New concept of designing Pu and MA containing fuel for fast reactors

New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO₂ powder of <50 μm fractions prepared by pyrochemical or other methods. In this way a high uranium content fuel meat metallurgically bonded to cladding forms a heat conducting framework, pores of which contain PuO₂ powder. Framework fuel element having porous meat is fabricated by capillary impregnation method with the use of Zr eutectic matrix alloys, which provides metallurgical bond between fuel and cladding and protects it from interaction. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio does not interact with fuel cladding and is more environmentally clean. Its principle advantage is a simple production process that is easily realized remotely, feasibility of involving high background Pu and MA isotopes into closed nuclear fuel cycle at the minimal influence on environment.     

Elemental analysis of mining wastes by energy dispersive X-ray fluorescence (EDXRF)

An energy dispersive X-ray fluorescence (EDXRF) tri-axial geometry experimental spectrometer has been employed to determine the concentrations of 13 different elements (K, Ca, Ti, Cr, Mn, Fe, Co, Ni, Cu, Zn, Rb, Sr and Pb) in mine wastes from different depths of two mine tailings from the Cartagena-La Union (Spain) mining district. The elements were determined and quantified using the fundamental parameters method. The concentrations of Cr, Ni, Cu, Zn and Pb were compared to the values from the European and Spanish legislation to evaluate the environmental risk and to classify the wastes as inert wastes or as wastes that have to be control land-filled. The results obtained demonstrate that these wastes can be considered as inert for the considered elements, apart from the concentration levels of Zn and Pb. Whilst Zn slightly overpasses the regulatory levels, Pb mean value exceeds three to six times the value to be considered as Class I potential land-filling material.     

Positron annihilation study of neutron irradiated model alloys and of a reactor pressure vessel steel

The hardening and embrittlement of reactor pressure vessel steels are of great concern in the actual nuclear power plant life assessment. This embrittlement is caused by irradiation-induced damage, and positron annihilation spectroscopy has been shown to be a suitable method for analysing most of these defects. In this paper, this technique (both positron annihilation lifetime spectroscopy and coincidence Doppler broadening) has been used to investigate neutron irradiated model alloys, with increasing chemical complexity and a reactor pressure vessel steel. It is found that the clustering of copper takes place at the very early stages of irradiation using coincidence Doppler broadening, when this element is present in the alloy. On the other hand, considerations based on positron annihilation spectroscopy analyses suggest that the main objects causing hardening are most probably self-interstitial clusters decorated with manganese in Cu-free alloys. In low-Cu reactor pressure vessel steels and in (Fe, Mn, Ni, Cu) alloys, the main effect is still due to Cu-rich precipitates at low doses, but the role of manganese-related features becomes pre-dominant at high doses.     

Compatibility of martensitic/austenitic steel welds with liquid lead bismuth eutectic environment

The high-chromium ferritic/martensitic steel T91 and the austenitic stainless steel 316L are to be used in contact with liquid lead-bismuth eutectic (LBE), under high irradiation doses. Both tungsten inert gas (TIG) and electron beam (EB) T91/316L welds have been examined by means of metallography, scanning electron microscopy (SEM-EDX), Vickers hardness measurements and tensile testing both in inert gas and in LBE. Although the T91/316L TIG weld has very good mechanical properties when tested in air, its properties decline sharply when tested in LBE. This degradation in mechanical properties is attributed to the liquid metal embrittlement of the 309 buttering used in TIG welding of T91/316L welds. In contrast to mixed T91/316L TIG welding, the mixed T91/316L EB weld was performed without buttering. The mechanical behaviour of the T91/316L EB weld was very good in air after post weld heat treatment but deteriorated when tested in LBE.     

Laser cladding technology to small diameter pipes

A laser cladding method which produces a highly corrosion-resistant material coating layers (cladding) on the austenitic stainless steel (Type 304 SS) pipe inner surface was developed to prevent SCC (stress corrosion cracking) occurrence. This technology is applicable to a narrow and long distance area from operators, because of the good accessibility of the YAG (Yttrium–Aluminum–Garnet) laser beam that can be transmitted through an optical fiber. In this method a mixed paste metallic powder and heating-resistive organic solvent are firstly placed on the inner surface of a small pipe, and then a YAG laser beam transmitted through an optical fiber irradiates to the pasted area. A mixed paste will be melted and form a cladding layer subsequently. A cladding layer shows as excellent corrosion resistance property. This laser cladding (LC) method had already applied to several domestic nuclear power plants and had obtained a good reputation. This report introduces the outline of laser cladding technology, the developed equipment for practical application in the field, and the circumstance in actual plant application.     

Heat load behaviors of plasma sprayed tungsten coatings on copper alloys with different compliant layers

Plasma sprayed tungsten (PS-W) coatings with the compliant layers of titanium (Ti), nickel-chromium-aluminum (NiCrAl) alloys and W/Cu mixtures were fabricated on copper alloys, and their properties of the porosity, oxygen content, thermal conductivity and bonding strength were measured. High heat flux tests of actively cooled W coatings were performed by means of an electron beam facility. The results indicated that APS-W coating showed a poorer heat transfer capability and thermo-mechanical properties than VPS-W coating, and the compliant layers improved W coating performance under the heat flux load. Among three compliant layers, W/Cu was the preferable because of its better effects on heat removal and stress alleviating. The optimization of W/Cu compliant layer found that 0.1 mm and 25 vol.%W was optimum compliant layer structure for 1 mm W coating, which induced a 23% reduction of the maximum stress compared to the sharp interface, and the plastic strain was reduced to 0.01% from 1.55%.     

Cytocompatibility of Ar plasma treated and Au nanoparticle-grafted PE

Polyethylene (PE) was irradiated with inert Ar plasma, and the chemically active PE surface was grafted with Au nanoparticles. The composition and the structure of the modified PE surface were studied using X-ray photoelectron spectroscopy (XPS) and Rutherford backscattering spectroscopy (RBS). Changes in the surface wettability were determined from the contact angle measured in a reflection goniometer. The changes in the surface roughness and morphology were followed by atomic force microscopy (AFM). The modified PE samples were seeded with rat vascular smooth muscle cells (VSMC) or mouse NIH 3T3 fibroblasts, and their adhesion and proliferation were studied. We found that plasma discharge and Au grafting lead to dramatic changes in the surface morphology and roughness of PE. The Au nanoparticles were found not only on the sample surface, but also in the sample interior up to the depth of about 100 nm. In addition, plasma modification of the PE surface, followed with grafting Au-nanoparticles, significantly increased the attractiveness of the PE surface for the adhesion and growth of VSMC, and particularly for mouse embryonic 3T3 fibroblasts.     

Laser cladding of Colmonoy 6 powder on AISI316L austenitic stainless steel

Stainless steels are widely used in nuclear power plant due to their good corrosion resistance, but their wear resistance is relatively low. Therefore, it is very important to improve this property by surface treatment. This paper investigates cladding Colmonoy 6 powder on AISI316L austenitic stainless steel by CO₂ laser. It is found that preheating is necessary for preventing cracking in the laser cladding procedure and 450 °C is the proper preheating temperature. The effects of laser power, traveling speed, defocusing distance, powder feed rate on the bead height, bead width, penetration depth and dilution are investigated. The friction and wear test results show that the friction coefficient of specimens with laser cladding is lower than that of specimens without laser cladding, and the wear resistance of specimens has been increased 53 times after laser cladding, which reveals that laser cladding layer plays roles on wear resistance. The microstructures of laser cladding layer are composed of Ni-rich austenitic, boride and carbide.     

0Cr18Ni10Ti不锈钢源壳TIG焊接温度场数值模拟及试验验证

针对0Cr18Ni10Ti不锈钢放射源源壳钨极氩弧焊(TIG)焊接过程,采用ANSYS有限元软件对焊接温度场进行数值模拟分析,建立了非稳态TIG焊接熔池形态的数值分析模型,分析中引入了热焓和表面分布高斯电弧热源模型,初步计算了焊接电流和焊接速度对焊接温度场分布的影响。通过比较焊缝有效熔深的测量结果和计算结果,验证了所建模型的正确性和可靠性。以计算结果为基础,对焊接工艺参数进行优化,建立了0Cr18Ni10Ti不锈钢放射源源壳的焊接工艺路线。     

Neutronic evaluation of coating and cladding materials for accident tolerant fuels

In severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding interaction. Most ATF concepts are expected to work within the design framework of current and future light water reactors, and for that reason they must match or exceed the performance of conventional fuel in normal conditions. This study analyzed the neutronic performance of ATF when employed in both pressurized and boiling water reactors. Two concepts were evaluated: (1) coating the exterior of zirconium-alloy cladding with thin ceramics to limit the zirconium available for reaction with high-temperature steam; (2) replacing zirconium alloys with alternative materials possessing slower oxidation kinetics and reduced hydrogen production. Findings show that ceramic coatings should remain 10–30 μm thick to limit the neutronic penalty. Alternative cladding materials, with the exception of SiC, enhance neutron loss compared to zirconium-alloys. An extensive parametric analysis concluded that reference performance metrics can be met by employing 300-μm or less thick cladding or increasing fuel enrichment by up to 1.74% depending on material and geometry.     

Influence of oxide layer morphology on hydrogen concentration in tin and niobium containing zirconium alloys after high temperature steam oxidation

The influence of the oxide layer morphology on the hydrogen uptake during steam oxidation of (Zr,Sn) and Zr-Nb nuclear fuel rod cladding alloys was investigated in isothermal separate-effect tests and large-scale fuel rod bundle simulation experiments. From both it can be concluded that the concentration of hydrogen in the remaining metal strongly depends on the existence of tangential cracks in the oxide layers formed by the tetragonal - monoclinic phase transition in the oxide, known as breakaway effect. In these cracks hydrogen is strongly enriched. It results in very local high hydrogen partial pressure at the oxide/metal interface and in an increase of the hydrogen concentration in the metal at local regions where such cracks in the oxide layer exist. Due to this effect the hydrogen uptake of the remaining zirconium alloy does not depend monotonically on temperature. Differences between (Zr,Sn) and Zr-Nb alloys are caused by differences in the hydrogen production due to different oxidation kinetics and in the crack forming phase transformation in the oxides as well as in the mechanical stability of the oxides.     

Quantification and local distribution of hydrogen within Zircaloy-4 PWR nuclear fuel cladding tubes at the nuclear microprobe of the Pierre Süe Laboratory from μ-ERDA

Hydrogen content and its distribution in in-core materials of nuclear plants are known to have a strong influence on their behaviour, especially on their mechanical properties but also on their corrosion resistance. This point has to be largely investigated in the case of the nuclear fuel cladding (Zr based alloys) of pressurized water reactors (PWR).Two situations have been considered here, with regards to the hydrogen content and its spatial distribution within the thickness of the tubes:

(1)

Irradiated fuel cladding tubes after a nominal period under working conditions in a PWR core.

(2)

Non-irradiated fuel cladding previously exposed to conditions representative of an hypothetical “loss of coolant accident” scenario (LOCA).

As far as micrometric distributions of H were required, μ-ERDA has been performed at the nuclear microprobe of the Pierre Süe Laboratory. This facility is fitted with two beam lines. In the first one, used for non-active sample analysis, the μ-ERDA configuration has been improved to reduce the limits of detection and the reliability of the results. The second one offers the unique feature of being dedicated to radioactive samples. We will present the nuclear microprobe and emphasize on the μ-ERDA configuration of the two beam lines. We will illustrate the performance of the setup by describing the results obtained for Zircaloy-4 cladding both on non-irradiated and irradiated samples.     

Deposition of gold nano-particles and nano-layers on polyethylene modified by plasma discharge and chemical treatment

Polyethylene (PE) was treated in Ar plasma discharge and then grafted from methanol solution of 1,2-ethanedithiol to enhance adhesion of gold nano-particles or sputtered gold layers. The modified PE samples were either immersed into freshly prepared colloid solution of Au nano-particles or covered by sputtered, 50 nm thick gold nano-layer. Properties of the plasma modified, dithiol grafted and gold coated PE were studied using XPS, UV-VIS, AFM, EPR, RBS methods and nanoindentation. It was shown that the plasma treatment results in degradation of polymer chain, creation of excessive free radicals and conjugated double bonds. After grafting with 1,2-ethanedithiol the concentration of free radicals declined but the concentration of double bonds remained unchanged. Plasma treatment changes PE surface morphology and increases surface roughness too. Another significant change in the surface morphology and roughness was observed after deposition of Au nano-particles. The presence of Au on the sample surface after the coating with Au nano-particles was proved by XPS and RBS methods. Nanoindentation measurements shown that the grafting of plasma activated PE surface with dithiol increases significantly adhesion of sputtered Au nano-layer.     

RNRA and neutron threshold analyses of thick lithium coatings deposited by sputter evaporation

Li coatings on various substrates have numerous applications: Boron neutron capture therapy, super conducting tokamak, etc.Unfortunately the main difficulty using Li is its reactivity in air and diffusion into metals. It is the only metal that reacts with nitrogen at room temperature and it tarnishes and oxidizes rapidly in air.In this work, we investigate how to profile thick Li layers (50 μm) deposited on SiO₂ substrates by a method based on plasma sputtering, involving both DC sputtering and evaporation simultaneously.A thick Li layer (≈10 μm) was covered with a thin stainless steel layer to prevent oxidation during transfer of the sample from the sputtering chamber and the accelerator. Li coatings were investigated by RNRA and neutron threshold reaction to obtain interdiffusion profiles of the different components and their concentration. The depth profile using the ⁷Li(p,γ)⁸Be^∗ resonance nuclear reaction occurring at 440 keV allows us to obtain Li concentration versus depth up to 50 μm.Preliminary results indicate that homogeneous Li layers can be obtained and protected against air, even though it diffuses into the encapsulated layers.     

Homogeneous near surface activity distribution by double energy activation for TLA

Thin layer activation (TLA) is a versatile tool for activating thin surface layers in order to study real-time the surface loss by wear, corrosion or erosion processes of the activated parts, without disassembling or stopping running mechanical structures or equipment. The research problem is the determination of the irradiation parameters to produce point-like or large area optimal activity-depth distribution in the sample. Different activity-depth profiles can be produced depending on the type of the investigated material and the nuclear reaction used. To produce activity that is independent of the depth up to a certain depth is desirable when the material removed from the surface by wear, corrosion or erosion can be collected completely. By applying dual energy irradiation the thickness of this quasi-constant activity layer can be increased or the deviation of the activity distribution from a constant value can be minimized. In the main, parts made of metals and alloys are suitable for direct activation, but by using secondary particle implantation the wear of other materials can also be studied in a surface range a few micrometers thick.In most practical cases activation of a point-like spot (several mm²) is enough to monitor the wear, corrosion or erosion, but for special problems relatively large surfaces areas of complicated spatial geometry need to be activated uniformly. Two ways are available for fulfilling this task, (1) production of large area beam spot or scanning the beam over the surface in question from the accelerator side, or (2) a programmed 3D movement of the sample from the target side. Taking into account the large variability of tasks occurring in practice, the latter method was chosen as the routine solution in our cyclotron laboratory.     

Radiation-damage in molybdenum-rhenium alloys for space reactor applications

Various Mo-Re alloys are attractive candidates for use as fuel cladding and core structural materials in spacecraft reactor applications. Molybdenum alloys with rhenium contents of 41-47.5% (wt%), in particular, have good creep resistance and ductility in both base metal and weldments. However, irradiation-induced changes such as transmutation and radiation-induced segregation could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to evaluate the performance of Mo-41Re and Mo-47.5Re after irradiation at space reactor relevant temperatures. Tensile specimens of Mo-41Re and Mo-47.5Re alloys were irradiated to ∼0.7 displacements per atom (dpa) at 1073, 1223, and 1373 K and ∼1.4 dpa at 1073 K in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Following irradiation, the specimens were strained to failure at a rate of 1 × 10⁻³ s⁻¹ in vacuum at the irradiation temperature. In addition, unirradiated specimens and specimens aged for 1100 h at each irradiation temperature were also tested. Fracture mode of the tensile specimens was determined. The tensile tests and fractography showed severe embrittlement and IG failure with increasing temperatures above 1100 K, even at the lowest fluence. This high temperature embrittlement is likely the result of irradiation-induced changes such as transmutation and radiation-induced segregation. These factors could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to examine the irradiation-induced degradation for these Mo-Re alloys under neutron irradiation.     

Mechanical properties at sub-microscale and macroscale of polycrystalline uranium mononitride

The indentation hardness, Vickers hardness, fracture toughness, and Young’s modulus of polycrystalline uranium mononitride (UN) at sub-microscale and macroscale were evaluated by an indentation test, Vickers hardness test, and the ultrasonic pulse echo method. The Young modulus and Vickers hardness were in good agreement with the literature values. The fracture toughness of UN was about three times that of UO₂. In addition, we revealed the indentation size effect on the indentation hardness of UN.     

Aging effects on microstructural and mechanical properties of select refractory metal alloys for space-reactor applications

Refractory alloys based on niobium, tantalum and molybdenum are potential candidate materials for structural applications in proposed space nuclear reactors. Long-term microstructural stability is a requirement of these materials for their use in this type of creep dominated application. Early work on refractory metal alloys has shown aging embrittlement occurring for some niobium and tantalum-base alloys at temperatures near 40% of their melting temperatures in either the base metal or in weldments. Other work has suggested microstructural instabilities during long-term creep testing, leading to decreased creep performance. This paper examines the effect of aging 1100 h at 1098, 1248 and 1398 K on the microstructural and mechanical properties of two niobium (Nb-1Zr and FS-85), tantalum (T-111 and ASTAR-811C) and molybdenum (Mo-41Re and Mo-47.5Re) base alloys. Changes in material properties are examined through mechanical tensile testing coupled with electrical resistivity changes and microstructural examination through optical and electron microscopy analysis.     

Quantitative elemental localisation in leaves and stems of nickel hyperaccumulating shrub Hybanthusfloribundus subsp. floribundus using micro-PIXE spectroscopy

Hybanthusfloribundus (Lindl.) F.Muell. subsp. floribundus is a native Australian nickel (Ni) hyperaccumulating shrub and a promising species for rehabilitation and phytoremediation of Ni tailings. Spatial localisation and quantification of Ni in leaf and stem tissues of H.floribundus subsp. floribundus was studied using micro-proton-induced X-ray emission (micro-PIXE) spectroscopy. Young plants, grown in a potting mix under controlled glasshouse conditions were exposed to Ni concentrations of 0 and 26 mM kg⁻¹ for 20 weeks. Leaf and stem samples were hand-sectioned and freeze-dried prior to micro-PIXE analysis. Elemental distribution maps of leaves revealed Ni concentration of 7800 mg kg⁻¹ dry weight (DW) in whole leaf sections, which was identical to the bulk tissue analysis. Elemental maps showed that Ni was preferentially localised in the adaxial epidermis (10,000 mg kg⁻¹ DW) and reached a maximum of up to 10,000 mg kg⁻¹ DW in the leaf margin. Freeze-dried stem sections from the same plants contained lower Ni than leaf tissues (1800 mg kg⁻¹ versus 7800 mg kg⁻¹ DW, respectively), however did not resolve a clear pattern of compartmentalisation across different anatomical regions. Our results suggest localisation in epidermal cells is an important physiological mechanism involved in Ni accumulation and tolerance in leaves of H.floribundus subsp. floribundus.