Simulations of unprotected loss of heat sink and combination of events accidents for a molten salt reactor

Zhang, Dalin; Zhang, Dalin; Xiao, Yao; Tian, Wenxi; Su, G.H.; Qiu, Suizheng

doi:10.1016/j.anucene.2012.09.009

Cited by 28 publications

(11 citation statements)

References 18 publications

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“…The total delayed neutron precursor fraction b is replaced by effective delayed neutron precursor fraction b eff , which can be defined as the delayed neutron precursor fraction without flow effect subtracts the loss of delayed neutron precursor fraction due to flow effect. And the DNPs can be separated into two parts: one is in reactor core, and the other is in external loop (Guo et al, 2013). Based on the conservation law, the neutron equations for liquid fueled reactor can be written as following form:…”

Section: Neutronic Modelmentioning

confidence: 99%

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Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

Shi

Cheng

2016

Nuclear Engineering and Design

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Section: Neutronic Modelmentioning

confidence: 99%

“…Wang et al (2006) focused on examining the thermo-hydraulic behavior of the MOSART with the SIMMER-III code by extending both thermo-hydraulic and neutronic models. Guo et al (2013) developed the safety analysis code by splitting the delayed neutron precursors (DNPs) into two parts: one is in reactor core and the other is in external loop.…”

Section: Introductionmentioning

confidence: 99%

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

Shi

Cheng

2016

Nuclear Engineering and Design

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“…28,43 COUPLE, a time-dependent coupled neutronics and thermal-hydraulics code, was developed on the basis of the previous work. 46,47 Using the conceptual design of the TMSR-LF as the reference case, a pump stop accident was simulated at nominal power of 2 MW th by Cinsf1D code. 40,44 Compared with the COUPLE, the coupling of MCNP with Multiple-channel Analysis Code can hardly simulate the transients with a proper speed.…”

Section: Neutronics and Thermal-hydraulics Coupling And Safety Analmentioning

confidence: 99%

Review of conceptual design and fundamental research of molten salt reactors in China

Zhang

Liu

et al. 2018

Int J Energy Res

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Summary Molten salt reactor (MSR) as 1 candidate of the generation IV advanced nuclear power systems attracted more attention in China due to its top ranked in fuel cycle and thorium utilization. Two types of MSR concepts were studied and developed in parallel, namely the MSR with liquid fuel and that with solid fuel. Abundant fundamental research including the neutronics modeling, thermal‐hydraulics modeling, safety analysis, material investigation, molten salts technologies etc. were carried out. Some analysis software such as COUPLE and FANCY were developed. Several experimental facilities like high‐temperature fluoride salt experiment loop have been constructed. Some passive residual heat removal systems were designed, and 1 test facility is under construction. The key MSR techniques including the extraction and separation of molten salt and construction of N‐base alloy have been mastered. Based on these fundamental research, Chinese Academy of Sciences has completed the design of thorium‐based MSRs with solid fuel and liquid fuel and is promoting their construction in the near future. In China, future efforts should be paid to the material, online fuel processing, Th‐U fuel cycle, component design, and construction and thermal‐hydraulic experiments for MSR, which are rather challenging nowadays.

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“…Křepel et al analyzed transient of unprotected pump coast‐down, reactivity insertion, and overcooling of the fuel by DYN3D‐MSR code for illustration of MSR specific dynamic features . Zhang et al and Guo et al studied the transient of unprotected loss of flow, unprotected overcooling accident and unprotected loss of heat sink based on coupled point kinetic model and single‐channel TH model . Shi et al investigated transient of load demand change, primary flow, and secondary flow for MSBR based on coupled point kinetic model and RELAP5 code .…”

Section: Introductionmentioning

confidence: 99%

“…18,19 Zhang et al and Guo et al studied the transient of unprotected loss of flow, unprotected overcooling accident and unprotected loss of heat sink based on coupled point kinetic model and single-channel TH model. 20,21 Shi et al investigated transient of load demand change, primary flow, and secondary flow for MSBR based on coupled point kinetic model and RELAP5 code. 22 Li and Wang focused on several accident conditions for MSFR by both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), such as loss of heat sink (ULOHS), unprotected loss of flow (ULOF), and unprotected transient over power (UTOP).…”

Section: Introductionmentioning

confidence: 99%

Transient analysis for liquid-fuel molten salt reactor based on MOREL2.0 code

et al. 2017

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SummaryA molten salt reactor (MSR) is characterized by simultaneously using liquid fuel salt as both the nuclear fuel and coolant. The redistribution of delayed neutron precursors (DNPs) makes the transient behavior of MSRs different from traditional solid-fuel reactors. In this study, a 3D coupled neutronics/thermal hydraulics code, MOREL2.0, was employed to analyze a liquid-fuel Thorium Molten Salt Reactor (TMSR-LF) under perturbations of fuel pump start-up and coast-down and by overheating and overcooling the inlet fuel temperature. Some transient processes were simulated to provide guidance for the future design and optimization of TMSR-LFs. In response to the perturbations, reactivity was lost and gained in the pump start-up and coast-down, respectively.Overheating the inlet fuel temperature introduced negative reactivity, and TMSR-LF maintained a safety state, while overcooling the inlet fuel temperature resulted in positive reactivity. Overcooling by 70 K produced a supercritical transient condition and a rapid increase in power within a short period, which was followed by a decrease in power due to negative temperature feedback. The transient results demonstrate that the negative temperature feedback coefficients guarantee TMSR-LF inherent safety and the variation range of temperature stay within the safety margin.

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Simulations of unprotected loss of heat sink and combination of events accidents for a molten salt reactor

Cited by 28 publications

References 18 publications

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

Review of conceptual design and fundamental research of molten salt reactors in China

Transient analysis for liquid-fuel molten salt reactor based on MOREL2.0 code

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