TY - GEN
T1 - Preliminary core and fuel design of bwr with multi-Axial fuel shuffling
AU - Tasaki, Yudai
AU - Yamaji, Akifumi
N1 - Publisher Copyright:
© 2018 American Nuclear Society. All rights reserved.
PY - 2018
Y1 - 2018
N2 - The concept of "multi-Axial fuel shuffling" has been recently proposed for a high breeding core design of supercritical-water cooled reactor. In this study, the same design principle is applied to boiling water reactor (BWR) condition. The results show that the Compound System Doubling Time (CSDT) can be reduced by increasing fuel batch number of the upper blanket layer, but more investigations may be necessary to consider further improvement of the core breeding performance. Moreover, fuel performance of the axially heterogeneous core has been evaluated with the power history obtained by the core calculations and modified FEMAXI-7 code which consider two different types of pellets (i.e., MOX and depleted uranium) within one fuel rod. The analysis results indicate that uncertainty in thermal conductivity of MOX pellets may be important in evaluating the peak pellet temperature, while relatively large plenum volume may be required at the bottom of the fuel rod to accommodate the large amount of fission gas release. Another potential design issue may be the cladding outer diameter increase by the MOX pellet swelling, which may have significant influence on evaluation of MCHFR.
AB - The concept of "multi-Axial fuel shuffling" has been recently proposed for a high breeding core design of supercritical-water cooled reactor. In this study, the same design principle is applied to boiling water reactor (BWR) condition. The results show that the Compound System Doubling Time (CSDT) can be reduced by increasing fuel batch number of the upper blanket layer, but more investigations may be necessary to consider further improvement of the core breeding performance. Moreover, fuel performance of the axially heterogeneous core has been evaluated with the power history obtained by the core calculations and modified FEMAXI-7 code which consider two different types of pellets (i.e., MOX and depleted uranium) within one fuel rod. The analysis results indicate that uncertainty in thermal conductivity of MOX pellets may be important in evaluating the peak pellet temperature, while relatively large plenum volume may be required at the bottom of the fuel rod to accommodate the large amount of fission gas release. Another potential design issue may be the cladding outer diameter increase by the MOX pellet swelling, which may have significant influence on evaluation of MCHFR.
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M3 - Conference contribution
AN - SCOPUS:85050081617
T3 - Proceedings of the 2018 International Congress on Advances in Nuclear Power Plants, ICAPP 2018
SP - 1144
EP - 1152
BT - Proceedings of the 2018 International Congress on Advances in Nuclear Power Plants, ICAPP 2018
PB - American Nuclear Society
T2 - 2018 International Congress on Advances in Nuclear Power Plants, ICAPP 2018
Y2 - 8 April 2018 through 11 April 2018
ER -