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DOE OSTI2022

Automatic gas sorption apparatus and method

Automated apparatuses for use in examining gas sorption materials are described. Devices utilize a noncontact magnetic induction heating approach for controlling the temperature of tested materials. The apparatuses can be used to generate single or multiple isotherms simultaneously. The apparatuses can examine nanogram or microgram-scale quantities of materials of interest and can do so automatically and unattended. Pressure-composition isotherms can be provided through use of disclosed apparatuses in a period of a few hours.

Hunyadi Murph, Simona E.↗

DOE OSTI2015

Trusted materials using orthogonal testing. 2015 Annual report

The purpose of this project is to prove (or disprove) that a reasonable number of simple tests can be used to provide a unique data signature for materials, changes in which could serve as a harbinger of material deviation, prompting further evaluations. The routine tests are mutually orthogonal to any currently required materials specification tests.

36 MATERIALS SCIENCE↗

DOE OSTI2022

Materials properties characterization in the most extreme environments

Abstract There is an ever-increasing need for material systems to operate in the most extreme environments encountered in space exploration, energy production, and propulsion systems. To effectively design materials to reliably operate in extreme environments, we need an array of tools to both sustain lab-scale extreme conditions and then probe the materials properties across a variety of length and time scales. Within this article, we examine the state-of-the-art experimental systems for testing materials under extreme environments and highlight the limitations of these approaches. We focus on three areas: (1) extreme temperatures, (2) extreme mechanical testing, and (3) chemically hostile environments. Within these areas, we identify six opportunities for instrument and technique development that are poised to dramatically impact the further understanding and development of next-generation materials for extreme environments. Graphical abstract

Schreiber, Daniel K.↗

DOE OSTI2016

Quantitative Three-Dimensional Imaging of Heterogeneous Materials by Thermal Tomography

Infrared thermal imaging based on active thermal excitations has been widely used for nondestructive evaluation ( NDE) of materials. While the experimental systems have remained essentially the same during the last few decades, development of advanced data-processing methods has significantly improved the capabilities of this technology. However, many limitations still exist. One fundamental limitation is the requirement, either explicitly or implicitly, of the tested material to be homogeneous such that detected thermal contrasts may be used to determine an average material property or attributed to flaws. In this paper, a new thermal tomography ( TT) method is introduced, which for the first time can evaluate heterogeneous materials by directly imaging their thermal-property variations with space. It utilizes one-sided flash thermal-imaging data to construct the three-dimensional ( 3D) distribution of thermal effusivity in the entire volume of a test sample. Theoretical analyses for single and multilayer material systems were conducted to validate its formulation and to demonstrate its performance. Experimental results for a ceramic composite plate and a thermal barrier coating ( TBC) sample are also presented. It was shown that thermal diffusion is the primary factor that degrades the spatial resolution with depth for TT; the spatial resolutions in the lateral and axial directions were quantitatively evaluated.

47 OTHER INSTRUMENTATION↗

DOE OSTI2022

Modeling atomically mixed graded density impactors

Graded density impactors (GDIs) are multi-material composite impactors used in gas gun experiments to tailor the drive conditions imparted to a sample test material. Previous graded density impactors generally rely on thin, but discrete, layers of different materials. The thinner and the greater number of layers will result in smoother compression. Taken to the limit of very thin layers would be pure material 1 at one surface, such as the front surface of an impactor, smoothly transitioning at the atomic scale to pure material 2 on the back surface. Such an impactor can initially shock, then smoothly compresses a material during a dynamic experiment. This type of experiment can serve to explore a larger region of thermodynamic space than a single or even multi-shock experiments. An overview of how graded density impactors are made is reviewed and sample results are given. Additionally, a strategy for modeling these kinds of impactors is presented. The length scales of constituent mixing are given from the experimental build through electrochemical-deposition. Equation of state models for pure constituents and their subsequent mixtures are presented. It is demonstrated that the time scales for pressure and temperature equilibration, for atomically mixed GDIs, are short enough to be a justifiable closure for the resulting multiphase flow. Furthermore, we present simulation results of dynamic shock followed by a ramp compression, utilizing a silver/gold graded density impactor, onto a tantalum sample.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

DOE OSTI2018

Soft Materials Solutions for Future Needs

Background: Rapid heating of polymer components can have catastrophic effects, such as contained vaporization, which are poorly understood. Hostile environment effects on stockpile-aged or new materials need study. Need: Elucidate chemical and physical changes using scientifically rigorous surrogate stimuli, and in situ techniques. Bulk and surface phenomena must be distinguished. Required diagnostics are impractical at large user facilities. Goal and Vision: Identify failure mechanisms and test materials of diverse pedigrees (stockpile, candidate replacement) using a local, adaptable, high throughput testing capability, integrated with modeling.

36 MATERIALS SCIENCE↗

DOE OSTI2012

Fusion Nuclear Schience Facility-AT: A Material And Component Testing Device

A Fusion Nuclear Science Facility (FNSF) is a necessary complement to ITER, especially in the area of materials and components testing, needed for DEMO design development. FNSF-AT, which takes advantage of advanced tokamak (AT) physics should have neutron wall loading of 1-2 MW/m2, continuous operation for periods of up to two weeks, a duty factor goal of 0.3 per year and an accumulated fluence of 3-6 MW-yr/m2 (~30-60 dpa) in ten years to enable the qualification of structural, blanket and functional materials, components and corresponding ancillary equipment necessary for the design and licensing of a DEMO. Base blankets with a ferritic steel structure and selected tritium blanket materials will be tested and used for the demonstration of tritium sufficiency. Additional test ports at the outboard mid-plane will be reserved for test blankets with advanced designs or exotic materials, and electricity production for integrated high fluence testing in a DT fusion spectrum. FNSF-AT will be designed using conservative implementations of all elements of AT physics to produce 150-300 MW fusion power with modest energy gain (Q<7) in a modest sized normal conducting coil device. It will demonstrate and help to select the DEMO plasma facing, structural, tritium breeding, functional materials and ancillary equipment including diagnostics. It will also demonstrate the necessary tritium fuel cycle, design and cooling of the first wall chamber and divertor components. It will contribute to the knowledge on material qualification, licensing, operational safety and remote maintenance necessary for DEMO design

Wong, C P↗

DOE OSTI2013

Compilation of Quality Assurance Documentation for Analyses Performed for the Resumption of Transient Testing Environmental Assessment

This is a companion document to the analyses performed in support of the environmental assessment for the Resumption of Transient Fuels and Materials Testing. It is provided to allow transparency of the supporting calculations. It provides computer code input and output. The basis for the calculations is documented separately in INL (2013) and is referenced, as appropriate. Spreadsheets used to manipulate the code output are not provided.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2011

Verifying disarmament: scientific, technological and political challenges

There is growing interest in, and hopes for, nuclear disarmament in governments and nongovernmental organizations (NGOs) around the world. If a nuclear-weapon-free world is to be achievable, verification and compliance will be critical. VerifYing disarmament would have unprecedented scientific, technological and political challenges. Verification would have to address warheads, components, materials, testing, facilities, delivery capabilities, virtual capabilities from existing or shutdown nuclear weapon and existing nuclear energy programs and material and weapon production and related capabilities. Moreover, it would likely have far more stringent requirements. The verification of dismantlement or elimination of nuclear warheads and components is widely recognized as the most pressing problem. There has been considerable research and development done in the United States and elsewhere on warhead and dismantlement transparency and verification since the early 1990s. However, we do not today know how to verifY low numbers or zero. We need to develop the needed verification tools and systems approaches that would allow us to meet this complex set of challenges. There is a real opportunity to explore verification options and, given any realistic time frame for disarmament, there is considerable scope to invest resources at the national and international levels to undertake research, development and demonstrations in an effort to address the anticipated and perhaps unanticipated verification challenges of disarmament now andfor the next decades. Cooperative approaches have the greatest possibility for success.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2017

MLEP-Fail Calibration for 1/8 inch thick plate of 13-8 PH steel in the H1150 condition

The purpose of the work presented in this memo was to calibrate the Sierra material model Multilinear Elastic-Plastic Hardening Model with Failure , (MLEP-Fail), for 1/8 inch thick plate of 13-8 PH steel in the H1150 condition. Material testing was pursued first, followed by the actual calibration.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Elevated Temperature Graphite Mechanical Testing

High purity graphite will be used for core components within most High Temperature Reactor (HTR) designs. Several "high tech” industries currently utilize synthetic, high-purity, commercially available graphite components to fabricate photovoltaic cells, semi-conductors, optical fibers, and other high value electronic industry products. New advanced HTR designs are also interested in using these graphite grades for long-term, internal core component applications. Consequently, the US Department of Energy, Advanced Reactor Technologies (DOE-ART) program has spent several years testing different high purity graphite grades for potential use within these new nuclear reactor designs. A significant part of that effort has been in the development and improvement of American Society for Testing and Materials (ASTM) test standards specifically for nuclear graphite grades. Nearly all ASTM test standards have either been developed or improved by the DOE-ART program over the past 25 years. The ART program continues to assist in the development of new ASTM test standards in support of the future commercial HTR fleet currently being designed and built in the USA. The newest effort undertaken by ART is the development of high temperature mechanical testing practices that may be acceptable for a future ASTM test standard (or guide) for this critical material property measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Decarburization and Carburization Behavior of Grade 91 Steel in Sodium Environments

Two heats of G91 steel have been investigated in sodium at 550-650°C to understand its corrosion behaviour, microstructural evolution, and tensile properties. Sodium exposure tests of subsized tensile and coupon specimens of G91 base metal and weldment were conducted in Argonne’s forced convection sodium materials testing loops. Maximum exposure times of ~39,000 h at 550°C, ~37,000 h at 600°C, and ~20,000 h at 650°C have been achieved, respectively. Thermal aging experiments of G91 steel were conducted in parallel, and the data of the sodium-exposed specimens were compared with the thermal aging data of the same heat to isolate the thermal aging effect from the effect of sodium exposure. It was found that the tensile strength of G91 steel can be reduced by more than 50% after sodium exposures at 650°C. The effect of a sodium environment on tensile properties is largely dependent on the carburization/decarburization behavior of G91 steel exposed to sodium. The carbon concentrations in sodium in Argonne’s SMT-1 and SMT-2 loops were determined by the equilibration method. The carbon activity in sodium was established by equilibrating high-purity nickel in sodium, analyzing the material for carbon, and using the reported data on the carbon solubility for this material. The carbon activity results were combined with the carbon solubility data in sodium to establish the carbon concentration in sodium. The carbon concentrations in the SMT-1 and SMT-2 loop were estimated to be 1.23 and 0.25 ppm, respectively. With this initial success, we will continue to monitor the carbon contents in the two loops to correlate the microstructural and mechanical property changes with the sodium purity. Future experiments will also consider sodium exposure tests in controlled carbon activities. Online monitoring and control of carbon contents in sodium in nuclear reactors would also be of interest. Thermodynamic analysis was conducted to understand the carburization-decarburization behavior of G91 steel exposed to sodium in the SMT-1 and SMT-2 loops. The equilibrium carbon activity in the alloy was evaluated by the thermodynamics of M 23 C 6 carbides. The calculated carburization-decarburization boundary showed that G91 would undergo decarburization at 650°C in the SMT-1 loop and carburization at 600°C in the SMT-2 loop, which is consistent with our findings from the sodium exposure tests. Future work will include thermodynamic analysis of the carbon concentration – activity relationship including the effect of carbide-forming elements, vanadium and niobium in G91 steel. Kinetics of carbon diffusion and the effect on microstructure and mechanical performance under various service conditions will also be investigated. The ultimate goal is to understand the mechanisms and the kinetics of carbon transfer in the sodium-steel system and to provide a basis for predicting the effect of long-term decarburization-carburization process on the integrity of reactor components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

SU-F-T-181: Proton Therapy Tissue-Equivalence of 3D Printed Materials

Purpose: This work investigated the proton tissue-equivalence of various 3D printed materials. Methods: Three 3D printers were used to create 5 cm cubic phantoms made of different plastics with varying percentages of infill. White resin, polylactic acid (PLA), and NinjaFlex plastics were used. The infills ranged from 15% to 100%. Each phantom was scanned with a CT scanner to obtain the HU value. The relative linear stopping power (RLSP) was then determined using a multi-layer ion chamber in a 200 MeV proton beam. The RLSP was measured both parallel and perpendicular to the print direction for each material. Results: The HU values of the materials ranged from lung-equivalent (−820 HU σ160) when using a low infill, to soft-tissue-equivalent 159 (σ12). The RLSP of the materials depended on the orientation of the beam relative to the print direction. When the proton beam was parallel to the print direction, the RLSP was generally higher than the RLSP in the perpendicular orientation, by up to 45%. This difference was smaller (less than 6%) for the materials with 100% infill. For low infill cubes irradiated parallel to the print direction, the SOBP curve showed extreme degradation of the beam in the distal region. The materials with 15–25% infill had wide-ranging agreement with a clinical HU-RLSP conversion curve, with some measurements falling within 1% of the curve and others deviating up to 45%. The materials with 100% infill all fell within 7% of the curve. Conclusion: While some materials tested fall within 1% of a clinical HU-RLSP curve, caution should be taken when using 3D printed materials with proton therapy, as the orientation of the beam relative to the print direction can result in a large change in RLSP. Further investigation is needed to measure how the infill pattern affects the material RLSP. This work was supported by PHS grant CA180803.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2014

PBX 9502 Multimode Damage Accumulation Cycles-to-Failure Study

In FY13, we performed a detailed cycles-to-failure study of PBX 9502 using 3- point bend specimens. In that study we examined the effect of temperature and cyclic frequency on the failure behavior of PBX 9502. We identified cycles-to-failure curves for samples tested at -20°C, ~19°C and 50°C and at cyclic frequencies of 0.0001 up to 1-Hz. We also did cyclic conditioning of parts in an effort to establish damage accumulation criteria that could then be used in predictive models. As follow-on to that work, in FY14 we have added multimode cycles-to-failure to our test suite. Work this year re-established the baseline cycles-to-failure behavior of the PBX 9502 material tested round ~19°C at 1-Hz and also looked at damage accumulation under a multimode loading environment. This fresh baseline development was performed to remove any time related or test rig related variables. The multimode loading was done because real-world applications are not mono-modal and we desire for our predictive capability to be robust and include more complex, realistic scenarios. In the following sections we briefly describe the test setup, we provide a summary of the new baseline results and we show the multimode cycles-to-failure data and how it compares to the mono-modal test data.

42 ENGINEERING↗

DOE OSTI2024

Swelling of TPX

The hygroscopic swelling and the thermal expansion behavior of TPX showed anomalous behavior relative to our observations for several other materials tested in the last few years. The hygroscopic swelling showed no correlation to the change in humidity, i.e., the material didn’t necessarily always swell when humidity was increased, it sometime shrank as well. In response to increase in temperature, the material consistently expanded. However, the rate of expansion with temperature changed between experiments. Hence, it was difficult to extract any reliable parameter values for this material.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2023

Predicting Safety Rod Reactivity Insertion in the Advanced Test Reactor

42 ENGINEERING↗

DOE OSTI2023

Local Power Impact Experiment Design for a New Fuel Type for use in the Advanced Test Reactor

The Advanced Test Reactor (ATR), and complimentary zero-power ATR Critical (ATRC) reactor, located at Idaho National Labs (INL), are undergoing conversion from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU). Both have a variety of testing locations that can receive large variations in flux due to its unique serpentine design, consisting of five lobes (see Figure 1). Initial criticality and power distribution throughout the core are controlled by core-external outer shim control cylinders (OSCCs). Distinct test loops allow for testing at specific temperatures, pressures, and irradiation conditions. The ATR is one of the key nuclear engineering research and testing facilities within the DOE National Laboratory Complex, and the ATRC supports its operation [1]. Currently, the Office of Material Management and Minimization (M3) within the National Nuclear Security Administration of the DOE is working to convert the remaining research reactors, including the ATR, from 93% HEU fuel to 19.75% LEU fuel (LEU) to support non-proliferation [2]. Extensive materials testing at INL and internationally has demonstrated that a high-density uranium molybdenum (U 10Mo) alloy can meet the performance requirements of the remaining high powered research reactors. However, there are many technical challenges to address before the conversion to LEU can be successful, including the accurate characterization of the reactor core physics with LEU fuel. Reactor physics safety evaluations currently use Monte Carlo for the 21st Century (MC21), a continuous-energy Monte Carlo radiation transport code [3]. Existing MC21 models of the ATR and ATRC cores have a validation basis for use in neutronics analyses with HEU fuel. The models are used to support safety analyses that include comparisons to the safety requirements for the reactors. However, the use of the LOWE element in the ATR and ATRC is not currently covered by the current model validation basis. To deploy the new fuel type, extensive computational reactor physics support is necessary to support the use of LOWE in the ATR and ATRC. Therefore, LOWE requires a rigorous validation basis, aligned with that of HEU fuel, that takes advantage of the existing software tools and processes currently used for the ATR and ATRC. The experiment to validate of the MC21 models for determining power, the Power Impact Validation Experiment, will consist of two flux runs in the ATRC, one with fully HEU loading and one with a single LOWE element. Both flux runs will be instrumented with 20 sets of azimuthal fission wires and 3 sets of axial fission wires, as shown in Figure 4. Standard flux run methodology will be used [4]. Power Impact Validation Experiment data will be compared against MC21 calculated data, both for absolute fission rate accuracy and to determine the relative change in fission rates between the two runs. The results of the Power Impact Validation Experiment and subsequent evaluations will provide the validation basis for MC21 for use with LOWE elements. Key features of the Power Impact Validation Experiment include: (1) Two flux runs to allow for LOWE perturbed measurements to be compared to already validated measurements taken from a full core of HEU fuel, (2) Optimization of instrumentation to balance analytical needs with practical considerations (e.g., limited time window to count beta particles from fission products), and (3) Standard ATRC core loading, including both driver positions and flux traps, to minimize cost while remaining representative of typical ATR core loading.

42 ENGINEERING↗

DOE OSTI2020

Degradation of impact toughness in cast stainless steels during long-term thermal aging

Cast austenitic stainless steels (CASSs) have been extensively used for the large components of nuclear reactor primary coolant systems. Since the cast steels inevitably contain degradable metastable phases and replacement of the large coolant system components is impractical, the thermal embrittlement of CASS components has been a serious concern in the extended-term operation of nuclear power plants. This study aimed to systematically measure and analyze the effect of long-term thermal aging on the Charpy impact toughness to provide a comprehensive understanding of thermal degradation behavior and a practical aging model to predict the degree of thermal degradation in the cast stainless steels. Additionally, the materials tested in the research include eight CASS alloys (two CF3s, one CF3M, three CF8s, and two CF8Ms) and two reference wrought materials (304L and 316L), in which the nominal δ-ferrite content ranges from ~2% to 33%. These stainless steels have been thermally aged at two light water reactor (LWR) temperatures (290 and 330 °C) and at two accelerated-aging temperatures (360 and 400 °C) for up to 30,000 h; these include both under-aged and over-aged conditions relative to the extended service lifetime (80 years). Charpy impact testing was performed for aged and non-aged specimens, and the impact (absorbed) energy parameters were correlated with a new aging parameter (A). Both the reduction of impact fracture toughness and the shift of ductile-brittle transition temperature were strongly dependent on the δ-ferrite content and degree of thermal aging. A linear relationship was found between the increasing rate of the index transition temperature T 41J and aging parameter A; base on which an empirical model was proposed for prediction of the transition temperature as a simple function of the aging parameter (A) and δ-ferrite content (F). Finally, the critical aging parameter for embrittlement (AC) was evaluated and compared with the existing δ-ferrite content criteria.

36 MATERIALS SCIENCE↗