Double diffusive dissolution model of UO pellet in molten Zr cladding

Ito, Ayumi*; Yamashita, Susumu; Tasaki, Yudai; Kakiuchi, Kazuo; Kobayashi, Yoshinao*

Journal of Nuclear Science and Technology, 60(4), p.450 - 459, 2023/04

https://doi.org/10.1080/00223131.2022.2114956

Improvement of intragranular fission gas behavior model for fuel performance code FEMAXI-8

Udagawa, Yutaka; Tasaki, Yudai

JAEA-Data/Code 2021-007, 56 Pages, 2021/07

JAEA-Data-Code-2021-007.pdf:5.05MB

Japan Atomic Energy Agency (JAEA) has released FEMAXI-8 in 2019 as the latest version of the fuel performance code FEMAXI, which has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly normal operation conditions and anticipated transient conditions. This report summarizes a newly developed model to analyze intragranular fission gas behaviors considering size distribution of gas bubbles and their dynamics. While the intragranular fission gas behavior models implemented in the previous FEMAXI versions have supported only treating single bubble size for a given fuel element, the new model supports multiple gas groups according to their size and treats their dynamic behaviors, making the code more versatile. The model was first implemented as a general module that takes arbitrary number of bubble groups and spatial mesh division for a given fuel grain system. An interface module to connect the model to FEMAXI-8 was then developed so that it works as a 2 bubble groups model, which is the minimum configuration of the multi-grouped model to be operated with FEMAXI-8 at the minimum calculation cost. FEMAXI-8 with the new intragranular model was subjected to a systematic validation calculation against 144 irradiation test cases and recommended values for model parameters were determined so that the code makes reasonable predictions in terms of fuel center temperature, fission gas release, etc. under steady-state and power ramp conditions.

Core and fuel design of BWR with multi-axial fuel shuffling

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.

Preliminary study of fission gas release model for MOX fuels with treatment of heterogeneous microstructure

Tasaki, Yudai; Udagawa, Yutaka; Amaya, Masaki

no journal, , 

no abstracts in English

Overview of research activities of the Fuel Safety Research Group

Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Tasaki, Yudai; Udagawa, Yutaka

no journal, , 

no abstracts in English