Materials and Processes for Nuclear Fuel Reprocessing

Metal Fuel–Cladding Interaction

Use of Pu/sup 239/ -- U/sup 235/-fueled fast critical assemblies to estimate properties of Pu/sup 239/-fueled assemblies is of interest because of safety considerations and limited plutonium availability. Bare and reflected homogeneous cores and reflected two-region cores are considered. The fuel, 5% by volume, is assumed to be Pu/sup 239/ and U/sup 235/ of various fuel composition ratios for the homogeneous cores. For the tworegion cores the 5% fuel volume is Pu/sup 235/ in the central region and U/sup 235/ in the outer core region. Core diluents, simulating fertile, structural, and coolant materials, are assumed identical in all cases. it is estimated that construction of the reflected two- region core with ratio of central core region volume to total core volume of 0.1 will theoretically decrease the calculated error in prediction of the critical size of a corresponding solely Pu/sup 239/-fueled assembly by a factor of about 10 to 20. (auth)

Metallic nuclear fuels were tested in fast reactor programs and performed well. However, metallic fuels have shown the phenomenon of FCCI that are due to deleterious reactions between lanthanide fission products and cladding material. As the burnup is increased, lanthanide fission products that contact with the cladding could react with cladding constituents such as iron and chrome. These reactions produce higher-melting intermetallic compounds and low-melting alloys, and weaken the mechanical integrity.

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.

This research involved developing an expert system that allows a nuclear fuel engineer to quickly provide answers to strategic nuclear fuel management questions, which are typically broad based. Current nuclear fuel analysis research concentrates on getting more accurate and precise answers at the expense of using large computer programs to get answers that are too specific to answer the broad based questions. The expert system brings together several artificial intelligence techniques to allow a nuclear fuel engineer to consider several scenarios in a general way in order to quickly answer the fuel management questions asked. The expert system is based upon a hierarchy of several abstraction levels using a constraint propagation system at the lowest level. The constraint propagation system prevents a novice nuclear fuel engineer from studying a scenario with input conditions that contradict standard nuclear fuel management relationships. The other abstraction levels include generic number representations, generic mathematical operators, and generic relationships for economic analysis. The highest level of the hierarchy is the knowledge base for nuclear fuel analysis of the equilibrium nuclear fuel cycle. The simplicity of adding other number representations to the expert system is demonstrated by implementing an interval number representation. Since the mathematical operators used at the knowledge domain level are generic, any new number representations, such as fuzzy numbers, could be added without having to change the basic domain knowledge. An example session shows how the system can be used to provide guidance to a nuclear fuel analyst in search of a good nuclear fuel management strategy. By using the interval number representation, the example session includes a simple sensitivity study on how some of the input variables' uncertainty affects the objective variable's value.

This report documents work performed supporting the Department of Energy (DOE) Office of Nuclear Energy (NE) Fuel Cycle Technologies Used Fuel Disposition Campaign (UFDC) under work breakdown structure element 1.02.08.10, ST Analysis. In particular, this report fulfills the M4 milestone M4FT- 15OR0810036, Quantify effects of power uncertainty on fuel assembly characteristics, within work package FT-15OR081003 ST Analysis-ORNL. This research was also supported by the Consortium for Advanced Simulation of Light Water Reactors (http://www.casl.gov), an Energy Innovation Hub (http://www.energy.gov/hubs) for Modeling and Simulation of Nuclear Reactors under U.S. Department of Energy Contract No. DE-AC05-00OR22725. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each evaluated WBN1 fuel rod. An alternate model for the amount of helium released from the zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layer is derived and applied to FRAPCON output data to quantify the RIP and CHS for these types of fuel rods. SCALE/Polaris is used to quantify fuel rodspecific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel pellets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd/MTU is determined to be the total fuel rod void volume and the amount of released fission gas in the fuel rod, respectively. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP and CHS predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceed a specified RIP or CHS limit. Results are separated into IFBA and standard rods so that the two groups may be analyzed individually. FRAPCON results are provided in sufficient detail to enable the recalculation of the RIP while considering any desired plenum gas temperature, total void volume, or total amount of gas present in the void volume. A method to predict the CHS from a determined or assumed RIP is also proposed, which is based on the approximately linear relationship between the CHS and the RIP. Finally, improvements to the computational methodology of FRAPCON are proposed.

The paper consists of three sections and is of an informational and generalizing nature, indicating promising areas for further research. The first section "R-functions method in mathematical modeling of convective heat transfer in fuel cartridges with fuel elements" is devoted to the use of new constructive tools of the R-functions method for mathematical and computer modeling of fuel elements packings with different types of symmetry, as well as the study of convective heat transfer in fuel elements grids and the effect of the type of packing on the distribution of velocity and temperature. An octahedral cartridge with 37 fuel elements packed according to three patterns (cyclic, checkerboard and in-line) is considered. It is noted that when constructing the equations of a cartridge with bundles of fuel elements using the new method, the number of R-operations and, accordingly, the calculation time are significantly reduced. An analysis of the obtained results allows to conclude that the maximum temperature is obtained with cyclic packing. The scheme of the reactor, the cartridges of which are hexagonal casings, with 91 fuel elements placed in each of them both with checkerboard and cyclic packing, is also considered. In the second section "Thermal-hydraulic calculation of fuel elements cartridges in case of violation of the rods packing symmetry", a hexagonal fuel cartridge with 169 fuel elements and checkerboard packing is considered. An increase in temperature is analyzed in case of violation of the packing symmetry while maintaining the parallelism of the rods, as well as in case of a curvature of one of them. The third section "R-functions, fuel element with polyzonal finning of the shell and heat transfer during fluid motion" is focused on the construction of equations for various finning surfaces of fuel elements and the study of hydrodynamic and temperature fields in case of polyzonal finning of the shell. At the same time, using the apparatus of tensor analysis, a transition to a curvilinear non-orthogonal (helical) coordinate system was made. It is noted that mathematical modeling and the associated computer experiment are indispensable in cases where a full-scale experiment is impossible or difficult to conduct for one reason or another. In addition, working with mathematical model of the process and the computational experiment make it possible to investigate the properties and behavior of the process in various situations relatively quickly and without significant expenses. The reliability of the methods, results and conclusions is confirmed by comparison with the information given in the references, the results of the analysis of the numerical convergence of solutions and the calculation of the residual.

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.

The Fort St. Vrain Early Validation Irradiation Experiment is an irradiation test of reference and of improved High-Temperature Gas-Cooled Reactor fuels in the Fort St. Vrain Reactor. The irradiation test includes fuel rods fabricated at ORNL on an engineering scale fuel rod molding machine. Fuel rods were nondestructively assayed for /sup 235/U content by a technique based on the detection of prompt-fission neutrons induced by thermal-neutron interrogation and were later chemically assayed by using the modified Davies Gray potentiometric titration method. The chemical analysis of the thorium content was determined by a volumetric titration method. The chemical assay method for uranium was evaluated and the results from the as-molded fuel rods agree with those from: (1) large samples of Triso-coated fissile particles, (2) physical mixtures of the three particle types, and (3) standard solutions to within 0.05%. Standard fuel rods were fabricated in order to evaluate and calibrate the nondestructive assay device. The agreement of the results from calibration methods was within 0.6%. The precision of the nondestructive assay device was established as approximately 0.6% by repeated measurements of standard rods. The precision was comparable to that estimated by Poisson statistics. A relative difference of 0.77 to 1.5% was found between the nondestructive and chemical determinations on the reactor grade fuel rods.

This document hasbeen prepared to assist research reactor operators possessing spent fuel containing enriched uranium ofUnited States origin to prepare part ofthe documentation necessary to ship this fuel to the United States. Data are included onthe nuclear mass inventory, photon dose rate, thermal decay heat, and fission product decay rate of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present inspent research reactor fuel are estimated for MTR,TRIGA andDIDO fuel assembly types. Theisotopic masses of each fuel assembly type are given as functions ofU-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg235U, and for fission product decay times ofup to 20 years. Thermal decay heat load and fission product radioactivity is estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

Nuclear Fuel Assembly Deformation, Reduced Mechanical Model Dedicated to FSI Simulation

Authors

Currently the problem of spent nuclear fuel (SNF) reprocessing is relevant and studied in many countries. Existing methods of radiochemical reprocessing can not fully meet all the economic, energy and environmental needs of humanity [1]. Recently, the physical principles of magnetoplasma (MP) method of SNF separation is actively developed and investigated in NSC KIPT. This method consists in separation of nuclear fuel (NF) from fission products (FP) in the plasma phase [2]. During MP method application it is also expected the preliminary purification of SNF from some FP at the thermal heating stage [3]. At the next stage - the ionization stage - the complex multicomponent substance is converted into the plasma. Then directly NF is separated from FP at the magnetoplasma separation stage in plasma rotated in crossed electric and magnetic fields. The motion trajectories of NF and FP charged particles are greatly differ, therefore, NF and FP can be spatially separated [4]. Particles of NF and FP will get to the respective collectors, located in different places of SNF separation plant. Subsequently NF can be reused in different types of nuclear reactors.

The GGA fast-flux experiment In EBR-II will study fuel-rod specimens with cladding temperatures in the range of 650°C to 800°C at higher temperatures not heretofore covered in the LMFBR program and at linear powergeneration rates of 14 to 16 kW/ft.The mixed oxide fuel is contained in 20% cold-worked 316 stainless steel and will achieve bumups ranging from 2.7 to 10.7% FIMA.The capsules are generally of the B-7B type with a stainless steel annular thermal barrier sodium bonded to the fuel rod and to the capsule tube.Silicon-carbide temperature monitors are placed in various positions along the thermal barrier to allow postirradiation determination of capsule temperatures utilizing dimensional changes during isochronal annealing of the SiC.Activated charcoal has been placed in all of the fuel rods to determine the effects of fast neutron irradiation on the sorption characteristics of the charcoal for fission products.All eight fuel rods contain sealed charcoal capsules for postirradiation sorption experiments.Three of the fuel rods also contain charcoal traps connected to the fueled region to study the fission-product distribution in the trap under reactor operating conditions.

A numerical study on fuel rod surface temperature distributions in liquid metal cooled sub-channel flow is presented in this paper. Special interests are focused on the effects of misaligned central fuel rod in typical fuel rod cluster geometries. Numerical models based on FLUENT, a Computational Fluid Dynamics (CFD) commercial software, are developed to model steady state fuel thermal conductions, fully-developed turbulent convective flows, and local flow mixing. Numerical results show that fuel rod surface temperatures are higher for surfaces facing the narrow gap of the sub-channel. The temperature differences are mainly dependent on the fuel pitch-to-diameter ratio, flow patterns, and other flow physical properties. Situations are worse for cases where misaligned fuel rods exist in tightly packed fuel rod clusters. Even for a small amount of rod center deviation, such temperature differences can be as high as 54 K around a fuel rod whose diameter is only 4.7 mm. The situations could be very serious and might generate unnecessary thermal stresses and cause fuel failure.