Zobrazeno 1 - 10
of 2 083
pro vyhledávání: '"sodium-cooled fast reactor"'
Publikováno v:
Nuclear Engineering and Technology, Vol 56, Iss 9, Pp 3626-3643 (2024)
One of the issues in the development of the sodium-cooled fast reactor (SFR) using transuranic (TRU) metallic fuel is the absence of criticality benchmark experiment that faithfully mocks up the nuclear characteristics of the target design for valida
Externí odkaz:
https://doaj.org/article/3aacc65c325543949da763d83f0bee1d
Autor:
S. Venkateswarlu, E. Hemanth Rao, G.V. Prasad Reddy, Sanjay Kumar Das, D. Ponraju, B. Venkatraman
Publikováno v:
Nuclear Engineering and Technology, Vol 56, Iss 9, Pp 3864-3871 (2024)
In case of a hypothetical severe accident in a Sodium-cooled Fast Reactor (SFR), coolability of the debris bed in the post-accident phase plays a vital role in mitigating the accident and ensuring the structural integrity of the reactor vessel. Few n
Externí odkaz:
https://doaj.org/article/1ceab43ae79747ef8a3596cd2800518d
Publikováno v:
Yuanzineng kexue jishu, Vol 58, Iss 5, Pp 1101-1108 (2024)
With the aim of “carbon peaking” and “carbon neutrality” in China, hydrogen energy will play a crucial role of in the future. The third-generation nuclear power (GEN-Ⅲ) features a relatively low reactor outlet temperature, with the primary
Externí odkaz:
https://doaj.org/article/c0a66c7d3a434b15b42ecb91f7f80876
Publikováno v:
Yuanzineng kexue jishu, Vol 58, Iss 4, Pp 825-835 (2024)
In the event of core disassembly accident (CDA) in a sodium-cooled fast reactor (SFR), the molten material from the core migrates and interacts with the low-temperature coolant in the lower plenum, resulting in the formation of core debris that event
Externí odkaz:
https://doaj.org/article/f7a1e40dabaa4d8790fa73f999bb23fa
Autor:
CHEN Qidong, GAO Fuhai
Publikováno v:
Yuanzineng kexue jishu, Vol 58, Iss 3, Pp 604-613 (2024)
For many years, sodium-cooled fast reactors have occupied the most important part of the closed fuel cycle. In order to improve the economy of sodium-cooled fast reactors, the nuclear industry around the world is actively increasing fuel burnup as mu
Externí odkaz:
https://doaj.org/article/14d73d6d09884e5ea4a83c8430427150
Publikováno v:
Yuanzineng kexue jishu, Vol 58, Iss 3, Pp 689-697 (2024)
The inherent safety of reactor is emphasized in the design of the sodium-cooled fast reactors (SFRs). For accident mitigation, inherent safety and passive measures are adopted to reduce the demand for power sources and enhance safety and economy. SFR
Externí odkaz:
https://doaj.org/article/bb7afbe75c31408cb9ba4e24d4c4600b
Publikováno v:
Nuclear Engineering and Technology, Vol 56, Iss 3, Pp 772-784 (2024)
The hexagonal nodal code RENUS has been enhanced to handle irregularly deformed hexagonal assemblies. The underlying RENUS methods involving triangle-based polynomial expansion nodal (T-PEN) and corner point balance (CPB) were extended in a way to us
Externí odkaz:
https://doaj.org/article/27727515345643ac9515888a0fc13431
Publikováno v:
Nuclear Engineering and Technology, Vol 56, Iss 3, Pp 873-879 (2024)
A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safet
Externí odkaz:
https://doaj.org/article/8c1fbc5d91eb4f30b55e255ba0808880
Autor:
Tomohiko Yamamoto, Atsushi Kato, Masato Hayakawa, Kazuhito Shimoyama, Kuniaki Ara, Nozomu Hatakeyama, Kanau Yamauchi, Yuhei Eda, Masahiro Yui
Publikováno v:
Nuclear Engineering and Technology, Vol 56, Iss 3, Pp 893-899 (2024)
In a secondary cooling system of a sodium-cooled fast reactor (SFR), rapid detection of hydrogen due to sodium-water reaction (SWR) caused by water leakage from a heat exchanger tube of a steam generator (SG) is important in terms of safety and prope
Externí odkaz:
https://doaj.org/article/687ef101b1b34ed29511af02d53f54f9
Publikováno v:
Nuclear Engineering and Technology, Vol 56, Iss 1, Pp 257-264 (2024)
The sodium cooled fast reactor (SFR) is one of the Gen-IV reactors with the most operating experience accumulated. Although the technology level is the most mature among the Gen-IV reactors, there is still a safety problem that has not been solved, w
Externí odkaz:
https://doaj.org/article/284cb6578d2f4990ab857c52f70d7342