The Development of RELAP5 for Reactor System Analysis and Simulation Use of Visualized Modularization Software (RELAP-MV)

2016 ◽  
Author(s):  
Eltayeb Yousif ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Haoran Ju
Keyword(s):  
Power System ◽  
Steam Generator ◽  
System Analysis ◽  
Computational Simulation ◽  
Modeling Method ◽  
Input File ◽  
Pressurized Water ◽  
Water Reactor ◽  
Modular Modeling ◽  
Relap5 Code

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.

Risk-Informed Analysis for Accident Tolerant Fuel in Pressurized Water Reactors

2020 ◽  
Author(s):  
Zhegang Ma ◽  
Carlo Parisi ◽  
Cliff Davis ◽  
Sai Zhang ◽  
Hongbin Zhang
Keyword(s):  
System Analysis ◽  
Hydrogen Generation ◽  
Cooling System ◽  
Pressurized Water ◽  
Water Reactor ◽  
Core Damage ◽  
Systems Research ◽  
Research Activities ◽  
Near Term ◽  
Accident Scenarios

Abstract This paper presents the research activities performed by Idaho National Laboratory (INL) for the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk-Informed System Analysis (RISA) Pathway, Enhanced Resilient Plant (ERP) Systems research, using the probabilistic risk assessment (PRA) tool SAPHIRE and the deterministic best estimate tool RELAP5-3D for risk-informed analysis. The ERP research supports DOE and industry initiatives by developing Accident Tolerant Fuel (ATF), the Diverse and Flexible Coping Strategy (FLEX), and passive cooling system designs to enhance existing reactors’ safety features (both active and passive) and to substantially reduce operating costs of nuclear power plants (NPPs) through risk-informed approaches to analyze the plant enhancements and their characterization. The risk-informed analysis used SAPHIRE and RELAP5-3D to evaluate the risk impacts from near-term ATF (FeCrAl and Chromium-coated clads) on a generic Westinghouse three-loop pressurized water reactor (PWR) under the following accident scenarios: station blackout (SBO), loss of feedwater (LOFW), steam generator tube rupture (SGTR), loss-of-coolant accidents (LOCAs), locked rotor transient, turbine trip transient, anticipated transient without scram (ATWS), and main steam line break (MSLB). The RELAP5-3D simulations included the time to core damage, time to 0.5 kilograms hydrogen generation, and total hydrogen generation. The simulation results show there are modest gains of coping time (delay of time to core damage) due to efficacy of the near-term ATF designs in various accident scenarios. The risk benefits on behalf of the core damage frequency (CDF) brought by the ATF designs would be small for most of the scenarios. However, results revealing much less hydrogen being produced at the time of core damage show a clear benefit in adopting ATFs.

Development of Steam Generator Model Using RELAP5 to Simulate Two-Phase Natural Circulation for Pressurized Water Reactor

2014 ◽  
Author(s):  
Shinya Miyata ◽  
Satoru Kamohara ◽  
Wataru Sakuma ◽  
Hiroaki Nishi
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Injection System ◽  
Pressurized Water Reactor ◽  
Two Phase ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loss Of Coolant ◽  
Mass Inventory ◽  
Core Cooling

In typical pressurized water reactor (PWR), to cope with beyond design basis events such as station black out (SBO) or small break loss of coolant accident with safety injection system failure, injection from accumulator sustains core cooling by compensating for loss of coolant. Core cooling is sustained by single- or two-phase natural circulation or reflux condensation depending on primary coolant mass inventory. Behavior of the natural circulation in PWR has been investigated in the facilities such as Large Scale Test Facility (LSTF) which is a full-height and full-pressure and thermal-hydraulic simulator of typical four-loop PWR. Two steady-state natural circulation tests were conducted in LSTF at both high and low pressure. These two tests were conducted changing the primary mass inventory as a test parameter, while keeping the other parameters such as core power, steam generator (SG) pressure, and steam generator water level as they are. Mitsubishi Heavy Industries (MHI) plans new natural circulation tests to cover wider range of core power and pressure as test-matrix (including the previous LSTF tests) to validate applicability of the model in wider range of core power and pressure conditions including the SBO conditions. In this paper, the previous LSTF natural circulation tests are reviewed and the new test plan will be described. Additionally, MHI also started a feasibility study to improve the steam generator tube and inlet/outlet plenum model using the M-RELAP5 code [4]. Newly developed model gives reasonable agreement with the previous LSTF tests and applies to the new test conditions. The feasibility findings will also be described in this paper.

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