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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 55 records · Page 3

Characterization of Heat Transfer in a Trapped-Vortex Combustor Designed for High-Temperature Material Testing

Next generation gas turbines will reduce carbon emissions by improving their cycle efficiencies and implementing hydrogen as a low-carbon fuel. The combined-cycle efficiency can be improved by increasing the combustor outlet temperature and decreasing the cooling air required for hot-section parts. Ultra-high temperature ceramic matrix composites (CMC) can survive increasingly extreme environments due to their strong mechanical properties at expected conditions. The introduction of CMCs in high-hydrogen flame environments requires testing at realistic conditions that mimic the thermochemical and fluid mechanic states in gas-turbine combustors. In this work, we describe a new experiment that has been designed to test high-temperature materials in a combustor-relevant environment. Its trapped-vortex combustor chamber design allows for high levels of fuel flexibility and wide flame stability limits. Testing of monolithic silicon carbide (SiC) samples in this environment was done to characterize the heat transfer to material samples over a wide range of operating conditions, including variations in fuel composition and thermal power. Heat flux measurements were verified in non-reacting and reacting environments after which they were benchmarked with previous literature to verify the operation of the facility with heated air. Testing in combustor-relevant conditions shows high levels of heat flux to the material samples in the combustor, particularly downstream of the anchored flame.

Richins, Porter [Pennsylvania State University , ,↗

Effect of surface oxidation on the onset of nucleate boiling in a materials test reactor coolant channel

The onset of nucleate boiling (ONB) serves as the thermal-hydraulic operating limit for many research and test reactors. However, boiling incipience under forced convection has not been well-characterized in narrow channel geometries or for oxidized surface conditions. This study presents experimental data for the ONB in vertical upflow of deionized (DI) water in a simulated materials test reactor (MTR) coolant channel. The channel gap thickness and aspect ratio were 1.96 mm and 29:1, respectively. Boiling surface conditions were carefully controlled and characterized, with both heavily oxidized and native oxide surfaces tested. Measurements were performed for mass fluxes ranging from 750 to 3000 kg/m 2 s and for subcoolings ranging from 10 to 45°C. ONB was identified using a combination of high-speed visual observation, surface temperature measurements, and channel pressure drop measurements. Surface temperature measurements were found to be most reliable in identifying the ONB. For the nominal (native oxide) surface, results indicate that the correlation of Bergles and Rohsenow, when paired with the appropriate single-phase heat transfer correlation, adequately predicts the ONB heat flux. Furthermore, incipience on the oxidized surface occurred at a higher heat flux and superheat than on the plain surface.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Dissolution of Material Test Reactor Fuel in an H-Canyon Dissolver

In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO 3 flowsheet. Since the development of the existing flow sheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

JHR Project: a future Material Testing Reactor working as an International user Facility: The key-role of instrumentation in support to the development of modern experimental capacity

Research and development on fuel and material behaviour under irradiation is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. These needs mainly deal with a constant improvement of performances and safety in order to optimize the fuel cycle and hence to reach nuclear energy sustainable objectives. A sustainable nuclear energy requires a high level of performances in order to meet specific needs such as: - Pursuing improvement of the performances and safety of present and coming water cooled reactor technologies. This will require a continuous R and D support following a long-term trend driven by the plant life management, safety demonstration, flexibility and economics improvement. Experimental irradiations of structure materials are necessary to anticipate these material behaviours and will contribute to their optimisation. - Upgrading continuously nuclear fuel technology in present and future nuclear power plants to achieve better performances and to optimise the fuel cycle keeping the best level of safety. Fuel evolution for generation II, III and III+ is a key stake requiring developments, qualification tests and safety experiments to ensure the competitiveness and safety: experimental tests exploring the full range of fuel behaviour determine fuel stability limits and safety margins, as a major input for the fuel reliability analysis. To perform such accurate and innovative progress and developments, specific and ad hoc instrumentation, irradiation devices, measurement methods are necessary to be set up inside or beside the material testing reactor (MTR) core. These experiments require beforehand in situ and on line sophisticated measurements to accurately determine different key parameters such as thermal and fast neutron fluxes and nuclear heating in order to precisely monitor and control the conducted assays. The new Material Testing Reactor JHR (Jules Horowitz Reactor) currently under construction at CEA Cadarache research centre in the south of France will represent a major Research Infrastructure for scientific studies regarding material and fuel behavior under irradiation. It will also be devoted to medical isotopes production. Hence JHR will offer a real opportunity to perform R and D programs regarding needs above and hence will crucially contribute to the selection, optimization and qualification of these innovative materials and fuels. The JHR reactor objectives, principles and main characteristics associated to specific experimental devices associated to measurement techniques and methodology, their performances, their limitations and field of applications will be presented and discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Nuclear characterization of a general-purpose instrumentation and materials testing location in TREAT

The Transient Reactor Test facility (TREAT) was constructed in the late 1950s, provided thousands of transient irradiations before being placed in standby in 1994, and resumed operations in 2017 in order to reclaim its crucial role in nuclear-heated safety research. The latter half of TREAT’s historic operation was best known for integral-scale testing of fuel specimens under postulated reactor plant accident conditions, while TREAT’s earlier history included extensive simpler phenomena identification tests that elucidated fundamental behaviors and paved the way for these integral-scale tests. Advances in modern computational capabilities and a resurgence of interest in novel reactor technology have created an opportunity for emphasizing modernized science-based and separate effects test capabilities once again at TREAT. An innovative approach to this type of testing has been developed to leverage minor radioactivity built-in during brief TREAT irradiations in low activation hardware to facilitate handling for materials and instrumentation testing. This capability, termed the Minimal Activation Retrievable Capsule Holder (MARCH) irradiation vehicle system, will be used for inaugural fueled experiments in TREAT’s modern era as well as novel approaches to study materials undergoing neutron irradiation and instrumentation development and qualifications. This paper describes a comprehensive nuclear characterization, obtained via computational modeling, of the test position in the MARCH system. Though directly applicable to the MARCH system, the results also provide general quantification of the nuclear performance of the reactor and potential test materials, crucial for evaluating potential experiment design and response in TREAT. The neutron and photon flux environment was calculated via MCNP with ENDF/B-VII.1 nuclear data. Wire heating rate and DPA calculations were also performed and evaluated using historic TREAT test data. Furthermore, these calculations were performed to provide steady-state baseline reference values typical in both half- and full-slotted TREAT core configurations, enabling design scoping analysis prior to development of more specific design testing needs and transient testing experimentation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Survey of U.S. Research Reactor Auxiliary Facilities Used for Material Testing and Basic Neutron Science

A catalogue of and attributes of the Materials Irradiation and Testing facilities (MIF) and the Basic Neutron Science facilities (BSF) of auxiliary (AUX) facilities of research reactors are compiled. The survey of these facilities is drawn from the set of U.S. university and DOE research reactor facilities and should be considered as a reference point when comparing commonalities in international AUX facilities. The size and shielding capabilities of the facilities, the typical characterization equipment used, and the neutron flux and irradiation capabilities of the facilities are listed. Short descriptions of experimental activities and current practices are detailed for MIFs and BSFs. The attributes are important in consideration of the reconfiguration of the facilities for purposes other than stated mission, i.e., for proliferation of weapons-usable nuclear material (WUNM). An evaluation of the capacity for production rate of WUNM from neutron beams-on-targets or a sample that has been placed in an irradiation position for a period is provided. Further evaluation of reconfiguration of these MIF and BSF are recommended to refine the proliferation risks. Specifically, MIF hot cells and potential configurations, additional modeling of facility throughput and development of methods for determining levels of concern for MIF and BSF have all been identified as steps for refinement of determination of the risks associated with these facilities. Future work will focus on: 1. A model of projected throughput for different MIF and BSF configurations that will be used to provide a window of potential operational misuse for a facility and better understand the capability of production and the rate of processing for WUNM in these facilities. 2. A misuse study for various configurations of MIF hot cell layouts detailing capacity, specifications, shielding constraints for MIF operations, identification of a maximum shielding thickness, and reasonable need associated with each activity performed in a MIF. 3. Development of a rule of thumb rating system utilizing data gathered from previous reports to show levels of concern at a given power level, flux, and experiment set.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Status and Planned Experiments of the Hiradmat Pulsed Beam Material Test Facility at CERN SPS

HiRadMat (High Irradiation to Materials) is a facility at CERN designed to provide high-intensity pulsed beams to an irradiation area where material samples as well as accelerator component assemblies (e.g. vacuum windows, shock tests on high power targets, collimators) can be tested. The beam parameters (SPS 440 GeV protons with a pulse energy of up to 3.4 MJ, or alternatively lead/argon ions at the proton equivalent energy) can be tuned to match the needs of each experiment. It is a test area designed to perform single pulse experiments to evaluate the effect of high-intensity pulsed beams on materials in a dedicated environment, excluding long-time irradiation studies. The facility is designed for a maximum number of 1016 protons per year, in order to limit the activation of the irradiated samples to acceptable levels for human intervention. This paper will demonstrate the possibilities for research using this facility and go through examples of upcoming experiments scheduled in the beam period 2015/2016.

43 PARTICLE ACCELERATORS↗

Integrated FHR technology development: Tritium management, materials testing, salt chemistry control, thermal hydraulics and neutronics, associated benchmarking and commercial basis

This report summarizes the results of a 3-year Integrated Research Project (IRP) sponsored by the U.S. Department of Energy to address major technical challenges in the development of the Fluoride-salt-cooled High-temperature Reactor (FHR). The IRP universities included the Massachusetts Institute of Technology (MIT), the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The FHR is a new reactor concept less than 20 years old that combines (1) a clean fluoride salt coolant, (2) the graphite-matrix coated-particle fuel originally developed for High- Temperature Gas-Cooled Reactors and (3) passive decay heat removal systems from sodium fast reactors. The base-line design is a pebble-bed FHR that delivers heat between 600 to 700°C with a 7 Li 2 BeF 4 (flibe) salt coolant to the power cycle or industrial users. The Executive Summary describes what was accomplished and summarizes major conclusions. The main report contains stand-alone sections in each major area of work. Highlights are summarized below. Tritium. The coolant salt produces tritium necessitating understanding and control of tritium. Experimental work included measuring hydrogen uptake on carbon, tritium uptake during irradiation of graphite in 700°C salt and the permeability of tritium through metals with and without coatings. An FHR systems model was developed to predict tritium behavior and enable design of tritium control systems. The leading candidate for tritium removal from the flibe salt is carbon bed outside the reactor core. Thermal hydraulics and Neutronics. A 10-meter high FHR integral effects test facility, using Dowtherm A®, has been built and used to simulate the integral response of a representative FHR to improve understanding of FHR thermal hydraulics and validate thermal hydraulics codes used for the FHR. Neutronic analysis included code-to-code benchmarking, multiphysics modeling, uncertainty analysis for nuclear data uncertainties and shielding analysis. Twisted-tube heat exchangers were examined that may provide significant cost reductions because of their improved heat-transfer performance with liquid salts. The impacts on safety and design caused by uncertainties in physical properties of the salt were assessed with recommendations on where added physical property measurements are needed. Work began on understanding radiative heat transfer that becomes important in salt systems at higher temperatures. Materials. Static corrosion tests were conducted on graphite and materials in 700°C salt -- both in the UW laboratory and in the MIT reactor. In addition to static corrosion testing, a circulating salt corrosion loop was built and is operating. Methods to measure salt properties including redox were developed. The presence of carbon increases corrosion. With good redox control, the evidence indicates stainless steel can be used as a material of construction. Benchmarking. An additional series of workshops where held to enable benchmarking and comparison of results between different investigators. Commercialization. The FHR delivers higher average-temperature (~650°C) heat to the power cycle or industry than other reactors. This enables power cycles with lower-cost heat storage for variable electricity to the grid while the reactor operates at base-load -- providing a major competitive advantage by maximizing revenue from electricity sales. In October, 2016, a new startup company named Kairos Power was formed, and opened its first office in Jack London Square, Oakland, California, in January 2017. Kairos Power exists largely due to the work that was accomplished by this IRP and the previous DOE-funded FHR IRP that performed research on FHR technology from 2012 to 2015. In September, 2018, as this IRP is concluding, Kairos Power was staffed at 60 full time employees, working to develop technology derived from this IRP. Many of these employees worked with the IRP before joining Kairos Power 1 (https://kairospower.com). Other students and researchers from this IRP have joined other startup companies, including the Terrapower Molten Chloride Fast Reactor and the Terrestrial Energy development efforts.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Markforged Continuous Fiber Composite Material Testing

Continuous fiber composite 3D printing promises to greatly expand the design space of polymer additive manufacturing (AM). A series of mechanical tests were performed to attempt to fully characterize the composite materials built from feedstock provided by the Markforged company. These tests included tensile, compression, Charpy impact, three point bending, coefficient of thermal expansion, and calorimetry tests. Specimens were prepared on the Mark Two 3D printer at LANL during the spring and summer of 2019. These samples were then shipped to New Mexico Institute of Mining and Technology’s Thermo-Mechanical Lab for testing during the summer and fall of 2019. Test data are reported and analyzed for each of the tests.

36 MATERIALS SCIENCE↗

Optimization of Advanced Steels for Cyclic Operation through an Integration of Material Testing, Modeling and Novel Component Test Validation

The efforts in this project were focused in three primary objectives, provided below: • To develop the needed microstructural processing and performance relationships and associated material models for specific constituents in fabricated weldments (such as the parent material, heat affected zone regions and weld metal), • Apply these metallurgical relationships through modeling of a welded pressure bearing power plant component subjected to cyclic operational conditions under both mechanical and thermal loading, and • Validate the model through novel structural feature and component tests. An attempt was made to optimize the heat treatment schedule to reduce or eliminate the presence of boron nitride (BN) in the matrix of the Grade 92 material. BN has been shown to be a primary nucleation source for creep damage. Proper evaluation of material behavior, and as assessed in this research, included the influence of geometric notches (such as those from poor fabrication or design) and metallurgical notches (such as weldments or poorly processed parent material). Better practice methodologies were implemented in the theoretical design of a high-temperature, critical component for cyclic operation. It was demonstrated that vessel testing can be reliably performed and provides a critical link between standard testing, structural assessment and validation. While the goal of an improved Grade 92 steel was not achieved in this project, there is considerable value in applying the documented testing and assessment methodologies to assess material and/or component behavior. This is especially true for materials and/or designs which are expected to be subjected to flexible operating modes. Design of a cycling state of the art thick wall steam header for long term operation with a state of the art ferritic steel was a task in this project. Implementation of that task showed how many factors including plant operating mode, creep strength, creep ductility, fatigue strength, and resistance to crack growth should be considered and evaluated to properly design high temperature components for long term service (200,000 operating hours or more) in a modern Fossil energy power plant. The current ASME non-nuclear boiler design code does not mandate this level of complex analysis nor does it provide all the metallurgical and mechanical properties data needed to perform this kind of design analysis.

20 FOSSIL-FUELED POWER PLANTS↗

Low Activity Waste Tuning Feed Material Testing; Qualification and Transition Testing (Final Report)

The primary objective of this work was to collect data to support processing of the proposed LAW tuning feed composition at the WTP. This was achieved through assessment of the modified LAWB83 feed and glass composition by feed rheology testing and by processing selected LAW feeds through a DM100 melter to confirm that the tuning feed material can be fed to the melter with acceptable commercial or hazards risk. The testing also confirmed that transitions from the startup frit to the LAWB83 tuning feed and then to the cold commissioning simulant composition can be performed without causing any significant melter performance issues.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Summary of Blast Shield and Material Testing for Development of Solid Debris Collection at the National Ignition Facility (NIF)

The ability to collect solid debris from the target chamber following a NIF shot has application for both capsule diagnostics, particularly for fuel-ablator mix, and measuring cross sections relevant to the Stockpile Stewardship program and nuclear astrophysics. Simulations have shown that doping the capsule with up to 10 15 atoms of an impurity not otherwise found in the capsule does not affect its performance. The dopant is an element that will undergo nuclear activations during the NIF implosion, forming radioactive species that can be collected and measured after extraction from the target chamber. For diagnostics, deuteron or alpha induced reactions can be used to probe the fuel-ablator mix. For measuring neutron cross sections, the dopant should be something that is sensitive to the 14 MeV neutrons produced through the fusion of deuterium and tritium. Developing the collector is a challenge due to the extreme environment of the NIF chamber. The collector surface is exposed to a large photon flux from x-rays and unconverted laser light before it is exposed to a debris wind that is formed from vaporized material from the target chamber center. The photons will ablate the collector surface to some extent, possibly impeding the debris from reaching the collector and sticking. In addition, the collector itself must be mechanically strong enough to withstand the large amount of energy it will be exposed to, and it should be something that will be easy to count and chemically process. In order to select the best material for the collector, a variety of different metals have been tested in the NIF chamber. They were exposed to high-energy laser shots in order to evaluate their postshot surface characterization, morphology, degree of melt, and their ability to retain debris from the chamber center. The first set of samples consisted of 1 mm thick pieces of aluminum that had been fielded in the chamber as blast shields protecting the neutron activation diagnostic. Ten of these pieces were fielded at the equator and one was fielded on the pole. The shields were analyzed using a combination of scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), x-ray fluorescence (XRF), neutron activation analysis (NAA) and chemical leaching followed by mass spectrometry. On each shield, gold debris originating from the gold hohlraum was observed, as well as large quantities of debris that were present in the center of the target chamber at the time of the shot (i.e., stainless steel, indium, copper, etc.) Debris was visible in the SEM as large blobs or splats of material that had encountered the surface of the aluminum and stuck. The aluminum itself had obviously melted and condensed, and some of the large debris splats arrived after the surface had already hardened. Melt depth was determined by cross sectioning the pieces and measuring the melted surface layers via SEM. After the SEM analysis was completed, the pieces were sent for NAA at the USGS reactor and were analyzed by U. Greife at the Colorado School of Mines. The NAA showed that the majority of gold mass present on the shields was not in the form of large blobs and splats, but was present as small particulates that had most likely formed as condensed vapor. Further analysis showed that the gold was entrained in the melted aluminum surface layers and did not extend down into the bulk of the aluminum. Once the gold mass was accounted for from the NAA, it was determined that the aluminum fielded at the equator was collecting a fraction of the total gold hohlraum mass equivalent to 120% ± 10% of the solid angle subtended by the shield. The attached presentation has more information on the results of the aluminum blast shield analysis. In addition to the information given in the presentation, the surfaces of the shields have been chemically leached and submitted for mass spectrometric analysis. The results from that analysis are expected to arrive after the due date of this report and will be written up at a later time. Based on the results of the aluminum blast shield analysis, it was determined that additional materials needed to be tested as potential collectors in the NIF chamber. 1-2 mm thick pieces of tantalum, niobium, vanadium, silver, titanium, molybdenum, and graphite foil were fielded in the Wedge Range Filter (WRF) mount at a distance of 50 cm from target chamber center during the shock timing campaign. The pieces were subsequently removed and analyzed in a similar fashion to the aluminum shields. As of this writing, the pieces are still under analysis, but initial results indicate that gold debris was collected on the various materials. Currently, the pieces are being cross-sectioned so that the melt depths of each material can be compared. In addition, NAA and/or mass spectrometry will be performed in order to determine the total gold mass that was collected on each surface.

36 MATERIALS SCIENCE↗

Low Activity Waste Tuning Feed Material Testing VSL-18R4350-1 (Final Report)

High level waste (HLW) from the Hanford tank farms will be retrieved and transferred to the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The waste will be pretreated to separate the soluble chemical salts (mostly sodium nitrate/nitrite and sodium hydroxide) from the bulk of the radioactive materials. This decontaminated salt solution is the low activity waste (LAW) stream which constitutes most of the total waste volume. The remaining HLW contains most of the radioactive materials but a fraction of the volume. Each of these waste streams is directed to an independent processing facility where the waste is mixed with glass forming chemicals (mostly silica and borax or boric acid) and fed into the melters for stabilization by conversion into glass. The molten glass is poured into stainless steel containers to produce packages for disposal: local shallow burial for the LAW containers and a future geologic repository for the HLW containers. The LAW facility melters produce significant quantities of NOx-laden off-gas that require abatement in accordance with 10 CFR 830 and air emission requirements. The NOx emissions also pose a safety risk. The LAW facility commissioning sequence requires that the melters be operated and process control loops tuned prior to introducing waste or waste feed materials that produce NOx. Therefore, a temporary or permanent system architecture is needed that provides a feed supply to the melters for the purpose of melter testing and off-gas tuning that does not result in the production of significantly hazardous off-gas products.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Nd and Sm isotopic composition of spent nuclear fuels from three material test reactors

Rare earth elements such as neodymium and samarium are ideal for probing the neutron environment that spent nuclear fuels are exposed to in nuclear reactors. The large number of stable isotopes can provide distinct isotopic signatures for differentiating the source material for nuclear forensic investigations. The rare-earth elements were isolated from the high activity fuel matrix via ion exchange chromatography in a shielded cell. The individual elements were then separated using cation exchange chromatography. In conclusion, the neodymium and samarium aliquots were analyzed via MC–ICP–MS, resulting in isotopic compositions with a precision of 0.01–0.3%.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Material Testing for Shear-Dominated Ductile Failure

An initial foray into the design of specimens that can be used to provide data about the quasistatic ductile failure of metals when subjected to shear-dominated (low triaxiality) states of stress was undertaken. Four specimen geometries made from two materials with different ductility (Al 7075, lower ductility and steel A286, higher ductility) were considered as candidates. Based on results from analysis and experimentation, it seems that two show promise for further consideration. Whereas preliminary results indicate that the Johnson-Cook model fit the failure data for Al 7075 well, it did not fit the data for steel A286. Further work is needed to consolidate the results and evaluate other failure models that may fit the steel data better, as well as to extend the results of this work to the dynamic loading regime.

36 MATERIALS SCIENCE↗

Enhancing activated-peroxide formulations for porous materials: Test methods and results

During an urban wide-area incident involving the release of a biological warfare agent, the recovery/restoration effort will require extensive resources and will tax the current capabilities of the government and private contractors. In fact, resources may be so limited that decontamination by facility owners/occupants may become necessary and a simple decontamination process and material should be available for this use. One potential process for use by facility owners/occupants would be a liquid sporicidal decontaminant, such as pHamended bleach or activated-peroxide, and simple application devices. While pH-amended bleach is currently the recommended low-tech decontamination solution, a less corrosive and toxic decontaminant is desirable. The objective of this project is to provide an operational assessment of an alternative to chlorine bleach for low-tech decontamination applications activated hydrogen peroxide. This report provides the methods and results for activatedperoxide evaluation experiments. The results suggest that the efficacy of an activated-peroxide decontaminant is similar to pH-amended bleach on many common materials.

36 MATERIALS SCIENCE↗

Material Testing of PH13-8MO H950 Steel for Xue-Wierzbicki Fracture Criterion Determination

Four tensile coupon designs of PH13-8Mo H950 steel were tested to failure using quasi-static rates to obtain data to calibrate the Xue-Wierzbicki failure model for ductile fracture. The tests recorded the force-displacement, location of the first crack, displacement to fracture, area reduction, and crack propagation path. The test method and coupon designs were adopted from Tomasz Wierzbicki’s “Calibration and evaluation of seven fracture models” report. The XueWierzbicki model predicts fracture based on accumulated equivalent plastic strain, stress triaxiality, and deviatoric state parameter. Calibrating the Xue-Wierzbicki failure model required testing four coupon designs to calculate the four free parameters in the model. The coupon designs tested a range of stress triaxialities with two axisymmetric tests, one shear test, and one plane stress test. The data obtained and presented in this report can be used to develop a Xue-Wierzbicki fracture model for PH13-8Mo H950 steel.

36 MATERIALS SCIENCE↗