Engineering topics

Meyer, Mitch

Publications and source records attributed to Meyer, Mitch.

10 records

DOE OSTI2017

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

LDRD Strategic R&D Fund – Fiscal Year 2018 Research Priorities

Proposals are now being accepted for the Fiscal Year 2018 Laboratory Directed Research and Development (LDRD) Director’s Strategic Research and Development (R&D) Fund. This call seeks high-quality proposals that address the Laboratory’s research priorities identified within the attached document LDRD Strategic R&D Fund - Fiscal Year 2018 Research Priorities.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2015

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

Low-enrichment (U-235 < 20%) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing was comprised of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates that were tested in INL's Advanced Test Reactor (ATR) were subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. Adjacent to the AA6061 cladding were Mg-rich precipitates, which was in close proximity to the region where Xe is observed to be enriched. In samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface were possible indications of porosity/debonding, which suggested that the interface in this location is relatively weak.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Thermal Properties Measurement Report

The Thermal Properties Measurement Report summarizes the research, development, installation, and initial use of significant experimental thermal property characterization capabilities at the INL in FY 2015. These new capabilities were used to characterize a U 3 Si 2 (candidate Accident Tolerant) fuel sample fabricated at the INL. The ability to perform measurements at various length scales is important and provides additional data that is not currently in the literature. However, the real value of the data will be in accomplishing a phenomenological understanding of the thermal conductivity in fuels and the ties to predictive modeling. Thus, the MARMOT advanced modeling and simulation capability was utilized to illustrate how the microstructural data can be modeled and compared with bulk characterization data. A scientific method was established for thermal property measurement capability on irradiated nuclear fuel samples, which will be installed in the Irradiated Material Characterization Laboratory (IMCL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Synthesis and sintering of UN-UO 2 fuel composites

In this study, the design and development of an economical, accident tolerant fuel (ATF) for use in the current light water reactor (LWR) fleet is highly desirable for the future of nuclear power. Uranium mononitride has been identified as an alternative fuel with higher uranium density and thermal conductivity when compared to the benchmark, UO 2 , which could also provide significant economic benefits. However, UN by itself reacts with water at reactor operating temperatures. In order to reduce its reactivity, the addition of UO 2 to UN has been suggested. In order to avoid carbon impurities, UN was synthesized from elemental uranium using a hydride-dehydride-nitride thermal synthesis route prior to mixing with up to 10 wt% UO 2 in a planetary ball mill. UN and UN – UO 2 composite pellets were sintered in Ar – (0–1 at%) N 2 to study the effects of nitrogen concentration on the evolved phases and microstructure. UN and UN-UO 2 composite pellets were also sintered in Ar – 100 ppm N 2 to assess the effects of temperature (1700–2000 °C) on the final grain morphology and phase concentration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

RERTR-12 Post-irradiation Examination Summary Report

The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-property changes of the fuel and cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Scanning Electron Microscopy Analysis of Fuel/Matrix Interaction Layers in Highly-Irradiated U–Mo Dispersion Fuel Plates with Al and Al–Si Alloy Matrices

In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U–7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifially, samples from irradiated U–7Mo dispersion fuel elements with pure Al, Al–2Si and AA4043 (~4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U–7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission-gas bubbles. Additionally, solid-fission-product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U–7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al–Si matrices.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Report on Potential Advantages and Uncertainties of the Thorium Seed Blanket Unit Fuel Concept

Thorium is a naturally occurring radioactive element with potential application in the nuclear fuel cycle. Thorium has no fissile isotopes, meaning that unlike uranium it cannot be enriched to produce a viable nuclear fuel. Natural thorium is almost entirely comprised of the isotope 232 Th and a microscopic amount of 228 Th that is created from the radioactive decay of 232 Th. Small amounts of other thorium isotopes can be detected from the decay of uranium that often occurs with thorium in nature. During the early stages of the development of nuclear power in the United States it was believed that uranium supplies would be insufficient and that breeding would be required to fuel the envisioned nuclear infrastructure. The use of 232 Th in the nuclear fuel cycle would play a role equivalent to 238 U. Both elements are fertile and can absorb a neutron to transmute into a fissile isotope capable of sustaining a chain reaction. In the case of 238 U the fissile isotope produced is 239 Pu. 232 Th produces the less well-known fissile isotope 233 U. This report provides a brief overview of light water reactor thorium technology, and an assessment of the technology proposed by Lightbridge based on available open literature and information provided by Lightbridge Corporation. It serves to identify gaps in the current understanding of this technology, and to identify areas where further work may be required or beneficial. Analysis of the fuel cycle was not included in the scope of this report. Based on the data available and the scope of this assessment, this document cannot draw definitive conclusions about this particular technology, and should not be understood to be comprehensive or to make a recommendation, either positive or negative, on the suitability of this fuel cycle for commercial use.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2011

Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting

The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2010

Materials challenges for nuclear systems

The safe and economical operation of any nuclear power system relies to a great extent, on the success of the fuel and the materials of construction. During the lifetime of a nuclear power system which currently can be as long as 60 years, the materials are subject to high temperature, a corrosive environment, and damage from high-energy particles released during fission. The fuel which provides the power for the reactor has a much shorter life but is subject to the same types of harsh environments. This article reviews the environments in which fuels and materials from current and proposed nuclear systems operate and then describes how the creation of the Advanced Test Reactor National Scientific User Facility is allowing researchers from across the U.S. to test their ideas for improved fuels and materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗