Pu Zr Research Articles

Pyrohydrolytic separation of fluoride and chloride in U–Zr and U–Pu–Zr metallic fuel alloys through oxidation accompanied by estimation using ion chromatography

Pyrohydrolytic separation of fluoride and chloride in U–Zr and U–Pu–Zr metallic fuel alloys through oxidation accompanied by estimation using ion chromatography

A comparison of the extraction behaviour of tris(2-methylbutyl) phosphate and tri-n-alkyl phosphates for the separation of metal ions for U–Zr and U–Pu–Zr systems by cross-current mode

Abstract Even though, pyroprocessing is considered as a suitable technique for metal fuel processing, attempts are being made in our laboratory to develop a solvent extraction based process as an interim method. Since, the metallic fuels contain considerable amount of Pu(IV) and Zr(IV), the use of conventional extractant tri-n-butyl phosphate (TBP) for reprocessing may pose problems due to third phase formation. In the exploration for the identification of an alternate extractant in the organophosphate family, tris(2-methylbutyl) phosphate (T2MBP), a branched isomer of tri-n-amyl phosphate (TAP) was found to be a potential extractant for nuclear fuel reprocessing. In this context, batch wise extraction and stripping studies with a feed solution containing U(VI) and 6 wt% Zr(IV) were carried out with unirradiated and irradiated 1.1 M solution of T2MBP in n-dodecane (n-DD) and the results were compared with corresponding solutions of TBP and TAP in n-DD. Among all these systems under identical conditions, third phase formation was observed only in the case of irradiated TBP system which makes U–Zr fuel reprocessing difficult using TBP as the extractant. Furthermore, studies have also been carried out with U–Pu–Zr feed solution to understand the extraction and stripping behaviour of these extractants. The stage wise and cumulative percentage of extraction and stripping for each metal ion of these systems were evaluated. Overall, these batch studies indicated that, T2MBP system has comparable extraction and stripping behaviour with U–Zr and U–Pu–Zr feed solutions and T2MBP exhibits better separation factor than TBP.

In search of θ-(Pu,Zr) in binary Pu–Zr: Thermal and microstructural analyses of Pu − 30Zr alloy

The existing Pu–Zr binary phase diagrams report the stability of the compound θ-(Pu,Zr) in the low temperature region between 300 and 0 °C. Furthermore, the current understanding is that θ-(Pu–Zr) is thermodynamically favored over the metastable δ-(Pu,Zr) phase. In an effort to shed light on the phases formed in Pu–Zr binary alloys and reduce uncertainties in the poorly defined boundary between the θ-(Pu,Zr), (θ+δ), δ-(Pu,Zr), and (θ+α-Zr) regions, Pu − 30Zr (in wt.%, equivalent 53 at.%) alloys were subjected to microstructural characterization, annealing, and differential scanning calorimetry (DSC). The results indicate that the alloy is composed of δ-(Pu,Zr) matrix, with a number of smaller, randomly distributed α-Zr precipitates. The phase transition temperatures (determined based on the DSC data) and phases identified in Pu − 30Zr alloys (based on crystallographic data) compare well to those predicted by the phase diagrams with the exception of the θ-(Pu–Zr) phase. Our data indicates that θ-(Pu–Zr) is metastable and can be observed only within a small temperature window (100–300 °C). The consecutive heating cycles remove θ-(Pu–Zr) from the system, and no traces of θ-(Pu–Zr) remain at room temperature, as evidenced by microstructural characterization. This calls for reevaluation of the binary Pu–Zr phase diagram, with particular attention paid to the existence of θ-(Pu–Zr) and its stability.

Basic Concepts of a Modular Fast Sodium Reactor with Metallic Fuel

The conceptual precepts of a modular fast sodium reactor with mixed uranium-plutonium zirconium-alloy metal fuel (U–Pu–Zr) related to Gen-IV are presented. The range of its parameters is determined, and the parameters are optimized in order to meet the characteristic requirements of the development of a nuclear power system in terms of safety, competitiveness, and resource availability. It was shown with the aid of computational studies that in a modular reactor, in spite of its small size, the systems requirements in terms of specific load and excess production are met and the minimum reactivity excess on burnup is realized. Loading metallic fuel into the core and thorium into the breeding zone and using fuel cycles compatible with thermal reactors it becomes much easier to solve the problems of supplying fuel for nuclear power and managing actinides and plutonium in the nuclear fuel cycle.

Evaluation of long chain monoamide extractants for the reprocessing of U–Pu–Zr metallic fuel solution

Long chain monoamide extractants, N,N-di-decyloctanamide(DDOA), N,N-di-hexyldecanamide(DHDA), N,N-di-2-ethylhexyloctanamide(D2EHOA) and N,N-dihexyl-2-ethylhexanamide(DH2EHA) were synthesized and studied for the recovery of U(VI), Pu(IV) and Zr(IV) from a simulated dissolver solution of un-irradiated U–Zr metallic fuel. The results were compared with the results of N,N-dihexyloctanamide(DHOA) and tri-n-butylphosphate(TBP) under similar conditions. Solvent extraction studies were carried out for comparing the extraction behavior of U(VI), Pu(IV) and Zr(IV) in monoamide extractants with TBP system. The influence of length and branching of alkyl chains on either side of the amidic group on the extraction efficiency, third phase behaviour and metal ion selectivity in long chain monoamides has been discussed based on the results of above studies.

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

The fast sodium reactor fuel assembly (FA) with U–Pu–Zr metallic fuel is described. In comparison with a “classical” fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.

Dissolution and characterisation studies on U–Zr and U–Pu–Zr alloys in nitric acid medium

Dissolution of metallic alloys, U–Zr and U–Pu–Zr has been investigated in HNO3 media. An electro oxidative dissolution technique was also examined. Explosive nature of metallic alloys during dissolution in nitric acid has been investigated. A method has been developed for the determination of zirconium in the presence of uranium and plutonium using a spectrophotometric technique. The influence of HNO3, uranium and plutonium during the estimation of Zr has been studied. Plutonium interferes in the estimation of Zr and it can be avoided by employing ascorbic acid. The method was employed for the estimation of Zr in samples generated during dissolution of Zr containing alloy fuels.

Studies Related to the Processing of U–Zr and U–Pu–Zr Metallic Fuels Using Tri-iso-amyl Phosphate (TiAP) as Extractant

ABSTRACTMetallic fuel is reprocessed by a compact nonaqueous pyrochemical process. The present study explores the possible uses of an aqueous reprocessing-based method as an alternative to the pyro-processing of metallic fuels. Solvent extraction studies in the batch mode were carried out for both U–Zr- and U–Pu–Zr-based metal alloy systems. Earlier studies carried out in our laboratory have established that Tri-iso-amyl Phosphate (TiAP) is a promising extractant for the reprocessing of spent fuels. In the present study, the distribution data has been generated for the extraction of uranium and zirconium as a function of equilibrium aqueous phase metal ion and nitric acid concentration with TiAP and Tri-n-butyl Phosphate (TBP)-based solvents. These studies indicate the formation of a third phase with zirconium in the presence of uranium with TBP under certain experimental conditions whereas it was not encountered with the TiAP system. Flow sheet for the co-extraction and co-stripping of heavy metal ions by 1.1M TiAP and 1.1M TBP in n-dodecane from U–Zr as well as U–Pu–Zr feed solutions in stage-wise mode has been evaluated. Percentage extraction and stripping of metal ions were calculated stage-wise and the results are discussed.

XANES analysis of a Cm-doped borosilicate glass under $$\alpha $$ α -self-irradiation effects

Industrial borosilicate glasses containing fission products and minor actinides can be subjected to structural damage caused mainly by \(\alpha \)-self-irradiation effects. In this field of glasses under extreme conditions, we present an X-ray Absorption Near-Edge Structure investigation of two six-oxide borosilicate curium-doped glasses (based on the International Simplified Glass (ISG) composition). The first sample is an 8-year ISG damaged glass, which has already accumulated an \(\alpha \)-decay dose greater than 6.10\(^{18}\) \(\alpha \) g\(^{-1}\), a value corresponding to a damaged but stabilized structural state. The second sample results from annealing of the latter ISG damaged glass. Three species, Cm, Pu and Zr were probed at \(L_{3}\)-edge, \(L_{3}\)-edge and K-edge, respectively. From the experimental results, Cm and Pu species appear respectively in +3 and +4 oxidation states in both glasses. No Cm local environment changes are observed. In contrast, a small variation in Pu local environment appears between the damaged and annealed glasses, reflecting a possible coordination variation or Pu–Zr substitution. A more drastic effect appears for Zr local environment, where a sevenfold coordinated site grows over time under \(\alpha \)-self-irradiation effects, at the expense of the initial major sixfold site of symmetry. Moreover, annealing the damaged glass does not permit to retrieve a similar structural state to the one of a just melted curium-doped ISG.

TEM identification of subsurface phases in ternary U–Pu–Zr fuel

Phases and microstructure in as-cast, annealed at 850 °C, and subsequently cooled U–23Pu–9Zr fuel were characterized using scanning and transmission electron microscopy techniques. SEM examination shows formation of three phases in the alloy that were identified in TEM using selective area diffraction pattern analysis: α-Zr globular and elongated δ-UZr2 inclusions and a thick oxide layer formed on top of β-Pu phase, which has been initially assumed to be ζ-(U, Pu). However, further examination of the cross-sectional TEM specimens identified the matrix phases as δ-UZr2, β-Pu, and (U, Zr)ht. Two types of inclusions were observed in the immediate vicinity of the specimen surface and they were consistent with α-Zr and ζ-(U, Pu).

Microstructural investigation of as-cast uranium rich U–Zr alloys

The present study evaluates the microstructure in as-cast uranium rich U–Zr alloys, an important subsystem of U–Pu–Zr ternary metallic nuclear reactor fuel, as a function of the Zr content, from 2wt.% to 15wt.%Zr. It has been previously suggested that the unique intermetallic compound δ phase in U–Zr alloys is only present in as-cast U–Zr alloys with a Zr content exceeding 10wt.%Zr. However, our analysis of transmission electron microscopy (TEM) data shows that the δ phase is common to all as-cast alloys studied in this work. Furthermore, specific coherent orientation relationship is found between the α and δ phases, consistent with previous findings, and a third variant is discovered in this paper.

TEM examination of phases formed between U–Pu–Zr fuel and Fe

Exposure to high temperatures and irradiation results in interaction and interdiffusion between fuel and cladding constituents that can lead to formation of undesirable brittle or low-melting point phases. A diffusion couple study has been conducted to understand fuel-cladding interaction occurring between U–22Pu–4Zr (in wt%) fuel and pure Fe at elevated temperatures. The phases formed within fuel cladding chemical interaction (FCCI) layer have been characterized in the transmission electron microscope (TEM). The phases formed within FCCI layer have been identified as Fe2U (Fd-3m), FeU6 (I4/mcm), Fe2Zr (Fd-3m), FeZr2 (I4/mcm), Fe2Pu (Fd-3m), UZr2 (P6/mmm), β-Zr (Im-3m), and ZrO2 (Fm-3m).

Atomistic modeling of high temperature uranium–zirconium alloy structure and thermodynamics

A semi-empirical Modified Embedded Atom Method (MEAM) potential is developed for application to the high temperature body-centered-cubic uranium–zirconium alloy (γ-U–Zr) phase and employed with molecular dynamics (MD) simulations to investigate the high temperature thermo-physical properties of U–Zr alloys. Uranium-rich U–Zr alloys (e.g. U–10Zr) have been tested and qualified for use as metallic nuclear fuel in U.S. fast reactors such as the Integral Fast Reactor and the Experimental Breeder Reactors, and are a common sub-system of ternary metallic alloys like U–Pu–Zr and U–Zr–Nb. The potential was constructed to ensure that basic properties (e.g., elastic constants, bulk modulus, and formation energies) were in agreement with first principles calculations and experimental results. After which, slight adjustments were made to the potential to fit the known thermal properties and thermodynamics of the system. The potentials successfully reproduce the experimental melting point, enthalpy of fusion, volume change upon melting, thermal expansion, and the heat capacity of pure U and Zr. Simulations of the U–Zr system are found to be in good agreement with experimental thermal expansion values, Vegard's law for the lattice constants, and the experimental enthalpy of mixing. This is the first simulation to reproduce the experimental thermodynamics of the high temperature γ-U–Zr metallic alloy system. The MEAM potential is then used to explore thermodynamics properties of the high temperature U–Zr system including the constant volume heat capacity, isothermal compressibility, adiabatic index, and the Grüneisen parameters.

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.

Thermal conductivity of U–20 wt.%Pu–2 wt.%Am–10 wt.%Zr alloy

The authors fabricated U–20wt.%Pu–2wt.%Am–10wt.%Zr alloys and evaluated the heat capacity and thermal conductivity. The heat capacity was measured between 335 and 827K by a drop calorimetry. The thermal diffusivity was measured between 300 and 1073K by a laser flash method. The thermal conductivity was evaluated from the measured thermal diffusivity and heat capacity. The thermal conductivity of U–20wt.%Pu–2wt.%Am–10wt.%Zr alloy was slightly higher than literature values of U–Pu–Zr and U–Pu–MA–Zr alloys. The thermal conductivity was formulated for use in practical design for Generation IV fast reactors.

Neutronics studies on the feasibility of developing fast breeder reactor with flexible breeding ratio

This paper investigates the feasibility of designing a flexible fast breeder reactor from the view of neutronics. It requires that the variable breeding ratio can be achieved in operating a fast reactor without significant changes of the core, including the minimum change of fuel assembly design, the minimum change of the core configuration and the same control system arrangement in the core. The sodium cooled fast reactor is investigated. Two difficulties are overcome: (1) the different excess reactivity is well controlled for different cores, especially for the one with small breeding ratio; (2) the maximum linear power density is well controlled while the breeding ratio changes. The optimizations are done to meet the requirements. The U–Pu–Zr alloy is applied to enhance the breeding. The enrichment-zoning technique with unfixed blanket assembly loading position is searched to get acceptable power distributions when the breeding ratio changes. And the control system is designed redundantly to fulfill the control needs. Then, the achieved breeding ratio can be adjusted from 1.1 to 1.4. The reactivity coefficients, temperature distributions and preliminary safety performances are evaluated to investigate the feasibility of the new concept. All the results show that it is feasible to develop the fast reactor with flexible breeding ratios, although it still highly relies on the advancement of the coolant flow control technology.

Modeling constituent redistribution in U–Pu–Zr metallic fuel using the advanced fuel performance code BISON

An improved robust formulation for constituent distribution in metallic nuclear fuels is developed and implemented into the advanced fuel performance framework BISON. The coupled thermal diffusion equations are solved simultaneously to reanalyze the constituent redistribution in post irradiation data from fuel tests performed in Experimental Breeder Reactor-II (EBR-II). Deficiencies observed in previously published formulation and numerical implementations are also improved. The present model corrects an inconsistency between the enthalpies of solution and the solubility limit curves of the phase diagram while also adding an artificial diffusion term when in the 2-phase regime that stabilizes the standard Galerkin finite element (FE) method used by BISON. An additional improvement is in the formulation of zirconium flux as it relates to the Soret term. With these new modifications, phase dependent diffusion coefficients are revaluated and compared with the previously recommended values.The model validation included testing against experimental data from fuel pins T179, DP16 and T459, irradiated in EBR-II. A series of viable material properties for U–Pu–Zr based materials was determined through a sensitivity study, which resulted in three cases with differing parameters that showed strong agreement with one set of experimental data, rod T179. Subsequently a full-scale simulation of T179 was performed to reduce uncertainties, particularly relating to the temperature boundary condition for the fuel. In addition a new thermal conductivity model combining all available data covering 0–100% zirconium concentration and a zirconium concentration dependent linear heat rate solution derived from Monte Carlo N-Particle (MCNP) simulations were developed. An iterative calibration process was applied to obtain optimized diffusion coefficients for U–Pu–Zr metallic fuels. Optimized diffusion coefficients suggest relative improvements in comparison to previous reported values. The most influential or uncertain phase is found to be the gamma phase, followed by alpha phase, and thirdly the beta phase; indicating separate effect testing should concentrate on these phases.

High- and low-Am RE inclusion phases in a U–Np–Pu–Am–Zr alloy

Structural, microstructural, and microchemical data were collected from rare-earth inclusions in an as-cast U–Pu–Zr alloy with ∼3at.% Am, 2% Np, and 9% rare-earth elements (La, Ce, Pr, and Nd). Two RE phases with different concentrations of Am were identified.The composition of high-Am RE inclusions is ∼2–5at.% La, 15–20% Ce, 5–10% Pr, 25–45% Nd, 1% Np, 5–10% Pu, and 10–20% Am. Some areas also have O, although this does not appear to be an essential part of the high-Am RE phase. The inclusions have a face-centered cubic structure with a lattice parameter a∼0.54nm.The composition of the only low-Am RE inclusion studied in detail is ∼35–40at.% O, 40–45% Nd, 1–2% Zr, 4–5% La, 9–10% Ce, and 6–7% Pr. This inclusion is an oxide with a crystal structure similar to the room-temperature structure of Nd2O3. Microstructural features suggest that oxidation occurred during casting, and that early crystallization of high-temperature oxides led to formation of two distinct RE phases.

Am phases in the matrix of a U–Pu–Zr alloy with Np, Am, and rare-earth elements

Phases and microstructures in the matrix of an as-cast U–Pu–Zr alloy with 3wt% Am, 2% Np, and 8% rare-earth elements were characterized by scanning and transmission electron microscopy. The matrix consists primarily of two phases, both of which contain Am: ζ-(U, Np, Pu, Am) (∼70at.% U, 5% Np, 14% Pu, 1% Am, and 10% Zr) and δ-(U, Np, Pu, Am)Zr2 (∼25% U, 2% Np, 10–15% Pu, 1–2% Am, and 55–60at.% Zr). These phases are similar to those in U–Pu–Zr alloys, although the Zr content in ζ-(U, Np, Pu, Am) is higher than that in ζ-(U, Pu) and the Zr content in δ-(U, Np, Pu, Am)Zr2 is lower than that in δ-UZr2. Nanocrystalline actinide oxides with structures similar to UO2 occurred in some areas, but may have formed by reactions with the atmosphere during sample handling.Planar features consisting of a central zone of ζ-(U, Np, Pu, Am) bracketed by zones of δ-(U, Np, Pu, Am)Zr2 bound irregular polygons ranging in size from a few micrometers to a few tens of micrometers across. The rest of the matrix consists of elongated domains of ζ-(U, Np, Pu, Am) and δ-(U, Np, Pu, Am)Zr2. Each of these domains is a few tens of nanometers across and a few hundred nanometers long. The domains display strong preferred orientations involving areas a few hundred nanometers to a few micrometers across.

Thermal conductivities of actinides (U, Pu, Np, Cm, Am) and uranium-alloys (U–Zr, U–Pu–Zr and U–Pu–TRU–Zr)

The thermal conductivity correlations of actinides (U, Pu, Np, Cm, Am) and alloys of U–Zr, U–Pu–Zr and U–Pu–TRU–Zr were developed for the use in TRU burning reactors as a function of temperature and alloy composition, using the available data in the literature and extrapolating information obtained in the literature. Because of the scarcity in the measured data for thermal conductivities of TRU elements in the literature, estimations for the thermal conductivities of Pu, Am, Np, and Cm were made. A correlation for U–Zr alloy was also developed and extended to U–Pu–Zr and U–Pu–TRU–Zr alloys.