Extension of fuel performance code package FEMAXI; Development and validation of reactivity-initiated accident analysis module RANNS for light water reactor fuels

Tasaki, Yudai; Udagawa, Yutaka

JAEA-Data/Code 2024-012, 76 Pages, 2024/12

JAEA-Data-Code-2024-012.pdf:9.25MB

Japan Atomic Energy Agency (JAEA) has been developing a fuel performance code, FEMAXI, to evaluate the behavior of LWR fuels under normal operation and transient conditions. In March 2019, FEMAXI-8, the first systematically validated and performance evaluated code, was released. Since then, the code has undergone various improvements. In parallel, since the 2000s, JAEA has been developing the RANNS module as a branch for design basis accident (DBA) analysis, with a particular emphasis on computational stability, so that fuel behavior can be tracked even for very steep transients, in this case mainly reactivity-initiated accidents (RIAs). The specific models include boiling heat transfer, fission gas release by grain boundary failure, and cladding failure determination based on fracture mechanics parameters, which are essential for predicting such transient behavior. In this report, prior to the release of RANNS, we present a description of the models for accident behavior analysis, the relationship with FEMAXI-8 in terms of the design and structure of the program, and the results of a large-scale validation using the extensive database of RIA experiments conducted and accumulated by JAEA, to evaluate the overall RIA analysis performance. The code will be made available to users as a packaged FEMAXI/RANNS, enabling them to analyze fuel behavior under various conditions. The model parameter sets determined through the above validation analyses are also presented in this report, and by referring to them, the analysis can be easily performed with almost no change in usability from the previously released FEMAXI-8.

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.

Preliminary analysis on fission gas release of MOX fuel in consideration of the heterogeneous structure

Tasaki, Yudai

no journal, , 

Fuel pellets contained in a fuel rod of a light water reactor release some degree of FP gas, and it raises inner gas pressure of the fuel rod. It is important for fuel design and safety evaluation to estimate fission gas release ratio (FGR), since excessive inner gas pressure may impair fuel integrity. Therefore, JAEA develops fuel performance code FEMAXI for evaluating various fuel behaviors. However, FEMAXI has limited capability for considering FGR in case of evaluating to compare MOX fuels which have different microstructures arised from the difference of production method. Because, FGR model of FEMAXI can apply only one kind of fuel grain. From the above, it is necessary to acquire sufficient understanding of MOX fuel behavior based on experimental data, and FGR modeling has to be improved accordingly. This study modified FGR model so that two kinds of grains with respect to Pu enrichment get applicable and released fission gas values from them are applied weighted average based on volume occupancy ratio of each structures. Upon analyzing two types of MOX fuels, the different homogeneity of microstructure, with and without the modified FGR model, then comparing the experimental data provided by a testing reactor, analytical results of FGR captured tendency of experimental data qualitatively. Hence, it was revealed that modified FGR model is valid for evaluating FGR in consideration of the heterogeneous structure of MOX fuel.

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.

2021 The 5th Summer Seminar on Nuclear Fuel, Materials and Water Chemistry Award Lecture for Young Scientists "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

Development of fuel performance code package FEMAXI/RANNS; Model development for reactivity-initiated accident analysis and systematic validation

Tasaki, Yudai; Udagawa, Yutaka

no journal, , 

no abstracts in English