Assessment of the severe accident code MIDAC based on FROMA, QUENCH-06&amp;16 experiments

Wu, Shihao; Zhang, Yapei; Wang, Dong; Tian, Wenxi; Qiu, Suizheng; Su, G.H.

doi:10.1016/j.net.2021.08.004

Cited by 17 publications

(2 citation statements)

References 23 publications

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“…The natural circulation characteristics in the lower head after a severe accident of the reactor are obtained, which provided a foundation for design optimization and heat transfer enhancement. A simulation and analysis code for severe accidents in PWR has been developed [48], which includes early behavior module, core degradation module, debris bed module, molten corium in-vessel retention module and thermal hydraulic module, providing effective support for analysis and mitigation of severe accidents in PWR.…”

Section: Safety Characteristics Analysis Under Severe Accidentmentioning

confidence: 99%

Analysis on the high-quality development of nuclear energy under the goal of peaking carbon emissions and achieving carbon neutrality

Chen

Zhang

2022

Carb Neutrality

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Striving to peak carbon emissions and achieve carbon neutrality (known as the "Dual-Carbon" goal) is an inevitable requirement for elevating the environmental resource constraints and realizing harmonious coexistence between the mankind and the earth. In the energy system, nuclear energy offers various advantages, such as high energy density, low carbon emission, strong environmental adaptability and large potential for energy co-generation and co-supply. It is one of the supporting energy sources for the transformation and upgradation of the energy system to a clean, efficient and low-carbon way. In this paper, the opportunities and challenges for innovation-driven nuclear energy development in the fields of electricity generation, hydrogen production, heat supply and seawater desalination under the goal of "Dual-Carbon" are discussed and analyzed. Besides, the relevant research on improving the safety and economy of the pressurized water reactor (PWR) and sodium-cooled fast reactor (SFR) conducted by the Nuclear THermal–hydraulic research Lab (NuTHeL) of Xi’an Jiaotong University is briefly introduced.

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Section: Safety Characteristics Analysis Under Severe Accidentmentioning

confidence: 99%

Analysis on the high-quality development of nuclear energy under the goal of peaking carbon emissions and achieving carbon neutrality

Chen

Zhang

2022

Carb Neutrality

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“…Many computer codes have been developed for the safety analysis including the severe accident, such as MELCOR [10,11], MAAP5 [12] in USA, and ASTEC [13] in France for a severe accident sequence analysis. Many accident analysis have been performed to analyze accident progression and general reactor safety problems [14][15][16][17][18] and to solve the specifc thermal hydraulic issues on the safety analysis including the severe accident, such as natural circulation, external reactor vessel cooling, reactor vessel cooling, and so on [19][20][21][22]. Tese results may be used to solve the nuclear license issues for a large power reactor and small reactors.…”

Section: Introductionmentioning

confidence: 99%

Simulations of Core Damage Progression for TMI-2 Severe Accident Using CINEMA Computer Code

Park

Son

Bae

et al. 2023

Science and Technology of Nuclear Installations

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As an integrated computer code development for severe accident sequence analysis in Korea, CINEMA has been developing from an initiation event to a containment failure. The CINEMA computer code is composed of CSPACE, SACAP, and SIRIUS, which are capable of simulating core melt progression with thermal hydraulic analysis of the RCS (reactor coolant system), severe accident analysis of the containment, and fission product analysis in the vessel and the containment, respectively. The severe accident progression in TMI unit 2 has been analyzed as a part of a validation of the CINEMA computer code. This analysis has been performed to validate CINEMA models on the core melt progression, in particular, RCS thermal hydraulic behavior during core melt progression, fuel cladding oxidation with hydrogen generation, and fuel melting with relocation to the lower part of the core. The CINEMA results on main parameters, such as RCS pressure and an integrated hydrogen generation mass are compared with the TMI-2 data. The CINEMA results have shown that the RCS pressure is very similar to the TMI-2 data. The CINEMA results and measured total hydrogen production are very similar, which were approximately 465 kg and 460 kg, respectively.

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Experimental study on turbulent natural convection of air between the open-ended parallel plates with one-side heating

Dong

Lu²,

Chen³

et al. 2023

Annals of Nuclear Energy

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Assessment of the severe accident code MIDAC based on FROMA, QUENCH-06&16 experiments

Cited by 17 publications

References 23 publications

Analysis on the high-quality development of nuclear energy under the goal of peaking carbon emissions and achieving carbon neutrality

Analysis on the high-quality development of nuclear energy under the goal of peaking carbon emissions and achieving carbon neutrality

Simulations of Core Damage Progression for TMI-2 Severe Accident Using CINEMA Computer Code

Experimental study on turbulent natural convection of air between the open-ended parallel plates with one-side heating

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