Phase characteristics of a number of U-Pu-Am-Np-Zr metallic alloys for use as fast reactor fuels

Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. A series of test designs for the Advanced Fuel Cycle Initiative (AFCI) have been irradiated in the Advanced Test Reactor (ATR), designated as the AFC-1 and AFC-2 designs. Metal fuel compositions in these designs have included varying amounts of U, Pu, Zr, and minor actinides (Am, Np). Investigations into the phase behavior and relationships based on the alloy constituents have been conducted using X-ray diffraction and differential thermal analysis. Results of these investigations, along with proposed relationships between observed behavior and alloy composition, are provided. In general, observed behaviors can be predicted by a ternary U-Pu-Zr phase diagram, with transition temperatures being most dependent on U content. Furthermore, the enthalpy associated with transitions is strongly dependent on the as-cast microstructural characteristics.     

Thermal conductivity of rare earth-uranium ternary oxides of the type RE6UO12

The knowledge of thermophysical properties of the rare earth uranium ternary oxides of the type RE₆UO₁₂ (RE=La, Gd and Dy) is essential to understand the fuel performance during reactor operation and for modeling fuel behavior. Literature on the high temperature properties of this compound is not available and there is no report at all on the thermal conductivity of these compounds. Hence a study of thermal conductivity of this compound has been taken up. The compounds were synthesized by a solution combustion method using metal nitrates and urea. Thermal diffusivity of these compounds was measured by the laser flash method in the temperature range 673-1373 K. The specific heat data was computed using Neumann-Kopp’s law. Thermal conductivity was calculated using the measured thermal diffusivity value, density and specific heat data for different temperatures. The temperature dependence of thermal conductivity and the implication of structural aspects of these compounds on the data are discussed here.     

Development of Fast Reactor Metal Fuels Containing Minor Actinides

Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ～10 at% in the Phénix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU.     

The possible influence of order-disorder transformations on the swelling of 85Ni-15Cr nichrome alloy

The binary alloy 85Ni-15Cr swells during neutron irradiation in a manner that is quite unrepresentative of either simple austenitic alloys or pure nickel. The nickel-chromium system is known to exhibit ordering out of reactor and this alloy was used to test the concept that alloys which develop order exhibit lower swelling rates. When the neutron irradiation data are compared to that of ion irradiations conducted by Hudson and Ashby on Ni-Cr alloys, it is shown that chromium additions appear to depress swelling in nickel initially but also appear to suppress the tendency for swelling to saturate at high exposure. Below a transition temperature near 550°C the swelling is relatively sluggish and is quite insensitive to irradiation temperature. Above the transition temperature the swelling behavior is more complex but typical of austenitic alloys. The swelling transition temperature is thought to be related to the critical temperature for order-disorder transformation.     

Application of electrochemical techniques in pyrochemical processes - Electrochemical behaviour of rare earths at W, Cd, Bi and Al electrodes

The electrochemical behaviour of some rare earths ions (REs) - from the light to heavy lanthanides (i.e. Ce, La, Pr, Gd, Er, Ho) and Y - were investigated in the eutectic LiCl-KCl at different substrates: (i) liquid metals Cd and Bi, (ii) aluminium, and (iii) tungsten. The electrode reaction of the RE(III)/RE couples at the Cd and Bi pool electrodes was elucidated by cyclic voltammetry. The differences between the equilibrium potential adopted by a RE electrode and the observed with the same RE(III) solution at the liquid electrodes were consistent with the activity coefficients of RE in the liquid metal phase. The relative partial molar Gibbs energies and activities of RE in the RE-Cd and RE-Bi intermetallic compounds could be estimated by the analysis of the open circuit chronopotentiograms using Cd and Bi coated tungsten electrodes. The Gibbs energies of formation of different intermetallic compounds, as well as their molar entropies and enthalpies of formation were also calculated from the temperature dependence of the emf. The redox potential of the RE(III)/RE couples at the Al electrode was observed at more positive potentials than that at the inert electrode (W). This potential shift was explained by a lowering of the activity of the REs in the Al phases due to the formation of intermetallic compounds. Electromotive force measurements for various intermetallic compounds in two-phase coexisting states were carried out. The activities and relative partial molar Gibbs energies of REs were also obtained. Moreover, the molar entropies and enthalpies of the aluminium-rich alloys were also calculated from the temperature dependence of the emf measurements.     

In situ XRD analysis of the oxide layers formed by oxidation at 743 K on Zircaloy 4 and Zr–1NbO

Two alloys, having different oxidation behaviour (Zy4 and Zr–1NbO), have been investigated during oxidation at high temperature (743 K) and low oxygen pressure (10 kPa) by in situ X-ray diffraction (XRD). Tetragonal phase content and ‘pseudo-stresses’ on the monoclinic phase have been measured as a function of the oxide layer thickness. The tetragonal phase contents are similar for both alloys and decreased with the oxide layer thickness. Pseudo-stresses were much more compressive on Zr–1NbO alloy, with limited changes at the corrosion kinetics transition. On cooling, the tetragonal fractions do not change, while ‘pseudo-stresses’ decreased in different ways for the two alloys. With respect to stress analysis, no correlation was found between ‘pseudo-stresses’ and tetragonal phase content. In addition, due to the thermoelastic properties of the highly anisotropic phases of the zirconia, large internal thermal stresses are expected to develop during any temperature changes. The orders of magnitude of them are similar to the stresses induced by swelling during oxidation from Zr to ZrO₂.     

Effects of oxygen, nitrogen and carbon additions on the mechanical properties of vanadium and V/Mo alloys

Effects of oxygen, nitrogen and carbon additions on the mechanical properties at room temperature of vanadium and V-Mo alloys containing up to 25 at% molybdenum were studied. The V-O, N or C and V-Mo-O, N or C alloy systems, respectively, were prepared by heating vanadium and the V-Mo alloys with oxygen, nitrogen or propane gas in sealed quartz capsules. The prepared alloys were homogenized, argon-quenched and analyzed for oxygen, nitrogen and carbon content. Then they were examined by tensile and hardness tests at room temperature, SEM observations and TEM studies. Oxygen and nitrogen additions to vanadium and the V-Mo alloys raise the hardness by solid solution, with nitrogen being more effective. Carbon additions to vanadium form coarse precipatates of V₂C along the grain boundaries, which do not raise the hardness, while those to the V-Mo alloys form fine precipitates of vanadium carbide, homogeneously distributed in the matrix, which raise the hardness. Threshold concentrations of oxygen, nitrogen and carbon for embrittlement (fracture without plastic deformation) of these alloy systems decrease with increasing molybdenum concentration, those of carbon being most detrimental. Also, it was shown that the embrittlement of the V-Mo alloys after exposure to liquid sodium may be explained not only by solid solution hardening due to absorbed oxygen but also by combined effects of absorbed oxygen, nitrogen and carbon.     

Mechanical and thermomechanical properties of commercially pure chromium and chromium alloys

The mechanical and thermal properties of commercially pure chromium and the chromium-based alloys Cr–5Fe–1Y₂O₃ and Cr–44Fe–5Al–0.3Ti–0.5Y₂O₃ have been investigated in order to determine the thermal stress factor of these materials and to assess their capability to withstand high-thermal loads in fusion applications. Especially the alloy Cr–5Fe–1Y₂O₃ combines sufficient mechanical strength at temperatures up to 1000 °C, high-thermal conductivity and a low-thermal expansion coefficient to yield the lowest thermal stress factor of all metallic candidate materials for first wall and blanket applications. The high-ductile-to-brittle transition temperature may lead to a rather high value for the lower operation-temperature limit.     

High temperature fracture characteristics of a nanostructured ferritic alloy (NFA)

The nanostructured ferritic alloys (NFAs) have been developed to improve high temperature strength and radiation resistance by refining grains and including nanoclusters. Among the key properties of NFAs needed to be assessed for advanced reactor applications the cracking resistance at high temperatures has not been well known. In this work, therefore, the high temperature fracture behavior has been investigated for the latest nanostructured ferritic alloy 14YWT (SM10). The fracture toughness of the alloy was above 140 MPa √m at low temperatures, room temperature (RT) and 200 °C, but decreased to a low fracture toughness range of 52-82 MPa √m at higher temperatures up to 700 °C. This behavior was explained by the fractography results indicating that the unique nanostructure of 14YWT alloy produced shallow plasticity layers at high temperatures and a low-ductility grain boundary debonding occurred at 700 °C. The discussion also proposes methods to improve resistance to cracking.     

Minor actinide transmutation in fast reactor metal fuels irradiated for 120 and 360 equivalent full-power days

An irradiation experiment on uranium–plutonium–zirconium (U–Pu–Zr) alloys containing 5 wt% or less minor actinides (MAs) and rare earths was carried out in the Phénix fast reactor. The isotope compositions of the fuel alloys irradiated for 120 and 360 equivalent full-power days (EFPDs) were chemically analyzed by inductively coupled plasma–mass spectrometry after 3.3–5.3 years of cooling. The results of chemical analysis indicated that the discharged burnups of the fuel alloys irradiated for 120 and 360 EFPDs were 2.1–2.5 and 5.3–6.4 at%, respectively. The changes in the isotopic abundances of plutonium, americium, and curium during the irradiation experiment were assessed to discuss the transmutation performance of MA nuclides added to U–Pu–Zr alloy fuel. Multigroup three-dimensional diffusion and burnup calculations accurately predicted the changes in these isotopic abundances after fuel fabrication. An evaluation of the MA transmutation ratio based on the results of chemical analysis revealed that the quantity of MA elements in the U–19Pu–10Zr–5MA (wt%) alloy decreased by about 20% during the irradiation experiment for 360 EFPDs.     

稀土元素Gd对2308双相不锈钢热变形行为的影响

为解决含Gd双相不锈钢热加工不足问题,本文以含2%Gd的双相不锈钢为研究对象,在不同温度下开展热模拟压缩实验,研究含Gd双相不锈钢热变形行为及组织演变。利用Gleeble-1500D热模拟试验机对含Gd双相不锈钢进行变形量为50%的单道次热变形试验。根据真应力-真应变曲线计算了该合金的热变形激活能Q_d,建立本构方程。同时对热变形后的组织进行了分析,探究稀土元素Gd对含Gd双相不锈钢热变形行为的影响,结果表明,在热变形过程中,合金的动态软化机制主要为动态再结晶。合金包含两种含Gd析出相,即条带状的脆性析出相M₃Gd相和M₁₇Gd₂相（M=Fe、Cr、Ni）,均为六方结构。当变形温度为1 050 ℃时,脆性M3Gd相破坏了基体的连续性,无法与基体协同变形,降低了合金的热塑性,导致合金在热变形过程中出现沿晶开裂。含Gd双相不锈钢适宜的热加工工艺区间的应变速率为0.01~0.1 s^-1,变形温度为950~1 000 ℃。     

中子辐照对耐事故燃料FeCrAl合金力学性能的影响研究

FeCrAl合金具有良好的抗高温氧化和力学性能,能够作为燃料包壳材料。为研究FeCrAl合金的辐照力学性能,开展了不同元素成分含量和2×1019 cm?2、8×1019 cm?2 2种中子注量辐照下的FeCrAl合金力学性能试验,并在室温和380℃下测试了FeCrAl合金的拉伸性能,获得了不同Cr和Al含量FeCrAl合金的抗拉强度和屈服强度,并研究了Al含量、Cr/Al含量配比及中子辐照对FeCrAl合金力学性能的影响。研究表明,FeCrAl合金强度随着Al含量增加大致呈增加趋势;经2×1019 cm?2中子辐照后,FeCrAl合金强度有较大提升;再经8×1019 cm?2中子辐照后,FeCrAl合金强度升高不明显。该研究结果为耐事故燃料（ATF）包壳材料的研发选型提供了重要的数据支撑。      

Zirconium alloys matrix as innovative material for composite fuel

A novel class of zirconium alloys is suggested as fuel matrix. They are “deep” ternary or quaternary eutectics having relatively low melting point i.e. from 963 to 1133 K in comparison with pure zirconium and intended for use as a matrix of dispersion high uranium content fuel (CERMET and METMET) particularly for thermal reactors. For fast reactors and MA burning Zr–Ti based alloys are proposed that have resistant metallurgical bonds between fuel and steel cladding. Investigations have been carried out on the structure and properties of the alloys as well as the specific technologies of their fabrication, in particular via induction furnace melting. The alloys may be also produced in the amorphous state as granules and strips. It is shown that thanks to their capillary properties they might be applied in brazing dissimilar materials. Based on novel zirconium matrix alloys high uranium content fuel compositions were produced. They have high thermal conductivity and compatibility as well as 25–50% higher uranium content than for VVER and PWR fuels.Fuel pins are fabricated by capillary impregnation method. The use of dispersion fuel with novel zirconium matrix alloys improves neutronics characteristics of reactor cores and might lead to extension of burn-up, low operating temperatures and serviceability under transient conditions.     

Crystalline phases in a devitrified simulated high-level waste glass containing the elements of the platinum group

A borosilicate glass containing 20 wt% simulated high-level waste oxides was subjected to heat treatment at 700°C for 1000 h. Seven crystalline phases were newly formed by the treatment in addition to two phases, (Ru, Rh)O₂ and (Pd, Rh, Te), which had already existed in the as-prepared simulated high-level waste glass. Among the new seven phases, five phases were certainly identified to be (RE)BSiO₅, CeO₂, SiO₂, (RE)PO₄ and (Sr, Ba, RE)MoO₄. Of two unidentified phases, one was rich in silicon, chromium and rare earth elements (RE), and the other was rich in nickel and chromium.The crystalline phases of the elements of the platinum group facilitated the occurrence of the other phases and suppressed crystal growth.     

Hf对ODS-FeCrAl合金微观组织及性能的影响

为研究少量添加的Hf对ODS-FeCrAl合金微观组织及性能的影响，通过机械合金化和后续的热等静压工艺研制两种14Cr ODS FeCrAl合金，对其进行微观结构、力学性能、1 200 ℃空气抗氧化性能和1 200 ℃水蒸气抗腐蚀性能测试。TEM结果表明，两种合金中YAlO3氧化物析出相细小弥散，而Hf的添加促进合金基体中生成颗粒尺寸更加细小的Y2Hf2O7纳米氧化物。由于纳米氧化物Orowan的强化机制，两种合金均具有优异的室温与高温力学性能。XRD和SEM结果表明，两种合金经过1 200 ℃空气氧化与1 200 ℃水蒸气腐蚀后，表面均形成均匀致密的α Al2O3膜，对合金起保护作用。Hf添加有利于合金在1 200 ℃水蒸气腐蚀条件下形成更加均匀的α Al2O3膜，抑制Al3+的向外扩散，避免合金内部晶界处形成空穴。     

Determination of Heat Capacity of Simulated Radioactive Waste Glasses by Drop Calorimetry

The enthalpies of two kinds of simulated radioactive waste glasses have been measured with an ice isothermal calorimeter at temperatures of 424~875 K by means of drop calorimetry. The fitting functions for the enthalpies per gram have been determined by the least squares fitting. Specific heat and average molar heat capacity have been obtained. It is likely that heat capacity of waste glass remarkably depends on its composition especially above its glass transition temperature.     

Nickel-based Gadolinium Alloy for Neutron Adsorption Application in Ram Packages

Abstract

The National Spent Nuclear Fuel Program, located at the Idaho National Laboratory (INL), coordinates and integrates national efforts in management and disposal of US Department of Energy (DOE)-owned spent nuclear fuel. These management functions include development of standardised systems for long-term disposal in the proposed Yucca Mountain repository. Nuclear criticality control measures are needed in these systems to avoid restrictive fissile loading limits because of the enrichment and total quantity of fissile material in some types of the DOE spent nuclear fuel. This need is being addressed by development of corrosion-resistant, neutron-absorbing structural alloys for nuclear criticality control. This paper outlines results of a metallurgical development programme that is investigating the alloying of gadolinium into a nickel–chromium–molybdenum alloy matrix. Gadolinium has been chosen as the neutron absorption alloying element due to its high thermal neutronabsorption cross section and low solubility in the expected repository environment. The nickel–chromium–molybdenum alloy family was chosen for its known corrosion performance, mechanical properties, and weldability. The workflow of this programme includes chemical composition definition, primary and secondary melting studies, ingot conversion processes, properties testing, and national consensus codes and standards work. The microstructural investigation of these alloys shows that the gadolinium addition is present in the alloy as a gadolinium-rich second phase. The mechanical strength values are similar to those expected for commercial Ni–Cr–Mo alloys. The alloys have been corrosion tested with acceptable results. The initial results of weldability tests have also been acceptable. Neutronic testing in a moderated critical array has generated favourable results. An American Society for Testing and Materials material specification has been issued for the alloy and a Code Case has been submitted to the American Society of Mechanical Engineers for code qualification.     

Delayed hydride cracking of spent fuel rods in dry storage

Failures of zirconium alloy cladding tubes during a long-term storage at room temperature were first reported by Simpson and Ells in 1974, which remains unresolved by the old delayed hydride cracking (DHC) models. Using our new DHC model, we examined failures of cladding tubes after their storage at room temperature. Stress-induced hydride phase transformation from γ to δ at a crack tip creates a difference in hydrogen concentration between the bulk region and the crack tip due to a higher hydrogen solubility of the γ-hydride, which is a driving force for DHC at low temperatures. Accounting for our new DHC model and the failures of zirconium alloy cladding tubes during long-term storage at room temperature, we suggest that the spent fuel rods to be stored either in an isothermal condition or in a slow cooling condition would fail by DHC during their dry storage upon cooling to below 180 °C. Further works are recommended to establish DHC failure criterion for the spent fuel rods that are being stored in dry storage.     

Elastic anisotropy of zirconium alloy fuel cladding

Some methods to determine the anisotropic elasticity coefficients of zirconium alloy fuel cladding are discussed together with the conventional elastic constants.A simplified method, which uses the f parameters, was proposed, and the validity and applicability of the method were also investigated. The integration method, which was originally proposed by Rosenbaum et al., was found to be in excellent agreement with the experimental values of our own twisting test from room temperature to 800°C. The proposed f parameter method was also found to agree well with the values obtained by the integration method or experiment, especially at high temperatures near 700°C. It became evident that the elastic property of the typical fuel cladding was roughly isotropic at room temperature, and that the elastic anisotropy monotonically increased with temperature. Some stress or strain distributions of the fuel cladding were also obtained using anisotropic elasticity constants. The stress induced in the fuel cladding with simulated ridge deformation was very little affected by the difference in texture, but was more influenced by the elastic constants employed.     

Reactions of Uranium-Plutonium Alloys with Iron

In metallic U-Pu-Zr fuel for fast reactors, metallurgical reactions occur between the fuel alloy and the stainless steel cladding, and a liquid phase may be formed in the reaction zone at a higher temperature. In order to clarify the condition for liquefaction at the fuel-cladding interface, the reactions of U-Pu alloys with Fe have been examined at 923 and 943 K. The test results confirmed that the liquid phase is not formed at 923 K in any region of the reaction zone when the maximum Pu content in the (U,Pu)₆Fe phase is less than the Pu solubility limit in this phase. Comparison of the present test results with the liquefaction data from the various tests on metallic fuel-cladding compatibility suggested that the liquefaction condition is independent of the Zr content in the fuel alloy and can be expressed as a function of the atom fraction ratio of Pu/(U+Pu) in the fuel alloy and the reaction temperature. At 923 K, liquefaction will occur when the Pu/(U+Pu) ratio is larger than 0.25.