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Studies on Capacity Expansion of Fuel Plants for Nuclear Research Reactors

World Journal of Nuclear Science and Technology 2018年第2期8卷 38-53页

作者： Miguel Luiz Miotto Negro Michelangelo Durazzo Marco Aurélio de mesquita Elita Fontenele Urano de Carvalho roberto navarro de mesquita delvonei Alves de Andrade Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) S&atilde o Paulo Brazil Escola Politecnica da Universidade de Sao Paulo (Polytechnic School of Sao Paulo University) S&atilde o Paulo Brazil

The demand for nuclear fuel for research reactors is rising worldwide. Thus, the production facilities of this kind of fuel need reliable guidance on how to augment their production in order to meet the increasing demand efficiently and safely. We proposed a specific procedure for increasing production capacity. That procedure was tested with data from a real plant, which produces plate-type fuel elements loaded with LEU U3Si2-Al fuel. The test was made by means of discrete event simulation, and the results indicated the proposed procedure is efficient in raising production capacity.

关键词： Fabrication of Uranium Silicide Fuel Plate-Type Fuel Elements Nuclear Research Reactors Production Capacity Expansion

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Thermal Hydraulic Analysis Improvement for the IEA-R1 Research Reactor and Fuel Assembly design Modification

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World Journal of Nuclear Science and Technology 2018年第2期8卷 54-69页

作者： 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) S&auml o 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 th

关键词： Nuclear Research Reactor Uranium Reduction Thermal Hydraulic Analysis Flow Measurement Dummy Fuel Assembly

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