PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation

Hadad, Kamal; Esmaeili-Sanjavanmareh, M.

doi:10.3139/124.110649

Cited by 1 publication

(1 citation statement)

References 9 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…21 Hadad and Esmaeili-Sanjavanmareh have analyzed large break loss of coolant accident for a VVER-1000 type nuclear power plant. 22 Chiang et al, have combined a number of other nuclear thermal hydraulic codes with PCTRAN to simulate Fukushima-like accident for Chinshan BWR/4 NPP. 23 Po has shown the use of PCTRAN for simulation of Indonesian conceptual HTR reactor.…”

Section: Introductionmentioning

confidence: 99%

An Investigation on the Possible Radioactive Contamination of Environment during a Steam-Line Break Accident in a VVER-1200 Nuclear Power Plant

Khan¹,

Ghosh

Rahman³

et al. 2019

Curr. World Environ

View full text Add to dashboard Cite

In this work, the possibility of contamination of environment by radioactive elements due to a steam-line break accident has been investigated for a VVER-1200 type nuclear power plant. Personal Computer Transient Analyzer (PCTRAN) has been used to generate the response data of the plant safety systems numerically for an accidental condition like such. A break of 1000 cm2 in the A-loop of the steam line has been considered. A break of the size is considered a “Large Break”, which is believed to be responsible for multiple serious accidents in the past. Also, it has also been assumed that off-site AC power supply is unavailable. Simulations were run for time duration of 300 seconds since most of the safety features of the plant should respond within 50 seconds from the initiation of the accident. Results show that SCRAM is initiated within 22.5 seconds from the emergence of the break, which limited the peak core thermal power to around 105% of the nominal value. The peak temperatures of fuel elements and fuel cladding are recorded to be around 1850oC and 620oC respectively, which are both within the safety limits. The pressure inside reactor pressure vessel has not undergone any significant changes, showing no sign of failure. Again, the pressure inside the reactor containment building is kept within 2.5 bar by the safety systems, indicating that there is no possibility of containment failure due to over-pressure. Finally, the readings from radiation monitor show that there is no noticeable release of radioactive elements to the environment during the accident. Therefore, it may be concluded that the release of radioactive elements in the surrounding environment during a steam-line break accident is very unlikely provided that the plant safety systems are fully functional.

show abstract

Section: Introductionmentioning

confidence: 99%