The development of Module In-vessel degraded severe accident Analysis Code MIDAC and the relevant research for CPR1000 during the station blackout scenario

Wang, Jun; Tian, Wenxi; Zhang, Yapei; Chen, Lie; Li, Longze; Zhang, Luteng; Zhou, Yukun; Su, G.H.; Qiu, Suizheng

doi:10.1016/j.pnucene.2014.05.015

Cited by 39 publications

(5 citation statements)

References 32 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Compared with SCDAP/ RELAP5, MIDAC demonstrated a high accuracy in the primary system thermal-hydraulic transient analysis in the CPR1000 station blackout, and the comparison results proved the validity of MIDAC code (Wang et al, 2014c).…”

Section: Severe Accident Review At Xjtumentioning

confidence: 93%

“…Xi'an Jiaotong University developed MIDAC for analysis of IVSA to satisfy the domestic requirement for independent scheduling of software (Wang et al, 2014c). Compared with SCDAP/ RELAP5, MIDAC demonstrated a high accuracy in the primary system thermal-hydraulic transient analysis in the CPR1000 station blackout, and the comparison results proved the validity of MIDAC code (Wang et al, 2014c).…”

Section: Severe Accident Review At Xjtumentioning

confidence: 98%

“…It is important to predict the probability, behavior, and the influence of core degradation to calculate the risk and corresponding mitigation (Wang et al, 2014c). However, even until now, all the models of severe accidents are not sufficient and lack precision (Sehgal, 2012).…”

Section: Severe Accident Review At Xjtumentioning

confidence: 99%

See 2 more Smart Citations

Review on Core Degradation and Material Migration Research in Light-Water Reactors

Wang

Shi

et al. 2018

Front. Energy Res.

Self Cite

View full text Add to dashboard Cite

Section: Severe Accident Review At Xjtumentioning

confidence: 93%

Section: Severe Accident Review At Xjtumentioning

confidence: 98%

See 1 more Smart Citation

Review on Core Degradation and Material Migration Research in Light-Water Reactors

Wang

Shi

et al. 2018

Front. Energy Res.

Self Cite

View full text Add to dashboard Cite

“…In recent years, with the public's increasing concerns over the safety of the nuclear power plant,full deterministic evidence of evidence of nuclear power plants (NPP) safety is required (Wang et al, 2014), more natural circulation systems were applied in the designs of the advanced nuclear plants, because of the inherent passive safety. For example, Hualong Pressurized water reactor, Chinese Generation III nuclear reactor, uses an open natural circulation as the passive containment cooling system (PCCS) (Lakshmanan and Pandey, 2009).…”

Section: Introductionmentioning

confidence: 99%

Experimental Study on the Condensation of Steam With Air Out of the Vertical Tube Bundles

et al. 2018

View full text Add to dashboard Cite

In this paper, experimental study was carried out in an open natural circulation loop, to figure out the condensation heat transfer characteristic on the outside of the vertical pipe tube bundle and single-tube. Condensation heat transfer coefficients have been obtained under the wall subcooling ranging from 5 to 22 • C, total pressure ranging from 0.3 to 0.6 MPa and air mass fraction ranging from 0.05 to 0.65. The influence of mixed gas pressure, condensate depression of walls, and content of non-condensable gas on the condensation heat transfer performance is analyzed. Under the condition of the same air mass fraction, the condensation heat transfer coefficient increases with the increase of pressure. Also, as the air mass fraction is more than 30%, the effect of pressure will be weakened. The heat transfer features of pipe bundle and single tube is compared and studied, and an empirical correlation of the pipe bundle for the heattransfer coefficient is developed, covered all data points within 10%. According to the research results, heat transfer coefficient for pipe bundle decreases with the increase in non-condensable gas quality and wall subcooling, but increases with the increase in pressure.

show abstract

“…Based on the traditional PWR technology, core degradation, melting materials migration and clad oxidation are all priority research questions (Wang et al, 2014). Currently, SMRs mainly adopt a set of passive safety devices for mitigating the effects of severe accidents (Yin et al, 2016).…”

Section: Introductionmentioning

confidence: 99%

Accident Process and Core Thermal Response During a Station Blackout Initiated Study for Small Modular Reactor

et al. 2018

Self Cite

View full text Add to dashboard Cite

Interest in evaluation of reactor response and off-site consequences following beyond design basis accidents has significantly increased after Fukushima. Considering the inherent system safety of nuclear plants, experts begin to refocus on the development of small and medium nuclear reactors (SMRs). Based on proven and widely utilized traditional pressurized water reactor (PWR) technology, there are many kinds of small modular reactor design concepts, which are actively being developed in the USA, China, Japan, Russia, Korea, and other countries. But due to the significant differences between the traditional distributed arrangement and integrated arrangement of SMRs, the variation trend of key parameters may be different during normal operation or accident process. Hence, in this paper, we simulated the small modular reactor severe accident by MELCOR. This paper summarizes the core thermal hydraulic response for a hypothetical severe accident caused by station blackout at a small modular reactor using MELCOR. Analytical results for temperature distribution of the fuel pellets, the fuel cladding, flow rate of the coolant, and hydrogen mass changing over time are presented. The analyses are focused on safety assessment of the reactor core for severe accidents and are a part of the overall evaluation of safety features of the small modular reactor for residual risk posed by severe accidents.

show abstract

The development of Module In-vessel degraded severe accident Analysis Code MIDAC and the relevant research for CPR1000 during the station blackout scenario

Cited by 39 publications

References 32 publications

Review on Core Degradation and Material Migration Research in Light-Water Reactors

Review on Core Degradation and Material Migration Research in Light-Water Reactors

Experimental Study on the Condensation of Steam With Air Out of the Vertical Tube Bundles

Accident Process and Core Thermal Response During a Station Blackout Initiated Study for Small Modular Reactor

Contact Info

Product

Resources

About