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Low enriched uranium and thorium fuel utilization under once‐through and offline reprocessing scenarios in small modular molten salt reactor
Author(s) -
Zhu Guifeng,
Zou Yang,
Yan Rui,
Tan Menglu,
Zou Chunyan,
Kang Xuzhong,
Chen Jingen,
Guo Wei,
Dai Ye
Publication year - 2019
Publication title -
international journal of energy research
Language(s) - English
Resource type - Journals
SCImago Journal Rank - 0.808
H-Index - 95
eISSN - 1099-114X
pISSN - 0363-907X
DOI - 10.1002/er.4676
Subject(s) - thorium fuel cycle , uranium , spent nuclear fuel , liquid fluoride thorium reactor , enriched uranium , nuclear engineering , thorium , breeder (animal) , molten salt reactor , molten salt , graphite , nuclear reactor core , nuclear fuel , spent fuel pool , waste management , radiochemistry , environmental science , materials science , chemistry , engineering , metallurgy , blanket , composite material
Summary Using low enrichment uranium as driver fuel under once‐through mode in molten salt reactor (MSR) attracts more and more attention because of its fuel availability, no new technology, and nuclear nonproliferation. It is regarded as a wise research and development road to shorten deployment time of MSRs and to prepare techniques and experiences for thorium‐uranium breeding of MSRs in the future. However, this fuel management is still faced with some different technical routes, such as the selection of carrier salts, the enrichment of uranium, with or without thorium, and the recycle necessary of spent nuclear fuel. Therefore, various fuel cycle schemes were compared and analyzed using an in‐house developed fuel management code MOBAT. Different graphite assemblies were optimized by changing the salt volume fraction in core and dimension to find a region for best fuel utilization and negative temperature reactivity coefficient. Prismatic block with 10% volume fraction of molten salt is considered as a good assembly type because of its significant space shielding effect of U‐238. For carrier salts, LiF‐BeF 2 with 99.995% enriched Li‐7 displays higher fuel utilization and lower cost of fuel cycle than NaF‐BeF 2 , while the tritium production at the beginning of life will be two orders of magnitude higher. For fuel enrichment, 20% enriched uranium is recommended because the background of neutron absorption from carrier salt and graphite will be more significant in lower enrichment condition. Importantly, it shows that thorium is a good breed and burned fuel in situ and could improve the fuel utilization by 20%. Also, offline reprocessing to recover the uranium is a commendable scheme when the cost of offline reprocessing is lower than 400 $/kgHN.