Daniel Artur Pinheiro Palma

The derivation of the doppler broadening function using frobenius method

An analytical approximation of the Doppler broadening function ψ(ξ,x) is proposed. This approximation is based on the solution of the differential equation for ψ(ξ,x) using the methods of Frobenius and parameters variation. The analytical form derived for ψ(ξ,x) in terms of elementary functions is very simple and precise. It can be useful for applications related to the treatment of nuclear resonances, mainly for calculations of multigroup parameters and resonances self-protection factors, the latter being used to correct microscopic cross section measurements by the activation technique.

Uncertainty assessment of the experimental thermal-hydraulic parameters for CDTN TRIGA research reactor

Experimental studies have been performed in the TRIGA Nuclear Reactor at Nuclear Technology Development Centre (CDTN), Brazil, to find out its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power. 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. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The uncertainty analysis on thermal hydraulics parameters is determined, basically, by the uncertainty of the reactors thermal power.

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.

Influence of geometry in TRIGA reactor criticality calculation and reactivity determination using Serpent 2 and MCNPX codes

The IPR-R1 TRIGA Mark I research reactor is located at the Nuclear Technology Development Centre (CDTN), in Belo Horizonte, Brazil. It is operating for more than 50 years and was successfully simulated before. However, new techniques and methods used in nuclear reactors analysis make a further simulation inevitable. In this manuscript, the computational model of an initial core of the IPR-R1 TRIGA reactor was developed employing two different Monte Carlo codes, MCNPX and Serpent 2, to simulate the neutronics behaviour. A new model is suggested, more complete, to improve the simulations results making the model more close the experimental data. This work explores how changes could be inserted in order to make the model closer to reality and if such participation would be noticeable in both codes used. The neutronic parameters obtained from these simulations performed in Serpent 2 are compared to MCNPX simulation results at the same conditions, and the results are compared with previous experimental data.

A human-machine interface for a TRIGA research reactor of Brazil

During seven years, the main operational parameters of the IPR-R1 TRIGA reactor of Nuclear Technology Development Centre (CDTN) at Belo Horizonte, Brazil, have been monitored and displayed online by using a data acquisition system developed for this reactor. Besides showing the real-time performance of the plant, the system stored the information in a computer hard disk, with an accessible historical database, in order to make the chronological information on reactor performance and its behaviour available to users. Some of the parameters stored are the control rod positions and reactivity, the reactor power, the fuel and water temperatures, the radiation levels, the primary cooling system flow, and the water pool level. Records of the reactor process variables are important for immediate or subsequent safety analyses to show the short and long-term trends, and to report the reactor operations to the organisation and external authorities. This paper describes the data acquisition system, and the electronic database developed for the IPR-R1 TRIGA reactor.

New Methods in Doppler Broadening Function Calculation

In all nuclear reactors some neutrons can be absorbed in the resonance region and, in the design of these reactors, an accurate treatment of the resonant absorptions is essential. Apart from that, the resonant absorption varies with fuel temperature, due to the Doppler broad‐ ening of the resonances (Stacey, 2001). The thermal agitation movement of the reactor core is adequately represented in microscopic cross-section of the neutron-core interaction through the Doppler Broadening function. This function is calculated numerically in modern sys‐ tems for the calculation of macro-group constants, necessary to determine the power distri‐ bution in a nuclear reactor. This function has also been used for the approximate calculations of the resonance integrals in heterogeneous fuel cells (Campos and Martinez, 1989). It can also be applied to the calculation of self-shielding factors to correct the meas‐ urements of the microscopic cross-sections through the activation technique (Shcherbakov and Harada, 2002). In these types of application we can point out the need to develop pre‐ cise analytical approximations for the Doppler broadening function to be used in the codes that calculates the values of this function. Tables generated from such codes are not conven‐ ient for some applications and experimental data processing.

An Approximation for the Interference Term Applied to the Calculation of the Average Scattering Cross Section Using Fourier Series

The calculation of the Doppler broadening function ψ(x,ξ) and of the interference term χ(x,ξ) are important in the generation of nuclear data. In a recent paper, Goncalves and Martinez proposed an analytical approximation for the calculation of both functions based in sine and cosine Fourier transforms. This paper presents new approximations for these functions, ψ(x,ξ) and χ(x,ξ), using expansions in Fourier series, generating expressions that are simple, fast and precise. Numerical tests applied to the calculation of scattering average cross section provided satisfactory accuracy.Copyright