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Kato, Masato
Encyclopedia of Nuclear Energy, Vol.2, p.298 - 307, 2021/00
Sato, Hiroyuki; Aoki, Takeshi; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 10 Pages, 2020/08
The present study aims to propose a guidance that facilitates to determine fuel design limits of commercial HTGR on the basis of licensing experience through the HTTR construction. The guidance consists of a set of FOMs and a process to determine their evaluation criteria. The FOMs are firstly identified to satisfy safety requirements and a basic concept of safety guides established in a special committee under the AESJ with the support of the Research Association of High Temperature Gas Cooled Reactor Plant. The development process for the evaluation criteria takes into account not only the top-level regulatory criteria but also design dependent constraints including the performance of fission product containment in physical barriers other than fuel, fuel qualification criteria, design specifications of an instrumentation and control system. As a result, a comprehensive and transparent procedure for designers of prismatic-type commercial HTGR has been developed.
gas cooled reactor; ; Mizuta, Shunji
JNC TN9400 2000-040, 41 Pages, 2000/03
JNC-TN9400-2000-040.pdf:0.85MB
The corrosion behavior of ferritic stainless steels applied to core components under C0 gas environment was investigated in order to be helpful to fuel design in C0 gas cooled reactor as the feasibility study for fast breeder reactor. The dependence of the corrosion behavior, before a breakaway occurs, on C0 gas temperature, Si and Cr contents of ferritic steels was determined quantitatively. The following correlations to calculate the metal loss thickness was established. X = 4.4w w = √(k
t) k = 
exp( - 5.45[Si]) exp( - 1.09[Cr]) exp( - 11253/T) = 1.65 10
4.40 10
X : metal loss thickness[
ml, w : corrosion weight gain [mg/cm
] k : parabola constant [(mg/cm)/hr], t : time [hr], : constant [Si] : Si content[wt.%], [Cr] : Cr content [wt.%], T : temperature [K]
; 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 7415, 124 Pages, 1977/12
no abstracts in English
; ; ;
JAERI-M 5423, 10 Pages, 1973/10
no abstracts in English
; ;
Genshiryoku Kogyo, 19(2), p.1 - 6, 1973/02
no abstracts in English
; ; ; ; ;
JAERI-M 4881, 297 Pages, 1972/07
no abstracts in English
Goto, Minoru; Ueta, Shohei; Aihara, Jun; Fukaya, Yuji; Inaba, Yoshitomo; Tachibana, Yukio; Kunitomi, Kazuhiko; Okamoto, Koji*
no journal, ,
In this study, a feasibility of the coated fuel particle for a plutonium burner High Temperature Gas-cooled Reactor (HTGR) is evaluated by calculating the inner pressure. Additionally, a feasibility of the reactor core is also evaluated by calculating the nuclear characteristics and the fuel temperature. This study is started in 2014 and will be completed in 2017. In 2014, the calculation codes and the input files for the evaluations were prepared.
Ota, Hirokazu*; Ohgama, Kazuya; Ogata, Takanari*; Ikusawa, Yoshihisa; Oki, Shigeo
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.
Tasaki, Yudai; Yamaji, Akifumi*; Amaya, Masaki
no journal, ,
In breeding core designs with light water, tight lattice fuel bundle design in which the coolant flow area is small is adopted to prevent the neutron moderating. Additionally, the core often consists of MOX fuel and blanket fuel, which aims to irradiate depleted uranium effectively. In preceding study, the concept of "Multi-axial fuel shuffling" has been proposed for a higher breeding core design of supercritical-water cooled reactor (SCWR), in which the core consists of multiple layers of MOX fuels and blanket fuels with independent fuel shuffling for the upper blanket layer where coolant density is lowest. As a result, the SCWR with multi-axial fuel shuffling has shown improvement of breeding performance. The same principle may be applied to BWR, since the coolant density gets low due to developing void fraction. However, the fuel rod included such a core design has two kinds of fuel pellets, and MOX fuel parts tend to get high power peaking. Therefore, it is necessary to investigate and mitigate the fuel maximum temperature and the shear stress of the boundary between MOX and blanket fuel parts which may occur by the difference of PCMI characteristics of two fuel parts. Moreover, it is possible that the cladding outer diameter change especially in MOX fuel parts may impact on the thermal-hydraulics, because the gap between rods is narrow owing to the tight lattice fuel bundle design. This study has shown the improvement of breeding performance of BWR with multi-axial fuel shuffling, and the fuel design which mitigates the above design issues. The cladding outer diameter change doesn't impact on critical heat flux ratio mostly, but depends on pressure drop of the flow channel. Therefore, this result suggests a design issue with respect to the core flow distribution.
