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

2019 ◽  
Vol 14 (2) ◽  
pp. 299-311
Author(s):  
Abid Hossain Khan ◽  
Angkush Kumar Ghosh ◽  
Md Sumon Rahman ◽  
S M Tazim Ahmed ◽  
C L Karmakar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Thermal Power ◽  
Steam Line ◽  
Radioactive Elements ◽  
Time Duration ◽  
Plant Safety ◽  
Safety Features ◽  
Safety Systems

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.

scholarly journals Station Black-Out Analysis with MELCOR 1.8.6 Code for Atucha 2 Nuclear Power Plant

2012 ◽  
Vol 2012 ◽  
pp. 1-17 ◽  
Author(s):  
Analia Bonelli ◽  
Oscar Mazzantini ◽  
Martin Sonnenkalb ◽  
Marcelo Caputo ◽  
Juan Matias García ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heavy Water ◽  
Nuclear Power ◽  
Failure Time ◽  
Thermal Power ◽  
Radiation Heat Transfer ◽  
Primary System ◽  
Water Release ◽  
Plant Safety

A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the Station Black-Out sequence the first pressurizer safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. This feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than that of other German PWRs.

Validation of the periodicity of checks of nuclear power plant safety systems using the bars system of programs

Atomic Energy ◽  
2010 ◽  
Vol 109 (2) ◽  
pp. 81-87 ◽  
Author(s):  
G. A. Ershov ◽  
Yu. L. Ermakovich ◽  
M. A. Kozlov ◽  
M. A. Parfentiev ◽  
A. I. Kalinkin ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Plant Safety ◽  
Nuclear Power Plant Safety ◽  
Safety Systems

Various MTC Studies of TRACE with RCS Pressure Predictions under ATWS for Maanshan

2013 ◽  
Vol 284-287 ◽  
pp. 1151-1155
Author(s):  
Che Hao Chen ◽  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Thermal Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Graphic User Interface ◽  
Pressurized Water ◽  
Plant Safety ◽  
Regulatory Commission

The objective of this study is to utilize TRACE (TRAC/RELAP Advanced Computational Engine) code to analyze the reactor coolant system (RCS) pressure transients under ATWS (Anticipated Transient Without Scram) for Maanshan PWR (Pressurized Water Reactor) in various MTC (Moderator Temperature Coefficient) conditions. TRACE is an advanced thermal hydraulic code for nuclear power plant safety analysis, which is currently under development by the United States Nuclear Regulatory Commission (USNRC). A graphic user interface program named SNAP (Symbolic Nuclear Analysis Package), which processes inputs and outputs for TRACE is also under development. Maanshan nuclear power plant (NPP) is the only Westinghouse PWR in Taiwan. The rated core thermal power of Maanshan with MUR (Measurement Uncertainty Recapture) is 2822 MWt. In document SECY-83-293, all initializing events were classified as either turbine trip or non-turbine trip events and their ATWS risks were also evaluated according to these two events. Loss of condenser vacuum (LOCV) and Loss of normal feedwater (LONF) ATWS were identified as limiting transients of turbine trip and non-turbine trip events in this study. According to ASME Code Level C service limit criteria, the RCS pressure for Maanshan NPP must be under 22.06 MPa. Furthermore, we select the LOCV transient to analyze various MTC effects on RCS pressure variations.

Kalman Filter Based Predictive Trip Detection for Nuclear Power Plant Safety Systems

2014 ◽  
Author(s):  
Drew J. Rankin ◽  
Jin Jiang
Keyword(s):  
Kalman Filter ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Estimation Model ◽  
Response Characteristics ◽  
Plant Safety ◽  
Error Covariance ◽  
Nuclear Power Plant Safety ◽  
Safety Systems

The primary aim of this work is to utilize a Kalman filter (KF) to predict reaching the trip set-point for a trip parameter in a nuclear power plant (NPP). To address uncertainty in the predicted measurements, prediction bounds are calculated by propagating the state error covariance. These predicted bounds enable the calculation of levels of confidence in making trip decisions. Further, to address uncertainty in the estimation model, the observed prediction error is used to offset the predicted measurements. The predictive trip detection routines are evaluated through simulations of a single NPP sub-system. More specifically, the water level parameter in a steam generator of a NPP is considered. The model of this sub-system is represented by the Irving linear parameter varying (LPV) model with inverse response characteristics. The simulations include a level low postulated initiating event (PIE) made to occur during two different common power transients for various estimation models. The results of this paper are a proof of concept for KF-based predictive trip detection which is demonstrated to achieve reduced time-to-trip when applied to a single sub-system.

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