Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 87

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Geochemical and grain composition analysis of embankment and debris flow deposits in the Izusan area, Atami City, Shizuoka Prefecture, central Japan

Kitamura, Akihisa*; Okazaki, Sota*; Kondo, Mitsuru*; Watanabe, Takahiro; Nakanishi, Toshimichi*; Hori, Rie*; Ikeda, Masayuki*; Ichimura, Koji*; Nakagawa, Yuki*; Mori, Hideki*

Shizuoka Daigaku Chikyu Kagaku Kenkyu Hokoku, (49), p.73 - 86, 2022/07

On July 3 2021, a debris flow caused by a landslide from a landfill occurred along the Aizome River in the Izusan area of Atami City, Shizuoka. In this study, debris flow deposits and soil samples were characterized in terms of their sedimentology and geochemically analyzed.

Journal Articles

High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950$$^{circ}$$C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.

Journal Articles

Research activities of Japan Nuclear Data Committee in fiscal years of 2003 and 2004

Igashira, Masayuki*; Watanabe, Yukinobu*; Fukahori, Tokio; Okumura, Keisuke; Katakura, Junichi; Chiba, Satoshi; Shibata, Keiichi; Yamano, Naoki*; Nakagawa, Tsuneo; Odano, Naoteru*; et al.

Nihon Genshiryoku Gakkai Wabun Rombunshi, 6(1), p.85 - 96, 2007/03

This technical note summarizes research activities on nuclear data carried out by Japanese Nuclear Data Committee (JNDC) during the fiscal years of 2003 and 2004. During this period, the nuclear data files for special purposes (JENDL-HE-2004 and JENDL-PD-2004) were released. Other activities are described: analysis of post nuclear fuel irradiation experiments, nuclear chart and nuclear data evaluation for astrophysics.

JAEA Reports

Development of the Unattended Spent Fuel Flow Monitoring Safeguards System (UFFM) for the High Temperature Engineering Test Reactor (HTTR) (Joint research)

Nakagawa, Shigeaki; Umeda, Masayuki; Beddingfield, D. H.*; Menlove, H. O.*; Yamashita, Kiyonobu

JAEA-Technology 2007-003, 24 Pages, 2007/02

JAEA-Technology-2007-003.pdf:3.61MB

As of the safeguards approach in the HTTR facility, an unattended spent fuel flow monitor (UFFM) was applied to carry out an item counting of spent fuel blocks. The UFFM is so designed and fabricated as to be the compact and unique monitor system to verify a movement of spent fuel blocks in "difficult to access" area and reduce inspection efforts. This system consists of two detector packages, electronics and computer. One package consists of two ionization chambers and a He-3 counter. The IAEA acceptance tests were performed and it was confirmed the followings: (1) All the detectors were functioning properly to measure a spent fuel block flow. (2) The time difference between detector signals was sufficient to determine the direction of the spent fuel blocks. (3) The UFFM was useful to carry out the item counting. The UFFM was approved as the IAEA safeguards equipment in the safeguards approach in the HTTR.

JAEA Reports

MVP/GMVP 2; General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki

JAERI 1348, 388 Pages, 2005/06

JAERI-1348.pdf:2.02MB

To realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector supercomputers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them.

Journal Articles

Web-based search and plot system for nuclear reaction data

Otsuka, Naohiko; Aikawa, Masayuki*; Suda, Takuma*; Naito, Kenichi*; Korennov, S.*; Arai, Koji*; Noto, Hiroshi*; Onishi, Akira*; Kato, Kiyoshi*; Nakagawa, Tsuneo; et al.

AIP Conference Proceedings 769, p.561 - 564, 2005/05

A web-based search and plot system for nuclear reaction data has been developed, covering experimental data in EXFOR format and evaluated data in ENDF format. The system is implemented for Linux OS, with Perl and MySQL used for CGI scripts and the database manager, respectively. Two prototypes for experimental and evaluated data are presented.

Journal Articles

Research activities of Japanese Nuclear Data Committee in the fiscal years of 2001 and 2002

Igashira, Masayuki*; Shibata, Keiichi; Takano, Hideki*; Yamano, Naoki*; Matsunobu, Hiroyuki*; Kitao, Kensuke*; Katakura, Junichi; Nakagawa, Tsuneo; Hasegawa, Akira; Iwasaki, Tomohiko*; et al.

Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.128 - 139, 2004/03

no abstracts in English

Journal Articles

Japanese evaluated nuclear data library version 3 revision-3; JENDL-3.3

Shibata, Keiichi; Kawano, Toshihiko*; Nakagawa, Tsuneo; Iwamoto, Osamu; Katakura, Junichi; Fukahori, Tokio; Chiba, Satoshi; Hasegawa, Akira; Murata, Toru*; Matsunobu, Hiroyuki*; et al.

Journal of Nuclear Science and Technology, 39(11), p.1125 - 1136, 2002/11

 Times Cited Count:673 Percentile:93.94(Nuclear Science & Technology)

Evaluation for JENDL-3.3 has been performed by considering the accumulated feedback information and various benchmark tests of the previous library JENDL-3.2. The major problems of the JENDL-3.2 data were solved by the new library: overestimation of criticality values for thermal fission reactors was improved by the modifications of fission cross sections and fission neutron spectra for $$^{235}$$U; incorrect energy distributions of secondary neutrons from important heavy nuclides were replaced with statistical model calculations; the inconsistency between elemental and isotopic evaluations was removed for medium-heavy nuclides. Moreover, covariance data were provided for 20 nuclides. The reliability of JENDL-3.3 was investigated by the benchmark analyses on reactor and shielding performances. The results of the analyses indicate that JENDL-3.3 predicts various reactor and shielding characteristics better than JENDL-3.2.

JAEA Reports

Measurement of $$gamma$$ ray from fuel of High Temperature Engineering Test Reactor; Method of measurement and results

Fujimoto, Nozomu; Nojiri, Naoki; Takada, Eiji*; Yamashita, Kiyonobu; Kikuchi, Takayuki; Nakagawa, Shigeaki; Kojima, Takao; Umeta, Masayuki; Hoshino, Osamu; Kaneda, Makoto*; et al.

JAERI-Tech 2001-002, 64 Pages, 2001/02

JAERI-Tech-2001-002.pdf:3.64MB

no abstracts in English

Journal Articles

Development of continuous energy Monte Carlo burn-up calculation code MVP-BURN

Okumura, Keisuke; Nakagawa, Masayuki; Kaneko, Kunio*;

JAERI-Conf 2000-018, p.31 - 41, 2001/01

Burnup calculation codes based on the conventional deterministic approach often encounter difficult problems because of the constraints on the geometry description, limit of approximation on the effective resonance cross-sections, failing of the diffusion approximation due to extremely strong anisotropic or heterogenity. They are, for example, the prediction of burn characteristics of plutonium spot, core design of ultra-small reactors, analysis of the sample material in an irradiation capsule of the research rector. To deal with these problems any time, a burn-up calculation code (MVP-BURN) was developed by using a continuous energy Monte Carlo code MVP. MVP-BURN was validated by comparison with the results of deterministic codes in the international benchmark problems, and by comparison with the measured values of the spent fuel composition irradiated in a commercial reactor.

Journal Articles

The Nuclear interaction at Oklo 2 billion years ago

Fujii, Yasunori*; Iwamoto, Akira; Fukahori, Tokio; Onuki, Toshihiko; Nakagawa, Masayuki; Hidaka, Hiroshi*; Oura, Yasutsugu*; M$"o$ller, P.*

Nuclear Physics B, 573(1-2), p.377 - 401, 2000/05

 Times Cited Count:191 Percentile:97.51(Physics, Particles & Fields)

no abstracts in English

JAEA Reports

Test program for NIS calibration to reactor thermal output in HTTR

Nakagawa, Shigeaki; Shinozaki, Masayuki; Tachibana, Yukio; Kunitomi, Kazuhiko

JAERI-Tech 2000-038, p.39 - 0, 2000/03

JAERI-Tech-2000-038.pdf:1.86MB

no abstracts in English

JAEA Reports

Maxwellian-averaged cross sections calculated from JENDL-3.2

Nakagawa, Tsuneo; Chiba, Satoshi; Osaki, Toshiro*; Igashira, Masayuki*

JAERI-Research 2000-002, p.93 - 0, 2000/02

JAERI-Research-2000-002.pdf:4.41MB

no abstracts in English

Journal Articles

Validation of a continuous-energy Monte Carlo burn-up code MVP-BURN and its application to analysis of post irradiation experiment

Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio*

Journal of Nuclear Science and Technology, 37(2), p.128 - 138, 2000/02

no abstracts in English

Journal Articles

Application of continuous energy Monte Carlo code MVP to burn-up and whloe core calculations using cross sections at arbitrary temperatures

Mori, Takamasa; Okumura, Keisuke; Nagaya, Yasunobu; Nakagawa, Masayuki

Mathematics and Computation, Reactor Physics and Environmental Analysis in Nuclear Applications, 2, p.987 - 996, 1999/09

no abstracts in English

Journal Articles

Application of neural network to multi-dimensional design window search in reactor core design

Kugo, Teruhiko; Nakagawa, Masayuki

Journal of Nuclear Science and Technology, 36(4), p.332 - 343, 1999/04

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

no abstracts in English

Journal Articles

Present status of Monte Carlo simulation for neutron and photon transport

Ueki, Kotaro*; Mori, Takamasa; Sakurai, Kiyoshi; Nakagawa, Masayuki;

Nihon Genshiryoku Gakkai-Shi, 41(6), p.614 - 627, 1999/00

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Benchmark analysis of experiments in fast critical assemblies using a continuous-energy monte carlo code MVP

Nagaya, Yasunobu; Nakagawa, Masayuki; Mori, Takamasa

Journal of Nuclear Science and Technology, 35(1), p.6 - 19, 1998/01

 Times Cited Count:3 Percentile:31.90(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Applicability of design window search procedure using neural network to neutronics

Kugo, Teruhiko; Nakagawa, Masayuki

Proc. of Int. Conf. on the Phys. of Nucl. Sci. and Technol., 1, p.704 - 711, 1998/00

no abstracts in English

Journal Articles

Development of intelligent code system to support conceptual design of nuclear reactor core

Kugo, Teruhiko; Nakagawa, Masayuki;

Journal of Nuclear Science and Technology, 34(8), p.760 - 770, 1997/08

 Times Cited Count:2 Percentile:22.91(Nuclear Science & Technology)

no abstracts in English

87 (Records 1-20 displayed on this page)