A thesis submitted to the Faculty of Science, University of the Witwatersrand, Johannesburg in fulfilment of the requirements for the degree of Doctor of Philosophy. Johannesburg, 2015 / Spent nuclear fuel storage is gradually becoming a nightmare for nuclear reactors which were commissioned in the 1980s. This leaves the nuclear facility management with the dilemma of having to choose between pursuing the cask storage option to relieve the demand pressure on the spent fuel pool, or to opt for the more radical but unpopular option of shutting down the reactor compromising the energy supply, and South Africa is no exception. In a bid to minimise the risk of reactor shut down, the Nuclear Analysis Section (NAS) of Eskom launched the present study of investigating the design requirements of spent fuel casks suitable for the storage and transportation of spent fuel assemblies that have an initial enrichment of up to 5 wt% and much higher burnup of between 50 and 60 GWD/MTU. The aim of the present study is to investigate the suitability of the existing casks for use in 5 wt% enriched fuel, given that they are licensed for a maximum enrichment of 3.5 wt%. As a result of the huge number of casks required, there is potentially a risk of shortage of cask storage space and, therefore, it was prudent that the study also investigates the most optimum storage array that will maximise the storage space, while keeping the effective neutron multiplication factor (keff) below the internationally recommended value of 0.95 [IAEA, 2014]. As such, it is also necessary to identify parameters which have the greatest effect on the neutron multiplication factor. These include determining the effect of changes in moderator and fuel temperature on the neutron multiplication factor and also what the effect of an increase in the concentration in 10B of the boral plate will have on the neutron multiplication factor. / M T 2016

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2012. / Cataloged from PDF version of thesis. / Includes bibliographical references (p. 100-101). / The MIT Research Reactor (MITR-II) is currently undergoing analysis for the planned conversion from high enriched uranium (HEU) to low enriched uranium (LEU), as part of a global effort to minimize the availability of weapons-grade uranium. In support of efficient fuel management analysis with the new LEU fuel, a core design optimization tool has been developed. Using a coarse model, the tool can quickly consider the large range of refueling options available, and identify a solution which minimizes power peaking with the least fuel shuffling possible. The selected scheme can then be examined in greater detail with a more robust simulation tool. The unique geometry of the MITR core makes it difficult to develop a model that both runs very quickly and provides detailed power distribution information. Therefore, a correlation-based approach has been employed. Relationships between burnup, critical control blade position, core Um mass, and power distribution are used to predict fuel element U²³⁵ depletion, critical control blade motion, and power peaking. The tool applies the correlations to identify an optimal loading pattern, defined as the core which has the lowest maximum radial peaking factor in the set of valid solutions with the minimum number of fuel shuffling actions. The correlations that are utilized by the optimization tool were developed using data from simulations with MCODE-FM, a fuel management wrapper for the MCNP-ORIGEN linkage code MCODE. The correlations have been verified with results from additional MCODE-FM runs, and the code logic has been verified with the core loading solutions for a variety of input parameters. The verification found that the code is able to predict radial peaking, core mass, and general control blade motion with sufficient accuracy to develop a good refueling scheme. The tool provides the output solution in an interactive format, which allows the user to quickly examine small perturbations on the identified loading pattern. In addition to the optimization tool development, loading patterns for the mixed HEU-LEU fuel transition cores have been evaluated. This analysis identified general behavioral trends of the mixed-fuel cores, which serve as an initial basis for future transition core analysis. / by Heather M. Connaway. / S.M.

Tese de mestrado integrado. Engenharia Química. Faculdade de Engenharia. Universidade do Porto. 2010

The Classical Model describing the effects of temperature on enzyme activity consists of two processes: the catalytic reaction defined by ΔG cat and irreversible inactivation defined by ΔG inact, this model however, does not account for the observed temperature- dependant behaviour of enzymes. The recent development of the Equilibrium Model is governed not only by ΔG cat and ΔG inact but also by two new intrinsic parameters ΔHeq and Teq, which describe the enthalpy and the temperature of the midpoint, respectively, of a active and reversibly inactive enzyme transition. Teq is central to the physiological adaptation of an enzyme to its environmental temperature and links the molecular, physiological and environmental aspects of life to temperature in a way that has not been previously possible. The Equilibrium Model is therefore a more complete and accurate description of the effects of temperature on enzymes, it has provided new tools for describing and investigating enzyme thermal adaptation and possibly new biotechnological tools. The effects of the incorporating in the new Model of the parameters Teq and ΔH eq yield major differences from the Classical Model, with simulated data calculated according to the Equilibrium Model fitting well to experimental data and showing an initial rate temperature optimum that is independent of assay duration. Simulated data simulated according to the Classical Model can not be fitted to experimental data. All enzymes so far studied (gt30) display behaviour predicted by the Equilibrium Model. The research described here has set out to: experimentally test observations made by Eisenthal et al., on the basis of enzyme reactor data simulated according to the Equilibrium Model; to test the Equilibrium Model using an unusual (rapidly renaturable) enzyme, RNAase; and to test the proposed molecular basis of the Equilibrium Model by examining the effect of a change at the enzymes active site. The experimental results gathered here on the effect of time and temperature on enzyme reactor output confirm the predictions made by Eisenthal et al. (2006) and indicate that the Equilibrium Model can be a useful aid in predicting reactor performance. The Equilibrium Model depends upon the acquisition of data on the variation of the Vmax of an enzyme with time and temperature, and the non-ideal behaviour of RNase A made it impossible to collect such data for this enzyme, as a result the Equilibrium Model could not be applied. The disulfide bond within the active site cleft of A.k 1 protease was cleaved as a probe of the mechanism of the Equilibrium Model, which is proposed to arise from molecular changes at the enzymes active site. Support for the proposed mechanism was gained through the comparison of experimentally determined temperature dependence of the native and reduced forms of the enzyme and application of this data to the Equilibrium Model.

<p>Generation IV (Gen IV) is a developing new generation of nuclear power reactors which is foreseen to bring about a safer and more sustainable production of nuclear power. A Swedish research program called GENIUS aims at developing the Gen IV technology, with emphasis on lead-cooled fast reactors. The present work is part of the GENIUS project, and deals with safeguards aspects for an envisioned future 100 MW Gen IV demonstration facility including storage and reprocessing plant. Also, the safeguards licensing aspects for the facilities have been investigated and results thereof are presented.</p><p>As a basis for the study, the changed usage and handling of nuclear fuel, as compared to that of today, have been examined in order to determine how today's safeguards measures can be modified and extended to meet the needs of the demonstration facility. Safeguards approaches have been considered for within and between each unit at the demonstration facility, with the main focus on system aspects rather than proposing safeguards instrumentation on a detailed level.</p><p>The proposed safeguards approach include the implementation of well-tried measures that are used at currently existing nuclear facilities as well as suggestions for new procedures. The former include, among others, regular inventory verifications, containment and surveillance measures as well as non-destructive and destructive measurements of nuclear materials. The traditional approaches may be improved and supplemented by modern techniques and approaches such as nuclear forensics, safeguards-by-design and improved on-line monitoring of streams of nuclear material. The safeguards approach for the demonstration facility should be outlined early in the licensing process, such that the facility units can be designed in a way that allows for implementation of adequate safeguards measures with minimal intrusion on the regular activities.</p><p>For operating a nuclear facility in Sweden, two separate permits are required. A license application for a new facility shall be handed both to the Swedish Radiation Safety Authority and to the environmental court, which in parallel prepare for decisions according to the Nuclear Activities Act and the Environmental Code, respectively. In terms of the Swedish legislation, there are no fundamental differences between Gen IV facilities and currently existing plants. However, comprehensive investigations and evaluations would be required in order to license new Gen IV facilities.</p>

Small graphite-moderated and gas-cooled reactors have been around since the beginning of the atomic age. Though their existence in the past has been associated with nuclear weapons programs, they are capable of being used in civilian power programs. The simpler design constraints associated with this type of reactor would make them ideal for developing nations to bolster their electricity generation and help promote a greater standard of living in those nations. However, the same benefits that make this type of reactor desirable also make it suspicious to the international community as a possible means to shorten that state?s nuclear latency. If a safeguards approach could be developed for a fuel cycle featuring one of these reactors, it would ease the tension surrounding their existence and possibly lead to an increased latency through engineered barriers. The development of this safeguards approach follows a six step procedure. First, the fuel cycle was analyzed for the types of facilities found in it and how nuclear material flows between facilities. The goals of the safeguards system were established next, using the normal IAEA standards for the non-detection and false alarm probabilities. The 5 MWe Reactor was modeled for both plutonium production and maximum power capacity. Each facility was analyzed for material throughput and the processes that occur in each facility were researched. Through those processes, diversion pathways were developed to test the proposed safeguards system. Finally, each facility was divided into material balance areas and a traditional nuclear material accountancy system was set up to meet the established safeguards goals for the facility. The DPRK weapons program is a great example of the type of fuel cycle that is the problem. The three major facilities in the fuel cycle, the Fuel Fabrication Facility, the 5 MWe Reactor, and the Radiochemical Laboratory, can achieve the two goals of safeguards using traditional methods. Each facility can be adequately safeguarded using methods and practices that are relatively inexpensive and can obtain material balance periods close to the timeliness limits set forth by the IAEA. The Fuel Fabrication Facility can be safeguarded at both its current needed capacity and its full design capacity using inexpensive measurements. The material balance period needed for both capacities are reasonable. For the 5 MWe reactor, plutonium production is simulated to be 6.7 kg per year and is on the high side of estimates. The Radiochemical Laboratory can also be safeguarded at its current capacity. In fact, the timeliness goal for the facility dictates what the material balance period must be for the chosen set of detectors which make it very reasonable.

Generation IV (Gen IV) is a developing new generation of nuclear power reactors which is foreseen to bring about a safer and more sustainable production of nuclear power. A Swedish research program called GENIUS aims at developing the Gen IV technology, with emphasis on lead-cooled fast reactors. The present work is part of the GENIUS project, and deals with safeguards aspects for an envisioned future 100 MW Gen IV demonstration facility including storage and reprocessing plant. Also, the safeguards licensing aspects for the facilities have been investigated and results thereof are presented. As a basis for the study, the changed usage and handling of nuclear fuel, as compared to that of today, have been examined in order to determine how today's safeguards measures can be modified and extended to meet the needs of the demonstration facility. Safeguards approaches have been considered for within and between each unit at the demonstration facility, with the main focus on system aspects rather than proposing safeguards instrumentation on a detailed level. The proposed safeguards approach include the implementation of well-tried measures that are used at currently existing nuclear facilities as well as suggestions for new procedures. The former include, among others, regular inventory verifications, containment and surveillance measures as well as non-destructive and destructive measurements of nuclear materials. The traditional approaches may be improved and supplemented by modern techniques and approaches such as nuclear forensics, safeguards-by-design and improved on-line monitoring of streams of nuclear material. The safeguards approach for the demonstration facility should be outlined early in the licensing process, such that the facility units can be designed in a way that allows for implementation of adequate safeguards measures with minimal intrusion on the regular activities. For operating a nuclear facility in Sweden, two separate permits are required. A license application for a new facility shall be handed both to the Swedish Radiation Safety Authority and to the environmental court, which in parallel prepare for decisions according to the Nuclear Activities Act and the Environmental Code, respectively. In terms of the Swedish legislation, there are no fundamental differences between Gen IV facilities and currently existing plants. However, comprehensive investigations and evaluations would be required in order to license new Gen IV facilities.

Sulphide containing streams must be treated before releases to environment due to the toxicity, corrosivity and unpleasant odour of sulphide. Anaerobic chemolithotrophic desulphurization under denitrifying conditions is the preferred process when compared with others like physicochemical processes, photoautotrophic and aerobic chemolithotrophic desulphurizations as the catalysts, high pressure, high temperature, light energy and oxygen are not needed. Another main advantage of this process is that the denitrification can be achieved with desulphurization simultaneously. In this work, the anaerobic chemolithotrophic desulphurization under denitrifying conditions (autotrophic denitrification) and heterotrophic denitrification processes were studied. Desulphurization under denitrifying conditions was studied in continuous stirred tank bioreactors (CSTB), while batch, continuous stirred tank and biofilm reactors were used to investigate the heterotrophic denitrification. The kinetics of desulphurization, autotrophic and heterotrophic denitrifications obtained in different systems and under various conditions were compared.<p> Using three different feed sulphide concentrations in the range 10-20 mM, a linear relationship between sulphide loading rates and sulphide removal rates was observed in continuous stirred tank reactors, regardless of initial sulphide concentration. The highest sulphide removal rate of 1.79 mM h-1 was obtained in CSTB fed with 15 mM sulphide. In these systems cell washout occurred at lower dilution rates as sulphide concentration in the feed was increased from 10 to 20 mM. The ratio of sulphide to nitrate loading rates influenced the composition of the sulphur oxidation end products where higher ratios favored the formation of elemental sulphur and lower ratios promoted the formation of sulphate.<p> In the batch system initial concentration of nitrate (5 to 50 mM) did not have a notable effect on denitrification process. Nitrate was converted to nitrite first and the produced nitrite was then converted to other gaseous end products such as nitrogen. Increases of temperature in the range of 15 to 35ºC increased the bacterial growth rate significantly with the value of apparent activation energy for specific growth rate being 60.6 kJ mol-1. Using the experimental data generated in two continuous bioreactors operated with feeds containing 10 and 30 mM nitrate biokinetic coefficients for heterotrophic denitrification were determined. The values of µm, Ks, ms, YMX/S, kd for initial nitrate concentrations of 10 and 30 mM were 0.087 and 0.082 h-1, 2.01 and 5.27 mM (NO3-), 1.441 and 1.096 mM (NO3-) (g biomass) -1 h-1, 0.011 and 0.013 g (biomass) (mM NO3-)-1, and 0.016 and 0.014 h-1 respectively. In the biofilm system the linear relationship between nitrate loading rate and nitrate removal rate was observed again for the whole range of tested nitrate loading rate range (up to 183 mM h-1), regardless of the approach used to increase the loading rate (increases in feed flow rate or feed nitrate concentration). The highest nitrate removal rate was 183 mM h-1 which was around 194 times higher than that achieved in the continuous stirred tank bioreactor with free cells.<p> A comparison of the autotrophic and heterotrophic denitrification processes studied in the CSTB system indicated that in case of autotrophic denitrification wash-out occurred suddenly and at a much lower loading rate of 0.75 to 0.96 mM (NO3-) h-1 for initial sulphide concentrations 10 to 20 mM, while in case of heterotrophic denitrification increase of nitrate loading rate did not have such a drastic effect and removal rate of nitrate decreased slowly with the increases of nitrate loading rate. A comparison of the kinetic data obtained in the biofilm reactor in the present work and those generated for autotrophic denitrification in an earlier work conducted at University of Saskatchewan (Tang, 2008) showed that the dependency of nitrate removal rate on its loading rate were linear in either case and somewhat similar. However, the maximum nitrate removal rate obtained in the heterotrophic system (183 mM h-1) was much higher than that obtained in the autotrophic system with sulphide.

The present work focuses on the study of two-phase flow behavior called “Intermittent Buoyancy Induced Flow” (IBIF) resulting from the loss of coolant circulation in a CANDU nuclear reactor core. The main objectives are to study steam bubble formation and migration through the pressure tube and into the feeder tubes and headers, and to study the effect of pressure tube sagging on the two-phase flow behavior during IBIF. Experiments are conducted using air and water flow at atmospheric pressure to qualitatively examine the IBIF phenomena. The test showed oscillating periodic behavior in the void fraction as the air vents. In addition to this, a mathematical model based on a simplified momentum balance for the liquid and gas phases was formulated. The model was further solved and compared to the experimental data. The model predictions showed a reasonable agreement within the investigated range of void fractions.

The present work focuses on the study of two-phase flow behavior called “Intermittent Buoyancy Induced Flow” (IBIF) resulting from the loss of coolant circulation in a CANDU nuclear reactor core. The main objectives are to study steam bubble formation and migration through the pressure tube and into the feeder tubes and headers, and to study the effect of pressure tube sagging on the two-phase flow behavior during IBIF. Experiments are conducted using air and water flow at atmospheric pressure to qualitatively examine the IBIF phenomena. The test showed oscillating periodic behavior in the void fraction as the air vents. In addition to this, a mathematical model based on a simplified momentum balance for the liquid and gas phases was formulated. The model was further solved and compared to the experimental data. The model predictions showed a reasonable agreement within the investigated range of void fractions.