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785 records · Page 44
All sourcesNASA NTRSDOE OSTI
NASA NTRS2010

Electrical Capacitance Volume Tomography with High-Contrast Dielectrics

The Electrical Capacitance Volume Tomography (ECVT) system has been designed to complement the tools created to sense the presence of water in nonconductive spacecraft materials, by helping to not only find the approximate location of moisture but also its quantity and depth. The ECVT system has been created for use with a new image reconstruction algorithm capable of imaging high-contrast dielectric distributions. Rather than relying solely on mutual capacitance readings as is done in traditional electrical capacitance tomography applications, this method reconstructs high-resolution images using only the self-capacitance measurements. The image reconstruction method assumes that the material under inspection consists of a binary dielectric distribution, with either a high relative dielectric value representing the water or a low dielectric value for the background material. By constraining the unknown dielectric material to one of two values, the inverse math problem that must be solved to generate the image is no longer ill-determined. The image resolution becomes limited only by the accuracy and resolution of the measurement circuitry. Images were reconstructed using this method with both synthetic and real data acquired using an aluminum structure inserted at different positions within the sensing region. The cuboid geometry of the system has two parallel planes of 16 conductors arranged in a 4 4 pattern. The electrode geometry consists of parallel planes of copper conductors, connected through custom-built switch electronics, to a commercially available capacitance to digital converter. The figure shows two 4 4 arrays of electrodes milled from square sections of copper-clad circuit-board material and mounted on two pieces of glass-filled plastic backing, which were cut to approximately square shapes, 10 cm on a side. Each electrode is placed on 2.0-cm centers. The parallel arrays were mounted with the electrode arrays approximately 3 cm apart. The open ends were surrounded by a metal guard to reduce the sensitivity of the electrodes to outside interference and to help maintain the spacing between the arrays. Other uses for this innovation potentially include quantifying the amount of commodity remaining in the fuel and oxidizer tanks while on-orbit without having to fire spacecraft engines. Another orbit application is moisture sensing in plant-growth experiments because microgravity causes moisture in soil to distribute itself in unusual ways. At the moment, the hardware and image reconstruction technique may only be of interest to people involved in nondestructive evaluation. The reconstructed image takes almost a full week to reproduce with existing computer power. However, because computer power and speeds follows Moore s Law, execution times are likely to become acceptable within the next five to eight years. The code was written in Mathematica for dedicated use with the ECVT system. In its present form, it is not suitable to be used directly as a consumer product. However, the code could be likely improved by rewriting it in a compiled language such as C or Fortran.

Nurge, Mark↗
DOE OSTI2016

Nanoscale Mechanical Behavior of Nuclear Materials

One of the primary performance objectives targeted for improved performance of nuclear fuels is fracture toughness. New materials or composite fuel architectures that offer greater resistance to cracking under the extreme environments encountered during nuclear reactor service would provide significant improvements to steady state (e.g. heat transfer, fuel redistribution) and transient (e.g. radionuclide release at elevated temperatures) conditions. Options currently under investigation include 'nontraditional' nuclear fuels designated around high uranium density. These include uranium silicides, uranium borides, and composite fuel materials constructed of these and uranium nitride or uranium dioxide. The increase uranium density compared with uranium dioxide (UO{sub 2}) has made them interesting to a new generation of nuclear fuels research driven by the renewed push for accident-tolerant light water reactor (LWR) fuels. Measurement of the thermophysical and thermodynamic properties of these constituent materials has progressed in recent years, but evaluation of mechanical properties at relevant temperatures is necessary to support further evaluations. While thermophysical properties such as thermal conductivity, heat capacity, and thermal expansion are generally the primary properties of interest, mechanical properties will play an important role in governing the deformation and fracture of the fuel during both normal operation and potential transients. It is therefore necessary to develop techniques capable of measuring these properties to predict the response of a candidate fuel form to anticipated thermal gradients. Nanoindentation techniques provide a demonstrated alternative to traditional mechanical test methods. Much work has also been carried out over the past few years to obtain stress-strain curves using a spherical nanoindentation approach. The techniques are expected to be useful for measurement of the mechanical properties of local structure of various materials. Nanomechanical test techniques have particular value in the area of nuclear materials, where numerous factors motivate limiting the sample volume to the extent possible. However, few attempts at application of nanoindentation to evaluate the mechanical properties of the nuclear fuels and cladding materials are reported in the literature. The effect of a limited set of variables (e.g. porosity, temperature dependence) on the mechanical properties of UO{sub 2} has been explored using early nanomechanical testing techniques. However, nanomechanical evaluation has not been extended to other nuclear fuels. In this work, we demonstrate the effectiveness of using instrumented spherical nanoindentation test techniques to capture the mechanical response of regions within individual grains in ion-irradiated W and Zr over a range of ion-irradiation conditions. As well as the nanomechanical response of U-Si and UO{sub 2} fuel materials Nanoindentation results obtained at room temperature will be used to express nano- and bulk- mechanical properties as a function of material stoichiometry. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Characterization of Mk-IV and EBR-II X441A Metallic Fuel Pins for the THOR-C-2 and THOR-M-TOP-1 Experiments: Results from Post-Transient Neutron Tomography of THOR-C-2 Capsule and Pre-transient Non-Destructive and Destructive Examination of DP 36 and DP 40 (Rev.1)

Current interest in sodium-cooled fast reactor designs, such as TerraPower’s Natrium Reactor, has highlighted the need for advanced-reactor fuel technology development. A Fuel Safety Research and Development (FSRD) program for metallic fast reactor fuels has been created to achieve comprehensive safety testing within the re-commissioned Transient Reactor Test (TREAT) facility at the Idaho National Laboratory. Despite over 60 years of metallic fuel irradiation, uncertainties exist in the performance of the fuel system, particularly under anticipated operational occurrences and severe accident scenarios. Throughout historical testing within the Experimental Breeder Reactor (EBR)-II and the Fast Flux Test Facility, fuel behavior has demonstrated benign response to transient reactor conditions; however, accurate predictions of failure thresholds rely heavily on fuel composition and irradiation history. In advancing the FSRD program, two planned transient heating experiments are at various stages of completion. The Temperature Heat sink Overpower Response (THOR)-C-2, fueled with an unirradiated Mk IV U-10Zr pin, has undergone transient irradiation in TREAT and post-transient three-dimensional neutron tomography. Additionally, pre-transient characterization of test and sibling U-19Pu-10Zr pins for THOR-M-TOP-1 was evaluated by both non-destructive and destructive methods. The test and sibling pins were selected from previously irradiated EBR-II experiment X441A. Both pins underwent visual examination, precise gamma spectrometry, two-dimensional neutron radiography, and element contact profilometry while the sibling pin was additionally subjected to sectioning and optical microscopy. THOR-C-2 radiography captured the fuel and cladding relocation during the intermediate transient at the top and bottom of the THOR capsule, allowing key features to be linked to the pin’s measured thermal response. For THOR M TOP 1, a solid baseline for steady-state behavior has been established. No anomalous features were identified in either the test or sibling pin. Defining characteristics and features were recorded for further comparisons.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Full core LOCA analysis for BWR/6 - Methodology and first results

As part of collaboration between the Paul Scherrer Institute (PSI) and the Swiss regulator ENSI, a study aimed at core-wide estimates of the fuel rod behaviour during large-break loss-of-coolant-accidents (LB-LOCA) in BWR/6 plants was recently conducted on the basis of a Swiss reactor. To perform such full core LOCA analysis, a methodology consisting of a multi-physics sequence of interfaced best-estimate models and codes was established: CASMO-4E/SIMULATE-3 (C4/S3) for the core modelling, TRACE for the plant system analysis and FALCON for the fuel thermomechanical calculations during base irradiation (BI) as well as for the transient analyses (TA). As starting point, reference C4/S3 models validated for all operated cycles of the given plant were used to select as situation target, a recent cycle representative of modern BWR core loadings and fuel assembly designs. Based on the selected C4/S3 cycle model, PSI in-house computational modules were developed or updated in order to set up the TRACE full core model with regards to assembly geometries, fuel rods thermophysical properties and initial 3-D power distributions. Also, an interface to FALCON BI models was established for the transfer of the irradiation history, including primarily linear heat generation rates (LHGR), for an equivalent fuel pin. For each assembly, the power of the equivalent pin is defined with the maximum power of all pins at each axial position. However, in order to reduce the computational time of the fuel performance analyses, the whole core was subdivided into groups of assemblies, each represented by the most powerful assembly within the group. On this basis, the FALCON BI analyses were conducted with coupling to the PSI advanced model for gas release and swelling (GRWS-A) and an interface was developed to transfer the calculated fuel rod characteristics at end of irradiation to initialize the core fuel rods heat structures in the TRACE model. Subsequently, double-ended recirculation line rupture scenarios at hot-full-power (HFP) conditions were investigated with TRACE in order to determine a conservative transient case as function of cycle burnup, single failure criterion and availability of safety systems. For the most conservative case of this study, an interface was set-up to transfer the TRACE results as boundary conditions to the FALCON TA models, including thus primarily time and axial distributions of the clad outer temperatures. As final step, the FALCON TA were conducted for the representative assemblies in order to infer core-wide estimates of the fuel rod behaviour with specific focus on equivalent cladding reacted levels, hoop strains as well as margins to failure criteria. The objective of this paper is to present this methodology along with the obtained first results. Regarding the latter, it is found that principally, even for the most conservative LOCA scenario of this study, the clad temperature excursion remains mild enough at all core locations such that no significant rod ballooning and no cladding failure is predicted, something in line with corresponding analyses conducted earlier for BWR/4 plants. (authors)

42 ENGINEERING↗
DOE OSTI2019

Thermal-hydraulic analyses of MIT reactor LEU transition cycles

The Massachusetts Institute of Technology Reactor (MITR) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. In the framework of non-proliferation policy, research and test reactors have started a program to convert highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. A new type of LEU fuel based on a high-density alloy of uranium and molybdenum (U-10Mo) is expected to allow conversion. A transition core plan, from 22 fresh LEU fuel elements to 24 in an equilibrium state, is proposed. This study evaluates the thermal-hydraulic safety margins of each transition cycle and state (i.e. beginning, middle and end). The STAT7 and RELAP5 codes are used in the study. STAT7 provides an integrated platform for statistical propagation of uncertainties in determining steady-state thermal-hydraulic margins. A RELAP5 model was created to verify MITR's STAT7 model, and provide transient analyses. The steady-state analyses determined that, during the entire transition from a fresh to equilibrium core, the Limiting System Safety Settings (LSSS) has a 10.2% margin to the minimum power at which onset of nucleate boiling (ONB) is precluded at 3σ confidence. The loss of flow (LOF) transient, which is the most limiting anticipated transient for the MITR, was simulated. With a conservative initial state of the reactor (operating at μ + 3σ power level and LSSS mass flow rate), the maximum calculated centerline fuel temperature was 115 degrees C, which is a significantly lower value than the blistering limit of 350 degrees C. The cladding wall temperatures did not exceed the ONB temperature throughout the transient. Therefore, during LOF, no nucleate boiling is expected, precluding any critical heat flux trigger. Overall, this study indicates that the proposed LEU transition core specifications have significant margin to thermal-hydraulic limits, during steady-state operation and the LOF transient.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Computational and Experimental Examination of Simulated Core Damage and Relocation Dynamics of a BWR Fuel Assembly - 16013

The Japan Atomic Energy Agency (JAEA) plans a large-scale test to evaluate damage and relocation behavior of Boiling Water Reactor (BWR) core materials consisting of fuel rods (zirconia as fuel simulant), channel boxes, control blade assemblies as well as lower core-support structures. Its purpose is to contribute to the understanding of the core material relocation behavior in the event of severe accidents with the BWR design conditions for which existing experimental database is quite limited. Prior to large-scale testing anticipated at JAEA in the future, JAEA desires preliminary investigations to examine melting test pieces, being composed of a mixture of zirconia and metals by a non-transferred thermal plasma heating system. The purpose of such tests is to verify the materials and test piece will be heated by plasma as a primary heat source to the target temperature of ca. 2900 K and to collect data about the relocation behavior of the high-temperature materials. This paper describes the first in a series of tests demonstrating the effective use of thermal plasma technology to recreate the dynamics of a typical fuel core bundle channel assembly section. Results from preliminary computational simulations of simplified fuel bundle assemblies are presented illustrating the effectiveness of a 150 kW, non transferred (NTR) configuration of a thermal plasma argon jet to heat the assembly within a highly-insulated containment vessel. An experimental test program using the computational analyses as a basis and a Phoenix Solutions Co (PSC) commercial model PT150, 150 kW plasma torch is described and the test results are presented illustrating the time versus temperature history of the fuel bundle assembly melting dynamics. Details on plasma operation are discussed including: power requirements, current flow, voltage levels achieved, argon gas flow rates, processing times and standoff distances of the torch discharge from the top of the fuel bundle assembly (achieved with remote control of the axial (vertical) translation of the torch within the containment processing chamber). Photos taken from the top of the processing chamber are presented illustrating the various phases of melting characteristics of the zircaloy-clad, zirconium oxide filled fuel tubes, all within a stainless steel tube bundle channel centered within an insulated zirconia crucible. Based on lessons learned from this initial test campaign, a brief discussion of the requirements and features of a follow-on Phase II test program is provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Portable Vibro-acoustic Testing System for In Situ Microstructure Characterization and Metrology

Portable Vibro-acoustic Testing System for In situ Microstructure Characterization and Metrology James A. Smith, Corrie I. Nichol Idaho National Laboratory, Idaho Falls ID Mostafa Fatemi, Mayo Clinic, Rochester MN There is a need in research reactors like the INL to inspect irradiated materials and structures. The goal of this work is to develop a portable scanning infrastructure for a material characterization technique called vibro-acoustography (VA) that has been developed by INL [1] for nuclear applications to characterize fuel, cladding materials, and structures. The proposed VA technology is based on ultrasound and acoustic waves; however, it provides information beyond what is available from the traditional ultrasound techniques and can expand the knowledge on nuclear material characterization and microstructure evolution. VA is a three-dimensional (3D) imaging modality based on ultrasound-stimulated acoustic emission. VA uses the force caused by the beating of two frequencies to generate an acoustic emission signal. Due to absorption or reflection, the energy density in the object at an acoustic focal point changes to produce a “radiation force.” This force locally vibrates the object, which results in an acoustic field that depends on the characteristics of the object at that point. This acoustic field is detected for every point in the object; the resulting data is used to make an image of the object’s mechanical properties. This technology is fundamentally different from traditional ultrasound imaging. More specifically, VA is: (a) based on the “radiation force,” a nonlinear phenomenon in sound propagation, and (b) sensitive to both high and low frequency dynamic responses of the object. The combination of these two response types provides information about characteristics of the object beyond what is possible by traditional ultrasound imaging. High spatial resolution is achieved, since the radiation force is localized to the focal point of the ultrasound beam. Low-frequency information, which is a representative of structural characteristics, is gained through the radiation force mechanism at a low (beat) frequency. This paper will report on the development of a portable scanning that can be set up to characterize materials and component in open water reactors and canals in situ. We will show some initial laboratory results of images generated by vibro-acoustics of surrogate fuel plates and graphite structures and discus the design of the portable system. 1. Smith, J. A., and Kunerth, D. C., “Vibro-acoustic Sensors for Materials Characterization and Related Methods and Systems,” Patent Application 13226280, 09/06/2011.

36 MATERIALS SCIENCE↗
DOE OSTI2025

A comparison of surrogate constitutive models for viscoplastic creep simulation of HT-9 steel

Mechanistic microstructure-informed constitutive models for the mechanical response of polycrystals are a cornerstone of computational materials science. However, as these models become increasingly more complex – often involving coupled differential equations describing the effect of specific deformation modes – their associated computational costs can become prohibitive, particularly in optimization or uncertainty quantification tasks that require numerous model evaluations. To address this challenge, surrogate constitutive models that balance accuracy and computational efficiency are highly desirable. Data-driven surrogate models, that learn the constitutive relation directly from data, have emerged as a promising solution. In this work, we develop two local surrogate models for the viscoplastic response of a steel: a piecewise response surface method and a mixture of experts model. These surrogates are designed to adapt to complex material behavior, which may vary with material parameters or operating conditions. The surrogate constitutive models are applied to creep simulations of HT-9 steel, an alloy of considerable interest to the nuclear energy sector due to its high tolerance to radiation damage, using training data generated from viscoplastic self-consistent (VPSC) simulations. In conclusion, we define a set of test metrics to numerically assess the accuracy of our surrogate models for predicting viscoplastic material behavior, and show that the mixture of experts model outperforms the piecewise response surface method in terms of accuracy.

36 MATERIALS SCIENCE↗
DOE OSTI2019

High temperature oxidation and microstructural evolution of cold spray chromium coatings on Zircaloy-4 in steam environments

Oxidation kinetics and microstructural evolution of cold sprayed Cr coatings on Zircaloy-4 at 1130–1310 °C in flowing steam at atmospheric pressure have been studied. The study is aimed at understanding of the response of Cr coated Zr-alloy under a steam environment and high temperatures pertinent to design basis accidents (DBAs) and beyond design basis accidents (BDBAs) in light water reactors (LWRs). Surface morphology, microstructure, and phases of post-oxidation test samples were characterized using Scanning Electron Microscopy (SEM), x-ray diffraction (XRD), and Scanning Transmission Electron Microscopy (STEM). Growth kinetics of the Cr-oxide scale and interdiffusion layers between the Cr coating and the Zr-alloy substrate were quantified from cross-sectional SEM images. Cross-sectional analysis showed that the Cr coatings offered a 50 times reduction in oxidation rate over bare Zircaloy-4 in a 1310 °C steam environment. Oxidation kinetics at 1130 °C followed parabolic law (i.e., n ~ 0.5 in power law kinetics) but at 1230 °C and 1310 °C the value of n was suppressed to below 0.5 possibly due to the volatilization of Cr species at the two highest temperatures. Interdiffusion at coating/substrate interface resulted in formation of a brittle Cr 2 Zr or Zr(Fe, Cr) 2 intermetallic compound layer on the order of micrometer in thickness and scattered Cr-rich precipitates were observed well below the interface within the Zr-alloy substrate after cooling. These experimental results could provide data to LWR system simulation codes for better estimation of coping time in the event of accidents.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Modeling DOE Standard Canister Configurations with Updated Surface Chemistry

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel currently specifies storage within helium backfilled DOE sealed standard canisters. The aluminum cladding of the ATR fuel contains an oxyhydroxide layer of boehmite/bayerite that generates hydrogen when subjected to irradiation. Understanding the effect of this hydrogen buildup over time to important for long term storage considerations. Previous modeling efforts have built a coupled CFD-chemical model to simulate the temperature gas phase concentrations within the DOE sealed standard canisters. These models have coupled the temperature conditions to both the gas phase radiolysis chemistry and the surface chemistry associated with the oxyhydroxide layer. Continued experimental work has identified trends for aluminum surrogate samples with oxyhydroxide layers for tests done at higher dose rates. At low initial doses a fast generation rate of hydrogen occurs which starts to roll over to a lower generation rate as the total dose applied in increased. A previous study created a small-scale chemical model was built to replicate a mini-canister surrogate system at SRNL as well as for the smaller capsule tests performed at INL. The previous iteration of the model utilized step function for its G-values for the hydrogen generation over a 200-year period. This model makes an update to the surface chemistry to account for theorized surface chemistry reactions allowing for oxygen to remain bounded to the oxyhydroxide layer. As additional experiments continue, the kinetic fits may be adjusted to adapt to new data. Three canister configurations are modeled – the base 18-inch, 15-foot DOR standard canister with 30 ATR fuel elements, an 18-inch, 10-foot DOE standard canister with 32 ATR fuel elements and a 24-inch, 10-foot DOE standard canister with 40 ATR fuel elements. In previous reports, the primary sensitivity of the canister conditions was identified as the decay heat of the fuel and the dried condition of the fuel. Only these parameters are studied in the report. For the nominal case with dried fuel, the hydrogen generation is only about 2%, so it is even less than the flammability limit if it were exposed to oxygen. For the densely packed 18-inch case the total hydrogen concentration is only around 4% with the nominal decay heat. For the densely packed 24-inch case the total hydrogen concentration is only 2.5%, roughly 20% higher than the original packing design, and still under flammability conditions if exposed to oxygen.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗