Development of the unified cross-section set ADJ2017R

Yokoyama, Kenji; Maruyama, Shuhei; Taninaka, Hiroshi; Oki, Shigeo

JAEA-Data/Code 2021-019, 115 Pages, 2022/03

JAEA-Data-Code-2021-019.pdf:6.21MB
JAEA-Data-Code-2021-019-appendix(CD-ROM).zip:435.94MB

In JAEA, several versions of unified cross-section set for fast reactors have been developed so far; we have developed a new unified cross-section set ADJ2017R, which is an improved version of the unified cross-section setADJ2017 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses in reactor core design via the cross-section adjustment methodology; the values are stored in the standard database for FBR core design. In the methodology, the cross-section set is adjusted by integrating the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. ADJ2017R basically has the same performance as ADJ2017, but we conducted an additional investigation on ADJ2017 and revised the following two points. The first is to unify the evaluation method of the correlation coefficient of uncertainty caused by experiments (hereinafter referred to as the experimental correlation coefficient). Because it was found that the common uncertainty used in the evaluation of the experimental correlation coefficient was evaluated by two different methods, the experimental correlation coefficients were revised for all experimental data, and the evaluation method was unified. The second is the review of the integral experiment data used for the cross-section adjustment calculation. It was found that one of the experimental values of composition ratio after irradiation of the Am-243 sample has a problem in uncertainty evaluation because its experimental uncertainty is extremely small compared to the others. The cross-section adjustment calculation was, therefore, redone by excluding the experimental value. In the creation of ADJ2017, a total of 719 data sets were analyzed and evaluated, and eventually adopted 620 integral experimental data sets. In contrast, a total of 61

A Design study on a mixed oxide fuel sodium-cooled fast reactor core partially loading highly concentrated MA-containing metal fuel

Ohgama, Kazuya; Ota, Hirokazu*; Oki, Shigeo; Iizuka, Masatoshi*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05

Tradeoff analysis of metal-fueled fast reactor design concepts

Stauff, N. E.*; Ohgama, Kazuya; Aliberti, G.*; Oki, Shigeo; Kim, T. K.*

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

Development of the versatile reactor analysis code system, MARBLE2

Yokoyama, Kenji; Jin, Tomoyuki; Hirai, Yasushi*; Hazama, Taira

JAEA-Data/Code 2015-009, 120 Pages, 2015/07

JAEA-Data-Code-2015-009.pdf:1.93MB

The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added inMARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, -ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability.

Proceedings of the 11th International Workshop on Ceramic Breeder Blanket Interactions; December 15 - 17, 2003, Tokyo, Japan

Enoeda, Mikio

JAERI-Conf 2004-012, 237 Pages, 2004/07

JAERI-Conf-2004-012.pdf:44.1MB

This report is the Proceedings of "the Eleventh International Workshop on Ceramic Breeder Blanket Interactions" which was held as a workshop on ceramic breeders Under the IEA Implementing Agreement on the Nuclear Technology of Fusion Reactors, and the Japan-US Fusion Collaboration Framework. In the workshop, information exchange was performed for designs of solid breeder blankets and test blankets in EU, Russia and Japan, recent results of irradiation tests, HICU, EXOTIC-8 and the irradiation tests by IVV-2M, modeling study on tritium release behavior of LiTiO and other breeders, fabrication technology developments and characterization of the LiTiO and LiSiO pebbles, research on measurements and modeling of thermo-mechanical behaviors of LiTiO and LiSiO pebbles, and interfacing issues, such as, fabrication technology for blanket box structure, neutronics experiments of blanket mockups by fusion neutron source and tritium recovery system.

Current status of the AGS spallation target experiment

Nakashima, Hiroshi; Takada, Hiroshi; Kasugai, Yoshimi; Meigo, Shinichiro; Maekawa, Fujio; Kai, Tetsuya; Konno, Chikara; Ikeda, Yujiro; Oyama, Yukio; Watanabe, Noboru; et al.

Proceedings of 6th Meeting of the Task Force on Shielding Aspects of Accelerators, Targets and Irradiation Facilities (SATIF-6), (OECD/NEA No.3828), p.27 - 36, 2004/00

no abstracts in English

Core neutronics module and database access module for intelligent reactor design system (IRDS)

Kugo, Teruhiko; Tsuchihashi, Keiichiro*; Nakakawa, Masayuki; Ido, Masaru*

JAERI-Data/Code 2000-011, p.138 - 0, 2000/02

JAERI-Data-Code-2000-011.pdf:7.41MB

no abstracts in English

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

Kugo, Teruhiko; Nakakawa, Masayuki

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

https://doi.org/10.3327/jnst.36.332

no abstracts in English

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

Kugo, Teruhiko; Nakakawa, Masayuki;

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

https://doi.org/10.3327/jnst.34.760

no abstracts in English

Design and techniques for fusion blanket neutronics experiments using an accelerator-based deuterium-tritium neutron source

Oyama, Yukio; Konno, Chikara; Ikeda, Yujiro; Maekawa, Fujio; Maekawa, Hiroshi; ; ; Nakamura, Tomoo; M.A.Abdou*; Bennett, E. F.*; et al.

Fusion Technology, 28(1), p.56 - 73, 1995/08

no abstracts in English