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JAEA Reports

Interim activity status report of "the group for investigation of reasonable safety assurance based on graded approach" (from September, 2019 to September, 2020)

Yonomoto, Taisuke; Nakashima, Hiroshi*; Sono, Hiroki; Kishimoto, Katsumi; Izawa, Kazuhiko; Kinase, Masami; Osa, Akihiko; Ogawa, Kazuhiko; Horiguchi, Hironori; Inoi, Hiroyuki; et al.

JAEA-Review 2020-056, 51 Pages, 2021/03


A group named as "The group for investigation of reasonable safety assurance based on graded approach", which consists of about 10 staffs from Sector of Nuclear Science Research, Safety and Nuclear Security Administration Department, departments for management of nuclear facility, Sector of Nuclear Safety Research and Emergency Preparedness, aims to realize effective graded approach (GA) about management of facilities and regulatory compliance of JAEA. The group started its activities in September, 2019 and has had discussions through 10 meetings and email communications. In the meetings, basic ideas of GA, status of compliance with new regulatory standards at each facility, new inspection system, etc were discussed, while individual investigation at each facility were shared among the members. This report is compiled with expectation that it will help promote rational and effective safety management based on GA by sharing contents of the activity widely inside and outside JAEA.

JAEA Reports

Mock-up test of the modified STACY (Performance check of water feed and drain system)

Seki, Masakazu; Maekawa, Tomoyuki; Izawa, Kazuhiko; Sono, Hiroki

JAEA-Technology 2017-038, 52 Pages, 2018/03


The Japan Atomic Energy Agency is conducting a reactor modification project of the Static Experiment Critical Facility (STACY). In the modification, STACY is to be converted from a thermal reactor using solution fuel into that using fuel rods and light water moderator. Reactivity of the modified STACY core is controlled by the water level fed in the core tank as well as the present STACY. In order to verify the basic design of the water feed and drain system of the modified STACY, we constructed a mockup test apparatus with almost the same structure and specifications as the modified STACY. In the mockup test, performance checks were pursued regarding limitation of maximum flow of water feeding, adjustment of the flow rate of water feeding, stop of water feeding and others. This report describes the outline and results of the mock-up test of the water feed and drain system of the modified STACY.

Journal Articles

Research facilities of Japan Atomic Energy Agency in Fukushima Prefecture; Overview

Yoshikawa, Hideki

Enerugi Rebyu, 37(10), p.13 - 14, 2017/10

We introduce the R&D activities of Japan Atomic Energy Agency immediately after the accident at Fukushima Daiichi Nuclear Power Station and some facilities established in Fukushima Prefecture.

Journal Articles

Validation of PHITS spallation models from the perspective of the shielding design of transmutation experimental facility

Iwamoto, Hiroki; Meigo, Shinichiro

EPJ Web of Conferences, 153, p.01016_1 - 01016_9, 2017/09

 Times Cited Count:2 Percentile:85.61

JAEA Reports

Solvent extraction and release behavior of ruthenium and europium in fire accident conditions in reprocessing plants (Contract research)

Amano, Yuki; Watanabe, Koji; Masaki, Tomoo; Tashiro, Shinsuke; Abe, Hitoshi

JAEA-Technology 2016-012, 21 Pages, 2016/06


To contribute to safety evaluation of fire accident in fuel reprocessing plants, solvent extraction behavior of ruthenium, which could form volatile species, was investigated. Distribution ratios of ruthenium at fire accident conditions were obtained by extraction experiments with several solvent composition at different temperature as parameters. In order to investigate release behavior of ruthenium and europium at fire accident, release ratios of ruthenium and europium were also obtained by solvent combustion experiments.

Journal Articles

Development of multi-group neutron activation cross-section library for decommissioning of nuclear facilities

Okumura, Keisuke; Kojima, Kensuke; Tanaka, Kenichi*

JAEA-Conf 2015-003, p.43 - 47, 2016/03

In the safety assessment concerning disposal of radioactive wastes generated in the decommissioning of nuclear facilities, it is necessary to evaluate the radionuclide inventory produced by the activation of structured materials. For this purpose, we have to pay much attention to the activation of many impurities irradiated in various neutron spectra depending on their positions and materials. Therefore, accurate activation cross-section data are necessary for many nuclides and reactions. A new multi-group neutron activation cross-section library (MAXS) was developed based on the recent nuclear data JENDL-4.0 and JEFF-3.0/A to apply it to the activation calculations for the decommissioning of nuclear facilities. The library contains cross-sections and isomeric ratios for many reactions such as (n,$$gamma$$), (n,f), (n,2n), (n,3n), (n,p), (n,$$alpha$$), (n,d), (n,t), (n,n$$alpha$$), (n,np), and so on, for 779 nuclides, in the 199-energy group structure of VITAMIN-B6.

Journal Articles

Scaling issues for the experimental characterization of reactor coolant system in integral test facilities and role of system code as extrapolation tool

Mascari, F.*; Nakamura, Hideo; Umminger, K.*; De Rosa, F.*; D'Auria, F.*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.4921 - 4934, 2015/08

JAEA Reports

Investigation of release behavior of volatile ruthenium species from thermal decomposition of ruthenium nitrosylnitrate

Abe, Hitoshi; Masaki, Tomoo; Amano, Yuki; Uchiyama, Gunzo

JAEA-Research 2014-022, 12 Pages, 2014/11


To contribute safety evaluation of boiling and drying accident of high active liquid waste (HALW) in fuel reprocessing plant, release behavior of Ru, which was considered as an important nuclide for evaluating public dose from the volatile viewpoint, has been investigated. It has been reported that release of Ru becomes conspicuously after HALW is dried up. In this work, to grasp the release behavior of Ru, release ratio of Ru with thermal decomposition of Ru nitrate, which would be in the dried HALW, was measured and release rate constant of Ru from the nitrate was estimated. It was found that the calculation result of release rate of Ru from the nitrate with rise of temperature by using the constant could well simulate the result acquired from the beaker-scale experiment.

Journal Articles

Concept of transmutation experimental facility

Oigawa, Hiroyuki; Sasa, Toshinobu; Kikuchi, Kenji; Nishihara, Kenji; Kurata, Yuji; Umeno, Makoto*; Tsujimoto, Kazufumi; Saito, Shigeru; Futakawa, Masatoshi; Mizumoto, Motoharu; et al.

Proceedings of 4th International Workshop on the Utilisation and Reliability of High Power Proton Accelerators, p.507 - 517, 2005/11

Under the framework of J-PARC, the Japan Atomic Energy Research Institute (JAERI) plans to construct the Transmutation Experimental Facility (TEF). The TEF consists of two facilities: the Transmutation Physics Experimental Facility (TEF-P) and the ADS Target Test Facility (TEF-T). The TEF-P is a critical facility which can accept a 600 MeV - 10 W proton beam. The TEF-T is a material irradiation facility using a 600 MeV - 200 kW proton beam, where a Pb-Bi target is installed, but neutron multiplication by nuclear fuel will not be attempted. This report describes the purposes of the facility, the present status of the conceptual design, and the expected experiments to be performed.

JAEA Reports

Annual report on experimental operation of mock-up model test facility with a full-scale reaction tube for HTTR hydrogen production system in 2001 fiscal year (Contract research)

Hayashi, Koji; Inagaki, Yoshiyuki; Kato, Michio; Fujisaki, Katsuo*; Aita, Hideki; Takeda, Tetsuaki; Nishihara, Tetsuo; Inaba, Yoshitomo; Ohashi, Hirofumi; Katanishi, Shoji; et al.

JAERI-Tech 2005-032, 46 Pages, 2005/06


This is annual report on the experimental operation of the mock-up test facility with a full-scale reaction tube for the HTTR hydrogen production system in 2001 fiscal year. The first experimental operation was performed during two weeks from March 1, 2002 to March 13, 2002 to test on the thermal hydraulic performance of the steam reformer and also to train the operators. The thermal hydraulic performance test of the steam reformer was performed to evaluate the heat transfer characteristics between helium gas and process gas in the steam reformer. This report is summarized with an overview of the test, the results and its operation records.

JAEA Reports

Study on system integration of robots Operated in nuclear fusion facility and nuclear power plant facilities

Oka, Kiyoshi

JAERI-Research 2004-009, 225 Pages, 2004/07


In the present paper, I describe the necessary approaches and elemental technologies to solve the issues on the system integration of the typical robot systems for maintenance in the nuclear fusion facility and rescue in the accident of the nuclear power plant facilities. These robots work under the intense radiation condition and restricted space in place of human. In particular, I propose a new approach to realize the system integration of the robot for actual use from the viewpoints of not only the environmental and working conditions but also the restructure and optimization of the required elemental technologies with a well balance in the robot system. Based on the above approach, I have a contribution to realize the robot systems working under the actual conditions for maintenance in the nuclear fusion facility and rescue in the accident of the nuclear power plant facilities.

JAEA Reports

Differential pressure analysis for ventilation filter by smoke under fire accident with CELVA-1D

Watanabe, Koji; Tashiro, Shinsuke; Abe, Hitoshi; Takada, Junichi; Morita, Yasuji

JAERI-Tech 2004-029, 48 Pages, 2004/03


In a part of building ventilating installation of a nuclear fuel facility, a reprocessing plant for example, the pre-filters are adopted as one of the ventilation filters. In a fire accident, it is supposed that, because of the pre-filter clogging by large smoke, its differential pressure ($$Delta$$p) is evolved up to the value at its breakage. Therefore, in regard to maintaining the confinement of radioactive materials, it is important to predict the time course of $$Delta$$p of the pre-filter accurately. In the current study, it was assumed that the empirical equation for the DF of the pre-filter as function of smoke particle diameter (SPD), between 0.7-2 $$mu$$m, could be applied to estimating its DF for SPD above 2 $$mu$$m. Under this assumption, the time corresponding to its $$Delta$$p at its breakage, calculated by CELVA-1D, was agreed well with the experimental result.

Journal Articles

Nuclear science and engineering expected in high-intensity Proton Accelerator facility (J-PARC)

Kiyanagi, Yoshiaki*; Nagamiya, Shoji*; Oyama, Yukio; Ikeda, Yujiro; Oigawa, Hiroyuki; Igashira, Masayuki*; Baba, Mamoru*; Iwasaki, Tomohiko*; Watanabe, Yukinobu*; Ishibashi, Kenji*

Nihon Genshiryoku Gakkai-Shi, 46(3), p.173 - 197, 2004/03

no abstracts in English

Journal Articles

Development of remote surveillance squads for nuclear facility accidents

Kobayashi, Tadayoshi; Miyajima, Kazutoshi; Yanagihara, Satoshi

Keisoku Jido Seigyo Gakkai Dai-3-Kai Shisutemu Integureshon Bumon Koenkai (SI2002) Koen Rombunshu, 2, p.105 - 106, 2002/00

no abstracts in English

Journal Articles

Application of infrared thermography to nondestructive evaluation of nuclear structures and components

Ishii, Toshimitsu; Ooka, Norikazu; Naka, Michihiro; Kanaya, Kunio; Saito, Junichi; Okamoto, Yoshizo*

Proceedings of 6th Far-East Conference on Nondestructive Testing (FENDT '02), p.273 - 278, 2002/00

no abstracts in English

Journal Articles

Development of radiation resistant robot for rescue operations in accident of nuclear facility

Shibanuma, Kiyoshi

Denki Kyokai-Ho, (925), p.20 - 23, 2001/11

no abstracts in English

Journal Articles

Procedure for segregation of non-radioactive waste related to decommissioning nuclear facilities; Experiences on the Japan Power Demonstration Reactor dismantling

Tachibana, Mitsuo; Hatakeyama, Mutsuo; Yanagihara, Satoshi

Nihon Genshiryoku Gakkai-Shi, 41(6), p.677 - 685, 1999/00

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Present status and activities of High energy nuclear data evaluation; For future radiation utilization facilities

Fukahori, Tokio; Mukaiyama, Takehiko; Oyama, Yukio; Chiba, Satoshi; Takada, Hiroshi; Maekawa, Hiroshi; Shibata, Tokushi*; Nakamura, Takashi*; *; *; et al.

Nihon Genshiryoku Gakkai-Shi, 40(1), p.3 - 28, 1998/00

 Times Cited Count:1 Percentile:11.76(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Radiation aging and degradation mechanism of polymer insulation for electric cables

*; *; *; Seguchi, Tadao; Yagi, Toshiaki

Proc. of 46th Int. Wire and Cable Symp., p.1 - 5, 1997/00

no abstracts in English

Journal Articles

Three-dimensional void fraction measurement of two-phase flow in a rod bundle by neutron radiography

*; *; Fujii, Terushige*; *; Matsubayashi, Masahito; Tsuruno, Akira

Nuclear Instruments and Methods in Physics Research A, 377, p.115 - 118, 1996/00

 Times Cited Count:12 Percentile:71.65(Instruments & Instrumentation)

no abstracts in English

39 (Records 1-20 displayed on this page)