About 60 percent of the electricity production in Belgium originates from nuclear power plants. Belgium owns 7 nuclear pressurized water reactors, which are located in two sites: 4 reactors in Doel and 3 reactors in Tihange. Together they have a capacity of approximately 5900 MWe. All these reactors use classical uranium oxide fuel assemblies. Two of them (Doel3, Tihange2) have also accepted a limited number of mixed (uranium and plutonium) oxide fuel assemblies. These mixed fuel assemblies came from the reprocessing of spent uranium oxide fuel assemblies in La Hague (France). The reprocessing of spent fuel gives birth to vitrified high-level waste, and to different isotopes of uranium and plutonium, which can be used in the manufacture of mixed oxide fuel assemblies. Each country producing radioactive waste must find a solution to dispose them safely. The internationally accepted solution is to dispose high-level radioactive waste in a deep and stable geological layer. This seems to be the most secure and environment-friendly way to get rid of the high-level radioactive waste. One of the few stable geological layers, which could accept radioactive waste in Belgium, is the Boom clay layer. Another possible layer is the Ypresian clay layer, but it is not the reference option for the moment. The Boom clay layer is quite thin (about 100 m thick) and is not at a large depth (about 240 m below the ground surface) at the proposed disposal site, beneath the SCK CEN Nuclear Research Centre in Mol. A large number of studies have already been performed on the Boom clay layer, and on the possibility of building a high-level radioactive waste repository in this geological medium. Since 1993, the Belgian government has promulgated a moratorium on the reprocessing of spent uranium oxide fuels in La Hague. Since then, spent fuel assemblies are considered as waste, and ONDRAF/NIRAS (the Belgium Agency for Radioactive Waste and Enriched Fissile Materials) has thus to deal with them as waste. This rises a number of questions on how to deal with this new kind of waste. A solution is to directly dispose these spent fuel assemblies in containers in a repository, just like the other high-level radioactive waste. This repository would be build in the Boom clay layer at a depth of about 240 m beneath the SCK CEN. One of the questions raised by this new kind of waste is: "could the direct disposal of the spent nuclear fuel assemblies lead to nuclear criticality risks in the future?". Nuclear criticality is the ability of a system to sustain a nuclear fission chain reaction. This question was not a key issue with vitrified high-level waste because these do not include fissile uranium and plutonium isotopes, which could lead to a criticality event. The spent fuel repository will be designed in order to totally avoid the occurrence of a criticality event at the closure time. But in the future history of the repository, external events could possibly affect this. These events could maybe lead to criticality inside the repository, and this has also to be avoided. This work tries to answer this question, and to determine how to avoid a long-term criticality event inside the repository. The only complete research work answering this question has been performed in the U.S. for the Yucca Mountain repository but this design is fully different from the Belgian one studied here: for example, the waste are not only spent fuel waste, and the geological layer is volcanic tuff. The main achievements of this work are: *A first set of in-package criticality scenarios for different design options for a Belgian spent fuel repository in the Boom clay layer. *A large number of criticality calculations with different parameters (fuel type, fuel burnup, fuel enrichment, distance between the fuel assemblies, distance between the fuel rods, water fraction inside the overpack) for the different design options. *A preliminary study of the effects of the spent fuel assemblies isotopic evolution with time on the multiplication factor. *For the first time, a coupling between the in-package criticality scenarios and the criticality calculations has been performed.

Several alternative strategies are being considered as spent nuclear fuel (SNF) management options. Transuranic nuclides (TRU) are responsible for the SNF long-term radiotoxicity beyond the first 500 years. One of the most viable approaches suggests creating new transmutation fuels containing TRUs for use in thermal and fast nuclear reactors. Irradiation of TRUs results in their transmutation and ultimate incineration by fission. The objective of this thesis is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled Very High Temperature Reactor (VHTR) systems. This effort was focused on the prismatic core configuration. The 3D core models were created for use in calculations with the SCALE 5.1 code system. As part of the research effort, basic nuclear characteristics of TRUs were taken into consideration. The potential variations of PWR spent fuel compositions were modeled with the International Atomic Energy Agency (IAEA) Nuclear Fuel Cycle Simulation System, VISTA. The VHTR configurations with varying TRU compositions were analyzed assuming a single-batch core operation. Their performance was compared to the VHTR cases with low enriched uranium (LEU). The analysis shows that TRUs can be effectively utilized in the VHTR systems. The TRU-fueled VHTRs exhibit favorable performance characteristics.

Spent acid retention in the near-wellbore region causes reduction of relative permeability to gas and eventually curtailed gas production. In low-permeability gas carbonate reservoirs, capillary forces are the key parameters that affect the trapping of spent acid in the formation. Capillarity is a function of surface tension at the gas-liquid interface and contact angle of the fluids in the rock. To weaken capillary forces, surface tension should be low and contact angle should be large. This work provides a comprehensive study on the effect of various common acid additives on carbonate rock wettability, and surface tension and contact angle, as the main parameters that control capillarity. Surface tension and contact angle experiments were conducted using Drop Shape Analysis (DSA) instrument at high temperature and pressure. Core flood experiments were also conducted to study the overall impact of the acid additives on wettability by analyzing irreducible fluid saturation in the rocks before and after spent acid exposure. Spontaneous water imbibition was conducted in each case to check for permanent or long-term wettability change as a result of using these additives. Acid additives such as methanol and corrosion inhibitors reduced both surface tension and contact angle. Iron control agents had no impact on surface tension, however, they decreased contact angle at the lower concentration used. Formic and acetic acids did not affect the surface tension, but they had a reducing impact on the contact angle. According to the core flood experiment results, formic acid decreased irreducible fluid saturation whereas methanol increased irreducible fluid saturation. On the other hand, the fluorochemical surfactant tested changed the rock wettability into more gas wetting. Use of this chemical would help in recovering spent acid. The results of the spontaneous water imbibition tests showed that organic acids and iron control chemicals did not have a permanent impact on wettability of the rocks. However, the wettability change as a result of using fluorochemical surfactant would persist for a long time as this chemical forms a film on the rock surface.

Several alternative strategies are being considered as spent nuclear fuel (SNF) management options. Transuranic nuclides (TRU) are responsible for the SNF long-term radiotoxicity beyond the first 500 years. One of the most viable approaches suggests creating new transmutation fuels containing TRUs for use in thermal and fast nuclear reactors. Irradiation of TRUs results in their transmutation and ultimate incineration by fission. The objective of this thesis is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled Very High Temperature Reactor (VHTR) systems. This effort was focused on the prismatic core configuration. The 3D core models were created for use in calculations with the SCALE 5.1 code system. As part of the research effort, basic nuclear characteristics of TRUs were taken into consideration. The potential variations of PWR spent fuel compositions were modeled with the International Atomic Energy Agency (IAEA) Nuclear Fuel Cycle Simulation System, VISTA. The VHTR configurations with varying TRU compositions were analyzed assuming a single-batch core operation. Their performance was compared to the VHTR cases with low enriched uranium (LEU). The analysis shows that TRUs can be effectively utilized in the VHTR systems. The TRU-fueled VHTRs exhibit favorable performance characteristics.

The development of non-destructive assay (NDA) capabilities to directly measure the fissile content in spent fuel is needed to improve the timely detection of the diversion of significant quantities of fissile material. Currently, the International Atomic Energy Agency (IAEA) does not have effective NDA methods to verify spent fuel and recover continuity of knowledge in the event of a containment and surveillance systems failure. This issue has become increasingly critical with the worldwide expansion of nuclear power, adoption of enhanced safeguards criteria for spent fuel verification, and recent efforts by the IAEA to incorporate an integrated safeguards regime. In order to address these issues, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been developed to improve existing nuclear safeguards and material accountability measurements. The following characteristics of SINRD were analyzed: (1) ability to measure the fissile content in Light Water Reactors (LWR) fuel assemblies and (2) sensitivity and penetrability of SINRD to the removal of fuel pins from an assembly. The Monte Carlo Neutral Particle eXtended (MCNPX) transport code was used to simulate SINRD for different geometries. Experimental measurements were also performed with SINRD and were compared to MCNPX simulations of the experiment to verify the accuracy of the MCNPX model of SINRD. Based on the results from these simulations and measurements, we have concluded that SINRD provides a number of improvements over current IAEA verification methods. These improvements include: 1) SINRD provides absolute measurements of burnup independent of the operator’s declaration. 2) SINRD is sensitive to pin removal over the entire burnup range and can verify the diversion of 6% of fuel pins within 3σ from LWR spent LEU and MOX fuel. 3) SINRD is insensitive to the boron concentration and initial fuel enrichment and can therefore be used at multiple spent fuel storage facilities. 4) The calibration of SINRD at one reactor facility carries over to reactor sites in different countries because it uses the ratio of fission chambers (FCs) that are not facility dependent. 5) SINRD can distinguish fresh and 1-cycle spent MOX fuel from 3- and 4-cycles spent LEU fuel without using reactor burnup codes.

Matrix acidizing is a well stimulation technique used to remove formation damage in the near wellbore region. But it comes with an associated set of challenges such as corrosion of the tubulars and iron precipitation in the formation. To counter these challenges, different chemicals, or additives, are added to the acid solution such as corrosion inhibitors and iron control agents. These additives may change the relative permeability of the spent acid, and formation wettability, and may either hinder or improve spent acid clean-up. Such effects of additives on the spent acid clean-up have not been documented. The aim of this research effort was to document the aforementioned change in the spent acid concentration (by using one additive at a time) before and after gas flowback. This was achieved by acidizing cores and creating wormholes halfway through them, then CT scanning them to observe the spent acid region. Later on, gas was flown through the core opposite to the direction of acid injection for 2 hours, and another CT scan was taken. The difference between the two CT scans was documented. Using a different additive each time, a series of such CT scans was obtained to develop an idea about whether the said additive was beneficial or detrimental to spent acid clean-up. It was found that the corrosion inhibitor FA-CI performed the best in terms of spent acid recovery after gas flowback for both Indiana Limestone and Texas Cream Chalk cores. Moreover, the corrosion inhibitor MI-CI was the worst for Indiana Limestone and the non-emulsifying agent M-NEA the worst for Texas Cream Chalk for spent acid recovery after gas flowback.

Title from PDF of title page (University of Missouri--Columbia, viewed on Feb 24, 2010). The entire thesis text is included in the research.pdf file; the official abstract appears in the short.pdf file; a non-technical public abstract appears in the public.pdf file. Dissertation supervisors: Dr. William H. Miller, Dr. Brady D. Hanson. Vita. Includes bibliographical references.

No description available.

No description available.

Made available in DSpace on 2018-08-01T21:58:48Z (GMT). No. of bitstreams: 1 tese_10827_Dissertação Jenifer Rigo Almeida - FINAL.pdf: 2105933 bytes, checksum: 17fccc5751be81765e75282388ce4b0c (MD5) Previous issue date: 2017-03-27 / Baterias de íon-Lítio (LIBs) exauridas são consideradas resíduos sólidos perigosos devido à presença de metais e compostos orgânicos em sua composição, representando desperdício de recursos naturais não renováveis e de metais valiosos quando descartadas. Este trabalho tem por objetivo fornecer dados quantitativos sobre a composição química do material ativo do cátodo (MAC) de diferentes LIBs exauridas visando monitorar variações com o passar dos anos e auxiliar nos processos de reciclagem do material. Os elementos Al, Co, Cr, Cu, Ga, Li, Mg, Mn, Ni, Ti e Zn foram determinados por espectrometria de emissão óptica com plasma indutivamente acoplado (ICP OES) após lixiviação ácida empregando 2,0 mol.L-1 de ácido cítrico (HCit) e H2O2 (0,25 mol.L-1) como alternativa ambientalmente favorável. As condições otimizadas para adequação do meio às curvas analíticas foram: para Al, Cu: Curva de HCit diluído 10 vezes sem padrão interno (PI); para Co, Li, Mn, Ni: Curva de HCit diluído 500 vezes sem PI; para Ga, Zn: Curva de HCit diluído 10 vezes com Y. O procedimento analítico empregado alcançou limites de detecção de 0,01 mg.L-1 para Al; 0,20 mg.L-1 para Co; 0,006 mg.L-1 para Cr; 0,02 mg.L-1 para Cu; 0,004 mg.L-1 para Ga; 0,02 mg.L-1 para Li; 0,0005 mg.L-1 para Mg; 0,07 mg.L-1 para Mn; 0,70 mg.L-1 para Ni; 0,0005 mg.L-1 para Ti e 0,007 mg.L-1 para Zn. A exatidão do procedimento foi confirmada por testes de adição e recuperação dos analitos obtendo-se valores entre 92-113 %. Os elementos majoritários Co (43-67 % m/m), Li (5,3-6,8 % m/m), Mn (0,8-8,2 % m/m), Ni (0,1-11,7 % m/m) e Al (0,06-3,2 % m/m) e os elementos minoritários Cr (0,0005-0,002 % m/m), Cu (0,01-0,05 % m/m), Mg (0,005-0,02 % m/m), Ti (0,001-0,07 % m/m), Ga (0,0009-0,03 % m/m) e Zn (0,009-0,05 % m/m) demonstraram que a composição do MAC pode variar de acordo com a capacidade e ano de fabricação. As baterias mais antigas foram as que apresentaram maiores teores de Co e Li. As baterias de menor capacidade foram as que continham os maiores teores de Mn e Ni, indicando que o Co foi substituído. O pó do MAC e o resíduo após lixiviação foram caracterizados por difratometria de raios X (DRX) obtendo-se LiCoO2 como composto principal, podendo ser reutilizado.