Author(s)

Pedro Ernesto Umbehaun, Walmir Maximo Torres, José Antonio Batista Souza, Mitsuo Yamaguchi, Antonio Teixeira e Silva, Roberto Navarro de Mesquita, Nikolas Lymberis Scuro, Delvonei Alves de Andrade

Nuclear and Energy Research Institute (Instituto de Pesquisas Energéticas e Nucleares), Säo Paulo, Brazil.

This paper presents the sequence of activities to improve the thermal hydraulic analysis of the IEA-R1 research reactor to operate in safe conditions after power upgrade from 2 to 5 MW and core size reduction from 30 to 24 fuel assemblies. A realistic analysis needs the knowledge of the actual operation conditions (heat flow, flow rates) beyond the geometric data and the uncertainties associated with manufacturing and measures. A dummy fuel assembly was designed and constructed to measure the actual flow rate through the core fuel assemblies and its pressure drop. First results showed that the flow distribution over the core is nearly uniform. Nevertheless, the values are below than the calculated ones and the core bypass flow rate is greater than those estimated previously. Based on this, several activities were performed to identify and reduce the bypass flow, such as reduction of the flow rate through the sample irradiators, closing some unnecessary secondary holes on the matrix plate, improvement in the primary flow rate system and better fit of the core components on the matrix plate. A sub-aquatic visual system was used as an important tool to detect some bypass flow path. After these modifications, the fuel assemblies flow rate increased about 13%. Additional tests using the dummy fuel assembly were carried out to measure the internal flow distribution among the rectangular channels. The results showed that the flow rate through the outer channels is 10% - 15% lower than the internal ones. The flow rate in the channel formed between two adjacent fuel assemblies is an estimated parameter and it is difficult to measure because this is an open channel. A new thermal hydraulic analysis of the outermost plates of the fuel assemblies takes into account all this information. Then, a fuel design modification was proposed with the reduction of 50% in the uranium quantity in the outermost fuel plates. In order to avoid the oxidation of the outermost plates by high temperature, low flow rate, a reduction of 50% in the uranium density in the same ones was shown to be adequate to solve the problem.

Nuclear Research Reactor, Uranium Reduction, Thermal Hydraulic Analysis, Flow Measurement, Dummy Fuel Assembly

Umbehaun, P. , Torres, W. , Souza, J. , Yamaguchi, M. , e Silva, A. , de Mesquita, R. , Scuro, N. and de Andrade, D. (2018) Thermal Hydraulic Analysis Improvement for the IEA-R1 Research Reactor and Fuel Assembly Design Modification. World Journal of Nuclear Science and Technology, 8, 54-69. doi: 10.4236/wjnst.2018.82006.