RELAP5/SCDAPSIM model development for AP1000 and verification for large break LOCA
| Autor: | Chris M. Allison, Ashok Khanna, A.K. Trivedi, Prabhat Munshi |
|---|---|
| Rok vydání: | 2016 |
| Předmět: |
Nuclear and High Energy Physics
Engineering 020209 energy Nuclear engineering Flow (psychology) 02 engineering and technology 01 natural sciences 010305 fluids & plasmas Deck law.invention Thermal hydraulics law Pressurizer 0103 physical sciences 0202 electrical engineering electronic engineering information engineering General Materials Science Safety Risk Reliability and Quality Waste Management and Disposal computer.programming_language business.industry Mechanical Engineering Pressurized water reactor TRAC Structural engineering Natural circulation Nuclear Energy and Engineering business computer Loss-of-coolant accident |
| Zdroj: | Nuclear Engineering and Design. 305:222-229 |
| ISSN: | 0029-5493 |
| Popis: | The AP1000 is a Westinghouse 2-loop pressurized water reactor (PWR) with all emergency core cooling systems based on natural circulation. Its core design is very similar to a 3-loop PWR with 157 fuel assemblies. Westinghouse has reported their results of the safety analysis in its design control document (DCD) for a large break loss of coolant accident (LOCA) using WCOBRA/TRAC and for a small break LOCA using NOTRUMP. The current study involves the development of a representative RELAP5/SCDASIM model for AP1000 based on publically available data and its verification for a double ended cold leg (DECL) break in one of the cold legs in the loop containing core makeup tanks (CMT). The calculated RELAP5/SCDAPSIM results have been compared to publically available WCOBRA–TRAC and TRACE results of DECL break in AP1000. The objective of this study is to benchmark thermal hydraulic model for later severe accident analyses using the 2D SCDAP fuel rod component in place of the RELAP5 heat structures which currently represent the fuel rods. Results from this comparison provides sufficient confidence in the model which will be used for further studies such as a station blackout. The primary circuit pumps, pressurizer and steam generators (including the necessary secondary side) are modeled using RELAP5 components following all the necessary recommendations for nodalization. The core has been divided into 6 radial rings and 10 axial nodes. For the RELAP5 thermal hydraulic calculation, the six groups of fuel assemblies have been modeled as pipe components with equivalent flow areas. The fuel including the gap and cladding is modeled as a 1d heat structure. The final input deck achieved all steady state thermal hydraulic conditions as reported in the DCD. The analysis has been performed for the primary safety criteria, the peak clad temperature (PCT) as it is well established for a PWR that oxidation and hydrogen generation do not violate the safety criteria as long as PCT is under the safe limit. Results from this study show that the calculated value for the PCT is 1229 K well below the acceptance criteria of 1477 K, lower than the DCD value of 1311 K and higher than the TRACE value of 1186 K. |
| Databáze: | OpenAIRE |
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