Patrícia A.L. Reis

Neutron production evaluation from a ADS target utilizing the MCNPX 2.6.0 code

Accelerators Driven Systems (ADS) are an innovative type of nuclear system, which is useful for long-lived fission product transmutation and fuel regeneration. The ADS consist of a coupling of a sub-critical nuclear core reactor and a proton beam produced by particle accelerator. These particles are injected into a target for the neutrons production by spallation reactions. This target is of utmost importance for an ADS, representing the coupling of the accelerator and the sub-critical core. The determination of optimal materials for these targets is fundamental for the design of an ADS. The main characteristic of an ideal target is the high production of neutrons per incident proton. In this work are shown results for the neutron production of various types of targets using the MCNPX 2.6.0 code. Furthermore, it is performed a comparative study of transport models to describe the spallation reactions available in this code.

Experimental Investigation of the Onset of Subcooled Nucleate Boiling in an Open-Pool Nuclear Research Reactor

The IPR-R1 TRIGA nuclear research reactor at the Nuclear Technology Development Center (CDTN), in Belo Horizonte (Brazil), is a pool type reactor. The maximum core power is 250 kW cooled by natural circulation of light water and an open surface. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel/cladding (gap) and the cladding to coolant interfaces. The objective of the thermal and hydrodynamic projects of the reactors is to remove the heat safely, without producing excessive temperature in the fuel elements. The regions of the reactor core where boiling occurs at many different power levels can be determined from the heat transfer coefficient data. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. Experimental results indicated that subcooled pool boiling occurs at the cladding surface in the reactor core central channels at power levels in excess of 60 kW. However, due to the high heat transfer coefficient in subcooled boiling the cladding temperature is quite uniform along most of the active fuel rod region and do not increase very much with the reactor power. An operational computer program and a data acquisition and signal processing system were developed as part of this research project to allow on line monitoring of the operational parameters.

Safety Studies and General Simulations of Research Reactors Using Nuclear Codes

Antonella L. Costa1, Patricia A. L. Reis1, Clarysson A. M. Silva1, Claubia Pereira1, Maria Auxiliadora F. Veloso1, Bruno T. Guerra1, Humberto V. Soares1 and Amir Z. Mesquita2 1Departamento de Engenharia Nuclear – Escola de Engenharia Universidade Federal de Minas Gerais Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq 2Centro de Desenvolvimento da Tecnologia Nuclear/Comissao Nacional de Energia Nuclear – CDTN/CNEN Brasil

Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature.

Experimental Investigation of Thermal Hydraulics in the IPR-R1 TRIGA Nuclear Reactor

Rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, leading to a “nuclear power renaissance” in countries the world over. In Brazil, the nuclear renaissance can be seen in the completion of construction of its third nuclear power plant and in the governments decision to design and build the Brazilian Multipurpose research Reactor (RMB). The role of nuclear energy in Brazil is complementary to others sources. Presently two Nuclear Power Plants are in operation (Angra 1 and 2) with a total of 2000 MWe that accounts for the generation of approximately 3% of electric power consumed in Brazil. A third unity (Angra 3) is under construction. Even though with such relatively small nuclear park, Brazil has one of the biggest world nuclear resources, being the sixth natural uranium resource in the world and has a fuel cycle industry capable to provide fuel elements. Brazil has four research reactors in operation: the MB-01, a 0.1 kW critical facility; the IEA-R1, a 5 MW pool type reactor; the Argonauta, a 500 W Argonaut type reactor and the IPR-R1, a 100 kW TRIGA Mark I type reactor. They were constructed mainly for using in education, radioisotope production and nuclear research.

Research Reactor Analysis Using Thermal Hydraulic and Neutron Kinetic Coupling

Simulations of complex scenarios in nuclear power plants have been improved by the utilization of coupled thermal hydraulic (TH) and neutron kinetics (NK) system codes with the development of computer technology and new calculation methodology which made it possible to perform transport calculation schemes with accurate solutions. This paper presents a model for the IPR-R1 TRIGA research reactor using the RELAP5-3D 3.0.0 code. By using this code, a multi-dimensional neutron kinetics model based on the NESTLE code can be implemented also. In this way, during a 3D TH/NK coupled simulation, RELAP5-3D calls the appropriate NESTLE subroutines to perform the calculations. The development and the assessment of the thermal hydraulic RELAP5 code model for the IPR-R1 TRIGA have been validated for steady state and transient situations and the results were published in preceding works. The model has been adapted to RELAP5-3D code and was verified to point kinetic calculations. After this, adequate cross sections to the NK code were supplied using the WIMSD5 code. The results of steady state and transient calculations using the 3D neutron modeling to the IPR-R1 are being presented in this paper.Copyright