This work presents the thermal hydraulic simulation of the Brazilian multipurpose reactor (RMB) using a RELAP5/MOD3.3 model. Beyond steady state calculations, three transient cases of loss of flow accident (LOFA) in the primary cooling system have been simulated. The RELAP5 simulations demonstrate that after all initiating events, the reactor reaches a safe new steady state keeping the integrity and safety of the core. Moreover, a sensitivity study was performed to verify the nodalization behavior due to the variation of the thermal hydraulic channels in the reactor core. Transient calculations demonstrate that both nodalizations follow approximately the same behavior. 1. Introduction The Brazilian Nuclear Energy Commission (Comiss？o Nacional de Energia Nuclear—CNEN) is leading the project of the Brazilian Multipurpose Reactor (RMB) envisaged to be projected, constructed, and operated to attend the present Brazilian need for a multipurpose neutron source, which will be able to supply the demand of radioisotopes, carry out material tests, and develop scientific, commercial, and medical applications with the use of neutron beams. The RMB will have three main functions: radioisotope production (mainly molybdenum), fuel, and material irradiation testing to support the Brazilian nuclear energy program. Moreover, it will provide neutron beams for scientific and applied research. Among the different types of research nuclear reactors, the open pool reactors are the most common and the most used, because of their great versatility, easy operation, and safety. The reactors Osiris in France, and mainly the Australian research reactor OPAL (Open Pool Australian Lightwater Reactor) projected by Argentina and built in Australia are being used as initial references for the RMB project. In the present work, a nodalization for the RMB core using the RELAP5/MOD3.3 and the most important components of the pool loop and core loop circuits are presented. Loss of flow accidents have been performed for two types of nodalization changing the number of core thermal hydraulic channels to verify the sensitivity study. 1.1. The Use of RELAP5 for Research Reactors Simulation The thermal hydraulic system code RELAP5 has been developed to best estimate transient simulation of light water power reactor systems during postulated accidents. Recent works have demonstrated that the code can be also used with good predictions for thermal hydraulic analysis of research reactors as it can be verified in the present literature [1–8]. Most of the research reactors operate at low-pressure
References
[1]
A. R. Antariksawan, M. Q. Huda, T. Liu, J. Zmitkova, C. M. Allison, and J. K. Hohorst, “Validation of RELAP/SCAPSIM/MOD3. 4 for research reactor applications,” in Proceedings of the 13th International Conference on Nuclear Engineering, pp. 1–8, Beijing, China, 2005.
[2]
C. B. Davis, “Applicability of RELAP5/MOD3. 2 to research reactors,” in Proceedings of the IAEA Regional Training Workshop on Safety Analysis Methodology and Computer Code Utilization, KINS, Daejeon, Republic of Korea, 2002.
[3]
B. Di Maro, F. Pierro, A. Bousbia Salah, and F. D'Auria, “Analysis of a pump trip in a typical research reactor by RELAP5/MOD 3.3,” in Proceedings of the International Congress on Advances in Nuclear Power Plants (ICAPP '03), Cordoba, Spain, 2003.
[4]
A. L. Costa, P. A. L. Reis, C. Pereira, M. A. F. Veloso, A. Z. Mesquita, and H. V. Soares, “Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model,” Nuclear Engineering and Design, vol. 240, no. 6, pp. 1487–1494, 2010.
[5]
T. Hamidouche, A. Bousbia-Salah, M. Adorni, and F. D'Auria, “Dynamic calculations of the IAEA safety MTR research reactor Benchmark problem using RELAP5/3.2 code,” Annals of Nuclear Energy, vol. 31, no. 12, pp. 1385–1402, 2004.
[6]
A. Khedr, M. Adorni, and F. d’Auria, “The effect of code user and boundary conditions on RELAP calculations of MTR research reactor transient scenarios,” Nuclear Technology and Radiation Protection, vol. 1, pp. 16–22, 2005.
[7]
B. Kon？ar and B. Mavko, “Modelling of low-pressure subcooled flow boiling using the RELAP5 code,” Nuclear Engineering and Design, vol. 220, no. 3, pp. 255–273, 2003.
[8]
P. A. L. Reis, A. L. Costa, C. Pereira et al., “Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor,” Annals of Nuclear Energy, vol. 37, no. 10, pp. 1341–1350, 2010.
[9]
IAEA, Research Reactor Core Conversion From the Use of High Enriched Uranium to the Use of Low Enriched Uranium Fuels Guidebook, IAEA-TECDOC-233, IAEA, Vienna, Austria, 1980.
[10]
R. B. Rothrock, “Thermal hydraulic bases for the safety limits and limiting safety system for the HFIR operating at 100 MW and 468 psig primary system, using specially selected fuel elements,” Tech. Rep. ORNL/TM-13694, 1998.
[11]
L. W. Hu and J. A. Bernard, “Thermal hydraulic criteria for the MIT research reactor safety limits calculations,” Transactions of the ANS, vol. 91, pp. 114–115, 1999.
[12]
J. S. Maulbetsch and P. Griffith, “A study of system-induced instabilities in forced convective flows with subcooled boiling,” MIT TR 5382-35, 1965.
[13]
R. H. Whittle and R. Forgan, “A correlation for the minima in the pressure drop versus flow-rate curves for sub-cooled water flowing in narrow heated channels,” Nuclear Engineering and Design, vol. 6, no. 1, pp. 89–99, 1967.
[14]
P. Saha and N. Zuber, “Point of net vapor generation and vapor void fraction in subcooled boiling,” in Proceedings of the 5th International Heat Transfer Conference, vol. 4, Tokyo, Japan, September 1974.
[15]
R. B. Duffey and E. D. Hughes, “Static flow instability onset in tubes, channels, annuli, and rod bundles,” International Journal of Heat and Mass Transfer, vol. 34, no. 10, pp. 2483–2496, 1991.
[16]
G. R. Warrier and V. K. Dhir, “Review of experimental and analytical studies on low pressure subcooled flow boiling,” in Proceedings of the 5th ASME/JSME Joint Thermal Engineering Conference, San Diego, Calif, USA, March 1999.
[17]
T. Hamidouche and A. Bousbia-Salah, “RELAP5/3.2 assessment against low pressure onset of flow instability in parallel heated channels,” Annals of Nuclear Energy, vol. 33, no. 6, pp. 510–520, 2006.
[18]
T. Hamidouche, A. Bousbia-Salah, E. K. Si-Ahmed, M. Y. Mokeddem, and F. D'Auria, “Application of coupled code technique to a safety analysis of a standard MTR research reactor,” Nuclear Engineering and Design, vol. 239, no. 10, pp. 2104–2118, 2009.
[19]
W. L. Woodruff, N. A. Hanan, R. S. Smith, and J. E. Matos, “A comparison of the PARET/ANL and RELAP5/MOD3. 3 codes for the analysis of IAEA benchmark transients,” in Proceedings of the International Meeting on Reduced Enrichment for Research and Test Reactors, Seoul, Republic of Korea, October 1996.
[20]
A. Hedayat, H. Davilu, and J. Jafari, “Loss of coolant accident analyses on Tehran research reactor by RELAP5/MOD3.2 code,” Progress in Nuclear Energy, vol. 49, no. 7, pp. 511–528, 2007.
[21]
C. Allison, J. K. Hohorst, A. Antariksawan, et al., “Validation of RELAP/SCDAPSIM/MOD3. 4 for research reactor applications,” Proceeding of the 13th International Conference on Nuclear Engineering (ICONE '05), Beijing, China, 2005.
[22]
IAEA, Safety Analysis of Research Reactors, Safety Reports Series no. 55, IAEA, Vienna, Austria, 2008, STI/PUB/1321.
[23]
ANSTO, “Preliminary Safety Analysis Report (PSAR). Replacement Research Reactor Facility,” Australian Nuclear Science and Technology Organisation (ANSTO), 2001.
[24]
IAEA, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, IAEA Safety Series no. 35-G1, IAEA, Vienna, Austria, 1994.
[25]
ANSTO, “Safety Analysis Report (SAR). Replacement Research Reactor Project,” Prepared by INVAP for Australian Nuclear Science and Technology Organisation (ANSTO), 2004.
[26]
ANSTO, “Probabilistic Safety Assessment (PSA) Summary for Public Release. Replacement Research Reactor Project,” Prepared by Australian Nuclear Science and Technology Organisation (ANSTO), 2005.
[27]
M. A. Navarro, F. L. A. Schweizer, and A. A. C. Santos, “Numerical investigation in the Brazilian multipurpose reactor pool,” in Proceedings of the International Nuclear Atlantic Conference (INAC '11), Belo Horizonte, Brazil, October 2011.
[28]
PLG, “Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor (HIFAR),” 1998.
[29]
IAEA, Component Reliability Data for Use in Probabilistic Safety Assessment, IAEA-TECDOC-478, IAEA, Vienna, Austria, 1988.
[30]
IAEA, Generic Component Reliability Data for Research Reactor PSA, IAEA-TECDOC-930, IAEA, Vienna, Austria, 1997.
[31]
CNEN, “Divis？o de Fabrica？？o e Projeto,” Restrict Document, CNEN, 2012.
[32]
CNEN, “Projeto preliminar do refletor do RMB,” Restrict Document RMB-10100-RD-010, 2011.
