ADIMIR DOS SANTOS

ADIMIR DOS SANTOS

(Fonte: Lattes)

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Possui graduação em Bacharelado Em Física pela Universidade de São Paulo (1975), mestrado em Reatores Nucleares de Potência e Tecnologia do Com pelo Instituto de Pesquisas Energéticas e Nucleares (1978) e doutorado em Nuclear Engineering pela University of Wisconsin – Madison (1984). Atualmente é PESQUISADOR TITULAR III do Instituto de Pesquisas Energéticas e Nucleares, professor titular da Universidade de São Paulo e Revisor de periódico da Progress in Nuclear Energy. Tem experiência na área de Engenharia Nuclear, com ênfase em Tecnologia dos Reatores. Atuando principalmente nos seguintes temas: SENSITIVITY ANALYSIS, TRANSMUTATION, THORIUM, U-233 BREEDING. (Texto extraído do Currículo Lattes em 28 set. 2021)

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Stochastic modeling of a neutron imaging center at the Brazilian Multipurpose Reactor
2024 - OLIVEIRA, L.P. de; SOUZA, A.P.S.; GENEZINI, F.A.; SANTOS, A. dos
Neutron imaging is a non-destructive technique for analyzing a wide class of samples, such as archaeological or industrial material structures. In recent decades, technological advances have had a great impact on the neutron imaging technique, which has meant an evolution from simple radiographs using films (2D) to modern tomography systems with digital processing (3D). The 5 MW research nuclear reactor IEA-R1, which is located at the Instituto de Pesquisas Energéticas e Nucleares (IPEN) in Brazil, possesses a neutron imaging instrument with 1.0 × 106 𝑛∕𝑐𝑚2 𝑠 in the sample position. IEA-R1 is over 60 years old and the future of neutron science in Brazil, including imaging, will be expanded to a new facility called the Brazilian Multipurpose Reactor (RMB, Portuguese acronym), which will be built soon. The new reactor will house a suite of instruments at the Neutron National Laboratory, including the neutron imaging facility, viz., Neinei. Inspired by recent author’s works, we model the Neinei instrument through stochastic Monte Carlo simulations. We investigate the sensitivity of the neutron imaging technique parameter (𝐿∕𝐷 ratio) with the neutron flux, and the results are compared to data from the Neutra (PSI), Antares (FRM II), BT2 (NIST) and DINGO (OPAL) instruments. The results are promising and provide avenues for future improvements.
A reactivity meter with uncertainties
2024 - SANTOS, ADIMIR dos
Three sets of effective kinetics parameters were measured and evaluated for the fuel rod core of the IPEN/MB-01 research reactor. A correlation matrix among all these effective kinetics parameters is taking into account to reduce the inferred reactivity uncertainties. This work considers the propagation of the uncertainties inherent in these sets of effective delayed neutron parameters to the inferred reactivities. The comparisons to the reactivity and its uncertainty values produced by the Inhour Equation came into a very good agreement and reinforces the validity of the whole procedure for the development of this reactivity meter with uncertainties. The reduction of the reactivity uncertainty values compared to previous works is significant. The analyses of some specific cases of the evaluation IPEN(MB01)-LWR-RESR-015 reveal that its reactivity values are in accordance with those of this work. However, its reactivity uncertainty values for positive reactivities seems to cancel the uncorrelated component of the uncertainties.
Multigroup covariance matrix self-shielding effects for thermal reactors fueled with slightly enriched uranium
2023 - SANTOS, ADIMIR dos; GALIA, MARCUS V.C.; LEAL, LUIZ
A numerical approach has been successfully developed to treat the self-shielding effects in the multigroup cross section covariance matrices of thermal reactors fueled with slightly enriched uranium. The procedure employs the coupled NJOY/AMPX-II systems developed at IPEN and the 238U resonance parameter covariance data from JENDL 3.3. Only the first two most important 238U resonances are under analyses. The direct and indirect effects of the 238U resonance self-shielding is taken into account. The effect of the change in the cross section is called the direct effect and that of the neutron flux due to change in the cross section is called the indirect effect. The keff uncertainty analyses applied to the IPEN/MB-01 reveal that the self-shielding effects both direct and indirect have an important bearing on the multigroup covariance matrix as well as on the keff uncertainty. Also, the indirect effects account for nearly 44% of the total uncertainty. The ERRORR module of the NJOY system is in severe disagreement to the developed method because it considers only the direct effect in the multigroup cross section covariance matrix. Such results underline the application dependence of multi-group cross section covariance matrix, and that ENDF FILE 33 content must be corrected due to the resonance self-shielding effects mainly for applications in thermal reactor fueled with slightly enriched uranium.
Comissionamento: obtenção experimental da curva de calibração e da reatividade integral das barras de controle do Reator IPEN/MB-01 - núcleo contendo elementos combustíveis tipo placa
2023 - BITELLI, ULYSSES D.; SANTOS, ADIMIR dos
Este relatório apresenta o procedimento experimental [1] e os resultados obtidos para calibração das quatro barras de controle do novo núcleo do Reator IPEN/MB-01 [2] contendo elementos combustíveis tipo placa, correlacionando reatividade com o trecho de barra retirado ou inserido no núcleo (reatividade diferencial) e, por conseguinte, obtendo a reatividade total inserida (valor integral das barras de controle), através do somatório dos valores diferenciais de reatividade. Os valores de reatividade integral das quatro barras de controle de háfnio do núcleo contendo elementos combustíveis tipo placa obtidos experimentalmente são dados a seguir. BC#1 = (3371,99 ± 34,38) pcm, BC#2 = (3780,89 ± 39,45) pcm, BC#3 = (3309,78 ± 32,18) pcm, BC#4 = (3771,08 ± 37,52) pcm. As curvas ajustadas de calibração das barras de controle podem ser vistas nos apêndices para as quatro barras de controle e foram obtidas através do ajuste [3] da equação de Boltzmann dada abaixo aos valores diferencias de reatividade obtidos para cada trecho de barra retirado/inserido. 𝑦 = 𝐴2 + (𝐴1 − 𝐴2)/(1 + 𝑒^((𝑥 − 𝑥0)/𝑑𝑥) ). Os coeficientes da Equação de Boltzmann são dados na Tabela 1.
Simulating Araponga
2022 - SOUZA, A.P.S.; OLIVEIRA, L.P. de; GENEZINI, F.A.; SANTOS, A. dos
The Brazilian Multipurpose Reactor (RMB) is a fundamental project that aims to turn Brazil into a self-sufficient country in the production of radioisotopes and radiopharmaceuticals to supply the Unified Health System (SUS) as much as the private institutions. In addition, the RMB project describes other applications as irradiation and testing of nuclear fuels and structural material analysis, for instance. There are many techniques in the project to study structural aspects of materials, where neutron diffraction represents one of the priorities for implementation. This technique will take place mainly on two diffractometers on Thermal Neutron Guide 1 (TG1), namely Araponga, a high-resolution diffractometer, and Flautim, a high-intensity diffractometer. In this work, we study the performance of the Araponga diffractometer through McStas simulations with input produced by the MCNP code of the RMB core. We investigate the neutron flux values considering a state-of-art high-resolution diffractometer, and the results are promising since some simulated scenarios present values compatible with high-intensity devices.
MEA - Modified Energy Amplifier proposal
2001 - MAIORINO, J.R.; PEREIRA, S.A.; SANTOS, A.; SILVA, A.T.
Recently Rubbia et al proposed a conceptual design of an Accelerator Driven System, known as Energy Amplifier (EA), as an advanced innovative reactor which utilizes a spallation neutron source induced by protons, from a Cyclotron or Linac, in a subcritical array imbibed in liquid lead coolant. Besides of being breeder and waste burner, the conceptual design generates energy and allows the use of Thorium as fuel. This paper introduces some qualitative changes in the Rubbia's concept. More than one point of spallation is proposed in order to reduce the requirement of proton energy and current of the accelerator, and mainly to make a flatter power density distribution. The subcritical core, which in the Rubbia's concept is an hexagonal array of pins immersed in a liquid lead coolant, is replaced by a concept of a solid lead calandria with the fuel elements in channels cooled by Helium. This concept allows on line refueling or shuffling, and the utilization of a direct thermodynamic cycle (Brayton), which is more efficient than a vapor cycle. Although the calculations to demonstrate the feasibility of the MEA concept are underway and not yet finished, these ideas do not violate the basic physics of the EA, but reduces requirement in the accelerator complex, which is more realistic and economical in today accelerators technology. Finally, the utilization of He as coolant, compared with liquid Pb, is more realistic since the gas cooled reactors technology is well established and more efficient from the thermodynamic view, allowing simplification and the utilization in high temperature process, like hydrogen generation.
Simulating Araponga
2021 - SOUZA, A.P.S.; OLIVEIRA, L.P. de; GENEZINI, F.A.; SANTOS, A. dos
Reactor noise experiments considering high frequencies in the IPEN/MB-01 reactor
2021 - SANTOS, ADIMIR dos; SANTOS, DIOGO F. dos
Reactor noise experiments at high frequencies (up to 100 kHz) in subcritical configurations have been performed at the IPEN/MB-01 research reactor facility. The core configuration considered a short version of the IPEN/MB-01 core in a 24 × 26 rectangular array of fuel rods. The subcritical configurations considered the control banks totally withdrawn and the moderator poisoned with H3BO3. The boron concentrations were: 286.8 ± 10 and 578.6 ± 10 ppm. The pulses of two 3He detectors in the reflector region were summed and inserted into the IPEN/MB-01 Correlator and the APSD (Auto Power Spectral Density) was inferred through a mathematical model. The analysis reveals that the APSD in this frequency range is best described by a four-mode decay model. According to the two-region two-group (thermal and fast) kinetic model developed in this work, the first two decay modes describe the thermal group and the other two describe the fast group. After a tedious and severe analyses of the least-square fit of the experimental data, it was concluded that the kinetic behavior of the thermal and fast neutrons can be considered uncoupled. The analysis of the experimental data is still in progress and only the thermal group of the case of 286.8 ppm of boron was analyzed and some parameters could be inferred. The most important one so far is the prompt neutron generation time in the core region that could be inferred with a good level of accuracy.
Prediction of the power peaking factor in a boron-free small modular reactor based on a Support Vector Regression model and control rod bank positions
2021 - SANCHEZ, PRISCILA P.; SANTOS, ADIMIR dos
In order to ensure safety in a nuclear power plant, operation and protection systems must take into account safety parameters, whether to guide operators or to trip the reactor in emergency cases. Especially in a boron-free small modular reactor (SMR) where reactivity and power are controlled exclusively by rod banks, the power distribution is mostly influenced by its movements affecting the power peaking factor (PPF), which is an important parameter to be considered. The PPF relates the maximum local linear power density to the average power density in a fuel rod indicating a high neutron flux that can cause fuel rod damage. In this technical note, 2117 samples from simulations of an idealized boron-free SMR controlled exclusively by rod banks were used to generate a Support Vector Machine (SVM) model capable of estimating the PPF as a function of control rod bank positions. Such model could be used to predict the maximum PPF in the reactor core by carrying out simple calculation. Residing in a SVM parameter grid search and a 10-cross-validation process in the training set to reach an optimized and robust model, the results have shown a root-mean- squared error of about 0.1% consistent for both training and testing sets.
Potential advantages of molten salt reactor for merchant ship propulsion
2021 - FREITAS NETO, LUIZ G. de; FREIRE, LUCIANO O.; SANTOS, ADIMIR dos; ANDRADE, DELVONEI A. de
Operating costs of merchant ships, related to fuel costs, has led the naval industry to search alternatives to the current technologies of propulsion power. A possibility is to employ nuclear reactors like the Russian KLT-40S, which is a pressurized water reactor (PWR) and has experience on civilian surface vessels. However, space and weight are critical factors in a nuclear propulsion project, in addition to operational safety and costs. This work aims at comparing molten salt reactors (MSR) with PWR for merchant ship propulsion. The present study develops a qualitative analysis on weight, volume, overnight costs, fuel costs and nuclear safety. This work compares the architecture and operational conditions of these two types of reactors. The result is that MSR may produce lower amounts of high-activity nuclear tailings and, if it adopts the U233-thorium cycle, it may have lower risks of proliferating nuclear weapons. Besides proliferation issues, this 4th generation reactor may have lower weight, occupy less space, and achieve the same levels of safety with less investment. Thus, molten salt regenerative reactors using the U233-thorium cycle are potential candidates for use in ship propulsion.