RELAP5 Modeling of a Siphon Break Effect on the Brazilian Multipurpose Reactor

Humberto Vitor Soares; Antonio Belchior Jr.; Roberto Longo Freitas

doi:10.15392/bjrs.v8i3A.1563

Authors

Humberto Vitor Soares Amazônia Azul Tecnologias de Defesa S.A. -AMAZUL https://orcid.org/0000-0001-7272-9026
Antonio Belchior Jr. Instituto de Energia Elétrica e Nucleares https://orcid.org/0000-0003-3023-9074
Roberto Longo Freitas Centro de Coordenação RMB – RMB/CNEN https://orcid.org/0000-0001-9189-2726

DOI:

https://doi.org/10.15392/bjrs.v8i3A.1563

Keywords:

RMB, RELAP5, LOCA, Siphon Breaker

Abstract

This work presents the thermal-hydraulic simulation of the Brazilian Multipurpose Reactor (RMB) using the RELAP5/Mod3 code. The RMB will provide Brazil with a fundamental infrastructure for the national development on activities of the nuclear sector in the areas of social, strategic, industrial applications and scientific and technological development. A RELAP5/Mod3 code model was developed for thermal-hydraulic simulation of the RMB to analyze the phenomenology of the Siphon Breakers device (four flap valves in the cold leg and one open tube for the atmosphere in the hot leg) during a Loss of Coolant Accident (LOCA) at different points in the primary circuit. The Siphon Breaker device is an important passive safety system for research reactors in order to guarantee the water level in the core under accidental conditions. Different simulations were carried out at different location in the Core Cooling System (CCS) of the RMB, for example: LOCA before the CCS pumps with and without pump trip and LOCA after the CCS pumps and the heat exchanger. In all RELAP5/Mod3 code simulations, the Siphon Breaker device's performance after a LOCA was effective to allow enough air to enter the outlet pipe of the CCS in order to break the siphon effect and preventing the pool level from reaching the riser (chimney) and the RMB core discovering. In all cases, the reactor pool level stabilized at about 5.5 m after the end of the LOCA simulation and the fuel elements were kept underwater and cooled.

Views: 179
PDF Downloads: 357

Downloads

Download data is not yet available.

Author Biographies

Humberto Vitor Soares, Amazônia Azul Tecnologias de Defesa S.A. -AMAZUL

researcher at Amazônia Azul Tecnologias de Defesa S.A.

Antonio Belchior Jr., Instituto de Energia Elétrica e Nucleares

Resercher at Nuclear and Energy Research institute

Roberto Longo Freitas, Centro de Coordenação RMB – RMB/CNEN

Consultant at RMB Coordination Center

References

NAVARRO, M. A.; SCHWEIZER, F. L. A.; SANTOS A. A. C. Numerical investigation in the Brazilian Multipurpose Reactor pool. Proceedings of the International Nuclear Atlantic Conference (INAC 2011), Belo Horizonte, Brazil, 2011.

US NRC, UNITED STATES NUCLEAR REGULATORY COMMISSION. RELAP5/MOD3.3 Code Manuals. Idaho National Engineering Laboratory, NUREG/CR-5535, 2001.

ANTARIKSAWAN, A. R.; HUDA, M. Q.; LIU, T.; ZMITKOVA, J.; ALLISON, C. M.; HOHORST J. K. Validation of RELAP/SCAPSIM/MOD3.4 for research reactor applications. Proceedings of the 13th International Conference on Nuclear Engineering, Beijing, China, 2005.

DAVIS, B. C. Applicability of RELAP5/MOD3.2 to research reactors. Proceedings of the IAEA Regional Training Workshop on Safety Analysis Methodology and Computer Code Utilization, KINS, Daejeon, Republic of Korea, 2002.

DIMARO, B.; PIERRO, F.; BOUSBIA SALAH, A.; D’AURIA, F. Analysis of a pump trip in a typical research reactor by RELAP5/MOD 3.3. Proceedings of the International Congress on Advances in Nuclear Power Plants (ICAPP-03), Cordoba, Spain, 2003.

COSTA, A. L.; REIS, P. A. L.; PEREIRA, C.; VELOSO, M. A. F.; MESQUITA, A. Z.; SOARES, H. V.; Thermal-hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model. Nuclear Engineering and Design, v. 240, no. 6, p. 1487–1494, 2010.

HAMIDOUCHE, T.; BOUSBIA-SALAH, A.; ADORNI, M.; D'AURIA, F. Dynamic calculations of the IAEA safety MTR research reactor Benchmark problem using RELAP5/3.2 code. Annals of Nuclear Energy, v. 31, no. 12, p. 1385–1402, 2004.

KHEDR, A.; ADORNI, M.; D’AURIA, F. The effect of code user and boundary conditions on RELAP calculations of MTR research reactor transient scenarios. Nuclear Technology and Radiation Protection, v. 1, p. 16-22, 2005.

REIS, P. A. L.; COSTA, A. L.; PEREIRA, C.; et al. Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor. Annals of Nuclear Energy, v. 37, no. 10, p. 1341–1350, 2010.

KONCAR, B.; MAVKO, B. Modeling of low-pressure sub cooled flow boiling using the RELAP5 code. Nuclear Engineering and Design, v. 220, no. 3, p. 255–273, 2003.

HAMIDOUCHE, T.; BOUSBIA-SALAH A. RELAP5/3.2 assessment against low pressure onset of flow instability in parallel heated channels. Annals of Nuclear Energy, v. 33, no. 6, p. 510–520, 2006.

HAMIDOUCHE, T., BOUSBIA-SALAH, A., SI-AHMED, E. K.; MOKEDDEM, M. Y.; D'AURIA, F. Application of coupled code technique to a safety analysis of a standard MTR research reactor. Nuclear Engineering and Design, v. 239, no. 10, p. 2104–2118, 2009.

WOODRUFF, W. L.; HANAN, N. A.; SMITH, R. S.; MATOS, J. E. A comparison of the PARET/ANL and RELAP5/MOD3.3 codes for the analysis of IAEA benchmark transients. Proceedings of the International Meeting on Reduced Enrichment for Research and Test Reactors, Seoul, Republic of Korea, 1996.

HEDAYAT, A.; DAVILU, H.; JAFARI, J. Loss of coolant accident analyses on Tehran research reactor by RELAP5/MOD3.2 code. Progress in Nuclear Energy, v. 49, no. 7, p. 511–528, 2007.

ALLISON, C.; HOHORST, J. K.; ANTARIKSAWAN, A.; et al. Validation of RELAP/SCDAPSIM/MOD3.4 for research reactor applications. Proceeding of the 13th International Conference on Nuclear Engineering (ICONE 05), China, 2005.

BELCHIOR JR., A.; SOARES, H. V.; FREITAS, R. L. Validation of the Relap5 Code for the Simulation of the Siphon Break Effect in Pool Type Research Reactors. 2019 International Nuclear Atlantic Conference - INAC 2019, Santos, SP, Brazil, 21-25 October 2019.

SOARES, H. V. Avaliações de Segurança do Reator Multipropósito Brasileiro. Tese de Doutorado em Ciências e Técnicas Nucleares, Escola de Engenharia, Universidade Federal de Minas Gerais (UFMG), 2014.