Pu Zr Research Articles

Physics aspects of metal fuelled fast reactors with thorium blanket

Metal fuelled fast breeder reactors (MFBR) with high breeding ratio will play a major role in meeting the high nuclear power growth envisaged in India. In this regard several conceptual reactor designs with alloys of U–Pu–Zr fuel have been suggested for commercial operations. This study focusses on the physics design aspects of a sodium cooled U–Pu–6%Zr fuelled 1000MWe fast breeder reactor, which can attain a breeding ratio of nearly 1.5. The calculation results on reactor kinetics and safety parameters of the 1000MWe MFBR are presented. The changes in the breeding ratio by introduction of thorium in the blankets of the MFBR are also investigated. Burnup analyses are carried out to compare the core burnup effects in MOX and metal fuelled FBRs. Since the MOX fuelled 500MWe prototype fast breeder is getting constructed at IGCAR, for burnup comparisons a MFBR of similar design is considered. The results of this study indicate that the loss of reactivity in the metal core with burnup is less than half that of a MOX core and its breeding ratio remains nearly constant. It is also found that the isotopic composition of plutonium (Pu-vector composition) remains more steady with burnup in a metal core.

Separate effects identification via casting process modeling for experimental measurement of U–Pu–Zr alloys

Computational simulations of gravity casting processes for metallic U–Pu–Zr nuclear fuel rods have been performed using a design-of-experiments technique to determine the fluid flow, liquid heat transfer, and solid heat transfer parameters which most strongly influence the process solidification speed and fuel rod porosity. The results are used to make recommendations for the best investment of experimental time and effort to measure process parameters.

Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660°C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.

Extended fuel swelling models and ultra high burn-up fuel behavior of U–Pu–Zr metallic fuel using FEAST-METAL

Computational models in FEAST-METAL U–Pu–Zr metallic fuel behavior code have been upgraded to improve fission gas, solid fission product swelling, and pore sintering behavior in a microstructure dependent form. First, fission gas bubble growth is modeled by selecting small and large bubble groups according to a fixed number of gas atoms per bubble group. Small bubbles nucleated at phase boundaries grow via gas migration and turn into large bubbles. Furthermore, bubble morphology for each phase structure is captured by selecting the number of atoms per bubble and the shape of the bubbles in a phase dependent form. The gas diffusion coefficients for the single gamma phase and effective dual (α+δ) and (β+γ) phase structures are modeled separately, using the activation energy of the corresponding phase structure. In this study, it is found that pressure sintering of the interconnected porosity in dual phases should be less effective than the reference model in order to match clad strain and fission gas release behavior. In addition to these improvements, a probabilistic approach is taken to verify the fission gas-swelling threshold at which interconnected porosity begins. This fracture problem is treated as a function of critical crack length formed via bubble coalescence. It was found that a 10% gas-swelling threshold is appropriate for a wide range of gas bubble sizes. The new version of FEAST-METAL predicts the burn-up, smear density, and axial variation of the clad hoop strain and fission gas release behavior satisfactorily for selected test pins under EBR-II conditions. The code is used to predict ultra-high burn-up U–Pu–6Zr vented metallic fuel behavior with HT9 cladding for fast breeder reactor applications for 20 years of irradiation. It is assumed that HT9 clad will retain its fracture toughness and creep properties for this simulation. It appears that the increased dose on the cladding, increased solid fission product swelling, and low operating fuel temperature requires lower fuel smear density (∼60%) in order to ensure acceptable clad hoop strain at high burn-up (∼35at.%). Furthermore, keeping the peak clad temperature below 550°C seems to control lanthanide migration and clad inner wastage at a reasonable level. Possible design concerns and improvements are discussed.

U–Mo alloy fuel for TRU-burning advanced fast reactors

The use of U–Mo instead of U–Zr as the base alloy fuel for transuranics (TRU)-burning advanced fast reactors is assessed in several aspects. While the replacement of Zr with Mo involves no significant differences in terms of neutron physics (core design), U–TRU–Mo does provide advantages. U–TRU–Mo has lower TRU migration to cladding because of its simpler phase diagram, is advantageous in safety margin due to its higher thermal conductivity and better fuel-cladding-chemical-interaction resistance. High fuel swelling data, obtained at low temperatures, available in the literature are not directly applicable to the TRU-burning advanced fast reactors. The potential high swelling can also be controlled when strong cladding and degassing are used as are adopted for typical U–Pu–Zr fuel. Results and detailed analysis are presented in this paper, indicating the benefits of U–Mo base alloy fuel in TRU-burning advanced fast reactors.

An initial assessment of a mechanistic model, GRASS-SST, in U–Pu–Zr metallic alloy fuel fission-gas behavior simulations

A mechanistic kinetic rate theory model originally developed for the prediction of fission gas behavior in oxide nuclear fuels under steady-state and transient conditions has been assessed to investigate its applicability to model fission gas behavior in U–Pu–Zr metallic alloy fuel. In order to capture and validate the underlying physics for irradiated U–Pu–Zr fuels, the mechanistic model was applied to evaluate fission gas release, fission gas and fission product induced swelling, and detailed gas bubble size distributions in three different fuel zones: the outer α-U, the intermediate, and the inner γ-U zones.Due to its special microstructural features, the α-U zone in U–Pu–Zr fuels is believed to contribute the largest fraction of fission gas release among the different fuel zones. It is shown that with the use of small effective grain sizes, the mechanistic model can predict fission gas release that is in reasonable consistence with (though slightly lower than) experimentally measured data. These simulation results are comparable to the experimentally measured fission gas release since the mechanism of fission gas transport through the densely distributed laminar porosity in the α-U zone is analogous to the mechanism of fission gas transport through the interconnected gas bubble porosity utilized in the mechanistic model.Detailed gas bubble size distributions predicted with the mechanistic model in both the intermediate zone and the high temperature γ-U zone of U–Pu–Zr fuel are also compared to experimental measurements from available SEM micrographs. These comparisons show good agreement between the simulation results and experimental measurements, and therefore provide crucial guidelines for the selection of key physical parameters required for modeling these two zones. Material properties such as fuel grain size and thermal diffusivity of gas and model parameters such as di-atom nucleation probability and gas bubble re-solution constant are predicted by these comparisons. In addition, the results of parametric studies for several parameters are presented for both the intermediate zone and the γ-U zone simulations in order to clarify the sensitivities of simulation results on these parameters.

Chemical compatibility of uranium based metallic fuels with T91 cladding

Studies related to development of fast reactor fuels based on ternary U–Pu–Zr and binary U–Pu alloys has been initiated in India for building a data base on thermo-physical and thermodynamic properties, fuel-clad compatibility etc. which are very useful to the fuel-designer to optimize the design feature and to predict the in-reactor fuel behaviour. Fuel-clad chemical compatibility is considered as one of the major concerns for metallic fuels. In the present investigation, the performance of Zr as fuel-clad chemical interaction (FCCI) barrier layer between U and T91 was evaluated by diffusion couple experiments. The growth kinetics of reaction layers at U/Zr and Zr/T91 interfaces were established. The growth kinetics of the reaction zone at both the U/Zr and Zr/T91 interfaces were determined at 973K from the plot of log (width) versus log (time). The value of reaction index n was found to be around 2 at both the U/Zr and Zr/T91 interfaces. The reaction constant (k) for the growth of reaction layer at the U/Zr interface was determined to be 2.07×10−8ms−1/2 at 973K. Similarly, the rate constant at the Zr/T91 interface was found to be 1.95×10−8ms−1/2 at 973K. The activation energy Q for the reaction at the Zr/T91 interface was determined and was found to be 54.7kJmole−1. The fuel-clad chemical compatibility between U–6Zr alloy and T91 steel was also investigated in the present study by diffusion couple experiments. The interdiffusion between U–6Zr and T91 at 973K resulted in the formation of three different layers at the interface. The mechanism of formation of these layers was analysed in detail.

Anodic behaviour of a metallic U–Pu–Zr alloy during electrorefining process

Electrorefining tests of the non-irradiated U–Pu–Zr alloy were performed in LiCl–KCl–UCl 3–PuCl 3–ZrCl 4 melts at 773 K, aiming at reduced Zr dissolution. The tests were carried out both using potentiostatic electrolysis at −1.0 V (vs. Ag +/Ag), i.e. at a more negative potential than the Zr dissolution potential, and galvanostatic electrolysis with a limited amount of Zr dissolution. The ICP–AES analysis of the anode residues confirmed that a high dissolution yield of actinides (U: >99.6%, Pu: 99.9%) was successfully demonstrated for both electrolyses.

Computational study of atomic mobility for the bcc phase of the U–Pu–Zr ternary system

Experimental diffusion data in literature has been evaluated to assess the atomic mobility for the bcc phase in the U–Pu–Zr system by means of the DICTRA-type (Diffusion Controlled TRAnsformation) phenomenological treatment. The developed mobility database has been validated by comprehensive comparisons made between the experimental and calculated diffusion coefficients, as well as other interesting details resulting from interdiffusion, e.g. the concentration profile and the diffusion path of diffusion couples.

As-cast microstructures in U–Pu–Zr alloy fuel pins with 5–8 wt.% minor actinides and 0–1.5 wt% rare-earth elements

The Idaho National Laboratory (INL) is investigating U–Pu–Zr alloys with low concentrations of minor actinides (Np and Am) and rare-earth elements (La, Ce, Pr, and Nd) as possible nuclear fuels to be used to transmute minor actinides. Alloys with compositions 60U–20Pu–3Am–2Np–15Zr, 42U–30Pu–5Am–3Np–20Zr, 59U–20Pu–3Am–2Np–1RE–15Zr, 58.5U–20Pu–3Am–2Np–1.5RE–15Zr, 41U–30Pu–5Am–3Np–1RE–20Zr, and 40.5U–30Pu–5Am–3Np–1.5RE–20Zr (where numbers represent weight percents of each element and RE is a rare-earth alloy consisting of 6% La, 16% Pr, 25% Ce, and 53% Nd by weight) were arc-melted and vacuum cast as fuel pins approximately 4 mm in diameter. The as-cast pins were sectioned, polished, and examined by scanning electron microscopy. Each alloy contains high-Zr inclusions surrounded by a high-actinide matrix. Alloys with rare-earth elements also contain inclusions that are high in these elements. Within the matrix, concentrations of U and Zr vary inversely, while concentrations of Np and Pu appear approximately constant. Am occurs in the matrix and with some high-rare-earth inclusions, and occasionally as high-Am inclusions in samples without rare-earth elements.

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

An engineering code to predict the irradiation behavior of U–Zr and U–Pu–Zr metallic alloy fuel pins and UO2–PuO2 mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel–clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios.FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal–fuel version is called FEAST-METAL, and is described in this paper. The oxide–fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel–clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors.FEAST-METAL was benchmarked against the open-literature EBR-II database for steady state and furnace tests (transients). The results show that the code is able to predict important phenomena such as clad strain, fission gas release, clad wastage, clad failure time, axial fuel slug deformation and fuel constituent redistribution, satisfactorily.

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.

Phase characteristics of a U–20Pu–3Am–2Np–15Zr metallic alloy containing rare earths

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. The current irradiation test series design, designated AFC2, includes minor additions of rare earth elements to simulate expected fission product carry-over from the electrochemical molten salt reprocessing technique. The metal fuel alloys have been fabricated by an arc casting technique. The as-cast fuel alloys have been investigated for phase and thermal properties, specifically, enthalpies of transition, transition temperatures, and room temperature phase characteristics. Results and observations related to these characteristics for the “fresh” fuel alloys are provided. The alloy compositions are based on a U–20Pu–3Am–2Np–15Zr alloy, along with additions of 1 and 1.5 wt% RE (at the expense of U) where RE denotes rare earth alloy of cerium, lanthanum, praseodymium and neodymium. Phase behavior and associated transitions have been compared to available U–Pu–Zr ternary diagrams with acceptable agreement. Enthalpies of transition were deconvoluted from heating and cooling thermal traces for relatively reliable values. The rare earth additions to the base alloy have a minimal influence on the room temperature phases present, but the room temperature phases present slightly impacted the enthalpies of transition and transition temperatures.

Density-functional study of Zr-based actinide alloys: 2. U–Pu–Zr system

Density-functional theory, previously used to describe phase equilibria in the U–Zr alloys [A. Landa, P. Söderlind, P.E.A. Turchi, L. Vitos, A. Ruban, J. Nucl. Mater. 385 (2009) 68], is applied to study ground-state properties of the bcc U–Pu–Zr solid solutions. Calculated heats of formation of the Pu–U and Pu–Zr alloys are in a good agreement with CALPHAD assessments. We found that account for spin–orbit coupling is important for successful description of Pu-containing alloys.

Migration of minor actinides and lanthanides in fast reactor metallic fuel

Minor actinides (MA) and lanthanides (LA) migration to the cladding in fast reactor metallic fuel is a concern because of their reaction with the cladding material. MA and LA migration test results are analyzed and reactions with the cladding are characterized. Remedies for reducing MA and LA migration and reaction with cladding are reviewed. A possible method is proposed that includes the addition of a compound forming element such as indium or thallium in fuel. These elements preferentially form stable compounds with MA and LA and should therefore reduce migration of MA and LA. As an intrinsic solution to the issue, the feasibility of the use of U–Pu–Mo alloy as fuel is studied. Unlike U–Pu–Zr, U–Pu–Mo consists of a single phase at typical fuel operation temperatures, and should have negligible fuel constituent redistribution and reduce MA and LA migration.

Properties of plutonium and its alloys for use as fast reactor fuels

Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu–Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher melting U–Pu–Zr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

Studies on physics parameters of metal (U–Pu–Zr) fuelled FBR cores

For faster growth of nuclear power in India, it is essential to shift to the use of metal-fuels in fast breeder reactors (FBR), which gives a higher breeding ratio (BR) and lower doubling time (DT). Also, future commercialization of the FBR fuel cycle necessitates the use of metallic fuel along with the pyro-process recycling, which can be less costly than oxide fuel reprocessing. Two-dimensional diffusion calculations have been performed to investigate the various physics parameters of metal (U–Pu–Zr) fuelled FBR cores as a function of reactor parameters like reactor power, smear density, zirconium content in the fuel and the number of rows in radial blankets. A 1000 MWe fast reactor with U–Pu fuel (i.e. metal-fuel with no zirconium – which is a theoretical possibility now, due to the lack of irradiation experience) can attain a breeding ratio of 1.61 and a reactor fuel doubling time of 6.6 yrs. Two methods to reduce the sodium void reactivity, which is high and positive in metal-fuelled FBR cores, are suggested.

Transformation and fragmentation behavior of molten metal drop in sodium pool

In order to clarify the fragmentation mechanism of a metallic alloy (U–Pu–Zr) fuel on liquid phase formed by metallurgical reactions (liquefaction temperature = 650 °C), which is important in evaluating the sequence of core disruptive accidents for metallic fuel fast reactors, a series of experiments was carried out using molten aluminum (melting point = 660 °C) and sodium mainly under the condition that the boiling of sodium does not occur. When the instantaneous contact interface temperature ( T i) between molten aluminum drop and sodium is lower than the boiling point of sodium ( T c,bp), the molten aluminum drop can be fragmented and the mass median diameter ( D m) of aluminum fragments becomes small with increasing T i. When T i is roughly equivalent to or higher than T c,bp, the fragmentation of aluminum drop is promoted by thermal interaction caused by the boiling of sodium on the surface of the drop. Furthermore, even under the condition that the boiling of sodium does not occur and the solid crust is formed on the surface of the drop, it is confirmed from an analytical evaluation that the thermal fragmentation of molten aluminum drop with solid crust has a potential to be caused by the transient pressurization within the melt confined by the crust. These results indicate the possibility that the metallic alloy fuel on liquid phase formed by the metallurgical reactions can be fragmented without occurring the boiling of sodium on the surface of the melt.

Modeling of constituent redistribution in U–Pu–Zr metallic fuel

A computer model was developed to analyze constituent redistribution in U–Pu–Zr metallic nuclear fuels. Diffusion and thermochemical properties were parametrically determined to fit the postirradiation data from a fuel test performed in the Experimental Breeder Reactor II (EBR-II). The computer model was used to estimate redistribution profiles of fuels proposed for the conceptual designs of small modular fast reactors. The model results showed that the level of redistribution of the fuel constituents of the designs was similar to the measured data from EBR-II.

Thermodynamic analysis of Np–Zr–H, Am–Zr–H, Pu–Zr–H systems

High-level radioactive wastes generated after reprocessing spent nuclear fuels from nuclear reactors include long-lived and highly toxic radioactive nuclides. Recently, the transmutation method using actinide hydride targets in fast reactor has been proposed to reduce the amount of actinide in nuclear wastes. The burning of Pu using Pu–Zr–H fuel in light water reactor has also been proposed. Given the potential for high-temperature state during routine or extraordinary operating circumstances, accurate knowledge or reliable predictions of materials behaviour under a variety of conditions is required in order to develop safe and effective nuclear handling and remediation technologies. Semi-empirical Miedema and Griessen–Driessen models on the enthalpy of formation coupling with the available experimental entropy information were utilized to estimate the thermodynamic stability of the selected actinide–zirconium hydrides. The calculations showed that at high temperature, stable hydrides might be expected to form for the Np–Zr–H, Am–Zr–H and Pu–Zr–H systems at selected compositions, which can be effectively used for the MA transmutation or plutonium burning.