Kyoto University (京都大学) / 0048 / 新制・課程博士 / 博士(工学) / 甲第15375号 / 工博第3254号 / 新制||工||1490(附属図書館) / 27853 / 京都大学大学院工学研究科原子核工学専攻 / (主査)教授 森山 裕丈, 教授 山名 元, 准教授 佐々木 隆之 / 学位規則第4条第1項該当

Light water reactors (LWRs) are the world’s dominant nuclear reactor system. Uranium (U)-fuelled LWRs produce long-lived transuranic (TRU) isotopes. TRUs can be recycled in LWRs or fast reactors. The thermal neutron spectrum in LWRs is less suitable for burning TRUs as this causes a build-up of TRU isotopes with low fission probability. This increases the fissile feed requirements, which tends to result in a positive void coefficient (VC) and hence the reactor is unsafe to operate. Use of reduced-moderation LWRs can improve TRU transmutation performance, but the VC is still severely limiting for these designs. Reduced-moderation pressurized water reactors (RMPWRs) and boiling water reactors (RBWRs) are considered in this study. Using thorium (Th) instead of U as the fertile fuel component can greatly improve the VC. However, Th-based transmutation is a much less developed technology than U-based transmutation. In this thesis, the feasibility and fuel cycle performance of full TRU recycle in Th-fuelled RMPWRs and RBWRs are evaluated. Neutronic performance is greatly improved by spatial separation of TRU and 233-6U, primarily implemented here using heterogeneous RMPWR and RBWR assembly designs. In a RMPWR, the water to fuel ratio must be reduced to around 50% of the normal value to allow full actinide recycle. If implemented by retrofitting an existing PWR, steady-state thermal-hydraulic constraints can still be satisfied. However, in a large break loss-of-coolant accident, the emergency core cooling system may not be able to provide water to the core quickly enough to prevent fuel cladding failure. A discharge burn-up of ~40 GWd/t is possible in RMPWRs. Reactivity control is a challenge due to the reduced worth of neutron absorbers in the hard neutron spectrum, and their detrimental effect on the VC, especially when diluted, as for soluble boron. Control rods are instead used to control the core. It appears possible to achieve adequate power peaking, shutdown margin and rod-ejection accident response. In RBWRs, it appears neutronically feasible to achieve very high burn-ups (~120 GWd/t) but the maximum achievable incineration rate is less than in RMPWRs. The reprocessing and fuel fabrication requirements of RBWRs are less than RMPWRs but more than fast reactors. A two-stage TRU burning cycle, where the first stage is Th-Pu MOX in a conventional PWR feeding a second stage continuous burn in a RBWR, is technically reasonable. It is possible to limit the core area to that of an ABWR with acceptable thermal-hydraulic performance. In this case, it appears that RBWRs are of similar cost to inert matrix incineration in LWRs, and lower cost than RMPWRs and Th- and U-based fast reactor recycle schemes.

This thesis is intended to contribute to policy analysis on nuclear energy planning, and also as a contribution to applied mathematics. From point of view of nuclear policy analysis, this thesis is not designed to offer realistic detail on nuclear engineering itself, which is of second order relative to our chosen problem. The goal is to address some large scale problems in the management of the world stocks of two important nuclear fuels, Uranium (an economically finite natural resource) and Plutonium (the result at first of policies for Uranium burning, and later of policies on fast reactor breeding). This thesis assumes, as a ‘political’ working hypothesis, that at some future time world governments will agree urgently to decarbonise the world economy. Up to that point, assuming no previous large progress towards decarbonisation, basic world electricity consumption will have continued to grow at its historic average of 1.9% compound. This rate is hypothetically a combination of slower growth in the developed world and faster growth in the developing world. On this hypothesis, a necessary but not sufficient condition for decarbonising the economy would be the complete decarbonisation of future basic electricity demand, plus the provision of sufficient extra decarbonised electricity supply to take over powering all land transport. The demand for electricity for land transport at any time is assumed to equal (in line with historical experience) an increment of approximately 20% above the contemporary basic world demand for electricity. The hypothetical scenario for achieving this model of decarbonisation, without major stress to the worlds economic and social system, is to expand nuclear power to meet the whole of basic electricity demand. This would leave intermittent renewable sources to power the intermittent electricity demands of road transport.This thesis explores the above hypothetical future in various ways. We first list published forecasts of future Uranium use and future Uranium supply. These suggest that presently known Uranium reserves can meet demand for many decades. However on extrapolating the cumulative demand for Uranium that results from the above working hypothesis, we find that if a dash to decarbonise world electricity supply begins immediately, this would consume a very large multiple of presently known Uranium reserves. Sustaining that decarbonisation for only a few more decades of demand growth would consume further large multiples of the known Uranium supply. A delay in the start of the dash for decarbonisation by only a few decades greatly increases the cumulative Uranium demand needed to reach decarbonisation even briefly.Therefore the sustained achievement of decarbonisation, in a world economy of the historical type, requires such large Uranium resources that a successor fuel cycle is required. This thesis models only the case of a Uranium-based fast reactor fuel cycle, since this cycle can in principle consume all the cumulative past and future Plutonium stockpile, and can then meet its own Plutonium needs for a long period (hundreds or thousands of years), allowing ample time for economic adjustment. However a commercially effective fast reactor technology is some decades away.Up to this point, the thesis has only added two physical factors to the existing debate on Uranium needs: namely cumulative growth of electricity demand at its historic rate, and a political choice for 100% physical decarbonisation of the electricity supply.The mathematical and economic contribution of the thesis then begins. We ask the following questions:1. Under what circumstances would profit-maximising investors (or an economically rational centralized economy) actually choose to build enough reactors to decarbonise the world electricity supply?2. Would the need for investors to make a profit increase or decrease the life of the economically accessible Uranium reserves?3. What is the effect of accelerating or delaying the technical availability of fast reactors?4. When if at all would there be shortages of Uranium or Plutonium?5. Under what circumstances would rational investors chose a smooth and physically feasible handover from Uranium burning to fast reactors, thus avoiding the need for a large but temporary return to fossil fuel?The above questions set a mathematically demanding problem: four interacting physical stocks and two physical flow variables ( control variables) must simultaneously be optimized, along with their economic effects. The two control variables are the rate of building or decommissioning Uranium burners, and the rate of building or decommissioning fast reactors. The first control variable drives the cumulative stock of Uranium burning reactors, and hence the resulting maximum physical supply of electricity (with sales income bounded by demand), less the costs of operating, and of new investment. This variable also drives the cumulative depletion of the finite economically extractable reserve of Uranium, and it simultaneously drives an increase in the free Plutonium stock (from Uranium burning). The second control variable, the rate of building or decommissioning fast reactors, drives a decrease in the Plutonium stock (from charging new fast reactors) and it drives a cumulative increase in the stock of fast reactors. This affects the resulting rate of supply of electricity and of income less operating costs and new investment costs. The combined sales of electricity from the two reactor systems is bounded by the total world demand for electricity.The thesis explores this problem in several stages. A fully stochastic form of the problem (stochastic in the price of electricity) is posed using the tools of contingent claims analysis, but this proves intractable to solve, even numerically. Fortunately the price increases needed to impose decarbonisation are very large, and they result from discrete and long lasting government actions. Hence for policy analysis it is adequate to assume a large one off change in electricity price, and observe the progress towards the resulting evolving equilibrium. This problem is also addressed in stages, firstly we optimise the Uranium burning and the fast reactor cycles in isolation from each other, then we allow some purely heuristic and manually controlled interaction between them. Finally we solve, and economically optimize, the total dynamic system of two physical control variables and the resulting four interacting dependent stock variables.

Conducting research experiments on plutonium electrorefining is difficult due to the significant hazards and regulations associated with nuclear materials. Finding a surrogate for plutonium electrorefining studies would enable more fundamental research to be conducted. Potential surrogates were identified by determining the physical properties required to conduct electrorefining using a molten metal and molten salt in CaCl2 at 1123 K. More potential surrogates were identified by changing the matrix salt to be a LiCl-KCl-CaCl2 eutectic salt with electrorefining conducted at 673-773 K. Ce-CeCl3, In-InCl3, Zn-ZnCl2, Sn-SnCl2, and Bi-BiCl¬3 were investigated as potential plutonium electrorefining surrogates. Ce electrorefining in molten CaCl2 resulted in a difficult to separate colloid mixture of Ce, Ca and Cl. Electrorefining rates for In were too slow due to InCl3 volatilizing out of the molten salt. Zn was successfully electrorefined, but the metal obtained did not coalesce into one piece. Sn and Bi were successfully electrorefined and coalesced into solid product rings with high yields and coulombic efficiencies. While a surrogate could not be identified using the same conditions as plutonium electrorefining, two possible surrogates, Sn-SnCl2 and Bi-BiCl3,¬ were found that could imitate the physical configuration (i.e., molten salt on top of molten metal) of plutonium electrorefining at a reduced temperature using a eutectic LiCl-KCl-CaCl2 salt in place of CaCl2. Using this surrogate enables fundamental studies of aspects of plutonium electrorefining. One aspect of plutonium electrorefining research is to improve its efficiency and yield. Plutonium electrorefining is a time-intensive process which generates radioactive waste. Improvements in efficiency and yield can reduce process time and waste. One possible way of improving the efficiency of plutonium electrorefining is to study the impact of using an AC superimposed DC waveform. Four AC superimposed DC and two DC electrorefining runs were performed using bismuth as a plutonium surrogate. All six runs showed a high level of yield and coulombic efficiency. All six cathode rings were confirmed to be high-purity bismuth using scanning electron microscopy with energy dispersive x-ray analysis (SEM-EDS). While the results were inconclusive about the ability of AC superimposed DC waveforms to increase the efficiency of bismuth electrorefining, applying an AC superimposed DC waveform did not appear to decrease the efficiency or yield of the process. The change in waveform also did not result in impurities being present in the product cathode ring. Bismuth, in addition to being identified as a viable plutonium surrogate, has been investigated as a potential liquid electrode for molten salt electrorefining. Because of this, its electrochemical properties are of interest. However, bismuth's electrochemical behavior has received scant attention in eutectic LiCl-KCl melts and no studies were found in the ternary LiCl-KCl-CaCl2 melts. LiCl-KCl-CaCl2 melts offer some advantages over eutectic LiCl-KCl, such as lower melting point and higher oxide solubility. Cyclic voltammetry, square wave voltammetry, chronoamperometry, chronopotentiometry and open-circuit chronopotentiometry were used to measure electrochemical parameters, such as diffusivity and standard redox potential of bismuth electrodeposition in LiCl-KCl and LiCl-KCl-CaCl2 eutectics.

No description available.

In the late 1970's, emphasis on non-proliferation forced suspension of all commercial spent-fuel reprocessing. The spent-fuel storage problem plaguing the nuclear industry can be alleviated by reprocessing. For commercial spentfuel reprocessing to again become a reality, a process is needed to reform reprocessing operations such that non-proliferation goals are satisfied. To satisfy these goals, the existing process which generates plutonium-nitrate solution must be altered to generate plutonium-uranium oxide powder. The COPRECAL process is designed to produce this solid. The COPRECAL process allows uranium and plutonium to be extracted from spent-fuel for reuse in commercial lightwater reactors. The COPRECAL process is unique in that no pure plutonium is ever present throughout the process, whether the COPRECAL process is intrinsically vulnerable to plutonium diversion is the object of this work. A simulation model of the COPRECAL process is presented which employs state-of-the-art instrumentation to measure in-process plutonium through the simulated passage of time. Plutonium diversion schemes are incorporated into the model. After simulated thefts, model output statistics are plotted on control charts and analyzed. Results show need for major design changes in the COPRECAL process. / Master of Science / incomplete_metadata

The absolute cumulative yields of forty-one light and heavy fragments farmed in the thermal neutron fission of Pu239 have been measured with a mass spectrometer twins the isotope dilution technique. These include the yields of isotopes of rubidium, strontium, zirconium, molybdenum, ruthenium, cesium, cerium, neodymium and samarium. The yields of twelve light fragments formed in the thermal neutron fission of U235 have also been reported. A comprehensive study of fission yields requires a knowledge of carrier-free separation techniques, accurate half-lives and neutron capture crocs sections of the various fission nuclides. Lew procedures for the carrier—free analysis of the elements of zirconium, molybdenum and ruthenium have been developed. Also the half-lives of Sr89 and Zr95 and the thermal neutron capture cross sections of Xe135 and Sm149 were determined. A comparison of the yields of the light and heavy mass fission products of Pu239 has made possible a more detailed understanding of neutron emission from the primary fragments of fission. / Thesis / Doctor of Philosophy (PhD)

Dans la section UO2-PuO2-Pu2O3, les équilibres de phases décrivent un domaine monophasé (U1-y,Puy)O2-x stable pour y<0,20 à 25°C et jusqu'à l'équilibre solide-liquide. Aux teneurs Pu supérieures, ils sont plus complexes avec l'apparition d'une démixtion et la précipitation de phase(s) additionnelle(s). L'objectif de la thèse a consisté à améliorer la représentation du système pour 0,15≤y≤0,65 et 25≤T(°C)≤1500.A 25°C, une lacune de miscibilité composée de deux phases (U1-y,Puy)O2-X a été observée pour y<0,45, dont l'une est de rapport Oxygène/Métal proche de la stœchiométrie et une autre très réduite. Pour la première fois, un domaine triphasé a été caractérisé à teneurs Pu supérieures avec deux phases (U1-y,Puy)O2-X de teneurs proches de y=0,45, et une phase (U1-y,Puy)2O3 comprenant une faible proportion d'uranium solubilisée.L'étude en fonction de la température a démontré que la température de démixtion augmente avec la teneur Pu. Plusieurs représentations ont été établies. A 200°C, les limites d'existence du domaine multiphasé évoluent peu par rapport à 25°C. A 400°C, la démixtion survient à une teneur Pu proche de 0,35, largement inférieure à celle suggérée par la littérature. A 600°C, les résultats précisent les équilibres de phases jusqu'alors très méconnus avec une démixtion apparaissant à partir de y=0,60.L'analyse microstructurale des échantillons a clairement démontré l'impact significatif de la démixtion sur le matériau se traduisant par des fissures au sein des échantillons, d'autant plus nombreuses que la teneur en Pu est élevée. / In the UO2-PuO2-Pu2O3 section, a monophasic (U1-y,Puy)O2-x domain is stable for y<0,20 at 25°C and up to solid-liquid equilibrium. At higher Pu content, phase equilibria are more unclear with a phase separation process. The main objective of this work consisted in upgrading the representation of this system for 0,15≤y≤0,65 and 25≤T(°C)≤1500.At 25°C, a miscibility gap composed by two different (U1-y,Puy)O2-X phases has been observed for y<0,45, with one very closed to stoichiometric state (Oxygen/Metal=2) and one other very reduced. For the first time, a triphasic domain has been characterized at higher Pu contents, with two (U1-y,Puy)O2-X phases near y=0,45 and one (U1-y,Puy)2O3 phase with a low U content inside. Concerning the study in function of temperature, we have demonstrated that phase separation temperature increase when Pu content grows. Several representations have been established. At 200°C, the representation is closed to that at 25°C. At 400°C, the phase separation have been specified at a lower Pu content than that of literature : y=0,35. At 600°C, our results have clarified the section, until then very unclear, with a phase separation appearing at y=0,60.The microstructural analysis has clearly demonstrated the significant impact of the phase separation on the material. Indeed many cracks have been observed in our samples, and quantity of these defects increases when Pu content grows.

Les phénomènes de diffusion se produisant lors du frittage des oxydes mixtes d’uranium et deplutonium (MOX) dépendent du potentiel d’oxygène de l’atmosphère du four, qui détermine lanature et la concentration des défauts ponctuels dans le matériau. Les travaux de thèse ont porté surune meilleure compréhension de l’influence du potentiel d’oxygène sur la densification, la formationde la solution solide et l’interdiffusion U-Pu lors du frittage des combustibles MOX. Pour cela, unlarge domaine de potentiel d’oxygène a été étudié, entre -600 et -100 kJ.mol-1 à 1700°C, afin demettre en évidence les différents mécanismes diffusionnels et leur impact sur la microstructurelorsqu’on s’éloigne de la composition stoechiométrique i.e. lorsque la concentration en défautsaugmente.Les études ont montré que plus le potentiel d’oxygène augmente, plus la densification du mélange70 % UO2+x + 30 % PuO2 s’effectue à basse température. Lors du chauffage, les oxydes de départ(UO2+x et PuO2-x) densifient dans un premier temps puis la solution solide se forme à une températureplus élevée d’environ 200°C. La solution solide apparaît à plus basse température quand le potentield’oxygène augmente, avec une cinétique de formation plus rapide. L’étude de l’interdiffusion U-Puindique qu’un traitement thermique avec un potentiel d’oxygène supérieur à -150 kJ.mol-1 à 1700°Cpermet d’obtenir un coefficient d’interdiffusion supérieur d’un à deux ordres de grandeur à ceuxobtenus entre -550 et -350 kJ.mol-1 à 1700°C et conduit donc à une homogénéisation U-Pu accrue.Cette étude permet de donner des recommandations sur le choix de l’atmosphère et de proposer uncycle de frittage optimisé en fonction de l’application ou de la caractéristique souhaitée. / Diffusion mechanisms occurring during the sintering of mixed uranium-plutonium oxides (MOX) areaffected by the oxygen potential of the atmosphere, as this latter imposes the nature and theconcentration of point defects in the material. This work is focused on a better knowledge of theinfluence of oxygen potential on densification, solid solution formation and U-Pu interdiffusionduring the sintering of MOX fuels. In this aim, a wide range of oxygen potential was studied, between-600 and -100 kJ.mol-1 at 1700°C, to highlight the various diffusional mechanisms and their impact onthe microstructure as oxygen deviates from stoichiometric composition and the defect concentrationincreases.As oxygen potential increases, the densification of the 70 % UO2+x + 30 % PuO2 mixture occurs at alower temperature. During the heating, the initial oxides (UO2+x et PuO2-x) densify first and then thesolid solution starts forming at about 200°C higher. The solid solution appears at a lowertemperature as the oxygen potential increases, with a faster kinetics of formation. The U-Puinterdiffusion study indicates that a heat treatment with an oxygen potential higher than -150kJ.mol-1 at 1700°C allows to obtain an interdiffusion coefficient higher by one up to two orders ofmagnitude compared with those obtained between -550 and -350 kJ.mol-1 at 1700°C and thereforeleads to an enhanced U-Pu homogenizationThis study enables to make recommendations on the atmosphere choice and to provide an optimizedsintering cycle depending on the desired application or required feature.

Cette thèse développe un système d'analyse en ligne compact, précis, rapide et autonome par spectrométrie de fluorescence X raies L, pour quantifier les transuraniens (uranium, plutonium et américium) présents dans les solutions et effluents des procédés du plutonium. Au cours de cette étude, la configuration, les paramètres de ce système et une méthode de quantification non destructive ont été définis puis la capacité du spectromètre à dissocier ces éléments dans de fortes concentrations relatives a été vérifiée. Des solutions contenant des éléments non radioactifs ont d'abord été étudiées afin de s'affranchir des effets dus à la désintégration des transuraniens et d'optimiser efficacement le spectromètre. Par simulations Monte-Carlo avec le code PENELOPE, une méthode de quantification a été mise au point et diverses configurations pour optimiser l'intensité d'excitation de l'échantillon et mesurer une raie de diffusion Compton de la source X ont été testées numériquement. Ces configurations ont été expérimentées avec un spectromètre prototype sur des solutions non radioactives pour valider et compléter la configuration la plus intéressante. Des solutions contenant des éléments en fortes concentrations relatives ont été analysées afin de déterminer une méthode de déconvolution des spectres avec le logiciel COLEGRAM et démontrer la faculté de cette méthode à quantifier de telles solutions. Quelques solutions d'uranium et de plutonium ont été analysées pour s'approcher au plus près des futures conditions expérimentales, caractériser et diminuer les effets dus à la désintégration des radioéléments. / This thesis deals with the development of a new compact, accurate, fast, without cooling liquid, fluorescence L X-ray spectrometer, with the aim to determinate online transuranic elements (uranium, plutonium and americium) in nuclear materials reprocessing. The objective was to define the configuration and the characteristics of this spectrometer and the method to quantify transuranic elements between 0.1 g/L and 20 g/L. To minimize as law as possible the manipulation of these elements, we have used an original approach, namely Monte-Carlo simulations and none radioactive surrogate elements. The study of these solutions allowed to eliminate the specific effects of transuranic elements decay (internal conversion) and to optimize the spectrometer. Monte-Carlo simulations with the PENELOPE code were used for two reasons. Firstly, we have developed an analytical method to correct the matrix effects. Then, we have selected three systems to produce a quasi-monochromatic X-rays beam from the X-ray generator, to optimize the intensity of the L X-ray fluorescence spectra, and to measure the Compton scatter peak. These systems were then tested with an adjustable spectrometer in order to select and optimize the best configuration. We have shown that, after peaks and left tails separation from spectra with COLEGRAM software, the spectrometer can be used to analyze solutions with various concentrations of thallium and bismuth. Several solutions of uranium or plutonium were analyzed to determine the effects of interferences from gamma rays and internal conversion with the fluorescence L X-rays. These effects have been corrected by subtracting a passive spectrum to an active one.