Please cite this article in press as: Hainoun, A., et al., International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor. Nucl. Eng. Des. (2014), http://dx.doi.org/10.1016/j.nucengdes.2014.06.041 ARTICLE IN PRESS G Model NED-7997; No. of Pages 18 Nuclear Engineering and Design xxx (2014) xxx–xxx Contents lists available at ScienceDirect Nuclear Engineering and Design jou rn al hom ep age: www.elsevier.com/locate/nucengdes International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor A. Hainoun a,∗ , A. Doval b , P. Umbehaun c , S. Chatzidakis d , N. Ghazi a , S. Park e , M. Mladin f , A. Shokr g a Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus, Syrian Arab Republic b Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro, Argentina c Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP, Brazil d School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907, United States e Research Reactor Design & Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute, Republic of Korea f Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges, Romania g Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna, Austria h i g h l i g h t s • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. a r t i c l e i n f o Article history: Received 14 January 2014 Received in revised form 26 June 2014 Accepted 29 June 2014 a b s t r a c t In the framework of the IAEA Coordination Research Project on “Innovative methods in research reac- tor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reac- tor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) determinis- tic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 mea- suring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety analysis codes that comprise of CATHARE, RELAP5, MERSAT and PARET. The code RELAP5 was used independently by four of the participating teams and therefore the user effect and its impact on the code results can be characterized. The benchmark results demonstrate that most of the codes have the capability to correctly predict the SS case. However, for the LOFA case the simulation results show discrepancies to the measurement although the majority of the applied codes predict a qualitative correct time evolution of the corresponding transients for the coolant and clad temperatures. It is noted that the peak temperatures and the gradients around them are predicted conservatively. The quantitative assessments of benchmark results indicate different amounts of discrepancy between ∗ Corresponding author. Tel.: +963 112132580; fax: +963 6111926/7. E-mail address: pscientific2@aec.org.sy (A. Hainoun). http://dx.doi.org/10.1016/j.nucengdes.2014.06.041 0029-5493/© 2014 Elsevier B.V. All rights reserved.