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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 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

AGC-2 Specimen Post Irradiation Data Package Report

This report documents results of the post-irradiation examination material property testing of the creep, control, and piggyback specimens from the irradiation creep capsule Advanced Graphite Creep (AGC)-2 are reported. This is the second of a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades. The AGC-2 capsule was irradiated in the Idaho National Laboratory Advanced Test Reactor at a nominal temperature of 600°C and to a peak dose of 5 dpa (displacements per atom). One-half of the creep specimens were subjected to mechanical stresses (an applied stress of either 13.8, 17.2, or 20.7 MPa) to induce irradiation creep. All post-irradiation testing and measurement results are reported with the exception of the irradiation mechanical strength testing, which is the last destructive testing stage of the irradiation testing program. Material property tests were conducted on specimens from 15 nuclear graphite grades using a similar loading configuration as the first AGC capsule (AGC-1) to provide easy comparison between the two capsules. However, AGC-2 contained an increased number of specimens (i.e., 487 total specimens irradiated) and replaced specimens of the minor grade 2020 with the newer grade 2114. The data reported include specimen dimensions for both stressed and unstressed specimens to establish the irradiation creep rates, mass and volume data necessary to derive density, elastic constants (Young’s modulus, shear modulus, and Poisson’s ratio) from ultrasonic time-of-flight velocity measurements, Young’s modulus from the fundamental frequency of vibration, electrical resistivity, and thermal diffusivity and thermal expansion data from 100–500°C. No data outliers were determined after all measurements were completed. A brief statistical analysis was performed on the irradiated data and a limited comparison between pre- and post-irradiation properties is presented. A more complete evaluation of trends in the material property changes, as well as irradiation-induced creep due to irradiation, temperature, and applied load on specimens will be discussed in later AGC-2 post-irradiation examination analysis reports.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRSDate not supplied

Polymeric Materials Application and Testing for Space Missions

The EM41 Non-metallics polymers team performs polymeric materials screening, characterization, development, qualification, verification and validation testing for flight hardware and in-space applications. They perform insight, testing and application of polymeric materials for launch vehicles and space environments. Personnel have a deep understanding of the technologies associated with cryogenic insulation, high temperature thermal protection materials, paints, primers, coatings, and adhesive materials. Numerous TPS material characterization tests can also be performed by the Lab. The team has both internal and partner test capabilities to subject polymeric materials to Aerospace and Space environments Material properties, response and performance are tested within environments from the launch pad, through the launch environment and into space. A primary role of the team is to select materials and organize test regimen then analyze test results. Test regiment are established to confirm the materials meet the performance requirements levied by customers within the environments that are expected for the mission.

Thermal↗
DOE OSTI2012

Method to calibrate fission chambers in Campbelling mode

Fission chambers are neutron detectors which are widely used to instrument experimental reactors such as material testing reactors or zero power reactors. In the presence of a high level mixed gamma and neutron flux, fission chambers can be operated in Campbelling mode (also known as 'fluctuation mode' or 'mean square voltage mode') to provide reliable and precise neutron related measurements. Fission chamber calibration in Campbelling mode (in terms of neutron flux) is usually done empirically using a calibrated reference detector. A major drawback of this method is that calibration measurements have to be performed in a neutron environment very similar to the one in which the calibrated detector will be used afterwards. What is proposed here is a different approach based on characterizing the fission chamber response in terms of fission rate. This way, the detector calibration coefficient is independent from the neutron spectrum and can be determined prior to the experiment. The fissile deposit response to the neutron spectrum can then be assessed independently by other means (experimental or numerical). In this paper, the response of CEA-made miniature fission chambers in Campbelling mode is studied. A theoretical model of the signal is used to calculate the calibration coefficient. The model's input parameters come from statistical distribution of individual pulses. Supporting measurements were made in the CEA Cadarache zero power reactor MINERVE and results are compared to an empirical Campbelling mode calibration. The tested fission chamber calibration coefficient is roughly 2*10{sup -26} A{sup 2}/Hz/(c/s). Both numerical and empirical methods give consistent results, however a deviation of about 15% was observed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Development, calibration and experimental results obtained with an innovative calorimeter (CALMOS) for nuclear heating measurements

Nuclear heating inside an MTR reactor has to be known in order to be able to control samples temperature during irradiation experiments. An R and D program has been carried out at CEA to design a new type of in-core calorimetric system. This new development, started in 2002, has for main objective to manufacture a calorimeter suitable to monitoring nuclear heating inside the 70 MWth OSIRIS material testing reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. An innovative calorimetric probe, associated to a specific handling system, has been designed to provide access to measurements both along the fissile height and on the upper part of the core, where nuclear heating still remains high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for process validation, while a displacement system has been especially studied to move the probe along a given axial measurement range. This paper deals with the development, tests on preliminary mock-ups and the finalization of the probe. Main modeling and experimental results are presented. Moreover, alternative methods to calibration for nuclear heating rate measurements which are now possible with this new calorimeter are presented and discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Development, calibration, and experimental results obtained with an innovative calorimeter (CALMOS) for nuclear heating measurements

Nuclear heating inside an MTR reactor has to be known in order to be able to control samples temperature during irradiation experiments. An R and D program has been carried out at CEA to design a new type of in-core calorimetric system. This new development, started in 2002, has for main objective to manufacture a calorimeter suitable to monitoring nuclear heating inside the 70 MWth OSIRIS material testing reactor operated by CEA's Nuclear Energy Division at the Saclay research center. An innovative calorimetric probe, associated to a specific handling system, has been designed to provide access to measurements both along the fissile height and on the upper part of the core, where nuclear heating still remains high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for process validation, while a displacement system has been especially studied to move the probe along a given axial measurement range. This paper deals with the development, tests on preliminary mock-ups and the finalization of the probe. Main modeling and experimental results are presented. Moreover, alternative methods to calibration for nuclear heating rate measurements which are now possible with this new calorimeter are presented and discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1965

Multiple test chamber exposes materials to various environments

Multiple compartment test chamber exposes several material specimens to various environmental conditions for prolonged periods. The specimens are individually mounted in chamber compartments, rotated to various positions, and measured through optical windows to determine progressive changes in the material properties.

Johnston, R. L.↗
DOE OSTI2011

Analysis of gamma-ray dosimetry experiments in the zero power MINERVE facility

The objective of this study is to develop nuclear heating measurement methods in zero power experimental reactors. These developments contribute to the qualification of photonics calculation schemes for the assessment of gamma heating in the future Jules Horowitz Material Testing Reactor. This paper presents the analysis of thermoluminescent detector (TLD) experiments in the UO{sub 2} core of the MINERVE Research Reactor at the French Alternative Energies and Atomic Energy Commission center in Cadarache. The experimental sources of uncertainty in the gamma dose have been reduced by improving the measurement conditions and the repeatability of the calibration step for each individual TLD. The interpretation of these measurements needs to take into account the calculation of cavity correction factors related to calibration and irradiation configurations, as well as neutron correction calculations. These calculations are based on Monte Carlo simulations of neutron-gamma and gamma-electron transport coupled particles. The comparison between calculated and measured integral gamma-ray absorbed doses in the aluminum material surrounding the TLD shows that calculations slightly overestimate the measurement, with a calculated versus experimental ratio equal to 1.04 {+-} 5.7 % (k=2). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Performance of Composite Insulators Used for Electric Transmission under Extreme Climatic Conditions

Composite or polymeric insulators are being used in various parts of the world for high-voltage transmission and distribution lines. In some parts of India, places in the states like Rajasthan and Gujarat experience an increase in the daytime temperature up to 50 °C, while the nighttime temperature reduces as low as 3-5 °C. Because of these extreme temperature variations along with electric stress and humidity conditions, composite insulators have to perform well over their lifetime to ensure the reliability of the transmission network. In the present work, experimental conditions are simulated with temperature variation from 50 to 3-5 °C along with electrical stress, humidity variation, pollution, and ultraviolet radiations. Accelerated aging studies with these stresses are conducted to understand the degradation on the insulator similar to the field conditions. Surface resistivity measurements, leakage current analysis, and surface degradation studies using thermogravimetric analysis, scanning electron microscopy, energy-dispersive x-ray, and Fourier transform infrared spectroscopy (FTIR) are conducted. Some interesting results obtained from the study are presented. Experimentation under extreme conditions with polluted insulators showed considerable degradation which was inferred from the material testing analysis. The insulator slabs on which electrical stress was not applied showed a considerable peak at 1500 cm{sup −1} in the FTIR spectra, indicating the generation of low molecular weight PDMS chains from the bulk due to UV aging and thermal stress, while the degradation pattern of non-polluted sample was severe compared to the slab treated but was considerably less severe in comparison with the polluted insulator sample exhibiting pollution can considerably increase degradation rate.

36 MATERIALS SCIENCE↗
NASA NTRSDate not supplied

Prometheus and Constellation Materials Research and Testing Overview

Outline - New Technologies and New Materials meet Flight Program - Reactor Materials Selection - Material Radiation Resistance - Testing in four different radiation environments - Superalloy Creep Testing - Superalloy-Refractory Alloy Joining - Superalloy, Stainless Steel, Copper Brazing - Superalloy-Refractory Cross Contamination - Na and NaK Compatibility - Radiator Panel Materials Selection

nuclear power↗
DOE OSTI2011

Effective Materials Property Information Management for the 21st Century

This paper discusses key principles for the development of materials property information management software systems. There are growing needs for automated materials information management in various organizations. In part these are fuelled by the demands for higher efficiency in material testing, product design and engineering analysis. But equally important, organizations are being driven by the needs for consistency, quality and traceability of data, as well as control of access to proprietary or sensitive information. Further, the use of increasingly sophisticated nonlinear, anisotropic and multi-scale engineering analyses requires both processing of large volumes of test data for development of constitutive models and complex materials data input for Computer-Aided Engineering (CAE) software. And finally, the globalization of economy often generates great needs for sharing a single gold source of materials information between members of global engineering teams in extended supply-chains. Fortunately material property management systems have kept pace with the growing user demands and evolved to versatile data management systems that can be customized to specific user needs. The more sophisticated of these provide facilities for: (i) data management functions such as access, version, and quality controls; (ii) a wide range of data import, export and analysis capabilities; (iii) data pedigree traceability mechanisms; (iv) data searching, reporting and viewing tools; and (v) access to the information via a wide range of interfaces. In this paper the important requirements for advanced material data management systems, future challenges and opportunities such as automated error checking, data quality characterization, identification of gaps in datasets, as well as functionalities and business models to fuel database growth and maintenance are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2011

Analysis of fission gas release kinetics by on-line mass spectrometry

The release of fission gas (Xe and Kr) and helium out of nuclear fuel materials in normal operation of a nuclear power reactor can constitute a strong limitation of the fuel lifetime. Moreover, radioactive isotopes of Xe and Kr contribute significantly to the global radiological source term released in the primary coolant circuit in case of accidental situations accompanied by fuel rod loss of integrity. As a consequence, fission gas release investigation is of prime importance for the nuclear fuel cycle economy, and is the driven force of numerous R and D programs. In this domain, for solving current fuel behavior understanding issues, preparing the development of new fuels (e.g. for Gen IV power systems) and for improving the modeling prediction capability, there is a marked need for innovations in the instrumentation field, mainly for: . Quantification of very low fission gas concentrations, released from fuel sample and routed in sweeping lines. Monitoring of quick gas release variations by quantification of elementary release during a short period of time. Detection of a large range of atomic masses (e.g. H{sub 2}, HT, He, CO, CO{sub 2}, Ne, Ar, Kr, Xe), together with a performing separation of isotopes for Xe and Kr elements. Coupling measurement of stable and radioactive gas isotopes, by using in parallel mass spectrometry and gamma spectrometry techniques. To fulfill these challenging needs, a common strategy for analysis equipment implementation has been set up thanks to a recently launched collaboration between the CEA and the Univ. of Provence, with the technological support of the Liverpool Univ.. It aims at developing a chronological series of mass spectrometer devices based upon mass filter and 2D/3D ion traps with Fourier transform operating mode and having increasing levels of performances to match the previous challenges for out-of pile and in-pile experiments. The final objective is to install a high performance online mass spectrometer coupled to a gamma spectrometer in the fission product laboratory of the future Jules Horowitz Material Test Reactor. An intermediate step will consist of testing first equipment on an existing experimental facility in the LECA-STAR Hot Cell Laboratory of the CEA Cadarache. This paper presents the scientific and operational stakes linked to fission gas issues, resumes the current state of art for analyzing them in nuclear facilities, then presents the skills gathered through this collaboration to overcome technological bottlenecks. Finally it describes the implementation strategy in nuclear research facilities of the CEA Cadarache. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of three years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Reflector modelling of small high leakage cores making use of multi-group nodal equivalence theory

This research focuses on modelling reflectors in typical material testing reactors (MTRs). Equivalence theory is used to homogenise and collapse detailed transport solutions to generate equivalent nodal parameters and albedo boundary conditions for reflectors, for subsequent use in full core nodal diffusion codes. This approach to reflector modelling has been shown to be accurate for two-group large commercial light water reactor (LWR) analysis, but has not been investigated for MTRs. MTRs are smaller, with much larger leakage, environment sensitivity and multi-group spectrum dependencies than LWRs. This study aims to determine if this approach to reflector modelling is an accurate and plausible homogenisation technique for the modelling of small MTR cores. The successful implementation will result in simplified core models, better accuracy and improved efficiency of computer simulations. Codes used in this study include SCALE 6.1, OSCAR-4 and EQUIVA (the last two codes are developed and used at Necsa). The results show a five times reduction in calculational time for the proposed reduced reactor model compared to the traditional explicit model. The calculated equivalent parameters however show some sensitivity to the environment used to generate them. Differences in the results compared to the current explicit model, require more careful investigation including comparisons with a reference result, before its implementation can be recommended. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1992

CPL Materials Life Cycle Test Facility

The Capillary Pumped Loop (CPL) Materials Life Cycle Test Facility at the Goddard Space Flight Center (GSFC) will identify the operational parameters controlling the performance of a CPL over an extended period of time. The primary purpose of the facility is to investigate the long-term chemical compatibility between the anhydrous ammonia working fluid and the CPL materials of construction. Chemical reactions occurring within the system may produce non-condensable gases or particulate debris that can lead to a degradation in system performance. Small liquid samples will be drawn from the system at specific time intervals and analyzed to check for the presence of non-condensable gases. Periodic maximum and minimum heat load tests will be performed on the CPL to monitor trends in the overall system performance.

Buchko, Matthew T.↗
DOE OSTI2010

Testing of Small Graphite Samples for Nuclear Qualification

Accurately determining the mechanical properties of small irradiated samples is crucial to predicting the behavior of the overal irradiated graphite components within a Very High Temperature Reactor. The sample size allowed in a material test reactor, however, is limited, and this poses some difficulties with respect to mechanical testing. In the case of graphite with a larger grain size, a small sample may exhibit characteristics not representative of the bulk material, leading to inaccuracies in the data. A study to determine a potential size effect on the tensile strength was pursued under the Next Generation Nuclear Plant program. It focuses first on optimizing the tensile testing procedure identified in the American Society for Testing and Materials (ASTM) Standard C 781-08. Once the testing procedure was verified, a size effect was assessed by gradually reducing the diameter of the specimens. By monitoring the material response, a size effect was successfully identified.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

New Remote Method for Estimation of Contamination Levels of Reactor Equipment - 13175

Projects for decommissioning of shutdown reactors and reactor facilities carried out in several countries, including Russia. In the National Research Centre 'Kurchatov Institute' decontamination and decommissioning of the research reactor MR (Material Testing Reactor) has been initiated. The research reactor MR has a long history and consists of nine loop facilities for experiments with different kinds of fuel. During the operation of main and auxiliary equipment of reactors it was subjected to strong radioactive contamination. The character of this contamination requires individual strategies for the decontamination work. This requires information about the character of the distribution of radioactive contamination of equipment in the premises. A detailed radiation survey of these premises using standard dosimetric equipment is almost impossible because of high levels of radiation and high-density of the equipment that does not allow identifying the most active fragments using standard tools of measurement. The problem can be solved using the method of remote measurements of distribution of radioactivity with help of the collimated gamma-ray detectors. For radiation surveys of the premises of loop installations remotely operated spectrometric collimated system was used [1, 2, 3]. As a result of the work, maps of the distribution of activity and dose rate for surveyed premises were plotted and superimposed on its photo. The new results of measurements in different areas of the reactor and at its loop installations, with emphasis on the radioactive survey of highly-contaminated samples, are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Operation of Bubblers in the Savannah River Site Defense Waste Processing Facility Melter - 12166

Savannah River Remediation (SRR) LLC acquired the liquid waste contract at the Savannah River Site (SRS) in the summer of 2009. In order to achieve the main goal of the contract, closing of High Level Waste (HLW) tanks, it was necessary to process more waste throughout the SRS liquid waste facilities. The Defense Waste Processing Facility (DWPF) would need to increase its production rate of radioactive waste glass filled canisters as a part of the plan to achieve this commitment. To attain the increased production rate, four bubblers were installed in the DWPF Melter in September 2010 to agitate the DWPF Melter glass pool. The four bubblers were designed to be installed in existing nozzles on the top-head of the DWPF Melter. The design and fabrication of the four (4) bubblers was accomplished through SRR critical subcontractor EnergySolutions LLC. In addition to the existing bubbler design, a new design concept has been approved and is in the process of fabrication. The new design will allow for the lower end (inside melter) of the bubbler to be replaced while the upper end (outside melter) of the bubbler is reused to minimize cost and waste at the DWPF. The bubblers have been operating in the DWPF Melter for approximately 1 year. The originally installed bubbler set was replaced in January 2011. The bubblers were visually examined once removed from the melter and showed minimal signs of wear. Material testing of the Inconel 690 is being performed to determine if the bubblers operational life can be extended. The use of the bubblers has changed the dynamics within the melter glass pool. This is shown through indications that the bubblers have increased the glass pool circulation. Overall, performance of the bubblers has been encouraging and the DWPF Melter has seen a significant improvement in its ability to process waste since the bubbler installation. The installation of the bubblers accomplished the goal of increasing the glass production capability of DWPF Melter. Currently the DWPF Melter is producing glass at a rate of 91 to 93 kg of glass/hour (200 to 205 lbs of glass/hour). As a result of this increased glass production rate the DWPF has been able to produce the most canisters since radioactive operations began in 1996. These accomplishments can be directly attributed to the bubblers ability to increase mixing/convection within the glass melt pool thereby improving the heat transfer from melt pool to the cold cap. Further, this has helped SRR take a significant step forward toward closing the HLW tanks at the aggressive pace set forth in the most recent liquid waste contract at the SRS. As a path forward, the DWPF is continuing work with SRNL to complete the final metallurgical examination of the bubbler samples to help determine methods to increase the design life of the bubblers beyond the current 6-month restriction. Also, the DWPF is continuing work with EnergySolutions on a new bubbler design concept to minimize cost and waste at the DWPF. The new design will allow for the lower end (inside melter) of the bubbler to be replaced while the upper end (outside melter) of the bubbler is reused. This new design is currently in the final design approval phase and EnergySolutions will be selecting a fabricator to begin the work to build bubbler sets 5 and 6 using the new design. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗