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Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08
An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.
Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In this paper, focusing on the ARKADIA-Design as a part of it, the progress in the development of optimization processes on the representative problems in the fields of the core design, the plant structure design, and the maintenance schedule planning are introduced.
Yokoyama, Kenji; Sugino, Kazuteru; Ishikawa, Makoto; Maruyama, Shuhei; Nagaya, Yasunobu; Numata, Kazuyuki*; Jin, Tomoyuki*
JAEA-Research 2018-011, 556 Pages, 2019/03
JAEA-Research-2018-011.pdf:19.53MB
JAEA-Research-2018-011-appendix1(DVD-ROM).zip:433.07MBJAEA-Research-2018-011-appendix2(DVD-ROM).zip:580.12MBJAEA-Research-2018-011-appendix3(DVD-ROM).zip:9.17MB
We have developed a new unified cross-section set ADJ2017, which is an improved version of the unified cross-section set ADJ2010 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; the values are stored in the standard database for FBR core design via the cross-section adjustment methodology, which integrates with the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. The ADJ2017 is based on Japan's latest nuclear data library JENDL-4.0 as in the previous version of ADJ2010, and it incorporates more information on integral experimental data sets related to minor actinides (MAs) and degraded plutonium (Pu). In the creation of ADJ2010, a total of 643 integral experimental data sets were analyzed and evaluated, and 488 of the integral experimental data sets were finally selected to be used for the cross-section adjustment. In contrast, we have evaluated a total of 719 data sets, and eventually adopted 620 integral experimental data sets to create ADJ2017. ADJ2017 shows almost the same performance as ADJ2010 for the main neutronic characteristics of conventional sodium-cooled MOX-fuel fast reactors. In addition, for the neutronic characteristics related to MA and degraded Pu, ADJ2017 improves the C/E values of the integral experimental data sets, and reduces the uncertainty induced by the nuclear data. ADJ2017 is expected to be widely used in the analysis and design research of fast reactors. Moreover, it is expected that the integral experimental data sets used for ADJ2017 can be utilized as a standard database of FBR core design.
Nakano, Yoshihiro; Ishikawa, Nobuyuki; Nakatsuka, Toru; Iwamura, Takamichi
JAERI-Conf 2002-012, 219 Pages, 2002/12
JAERI-Conf-2002-012.pdf:17.4MB
no abstracts in English
Kunitomi, Kazuhiko; Katanishi, Shoji; Shiozawa, Shusaku
Nihon Genshiryoku Gakkai-Shi, 43(11), p.1085 - 1099, 2001/11
no abstracts in English
Odano, Naoteru; Ishida, Toshihisa; Wada, Koji*; Imai, Hiroshi*
JAERI-Research 2001-039, 59 Pages, 2001/07
JAERI-Research-2001-039.pdf:4.47MB
A very small reactor, SCR (Submersible Compact Reactor), whose thermal output is 1250 kW, is an integral-pressurized type reactor to be used as a power source for a scientific research vessel in medium depth region of the Arctic Ocean. Neutronic study has been carried out for design of the SCR core of which could achieve continuous long-term operation without refueling for 10 years considering 50 % of load factor of the core. In the present study, arrangement of fuel rods, 
U enrichment of UO
fuel rods and reflector materials were surveyed. The U enrichment has been determined to be 9.5 wt% to satisfy design criteria. In the present study Be metal was adopted as a reflector material. Reactor physics parameters including reactivity coefficients and power distributions were evaluated for the determined core specifications. Reactor physics parameters related to core safety were also confirmed and the evaluated parameters indicated that the determined core specifications in this study satisfied design conditions.
Okubo, Tsutomu; Takeda, Renzo*; Iwamura, Takamichi
JAERI-Research 2001-021, 84 Pages, 2001/03
JAERI-Research-2001-021.pdf:11.26MB
no abstracts in English
Research Group for Advanced Reactor System; Research Group for Reactor Physics; Research Group for Thermal and Fluid Engineering
JAERI-Research 2000-035, 316 Pages, 2000/09
JAERI-Research-2000-035.pdf:19.81MB
no abstracts in English
Yamashita, Toshiyuki
Nichiro Hatsudenro Nenryo Semmonka Kaigi Hobunshu, p.148 - 151, 1998/00
no abstracts in English
Hunter
PNC TN9410 97-057, 106 Pages, 1997/05
This study was based on a 'pancake' type fast reactor core design of 600 MW(e), which had been optimized for Pu burning with a feed Pu vector appropriate to once-through irradiation of MOX fuel in a PWR. The purpose of the study was to investigate the effects of varying the Pu vector, examining various methods of offsetting the effects of such a change, and finally to produce fuel cycles optimized for the different qualities of Pu vector within the same basic design. In addition to the reference (once-through) Pu vector, two extreme Pu vectors were examined: high quality Pu from military stockpiles; low quality Pu corresponding to the equilibrium point of multiple recycling in a Pu burning fast reactor. Variations in Pu quality were overcome by changing the fuel inventory - replacing some of the fuel by diluent material, and altering the fuel pin size. Using absorber material (
B
C) as diluent improves the rod worth shutdown margin but degrades the Na void and Doppler safety parameters, a non-absorber diluent has the opposite effects, so a mix of the 2 material types was used to optimize the core characteristics. Of the non-absorber diluent materials examined, ZrH gave significantly better performance than all others; 
BC was the second choice for non-absorber diluent, because of its compatibility with BC absorber. It was not possible to accommodate the lower quality (multi-recycled) Pu vector without a significant increase in the fuel pin volume. It was not generally possible, especially with the increased fuel pin size, to achieve positive rod worth shutdown margins - this was overcome by increasing the number of control rods. For the higher quality Pu vectors to maintain ratings within limits, it was necessary to adopt hollow fuel pellets, or else to use the diluent material as an inert matrix in the fuel pellets. It proved possible to accommodate both extremes of Pu vector within a single basic design, maintaining ...
*; *; Fukumura, Nobuo*; *; *; *; *
PNC TN1410 91-063, 239 Pages, 1991/08
no abstracts in English
; Takizuka, Takakazu
Journal of Nuclear Science and Technology, 24(7), p.516 - 525, 1987/07
Times Cited Count：8 Percentile：63.51(Nuclear Science & Technology)no abstracts in English
; ; *; *; *
JAERI-M 82-183, 68 Pages, 1982/12
no abstracts in English
; ; ; ; ; ; ; ; ; Suzuki, Katsuo; et al.
JAERI-M 8064, 255 Pages, 1979/03
no abstracts in English
Moriwaki, Hiroyuki*; Kan, Taro*; Oki, Shigeo
no journal, ,
no abstracts in English
Idomura, Yasuhiro
no journal, ,
The research plan and the current status of fusion plasma code development in the Post-K priority issue 6 "Development of Innovative Clean Energy" [Core Design of Fusion Reactor] are reviewed. This project develops first principles based plasma analysis codes on plasma turbulence phenomena and MHD phenomena, which are taken into account in the design of operation scenarios of fusion reactors. Because of the spatio-temporal scales in ITER core plasmas and the complexity of their physics models including multi-species ions and burning processes, these fusion plasma analysis codes require new simulation techniques. In this talk, we discuss the development of Exa-scale computating technologies such as many-core optimization, and new physics and numerical models to enable long time scale simulations.
Oki, Shigeo; Chikazawa, Yoshitaka; Kubo, Shigenobu; Hibi, Koki*; Kan, Taro*
no journal, ,
no abstracts in English
Goto, Minoru; Inaba, Yoshitomo; Ueta, Shohei; Fukaya, Yuji; Tachibana, Yukio; Okamoto, Koji*
no journal, ,
In addition to the high nuclear proliferation resistance, in order to enhance safety at high burn-up, The Univ. of Tokyo, JAEA, Fuji Electric. and NFI propose in a framework of contracted study to introduce a PuO-YSZ fuel kernel with ZrC coating to the plutonium burner HTGR. In this study, JAEA conducts design of the coated fuel particle and of the reactor core to confirm the feasibility of the plutonium burner HTGR. This paper describes the investigation result of the ZrC layer thickness that enables to absorb all of the free-oxygen emitted from the fuel kernel and the calculation results of nuclear characteristics of the reactor core and fuel temperature during the normal operation condition.
Nagaya, Yasunobu
no journal, ,
We review Monte Carlo capabilities developed recently and Monte Carlo applications to reactor analysis and reactor design. Finally, we describe future prospects of the Monte Carlo methodological development and the applicability of the Monte Carlo methods to reactor design.
