Uncertainty Analysis of SB-LOCA in AP1000 Nuclear Power Plant

2017 ◽  
Author(s):  
Hao Shi ◽  
Qi Cai ◽  
Yuqing Chen ◽  
Lizhi Jiang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Hydraulic System ◽  
Nuclear Power ◽  
Safety Margin ◽  
Analysis Model ◽  
The Core ◽  
Best Estimate ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Best estimate plus uncertainty methods (BEPUs) could be used to eliminate the over-conservatism and gain more safety margin in the analysis of thermal-hydraulic transient process at nuclear power plant. Based on the Best estimate thermal-hydraulic system code RELAP5/MOD3.2 platform, the best estimate plus uncertainty methods (BEPUs) proposed by GRS (Gesellschaft fur Anlagen- and Reaktorsicherheit) are presented together with applications to a small break loss of coolant accident (SB-LOCA) on the AP1000 Nuclear Power Plant best estimate analysis model. According to the results of uncertainty calculations, the dispersion bands of maximum cladding temperature and the core outlet void fraction are displayed and assessed.

A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV

2017 ◽  
Author(s):  
Eltayeb Yousif ◽  
Zhang Zhijian ◽  
Tian Zhao-fei ◽  
A. M. Mustafa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Systems ◽  
Nuclear Power ◽  
Power Plants ◽  
Computational Simulation ◽  
Primary System ◽  
Transient Time ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

To ensure effective operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, Loss of Coolant Accident (LOCA) is one of the main accidents. RELAP-MV Visualized Modularization software technology is recognized as one of the best estimated transient simulation programs of light water reactors, and also has the options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. In this study, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SB-LOCA accident in AP1000 nuclear power plant (NPP) is carried out by Relap5-MV thermo-hydraulics code. While focusing on LOCA analysis in this study, effort was also made to test the effectiveness of the RELAP5-MV software already developed. The accuracy and reliability of RELAP5-MV have been successfully confirmed by simulating LOCA. The calculation was performed up to a transient time of 4,500.0s. RELAP5-MV is able to simulate a nuclear power system accurately and reliably using this modular modeling method. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially for the beginners.

Performance of a Spray Pond for Nuclear Power Plant Ultimate Heat Sink

1977 ◽  
Vol 99 (4) ◽  
pp. 650-656
Author(s):  
V. E. Schrock ◽  
G. J. Trezek ◽  
L. R. Keilman
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Performance ◽  
Heat Sink ◽  
Nuclear Power ◽  
Field Tests ◽  
Computer Code ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Transient Thermal

Spray ponds have become an attractive method of providing the “ultimate heat sink”, i.e., the assured means of dissipating heat from a nuclear power plant. Two redundant spray ponds were the choice for this service in the Rancho Seco Nuclear Generating Station owned by Sacramento Municipal Utility District. This paper describes the results of full scale field tests of the Rancho Seco ponds which were conducted to verify the thermal performance, drift loss characteristics, and the capability to sustain the cooling requirements for a period of 30 days following a loss-of-coolant accident (LOCA). Correlations of local and average nozzle efficiency and of the drift loss are presented. A computer code was developed for the transient thermal performance of the pond. After verification the code was used to predict performance following LOCA under adverse meteorological conditions based on weather records.

Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

2015 ◽  
Vol 285 ◽  
pp. 1-14 ◽  
Author(s):  
Asko Arkoma ◽  
Markku Hänninen ◽  
Karin Rantamäki ◽  
Joona Kurki ◽  
Anitta Hämäläinen
Keyword(s):  
Statistical Analysis ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant
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