Metal Fuel–Cladding Interaction

Authors

This paper compares measured results with simulation results of neutron beam transmission through an irradiated fuel assembly. The main objective of the comparison is to establish the technical foundation for using Monte Carlo simulation to evaluate the feasibility of using neutron computed tomography for irradiated fuel assembly inspection. The measured results were obtained from an irradiated fuel assembly from the Experimental Breeder Reactor II (EBR-II), and the neutron beam was produced by the Argonne National Laboratory Neutron Radiography Reactor (NRAD). The measurements consist of a projection profile representing the relative neutron beam attenuation at a specific fuel assembly axial elevation obtained from digitized neutron radiography film. Simulation of the neutron beam and fuel assembly was performed using the Monte Carlo code MCNP5. Results presented include the measured beam attenuation projection profile, simulated neutron beam attenuation projection profiles, parametric study of simulation results, and comparison of the projection results. Comparison of the radiography-based measurement with the simulation results shows good agreement, thereby confirming that Monte Carlo simulation of neutron transmission through an irradiated fuel assembly using MCNP5 is a reliable method for evaluating the use of neutron computed tomography as a means of inspecting irradiated fuel assemblies.

Metallic nuclear fuels have generated renewed interest for advanced liquid metal reactors (LMRs) due to their physical properties, ease of fabrication, irradiation behavior, and simple reprocessing. Irradiation performance for both steady-state and transient operations is excellent. Ongoing irradiation tests in Argonne-West’s Idaho-based Experimental Breeder Reactor II (EBR-II) have surpassed 100,000 MWd/T burnup and are on their way to a lifetime burnup of 150,000 MWd/T or greater. Metallic fuel also has a unique neutronic characteristic that enables benign reactor responses to loss-of-flow without scram and loss-of-heat-sink without scram accident conditions. This inherent safety potential of metallic fuel was demonstrated in EBR-II just one year ago. Safety tests performed in the reactor have also demonstrated that there is ample margin to fuel element cladding failure under transient overpower conditions. These metallic fuel attributes are key ingredients of the integral fast reactor (IFR) concept being developed at Argonne National Laboratory.

Reliable knowledge of thermal conductivity is needed to assess reactor fuel performance. In normal operation, fuel pin linear power and fuel thermal conductivity together largely determine the fuel's peak temperature, and under severe accident conditions, fuel thermal conductivity largely determines the power to melt. Unfortunately, assessing the thermal conductivity of modern high-swelling sodium-bonded metallic fuel is neither theoretically straightforward nor amenable to laboratory measurements. At issue is whether the high swelling that takes place in metallic fuel during the first few atomic percent of burnup could lead to drastic conductivity reductions by factors of as much as 0.4 to 0.5. Such reduction may be mitigated by the infiltration of fuel porosity by high-conductivity liquid bond sodium or later by the solid and liquid fission products that accumulate with burnup. In this paper, the authors utilize measurements of maximum melting from intact irradiated metal fuel pins tested in the recent M-series in-pile test program in TREAT to estimate metal fuel thermal conductivity directly and in situ. Of six irradiated integral fast reactor prototype fuel pins ubjected to severe overpower and melting, four pins remained intact. Three of the test pins were of ternary alloy (U-19 Pu-10 Zr) and one of binarymore » (U-10 Zr) alloy. Importantly, these test pins span key range of low burnup where minimum values of conductivity would be expected.« less

Many of the new advanced reactor designs being proposed by the nuclear private sector require uranium fuel enriched between 5 and 20%, otherwise known as High Assay Low Enriched Uranium (HALEU), to support their designs. There are currently no commercial facilities based in the United States which have the capability to produce this material, and as a result the nuclear industry has requested support from the Department of Energy (DOE) to meet this need. One approach being considered by DOE to respond to this demand is to recover highly enriched uranium from used nuclear fuels and downblend it to enrichments less than 20 % U-235. DOE’s past research related to liquid metal fast breeder reactors resulted in the irradiation of more than 3 metric tons (MT) of nuclear fuel composed of highly enriched metallic uranium which resides at the Idaho National Laboratory (INL). This inventory of used nuclear fuel is currently being conditioned using an electrometallurgical treatment process (EMT) to facilitate its eventual disposition to a geologic repository. As part of this treatment process, the highly enriched component of this used fuel is separated, recovered, downblended to <20% U-235, and stored as a metallic uranium ingot. The majority of this fuel was irradiated in the Experimental Breeder Reactor II (EBR-II) over the course of its 30 years of operation. This sodium cooled fast reactor utilized a core composed of highly enriched, sodium-bonded, metallic fuel to sustain or “drive” the reactor. This EBR-II driver fuel is currently being treated for sodium neutralization in the Fuel Conditioning Facility (FCF) at the Materials and Fuels Complex (MFC) and the metallic uranium recovered as a component of this process has been proposed as a potential source of supply of high assay low enriched uranium. The irradiation of the metallic uranium fuel results in the creation of fission products and some minor actinides which are not compatible with many reuse scenario’s, and introduces significant challenges associated with fabrication of new fuel elements using this material. However, it is believed that the current EMT process can be slightly enhanced to address these concerns and result in the recovery of a product that can be considered as a source of supply for HALEU applications. Based on this, DOE requested the INL research methods to enhance the EMT process to produce a recovered uranium material that has a lower contaminant level and lower radiological dose rate and could be utilized in glovebox based fuel fabrication scenarios. The report that follows documents the progress toward meeting this goal.

Vanadium diffusion coating on HT-9 cladding for mitigating the fuel cladding chemical interactions

Metal fuel relocation experiments with pressure injection

A Multiphysics approach to calculate the constituent redistribution in U-Pu-Zr metallic fuel is developed in this work and the fuel performance of U-Pu-Zr metallic fuel under normal operating condition is investigated by CAMPUS code where the model of constituent redistribution is applied. The metallic nuclear fuel is proposed to be used as a candidate of Sodium cooled Fast Neutron Reactor fuel due to its superior thermodynamic properties, such as thermal conductivity and heat capacity. Recently, the constituent redistribution is found to be an important phenomenon in the metallic fuel pellet, which influences the thermodynamic properties of metallic fuel greatly. In this work, the properties of metallic fuel and the calculation method of constituent redistribution are introduced firstly. And the model of constituent redistribution in U-Pu-Zr metallic fuel is developed and implemented into CAMPUS code, and then the metallic fuel performance in the sodium-cooled fast reactor is further studied by CAMPUS code under normal operating condition. Thirdly, the model of constituent redistribution is verified and the fuel performance of U-Pu-Zr metallic fuel is presented and discussed. The Zirconium element is found to migrate to the fuel centerline and fuel surface and the performance of the U-Pu-Zr fuel is found to be influenced by the constituent redistribution dramatically.

Experience with advanced driver fuels in EBR-II

Effects of intermetallic compounds on the thermophysical properties of uranium –palladium alloys

Improvement of the Decay Heat Removal Characteristics of the Generation IV Gas-cooled Fast Reactor

HT9 ferritic-martensitic (FM) steel has served as a leading candidate for sodium-cooled fast reactor (SFR) cladding due to its favorable resistance to irradiation-induced swelling and good thermal and chemical properties.However, its limited creep strength at temperatures above 600 C and susceptibility to phase embrittlement under specific conditions could limit its application in next-generation SFRs.Optimized Grade 92 (Gr92) steels with thermal mechanical treatment (TMT) have demonstrated superior creep performance and reduced embrittlement compared to HT9, due to the reduced Cr content.This work aims to design new FM steels that retain or surpass the creep performance of Gr92 while improving resistance to fuel-cladding chemical interaction (FCCI) and embrittlement.Using thermodynamic modeling with the TCFe11 database, we analyzed three TMT-treated Gr92 variants, including ORNL-developed 41A2, 48A2, and 64-2 heats, to identify microstructural factors correlating with improved creep life.These insights guided the design of new Gr92-based alloy.A new alloys based on the composition of OP92M3 is fabricated with the intention to maximize MX phase fraction, maintain Cr levels above 9.5 wt%, preserve a favorable austenite transformation window, and ensure balanced precipitate stability and grain size.Thermodynamic simulations show that the best-designed alloy, OP92M3, achieves a high MX content with reduced Laves phase and -susceptible features, promising improved performance over current FM cladding materials.

Characterization of unirradiated and irradiated LMFBR fuels by analytical chemistry methods will continue, and additional methods will be modified and mechanized for hot cell application. Macro- and microexaminations will be made on fuel and cladding using the shielded electron microprobe, emission spectrograph, radiochemistry, gamma scanner, mass spectrometers, and other analytical facilities. New capabilities will be developed in gamma scanning, analyses to assess spatial distributions of fuel and fission products, mass spectrometric measurements of burnup and fission gas constituents and other chemical analyses. Microstructural analyses of unirradiated and irradiated materials will continue using optical and electron microscopy and autoradiographic and x-ray techniques. Analytical quality assurance standards tasks are designed to assure the quality of the chemical characterizations necessary to evaluate reactor components relative to specifications. Tasks include: (1) the preparation and distribution of calibration materials and quality control samples for use in quality assurance surveillance programs, (2) the development of and the guidance in the use of quality assurance programs for sampling and analysis, (3) the development of improved methods of analysis, and (4) the preparation of continuously updated analytical method manuals. Reliable analytical methods development for the measurement of burnup, oxygen-to-metal (O/M) ratio, and various gases in irradiated fuels is described.

Metallic Fuels: The EBR-II Legacy and Recent Advances

Effect of CrN barrier on fuel-clad chemical interaction