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DOE OSTI2013

High energy X-ray diffraction measurement of residual stresses in a monolithic aluminum clad uranium–10 wt% molybdenum fuel plate assembly

Residual stresses are expected in monolithic, aluminum clad uranium 10 wt% molybdenum (U–10Mo) nuclear fuel plates because of the large mismatch in thermal expansion between the two bonded materials. The full residual stress tensor of the U–10Mo foil in a fuel plate assembly was mapped with 0.1 mm resolution using high-energy (86 keV) X-ray diffraction. The in-plane stresses in the U–10Mo foil are strongly compressive, roughly -250 MPa in the longitudinal direction and -140 MPa in the transverse direction near the center of the fuel foil. The normal component of the stress is weakly compressive near the center of the foil and tensile near the corner. The disparity in the residual stress between the two in-plane directions far from the edges and the tensile normal stress suggest that plastic deformation in the aluminum cladding during fabrication by hot isostatic pressing also contributes to the residual stress field. A tensile in-plane residual stress is presumed to be present in the aluminum cladding to balance the large in-plane compressive stresses in the U–10Mo fuel foil, but cannot be directly measured with the current technique due to large grain size.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2017

Effects of fuel thickness on thermal and mechanical performance of U10-Mo monolithic fuel plates

Monolithic fuels are being considered for conversion of high performance research and test reactors. These plate-type fuels are comprised of a low enrichment, high density U-Mo alloy fuels within an aluminum cladding. Although the plates have demonstrated satisfactory performance; still, use of a high density fuel in a foil form retains technical challenges that could affect the overall performance. To understand performance of this new design during fabrication and irradiation, the plates have been evaluated for various geometrical and operational variables. As a part of these set of parametric sensitivity studies, effects of foil thickness on performance were studied. Because high performance research reactors will utilize different fuel thicknesses, it is necessary to evaluate possible effects for a range of thicknesses that are being considered for various designs. Based on the preliminary design specifications, the fuel thicknesses were varied between the limiting cases. The bounding fuel thicknesses were 0.203 mm as minimum that is projected for ATR, and 0.635 mm as maximum that is projected for MITR. To study possible effects, a behavioral model was developed for a selected plate from RERTR-12 experiments and the plate was simulated with as run irradiation history. The simulations were repeated for a range of fuel thicknesses, while keeping the cladding thickness and the operational parameters the same. The results have indicated that the plates with thicker fuels would have higher temperatures, deformations and shutdown stresses. To investigate the effects of fuel thickness exclusively, operational parameters were scaled. In particular, a second set of simulations with prorated volumetric heat generation rates were performed. The results indicated that the fabrication stresses in the fuel foil decrease with an increasing fuel thickness. On contrary, irradiation stresses of the fuel at shutdown are higher for the plates with thicker fuels. Greater peak deformations occurred in plates with thicker fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Assessment of Cladding-Cladding Delamination Failures in U-Mo Fuel Plates

Monolithic fuel is a candidate fuel form for the conversion of high-performance research reactors. This plate-type fuel consists of a high-density, U-Mo alloy-based fuel in a monolithic form that is sandwiched between zirconium diffusion barriers and encapsulated in an aluminum cladding. To qualify this fuel system, the program must demonstrate that the fuel meets the safety standards, has a predictable behavior, and performs well in a reactor. Although many plates have shown satisfactory thermo-mechanical performance in reactor, a small number of experimental plates exhibited failures, rendering performance of these plates unacceptable. Post-irradiation examinations of failed plates have revealed that various failures modes are present, including pillowing, blistering, buckling, warping, delamination, fracture, rupture, oxide spallation, etc. Previous studies have implied that such various operational and loading conditions could generate a complex stress-strain state, facilitating distinct failure modes that could impact the structural integrity. To understand such various failure modes, a comprehensive set of studies focusing on failure modes is being performed. As a part of these studies, this work studies the cladding-cladding separation failures that were observed in some of previous irradiation tests. To understand the root cause of cladding separation failures, selected plates from previous irradiation experiments were simulated under comparable operational conditions. Stress-strain state of the plates at the interface were examined. The results were comparatively evaluated to understand possible effects of stress state of the bond-line on delamination failures. This paper presents these findings, discusses such failure modes, and examines the influence of various factors on previously observed cladding delamination failures.

42 ENGINEERING↗

DOE OSTI2016

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation – Non-destructive analysis of the AFIP-1 fuel plates

Not Available

Wachs, D. M.↗

DOE OSTI2011

Microstructural Characterization of Burnable Absorber Materials Being Evaluated for Application in LEU U-Mo Fuel Plates

The starting microstructure of a fuel plate will impact how it performs during irradiation. As a result, microstructural characterization has been performed on as-fabricated monolithic fuel plates to determine the changes in fuel plate microstructure that may result from changes in fabrication parameters. Particular focus has been given to the fuel plate U-10Mo/Zr and Zr/AA6061 cladding interfaces, since the integrity of these interfaces will play a big role in determining the overall performance of the fuel plate during irradiation. In addition, burnable absorber materials for potential incorporation into monolithic fuel plates have been characterized to identify their as-fabricated microstructures. This information will be important when trying to understand the PIE data from fuel plates with burnable absorbers that are irradiated in future irradiation experiments. This paper will focus on the microstructures observed using optical metallography, X-ray diffraction, and scanning and transmission electron microscopy for monolithic fuel plates exposed to different fabrication parameters and for as-fabricated burnable absorber materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Thermal Model of Oxide Growth for Full Size Fuel Plate Experiments in the Advanced Test Reactor

The full-size plate number one (FSP-1) irradiation test designed for irradiation testing of plate type fuel is planned to be irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). This FSP-1 experiment is a non-instrumented drop-in test where aluminum-clad fuel plates are cooled directly by the ATR Primary Coolant System (PCS) water. This experiment is one of a series of experiments being irradiated at ATR for the United States High Performance Research Reactor (USHPRR) program. This fueled experiment contains aluminum-clad fuel full-size plates consisting of monolithic U-10Mo. Previous thermal analyses have been documented concerning oxide growth [1] and thermal safety margins [2] for similar U-10Mo mini plate experiments. This analysis is part of the thermal safety analysis of the experiment to be irradiated in ATR. Departure from nucleate boiling ratio (DNBR) and flow instability ratio (FIR) have been calculated for the flow coastdown condition 2 and the reactivity insertion accident condition-2 (RIA2) transient. The purpose of this paper is to investigate the thermal margin of the fuel meat during a RIA2 transient when oxide has accumulated on the plate surfaces acting as an insulator.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

STAT7 v1.2 User Guide: The STAT7 Code for Statistical Propagation of Uncertainties in Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

The STAT7 software was developed to perform steady-state, single-phase thermal hydraulics analysis of plate-fueled reactors based on statistical propagation of uncertainties. Application of the software is for non-power research and test reactors, including conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors such as Massachusetts Institute of Technology Research Reactor. Since it can be necessary to repeat analysis during fuel reloading, STAT7 accommodates flexibility in analyzing many realistic aspects of reactor fuel management. STAT7 uses a Monte Carlo approach to model uncertainty in common fuel fabrication parameters and other key reactor operating parameters required for thermal hydraulics analyses of research and test reactors. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux, or onset of flow instability. STAT7 supports water properties based on the IAPWS-IF97 functions (The International Association for the Properties of Water and Steam Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam) in addition to the fit functions. STAT7 predicts axial profiles of fuel, cladding, and coolant temperature along a lateral stripe that runs the full length of the fuel plate from the bottom to the top. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one latera stripe of a fuel element. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis can be performed by analyzing the margin to onset of nucleate boiling and onset of flow instability for each axial node of each stripe of each plate of each fuel element in the core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

The porosity surrounding carbides and second phase stringers in monolithic U-10Mo fuel plate after irradiation

Post-irradiation microstructure characterization plays an important role in qualifying the low-enriched uranium (LEU) monolithic U-10 wt%Mo plate-type fuel for United States high-performance research reactors (USHPRRs) program. Inhomogeneous features resulting from manufacturing and irradiation processes, including carbides, second phase stringers, and extensive void spaces caused by the combining of small porosities, may increase the risk of heat concentration in local regions of the fuel plate over the operating conditions. In this study, characteristics of carbides, stringers, and porosity after multiple levels of irradiation at varying fission densities were studied by electron microscopes to decipher the morphology of pores and the porosity evolution in U-10 wt%Mo. For carbides, the result shows that porosities start forming on UMo grain boundaries, then on UMo/carbides interfaces as the burn-up going higher. However, the porosities surrounding carbides grow larger than the ones on UMo grain boundaries. The porosities around the uranium carbides could interconnect to form larger void space. The study revealed that the void spaces larger than 5 µm were found around uranium carbides after high burnup, while no evidence was observed to support the similar voids formed near second phase stringers even though the size of the stringers (> 50 µm) was much larger than uranium carbides (< 20 µm). The evolution of porosities suggests that the formation of second phase stringers may not create more significant porosities compared to regular uranium carbides regions during fuel operating conditions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Failure investigation and mitigation after experimental research reactor fuel plate deformation in an irradiation device

Experimental research reactor fuel testing is conducted in the Belgian Reactor 2 (BR2) of the Belgian Nuclear Research Centre (SCK CEN) in dedicated irradiation vehicles or rigs. One such vehicle allows flat full-size fuel plates to be irradiated by inserting them into slotted baskets that captures a narrow portion of the longitudinal edges of the plates. The motion of the fuel plates within the baskets is possible within the narrow slots and thus, the plate is considered to be unattached. The design intentionally omits fixing mechanisms of the fuel plates to the baskets to facilitate the inspection and repositioning of the plates between the irradiation cycles and the accommodation of thermal expansion of the plates in the lateral direction. However, loosely inserted fuel plates have weak structural boundary conditions allowing for larger out-of-plane deflections caused by hydrodynamic loads exerted by the flowing coolant, as compared to those of fixed plates. Unexpected large deformations of plates occurred in several irradiation cycles that further resulted in a loss of cladding integrity. These deformations could not be attributed to a single source. This triggered a series of thermal hydraulic, structural, and fluid-structure interaction analyses aiming at understanding the observed phenomenon. The analyses revealed that, for a certain combination of unfavorable manufacturing and assembly tolerances, fuel plate edges could escape out of the slots in the irradiation basket due to the hydrodynamic load. Subsequently, the plate could become wedged inside the basket coolant channel opening. This resulted in reduced coolant flow and accelerated temperature increase and thermal expansion of the plate while under irradiation. This unfavorable feedback loop could then lead to excessive plate surface temperatures, deformed plates and cladding failure, as was observed in the experiments. These analyses not only provided a probable cause of the fuel plate failures, but also resulted in a new and improved design of the irradiation basket to avoid these issues in the future. In conclusion, a series of recent successful irradiations confirm that the sources of failures were identified correctly, and the implemented mitigations were adequate.

BR2↗

DOE OSTI2013

Laser-Based Characterization of Nuclear Fuel Plates

Ensuring the integrity of fuel-clad and clad-clad bonding in nuclear fuels is important for safe reactor operation and assessment of fuel performance, yet the measurement of bond strengths in actual fuels has proved challenging. The laser shockwave technique (LST) originally developed to characterize structural adhesion in composites is being employed to characterize interface strength in a new type of plate fuel being developed at Idaho National Laboratory (INL). LST is a non-contact method that uses lasers for the generation and detection of large-amplitude acoustic waves and is well suited for application to both fresh and irradiated nuclear-fuel plates. This paper will report on initial characterization results obtained from fresh fuel plates manufactured by different processes, including hot isostatic pressing, friction stir welding, and hot rolling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Laser-based characterization of nuclear fuel plates

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Microstructural Analysis of Irradiated U-Mo Fuel Plates: Recent Results

Microstructural characterization of irradiated dispersion and monolithic RERTR fuel plates using scanning electron microscopy (SEM) is being performed in the Electron Microscopy Laboratory at the Idaho National Laboratory. The SEM analysis of samples from U-Mo dispersion fuel plates focuses primarily on the behavior of the Si that has been added to the Al matrix to improve the irradiation performance of the fuel plate and on the overall behavior of fission gases (e.g., Xe and Kr) that develop as bubbles in the fuel microstructure. For monolithic fuel plates, microstructural features of interest, include those found in the U-Mo foil and at the U-Mo/Zr and Zr/6061 Al cladding interfaces. For both dispersion and monolithic fuel plates, samples have been produced using an SEM equipped with a Focused Ion Beam (FIB). These samples are of very high quality and can be used to uncover some very unique microstructural features that are typically not observed when characterizing samples produced using more conventional techniques. Overall, for the dispersion fuel plates with matrices that contained Si, narrower fuel/matrix interaction layers are typically observed compared to the fuel plates with pure Al matrix, and for the monolithic fuel plates microstructural features have been observed in the U-10Mo foil that are similar to what have been observed in the fuel particles found in U-Mo dispersion fuels. Most recently, more prototypic monolithic fuel samples have been characterized and this paper describes the microstructures that have been observed in these samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 mu m in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates. (C) 2014 Elsevier B.V. All rights reserved.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Interface Characterization within a Nuclear Fuel Plate

To predict the performance of nuclear fuels and materials, irradiated fuel plates must be characterized efficiently and accurately in highly radioactive environments. The characterization must take place remotely and work in settings largely inhospitable to modern digital instrumentation. Characterization techniques based on non-contacting laser sensing methods enable remote operation in a robust manner within a hot-cell environment. Laser characterization instrumentation can offer high spatial resolution and remain effective for scanning large areas. A laser shock (LS) system is currently being developed as a post-irradiation examination (PIE) technique in the hot fuel examination facility (HFEF) at the Idaho National Laboratory (INL). The laser shock technique will characterize material properties and failure loads/mechanisms in various composite components and materials such as plate fuel and next-generation fuel forms in high radiation areas. The laser shock-technique induces large amplitude shock waves to mechanically characterize interfaces such as the fuel–clad bond. As part of the laser shock system, a laser-based ultrasonic C-scan system will be used to detect and characterize debonding caused by the application of the laser shock. The laser shock system has been used to characterize the resulting bond strength within plate fuels which have been fabricated using different fabrication processes. The results of this study will be to select the fabrication process that provides the strongest interface.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2011

Microstructural Characterization of U-7Mo/Al-Si Alloy Matrix Dispersion Fuel Plates Fabricated at 500°C

The starting microstructure of a dispersion fuel plate will impact the overall performance of the plate during irradiation. To improve the understanding of the as-fabricated microstructures of U–Mo dispersion fuel plates, particularly the interaction layers that can form between the fuel particles and the matrix, scanning electron microscopy (SEM) and transmission electron microscopy (TEM) analyses have been performed on samples from depleted U–7Mo (U–7Mo) dispersion fuel plates with either Al–2 wt.% Si(Al–2Si) or AA4043 alloy matrix. It was observed that in the thick interaction layers, U(Al, Si)3 and U6Mo4Al43 were present, and in the thin interaction layers, (U, Mo) (Al, Si)3, U(Al, Si)4, U3Si3Al2, U3Si5, and possibly USi-type phases were observed. The U3Si3Al2 phase contained some Mo. Based on the results of this investigation, the time that a dispersion fuel plate is exposed to a relatively high temperature during fabrication will impact the nature of the interaction layers around the fuel particles. Uniformly thin, Si-rich layers will develop around the U–7Mo particles for shorter exposure times, and thicker, Si-depleted layers will develop for the longer exposure times.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

The Impact of Furnace-Testing on the Microstructure of an Irradiated U-Mo Fuel Plate

The Material Management and Minimization Program (formerly known as the Reduced Enrichment for Research and Test Reactor Program) is responsible for converting research reactors that use highly enriched uranium (HEU) fuels to ones that use low-enriched uranium (LEU) fuels. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. In addition to the reactor tests, there is interest in generating information about the transient behavior of irradiated fuel plates. For example, it is of interest to determine the temperature at which a blister will form on the surface of a fuel plate, due to fission gas release from the fuel meat, for a particular fuel plate design. Since in-pile testing to determine the 'blister threshold' (the temperature where a blister forms on the surface of a fuel plate) in plate-type fuel is very difficult and expensive and has historically been deemed impractical for fuel qualification testing, an out-of-pile 'blister test' has been developed and employed. This post-irradiation blister anneal test was developed as a conservative simulation of in-pile behavior and has been routinely used for many research reactor fuel designs currently in use. In this test, the irradiated fuel plate is inserted into a furnace at a given temperature, soaked for a period of time, and removed for inspection. If no visible blisters appear, the furnace temperature is increased, and the plate is soaked and inspected again. This process continues until blisters are observed. Recently, a section from an irradiated U-7Mo dispersion fuel plate with Al-4043 matrix, which has a nominal composition of 4.81Si-0.20Fe-0.14Ti-0.16Cu-0.01Cr 0.01Mn-bal Al, was tested in a furnace located in the Hot Fuel Examination Facility, located at the Idaho National Laboratory, up to 500 deg. C, at which point blisters were observed on the surface of the fuel plate section. A sample near the blistered region of the fuel plate section was then characterized using scanning electron (SEM) and transmission electron microscopy (TEM). This paper discusses the results of this characterization and how they compare to those generated for a similar sample characterized after fabrication but before irradiation and for a sample characterized after irradiation but before furnace testing. One goal of this work is to identify the changes in the microstructure during blister testing that contribute to the eventual release of fission gas that results in the development of blisters. SEM analysis of the furnace-tested sample showed that in some locations of the U-7Mo particles, Al-4043 matrix constituents (primarily Si) had penetrated the microstructure along grain boundaries. This provided potential pathways for increased fission gas migration in the fuel meat microstructure during the furnace heating process. Some attack of these boundaries was also observed in the as-fabricated U-7Mo/Al-4043 sample, but there appeared to be additional attack during the furnace annealing. Besides the changes observed in grain boundary features, the TEM analysis showed that the fission gas bubbles that are typically present in the intra-granular regions of the as-irradiated U-7Mo microstructure as a fission gas bubble superlattice had changed in size, morphology, and distribution. The result of this may be an increase in the mobility of the fission gas in intra-granular regions of the fuel. The complete implications of the changes in fission gas morphology and distribution in the fuel meat microstructure and their impact on blistering continues to be investigated. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Research reactor plate fuel multi-physics hotspot model for the HFIR

Ongoing development of a multi-physics hotspot model for nuclear research reactor plate fuel is presented with a focus on the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The background of plate fueled research reactors in general and the HFIR in particular are discussed along with the need to convert from high enriched uranium (HEU) to low enriched uranium (LEU) fuel, which drives the need to improve the analysis capability of fuel thermal-fluid performance. HFIR HEU fuel plate fabrication is described to explain the origin of fuel segregations and cladding non-bonds that currently contribute to the hotspot flux peaking factors. Previous efforts at modeling hotspots caused by these defects are reviewed and areas for improvement associated with high resolution multi-physics simulation are identified. Modes of heat transfer through non-bonds, including thermal contact conductance, are examined. Conjugate heat transfer over hotspots is also explored to high resolution up to and through the wall boundary. The challenges associated with multi-physics verification and validation (VV), particularly the requirements beyond VV 20, are presented. An approach to Validation within the limits of existing data for conduction heat transfer through the fuel plate, and convection into the cooling fluid is set forth. Preliminary simulations are presented assessing hot spot modeling approaches for contact conductance, and fluid turbulence modeling within COMSOL. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

On the Steady Mechanical Response of a Heterogeneous Fuel Plate

Herein, an exact solution for a heterogeneous, discontinuous wide-beam is developed having three unique boundary conditions sets. These boundary condition sets were chosen based on their relevance to plate-type fueled reactors. This solution, with respective boundary condition sets contributes new insights to the field of engineering mechanics as applied within the nuclear engineering discipline. Herein, the exact solution is validated under a set of test cases. A comparison is then made against other relations that have been developed to provide similar engineering insights. Lastly, the solution is applied toward a presently relevant engineering problem regarding the design of a monolithic plate-type fuel. The outcome of this work provides a solution form for the computation of out-of-plane deflection of a heterogeneous, discontinuous wide-beam that can be easily applied in engineering mechanics and is flexible in use.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗