Quando um país se torna auto-suficiente em uma parte do ciclo nuclear, quanto à produção de combustível que será usado em suas centrais nucleares para a geração de energia, precisa voltar sua atenção para a melhor forma de armazenar este combustível após a sua utilização. A armazenagem do combustível nuclear queimado é uma prática necessária e utilizada nos dias atuais em todo o mundo como temporária, tanto por países que não têm definido o plano de destinação final, isto é, o repositório definitivo, como também por aqueles que já o possuem. Existem dois aspectos principais que envolvem os combustíveis queimados: um referente à armazenagem do combustível nuclear queimado destinado ao reprocessamento e o outro ao que será enviado para deposição final quando o sítio de deposição definitiva estiver definido, corretamente localizado, adequadamente caracterizado quanto aos diversos aspectos técnicos, e licenciado. Este último aspecto pode envolver décadas de estudos por causa das definições técnicas e normativas em um dado país. No Brasil, o interesse está voltado para a armazenagem dos combustíveis queimados que não serão reprocessados. Este trabalho analisa os tipos possíveis de armazenagem, o panorama internacional e a possível proposta para a futura construção de um sítio de armazenagem temporária no país. / When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that won't be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country.

Quando um país se torna auto-suficiente em uma parte do ciclo nuclear, quanto à produção de combustível que será usado em suas centrais nucleares para a geração de energia, precisa voltar sua atenção para a melhor forma de armazenar este combustível após a sua utilização. A armazenagem do combustível nuclear queimado é uma prática necessária e utilizada nos dias atuais em todo o mundo como temporária, tanto por países que não têm definido o plano de destinação final, isto é, o repositório definitivo, como também por aqueles que já o possuem. Existem dois aspectos principais que envolvem os combustíveis queimados: um referente à armazenagem do combustível nuclear queimado destinado ao reprocessamento e o outro ao que será enviado para deposição final quando o sítio de deposição definitiva estiver definido, corretamente localizado, adequadamente caracterizado quanto aos diversos aspectos técnicos, e licenciado. Este último aspecto pode envolver décadas de estudos por causa das definições técnicas e normativas em um dado país. No Brasil, o interesse está voltado para a armazenagem dos combustíveis queimados que não serão reprocessados. Este trabalho analisa os tipos possíveis de armazenagem, o panorama internacional e a possível proposta para a futura construção de um sítio de armazenagem temporária no país. / When a country becomes self-sufficient in part of the nuclear cycle, as production of fuel that will be used in nuclear power plants for energy generation, it is necessary to pay attention for the best method of storing the spent fuel. Temporary storage of spent nuclear fuel is a necessary practice and is applied nowadays all over the world, so much in countries that have not been defined their plan for a definitive repository, as well for those that already put in practice such storage form. There are two main aspects that involve the spent fuels: one regarding the spent nuclear fuel storage intended to reprocessing and the other in which the spent fuel will be sent for final deposition when the definitive place is defined, correctly located, appropriately characterized as to several technical aspects, and licentiate. This last aspect can involve decades of studies because of the technical and normative definitions at a given country. In Brazil, the interest is linked with the storage of spent fuels that won't be reprocessed. This work analyses possible types of storage, the international panorama and a proposal for future construction of a spent nuclear fuel temporary storage place in the country.

<p>Nuclear energy is currently one of the world’s main sources of electricity. Closely connected to the use of nuclear energy are important issues such as the nonproliferation of fissile material that may potentially used in nuclear weapons (safeguards), and the management of the highly radioactive nuclear waste. This thesis addresses both these issues by contributing to the development of new experimental methods for ensuring safe and secure handling of the waste, with focus on methods to be used prior to encapsulation and final storage.</p><p>The methods rely on high resolution gamma ray spectroscopy (HRGS), involving the measurement and analysis of emitted gamma radiation from the fission products ¹³⁷Cs, ¹³⁴Cs and ¹⁵⁴Eu. This technique is nondestructive, making it relatively nonintrusive with respect to the normal operation of the nuclear facilities.</p><p>For the safeguards issue, it is important to experimentally verify the presence and identity of nuclear fuel assemblies and also that the fuel has experienced normal, civilian reactor operation. It has been shown in this thesis that the HRGS method may be used for verifying operator declared fuel parameters such as burnup, cooling time and irradiation history. In the experimental part of the work, the burnup and the cooling time has been determined with an accuracy of 1.6% and 1.5%, respectively (1 σ).</p><p>A technique has also been demonstrated, utilizing the ratio ¹³⁴Cs/¹⁵⁴Eu, with which it is possible to determine whether a fuel assembly is of MOX or LEU type. This is of interest for safeguards as well as for the safe operation of a final storage facility.</p><p>As an improvement to the HRGS technique, measuring a part of the fuel assembly length in order to reduce measurement time has been suggested and investigated. A theoretical case for partial defect verification has also been studied as an extension of the HRGS technique. </p><p>Finally, HRGS has been used for determining the decay heat in spent nuclear fuel assemblies, which is of importance for the safe operation of a final storage facility. This application is based on the radiation from ¹³⁷Cs, and the accuracy demonstrated was within 3% (1 σ).</p>

Nuclear energy is currently one of the world’s main sources of electricity. Closely connected to the use of nuclear energy are important issues such as the nonproliferation of fissile material that may potentially used in nuclear weapons (safeguards), and the management of the highly radioactive nuclear waste. This thesis addresses both these issues by contributing to the development of new experimental methods for ensuring safe and secure handling of the waste, with focus on methods to be used prior to encapsulation and final storage. The methods rely on high resolution gamma ray spectroscopy (HRGS), involving the measurement and analysis of emitted gamma radiation from the fission products 137Cs, 134Cs and 154Eu. This technique is nondestructive, making it relatively nonintrusive with respect to the normal operation of the nuclear facilities. For the safeguards issue, it is important to experimentally verify the presence and identity of nuclear fuel assemblies and also that the fuel has experienced normal, civilian reactor operation. It has been shown in this thesis that the HRGS method may be used for verifying operator declared fuel parameters such as burnup, cooling time and irradiation history. In the experimental part of the work, the burnup and the cooling time has been determined with an accuracy of 1.6% and 1.5%, respectively (1 σ). A technique has also been demonstrated, utilizing the ratio 134Cs/154Eu, with which it is possible to determine whether a fuel assembly is of MOX or LEU type. This is of interest for safeguards as well as for the safe operation of a final storage facility. As an improvement to the HRGS technique, measuring a part of the fuel assembly length in order to reduce measurement time has been suggested and investigated. A theoretical case for partial defect verification has also been studied as an extension of the HRGS technique. Finally, HRGS has been used for determining the decay heat in spent nuclear fuel assemblies, which is of importance for the safe operation of a final storage facility. This application is based on the radiation from 137Cs, and the accuracy demonstrated was within 3% (1 σ).

The topic of this doctoral thesis is the potential near field immobilization of the radionuclide 79Se after intrusion of groundwater into a spent nuclear fuel canister in a repository. 79Se is a non naturally occurring long lived selenium isotope formed as a result of fission in nuclear fuel. Given the long half life (~3 x 105 y) and that the oxyanions of selenium are expected to be highly mobile and potentially difficult toimmobilize the isotope is of interest for the long term safety assessment of high level waste repositories. In this work the near field has been limited to the study of processes at or near the UO2 surface of (simulated) spent nuclear fuel and to processes occurring at or near the surface of iron (canister material) corroding under anoxic conditions. Selenite (HSeO32-) was found to adsorb onto palladium (simulated noble metal inclusion in spent nuclear fuel). Under hydrogen atmosphere selenite was reduced to elemental selenium with a rate constant of ~2 x 10-9 m s-1 (with respect to the Pd surface, 24 bar H2) forming colloidal particles. The rate constant of selenite reduction was increased by about two orders of magnitude to ~2.5 x 10-7 m s-1 (with respect to the Pd surface, 10 bar H2) for a UO2 surface doped with Pd particles, indicating that UO2 is an efficient co-catalyst to Pd. Selenate (SeO42-) was neither adsorbed nor reduced in the presence of Pd, UO2 and hydrogen. In the iron corrosion studies selenate was found to become reduced to predominantly elemental Se in the presence of a pristine iron surface. Iron covered by a corrosion layer of magnetite did however appear inert with respect to selenate whereas selenite was reduced. The reduction of dissolved uranyl into UO2 by the corroding iron surfaces was found to significantly increase the removal rate of selenite as well as selenate. The uranyl was found to transiently transform the outer iron oxide layers on the iron, forming a reactive mixed Fe(II)/Fe(III) oxyhydroxide (Green rust). Exchanging the solution and increasing the carbonate content (from 2 mM to 20 mM NaHCO3) only resulted in a minor, transient remobilization of uranium. Addition of H2O2 did however result in a significant release of uranium as well as selenium from the iron oxide surfaces. An irradiation experiment was also performed confirming the one electron reduction barrier of selenate as an important factor in systems where selenate reduction would be thermodynamically favorable. / QC 20101208

Spent nuclear fuel (SNF) resultant from the generation of nuclear power is a chemically and radiologically diverse system which is advantageous to chemically process prior to geologic disposal. Hydrometallurgy is the primary technology for chemical processing for light water reactor spent fuels, where spent fuel is dissolved in an acid for liquid based separations. The primary means for recovery of desired metals from the SNF solution is liquid-liquid extraction which is based on distribution (partitioning) of the metal ions between two immiscible phases based on thermodynamic favorability. One of the means of increasing this favorability is by designing extractant molecules to be either "harder" or "softer" bases, which will more preferentially extract harder or softer metal ions respectively. This technique is used in designing extractant molecules for targeted extraction as actinides are slightly softer than lanthanides, and precious metals produced in significant quantities from the fission process are especially soft metals. The work performed in this thesis involved the synthesis of a novel soft electron donor organic extractant molecule for testing of targeted soft metal extraction. The molecule synthesized was bis-dibutanethiolthiophosphinato-methane, or S6, a bidentate neutral extractant molecule with significant thiolysis for a softer electron environment. The synthesis technique was refined and the molecule composition and structure was confirmed by ¹H NMR, ³¹P NMR, and elemental analysis. Two metal groups, f-elements (actinides and lanthanides) and soft transition metals were tested for their extractability from nitric acid solutions into an S6 solution in n-dodecane. Aqueous solutions of nitric acid and n-dodecane as an organic diluent are typical liquid-liquid extraction conditions in spent nuclear fuel reprocessing. As extraction experiments were performed with radiotracers, for the soft metal extraction experiment, a mixture of the selected metals was neutron-activated in the OSU TRIGA reactor, as was europium to create a lanthanide radiotracer. Actinides and lanthanides were not seen to effectively extract into the organic or form a precipitate at all, making their partitioning with this extractant seemingly ineffective. Through gamma spectroscopy of an irradiated metal solution post-extraction, it is seen that only silver and palladium preferentially complex in the mixed metal samples into an insoluble organic ligand, dropping out of solution. This effect was more pronounced at higher acid concentrations, but silver was seen to slightly extract to the organic phase at all acid concentrations as well. This testing has shown that the S6 extractant can be used to recover silver and palladium from a mixed metal aqueous solution, such as one resultant from advanced spent nuclear fuel reprocessing operations. This result shows promise for future development of sulfur based organophosphate ligands for targeted extraction of precious metals from solutions. / Graduation date: 2013

<p>The kinetics for radiation induced dissolution of spent nuclear fuel is a key issue in the safety assessment of a future deep repository. Spent nuclear fuel mainly consists of UO₂ and therefore the release of radionuclides (fission products and actinides) is assumed to be governed by the oxidation and subsequent dissolution of the UO₂ matrix. The process is influenced by the dose rate in the surrounding groundwater (a function of fuel age and burn up) and on the groundwater composition. In this licentiate thesis the effects of HCO₃- (a strong complexing agent for UO2²⁺) and ionic strength on the kinetics of UO₂ oxidation and dissolution of oxidized UO₂ have been studied experimentally.</p><p>The experiments were performed using aqueous UO₂particle suspensions where the oxidant concentration was monitored as a function of reaction time. These reaction systems frequently display first order kinetics. Second order rate constants were obtained by varying the solid UO₂surface area to solution volume ratio and plotting the resulting pseudo first order rate constants against the surface area to solution volume ratio. The oxidants used were H₂O₂(the most important oxidant under deep repository conditions), MnO₄- and IrCl₆^2-. The kinetics was studied as a function of HCO₃- concentration and ionic strength (using NaCl and Na₂SO₄as electrolytes).</p><p>The rate constant for the reaction between H₂O₂ and UO₂ was found to increase linearly with the HCO3- concentration in the range 0-1 mM. Above 1 mM the rate constant is independent of the HCO3- concentration. The HCO₃- concentration independent rate constant is interpreted as being the true rate constant for oxidation of UO₂ by H₂O₂ [(4.4 ± 0.3) x 10-6 m min-1] while the HCO3- concentration dependent rate constant is used to estimate the rate constant for HCO₃- facilitated dissolution of UO₂2+ (oxidized UO₂) [(8.8 ± 0.5) x 10-3 m min-1]. From experiments performed in suspensions free from HCO₃- the rate constant for dissolution of UO₂2+ was also determined [(7 ± 1) x 10^-8mol m^-2 s^-1]. These rate constants are of significant importance for simulation of spent nuclear fuel dissolution.</p><p>The rate constant for the oxidation of UO₂ by H₂O₂ (the HCO₃- concentration independent rate constant) was found to be independent of ionic strength. However, the rate constant for dissolution of oxidized UO₂ displayed ionic strength dependence, namely it increases with increasing ionic strength.</p><p>The HCO₃- concentration and ionic strength dependence for the anionic oxidants is more complex since also the electron transfer process is expected to be ionic strength dependent. Furthermore, the kinetics for the anionic oxidants is more pH sensitive. For both MnO₄- and IrCl₆2- the rate constant for the reaction with UO₂was found to be diffusion controlled at higher HCO3- concentrations (~0.2 M). Both oxidants also displayed ionic strength dependence even though the HCO₃- independent reaction could not be studied exclusively.</p><p>Based on changes in reaction order from first to zeroth order kinetics (which occurs when the UO₂ surface is completely oxidized) in HCO₃- deficient systems the oxidation site density of the UO₂ powder was determined. H₂O₂and IrCl₆2- were used in these experiments giving similar results [(2.1 ± 0.1) x 10-4 and (2.7 ± 0.5) x 10^-4 mol m^-2, respectively].</p>

<p>Denna licensiatavhandling behandlar oxidativ upplösning av UO2. Upplösning av UO2 studeras huvudsakligen då UO2-matrisen hos använt kärnbränsle förväntas fungera som en barriär mot frigörande av radionuklider i ett framtida djupförvar. Lösligheten av U(IV) är mycket låg under i djupförvaret rådande förhållanden emedan U(VI) har betydligt högre löslighet. Oxidation av UO2-matrisen kommer därför att påverka dess löslighet och därmed dess funktion som barriär. I denna avhandling studeras den relativa effektiviteten av en- och två-elektronoxidanter för upplösning av UO2. Vid låga oxidantkoncentrationer är utbytet för upplösningen för en-elektronoxidanter signifikant lägre än för två-elektronoxidanter. För en-elektronoxidanter ökar dock utbytet med ökande oxidanthalt, vilket kan förklaras av den ökade sannolikheten för två konsekutiva en-elektronoxidationer av samma reaktionssite och den ökade möjligheten till disproportionering.</p><p>Radikaler och molekylära radiolysprodukters relativa inverkan på oxidativ upplösning av UO2 studeras också i denna avhandling genom mätning av mängden upplöst U(VI) i γ-bestrålade system som dominerades av olika oxidanter. Dessa studier visade att upplösningshastigheten av UO2 kan uppskattas från oxidantkoncentrationer framtagna genom simuleringar av radiolys i motsvarande homogena system och hastighetskonstanterna för ytreaktionerna. Simuleringarna visar att de molekylära oxidanterna kommer vara de viktigaste oxidanterna i alla system i denna studie vid långa bestrålningstider (>10 timmar). Vid liknande simuleringar av α-bestrålade system fanns att vid förhållanden relevanta för ett djupförvar för använt kärnbränsle, är det endast de molekylära oxidanterna (i huvudsak H2O2) som är av betydelse för upplösningen av bränslematrisen.</p><p>Då använt kärnbränsle innehåller en mängd radionuklider som utsätter UO2-matrisen för kontinuerlig bestrålning, är det av vikt att undersöka hur bestrålning påverkar reaktiviteten av UO2. Bestrålningseffekten på reaktionen mellan UO2 och MnO4- studerades. Dessa försök visade att bestrålning av UO2 vid doser >40 kGy leder till att reaktiviteten ökar upp till 1.3 gånger reaktiviteten av obestrålad UO2. Den ökade reaktiviteten kvarstår efter bestrålningen och effekten kan därför möjligen tillskrivas permanenta förändringar i materialet. Vid uppskattning av reaktiviteten hos använt kärnbränsle måste hänsyn tas till denna effekt då bränslet redan efter ett par dagar i reaktor blivit utsatt för doser >40 kGy.</p><p>Det har tidigare föreslagits att hastigheten för en heterogen västka/fast-fas reaktion är beroende av partikelstorleken hos det fasta materialet, vilket har studerats för UO2-partiklar i denna avhandling. Experimentellt bestämda kinetiska parametrar jämförs med de föreslagna ekvationerna för fyra storleksfraktioner av UO2-pulver och en UO2-pellet. Studien visade partikelstorleksberoendet av andra ordningens hastighetskonstant och aktiveringsenergin för oxidation av UO2 med MnO4- beskrivs relativt väl av de föreslagna ekvationerna.</p> / <p>The general subject of this thesis is oxidative dissolution of UO₂. The dissolution of UO₂ is mainly investigated because of the importance of the UO₂ matrix of spent nuclear fuel as a barrier against radionuclide release in a future deep repository. U(IV) is extremely insoluble under the reducing conditions prevalent in a deep repository, whereas U(VI) is more soluble. Hence, oxidation of the UO₂-matrix will affect its solubility and thereby its function as a barrier. In this thesis the relative efficiency of one- and two electron oxidants in dissolving UO₂is studied. The oxidative dissolution yield of UO₂was found to differ between one- and two-electron oxidants. At low oxidant concentrations the dissolution yields for one-electron oxidants are significantly lower than for two-electron oxidants. However, the dissolution yield for one-electron oxidants increases with increasing oxidant concentration, which could be rationalized by the increased probability for two consecutive one-electron oxidations at the same site and the increased possibility for disproportionation.</p><p>Furthermore, the relative impact of radical and molecular radiolysis products on oxidative dissolution of UO₂is investigated. Experiments were performed where the amount of dissolved U(VI) was measured in γ-irradiated systems dominated by different oxidants. We have found that the UO₂dissolution rate in systems exposed to γ-irradiation can be estimated from oxidant concentrations derived from simulations of radiolysis in the corresponding homogeneous systems and rate constants for the surface reactions. These simulations show that for all systems studied in this work, the molecular oxidants will be the most important oxidants for long irradiation times (>10 hours). Similar simulations of α-irradiated systems show that in systems relevant for a deep repository for spent nuclear fuel, only the molecular oxidants (mainly H₂O₂) are of importance for the dissolution of the fuel matrix.</p><p>The effect on UO₂ reactivity by irradiation of the material is of importance when predicting the spent fuel dissolution rate since the fuel, due to its content of radionuclides, is exposed to continuous self-irradiation. The effect of irradiation on the reaction between solid UO₂and MnO₄^- in aqueous solutions was studied. It was found that irradiation of UO2 at doses >40 kGy increases the reactivity of the material up to ~1.3 times the reactivity of unirradiated UO₂. The increased reactivity remains after the irradiation and can possibly be attributed to permanent changes in the material. This issue must be taken into account when predicting the reactivity of spent nuclear fuel since the fuel is exposed to doses >40 kGy after only a few days in the reactor.</p><p>It has earlier been suggested that the rate of a heterogeneous liquid-solid reaction depends on the size of the solid particles. This was investigated for UO₂particles in this thesis. Experimental kinetic parameters are compared to the previously proposed equations for UO₂ powder of four size fractions and a UO₂ pellet. We have found that the particle size dependence of the second order rate constant and activation energy for oxidation of UO₂ by MnO₄^- is described quite well by the proposed equations.</p>

The ability to quickly and accurately quantify the plutonium (Pu) content in pressurized water reactor (PWR) spent nuclear fuel (SNF) is critical for nuclear forensics purposes. One non-destructive assay (NDA) technique being investigated to detect bulk Pu in SNF is measuring the self-induced x-ray fluorescence (XRF). Previous XRF measurements of Three Mile Island (TMI) PWR SNF taken in July 2008 and January 2009 at Oak Ridge National Laboratory (ORNL) successfully illustrated the ability to detect the 103.7 keV x ray from Pu using a planar high-purity germanium (HPGe) detector. This allows for a direct measurement of Pu in SNF. Additional gamma ray and XRF measurements were performed on TMI SNF at ORNL in October 2011 to measure the signal-to-noise ratio for the 103.7 keV peak. Previous work had shown that the Pu/U peak ratio was directly proportional to the Pu/U content and increased linearly with burnup. However, the underlying Compton background significantly reduced the signal-to-noise ratio for the x-ray peaks of interest thereby requiring a prolonged count time. Comprehensive SNF simulations by Stafford et al showed the contributions to the Compton continuum were due to high-energy gamma rays scattering in the fuel, shipping tube, cladding, collimator and detector1. The background radiation was primarily due to the incoherent scattering of the 137Cs 661.7 keV gamma. In this work methods to reduce the Compton background and thereby increase the signal-to-noise ratio were investigated. To reduce the debilitating effects of the Compton background, a crystal x-ray spectrometer system was designed. This wavelength-dispersive spectroscopy technique isolated the Pu and U x rays according to Bragg's law by x-ray diffraction through a crystal structure. The higher energy background radiation was blocked from reaching the detector using a customized collimator and shielding system. A flat quartz-crystal x-ray spectrometer system was designed specifically to fit the constraints and requirements of detecting XRF from SNF. Simulations were performed to design and optimize the collimator design and to quantify the improved signal-to-noise ratio of the Pu and U x-ray peaks. The proposed crystal spectrometer system successfully diffracted the photon energies of interest while blocking the high-energy radiation from reaching the detector and contributing to background counts. The spectrometer system provided a higher signal-to-noise ratio and lower percent error for the XRF peaks of interest from Pu and U. Using the flat quartz-crystal x-ray spectrometer and customized collimation system, the Monte Carlo N-Particle (MCNP) simulations showed the 103.7 keV Pu x-ray peak signal-to-noise ratio improved by a factor of 13 and decreased the percent error by a factor of 3.3.

The aim of diploma thesis is to analyse the overall technical base and to monitor logistic processes which are from the very beginning associated with radioactive waste and spent nuclear fuel. The main emphasis is put on spent nuclear fuel from nuclear power plants. The current situation in this area is illustrated by an example of the Czech Republic and Sweden. The comparison of Czech and Swedish system leads to demonstration of strengths and weaknesses of the used methods.