Estimativa de fontes externas uniformes de nêutrons em regiões combustíveis para uma distribuição prescrita de potência

Projetos de reatores nucleares foram classificados em quatro gerações (Gen) pelo Departamento de Energia dos Estados Unidos da América (DOE), quando o DOE introduziu o conceito de reatores de geração IV (Gen IV). Reatores Gen IV são um conjunto de projetos de reator nuclear, em sua maioria teóricos, atualmente sendo pesquisados. Entre os projetos Gen IV, incluem-se os projetos dos ADS (Accelerator Driven Systems), que são sistemas subcríticos estabilizados por fontes externas estacionárias de nêutrons. Estas fontes externas de nêutrons são normalmente geradas a partir da colisão de prótons com alta energia contra os núcleos de metais pesados presentes no núcleo do reator, fenômeno que é conhecido na literatura como spallation, e os prótons são acelerados num acelerador de partículas que é alimentado com parte da energia gerada pelo reator. A criticalidade de um sistema mantido por reações de fissão em cadeia depende do balanço entre a produção de nêutrons por fissão e a remoção por fuga pelos contornos e absorção de nêutrons. Um sistema está subcrítico quando a remoção por fuga e absorção ultrapassa a produção por fissão e, portanto, tende ao desligamento. Entretanto, qualquer sistema subcrítico pode ser estabilizado pela inclusão de fontes estacionárias de nêutrons em seu interior. O objetivo central deste trabalho é determinar as intensidades dessas fontes uniformes e isotrópicas de nêutrons, que se deve inserir em todas as regiões combustíveis do sistema, para que o mesmo estabilize-se gerando uma distribuição prescrita de potência elétrica. Diante do exposto, foi desenvolvido neste trabalho um aplicativo computacional em linguagem Java que estima as intensidades dessas fontes estacionárias de nêutrons, que devem ser inseridas em cada região combustível para que estabilizem o sistema subcrítico com uma dada distribuição de potência definida pelo usuário. Para atingir este objetivo, o modelo matemático adotado foi a equação unidimensional de transporte de nêutrons monoenergéticos na formulação de ordenadas discretas (SN) e o convencional método de malha fina diamond difference (DD) foi utilizado para resolver numericamente os problemas SN físicos e adjuntos. Resultados numéricos para dois problemas-modelos típicos são apresentados para ilustrar a acurácia e eficiência da metodologia proposta.

The effects of operational and design parameters on the tolerance of ADSR fuel pin cladding to beam interruptions

This paper investigates the effects of design parameters, such as cladding and coolant material choices, and operational phenomena, such as creep and fission product decay heat, on the tolerance of Accelerator Driven Subcritical Reactor (ADSR) fuel pin cladding to beam interruptions. This work aims to provide a greater understanding of the integration between accelerator and nuclear reactor technologies in ADSRs. The results show that an upper limit on cladding operating temperature of 550 °C is appropriate, as higher values of temperature tend to accelerate creep, leading to cladding failure much sooner than anticipated. The effect of fission product decay heat is to reduce significantly the maximum stress developed in the cladding during a beam-trip-induced transient. The potential impact of irradiation damage and the effects of the liquid metal coolant environment on the cladding are discussed. © 2013 Elsevier Ltd. All rights reserved.

Implementation of multi-group cross-section methodology in BGCore MC-depletion code

BGCore is a software package for comprehensive computer simulation of nuclear reactor systems and their fuel cycles. The BGCore interfaces Monte Carlo particles transport code MCNP4C with a SARAF module - an independently developed code for calculating in-core fuel composition and spent fuel emissions following discharge. In BGCore system, depletion coupling methodology is based on the multi-group approach that significantly reduces computation time and allows tracking of large number of nuclides during calculations. In this study, burnup calculation capabilities of BGCore system were validated against well established and verified, computer codes for thermal and fast spectrum lattices. Very good agreement in k eigenvalue and nuclide densities prediction was observed for all cases under consideration. In addition, decay heat prediction capabilities of the BGCore system were benchmarked against the most recent edition of ANS Standard methodology for UO2 fuel decay power prediction in LWRs. It was found that the difference between ANS standard data and that predicted by the BGCore does not exceed 5%.

Behaviour of laterised normal and self compacting concrete subjected to elevated temperatures

Concrete is a universal material in the construction industry. With natural resources like sand and aggregate, fast depleting, it is time to look for alternate materials to substitute these in the process of making concrete. There are instances like exposure to solar radiation, fire, furnaces, and nuclear reactor vessels, special applications like missile launching pads etc., where concrete is exposed to temperature variations In this research work, an attempt has been made to understand the behaviour of concrete when weathered laterite aggregate is used in both conventional and self compacting normal strength concrete. The study has been extended to understand the thermal behaviour of both types of laterised concretes and to check suitability as a fire protection material. A systematic study of laterised concrete considering parameters like source of laterite aggregate, grades of Ordinary Portland Cement (OPC) and types of supplementary cementitious materials (fly ash and GGBFS) has been carried out to arrive at a feasible combination of various ingredients in laterised concrete. A mix design methodology has been proposed for making normal strength laterised self compacting concrete based on trial mixes and the same has also been validated. The physical and mechanical properties of laterised concretes have been studied with respect to different variables like exposure temperature (200°C, 400°C and 600°C) and cooling environment (air cooled and water cooled). The behaviour of ferrocement elements with laterised self compacting concrete has also been studied by varying the cover to mesh reinforcement (10mm to 50mm at an interval of 10mm), exposure temperature and cooling environment.

COCO-2: a model to assess the economic impact of an accident

COCO-2 is a model for assessing the potential economic costs likely to arise off-site following an accident at a nuclear reactor. COCO-2 builds on work presented in the model COCO-1 developed in 1991 by considering economic effects in more detail, and by including more sources of loss. Of particular note are: the consideration of the directly affected local economy, indirect losses that stem from the directly affected businesses, losses due to changes in tourism consumption, integration with the large body of work on recovery after an accident and a more systematic approach to health costs. The work, where possible, is based on official data sources for reasons of traceability, maintenance and ease of future development. This report describes the methodology and discusses the results of an example calculation. Guidance on how the base economic data can be updated in the future is also provided.

Investigation of whole blood of SJL/J mice using neutron activation analysis

The Br (0.0022 +/- A 0.0006 gL(-1)), Ca (0.113 +/- A 0.012 gL(-1)), Cl (3.07 +/- A 0.36 gL(-1)), K (2.63 +/- A 0.14 gL(-1)), Mg (0.045 +/- A 0.002 gL(-1)) and Na (2.09 +/- A 0.10 gL(-1)) concentrations were determined in whole blood of SJL/J mice using the Neutron Activation Analysis (NAA) technique. Eleven whole blood samples were analyzed in the IEA-R1 nuclear reactor at IPEN (So Paulo, Brazil). These data contribute for applications in veterinary medicine related to biochemistry analyses using whole blood. Moreover, the correlation with human blood estimation allows to checking the similarities for studying muscular dystrophy using this model animal.

A física da terapia de captura de nêutrons pelo boro

Fundação de Amparo à Pesquisa do Estado de São Paulo (FAPESP)

Experimental study of a methodology for Fission-track Dating without neutron irradiation

To carry out the dating by the Fission Track Method (FTM) the international community that works with this method employs methodologies in which the mineral to be dated must be irradiated with neutrons. Such irradiation, performed in a nuclear reactor, demand a relatively long waiting time so that the activity of the sample attain a proper level for handling. The present work aims to establish a methodology that makes possible the dating by FTM using a mass spectrometer instead of a nuclear reactor. This methodology was applied to apatite samples from Durango, Mexico. © 2009 Elsevier Ltd. All rights reserved.

The licensing of a laboratory for manipulating radionuclides in Brazil

Radionuclides take a major role in guidelines of environmental agencies/national organizations of countries worldwide. In Brazil, CNEN-Comissão Nacional de Energia Nuclear is responsible for managing all subjects related to nuclear energy in the country. Thus, laboratories employing radionuclides for the development of their activities must submit a Radioprotection Plan to CNEN in order to get an operation license. Such plan must indicate that the laboratory is exempt of risks to the people involved and designed to fit all related environmental aspects. This was the case of LABIDRO-Hydrochemical and Isotopes Laboratory that belongs to IGCE-Geosciences and Exact Sciences Institute from UNESP - the University of the State of São Paulo Júlio de Mesquita Filho, located at Rio Claro city, São Paulo State, Brazil. The total monthly activity of the radionuclides utilized during the laboratorial activities held at LABIDRO corresponds to 0.01 μCi. This paper describes all information provided by LABIDRO in order to get the CNEN license. The LABIDRO plan also showed the expected radioactive waste released when the experiments take place and CNEN decided that it fits the guidelines established by Brazilian legislation. Therefore, LABIDRO received its license for utilizing radionuclides, which is valid until September 2016. © 2013 WIT Press.

Método de traços de fissão em zircão: estudos geocronológicos no grupo Bauru

Coordenação de Aperfeiçoamento de Pessoal de Nível Superior (CAPES)

Boundedness of solutions of retarded functional differential equations with variable impulses via generalized ordinary differential equations

In this paper, we give sufficient conditions for the uniform boundedness and uniform ultimate boundedness of solutions of a class of retarded functional differential equations with impulse effects acting on variable times. We employ the theory of generalized ordinary differential equations to obtain our results. As an example, we investigate the boundedness of the solution of a circulating fuel nuclear reactor model.

Optische Spektroskopie an Fermium (Z=100)

Im Rahmen der vorliegenden Arbeit wurde erstmals Laser-Atomspektroskopie an einem Element durchgeführt, für das bisher keine atomaren Niveaus bekannt waren. Die Experimente wurden am Element Fermium mit der Ordnungszahl Z=100 mit der Resonanzionisationsspektroskopie (RIS) in einer Puffergaszelle durchgeführt. Verwendet wurde das Isotop 255Fm mit einer Halbwertszeit von 20.1 h, das im Hochflusskernreaktor des ORNL, Oak Ridge, USA, hergestellt wurde. Die von einem elektrochemischen Filament in das Argon-Puffergas bei einer Temperatur von 960(20)°C abgedampften Fm-Atome wurden mit Lasern in einem Zweistufenprozess resonant ionisiert. Dazu wurde das Licht eines Excimerlaser gepumpten Farbstofflasers für den ersten Anregungsschritt um die Wellenlänge 400 nm durchgestimmt. Ein Teil des Excimer (XeF) Laser Pumplichtes mit den Wellenlänge 351/353 nm wurde für die nicht-resonante Ionisation verwendet. Die Ionen wurden mit Hilfe elektrischer Felder aus der optischen Zelle extrahiert und nach einem Quadrupol Massenfilter mit einem Channeltron-Detektor massenselektiv nachgewiesen. Trotz der geringen Probenmenge von 2.7 x 10^10 eingesetzten Atomen wurden zwei atomare Resonanzen bei Energien von 25099.8(2) cm-1 und 25111.8(2) cm-1 gefunden und das Sättigungsverhalten dieser Linien gemessen. Es wurde ein theoretisches Modell entwickelt, dass sowohl das spektrale Profil der sättigungsverbreiterten Linien als auch die Sättigungskurven beschreibt. Durch Anpassung an die Messdaten konnten die partiellen Übergangsraten in den 3H6 Grundzustand Aki=3.6(7) x 10^6/s und Aki=3.6(6) x 10^6/s bestimmt werden. Der Vergleich der Niveauenergien und Übergangsraten mit Multikonfigurations Dirac-Fock Rechnungen legt die spektroskopische Klassifizierung der beobachteten Niveaus als 5f12 7s7p 5I6 und 5G6 Terme nahe. Weiterhin wurde ein Übergang bei 25740 cm-1 gefunden, der aufgrund der beobachteten Linienbreite von 1000 GHz als Rydbergzustand Zustand mit der Niveauenergie 51480 cm-1 interpretiert wurde und über einen Zweiphotonen Prozess angeregt werden kann. Basierend auf dieser Annahme wurde die Obergrenze für die Ionisationsenergie IP = 52140 cm-1 = 6.5 eV abgeschätzt. In den Messungen wurden Verschiebungen in den Zeitverteilungsspektren zwischen den mono-atomaren Ionen Fm+ und Cf+ und dem Molekül-Ion UO+ festgestellt und auf Driftzeitunterschiede im elektrischen Feld der gasgefüllten optischen Zelle zurückgeführt. Unter einfachen Modellannahmen wurde daraus auf die relativen Unterschiede Delta_r(Fm+,Cf+)/r(Cf+) ˜ -0.2 % und Delta_r(UO+,Cf+)/r(Cf+) ˜ 20 % in den Ionenradien geschlossen. Über die Bestimmung der Abnahme der Fm-a Aktivität des Filamentes auf der einen Seite und die Messung der Resonanzzählrate auf der anderen Seite, wurde die Nachweiseffizienz der Apparatur zu 4.5(3) x 10^-4 bestimmt. Die Nachweisapparatur wurde mit dem Ziel weiterentwickelt, Laserspektroskopie am Isotop 251Fm durchzuführen, das über die Reaktion 249Cf(a,2n)251Fm direkt in der optischen Zelle erzeugt werden soll. Das Verfahren wurde am chemischen Homolog Erbium getestet. Dabei wurde das Isotop 163Er über die Reaktion 161Dy(a,2n)163Er erzeugt und nach Resonanzionisation nachgewiesen. Die Nachweiseffizienz der Methode wurde zu 1 x 10^-4 bestimmt.

Radiochemical aspects of production and processing of radiometals for preparation of metalloradiopharmaceuticals

Radiometals play an important role in nuclear medicine as involved in diagnostic or therapeutic agents. In the present work the radiochemical aspects of production and processing of very promising radiometals of the third group of the periodic table, namely radiogallium and radiolanthanides are investigated. The 68Ge/68Ga generator (68Ge, T½ = 270.8 d) provides a cyclotron-independent source of positron-emitting 68Ga (T½ = 68 min), which can be used for coordinative labelling. However, for labelling of biomolecules via bifunctional chelators, particularly if legal aspects of production of radiopharmaceuticals are considered, 68Ga(III) as eluted initially needs to be pre-concentrated and purified. The first experimental chapter describes a system for simple and efficient handling of the 68Ge/68Ga generator eluates with a cation-exchange micro-chromatography column as the main component. Chemical purification and volume concentration of 68Ga(III) are carried out in hydrochloric acid – acetone media. Finally, generator produced 68Ga(III) is obtained with an excellent radiochemical and chemical purity in a minimised volume in a form applicable directly for the synthesis of 68Ga-labelled radiopharmaceuticals. For labelling with 68Ga(III), somatostatin analogue DOTA-octreotides (DOTATOC, DOTANOC) are used. 68Ga-DOTATOC and 68Ga-DOTANOC were successfully used to diagnose human somatostatin receptor-expressing tumours with PET/CT. Additionally, the proposed method was adapted for purification and medical utilisation of the cyclotron produced SPECT gallium radionuclide 67Ga(III). Second experimental chapter discusses a diagnostic radiolanthanide 140Nd, produced by irradiation of macro amounts of natural CeO2 and Pr2O3 in natCe(3He,xn)140Nd and 141Pr(p,2n)140Nd nuclear reactions, respectively. With this produced and processed 140Nd an efficient 140Nd/140Pr radionuclide generator system has been developed and evaluated. The principle of radiochemical separation of the mother and daughter radiolanthanides is based on physical-chemical transitions (hot-atom effects) of 140Pr following the electron capture process of 140Nd. The mother radionuclide 140Nd(III) is quantitatively absorbed on a solid phase matrix in the chemical form of 140Nd-DOTA-conjugated complexes, while daughter nuclide 140Pr is generated in an ionic species. With a very high elution yield and satisfactory chemical and radiolytical stability the system could able to provide the short-lived positron-emitting radiolanthanide 140Pr for PET investigations. In the third experimental chapter, analogously to physical-chemical transitions after the radioactive decay of 140Nd in 140Pr-DOTA, the rapture of the chemical bond between a radiolanthanide and the DOTA ligand, after the thermal neutron capture reaction (Szilard-Chalmers effect) was evaluated for production of the relevant radiolanthanides with high specific activity at TRIGA II Mainz nuclear reactor. The physical-chemical model was developed and first quantitative data are presented. As an example, 166Ho could be produced with a specific activity higher than its limiting value for TRIGA II Mainz, namely about 2 GBq/mg versus 0.9 GBq/mg. While free 166Ho(III) is produced in situ, it is not forming a 166Ho-DOTA complex and therefore can be separated from the inactive 165Ho-DOTA material. The analysis of the experimental data shows that radionuclides with half-life T½ < 64 h can be produced on TRIGA II Mainz nuclear reactor, with specific activity higher than any available at irradiation of simple targets e.g. oxides.

Validation of the CFD code NEPTUNE for a full scale simulator for decay heat removal systems with in-pool heat exchangers

In the present work, a multi physics simulation of an innovative safety system for light water nuclear reactor is performed, with the aim to increase the reliability of its main decay heat removal system. The system studied, denoted by the acronym PERSEO (in Pool Energy Removal System for Emergency Operation) is able to remove the decay power from the primary side of the light water nuclear reactor through a heat suppression pool. The experimental facility, located at SIET laboratories (PIACENZA), is an evolution of the Thermal Valve concept where the triggering valve is installed liquid side, on a line connecting two pools at the bottom. During the normal operation, the valve is closed, while in emergency conditions it opens, the heat exchanger is flooded with consequent heat transfer from the primary side to the pool side. In order to verify the correct system behavior during long term accidental transient, two main experimental PERSEO tests are analyzed. For this purpose, a coupling between the mono dimensional system code CATHARE, which reproduces the system scale behavior, with a three-dimensional CFD code NEPTUNE CFD, allowing a full investigation of the pools and the injector, is implemented. The coupling between the two codes is realized through the boundary conditions. In a first analysis, the facility is simulated by the system code CATHARE V2.5 to validate the results with the experimental data. The comparison of the numerical results obtained shows a different void distribution during the boiling conditions inside the heat suppression pool for the two cases of single nodalization and three volume nodalization scheme of the pool. Finaly, to improve the investigation capability of the void distribution inside the pool and the temperature stratification phenomena below the injector, a two and three dimensional CFD models with a simplified geometry of the system are adopted.

Nicht geträgerte Radioarsenisotope: ihre Herstellung, Abtrennung und Markierung von Proteinen

Das Element Arsen besitzt eine Reihe von Isotopen, die in nahezu trägerfreier Form (nca) produziert werden können und deshalb in der Radiopharmazie für die Diagnose oder Endoradiotherapie Verwendung finden können. Bei der Positronenemissionstomographie (PET) gibt es eine gewisse Lücke bei der Versorgung mit langlebigen Positronenemittern, die zur Untersuchung von langsamen physiologischen Prozessen wie z.B. der Biodistribution und Anreicherung von Antikörpern in Tumorgewebe eingesetzt werden können. Die beiden Arsenisotope 72As (T1/2 = 26 h, 88 % beta+) und 74As (T1/2 = 17,8 d, 29 % beta+) vereinen eine lange physikalische Halbwertszeit mit einer hohen Positronenemissionsrate und sind daher geeignete Kandidaten. Da das Verhalten von radioaktivem Arsen und seine Verwendung in der molekularen Bildgebung international relativ wenig bearbeitet sind, wurde die Radiochemie des Arsens von der Isotopenproduktion an Kernreaktor und Zyklotron, über die Entwicklung von Abtrennungsmethoden für Germanium und Arsen, bis hin zur Entwicklung einer soliden Markierungschemie für Antikörper weiterentwickelt. Die in dieser Arbeit bearbeiteten Felder sind: 1. Die Isotopenproduktion der relevanten Arsenisotope (72/74/77As) wurde an Kernreaktor und Zyklotron durch Bestrahlung von GeO2- und Germaniummetalltargets durchgeführt. Pro 6 h Bestrahlung von 100 mg Germanium konnten ca. 2 MBq 77As am TRIGA Reaktor in Mainz hergestellt werden. Am Zyklotron des DKFZ in Heidelberg konnten unter optimierten Bedingungen bei der Bestrahlung von Germaniummetall (EP = 15 Mev, 20 µA, 200 µAh) ca. 4 GBq 72As und ca. 400 MBq 74As produziert werden. 2. Die Entwicklung neuer Abtrennungsmethoden für nca 72/74/77As von makroskopischen Mengen Germanium wurde vorangetrieben. Für die Aufarbeitung von GeO2- und Germaniummetalltargets kamen insgesamt 8 verschiedene Methoden wie Festphasenextraktion, Flüssig-Flüssig-Extraktion, Destillation, Anionenaustauschchromatographie zum Einsatz. Die erzielten Ausbeuten lagen dabei zwischen 31 und 56 %. Es wurden Abtrennungsfaktoren des Germaniums zwischen 1000 und 1•10E6 erreicht. Alle erfolgreichen Abtrennungsmethoden lieferten *As(III) in 500 µl PBS-Puffer bei pH 7. Diese Form des Radioarsens ist für die Markierung von SH-modifizierten Molekülen, wie z.B. Antikörpern geeignet. 3. Die Entwicklung von Methoden zur Bestimmung des Oxidationszustandes von nca *As in organischem, neutralem wässrigen, oder stark sauren Medium mittels Radio-DC und Anionenaustauschchromatographie wurde durchgeführt und führte zu einem besseren Verständnis der Redoxchemie des nca *As. 4. SH-modifizierte Antikörper wurden mit 72/74/77As(III) markiert. Dabei wurden zwei Methoden (Modifizierung mit SATA und TCEP) miteinander verglichen. Während das *As(III) bei Verwendung von TCEP in Ausbeuten > 90 % mit dem Antikörper reagierte, wurde für SATA-modifizierte Antikörper in Abhängigkeit von der verwendeten Abtrennungsmethode eine breite Spanne von 0 % bis > 90 % beobachtet. 5. Es wurden Phantommessungen mit 18F, 72As und 74As am µ-PET-Scanner durchgeführt, um erste Aussagen über die zu erwartende Auflösung der Arsenisotope zu erhalten. Die Auflösung von 74As ist mit 18F vergleichbar, während die von 72As erkennbar schlechter ist.