Many reactor safety simulation codes for nuclear power plants (NPPs) have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA) is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR). RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches) is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

The RELAP5 code, designed to predict the behavior of reactor systems during accident conditions, is used widespread over the world. This work aims to show and describe the RELAP-MV graphical software developed using computer language (XML) and Visualized Modularization software technology to recognized best estimated transient simulation program of Light water reactor, in combination with new options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. RELAP5 code is complex and inconvenient for utilizing method of data cards and close logic relationship of data in input file. The main purpose of developing RELAP5-MV is to simplify progress and increase simulation efficiency. Traditional modeling method and modular modeling method are supported with RELAP5-MV to achieve aims of device and system simulation. For traditional modeling method, all kinds of components are developed such as single volume, single junction, pipe, branch, time dependent volume, etc. For modular modeling method, the module library is established in the software. The library packages include the main system equipments of primary and secondary loops such as reactor core, U tube steam generator, once through steam generator, pump, pressurizer, steam turbine, condenser, heat exchanger, deaerator, etc. in a pressurized water reactor, which can be analyzed and modeled in details. From the library the capabilities are easy to select icons interface from the library packages. The analysis results show that the software can effectively simulate nuclear power system by RELAP5. Plot and data binding function is supported for post-processing of calculation result. Personal computer interface of RELAP5-MV makes it more convenient, fast and visualized in simulation system establishing process. Performance Relap5 related analysis activities, such as creating and modifying input file, viewing component division figures and generating output files can be realized by RELAP5-MV. The interactive simulation interface feature allows the users to simulate specific reactor transients and accidents — such as LOCA, LOFA, scram, etc. Accuracy and reliability of RELAP5-MV have already been confirmed by simulating main coolant system of Pressurize Water Reactor (PWR) and modeling efficiency increases significantly by using RELAP5-MV. Visualization modeling, analysis and computational simulation for thermal hydraulic analysis of nuclear reactor can not only lower the RELAP5 threshold but also improve the efficiency of nuclear science research greatly, and also promoting the development of related research in RELAP5 safety analysis. RELAP5-MV can give an approach to build, verify and assess simulation design of reactor power system.

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.