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This paper presents the results of the studies about the Fourth Generation Reactors developed at DENU-UFMG, such as the Very High Temperature Reactor (VHTR), Molten Salt Breeder Reactor (MSBR), Sodium Fast Reactor (SFR) and Gas-cooled Fast Reactor (GFR). These studies evaluate neutronic parameters of the nuclear system at steady state and during the burnup using computer models developed at DENU-UFMG using SCALE 6.0 (KENO-VI/ORIGENS), MCNPX 2.6.0 and MCNP5. The results show that the adopted methodologies confirm the capabilities of the codes to simulate specific situations in steady state or transient operating conditions.

ABSTRACT Purpose: To study the uptake capacity of cells from the reticuloendothelial system after irradiation with high-energy X-rays. Methods: Eighteen male Wistar rats were distributed in three groups: group A (n = 6): control, unirradiated animals studied alongside animals from group B; group B (n = 6) and group C (n = 6): animals irradiated and studied after 24 and 48 hours, respectively. The rats were anesthetized and placed on a 10 MV linear accelerator. Next, they were irradiated in the abdominal region, with 8 Gy. Twenty-four (groups A and B) and 48 hours later (group C), a colloidal carbon solution (1 mL/kg) was intravenously injected in the tail vein. Fifty minutes later, the spleens and livers were withdrawn and prepared to be studied. Kupffer cells and splenic macrophages containing carbon pigments were counted in an optical microscope. Arithmetic means were calculated for each group and compared among them. Results: X-rays were associated with a reduced number of Kupffer cells containing colloidal carbon, proliferation and enlargement of biliary ducts, hypoplasia, and hepatocyte necrosis. In the irradiated spleen, the colloidal carbon uptake was concentrated in the marginal zone around the white pulp, with an inexpressive uptake of pigments by macrophages from white and red pulps. Conclusions: The X-rays in the rat abdomen are associated with a reduction in the Kupffer cells uptake of colloidal carbon, hepatocyte disorders, bile duct proliferation, and splenic uptake of colloidal carbon concentrated in the marginal zone.

This study presented the recovery of uranium oxide and thorium oxide from aluminothermic slag from the metallurgical processing of columbite – a niobium/tantalum mineral with the presence of U3O8 and ThO2 – from a mining-industrial facility. The methodology consisted of sampling, comminution, and leaching using sulfuric acid. The head sample showed contents of (1.78 + 0.14) % for U3O8 and (3.66 + 0.04) % for ThO2. The metallurgical recovery reached values above 80% for the uranium oxide and above 70% for the thorium oxide for pH < 1.5 and process time greater than eight hours.

The need for energy by modern society is increasing. On the other hand, it is necessary to reduce costs and environmental impact. In this perspective, the recovery of uranium present in industrial waste from the processing of naturally occurring radioactive material (NORM) appears as a possible complement to the mining stage of the Nuclear Fuel Cycle. NORM's uranium recovery can reduce environmental liabilities and mineral processing costs (especially blasting, crushing, and grinding). The industrial residue of this study, a type of aluminothermic slag, comes from the metallurgical processing of columbite (niobium and tantalum mineral) and has a content, measured by X-ray fluorescence, of 1.78% of U3O8. This content is higher, for example than those found in Lagoa Real-BA (0.2% in rock) and Santa Quitéria-CE (0.1% in rock). Another material that will be studied is ThO2, which is also present in the slag with a content, measured by X-ray fluorescence, around 3.66%. The process parameters analyzed were pH of the solution, time, granulometry and percentage of solids. The metallurgical recovery of U3O8 reached a maximum value of 71,3% with pH = 1, time of 8 hours, 65 % percentage of solids, and 200 µm of granulometry. The metallurgical recovery of ThO2 reached a maximum value of 69,7% with pH = 1, time of 8 hours, 65 % percentage of solids, and 200 µm of granulometry.

Based on the idea of adopting closed fuel cycle in current pressurized water reactors (PWR) in order to reduce the use of natural uranium and recycle the spent fuel accumulated in the world inventory, this paper aims to compare two closed nuclear fuel cycles simulated at Model for Energy Supply Strategy Alternatives and their General Environmental Impacts (MESSAGE). The nuclear fuel cycles compared are: i) a closed fuel cycle with recovering of plutonium (Pu) to fabricate the mixed oxide (MOX) fuel; ii) a closed fuel cycle with recovering of a transuranic matrix to fabricate the transuranic fuel spiked with depleted uranium (TRU-U)O2. The comparison is based on the Brazilian nuclear energy system. They consider the time frame of 2019-2060 and the introduction of Angra 3 in the system. Advantages and disadvantages of using the strategy of operating with the different nuclear fuel cycles are shown, which include results regarding natural uranium consumption, spent fuel accumulation or utilization, nuclear waste and the nuclear fuel costs for both fuels.

Based on the BEAVRS Benchmark (Benchmark for Evaluation And Validation of Reactor Simulations) which contains reference specifications for operational data and measurements for a pressurized water reactor (PWR), it presents a model for obtaining macroscopic cross sections for the different fuels assemblies of BEAVRS using the WIMS-ANL. The microscopic cross section data library used was ENDF/B-VI. For the cross-section calculation method, the multicellular model approach was considered, and the results were compared with other reference works. This methodology aims to validate the model of calculation of the cross sections, and thus, evaluate whether such methodology generates valid results. Also, in this work, the macroscopic cross-sections generated by WIMS were used to simulate the reactor core of a typical PWR in PARCS. All calculations were made for the reactor's Hot Zero Power status and the first cycle of operation.

The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced power plant being a coupling between a modular helium cooled reactor and a gas turbine. The gas-cooled reactor types are of great interest due to their potential to provide high-temperature process heat in addition to their high thermal-to-electric power conversion efficiency and inherent safety features. The MHTGR is helium-cooled, graphite-moderated and uses Triso-coated fuel particles immersed in a cylindrical-shaped graphite matrix. The annular core and fuel blocks are based on the FSV (Fort Saint Vrain) reactor design. In this work, a MHTGR model developed in the RELAP5-3D code is presented, as well as its verification for steady state calculations. The core simulations have been performed to three power values: 350, 450 and 600 MWth. The heat transfer along the blocks was investigated taking into account the diversion flow. The radial power distribution within the compact fuel was assumed to be uniform for the analysis. Radial and axial normalized power factors were specified for each axial segment in the heat structures that are coupled to the thermal hydraulic channels that shape the reactor core. In the analyses, coolant and fuel temperatures, pressure drop and mass flow rate were verified at steady state conditions. The results were very close to the reference ones demonstrating that the developed model is capable to reproduce the MHTGR core in steady state operation.

The IPR-R1 Triga research reactor, located at CDTN (Nuclear Technology Development Center) in Brazil, is one of the oldest reactors in operation worldwide. The Brazilian Nuclear Energy Commission (Cnen) has granted a Permit for Permanent Operation license for the CDTN Triga reactor. This authorization includes some conditions to be settled within a period of up to two years, that is, until the end of 2019, which has not yet been carried out and is currently being prepared by CDTN. For this purpose, Project 0006.23 – Maintenance of Licensing of the IPR-R1 Triga Reactor was created in CDTN's Multiannual Program. In this project is included the Subproject: 6.23.3 – Preparation of the system to manage the aging of the IPR-R1 Triga reactor. Inserted in this context, the main objective of this paper is the proposal of good practices to be adopted in the future by CDTN, related to the Projects of Licensing and Aging of Triga reactor, for the management and physicochemical control of its cooling water, considering its electrical conductivity and pH, following IAEA recommendations. So, the present work is relevant, as it addresses a proposal for monitoring and controlling pH and electrical conductivity of the cooling water of Triga reactor. As a result of this proposal, the useful life of Triga reactor may be extended and there will be greater reliability and safety in its operation.

Nuclear Thermal Propulsion (NTP) is a successfully developed technology, though it hasn’t been officially used in the propulsion system of an astronautical mission. It is the technology most likely to make long distance journeys in space possible. It is considerably more efficient than the traditional chemical rocket engines, regarding propellant consumption for each unit of thrust generated. The reason for that is the greater number of choices regarding the propellant composition. Therefore, in this work, Hydrogen has been chosen as propellant due to its low molecular mass compared to other possible substances. The main aim is to explain how the propellant molecular mass impacts a rocket performance, and to show, by Computational Fluid Dynamics (CFD) simulations that, indeed, hydrogen seems to be the best choice available among possible nuclear thermal rocket propellants.

The concept of Molten Salt Reactor use Th to breed fissile 233U, where an initial source of fissile material needs to be provided. However, there is no available 233U and so; the fissile fuel supply is one of the unresolved problems. Thus, it is necessary to use existing fissile materials such as 235U or Pu to produce 233U. Current studies analyze the fuel transition from 235U/Th or Pu/Th to 233U/Th and, in this context, the present work evaluates the criticality and the neutron flux of MSBR (Molten Salt Breeder Reactor) considering the fuel: (i) mix of Th and enriched U; (ii) the combination of Th and reprocessed Pu; and (iii) matrix of reprocessed Pu/minor actinides (MAs) and Th. The goal is to verify which of these fuels can be used as initial fissile supply. The MSBR core was simulated by MCNPX 2.6.0 code and the criticality model presents similar behavior of previous studies. The results show that reprocessed fuels could have a potential to be used as initial fissile supply, but these fuels present a neutron flux profile less flattens than traditional 233U/Th. It is possible that a new distribution of fuel elements may improve this profile and future simulations will be performed to evaluate this behavior. The uranium, must has high enrichment value to be used as initial seed. Other studies need be performed to evaluates the uranium enrichment and the U/Th ratio that produces similar core criticality to traditional fuel.

NORM minerals generate, after metallurgical processing, radioactive slags that can be controlled by CNEN. Thus, it is necessary to establish a reprocessing route to make the material exemption. The route necessarily passes through the characterization of the slag from its physical properties. In this paper the results of the variation of the concentrations of U and Th oxides with the electrical, and magnetic properties of the slag, with their granulometry and with their density are presented. It has been observed that the metallurgical processing produces as slag a slag with molecular connections between the oxides that constitute it. In this way, the separation of the oxides will only be possible with the leaching of the slag.

The spent fuel from a PWR still contains some amount of fissile materials depending on their initial enrichment and the burnup. Thus, spent fuel from PWRs containing about 1.5% of fissile material could be used as fuel for CANDU reactors after some fission products are removed from it. Thus, an important proposal is the DUPIC cycle, where spent fuels from a PWR are packaged into a CANDU fuel bundle using mechanical reprocessing but without the need of chemical reprocessing. When it is refueled with reprocessed fuel, the reactivity of the system increases, and this behavior may affect the safety parameters of the reactor. Therefore, this work studies the neutronic parameters of two reprocessing fuel techniques: AIROX and OREOX, which are evaluated for two different cores configuration. The first one considers heavy water as a moderator and coolant. The second one considers heavy water and light water as moderator and coolant respectively. These studies evaluate the core behavior based on the different number of reprocessed fuels channels and compare them with the reference core. To perform the simulation the MCNPX was used to calculate the effective multiplication factor, fuel temperature coefficient of reactivity, void reactivity coefficient, and neutron flux, which were evaluated at steady state condition for the different cases. The results show that the presence of parasitic absorbers in the reprocessed fuels hardens the neutron spectrum. This behavior provokes an increase in the core reactivity, in the fuel temperature coefficient and in the void reactivity coefficient. Among these parameters, the use of light water reduces the core reactivity but do not improve the fuel temperature coefficient and the void reactivity coefficient.

The country with the highest nuclear power contribution to its energy matrix is France with 72.28%. The French nuclear history reveals that they trust in reprocessing option since an early stage of the nuclear power plants. Therefore, this work is devoted to studying the two options of a fuel cycle, i.e., Open Fuel Cycle (OFC) and Closed Fuel Cycle (CFC) for this country and the economics of each scenario. The assessment begins using the MESSAGE, since the first official registration by the International Atomic Energy Agency (IAEA), which is in 1970 and tries to follow the model until 2016. After that, the MESSAGE adjusts the energy planning to fit best until 2034 remaining the nuclear activity in this country. The results show the best reactor option to supply 50000 MWyr until 2034 using the two fuel cycle models. Finally, it is shown the amount of resources needed to maintain the nuclear power industry in France and justify the French option of the fuel cycle based on the economic results. The results show the use of nuclear resources taking advantage of a closed fuel cycle. The idea is to show the sustainability of a closed fuel cycle, applying this methodology to countries in nuclear development with a different scenario.

An artificial neural network methodology is being developed in order to find an optimum spatial distribution of the fuel assemblies in a nuclear reactor core during reloading. The main bounding parameter of the modeling was the neutron multiplication factor, keff. The characteristics of the network are defined by the nuclear parameters: cycle, burnup, enrichment, fuel type, and average power peak of each element. As for the artificial neural network, the ANN Feedforward Multi_Layer_Perceptron with various layers and neurons were constructed. Three algorithms were used and tested: LM (Levenberg-Marquardt), SCG (Scaled Conjugate Gradient) and BayR (Bayesian Regularization). The artificial neural network has implemented using MATLAB 2015a version. As preliminary results, the spatial distribution of the fuel assemblies in the core using a neural network was slightly better than the standard core.

The goal was to evaluate the neutronic behavior during the fuel burnup changing the amount of burnable poison and fuel enrichment. For these analyses, it was used a 17 x 17 PWR fuel element, simulated using the 238 groups library cross-section collapsed from ENDF/BVII.0 and TRITON module of SCALE 6.0 code system. The results confirmed the effective action of the burnable poison in the criticality control, especially at Beginning Of Cycle (BOC) and in the burnup kinetics, because at the end of the fuel cycle there was a minimal residual amount of neutron absorbers (155Gd and 157Gd), as expected. At the end of the cycle, the fuel element was still critical in all simulated situations, indicating the possibility of extending the fuel burn.

Recently, the interest in the study of various types of transients involving changes in the boron concentration inside the reactor, has led to an increase in the interest of developing and studying new models and tools that allow a correct study of boron transport. Therefore, a significant variety of different boron transport models and spatial difference schemes are available in the thermal-hydraulic codes. According to this interest, in this work it will be compared the results obtained using the different boron transport models implemented in the NRC thermal-hydraulic code TRACE. To do this, a set of models has been created using the different options and configurations that could have influence in boron transport. These models allow us to reproduce a simple event of filling or emptying the boron concentration in a long pipe. Moreover, with the aim to compare the differences obtained when one-dimensional or three-dimensional components are chosen, it has modeled many different cases using only pipe components or a mix of pipe and vessel components. In addition, the influence of the void fraction in the boron transport has been studied and compared under close conditions to a BWR commercial model. A final collection of the different cases and boron transport models are compared between them and those corresponding to the analytical solution provided by the Burgers equation. From this comparison, important conclusions are drawn that will be the basis of modeling the boron transport in TRACE adequately.

In radiotherapy treatment planning, tools that improve the accuracy and quality of the radiation treatment for cancer are important for the decreased death of healthy tissue and the probability to produce cancer cells. The purpose is to establish the simulation principles in MCNP reproducing the x-ray generation from the electron beam to the resultant photon beam used for radiotherapy in a LINAC VARIAN 600 C/D, being the main goal to approximate the x-ray spectrum, the profile dose and the relative dose to experimental ones. This analysis shows the difficulties to simulate the linear accelerator and how they are represented in the profile dose and relative dose and the probable margin of error obtained compared to the real case. These difficulties are due to the lack of information from the suppliers to the academic public. Therefore, some parts of the model should be inferred, such as the electron beam source and its angle. The results show the experimentally measured data comparison to the Monte Carlo results, where the measurements of PDD are inside the margin of error for buildup region and the flat region for the beam profile dose according to reference criteria. The developed model would help to optimize the simulation of patient dosimetry in radiotherapy treatment planning.

O processo de escolha de um local para a instalação de um repositório geológico final do combustível queimado é essencial para garantir o isolamento adequado de longo prazo dos radionuclídeos da biosfera. Entretanto, o grande número de fatores envolvidos – sejam eles: segurança de longo prazo, viabilidade sócio-econômica-ambiental e viabilidade técnica; – associados ao enorme espaço geográfico disponível para a implantação do repositório, torna o processo decisório complexo. O estudo aplicou, ao caso particular do Estado de Minas Gerais, uma metodologia de Análise de Decisão Multicritério Espacial, com o objetivo de identificar, classificar e excluir regiões de interesse para instalação de um repositório geológico. Os resultados indicam boas possibilidades de áreas adequadas para um repositório e os mesmos foram comparados com um estudo semelhante para o estado do Rio de Janeiro. Concluiu-se que é necessário novos trabalhos de levantamentos geológicos e a inclusão de novos dados para uma avaliação mais confiável.

A spent fuel pool of a typical Pressurized Water Reactor (PWR) was evaluated for criticality studies when it uses reprocessed fuels. PWR nuclear fuel assemblies with four types of fuels were considered: standard PWR fuel, MOX fuel, thorium-uranium fuel and reprocessed transuranic fuel spiked with thorium. The MOX and UO2 benchmark model was evaluated using SCALE 6.0 code with KENO-V transport code and then, adopted as a reference for other fuels compositions. The four fuel assemblies were submitted to irradiation at normal operation conditions. The burnup calculations were obtained using the TRITON sequence in the SCALE 6.0 code package. The fuel assemblies modeled use a benchmark 17x17 PWR fuel assembly dimensions. After irradiation, the fuels were inserted in the pool. The criticality safety limits were performed using the KENO-V transport code in the CSAS5 sequence. It was shown that mixing a quarter of reprocessed fuel withUO2 fuel in the pool, it would not need to be resized

O gerenciamento de recarga de combustível em um reator de potência tem como principal foco o padrão de recarga no núcleo de modo a alcançar melhor rendimento no ciclo observando todos os parâmetros de segurança adotados. Neste artigo apresentamos a estratégia de um algoritmo baseado em redes neurais artificiais que poderia ser usado para otimizar a recarga de combustível num reator nuclear. A ideia é apresentar uma metodologia baseada em rede neural feedforward multicamadas baseada em neurônios multi-valorados, que poderá ser usada para desenvolver uma metodologia capaz de escolher as melhores combinações que satisfaçam o fator de pico de potência radial e maximizem o fator de multiplicação efetivo no início do ciclo, e também satisfaçam a relação de potência crítica mínima e taxa máxima de geração de calor no final do ciclo.

The liquid-salt-cooled very high-temperature reactor (LS-VHTR) is a reactor that presents very good characteristics in terms of energy production and safety aspects. It uses as fuel the TRISO particles immersed in a graphite matrix with a cylindrical shape called fuel compact, as moderator graphite and as coolant liquid salt Li2BeF4 called Flibe. This work evaluates the thermal hydraulic performance of the heat removal system and the reactor core by performing different simplifications to represent the reactor core and the fuel compact under steady-state conditions, starting the modeling from a single fuel element, until complete the studies with the entire core model developed in the RELAP5-3D code. Two models were considered for representation of the fuel compact, homogeneous and non-homogeneous models, as well as different geometries of the heat structures was considered. The aim to develop several models was to compare the thermal hydraulic characteristics resulting from the construction of a more economical and less discretized model with much more refined models that can lead to more complexes analyzes to representing TRISO effect particles in the fuel compact. The different results found, mainly, for the core temperature distributions are presented and discussed.

Nuclear Power Plants (NPPs) based on Pressurized Water Reactors (PWRs) technology are considered an alternative to fossil fuels plants due to their reliability with low operational cost and low CO2 emissions. An example of PWR plant is Angra 2 built in Brazil. This NPP has a nominal electric power output of 1300 MW and made it possible for the country save its water resources during electricity generation from hydraulic plants, and improved Brazilian knowledge and technology in nuclear research area. Despite all these benefits, PWR plants generally have a relatively low thermal efficiency combined with a large amount of irreversibility generation or exergy destruction in their components, reducing their capacity to produce work. Because of that, it is important to assess such systems to understand how each component impacts on system efficiency. Based on that, the aim of this work is to evaluate Angra 2 by performing energy and exergy analyses to quantify the thermodynamic performance of this PWR plant and its components. The methodology consists in the development of a mathematical model in EES (Engineering Equation Solver) software based on thermodynamic states in addition to energy and exergy balance equations. According to the results, Angra 2 has energy efficiency of 36.18% and exergy efficiency of 49.24%. Reactor core is the most inefficient device in the NPP; it has exergy efficiency of 67.16% and is responsible for 63.88% of all exergy destroyed in Angra 2.

Radionuclides of natural origin are present in rocks and soils but have no homogeneous distribution in the soil of the Earth. Volcanic rocks have a higher concentration of uranium and thorium. Thus, sites with this type of origin present as radiological abnormalities. An example is the Planalto de Poços de Caldas in Minas Gerais. Another example is the Granito Madeira - a rock formation located in the Amazonas State, 350 km north of Manaus. This work presents the results of the natural radiation monitoring of Granito Madeira. 30,000 dose rates were collected with a sodium iodide scintillator transported on tracks within the Floresta Amazônica. These data were treated and georeferenced. As a result, an annual dose relative to gamma radiation of 6.0 mSv.year-1 was obtained. This information was represented in a frequency histogram and in a map of Granito Madeira.

The aim is to analyse the neutron spectrum influence in a hybrid system ADS-fission inducing transuranics (TRUs) transmutation. A simple model consisting of an Accelerator-Driven Subcritical (ADS) system containing spallation target, moderator or coolant, and spheres of actinides, “fuel,” at different locations in the system was modelled. The simulation was performed using the MCNPX 2.6.0 particles transport code evaluating capture (n,γ) and fission (n,f) reactions, as well as the burnup of actinides. The goal is to examine the behaviour and influences of the hard neutron spectrum from spallation reactions in the transmutation, without the contribution or interference of multiplier subcritical medium, and compare the results with those obtained from the neutron fission spectrum. The results show that the transmutation efficiency is independent of the spallation target material used, and the neutrons spectrum from spallation does not contribute to increased rates of actinides transmutation even in the vicinity of the target.

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.

Accelerator-driven systems (ADSs) are investigated for long-lived fission product transmutation and fuel regeneration. The aim of this paper is to investigate the nuclear fuel evolution and the neutronic parameters of a lead-cooled accelerator-driven system used for fuel breeding. The fuel used in some fuel rods was T232hO2 for U233 production. In the other fuel rods was used a mixture based upon Pu-MA, removed from PWR-spent fuel, reprocessed by GANEX, and finally spiked with thorium or depleted uranium. The use of reprocessed fuel ensured the use of T232hO2 without the initial requirement of U233 enrichment. In this paper was used the Monte Carlo code MCNPX 2.6.0 that presents the depletion/burnup capability, combining an ADS source and kcode-mode (for criticality calculations). The multiplication factor (keff) evolution, the neutron energy spectra in the core at BOL, and the nuclear fuel evolution during the burnup were evaluated. The results indicated that the combined use of T232hO2 and reprocessed fuel allowed U233 production without the initial requirement of U233 enrichment.