Assessment of transient characteristics of fast reactors and influences of minor actinides using neutron transport method-Reference-Cited by-同舟云学术

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Assessment of transient characteristics of fast reactors and influences of minor actinides using neutron transport method

Published:2017-06-14 Issue:14 Volume:41 Page:2194-2205
ISSN:0363-907X
Container-title:International Journal of Energy Research
language:en
Short-container-title:Int J Energy Res

Author:

He Mingtao¹,Zheng Youqi¹^ORCID,Wu Hongchun¹,Li Zhuo¹,Du Xianan¹

Affiliation:

1. School of Nuclear Science and Technology; No. 5, Xianning West Road, Xi'an Jiaotong University; Xi'an Shaanxi 710049 China

Funder

National Natural Science Foundation of China

Publisher

Wiley

Subject

Energy Engineering and Power Technology,Fuel Technology,Nuclear Energy and Engineering,Renewable Energy, Sustainability and the Environment

Reference27 articles.

1. Performance modeling and analysis of spent nuclear fuel recycling;Gao;International Journal of Energy Research,2015

2. Radioactive waste partitioning and transmutation within advanced fuel cycles: achievements and challenges;Salvatores;Progress in Particle and Nuclear Physics,2011

3. Feasibility study on ultralong-cycle operation and material performance for compact liquid metal-cooled fast reactors: a review work;Tak;International Journal of Energy Research,2015

4. Coupling of an innovative small PWR and advanced sodium-cooled fast reactor for incineration of TRU from once-through PWRs;Kim;International Journal of Energy Research,2016

5. Physics of Americium transmutation;Wallenius;Nuclear Engineering and Technology,2012

Cited by 3 articles. 订阅此论文施引文献订阅此论文施引文献，注册后可以免费订阅5篇论文的施引文献，订阅后可以查看论文全部施引文献

1. Multi-physics numerical analysis of the fuel-addition transients in the liquid-fuel molten salt reactor;Annals of Nuclear Energy;2020-09

2. Monte Carlo neutron-kinetics analysis in continuous time and energy;Annals of Nuclear Energy;2020-06

3. Transient analysis of MOX-3600 and MET-1000 sodium-cooled fast reactor using SARAX code system;Annals of Nuclear Energy;2018-11

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