ROM), 9 Pages, 2007/00

Crack growth rate (CGR) tests have been conducted with neutron irradiated compact tension (CT) specimens. The CGR tests of 316L and 304L base metals irradiated from 0.516 to 1.0710n/m (E1MeV), and of 316L and 308L weld metals irradiated from 0.523 to 0.54110n/m (E1MeV) were performed using the reversing dc potential drop (DCPD) method under constant load at a few average stress intensity factors (K) and electrochemical corrosion potential (ECP) conditions at 288C in water. CGRs of base metals were increased with increasing neutron fluence. Clear reductions in CGRs of base metals and weld metals were measured with decreasing ECP levels.

Study on Thermal Electric Conversion System for Sodium cooled FBR; Investigation for development of thermoelectric materials and systematic technology

Suzuki, Ryosuke*; Tanabe, Kentaro*; kondo, koki*; Ono, Katsutoshi*; Toda, Shinichi; Kasagawa, Yusuke; Tamayama, Kiyoshi; Oketani, Kazuhiro*

JNC TY4400 2003-004, 214 Pages, 2003/08

JNC-TY4400-2003-004.pdf:19.93MB

Recently, it has been important to reuse discharged heat energy from present nuclear plants in the view of reduction of environmental burden and improvement of heat efficiency for plant. For practical use in future of sodium cooled FBRs, which are typical high temperature system, this issue must be given priority. The thermal electric conversion system has been applied to the limited uses such as space or military, however, that results show good merits for reliability, maintenance free, and so on. Recently, this technology has been reconsidered in the view of saving energy in general industry. In this study, we made an investigation for applicability of the thermal electric conversion system to sodium cooled FBR as a heat recovery techbnology. Exactly, We have carried out the fundamental research and development for thermoelectric materials and elements, development of modules, and sodium tests with those modules, and then, we acquired the fundamental knowledge to estimate the efficiencies of thermal electric conversion system or modules for a sodium cooled FBR.

Improvement of the MSG code for the MONJU Evaporators; Additional function of reverse flow calculation on water/steam model and animation for post processing

Toda, Shinichi; Yoshikawa, Shinji; Watanabe, Osamu*; Kishida, Masako*; Oketani, Kazuhiro*

JNC TN4400 2003-005, 106 Pages, 2003/05

JNC-TN4400-2003-005.pdf:3.82MB

The improved version of the MSG code(Multi-dimensional Thermal-hydroulic Analysis Code for Steam Generators) has been released. It has been carried out to improve based on the original version in order to calculate reverse flow on water/steam side, and to animate the post-processing data. To calculate reverse flow locally, modification to set pressure at each divided node point of water/steam region in the helical-coil heat tansfer tubes has been carried out. And the matrix solver has been also improved to treat a problem within practical calculation time against increasing the pressure points. In this case pressure and enthalpy have to be calculated simultaneously, however, it was found out that using the bloci-Jacobean method make a diagonal-dominant matrix, and solve the matriz efficiently with a relaxation method. As the result of calculations of a steady-state condition and a transient of SG blow down with manual tip operation, the improvement on calculation function of the MSG code was confirmed. And an animation function of temperature contour in the sodium shell side as a post procssing has been added. Since the animation is very effective to understand themal-hydraulic behavior on the sodium shell side of the SG, especially in case of transient condition, the analysis and evalution of the calculation results will be enabled to be more quickly and effectively.

Safety design concepts for ITER-tritium facility; Toward construction in Japan

Ohira, Shigeru; Tada, Eisuke; Hada, Kazuhiko; Neyatani, Yuzuru; Maruo, Takeshi; Hashimoto, Masayoshi*; Araki, Takao*; Nomoto, Kazuhiro*; Tsuru, Daigo; Ishida, Toshikatsu*; et al.

Fusion Science and Technology, 41(3), p.642 - 646, 2002/05

no abstracts in English