Evaluation of chemical effects on fuel assembly blockage following a loss of coolant accident in nuclear power plants

2020 ◽  
Vol 44 (7) ◽  
pp. 5488-5499
Author(s):  
Da Wang ◽  
Baochen Chang ◽  
Tianyi Zhang ◽  
Sichao Tan ◽  
Fenglei Niu
Keyword(s):  
Fuel Assembly ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Chemical Effects ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

scholarly journals SUBSTANTIATION OF MODERNIZED BLACKOUT & LOSS-OF-COOLANT ACCIDENT MANAGEMENT STRATEGY AT NUCLEAR POWER PLANTS WITH WWER

2020 ◽  
Vol 2 (61) ◽  
pp. 70-77
Author(s):  
V. Skalozubov ◽  
◽  
V. Spinov ◽  
D. Spinov ◽  
Т. Gablaya ◽  
...  
Keyword(s):  
Management Strategy ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Management Strategies ◽  
Computational Modelling ◽  
Loss Of Coolant Accident ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Safety Systems

The analysis of the known results of RELAP5/V.3.2 simulation for loss of coolant & blackout accidents at WWER nuclear power plants showed that the design accident management strategies with design passive safety systems do not provide the necessary safety conditions for the maximum permissible temperature of fuel claddings, the minimum permissible level of coolant in the reactor and feed water in the steam generators. A conservative thermohydrodynamic model for a design and modernized blackout & loss-of-coolant accident management strategy at a nuclear power plant with WWER has been developed. Design passive safety systems carry out the design accident management strategy: pressurizer safety valves, secondary steam relief valves, and hydraulic reservoirs of the emergency core cooling system of the reactor. Promising afterheat removal passive systems and the reactor level and steam generator water level control systems carry out the modernized blackout & loss-of-coolant accident management strategy. The main conservative assumptions of the presented model of blackout & loss-of-coolant accidents: complete long-term failure of all electric pumps of active safety systems, the temperature of nuclear fuel in the central part of the fuel matrix is assumed as the maximum allowable one, effect of “run down” flow of a turbine feed pump and the coolant level in pressurizer on accident process is not considered. Computational modelling has found that violations of the safety conditions are over the entire range of leak sizes for the design blackout & loss-of-coolant accident management strategy. For the modernized blackout & loss-of-coolant accident management strategy, safety conditions are provided for 72 hours of the accident and more. The presented results of computational modelling of blackout accident management strategies for nuclear power plants can be used to modernize and improve symptom-informed emergency instructions and guidelines for the severe accident management at nuclear power plants with WWER. Application of the results of computational modelling of blackout accident management strategies is generally not substantiated for other types of reactor facilities. In this case, it is necessary to develop calculated models for blackout accident management taking into account the specifics of the structural and technical characteristics and operating conditions for safety related systems of nuclear power plants.

An advanced method for determination of loss of coolant accident in nuclear power plants

2011 ◽  
Vol 241 (6) ◽  
pp. 2013-2019 ◽  
Author(s):  
R. Mahmoodi ◽  
M. Shahriari ◽  
A. Zolfaghari ◽  
A. Minuchehr
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Advanced Method ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

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.

Flow-Induced Vibration Analysis of an Integral Economizer Once-Through Steam Generator

1989 ◽  
Vol 111 (4) ◽  
pp. 501-506
Author(s):  
M. K. Au-Yang ◽  
B. Brenneman
Keyword(s):  
Fuel Assembly ◽  
Steam Generator ◽  
Flow Rate ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
First Generation ◽  
Second Generation ◽  
Flow Induced Vibration ◽  
Operational Test

The integral economizer once-through steam generator is a second-generation steam generator used in B&W’s 205-fuel assembly nuclear power plants. Besides having an integral economizer, this steam generator differs from the first generation units, sixteen of which have been operating with B&W’s 177 fuel assembly nuclear power plants for more than ten years, in having a much higher flow rate. This higher flow rate induces a correspondingly higher fluid-dynamic load on all of the steam generator internal components, particularly the tube bundle. This paper describes the flow-induced vibration design analysis of this second-generation nuclear steam generator. The three most commonly known flow-induced vibration phenomena were considered: fluid-elastic instability, turbulence-induced vibration and vortex-induced vibration. To minimize uncertainties in the many experimentally determined input parameters such as damping ratios, Connors’ constant and the dynamic pressure power spectral densities, a parallel analysis was carried out on the operating first-generation steam generator, and the results compared. The analytical results were verified by the recent start-up of B&W’s first 205-fuel assembly nuclear plant. No vibration problems were encountered during either the pre-operational test or in several months of full power operations.

A Neural Networks Design Methodology for Detecting Loss of Coolant Accidents in Nuclear Power Plants

2018 ◽  
pp. 43-61 ◽  
Author(s):  
David Tian ◽  
Jiamei Deng ◽  
Gopika Vinod ◽  
T. V. Santhosh ◽  
Hissam Tawfik
Keyword(s):  
Neural Networks ◽  
Nuclear Power ◽  
Design Methodology ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Loss Of Coolant
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