Core and safety design for France-Japan common concept on sodium-cooled fast reactor

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

https://doi.org/10.1051/epjn/2022040

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.

Coolability evaluation of debris bed on core catcher in a sodium-cooled fast reactor

Matsuo, Eiji*; Sasa, Kyohei*; Koyama, Kazuya*; Yamano, Hidemasa; Kubo, Shigenobu; Hourcade, E.*; Bertrand, F.*; Marie, N.*; Bachrata, A.*; Dirat, J. F.*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05

Discharged molten-fuel from the core during Core Disruptive Accident (CDA) could become solidified particle debris by fuel-coolant interaction in the lower sodium plenum, and then the debris could form a bed on a core catcher located at the bottom of the reactor vessel. Coolability evaluations for the debris bed are necessary for the design of the core catcher. The purpose of this study is to evaluate the coolability of the debris bed on the core catcher for the ASTRID design. For this purpose, as a first step, the coolability calculations of the debris beds formed both in short term and later phase have been performed by modeling only the debris bed itself. Thus, details of core catcher design and decay heat removal system are not described in this paper. In all the calculations, coolant temperature around the debris bed is a parameter. The calculation tool is the debris bed module implemented into a one-dimensional plant dynamics code, Super-COPD. The evaluations have shown that the debris beds formed both in short term and later phase are coolable by the design which secures sufficient coolant flow around the core catcher located in the cold pool.

Photon transmittance data collection of building materials for evaluating radiation protection capability of buildings (Contract research)

Ishizaki, Azusa; Futemma, Akira; Takubo, Kazuya*; Nakanishi, Chika*; Munakata, Masahiro

JAEA-Data/Code 2018-022, 20 Pages, 2019/03

JAEA-Data-Code-2018-022.pdf:2.05MB

If a nuclear disaster occurs, we may evacuate indoor escape facilities and buildings such as houses as avoid extra exposure doses. In order to evaluate exposure doses, it is necessary to estimate shielding capabilities of the building materials constituting the sheltering facility. Therefore, photon irradiation tests with three kinds of photon energy were carried out for Japanese familiar building materials in Japan, and photon transmittance of each building material is acquired and summarized. As a result, it was found that the shielding capabilities of composite walls and roofs which are widely used in a tree structure and a steel structure were relatively low. And, difference of materials used for composite walls and roofs resulted in a difference in shielding capabilities. For example, in the case of composite walls, compared with the photon transmittance of wall with ceramic-based siding materials, those of wall with lightweight concrete were lower. Furthermore, photon transmittance was also measured for building materials with relatively low shielding performance added shielding materials as additional measures to enhance shielding capabilities.

Status report of the chopper spectrometer 4SEASONS

Kajimoto, Ryoichi; Nakamura, Mitsutaka; Inamura, Yasuhiro; Kamazawa, Kazuya*; Ikeuchi, Kazuhiko*; Iida, Kazuki*; Ishikado, Motoyuki*; Murai, Naoki; Kira, Hiroshi*; Nakatani, Takeshi; et al.

Journal of Physics; Conference Series, 1021(1), p.012030_1 - 012030_6, 2018/06

BB2016-1727.pdf:0.39MB

https://doi.org/10.1088/1742-6596/1021/1/012030

Evaluation for influence of new volcanic eruption on geological disposal site

Shimada, Taro; Takeda, Seiji; Sakai, Ryutaro*; Takubo, Kazuya; Tanaka, Tadao

MRS Advances (Internet), 1(61), p.4081 - 4086, 2016/00

https://doi.org/10.1557/adv.2017.193

Volcanic eruption which may affect geological disposal system directly depends on the regional location in Japan. It is required that the disposal site should be located far from existing volcanos. However, there are regions where it is impossible to exclude the possibility of appearance of new volcanic activity on the site even if the site is located far from existing volcanos. In order to identify the influence of volcanic eruption at disposal site to public if it occurs public exposure doses were evaluated based on the two scenarios considering types of eruption at new volcanic activity in Japan. One is the exposure by tephra widespread by Strombolian eruption and deposited on the ground surface, including radionuclides from vitrified waste forms after a volcanic conduit penetrated disposal galleries. The other is that by waste forms appeared at the surface by Merapi type pyroclastic flow. Exposure doses of the residents living on the tephra do not exceed 1mSv/y even when the eruption occurs at 1,000 years after closure of disposal site. Dose rate for the volcanic researchers temporarily approaching waste forms becomes less than 1mSv/h when the eruption occurs 100,000 years after. It indicated that attention should be paid to the impact by Merapi type pyroclastic flow on researchers approaching waste forms appeared rather than that by Strombolian eruption on residents living on the tephra widespread.

Development of divertor simulation research in the GAMMA 10/PDX tandem mirror

Nakashima, Yosuke*; Sakamoto, Mizuki*; Yoshikawa, Masayuki*; Oki, Kensuke*; Takeda, Hisahito*; Ichimura, Kazuya*; Hosoi, Katsuhiro*; Hirata, Mafumi*; Ichimura, Makoto*; Ikezoe, Ryuya*; et al.

Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 8 Pages, 2014/10

Comparison and sensitivity analysis of the core characteristics of a sodium-cooled fast breeder reactor with 750 MWe output evaluated by JENDL-4.0 and ADJ2000R

Ohgama, Kazuya; Oki, Shigeo; Sugino, Kazuteru; Okubo, Tsutomu

Journal of Nuclear Science and Technology, 51(4), p.558 - 567, 2014/04

https://doi.org/10.1080/00223131.2014.882800

Core characteristics of a sodium-cooled fast breeder reactor (FBR) with 750 MWe output using highly decontaminated uranium and plutonium and highly minor-actinide-containing compositions were evaluated using the fast reactor cross-section set generated by the new Japanese nuclear data library JENDL-4.0. The core characteristics were compared with those obtained using the unified cross-section set ADJ2000R in order to investigate the differences between both the results. The effects on the core characteristics caused by the differences in the nuclear data of important reactions and nuclides in the cross-section sets were analyzed by a burnup sensitivity analysis. It was confirmed that adopting JENDL-4.0 to the FBR core design improves the breeding ratio, the burnup reactivity, and the reactivity control balance, because of the differences in the capture cross-sections of U-238 and Pu-239 of both the libraries. The difference in the sodium void reactivity evaluated with both the libraries was less than 1% because the increase caused by the differences in the elastic scattering cross-sections of sodium, the inelastic scattering cross-section, and the -average value of U-238 was practically cancelled out by the decrease caused by the differences in the capture cross-sections of Pu-239, the inelastic scattering cross-section of iron, and the capture cross- sections of Am-241.

Numerical simulation of melt-down behavior in SFR severe accidents by the MUTRAN code

Kubota, Ryuzaburo*; Yamada, Yumi*; Koyama, Kazuya*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kubo, Shigenobu; Suzuki, Toru; Tobita, Yoshiharu

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12

This paper describes a melt-down event progression revealed by a numerical simulation in the protected loss of heat sink (PLOHS) event for Japan Sodium-cooled Fast Reactor (JSFR). A multi-component multi-field computer code, MUTRAN, has been applied in order to simulate complicated core material motions and associated heat-transfer phenomena among the materials in a degradation core. The analyses with MUTRAN covered core degradation behaviors from the intact geometry and addressed the two initial states: one was the core without the coolant as the leakage type, and the other was the core covered by the coolant only up to the top of the fissile fuel as the boiling type. The analyses revealed representative event progression.

Fast Breeder Reactor core design study using JENDL-4.0

Ohgama, Kazuya; Oki, Shigeo; Okubo, Tsutomu

JAEA-Conf 2012-001, p.21 - 26, 2012/07

https://jopss.jaea.go.jp/pdfdata/JAEA-Conf-2012-001.pdf#page=31

The Result of a wall failure in-pile experiment under the EAGLE project

Konishi, Kensuke; Toyooka, Junichi; Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vurim, A. D.*; Gaidaichuk, V. A.*; Pakhnits, A. V.*; et al.

Nuclear Engineering and Design, 237(22), p.2165 - 2174, 2007/11

https://doi.org/10.1016/j.nucengdes.2007.03.012

The WF (Wall Failure) test of the EAGLE program, in which 2kg of uranium dioxide fuel-pins were melted by nuclear heating, was successfully conducted in the IGR of NNC/Kazakhstan. In this test, a 3mm-thick stainless steel (SS) wall structure was placed between fuel pins and a 10mm-thick sodium-filled channel (sodium gap). During the transient, fuel pins were heated, which led to the formation of a fuel-steel mixture pool. Under the transient nuclear heating condition, the SS wall was strongly heated by the molten pool, leading to wall failure. The time needed for fuel penetration into the sodium-filled gap was very short (less than 1 second after the pool formation). The result suggests that molten core materials formed in hypothetical LMFBR core disruptive accidents have a certain potential to destroy SS-wall boundaries early in the accident phase, thereby providing fuel escape paths from the core region. The early establishment of such fuel escape paths is regarded as a favorable characteristic in eliminating the possibility of severe re-criticality events.