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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

In-air and pressurized water reactor environment fatigue experiments of 316 stainless steel to study the effect of environment on cyclic hardening

Argonne National Laboratory (ANL), under the sponsorship of Department of Energy’s Light Water Reactor Sustainability (LWRS) program, is trying to develop a mechanistic approach for more accurate life estimation of LWR components. In this context, ANL has conducted many fatigue experiments under different test and environment conditions on type 316 stainless steel (316SS) material which is widely used in the US reactors. Contrary to the conventional S~N curve based empirical fatigue life estimation approach, the aim of the present DOE sponsored work is to develop an understanding of the material ageing issues more mechanistically (e.g. time dependent hardening and softening) under different test and environmental conditions. Better mechanistic understanding will help develop computer-based advanced modeling tools to better extrapolate stress-strain evolution of reactor components under multi-axial stress states and hence help predict their fatigue life more accurately. In this paper (part-I) the fatigue experiments under different test and environment conditions and related stress-strain results for 316 SS are discussed. In a second paper (part-II) the related evolutionary cyclic plasticity material modeling techniques and results are discussed.

Cyclic plasticity material modeling↗
DOE OSTI2025

Bio-Oil Impact on Water Diffusion and Durability of Bitumen: Influence of Aging and Salinity

Bio-oils derived from inexpensive biomass offer environmentally friendly options for modifying bitumen for improved durability or rejuvenating aged material, although their impact on bitumen moisture resistance can be mixed. A better understanding of water diffusion into bitumen and its interactions with bio-oil compounds would aid the development of effective bio-oils. In this study, the effect of water exposure on bio-oil-modified bitumen blends was examined by de-wetting contact angle measurements and differential FTIR spectroscopy. The bio-oils improved the anti-stripping behavior of the bitumen but also increased water absorption which could weaken cohesive strength. Short-term thermal aging increased water diffusion for all bitumen blends including the control, probably due to the presence of more oxidized compounds. Negative peaks of alkanes and polar groups in the differential FTIR data suggest the co-diffusion of surfactant molecules towards the bitumen-water interface. Basic pH increased de-wetting of some bitumen blends from silica, possibly by attacking the silica surface itself. The presence of salt in solution altered the bitumen surface composition through the formation of salt complexes with bitumen or bio-oil compounds. In particular, the formation of calcium-carboxylate complexes appeared to greatly improve anti-stripping effects of bio-oils. Water diffusion into most of the bitumen blends was insensitive to pH or salt concentration except for a few notable outliers. Identifying the compounds responsible for increasing or decreasing water diffusion in these outlier cases would be valuable developing future bio-oil formations that avoid or promote those compounds.

Hung, Albert M.↗
DOE OSTI2017

Accelerated materials evaluation for nuclear applications

This report addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Predicted Radiation Output from Several Kilograms of Plutonium Oxide

Neutron and gamma signatures from 3.44 kg of plutonium oxide were predicted with MCNP6 in anticipation of a passive diagnostic measurement campaign. Calculated radiation fields are visualized through use of “FMESH” type tallies. This work is effectively an extension of previous efforts and application of MISC for the photon source term is the main advancement. Large sources of modeling uncertainty include physical can dimensions and thus internal volume and density estimates. Material age is unknown and will also affect the passive gamma signature. The measurement campaign will provide information for model refinement and subsequent simulations. The authors have requested passive data collection on the pair of cans with the largest and smallest mass and such data will be valuable for benchmarking purposes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Layers Can Be Deceiving: A Hopping Model for Small Molecule Diffusion in TATB Crystal

Sorption of small molecule gases in materials can play a significant role in their long‐term stability and compatibility within multi‐material assemblies. While many material properties of the insensitive high explosive TATB (1,3,5‐triamino‐2,4,6‐trinitrobenzene) are well understood, very little is known regarding its permeability to gases. TATB crystal exhibits a graphitic‐like layered packing structure that evokes a mental schema in which the layers form nanoscopic channels, but it is unclear whether this structure promotes gas transport. Here, we use molecular dynamics (MD) simulations to predict transport of small molecules through TATB single crystal. An approach to fit classical force fields is developed to model TATB interactions with H 2 O, He, Ne, and Ar, which is then combined with steered MD to probe gas transport along selected directions in the crystal. We find that small molecule transport occurs via a hopping mechanism that exhibits distinct jumps between interstitial sites and is substantially faster normal to the layers as compared to through them. This result stems from the finding that intralayer junctions between adjacent TATB molecules are the most stable interstitial sites and that energetic barriers are lower for hopping between adjacent layers. An empirical model for diffusion rate based on the MD data shows that the rate decays exponentially with increasing molecular radius and is negligibly small for all molecules larger than He, including common atmospheric gases. These findings have implications for the interpretation of experiments that measure surface area, material response to extreme conditions, and are expected to help constrain models for material aging.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Corrigendum to “Accelerated materials evaluation for nuclear applications” [J. Nucl. Mater. 488 (2017) 46–62]

The original article addresses the opportunities and complexities of using materials test reactors with high neutron fluxes to perform accelerated studies of material aging in power reactors operating at lower neutron fluxes and with different neutron flux spectra. Radiation damage and gas production in different reactors have been compared using the code, SPECTER. This code provides a common standard from which to compare neutron damage data generated by different research groups using a variety of reactors. This Corrigendum identifies a few typographical errors. Tables 2 and 3 are included in revised form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1972

Studies of the performance of W-Re type thermocouples.

The effect of exposure of bare-wire and BeO-insulated commercial W-Re thermocouple materials to high temperatures in gaseous environments was investigated. The temperature range of interest was primarily 2000 to 2400 K, and the investigations were confined to thermocouple wires of 0.25 mm diameter. Thermoelements were examined for metallurgical structural changes and chemical changes by conventional methods. In the temperature range of interest, grain growth was inhibited in the chemically doped W-3% Re alloy, and excellent room-temperature ductility was retained subsequent to the exposure. The compatibility of high purity (in excess of 99.8%), sintered BeO insulators with the thermoelements differed, depending upon whether the BeO-insulated thermoelement assemblies were self-heated electrically or heated in a furnace; in tests in argon environments, highly reliable performance occurred when the assemblies were heated in a furnace. BeO-insulated W-3% Re vs W-25% Re thermocouples, constructed with degassed and aged materials, exhibited drifts equivalent to about 3 mK/hr during 1000 hours of exposure at 2073 K in argon while in the presence of tantalum.

Burns, G. W.↗
NASA NTRS2016

Ground Truthing Orbital Clay Mineral Observations with the APXS Onboard Mars Exploration Rover Opportunity

NASA's Mars Exploration Rover Opportunity has been exploring approximately 22 km diameter Endeavour crater since 2011. Its rim segments predate the Hesperian-age Burns formation and expose Noachian-age material, which is associated with orbital Fe3+-Mg-rich clay mineral observations [1,2]. Moving to an orders of magnitude smaller instrumental field of view on the ground, the clay minerals were challenging to pinpoint on the basis of geochemical data because they appear to be the result of near-isochemical weathering of the local bedrock [3,4]. However, the APXS revealed a more complex mineral story as fracture fills and so-called red zones appear to contain more Al-rich clay minerals [5,6], which had not been observed from orbit. These observations are important to constrain clay mineral formation processes. More detail will be added as Opportunity is heading into her 10th extended mission, during which she will investigate Noachian bedrock that predates Endeavour crater, study sedimentary rocks inside Endeavour crater, and explore a fluid-carved gully. ESA's ExoMars rover will land on Noachian-age Oxia Planum where abundant Fe3+-Mg-rich clay minerals have been observed from orbit, but the story will undoubtedly become more complex once seen from the ground.

Schroeder, C.↗
DOE OSTI2020

Modular high temperature gas reactor core modeling with RELAP5-3D/PHISICS – Optimization schemes for load following

The objective of this study is to develop a 3D/NK model to study the load following operation of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The selected design is the 350 MW prismatic, graphite moderated, helium cooled thermal reactor based on the MHTGR-350 transient benchmark led by the High-Temperature Gas-Cooled Reactor (HTGR) Methods Core Simulation Group at Idaho National Laboratory (INL) in the framework of the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD/NEA). This kind of reactor usually reacts very slowly to the perturbation of the core parameters due to the large amount of graphite in the core. This kind of behavior promotes the reactor stability but on the other hand limits the reactor load-following operability. Recently advances in gas reactor designs have made load-following in gas reactors a feasible and achievable goal. Modern nuclear reactors (such as the MHTGR-350) are designed to withstand the structural stresses associated with load-following. Operating the reactor in load-following mode will inevitably reduce the load factor. Although a higher load factor means more revenue and the best usage of the fuel, following the variable energy demand will increase the maximum achievable fraction of generated nuclear power, being no more limited to the base load power generation. In general, if low impact on material aging and the safe operability are demonstrated the plant economics will be minimally affected. In order to ensure that the system can be safely operated in a load-following mode an extensively study has been carried out. Some tests with linear coolant mass flow rate change demonstrated that the reactor behavior is suitable for operation in load following mode. This assumption has been confirmed by a 4 days load following transient test in which the reactor supplied the requested power with a negligible error.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2007

Strain Measurement Using FBG on COPV in Stress Rupture Test

White Sands Test Facility (WSTF) was requested to perform ambient temperature hydrostatic pressurization testing of a Space Transportation System (STS) 40-in. Kevlar Composite Overwrapped Pressure Vessel (COPV). The 40-in. vessel was of the same design and approximate age as the STS Main Propulsion System (MPS) and Orbiter Maneuvering System (OMS) vessels. The NASA Engineering Safety Center (NESC) assembled a team of experts and conducted an assessment that involved a review of national Kevlar COPY data. During the review, the STS COPVs were found to be beyond their original certification of ten years. The team observed that the likelihood of STS COPV Stress rupture, a catastrophic burst before leak failure mode, was greater than previously believed. Consequently, a detailed assessment of remaining stress rupture life became necessary. Prior to STS-114, a certification deviation was written for two flights of OV-103 (Discovery) and OV-104 (Atlantis) per rationale that was based on an extensive review of the Lawrence Livermore National Laboratories, COPV data, and revisions to the STS COPV stress levels. In order to obtain flight rationale to extend the certification deviation through the end of the Program, the Orbiter Project Office has directed an interagency COPV team to conduct further testing and analysis to investigate conservatism in the stress rupture model and evaluate material age degradation. Additional analysis of stress rupture life requires understanding the fiber stresses including stress that occurs due to thru-wall composite compression in COPV components. Data must be obtained at both zero gauge pressure (pre-stress) and at the component operating pressure so that this phenomenon can be properly evaluated. The zero gauge pressure stresses are predominantly a result of the autofrettage process used during vessel manufacture. Determining these pre-stresses and the constitutive behavior of the overwrap at pressure will provide necessary information to better predict the remaining life of the STS COPVs. The primary test objective is obtaining data to verify the hypothesis of a radially oriented thru-thickness stress-riser in the COPV composite whose magnitude is a function of the applied pressure and the load history. The anticipated load dependent response follows from the constitutive behavior of the composite overwrap so data to quantify its nonlinear and time dependent response will be sought. The objective of the Fiber Braggs Gratings (FBGs) were to advance the state-of-the-art by developing techniques using FBG sensors that are capable of assessing stress-rupture degradation in Kevlar COPVs in a health monitoring mode (1). Moreover, they sought to answer questions of how embedded sensors affect overall integrity of the structure. And lastly, they sought to provide an important link in the overall stress rupture study that will help close the loop on the COPV fabrication process. NDE inspection methods will be used from start to finish and FBG will be an integral link within the overall chain.

Banks, Curtis↗
DOE OSTI2022

The electronic Raman scattering spectrum of PuO 2

Here the Raman spectrum of PuO 2 was measured up to 13,000 cm –1 with three different laser excitation wavelengths spanning the resonance (405 nm), near-resonance (457 nm), and preresonance (514 nm) energy range. Approximately 26 never-before-seen bands were observed between 3500 and 13,000 cm –1 . Given the very high energy of the Raman shifts of these bands and the relative insensitivity of their spectral position to the interrogating laser wavelength, they are believed to arise from an electronic origin. These bands are present in both freshly calcined and radiolytically aged PuO 2 , although a broad luminescence is observed in the aged material, which obscures many of the high frequency features. In situ laser annealing of the material attenuated this luminescence and allowed for clear observation of these never-before-seen spectral features. Discovery of these high-energy bands presents a new way of identifying PuO 2 for nuclear nonproliferation and forensics purposes.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Liquid Chromatography Mass Spectrometry study of eutectic bis(2,2-dinitropropyl) acetal/formal

This study is the first attempt to document methodology development undergone using liquid chromatography tandem quadrupole time of flight mass spectrometry (LC-QTOF) to investigate degradation products of eutectic bis(2,2-dinitropropyl) acetal/formal nitroplasticizer (called NP here). Method properties investigated are: desolvation temperature (°C) and spray voltage (V) of the electrospray ionization source, and the development of an acetone system rinse to prevent any residual contamination between sample injections. Details are given on why it is essential to investigate method optimization with changes shown in MS/MS analysis in addition to MS results. Trends in MS/MS analytic results reveal important relationships between baseline and aged materials. In addition to verification of previously proposed fragments, insights offered by this newly developed methodology will also identify new degradation products and shed light on the complexity of NP degradation chemistry.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Fine–scale structural features of intercritically aged HSLA-100 plate steel and their influence on yield strength and low-temperature impact toughness

Highlights: • The fine-scale microstructure of HSLA-100 steel intercritically aged at 677 °C was characterized. • Continuous austenite films evolved into isolated islands; large islands transformed to brittle martensite upon quenching. • Copper precipitates in martensite were coarse, and some sub-grain coarsening occurred within martensite. • Intercritically aged steel possessed an inferior strength-toughness balance as compared with sub-critically aged steel. - Abstract: Fine-scale structural features of an HSLA-100 steel intercritically aged at 950 K (677 °C) for 4.2 ks (70 min.) were investigated via transmission electron microscopy. The microstructure consisted of approximately 90 vol. pct. of low‑carbon lath martensite. Average lath width was about 0.25 μm, unchanged from the as-quenched condition, owing to a high number density of epsilon‑copper precipitates at these boundaries. About 10 vol. pct. of medium‑carbon austenite existed as coarsened, broken-up islands between laths; martensite lath boundary coverage by austenite was notably reduced compared to the as-quenched condition. Greater continuity between adjacent low‑carbon martensite laths was accompanied by coarse copper precipitates within laths and an order of magnitude lower dislocation density than the as-quenched martensite. Collectively, these features produced a sub-grain coarsening effect. Austenite islands larger than about 0.5 μm were unstable with respect to the formation of fresh, medium‑carbon martensite regions. Copper precipitation within austenite islands (owing to carbon and copper enrichment) might also have contributed to austenite transformation upon quenching. Intercritical aging was associated with a strong-and-undesirable deviation from strength-toughness behavior at sub-critical aging temperatures. Two key factors were identified as being responsible for this negative outcome: (i) a sub-grain coarsening effect associated with “delamellarization” of low‑carbon martensite and austenite and (ii) the formation of brittle martensite upon quenching. Quantitative estimates of the changes in strength and impact toughness are provided for microstructural differences between intercritically aged and subcritically aged material.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Standoff enrichment analysis of UF 6 cylinders

We investigate a new safeguards approach for verification of uranium processing operations via standoff measurement of UF 6 cylinder transports entering and leaving a facility. Current safeguards verification at facilities with UF 6 cylinders includes attended enrichment measurements of several randomly-selected cylinders using handheld instruments with inspection of all the cylinders typically only once a year. This approach requires significant inspector effort and can only be performed intermittently. Here we explore an unattended, multi-sensor enrichment measurement system that can operate continuously on cylinders in motion. Such an approach could be used to enhance continuity of knowledge by verifying that all shipments entering and leaving a facility are consistent with declared operations. The challenges in developing a standoff enrichment technique include limited statistical significance of the collected data, and interpreting the complex signatures from a shipment of UF 6 cylinders. These signatures vary with cylinder properties (e.g. the fuel cycle history, cylinder history, material distribution, material age, etc.) and measurement conditions (e.g. speed and distance of vehicle, detector response, cylinder orientation, attenuation from overpack and vehicle, etc.). We demonstrate that we can extract useful enrichment signatures from standoff measurements of UF 6 cylinders entering and leaving a facility. Our initial analysis demonstrates a measurable difference in enrichment-related signatures from a cylinder in motion, measured with modest detection efficiency from several meters away. We also demonstrate that the inclusion of vehicle tracking increases the signal-to-noise ratio (SNR) of our radiation signatures, and decreases uncertainty of parameters needed for enrichment analysis, e.g. distance and velocity.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2023

Irradiation of ultrasonic sensors and adhesive couplants for application in light water reactor primary loop piping and components

The Electric Power Research Institute (EPRI) Nuclear Sector and US Department of Energy Light Water Reactor Sustainability Program are committed to engaging in research and development endeavors to address materials aging issues specific to long term operation of light water power reactors. To this effect, EPRI launched an industry initiative to develop nondestructive evaluation systems for online monitoring of existing cracks in light water reactor primary coolant loop piping and components. One of the goals of this initiative is to develop a sensor system (or systems) that can determine nondestructively if cracks are growing or arrested and, in the case of the former, to characterize their growth rates. A missing component of this initiative is an experimental assessment of how sensors and adhesive couplants will perform in service when exposed to chronic energetic neutron radiation, particularly at the primary coolant loop hot and cold leg dissimilar metal welds, which join the primary loop piping to the reactor pressure vessel and reside in the vicinity of the reactor core. The objective of this experimental study was to determine how ultrasonic transducers and adhesive couplants perform when exposed to irradiation in a test reactor to simulate and accelerate in-service exposure. Further, to achieve this objective, the signal stability of piezoelectric transducers and performance of adhesive couplants as a function of accumulated fast neutron fluence were characterized by collecting ultrasonic data in-situ during irradiation. Of particular interest were the ultrasonic signal quality and time decay of the amplitude of acoustic reflections as a function of fast neutron fluence. The results of the study showed that, of the 8 transducer/substrate sample assemblies tested, only 3 generated usable ultrasonic signals through the conclusion of the irradiation campaign. It was found that high temperature epoxy tends to ultrasonically couple the sensors to the substrates better than three types of refractory ceramic cements studied, as is supported by post irradiation examination. The results obtained through this experimental study will be utilized in the achievement of the overall goal of development of a sensor system to perform online monitoring of primary loop components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Light Water Reactor Sustainability Accomplishments Report

Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you have any questions about the information in the report, or about the LWRS Program, please contact me, Richard A. Reister (the Federal Program Manager), or the respective research pathway leader (noted on pages 26 and 27), or visit the LWRS Program website (www.inl.gov/lwrs). The annually updated Integrated Program Plan and Pathway Technical Program Plans are also available for those seeking more detailed technical Information.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Light Water Reactor Sustainability Program (2015 Accomplishments Report)

This report covers selected highlights from the four research pathways in the LWRS Program: Materials Aging and Degradation; Risk-Informed Safety Margin Characterization; Advanced Instrumentation, Information, and Control Systems Technologies; and Reactor Safety Technologies, as well as a look-ahead at planned activities for 2016.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗