Steam Line Break investigation at full power reactor for VVER-1000/V320

Pavlova, M.P.; Andreeva, M.; Groudev, Pavlin

doi:10.1016/j.nucengdes.2015.01.006

Cited by 8 publications

(3 citation statements)

References 4 publications

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“…Summarizing of MSLB [5] scenario and assumptions: 1. Size break (with ID:580 mm) at loop #4 between the steam generator (SG) and the steam isolation valve (SIV), outside of the containment (in front of (SIV)); 2.…”

Section: Investigated Scenariomentioning

confidence: 99%

Comparison of VVER1000 plant respond during MSLB scenario using TRACE and RELAP computer codes

Stefanova,

Groudev

2024

IOP Conf. Ser.: Earth Environ. Sci.

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This paper presents a comparison of the results from simulation of „Main steam line break (MSLB) “accident using RELAP5/mod3.3 and TRACE50p5 computer codes. The work is based on the VVER-1000 MSLB OECD benchmark. The main purpose of this work was to demonstrate the capability of the TRACE computer code for modelling of 3D thermal hydraulic reactor vessel at the integral VVER 1000 model and to predict correctly the plant respond and spatial disturbance in the reactor vessel. In addition, this investigation was focused on validation of the correct predictability of the important integral parameters in VVER1000 TRACE 50p5 models. The other objectives were to provide the elements for the safety, to validate the safety analysis tools and to improve codes and methods for VVER comprehensive safety assessment. The investigated scenario is a “Main steam line break (MSLB) with inside diameter (ID) 580 mm” in a VVER1000 between the steam generator (SG) and the steam isolation valve (SIV), outside of the containment. This event is characterized with significant spatial effects in the core caused by asymmetric cooling and assumed stuck-out of control rods after SCRAM. One of the major concerns for this case is the possible return of power and criticality after SCRAM, as a result of the reactor core overcooling. The reactor state is at the beginning of 1st fuel campaign. For the purpose of the investigation, the simulation of MSLB accident with RELAP5/MOD3.3 and TRACE50p5 computer codes with modelling of reactor physics using point kinetics has been performed initially. Additionally, the 3D thermal hydraulic reactor vessel model with a point kinetic physics has been developed at the integral VVER1000 TRACE 50p5 model. The newly developed integral model of VVER1000 with 3D reactor vessel of TRACE50p5 has been validated with simulation of MSLB accident by code to code comparison with already validated VVER1000 models of RELAP5/MOD3.3 and 1D TRACE50p5 codes. The comparison between the results predicted by the codes demonstrates good agreement, except some of the parameters. The newly developed integral VVER1000 with 3D vessel model of TRACE50p5 predicts reasonable results. The predicted results by used computer codes demonstrate the stability of the reactor system during the accident progression of “Main steam line break”.

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Section: Investigated Scenariomentioning

confidence: 99%

Comparison of VVER1000 plant respond during MSLB scenario using TRACE and RELAP computer codes

Stefanova,

Groudev

2024

IOP Conf. Ser.: Earth Environ. Sci.

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“…The analysis of the nuclear power plant's behavior with thermo-hydraulic code is carried out for its safety justification in case of design disturbances during the processes and malfunctions or failures of the equipment. Several studies related to the WWER-1000 nuclear power plant accident, have been modeled with RELAP5/MOD3.2 [6]- [11]. Most of the publications present accident analyses in the full power operation of the plant.…”

Section: Description Of the Knpp And Relap5 Modelmentioning

confidence: 99%

Study of Accident Progression in Unsealed WWER-1000/V320 Reactor during Maintenance

Groudev¹,

Andreeva²

2016

JPEE

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“…15 Pavlova et al, have investigated steam-line break accident for VVER-1000 type nuclear power plant under full load condition. 16 Shawxin et al, have studied the transient response of a high-power passive reactor during a steamline break accident. 17 Kliem et al, have tested the performance NURESIM platform for analyzing main steam-line break accident of a pressurized water 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

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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.

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Steam Line Break investigation at full power reactor for VVER-1000/V320

Cited by 8 publications

References 4 publications

Comparison of VVER1000 plant respond during MSLB scenario using TRACE and RELAP computer codes

Comparison of VVER1000 plant respond during MSLB scenario using TRACE and RELAP computer codes

Study of Accident Progression in Unsealed WWER-1000/V320 Reactor during Maintenance

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

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