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JAEA Reports

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

Journal Articles

Criticality configuration design methodology applied to the design of fuel debris experiment in the new STACY

Gunji, Satoshi; Tonoike, Kotaro; Clavel, J.-B.*; Duhamel, I.*

Journal of Nuclear Science and Technology, 58(1), p.51 - 61, 2021/01

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

The new critical assembly STACY will be able to contribute to the validation of criticality calculations related to the fuel debris. The experimental core designs are in progress in the frame of JAEA/IRSN collaboration. This paper presents the method applied to optimize the design of the new STACY core to measure the criticality characteristics of pseudo fuel debris that simulated Molten Core Concrete Interaction (MCCI) of the fuel debris. To ensure that a core configuration is relevant for code validation, it is important to evaluate the reactivity worth of the main isotopes of interest and their k$$_{rm eff}$$ sensitivity to their cross sections. In the case of the fuel debris described in this study, especially for the concrete composition, silicon is the nucleus with the highest k$$_{rm eff}$$ sensitivity to the cross section. For this purpose, some parameters of the core configuration, as for example the lattice pitches or the core dimensions, were adjusted using optimization algorithm to find efficiently the optimal core configurations to obtain high sensitivity of silicon capture cross section. Based on these results, realistic series of experiments for fuel debris in the new STACY could be defined to obtain an interesting feedback for the MCCI. This methodology is useful to design other experimental conditions of the new STACY.

Journal Articles

ASTRID nuclear island design; Advances in French-Japanese joint team development of decay heat removal systems

Hourcade, E.*; Curnier, F.*; Mihara, Takatsugu; Farges, B.*; Dirat, J.-F.*; Ide, Akihiro*

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1740 - 1745, 2016/04

In the framework of the French-Japanese agreement signed in 2014, CEA, AREVA NP, JAEA, and MHI/MFBR is jointly performing components design of ASTRID such as Decay Heat Removal Systems (DHRS). This paper is giving highlights of ASTRID DHRS current strategy. Focus is made on operating temperature diversification for in-vessel heat exchanger as well as core catcher coolability by an original features such as heat exchanger located within reactor cold pool, whose design was taken over by Japan team since 2014.

Journal Articles

JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors, 4; Balance of plant

Chikazawa, Yoshitaka; Kato, Atsushi; Nabeshima, Kunihiko; Otaka, Masahiko; Uzawa, Masayuki*; Ikari, Risako*; Iwasaki, Mikinori*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05

Design study and evaluation for SDC and safety SDG on the BOP of the demonstration JSFR including fuel handling system, power supply system, component cooling water system, building arrangement are reported. For the fuel handling system, enhancement of storage cooling system has been investigated adding diversified cooling systems. For the power supply, existing emergency power supply system has been reinforced and alternative emergency power supply system is added. For the component cooling system and air conditioning, requirements and relation between safety grade components are investigated. Additionally for the component cooling system, design impact when adding decay heat removal system by sea water has been investigated. For reactor building, over view of evaluation on the external events and design policy for distributed arrangement is reported. Those design study and evaluation provides background information of SDC and SDG.

Journal Articles

Development of supercritical pressure water cooled solid breeder blanket in JAERI

Akiba, Masato; Ishitsuka, Etsuo; Enoeda, Mikio; Nishitani, Takeo; Konishi, Satoshi

Purazuma, Kaku Yugo Gakkai-Shi, 79(9), p.929 - 934, 2003/09

no abstracts in English

JAEA Reports

Nuclear power plant database internet version

Yamamoto, Nobuo

JAERI-Data/Code 99-023, 65 Pages, 1999/04

JAERI-Data-Code-99-023.pdf:2.6MB

no abstracts in English

Journal Articles

Shielding design of obtain compact marine reactor

;

Journal of Nuclear Science and Technology, 31(6), p.510 - 520, 1994/06

 Times Cited Count:9 Percentile:64.53(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Development of Nuclear Power Plants Database System -Part2- PPD8・User'sGuide

; Ichikawa, Michio

JAERI-M 84-113, 35 Pages, 1984/06

JAERI-M-84-113.pdf:1.35MB

no abstracts in English

JAEA Reports

Data List of Nuclear Power Plants in Japan, 1981 Edition

;

JAERI-M 82-132, 481 Pages, 1982/10

JAERI-M-82-132.pdf:12.63MB

no abstracts in English

JAEA Reports

JAEA Reports

JAEA Reports

JAEA Reports

JAEA Reports

JAEA Reports

JAEA Reports

Design Data and Safety Features of Commercial Nuclear Power Plants, Japan

; ;

JAERI-M 5959, 118 Pages, 1975/01

JAERI-M-5959.pdf:3.71MB

no abstracts in English

JAEA Reports

Conceptual design of a gas cooled tokamak reactor

; ; ; *; ; Tanaka, Kichizo; *;

JAERI-M 5502, 17 Pages, 1973/12

JAERI-M-5502.pdf:0.52MB

no abstracts in English

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