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Takano, Kazuya; Oki, Shigeo; Ozawa, Takayuki; Yamano, Hidemasa; Kubo, Shigenobu; Ogura, Masashi*; Yamada, Yumi*; Koyama, Kazuya*; Kurita, Koichi*; Costes, L.*; et al.
EPJ Nuclear Sciences & Technologies (Internet), 8, p.35_1 - 35_9, 2022/12
The France and Japan teams have carried out collaborative works to have common technical views regarding a sodium-cooled fast reactor concept. Japan has studied the feasibility of an enhanced high burnup low-void effect (CFV) core and fuel using oxide dispersion-strengthened steel cladding in ASTRID 600. Regarding passive shutdown capabilities, Japan team has performed a preliminary numerical analysis for ASTRID 600 using a complementary safety device, called a self-actuated shutdown system (SASS), one of the safety approaches of Japan. The mitigation measures of ASTRID 600 against a severe accident, such as a core catcher, molten corium discharge assembly, and the sodium void reactivity features of the CFV core, are promising to achieve in-vessel retention for both countries. The common design concept based on ASTRID 600 is feasible to demonstrate the SFR core and safety technologies for both countries.
Kato, Atsushi; Yamamoto, Tomohiko; Ando, Masato; Chikazawa, Yoshitaka; Murakami, Hisatomo*; Oyama, Kazuhiro*; Kaneko, Fumiaki*; Higurashi, Koichi*; Chanteclair, F.*; Chenaud, M.-S.*; et al.
EPJ Nuclear Sciences & Technologies (Internet), 8, p.11_1 - 11_10, 2022/06
This paper provides an overview of plant system studies to establish a common technical view for Sodium-cooled Fast Reactor concept between France and Japan based on ASTRID600 and the new concept with downsized output called ASTRID150. One of important issues on a reactor structure design is to enhance seismic resistance to be tolerable against strong earthquake such that postulated in Japan. A concept of High Frequency Design is shared, and the design options related to HFD have been examined and design recommendations are established. In addition, this paper include results of studies for a steam generator, a decay heat removal system, a fuel handling system and a containment vessel.
Okamura, Tomohiro*; Katano, Ryota; Oizumi, Akito; Nishihara, Kenji; Nakase, Masahiko*; Asano, Hidekazu*; Takeshita, Kenji*
EPJ Nuclear Sciences & Technologies (Internet), 7, p.19_1 - 19_13, 2021/11
Nuclear Material Balance code version 4.0 (NMB4.0) has been developed through collaborative R&D between Tokyo Institute of Technology and JAEA. Conventional nuclear fuel cycle simulation codes mainly analyze actinides and are specialized for front-end mass balance analysis. However, quantitative back-end simulation has recently become necessary for considering R&D strategies and sustainable nuclear energy utilization. Therefore, NMB4.0 was developed to realize the integrated nuclear fuel cycle simulation from front- to back-end. There are three technical features in NMB4.0: 179 nuclides are tracked, more than any other code, throughout the nuclear fuel cycle; the Okamura explicit method is implemented, which contributes to reducing the numerical cost while maintaining the accuracy of depletion calculations on nuclides with a shorter half-life; and flexibility of back-end simulation is achieved. The main objective of this paper is to show the newly developed functions, made for integrated back-end simulation, and verify NMB4.0 through a benchmark study to show the computational performance.
for MYRRHA using a Monte Carlo techniqueIwamoto, Hiroki; Stankovskiy, A.*; Fiorito, L.*; Van den Eynde, G.*
EPJ Nuclear Sciences & Technologies (Internet), 4, p.42_1 - 42_7, 2018/11
This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo transport code MCNP. The sensitivities are calculated by the modified 
-ratio method proposed by Chiba. Comparing the sensitivities obtained with different scaling factors 
introduced by Chiba shows that a value of 
is the most suitable for the uncertainty quantification of . Using the calculated sensitivities and the JENDL-4.0u covariance data, the uncertainties for the critical and subcritical cores are determined to be 2.2 
0.2% and 2.0 0.2%, respectively, which are dominated by delayed neutron yield of 
Pu and 
U.
Watanabe, So; Sano, Yuichi; Nomura, Kazunori; Koma, Yoshikazu; Okamoto, Yoshihiro
EPJ Nuclear Sciences & Technologies (Internet), 1, p.9_1 - 9_8, 2015/12
