Atomic distribution and local structure in ice VII from in situ neutron diffraction

Yamashita, Keishiro*; Komatsu, Kazuki*; Klotz, S.*; Fabelo, O.*; Fernndez-Daz, M. T.*; Abe, Jun*; Machida, Shinichi*; Hattori, Takanori; Irifune, Tetsuo*; Shimmei, Toru*; et al.

Proceedings of the National Academy of Sciences of the United States of America, 119(40), p.e2208717119_1 - e2208717119_6, 2022/10

https://doi.org/10.1073/pnas.2208717119

Here we present the first elucidation of the disordered structure of ice VII, the dominant high-pressure form of water, at 2.2 GPa and 298 K from both single-crystal and powder neutron diffraction techniques. We reveal the three-dimensional atomic distributions from the maximum entropy method and unexpectedly find a ring-like distribution of hydrogen in contrast to the commonly-accepted discrete sites. In addition, total scattering analysis at 274 K clarified the difference in the intermolecular structure from ice VIII, the ordered counterpart of ice VII, despite an identical molecular geometry. Our complementary structure analyses robustly demonstrate the unique disordered structure of ice VII. Furthermore, these noble findings are related to the proton dynamics which drastically vary with pressure, and will contribute to an understanding of the structural origin of anomalous physical properties of ice VII under pressures.

Annual report of Nuclear Emergency Assistance and Training Center (April 1, 2013 - March 31, 2014)

Sato, Takeshi; Muto, Shigeo; Akiyama, Kiyomitsu; Aoki, Kazufumi; Okamoto, Akiko; Kawakami, Takeshi; Kume, Nobuhide; Nakanishi, Chika; Koie, Masahiro; Kawamata, Hiroyuki; et al.

JAEA-Review 2014-048, 69 Pages, 2015/02

JAEA-Review-2014-048.pdf:13.91MB

JAEA was assigned as a designated public institution under the Disaster Countermeasures Basic Act and under the Armed Attack Situations Response Act. Based on these Acts, the JAEA has the responsibility of providing technical support to the national government and/or local governments in case of disaster responses or response in the event of a military attack, etc. In order to fulfill the tasks, the JAEA has established the Emergency Action Plan and the Civil Protection Action Plan. In case of a nuclear emergency, NEAT dispatches specialists of JAEA, supplies the national government and local governments with emergency equipment and materials, and gives technical advice and information. In normal time, NEAT provides various exercises and training courses concerning nuclear disaster prevention to those personnel taking an active part in emergency response institutions of the national and local governments, police, fire fighters, self-defense forces, etc. in addition to the JAEA itself. The NEAT also researches nuclear disaster preparedness and response, and cooperates with international organizations. In the FY2013, the NEAT accomplished the following tasks: (1) Technical support activities as a designated public institution in cooperation with the national and local governments, etc. (2) Human resource development, exercise and training of nuclear emergency response personnel for the national and local governments, etc. (3) Researches on nuclear disaster preparedness and response, and sending useful information (4) International contributions to Asian countries on nuclear disaster preparedness and response in collaboration with the international organizations

Conceptual design of multipurpose compact research reactor; Annual report FY2010 (Joint research)

Imaizumi, Tomomi; Miyauchi, Masaru; Ito, Masayasu; Watahiki, Shunsuke; Nagata, Hiroshi; Hanakawa, Hiroki; Naka, Michihiro; Kawamata, Kazuo; Yamaura, Takayuki; Ide, Hiroshi; et al.

JAEA-Technology 2011-031, 123 Pages, 2012/01

JAEA-Technology-2011-031.pdf:16.08MB

The number of research reactors in the world is decreasing because of their aging. However, the planning to introduce the nuclear power plants is increasing in Asian countries. In these Asian countries, the key issue is the human resource development for operation and management of nuclear power plants after constructed them, and also the necessity of research reactor, which is used for lifetime extension of LWRs, progress of the science and technology, expansion of industry use, human resources training and so on, is increasing. From above backgrounds, the Neutron Irradiation and Testing Reactor Center began to discuss basic concept of a multipurpose low-power research reactor for education and training, etc. This design study is expected to contribute not only to design tool improvement and human resources development in the Neutron Irradiation and Testing Reactor Center but also to maintain and upgrade the technology on research reactors in nuclear power-related companies. This report treats the activities of the working group from July 2010 to June 2011 on the multipurpose low-power research reactor in the Neutron Irradiation and Testing Reactor Center and nuclear power-related companies.

Handling of HTTR second driver fuel elements in assembling and storage working

Tomimoto, Hiroshi; Kato, Yasushi; Owada, Hiroyuki; Sato, Nao; Shimazaki, Yosuke; Kozawa, Takayuki; Shinohara, Masanori; Hamamoto, Shimpei; Tochio, Daisuke; Nojiri, Naoki; et al.

JAEA-Technology 2009-025, 29 Pages, 2009/06

JAEA-Technology-2009-025.pdf:21.78MB

The first driver fuel of the HTTR (High Temperature Engineering test Reactor) was loaded in 1998 and the HTTR reached first criticality state in the same year. The HTTR has been operated using the first driver fuel for a decade. In Fuel elements assembling, 4770 of fuel rods which consist of 12 kinds of enrichment uranium are loaded into 150 fuel graphite blocks for HTTR second driver fuel elements. Measures of prevention of fuel rod miss loading, are employed in fuel design. Additionally, precaution of fuel handling on assembling are considered. Reception of fuel rods, assembling of fuel elements and storage of second driver fuels in the fresh fuel storage rack in the HTTR were started since June, 2008. Assembling, storage and pre-service inspection were divided into three parts. The second driver fuel assembling was completed in September, 2008. This report describes concerns of fuel handling on assembling and storage work for the HTTR fuel elements.

Fabrication of irradiation capsule for IASCC irradiation tests, 1; Irradiation capsule for crack growth test (Joint research)

Ide, Hiroshi; Matsui, Yoshinori; Kawamata, Kazuo; Taguchi, Taketoshi; Kanazawa, Yoshiharu; Onuma, Yuichi; Watanabe, Hiroyuki; Inoue, Shuichi; Izumo, Hironobu; Ishida, Takuya; et al.

JAEA-Technology 2008-011, 46 Pages, 2008/03

JAEA-Technology-2008-011.pdf:19.39MB

It is known that Irradiation Assisted Stress Corrosion Cracking (IASCC) occurs when austenitic stainless steel components used for light water reactor (LWR) are irradiated for a long period. In order to evaluate the high aging of the nuclear power plant, the study of IASCC becomes the important problem. The specimens irradiated in the reactor were evaluated by post irradiation examination in the past study. For the appropriate evaluation of IASCC, It is necessary to test it under the simulated LWR conditions; temperature, water chemistry and irradiation conditions. In order to perform in-pile SCC test, saturated temperature capsule (SATCAP) was developed. There are crack growth test, crack propagation test and so on for in-pile SCC test. In this report, SATCAP for crack growth test is reported.

Study on mixing due to transversal flow in a fuel subassembly of fast reactor; Sodium experiment using a 37-pin subassembly model

Kamide, Hideki; Miyakoshi, Hiroyuki; Kawamata, Nobuhiro; Ohshima, Hiroyuki

Proceedings of 12th International Conference on Nuclear Engineering (ICONE-12) (CD-ROM), 49199 Pages, 2004/04

High burnup core is one of significant ways to increase cost performance of a fuel cycle. Deformation of a pin bundle or wrapper tube due to irradiation has been studied for the high burnup core. Here, mixing phenomena with cross flow was investigated for a prediction method of thermohydraulics in a deformed pin bundle in a fast reactor. Sodium experiments were carried out using a wire wrapped 37-pin subassembly model, where a blockage was installed in order to make cross flow around it. The subchannel analysis code, ASFRE, was applied to the experimental analyses. Axial and transverse temperature distributions in the subassembly were correlated with the cross flow and the wire spacer. The calculated temperature profiles including a decrease due to the cross flow were in good agreement with the experimental results. ASFRE simulated also influence of the wire spacer. Database for mixing phenomena with the cross flow was obtained for a subassembly of fast reactor

Detailed inspection of samples taken from un-irradiated fuel assembly in 1F4, 3; Results of cross sectional observation by optical microscope and SEM/EPMA

Endo, Shinya; Motooka, Takafumi; Uehara, Hiroyuki; Obata, Hiroki; Kawamata, Yutaka; Ueno, Fumiyoshi

no journal, , 

no abstracts in English