Coupling of RELAP and LabVIEW to Model a Nuclear Power Plant System Realistically

2014 ◽  
Author(s):  
J. C. Pack ◽  
Z. Fu ◽  
F. Aydogan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Virtual Instrument ◽  
Nuclear Power ◽  
System Modeling ◽  
Loop Model ◽  
Secondary System ◽  
Loss Of Coolant Accident ◽  
Accident Scenarios ◽  
Computer Simulators

Within the study and design of a nuclear power plant extensive system modeling is necessary to determine how the reactor is going to perform in any given situation, not only in the normal performance of the reactor but also transients including unanticipated transients without scram and hypothetical accidents. One of the difficulties in the performance of this modeling is that there are often separate programs used to model the primary and other loops in multiple loop systems. When the modeling requires no interaction between the loops, this method is adequate but in many of these scenarios an understanding of the interaction loops is crucial especially in the case of transients including accident scenarios. However, each loop is generally modeled individually and there is no feedback effect between loops. The purpose of this article is to demonstrate how this coupling between the primary and secondary system of a typical PWR can be performed. The primary and secondary sides of the PWR are modeled with Reactor Excursion and Leak Analysis Program (RELAP5) and Laboratory Virtual Instrument Engineering Workbench (LabVIEW) computer simulators respectively. Primary loop model includes a four loops PWR. The coupling between RELAP5 and LabVIEW has been executed with steady state and transients, in this case a loss of coolant accident (LOCA). The results of the coupling have been compared with the typical RELAP5 results without coupling.

Iconic Representation Extension of Event Sequence Diagram Framework for Nuclear Power Plant

2010 ◽  
Author(s):  
Haiyan Xie ◽  
Qi Cai ◽  
Yangwei Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Sequence Diagram ◽  
Event Sequence ◽  
Dynamic Reliability ◽  
Research Fields ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Accident Scenarios

As a dynamic reliability methodology, current Event Sequence Diagram (ESD) framework becomes an important tool to model the dynamic situations. It has been employed in the risk analysis in nuclear and space industries. It has been proved that ESD can be helpful for the operators of nuclear power plant to understand the accident scenarios and monitor progress during accident evolution. Despite this, there are some limits for the ESD framework in the practical application, such as the icons in the ESD framework are inadequate, so many dynamic situations are difficulty or complicated to describe. As for nuclear power plant, this paper proposes to extend its current iconic system to simpler and better model the dynamic process. Based on current ESD icons, we have created many other icons to represent the dynamic situations of the system and the component according to the characteristics of nuclear power plant. Some complex situations and interactions in the operation and accident are expressed by only one or two icons, so the system dynamics and accident evolution can be clearly described and simply translated into ESD models. Then the loss-of-coolant accident is used as the example to show the application of extended icons. These new icons greatly reduce the size of ESD model, which is well for its building and quantitative analysis. Most of the extensional ESD icons are also suitable for complex systems in other research fields and the ESD methodology may become popular for dynamic reliability analysis in the industry application.

scholarly journals Estimation of Dietary Intake of Radionuclides and Effectiveness of Regulation after the Fukushima Accident and in Virtual Nuclear Power Plant Accident Scenarios

2018 ◽  
Vol 15 (8) ◽  
pp. 1589 ◽  
Author(s):  
Michio Murakami ◽  
Takao Nirasawa ◽  
Takao Yoshikane ◽  
Keisuke Sueki ◽  
Kimikazu Sasa ◽  
...  
Keyword(s):  
Cost Effectiveness ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Dietary Intake ◽  
Emergency Preparedness ◽  
Nuclear Power ◽  
Food Distribution ◽  
Fukushima Accident ◽  
Accident Scenarios ◽  
Effectiveness Of Regulation

Evaluation of radiation exposure from diet is necessary under the assumption of a virtual accident as a part of emergency preparedness. Here, we developed a model with complete consideration of the regional food trade using deposition data simulated by a transport model, and estimated the dietary intake of radionuclides and the effectiveness of regulation (e.g., restrictions on the distribution of foods) after the Fukushima accident and in virtual accident scenarios. We also evaluated the dilution factors (i.e., ratios of contaminated foods to consumed foods) and cost-effectiveness of regulation as basic information for setting regulatory values. The doses estimated under actual emission conditions were generally consistent with those observed in food-duplicate and market-basket surveys within a factor of three. Regulation of restricted food distribution resulted in reductions in the doses of 54–65% in the nearest large city to the nuclear power plant. The dilution factors under actual emission conditions were 4.4% for radioiodine and 2.7% for radiocesium, which are ~20 times lower than those used in the Japanese provisional regulation values after the Fukushima accident. Strict regulation worsened the cost-effectiveness for both radionuclides. This study highlights the significance and utility of the developed model for a risk analysis of emergency preparedness and regulation.

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.

The Sensitivity of the Addition of Low-Low Pressurizer Pressure in Coincidence With High Containment Pressure on M310 Type Nuclear Power Plant

2016 ◽  
Author(s):  
Qiu Yanfei
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Deterministic Analysis ◽  
Limit Value ◽  
Loss Of Coolant Accident ◽  
Stop Time ◽  
Design Basis Accident ◽  
Operator Action ◽  
Containment Pressure

Due to the new security system that the operator intervention is assumed to occur in 20 minutes is not acceptable, for the current M310 type nuclear power plant. The loss of coolant accident with Intermediate breaks in primary loop is the only one of design basis accident which need operator action in 20 minutes. For certain size break, the consequences are very sensitive to the pump stop time. According to deterministic analysis that for a certain size break, if stop the pump in 20 minutes after accident, the peak cladding temperature will exceed the limit value of 1204°C. Therefore, it is necessary to add low-low pressurizer pressure in coincidence with high containment pressure signal to stop pump automatically on M310 type nuclear power plant.

Study and Design of Monitoring System of Reactor Coolant Pump in Nuclear Power Plant

2010 ◽  
Author(s):  
Lei You ◽  
Fuchun Sun ◽  
Pan He ◽  
Hongkun Xu ◽  
Fang Fang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Monitoring System ◽  
Virtual Instrument ◽  
Nuclear Power ◽  
Failure Process ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Instrument Technology ◽  
The Fourier Transform

In this paper, we develop a monitoring system of reactor coolant pumps in nuclear power plant (CPS). The safe running of reactor coolant pump is important for nuclear power plant. Based on the Fourier transform (FT) and some algorithm, The data collected from the pump are analyzed. Once the accident happens, it would cause unimaginable outcome. The system will be jumped to failure process mode when the pump has something wrong. The advanced VXI and virtual instrument technology are applied to system, and the reactor coolant pump will be monitored overall so as to assure that the reactor coolant pump runs in safe, which has a significant value to secure the safe operation and reliability of the nuclear plant. The monitoring system will help the operators find fault of reactor coolant pump.

Sign in / Sign up

Export Citation Format

Share Document