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Journal Articles

Implementation of random sampling for ACE-format cross sections using FRENDY and application to uncertainty reduction

Kondo, Ryoichi*; Endo, Tomohiro*; Yamamoto, Akio*; Tada, Kenichi

Proceedings of International Conference on Mathematics and Computational Methods applied to Nuclear Science and Engineering (M&C 2019) (CD-ROM), p.1493 - 1502, 2019/00

A perturbation capability of ACE formatted cross section files was developed using the modules of FRENDY. Uncertainty quantification using MCNP was carried out for the Godiva critical experiment by the RS method. We verified the results of the RS method by comparing with those obtained by the conventional sensitivity analyses. Moreover, uncertainty reduction using the bias factor method with the RS technique was applied to kinetic parameter, i.e., neutron generation time.

Journal Articles

Present status and future plan of JENDL

Iwamoto, Osamu

JAEA-Conf 2018-001, p.87 - 91, 2018/12

Status and plan of JENDL will be presented. After the release of JENDL-4.0 in 2010, six special purpose files have been developed. Four of them were already released and two are under preparation for the release. New decay and yield data for fission products were released as JENDL/FPD-2011 and JENDL/FPY-2011 in 2011, respectively. JENDL-4.0/HE released in 2015 includes proton and neutron induced reaction data up to 200 MeV. Comprehensive decay data were released as JENDL/DDF-2015 which contains data for 3,237 nuclides. New photonuclear reaction data JENDL/PD-2016 and an activation file JENDL/AD-2017 are under preparation for release. Regarding general purpose file, two activities are in progress. One is JENL-4.0u which is created for maintenance of JENDL-4.0 and the other is development of next version of JENDL. For the next JENDL, evaluation for light nuclei and structure material are in progress. It is planed that next version of JENDL will be JENDL-5 which contains nuclear data for all nuclei having natural abundance. Addition of covariance data will be one of the main targets.

Journal Articles

2018 Annual Meeting of Japan Atomic Energy Society, Joint Session of Nuclear Data Subcommittee and Sigma Special Advisory Committee; Present status and future of nuclear data evaluation code in Japan, 4; Role and improvement of nuclear reaction models in the PHITS code

Hashimoto, Shintaro; Sato, Tatsuhiko; Iwamoto, Yosuke; Ogawa, Tatsuhiko; Furuta, Takuya; Abe, Shinichiro; Niita, Koji*

Kaku Deta Nyusu (Internet), (120), p.26 - 34, 2018/06

Particle and heavy-ion transport code system PHITS has been used for calculations of radiation shielding in accelerator facilities. PHITS describes physical phenomena induced by radiation as combination of transport and collision processes. The collision process including nuclear reactions is simulated by the three-step calculation: a generation of a reaction, pre-equilibrium, and compound processes. In the simulation, many physics models are used. This report explains roles of the models in PHITS and shows their developments we recently performed.

Journal Articles

Cutting-edge studies on nuclear data for continuous and emerging need, 6; Processing and validation of nuclear data

Tada, Kenichi; Kosako, Kazuaki*; Yokoyama, Kenji; Konno, Chikara

Nippon Genshiryoku Gakkai-Shi, 60(3), p.168 - 172, 2018/03

The neutronics calculation codes cannot treat the evaluated nuclear data file directly. The nuclear data processing is required to use the nuclear data file in the neutronics calculation codes. The nuclear data processing is not just a converter but also many processes to evaluate the physical values for the neutronics calculation codes. In this paper, we describe the overview of the nuclear data processing and validation of the nuclear data.

Journal Articles

Another important piece; One point burnup calculation code as a Killer Application

Suyama, Kenya; Yokoyama, Kenji

Kaku Deta Nyusu (Internet), (119), p.38 - 47, 2018/02

We have developed numerous neutronics calculation codes in Japan. However, development of the one-point burnup calculation code which replaces the still widely used ORIGEN2 code has not been successful. The one point burnup code is indispensable to evaluate the characteristics of the used nuclear fuel increasing in Japan, and it uses all evaluated nuclear data including the fission yield and decay data as well as cross section data. It means that it could be the Killer Application in the field of the nuclear data and neutronics code. This report describes the necessity of the one point burnup calculation code development in Japan and required function and performance which have been considered by authors.

Journal Articles

Status of the JENDL project

Iwamoto, Osamu; Shibata, Keiichi; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Ichihara, Akira; Nakayama, Shinsuke

EPJ Web of Conferences (Internet), 146, p.02005_1 - 02005_6, 2017/09

 Times Cited Count:0 Percentile:100

Journal Articles

Prospective features for integration of nuclear forensics capability in national framework

Tamai, Hiroshi; Okubo, Ayako; Kimura, Yoshiki; Shinohara, Nobuo; Tazaki, Makiko; Shimizu, Ryo; Suda, Kazunori; Tomikawa, Hirofumi

Proceedings of INMM 58th Annual Meeting (Internet), 6 Pages, 2017/07

Nuclear forensics is a technical measure to analyse and collate samples of illegally used nuclear materials, etc., to clarify their origins, routes, etc. and contribute to criminal identifications. Close collaboration with police and judicial organizations is essential. The national response framework is being built up with international cooperation. Discussions on promoting technical capability and regional cooperation are presented.

Journal Articles

Atomic Energy Society of Japan 2017 Annual Meeting, joint session of "sigma advisory committee", "subcommittee on nuclear data" and "subcommittee on reactor physics"; Current status and future perspective of the Verification and Validation (V&V) of JENDL and neutronics calculation codes by use of the benchmark problems and integral experiments, 2; International benchmarks of OECD/NEA in the field of the neutronics calculation

Suyama, Kenya

Kaku Deta Nyusu (Internet), (117), p.5 - 14, 2017/06

The benchmark calculation is one of the main activities of the Nuclear Science Committee under the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA/NSC). The international benchmark relatively frequently means the benchmark activity carried out by the NEA. In this manuscript, the author discusses the significance of the international benchmark by describing (i) the current status of the benchmark in the field of the reactor physics conducted by the OECD/NEA/NSC, (ii) revision of the neutronics calculation code system to reflect the results of the benchmark, (iii) the benchmark calculation as the asset for the future research and development, (iv) examples of the benchmark calculation based on the experimental data, and (v) how to propose the benchmark in the OECD/NEA/NSC.

JAEA Reports

Study on nuclear analysis method for high temperature gas-cooled reactor and its nuclear design (Thesis)

Goto, Minoru

JAEA-Review 2014-058, 103 Pages, 2015/03

JAEA-Review-2014-058.pdf:22.36MB

The following issues were investigated using experimental data of HTTR, which is a Japan's HTGR with 30 MW thermal power. (1)Applicability of nuclear data libraries to nuclear analysis for HTGR, (2) Applicability of the improved nuclear analysis method for HTGR, (3) Effectiveness of a rod-type burnable poison on HTGR reactivity control. Using these investigation results, a nuclear design of a small-sized HTGR with 50 MW thermal power (HTR50S) was performed. In the nuclear design of HTR50S, we challenged to decrease the number of the fuel enrichments and to increase the power density compared with HTTR. As a result, the nuclear design was completed successfully by reducing the number of the fuel enrichment to only three from twelve of HTTR and increasing the power density by 1.4 times of HTTR.

Journal Articles

Practical integrated simulation systems for coupled numerical simulations in parallel

Hazama, Osamu; Guo, Z.

Proceedings of International Conference on Supercomputing in Nuclear Applications (SNA 2003) (CD-ROM), p.119 - 120, 2003/09

In order for the numerical simulations to reflect textquotedblleft real-worldtextquotedblright phenomena and occurrences, incorporation of multidisciplinary and multi-physics simulations considering various physical models and factors are becoming essential. However, there still exist many obstacles which inhibit such numerical simulations. For example, it is still difficult in many instances to develop satisfactory software packages which allow for such coupled simulations and such simulations will require more computational resources. A precise multi-physics simulation today will require parallel processing which again makes it a complicated process. Under the international cooperative efforts between CCSE/JAERI and Fraunhofer SCAI, a German institute, a library called the MpCCI, or Mesh-based Parallel Code Coupling Interface, has been implemented together with a library called STAMPI to couple two existing codes to develop an textquotedblleft integrated numerical simulation systemtextquotedblright intended for meta-computing environments.

Journal Articles

Development of an integrated numerical simulation infrastructure for fluid-structure coupled problems

Hazama, Osamu; Guo, Z.

Keisan Kogaku Koenkai Rombunshu, 8(2), p.759 - 760, 2003/05

no abstracts in English

Journal Articles

Recent changes of JT-60 data processing system

Matsuda, Toshiaki; Tsugita, Tomonori; Oshima, Takayuki; Sakata, Shinya; Sato, Minoru; Iwasaki, Keita*

Fusion Engineering and Design, 60(3), p.235 - 239, 2002/06

 Times Cited Count:5 Percentile:62.04

no abstracts in English

Journal Articles

PARCEL; Linear equation solvers

Yamada, Susumu; Shimizu, Futoshi; Kaji, Yoshiyuki; Kaburaki, Hideo

Keisan Kogaku Koenkai Rombunshu, 7(1), p.167 - 170, 2002/05

no abstracts in English

Journal Articles

Revised burnup code system SWAT; Description and validation using postirradiation examination data

Suyama, Kenya; Mochizuki, Hiroki*; Kiyosumi, Takehide*

Nuclear Technology, 138(2), p.97 - 110, 2002/05

 Times Cited Count:20 Percentile:20.3(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Activity report of Japanese Nuclear Data Committee in period of April 1999 to March 2001

Japanese Nuclear Data Committee

Nippon Genshiryoku Gakkai-Shi, 44(1), p.106 - 114, 2002/01

no abstracts in English

JAEA Reports

JENDL FP decay data file 2000

Katakura, Junichi; Yoshida, Tadashi*; Oyamatsu, Kazuhiro*; Tachibana, Takahiro*

JAERI 1343, 79 Pages, 2001/07

JAERI-1343.pdf:4.94MB

no abstracts in English

Journal Articles

Implementation of emergency system for radioactive source term estimation on global meta-computing environment and its real-time visualization

Muramatsu, Kazuhiro; Imamura, Toshiyuki; Kitabata, Hideyuki; Kaneko, Isamu; Takemiya, Hiroshi*; Hasegawa, Yukihiro*; Yamagishi, Nobuhiro*; Hirayama, Toshio

Keisan Kogaku Koenkai Rombunshu, 6(1), p.241 - 244, 2001/05

no abstracts in English

Journal Articles

Construction of integration system with MpCCI for heterogeneous computational methods

Guo, Z.; Onishi, Ryoichi*; Kimura, Toshiya*; Hirayama, Toshio

Proceedings of International Conference on Computational Engineering & Science (ICES 2001) (CD-ROM), 6 Pages, 2001/00

no abstracts in English

JAEA Reports

ORIGEN2 libraries based on JENDL-3.2 for LWR-MOX Fuels

Suyama, Kenya; Onoue, Masaaki*; Matsumoto, Hideki*; Sasahara, Akihiro*; Katakura, Junichi

JAERI-Data/Code 2000-036, 35 Pages, 2000/11

JAERI-Data-Code-2000-036.pdf:1.19MB

no abstracts in English

Journal Articles

MCNP高温ライブラリーの信頼性チェックの考え方

Sakurai, Kiyoshi; Maekawa, Fujio; Yamamoto, Toshihiro; Mori, Takamasa; Naito, Yoshitaka*

Kaku Deta Nyusu (Internet), (66), p.91 - 92, 2000/06

no abstracts in English

81 (Records 1-20 displayed on this page)