Abstract: Brazil has one of the largest reserves of thorium in the world, including rare earth minerals. It has developed a great program in the field of nuclear technology for decades, including facilities to produced oxides to microspheres and thorium nitrates. Nowadays, with the current climate change, it is necessary to reduce greenhouse gas emissions, one of this way is exploring the advent of IV Generation reactors, molten salt reactors, that using Thorium and Lithium. Thorium's technology is promising and has been awaiting the return of one nuclear policy that incorporates its relevance to the necessary levels, since countries like the BRICS (without Brazil) have been doing so for years. Brazil has also been developing studies on the purification of lithium, and this one associated to thorium, are the raw material of the molten salt reactors. This paper presents a summary of the thorium and lithium technology that the country already has, and its perspectives to the future.

Palavras-Chave: lithium; molten salt reactors; nuclear fuels; public policy; purification; thorium; uranium; brazil

OLIVEIRA, GLAUCIA A.C. de; LAINETTI, PAULO E.O.; BUSTILLOS, JOSE O.W.V.; PIRANI, DEBORA A.; BERGAMASCHI, VANDERLEI S.; FERREIRA, JOAO C.; SENEDA, JOSE A. Thorium and lithium in Brazil. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5915-5922. Disponível em: http://repositorio.ipen.br/handle/123456789/30738. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: In the early days of nuclear energy in Brazil, a reactor designed at the Argonne National Laboratory, originating the name ARGONAUT from the combination of the name of the Laboratory with the initials of Nuclear Assembly for University Training, reached criticality at the Institute of Nuclear Engineering. The Argonaut is a water moderated research reactor, which uses uranium enriched to 20% (235U) with prismatic graphite reflectors, designed to provide a thermal neutron flux up to 1010 n.cm-2.s-1 at an operating power of 5 kW. The presence of a nuclear research facility at the campus of Federal University of Rio de Janeiro (UFRJ) still cause concerns about radiological safety of the community around, even though this facility has been securely operating for more than fifty years. Besides, there were questioning about the potential risk of this facility to the IEN´s workforce by the Central of Harmonization Unit of Brazil (CGU). Thus, the present work aims to assess the potential risk of radiological accidents. Previously, the potential accidents evolving Argonaut reactor were considered to be the insertion of excess reactivity, catastrophic rearrangement of the core, graphite fire and fuel-handling accident. However, a recent accident scenario reassessment concluded that a severe physical damage of the core after reactor shutdown should be the emergency situation with the greater potential risk among the feasible postulated accidents. According with the shutdown procedure, the water, used as moderator and coolant, drains out of the core and the concrete covers (each weighing 2.5 tons) are routinely removed from the top of reactor using a crane. The damage caused by the failure of the crane dropping the covers on the core would lead to breaking of the aluminum coating and the nuclear fuel plates with their release to the reactor room. This study assesses the radiological impact to workers and members of the public caused by partial inventory release to the atmosphere. Generic gaussian model was used to estimate the relative concentrations of air at ground level through the calculation of dispersion factors derived from wind data. For the dose calculation, the conversion coefficients by inhalation and plume immersion established by the ICRP were used. The results show that potential risk is above 1/10 of the limit of annual dose for workers, while they stay below the limit for members of the public, within a radius greater than 1 km.

Palavras-Chave: argonaut reactor; dose rates; fission product release; fuel elements; gaussian processes; ionizing radiations; personnel; radiation accidents; radiation doses; reactor accidents; risk assessment; volatile matter

SOUZA, PAULA C.A. de; AGUIAR, ANDRE S.; HEIMLICH, ADINO; LAPA, CELSO M.F.; LAMEGO, FERNANDO. Assessment of potential risk and radiological impact of accidental release from the ARGONAUT reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5877-5885. Disponível em: http://repositorio.ipen.br/handle/123456789/30736. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: A fundamental concept in nuclear reactor operation is that safety is the result of interactions between human, technological and organizational factors. The National Nuclear Energy Commission understands how human factors from psychological, physiological, behavioral and emotional origin can affect the reactor operation. For that reason, reactor operators are submitted to rigorous evaluations every ye ar. When conducting case study du ring these sixty years of IEA R1, three of them hypothetical and possible related to the reactor operation illustrates the co ncern about the safety and security : Case 1 Operator had a stroke during reactor operation in the control room. C ase 2 Operator suffered stress in traffic in his going to the reactor facility; when performing test in the emergency cooling system for reactor start up, he didn’t close a valve completely; changing the pool water technical quality causing a week delay in the reactor op eration . Case 3 Operator just arrived to afternoon shift in the control room, after a few minutes his co worker noticed that his cognition and behavior has changed, later in the hospital he was diagnosed with head cancer. This interdisciplinary work aims to include human factors of psychological , physiological and behavioral origin in 'reactor trip'. The ‘reactor trip’ (also know n as ‘scram’) usually applies to technical factors to avoid high consequence event, are protection circuits that can assume the s tatus of alert, hazard and essentially shut down the reactor automatically; when temperature, radioactivity, pressure, water flow, voltage and so on ; are out of the operating limits. Technologies associated with neuroscience and psychological assessments s uch as: Face Reader, Analogue Visual Mood Scale and Back Depression Inventory ; allows the evaluation of the operator in the control room. However, problems li ke described in the case study should be minimized. This inter disciplinary theoretical work is based on empirical doctoral thesis in progress.

Palavras-Chave: control rooms; human factors; iear-1 reactor; radiation protection; reactor accidents; reactor operators; reactor safety; scram; security

VAZ, ANTONIO C.A.; RODRIGUES, VALDEMIR G.; TOYODA, EDUARDO Y.; SAXENA, RAJENDRA N. Human factors inclusion proposal in “reactor trip” to increase safety in operation. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5819-5826. Disponível em: http://repositorio.ipen.br/handle/123456789/30734. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The thermal-hydraulic safety analysis of research reactors establishes the safety criteria to ensure the integrity of the fuel elements in the reactor core. It assures that all core components are being adequately cooled during operation. It is necessary to know if the average mass flow rate (and their standard deviation) among the fuel assemblies are enough to cool the power generated during operation. Once satisfied such condition, it allows the calculation of the maximum heat flux transferred from fuel assemblies to the coolant, and if the maximum cladding temperatures are below the limits set by the safety criteria. Among the objectives, this study presents a methodology for a preliminary three-dimensional numerical analysis of the flow distribution in the core of the IEA-R1 research reactor, under steady state condition. For this, the ANSYS-CFX® commercial code was used to analyze the flow dynamics in the core, and to visualize the velocity field. It was possible to conclude that a homogeneous flow distribution for all standard fuel assemblies were found, with 2.7% deviation from the average mass flow. What turned out to be negligible and can be assumed that there is a homogeneous distribution in the core. Complex structures were find in the computational domain. Once known the core flow dynamics, it allows future studies to determine whether the heat flux and temperature conditions abbeys thermal-hydraulic safety criteria.

Palavras-Chave: boundary conditions; c codes; flow models; fuel assemblies; iear-1 reactor; numerical analysis; reactor cores; research reactors; safety; steady-state conditions; thermal hydraulics

SCURO, NIKOLAS L.; ANGELO, GABRIEL; ANGELO, E.; TORRES, WALMIR M.; UMBEHAUN, PEDRO E.; ANDRADE, DELVONEI A. de. Preliminary numerical analysis of the flow distribution in the core of a research reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5667-5674. Disponível em: http://repositorio.ipen.br/handle/123456789/30732. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Epistemic uncertainty (uncertainty related to lack of knowledge), often found in the documentation of nuclear facility engineering projects, can affect the decision-making process of managers and analysts on safety and risk issues. This article conceptualizes the nature of the major uncertainties involved in engineering projects and describes a knowledge base developed in order to gather data and information related to the project of an Open-Pool Light-water Research Rector (OPLRR) and whose purpose is to assist professionals who work in the áreas of safety, design, operation, and maintenance of nuclear facilities. In order to reduce the epistemic uncertainties that may rise in the project, the OPLRR knowledge base is designed to contain a set of information that allows identifying and facilitating the forwarding of solutions to address inconsistencies, and/or pending issues that may exist in the project. In this sense, the information and the documents related to the project are organized in a graphical and hierarchical architecture, allowing the knowledge base users to quickly and easily obtain information regarding the systems, processes, equipment, and components of the Project. Besides that, a set of documents containing descriptions, reliability data and some other important information about the systems and components are specially created to the knowledge base and it is crucial to reduce epistemic uncertainties, once it raises the issues and the inconsistencies of the project, as well as it clarifies the interrelations between the systems, the functioning of the equipment, their failures modes and the consequences of their failures, and some other data, which are not originally contained in the documents of the project.

Palavras-Chave: data covariances; decision making; design; information dissemination; knowledge base; maintenance; nuclear facilities; personnel; pool type reactors; reactor cores; risk assessment; safety

VIEIRA NETO, ANTONIO S.; OLIVA, AMAURY M.; SAUER, MARIA E.L.J.; HUNOLD, MARCOS C.; OLIVEIRA, PATRICIA da S.P.de; ANDREA, VINICIUS. Knowledge base about risk and safety of nuclear facilities to support analysts and decision makers. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5513-5522. Disponível em: http://repositorio.ipen.br/handle/123456789/30730. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The aim of this paper is evaluated the consequences to ANGRA 2 nuclear power reactor and to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 200cm2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. The results obtained for ANGRA 2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core.

Palavras-Chave: angra-2 reactor; cladding; eccs; heat transfer; primary coolant circuits; reactor accident simulation; reactor cores; sbloca; steady-state conditions; two-phase flow; void fraction

BORGES, EDUARDO M.; SABUNDJIAN, GAIANE. Small break loss of coolant accident of 200 cm² in cold leg of primary loop of ANGRA 2 nuclear power reactor evaluation. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5479-5490. Disponível em: http://repositorio.ipen.br/handle/123456789/30728. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: This work presents the thermo-hydraulic simulation of the Brazilian Multipurpose Reactor (RMB) using the RELAP5/Mod3 code. The RMB will provide Brazil with a fundamental infrastructure for the national development on activities of the nuclear sector in the areas of social, strategic, industrial applications and scientific and technological development. A RELAP5/Mod3 code model was developed for thermo-hydraulic simulation of the RMB to analyze the phenomenology of the Siphon Breakers device (four flap valves in the cold leg and one open tube for the atmosphere in the hot leg) during a Loss of Coolant Accident (LOCA) at different points in the primary circuit. The Siphon Breaker device is an important passive safety system for research reactors in order to guarantee the water level in the core under accidental conditions. Different simulations were carried out at different location in the Core Cooling System (CCS) of the RMB, for example: LOCA before the CCS pumps with and without pump trip and LOCA after the CCS pumps and the heat exchanger. In all RELAP5/Mod3 code simulations, the Siphon Breaker device's performance after a LOCA was effective to allow enough air to enter the outlet pipe of the CCS in order to break the siphon effect and preventing the pool level from reaching the riser (chimney) and the RMB core discovering. In all cases, the reactor pool level stabilized at about 5.5 m after the end of the LOCA simulation and the fuel elements were kept underwater and cooled.

Palavras-Chave: cooling systems; fuel elements; loss of coolant; r codes; reactor accident simulation; reactor cores; reactor safety; rmb reactor; thermal hydraulics; transients

SOARES, HUMBERTO V.; BELCHIOR JUNIOR, ANTONIO; FREITAS, ROBERTO L. RELAP5 modeling of a siphon break effect on the Brazilian Multipurpose Reactor. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5443-5456. Disponível em: http://repositorio.ipen.br/handle/123456789/30726. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: According to international plans, the nuclear reactor fleet should reduce operational risk and avoid severe accidents. Around the world, there are 450 nuclear power reactors in operation, which supply about 11% of the electricity consumed. There are programs, such as Advanced Fuels Campaign (AFC), that plan to develop a more tolerant fuel system by 2025. These plans follow security concepts that present two options capable of replacing zirconium alloys used as cladding. The better candidates are metallic alloys and ceramic materials. Until the mid-1970s, austenitic steel was the main coating option. Recently, iron-based alloys have become short-term solutions composed of iron-chromium-aluminum (FeCrAl) alloys. However, there are various advantages from using multilayer of silicon carbide (SIC) and ceramic composites. Silicon carbide has higher corrosion resistance, coupled with higher mechanical strength compared to zirconium alloys. Upon steam contact, ceramic cladding mitigates hydrogen buildup, avoiding explosion risk. This study presents a comparison of the thermal and mechanical properties between zirconium alloys and ceramic alternatives. Ceramic materials show desirable mechanical strength, such as high initial crack resistance, stiffness, ultimate strength, impact response, and high corrosion resistance. SIC has a lower neutron cross-section with significant safety margins.

Palavras-Chave: ceramics; cladding; comparative evaluations; corrosion protection; cross sections; f codes; fuel rods; mechanical properties; nuclear fuels; physical properties; silicon carbides; steady-state conditions; thermal expansion; zirconium alloys

GOMES, DANIEL de S.; GIOVEDI, CLAUDIA. Comparative analysis of silicon carbide with zirconium-based alloys. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5417-5428. Disponível em: http://repositorio.ipen.br/handle/123456789/30724. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: In an open pool type reactor, the pool water inventory should act as a heat sink to provide emergency reactor core cooling. In the Brazilian Multipurpose Reactor – RMB, to avoid the loss of pool water inventory, all the Core Cooling System (CCS) lines penetrate at the pool top, far above the reactor core level. However, as most of CCS equipment and lines are located below the reactor core level, in the case of a Loss of Coolant Accident (LOCA), a large amount of pool water could be lost drained by siphon effect. To avoid RMB research reactor core discovering in the case of a LOCA, siphon breakers, that allow CCS line air intake, are installed in the CCS lines in order to stop the reactor pool draining due to siphon effect. As siphon breakers are important passive safety devices, their effectiveness should be verified. Several previous numerical and experimental studies about siphon break effect were found in the literature. Some of them comment about the effectiveness of the siphon breakers based on their air intake area. Others state that one-dimensional thermo-hydraulic system codes such as RELAP5 code would fail when modeling the siphon break effect. This work shows the RELAP5/MOD3.3 code capability in modeling the siphon break effect. A nodalization for RELAP5/MOD3.3 code of a Siphon Breaker Test Facility located at POSTECH University in Korea was developed. Experiments considering several siphon breakers device intake areas were simulated. A very good agreement between numerical and experimental results was obtained. As siphon breakers intake areas decrease, the siphon breaker effectiveness also decreases and more water is drained from the reactor pool. For smaller siphon breaker intake areas, RELAP5/MOD3.3 code showed conservative results, overestimating the reactor pool water losses.

Palavras-Chave: computerized simulation; loss of coolant; pipes; pool type reactors; r codes; reactor cores; ruptures; safety analysis; tanks; test facilities; test facilities; validation; void fraction

BELCHIOR JUNIOR, ANTONIO; SOARES, HUMBERTO V.; FREITAS, ROBERTO L. Validation of the RELAP5 code for the simulation of the Siphon Break effect in pool type research reactors. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5383-5392. Disponível em: http://repositorio.ipen.br/handle/123456789/30722. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: This work presents the simulation of a severe accident scenario in a referential model of pressurized water reactor, which came about from a rupture of 20cm2 in a cold leg of a reactor cooling system. The simulation was carried out on the MELCOR code using a model elaborated by the Global Research for Safety – Germany, with the passive autocatalytic recombiners implemented in almost every compartment in the containment. The efficacy and effectiveness of this well-known mitigating measure of severe accident management are demonstrated by means of a comparison with the case previously simulated without this measure using the same model. This referential reactor is important and very useful for the independent analysis of severe accidents in the Brazilian Angra 2 nuclear power plant in virtue of the similarity between both of them, so that after some proper modifications on this referential reactor’s model, it could be applied for the study of severe accidents in the other. In this sense, the result presented in this work is to be taken as an important reference for the severe accident analysis of Angra 2.

Palavras-Chave: boundary conditions; cladding; loss of coolant; m codes; melt-through; pwr type reactors; radiation protection; reactor accident simulation; reactor cooling systems

LEE, SEUNG M.; LAPA, NELBIA S.; SABUNDJIAN, GAIANE. MELCOR simulation of a severe accident scenario derived from a small break loca in a typical PWR with passive autocatalytic recombiners. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5346-5359. Disponível em: http://repositorio.ipen.br/handle/123456789/30720. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Neutron irradiation in nuclear power plants (NPPs) lead to microstructural changes in structural materials which induce a shift of the ductile to brittle transition temperature (DBTT) towards higher temperatures. Monitoring of the DBTT in NPP components receives therefore considerable attention. Small specimen testing techniques are developed for characterizing structural components with a limited amount of materials. One of the most used of these miniature testing is the small punch test (SPT) which is based on disc or square shaped specimens. SPTs may be performed from room to cryogenic temperatures, plotting the absorbed energy until rupture, against the test temperature. A ductile region (high energy) and a brittle region (low energy) with a transition between both zones are usually reported. The transition temperature thus obtained, DBTTSPT, is also related through empirical expressions to the transition temperature obtained in CVN tests, DBTTCVN, or in fracture toughness testing. Linear expressions such as DBTTSPT = α DBTTCVN have been used where α is a material characteristic constant. In all cases, the DBTTSPT temperature is much lower than that obtained in the CVN tests. In this paper, we present a short review of the literature on the determination of the DBTT for nuclear reactors pressure vessels steels by those two techniques analyzing the reason for the difference in their value as mentioned before. In dealing with irradiated materials, is a high priority to limit the exposure of the professional to irradiation. Therefore, the use of miniature specimens receives significant attention in the nuclear community. The high cost of irradiation experiments is a further incentive for using small specimen testing techniques.

Palavras-Chave: ductile-brittle transitions; embrittlement; fracture properties; irradiation; materials testing; miniaturization; monitoring; reactor vessels; steels

ANDRADE, ARNALDO H.P. de; MIRANDA, CARLOS A.J.; LOBO, RAQUEL de M. Monitoring of the ductile to brittle transition temperature of reactor pressure vessel steels by means of small specimens. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5322-5338. Disponível em: http://repositorio.ipen.br/handle/123456789/30718. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The Water Experimental Circuit (CEA) was built in IPEN in eighties and had the aim to perform thermal hydraulic experiments, simulating operational condition of Pressurized Water Reactors and Boiling Water Reactors. The CEA operated until 1984 and since then it was decommissioned. In order to do hydrodynamics tests in MTR fuel type elements of nuclear research reactor, in the years 2015, was conceived an experimental circuit named Orquidea, which shall operate with low pressure and temperature. This paper assess the mechanical and structural suitability of the Pressurizer V-102, that was used in the former Water Experimental Circuit (CEA) aiming reuse this vessel in new the circuit. The methodology applied to evaluate the vessel was based on ASME code, Section VIII, Division 1 & 2.

Palavras-Chave: a codes; flanges; fuel elements; hydrodynamics; mechanical properties; nozzles; numerical solution; pressurizers; pwr type reactors; reactor vessels; stress analysis; thermal hydraulics

FAINER, GERSON; FALOPPA, ALTAIR A.; ALMEIDA, JOEDSON T. de; FIGUEIREDO, CAROLINA D.R.; CARVALHO, DANIEL S.M.; MATTAR NETO, MIGUEL. Structural assessment of pressurizer V-102 of the circuit Orquídea. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5290-5305. Disponível em: http://repositorio.ipen.br/handle/123456789/30716. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: PWR reactor type, when subjected to the power ramp regime, a mechanical interaction between the cladding and the UO2 pellet (PCMI) may occur in the fuel rod. To investigate this phenomenon were used two softwares, the first was a modified fuel performance code to verify the behavior of fuel rod with steel cladding and another to analyze structural mechanical behavior. The fuel performance code results show that there is no contact between the pellet and the cladding in the fuel rod, considering the estimated burning under normal conditions of reactor operation. Thus, it was adopted the hypothesis of the interaction pellet-cladding occurrence, generated by pellet fragmentation and relocation, and power ramp simulation conditions independent of the ramp time. The simulations results show that the fuel rod maintains its integrity under the conditions of the adopted hypothesis.

Palavras-Chave: c codes; design; finite element method; fuel rods; fuel-cladding interactions; mechanical properties; numerical solution; pwr type reactors; stainless steels; steady-state conditions; stress intensity factors; thermal hydraulics

BERRETTA, JOSE R.; LIMA, LEONARDO S.; REIS, REGIS; AGUIAR, AMANDA A. PCMI effect study in the fuel rod of a PWR reactor type subjected to power ramps. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5270-5275. Disponível em: http://repositorio.ipen.br/handle/123456789/30714. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: This work aims to develop a suitable neutron shielding for an Inertial Electrostatic Confinement Nuclear Fusion device (IECF). Neutrons are generated in the IECF device as results of nuclear fusion reactions and their detection is fundamental for the development of the IECF device, because experimental data is needed to perform efficiency analysis and model validation. Nevertheless, it is essential to moderate the neutrons down to the thermal state to make it possible to detect those using conventional detectors. Therefore, to properly measure the fast neutron generation rate by the IECF device it is necessary to previously elaborate a detailed neutron transport model between the IECF device and the radiation shielding, where the neutron detector will be located. In this work, a model is elaborated using the Monte Carlo N-Particle Code and is used to design the required radiation shielding for the device. Later, the same model will be used to determine the proportionality factor between the fast neutron generation in the IECF device and the thermal neutron population in the shielding.

Palavras-Chave: dose equivalents; dose rates; electrostatics; fast neutrons; icf devices; inertial confinement; monte carlo method; neutron flux; neutron transport; shielding; thermal neutrons

LEE, SEUNG M.; YORIYAZ, HELIO; CABRAL, EDUARDO L.L. Development of neutron shielding for an inertial electrostatic confinement nuclear fusion device. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5088-5095. Disponível em: http://repositorio.ipen.br/handle/123456789/30712. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The search for new materials for nuclear fuels has been developed over the last 50 years, with the main aim of increasing the fuel efficiency during the operation of the reactors. The need to increase the uranium density in fuels to compensate the reduction of enrichment proposes that the UMo alloy is one of the materials that presents better characteristics to be used as fuel: molybdenum is a material that retains the gamma phase of the uranium in low concentrations, which is the only stable phase of uranium under the irradiation conditions, besides having low thermal neutron absorption. Although more advanced studies already provide information on the interaction between UMo and the Al matrix, we still need to study how this material behaves during all processing steps for fuel fabrication. The present work has the objective of to deepen the technological knowledge about the stages of production of dispersion type nuclear fuel, including the comminution process of the UMo alloy. The alloy pulverization made by the hydriding-grinding-dehydriding technique still reveals a large number of unknowns in the process variables. Knowing some parameters already existent in the literature, it is possible to discuss the behavior of the hydriding process and envision improvements to optimize it as well as make it reproducible. Subsequent manufacturing steps for briquette and rolling were performed according to IPEN's expertise and the results indicate that the UMo alloy is mechanically doable and may prove to be a substitute fuel for the current U3Si2 with a higher uranium density.

Palavras-Chave: briquets; comminution; dispersion nuclear fuels; fuel plates; fuel substitution; hydridation; molybdenum alloys; production; rolling; uranium alloys

PIRES, MARINA C.; MARQUES, JOSE R. de O.; LEAL NETO, RICARDO M.; DURAZZO, MICHELANGELO. Study of the manufacturing process of gamma-U7%wtMo dispersion fuel plates. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5024-5035. Disponível em: http://repositorio.ipen.br/handle/123456789/30710. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: This work discusses a parametric study applied to nuclear power generation based on a mixed fuel formed by the composition of thorium-uranium oxide (Th-U)O2. Also, approached in this study the physical neutrons models of a fuel system composed of ThO2 75 wt% and UO2 25 wt%, with 19.5% enrichment of U-235. The thermodynamic features of the thorium-uranium fuel system compared with the properties of uranium dioxide. Thorium-based fuel operating extended fuel cycles reach of over 80 GWd/MTU in a pressurized water reactor (PWR). Homogenous distribution of thorium-based fuel, used on the reactor core, could reduce Pu-239, once U-233 production capacity dependent on Th-232 replacing U-238 in the fuel matrix. The mixed oxide fuel has a lower buildup of Pu-239, causing the linear heat rate distribution slope to flatten and lowering fuel porosity. The release of gaseous fission products models for (Th-U)O2 could have different diffusion coefficients when compared to uranium oxide models. Besides, resulting in lower thermal gradients than UO2 and a reduction in fuel swelling. This parametric study reviews the aspects of radioactive decay chains of uranium and thorium. It founded the simulation using approved nuclear codes, such as SERPENT for neutron physics calculations and the FRAPCON code, which defines the licensing process. The results show that thoria based fuel has a higher performance than UO2 fuel in regular operation and can improve safety margins.

Palavras-Chave: comparative evaluations; enthalpy; f codes; mixed oxide fuels; performance; pwr type reactors; s codes; thermal conductivity; thorium; uranium oxides

GOMES, DANIEL de S.; STEFANI, GIOVANNI L. de; OLIVEIRA, FABIO B.V. de. Analysis of a pressurized power reactor using thorium mixed fuel under regular operation. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4996-5009. Disponível em: http://repositorio.ipen.br/handle/123456789/30708. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The uranium molybdenum (U-Mo) alloys have potential to be used as low enriched uranium nuclear fuel in research, test and power nuclear reactors. U-Mo alloy with composition between 7 and 10 wt% molybdenum shows excellent body centered cubic phase (γ phase) stabilization and presents a good nuclear fuel testing performance. Hot rolling is commonly utilized to produce parallel fuel plate where it promotes bonding the cladding and the fuel alloy. The mechanical deformation generates crystallographic preferential orientation, the texture, which influences the material properties. This work studied the texture evolution in hot rolled U-Mo alloys. The U7.4Mo and U9.5Mo alloys were melted in a vacuum induction furnace, homogenized at 1000°C for 5 h and then hot rolled at 650°C in three height reductions: 50, 65 and 80%. The as-cast and processed alloys microstructures were characterized by optical and electronic microscopies. The crystalline phases and the texture were evaluated by X-ray diffraction (XRD). The as-cast, homogenized and deformed alloys have γ phase. It was found microstructural differences between the U7.4Mo and U9.5Mo alloys. The homogenized treatment showed effective for microsegregation reduction and were not observed substantial grain size increasing. The deformed uranium molybdenum alloys presented strong γ fiber texture (111) <uvw> and moderated α-fiber texture (hkl) <110>.

Palavras-Chave: chemical composition; crystal structure; deformation; microstructure; molybdenum alloys; nuclear fuels; optical microscopy; rolling; scanning electron microscopy; texture; uranium alloys; x-ray diffraction

NIELSEN, GUILHERME F.; MORAIS, NATHANAEL W.S.; SILVA, SELMA L.; LIMA, NELSON B. de. Crystallographic texture of hot rolled uranium-molybdenum alloys. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4962-4970. Disponível em: http://repositorio.ipen.br/handle/123456789/30706. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Image processing has been an increasing research area in the last decades, especially due to crescent technological growth allied with lowering production costs. Many scientific applications have searched for establishment of quality norms associated with possible information obtainment from images. A common need from different applications has been the standardization of focus quality metric. The development of new methods for measuring the focus adjustment in order to obtain image quality metric analysis has enabled more reliable and precise data in many different industry and science sectors. Some examples are industrial equipment parts inspection using computational vision to defects classification. This work presents the initial steps to develop a methodology to estimate focus in real time in two-phase flow experiments inside tube with cylindrical geometry. This methodology is initially based on a software module using artificial intelligence methods to estimate image focus. This module is developed in LabVIEW platform using Fuzzy Logic inference base in different traditional digital focus metrics and integrated with digital cameras to increment precision on focus adjustment during two-phase flow experiments. This method will be calibrated to be used on void fraction estimation through image analysis in the natural circulation loop located at the Nuclear Engineering Center (CEN) do Instituto de Pesquisas Energéticas e Nucleares (IPEN). A set of the initial developed software modules will be presented with their respective functionalities, initial results and experimental focus estimated errors.

Palavras-Chave: artificial intelligence; defects; focusing; focusing; fuzzy logic; image processing; l codes; m codes; natural convection; quality assurance; real time systems; tubes; two-phase flow; void fraction; brazilian cnen

SILVA, MARCONES C.B. da; SCHOTT, SANDRO M.C.; MESQUITA, ROBERTO N. de. Development of a real-time focus estimaton software to be applied in two-phase flow imaging using intelligent processing. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4887-4902. Disponível em: http://repositorio.ipen.br/handle/123456789/30704. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The aim of this paper is to provide students, educators, and the general public with a virtual tool for learning about the peaceful use of nuclear technology and its importance to humanity. Using new technologies available in the market such as smartphones, software for the development of electronic games, virtual reality glasses, among others, we will virtually reproduce the facilities of the IEA-R1 nuclear research reactor, allowing anyone to perform a virtual and interactive visit to these facilities in a safe and didactic way. The use of virtual reality glasses and applications has been shown to be adequate in relation to the objectives proposed here.

Palavras-Chave: computer codes; computerized simulation; data visualization; educational tools; iear-1 reactor; mobile phones; real time systems; training

SILVA, LEANDRO G.M. e; SABUNDJIAN, GAIANE. Virtual visit to nuclear research reactor IEA-R1. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4839-4846. Disponível em: http://repositorio.ipen.br/handle/123456789/30702. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Modern and high power pool type research reactors generally operate with upward flow in the core. They have a chimney above the core, where the heated fluid is suctioned by the pumps. It passes through the decay tank and is sent to the heat exchangers for the cooling and returns to the core. The pipes inside the reactor pool have passive valves (natural circulation valves) that allow the establishment of natural circulation between the core and the pool for the decay heat removal, when the pumps are inoperative. These valves also have the siphon-breaker function in case of Loss of Coolant Accidents (LOCA), avoiding the pool emptying. In some reactors, these valves are located above the core chimney to facilitate the maintenance. When a LOCA causes a water level below these valves, they loose the natural circulation function. If the water level is the same of the chimney top, the available fluid for the core cooling is only that contained in the chimney and core, and a significant quantity of water in the pool is unavailable for core cooling. To bypass this problem during the reactor design phase, the inclusion of small holes of 10 mm of diameter on the lower plenum lateral side is proposed. These holes will allow a flow path between the pool and the core. Theoretical calculations were performed and analyzed for different drilling configurations: 4, 6 8, and 10 holes. A theoretical analysis of the estimated leakage rate during normal operation and evaporation and replacement rates during a hypothetical LOCA were performed. The calculation results showed that the four configurations analyzed are able to supply the water evaporated from chimney. An experiment is being proposed to validate the theoretical calculations and the considered hypotheses.

Palavras-Chave: core flooding systems; experimental data; flow rate; holes; leaks; loss of coolant; natural convection; pool type reactors; primary coolant circuits; reactor cores; reactor safety; research reactors; theoretical data; valves

MAPRELIAN, EDUARDO; BELCHIOR JUNIOR, ANTONIO; TORRES, WALMIR M. Lower plenum holes for research reactor core flooding: a proposal to improve the safety in design. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4631-4639. Disponível em: http://repositorio.ipen.br/handle/123456789/30700. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: This work aims to investigate the thermophysical properties of T iO 2 nanofluids in water base experimentally and also comparing results to the literature. Exis ting studies indicate that nanofluids presents increase in thermal conductivity compared to the base fluid which in this study will be water, thus, can be classified as promising fluids for heat transport applications. As the proposal is to use it in nuclea r applications, the survey of experimen tal measurements was performed before and after irradiation in the IPEN installations to verify the effect of ionizing radiation on the properties of nanofluids. Thermal conductivity , viscosity and some visualization of nanopar ticles in SEM were carried out in order to understand the behavior of radiation influence on nanofluids and it properties.

Palavras-Chave: heat transfer; ionizing radiations; nanofluids; nanoparticles; radiation effects; scanning electron microscopy; thermal conductivity; titanium oxides; viscosity; water

MOREIRA, PRISCILA G.; STEFANIAK, IZABELA; ROCHA, MARCELO S. Analysis of the thermal conductivity of the aqueous-based TiO2 nanofluid for nuclear applications. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4515-4524. Disponível em: http://repositorio.ipen.br/handle/123456789/30698. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: This paper describes the activities related to the work planned within Project BRA3.01/12 between CNEN and the European Community, relatively to its Task 2.1 (independent uncertainty quantification and sensitivity analysis utilizing the computational tool SUSA for the calculus related to LOCA simulation for licensing matter). SUSA software has been applied to the reference case, a double-ended LBLOCA in Angra 2, simulated with a RELAP5 code nodalization developed by the thermal hydraulic technicians of CNEN and its research institutes. This original nodalization has been improved for the development of the main objective of Task 2.1. The recommendations that our European counterparts provided on the last workshop, held at CNEN in Rio de Janeiro from January 28th to February 2nd, 2018, have been implemented as far as feasible.

Palavras-Chave: angra-2 reactor; boundary conditions; cladding; data covariances; lbloca; pressure vessels; r codes; reactor accident simulation; reactor cores; s codes; steady-state conditions

MADEIRA, ALZIRA A.; PEREIRA, LUIZ C.M.; SABUNDJIAN, GAIANE. An Angra 2 LBLOCA simulation model for RELAP5MOD3.3 code with uncertainty analysis. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4476-4502. Disponível em: http://repositorio.ipen.br/handle/123456789/30696. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: In the validation and verification processes of calculation methodologies and associated nuclear data libraries, the existence of experiments that can be considered benchmarks is of fundamental importance. For this purpose, a set of experiments with heavy material nuclear reflector was performed in the IPEN/MB-01 reactor using nickel plates properly inserted in the west face of the reactor core. A total of 32 plates around 3 mm thick were used in the experiment. The axial width and length were sufficient to cover the entire active reactor core. For each plate placement step, reactivity measurements were made due to their insertion in the core; as well as of the critical position of the equally removed BC1 and BC2 control rods. It could be observed that the increase of neutron absorption and consequent decrease of neutron moderation dominated the whole physics of the problem when few plates of reflective material were inserted (about 3 plates). Thereafter, neutron reflection became important overcoming neutron absorption; the reactivity increased until it surpassed the situation without plate (excess reactivity zero) obtaining an increase (net gain) of reactivity with the 32 plates inserted (about 295 pcm). Therefore, it was observed that the reflected nucleus became more reactive than the nucleus without reflective material. The theoretical analysis using MCNP-5 and ENDF/B-VII.0 nuclear data library showed the physical aspects of neutron absorption and reflection in the heavy reflector considered; however, it presented a discrepancy when fast neutron reflection dominates the physical phenomenon of neutron transport. In order to verify the impact of other models of thermal scattering of hydrogen in water for the computational simulations of the experiments, three models were considered, besides the one used by the ENDF/B-VII.0 library: ENDF/B-VII.0 scattering law; new evaluation of the S (alpha, beta) for hydrogen bound in water performed in Bariloche Atomic Center, Argentina; and the calculated with new released evaluations for (235)U, (238)U and (16)O.

Palavras-Chave: absorption; benchmarks; control elements; fast neutrons; fuel rods; ipen-mb-1 reactor; monte carlo method; nickel; nuclear data collections; plates; reactivity; reactor cores; reflection; thermal neutrons; thickness

SILVA, GRACIETE S. de A. e; MURA, LUIS F.L.; FUGA, RINALDO; SANTOS, ADIMIR dos. IPEN/MB-01 reactor experiments with nickel reflectors. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4350-4361. Disponível em: http://repositorio.ipen.br/handle/123456789/30694. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Ex-core neutron detectors are commonly referred as a detector placed outside the reactor pressure vessel and in a typical SMR design its use is employed to reactor control. Due to its position (far from core) neutron flux calculation for ex-core detector purposes is challenging when using Monte Carlo codes, therefore this work presents an intercomparison between two Monte Carlo codes and also a neutron flux analysis (axially and radially) to better positioning the ex-core neutron detectors. Discrepancies regarding energy treatment can be evaluated as the MAVRIC sequence uses a set of cross sections in a multigroup energy structure while MCNP6 uses continuous energy. In this work, neutron flux intercomparison is mostly focused on variance reduction techniques since these codes presents different approaches, mainly because the MAVRIC sequence uses a hybrid approach combining deterministic and probabilistic methods and MCNP6 code uses traditional variance reduction methods. Some Monte Carlo variables such as figure-of-merit, CPU-time and error distributions maps are evaluated, and neutron flux magnitudes compared. To do so, a typical small modular reactor is modeled with the aid of MCNP6 code and the MAVRIC sequence in two different situations: one being a deep subcritical state with an external neutron source for variance reduction techniques comparison and the other a generic start up procedure (control rods removal) for detector position optimization.

Palavras-Chave: comparative evaluations; control elements; cross sections; finite difference method; graphite moderated reactors; m codes; monte carlo method; neutron detectors; neutron flux; neutron sources; optimization; reactor cores

SOUZA, GREGÓRIO; CARLUCCIO, THIAGO; SANCHEZ, PRISCILA; ABE, ALFREDO. Neutron flux intercomparison and ex-core neutron detector optimization in a SMR reactor using MCNP6 code and MAVRIC sequence. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4144-4163. Disponível em: http://repositorio.ipen.br/handle/123456789/30692. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The Brazilian Multipurpose Reactor (RMB) is a project funded by the Brazilian Government by means of the Ministry of Science Technology Innovation and Communication. RMB will be the new Brazilian research reactor, constructed to attend three main purposes: radioisotope production, R&D and material testing. It will be sited 125 km away from S~ao Paulo, strategically, at a Nuclear Compound, where a state owned pole of nuclear technology is located. To construct and operate the RMB facilities, as required by the National Environmental Policy, it is necessary, in addition to the nuclear licensing process of the National Nuclear Energy Commission (CNEN), to conduct all the environmental licensing stages with the Brazilian Environmental Agency (IBAMA). Given this regulatory scenario, based on the standards, guidelines and legal requirements of the IAEA, CNEN, IBAMA and other Brazilian o cial agencies, since 2012, the activities required to comply with the protocol for obtaining initial environmental and construction licenses is being implemented. This paper aims to show a timeline about this process, update the community and register further steps. The RMB entrepreneurs carried out the Environmental Impact Assessment issued the Local report for the radioprotection directory and held three public hearings. Those, among other e orts, resulted on the Local Approval License, which was issued by CNEN Deliberative Commission and on the Initial Environmental License issued by IBAMA. Both of these permits were placed in 2015. Since then some activities for complying with the permit conditions is being performed at the site and properly reported in order to obtain the installation license from the agency.

Palavras-Chave: environmental impacts; environmental protection; hearings; licensing; limiting values; nuclear facilities; rmb reactor; brazilian organizations

HONÓRIO, DANIEL H.; JESUS, MARCELO Z.; PERROTTA, JOSE A.; MOLNARY, LESLIE de; AQUINO, AFONSO R. Licensing aspects of the Brazilian Multipurpose Reactor (RMB). In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4078-4091. Disponível em: http://repositorio.ipen.br/handle/123456789/30690. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: In the second half of the twentieth century in Brazil, several nuclear facilities were built for the most varied objectives. The largest number of such facilities is at the Nuclear and Energy Research Institute in São Paulo (IPEN-CNEN/SP). For different reasons, some of these facilities had their projects finalized and were deactivated. Some of the equipment was then dismantled, but the respective nuclear and radioactive material remained isolated in the original sites awaiting the proper decommissioning procedures. The Celeste Project is an example of a facility where the nuclear material has been kept, and is subject to Argentine-Brazilian Agency for Accounting and Control of Nuclear Materials (ABACC) periodic inspections. Because of a number of interests, including financial and/or budgeting situations at the institutions, decades have passed without any further action, and the people who withold information and knowledge about these facilities have already moved away from the area or are in the process of. Therefore, this work proposes an analysis about the knowledge management reflecting on the possible consequences for the decommissioning processes, in case of loss of the knowledge acquired.

Palavras-Chave: decommissioning; historical aspects; information dissemination; information needs; knowledge management; nuclear facilities; radioactive materials; radioactive waste management; safety; brazilian cnen; brazil

SMITH, RICARDO B.; SALVETTI, TEREZA C.; TESSARO, ANA P.G.; MARUMO, JULIO T.; VICENTE, ROBERTO. Knowledge management in the decommissioning of nuclear facilities in Brazil. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3997-4007. Disponível em: http://repositorio.ipen.br/handle/123456789/30688. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The process of standardizing or normalizing "something" is a reality in various segments of society, from industry, commerce and even services require Technical Standards to confer a quality standard on any and all goods that are produced. The objective of this research was to verify the adherence of the Dissertations defended in the IPEN/USP Post-Graduation Program to the technical standards of ABNT documentation. We analyzed 85 dissertations made available in the Institutional Repository of the Institute, from 2007 to 2016; we chose to evaluate the adhesion of the Abstract, Literature Review, List of References and Page Formatting by means of a Likert Scale standard form. It was observed that 87% of the Abstracts presented were very adequate to the standards, against 12% that were very inadequate. The Literature Review was very adequate in 51% of the projects, although 27% presented as neither very adequate nor very inadequate (neutral). However, the List of References was inadequate to the norms in 69% of the projects. Finally, in the formatting format it was possible to observe that 56% of the projects were in agreement with the rules presented for paging. In this evaluation it was evidenced that the guide of the Institute has exerted a strong influence on the quality of the assignments, thus guaranteeing greater quality in the physical presentation of the dissertations of the IPEN Program.

Palavras-Chave: knowledge management; nuclear energy; education; document types; quality control; recommendations; standardization; brazilian cnen

LEOCADIO, MEIRILANE S.; IGAMI, MERY P.Z.; ANDRADE, DELVONEI A. de. Adherence of the IPEN post-graduation program dissertations to the ABNT norms. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3964-3969. Disponível em: http://repositorio.ipen.br/handle/123456789/30686. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Safeguards agencies are concerned with the safety of nuclear installations and the security of nuclear materials. Material protection, control, and accountancy are the first steps towards maintaining continuity of knowledge of these materials and preventing illicit trafficking or diversion of these materials for illicit purposes. Related concerns also exist in arms control, where the item chain of custody is important. In order to strengthen and improve the efficiency and effectiveness of existing safeguards measures, tamperproof devices and materials are needed capable of determining elapsed time since the undeclared movement of a source. Our group developed a new approach for surveillance based on passive, solid-state devices. Relying on a non-electronic detection mechanism is highly desirable because complex electronic components and circuits are potentially vulnerable to tampering and snooping. The device is a set of passive optically stimulated luminescent detectors based on calcium sulfate doped with various rare earths. The different doping produces different temporal fading profiles. When a source causes energy deposition in the detectors, the latter accumulate trapped electrons that undergo de-trapping at different rates. Thus, reading them out produces a set of signals that correlates both with the strength of the source and with the time of its passage.

Palavras-Chave: calcium sulfates; doped materials; radiation detectors; radioactive materials; rare earths; safeguards; security; thermoluminescent dosemeters

D’ERRICO, FRANCESCO; JUNOT, DANILO O.; POLO, IVON O.; CHIERICI, ANDREA; CIOLINI, RICCARDO; SOUZA, DIVANIZIA N.; CALDAS, LINDA V. E.; SOUZA, SUSANA O. Differential-fading optically stimulable materials for nuclear safeguards. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3838-3843. Disponível em: http://repositorio.ipen.br/handle/123456789/30684. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Despite all peaceful applications of nuclear technology, it is still addressed with prejudice. Prejudices may be explicit (conscious) or implicit (unconscious). However, either explicit or implicit, they interfere with individuals’ behavior and attitudes. Prejudices against any theme may be reduced and even reversed by new learning on the theme. Multisensory techniques have proven to make learning richer and more motivating. This work aims to present the development of a multisensory program designed for learning about the beneficial applications of nuclear technology and compare this program to a 12-week traditional teaching program with lecture classes about the nuclear technology. The multisensory program was held at the Instituto de Pesquisas Energéticas e Nucleares (IPEN) for a group of teachers. Assisted tours to the IEA-R1 and to the Centro da Tecnologia das Radiações (CTR) as well as a coffee break serving a variety of commercially-available foods containing irradiated ingredients were part of the multisensory approach. The Implicit Association Test (IAT) was administered before and after the program to identify and measure the implicit associations towards the nuclear technology. This multisensory program was compared to a 12-week traditional teaching program with lecture classes about the nuclear technology held at IPEN. Unlike the multisensory program, the IAT results from the traditional program demonstrated that the lecture classes were not effective for changing the implicit associations. The multisensory program was an effective tool for changing the implicit associations and can be useful for disseminating the beneficial applications of the nuclear technology.

Palavras-Chave: education; f codes; learning; nuclear energy; public opinion; technology impacts; testing; brazilian cnen

BARABAS, ROBERTA de C.; BARABÁS, CARLOS; SABUNDJIAN, GAIANE. The development of a multisensory program for the dissemination of the beneficial applications of the nuclear technology. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3770-3781. Disponível em: http://repositorio.ipen.br/handle/123456789/30682. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: One of the great advances in the current evolution of nuclear power reactors is occurring in India, with the Advanced Heavy Water Reactor (AHWR). It is a reactor that uses thorium as part of its fuel, which in its two fueling cycle options, in conjunction with plutonium or low enriched uranium, produces energy at the commercial level, generating less actinides of long half-life and inert thorium oxide, which leads to an optimization in the proportion of energy produced versus the production of burnt fuels of the order of up to 50%. The objective of this work is to present the most recent research and projects in progress in India, and how the expected results should be in compliance with the current sustainability models and programs, especially the "Green Chemistry", a program developed since the 1990s in the United States and England, which defines sustainable choices in its twelve principles and that can also be mostly related to the nuclear field. Nevertheless, in Brazil, for more than 40 years there has been the discontinuation of research for a thorium-fueled reactor, and so far there has been no prospect of future projects. The AHWR is an important example as an alternative way of producing energy in Brazil, as the country has the second largest reserve of thorium on the planet.

Palavras-Chave: fuel element clusters; hwlwr type reactors; india; nuclear fuels; radioactive wastes; reactor design; sustainable development; thorium

SMITH, RICARDO B.; SACHDEVA, MAHIMA; BISURI, INDRANIL; VICENTE, ROBERTO. Advanced heavy water reactor: a new step towards sustainability. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3567-3579. Disponível em: http://repositorio.ipen.br/handle/123456789/30680. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: One of the main purposes of nuclear technology is to produce electricity, with the advantage of producing a lower volume of radioactive waste. The expansion of nuclear energy in the electrical system has been positive, as it is one of the types of energy that is available at any time and in the desired amount. Considered a reliable source and safe alternative to compose a country's energy matrix. In the case of Brazil, it has enough reserves of Uranium and Thorium to compose the energy matrix over many years. The increase in demand, and the need for energy from renewable sources has caused changes in the world's electric power generation. According to World Nuclear Association (WNA), 14% of the energy is generated by nuclear energy sources, and this percentage tends to increase with the construction of new plants. According to the International Atomic Energy Agency (IAEA), the goal for nuclear energy is to provide 25% of electricity in 2050. Other technologies are applied in the nuclear area, for example nuclear medicine, in which radioactive materials are used with low doses of radiation for treatment and diagnosis of diseases, even in development are effective and safe, especially in the areas of cardiological, neurological and oncological diagnosis. Despite the knowledge acquired with the development of Brazilian nuclear projects, many are partly lost and discontinuity investments of successive governments, therefore, this work intends to study an overview of nuclear energy in Brazil in recent years and its prospects.

Palavras-Chave: electric power; energy security; global aspects; nuclear energy; nuclear medicine; power generation; brazil

MOREIRA, RENAN P.; TATEI, TATIANE Y.; ARAUJO, DANIELLE G.; DUQUE, MARCO A. da S.; OLIVEIRA, IVAN C. de; AYOUB, JAMIL M.S.; SENEDA, JOSE A. Prospects for nuclear energy in Brazil. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3511-3516. Disponível em: http://repositorio.ipen.br/handle/123456789/30678. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Uma das formas de identificar algum tipo de preconceito é por meio do uso de softwares de técnicas de neurociências aplicadas ao uso de medida da memória implícita (Testes de Associação Implícita –TAI), que não depende da atenção consciente do participante, sendo suas respostas automáticas e espontâneas. Os seguintes testes de associação implícita foram encontrados na literatura: o Teste de Associação Implícita, o Priming, o Visual Organization Test (VOT) e o Inquisit. Dentre todos os softwares de associação implícitas apresentados, o FreeIAT será utilizado neste trabalho pelo fato de ser um programa largamente usado e validado em diversas pesquisas. Pelo fato desse programa apresentar resultados bem consistentes quanto à identificação de possíveis preconceitos em vários temas, viu-se a necessidade de elaborar um pós-processador matemático a fim de automatizar os resultados em forma de gráficos. Portanto, o objetivo desse trabalho é o de desenvolver um pós-processador matemático com interface amigável, que facilitará a apresentação e interpretação dos resultados dos usuários do FreeIAT e poderá ser utilizado em qualquer área de interesse. A linguagem utilizada para o desenvolvimento desse pós-processador é o C#. Os resultados preliminares desse novo pós-processador mostraram-se eficientes.

Palavras-Chave: automation; data analysis; f codes; g codes; graphical user interface; programming languages

SILVESTRE, LARISSA J.B.; SOUSA, EMERSON L.; SABUNDJIAN, GAIANE. Pós-processador matemático para o software de teste de associação implicita – FreeIAT. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3368-3374. Disponível em: http://repositorio.ipen.br/handle/123456789/30676. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: O siliceto de urânio é um intermetálico usado como combustível nuclear na maioria dos reatores de pesquisa modernos, incluindo os reatores MB-01 e IEA-R1 do IPEN. Durante a produção, o material é submetido a um rigoroso controle de qualidade, que inclui análises de tamanho de partícula, densidade, caracterização e composição da fase cristalina. A quantificação das fases cristalinas presentes é realizada por difração de raios X (DRX) e refinamento dos dados usando o método Rietveld. No entanto, devido à alta absorção de raios X por esse material, no que diz respeito ao método de quantificação adotado, é muito importante reduzir o tamanho das partículas. Para este objetivo, um moinho vibratório dedicado é usado antes da análise de DRX, reduzindo o diâmetro médio das partículas para poucos micrômetros. Para evitar a oxidação das amostras, o processo de moagem ocorre em meio isopropanóico, o qual é seco posteriormente, em vácuo a 80 ºC. Porém, em muitos casos, verifica-se que as massas das amostras moídas são maiores do que as iniciais. Nesse sentido, esse trabalho propõe analisar a causa dessa diferença de massa. Granulometria a laser, termogravimetria (TG). Os resultados de TG sugerem que uma camada é fortemente adsorvida ao material, protegendo o pó de oxidação em temperaturas acima de 4000C.

Palavras-Chave: adsorption; crystal structure; fuel elements; milling; particle size; thermal gravimetric analysis; uranium silicides; x-ray diffraction

OLIVEIRA, VITORIA A.; CARVALHO, ELITA U.; DURAZZO, MICHELANGELO; SAKATA, SOLANGE K.; GARCIA, RAFAEL H.L. Adsorção líquida no siliceto de urânio. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3193-3198. Disponível em: http://repositorio.ipen.br/handle/123456789/30674. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: A maioria dos países enfrenta grandes desafios para controlar e organizar a geração e a disposição dos resíduos sólidos urbanos. Milhões de toneladas desses resíduos são gerados anualmente pela população e pelas indústrias. A eliminação destes resíduos sólidos é um problema mundial crescente. Os materiais poliméricos (plásticos e borrachas) compreendem uma proporção cada vez maior de resíduos industriais que entram em aterros sanitários e ambientais. Devido à capacidade da radiação ionizante alterar a estrutura e as propriedades dos materiais poliméricos, e o fato de que ela é aplicável a todos os tipos de polímeros, a irradiação é promissora para tratar do problema de resíduos poliméricos. O objetivo deste trabalho foi utilizar a radiação gama proveniente de uma fonte de 60Co para reciclar o poli(tetrafluoroetileno) (PTFE) que é um polímero de difícil decomposição quando descartado no meio ambiente. Aparas industriais deste polímero foram selecionadas e submetidas ao processo de moagem. Posteriormente, as amostras foram submetidas ao processo de irradiação com uma dose de 200 kGy. Após a irradiação o material obtido foi micronizado obtendo-se um pó muito fino de PTFE o qual foi classificado de acordo com os tamanhos de partículas com características especiais para diferentes possibilidades de utilização industrial (aditivos para tintas, massas lubrificantes, óleos e como carga em polímero para diminuir o coeficiente de atrito).

Palavras-Chave: cobalt 60; grinding; ionizing radiations; particle size; polytetrafluoroethylene; recycling; scrap

SOUZA, ERIC W. de; VIEIRA, JOSE M.; SILVA, LEONARDO G. de A. e. Uso da radiação ionizante na reciclagem de poli (tetrafluoroetileno) (PTFE). In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3167-3171. Disponível em: http://repositorio.ipen.br/handle/123456789/30672. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Lithium Compounds have applications in strategic areas for intern consumption of a country as well as international commerce. In nuclear industry, the lithium is used for the cooling of PWR reactors as a pH stabilizer. Based on this assumption, the generation of knowledge to master the processing cycle of these compounds is essential. The high degree of purity of lithium compounds is determinant to have success in these applications. Lithium hydroxide LiOH and lithium carbonate Li2CO3 are the main forms in which lithium is used industrially. To improve the quality of the starting product, purifying process were used until obtaining an adequate purity level of raw material (> 99%). The present work aims to make feasible a purification of Li2CO3 through ion-exchange chromatography from a 98.5% purity compound. The impurities present in higher content are sodium and calcium. To separate these two elements from lithium or at least to lower their concentrations, a column with cationic resin was used to fix lithium. The determination of lithium, sodium and calcium contents in the solutions was performed by inductively coupled plasma optical emission spectrometry, ICP-OES. The experiments performed to evaluate the best lithium purification condition were based on the variation of the main operational parameters: pH, flow and elution solution. The results indicate increased purity from the application of ion exchange operations obtaining a suitable condition for nuclear uses.

Palavras-Chave: aqueous solutions; calcium; impurities; ion exchange chromatography; lithium carbonates; purification; sodium

ANDRADE, MARIANA N.; OLIVEIRA, GLAUCIA A.C.; PIRANI, DEBORA A.; COUTINHO, JOAO F.; BERGAMASCHI, VANDERLEI S.; SENEDA, JOSE A.; BUSTILLOS, JOSE O.V. Purification of lithium carbonate by ion-exchange processes for application in nuclear reactors. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3153-3157. Disponível em: http://repositorio.ipen.br/handle/123456789/30670. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: A liga de prata-indio-cádmio (Ag/In/Cd) é utilizada como material absorvedor em elementos de controle de reatores nucleares devido à alta seção de choque para absorção de nêutrons de seus componentes. Em Reatores Refrigerados a Água Pressurizada (PWR - Pressurized Water Reactor), a liga Ag/In/Cd é utilizada na forma de barra contendo 80% de prata, 15% de índio e 5% de cádmio em massa com tolerâncias, máxima e mínima, bastante rigorosas em sua composição. A liga na forma de barra é encapsulada em tubos metálicos, os quais compõem o conjunto do elemento de controle no reator nuclear. Para ser aplicada com este propósito, a barra de liga Ag/In/Cd deve apresentar uma composição homogênea ao longo de toda a sua extensão, a fim de assegurar seu comportamento adequado dentro do reator. O objetivo deste projeto é desenvolver e qualificar a metodologia de análise química aplicada à caracterização da liga Ag/In/Cd para ser usada em barras de controle em reatores do tipo PWR. A metodologia padronizada para determinar o teor de prata, índio e cádmio na liga de grau nuclear é a titulação potenciométrica para prata e a titulação de complexação para o índio e o cádmio. A precisão dos resultados obtidos depende da prévia calibração dos materiais volumétricos e equipamentos utilizados, bem como da calibração dos reagentes titulantes a serem utilizados na titulação. Além disso, a qualificação desse processo para fins nucleares requer a elaboração de todos os documentos relacionados a cada uma das etapas do processo, incluindo práticas operacionais e registros da qualidade. O desenvolvimento e a qualificação da metodologia representam passos fundamentais no sentido de tornar o Brasil autossuficiente na produção desse material aplicado à área nuclear.

Palavras-Chave: accuracy; cadmium alloys; calibration; control elements; indium alloys; pwr type reactors; quality assurance; silver alloys; ternary alloy systems; titration

PEREIRA, DEBORA A.; FERREIRA, DOUGLAS A.; FATTE, MARIO; SOUZA, NATALIA DE O.; GIOVEDI, CLAUDIA; COTRIM, MARYCEL E.B.; PIRES, MARIA A. Análise química de liga de grau nuclear aplicada como material de controle em reatores nucleares. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3117-3129. Disponível em: http://repositorio.ipen.br/handle/123456789/30668. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Atualmente, os tratamentos mais utilizados para o combate ao câncer são cirurgia, quimioterapia e radioterapia, o aperfeiçoamento desses métodos são cada vez mais solicitados. Diante deste cenário de crescente incidência do câncer de mama há a necessidade de que os serviços de radioterapia se adaptem à grande demanda de pacientes. Estudos têm apontado que o hipofracionamento pode ser uma técnica promissora para a redução da quantidade de sessões por paciente, consequentemente diminuindo o tempo total de tratamento. O hipofracionamento de dose é uma técnica utilizada na radioterapia, seu escopo é reduzir a quantidade de sessões por paciente, sem comprometer o tratamento. A proposta deste estudo foi destacar o impacto do uso dessa técnica no câncer de mama no Brasil, a fim de que os profissionais envolvidos estejam efetivamente bem preparados para utilizar essa técnica com excelência, tratando um número maior de pacientes em um período menor de tempo sem diminuir a qualidade da terapia.

Palavras-Chave: fractionated irradiation; mammary glands; neoplasms; radiation doses; radiotherapy; side effects

SUKADOLNIK, MIKAELL P.; SILVA, RODRIGO A. da; SOUZA, DAIANE C.B. de; MUNOZ, BERGMAN N.S. O impacto do hipofracionamento de dose na saúde da mulher brasileira acometida com câncer de mama. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3064-3068. Disponível em: http://repositorio.ipen.br/handle/123456789/30665. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Na extração de petróleo, há presença significativa de material radioativo de origem natural e, por isso, as empresas que realizam esse trabalho devem atender as normas de proteção radiológica estabelecidas pela Comissão Nacional de Energia Nuclear (CNEN). Determinar a concentração radioisotópica em rejeitos radioativos é um passo fundamental no processo de caracterização dos rejeitos e é essencial no tratamento, no transporte e na eliminação deles. Este estudo consistiu na utilização de medidas das taxas de dose e cálculos para estimativa do conteúdo radioativo presente em tambores com rejeitos provenientes da indústria de petróleo. Foi utilizado o programa para cálculo de blindagem Microshield. Os principais resultados obtidos foram os valores de taxa de dose e a espectrometria de emissão gama. O método baseado na medição das taxas de exposição em torno de embalados fornece boas aproximações quando as informações sobre emissores gama presentes nos embalados de rejeitos são obtidos por espectrometria gama.

Palavras-Chave: containers; dose rates; gamma radiation; gamma spectroscopy; m codes; petroleum industry; point kernels; radiation doses; radiation monitoring; radioactive wastes; radioactivity; tailings

SILVA, RODRIGO A. da; SUKADOLNIK, MIKAELL P.; TESSARO, ANA P.G.; SOUZA, DAIANE C.B. de; VICENTE, ROBERTO. Caracterização de embalados de rejeitos radioativos utilizando Microshield. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 3027-3033. Disponível em: http://repositorio.ipen.br/handle/123456789/30663. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: O processo desordenado da industrialização originou graves problemas de poluição para o meio aquático. Inúmeros episódios de contaminação de regiões costeiras, ocorridos no mundo todo, levaram muitos países a estabelecer extensos programas de monitoração, que incluem análises de águas, sedimentos e organismos marinhos, para diversos contaminantes orgânicos e inorgânicos. As concentrações de substâncias potencialmente tóxicas em água do mar são extremamente baixas e consideravelmente diversificadas no espaço e no tempo, tornando assim suas determinações complexas. O presente trabalho consiste em avaliar a exposição, os efeitos e a bioacumulação de contaminantes em conchas de mexilhões Perna perna e vôngole Anomalocardia brasiliana nativos, por um período de um ano (4 estações) em Praias de Caraguatatuba pela avaliação da bioacumulação sazonal dos seguintes elementos: As, Co, Cr, Fe, Se e Zn pelo métodos de análise por ativação com nêutrons instrumental (INAA), Cd e Pb chumbo por espectrometria de absorção atômica com forno de grafite (GF-AAS) e Hg, por espectrometria de absorção atômica com geração de vapor frio (CV-AAS).

Palavras-Chave: absorption spectroscopy; biological accumulation; chemical composition; elements; mussels; neutron activation analysis; pollution

PAULA, JOSE H. de; CATHARINO, MARILIA G.M.; THEOPHILO, CAROLINA Y.S.; SOUSA, EDUINETTY C.P.M. de; GASPARIO, MARCIA R.; LINS, CLAUDIO G.; SILVA, PAULO S.C. da. Avaliação da qualidade da área costeira da região de Caraguatatuba utilizando conchas de organismos bivalves. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 2934-2941. Disponível em: http://repositorio.ipen.br/handle/123456789/30661. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Nas últimas décadas, com o conhecimento do papel dos elementos químicos na saúde humana e o aperfeiçoamento de métodos analíticos, análises de impurezas de elementos químicos em medicamentos sintéticos se tornaram de grande interesse. O objetivo deste estudo foi determinar pelo método de Análise por Ativação com Nêutrons (NAA) elementos químicos presentes em dois medicamentos codificados de D e R, muito utilizados pela população. O medicamento D é utilizado como analgésico e atua aliviando a dor e a febre e o R é utilizado para a redução dos níveis de colesterol e triglicérides. Para análise, os dois medicamentos foram triturados na forma de pó e, alíquotas de cada uma das amostras foram pesadas e irradiadas com os padrões sintéticos dos elementos no reator nuclear IEA R1, sob fluxo de nêutrons térmicos por 16 h. Após adequados tempos de decaimento, as atividades gama das amostras e padrões foram medidas usando um detector de germânio hiperpuro ligado a um analisador digital de espectro. Nos medicamentos D e R foram determinados os elementos Br, Co, Cr, Cs, Fe, K, Na, Sb, Sc e Zn e, no medicamento R, além destes elementos, foi determinado o CA. Para avaliar a qualidade analítica dos resultados, foi analisado o material de referência certificado INCT MPH 2 (Mixed Polish Herbs) e os resultados obtidos indicaram uma boa reprodutibilidade e concordância com os valores do certificado. O estudo demonstrou a viabilidade de aplicar o método NAA na avaliação dos elementos presentes em medicamentos sintéticos. Além disso, os resultados obtidos pelos limites de detecção e de quantificação indicaram uma alta sensibilidade do método para análise.

Palavras-Chave: elements; gamma radiation; medicinal plants; neutron activation analysis; neutron flux; thermal neutrons

ARAUJO, MARINA NANO; SAIKI, MITIKO. Determinação de elementos químicos em medicamentos sintéticos pelo método de análise por ativação com nêutrons. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 2869-2878. Disponível em: http://repositorio.ipen.br/handle/123456789/30659. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: A radiologically significant nuclides catalog is paramount to the classification of radioactive waste. In order to produce such a catalog, it is required to know the isotopic composition of the radioactive waste produced in the nuclear power plants, their isotopic inventories and both short and long term toxicity for each relevant nuclide in each exposure scenario. Estimating the waste produced in a power plant is an old problem that still poses a great challenge, even with current technology and methods. This paper describes an attempt at estimating the radionuclide concentration levels of the radioactive waste produced in the Angra dos Reis nuclear power plant. A review on the various methods used around the world to estimate these isotopic compositions was needed in order to achieve such a result. Alongside the review was used a computer simulation with the Origen 6.0 code and calculations to find out the future activities and toxicities for each analyzed radionuclide in each considered exposure scenario. The resulting data is used to build on a radiologically significant nuclides catalog that can be used as guiding tool for the development of radiological containment policies. This data will be helpful for the long term storage of the studied radionuclides.

Palavras-Chave: angra-1 reactor; angra-2 reactor; catalogs; classification; computerized simulation; isotope ratio; o codes; radioactive wastes; radioactivity; radioisotopes

OLIVEIRA JUNIOR, JULIO de; VICENTE, ROBERTO. Identification of potentially relevant radionuclides in the nuclear central of Angra dos Reis. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 2855-2864. Disponível em: http://repositorio.ipen.br/handle/123456789/30657. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The objective of this work is to analyze Cl in whole blood of newborns, concomitant with the traditional collection of “the heel prick test” (fourth drop). The dosage of Cl in whole blood samples of twenty newborns were determined by Energy Disperse X-Ray Fluorescence (EDXRF) and the Instrumental Neutron Activation Analyses (INAA) analytical techniques. Particularly, the alternative methodology based on EDXRF technology, using a portable XRF spectrometer, showed to be a fast and efficient procedure for Cl dosage in whole blood. We intend to introduce benefits to clinical practice in children, especially newborns and premature infants using this alternative procedure.

Palavras-Chave: blood; chlorine; concentration ratio; gamma radiation; high-purity ge detectors; neonates; neutron activation analysis; optimization; thermal neutrons; x-ray fluorescence analysis

MELO, JULIANA S.; ZAMBONI, CIBELE B.; AZEVEDO, MARIA R.; KONSTANTYNER, TULIO. Evaluation of Cl in blood by NAA and XRF techniques: an alternative for pediatric practice. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 2825-2828. Disponível em: http://repositorio.ipen.br/handle/123456789/30655. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: In southwest Amazonia, in the region of the Upper Madeira River, Rondônia, archaeological research has shown that communities with diverse cultures co-existed in the pre-colonial past (ca. 1,000 BP). Archaeological sites from this period located on river banks and islands consist of large extensions of ceramic deposits which reflect different daily activities and social positions that existed within these groups. The complexity of these societies is attested to by the diversity of both ceramic forms and iconography. In this work, 140 ceramic fragments from eight archaeological sites were studied by means of instrumental neutron activation analysis (INAA) to determine Na, K, La, Sm, Yb, Lu, U, Sc, Cr, Fe, Co, Zn, Rb, Cs, Ce, Eu, Hf, Ta and Th mass fractions, with the purpose of classifying and ordering artifacts which are related to one another in their chemical compositions. The analytical method used is adequate for this type of study because it is a semi-destructive technique with high sensitivity and precision that can determine chemical elements in trace and ultra-trace levels, essential for studying small variations in elemental concentrations. Multivariate statistical analyses were used to evaluate the dataset. Initially the mass fractions were normalized to compensate for the large difference in magnitude among elements determined in percentage and in trace level. Subsequently, the mass fraction data were interpreted through cluster analysis, discriminant analysis and a log-log scatterplot. The results show the existence of four compositional groups, indicating different clay sources.

Palavras-Chave: archaeology; archaeological sites; cultural objects; ceramics; chemical composition; archaeological specimens; neutron activation analysis; clays; raw materials; resources; brazil

COSTA, ANGISLAINE F.; MUNITA, CASIMIRO S.; ZUSE, SILVANA; KIPNIS, RENATO. Archaeometry and archaeology: preliminary studies of the ceramics from archaeological sites of the Upper Madeira River/Rondônia - Brazil. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 2747-2760. Disponível em: http://repositorio.ipen.br/handle/123456789/30653. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: This study measured the chemical composition of 45 ceramic samples from the Jericho archaeological site, Palestine, by means of instrumental neutron activation analysis (INAA). The mass fraction of Na, K, La, Sm, Yb, Lu, U, Sc, Cr, Fe, Co, Zn, Rb, Cs, Ce, Eu, Hf and Th was determined with the purpose to detect the presence of ceramic groupings based on their composition. The analytical method is appropriate for this type of study because it is a non-destructive technique with high sensitivity, accuracy and precision, and determines chemical elements in trace and ultra-trace levels. These characteristics are essential to study small concentration variations. Initially the mass fractions were normalized to compensate for the large difference in magnitude among elements determined in percentage and trace level. Subsequently, the dataset was interpreted through cluster and discriminant analysis. The results showed the existence of three different chemical groups.

Palavras-Chave: archaeological sites; chemical composition; neutron activation analysis; middle east; historical aspects; cultural objects; gasers; spectroscopy; multivariate analysis; ceramics

BARROS, JOANNA F.; SILVA, RODRIGO P. da; MUNITA, CASIMIRO S. Preliminary chemical studies at the Jericho archaeological site. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 2709-2717. Disponível em: http://repositorio.ipen.br/handle/123456789/30651. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: Conservation and preservation strategies are essential to manage botanical collections specially for dried herbarium specimens also known as exsiccates, usually referring to a set of identified specimens belonging to taxa and distributed among all herbaria around the world. Particularly, these collections are very sensitive to the attack of fungi and insects. In recent years, disinfection by ionizing radiation has become an effective strategy to preserve cultural heritage objects and archived materials with excellent results. In this work, the effects on color properties of gamma radiation on exsiccates samples were studied. Thus, six exsiccates, botanical pressed and dehydrated samples were selected from the Dom Bento José Pickel Herbarium (SPSF), situated at São Paulo (Brazil). Three of these samples comes from Asteraceae family and were collected in 1946, 1984 and 1986, while three other samples belong to Solanaceae family and were collected in 1953, 1984 and 2007. Families of selected botanical collections are very susceptible to biodegradation. The irradiation was performed at the Multipurpose Gamma Irradiation Facility at IPEN applying absorbed doses of 1 kGy, 6 kGy and 10 kGy, which are values of absorbed dose for disinfestation and disinfection. Results were analyzed using colorimetry with CIELAB color space scale and scanning electron microscopy. The results showed that there were no significant changes on colorimetric morphological properties of the samples.

Palavras-Chave: absorbed radiation doses; botany; calorimetry; cobalt 60; disinfestation; fungi; gamma radiation; morphological changes; pest control; plants; radiation dose units; scanning electron microscopy

LIMA, LENI M.P.R.; KODAMA, YASKO; OTUBO, LARISSA; SANTOS, PAULO de S.; VASQUEZ, PABLO A. Effect of ionizing radiation on the color of botanical collections: exsiccata. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 2650-2661. Disponível em: http://repositorio.ipen.br/handle/123456789/30649. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item