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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Predictive characterization of aging and degradation of reactor materials in extreme environments. Final report, December 20, 2013 - September 20, 2017

Understanding of reactor material behavior in extreme environments is vital not only to the development of new materials for the next generation nuclear reactors, but also to the extension of the operating lifetimes of the current fleet of nuclear reactors. To this end, this project conducted a suite of unique experimental techniques, augmented by a mesoscale computational framework, to understand and predict the long-term effects of irradiation, temperature, and stress on material microstructures and their macroscopic behavior. The experimental techniques and computational tools were demonstrated on two distinctive types of reactor materials, namely, Zr alloys and high-Cr martensitic steels. These materials are chosen as the test beds because they are the archetypes of high-performance reactor materials (cladding, wrappers, ducts, pressure vessel, piping, etc.). To fill the knowledge gaps, and to meet the technology needs, a suite of innovative in situ transmission electron microscopy (TEM) characterization techniques (heating, heavy ion irradiation, He implantation, quantitative small-scale mechanical testing, and various combinations thereof) were developed and used to elucidate and map the fundamental mechanisms of microstructure evolution in both Zr and Cr alloys for a wide range environmental boundary conditions in the thermal-mechanical-irradiation input space. Knowledge gained from the experimental observations of the active mechanisms and the role of local microstructural defects on the response of the material has been incorporated into a mathematically rigorous and comprehensive three-dimensional mesoscale framework capable of accounting for the compositional variation, microstructural evolution and localized deformation (radiation damage) to predict aging and degradation of key reactor materials operating in extreme environments. Predictions from this mesoscale framework were compared with the in situ TEM observations to validate the model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Intercomparison of the Radio-Chronometric Ages of Plutonium-Certified Reference Materials with Distinct Isotopic Compositions

In this article, an intercomparison of the radio-chronometric ages of four distinct plutonium-certified reference materials varying in chemical form, isotopic composition, and period of production are presented. The cross-comparison of the different 234 U/ 238 Pu, 235 U/ 239 Pu, 236 U/ 240 Pu, and 241 Am/ 241 Pu model purification ages obtained at four independent analytical facilities covering a range of laboratory environments from bulk sample processing to clean facilities dedicated to nuclear forensic investigation of environmental samples enables a true assessment of the state-of-practice in “age dating capabilities” for nuclear materials. The analytical techniques evaluated used modern mass spectrometer instrumentation including thermal ionization mass spectrometers and inductively coupled plasma mass spectrometers for isotopic abundance measurements. Both multicollector and single collector instruments were utilized to generate the data presented here. Consensus values established in this study make it possible to use these isotopic standards as quality control standards for radio-chronometry applications. Results highlight the need for plutonium isotopic standards that are certified for 234 U/ 238 Pu, 235 U/ 239 Pu, 236 U/ 240 Pu, and 241 Am/ 241 Pu model purification ages as well as other multigenerational radio-chronometers such as 237 Np/ 241 Pu. Due to the capabilities of modern analytical instrumentation, analytical laboratories that focus on trace level analyses can obtain model ages with marginally larger uncertainties than laboratories that handle bulk samples. Lastly, when isotope ratio measurement techniques like thermal ionization mass spectrometry and inductively coupled plasma mass spectrometry with comparable precision are utilized, model purification ages with similar uncertainties are obtained.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Fracture Resistance of Cast Stainless Steels after Thermal Aging for up to 10000 Hours

This work package, Cast Stainless Steel Aging, aims to achieve a comprehensive scientific understanding on the aging and failure phenomena in cast austenitic stainless steels (CASSs) using holistic experimental and modeling means and to provide a practical and science-based model to predict the degree of thermal degradation of CASS components in extended-term operations. The test materials in the project include four model CASSs (CF3, CF3M, CF8, and CF8M), four EPRI-provided CASSs (CF3, two CF8s, and CF8M), and two reference wrought materials (304L and 316L), which contain a wide variety of d-ferrite contents (2–33%). These materials have been thermally aged at two light water reactor (LWR)-relevant temperatures (290 and 330°C) and at two accelerated-aging temperatures (360 and 400°C) for up to three years. In the fiscal year, the fracture toughness testing and J-R curve calculation were completed for the ten CASS and wrought materials aged up to 10,000 hours. This report is to present the results of the static fracture (J-R) testing for the model and EPRI-provided CASS materials after thermal aging up to 10,000 hours. First, a new aging parameter (A) was defined to present the aging degradation of mechanical properties against the common variable. This definition is used to scale an aging time at a temperature to the effective aging time at a reference temperature based on Arrhenius equation, which is a rate theory equation for thermally activated mechanisms.

36 MATERIALS SCIENCE↗
NASA NTRS1992

Isotopic signatures of black tektites from the K-T boundary on Haiti - Implications for the age and type of source material

An isotopic study was carried out to characterize the type of black tektites from the Cretaceous-Tertiary (K-T) boundary on Haiti (the first reasonably well-preserved impact-derived material recovered from the K-T boundary), in order to help characterize the tektite source material (i.e., the type of rocks that were melted and ejected during the impact event(s) at around 64.5 Ma). Results show that the isotopic data and all of the element concentration data obtained are consistent with an andesitic-dacitic composition for the tektites and their source material. The Nd isotopic data suggest that the source rocks were not older than Silurian (T(chur) = 400 Ma) in age, and were composed largely of young (less than 1080 Ma) crustal material. Of the suspected K-T boundary impact sites, both the Manson (Iowa) and Chicxulub (Yucatan) structures occur in suitable lithologies to yield the Haitian black tektites.

Premo, W. R.↗
DOE OSTI2017

Investigation of nanostructured Al-10 wt.% Zr material prepared by ball milling for high temperature applications

Ground chips of as-cast Al-10 wt.% Zr alloy were subjected to mechanical alloying (MA) with 5 vol.% of nanodiamond addition in a high energy planetary ball-mill. The aim of this work was to investigate the microstructure, phase transformation and mechanical properties of the material both after MA and after subsequent annealing. Optical and transmission electron microscopes were used for morphological and microstructural analysis. The effect of milling time on powder microhardness, Al lattice parameter, lattice microstrain and crystallite size was determined. It was shown that mechanical alloying of as-cast Al-10wt.%Zr alloy during 20 h leads to a complete dissolution of the primary tetragonal Al{sub 3}Zr crystals in aluminum. At the same time, the powder microhardness increases to 370 HV. Metastable cubic Al{sub 3}Zr phase nanoparticles precipitate from the Al solution due to its decomposition after annealing, however, the Al solid solution remains supersaturated and nanocrystalline. Compression tests at room temperature and at 300 °C showed that the strength values of the hot-pressed samples reach 822 MPa and 344 MPa, respectively. - Highlights: •As-cast Al-10 wt.% Zr alloy was mechanically alloyed with 5 vol.% nanodiamond. •The primary tetragonal Al{sub 3}Zr crystals were completely dissolved in Al after 20 h. •Cubic Al{sub 3}Zr phase nanoparticles precipitated from Al solution after aging. •The aged bulk material showed a high strength at room and elevated temperatures.

36 MATERIALS SCIENCE↗
NASA NTRS1978

Effect of shelf aging on O-ring materials

Commercial O-rings made from 13 different rubber compounds were tested for physical properties after 7 and 12 years of shelf aging. No gross changes were observed in tensile strength, elongation, or compression deflection characteristics.

Wood, T. E.↗
DOE OSTI2018

Insight into Thermal Aging of Jacket-bonded Ethylene-propylene Rubber Cable Insulation

Electrical cables are integral to nuclear power plant operation and control during power generation, during planned outages, and during design basis events. Polymeric insulation and jacketing used in cable construction does age over time as a function of environmental stresses including heat, moisture, and radiation. Degradation of polymer mechanical properties can lead to breaks in the insulation and consequent electrical shorting. In this work we consider the thermal aging behavior of one of the two most common insulation polymer classes in nuclear power plants, ethylene-propylene rubber (EPR), both as an independent material and within a cable in which the EPR is bonded to a chlorinated polyethylene (CPE) jacket. Understanding the changes in these two materials with aging, both independently and in concert, will allow the establishment of aging lifetime curves and enable the development of effective non-destructive methods for tracking the condition of installed cables composed of similar material systems. Additional and more effective methods for monitoring the condition of installed cables are needed as a technical basis for continued safe use of aging cables in long term operation and to minimize operating costs for cable aging management through better informed retain/repair/replace decisions.

Light Water Reactors, electrical cables, aging, EP↗
DOE OSTI2012

RELIABILITY MODELS OF AGING PASSIVE COMPONENTS INFORMED BY MATERIALS DEGRADATION METRICS TO SUPPORT LONG-TERM REACTOR OPERATIONS

Paper describes a methodology for the synthesis of nuclear power plant service data with expert-elicited materials degradation information to estimate the future failure rates of passive components. This method should be an important resource to long-term plant operations and reactor life extension. Conventional probabilistic risk assessments (PRAs) are not well suited to addressing long-term reactor operations. Since passive structures and components are among those for which replacement can be least practical, they might be expected to contribute increasingly to risk in an aging plant; yet, passives receive limited treatment in PRAs. Furthermore, PRAs produce only snapshots of risk based on the assumption of time-independent component failure rates. This assumption is unlikely to be valid in aging systems. The treatment of aging passive components in PRA presents challenges. Service data to quantify component reliability models are sparse, and this is exacerbated by the greater data demands of age-dependent reliability models. Another factor is that there can be numerous potential degradation mechanisms associated with the materials and operating environment of a given component. This deepens the data problem since risk-informed management of component aging will demand an understanding of the long-term risk significance of individual degradation mechanisms. In this paper we describe a Bayesian methodology that integrates metrics of materials degradation susceptibility with available plant service data to estimate age-dependent passive component reliabilities. Integration of these models into conventional PRA will provide a basis for materials degradation management informed by predicted long-term operational risk.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Phase changes in fluoropolymer binders

Motivation: Understanding fluoropolymer binder structure and aging response continues to be an important goal for the prediction and assessment of polymer bonded explosive compositions within the stockpile. As part of ongoing efforts within NNM, we are developing new methods to quantify crystallization and chemical degradation within aged Fluoropolymer materials. Significance: Time, temperature and radiation are known to alter the mechanical properties of fluoropolymers as crystalline domains nucleate and grow within the material. However, current methods do not accurately quantify and assess the degree or rapidity of these physical aging processes. For this reason, we are developing methods that directly measure crystalline domain, composition, size and shape and cross-correlating them with traditional techniques. Progress: Most recently, we thermally aged fluoropolymer binder, FK800, for various durations up to 23 days at several temperatures including: 25C (red), 35C (green), and 60C (blue). We used wide angle X-ray scattering (WAXS) to observe when crystalline peaks emerged, signaling a phase transition from amorphous to partially crystalline. Some of these data are summarized in Figure 1. As expected, samples held at 25C (red), which is below the glass transition temperature at 28C, remain amorphous – the WAXS data (Figure 1b) lacks sharp peaks indicative of crystallinity and the ultra-small angle X-ray scattering (USAXS) also has a smooth background without peaks indicative of structure (Figure 1a). By contrast, samples held at 35C and 60C have new peaks in both the USAXS and WAXS data. The structure is clearer after subtracting off the amorphous background from the WAXS data (Figure 1c). These data show that samples aged at 60C are more crystalline as determined by the more defined peaks and also have a larger lattice constant as seen by the shift in peak position as compared to the 35C samples. The corresponding USAXS data (Figure 1a) indicate that the size of the crystals was larger at 60C than at 35C as seen by the shift in the knee position to lower q-values. Interestingly, the shape of the scatter curve also changed, which indicates a change in morphology. Atomic force microscopy (AFM) was used to image the fluoropolymer (Figure 1c-e). At 35C, ~50nm long stacked crystal plates are observed. High resolution imaging shows that the stacks consist of 2-6 lamellae that are spaced approximately 3nm apart. At 60C, the individual lamella appear both thicker and longer and are intertwined to fill the surface. Morphological data such as these images help to interpret & model volume averaging techniques like X-ray scattering and NMR. We also expect the shift from isolated islands to intertwined clusters to impact mechanical properties.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Expanded Materials Degradation Analysis (EMDA) Volume 3: Aging of Reactor Pressure Vessels

In NUREG/CR-6923 ORNL/TM-2013/532, "Expert Panel Report on Proactive Materials Degradation Assessment," referred to as the PMDA report, NRC conducted a comprehensive evaluation of potential aging-related degradation modes for core internal components, as well as primary, secondary, and some tertiary piping systems, considering operation up to 40 years. This document has been a very valuable resource, supporting NRC staff evaluations of licensees' aging management programs and allowing for prioritization of research needs.This report describes an expanded materials degradation assessment (EMDA), which significantly broadens the scope of the PMDA report. The analytical timeframe is expanded to 80 years to encompass a potential second 20-year license-renewal operating-period, beyond the initial 40-year licensing term and a first 20-year license renewal. Further, a broader range of structures, systems, and components (SSCs) was evaluated, including core internals, piping systems, the reactor pressure vessel (RPV), electrical cables, and concrete and civil structures. The EMDA uses the approach of the phenomena identification and ranking table (PIRT), wherein an expert panel is convened to rank potential degradation scenarios according to their judgment of susceptibility and current state of knowledge.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Guidance on Radiochronometry

Radiochronometry is the measurement of radioactive materials and their decay products to determine the “age” of the material. It differs from the practice of geochronology in that the materials (samples) are man-made, and in that the timespan of interest is confined to the nuclear era, i.e., since the discovery of fission in 1939. In radiochronometry, the “age” of the material ideally records the time when the sample was manufactured, or produced in the final form that is analyzed in the laboratory. It is an important predictive signature in a nuclear forensic investigation which can help in the attribution to a source. The calculation of this “age” can only be made if a model is invoked that allows simplification of the parent radionuclide to daughter isotope decay equation. The simplification required for this model is that the parent was purified completely from all traces of its daughter product at the time of sample preparation. Then, this “age” should be described as the “model age” from which a “model date” in the past can then be calculated. For this model date to represent the date of sample preparation, two other assumptions are required to be true. The first is that the material has remained a closed system since that time (i.e., that there has been no loss or gain of either parent or daughter), and the second is that the analyses are accurate.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
NASA NTRS1995

Physical aging effects on the compressive linear viscoelastic creep of IM7/K3B composite

An experimental study was undertaken to establish the viscoelastic behavior of 1M7/K3B composite in compression at elevated temperature. Creep compliance, strain recovery and the effects of physical aging on the time dependent response was measured for uniaxial loading at several isothermal conditions below the glass transition temperature (T(g)). The IM7/K3B composite is a graphite reinforced thermoplastic polyimide with a T(g) of approximately 240 C. In a composite, the two matrix dominated compliance terms associated with time dependent behavior occur in the transverse and shear directions. Linear viscoelasticity was used to characterize the creep/recovery behavior and superposition techniques were used to establish the physical aging related material constants. Creep strain was converted to compliance and measured as a function of test time and aging time. Results included creep compliance master curves, physical aging shift factors and shift rates. The description of the unique experimental techniques required for compressive testing is also given.

Veazie, David R.↗
NASA NTRS1977

Estimating surface temperature in forced convection nucleate boiling: A simplified method

During a test program to investigate low-cycle thermal fatigue, 21 of 22 cylindrical test sections of a cylindrical rocket thrust chamber were thermally cycled to failure. Cylinder liners were fabricated from OFHC copper, Amzirc, and NARloy-Z. The cylinders were fabricated by milling cooling channels into the liner and closing out the backside with electrodeposited copper. The tests were conducted at a chamber pressure of 4.14 MN/sq m (600 psia) and an oxidant-fuel ratio of 6.0 using hydrogen-oxygen as propellants. The average throat heat flux was 54 MW/sq m (33 Btu/sq in./sec). All of the failures were characterized by a thinning of the cooling channel wall and eventual failure by tensile rupture. The 1/2-hard Amzirc material showed little improvement in cyclic life when compared with OFHC copper; while the NARloy-Z and aged Amzirc materials had the best cyclic life characteristics. One OFHC copper cylinder was thermall cycled 2044 times at a steady-state hot-gas-side wall temperature of 514 K (925 R) without failing.

Hendricks, R. C.↗
NASA NTRS2003

Propulsion Research at the Propulsion Research Center of the NASA Marshall Space Flight Center

The Propulsion Research Center of the NASA Marshall Space Flight Center is engaged in research activities aimed at providing the bases for fundamental advancement of a range of space propulsion technologies. There are four broad research themes. Advanced chemical propulsion studies focus on the detailed chemistry and transport processes for high-pressure combustion, and on the understanding and control of combustion stability. New high-energy propellant research ranges from theoretical prediction of new propellant properties through experimental characterization propellant performance, material interactions, aging properties, and ignition behavior. Another research area involves advanced nuclear electric propulsion with new robust and lightweight materials and with designs for advanced fuels. Nuclear electric propulsion systems are characterized using simulated nuclear systems, where the non-nuclear power source has the form and power input of a nuclear reactor. This permits detailed testing of nuclear propulsion systems in a non-nuclear environment. In-space propulsion research is focused primarily on high power plasma thruster work. New methods for achieving higher thrust in these devices are being studied theoretically and experimentally. Solar thermal propulsion research is also underway for in-space applications. The fourth of these research areas is advanced energetics. Specific research here includes the containment of ion clouds for extended periods. This is aimed at proving the concept of antimatter trapping and storage for use ultimately in propulsion applications. Another activity in this involves research into lightweight magnetic technology for space propulsion applications.

Blevins, John↗
DOE OSTI2019

Report on FY19 scoping creep and creep-fatigue testing on heat treated Alloy 709 base metal

Alloy 709 is a high temperature austenitic steel with potential for replacing Type 316 stainless steel. Efforts have been underway to lay the groundwork for an ASME code case by first refine the processing conditions to manufacture Alloy 709 plate. This effort continued into this year, as well as work examining the range of allowable solution annealing temperatures, the ability to recover material that was solution annealed at a temperature that was found to produce poor mechanical properties, and to examine the effects of pre-aging the material prior to testing.

42 ENGINEERING↗
DOE OSTI2016

Perspective of Dry Interim Storage of spent nuclear fuel (SNF) in Germany - 16504

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office of Radiation Protection (BfS). In 1995 a license to store spent nuclear fuel (SNF) and vitrified high level waste (HAW) for the interim storage facility in Gorleben was granted. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the NPP Lingen. Since this time each NPP erected its own storage facility and three central storage facilities have been built. All of these facilities use dry storage in metallic dual purpose casks (DPCs). The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2035. During the last two years in Germany preparatory consultations to restart the process towards finding a repository took place, organised by a commission. This commission consists out of members of the scientific community, religious parties, social communities and environmental organizations as well as members of federal states governments and members of the German parliament (Bundestag). Recently the commission decided to extend its working period until summer 2016. Taking this background into account the BfS plans to intensify its research into extended interim storage with respect to potential material degradation leading to interference with safety functions (e.g. neutron shielding material). This includes topics like ageing of gasket materials which are already studied in ongoing experiments (e.g. at the Federal Institute for Materials Research and Testing) as well as studies of long term relaxation processes. Because of the need for transportation towards a repository another important field is the behaviour of the inventory under mechanical loads. For SNF, first calculations showed that because of the lower decay heat load after 60 years due to hydride reorientation the cladding might become much less ductile. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗