Development of New MHI LOCA Analysis Code: MCOBRA/RELAP5-GOTHIC

2016 ◽  
Author(s):  
Masaaki Katayama ◽  
Tetsuya Teramae ◽  
Masatsugu Mizokami ◽  
Shinya Kosaka
Keyword(s):  
Heat Transfer ◽  
Current Flow ◽  
Safety Analysis ◽  
Code System ◽  
Steam Generators ◽  
Code Model ◽  
Integral Effect ◽  
Counter Current ◽  
Relap5 Code ◽  
Containment Integrity

Mitsubishi Heavy Industries, Ltd. (MHI) has developed the safety analysis code system, MCOBRA/RELAP5-GOTHIC, for large break LOCA, small break LOCA and containment integrity analyses of PWR. The code system contains three codes, MCOBRA code, M-RELAP5 code and GOTHIC code. MHI implemented some models and correlations to MCOBRA and M-RELAP5 code to improve prediction of important phenomena of LOCA event. MHI conducted a lot of analyses for separate effect tests and integral effect tests to confirm the applicability of the MHI’s safety analysis code to LOCA key phenomena, like core heat transfer, DNB, entrainment, break flow, counter-current flow limitation (CCFL), core froth level, and so on. The code model uncertainties are quantified to calculate core heat transfer break flow rate, loop seal phenomena; heat transfer in steam generators, and so on. The results show that the code system can predict the key phenomena well and can calculate LOCA event scenario. Consequently, it is confirmed that the new code system is applicable to the LOCA licensing analysis.

Comparative Thermal Analyses Between Theoretical Mode and RELAP5 Code Simulation for OTSG of a Small PWR

2020 ◽  
Author(s):  
Baihui Jiang ◽  
Zhiwei Zhou ◽  
Zhaoyang Xia ◽  
Qian Sun
Keyword(s):  
Heat Transfer ◽  
Heat Exchange ◽  
Nuclear Reactors ◽  
Heat Transfer Analysis ◽  
Tube Length ◽  
Outlet Temperature ◽  
Steam Generators ◽  
Inclined Tube ◽  
Heat Transfer Calculation ◽  
Relap5 Code

Abstract Due to the low nuclear safety risk, low initial investment cost and short construction period, integrated small nuclear reactors have received wide attention from all over the world. As advanced new type of nuclear reactors, the technologies of integrated small nuclear reactors are in the process of exploration and development. Steam generators are used as the heat transfer system for energy exchange between the primary and secondary circuit in reactors, and their heat transfer analysis is very important for reactor design and development. Due to the simple structure, strong heat exchange capacity and timely load following, Once-Through Steam Generators (OTSGs) are the mainly used steam generators in the design of integrated small nuclear reactors. RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LLC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. However, when RELAP5 models steam generators, only structural models related to straight pipes can be established, which is very inconvenient for the heat transfer research of Once-Through Steam Generators. In this paper, Once-Through Steam Generators with specified structural parameters are taken as the research object. The heat transfer calculation is performed on the simplified inclined tube models by RELAP5 code and the theoretical calculation of the spiral tube heat transfer models is also carried out. Comparing the steam outlet temperature on the primary and secondary sides, the heat exchange power, the average heat transfer coefficient and the tube length of different heat exchange zones under given primary and secondary side inlet fluid conditions, it is confirmed that the RELAP5 heat transfer calculation is verified for simplifying Once-Through Steam Generators with inclined tube models.

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