Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22


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Collins, R. Pawlowski, A. Toth, R. Gardner, S. Novascone, S. Pitts, J. Hales, G. Kang Seog Kim, Mark L. Williams, Andrew M. Asymptotic Convergence of the angular discretization error of the uncollided scalar flux in the Discrete Ordinates Transport Equation. Aadaptive optimal logarithm grid method for energy lookup optimization in Monte Carlo simulation. Mathematics derivation of the adjoint-weighted tally value for geometric perturbation of k-eigenvalue based on continuous-energy Monte Carlo method.

Sensitivity analysis of the effect of spent fuel out-of-pile residence time and actinides recovery ratio on the performance of nuclear energy system with closed fuel cycle. Subcriticality analysis of neutron-multiplying media with unknown composition and localization of nuclear materials. Ernoult, X. Doligez, N. Thiolliere, A. Zakari-Issoufou, A. Bidaud, S. Bouneau, J. Clavel, F. Courtin, S. David, A. Percher, N. Killingsworth, S. Kim, J. Scorby, D. Heinrichs, R. Sanchez, T. LIN, J. ALT, L. Kochetkov, A. Messaoudi, P. Baeten, V. Billebaud , S. Chabod, T.

Chevret, F. Lecolley, J. Lecouey, G. Lehaut, N. Marie, D. Villamarin, G. Vittiglio and J. Leconte, P. Archier, C. De Saint Jean, R. Diniz, A. Dos Santos, L. Fautrat, D. Among other features, MSRs envisage increased safety, reduced costs and compactness [2]. These, combined with the potential of using thorium as fuel as well as reusing the waste of the existing nuclear power plants, lead to active research in MSR technology during the past two decades.

However, the conventional approach to regulating nuclear technologies makes it challenging to share the knowledge and the technical experience accumulated in this field among all involved parties. Establishing a platform which would enable decentralized and highly secure data collaboration as well as allow providing services relevant to the MSR development, could become ground-breaking for the nuclear power technologies and open up a path for making nuclear a publicly accepted solution for producing clean energy worldwide. In this article we explain how the Thorium Salt Network is planning to build a blockchain based platform which would allow MSR experts and supporters to not only collaborate and share knowledge but also provide and receive services in MSR material research, fuel carrier salt purchase, etc.

Resume : A promising method of reprocessing of advanced nitride spent nuclear fuel SNF is the application of pyroelectrochemical technologies where SNF is dissolved in the LiCl-KCl eutectic with subsequent electrolytic deposition of uranium and plutonium. A salt residue containing fission products and minor actinides is generated in this process as a waste.

Such material have to be immobilized in a stable waste forms among which crystalline matrices are the most promising materials. We have used an argillaceous material that has high sorption affinity towards alkali metal ions as a base of crystalline matrix. The optimal temperature and annealing time were established to produce mechanically stable matrix with high load of LiCl-KCl eutectic.

The compressive strength of the material is higher than 10 MPa. X-ray diffraction revealed the formation of new phases containing lithium after annealing. Morphology and crystallinity of the samples were investigated with scanning electron microscopy and high-resolution transmission electron microscopy. The samples were irradiation by different sources: gamma-quanta and electron beam.

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Irradiation of the samples up to a dose of MGy did not result in any structural changes or compressive strength decrease regardless of the type of radiation. Leaching test demonstrate high chemical durability of the material. Resume : With the development of nuclear power in the world and the gradual industrialization of the country, the demand for uranium resources is also increasing.

Capturing U VI from wastewater and seawater is highly attractive for the environment and clean energy with the increasing deficiency of land sources. Howbeit, the huge volume of water and the very low concentration of U VI pose a substantial challenge to the industrial application. The synergistic reaction between Co-N bond and amidoximation improves the adsorption performance in a wide pH range, which is favorable for U VI capture under nuclear wastewater and seawater.

Furthermore, the amount of U VI reached 2. This includes a huge number of applications, such as optical materials, phosphors, laser active elements, dosimeter, imaging plates, materials for fusion technology etc. After introducing some basics on the radiation point defects in halides, binary oxides and oxide perovskites [1] as well as the mechanisms of point defect formation under particle irradiation neutron, ion, proton, electron , the current understanding of the point defect thermal annealing processes will be briefly reviewed.

We will present recent results and discuss current understanding of the kinetics of the F-type center annealing in above-mentioned compounds after electron, heavy ions or neutron irradiation, which were treated as the bimolecular process with equal concentrations of the complementary F and Oi defects. The process is controlled by the interstitial oxygen ion mobility, which is much higher than that of the F centers.

The results obtained are used for the evaluation of interstitial oxygen migration parameters and are compared with the available ab initio calculations. References 1. Popov, E. Kotomin, J. Maier, Nucl. B Kotomin, V. Kuzovkov, A. Popov, R. Vila, Nucl. Popov, J. Maier, R. Vila, J. Phys Chem A Kuzovkov, E. Kotomin, A. Resume : An analytical expression that evaluates the effect of the pH and the redox potential Eh was developed for studying the sorption of actinides onto substrates mimicking inorganic organic and bioorganic absorbers in seawater conditions. It includes surface complexation with one type of surface site and its formulation yields to a distribution coefficient Kd as function of the pH hydrolysis and Eh redox sensitive species.

The formulation considers also the values of the stability and hydrolysis constants for all species in solution and sorbed at the surface, and makes use of semi-empirical correlations between hydrolysis and surface complexation constants, for each surface. The presence of complexing ligands in solution such as carbonates was also taken into account. The model was applied to the sorption of uranium onto aluminol, hydrated iron oxide and silanol, as well as carboxylic and phenolic both mono- and bi- in the presence and in absence of carbonates in solution. The primary driver on this approach is to simulate the absorption of uranium species in seawater.

When carbonates are present in solution the calculated values of the distribution coefficient were lower than those calculated in the absence of carbonates, and no redox effect was detected. The distribution coefficient Kd values obtained with the developed model are compared with values reported for the sorption of uranium onto specific absorbants from the literature.

It is known that in the water stability region U IV and their hydroxides are the primary stable species in surface waters; however, some reduction effects are possible when interacting with the surface, which must be taken into account in the model. All the possible redox reactions of uranium were considered as a consequence. Moreover, this model is applicable to study the sorption of other redox sensitive elements of interest. This model will help to derive the best conditions for absorption of uranium from seawater.

Resume : Nanosized hillock-like surface defects produced by swift heavy ions have been registered in Al2O3 and MgO, having a relatively high threshold of specific ionizing energy loss for structural disorder enhancement. It was found that all hillocks in TiO2 are crystalline and epitaxial with the original crystal surface while the hillocks in Al2O3 are amorphous.

Analysis of rutile crystals irradiated in the temperature range K has revealed that average hillock height increased with irradiation temperature although large specimen-to-specimen and hillock-to-hillock size differences were also recorded. These experimental observations were discussed in the framework of a simple model based on the inelastic thermal spike i-TS model. Resume : Metals and alloys, such as stainless steels and zirconium alloys, used as structural materials in the nuclear core of pressurized water undergo irradiation creep deformation.

At a macroscopic level, the mechanical behavior is well characterized. Yet, the underlying microscopic mechanisms are still unclear. Many theoretical mechanisms have been proposed in the literature, but only few experimental results were conclusive. Recent in situ TEM straining experiments under ion irradiation conducted on Zircaloy-4 have demonstrated that, at high stress levels, dislocations pinned on irradiation induced point defects clusters start to glide once under irradiation.

One of the proposed hypotheses was that the observed dislocation glide assisted by irradiation was due to a direct interaction between the displacement cascade and the pinned dislocation. The interaction of a screw dislocation with an interstitial loop in the prismatic plane was first studied.

The objective was to pin the dislocation on the irradiation defect and form a helical turn, which is the configuration the most difficult to unpin. The atomic positions corresponding to a stress below the unpinning stress were then extracted and displacement cascade calculations were performed with a PKA Primary Knock-on Atom energy of 20 keV.

It was shown that for a high applied stress and for particular PKA positions the displacement cascade can directly unpin the dislocation in agreement with the experiments. Based on these numerical simulations, a simple analytical probabilistic model was then proposed to explain the irradiation creep deformation under high applied stress.

Resume : The radiation-resistant binary oxides are very important materials for applications in fusion reactors. In this work, we analyzed the kinetics of the F-type and V center annealing after neutron irradiation. F-type center kinetics were treated as the bimolecular process with equal concentrations of the complementary F and Oi defects. Such process is controlled by the interstitial oxygen ion mobility, which is much higher than that of the F centers.

It is demonstrated how the shape of the F-annealing curves is determined by two control parameters: Ea and effective pre-exponential factor and strongly depends on irradiation conditions. The results obtained are used for the evaluation of interstitial oxygen migration parameters and are compared with the available ab initio calculations for MgO and Al2O3.

Finally, we also treated the kinetics of F-center annealing in thermochemically-reduced crystals. The obtained activation energy allows to evaluate both the intrinsic F-center migration energy and also the conditions for metal colloid formation in BeO and ZnO. References: 1. Resume : With the rapid growth of industrialization, water pollution caused by heavy metals has become one of the most serious environmental problems, and attracted considerable attention. Heavy metals are non-degradable and can accumulate in living tissues, so they must be removed from wastewater.

Among these methods, adsorption stand out from the aforementioned methods due to its simple operation and cost-effectiveness. Thus far, the most efficient adsorbent for adsorption heavy metals have high adsorption capacity and removal rate. However, in recent years, the adsorbents designed for the capture of heavy metal have not been suitable for application at seawater pH. Currently, the design of a suitable wastewater adsorbent shows more promise for the extraction of heavy metals. In this report, we report a facile approach to construct a suitable wastewater pH and large surface area material that polyaniline PANI covalent grafted onto the surface of GO nanosheets.

Finally, the result also exhibited outstanding adsorption efficiency and adsorption capacity under the operating conditions for the adsorption-desorption of U VI from aqueous solution, which indicated a promising potential in the application of the absorbent in wastewater. Resume : Atomic displacement is one of the key factors that influence the behaviors of material properties during and after irradiation. For example, the life of a nuclear reactor is determined by the irradiation resistance of the reactor pressure vessel.

Our DPA calculation results utilizing the improved efficiency function are validated against the experimental data for the Fe, Ni, and Cu. Resume : Fuel-cladding chemical interaction FCCI occurs when the nuclear fuel or fission products react with the cladding material. A major cause of FCCI in metallic fuels during irradiation is fission product lanthanides, which tend to migrate to the fuel periphery, coming in contact with the cladding. The result of this interaction is degradation of the cladding, and will eventually lead to rupture of the fuel assembly.

In order to extend fuel life and safely reach higher burn-up, a method to control the lanthanides, to either decrease or mitigate FCCI, is needed. Fuel additives are one method being investigated for this purpose. The rationale behind fuel additives is to have an element dispersed throughout the fuel matrix that will react with the lanthanides as they are produced. A comparison of these additives, in both U-Zr and U-Pu-Zr based fuels, will be presented, along with post-irradiation examination results for the Pd additive in a U-Zr fuel. Resume : Herein, a design of bifunctional hybrid adsorbent made of a magnetic core and a zeolite imidazolate framework ZIF-8 shell embedded with carbon dots CDs is synthesized for U VI efficient removal.

The carbon dots and magnetic core Fe3O4 were prepared by one-pot hydrothermal approach using carboxymethyl cellulose CMC as precursor and stabilizer. The adsorbents retained the strong superparamagnetic behavior of Fe3O4 nanoparticles and favorable biocompatibility. Interestingly, the adsorbents not only clearly display a typical photoluminescence effect of carbon dots, but also obviously enhance U VI adsorption performance via assisting with the carbon dots.

Resume : Over the past two decades, nanoscience and nanotechnology have attracted increasing attention of the scientist community and industry. As a consequence of nanoscale, this type of materials exhibit size- and shape-dependent properties. Indeed, the physical and chemical properties show a lot of difference between the bulk and the nanoparticles electronic, magnetic, optical, catalytic…. In , Murray and al have reported a surfactant-assisted method in organic medium enable to produce highly monodisperse nanocrystals [1]. Since, scientists have devoted a significant effort on the improvement of this synthetic method for the production of size- and shape-controlled nano-objects witch can be envisaged as novel building blocks.

These parameters are crucial for the organization of nano-objects in 2D or 3D architectures and for the study of their intrinsic and collective properties. This project aims to elaborate and characterize new nanostructured micrometric objects using two bottom-up approaches.

The second one need two steps: the first step consists in nanoparticles synthesis and the second one involves nanoparticle surface modification by reactive molecules able to promote self-assembling by formation of covalent bridges Click Chemistry or by electrostatic interactions. Using the first synthesis method, hybrid nanomaterial was produced as highly organized sheets presenting a lamellar structure. The distance between sheets can be controlled by temperature and the inter-lamellar distance can be modulated by the nature of the organic linker. The self-assembly synthesis was used to obtain 3D nanostructured hybrid materials.

The first part of this work has been done on uranium oxide. The next goal for this work is to assemble mixed actinide nanoparticles with platinoid nanoparticles in order to produce model materials. Murray, D. Norris, et M. Bawendi, J. Resume : The current fleet of operating nuclear power reactors in India consists of 22 number of - small, medium and large size reactors.

Over the years, different types of fuel combinations have been used in these reactors. This paper talks about the experiences gained and future outlook with regard to fuel cycle management of Light Water Reactors. The paper focuses on the strategies adopted for the achievement of this goal. The better design of fuelling patterns and optimization of control rod withdrawal sequences have tremendously improved the fuel performance in these reactors.

The paper discusses experience gained with loading of MOX fuel in these reactors. These units viz. Currently Unit-1 is operating in 4th cycle while unit-2 is operating in 2nd cycle. Commissioning tests performed during first physical startups and subsequent operation of both units demonstrated that fuel and core behaviour were in close agreement with theoretical estimations.

In unit-1, the first set of thermal surveillance specimens coupons have been dismantled after Cycle-3 for material analysis. The paper brings out salient features of fuel performance for both units and future outlook with regard to further improvements. The damage effects caused by different ions were investigated by Raman spectroscopy. The significant red-shifts of LO mode were observed, which reveals the presence of lattice defects and stress induced during the irradiation process.

The presence of this mode also provides a new way to identify the nature defects in InP crystalline. Resume : F centers play an important role in characterization and determination of the radiation resistance of functional materials for nuclear materials science.

Dose, Rate, Time Calculations

In this work the influence of the induced F-center to the atomic structure and electronic properties of Mn ion substituted for the host Al at YAlO3 have been studied from the first principles. To perform ab initio modeling of Mn-doped YAlO3 we were using approach of hybrid exchange-correlation functional HSE within density functional theory.

Calculated atomic coordinates and lattice parameters, as well as bulk modulus and band gap for perfect YAlO3 are in a very good agreement with the most recent experimental observations. The electronic charge redistribution calculated for perfect orthorhombic YAlO3 crystal suggests notable covalency of the Al-O bond, that can be validated from further X-ray powder difraction analysis. This covalency is somewhat decreased in the Mn-O bond after doping. Formation of F0-center in the vicinity of the Mn dopant is accompanied by a well pronounced shift 0. The F-center traps 0.

The remaining 0. Funding from Latvian-Ukranian biletaral project is greatly acknowledged. As a key manufacturing process, electron beam welding is applied to seal the gap between the cladding and the end plug of the fuel rod. KAERI installed the new electron beam welding machine and the welding jig to weld multiple fuel rods at the same time. Up to seven fuel rods can be mounted on the new jig and welded at one time. Welded fuel rods were checked the welding integrity such as the presence of pores or cracks by X-ray inspection and liquid penetration test. Also, the microstructural analysis of the cross-section was conducted to inspect the penetration depth and the soundness of weldment.

X-ray and microstructural analysis results showed that the rods was fully welded and satisfied with the manufacturing specification. As a result, the fabrication efficiency was remarkably improved. Resume : Calculations and measurements of the isotopic fraction of the fission gases released during burn-up in light water reactor have been successfully achieved as function of the burn up for the fuel and in the plenum. The burn-up explored are ranging from 0 to MW h kg The model calculation takes into account all isotope inventory considering the neutron absorption, fission, fission product build-up as well as precursor and fission gas isotope decay and their reactivity in the neutron flux of 0.

The experimental data are gained by puncture and mass spectroscopy analysis. Comparison of experimental and modeled data fit reasonably well for both Kr and Xe isotopes. Resume : In order to qualify innovative materials for next generation nuclear systems, their radiation resistance must be assured. Irradiations studies with neutrons present overwhelming complications related to cost, availability and duration.

Ions have been found as a valid alternative to simulate neutrons and gather useful data. Previously, we reported on amorphous Alumina Al2O3 coatings under heavy ions irradiation up to dpa, showing a general radiation-induced crystallization. In this work, we use different ions to irradiate up to 3 dpa.

Here, we concentrate on the low dpa regime, to study the first stages of crystallization and obtain dpa values compatible with neutrons. Irradiations are now performed at different temperatures in order to decouple the thermal contribution from the radiation-induced effects. Based on the experimental data, a preliminary kinetic model is proposed. From a mechanical point of view, an evident size-effect is manifested. The growth of nano-crystalline domains increase rapidly the hardness, in accordance with the inverse Hall-Petch trend.

A second irradiation campaign is carried out with in-situ TEM tandem apparatus. Tests are repeated on free-standing films, to collect dynamically microstructural changes and phases transformation.

PHYSOR "Reactor Physics Paving The Way Towards More Efficent Systems"

To conclude, a consistent and coherent picture of the evolution of amorphous Al2O3 under irradiation is presented. It would operate maintenance-free for 10 to 20 years and choice of materials is thus of paramount importance. One notable problem of molten salts is that they can corrode steel and melt aluminium, so non-traditional materials and manufacturing methods must be used. Attention must therefore be directed to formulating salt mixes generally fluorides which have the least corrosive properties and introducing low-corrosion materials from which the various parts of the reactor are constructed.

With regard to quantity rather than quality of materials, nuclear reactors, including thorium MSRs, use only a fraction of the materials such as concrete and steel used by other forms of sustainable energy. Small reactors such as ThEA could do this. Of the nations recognized by the UNO, have access to the sea. ThEA would be suitable for powering these since it is light enough to be taken by rocket. It could also power orbiting space stations, permitting a wider range of zero-gravity industries. Grazing and symmetrical incidence X-ray diffraction investigations showed that films were nanocrystalline, with grain size around 10 nm, dense and textured.

Rutherford backscattering spectrometry investigations showed that films were slightly Zr rich and their chemical composition was not changed by Au irradiation. X-ray photoelectron spectroscopy investigations found a low. Kauric a, A. Smith b, S. Bordier a, S. The high temperature gradient between the centre and the periphery of the fuel pellet induces the migration of the most volatile fission products Cs, I, etc towards the pellet rim. In the potential event of a clad breach, the metallic sodium would interact with the fuel, and hence with this layer, mainly composed by the fission products: cesium, barium, molybdenum, iodine, tellurium, with phases such as Cs2MoO4, CsI, Cs2Te, BaMoOx, etc.

For a thorough safety assessment of the SFR, the possible reactions between sodium and the compounds of the irradiated fuel, in particular cesium molybdate, need to be investigated. In fact, cesium and molybdenum are produced with high fission yields and cesium molybdate is one of the most stable compound formed at high burnup in the JOG under operating conditions of the reactor.

Therefore, the interaction between sodium, cesium and molybdenum will start at the beginning of the accident. Moreover, in case of a severe accident, the fast increase of the fuel temperature can lead to the fission product release and to the fuel melting. However, these extreme temperature conditions are challenging to reproduce in the laboratory.

Therefore, a thermodynamic model is needed to be able to predict the phases formed under these extreme conditions. In this work, the investigation of the system Cs-Na-Mo-O is presented. For such multicomponent systems and large scale of compositions and temperatures, the Calphad method is the most suitable to analyse the chemical interaction between fission products, sodium and the phases formed in solid, liquid and gas states. CALPHAD, which stands for CALculation of PHAse Diagram, is a semi-empirical method that enables to develop a thermodynamic model based on the Gibbs free energy of the gas, liquid and solid phases as a function of the temperature, composition and pressure of the system.

It describes the equilibrium state of the system by modelling the total Gibbs energy of the system, expressed as a linear function of the Gibbs energy of the phases in the whole system. To do so, adjustable parameters are optimised to fit the experimental data available structural and thermodynamic properties or phase diagram data. Recently, Zolotova et al. While studying the thermodynamic properties of the newly determined quaternary compound, Smith et al. As this system is key for the safety assessment of the SFR, a reinvestigation has been performed in this work in order to develop an accurate model of the pseudo-binary section and then of the quaternary system Cs-Na-Mo-O.

The results are consistent with the ones obtained by Zolotova et al. Moreover, we did not observe the low temperature phase transition of the quaternary compound in agreement with the study of Smith et al. In addition, the enthalpy of fusion of the quaternary compound Cs3Na MoO4 2 is reported for the first time, and the hexavalent state of molybdenum in Cs3Na MoO4 2 has been confirmed with an X-ray absorption spectroscopy experiment.

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These data combined with the ones measured by Smith et al. In this work, an X-Ray diffraction pattern of the compound has been reported for the first time. To refine the structure by the Rietveld method, the compound RbNaMo3O10 space group Pnma 62 was used as starting model. The isostructurality of the two compounds can be expected from the fact that the ionic radii of rubidium and cesium are very close [6]. Moreover, to determine precisely the Mo-O bond distances, an X-Ray Absorption Spectroscopy experiment has been performed. As the compound was pure enough to measure its thermodynamic properties, a DSC analysis has also been performed.

Thanks to this measurement, the melting temperature and enthalpy of fusion have been determined and no phase transition has been observed. The enthalpies of fusion of the two quaternary compounds have been measured and the melting temperature of the CsNaMo3O10 compound has been determined for the first time. Thanks to this study and the data reported in the literature, an accurate model of the Cs-Na-Mo-O phase diagram has been developed for the first time.

Zolotova, Z. Solodovnikova, V. Yudin, S. Solodovnikov, E. Khaikina, O. Basovich, I. Korolkov, I. Solid State Chem. Smith, G. Kauric, L. Goubitz, N. Clavier, G. Wallez, R. V Korolkov, I. Smith, M. Griveau, E. Colineau, R. Konings, Thermodynamic study of Cs3Na MoO4 2: Determination of the standard enthalpy of formation and standard entropy at But the plutonium concentration is low and thus the system can be considered as a five-component one.

This system has been considered early [1]. Modeling of the system includes few points: 1. The application of mole fractions and volume ones and rational activity coefficients, 2. The calculation of concentration of free unbounded with solvates water. The methods of activity coefficient calculation. The methods to estimate errors during the calculation of various coefficients. The main system is usually divided into sub-systems and every sub-system is calculated through thermodynamic activities.

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Activities of nitric acid and uranyl nitrate were calculated in []. Ochkin, A. Procedia Chemistry 7, Theoretical Foundations of Chemical Engineering, , Vol. Davis W. Nuclear Chem. Ochkin A. Radiochemistry, , Vol. In this process, uranium, plutonium, and neptunium are concentrated in the organic phase, and americium, curium, and fission products are concentrated in the aqueous phase.

The concentration of target elements in HLRWs does not exceed 0. HLRW specific activities were determined by 90Sr and Cs fission products and americium and curium radionuclides.


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Now it is necessary to isolate americium and curium radionuclides. Here, there are the two possible variants: 1 the coextraction of americium and curium — DIAMEX — process; 2 the extraction of americium only, whereas curium remains in the aqueous phase EXAm -process. Herein, a facile yet versatile strategy for fabrication of adsorbents were reported. The fabricated mesoporous magnesium hydroxide Mgs material was obtained via the in situ conversion of the natural ore powder magnesite precursor, MgCO3.

The unique internal pore structure to be well suited as a platform for the deployment of highly efficient sorbents, and combined with carboxymethyl cellulose CMC lock water features that they exhibit remarkable performance and production cost advantages. Resume : X-ray absorption spectroscopy was applied to understand the speciation of elements relevant to the immobilization and disposal of radioactive plutonium bearing wastes, utilizing Ce as a Pu surrogate.

Cl K-edge XANES studies of zirconolite glass-ceramics, designed for immobilization of Pu residues, demonstrated incorporation within the aluminosilicate glass phase as the chloride anion, likely co-ordinate to Na as a network modifying species, below the Cl solubility limit of 1. In each of these examples, X-ray Absorption Spectroscopy has provided a pivotal understanding of element speciation in relation to the mechanism of incorporation within the host waste form intended for geological disposal. Resume : Mo decays to Tcm, which is the most widely used radiopharmaceutical isotope for medical diagnostic purposes.

Recently, Mo producers have been attempting to replace conventional highly enriched uranium HEU targets with low enriched uranium LEU targets by international non-proliferation policies. As a result, it is necessary to develop high-uranium-density targets with LEU to improve the Mo production efficiency of LEU targets.

Atomization is a key technology for achieving high-uranium-density because atomized powder is able to have various U-Al compositions and a high uranium content. We successfully mass-produced uranium alloy powder for high-density targets through centrifugal atomization and fabricated a high density target with a uranium density of 3. The development of higher density target fabrication technology over 4.

Electrostatic interactions and possible "ligand separation" mechanism was proposed to explain this phenomenon. This paper provided a new insight for designing the effective adsorbent for U VI recovery at seawater pH. Primary radiation defects in ionic solids consist of Frenkel defects—pairs of anion vacancies with trapped electrons F-type centers and interstitial ions. Using phenomenological theory of diffusion-controlled recombination of the F-type centers with much more mobile interstitial ions complementary hole centers [1], we suggested theory, how to extract from experimental data the migration energy of interstitials and pre-exponential factor of diffusion.

The correlation of these two parameters satisfies the so-called Meyer—Neldel rule MNR [2] observed more than once earlier in glasses, liquids, and disordered materials, but not in the irradiated materials. Our results [2] allow to establish a direct relation between irradiation fluence and its post-irradiation thermal annealing. We compare results for corundum and YAG -- both wide gap insulating materials but with different crystalline structures. As the result, it is demonstrated for corundum that with the increase of radiation fluence both the migration energy and pre-exponent are decreasing, irrespective of the type of irradiation.

This is MNR with normal dose dependence. We discuss the cause of this phenomenon. Thus, in this study, we demonstrated that the dependence of defect migration parameters on the radiation fluence plays an important role in the quantitative analysis of the radiation damage of real materials and cannot be neglected. Villa, Nucl. B , A, , 28 Izerrouken, A. Meftah, M. Nekkab, Nucl. Detailed crystallographic analysis carried out in the present study has shown that all the three phase display structural relationship among them. This uniquely defined structural relationship is a consequence of the strain associated with the phase transformation.

Systematic study of the decomposition behaviour of three alloys viz. Many such transformation related issues will be discussed in the present paper. Resume : China has been committed to peaceful use of nuclear energy for decades. Currently China has the third largest nuclear power industry in the world, with 45 units in operation and 11 units under construction. Resume : Materials employed in nuclear environments should be able to exhibit the highest radiation tolerance.

Mainly the materials issues such as embrittlement and swelling in fuel cladding and structural components are responsible for the lifetimes of current and even more of new-build and future reactors. Radiation tolerance-related requirements are further boosted by the higher performance expected for Generation IV fission and fusion reactors, which will expose materials to much higher numbers of atomic displacements then current and near-feature reactors.

During nuclear energy production, both fuel components and structural materials are subject to substantial radiation damage. They initially appears in the form of local intrinsic point defects within the material i. Then, these point defects agglomerate, interact with the underlying microstructure and lead to undesirable effects such as blistering and radiation-induced embrittlement, which render the materials.

Another important factor is represented by helium embrittlement. He originates from the transmutation of reactor elements which can release alpha particles that acquire electrons to become helium atoms. He is insoluble and mobile in most metals and migrates to grain boundaries and interfaces where it forms bubbles leading to embrittlement.

Many of the materials being developed for deployment in these environments are based on incremental improvements to existing materials such as Oxide Dispersion Strengthened steel. Recently, as a new material to be used in nuclear structures, we prepared Zr-Nb metallic multilayer composites using magnetron sputtering technique and showed both theoretically and experimentally that they are mechanically durable under the harsh environments we mentioned above.

Here, we present the density functional simulation results of our further investigation, vacancy-interface-He interaction and migration of He in nanoscale Zr-Nb metallic multilayer composite. Our results enlighten where He atoms would be positioned within the material and why as well as how they would move within the material. Resume : Nuclear fuel cladding materials that possess higher performance attributes than the currently used cladding materials are always desired in order to improve nuclear fuel performance in reactors.

In order to improve the properties of the existing light water fuel cladding material, laser shock processing LSP has been applied to commercial zirconium alloy. After LSP a thin oxide layer formed on the surface of the specimen. Meanwhile, the surface residual stress and microstructure of the specimen was altered.

After being exposed to high temperature steam for a long time, the corrosion weight gains of samples processed by the pulsed laser were less than those that were untreated, and the transition time of corrosion kinetics has been delayed. We also studied the mechanism of zirconium alloy corrosion process, especially the effect of macroscopic compressive stress on corrosion of zirconium alloy. From the above experiments we concluded that LPS could enhance the corrosion resistance of zirconium alloy.

Resume : As the Loss of coolant accident in Fukushima showed, it is vital for public safety and fuel integrity to understand and continuously improve the performance of fuel cladding materials for light water reactor applications, both during in-core use and in long-term post-irradiation storage. One of the major causes of embrittlement and failure in reactor structural materials including fuel cladding materials is the formation and evolution of precipitates by the formation of hydrides or radiation-induced phase separation.

In this presentation, we will summarize our recent work to characterize hydride formation in zircaloy 4 fuel cladding and the formation and thermodynamic stability of radiation-induced precipitates in FeCrAl fuel cladding candidate alloys by small-angle neutron scattering. Small-angle neutron scattering SANS is a powerful technique for measuring bulk averaged nanometer length scale structures in a variety of materials nondestructively, providing complementary information to direct-geometry small-volume or surface techniques like electron microscopy and atom probe tomography.

This technique is well-suited for use in studying the properties of alloys due to the widely varying contrasts of different transition metal isotopes when viewed with neutrons, the sensitivity of neutrons to magnetic structure, and the high penetrating power of neutrons in many materials regardless of atomic number. The large and opposite neutron scattering cross sections of hydrogen and deuterium make SANS and other neutron techniques extremely well-suited to the characterization of hydrides in alloys.

These properties, together with the high flux available at modern neutron scattering user facilities and the existing infrastructure for working with radioactive materials, make SANS a uniquely powerful technique for studying reactor structural alloys. The first work that we will describe is a nondestructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor LWR fuel cladding was developed.

In the second part of the talk we will describe how SANS is used to characterize the growth of irradiation induced precipitates in Fe-Cr-Al model alloys ranging in composition from wt. We also describe the procedures used to perform measurements using on highly radioactive samples shielded to minimize personnel radiation dose and the risk of contamination.

One challenge faced when studying alloy systems with SANS or when using thick metallic shielding for irradiated samples is the possibility of contamination of the SANS signal by multiple Bragg diffraction, particularly when working with shorter neutron wavelengths. In this talk, we will show examples of how this can appear and describe a strategy for reducing it impact. Methods such as this will be particularly powerful and valuable as the focus in SANS instrumentation shifts from the traditional, monochromatic steady-state instruments typical of reactor scattering facilities to polychromatic time-of-flight sources like those at pulsed sources like most spallation sources and now being introduced as an option at steady-state sources such as the ILL.

Department of Energy. Department of Energy under Contract No. Resume : MAX phases have been considered for various applications, such as contact and structural materials in nuclear applications among others. Particularly, Zr- and V-containing MAX phases, exhibit a remarkable resistance towards oxidation, irradiation and thermal solicitation.

The coatings were obtained using cathodic arc deposition using elemental targets. The films were systematically analyzed to track their growth properties, in terms of orientation and microstructure with respect to the substrate. Effects of pressure on oxides formation at elevated temperature were studied as compared to commercial FeCrAl alloy. Problems of interaction between the liquid metal and the fuel cladding materials, as well as the development of methods to improve the corrosion resistance of steels are important tasks to be solved.

During last decades different methods were shown to be effective in fabrication of additional surface alloyed layer with elements that form thin and dense protective oxide layers on the surface of steel such as aluminum and chromium thus preventing steel from corrosion. Samples of EP ferritic-martensitic steel were pre-oxidized in flowing liquid lead loop with controlled oxygen content and irradiated by 4 MeV Au ions at room temperature and at C. Resume : The development of safer and more efficient nuclear reactors calls for new materials with better radiation tolerance.

In recent years, we have made a renewed exploration on the controlling mechanisms of radiation tolerance of alloy matrix by deliberately taking out the effects of complicated microstructure features or the pre-existing defect sinks and focusing only on the response of alloy matrices of various composition and crystal structure.

Through a systematic study of a large group of nickel-based fcc single-phase concentrated solid solution alloys with ion beam irradiation and TEM analysis, we have found that the resistance to void swelling generally increases with increased number of alloying elements, but similar enhancement in swelling resistance can also be achieved in selected binary alloys by increasing the concentration or changing the species of the alloying element.

The result can be explained by the change in the sluggishness of interstitial cluster motion that may affect the recombination rate of the Frenkel defect pairs.

MATERIALS FOR ENERGY

The results from the fcc alloys are also compared with that obtained from several bcc alloys, such as FeCrAl and Mo, under the similar irradiation conditions. The effect of alloying composition on radiation hardening studied with nano-indentation will also be presented and discussed in terms of the characteristics of dislocation loops. The effect of Si on the microstructure, mechanical properties as well as on the oxidation behavior was investigated.

No major changes in sinterability have been observed in both the alloys. Synergistic effect of Cr and Si addition and finer microstructure helped to form thin, dense and protective oxide layers, which prevented outward cations and inward anion transport, as a result, excellent oxidation resistance of Si-containing ODS steel.

Keywords: Nanostructured ODS ferritic steel; Mechanical alloying; Spark plasma sintering Rietveld refinement; morphology, oxidation resistance. Resume : The advanced oxide dispersion strengthened ODS 14YWT ferritic alloy was developed for resisting microstructural changes due to exposure to harsh conditions in future reactor concepts, requiring resistance to high dose neutron irradiations at high temperatures. Resistance to microstructural degradation of ODS alloys to ion irradiation has been demonstrated so far but, unfortunately, there are scarce studies focusing on the microstructural response of 14YWT to neutron irradiation.

The stability of the microstructure after irradiation is stated by the presence of Y-rich nanoparticles and no evidence of dislocation loops or cavities formation. The main irradiation effect observed on these samples is the alpha prime formation at both temperatures, which was not observed in previous ion irradiation experiments. A detailed study of these events has been performed in terms of atom probe tomography APT and transmission electron microscopy TEM techniques. These results provide essential data as a baseline to the aim of establishing a correlation between ion and neutron irradiation effects in structural materials.

Resume : The forthcoming Generation IV nuclear reactors will operate at much higher temperatures than the existing nuclear power plants. This feature makes the selection of structural materials to be used in fuel assemblies employed in the reactor core much more challenging. One of the most promising candidates are oxide dispersion strengthened ODS steels that are well known for their advanced mechanical properties and radiation resistance. The oxide nanoparticles are found to be excellent nucleation sites for helium bubbles, yet their contribution to swelling is found to remain relatively minor as compared to other microstructural defects, especially grain boundaries.

However, the bubble growth on oxide particles is shown to be potentially risky in terms of the loss of oxide particle efficiency as dislocation pinning centers and of triggering the bubble-to-void transition, both effects resulting in the severe degradation of steel mechanical properties. Resume : Steels have important applications in current and advanced nuclear reactors, however, their irradiation tolerance and mechanical properties need to be improved. Bulk ultrafine-grained metals possess drastically higher strength than their conventional coarse-grained counterparts, and may have significantly enhanced irradiation tolerance.

In this study, ultrafine-grained austenitic and ferritic-martenstic steels were manufactured by equal-channel angular pressing ECAP and high-pressure torsion HPT. Advanced microstructural characterization techniques were utilized to investigate the microstructures and chemistry of the steels before irradiation. The thermal stability of the ultrafine-grained steels was also investigated.

Neutron irradiation was designed and is being performed to study irradiation behavior of the steels. Limited ion irradiation was also conducted to compared with the neutron irradiation. Resume : The expected degradation of the mechanical properties of structural materials under irradiation i. Radiation-induced embrittlement is caused by nano-scale defects that obstruct plasticity mediated by dislocations. In this work, we summarize and discuss recent results obtained using several fine-scale experimental techniques namely: transmission and scanning electron microscopy, small angle neutron scattering, internal friction and magnetic after effect measurements, atom probe tomography applied to the Fe-Cr model alloys from a single dedicated irradiation campaign.

Recent experiments were performed to investigate the impact of such important alloying elements as Ni, Si and P on the microstructural evolution. We provide new microscopy information on the pattern of irradiation damage in the Fe-Cr alloys doped with the above noted impurities. To rationalize and bridge experimental information, we apply material's modelling framework, involving several up-scaling methods from atomistic simulations to the full-scale Monte Carlo model to predict nano-scale irradiation defect evolution. As a result, we distinguish several types of microstructural features created by neutrons which are dislocation loops, voids and solute rich clusters.

The results showed that the concentration of the alloying elements Si, Mn impacted both the oxidation kinetics and the oxide scales formed in a synergistic manner. Nevertheless, the increase of Mn content induced in the formation of a non-uniform oxide scale on the steel surface. The corrosion mechanism affected by the alloying elements Si, Mn was discussed in this study.

Resume : We conducted reactor performance calculations to assess the potential design basis accident performance of HTGR fuel designs. The maximum fuel temperature in the cores fueled by these three FCM fuels was predicted to be higher than that in the reference MWt mHTGR core in both normal operating conditions and during representative design basis accidents. To better understand the potential safety margins in mHTGR design basis accidents, we performed thermal-hydraulics sensitivity studies to investigate how maximum fuel temperature varies considering various parameters, e.

We found that the difference in the steady-state axial power distribution contributed the most to the difference in the maximum fuel temperature, in both normal operation and design basis accidents. Experimental data suggested that the annealing process of irradiation defects in SiC would be rapid at mHTGR relevant fuel temperatures. The bounding potential impact of the SiC annealing on the maximum fuel temperature was analyzed considering both the thermal conductivity recovery and the Wigner energy release due to the annealing of SiC.

Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22 Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22
Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22 Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22
Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22 Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22
Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22 Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22
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Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22 Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22

Related Spatial Kinetics Calculations of MOX-Fuelled Core, var. 22



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