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Sample records for agr-1 irradiation test

  1. AGR-1 Irradiation Experiment Test Plan

    SciTech Connect

    John T. Maki

    2009-10-01

    This document presents the current state of planning for the AGR-1 irradiation experiment, the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment will be irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The test will contain six independently controlled and monitored capsules. Each capsule will contain a single type, or variant, of the AGR coated fuel. The irradiation is planned for about 700 effective full power days (approximately 2.4 calendar years) with a time-averaged, volume-average temperature of approximately 1050 °C. Average fuel burnup, for the entire test, will be greater than 17.7 % FIMA, and the fuel will experience fast neutron fluences between 2.4 and 4.5 x 1025 n/m2 (E>0.18 MeV).

  2. AGR-1 Irradiation Test Final As-Run Report

    SciTech Connect

    Blaise P. Collin

    2012-06-01

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 ?1025 n/m2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below 10-7 with only one

  3. Validation of the Physics Analysis used to Characterize the AGR-1 TRISO Fuel Irradiation Test

    SciTech Connect

    Sterbentz, James W.; Harp, Jason M.; Demkowicz, Paul A.; Hawkes, Grant L.; Chang, Gray S.

    2015-05-01

    The results of a detailed physics depletion calculation used to characterize the AGR-1 TRISO-coated particle fuel test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory are compared to measured data for the purpose of validation. The particle fuel was irradiated for 13 ATR power cycles over three calendar years. The physics analysis predicts compact burnups ranging from 11.30-19.56% FIMA and cumulative neutron fast fluence from 2.21?4.39E+25 n/m2 under simulated high-temperature gas-cooled reactor conditions in the ATR. The physics depletion calculation can provide a full characterization of all 72 irradiated TRISO-coated particle compacts during and post-irradiation, so validation of this physics calculation was a top priority. The validation of the physics analysis was done through comparisons with available measured experimental data which included: 1) high-resolution gamma scans for compact activity and burnup, 2) mass spectrometry for compact burnup, 3) flux wires for cumulative fast fluence, and 4) mass spectrometry for individual actinide and fission product concentrations. The measured data are generally in very good agreement with the calculated results, and therefore provide an adequate validation of the physics analysis and the results used to characterize the irradiated AGR-1 TRISO fuel.

  4. AGR-1 Irradiation Test Final As-Run Report, Rev. 3

    SciTech Connect

    Collin, Blaise P.

    2015-01-01

    This document presents the as-run analysis of the AGR-1 irradiation experiment. AGR-1 is the first of eight planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the US Department of Energy (DOE) as part of the Next-Generation Nuclear Plant (NGNP) project. The objectives of the AGR-1 experiment are: 1. To gain experience with multi-capsule test train design, fabrication, and operation with the intent to reduce the probability of capsule or test train failure in subsequent irradiation tests. 2. To irradiate fuel produced in conjunction with the AGR fuel process development effort. 3. To provide data that will support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. In order to achieve the test objectives, the AGR-1 experiment was irradiated in the B-10 position of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) for a total duration of 620 effective full power days of irradiation. Irradiation began on December 24, 2006 and ended on November 6, 2009 spanning 13 ATR cycles and approximately three calendar years. The test contained six independently controlled and monitored capsules. Each capsule contained 12 compacts of a single type, or variant, of the AGR coated fuel. No fuel particles failed during the AGR-1 irradiation. Final burnup values on a per compact basis ranged from 11.5 to 19.6 %FIMA, while fast fluence values ranged from 2.21 to 4.39 x 1025 n/m2 (E >0.18 MeV). We’ll say something here about temperatures once thermal recalc is done. Thermocouples performed well, failing at a lower rate than expected. At the end of the irradiation, nine of the originally-planned 19 TCs were considered functional. Fission product release-to-birth (R/B) ratios were quite low. In most capsules, R/B values at the end of the irradiation were at or below

  5. AGR-1 Post Irradiation Examination Final Report

    SciTech Connect

    Demkowicz, Paul Andrew

    2015-08-01

    The post-irradiation examination (PIE) of the Advanced Gas Reactor (AGR)-1 experiment was a multi-year, collaborative effort between Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) to study the performance of UCO (uranium carbide, uranium oxide) tristructural isotropic (TRISO) coated particle fuel fabricated in the U.S. and irradiated at the Advanced Test Reactor at INL to a peak burnup of 19.6% fissions per initial metal atom. This work involved a broad array of experiments and analyses to evaluate the level of fission product retention by the fuel particles and compacts (both during irradiation and during post-irradiation heating tests to simulate reactor accident conditions), investigate the kernel and coating layer morphology evolution and the causes of coating failure, and explore the migration of fission products through the coating layers. The results have generally confirmed the excellent performance of the AGR-1 fuel, first indicated during the irradiation by the observation of zero TRISO coated particle failures out of 298,000 particles in the experiment. Overall release of fission products was determined by PIE to have been relatively low during the irradiation. A significant finding was the extremely low levels of cesium released through intact coatings. This was true both during the irradiation and during post-irradiation heating tests to temperatures as high as 1800°C. Post-irradiation safety test fuel performance was generally excellent. Silver release from the particles and compacts during irradiation was often very high. Extensive microanalysis of fuel particles was performed after irradiation and after high-temperature safety testing. The results of particle microanalysis indicate that the UCO fuel is effective at controlling the oxygen partial pressure within the particle and limiting kernel migration. Post-irradiation examination has provided the final body of data that speaks to the quality of the AGR-1 fuel, building

  6. Performance of AGR-1 High-Temperature Reactor Fuel During Post-Irradiation Heating Tests

    SciTech Connect

    Morris, Robert Noel; Baldwin, Charles A; Hunn, John D; Demkowicz, Paul; Reber, Edward

    2014-01-01

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide TRISO fuel compacts from the AGR-1 experiment has been evaluated at temperatures of 1600 1800 C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4 to 19.1% FIMA have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 10-6 after 300 h at 1600 C or 100 h at 1800 C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 C, and 85Kr release was very low during the tests (particles with breached SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 C in one compact. Post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.

  7. Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

    SciTech Connect

    Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; Hunn, John D.; Reber, Edward L.

    2016-05-18

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating

  8. Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

    DOE PAGES

    Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; ...

    2016-05-18

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium,more » and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers

  9. AGR-1 Irradiated Test Train Preliminary Inspection and Disassembly First Look

    SciTech Connect

    Paul Demkowicz; Lance Cole; Scott Ploger; Philip Winston; Binh Pham; Michael Abbott

    2011-01-01

    The AGR-1 irradiation experiment ended on November 6, 2009, after 620 effective full power days in the Advanced Test Reactor, achieving a peak burnup of 19.6% FIMA. The test train was shipped to the Materials and Fuels Complex in March 2010 for post-irradiation examination. The first PIE activities included non-destructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and the graphite fuel holders. Dimensional measurements of the compacts, graphite holders, and steel capsules shells were performed using a custom vision measurement system (for outer diameters and lengths) and conventional bore gauges (for inner diameters). Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Neutron radiography of the intact Capsule 2 showed a high degree of detail of interior components and confirmed the observation that there was no major damage to the capsule. Disassembly of the capsules was initiated using procedures qualified during out-of-cell mockup testing. Difficulties were encountered during capsule disassembly due to irradiation-induced changes in some of the capsule components’ properties, including embrittled niobium and molybdenum parts that were susceptible to fracture and swelling of the graphite fuel holders that affected their removal from the capsule shells. This required various improvised modifications to the disassembly procedure to avoid damage to the fuel compacts. Ultimately the capsule disassembly was successful and only one compact from Capsule 4 (out of 72 total in the test train) sustained damage during the disassembly process, along with the associated graphite holder. The compacts were generally in very good condition upon removal. Only relatively minor

  10. Completion of the first NGNP Advanced Gas Reactor Fuel Irradiation Experiment, AGR-1, in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover; John Maki; David Petti

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The design of AGR-1 test train and support systems used to monitor and control the experiment during

  11. First elevated-temperature performance testing of coated particle fuel compacts from the AGR-1 irradiation experiment

    SciTech Connect

    Charles A. Baldwin; John D. Hunn; Robert N. Morris; Fred C. Montgomery; Chinthaka M. Silva; Paul A. Demkowicz

    2014-05-01

    In the AGR-1 irradiation experiment, 72 coated-particle fuel compacts were taken to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures. This paper discusses the first post-irradiation test of these mixed uranium oxide/uranium carbide fuel compacts at elevated temperature to examine the fuel performance under a simulated depressurized conduction cooldown event. A compact was heated for 400 h at 1600 degrees C. Release of 85Kr was monitored throughout the furnace test as an indicator of coating failure, while other fission product releases from the compact were periodically measured by capturing them on exchangeable, water-cooled deposition cups. No coating failure was detected during the furnace test, and this result was verified by subsequent electrolytic deconsolidation and acid leaching of the compact, which showed that all SiC layers were still intact. However, the deposition cups recovered significant quantities of silver, europium, and strontium. Based on comparison of calculated compact inventories at the end of irradiation versus analysis of these fission products released to the deposition cups and furnace internals, the minimum estimated fractional losses from the compact during the furnace test were 1.9 x 10-2 for silver, 1.4 x 10-3 for europium, and 1.1 x 10-5 for strontium. Other post-irradiation examination of AGR-1 compacts indicates that similar fractions of europium and silver may have already been released by the intact coated particles during irradiation, and it is therefore likely that the detected fission products released from the compact in this 1600 degrees C furnace test were from residual fission products in the matrix. Gamma analysis of coated particles deconsolidated from the compact after the heating test revealed that silver content within each particle varied considerably; a result that is probably not related to the furnace test, because it has also been observed in other as-irradiated AGR-1 compacts. X

  12. AGR-1 Safety Test Predictions using the PARFUME code

    SciTech Connect

    Blaise Collin

    2012-05-01

    The PARFUME modeling code was used to predict failure probability of TRISO-coated fuel particles and diffusion of fission products through these particles during safety tests following the first irradiation test of the Advanced Gas Reactor program (AGR-1). These calculations support the AGR-1 Safety Testing Experiment, which is part of the PIE effort on AGR-1. Modeling of the AGR-1 Safety Test Predictions includes a 620-day irradiation followed by a 300-hour heat-up phase of selected AGR-1 compacts. Results include fuel failure probability, palladium penetration, and fractional release of fission products. Results show that no particle failure is predicted during irradiation or heat-up, and that fractional release of fission products is limited during irradiation but that it significantly increases during heat-up.

  13. Microscopic analysis of irradiated AGR-1 coated particle fuel compacts

    SciTech Connect

    Scott A. Ploger; Paul A. Demkowicz; John D. Hunn; Jay S. Kehn

    2014-05-01

    The AGR-1 experiment involved irradiation of 72 TRISO-coated particle fuel compacts to a peak compact-average burnup of 19.5% FIMA with no in-pile failures observed out of 3 x 105 total particles. Irradiated AGR-1 fuel compacts have been cross-sectioned and analyzed with optical microscopy to characterize kernel, buffer, and coating behavior. Six compacts have been examined, spanning a range of irradiation conditions (burnup, fast fluence, and irradiation temperature) and including all four TRISO coating variations irradiated in the AGR-1 experiment. The cylindrical specimens were sectioned both transversely and longitudinally, then polished to expose from 36 to 79 individual particles near midplane on each mount. The analysis focused primarily on kernel swelling and porosity, buffer densification and fracturing, buffer–IPyC debonding, and fractures in the IPyC and SiC layers. Characteristic morphologies have been identified, 981 particles have been classified, and spatial distributions of particle types have been mapped. No significant spatial patterns were discovered in these cross sections. However, some trends were found between morphological types and certain behavioral aspects. Buffer fractures were found in 23% of the particles, and these fractures often resulted in unconstrained kernel protrusion into the open cavities. Fractured buffers and buffers that stayed bonded to IPyC layers appear related to larger pore size in kernels. Buffer–IPyC interface integrity evidently factored into initiation of rare IPyC fractures. Fractures through part of the SiC layer were found in only four classified particles, all in conjunction with IPyC–SiC debonding. Compiled results suggest that the deliberate coating fabrication variations influenced the frequencies of IPyC fractures and IPyC–SiC debonds.

  14. Microscopic analysis of irradiated AGR-1 coated particle fuel compacts

    SciTech Connect

    Scott Ploger; Paul Demkowicz; John Hunn; Robert Morris

    2012-10-01

    The AGR-1 experiment involved irradiation of 72 TRISO-coated particle fuel compacts to a peak burnup of 19.5% FIMA with no in-pile failures observed out of 3×105 total particles. Irradiated AGR-1 fuel compacts have been cross-sectioned and analyzed with optical microscopy to characterize kernel, buffer, and coating behavior. Five compacts have been examined so far, spanning a range of irradiation conditions (burnup, fast fluence, and irradiation temperature) and including all four TRISO coating variations irradiated in the AGR-1 experiment. The cylindrical specimens were sectioned both transversely and longitudinally, then polished to expose between approximately 40-80 individual particles on each mount. The analysis focused primarily on kernel swelling and porosity, buffer densification and fracturing, buffer-IPyC debonding, and fractures in the IPyC and SiC layers. Characteristic morphologies have been identified, over 800 particles have been classified, and spatial distributions of particle types have been mapped. No significant spatial patterns were discovered in these cross sections. However, some trends were found between morphological types and certain behavioral aspects. Buffer fractures were found in approximately 23% of the particles, and these fractures often resulted in unconstrained kernel swelling into the open cavities. Fractured buffers and buffers that stayed bonded to IPyC layers appear related to larger pore size in kernels. Buffer-IPyC interface integrity evidently factored into initiation of rare IPyC fractures. Fractures through part of the SiC layer were found in only three particles, all in conjunction with IPyC-SiC debonding. Compiled results suggest that the deliberate coating fabrication variations influenced the frequencies of IPyC fractures, IPyC-SiC debonds, and SiC fractures.

  15. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    SciTech Connect

    Paul Demkowicz; Scott Ploger; John Hunn

    2012-05-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Five irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These five compacts also included all four TRISO coating variations irradiated in the AGR experiment. The five compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. Approximately 40 to 80 particles within each cross section were exposed near enough to mid-plane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 830 classified particles allowed other relationships among morphological types to be established.

  16. Ceramographic Examinations of Irradiated AGR-1 Fuel Compacts

    SciTech Connect

    Paul Demkowicz; Scott Ploger; John Hunn; Jay S. Kehn

    2012-09-01

    The AGR 1 experiment involved irradiating 72 cylindrical fuel compacts containing tri-structural isotropic (TRISO)-coated particles to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures observed out of almost 300,000 particles. Six irradiated AGR 1 fuel compacts were selected for microscopy that span a range of irradiation conditions (temperature, burnup, and fast fluence). These six compacts also included all four TRISO coating variations irradiated in the AGR experiment. The six compacts were cross-sectioned both transversely and longitudinally, mounted, ground, and polished after development of careful techniques for preserving particle structures against preparation damage. From 36 to 79 particles within each cross section were exposed near enough to midplane for optical microscopy of kernel, buffer, and coating behavior. The microstructural analysis focused on kernel swelling and porosity, buffer densification and fracture, debonding between the buffer and inner pyrolytic carbon (IPyC) layers, and fractures in the IPyC and SiC layers. Three basic particle morphologies were established according to the extent of bonding between the buffer and IPyC layers: complete debonding along the interface (Type A), no debonding along the interface (Type B), and partial debonding (Type AB). These basic morphologies were subdivided according to whether the buffer stayed intact or fractured. The resulting six characteristic morphologies were used to classify particles within each cross section, but no spatial patterns were clearly observed in any of the cross-sectional morphology maps. Although positions of particle types appeared random within compacts, examining a total of 931 classified particles allowed other relationships among morphological types to be established.

  17. Post-irradiation Examination and Fission Product Inventory Analysis of AGR-1 Irradiation Capsules

    SciTech Connect

    J M Harp; P D Demkowicz; S A Ploger

    2012-10-01

    The AGR-1 experiment was the first in a series of Advanced Gas Reactor (AGR) experiments designed to test TRISO fuel under High Temperature Gas Reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL’s Materials and Fuels Complex (MFC). The inventory and distribution of fission products, especially Ag-110m, was assessed and analyzed for all the components of the AGR-1 capsules. This data should help inform the study of fission product migration in coated particle fuel. Gamma spectrometry was used to measure the activity of various different fission products in the different components of the AGR-1 test train. Each capsule contained: 12 fuel compacts, a graphite holder that kept the fuel compacts in place, graphite spacers that were above and below the graphite holders and fuel compacts, gas lines through which a helium neon gas mixture flowed in and out of each capsule, and the stainless steel shell that contained the experiment. Gamma spectrometry results and the experimental techniques used to capture these results will be presented for all the capsule components. The components were assayed to determine the total activity of different fission products present in or on them. These totals are compared to the total expected activity of a particular fission product in the capsule based on predictions from physics simulation. Based on this metric, a significant fraction of the Ag-110m was detected outside the fuel compacts, but the amount varied highly between the 6 capsules. Very small fractions of Cs-137 (<2E-5), Cs-134 (<1e-5), and Eu-154 (<4e-4) were detected outside of the fuel compacts. Additionally, the distribution of select fission products in some of the components including the fuel compacts and the graphite holders were measured and will be discussed.

  18. Irradiation performance of AGR-1 high temperature reactor fuel

    DOE PAGES

    Demkowicz, Paul A.; Hunn, John D.; Ploger, Scott A.; ...

    2015-10-23

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that itmore » was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10–4 to 5 × 10–4 for 154Eu and 8 × 10–7 to 3 × 10–5 for 90Sr. The average 134Cs fractional release from compacts was <3 × 10–6 when all particles maintained intact SiC. An estimated four particles out of 2.98 × 105 in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs fractional release in two capsules to approximately 10–5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. In conclusion, palladium, silver, and

  19. Irradiation performance of AGR-1 high temperature reactor fuel

    SciTech Connect

    Demkowicz, Paul A.; Hunn, John D.; Ploger, Scott A.; Morris, Robert N.; Baldwin, Charles A.; Harp, Jason M.; Winston, Philip L.; Gerczak, Tyler J.; van Rooyen, Isabella J.; Montgomery, Fred C.; Silva, Chinthaka M.

    2015-10-23

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10–4 to 5 × 10–4 for 154Eu and 8 × 10–7 to 3 × 10–5 for 90Sr. The average 134Cs fractional release from compacts was <3 × 10–6 when all particles maintained intact SiC. An estimated four particles out of 2.98 × 105 in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs fractional release in two capsules to approximately 10–5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that

  20. Irradiation performance of AGR-1 high temperature reactor fuel

    SciTech Connect

    Paul A. Demkowicz; John D. Hunn; Robert N. Morris; Charles A. Baldwin; Philip L. Winston; Jason M. Harp; Scott A. Ploger; Tyler Gerczak; Isabella J. van Rooyen; Fred C. Montgomery; Chinthaka M. Silva

    2014-10-01

    The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO-coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.5% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel–including the extent of fission product release and the evolution of kernel and coating microstructures–was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110mAg from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocrabon and compact matrix. The capsule-average fractional release from the compacts was 1×10 4 to 5×10 4 for 154Eu and 8×10 7 to 3×10 5 for 90Sr. The average 134Cs release from compacts was <3×10 6 when all particles maintained intact SiC. An estimated four particles out of 2.98×105 experienced partial cesium release due to SiC failure during the irradiation, driving 134Cs release in two capsules to approximately 10 5. Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. Palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization

  1. PIE on Safety-Tested AGR-1 Compact 5-1-1

    SciTech Connect

    Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.; Montgomery, Fred C.; Gerczak, Tyler J.

    2015-08-01

    Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.

  2. DETERMINATION OF THE AGR-1 CAPSULE TO FPMS SPECTROMETER TRANSPORT VOLUMES FROM LEADOUT FLOW TEST DATA

    SciTech Connect

    J. K. Hartwell; J. B. Walter; D. M. Scates; M. W. Drigert

    2007-05-01

    The AGR-1 experiment is a fueled multiple-capsule irradiation experiment being conducted in the Advanced Test Reactor (ATR) in support of the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. A flow experiment conducted during the AGR-1 irradiation provided data that included the effect of flow rate changes on the decay of a short-lived radionuclide (23Ne). This data has been analyzed to determine the capsule-specific downstream transport volume through which the capsule effluents must pass before arrival at the fission product monitoring system spectrometers. These resultant transport volumes when coupled with capsule outlet flow rates determine the transport times from capsule-to-detector. In this work an analysis protocol is developed and applied in order to determine capsule-specific transport volumes to precisions of better than +/- 7%.

  3. Data Compilation for AGR-1 Pre-Production Test: NUCO350-75T-Z

    SciTech Connect

    Hunn, John D; Lowden, Richard Andrew; Pappano, Peter J

    2006-03-01

    This document is a compilation of characterization data for compact lot NUCO350-75T-Z. This compact lot was fabricated using particle composite NUCO350-75T, which was a composite of three batches of TRISO-coated 350 m natural uranium oxide/uranium carbide kernels (NUCO). The compacts and coated particles were produced as part of a development effort at ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program. The kernels were obtained from BWXT and were identified as composite G73B-NU-69300. The BWXT kernel lot G73B-NU-69300 was riffled into sublots for characterization and coating. The ORNL identification for these kernel sublots was NUCO350-## (where ## were a series of integers beginning with 01). NUCO350-75T-Z was produced as part of the ORNL AGR development effort and is not fully representative of a final product. This compact lot was the first run through of the entire ORNL AGR-1 irradiation test fuel production process involving coating, characterization, and compacting of TRISO-coated 350 m NUCO. The results of this exercise were used to fine tune the irradiation test fuel production process and as a basis for the decision to proceed with the production of the baseline fuel for the AGR-1 irradiation test.

  4. Comparison of fission product release predictions using PARFUME with results from the AGR-1 irradiation experiment

    SciTech Connect

    Blaise Collin

    2014-09-01

    This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive

  5. Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    SciTech Connect

    Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; Maki, John T.

    2016-04-07

    Safety tests were conducted on fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800 °C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during 15 of these safety tests. Comparisons between PARFUME predictions and post-irradiation examination results of the safety tests were conducted on two types of AGR-1 compacts: compacts containing only intact particles and compacts containing one or more particles whose SiC layers failed during safety testing. In both cases, PARFUME globally over-predicted the experimental release fractions by several orders of magnitude: more than three (intact) and two (failed SiC) orders of magnitude for silver, more than three and up to two orders of magnitude for strontium, and up to two and more than one orders of magnitude for krypton. The release of cesium from intact particles was also largely over-predicted (by up to five orders of magnitude) but its release from particles with failed SiC was only over-predicted by a factor of about 3. These over-predictions can be largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. The integral release nature of the data makes it difficult to estimate the individual over-estimations in the kernel or each coating layer. Nevertheless, a tentative assessment of correction factors to these diffusivities was performed to enable a better match between the modeling predictions and the safety testing results. The method could only be successfully applied to silver and cesium. In the case of strontium, correction factors could not be assessed because

  6. Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    DOE PAGES

    Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; ...

    2016-04-07

    Safety tests were conducted on fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800 °C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during 15 of these safety tests. Comparisons between PARFUME predictions and post-irradiation examination results of the safety tests were conducted on two types of AGR-1 compacts: compactsmore » containing only intact particles and compacts containing one or more particles whose SiC layers failed during safety testing. In both cases, PARFUME globally over-predicted the experimental release fractions by several orders of magnitude: more than three (intact) and two (failed SiC) orders of magnitude for silver, more than three and up to two orders of magnitude for strontium, and up to two and more than one orders of magnitude for krypton. The release of cesium from intact particles was also largely over-predicted (by up to five orders of magnitude) but its release from particles with failed SiC was only over-predicted by a factor of about 3. These over-predictions can be largely attributed to an over-estimation of the diffusivities used in the modeling of fission product transport in TRISO-coated particles. The integral release nature of the data makes it difficult to estimate the individual over-estimations in the kernel or each coating layer. Nevertheless, a tentative assessment of correction factors to these diffusivities was performed to enable a better match between the modeling predictions and the safety testing results. The method could only be successfully applied to silver and cesium. In the case of strontium, correction factors could not be assessed

  7. AGR-1 Compact 4-1-1 Post-Irradiation Examination Results

    SciTech Connect

    Demkowicz, Paul Andrew; Harp, Jason M.; Winston, Philip L.; Ploger, Scott A.; van Rooyen, Isabella J.

    2016-02-01

    Destructive post-irradiation examination was performed on AGR-1 fuel Compact 4-1-1, which was irradiated to a final compact-average burnup of 19.4% FIMA (fissions per initial metal atom) and a time-average, volume-average temperature of 1072°C. The analysis of this compact focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy, electron microscopy, and elemental analysis. Deconsolidation-leach-burn-leach (DLBL) analysis revealed no particles with failed TRISO or failed SiC layers (as indicated by very low uranium inventory in all of the leach solutions). The total fractions of the predicted compact inventories of fission products Ce-144, Cs-134, Cs-137, and Sr-90 that were present in the compact outside of the SiC layers were <2×10-6, based on DLBL data. The Ag-110m fraction in the compact outside the SiC layers was 3.3×10-2, indicating appreciable release of silver through the intact coatings and subsequent retention in the OPyC layers or matrix. The Eu-154 fraction was 2.4×10-4, which is equivalent to the inventory in one average particle, and indicates a small but measurable level of release from the intact coatings. Gamma counting of 61 individual particles indicated no particles with anomalously low fission product retention. The average ratio of measured inventory to calculated inventory was close to a value of 1.0 for several fission product isotopes (Ce-144, Cs-134, and Cs-137), indicating good retention and reasonably good agreement with the predicted inventories. Measured-to-calculated (M/C) activity ratios for fission products Eu-154, Eu-155, Ru-106, Sb

  8. AGR-1 Fuel Compact 6-3-2 Post-Irradiation Examination Results

    SciTech Connect

    Paul demkowicz; jason Harp; Scott Ploger

    2012-12-01

    Destructive post-irradiation examination was performed on fuel Compact 6-3-2, which was irradiated in the AGR-1 experiment to a final compact average burnup of 11.3% FIMA and a time-average, volume-average temperature of 1070°C. The analysis of this compact was focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, measurement of fuel burnup by several methods, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy. A single particle with a defective SiC layer was identified during deconsolidation-leach-burn-leach analysis, which is in agreement with previous measurements showing elevated cesium in the Capsule 6 graphite fuel holder associated with this fuel compact. The fraction of the compact europium inventory released from the particles and retained in the matrix was relatively high (approximately 6E-3), indicating release from intact particle coatings. The Ag-110m inventory in individual particles exhibited a very broad distribution, with some particles retaining =80% of the predicted inventory and others retaining less than 25%. The average degree of Ag-110m retention in 60 gamma counted particles was approximately 50%. This elevated silver release is in agreement with analysis of silver on the Capsule 6 components, which indicated an average release of 38% of the Capsule 6 inventory from the fuel compacts. In spite of the relatively high degree of silver release from the particles, virtually none of the Ag-110m released was found in the compact matrix, and presumably migrated out of the compact and was deposited on the irradiation capsule components. Release of all other fission products from the particles appears to be less than a single

  9. Preliminary results of post-irradiation examination of the AGR-1 TRISO fuel compacts

    SciTech Connect

    Paul Demkowicz; John Hunn; Robert Morris; Jason Harp; Philip Winston; Charles Baldwin; Fred Montgomery; Scott Ploger; Isabella van Rooyen

    2012-10-01

    Five irradiated fuel compacts from the AGR-1 experiment have been examined in detail in order to assess in-pile fission product release behavior. Compacts were electrolytically deconsolidated and analyzed using the leach-burn-leach technique to measure fission product inventory in the compact matrix and identify any particles with a defective SiC layer. Loose particles were then gamma counted to measure the fission product inventory. One particle with a defective SiC layer was found in the five compacts examined. The fractional release of Ag 110m from the particles was significant. The total fraction of silver released from all the particles within a compact ranged from 0-0.63 and individual particles within a single compact often exhibited a very wide range of silver release. The average fractional release of Eu-154 from all particles in a compact was 2.4×10-4—1.3×10-2, which is indicative of release through intact coatings. The fractional Cs-134 inventory in the compact matrix was <2×10-5 when all coatings remained intact, indicating good cesium retention. Approximately 1% of the palladium inventory was found in the compact matrix for two of the compacts, indicating significant release through intact coatings.

  10. Identification of Silver and Palladium in Irradiated TRISO Coated Particles of the AGR-1 Experiment

    SciTech Connect

    van Rooyen, Y. J.; Lillo, T. M.; Wu, Y. Q.

    2014-03-01

    Evidence of the release of certain metallic fission product through intact tristructural isotropic (TRISO) particles has been seen for decades around the world, as well as in the recent AGR-1 experiment at Idaho National Laboratory (INL). However, understanding the basic mechanism of transport is still lacking. This understanding is important because the TRISO coating is part of the high temperature gas reactor functional containment and critical for the safety strategy for licensing purposes. Our approach to identify fission products in irradiated AGR-1 TRISO fuel using scanning transmission electron microscopy (STEM), Electron Energy Loss Spectroscopy (EELS) and Energy Filtered TEM (EFTEM), has led to first-of-a-kind data at the nano-scale indicating the presence of silver at triple points and grain boundaries of the SiC layer in the TRISO particle. Cadmium was also found in the triple junctions. In this initial study, the silver was only identified in SiC grain boundaries and triple points on the edge of the SiC-IPyC interface up to a depth of approximately 0.5 um. Palladium was identified as the main constituent of micron-sized precipitates present at the SiC grain boundaries. Additionally spherical nano-sized palladium rich precipitates were found inside the SiC grains. These nano-sized Pd precipitates were distributed up to a depth of 5 um away from the SiC-IPyC interlayer. No silver was found in the center of the micron-sized fission product precipitates using these techniques, although silver was found on the outer edge of one of the Pd-U-Si containing precipitates which was facing the IPyC layer. Only Pd-U containing precipitates were identified in the IPyC layer and no silver was identified in the IPyC layer. The identification of silver alongside the grain boundaries and the findings of Pd alongside grain boundaries as well as inside the grains, provide significant knowledge for understanding silver and palladium transport in TIRSO fuel, which has been

  11. Comparison of fission product release predictions using PARFUME with results from the AGR-1 safety tests

    SciTech Connect

    Blaise Collin

    2014-09-01

    Safety tests were conducted on fourteen fuel compacts from AGR-1, the first irradiation experiment of the Advanced Gas Reactor (AGR) Fuel Development and Qualification program, at temperatures ranging from 1600 to 1800°C to determine fission product release at temperatures that bound reactor accident conditions. The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, strontium, and krypton from fuel compacts containing tristructural isotropic (TRISO) coated particles during the safety tests, and the predicted values were compared with experimental results. Preliminary comparisons between PARFUME predictions and post-irradiation examination (PIE) results of the safety tests show different trends in the prediction of the fractional release depending on the species, and it leads to different conclusions regarding the diffusivities used in the modeling of fission product transport in TRISO-coated particles: • For silver, the diffusivity in silicon carbide (SiC) might be over-estimated by a factor of at least 102 to 103 at 1600°C and 1700°C, and at least 10 to 102 at 1800°C. The diffusivity of silver in uranium oxy-carbide (UCO) might also be over-estimated, but the available data are insufficient to allow definitive conclusions to be drawn. • For cesium, the diffusivity in UCO might be over-estimated by a factor of at least 102 to 103 at 1600°C, 105 at 1700°C, and 103 at 1800°C. The diffusivity of cesium in SiC might also over-estimated, by a factor of 10 at 1600°C and 103 at 1700°C, based upon the comparisons between calculated and measured release fractions from intact particles. There is no available estimate at 1800°C since all the compacts heated up at 1800°C contain particles with failed SiC layers whose release dominates the release from intact particles. • For strontium, the diffusivity in SiC might be over-estimated by a factor of 10 to 102 at 1600 and 1700°C, and 102 to 103 at 1800°C. These

  12. AGR-1 Thermocouple Data Analysis

    SciTech Connect

    Jeff Einerson

    2012-05-01

    This report documents an effort to analyze measured and simulated data obtained in the Advanced Gas Reactor (AGR) fuel irradiation test program conducted in the INL's Advanced Test Reactor (ATR) to support the Next Generation Nuclear Plant (NGNP) R&D program. The work follows up on a previous study (Pham and Einerson, 2010), in which statistical analysis methods were applied for AGR-1 thermocouple data qualification. The present work exercises the idea that, while recognizing uncertainties inherent in physics and thermal simulations of the AGR-1 test, results of the numerical simulations can be used in combination with the statistical analysis methods to further improve qualification of measured data. Additionally, the combined analysis of measured and simulation data can generate insights about simulation model uncertainty that can be useful for model improvement. This report also describes an experimental control procedure to maintain fuel target temperature in the future AGR tests using regression relationships that include simulation results. The report is organized into four chapters. Chapter 1 introduces the AGR Fuel Development and Qualification program, AGR-1 test configuration and test procedure, overview of AGR-1 measured data, and overview of physics and thermal simulation, including modeling assumptions and uncertainties. A brief summary of statistical analysis methods developed in (Pham and Einerson 2010) for AGR-1 measured data qualification within NGNP Data Management and Analysis System (NDMAS) is also included for completeness. Chapters 2-3 describe and discuss cases, in which the combined use of experimental and simulation data is realized. A set of issues associated with measurement and modeling uncertainties resulted from the combined analysis are identified. This includes demonstration that such a combined analysis led to important insights for reducing uncertainty in presentation of AGR-1 measured data (Chapter 2) and interpretation of

  13. Comparison of silver, cesium, and strontium release predictions using PARFUME with results from the AGR-1 irradiation experiment

    NASA Astrophysics Data System (ADS)

    Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; Maki, John T.

    2015-11-01

    The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, and strontium from tristructural isotropic coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination (PIE) measurements provided data on release of these fission products from fuel compacts and fuel particles, and retention of silver in the compacts outside of the silicon carbide (SiC) layer. PARFUME-predicted fractional release of silver, cesium, and strontium was determined and compared to the PIE measurements. For silver, comparisons show a trend of over-prediction at low burnup and under-prediction at high burnup. PARFUME has limitations in the modeling of the temporal and spatial distributions of the temperature and burnup across the compacts, which affects the accuracy of its predictions. Nevertheless, the comparisons on silver release lie in the same order of magnitude. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed SiC layers, the over-prediction is by a factor of up to 3, corresponding to a potential over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of up to 250. For intact particles, whose release is much lower, the over-prediction is by a factor of up to 100, which could be attributed to an over-estimated diffusivity in SiC by about 40% on average. The release of strontium from intact particles is also over-predicted by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release

  14. Comparison of silver, cesium, and strontium release predictions using PARFUME with results from the AGR-1 irradiation experiment

    DOE PAGES

    Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; ...

    2015-08-22

    The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, and strontium from tristructural isotropic coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination measurements provided data on release of these fission products from fuel compacts andmore » fuel particles, and retention of silver in the compacts outside of the silicon carbide (SiC) layer. PARFUME-predicted fractional release of silver, cesium, and strontium was determined and compared to the PIE measurements. For silver, comparisons show a trend of over-prediction at low burnup and under-prediction at high burnup. PARFUME has limitations in the modeling of the temporal and spatial distributions of the temperature and burnup across the compacts, which affects the accuracy of its predictions. Nevertheless, the comparisons on silver release lie in the same order of magnitude. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed SiC layers, the over-prediction is by a factor of up to 3, corresponding to a potential over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of up to 250. For intact particles, whose release is much lower, the over-prediction is by a factor of up to 100, which could be attributed to an over-estimated diffusivity in SiC by about 40% on average. The release of strontium from intact particles is also over-predicted by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release

  15. Comparison of silver, cesium, and strontium release predictions using PARFUME with results from the AGR-1 irradiation experiment

    SciTech Connect

    Collin, Blaise P.; Petti, David A.; Demkowicz, Paul A.; Maki, John T.

    2015-08-22

    The PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, and strontium from tristructural isotropic coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination measurements provided data on release of these fission products from fuel compacts and fuel particles, and retention of silver in the compacts outside of the silicon carbide (SiC) layer. PARFUME-predicted fractional release of silver, cesium, and strontium was determined and compared to the PIE measurements. For silver, comparisons show a trend of over-prediction at low burnup and under-prediction at high burnup. PARFUME has limitations in the modeling of the temporal and spatial distributions of the temperature and burnup across the compacts, which affects the accuracy of its predictions. Nevertheless, the comparisons on silver release lie in the same order of magnitude. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed SiC layers, the over-prediction is by a factor of up to 3, corresponding to a potential over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of up to 250. For intact particles, whose release is much lower, the over-prediction is by a factor of up to 100, which could be attributed to an over-estimated diffusivity in SiC by about 40% on average. The release of strontium from intact particles is also over-predicted by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release from

  16. Electron microscopic evaluation and fission product identification of irradiated TRISO coated particles from the AGR-1 experiment: A preliminary Study

    SciTech Connect

    I J van Rooyen; D E Janney; B D Miller; J L Riesterer; P A Demkowicz

    2012-10-01

    ABSTRACT Post-irradiation examination of coated particle fuel from the AGR-1 experiment is in progress at Idaho National Laboratory and Oak Ridge National Laboratory. In this presentation a brief summary of results from characterization of microstructures in the coating layers of selected irradiated fuel particles with burnup of 11.3% and 19.3% FIMA will be given. The main objective of the characterization were to study irradiation effects, fuel kernel porosity, layer debonding, layer degradation or corrosion, fission-product precipitation, grain sizes, and transport of fission products from the kernels across the TRISO layers. Characterization techniques such as scanning electron microscopy, transmission electron microscopy, energy dispersive spectroscopy, and wavelength dispersive spectroscopy were used. A new approach to microscopic quantification of fission-product precipitates is also briefly demonstrated. The characterization emphasized fission-product precipitates in the SiC-IPyC interface, SiC layer and the fuel-buffer interlayer, and provided significant new insights into mechanisms of fission-product transport. Although Pd-rich precipitates were identified at the SiC-IPyC interlayer, no significant SiC-layer thinning was observed for the particles investigated. Characterization of these precipitates highlighted the difficulty of measuring low concentration Ag in precipitates with significantly higher concentrations of contain Pd and U. Different approaches to resolving this problem are discussed. Possible microstructural differences between particles with high and low releases of Ag particles are also briefly discussed, and an initial hypothesis is provided to explain fission-product precipitate compositions and locations. No SiC phase transformations or debonding of the SiC-IPyC interlayer as a result of irradiation were observed. Lessons learned from the post-irradiation examination are described and future actions are recommended.

  17. AGR-1 Data Qualification Report

    SciTech Connect

    Michael Abbott

    2010-03-01

    ABSTRACT Projects for the very high temperature reactor (VHTR) Technology Development Office (TDO) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor experiment (AGR-1), the processing of these data within NDMAS, and reports the qualification status of the data. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category assigned by the data generator. They include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent quality assurance program. The NDMAS database processing and qualification status of the following five data streams is reported in this document: 1. Fuel fabrication data. All data have been processed into the NDMAS database and qualified (1,819 records). 2. Fuel irradiation data. Data from all 13 AGR-1 reactor cycles have been processed into the NDMAS database and tested. Of these, 85% have been qualified and 15% have failed NDMAS accuracy testing. 3. FPMS data. Reprocessed (January 2010) data from all 13 AGR-1 reactor cycles have been processed into the database and capture tested. Final qualification of these data will be recorded after QA approval of an Engineering Calculations and Analysis Report

  18. First high temperature safety tests of AGR-1 TRISO fuel with the Fuel Accident Condition Simulator (FACS) furnace

    NASA Astrophysics Data System (ADS)

    Demkowicz, Paul A.; Reber, Edward L.; Scates, Dawn M.; Scott, Les; Collin, Blaise P.

    2015-09-01

    Three TRISO fuel compacts from the AGR-1 irradiation experiment were subjected to safety tests at 1600 and 1800 °C for approximately 300 h to evaluate the fission product retention characteristics. Silver behavior was dominated by rapid release of an appreciable fraction of the compact inventory (3-34%) at the beginning of the tests, believed to be from inventory residing in the compact matrix and outer pyrocarbon (OPyC) prior to the safety test. Measurable release of silver from intact particles appears to become apparent only after ∼60 h at 1800 °C. The release rate for europium and strontium was nearly constant for 300 h at 1600 °C (reaching maximum values of approximately 2 × 10-3 and 8 × 10-4 respectively), and at this temperature the release may be mostly limited to inventory in the compact matrix and OPyC prior to the safety test. The release rate for both elements increased after approximately 120 h at 1800 °C, possibly indicating additional measurable release through the intact particle coatings. Cesium fractional release from particles with intact coatings was <10-6 after 300 h at 1600 °C or 100 h at 1800 °C, but release from the rare particles that experienced SiC failure during the test could be significant. However, Kr release was still very low for 300 h 1600 °C (<2 × 10-6). At 1800 °C, krypton release increased noticeably after SiC failure, reflecting transport through the intact outer pyrocarbon layer. Nonetheless, the krypton and cesium release fractions remained less than approximately 10-3 after 277 h at 1800 °C.

  19. First high temperature safety tests of AGR-1 TRISO fuel with the Fuel Accident Condition Simulator (FACS) furnace

    SciTech Connect

    Demkowicz, Paul A.; Reber, Edward L.; Scates, Dawn M.; Scott, Les; Collin, Blaise P.

    2015-09-01

    Three TRISO fuel compacts from the AGR-1 irradiation experiment were subjected to safety tests at 1600 and 1800 °C for approximately 300 h to evaluate the fission product retention characteristics. Silver behavior was dominated by rapid release of an appreciable fraction of the compact inventory (3–34%) at the beginning of the tests, believed to be from inventory residing in the compact matrix and outer pyrocarbon (OPyC) prior to the safety test. Measurable release of silver from intact particles appears to become apparent only after ~60 h at 1800 °C. The release rate for europium and strontium was nearly constant for 300 h at 1600 °C (reaching maximum values of approximately 2×10⁻³ and 8×10⁻⁴ respectively), and at this temperature the release may be mostly limited to inventory in the compact matrix and OPyC prior to the safety test. The release rate for both elements increased after approximately 120 h at 1800 °C, possibly indicating additional measurable release through the intact particle coatings. Cesium fractional release from particles with intact coatings was <10⁻⁶ after 300 h at 1600 °C or 100 h at 1800 °C, but release from the rare particles that experienced SiC failure during the test could be significant. However, Kr release was still very low for 300 h 1600 °C (<2 × 10⁻⁶). At 1800 °C, krypton release increased noticeably after SiC failure, reflecting transport through the intact outer pyrocarbon layer. Nonetheless, the krypton and cesium release fractions remained less than approximately 10⁻³ after 277 h at 1800 °C.

  20. Uncertainty Quantification of Calculated Temperatures for the AGR-1 Experiment

    SciTech Connect

    Binh T. Pham; Jeffrey J. Einerson; Grant L. Hawkes

    2012-04-01

    This report documents an effort to quantify the uncertainty of the calculated temperature data for the first Advanced Gas Reactor (AGR-1) fuel irradiation experiment conducted in the INL's Advanced Test Reactor (ATR) in support of the Next Generation Nuclear Plant (NGNP) R&D program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR-1 test, the results of the numerical simulations can be used in combination with the statistical analysis methods to improve qualification of measured data. Additionally, the temperature simulation data for AGR tests can be used for validation of the fuel transport and fuel performance simulation models. The crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. The report is organized into three chapters. Chapter 1 introduces the AGR Fuel Development and Qualification program and provides overviews of AGR-1 measured data, AGR-1 test configuration and test procedure, and thermal simulation. Chapters 2 describes the uncertainty quantification procedure for temperature simulation data of the AGR-1 experiment, namely, (i) identify and quantify uncertainty sources; (ii) perform sensitivity analysis for several thermal test conditions; (iii) use uncertainty propagation to quantify overall response temperature uncertainty. A set of issues associated with modeling uncertainties resulting from the expert assessments are identified. This also includes the experimental design to estimate the main effects and interactions of the important thermal model parameters. Chapter 3 presents the overall uncertainty results for the six AGR-1 capsules. This includes uncertainties for the daily volume-average and peak fuel temperatures, daily average temperatures at TC locations, and time-average volume-average and time-average peak fuel temperatures.

  1. Uncertainty Quantification of Calculated Temperatures for the AGR-1 Experiment

    SciTech Connect

    Binh T. Pham; Jeffrey J. Einerson; Grant L. Hawkes

    2013-03-01

    This report documents an effort to quantify the uncertainty of the calculated temperature data for the first Advanced Gas Reactor (AGR-1) fuel irradiation experiment conducted in the INL’s Advanced Test Reactor (ATR) in support of the Next Generation Nuclear Plant (NGNP) R&D program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR-1 test, the results of the numerical simulations can be used in combination with the statistical analysis methods to improve qualification of measured data. Additionally, the temperature simulation data for AGR tests can be used for validation of the fuel transport and fuel performance simulation models. The crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. The report is organized into three chapters. Chapter 1 introduces the AGR Fuel Development and Qualification program and provides overviews of AGR-1 measured data, AGR-1 test configuration and test procedure, and thermal simulation. Chapters 2 describes the uncertainty quantification procedure for temperature simulation data of the AGR-1 experiment, namely, (i) identify and quantify uncertainty sources; (ii) perform sensitivity analysis for several thermal test conditions; (iii) use uncertainty propagation to quantify overall response temperature uncertainty. A set of issues associated with modeling uncertainties resulting from the expert assessments are identified. This also includes the experimental design to estimate the main effects and interactions of the important thermal model parameters. Chapter 3 presents the overall uncertainty results for the six AGR-1 capsules. This includes uncertainties for the daily volume-average and peak fuel temperatures, daily average temperatures at TC locations, and time-average volume-average and time-average peak fuel temperatures.

  2. AGR-1 Data Qualification Interim Report

    SciTech Connect

    Machael Abbott

    2009-08-01

    Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the first NDMAS objective. It describes the data streams associated with the first Advanced Gas Reactor (AGR-1) experiment, the processing of these data within NDMAS, and reports the interim FY09 qualification status of the AGR-1 data to date. Data qualification activities within NDMAS for specific types of data are determined by the data qualification category, which is assigned by the data generator, and include: (1) capture testing, to confirm that the data stored within NDMAS are identical to the raw data supplied, (2) accuracy testing, to confirm that the data are an accurate representation of the system or object being measured, and (3) documentation that the data were collected under an NQA-1 or equivalent QA program. The interim qualification status of the following four data streams is reported in this document: (1) fuel fabrication data, (2) fuel irradiation data, (3) fission product monitoring system (FPMS) data, and (4) Advanced Test Reactor (ATR) operating conditions data. A final report giving the NDMAS qualification status of all AGR-1 data (including cycle 145A) is planned for February 2010.

  3. STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

    NASA Astrophysics Data System (ADS)

    Leng, B.; van Rooyen, I. J.; Wu, Y. Q.; Szlufarska, I.; Sridharan, K.

    2016-07-01

    Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.

  4. Sensitivity Evaluation of the Daily Thermal Predictions of the AGR-1 Experiment in the Advanced Test Reactor

    SciTech Connect

    Grant Hawkes; James Sterbentz; John Maki

    2011-05-01

    A temperature sensitivity evaluation has been performed for the AGR-1 fuel experiment on an individual capsule. A series of cases were compared to a base case by varying different input parameters into the ABAQUS finite element thermal model. These input parameters were varied by ±10% to show the temperature sensitivity to each parameter. The most sensitive parameters are the outer control gap distance, heat rate in the fuel compacts, and neon gas fraction. Thermal conductivity of the compacts and graphite holder were in the middle of the list for sensitivity. The smallest effects were for the emissivities of the stainless steel, graphite, and thru tubes. Sensitivity calculations were also performed varying with fluence. These calculations showed a general temperature rise with an increase in fluence. This is a result of the thermal conductivity of the fuel compacts and graphite holder decreasing with fluence.

  5. Data Compilation for AGR-1 Variant 3 Compact Lot LEU01-49T-Z

    SciTech Connect

    Hunn, John D; Montgomery, Fred C; Pappano, Peter J

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 vriant 3 fuel compact lot LEU01-49T-Z. The compacts were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-49T, which was a composite of three batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrocarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 variant 3 coated particle composite LEU01-49t CAN BE FOUND IN ornl/tm-2006/022.

  6. Data Compilation for AGR-1 Baseline Compact Lot LEU01-46T-Z

    SciTech Connect

    Hunn, John D; Montgomery, Fred C; Pappano, Peter J

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 baseline compact lot LEU01-46T-Z. The compacts were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-46T, which was a composite of four batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrocarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 baseline coated particle composite LEU01-46T can be found in ORNL/TM-2006/019. The AGR-1 Fuel product Specification and Characterization Guidance (INL EDF-4380) provides the requirements necessary for acceptance of the fuel manufactured for the AGR-1 irradiation test. Section 6.2 of EDF-4380 provides the property requirements for the heat treated compacts. The Statistical Sampling Plan for AGR Fuel materials (INL EDF-4542) provides additional guidance regarding statistical methods for product acceptance and recommended sample sizes. The procedures for characterizing and qualifying the compacts are outlined in ORNL product inspection plan AGR-CHAR-PIP-05. the inspection report forms generated by this product inspection plan document the product acceptance for the property requirements listed in section 6.2 of EDF-4380.

  7. Data Compilation for AGR-1 Variant 2 Compact Lot LEU01-48T-Z

    SciTech Connect

    Hunn, John D; Montgomery, Fred C; Pappano, Peter J

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 variant 2 compact lot LEU01-48T-Z. The compacts were produced by ORNL for the Advanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-48T, which was a composite of three batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrocarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified as LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 variant 2 coated particle composite LEU01-48T can be found in ORNL/TM-2006/021. The AGR-1 Fuel Product Specification and Characterization Guidance (INL EDF-4380) provides the requirements necessary for acceptance of the fuel manufactured for the AGR-1 irradiation test. Section 6.2 of EDF-4380 provides the property requirements for the heat treated compacts. The Statistical Sampling Plan for AGR Fuel materials (INL EDF-4542) provides additional guidance regarding statistical methods for product acceptance and recommended sample sizes. The procedures for characterizing and qualifying the compacts are outlined in ORNL product inspection plan AGR-CHAR-PIP-05. The inspection report forms generated by this product inspection plan document the product acceptance for the property requirements listed in section 6.2 of EDF-4380.

  8. Data Compilation for AGR-1 Variant 1 Compact Lot LEU01-47T-Z

    SciTech Connect

    Hunn, John D; Montgomery, Fred C; Pappano, Peter J

    2006-08-01

    This document is a compilation of characterization data for the AGR-1 variant 1 compact lot LEU01-47T-Z. The compacts were produced by ORNL for the ADvanced Gas Reactor Fuel Development and Qualification (AGR) program for the first AGR irradiation test train (AGR-1). This compact lot was fabricated using particle composite LEU01-47T, which was a composite of three batches of TRISO-coated 350 {micro}m diameter 19.7% low enrichment uranium oxide/uranium carbide kernels (LEUCO). The AGR-1 TRISO-coated particles consist of a spherical kernel coated with an {approx} 50% dense carbon buffer layer (100 {micro}m nominal thickness), followed by a dense inner pyrocarbon layer (40 {micro}m nominal thickness), followed by a SiC layer (35 {micro}m nominal thickness), followed by another dense outer pyrcoarbon layer (40 {micro}m nominal thickness). The kernels were obtained from BWXT and identified as composite G73D-20-69302. The BWXT kernel lot G73D-20-69302 was riffled into sublots for characterization and coating by ORNL and identified at LEU01-?? (where ?? is a series of integers beginning with 01). A data compilation for the AGR-1 variant 1 coated particle composite LEU01-47T can be found in ORNL/TM-2006/020. The AGR-1 Fuel Product Specification and Characterization Guidance (INL EDF-4380) provides the requirements necessary for acceptance of the fuel manufactured for the AGR-1 irradiation test. Section 6.2 of EDF-4380 provides the property requirements for the heat treated compacts. The Statistical Sampling Plan for AGR Fuel Materials (INL EDF-4542) provides additional guidance regarding statistical methods for product acceptance and recommended sample sizes. The procedures for characterizing and qualifying the compacts are outlined in ORNL product inspection plan AGR-CHAR-PIP-05. The inspection report forms generated by this product inspection plan document the product acceptance for the property requirements listed in section 6.2 of EDF-4380.

  9. The effect of birthrate granularity on the release-to-birth ratio for the AGR-1 in-core experiment

    SciTech Connect

    D. M. Scates; J. B. Walter; J. T. Maki; J. W. Sterbentz; J. R. Parry

    2014-05-01

    The AGR-1 Advanced Gas Reactor (AGR) tristructural-isotropic-particle fuel experiment underwent 13 irradiation intervals from December 2006 until November 2009 within the Idaho National Laboratory Advanced Test Reactor in support of the Next Generation Nuclear Power Plant program. During this multi-year experiment, release-to-birth rate ratios were computed at the end of each operating interval to provide information about fuel performance. Fission products released during irradiation were tracked daily by the Fission Product Monitoring System using 8-h measurements. Birth rate calculated by MCNP with ORIGEN for as-run conditions were computed at the end of each irradiation interval. Each time step in MCNP provided neutron flux, reaction rates and AGR-1 compact composition, which were used to determine birth rate using ORIGEN. The initial birth-rate data, consisting of four values for each irradiation interval at the beginning, end, and two intermediate times, were interpolated to obtain values for each 8-h activity. The problem with this method is that any daily changes in heat rates or perturbations, such as shim control movement or core/lobe power fluctuations, would not be reflected in the interpolated data and a true picture of the system would not be presented. At the conclusion of the AGR-1 experiment, great efforts were put forth to compute daily birthrates, which were reprocessed with the 8-h release activity. The results of this study are presented in this paper.

  10. The Effect of Birthrate Granularity on the Release- to- Birth Ratio for the AGR-1 In-core Experiment

    SciTech Connect

    Dawn Scates; John Walter

    2012-10-01

    The AGR-1 Advanced Gas Reactor (AGR) tristructural-isotropic-particle fuel experiment underwent 13 irradiation intervals from December 2006 until November 2009 within the Idaho National Laboratory Advanced Test Reactor in support of the Next Generation Nuclear Power Plant program. During this multi-year experiment, release-to-birth rate ratios were computed at the end of each operating interval to provide information about fuel performance. Fission products released during irradiation were tracked daily by the Fission Product Monitoring System using 8-hour measurements. Birth rates calculated by MCNP with ORIGEN for as-run conditions were computed at the end of each irradiation interval. Each time step in MCNP provided neutron flux, reaction rates and AGR-1 compact composition, which were used to determine birth rates using ORIGEN. The initial birth-rate data, consisting of four values for each irradiation interval at the beginning, end, and two intermediate times, were interpolated to obtain values for each 8-hour activity. The problem with this method is that any daily changes in heat rates or perturbations, such as shim control movement or core/lobe power fluctuations, would not be reflected in the interpolated data and a true picture of the system would not be presented. At the conclusion of the AGR-1 experiment, great efforts were put forth to compute daily birthrates, which were reprocessed with the 8-hour release activity. The results of this study are presented in this paper.

  11. Fabrication and Comparison of Fuels for Advanced Gas Reactor Irradiation Tests

    SciTech Connect

    Jeffrey Phillips; Charles Barnes; John Hunn

    2010-10-01

    As part of the program to demonstrate TRISO-coated fuel for the Next Generation Nuclear Plant, a series of irradiation tests of Advanced Gas Reactor (AGR) fuel are being performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 until November 2009. Development of AGR-1 fuel sought to replicate the properties of German TRISO-coated particles. No particle failures were seen in the nearly 3-year irradiation to a burn up of 19%. The AGR-1 particles were coated in a two-inch diameter coater. Following fabrication of AGR-1 fuel, process improvements and changes were made in each of the fabrication processes. Changes in the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a six-inch diameter coater using a change size about twenty-one times that of the two-inch diameter coater used to coat AGR-1 particles. Changes were also made in the compacting process, including increasing the temperature and pressure of pressing and using a different type of press. Irradiation of AGR-2 fuel began in late spring 2010. Properties of AGR-2 fuel compare favorably with AGR-1 and historic German fuel. Kernels are more homogeneous in shape, chemistry and density. TRISO-particle sphericity, layer thickness standard deviations, and defect fractions are also comparable. In a sample of 317,000 particles from deconsolidated AGR-2 compacts, 3 exposed kernels were found in a leach test. No SiC defects were found in a sample of 250,000 deconsolidated particles, and no IPyC defects in a sample of 64,000 particles. The primary difference in properties between AGR-1 and AGR-2 compacts is that AGR-2 compacts have a higher matrix density, 1.6 g/cm3 compared to about 1.3 g/cm3 for AGR-1 compacts. Based on

  12. PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2

    SciTech Connect

    Hunn, John D.; Gerczak, Tyler J.; Morris, Robert Noel; Baldwin, Charles A.; Montgomery, Fred C.

    2016-04-01

    Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.

  13. Analysis of Fission Products on the AGR-1 Capsule Components

    SciTech Connect

    Paul A. Demkowicz; Jason M. Harp; Philip L. Winston; Scott A. Ploger

    2013-03-01

    The components of the AGR-1 irradiation capsules were analyzed to determine the retained inventory of fission products in order to determine the extent of in-pile fission product release from the fuel compacts. This includes analysis of (i) the metal capsule components, (ii) the graphite fuel holders, (iii) the graphite spacers, and (iv) the gas exit lines. The fission products most prevalent in the components were Ag-110m, Cs 134, Cs 137, Eu-154, and Sr 90, and the most common location was the metal capsule components and the graphite fuel holders. Gamma scanning of the graphite fuel holders was also performed to determine spatial distribution of Ag-110m and radiocesium. Silver was released from the fuel components in significant fractions. The total Ag-110m inventory found in the capsules ranged from 1.2×10 2 (Capsule 3) to 3.8×10 1 (Capsule 6). Ag-110m was not distributed evenly in the graphite fuel holders, but tended to concentrate at the axial ends of the graphite holders in Capsules 1 and 6 (located at the top and bottom of the test train) and near the axial center in Capsules 2, 3, and 5 (in the center of the test train). The Ag-110m further tended to be concentrated around fuel stacks 1 and 3, the two stacks facing the ATR reactor core and location of higher burnup, neutron fluence, and temperatures compared with Stack 2. Detailed correlation of silver release with fuel type and irradiation temperatures is problematic at the capsule level due to the large range of temperatures experienced by individual fuel compacts in each capsule. A comprehensive Ag 110m mass balance for the capsules was performed using measured inventories of individual compacts and the inventory on the capsule components. For most capsules, the mass balance was within 11% of the predicted inventory. The Ag-110m release from individual compacts often exhibited a very large range within a particular capsule.

  14. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  15. Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

    SciTech Connect

    S. Blaine Grover; David A. Petti

    2014-05-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and sup

  16. AGR-2 Irradiation Test Final As-Run Report

    SciTech Connect

    Collin, Blaise P.

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel.

  17. AGR-2 Irradiation Test Final As-Run Report

    SciTech Connect

    Collin, Blaise P.

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technology Development Office (TDO) program. The objectives of the AGR-2 experiment are to: 1. Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. 2. Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. 3. Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tristructural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S.-produced fuel.

  18. AGR-2 Irradiated Test Train Preliminary Inspection and Disassembly First Look

    SciTech Connect

    Ploger, Scott; Demkowciz, Paul; Harp, Jason

    2015-05-01

    The AGR 2 irradiation experiment began in June 2010 and was completed in October 2013. The test train was shipped to the Materials and Fuels Complex in July 2014 for post-irradiation examination (PIE). The first PIE activities included nondestructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and their graphite fuel holders. Dimensional metrology was then performed on the compacts, graphite holders, and steel capsule shells. AGR 2 disassembly and metrology were performed with the same equipment used successfully on AGR 1 test train components. Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Disassembly of the AGR 2 test train and its capsules was conducted rapidly and efficiently by employing techniques refined during the AGR 1 disassembly campaign. Only one major difficulty was encountered while separating the test train into capsules when thermocouples (of larger diameter than used in AGR 1) and gas lines jammed inside the through tubes of the upper capsules, which required new tooling for extraction. Disassembly of individual capsules was straightforward with only a few minor complications. On the whole, AGR 2 capsule structural components appeared less embrittled than their AGR 1 counterparts. Compacts from AGR 2 Capsules 2, 3, 5, and 6 were in very good condition upon removal. Only relatively minor damage or markings were visible using high resolution photographic inspection. Compact dimensional measurements indicated radial shrinkage between 0.8 to 1.7%, with the greatest shrinkage observed on Capsule 2 compacts that were irradiated at higher temperature. Length shrinkage ranged from 0.1 to 0.9%, with by far the lowest axial shrinkage on Capsule 3 compacts

  19. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    SciTech Connect

    Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; Sterbentz, James W.

    2014-09-03

    AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma spectrometry

  20. An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

    DOE PAGES

    Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; ...

    2014-09-03

    AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methodsmore » for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma

  1. FFTF utilization for irradiation testing

    SciTech Connect

    Corrigan, D.C.; Julyk, L.J.; Hoth, C.W.; McGuire, J.C.; Sloan, W.R.

    1980-01-01

    FFTF utilization for irradiation testing is beginning. Two Fuels Open Test Assemblies and one Vibration Open Test Assembly, both containing in-core contact instrumentation, are installed in the reactor. These assemblies will be used to confirm plant design performance predictions. Some 100 additional experiments are currently planned to follow these three. This will result in an average core loading of about 50 test assemblies throughout the early FFTF operating cycles.

  2. Ultrasonic Transducer Irradiation Test Results

    SciTech Connect

    Daw, Joshua; Palmer, Joe; Ramuhalli, Pradeep; Keller, Paul; Montgomery, Robert; Chien, Hual-Te; Kohse, Gordon; Tittmann, Bernhard; Reinhardt, Brian; Rempe, Joy

    2015-02-01

    Ultrasonic technologies offer the potential for high-accuracy and -resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other ongoing efforts include an ultrasonic technique to detect morphology changes (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. For this reason, the Pennsylvania State University (PSU) was awarded an ATR NSUF project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2. The goal of this research is to characterize and demonstrate magnetostrictive and piezoelectric transducer operation during irradiation, enabling the development of novel radiation-tolerant ultrasonic sensors for use in Material Testing Reactors (MTRs). As such, this test is an instrumented lead test and real-time transducer performance data is collected along with temperature and neutron and gamma flux data. The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers. To date, one piezoelectric

  3. Fuel Accident Condition Simulator (FACS) Furnace for Post-Irradiation Heating Tests of VHTR Fuel Compacts

    SciTech Connect

    Paul A Demkowicz; Paul Demkowicz; David V Laug

    2010-10-01

    Abstract –Fuel irradiation testing and post-irradiation examination are currently in progress as part of the Next Generation Nuclear Plant Fuels Development and Qualification Program. The PIE campaign will include extensive accident testing of irradiated very high temperature reactor fuel compacts to verify fission product retention characteristics at high temperatures. This work will be carried out at both the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory, beginning with accident tests on irradiated fuel from the AGR-1 experiment in 2010. A new furnace system has been designed, built, and tested at INL to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000°C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, Eu, and I) and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator (FACS) furnace system, as well as preliminary system calibration results.

  4. Irradiation Testing of Ultrasonic Transducers

    SciTech Connect

    Daw, Joshua; Tittmann, Bernhard; Reinhardt, Brian; Kohse, Gordon E.; Ramuhalli, Pradeep; Montgomery, Robert O.; Chien, Hual-Te; Villard, Jean-Francois; Palmer, Joe; Rempe, Joy

    2014-07-30

    Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of single, small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other efforts include an ultrasonic technique to detect morphology changes (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of existing knowledge of ultrasonic transducer material survivability under irradiation conditions. For this reason, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate promising magnetostrictive and piezoelectric transducer performance in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2 (E> 0.1 MeV). The goal of this research is to characterize magnetostrictive and piezoelectric transducer survivability during irradiation, enabling the development of novel radiation tolerant ultrasonic sensors for use in Material and Test Reactors (MTRs). As such, this test will be an instrumented lead test and real-time transducer performance data will be collected along with temperature and neutron and gamma flux data. The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers.

  5. Irradiation Testing of Ultrasonic Transducers

    SciTech Connect

    Daw, Joshua; Tittmann, Bernhard; Reinhardt, Brian; Kohse, Gordon E.; Ramuhalli, Pradeep; Montgomery, Robert O.; Chien, Hual-Te; Villard, Jean-Francois; Palmer, Joe; Rempe, Joy

    2013-12-01

    Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of single, small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. Other efforts include an ultrasonic technique to detect morphology changes (such as crack initiation and growth) and acoustic techniques to evaluate fission gas composition and pressure. These efforts are limited by the lack of existing knowledge of ultrasonic transducer material survivability under irradiation conditions. For this reason, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate promising magnetostrictive and piezoelectric transducer performance in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 1021 n/cm2 (E> 0.1 MeV). The goal of this research is to characterize magnetostrictive and piezoelectric transducer survivability during irradiation, enabling the development of novel radiation tolerant ultrasonic sensors for use in Material and Test Reactors (MTRs). As such, this test will be an instrumented lead test and real-time transducer performance data will be collected along with temperature and neutron and gamma flux data. The current work bridges the gap between proven out-of-pile ultrasonic techniques and in-pile deployment of ultrasonic sensors by acquiring the data necessary to demonstrate the performance of ultrasonic transducers.

  6. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  7. HRB-22 irradiation phase test data report

    SciTech Connect

    Montgomery, F.C.; Acharya, R.T.; Baldwin, C.A.; Rittenhouse, P.L.; Thoms, K.R.; Wallace, R.L.

    1995-03-01

    Irradiation capsule HRB-22 was a test capsule containing advanced Japanese fuel for the High Temperature Test Reactor (HTTR). Its function was to obtain fuel performance data at HTTR operating temperatures in an accelerated irradiation environment. The irradiation was performed in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL). The capsule was irradiated for 88.8 effective full power days in position RB-3B of the removable beryllium (RB) facility. The maximum fuel compact temperature was maintained at or below the allowable limit of 1300{degrees}C for a majority of the irradiation. This report presents the data collected during the irradiation test. Included are test thermocouple and gas flow data, the calculated maximum and volume average temperatures based on the measured graphite temperatures, measured gaseous fission product activity in the purge gas, and associated release rate-to-birth rate (R/B) results. Also included are quality assurance data obtained during the test.

  8. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  9. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2009-09-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  10. BPX insulation irradiation program test results

    SciTech Connect

    McManamy, T.J. ); Kanemoto, G. ); Snook, P.G. . Plasma Physics Lab.)

    1991-01-01

    The toroidal field coil insulation for the Burning Plasma Experiment (BPX) is expected to receive a radiation dose of nearly 10{sup 10} rad and to withstand significant mechanical stresses. An irradiation test program was performed at the Idaho National Engineering Laboratory (INEL) using the Advanced Technology Reactor (ATR) for irradiations to doses on the order of 3 {times} 10{sup 10} rad. The flexure and shear strength with compression of commercially procured sheet material were reported earlier. A second series of tests has been performed to slightly higher dose levels with vacuum impregnated materials, glass strand material, and Spaulrad-S sheet samples. Vacuum impregnation with a Shell 9405 resin and 9470 hardener was used to produce bonded copper squares and flexure samples of both pure resin and resin with S-glass. A new test fixture was developed to test the bonded samples in shear without applied compression. The Spaulrad-S flexure samples demonstrated a loss of strength with irradiation, similar to previous results. The pure resin lost nearly all flexibility, while the S-glass-reinforced samples retained between 30% and 40% of the initial flexure strength. The S-glass strands showed a 30% loss of strength at the higher dose level when tested in tension. The bonded copper squares had a low room-temperature shear strength of approximately 17 MPa before irradiation, which was unchanged in the irradiated samples. Shear testing of unirradiated bonded copper squares with ten different types of surface treatment revealed that the low shear strength resulted from the polyurethane primer used. In the later series of test, the epoxy-based primers and DZ-80 from Ciba-Geigy did much better, with shear strengths on the order of 40 MPa. These samples also demonstrated a resistance to cryogenic shock. One irradiated bonded sample was tested up 10 210 MPa in compression, the limit of the test fixture, without failure.

  11. Measurement of Diameter Changes during Irradiation Testing

    SciTech Connect

    Davis, K. L.; Knudson, D. L.; Crepeau, J. C.; Solstad, S.

    2015-03-01

    New materials are being considered for fuel, cladding, and structures in advanced and existing nuclear reactors. Such materials can experience significant dimensional and physical changes during irradiation. Currently in the US, such changes are measured by repeatedly irradiating a specimen for a specified period of time and then removing it from the reactor for evaluation. The time and labor to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data and handling may disturb the phenomena of interest. In-pile detection of changes in geometry is sorely needed to understand real-time behavior during irradiation testing of fuels and materials in high flux US Material and Test Reactors (MTRs). This paper presents development results of an advanced Linear Variable Differential Transformer-based test rig capable of detecting real-time changes in diameter of fuel rods or material samples during irradiation in US MTRs. This test rig is being developed at the Idaho National Laboratory and will provide experimenters with a unique capability to measure diameter changes associated with fuel and cladding swelling, pellet-clad interaction, and crud buildup.

  12. AGR-2 Irradiation Test Final As-Run Report, Rev 2

    SciTech Connect

    Collin, Blaise P.

    2014-08-01

    This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a

  13. Updated Results of Ultrasonic Transducer Irradiation Test

    SciTech Connect

    Daw, Joshua; Palmer, Joe; Ramuhalli, Pradeep; Keller, Paul; Montgomery, Robert; Chien, Hual-Te; Tittmann, Bernhard; Reinhardt, Brian; Kohse, Gordon; Rempe, Joy; Villard, J.F.

    2015-07-01

    Ultrasonic technologies offer the potential for high accuracy and resolution in-pile measurement of a range of parameters, including geometry changes, temperature, crack initiation and growth, gas pressure and composition, and microstructural changes. Many Department of Energy-Office of Nuclear Energy (DOE-NE) programs are exploring the use of ultrasonic technologies to provide enhanced sensors for in-pile instrumentation during irradiation testing. For example, the ability of small diameter ultrasonic thermometers (UTs) to provide a temperature profile in candidate metallic and oxide fuel would provide much needed data for validating new fuel performance models. These efforts are limited by the lack of identified ultrasonic transducer materials capable of long term performance under irradiation test conditions. To address this need, the Pennsylvania State University (PSU) was awarded an Advanced Test Reactor National Scientific User Facility (ATR NSUF) project to evaluate the performance of promising magnetostrictive and piezoelectric transducers in the Massachusetts Institute of Technology Research Reactor (MITR) up to a fast fluence of at least 10{sup 21} n/cm{sup 2}. A multi-National Laboratory collaboration funded by the Nuclear Energy Enabling Technologies Advanced Sensors and Instrumentation (NEET-ASI) program also provided initial support for this effort. This irradiation, which started in February 2014, is an instrumented lead test and real-time transducer performance data are collected along with temperature and neutron and gamma flux data. The irradiation is ongoing and will continue to approximately mid-2015. To date, very encouraging results have been attained as several transducers continue to operate under irradiation. (authors)

  14. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    SciTech Connect

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; Allen, Todd R.

    2016-08-15

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy’s Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representative of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Finally, insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.

  15. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    DOE PAGES

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; ...

    2016-08-15

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy’s Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representativemore » of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Finally, insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.« less

  16. SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

    NASA Astrophysics Data System (ADS)

    Gerczak, Tyler J.; Hunn, John D.; Lowden, Richard A.; Allen, Todd R.

    2016-11-01

    Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy's Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representative of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.

  17. Irradiation Environment of the Materials Test Station

    SciTech Connect

    Pitcher, Eric John

    2012-06-21

    Conceptual design of the proposed Materials Test Station (MTS) at the Los Alamos Neutron Science Center (LANSCE) is now complete. The principal mission is the irradiation testing of advanced fuels and materials for fast-spectrum nuclear reactor applications. The neutron spectrum in the fuel irradiation region of MTS is sufficiently close to that of fast reactor that MTS can match the fast reactor fuel centerline temperature and temperature profile across a fuel pellet. This is an important characteristic since temperature and temperature gradients drive many phenomena related to fuel performance, such as phase stability, stoichiometry, and fission product transport. The MTS irradiation environment is also suitable in many respects for fusion materials testing. In particular, the rate of helium production relative to atomic displacements at the peak flux position in MTS matches well that of fusion reactor first wall. Nuclear transmutation of the elemental composition of the fusion alloy EUROFER97 in MTS is similar to that expected in the first wall of a fusion reactor.

  18. LWRS ATR Irradiation Testing Readiness Status

    SciTech Connect

    Kristine Barrett

    2012-09-01

    The Light Water Reactor Sustainability (LWRS) Program was established by the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors. The LWRS Program is divided into four R&D Pathways: (1) Materials Aging and Degradation; (2) Advanced Light Water Reactor Nuclear Fuels; (3) Advanced Instrumentation, Information and Control Systems; and (4) Risk-Informed Safety Margin Characterization. This report describes an irradiation testing readiness analysis in preparation of LWRS experiments for irradiation testing at the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) under Pathway (2). The focus of the Advanced LWR Nuclear Fuels Pathway is to improve the scientific knowledge basis for understanding and predicting fundamental performance of advanced nuclear fuel and cladding in nuclear power plants during both nominal and off-nominal conditions. This information will be applied in the design and development of high-performance, high burn-up fuels with improved safety, cladding integrity, and improved nuclear fuel cycle economics

  19. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. B. Grover

    2007-05-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing.1,2 The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment

  20. Instrumentation to Enhance Advanced Test Reactor Irradiations

    SciTech Connect

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  1. Thermal analysis of the FSP-1RR irradiation test

    SciTech Connect

    Webb, R.H.; Lyon, W.F. III

    1992-10-14

    The thermal analysis of four unirradiated fuel pins to be tested in the FSP-1RR fuels irradiation experiment was completed. This test is a follow-on experiment in the series of fuel pin irradiation tests conducted by the SP-100 Program in the Fast Flux Test Facility. One of the pins contains several meltwire temperature monitors within the fuel and the Li annulus. A post-irradiation examination will verify the accuracy of the pre-irradiation thermal analysis. The purpose of the pre-irradiation analysis was to determine the appropriate insulating gap gas compositions required to provide the design goal cladding operating temperatures and to ensure that the meltwire temperature ranges in the temperature monitored pin bracket peak irradiation temperatures. This paper discusses the methodology and summarizes the results of the analysis.

  2. Thermal analysis of the FSP-1RR irradiation test

    SciTech Connect

    Webb, R.H.; Lyon, W.F. III )

    1993-01-10

    The thermal analysis of four unirradiated fuel pins to be tested in the FSP-1RR fuels irradiation experiment was completed. This test is a follow-on experiment in the series of fuel pin irradiation tests conducted by the SP-100 Program in the Fast Flux Test Facility. One of the pins contains several meltwire temperature monitors within the fuel and the Li annulus. A post-irradiation examination will verify the accuracy of the pre-irradiation thermal analysis. The purpose of the pre-irradiation analysis was to determine the appropriate insulating gap gas compositions required to provide the design goal cladding operating temperatures and to ensure that the meltwire temperature ranges in the temperature monitored pin bracket peak irradiation temperatures. This paper discusses the methodology and summarizes the results of the analysis.

  3. Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

    DOE PAGES

    Hunn, John D.; Baldwin, Charles A.; Gerczak, Tyler J.; ...

    2016-04-06

    As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of 134Cs and 137Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compactmore » containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800 °C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. In addition, all three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking.« less

  4. Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

    SciTech Connect

    Hunn, John D.; Baldwin, Charles A.; Gerczak, Tyler J.; Montgomery, Fred C.; Morris, Robert N.; Silva, Chinthaka M.; Demkowicz, Paul A.; Harp, Jason M.; Ploger, Scott A.

    2016-04-06

    As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of 134Cs and 137Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800 °C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. In addition, all three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking.

  5. Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

    SciTech Connect

    Tsai, H.; Gomes, I.C.; Smith, D.L.; Palmer, A.J.; Ingram, F.W.; Wiffen, F.W.

    1998-09-01

    The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

  6. Discuss the testing problems of ultraviolet irradiance meters

    NASA Astrophysics Data System (ADS)

    Ye, Jun'an; Lin, Fangsheng

    2014-09-01

    Ultraviolet irradiance meters are widely used in many areas such as medical treatment, epidemic prevention, energy conservation and environment protection, computers, manufacture, electronics, ageing of material and photo-electric effect, for testing ultraviolet irradiance intensity. So the accuracy of value directly affects the sterile control in hospital, treatment, the prevention level of CDC and the control accuracy of curing and aging in manufacturing industry etc. Because the display of ultraviolet irradiance meters is easy to change, in order to ensure the accuracy, it needs to be recalibrated after being used period of time. By the comparison with the standard ultraviolet irradiance meters, which are traceable to national benchmarks, we can acquire the correction factor to ensure that the instruments working under accurate status and giving the accurate measured data. This leads to an important question: what kind of testing device is more accurate and reliable? This article introduces the testing method and problems of the current testing device for ultraviolet irradiance meters. In order to solve these problems, we have developed a new three-dimensional automatic testing device. We introduce structure and working principle of this system and compare the advantages and disadvantages of two devices. In addition, we analyses the errors in the testing of ultraviolet irradiance meters.

  7. Irradiation testing of high density uranium alloy dispersion fuels

    SciTech Connect

    Hayes, S.L.; Trybus, C.L.; Meyer, M.K.

    1997-10-01

    Two irradiation test vehicles have been designed, fabricated, and inserted into the Advanced Test Reactor in Idaho. Irradiation of these experiments began in August 1997. These irradiation tests were designed to obtain irradiation performance information on a variety of potential new, high-density dispersion fuels. Each of the two irradiation vehicles contains 32 microplates. Each microplate is aluminum clad, having an aluminum matrix phase and containing one of the following compositions as the fuel phase: U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru, U-10Mo-0.05Sn, U{sub 2}Mo, or U{sub 3}Si{sub 2}. These experiments will be discharged at peak fuel burnups of 40% and 80%. Of particular interest is the fission gas retention/swelling characteristics of these new fuel alloys. This paper presents the design of the irradiation vehicles and the irradiation conditions.

  8. Daily Thermal Predictions of the AGR-1 Experiment with Gas Gaps Varying with Time

    SciTech Connect

    Grant Hawkes; James Sterbentz; John Maki; Binh Pham

    2012-06-01

    A new daily as-run thermal analysis was performed at the Idaho National Laboratory on the Advanced Gas Reactor (AGR) test experiment number one at the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment. The purpose of this analysis was to calculate the daily average temperatures of each compact to compare with experimental results. Post irradiation examination (PIE) measurements of the graphite holder and fuel compacts showed the gas gaps varying from the beginning of life. The control temperature gas gap and the fuel compact – graphite holder gas gaps were linearly changed from the original fabrication dimensions, to the end of irradiation measurements. A steady-state thermal analysis was performed for each daily calculation. These new thermal predictions more closely match the experimental data taken during the experiment than previous analyses. Results are presented comparing normalized compact average temperatures to normalized log(R/B) Kr-85m. The R/B term is the measured release rate divided by the predicted birth rate for the isotope Kr-85m. Correlations between these two normalized values are presented.

  9. Daily thermal predictions of the AGR-1 experiment with gas gaps varying with time

    SciTech Connect

    Hawkes, G.; Sterbentz, J.; Maki, J.; Pham, B.

    2012-07-01

    A new daily as-run thermal analysis was performed at the Idaho National Laboratory on the Advanced Gas Reactor (AGR) test experiment number one at the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment. The purpose of this analysis was to calculate the daily average temperatures of each compact to compare with experimental results. Post irradiation examination (PIE) measurements of the graphite holder and fuel compacts showed the gas gaps changed from the beginning of life. The control temperature gas gap and the fuel compact - graphite holder gas gaps were modeled with a linear change from the original fabrication gap dimensions to the end of irradiation measurements. A steady-state thermal analysis was performed for each daily calculation with the commercial finite element heat transfer code ABAQUS. These new thermal predictions more closely match the experimental data taken during the experiment than previous analyses. Results are presented comparing normalized compact average temperatures to normalized log(R/B) Kr-85m. The R/B term is the measured release rate divided by the predicted birth rate for the isotope Kr-85m. Correlations between these two normalized values are presented. (authors)

  10. Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

    SciTech Connect

    Daniel M. Wachs; Richard G. Ambrosek; Gray Chang; Mitchell K. Meyer

    2006-10-01

    Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progress toward element testing will be reviewed.

  11. USE OF SILICON CARBIDE MONITORS IN ATR IRRADIATION TESTING

    SciTech Connect

    K. L. Davis; B. Chase; T. Unruh; D. Knudson; J. L. Rempe

    2012-07-01

    In April 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) a National Scientific User Facility (NSUF) to advance US leadership in nuclear science and technology. By attracting new users from universities, laboratories, and industry, the ATR will support basic and applied nuclear research and development and help address the nation's energy security needs. In support of this new program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced temperature sensors for irradiation testing. Although most efforts emphasize sensors capable of providing real-time data, selected tasks have been completed to enhance sensors provided in irradiation locations where instrumentation leads cannot be included, such as drop-in capsule and Hydraulic Shuttle Irradiation System (HSIS) or 'rabbit' locations. For example, silicon carbide (SiC) monitors are now available to detect peak irradiation temperatures between 200°C and 800°C. Using a resistance measurement approach, specialized equipment installed at INL's High Temperature Test Laboratory (HTTL) and specialized procedures were developed to ensure that accurate peak irradiation temperature measurements are inferred from SiC monitors irradiated at the ATR. Comparison examinations were completed by INL to demonstrate this capability, and several programs currently rely on SiC monitors for peak temperature detection. This paper discusses the use of SiC monitors at the ATR, the process used to evaluate them at the HTTL, and presents representative measurements taken using SiC monitors.

  12. X-Ray-Diffraction Tests Of Irradiated Electronic Devices: I

    NASA Technical Reports Server (NTRS)

    Shaw, David C.; Lowry, Lynn E.; Barnes, Charles E.

    1993-01-01

    X-ray-diffraction tests performed on aluminum conductors in commercial HI1-507A complementary metal oxide/semiconductor (CMOS) integrated-circuit analog multiplexers, both before and after circuits exposed to ionizing radiation from Co(60) source, and after postirradiation annealing at ambient and elevated temperatures. Tests in addition to electrical tests performed to determine effects of irradiation and of postirradiation annealing on electrical operating characteristics of circuits. Investigators sought to determine whether relationship between effects of irradiation on devices and physical stresses within devices. X-ray diffraction potentially useful for nondestructive measurement of stresses.

  13. Engineering test plan for US/UK higher actinides irradiations tests

    SciTech Connect

    Basmajian, J A

    1981-03-01

    The objective of the Higher Actinides Irradiations Program is to verify the neutronic and irradiation performance of americium and curium oxides in a fast reactor. The data obtained from the irradiation will be used to assess the basic neutronics parameters for actinide elements and determine the irradiation potential of the oxides of {sup 241}Am and {sup 244}Cm. This information has application in breeder reactor physics, fuel cycle analysis and assessment of waste management options. The irradiation test program is a cooperative effort wherein the US is supplying the completed irradiation test pins, while the UK will perform the irradiation in their Prototype Fast Reactor (PFR). Postirradiation examination and data analyses will be conducted on a cooperative basis, with some examinations performed in the UK and others in the US. 5 figs., 5 tabs.

  14. Irradiation Facilities at the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-12-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC – formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world’s data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities1. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens.

  15. FFTF (Fast Flux Test Facility) as an irradiation test bed for fusion materials and components

    SciTech Connect

    Greenslade, D.L.; Puigh, R.J.; Hollenberg, G.W.; Grover, J.M.

    1986-03-01

    The relatively large irradiation volume, instrumentation capabilities, and fast neutron flux associated with the Fast Flux Test Facility (FFTF) make this reactor an ideal test bed for fusion materials and components irradiations. Significant fusion materials irradiations are presently being performed in the Materials Open Test Assembly (MOTA) in FFTF. The MOTA is providing a controlled temperature and high neutron flux environment for such materials as the low activation alloys, copper alloys, ceramic insulators, and high heat flux materials. Conceptual designs utilizing the versatile MOTA irradiation vehicle have been developed to investigate irradiation effects on the mechanical and tritium breeding behaviors of solid breeder materials. More aggressive conceptual designs have also been developed to irradiate solid breeder blanket submodules in the FFTF. These specific component test designs will be presented and their potential roles in the development of fusion technology discussed.

  16. Meso-scale modeling of irradiated concrete in test reactor

    SciTech Connect

    Giorla, Alain B.; Vaitová, M.; Le Pape, Yann; Štemberk, P.

    2015-10-18

    In this paper, we detail a numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale. Irradiation experiments in test reactor (Elleuch et al.,1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al.,2015). In conclusion, the proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  17. Meso-scale modeling of irradiated concrete in test reactor

    DOE PAGES

    Giorla, Alain B.; Vaitová, M.; Le Pape, Yann; ...

    2015-10-18

    In this paper, we detail a numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale. Irradiation experiments in test reactor (Elleuch et al.,1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damagemore » around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al.,2015). In conclusion, the proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.« less

  18. TEST RESULTS FROM GAMMA IRRADIATION OF ALUMINUM OXYHYDROXIDES

    SciTech Connect

    Fisher, D.; Westbrook, M.; Sindelar, R.

    2012-02-01

    Hydrated metal oxides or oxyhydroxides boehmite and gibbsite that can form on spent aluminum-clad nuclear fuel assemblies during in-core and post-discharge wet storage were exposed as granular powders to gamma irradiation in a {sup 60}Co irradiator in closed laboratory test vessels with air and with argon as separate cover gases. The results show that boehmite readily evolves hydrogen with exposure up to a dose of 1.8 x 10{sup 8} rad, the maximum tested, in both a full-dried and moist condition of the powder, whereas only a very small measurable quantity of hydrogen was generated from the granular powder of gibbsite. Specific information on the test setup, sample characteristics, sample preparation, irradiation, and gas analysis are described.

  19. In situ nano-compression testing of irradiated copper

    PubMed Central

    Kiener, D.; Hosemann, P.; Maloy, S. A.; Minor, A. M.

    2011-01-01

    Increasing demand for energy and reduction of CO2 emissions has revived interest in nuclear energy. Designing materials for radiation environments necessitates fundamental understanding of how radiation-induced defects alter mechanical properties. Ion beams create radiation damage efficiently without material activation, but their limited penetration depth requires small-scale testing. However, strength measurements of nano-scale irradiated specimens have not been previously performed. Here we show that yield strengths approaching macroscopic values are measured from irradiated ~400 nm diameter copper specimens. Quantitative in situ nano-compression testing in a transmission electron microscope reveals that the strength of larger samples is controlled by dislocation-irradiation defect interactions, yielding size-independent strengths. Below ~400 nm, size-dependent strength results from dislocation source limitation. This transition length-scale should be universal, but depend on material and irradiation conditions. We conclude that for irradiated copper, and presumably related materials, nano-scale in situ testing can determine bulk-like yield strengths and simultaneously identify deformation mechanisms. PMID:21706011

  20. Temperature controlled material irradiation in the advanced test reactor

    NASA Astrophysics Data System (ADS)

    Ingram, F. W.; Palmer, A. J.; Stites, D. J.

    1998-10-01

    The United States Department of Energy (US DOE) has initiated the development of an Irradiation Test Vehicle (ITV) for fusion materials irradiation at the Advanced Test Reactor (ATR) in Idaho Falls, Idaho, USA. The ITV is capable of providing neutron spectral tailoring and individual temperature control for up to 15 experiment capsules simultaneously. The test vehicle consists of three In-Pile Tubes (IPTs) running the length of the reactor vessel. These IPTs are kept dry and test trains with integral instrumentation are inserted and removed through a transfer shield plate above the reactor vessel head. The test vehicle is designed to irradiate specimens as large as 2.2 cm in diameter, at temperatures of 250-800°C, achieving neutron damage rates as high as 10 displacements per atom per year. The high fast to thermal neutron flux ratio required for fusion materials testing is accomplished by using an aluminum filler to displace as much water as possible from the flux trap and surrounding the filler piece with a ring of replaceable neutron absorbing material. The gas blend temperature control system remains in place from test to test, thus hardware costs for new tests are limited to the experiment capsule train and integral instrumentation.

  1. Production of LEU Fully Ceramic Microencapsulated Fuel for Irradiation Testing

    SciTech Connect

    Terrani, Kurt A; Kiggans Jr, James O; McMurray, Jake W; Jolly, Brian C; Hunt, Rodney Dale; Trammell, Michael P; Snead, Lance Lewis

    2016-01-01

    Fully Ceramic Microencapsulated (FCM) fuel consists of tristructural isotropic (TRISO) fuel particles embedded inside a SiC matrix. This fuel inherently possesses multiple barriers to fission product release, namely the various coating layers in the TRISO fuel particle as well as the dense SiC matrix that hosts these particles. This coupled with the excellent oxidation resistance of the SiC matrix and the SiC coating layer in the TRISO particle designate this concept as an accident tolerant fuel (ATF). The FCM fuel takes advantage of uranium nitride kernels instead of oxide or oxide-carbide kernels used in high temperature gas reactors to enhance heavy metal loading in the highly moderated LWRs. Production of these kernels with appropriate density, coating layer development to produce UN TRISO particles, and consolidation of these particles inside a SiC matrix have been codified thanks to significant R&D supported by US DOE Fuel Cycle R&D program. Also, surrogate FCM pellets (pellets with zirconia instead of uranium-bearing kernels) have been neutron irradiated and the stability of the matrix and coating layer under LWR irradiation conditions have been established. Currently the focus is on production of LEU (7.3% U-235 enrichment) FCM pellets to be utilized for irradiation testing. The irradiation is planned at INL s Advanced Test Reactor (ATR). This is a critical step in development of this fuel concept to establish the ability of this fuel to retain fission products under prototypical irradiation conditions.

  2. X-Ray-Diffraction Tests Of Irradiated Electronic Devices: II

    NASA Technical Reports Server (NTRS)

    Shaw, David C.; Lowry, Lynn E.; Barnes, Charles E.

    1993-01-01

    Report describes research on use of x-ray diffraction to measure stresses in metal conductors of complementary metal oxide/semiconductor (CMOS) integrated circuits exposed to ionizing radiation. Expanding upon report summarized in "X-Ray-Diffraction Tests Of Irradiated Electronic Devices: I" (NPO-18803), presenting data further suggesting relationship between electrical performances of circuits and stresses and strains in metal conductors.

  3. The materials test station: a fast spectrum irradiation facility

    SciTech Connect

    Pitcher, Eric J.

    2007-07-01

    The Materials Test Station is a fast-neutron spectrum irradiation facility under design at the Los Alamos National Laboratory in support of the United States Department of Energy's Global Nuclear Energy Partnership. The facility will be capable of rodlets-scale irradiations of candidate fuel forms being developed to power the next generation of fast reactors. Driven by a powerful proton beam, the fuel irradiation region exhibits a neutron spectrum similar to that seen in a fast reactor, with a peak neutron flux of 1.6 x 10{sup 15} n.cm{sup -2}.s{sup -1}. Site preparation and construction are estimated to take four years, with a cost range of $60 M to $90 M. (author)

  4. Evaluation of irradiated fuel during RIA simulation tests. Final report

    SciTech Connect

    Montgomery, R.O.; Rashid, Y.R.

    1996-08-01

    A critical assessment of the RIA-simulation experiments performed to date on previously irradiated test rods is presented. Included in this assessment are the SPERT-CDC, the NSRR, and the CABRI REP Na experimental programs. Information was collected describing the base irradiation, test rod characterization, and test procedures and conditions. The representativeness of the test rods and test conditions to anticipated LWR RIA accident conditions was evaluated using analysis results from fuel behavior and three-dimensional spatial kinetics simulations. It was shown that the pulse characteristics and coolant conditions are significantly different from those anticipated in an LWR-Furthermore, the unrepresentative test conditions were found to exaggerate the mechanisms that caused cladding failure. The data review identified several test rods which contained unusual cladding damage incurred prior to the RIA-simulation test that produced the observed failures. The mechanisms responsible for the observed test rod failures have been shown to result from processes that have a second order effect of burnup. A correlation with burnup could not be appropriately established for the fuel enthalpy at failure. However, the successful test rods can be used to construct a conservative region of success for fuel rod behavior during an RIA event.

  5. Design of a Compact Fatigue Tester for Testing Irradiated Materials

    SciTech Connect

    Hartsell, Brian; Campbell, Michael; Fitton, Michael; Hurh, Patrick; Ishida, Taku; Nakadaira, Takeshi

    2015-06-01

    A compact fatigue testing machine that can be easily inserted into a hot cell for characterization of irradiated materials is beneficial to help determine relative fatigue performance differences between new and irradiated material. Hot cell use has been carefully considered by limiting the size and weight of the machine, simplifying sample loading and test setup for operation via master-slave manipulator, and utilizing an efficient design to minimize maintenance. Funded from a US-Japan collaborative effort, the machine has been specifically designed to help characterize titanium material specimens. These specimens are flat cantilevered beams for initial studies, possibly utilizing samples irradiated at other sources of beam. The option to test spherically shaped samples cut from the T2K vacuum window is also available. The machine is able to test a sample to $10^7$ cycles in under a week, with options to count cycles and sense material failure. The design of this machine will be presented along with current status.

  6. AGR-1, AGR-2 and AGR-3/4 Dimensional Change Data Analysis

    SciTech Connect

    Herberger, Sarah E.

    2016-02-01

    A series of Advanced Gas Reactor (AGR) experiments have been completed in the Advanced Test Reactor at Idaho National Laboratory in support of qualification and development of tristructural isotropic fuel. Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder. These capsules are instrumented with thermocouples embedded in the graphite, enabling temperature control. The fuel compacts are composed of fuel particles surrounded by a graphitic A3 matrix material. Dimensional change in AGR fuel compacts is vital because the swelling or shrinkage affects the size of the gas gaps that are used to control temperatures. Analysis of dimensional change in the AGR fuel compacts is needed to establish the variables directly relating to compact shrinkage. The variables initially identified for consideration were matrix density, compact density, fuel packing fraction, uranium loading, fuel particle diameter, cumulative fast neutron fluence, and volume average time average fuel temperature. In addition to the data available from the AGR experiments, the analysis included specimens formed from the same A3 matrix material used in Advanced Graphite Creep (AGC) experiments, which provide graphite creep data during irradiation for design and licensing purposes. The primary purpose of including the AGC specimens was to encompass dimensional behavior at zero packing fraction, zero uranium loading, and zero particle diameter. All possible combinations of first-order variable regressions were considered in the analysis. The study focused on identifying the best regression models for percent change in diameter, length, and volume. Bootstrap analysis was used to ensure the resulting regression models were robust and well-performing. The variables identified as very significant in predicting change in one or more dimensions (diameter, length, and volume) are volume average time average temperature, fast fluence

  7. Test of radiation hardness of CMOS transistors under neutron irradiation

    SciTech Connect

    Sadrozinski, H.F.W.; Rowe, W.A.; Seiden, A.; Spencer, E.; Hoffman, C.M.; Holtkamp, D.; Kinnison, W.W.; Sommer, W.F. Jr.; Ziock, H.J.

    1989-01-01

    We have tested 2 micron CMOS test structures from various foundries in the LAMPF Beam stop for radiation damage under prolongued neutron irradiation. The fluxes employed covered the region expected to be encountered at the SSC and led to fluences of up to 10/sup 14/ neutrons/cm/sup 2/ in about 500 hrs of running. We show that test structures which have been measured to survive ionizing radiation of the order MRad also survive these high neutron fluences. 5 refs., 4 figs.

  8. Photopatch and UV-irradiated patch testing in photosensitive dermatitis

    PubMed Central

    Rai, Reena; Thomas, Maria

    2016-01-01

    Background: The photopatch test is used to detect photoallergic reactions to various antigens such as sunscreens and drugs. Photosensitive dermatitis can be caused due to antigens like parthenium, fragrances, rubbers and metals. The photopatch test does not contain these antigens. Therefore, the Indian Standard Series (ISS) along with the Standard photopatch series from Chemotechnique Diagnostics, Sweden was used to detect light induced antigens. Aim: To detect light induced antigens in patients with photosensitive dermatitis. Methods: This study was done in a descriptive, observer blinded manner. Photopatch test and ISS were applied in duplicate on the patient's back by the standard method. After 24 hours, readings were recorded according to ICDRG criteria. One side was closed and other side irradiated with 14 J/cm2 of UVA and a second set of readings were recorded after 48 hrs. Result: The highest positivity was obtained with parthenium, with 18 out of 35 (51%) patients showing a positive patch test reaction with both photoallergic contact dermatitis and photoaggravation. Four patients (11%) showed positive patch test reaction suggestive of contact dermatitis to potassium dichromate and fragrance mix. Six patients had contact dermatitis to numerous antigens such as nickel, cobalt, chinoform and para-phenylenediamine. None of these patients showed photoaggravation on patch testing. Conclusion: Parthenium was found to cause photoallergy, contact dermatitis with photoaggravation and contact allergy. Hence, photopatch test and UV irradiated patch test can be an important tool to detect light induced antigens in patients with photosensitive dermatitis. PMID:26955581

  9. AGR 3/4 Irradiation Test Final As Run Report

    SciTech Connect

    Collin, Blaise P.

    2015-06-01

    Several fuel and material irradiation experiments have been planned for the Idaho National Laboratory Advanced Reactor Technologies Technology Development Office Advanced Gas Reactor Fuel Development and Qualification Program (referred to as the INL ART TDO/AGR fuel program hereafter), which supports the development and qualification of tristructural-isotropic (TRISO) coated particle fuel for use in HTGRs. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development and validation of fuel performance and fission product transport models and codes, and provide irradiated fuel and materials for post irradiation examination and safety testing (INL 05/2015). AGR-3/4 combined the third and fourth in this series of planned experiments to test TRISO coated low enriched uranium (LEU) oxycarbide fuel. This combined experiment was intended to support the refinement of fission product transport models and to assess the effects of sweep gas impurities on fuel performance and fission product transport by irradiating designed-to-fail fuel particles and by measuring subsequent fission metal transport in fuel-compact matrix material and fuel-element graphite. The AGR 3/4 fuel test was successful in irradiating the fuel compacts to the burnup and fast fluence target ranges, considering the experiment was terminated short of its initial 400 EFPD target (Collin 2015). Out of the 48 AGR-3/4 compacts, 42 achieved the specified burnup of at least 6% fissions per initial heavy-metal atom (FIMA). Three capsules had a maximum fuel compact average burnup < 10% FIMA, one more than originally specified, and the maximum fuel compact average burnup was <19% FIMA for the remaining capsules, as specified. Fast neutron fluence fell in the expected range of 1.0 to 5.5×1025 n/m2 (E >0.18 MeV) for all compacts. In addition, the AGR-3/4 experiment was globally successful in keeping the

  10. Updated FY12 Ceramic Fuels Irradiation Test Plan

    SciTech Connect

    Nelson, Andrew T.

    2012-05-24

    The Fuel Cycle Research and Development program is currently devoting resources to study of numerous fuel types with the aim of furthering understanding applicable to a range of reactors and fuel cycles. In FY11, effort within the ceramic fuels campaign focused on planning and preparation for a series of rabbit irradiations to be conducted at the High Flux Isotope Reactor located at Oak Ridge National Laboratory. The emphasis of these planned tests was to study the evolution of thermal conductivity in uranium dioxide and derivative compositions as a function of damage induced by neutron damage. Current fiscal realities have resulted in a scenario where completion of the planned rabbit irradiations is unlikely. Possibilities for execution of irradiation testing within the ceramic fuels campaign in the next several years will thus likely be restricted to avenues where strong synergies exist both within and outside the Fuel Cycle Research and Development program. Opportunities to augment the interests and needs of modeling, advanced characterization, and other campaigns present the most likely avenues for further work. These possibilities will be pursued with the hope of securing future funding. Utilization of synthetic microstructures prepared to better understand the most relevant actors encountered during irradiation of ceramic fuels thus represents the ceramic fuel campaign's most efficient means to enhance understanding of fuel response to burnup. This approach offers many of the favorable attributes embraced by the Separate Effects Testing paradigm, namely production of samples suitable to study specific, isolated phenomena. The recent success of xenon-imbedded thick films is representative of this approach. In the coming years, this strategy will be expanded to address a wider range of problems in conjunction with use of national user facilities novel characterization techniques to best utilize programmatic resources to support a science-based research program.

  11. Characterization of nuclear transmutations in materials irradiated test facilities

    SciTech Connect

    Gomes, I.C.; Smith, D.L.

    1994-05-01

    This study presents a comparison of nuclear transmutation rates for candidate fusion first wall/blanket structural materials in available, fission test reactors with those produced in a typical fusion spectrum. The materials analyzed in this study include a vanadium alloy (V-4Cr-4Ti), a reduced activation martensitic steel (Fe-9Cr-2WVTa), a high conductivity copper alloy (Cu-Cr-Zr), and the SiC compound. The fission irradiation facilities considered include the EBR-II fast reactor, and two high flux mixed spectrum reactors, HFIR (High Flux Irradiation Reactor) and SM-3 (Russian reactor). The transmutation and dpa rates that occur in these test reactors are compared with the calculated transmutation and dpa rates characteristic of a D-T fusion first wall spectrum. In general, past work has shown that the displacement damage produced in these fission reactors can be correlated to displacement damage in a fusion spectrum; however, the generation of helium and hydrogen through threshold reactions [(n,x,{alpha}) and (n,xp)] are much higher in a fusion spectrum. As shown in this study, the compositional changes for several candidate structural materials exposed to a fast fission reactor spectrum are very low, similar to those for a characteristic fusion spectrum. However, the relatively high thermalized spectrum of a mixed spectrum reactor produces transmutation rates quite different from the ones predicted for a fusion reactor, resulting in substantial differences in the final composition of several candidate alloys after relatively short irradiation time.

  12. Temperature controlled material irradiation in the advanced test reactor

    SciTech Connect

    Furstenau, R.V.; Ingrahm, F.W.

    1995-12-31

    The Advanced Test Reactor (ATR) is located at the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho, USA and is owned and regulated by the U.S. Department of Energy (US DOE). The ATR is operated for the US DOE by Lockheed Martin Idaho Technologies. In recent years, prime irradiation space in the ATR has been made available for use by customers having irradiation service needs in addition to the reactor`s principal user, the U.S. Naval Nuclear Propulsion Program. To enhance the reactor`s capabilities, the US DOE has initiated the development of an Irradiation Test Vehicle (ITV) capable of providing neutron spectral tailoring and temperature control for up to 28 experiments. The ATR-ITV will have the flexibility to simultaneously support a variety of experiments requiring fast, thermal or mixed spectrum neutron environments. Temperature control is accomplished by varying the thermal conductivity across a gas gap established between the experiment specimen capsule wall and the experiment `in-pile tube (IPT)` inside diameter. Thermal conductivity is adjusted by alternating the control gas mixture ratio of two gases with different thermal conductivities.

  13. Complex regulation of Arabidopsis AGR1/PIN2-mediated root gravitropic response and basipetal auxin transport by cantharidin-sensitive protein phosphatases

    NASA Technical Reports Server (NTRS)

    Shin, Heungsop; Shin, Hwa-Soo; Guo, Zibiao; Blancaflor, Elison B.; Masson, Patrick H.; Chen, Rujin

    2005-01-01

    Polar auxin transport, mediated by two distinct plasma membrane-localized auxin influx and efflux carrier proteins/complexes, plays an important role in many plant growth and developmental processes including tropic responses to gravity and light, development of lateral roots and patterning in embryogenesis. We have previously shown that the Arabidopsis AGRAVITROPIC 1/PIN2 gene encodes an auxin efflux component regulating root gravitropism and basipetal auxin transport. However, the regulatory mechanism underlying the function of AGR1/PIN2 is largely unknown. Recently, protein phosphorylation and dephosphorylation mediated by protein kinases and phosphatases, respectively, have been implicated in regulating polar auxin transport and root gravitropism. Here, we examined the effects of chemical inhibitors of protein phosphatases on root gravitropism and basipetal auxin transport, as well as the expression pattern of AGR1/PIN2 gene and the localization of AGR1/PIN2 protein. We also examined the effects of inhibitors of vesicle trafficking and protein kinases. Our data suggest that protein phosphatases, sensitive to cantharidin and okadaic acid, are likely involved in regulating AGR1/PIN2-mediated root basipetal auxin transport and gravitropism, as well as auxin response in the root central elongation zone (CEZ). BFA-sensitive vesicle trafficking may be required for the cycling of AGR1/PIN2 between plasma membrane and the BFA compartment, but not for the AGR1/PIN2-mediated root basipetal auxin transport and auxin response in CEZ cells.

  14. Design considerations of the irradiation test vehicle for the advanced test reactor

    SciTech Connect

    Tsai, H.; Gomes, I.C.; Smith, D.L.

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  15. Transcriptome profiling of mice testes following low dose irradiation

    PubMed Central

    2013-01-01

    Background Radiotherapy is used routinely to treat testicular cancer. Testicular cells vary in radio-sensitivity and the aim of this study was to investigate cellular and molecular changes caused by low dose irradiation of mice testis and to identify transcripts from different cell types in the adult testis. Methods Transcriptome profiling was performed on total RNA from testes sampled at various time points (n = 17) after 1 Gy of irradiation. Transcripts displaying large overall expression changes during the time series, but small expression changes between neighbouring time points were selected for further analysis. These transcripts were separated into clusters and their cellular origin was determined. Immunohistochemistry and in silico quantification was further used to study cellular changes post-irradiation (pi). Results We identified a subset of transcripts (n = 988) where changes in expression pi can be explained by changes in cellularity. We separated the transcripts into five unique clusters that we associated with spermatogonia, spermatocytes, early spermatids, late spermatids and somatic cells, respectively. Transcripts in the somatic cell cluster showed large changes in expression pi, mainly caused by changes in cellularity. Further investigations revealed that the low dose irradiation seemed to cause Leydig cell hyperplasia, which contributed to the detected expression changes in the somatic cell cluster. Conclusions The five clusters represent gene expression in distinct cell types of the adult testis. We observed large expression changes in the somatic cell profile, which mainly could be attributed to changes in cellularity, but hyperplasia of Leydig cells may also play a role. We speculate that the possible hyperplasia may be caused by lower testosterone production and inadequate inhibin signalling due to missing germ cells. PMID:23714422

  16. Safety Testing of AGR-2 UCO Compacts 5-2-2, 2-2-2, and 5-4-1

    SciTech Connect

    Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.; Montgomery, Fred C.

    2016-08-01

    Post-irradiation examination (PIE) is being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2). This effort builds upon the understanding acquired throughout the AGR-1 PIE campaign, and is establishing a database for the different AGR-2 fuel designs. The AGR-2 irradiation experiment included TRISO fuel particles coated at BWX Technologies (BWXT) with a 150-mm-diameter engineering-scale coater. Two coating batches were tested in the AGR-2 irradiation experiment. Batch 93085 had 508-μm-diameter uranium dioxide (UO2) kernels. Batch 93073 had 427-μm-diameter UCO kernels, which is a kernel design where some of the uranium oxide is converted to uranium carbide during fabrication to provide a getter for oxygen liberated during fission and limit CO production. Fabrication and property data for the AGR-2 coating batches have been compiled and compared to those for AGR-1. The AGR-2 TRISO coatings were most like the AGR-1 Variant 3 TRISO deposited in the 50-mm-diameter ORNL lab-scale coater. In both cases argon-dilution of the hydrogen and methyltrichlorosilane coating gas mixture employed to deposit the SiC was used to produce a finer-grain, more equiaxed SiC microstructure. In addition to the fact that AGR-1 fuel had smaller, 350-μm-diameter UCO kernels, notable differences in the TRISO particle properties included the pyrocarbon anisotropy, which was slightly higher in the particles coated in the engineering-scale coater, and the exposed kernel defect fraction, which was higher for AGR-2 fuel due to the detected presence of particles with impact damage introduced during TRISO particle handling.

  17. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    SciTech Connect

    Sienicki, James J.; Grandy, Christopher

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  18. TEMPERATURE DEPENDANT BEHAVIOUR OBSERVED IN THE AFIP-6 IRRADIATION TEST

    SciTech Connect

    A. B. Robinson; D. M. Wachs; P. Medvedev; S.J. Miller; F. J. Rice; M. K. Meyer; D. M. Perez

    2012-03-01

    The AFIP-6 test assembly was irradiated for one cycle in the Advanced Test Reactor at Idaho National Laboratory. The experiment was designed to test two monolithic fuel plates at power and burn-ups which bounded the operating conditions of both ATR and HFIR driver fuel. Both plates contained a solid U-Mo fuel foil with a zirconium diffusion barrier between 6061-aluminum cladding plates bonded by hot isostatic pressing. The experiment was designed with an orifice to restrict the coolant flow in order to obtain prototypic coolant temperature conditions. While these coolant temperatures were obtained, the reduced flow resulted in a sufficiently low heat transfer coefficient that failure of the fuel plates occurred. The increased fuel temperature led to significant variations in the fission gas retention behaviour of the U-Mo fuel. These variations in performance are outlined herein.

  19. Irradiation testing of a niobium-molybdenum developmental thermocouple

    SciTech Connect

    Knight, R.C.; Greenslade, D.L.

    1991-10-01

    A need exists for a radiation-resistant thermocouple capable of monitoring temperatures in excess of the limits of the chromel/alumel system. Tungsten/rhenium and platinum/rhodium thermocouples have sufficient temperature capability but have proven to be unstable because of irradiation-induced decalibration. The niobium/molybdenum system is believed to hold great potential for nuclear applications at temperatures up to 2000 K. However, the fragility of pure niobium and fabrication problems with niobium/molybdenum alloys have limited development of this system. Utilizing the Fast Flux Test Facility, a developmental thermocouple with a thermoelement pair consisting of a pure molybdenum and a niobium-1%zirconium alloy wire was irradiated fro 7200 hours at a temperature of 1070 K. The thermocouple performed flawlessly for the duration of the experiment and exhibited stability comparable to a companion chromel/alumel unit. A second thermocouple, operating at 1375 K, is currently being employed to monitor a fusion materials experiment in the Fast Flux Test Facility. This experiment, also scheduled for 7200 hours, will serve to further evaluate the potential of the niobium-1%zirconium/molybdenum thermoelement system. 7 refs., 7 figs.

  20. Evaluation of the advanced mixed oxide fuel test FO-2 irradiated in Fast Flux Test Facility

    SciTech Connect

    Gilpin, L.L.; Baker, R.B.; Chastain, S.A.

    1989-05-01

    The advanced mixed-oxide (UO/sub 2/-PuO/sub 2/) test assembly, FO-2, irradiated in the Fast Flux Test Facility (FFTF), is undergoing postirradiation examination (PIE). This is one of the first FFTF tests examined that used the advanced ferrite-martensite alloy, HT9, which is highly resistant to irradiation swelling. The FO-2 includes the first annular fueled pins irradiated in FFTF to undergo destructive examination. The FO-2 is a lead assembly for the ongoing FFTF Core Demonstration Experiment (CDE) (Leggett and Omberg 1987) and was designed to evaluate the effects of fuel design variables, such as pellet density, smeared density, and fuel form (annular or solid fuel), on advanced pin performance. The assembly contains a total of 169 fuel pins of twelve different types. The test was irradiated for 312 equivalent full power days (EFPD) in FFTF. It had a peak pin power of 13.7 kW/ft and reached a peak burnup of 65.2 MWd/kgM with a peak fast fluence of 9.9 /times/ 10/sup 22/ n/cm/sup 2/ (E > 0.1 MeV). This document discusses the test and its results. 6 refs., 19 figs., 4 tabs.

  1. Improving Thermal Model Prediction Through Statistical Analysis of Irradiation and Post-Irradiation Data from AGR Experiments

    SciTech Connect

    Binh T. Pham; Grant L. Hawkes; Jeffrey J. Einerson

    2014-05-01

    As part of the High Temperature Reactors (HTR) R&D program, a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. While not possible to obtain by direct measurements in the tests, crucial fuel conditions (e.g., temperature, neutron fast fluence, and burnup) are calculated using core physics and thermal modeling codes. This paper is focused on AGR test fuel temperature predicted by the ABAQUS code's finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for qualification of AGR-1 thermocouple data. Abnormal trends in measured data revealed by the statistical analysis are traced to either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. The main thrust of this work is to exploit the variety of data obtained in irradiation and post-irradiation examination (PIE) for assessment of modeling assumptions. As an example, the uneven reduction of the control gas gap in Capsule 5 found in the capsule metrology measurements in PIE helps identify mechanisms other than TC drift causing the decrease in TC readings. This suggests a more physics-based modification of the thermal model that leads to a better fit with experimental data, thus reducing model uncertainty and increasing confidence in the calculated fuel temperatures of the AGR-1 test.

  2. Improving Thermal Model Prediction Through Statistical Analysis of Irradiation and Post-Irradiation Data from AGR Experiments

    SciTech Connect

    Dr. Binh T. Pham; Grant L. Hawkes; Jeffrey J. Einerson

    2012-10-01

    As part of the Research and Development program for Next Generation High Temperature Reactors (HTR), a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. The data representing the crucial test fuel conditions (e.g., temperature, neutron fast fluence, and burnup) while impossible to obtain from direct measurements are calculated by physics and thermal models. The irradiation and post-irradiation examination (PIE) experimental data are used in model calibration effort to reduce the inherent uncertainty of simulation results. This paper is focused on fuel temperature predicted by the ABAQUS code’s finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for improving qualification of AGR-1 thermocouple data. The present work exercises the idea that the abnormal trends of measured data observed from statistical analysis may be caused by either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. As an example, the uneven reduction of the control gas gap in Capsule 5 revealed by the capsule metrology measurements in PIE helps justify the reduction in TC readings instead of TC drift. This in turn prompts modification of thermal model to better fit with experimental data, thus help increase confidence, and in other word reduce model uncertainties in thermal simulation results of the AGR-1 test.

  3. Irradiation Test of Advanced PWR Fuel in Fuel Test Loop at HANARO

    SciTech Connect

    Yang, Yong Sik; Bang, Je Geon; Kim, Sun Ki; Song, Kun Woo; Park, Su Ki; Seo, Chul Gyo

    2007-07-01

    A new fuel test loop has been constructed in the research reactor HANARO at KAERI. The main objective of the FTL (Fuel Test Loop) is an irradiation test of a newly developed LWR fuel under PWR or Candu simulated conditions. The first test rod will be loaded within 2007 and its irradiation test will be continued until a rod average their of 62 MWd/kgU. A total of five test rods can be loaded into the IPS (In-Pile Section) and fuel centerline temperature, rod internal pressure and fuel stack elongation can be measured by an on-line real time system. A newly developed advanced PWR fuel which consists of a HANA{sup TM} alloy cladding and a large grain UO{sub 2} pellet was selected as the first test fuel in the FTL. The fuel cladding, the HANA{sup TM} alloy, is an Nb containing Zirconium alloy that has shown better corrosion and creep resistance properties than the current Zircaloy-4 cladding. A total of six types of HANA{sup TM} alloy were developed and two or three of these candidate alloys will be used as test rod cladding, which have shown a superior performance to the others. A large-grain UO{sub 2} pellet has a 14{approx}16 micron 2D diameter grain size for a reduction of a fission gas release at a high burnup. In this paper, characteristics of the FTL and IPS are introduced and the expected operation and irradiation conditions are summarized for the test periods. Also the preliminary fuel performance analysis results, such as the cladding oxide thickness, fission gas release and rod internal pressure, are evaluated from the test rod safety analysis aspects. (authors)

  4. Recent Accomplishments in the Irradiation Testing of Engineering-Scale Monolithic Fuel Specimens

    SciTech Connect

    N.E. Woolstenhulme; D.M. Wachs; M.K. Meyer; H.W. Glunz; R.B. Nielson

    2012-10-01

    The US fuel development team is focused on qualification and demonstration of the uranium-molybdenum monolithic fuel including irradiation testing of engineering-scale specimens. The team has recently accomplished the successful irradiation of the first monolithic multi-plate fuel element assembly within the AFIP-7 campaign. The AFIP-6 MKII campaign, while somewhat truncated by hardware challenges, exhibited successful irradiation of a large-scale monolithic specimen under extreme irradiation conditions. The channel gap and ultrasonic data are presented for AFIP-7 and AFIP-6 MKII, respectively. Finally, design concepts are summarized for future irradiations such as the base fuel demonstration and design demonstration experiment campaigns.

  5. The Fuel Accident Condition Simulator (FACS) furnace system for high temperature performance testing of VHTR fuel

    SciTech Connect

    Paul A. Demkowicz; David V. Laug; Dawn M. Scates; Edward L. Reber; Lyle G. Roybal; John B. Walter; Jason M. Harp; Robert N. Morris

    2012-10-01

    The AGR-1 irradiation of TRISO-coated particle fuel specimens was recently completed and represents the most successful such irradiation in US history, reaching peak burnups of greater than 19% FIMA with zero failures out of 300,000 particles. An extensive post-irradiation examination (PIE) campaign will be conducted on the AGR-1 fuel in order to characterize the irradiated fuel properties, assess the in-pile fuel performance in terms of coating integrity and fission metals release, and determine the fission product retention behavior during high temperature safety testing. A new furnace system has been designed, built, and tested to perform high temperature accident tests. The Fuel Accident Condition Simulator furnace system is designed to heat fuel specimens at temperatures up to 2000 degrees C in helium while monitoring the release of volatile fission metals (e.g. Cs, Ag, Sr, and Eu), iodine, and fission gases (Kr, Xe). Fission gases released from the fuel to the sweep gas are monitored in real time using dual cryogenic traps fitted with high purity germanium detectors. Condensable fission products are collected on a plate attached to a water-cooled cold finger that can be exchanged periodically without interrupting the test. Analysis of fission products on the condensation plates involves dry gamma counting followed by chemical analysis of selected isotopes. This paper will describe design and operational details of the Fuel Accident Condition Simulator furnace system and the associated fission gas monitoring system, as well as preliminary system calibration results.

  6. Graphite irradiation testing for HTR technology at the High Flux Reactor in Petten

    NASA Astrophysics Data System (ADS)

    Vreeling, J. A.; Wouters, O.; Laan, J. G. van der

    2008-10-01

    In 2001 the Nuclear Research and Consultancy Group started a large graphite irradiation program for the development of High Temperature Reactor technology in the European framework. The irradiation experiments, containing present day available graphite grades, are performed at the High Flux Reactor in Petten. The grades are NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, LPEB, IG-110 and IG-430. In the fifth framework programme (2001-2004) and sixth framework programme (2005-2009) four irradiation experiments are foreseen, resulting in design curves at irradiation temperatures between 650 °C and 950 °C. The post-irradiation testing is focused on dimensional changes, dynamic Young's modulus, coefficient of thermal expansion and coefficient of thermal conductivity. The irradiation programme and preliminary results from the first irradiation experiment at 750 °C to 8 dpa will be discussed in this paper.

  7. Ion irradiation testing of Improved Accident Tolerant Cladding Materials

    SciTech Connect

    Anderoglu, Osman; Tesmer, Joseph R.; Maloy, Stuart A.

    2014-01-14

    This report summarizes the results of ion irradiations conducted on two FeCrAl alloys (named as ORNL A&B) for improving the accident tolerance of LWR nuclear fuel cladding. After irradiation with 1.5 MeV protons to ~0.5 to ~1 dpa and 300°C nanoindentations were performed on the cross-sections along the ion range. An increase in hardness was observed in both alloys. Microstructural analysis shows radiation induced defects.

  8. Initiate test loop irradiations of ALSEP process solvent

    SciTech Connect

    Peterman, Dean R.; Olson, Lonnie G.; McDowell, Rocklan G.

    2014-09-01

    This report describes the initial results of the study of the impacts of gamma radiolysis upon the efficacy of the ALSEP process and is written in completion of milestone M3FT-14IN030202. Initial irradiations, up to 100 kGy absorbed dose, of the extraction section of the ALSEP process have been completed. The organic solvent used for these experiments contained 0.05 M TODGA and 0.75 M HEH[EHP] dissolved in n-dodecane. The ALSEP solvent was irradiated while in contact with 3 M nitric acid and the solutions were sparged with compressed air in order to maintain aerated conditions. The irradiated phases were used for the determination of americium and europium distribution ratios as a function of absorbed dose for the extraction and stripping conditions. Analysis of the irradiated phases in order to determine solvent composition as a function of absorbed dose is ongoing. Unfortunately, the failure of analytical equipment necessary for the analysis of the irradiated samples has made the consistent interpretation of the analytical results difficult. Continuing work will include study of the impacts of gamma radiolysis upon the extraction of actinides and lanthanides by the ALSEP solvent and the stripping of the extracted metals from the loaded solvent. The irradiated aqueous and organic phases will be analyzed in order to determine the variation in concentration of solvent components with absorbed gamma dose. Where possible, radiolysis degradation product will be identified.

  9. AFCI Fuel Irradiation Test Plan, Test Specimens AFC-1Æ and AFC-1F

    SciTech Connect

    D. C. Crawford; S. L. Hayes; B. A. Hilton; M. K. Meyer; R. G. Ambrosek; G. S. Chang; D. J. Utterbeck

    2003-11-01

    The U. S. Advanced Fuel Cycle Initiative (AFCI) seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby dramatically decreasing the volume of material requiring disposition and the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository (DOE, 2003). One important component of the technology development is actinide-bearing transmutation fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. There are little irradiation performance data available on non-fertile fuel forms, which would maximize the destruction rate of plutonium, and low-fertile (i.e., uranium-bearing) fuel forms, which would support a sustainable nuclear energy option. Initial scoping level irradiation tests on a variety of candidate fuel forms are needed to establish a transmutation fuel form design and evaluate deployment of transmutation fuels.

  10. Irradiation Programs and Test Plans to Assess High-Fluence Irradiation Assisted Stress Corrosion Cracking Susceptibility.

    SciTech Connect

    Teysseyre, Sebastien

    2015-03-01

    . Irradiation assisted stress corrosion cracking (IASCC) is a known issue in current reactors. In a 60 year lifetime, reactor core internals may experience fluence levels up to 15 dpa for boiling water reactors (BWR) and 100+ dpa for pressurized water reactors (PWR). To support a safe operation of our fleet of reactors and maintain their economic viability it is important to be able to predict any evolution of material behaviors as reactors age and therefore fluence accumulated by reactor core component increases. For PWR reactors, the difficulty to predict high fluence behavior comes from the fact that there is not a consensus of the mechanism of IASCC and that little data is available. It is however possible to use the current state of knowledge on the evolution of irradiated microstructure and on the processes that influences IASCC to emit hypotheses. This report identifies several potential changes in microstructure and proposes to identify their potential impact of IASCC. The susceptibility of a component to high fluence IASCC is considered to not only depends on the intrinsic IASCC susceptibility of the component due to radiation effects on the material but to also be related to the evolution of the loading history of the material and interaction with the environment as total fluence increases. Single variation type experiments are proposed to be performed with materials that are representative of PWR condition and with materials irradiated in other conditions. To address the lack of IASCC propagation and initiation data generated with material irradiated in PWR condition, it is proposed to investigate the effect of spectrum and flux rate on the evolution of microstructure. A long term irradiation, aimed to generate a well-controlled irradiation history on a set on selected materials is also proposed for consideration. For BWR, the study of available data permitted to identify an area of concern for long term performance of component. The efficiency of

  11. Development of New Irradiation Facilities for Material Testing

    NASA Astrophysics Data System (ADS)

    Eck, J.; Lavielle, D.; Millera, T.

    2014-06-01

    JUICE (JUpiter ICy moons Explorer) and Solar Orbiter are very challenging from scientific and technical point of views.These two spacecrafts will have to face harsh environmental conditions (particle and UV radiation, wide temperature ranges,...) and considered materials have to withstand these constraints to ensure the proper operation of the instruments of the payload.New facilities are currently under development at TRAD, in order to evaluate the behavior of the candidate materials in representative conditions. VEISpa (Vacuum Electron Irradiation facility for Spatialization) will allow electron irradiations up to an energy of 4 MeV, under high vacuum whereas SWiPI (Solar Wind Proton Irradiation) facility will focus on the effects of solar wind protons.Both facilities will allow the study of synergistic effects of radiation and temperature constraints (typically -150/+200°C) that can lead to different material behaviour compared to cumulative effects.

  12. IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR

    SciTech Connect

    I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

    2010-10-01

    Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

  13. Microstructure and fracture behavior of F82H steel under different irradiation and tensile test conditions

    NASA Astrophysics Data System (ADS)

    Wang, K.; Dai, Y.; Spätig, P.

    2016-01-01

    Specimens of martensitic steel F82H were irradiated to doses ranging from 10.7 dpa/850 appm He to 19.6 dpa/1740 appm He at temperatures between 165 and 305 °C in the second experiment of SINQ Target Irradiation Program (STIP-II). Tensile tests were conducted at different temperatures and various fracture modes were observed. Microstructural changes including irradiation-induced defect clusters, dislocation loops and helium bubbles under different irradiation conditions were investigated using transmission electron microscopy (TEM). The deformation microstructures of tensile tested specimens were carefully examined to understand the underlying deformation mechanisms. Deformation twinning was for the first time observed in irradiated martensitic steels. A change of deformation mechanism from dislocation channeling to deformation twinning was observed when the fracture mode changed from rather ductile (quasi-cleavage) to brittle (intergranular or cleavage and intergranular mixed).

  14. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of

  15. Proposed new damp heat test standards for commercial CIGS modules with bias application or light irradiation

    NASA Astrophysics Data System (ADS)

    Sakurai, Keiichiro; Tomita, Hiroshi; Ogawa, Kinichi; Schmitz, Darshan; Shibata, Hajime; Tokuda, Shuuji; Masuda, Atsushi

    2016-09-01

    Based on our results that conventional damp-heat (DDH) test on a commercial CCIGS (a.k.a. CCIS, CIGSS) module causes an irreversible "Test-specific" degradation (TSD) that is not observed in modules deployed in fields, we propose a new option for DDH testing of CIGS modules. We have tested full-size CIGS modules with/without forward bias, light irradiation and humidity during heat tests. The results clearly show that adding forward bias, or white light irradiation during DH tests suppresses this irreversible degradation. Based on these results, we have proposed to add forward bias and/or light irradiation during DH tests of CIGS modules, to make the test condition closer to real fields and suppress degradations not observed in the field.

  16. Test system accurately determines tensile properties of irradiated metals at cryogenic temperatures

    NASA Technical Reports Server (NTRS)

    Levine, P. J.; Skalka, R. J.; Vandergrift, E. F.

    1967-01-01

    Modified testing system determines tensile properties of irradiated brittle-type metals at cryogenic temperatures. The system includes a lightweight cryostat, split-screw grips, a universal joint, and a special temperature control system.

  17. Development of a small specimen test machine to evaluate irradiation embrittlement of fusion reactor materials

    NASA Astrophysics Data System (ADS)

    Ishii, T.; Ohmi, M.; Saito, J.; Hoshiya, T.; Ooka, N.; Jitsukawa, S.; Eto, M.

    2000-12-01

    Small specimen test techniques (SSTT) are essential to use an accelerator-driven deuterium-lithium stripping reaction neutron source for the study of fusion reactor materials because of the limitation of the available irradiation volume. A remote-controlled small punch (SP) test machine was developed at the hot laboratory of the Japan Materials Testing Reactor (JMTR) in the Japan Atomic Energy Research Institute (JAERI). This report describes the SP test method and machine for use in a hot cell, and test results on irradiated ferritic steels. The specimen was either a coupon 10×10×0.25 mm 3 or a TEM disk 3 mm in diameter by 0.25 mm in thickness. Tests can be performed at temperatures ranging from 93 to 1123 K in a vacuum or in an inert gas environment. The ductile to brittle transition temperature of the irradiated ferritic steel as determined by the SP test is also evaluated.

  18. JOYO-1 Irradiation Test Campaign Technical Close-out, For Information

    SciTech Connect

    G. Borges

    2006-01-31

    The JOYO-1 irradiation testing was designed to screen the irradiation performance of candidate cladding, structural and reflector materials in support of space reactor development. The JOYO-1 designation refers to the first of four planned irradiation tests in the JOYO reactor. Limited irradiated material performance data for the candidate materials exists for the expected Prometheus-1 duration, fluences and temperatures. Materials of interest include fuel element cladding and core materials (refractory metal alloys and silicon carbide (Sic)), vessel and plant structural materials (refractory metal alloys and nickel-base superalloys), and control and reflector materials (BeO). Key issues to be evaluated were long term microstructure and material property stability. The JOYO-1 test campaign was initiated to irradiate a matrix of specimens at prototypical temperatures and fluences anticipated for the Prometheus-1 reactor [Reference (1)]. Enclosures 1 through 9 describe the specimen and temperature monitors/dosimetry fabrication efforts, capsule design, disposition of structural material irradiation rigs, and plans for post-irradiation examination. These enclosures provide a detailed overview of Naval Reactors Prime Contractor Team (NRPCT) progress in specific areas; however, efforts were in various states of completion at the termination of NRPCT involvement with and restructuring of Project Prometheus.

  19. Reweldability test of irradiated SS316 by the TIG welding method

    NASA Astrophysics Data System (ADS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Oyamada, Rokuro

    1996-10-01

    Stainless steel is a candidate material for the structural material in fusion reactors. Rewelding of irradiated materials will have a large impact on the design and the maintenance of in-vessel components. In the present work, the welding specimens made of type 316 stainless steel were irradiated in JMTR (Japan materials testing reactor) to a fast neutron fluence of ˜2.0 × 10 20 n/cm 2 ( E > 1 MeV) at a temperature of ˜200°C. The rewelding of unirradiated and/or irradiated stainless steel was performed by the tungsten inert gas (TIG) welding method and the weldments of unirradiated and/or irradiated SS316 were characterized by tensile testing (test temp.: 20°C and 200°C), hardness, metallographical observation and SEM/XMA analyses.

  20. Pre-irradiation testing of actively cooled Be-Cu divertor modules

    SciTech Connect

    Linke, J.; Duwe, R.; Kuehnlein, W.

    1995-09-01

    A set of neutron irradiation tests is prepared on different plasma facing materials (PFM) candidates and miniaturized components for ITER. Beside beryllium the irradiation program which will be performed in the High Flux Reactor (HFR) in Petten, includes different carbon fiber composites (CFQ) and tungsten alloys. The target values for the neutron irradiation will be 0.5 dpa at temperatures of 350{degrees}C and 700{degrees}C, resp.. The post irradiation examination (PIE) will cover a wide range of mechanical tests; in addition the degradation of thermal conductivity will be investigated. To determine the high heat flux (HHF) performance of actively cooled divertor modules, electron beam tests which simulate the expected heat loads during the operation of ITER, are scheduled in the hot cell electron beam facility JUDITH. These tests on a selection of different actively cooled beryllium-copper and CFC-copper divertor modules are performed before and after neutron irradiation; the pre-irradiation testing is an essential part of the program to quantify the zero-fluence high heat flux performance and to detect defects in the modules, in particular in the brazed joints.

  1. Neutron-Irradiated Samples as Test Materials for MPEX

    SciTech Connect

    Ellis, Ronald James; Rapp, Juergen

    2015-10-09

    Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility.

  2. Neutron-Irradiated Samples as Test Materials for MPEX

    DOE PAGES

    Ellis, Ronald James; Rapp, Juergen

    2015-10-09

    Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of themore » samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility.« less

  3. Advanced Test Reactor In-Canal Ultrasonic Scanner: Experiment Design and Initial Results on Irradiated Plates

    SciTech Connect

    D. M. Wachs; J. M. Wight; D. T. Clark; J. M. Williams; S. C. Taylor; D. J. Utterbeck; G. L. Hawkes; G. S. Chang; R. G. Ambrosek; N. C. Craft

    2008-09-01

    An irradiation test device has been developed to support testing of prototypic scale plate type fuels in the Advanced Test Reactor. The experiment hardware and operating conditions were optimized to provide the irradiation conditions necessary to conduct performance and qualification tests on research reactor type fuels for the RERTR program. The device was designed to allow disassembly and reassembly in the ATR spent fuel canal so that interim inspections could be performed on the fuel plates. An ultrasonic scanner was developed to perform dimensional and transmission inspections during these interim investigations. Example results from the AFIP-2 experiment are presented.

  4. Evaluation of irradiation effects of 16 MeV proton-irradiated 12Cr-1MoV steel by small punch (SP) tests

    SciTech Connect

    Chi, S.H.; Hong, J.H. ); Kim, I.S. . Dept. of Nuclear Engineering)

    1994-06-15

    Recently, interest in small-scale specimens for testing irradiated materials has arisen in conjunction with the need to develop materials for fusion reactor materials and to study irradiation effects using an ion irradiation facility. Several attempts have been made to evaluate material property changes due to irradiation using a small specimen technique. The SP (small punch) test is an example of small-scale specimen test techniques, originally developed by Baik et al. to estimate DBTT (ductile-to-brittle transition temperature) using broken standard CVN (Charpy 5-notch) specimens. The objective of the present study is to evaluate 16 MeV proton irradiation effects on a fusion reactor candidate material in terms of changes in energy up to failure and J[sub IC] fracture toughness (SP J[sub IC]) by using a SP test technique and a J[sub IC] - [bar [epsilon

  5. A new integrated test structure for on-chip post-irradiation annealing in MOS devices

    SciTech Connect

    Chabrerie, C.; Flament, O.; Boudenot, J.C.

    1998-06-01

    The authors have developed a prototype test structure (named THERMOS) demonstrating the feasibility and the interest of the on-chip heating in a Silicon-On-Insulator technology. This circuit has been specially designed for the study of post-irradiation effects in a radiation-hardened CMOS technology. Preliminary results are presented here for the on-chip annealing of irradiated n-channel transistors.

  6. Properties of precipitation hardened steel irradiated at 323 K in the Japan materials testing reactor

    NASA Astrophysics Data System (ADS)

    Niimi, M.; Matsui, Y.; Jitsukawa, S.; Hoshiya, T.; Tsukada, T.; Ohmi, M.; Mimura, H.; Ooka, N.; Hide, K.

    A precipitation hardening type 630 stainless steel was irradiated in the Japan Materials Testing Reactor (JMTR) in contact with the reactor primary coolant. The temperature of the irradiated specimens was about 330 K. The fast neutron ( E > 1 MeV) fluence for the specimens ranged from 10 24 to 10 26 m -2. Tension tests and fracture toughness tests were carried out at room temperature, while Charpy impact tests were done at temperatures of 273-453 K. Tensile strength data showed a peak of 1600 MPa at around 7 × 10 24 m -2, then gradually decreased to about 1500 MPa at 1.2 × 10 26 m -2. The elongation decreased with irradiation from 12% for unirradiated material to 6% at 1.2 × 10 26 m -2. The fractography after the tension test revealed that the fracture was ductile. Fracture toughness decreased to about a half of the value for unirradiated material with irradiation. The cleavage fracture was dominant on the fractured surface. Charpy impact tests showed an increase of ductile-brittle transition temperature (DBTT) by 60 K with irradiation.

  7. MELT WIRE SENSORS AVAILABLE TO DETERMINE PEAK TEMPERATURES IN ATR IRRADIATION TESTING

    SciTech Connect

    K. L. Davis; D. Knudson; J. Daw; J. Palmer; J. L. Rempe

    2012-07-01

    In April 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) a National Scientific User Facility (NSUF) to advance US leadership in nuclear science and technology. By attracting new users from universities, laboratories, and industry, the ATR will support basic and applied nuclear research and development and help address the nation's energy security needs. In support of this new program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced temperature sensors for irradiation testing. Although most efforts emphasize sensors capable of providing real-time data, selected tasks have been completed to enhance sensors provided in irradiation locations where instrumentation leads cannot be included, such as drop-in capsule and Hydraulic Shuttle Irradiation System (HSIS) or 'rabbit' locations. To meet the need for these locations, the INL has developed melt wire temperature sensors for use in ATR irradiation testing. Differential scanning calorimetry and environmental testing of prototypical sensors was used to develop a library of 28 melt wire materials, capable of detecting peak irradiation temperatures ranging from 85 to 1500°C. This paper will discuss the development work and present test results.

  8. TEM in situ micropillar compression tests of ion irradiated oxide dispersion strengthened alloy

    NASA Astrophysics Data System (ADS)

    Yano, K. H.; Swenson, M. J.; Wu, Y.; Wharry, J. P.

    2017-01-01

    The growing role of charged particle irradiation in the evaluation of nuclear reactor candidate materials requires the development of novel methods to assess mechanical properties in near-surface irradiation damage layers just a few micrometers thick. In situ transmission electron microscopic (TEM) mechanical testing is one such promising method. In this work, microcompression pillars are fabricated from a Fe2+ ion irradiated bulk specimen of a model Fe-9%Cr oxide dispersion strengthened (ODS) alloy. Yield strengths measured directly from TEM in situ compression tests are within expected values, and are consistent with predictions based on the irradiated microstructure. Measured elastic modulus values, once adjusted for the amount of deformation and deflection in the base material, are also within the expected range. A pillar size effect is only observed in samples with minimum dimension ≤100 nm due to the low inter-obstacle spacing in the as received and irradiated material. TEM in situ micropillar compression tests hold great promise for quantitatively determining mechanical properties of shallow ion-irradiated layers.

  9. Test plan for the irradiation of nonmetallic materials.

    SciTech Connect

    Brush, Laurence H.; Farnum, Cathy Ottinger; Gelbard, Fred; Dahl, M.; Joslyn, C. C.; Venetz, T. J.

    2013-03-01

    A comprehensive test program to evaluate nonmetallic materials use in the Hanford Tank Farms is described in detail. This test program determines the effects of simultaneous multiple stressors at reasonable conditions on in-service configuration components by engineering performance testing.

  10. Test plan for the irradiation of nonmetallic materials.

    SciTech Connect

    Brush, Laurence H.; Farnum, Cathy Ottinger; Dahl, M.; Joslyn, C. C.; Venetz, T. J.

    2013-05-01

    A comprehensive test program to evaluate nonmetallic materials use in the Hanford tank farms is described in detail. This test program determines the effects of simultaneous multiple stressors at reasonable conditions on in-service configuration components by engineering performance testing.

  11. SUMMARY OF ‘AFIP’ FULL SIZED PLATE IRRADIATIONS IN THE ADVANCED TEST REACTOR

    SciTech Connect

    Robinson, Adam B; Wachs, Daniel M

    2010-03-01

    Recent testing at the Idaho National Laboratory has included four AFIP (ATR Full Size plate In center flux trap Position) experiments. These experiments included both dispersion plates and monolithic plates fabricated by both hot isostatic pressing and friction bonding utilizing both thermally sprayed inter-layers and zirconium barriers. These plates were tested between 100 and 350 w/cm2 at low temperatures and high burn-ups. The post irradiation exams performed have indicated good performance under the conditions tested and a summary of the findings and irradiation history are included herein.

  12. Comparison of proton microbeam and gamma irradiation for the radiation hardness testing of silicon PIN diodes

    NASA Astrophysics Data System (ADS)

    Jakšić, M.; Grilj, V.; Skukan, N.; Majer, M.; Jung, H. K.; Kim, J. Y.; Lee, N. H.

    2013-09-01

    Simple and cost-effective solutions using Si PIN diodes as detectors are presently utilized in various radiation-related applications in which excessive exposure to radiation degrades their charge transport properties. One of the conventional methods for the radiation hardness testing of such devices is time-consuming irradiation with electron beam or gamma-ray irradiation facilities, high-energy proton accelerators, or with neutrons from research reactors. Recently, for the purpose of radiation hardness testing, a much faster nuclear microprobe based approach utilizing proton irradiation has been developed. To compare the two different irradiation techniques, silicon PIN diodes have been irradiated with a Co-60 gamma radiation source and with a 6 MeV proton microbeam. The signal degradation in the silicon PIN diodes for both irradiation conditions has been probed by the IBIC (ion beam induced charge) technique, which can precisely monitor changes in charge collection efficiency. The results presented are reviewed on the basis of displacement damage calculations and NIEL (non-ionizing energy loss) concept.

  13. Post-irradiation mechanical tests on F82H EB and TIG welds

    NASA Astrophysics Data System (ADS)

    Rensman, J.; van Osch, E. V.; Horsten, M. G.; d'Hulst, D. S.

    2000-12-01

    The irradiation behaviour of electron beam (EB) and tungsten inert gas (TIG) welded joints of the reduced-activation martensitic steel IEA heat F82H-mod. was investigated by neutron irradiation experiments in the high flux reactor (HFR) in Petten. Mechanical test specimens, such as tensile specimens and KLST-type Charpy impact specimens, were neutron irradiated up to a dose level of 2-3 dpa at a temperature of 300°C in the HFR reactor in Petten. The tensile results for TIG and EB welds are as expected with practically no strain hardening capacity left. Considering impact properties, there is a large variation in impact properties for the TIG weld. The irradiation tends to shift the DBTT of particularly the EB welds to very high values, some cases even above +250°C. PWHT of EB-welded material gives a significant improvement of the DBTT and USE compared to the as-welded condition.

  14. Application of half-embryo test to irradiated apples and cherries

    NASA Astrophysics Data System (ADS)

    Kawamura, Yoko; Miura, Aya; Sugita, Takiko; Yamada, Takashi; Saito, Yukio

    1995-09-01

    The half-embryo test was applied to irradiated apples and cherries. The optimum incubation temperature for apples and cherries was 30°C and 25°C, respectively. Benzyladenine stimulated the shooting of cherry half-embryos, therefore, they were incubated with 10 μM benzyladenine. The irradiation of apples and cherries caused obvious changes in the growth of the half-embryos. A dose of 0.15 kGy or more almost totally retarded shoot elongation. If shooting is less than 50%, the apples and cherries are identified as "irradiated". An assessment could be made after I to 4 days and the detection limit of the irradiation dose is 0.15 kGy.

  15. The Advanced Test Reactor Irradiation Capabilities Available as a National Scientific User Facility

    SciTech Connect

    S. Blaine Grover

    2008-09-01

    The Advanced Test Reactor (ATR) is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These capabilities include simple capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. Monitoring systems have also been utilized to monitor different parameters such as fission gases for fuel experiments, to measure specimen performance during irradiation. ATR’s control system provides a stable axial flux profile throughout each reactor operating cycle, and allows the thermal and fast neutron fluxes to be controlled separately in different sections of the core. The ATR irradiation positions vary in diameter from 16 mm to 127 mm over an active core height of 1.2 m. This paper discusses the different irradiation capabilities with examples of different experiments and the cost/benefit issues related to each capability. The recent designation of ATR as a national scientific user facility will make the ATR much more accessible at very low to no cost for research by universities and possibly commercial entities.

  16. Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 200/sup 0/C

    SciTech Connect

    Conlin, J.A.

    1985-08-01

    A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 200/sup 0/C. The design description and results of the prototype capsule performance are presented.

  17. Evaluation of the advanced mixed-oxide fuel test FO-2 irradiated in the FFTF (Fast Flux Test Facility)

    SciTech Connect

    Burley Gilpin, L.L.; Chastain, S.A.; Baker, R.B.

    1989-01-01

    The advanced mixed-oxide (UO{sub 2}-PuO{sub 2}) test assembly, FO-2, irradiated in the Fast Flux Test Facility (FFTF) is undergoing postirradiation examination. This is one of the first FFTF tests examined that used the advanced ferrite-martensite alloy, HT9, which is highly resistant to irradiation swelling. The FO-2 includes the first annular fueled pins irradiated in FFTF to undergo destructive examination. The FO-2 is a lead assembly for the ongoing FFTF Core Demonstration Experiment (CDE) and was designed to evaluate the effects of fuel design variables, such as pellet density, smeared density, and fuel form (annular or solid fuel), on advanced pin performance. The assembly contains a total of 169 fuel pins of 12 different types. Two L (annular) fuel pins, GF02L04 (FFTF and transient tested) and GF02L09 (FFTF only), were destructively examined. Evaluation of the FO-2 fuel pins and assembly shows the excellent and predictable performance of the mixed-oxide fuels with HT9 structural material. This, combined with the robust behavior of the pins in transient tests, and the continued excellent performance of the CDE indicate this is a superior fuel system for liquid-metal reactors. It offers greatly reduced deformation during irradiation, while maintaining good operating characteristics.

  18. Irradiation Test of Fuel Containing Minor Actinides in the Experimental Fast Reactor Joyo

    NASA Astrophysics Data System (ADS)

    Soga, Tomonori; Sekine, Takashi; Tanaka, Kosuke; Kitamura, Ryoichi; Aoyama, Takafumi

    The mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast reactor Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted in the Joyo MK-III 3rd operational cycle. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX), and reference MOX fuel. The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes in order to confirm whether or not fuel melting occurred. After 10 minutes irradiation in May 2006, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins including neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. Linear heat rates for each pin were calculated using MCNP accounting for both prompt and delayed heating components, and then adjusted using E/C for 10B (n, α) reaction rates measured in the MK-III core neutron field characterization test. Post irradiation examination of these pins to confirm the fuel melting and the local concentration under irradiation of NpO2-x or AmO2-x in the (U, Pu)O2-x fuel are underway. The test results are expected to reduce uncertainties on the design margin in the thermal design for MA-MOX fuel.

  19. Identification of irradiated wheat by germination test, DNA comet assay and electron spin resonance

    NASA Astrophysics Data System (ADS)

    Barros, Adilson C.; Freund, Maria Teresa L.; Villavicencio, Ana Lúcia C. H.; Delincée, Henry; Arthur, Valter

    2002-03-01

    In several countries, there has been an increase in the use of radiation for food processing thus improving the quality and sanitary conditions, inhibiting pathogenic microorganisms, delaying the natural aging process and so extending product lifetime. The need to develop analytical methods to detect these irradiated products is also increasing. The goal of this research was to identify wheat irradiated using different radiation doses. Seeds were irradiated with a gamma 60Co source (Gammacell 220 GC) in the Centro de Energia Nuclear na Agricultura and the Instituto de Pesquisas Energéticas e Nucleares. Dose rate used were 1.6 and 5.8kGy/h. Applied doses were 0.0, 0.10, 0.25, 0.50, 0.75, 1.0, and 2.0kGy. After irradiation, seeds were analysed over a 6 month period. Three different detection methods were employed to determine how irradiation had modified the samples. Screening methods consisted of a germination test measuring the inhibition of shooting and rooting and analysis of DNA fragmentation. The method of electron spin resonance spectroscopy allowed a better dosimetric evaluation. These techniques make the identification of irradiated wheat with different doses possible.

  20. Irradiation testing of 316L(N)-IG austenitic stainless steel for ITER

    NASA Astrophysics Data System (ADS)

    van Osch, E. V.; Horsten, M. G.; de Vries, M. I.

    1998-10-01

    In the frame work of the European Fusion Technology Programme and the International Thermonuclear Experimental Reactor (ITER), ECN is investigating the irradiation behaviour of the structural materials for ITER. The main structural material for ITER is austenitic stainless steel Type 316L(N)-IG. The operating temperatures of (parts of) the components are envisaged to range between 350 and 700 K. A significant part of the dose-temperature domain of irradiation conditions relevant for ITER has already been explored, there is, however, very little data at about 600 K. Available data tend to indicate a maximum in the degradation of the mechanical properties after irradiation at this temperature, e.g. a minimum in ductility and a maximum of hardening. Therefore an irradiation program for plate material 316L(N)-IG, its Electron Beam (EB) weld and Tungsten Inert Gas (TIG) weld metal, and also including Hot Isostatically Pressed (HIP) 316L(N) powder and solid-solid joints, was set up in 1995. Irradiations have been carried out in the High Flux Reactor (HFR) in Petten at a temperature of 600 K, at dose levels from 1 to 10 dpa. The paper presents the currently available post-irradiation test results. Next to tensile and fracture toughness data on plate, EB and TIG welds, first results of powder HIP material are included.

  1. Tensile tests and metallography of brazed AISI 316L specimens after irradiation

    NASA Astrophysics Data System (ADS)

    Groot, P.; Franconi, E.

    1994-08-01

    Stainless steel type 316L tensile specimens were vacuum brazed with three kinds of alloys: BNi-5, BNi-6, and BNi-7. The specimens were irradiated up to 0.7 dpa at 353 K in the High Flux Reactor at JRC Petten, the Netherlands. Tensile tests were performed at a constant displacement rate of 10 -3 s -1 at room temperature in the ECN hot cell facility. BNi-5 brazed specimens showed ductile behaviour. Necking and fractures were localized in the plate material. BNi-6 and BNi-7 brazed specimens failed brittle in the brazed zone. This was preceded by uniform deformation of the plate material. Tensile test results of irradiated specimens showed higher stresses due to radiation hardening and a reduction of the elongation of the plate material compared to the reference. SEM examination of the irradiated BNi-6 and BNi-7 fracture surfaces showed nonmetallic phases. These phases were not found in the reference specimens.

  2. Gamma-irradiation tests of IR optical fibres for ITER thermography--a case study

    SciTech Connect

    Reichle, R.; Pocheau, C.; Jouve, M.

    2008-03-12

    In the course of the development of a concept for a spectrally resolving infrared thermography diagnostic for the ITER divertor we have tested 3 types of infrared (IR) fibres in Co{sup 60} irradiation facilities under {gamma} irradiation. The fibres were ZrF{sub 4} (and HfF{sub 4}) fibres from different manufacturers, hollow fibres (silica capillaries with internal Ag/AgJ coating) and a sapphire fibre. For the IR range, only the latter fibre type encourages to go further for neutron tests in a reactor. If one restricted the interest onto the near infrared range, high purity core silica fibres could be used. This study might be seen as a typical example of the relation between diagnostic development for a nuclear environment and irradiation experiments.

  3. Gamma-irradiation tests of IR optical fibres for ITER thermography—a case study

    NASA Astrophysics Data System (ADS)

    Reichle, R.; Brichard, B.; Pocheau, C.; Jouve, M.; van Ierschot, S.; Martinez, S.; Ooms, H.; Berghmans, F.; Decréton, M.

    2008-03-01

    In the course of the development of a concept for a spectrally resolving infrared thermography diagnostic for the ITER divertor we have tested 3 types of infrared (IR) fibres in Co60 irradiation facilities under γ irradiation. The fibres were ZrF4 (and HfF4) fibres from different manufacturers, hollow fibres (silica capillaries with internal Ag/AgJ coating) and a sapphire fibre. For the IR range, only the latter fibre type encourages to go further for neutron tests in a reactor. If one restricted the interest onto the near infrared range, high purity core silica fibres could be used. This study might be seen as a typical example of the relation between diagnostic development for a nuclear environment and irradiation experiments.

  4. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    SciTech Connect

    Carmack, William Jonathan; Barrett, Kristine Eloise; Chichester, Heather Jean MacLean

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  5. Flexural strength of proof-tested and neutron-irradiated silicon carbide

    NASA Astrophysics Data System (ADS)

    Price, R. J.; Hopkins, G. R.

    1982-08-01

    Proof testing before service is a valuable method for ensuring the reliability of ceramic structures. Silicon carbide has been proposed as a very low activation first-wall and blanket structural material for fusion devices, where it would experience a high flux of fast neutrons. Strips of three types of silicon carbide were loaded in four-point bending to a stress sufficient to break about a third of the specimens. Groups of 16 survivors were irradiated to 2 × 10 26n/ m2 ( E>0.05 MeV) at 740°C and bend tested to failure. The strength distribution of chemically vapor-deposited silicon carbide (Texas Instruments) was virtually unchanged by irradiation. The mean strength of sintered silicon carbide (Carborundum Alpha) was reduced 34% by irradiation, while the Weibull modulus and the truncated strength distribution characteristic of proof-tested material were retained. Irradiation reduced the mean strength of reaction-bonded silicon carbide (Norton NC-430) by 58%, and the spread in strength values was increased. We conclude that for the chemically vapor-deposited and the sintered silicon carbide the benefits of proof testing to eliminate low strength material are retained after high neutron exposures.

  6. Evaluation of irradiated pressure vessel steel by mechanical tests and positron annihilation lineshape analysis

    SciTech Connect

    Nakamura, Noriko; Ohta, Yoshio; Yoshida, Kazuo; Maeda, Noriyoshi

    1999-10-01

    Mechanical test and positron annihilation lineshape analysis have been performed on neutron irradiated pressure vessel steels, A533B1 steel and the weld metal. Marked changes in the mechanical properties were observed for both metals after the neutron exposure. S-parameters, the positron annihilation parameters, also increased after the neutron irradiation but only the small change was observed in the different levels of neutron fluence. The change in S-parameter and the mechanical properties were well correlated. It is concluded that changes in embrittlement induced by radiation can be monitored by positron annihilation lineshape analysis but detectability is dependent on the materials.

  7. Test Temperature Dependence of Transesterification of Triolein under Low-Frequency Ultrasonic Irradiation Condition

    NASA Astrophysics Data System (ADS)

    Hanh, Hoang Duc; Dong, Nguyen The; Okitsu, Kenji; Maeda, Yasuaki; Nishimura, Rokuro

    2007-07-01

    The test temperature dependence of the transesterification of triolein with methanol and a base catalyst (NaOH and KOH) was investigated using a molar ratio of methanol to triolein of 6:1, a catalyst concentration of 1% and a temperature range of 3-50 °C under a low-frequency ultrasonic irradiation condition (40 kHz). It was found that the methyl ester concentration at an irradiation time of 5 min increased with increasing temperature, where it tended to level off at temperatures higher than 20 °C. Furthermore, apparent activation energy was estimated from the relationship between the rate and the reciprocal of temperature.

  8. Charpy impact test results for low activation ferritic alloys irradiated to 30 dpa

    SciTech Connect

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

    1996-04-01

    Miniature specimens of six low activation ferritic alloys have been impact field tested following irradiation at 370{degrees}C to 30 dpa. Comparison of the results with those of control specimens and specimens irradiated to 10 dpa indicates that degradation in the impact behavior appears to have saturated by {approx}10 dpa in at least four of these alloys. The 7.5Cr-2W alloy referred to as GA3X appears most promising for further consideration as a candidate structural material in fusion reactor applications, although the 9Cr-1V alloy may also warrant further investigation.

  9. Status of the Norwegian thorium light water reactor (LWR) fuel development and irradiation test program

    SciTech Connect

    Drera, S.S.; Bjork, K.I.; Kelly, J.F.; Asphjell, O.

    2013-07-01

    Thorium based fuels offer several benefits compared to uranium based fuels and should thus be an attractive alternative to conventional fuel types. In order for thorium based fuel to be licensed for use in current LWRs, material properties must be well known for fresh as well as irradiated fuel, and accurate prediction of fuel behavior must be possible to make for both normal operation and transient scenarios. Important parameters are known for fresh material but the behaviour of the fuel under irradiation is unknown particularly for low Th content. The irradiation campaign aims to widen the experience base to irradiated (Th,Pu)O{sub 2} fuel and (Th,U)O{sub 2} with low Th content and to confirm existing data for fresh fuel. The assumptions with respect to improved in-core fuel performance are confirmed by our preliminary irradiation test results, and our fuel manufacture trials so far indicate that both (Th,U)O{sub 2} and (Th,Pu)O{sub 2} fuels can be fabricated with existing technologies, which are possible to upscale to commercial volumes.

  10. Results of the Irradiation of R6R018 in the Advanced Test Reactor

    SciTech Connect

    Adam B Robinson; Daniel Wachs; Pavel Medvedev; Curtis Clark; Gray Chang; Misti Lillo; Jan-Fong Jue; Glenn Moore; Jared Wight

    2010-04-01

    For over 30 years the Reduced Enrichment for Research and Test Reactors (RERTR) program has worked to provide the fuel technology and analytical support required to convert research and test reactors from nuclear fuels that utilize highly enriched uranium (HEU) to fuels based on low-enriched uranium (LEU) (defined as <20% U-235). This effort is driven by a desire to minimize international civilian commerce in weapons usable materials. The RERTR fuel development program has executed a wide array of fuel tests over the last decade that clearly established the viability of research reactor fuels based on uranium-molybdenum (U-Mo) alloys. Fuel testing has included a large number of dispersion type fuels capable of providing uranium densities up to approximately 8.5 g U/cc (~1.7 g U-235/cc at 20% enrichment). The dispersion fuel designs tested are very similar to existing research test reactor fuels in that the U-Mo particles simply replace the current fuel phase within the matrix. In 2003 it became evident that the first generation U-Mo-based dispersion fuel within an aluminum matrix exhibited significant fuel performance problems at high power and burn-up. These issues have been successfully addressed with a modest modification to the matrix material composition. Testing has shown that small additions of silicon (2–5 wt%) to the aluminum (Al) matrix stabilizes the fuel performance. The fuel plate R6R018 which was irradiated in the Advanced Test Reactor (ATR) as part of the RERTR-9B experiment was part of an investigation into the role of the silicon content in the matrix. This plate consisted of a U-7Mo fuel phase dispersed in an Al-3.5Si matrix clad in Al-6061. This report outlines the fabrication history, the as fabricated analysis performed prior to irradiation, the irradiation conditions, the post irradiation examination results, and an analysis of the plates behavior.

  11. Small-Scale Mechanical Testing on Proton Beam-Irradiated 304 SS from Room Temperature to Reactor Operation Temperature

    NASA Astrophysics Data System (ADS)

    Vo, H.; Reichardt, A.; Howard, C.; Abad, M. D.; Kaoumi, D.; Chou, P.; Hosemann, P.

    2015-12-01

    Austenitic stainless steels are common structural components in light water reactors. Because reactor components are subjected to harsh conditions such as high operating temperatures and neutron radiation, they can undergo irradiation-induced embrittlement and related failure, which compromises reliable operation. Small-scale mechanical testing has seen widespread use as a testing method for both ion- and reactor-irradiated materials because it allows access to the mechanical properties of the ion beam-irradiated region, and for safe handling of a small amount of activated material. In this study, nanoindentation and microcompression testing were performed on unirradiated and 10 dpa proton-irradiated 304 SS, from 25°C to 300°C. Increases in yield stress (YS), critical resolved shear stress (CRSS) and hardness ( H) were seen in the irradiated region relative to the unirradiated region. Relationships between H, YS, and CRSS of irradiated and unirradiated materials are discussed over this temperature range.

  12. Assessment of Initial Test Conditions for Experiments to Assess Irradiation Assisted Stress Corrosion Cracking Mechanisms

    SciTech Connect

    Busby, Jeremy T; Gussev, Maxim N

    2011-04-01

    Irradiation-assisted stress corrosion cracking is a key materials degradation issue in today s nuclear power reactor fleet and affects critical structural components within the reactor core. The effects of increased exposure to irradiation, stress, and/or coolant can substantially increase susceptibility to stress-corrosion cracking of austenitic steels in high-temperature water environments. . Despite 30 years of experience, the underlying mechanisms of IASCC are unknown. Extended service conditions will increase the exposure to irradiation, stress, and corrosive environment for all core internal components. The objective of this effort within the Light Water Reactor Sustainability program is to evaluate the response and mechanisms of IASCC in austenitic stainless steels with single variable experiments. A series of high-value irradiated specimens has been acquired from the past international research programs, providing a valuable opportunity to examine the mechanisms of IASCC. This batch of irradiated specimens has been received and inventoried. In addition, visual examination and sample cleaning has been completed. Microhardness testing has been performed on these specimens. All samples show evidence of hardening, as expected, although the degree of hardening has saturated and no trend with dose is observed. Further, the change in hardening can be converted to changes in mechanical properties. The calculated yield stress is consistent with previous data from light water reactor conditions. In addition, some evidence of changes in deformation mode was identified via examination of the microhardness indents. This analysis may provide further insights into the deformation mode under larger scale tests. Finally, swelling analysis was performed using immersion density methods. Most alloys showed some evidence of swelling, consistent with the expected trends for this class of alloy. The Hf-doped alloy showed densification rather than swelling. This observation may be

  13. Thermal analysis of the FSP-1 fuel pin irradiation test. [for SP-100 space power reactor

    NASA Technical Reports Server (NTRS)

    Lyon, William F., III

    1991-01-01

    Thermal analysis of a pin from the FSP-1 fuels irradiation test has been completed. The purpose of the analysis was to provide predictions of fuel pin temperatures, determine the flow regime within the lithium annulus of the test assembly, and provide a standardized model for a consistent basis of comparison between pins within the test assembly. The calculations have predicted that the pin is operating at slightly above the test design temperatures and that the flow regime within the lithium annulus is a laminar buoyancy driven flow.

  14. Doses from external irradiation to Marshall Islanders from Bikini and Enewetak nuclear weapons tests.

    PubMed

    Bouville, André; Beck, Harold L; Simon, Steven L

    2010-08-01

    Annual doses from external irradiation resulting from exposure to fallout from the 65 atmospheric nuclear weapons tests conducted in the Marshall Islands at Bikini and Enewetak between 1946 and 1958 have been estimated for the first time for Marshallese living on all inhabited atolls. All tests that deposited fallout on any of the 23 inhabited atolls or separate reef islands have been considered. The methodology used to estimate the radiation doses at the inhabited atolls is based on test- and location-specific radiation survey data, deposition density estimates of 137Cs, and fallout times-of-arrival provided in a companion paper (Beck et al.), combined with information on the radionuclide composition of the fallout at various times after each test. These estimates of doses from external irradiation have been combined with corresponding estimates of doses from internal irradiation, given in a companion paper (Simon et al.), to assess the cancer risks among the Marshallese population (Land et al.) resulting from exposure to radiation from the nuclear weapons tests.

  15. Charpy impact tests on martensitic/ferritic steels after irradiation in SINQ target-3

    NASA Astrophysics Data System (ADS)

    Dai, Yong; Marmy, Pierre

    2005-08-01

    Charpy impact tests were performed on martensitic/ferritic (MF) steels T91, F82H, Optifer-V and Optimax-A/-C irradiated in SINQ Target-3 up to 7.5 dpa and 500 appm He in a temperature range of 120-195 °C. Results demonstrate that for all the four kinds of steels, the ductile-to-brittle transition temperature (DBTT) increases with irradiation dose. The difference in the DBTT shifts (ΔDBTT) of the different steels is not significant after irradiation in the SINQ target. The ΔDBTT data from the previous small punch (Δ DBTT SP) and the present Charpy impact (ΔDBTT CVN) tests can be correlated with the expression: Δ DBTT SP = 0.4ΔDBTT CVN. All the ΔDBTT data fall into a linear band when they are plotted versus helium concentration. The results indicate that helium effects on the embrittlement of MF steels are significant, particularly at higher concentrations. It suggests that MF steels may not be very suitable for applications at low temperatures in spallation irradiation environments where helium production is high.

  16. Studies on the methods of identification of irradiated food I. Seedling growth test

    NASA Astrophysics Data System (ADS)

    Qiongying, Liu; Yanhua, Kuang; Yuemei, Zheng

    1993-07-01

    A seedling growth test for the identification of gamma irradiated edible vegetable seeds was described. The identification of gamma irradiated grape and the other seeds has been investigated. The purpose of this study was to develop an easy, rapid and practical technique for the identification of irradiated edible vegetable seeds. Seven different irradiated edible vegetable seeds as: rice ( Oryza sativa), peanut ( Arachis hypogaea), maize ( Zeamays), soybean ( Glycine max), red bean ( Phaseolus angularis), mung bean ( Phaseolus aureus) and catjang cowpea ( Vigna cylindrica) were tested by using the method of seedling growth. All of the edible vegetable seeds were exposed to gamma radiation on different doses, O(CK), 0.5, 1.0, 1.5, 2.0, 3.0, 5.0 kGy. After treatment with above 1.0 kGy dose to the seeds, the seedling rate was less than 50% compared with the control. Although the seedling rate of rice seeds can reached 58%, the seedling growth was not normal and the seedling leaves appeared deformed. The results by this method were helpful to identify gamma treatment of the edible vegetable seeds with above 1.0 kGy dose.

  17. The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions.

    PubMed

    Garrison, L M; Zenobia, S J; Egle, B J; Kulcinski, G L; Santarius, J F

    2016-08-01

    The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ion gun can irradiate the samples with ion currents of 20 μA-500 μA; the typical current used is 72 μA, which is an average flux of 9 × 10(14) ions/(cm(2) s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.

  18. The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

    NASA Astrophysics Data System (ADS)

    Garrison, L. M.; Zenobia, S. J.; Egle, B. J.; Kulcinski, G. L.; Santarius, J. F.

    2016-08-01

    The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000 °C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ion gun can irradiate the samples with ion currents of 20 μA-500 μA; the typical current used is 72 μA, which is an average flux of 9 × 1014 ions/(cm2 s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. The MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.

  19. The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

    SciTech Connect

    Garrison, L. M.; Zenobia, Samuel J.; Egle, Brian J.; Kulcinski, Gerald L.; Santarius, John F.

    2016-08-01

    The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000°C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ion gun can irradiate the samples with ion currents of 20 μA–500 μA; the typical current used is 72 μA, which is an average flux of 9 × 1014 ions/(cm2 s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. In conclusion, the MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.

  20. The materials irradiation experiment for testing plasma facing materials at fusion relevant conditions

    DOE PAGES

    Garrison, L. M.; Zenobia, Samuel J.; Egle, Brian J.; ...

    2016-08-01

    The Materials Irradiation Experiment (MITE-E) was constructed at the University of Wisconsin-Madison Inertial Electrostatic Confinement Laboratory to test materials for potential use as plasma-facing materials (PFMs) in fusion reactors. PFMs in fusion reactors will be bombarded with x-rays, neutrons, and ions of hydrogen and helium. More needs to be understood about the interactions between the plasma and the materials to validate their use for fusion reactors. The MITE-E simulates some of the fusion reactor conditions by holding samples at temperatures up to 1000°C while irradiating them with helium or deuterium ions with energies from 10 to 150 keV. The ionmore » gun can irradiate the samples with ion currents of 20 μA–500 μA; the typical current used is 72 μA, which is an average flux of 9 × 1014 ions/(cm2 s). The ion gun uses electrostatic lenses to extract and shape the ion beam. A variable power (1-20 W), steady-state, Nd:YAG laser provides additional heating to maintain a constant sample temperature during irradiations. The ion beam current reaching the sample is directly measured and monitored in real-time during irradiations. The ion beam profile has been investigated using a copper sample sputtering experiment. In conclusion, the MITE-E has successfully been used to irradiate polycrystalline and single crystal tungsten samples with helium ions and will continue to be a source of important data for plasma interactions with materials.« less

  1. Independent Review of AFC 2A, 2B, and 2E ATR Irradiation Tests

    SciTech Connect

    M. Cappiello; R. Hobbins; K. Penny; L. Walters

    2014-01-01

    As part of the Department of Energy Advanced Fuel Cycle program, a series of fuels development irradiation tests have been performed in the Advanced Test Reactor (ATR) at the Idaho National Laboratory. These tests are providing excellent data for advanced fuels development. The program is focused on the transmutation of higher actinides which best can be accomplished in a sodium-cooled fast reactor. Because a fast test reactor is no longer available in the US, a special test vehicle is used to achieve near-prototypic fast reactor conditions (neutron spectra and temperature) for use in ATR (a water-cooled thermal reactor). As part of the testing program, there were many successful tests of advanced fuels including metals and ceramics. Recently however, there have been three experimental campaigns using metal fuels that experienced failure during irradiation. At the request of the program, an independent review committee was convened to review the post-test analyses performed by the fuels development team, to assess the conclusions of the team for the cause of the failures, to assess the adequacy and completeness of the analyses, to identify issues that were missed, and to make recommendations for improvements in the design and operation of future tests. Although there is some difference of opinion, the review committee largely agreed with the conclusions of the fuel development team regarding the cause of the failures. For the most part, the analyses that support the conclusions are sufficient.

  2. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  3. Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra.

    PubMed

    Ďuran, I; Bolshakova, I; Viererbl, L; Sentkerestiová, J; Holyaka, R; Lahodová, Z; Bém, P

    2010-10-01

    We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10(16) cm(-2) was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

  4. Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra

    SciTech Connect

    Duran, I.; Viererbl, L.; Lahodova, Z.; Sentkerestiova, J.; Bem, P.

    2010-10-15

    We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10{sup 16} cm{sup -2} was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

  5. Status of the NGNP Graphite Creep Experiments AGC-1 and AGC-2 Irradiated in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.

  6. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    SciTech Connect

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed

  7. Long-term laser irradiation tests of optical elements for ESA mission ADM-Aeolus

    NASA Astrophysics Data System (ADS)

    Leinhos, Uwe; Mann, Klaus; Bayer, Armin; Endemann, Martin; Wernham, Denny; Pettazzi, Federico; Thibault, Dominique

    2010-08-01

    The European Space Agency ESA is running a series of earth observation missions. In order to perform global windprofile observation based on Doppler-LIDAR, the satellite ADM-Aelolus will be launched in April 2011 and injected into an orbit 400 km above Earth's surface. ADM-Aeolus will be the first satellite ever that is equipped with a UV-laser (emitting at 355 nm) and a reflector telescope. At LLG, a setup was developed that allows monitoring transmission, reflection and fluorescence of laser-irradiated optical components, in order to assess their possible optical degradation due to radiation-induced contaminant deposition in orbit. For both a high-reflecting mirror and an anti-reflective coated window long-term irradiation tests (up to 500 million laser pulses) were performed at a base pressure < 10-9 mbar, using a XeF excimer laser (wavelength 351 nm, repetition rate 1kHz). At this, samples of polymers used inside the satellite (insulators for cabling, adhesives, etc.) were installed into the chamber, and the interaction of their degassing with the sample surfaces under laser irradiation was investigated. Various paramters were varied including pulse repetition rate, view factor and coatings. Optical degradation associated with contaminant adsorption was detected on the irradiated sample sites.

  8. Results of crack-arrest tests on two irradiated high-copper welds

    SciTech Connect

    Iskander, S.K.; Corwin, W.R.; Nanstead, R.K. )

    1990-12-01

    The objective of this study was to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288{degree}C to an average fluence of 1.9 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV). Evaluation of the results shows that the neutron-irradiation-induced crack-arrest toughness temperature shift is about the same as the Charpy V-notch impact temperature shift at the 41-J energy level. The shape of the lower-bound curves (for the range of test temperatures covered) did not seem to have been altered by irradiation compared to those of the ASME K{sub Ia} curve. 9 refs., 21 figs., 10 tabs.

  9. Radiation effects on rat testes. IX. Studies on oxidative enzymes after partial body gamma irradiation.

    PubMed

    Gupta, G S; Bawa, S R

    1975-08-01

    Oxidative enzymes in the rat testes have been studied after gamma irradiation. The role of these enzymes in relation to spermatogenesis and steroidogenesis after radiation injury to testis has been discussed. Loss of succinic dehydrogenase and sorbitol dehydrogenase reflects the losts of germ cell population. Malic enzyme and malic dehydrogenase seem to the related to the deficiency of steroid hormones, whereas increase in glucose-6-phosphate dehydrogenase and NADP isocitric dehydrogenase is due to secondary stimulation of pituitary.

  10. Minimum bar size for flexure testing of irradiated SiC/SiC composite

    SciTech Connect

    Youngblood, G.E.; Jones, R.H.

    1998-03-01

    This report covers material presented at the IEA/Jupiter Joint International Workshop on SiC/SiC Composites for Fusion structural Applications held in conjunction with ICFRM-8, Sendai, Japan, Oct. 23-24, 1997. The minimum bar size for 4-point flexure testing of SiC/SiC composite recommended by PNNL for irradiation effects studies is 30 {times} 6 {times} 2 mm{sup 3} with a span-to-depth ratio of 10/1.

  11. Status Report on the Fabrication of Fuel Cladding Chemical Interaction Test Articles for ATR Irradiations

    SciTech Connect

    Field, Kevin G.; Howard, Richard H.

    2015-09-28

    FeCrAl alloys are a promising new class of alloys for light water reactor (LWR) applications due to their superior oxidation and corrosion resistance in high temperature environments. The current R&D efforts have focused on the alloy composition and processing routes to generate nuclear grade FeCrAl alloys with optimized properties for enhanced accident tolerance while maintaining properties needed for normal operation conditions. Therefore, the composition and processing routes must be optimized to maintain the high temperature steam oxidation (typically achieved by increasing the Cr and Al content) while still exhibiting properties conducive to normal operation in a LWR (such as radiation tolerance where reducing Cr content is favorable). Within this balancing act is the addition of understanding the influence on composition and processing routes on the FeCrAl alloys for fuel-cladding chemical interactions (FCCI). Currently, limited knowledge exists on FCCI for the FeCrAl-UO2 clad-fuel system. To overcome the knowledge gaps on the FCCI for the FeCrAl-UO2 clad-fuel system a series of fueled irradiation tests have been developed for irradiation in the Advanced Test Reactor (ATR) housed at the Idaho National Laboratory (INL). The first series of tests has already been reported. These tests used miniaturized 17x17 PWR fuel geometry rodlets of second-generation FeCrAl alloys fueled with industrial Westinghouse UO2 fuel. These rodlets were encapsulated within a stainless steel housing.To provide high fidelity experiments and more robust testing, a new series of rodlets have been developed deemed the Accident Tolerant Fuel Experiment #1 Oak Ridge National Laboratory FCCI test (ATF-1 ORNL FCCI). The main driving factor, which is discussed in detail, was to provide a radiation environment where prototypical fuel-clad interface temperatures are met while still maintaining constant contact between industrial fuel and the candidate cladding alloys

  12. First tests of the ion irradiation and implantation beamline at the CMAM

    NASA Astrophysics Data System (ADS)

    Jiménez-Rey, D.; Benedicto, M.; Muñoz-Martín, A.; Bachiller-Perea, D.; Olivares, J.; Climent-Font, A.; Gómez-Ferrer, B.; Rodríguez, A.; Narros, J.; Maira, A.; Álvarez, J.; Nakbi, A.; Zucchiatti, A.; de Aragón, F.; García, J. M.; Vila, R.

    2014-07-01

    The implantation and irradiation beamline of the Tandem ion accelerator of the Centro de Micro Análisis de Materiales (CMAM), in Madrid, has been recently completed with a beam sweep and monitoring system, and a cryostat/furnace. These new implementations convert the beamline into a versatile tool to implant ions, between H and Au2, in different materials with precise control of the sample temperature, which may be varied between -180 °C and 600 °C. The size of the swept area on target may be as large as 10 × 10 cm2. The implantation chamber also allows carrying out in situ or/and on line analyses during the irradiations by means of advanced optical measurements, as well as ion beam analyses (IBA). These advancements can be employed in novel applications such as the fabrication of optical waveguides and irradiation tests of structural and functional materials for future fusion reactors. The results of beam tests and first experiments are shown.

  13. Histopathological changes of testes and eyes by neutron irradiation with boron compounds in mice.

    PubMed

    Kim, Yeon-Joo; Yoon, Won-Ki; Ryu, Si-Yun; Chun, Ki-Jung; Son, Hwa-Young; Cho, Sung-Whan

    2006-03-01

    This study was performed to investigate the biological effects of boron neutron capture therapy (BNCT) on the testes and eyes in mice using HANARO Nuclear Reactor, Korea Atomic Energy Research Institute. BNCT relies on the high capacity of 10B in capturing thermal neutrons. Sodium borocaptate (BSH, 75 ppm, iv) and boronophenylalanine (BPA, 750 ppm, ip) have been used as the boron delivery agents. Mice were irradiated with neutron (flux: 1.036739E +09, Fluence 9.600200E+12) by lying flat pose for 30 (10 Gy) or 100 min (33 Gy) with or without boron carrier treatment. In 45 days of irradiation, histopathological changes of the testes and eyes were examined. Thirty-three Gy neutron irradiation for 100 min induced testicular atrophy in which some of seminiferous tubules showed complete depletion of spermatogenic germ cells. Lens epithelial cells and lens fiber were swollen and showed granular changes in an exposure time dependent manner. However, boron carrier treatment had no significant effect on the lesions. These results suggest that the examination of histopathological changes of lens and testis can be used as "biological dosimeters" for gauging radiation responses and the HANARO Nuclear Reactor has sufficient capacities for the BNCT.

  14. Ion irradiation testing and characterization of FeCrAl candidate alloys

    SciTech Connect

    Anderoglu, Osman; Aydogan, Eda; Maloy, Stuart Andrew; Wang, Yongqiang

    2014-10-29

    The Fuel Cycle Research and Development program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels. This effort involves development of fuel cladding materials that will be resistant to oxidizing environments for extended period of time such as loss of coolant accident. Ferritic FeCrAl alloys are among the promising candidates due to formation of a stable Al₂O₃ oxide scale. In addition to being oxidation resistant, these promising alloys need to be radiation tolerant under LWR conditions (maximum dose of 10-15 dpa at 250 – 350°C). Thus, in addition to a number of commercially available alloys, nuclear grade FeCrAl alloys developed at ORNL were tested using high energy proton irradiations and subsequent characterization of irradiation hardening and damage microstructure. This report summarizes ion irradiation testing and characterization of three nuclear grade FeCrAl cladding materials developed at ORNL and four commercially available Kanthal series FeCrAl alloys in FY14 toward satisfying FCRD campaign goals.

  15. Facility for fast neutron irradiation tests of electronics at the ISIS spallation neutron source

    SciTech Connect

    Andreani, C.; Pietropaolo, A.; Salsano, A.; Gorini, G.; Tardocchi, M.; Paccagnella, A.; Gerardin, S.; Frost, C. D.; Ansell, S.; Platt, S. P.

    2008-03-17

    The VESUVIO beam line at the ISIS spallation neutron source was set up for neutron irradiation tests in the neutron energy range above 10 MeV. The neutron flux and energy spectrum were shown, in benchmark activation measurements, to provide a neutron spectrum similar to the ambient one at sea level, but with an enhancement in intensity of a factor of 10{sup 7}. Such conditions are suitable for accelerated testing of electronic components, as was demonstrated here by measurements of soft error rates in recent technology field programable gate arrays.

  16. Lead test assembly irradiation and analysis Watts Bar Nuclear Plant, Tennessee and Hanford Site, Richland, Washington

    SciTech Connect

    1997-07-01

    The U.S. Department of Energy (DOE) needs to confirm the viability of using a commercial light water reactor (CLWR) as a potential source for maintaining the nation`s supply of tritium. The Proposed Action discussed in this environmental assessment is a limited scale confirmatory test that would provide DOE with information needed to assess that option. This document contains the environmental assessment results for the Lead test assembly irradiation and analysis for the Watts Bar Nuclear Plant, Tennessee, and the Hanford Site in Richland, Washington.

  17. Results of Uranium Dioxide-Tungsten Irradiation Test and Post-Test Examination

    NASA Technical Reports Server (NTRS)

    Collins, J. F.; Debogdan, C. E.; Diianni, D. C.

    1973-01-01

    A uranium dioxide (UO2) fueled capsule was fabricated and irradiated in the NASA Plum Brook Reactor Facility. The capsule consisted of two bulk UO2 specimens clad with chemically vapor deposited tungsten (CVD W) 0.762 and 0.1016 cm (0.030-and 0.040-in.) thick, respectively. The second specimen with 0.1016-cm (0.040-in.) thick cladding was irradiated at temperature for 2607 hours, corresponding to an average burnup of 1.516 x 10 to the 20th power fissions/cu cm. Postirradiation examination showed distortion in the bottom end cap, failure of the weld joint, and fracture of the central vent tube. Diametral growth was 1.3 percent. No evidence of gross interaction between CVD tungsten or arc-cast tungsten cladding and the UO2 fuel was observed. Some of the fission gases passed from the fuel cavity to the gas surrounding the fuel specimen via the vent tube and possibly the end-cap weld failure. Whether the UO2 loss rates through the vent tube were within acceptable limits could not be determined in view of the end-cap weld failure.

  18. Recent irradiation tests of uranium-plutonium-zirconium metal fuel elements

    SciTech Connect

    Pahl, R.G.; Lahm, C.E.; Villarreal, R.; Hofman, G.L.; Beck, W.N.

    1986-09-01

    Uranium-Plutonium-Zirconium metal fuel irradiation tests to support the ANL Integral Fast Reactor concept are discussed. Satisfactory performance has been demonstrated to 2.9 at.% peak burnup in three alloys having 0, 8, and 19 wt % plutonium. Fuel swelling measurements at low burnup in alloys to 26 wt % plutonium show that fuel deformation is primarily radial in direction. Increasing the plutonium content in the fuel diminishes the rate of fuel-cladding gap closure and axial fuel column growth. Chemical redistribution occurs by 2.1 at.% peak burnup and generally involves the inward migration of zirconium and outward migration of uranium. Fission gas release to the plenum ranges from 46% to 56% in the alloys irradiated to 2.9 at.% peak burnup. No evidence of deleterious fuel-cladding chemical or mechanical interaction was observed.

  19. Developments, characterization and proton irradiation damage tests of AlN detectors for VUV solar observations

    NASA Astrophysics Data System (ADS)

    BenMoussa, A.; Soltani, A.; Gerbedoen, J.-C.; Saito, T.; Averin, S.; Gissot, S.; Giordanengo, B.; Berger, G.; Kroth, U.; De Jaeger, J.-C.; Gottwald, A.

    2013-10-01

    For next generation spaceborne solar ultraviolet radiometers, innovative metal-semiconductor-metal detectors based on wurtzite aluminum nitride are being developed and characterized. A set of measurement campaigns and proton irradiation damage tests was carried out to obtain their ultraviolet-to-visible characterization and degradation mechanisms. First results on large area prototypes up to 4.3 mm diameter are presented here. In the wavelength range of interest, this detector is reasonably sensitive and stable under brief irradiation with a negligible low dark current (3-6 pA/cm2). No significant degradation of the detector performance was observed after exposure to protons of 14.4 MeV energy, showing a good radiation tolerance up to fluences of 1 × 1011 protons/cm2.

  20. Further Charpy impact test results of low activation ferritic alloys, irradiated at 430{degrees}C to 67 dpa

    SciTech Connect

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

    1997-04-01

    Miniature CVN specimens of four ferritic alloys, GA3X, F82H, GA4X and HT9, have been impact tested following irradiation at 430{degrees}C to 67 dpa. Comparison of the results with those of the previously tested lower dose irradiation condition indicates that the GA3X and F82H alloys, two primary candidate low activation alloys, exhibit virtually identical behavior following irradiation at 430{degrees}C to {approximately}67 dpa and at 370{degrees}C to {approximately}15 dpa. Very little shift is observed in either DBTT or USE relative to the unirradiated condition. The shifts in DBTT and USE observed in both GA4X and HT9 were smaller after irradiation at 430{degrees}C to {approximately}67 dpa than after irradiation at 370{degrees}C to {approximately}15 dpa.

  1. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    SciTech Connect

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  2. Four-point Bend Testing of Irradiated Monolithic U-10Mo Fuel

    SciTech Connect

    Rabin, B. H.; Lloyd, W. R.; Schulthess, J. L.; Wright, J. K.; Lind, R. P.; Scott, L.; Wachs, K. M.

    2015-03-01

    This paper presents results of recently completed studies aimed at characterizing the mechanical properties of irradiated U-10Mo fuel in support of monolithic base fuel qualification. Mechanical properties were evaluated in four-point bending. Specimens were taken from fuel plates irradiated in the RERTR-12 and AFIP-6 Mk. II irradiation campaigns, and tests were conducted in the Hot Fuel Examination Facility (HFEF) at Idaho National Laboratory (INL). The monolithic fuel plates consist of a U-10Mo fuel meat covered with a Zr diffusion barrier layer fabricated by co-rolling, clad in 6061 Al using a hot isostatic press (HIP) bonding process. Specimens exhibited nominal (fresh) fuel meat thickness ranging from 0.25 mm to 0.64 mm, and fuel plate average burnup ranged from approximately 0.4 x 1021 fissions/cm3 to 6.0 x 1021 fissions/cm3. After sectioning the fuel plates, the 6061 Al cladding was removed by dissolution in concentrated NaOH. Pre- and post-dissolution dimensional inspections were conducted on test specimens to facilitate accurate analysis of bend test results. Four-point bend testing was conducted on the HFEF Remote Load Frame at a crosshead speed of 0.1 mm/min using custom-designed test fixtures and calibrated load cells. All specimens exhibited substantially linear elastic behavior and failed in a brittle manner. The influence of burnup on the observed slope of the stress-strain curve and the calculated fracture strength is discussed.

  3. Irradiation performance of Fast Flux Test Facility drivers using D9 alloy

    SciTech Connect

    Pitner, A.L.; Gneiting, B.C.; Bard, F.E.

    1995-11-01

    In comparison with the Fast Flux Test Facility Type 316 stainless steel driver design, six test assemblies employing D9 alloy in place of stainless steel for duct, cladding, and wire wrap material were irradiated to demonstrate the improved performance and lifetime capability of an advanced D9 alloy driver design. A single pinhole-type breach occurred in one of the high-exposure tests after a peak fuel burnup of 155 MWd/kg metal (M) and peak fast neutron fluence of 25 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV). Postirradiation examinations were performed on four of the test assemblies and measured results were compared with analytical evaluations. A revised swelling correlation for D9 alloy was developed to provide improved agreement between calculated and measured cladding deformation results. A fuel pin lifetime design criterion of 5% calculated hoop strain was derived from these results. Alternatively, fuel pin lifetimes were developed for two irradiation parameters using statistical failure analyses. For a 99.99% reliability, the analyses indicated a peak fast-fluence lifetime of 21.0 {times} 10{sup 22} n/cm{sup 2}, or a peak fuel burnup >120 MWd/kg M. In comparison with the Fast Flux Test Facility reference driver design, the extended lifetime capability of D9 alloy would reduce fuel supply requirements for the liquid-metal reactor by a third.

  4. Status of the NGNP graphite creep experiments AGC-1 and AGC-2 irradiated in the advanced test reactor

    SciTech Connect

    S. Blaine Grover

    2014-05-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the next generation nuclear plant (NGNP) very high temperature gas reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have three different compressive loads applied to the top half of three diametrically opposite pairs of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment.

  5. Mechanical behavior of AISI 304SS determined by miniature test methods after neutron irradiation to 28 dpa

    SciTech Connect

    Ellen M. Rabenberg; Brian J. Jaques; Bulent H. Sencer; Frank A. Garner; Paula D. Freyer; Taira Okita; Darryl P. Butt

    2014-05-01

    The mechanical properties of AISI 304 stainless steel irradiated for over a decade in the Experimental Breeder Reactor (EBR-II) were measured using miniature mechanical testing methods. The shear punch method was used to evaluate the shear strengths of the neutron-irradiated steel and a correlation factor was empirically determined to predict its tensile strength. The strength of the stainless steel slightly decreased with increasing irradiation temperature, and significantly increased with increasing dose until it saturated above approximately 5 dpa. Ferromagnetic measurements were used to observe and deduce the effects of the stress-induced austenite to martensite transformation as a result of shear punch testing.

  6. Comparison of irradiated 15Kh2MFA material mechanical properties using conventional testing methods and innovative approach of small punch testing (SPT) and automated ball indentation (ABIT)

    NASA Astrophysics Data System (ADS)

    Kopriva, R.; Petelova, P.; Eliasova, I.; Kytka, M.; Culek, M.

    2017-02-01

    Article describes two innovative testing methods – Small Punch Testing (SPT) and Automated Ball Indentation Test (ABIT) – which are based on the determination and evaluation of material properties from miniaturized testing specimens. These methods are very promising due to minimum material needed for testing and also in case of testing highly irradiated materials of components that are not included in standard surveillance programs. The test results were obtained for reactor pressure vessel (RPV) base material 15Ch2MFA in both states - initial unirradiated and irradiated. Subsequently results were compared with standard tensile tests to prove applicability of these testing methods for the evaluation of degradation of irradiated structural materials of nuclear power plants.

  7. Results of crack-arrest tests on irradiated a 508 class 3 steel

    SciTech Connect

    Iskander, S.K.; Milella, P.P.; Pini, M.A.

    1998-02-01

    Ten crack-arrest toughness values for irradiated specimens of A 508 class 3 forging steel have been obtained. The tests were performed according to the American Society for Testing and Materials (ASTM) Standard Test Method for Determining Plane-Strain Crack-Arrest Fracture Toughness, K{sub la} of Ferritic Steels, E 1221-88. None of these values are strictly valid in all five ASTM E 1221-88 validity criteria. However, they are useful when compared to unirradiated crack-arrest specimen toughness values since they show the small (averaging approximately 10{degrees}C) shifts in the mean and lower-bound crack-arrest toughness curves. This confirms that a low copper content in ASTM A 508 class 3 forging material can be expected to result in small shifts of the transition toughness curve. The shifts due to neutron irradiation of the lower bound and mean toughness curves are approximately the same as the Charpy V-notch (CVN) 41-J temperature shift. The nine crack-arrest specimens were irradiated at temperatures varying from 243 to 280{degrees}C, and to a fluence varying from 1.7 to 2.7 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV). The test results were normalized to reference values that correspond to those of CVN specimens irradiated at 284{degrees}C to a fluence of 3.2 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) in the same capsule as the crack-arrest specimens. This adjustment resulted in a shift to lower temperatures of all the data, and in particular moved two data points that appeared to lie close to or lower than the American Society of Mechanical Engineers K{sub la} curve to positions that seemed more reasonable with respect to the remaining data. A special fixture was designed, fabricated, and successfully used in the testing. For reasons explained in the text, special blocks to receive the Oak Ridge National Laboratory clip gage were designed, and greater-than-standard crack-mouth opening displacements measured were accounted for. 24 refs., 13 figs., 12 tabs.

  8. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2009-05-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any

  9. Spherical fuel elements for advanced HTR manufacture and qualification by irradiation testing

    NASA Astrophysics Data System (ADS)

    Mehner, A.-W.; Heit, W.; Röllig, K.; Ragoss, H.; Müller, H.

    1990-04-01

    The reference fuel cycle for future pebble bed HTRs uses low enriched uranium fuel. The spherical fuel element for these HTRs is a 60 mm diameter sphere containing TRISO-coated particles with UO 2 kernels. Qualification of this fuel was performed by production and quality control experience, irradiation testing and accident simulation experiments. The results of the qualification programme fully support the new safety concepts of advanced HTR designs. Further work concentrates on consolidating performance data sets and on quantifying the endurance limits of reference fuel elements under normal and accident conditions.

  10. New electron beam facility for irradiated plasma facing materials testing in hot cell

    SciTech Connect

    Sakamoto, N.; Kawamura, H.; Akiba, M.

    1995-09-01

    Since plasma facing components such as the first wall and the divertor for the next step fusion reactors are exposed to high heat loads and high energy neutron flux generated by the plasma, it is urgent to develop of plasma facing components which can resist these. Then, we have established electron beam heat facility ({open_quotes}OHBIS{close_quotes}, Oarai Hot-cell electron Beam Irradiating System) at a hot cell in JMTR (Japan Materials Testing Reactor) hot laboratory in order to estimate thermal shock resistivity of plasma facing materials and heat removal capabilities of divertor elements under steady state heating. In this facility, irradiated plasma facing materials (beryllium, carbon based materials and so on) and divertor elements can be treated. This facility consists of an electron beam unit with the maximum beam power of 50kW and the vacuum vessel. The acceleration voltage and the maximum beam current are 30kV (constant) and 1.7A, respectively. The loading time of electron beam is more than 0.1ms. The shape of vacuum vessel is cylindrical, and the mainly dimensions are 500mm in inner diameter, 1000mm in height. The ultimate vacuum of this vessel is 1 x 10{sup -4}Pa. At present, the facility for thermal shock test has been established in a hot cell. And performance estimation on the electron beam is being conducted. Presently, the devices for heat loading tests under steady state will be added to this facility.

  11. Irradiation performance of Fast Flux Test Facility drivers using D9 alloy

    SciTech Connect

    Pitner, A.L.; Gneiting, B.C.; Bard, F.E.

    1994-06-01

    Six test assemblies similar in design to the FFTF driver assembly but employing the advanced alloy D9 in place of Type 316 stainless steel for duct, cladding, and wire wrap material were irradiated to demonstrate the improved performance and lifetime capability of this design. A single pinhole-type breach was incurred in one of the high exposure tests after a peak fuel burnup of 155 MWd/kgM and peak fast neutron fluence of 25 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV). Postirradiation examinations were performed on four of the test assemblies and measured results were compared to analytical evaluations. A revised swelling correlation for D9 Alloy was developed to provide improved agreement between calculated and measured cladding deformation results. A fuel pin lifetime design criterion of 5% calculated hoop strain was derived. Alternatively, fuel pin lifetimes were developed for two irradiation parameters using statistical failure analyses. For a 99.99% reliability, the analyses indicated a peak fast fluence lifetime of 21.0 {times} 10{sup 22} n/cm{sup 2}, or a peak fuel burnup greater than 120 MWd/kgM. The extended lifetime capability of this design would reduce fuel supply requirements for the FFTF by a third relative to the reference driver design.

  12. Irradiation testing of full-sized, reduced-enrichment fuel elements

    SciTech Connect

    Snelgrove, J.L.; Copeland, G.L.

    1983-01-01

    The current status of the irradiation testing of full-sized, reduced-enrichment fuel elements and fuel rods under the US Reduced Enrichment Research and Test Reactor Program is reported. Being tested are UAl/sub x/-Al, U/sub 3/O/sub 8/-Al, U/sub 3/Si/sub 2/-Al, and U/sub 3/Si-Al dispersion fuels and UZrH/sub x/ (TRIGA) fuel at uranium densities in the fuel meat ranging from 1.7 to 6.0 Mg/m/sup 3/. Generally good performance has been experienced to date. Some preliminary results of postirradiation examinations are also included. A whole-core demonstration in the Oak Ridge Research Reactor is planned. Some details of this demonstration are provided.

  13. Autologous Transplantation of Adult Mice Spermatogonial Stem Cells into Gamma Irradiated Testes

    PubMed Central

    Koruji, Morteza; Movahedin, Mansoureh; Mowla, Seyed Javad; Gourabi, Hamid; Pour-Beiranvand, Shahram; Jabbari Arfaee, Ali

    2012-01-01

    Objective: We evaluated structural and functional changes of fresh and frozen-thawed adult mouse spermatogonial stem cells following auto-transplantation into gamma-irradiated testes. Materials and Methods: In this experimental research, the right testes from adult mice (n=25) were collected, then Sertoli and spermatogonial cells were isolated using two-step enzymatic digestion, lectin immobilization and differential plating. Three weeks after cultivation, the Bromodeoxyuridine (BrdU)-labeled spermatogonial cells were transplanted, via rete testis, into the other testis of the same mouse, which had been irradiated with 14Gy. The mice were transplanted with: fresh cells (control 1), fresh cells co-cultured with Sertoli cells (control 2), the frozen-thawed cells (experimental 1) and frozen-thawed cells co-cultured with Sertoli cells (experimental 2). The morphological changes between different transplanted testes groups were compared in 8 weeks after transplantation. The statistical significance between mean values was determined by Kruskal Wallis and one-way analysis of variance in efficiency of transplantation. Results: The statistical analysis revealed significant increases in the mean percentage of testis weight and normal seminiferous tubules following spermatogonial stem cells transplantation in the recipient'fs testes. The normal seminiferous tubules percentage in the co-culture system with fresh cells and frozen-thawed groups were more than those in non-transplanted and fresh cell transplanted groups (p≤0.001). Conclusion: Our results demonstrated that spermatogonial stem cells in the colonies could result sperm production in the recipient’s testes after autologous transplantation. PMID:23507977

  14. Irradiation and compatibility testing of Li/sub 2/O materials at EBR-II

    SciTech Connect

    Porter, D.L.; Krsul, J.R.; Laug, M.T.; Tetenbaum, M.; Walters, L.C.

    1982-12-01

    A study was made of the neutron-irradiation behavior of /sup 6/Li-enriched Li/sub 2/O material in EBR-II. In addition, a stress corrosion study was performed ex-reactor to test compatibility of Li/sub 2/O materials with a variety of stainless steels. Results of the irradiation testing showed that tritium and helium retention in the Li/sub 2/O (approx. 89% dense) lessened with neutron exposure. Helium tritium retention appeared to approach steady-state after approx. 1% /sup 6/Li burnup. The effect was likely caused by the formation of open porosity in the pellets. The stress corrosion studies, using a 316 stainless steel (Ti-modified) and a 35% Ni alloy, showed that stress does not enhance the corrosion, and that dry Li/sub 2/O is not significantly corrosive, the LiOH content producing the corrosive effects. Corrosion, in general, was not severe as a passivation in sealed capsules seemed to occur after a time greatly reducing corrosion rates.

  15. 3D-FBK Pixel Sensors: Recent Beam Tests Results with Irradiated Devices

    SciTech Connect

    Micelli, A.; Helle, K.; Sandaker, H.; Stugu, B.; Barbero, M.; Hugging, F.; Karagounis, M.; Kostyukhin, V.; Kruger, H.; Tsung, J.W.; Wermes, N.; Capua, M.; Fazio, S.; Mastroberardino, A.; Susinno, G.; Gallrapp, C.; Di Girolamo, B.; Dobos, D.; La Rosa, A.; Pernegger, H.; Roe, S.; /CERN /Prague, Tech. U. /Prague, Tech. U. /Freiburg U. /Freiburg U. /Freiburg U. /INFN, Genoa /Genoa U. /INFN, Genoa /Genoa U. /INFN, Genoa /Genoa U. /INFN, Genoa /Genoa U. /INFN, Genoa /Genoa U. /Glasgow U. /Glasgow U. /Glasgow U. /Hawaii U. /Barcelona, IFAE /Barcelona, IFAE /LBL, Berkeley /Barcelona, IFAE /LBL, Berkeley /LBL, Berkeley /Manchester U. /Manchester U. /Manchester U. /Manchester U. /Manchester U. /Manchester U. /Manchester U. /Manchester U. /Manchester U. /New Mexico U. /New Mexico U. /Oslo U. /Oslo U. /Oslo U. /Oslo U. /Oslo U. /SLAC /SLAC /SLAC /SLAC /SLAC /SLAC /SLAC /SLAC /SLAC /SUNY, Stony Brook /SUNY, Stony Brook /SUNY, Stony Brook /INFN, Trento /Trento U. /INFN, Trento /Trento U. /INFN, Trento /Trento U. /INFN, Trieste /Udine U. /INFN, Trieste /Udine U. /INFN, Trieste /Udine U. /INFN, Trieste /Udine U. /INFN, Trieste /Udine U. /INFN, Trieste /Udine U. /Barcelona, Inst. Microelectron. /Barcelona, Inst. Microelectron. /Barcelona, Inst. Microelectron. /Fond. Bruno Kessler, Trento /Fond. Bruno Kessler, Trento /Fond. Bruno Kessler, Trento /Fond. Bruno Kessler, Trento /Fond. Bruno Kessler, Trento /SINTEF, Oslo /SINTEF, Oslo /SINTEF, Oslo /SINTEF, Oslo /VTT Electronics, Espoo /VTT Electronics, Espoo

    2012-04-30

    The Pixel Detector is the innermost part of the ATLAS experiment tracking device at the Large Hadron Collider, and plays a key role in the reconstruction of the primary vertices from the collisions and secondary vertices produced by short-lived particles. To cope with the high level of radiation produced during the collider operation, it is planned to add to the present three layers of silicon pixel sensors which constitute the Pixel Detector, an additional layer (Insertable B-Layer, or IBL) of sensors. 3D silicon sensors are one of the technologies which are under study for the IBL. 3D silicon technology is an innovative combination of very-large-scale integration and Micro-Electro-Mechanical-Systems where electrodes are fabricated inside the silicon bulk instead of being implanted on the wafer surfaces. 3D sensors, with electrodes fully or partially penetrating the silicon substrate, are currently fabricated at different processing facilities in Europe and USA. This paper reports on the 2010 June beam test results for irradiated 3D devices produced at FBK (Trento, Italy). The performance of these devices, all bump-bonded with the ATLAS pixel FE-I3 read-out chip, is compared to that observed before irradiation in a previous beam test.

  16. Fabrication Control Plan for ORNL RH-LOCA ATF Test Specimens to be Irradiated in the ATR

    SciTech Connect

    Field, Kevin G.; Howard, Richard; Teague, Michael

    2014-06-01

    The purpose of this fabrication plan is (1) to summarize the design of a set of rodlets that will be fabricated and then irradiated in the Advanced Test Reactor (ATR) and (2) provide requirements for fabrication and acceptance criteria for inspections of the Light Water Reactor (LWR) – Accident Tolerant Fuels (ATF) rodlet components. The functional and operational (F&OR) requirements for the ATF program are identified in the ATF Test Plan. The scope of this document only covers fabrication and inspections of rodlet components detailed in drawings 604496 and 604497. It does not cover the assembly of these items to form a completed test irradiation assembly or the inspection of the final assembly, which will be included in a separate INL final test assembly specification/inspection document. The controls support the requirements that the test irradiations must be performed safely and that subsequent examinations must provide valid results.

  17. Effects of sterilizing gamma irradiation on bloodspot newborn screening tests and whole blood cyclosporine and tacrolimus measurements.

    PubMed

    Stickle, Douglas F; McKenzie, David A; Landmark, James D; Pirruccello, Samuel J; Post, Gregory R; Iwen, Peter C; Thompson, Robert B; Hinrichs, Steven H

    2003-02-01

    Sterilizing irradiation of the US mail has been proposed as a method to prevent delivery of viable anthrax spores. Because newborn screening samples (bloodspots) and cyclosporine and tacrolimus specimens (whole blood) are delivered routinely through the mail, we studied whether sterilizing gamma irradiation could affect these test results. Specimens were exposed to 18 kGy gamma irradiation (100 hours x 18,000 rad/h), a "kill dose" for Bacillus pumilus spore strips. Irradiation had no significant effect on whole blood cyclosporine or tacrolimus results, but it had a degradative effect on bloodspot phenylalanine, hemoglobins, biotinidase, galactose-1-phosphate uridyltransferase, thyroxine, and thyrotropin. Such irradiation potentially could cause false-negative results for the detection of phenylketonuria and likely would lead to an increase in secondary testing for hemoglobin variants, but it is unlikely to lead to false-negative or false-positive results for the remaining newborn screening tests. These experiments cannot rule out possible greater effects by larger doses or other types of irradiation.

  18. Crack-arrest tests on two irradiated high-copper welds. Phase 2: Results of duplex-type experiments

    SciTech Connect

    Iskander, S.K.; Corwin, W.R.; Nanstad, R.K.

    1994-03-01

    The objective of the Heavy-Section Steel Irradiation Program Sixth Irradiation Series is to determine the effect of neutron irradiation on the shift and shape of the lower-bound curve to crack-arrest toughness data. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Crack-arrest specimens fabricated from these welds were irradiated at a nominal temperature of 288{degrees}C to an average fluence of 1.9 {times} 10{sup 19} neutrons/cm{sup 2} (>1 MeV). This is the second report giving the results of the tests on irradiated duplex-type crack-arrest specimens. A previous report gave results of tests on irradiated weld-embrittled-type specimens. Charpy V-notch (CVN) specimens irradiated in the same capsules as the crack-arrest specimens were also tested, and a 41-J transition temperature shift was determined from these specimens. {open_quotes}Mean{close_quote} curves of the same form as the American Society of Mechanical Engineers (ASME) K{sub la} curve were fit to the data with only the {open_quotes}reference temperature{close_quotes} as a parameter. The shift between the mean curves agrees well with the 41-J transition temperature shift obtained from the CVN specimen tests. Moreover, the four data points resulting from tests on the duplex crack-arrest specimens of the present study did not make a significant change to mean curve fits to either the previously obtained data or all the data combined.

  19. Steady-state irradiation testing of U-Pu-Zr fuel to >18% burnup

    SciTech Connect

    Pahl, R.G.; Wisner, R.S. ); Billone, M.C.; Hofman, G.L. )

    1990-01-01

    Tests of austenitic stainless steel clad U-xP-10Zr fuel (x=o, 8, 19 wt. %) to peak burnups as high as 18.4 at. % have been completed in the EBR-II. Fuel swelling and fractional fission gas release are slowly increasing functions of burnup beyond 2 at. % burnup. Increasing plutonium content in the fuel reduces swelling and decreases the amount of fission gas which diffuses from fuel to plenum. LIFE-METAL code modelling of cladding strains is consistent with creep by fission gas loading and irradiation-induced swelling mechanisms. Fuel/cladding chemical interaction involves the ingress of rare-earth fission products. Constituent redistribution in the fuel had not limited steady-state performance. Cladding breach behavior at closure welds, in the gas plenum, and in the fuel column region have been benign events. 3 refs., 5 figs.

  20. Comparative effects of X irradiation on the testes of adult Sprague-Dawley and Wistar rats.

    PubMed

    Delic, J I; Schlappack, O K; Harwood, J R; Stanley, J A

    1987-10-01

    The response of the testes of two strains of adult rats (Sprague-Dawley and Wistar) to graded single doses and split doses of 230 kVp X rays has been investigated. A marked difference was noted between the strains in the response of the clonogenic spermatogonia to irradiation, as measured histologically by the repopulation index. Single-dose response curves derived for these cells in the Sprague-Dawley strain had a much larger shoulder (up to about 4-5 Gy) than for the Wistar (less than 2 Gy). Split-dose studies revealed that this difference may partly be explained by a greater repair capacity in the cells of the Sprague-Dawley strain. Changes in serum FSH concentrations mirrored the changes in clonogenic spermatogonial survival following split doses of radiation.

  1. AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

    SciTech Connect

    Stempien, John Dennis; Rice, Francine Joyce; Harp, Jason Michael; Winston, Philip Lon

    2016-03-01

    The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite

  2. High intensity solar irradiation testing of UV optics. [OSO-8 instruments

    NASA Technical Reports Server (NTRS)

    Greyerbiehl, J. M.; Oberright, J. E.

    1976-01-01

    The Orbiting Solar Observatory-I (OSO-8 in orbit) incorporates two high resolution solar pointing spectrometers operating from 1000 A to 4000 A. Energy from the sun enters a Cassegrainian telescope and is focused on a slit while the solar disk is scanned to one arc-second resolution. The stability of the secondary mirrors reflectance was of concern since they would be exposed to intense focused solar energy up to 27 suns. A test program was initiated to simulate this energy input on sample UV mirrors of the MgF2 and LiF types and to evaluate their performance after irradiation. Tests were conducted to simulate the solar spectrum at high intensities (25 suns) and at a single wavelength near Lyman-alpha, but with twenty times the solar intensity at Lyman-alpha. Post-test measurements after every exposure were made at wavelengths from 1025 A to 1849 A. After 75 simulated 'orbits', reflectance changes due to temperature effects were noted to be less than 10%. Reductions in reflectance under high intensity solar radiation were generally greater than 10%. Polymerization of surface contaminations on the LiF mirrors reduced reflectances at short wavelengths by 40%.

  3. First Tests for the Detection of the LINAC Irradiation Field Using PIN Diodes

    SciTech Connect

    Nava, C. E. Ojeda; Ramirez-Jimenez, F. J.; Navarro, L. F. Villasenor; Cruz, M. Duran

    2008-08-11

    The employment of the technology of semiconductor detectors, in the medical physics environment is of great importance due to its versatility and dependability. In this work we present the first results and the experimental arrangement employed with PIN diodes that are conditioned for the measurement of the field of irradiation of a lineal accelerator (LINAC) used in radiotherapy. In our tests we used a PIN photodiode. In former experiments, this diode presented a response to the intensity of the applied field when it was exposed to an X-ray beam in medical and industrial radiography equipments. This diode is a low cost and easy acquisition one in the field. These characteristics transform it into a serious candidate as detector to be used in electronic arrangements for the detection of radiation fields in radio-therapy with X-rays. Experiments were designed to obtain the response of this diode when it was exposed to X-ray beams of a LINAC used in radiotherapy. Firstly the tests were carried out for a 6 MeV photon beam with a source to surface distance (SSD) of 100 cm, obtaining very encouraging results. We seek to carry out tests for more energy values in order to obtain the energy response of this detector as a radiation sensor device. This device could be applied in the design of working tools, for example, for the quality control in procedures of radiotherapy.

  4. MOX capsule post-irradiation examination. Volume 2: Test plan for 30-GWd/MT burnup fuel

    SciTech Connect

    Morris, R.N.

    1997-12-01

    This test plan is a Level-2 document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-Oxide Fuel Irradiation Test Project Plan. The planned post-irradiation examination (PIE) work to be performed on the mixed uranium and plutonium oxide fuel capsules that have received burnups of approximately 30 GWd/MT is described. The major emphasis of this PIE task will be material interactions, particularly indications of gallium transport and interactions. This PIE will include gamma scanning, ceramography, metallography, pellet radial gallium analysis, and clad gallium analysis. A preliminary PIE report will be generated before all the work is completed so that the progress of the fuel irradiation may be known in a timely manner.

  5. Test results of an ITER relevant FPGA when irradiated with neutrons

    SciTech Connect

    Batista, Antonio J. N.; Santos, Bruno; Fernandes, Ana; Goncalves, Bruno; Leong, Carlos; Teixeira, Joao P.; Ramos, Ana Rita; Santos, Joana P.; Marques, Jose G.

    2015-07-01

    The data acquisition and control instrumentation cubicles room of the ITER tokamak will be irradiated with neutrons during the fusion reactor operation. A Virtex-6 FPGA from Xilinx (XC6VLX365T-1FFG1156C) is used on the ATCA-IO-PROCESSOR board, included in the ITER Catalog of I and C products - Fast Controllers. The Virtex-6 is a re-programmable logic device where the configuration is stored in Static RAM (SRAM), functional data stored in dedicated Block RAM (BRAM) and functional state logic in Flip-Flops. Single Event Upsets (SEU) due to the ionizing radiation of neutrons causes soft errors, unintended changes (bit-flips) to the values stored in state elements of the FPGA. The SEU monitoring and soft errors repairing, when possible, were explored in this work. An FPGA built-in Soft Error Mitigation (SEM) controller detects and corrects soft errors in the FPGA configuration memory. Novel SEU sensors with Error Correction Code (ECC) detect and repair the BRAM memories. Proper management of SEU can increase reliability and availability of control instrumentation hardware for nuclear applications. The results of the tests performed using the SEM controller and the BRAM SEU sensors are presented for a Virtex-6 FPGA (XC6VLX240T-1FFG1156C) when irradiated with neutrons from the Portuguese Research Reactor (RPI), a 1 MW nuclear fission reactor operated by IST in the neighborhood of Lisbon. Results show that the proposed SEU mitigation technique is able to repair the majority of the detected SEU errors in the configuration and BRAM memories. (authors)

  6. Fabrication of U-10 wt.%Zr Metallic Fuel Rodlets for Irradiation Test in BOR-60 Fast Reactor

    DOE PAGES

    Kim, Ki-Hwan; Kim, Jong-Hwan; Oh, Seok-Jin; ...

    2016-01-01

    The fabrication technology for metallic fuel has been developed to produce the driver fuel in a PGSFR in Korea since 2007. In order to evaluate the irradiation integrity and validate the in-reactor of the starting metallic fuel with FMS cladding for the loading of the metallic fuel, U-10 wt.%Zr fuel rodlets were fabricated and evaluated for a verification of the starting driver fuel through an irradiation test in the BOR-60 fast reactor. The injection casting method was applied to U-10 wt.%Zr fuel slugs with a diameter of 5.5 mm. Consequently, fuel slugs per melting batch without casting defects were fabricated through the developmentmore » of advanced casting technology and evaluation tests. The optimal GTAW welding conditions were also established through a number of experiments. In addition, a qualification test was carried out to prove the weld quality of the end plug welding of the metallic fuel rodlets. The wire wrapping of metallic fuel rodlets was successfully accomplished for the irradiation test. Thus, PGSFR fuel rodlets have been soundly fabricated for the irradiation test in a BOR-60 fast reactor.« less

  7. High-heat-flux testing of irradiated tungsten-based materials for fusion applications using infrared plasma arc lamps

    SciTech Connect

    Sabau, Adrian S.; Ohriner, Evan K.; Kiggans, Jim; Schaich, Charles R.; Ueda, Yoshio; Harper, David C.; Katoh, Yutai; Snead, Lance L.; Byun, Thak S.

    2014-11-01

    Testing of advanced materials and component mock-ups under prototypical fusion high-heat-flux conditions, while historically a mainstay of fusion research, has proved to be quite challenging, especially for irradiated materials. A new high-heat-flux–testing (HHFT) facility based on water-wall plasma arc lamps (PALs) is now introduced for materials and small-component testing. Two PAL systems, utilizing a 12 000°C plasma arc contained in a quartz tube cooled by a spiral water flow over the inside tube surface, provide maximum incident heat fluxes of 4.2 and 27 MW/m2 over areas of 9×12 and 1×10 cm2, respectively. This paper will present the overall design and implementation of a PAL-based irradiated material target station (IMTS). The IMTS is primarily designed for testing the effects of heat flux or thermal cycling on material coupons of interest, such as those for plasma-facing components. Temperature results are shown for thermal cycling under HHFT of tungsten coupon specimens that were neutron irradiated in HFIR. Finally, radiological surveys indicated minimal contamination of the 36×36×18 cm test section, demonstrating the capability of the new facility to handle irradiated specimens at high temperature.

  8. High-heat-flux testing of irradiated tungsten-based materials for fusion applications using infrared plasma arc lamps

    DOE PAGES

    Sabau, Adrian S.; Ohriner, Evan K.; Kiggans, Jim; ...

    2014-11-01

    Testing of advanced materials and component mock-ups under prototypical fusion high-heat-flux conditions, while historically a mainstay of fusion research, has proved to be quite challenging, especially for irradiated materials. A new high-heat-flux–testing (HHFT) facility based on water-wall plasma arc lamps (PALs) is now introduced for materials and small-component testing. Two PAL systems, utilizing a 12 000°C plasma arc contained in a quartz tube cooled by a spiral water flow over the inside tube surface, provide maximum incident heat fluxes of 4.2 and 27 MW/m2 over areas of 9×12 and 1×10 cm2, respectively. This paper will present the overall design andmore » implementation of a PAL-based irradiated material target station (IMTS). The IMTS is primarily designed for testing the effects of heat flux or thermal cycling on material coupons of interest, such as those for plasma-facing components. Temperature results are shown for thermal cycling under HHFT of tungsten coupon specimens that were neutron irradiated in HFIR. Finally, radiological surveys indicated minimal contamination of the 36×36×18 cm test section, demonstrating the capability of the new facility to handle irradiated specimens at high temperature.« less

  9. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    SciTech Connect

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    2014-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

  10. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    SciTech Connect

    J. L. Rempe; D. L. Knudson; J. E. Daw

    2011-03-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  11. Small punch tests on martensitic/ferritic steels F82H, T91 and Optimax-A irradiated in SINQ Target-3

    NASA Astrophysics Data System (ADS)

    Jia, X.; Dai, Y.

    2003-12-01

    Small punch (SP) tests were conducted in a temperature range from -190 to 80 °C on martensitic/ferritic steels F82H, T91 and Optimax-A irradiated in SINQ Target-3 up to 9.4 dpa in a irradiation temperature range of 90-275 °C. Results demonstrate: (a) the irradiation hardening deduced from SP tests is reasonably consistent with the results obtained by tensile tests; (b) with increasing irradiation dose, the SP yield load increases at all test temperatures, while the displacement at the maximum load and the total displacement at failure decrease; (c) the ductile-to-brittle transition temperature (DBTT SP) increases with increasing irradiation dose, and does so more quickly at irradiation doses above ˜6-7 dpa; in addition, the ΔDBTT SP increases linearly with helium content.

  12. A proton irradiation test facility for space research in Ankara, Turkey

    NASA Astrophysics Data System (ADS)

    Gencer, Ayşenur; Yiǧitoǧlu, Merve; Bilge Demirköz, Melahat; Efthymiopoulos, Ilias

    2016-07-01

    Space radiation often affects the electronic components' performance during the mission duration. In order to ensure reliable performance, the components must be tested to at least the expected dose that will be received in space, before the mission. Accelerator facilities are widely used for such irradiation tests around the world. Turkish Atomic Energy Authority (TAEA) has a 15MeV to 30MeV variable proton cyclotron in Ankara and the facility's main purpose is to produce radioisotopes in three different rooms for different target systems. There is also an R&D room which can be used for research purposes. This paper will detail the design and current state of the construction of a beamline to perform Single Event Effect (SEE) tests in Ankara for the first time. ESA ESCC No.25100 Standard Single Event Effect Test Method and Guidelines is being considered for these SEE tests. The proton beam kinetic energy must be between 20MeV and 200MeV according to the standard. While the proton energy is suitable for SEE tests, the beam size must be 15.40cm x 21.55cm and the flux must be between 10 ^{5} p/cm ^{2}/s to at least 10 ^{8} p/cm ^{2}/s according to the standard. The beam size at the entrance of the R&D room is mm-sized and the current is variable between 10μA and 1.2mA. Therefore, a defocusing beam line has been designed to enlarge the beam size and reduce the flux value. The beam line has quadrupole magnets to enlarge the beam size and the collimators and scattering foils are used for flux reduction. This facility will provide proton fluxes between 10 ^{7} p/cm ^{2}/s and 10 ^{10} p/cm ^{2}/s for the area defined in the standard when completed. Also for testing solar cells developed for space, the proton beam energy will be lowered below 10MeV. This project has been funded by Ministry of Development in Turkey and the beam line construction will finish in two years and SEE tests will be performed for the first time in Turkey.

  13. On the (in)adequacy of the Charpy impact test to monitor irradiation effects of ferritic/martensitic steels

    NASA Astrophysics Data System (ADS)

    Chaouadi, R.

    2007-02-01

    Irradiation embrittlement studies rely very often on Charpy impact data, in particular the ductile-to-brittle transition temperature (DBTT). However, while the DBTT-shift is equivalent to the increase of the fracture toughness transition temperature of ferritic steels, it is not the case for ferritic/martensitic steels. The aim of this study is to critically assess experimental data obtained on a 9%Cr-ferritic/martensitic steel, Eurofer-97, to better understand the underlying mechanisms involved during the fracture process. More specifically, a dedicated analysis using the load diagram approach allows to unambiguously reveal the actual effects of irradiation on physically rather than empirically based parameters. A comparison is made between a ferritic and ferritic/martensitic steel to better identify the possible similarities and differences. Tensile, Charpy impact and fracture toughness tests data are examined in a global approach to assess the actual rather than apparent irradiation effects. The adequacy or inadequacy of the Charpy impact test to monitor irradiation effects is extensively discussed.

  14. Evaluation of the limulus amoebocyte lysate test in conjunction with a gram negative bacterial plate count for detecting irradiation of chicken

    NASA Astrophysics Data System (ADS)

    Scotter, Susan L.; Wood, Roger; McWeeny, David J.

    A study to evaluate the potential of the Limulus amoebocyte lysate (LAL) test in conjuction with a Gram negative bacteria (GNB) plate count for detecting the irradiation of chicken is described. Preliminary studies demonstrated that chickens irradiated at an absorbed dose of 2.5 kGy could be differentiated from unirradiated birds by measuring levels of endotoxin and of numbers of GNB on chicken skin. Irradiated birds were found to have endotoxin levels similar to those found in unirradiated birds but significantly lower numbers of GNB. In a limited study the test was found to be applicable to birds from different processors. The effect of temperature abuse on the microbiological profile, and thus the efficacy of the test, was also investigated. After temperature abuse, the irradiated birds were identifiable at worst up to 3 days after irradiation treatment at the 2.5 kGy level and at best some 13 days after irradiation. Temperature abuse at 15°C resulted in rapid recovery of surviving micro-organisms which made differentiation of irradiated and unirradiated birds using this test unreliable. The microbiological quality of the bird prior to irradiation treatment also affected the test as large numbers of GNB present on the bird prior to irradiation treatment resulted in larger numbers of survivors. In addition, monitoring the developing flora after irradiation treatment and during subsequent chilled storage also aided differentiation of irradiated and unirradiated birds. Large numbers of yeasts and Gram positive cocci were isolated from irradiated carcasses whereas Gram negative oxidative rods were the predominant spoilage flora on unirradiated birds.

  15. Test irradiations of full-sized U 3Si 2-Al fuel plates up to very high fission densities

    NASA Astrophysics Data System (ADS)

    Böning, K.; Petry, W.

    2009-01-01

    In the course of the licensing procedure of the 'Forschungsneutronenquelle Heinz Maier-Leibnitz', i.e. the new 20 MW high-flux research reactor FRM II in Garching near Munich, extensive test irradiations have been performed to qualify the U 3Si 2-Al dispersion fuel with a relatively high density of highly enriched uranium (93 wt% of 235U) up to very high fission densities. Two of the three FRM II type fuel plates used in the irradiation tests contained U 3Si 2-Al dispersion fuel with HEU densities of 3.0 gU/cm 3 or 1.5 gU/cm 3 ('homogeneous plates') and one plate had two adjacent zones of either density ('mixed plate'). They were irradiated in the French MTR reactors SILOE and OSIRIS in the years before 2002. The local plate thickness was measured on certain tracks along the plates during interruptions of the irradiation. The maximum fission density obtained in the U 3Si 2 fuel particles was 1.4 × 10 22 f/cm 3 and 1.1 × 10 22 f/cm 3 in the 1.5 gU/cm 3 and 3.0 gU/cm 3 fuel zones, respectively. In the course of the irradiations, the plate thickness increased monotonically and approximately linearly, leading to a maximum plate thickness swelling of 14% and 21% and a corresponding volume increase of the fuel particles of 106% and 81%, respectively. Our results are discussed and compared with the data from the literature.

  16. Effect of gamma irradiation on reactivity of rinderpest virus antigen with bovine immune serum in enzyme-linked immunosorbent assays and virus neutralization and indirect fluorescent-antibody tests.

    PubMed Central

    Saliki, J T; Berninger, M L; Torres, A; House, J A; Mebus, C A; Dubovi, E J

    1993-01-01

    Gamma irradiation effectively inactivated gradient-purified rinderpest virus. Irradiated antigen and sera remained functional in enzyme-linked immunosorbent assays, virus neutralization tests, and indirect fluorescent-antibody tests. Irradiation, however, led to a dose-dependent decrease in reactivity, particularly significant (P < 0.05) when both reagents were irradiated. To avoid false-positive reactions, only one reagent (serum or antigen) may be irradiated. PMID:8432831

  17. PIREX II — A new irradiation facility for testing fusion first wall materials

    NASA Astrophysics Data System (ADS)

    Marmy, P.; Daum, M.; Gavillet, D.; Green, S.; Green, W. V.; Hegedus, F.; Proennecke, S.; Rohrer, U.; Stiefel, U.; Victoria, M.

    1990-03-01

    A new irradiation facility, PIREX II (Proton Irradiation Experiment), became operational in March 1987. It is located on a dedicated beam line split from the main beam of the 590 MeV proton accelerator at the Paul Scherrer Institute (PSI). Irradiation with protons of this energy introduces simultaneously displacement damage, helium and other impurities. Because of the penetration range of 590 MeV protons, both damage and impurities are homogeneously distributed in the target material. The installation has its own beam line optics that can support a proton current of up to 50 μA. At a typical beam density of 4 {μA}/{mm 2}, the damage rate in steel is 0.7 × 10 -5{dpa}/{s} (dpa: displacements per atom), and the helium production rat He/dpa. Both flat tensile specimens of up to 0.4 mm thickness and tubular fatigue samples of 3 mm diameter can be irradiated. Cooling of the sample is performed by flowing pressurized helium gas over the sample. Irradiation temperatures can be controlled between 100 ° C and 800 ° C. Installation of an in situ low cycle fatigue device is foreseen. Beams of up to 20 μA have been obtained, the beam having an approximately Gaussian distribution of elliptical cross section with 4σ xbetween 0.8 and 8 nun by 4σ y of up to 10 mm. Irradiations for a dosimetry program have been completed on samples of Al, Cu, Fe, Ni, Au, W, and 1.4914 ferritic steel. The evaluation of results allows the correct choice of reactions to be used for determining total dose, from the standpoint of half life and gamma energy. A program of irradiations on candidate materials for the Next European Torus (NET) design (Cu and Cu alloys, 1.4914 ferritic martensitic steel, W and W-Re alloys and Mo and Mo alloys), where the above mentioned characteristics of this type of irradiation can be used advantageously, is now under way.

  18. Critical processes and parameters in the development of accident tolerant fuels drop-in capsule irradiation tests

    SciTech Connect

    Barrett, K. E.; Ellis, K. D.; Glass, C. R.; Roth, G. A.; Teague, M. P.; Johns, J.

    2015-12-01

    The goal of the Accident Tolerant Fuel (ATF) program is to develop the next generation of Light Water Reactor (LWR) fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. An irradiation test series has been defined to assess the performance of proposed ATF concepts under normal LWR operating conditions. The Phase I ATF irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL). Design, analysis, and fabrication processes for ATR drop-in capsule experiment preparation are presented in this paper to demonstrate the importance of special design considerations, parameter sensitivity analysis, and precise fabrication and inspection techniques for figure innovative materials used in ATF experiment assemblies. A Taylor Series Method sensitivity analysis approach was used to identify the most critical variables in cladding and rodlet stress, temperature, and pressure calculations for design analyses. The results showed that internal rodlet pressure calculations are most sensitive to the fission gas release rate uncertainty while temperature calculations are most sensitive to cladding I.D. and O.D. dimensional uncertainty. The analysis showed that stress calculations are most sensitive to rodlet internal pressure uncertainties, however the results also indicated that the inside radius, outside radius, and internal pressure were all magnified as they propagate through the stress equation. This study demonstrates the importance for ATF concept development teams to provide the fabricators as much information as possible about the material properties and behavior observed in prototype testing, mock-up fabrication and assembly, and chemical and mechanical testing of the materials that may have been performed in the concept development phase. Special

  19. Critical processes and parameters in the development of accident tolerant fuels drop-in capsule irradiation tests

    DOE PAGES

    Barrett, K. E.; Ellis, K. D.; Glass, C. R.; ...

    2015-12-01

    The goal of the Accident Tolerant Fuel (ATF) program is to develop the next generation of Light Water Reactor (LWR) fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. An irradiation test series has been defined to assess the performance of proposed ATF concepts under normal LWR operating conditions. The Phase I ATF irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL). Design, analysis, and fabrication processes formore » ATR drop-in capsule experiment preparation are presented in this paper to demonstrate the importance of special design considerations, parameter sensitivity analysis, and precise fabrication and inspection techniques for figure innovative materials used in ATF experiment assemblies. A Taylor Series Method sensitivity analysis approach was used to identify the most critical variables in cladding and rodlet stress, temperature, and pressure calculations for design analyses. The results showed that internal rodlet pressure calculations are most sensitive to the fission gas release rate uncertainty while temperature calculations are most sensitive to cladding I.D. and O.D. dimensional uncertainty. The analysis showed that stress calculations are most sensitive to rodlet internal pressure uncertainties, however the results also indicated that the inside radius, outside radius, and internal pressure were all magnified as they propagate through the stress equation. This study demonstrates the importance for ATF concept development teams to provide the fabricators as much information as possible about the material properties and behavior observed in prototype testing, mock-up fabrication and assembly, and chemical and mechanical testing of the materials that may have been performed in the concept development phase

  20. Initial tensile test results from J316 stainless steel irradiated in the HFIR spectrally tailored experiment

    SciTech Connect

    Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.

    1995-04-01

    The objective of this work is to determine the effects of neutron irradiation on the mechanical properties of austenitic stainless steel alloys. In this experiment, the spectrum has been tailored to reduce the thermal neutron flux and achieve a He/dpa level near that expected in a fusion reactor.

  1. Microbial analysis and survey test of gamma-irradiated freeze-dried fruits for patient's food

    NASA Astrophysics Data System (ADS)

    Park, Jae-Nam; Sung, Nak-Yun; Byun, Eui-Hong; Byun, Eui-Baek; Song, Beom-Seok; Kim, Jae-Hun; Lee, Kyung-A.; Son, Eun-Joo; Lyu, Eun-Soon

    2015-06-01

    This study examined the microbiological and organoleptic qualities of gamma-irradiated freeze-dried apples, pears, strawberries, pineapples, and grapes, and evaluated the organoleptic acceptability of the sterilized freeze-dried fruits for hospitalized patients. The freeze-dried fruits were gamma-irradiated at 0, 1, 2, 3, 4, 5, 10, 12, and 15 kGy, and their quality was evaluated. Microorganisms were not detected in apples after 1 kGy, in strawberries and pears after 4 kGy, in pineapples after 5 kGy, and in grapes after 12 kGy of gamma irradiation. The overall acceptance score, of the irradiated freeze-dried fruits on a 7-point scale at the sterilization doses was 5.5, 4.2, 4.0, 4.1, and 5.1 points for apples, strawberries, pears, pineapples, and grapes, respectively. The sensory survey of the hospitalized cancer patients (N=102) resulted in scores of 3.8, 3.7, 3.9, 3.9, and 3.7 on a 5-point scale for the gamma-irradiated freeze-dried apples, strawberries, pears, pineapples, and grapes, respectively. The results suggest that freeze-dried fruits can be sterilized with a dose of 5 kGy, except for grapes, which require a dose of 12 kGy, and that the organoleptic quality of the fruits is acceptable to immuno-compromised patients. However, to clarify the microbiological quality and safety of freeze-dried fruits should be verified by plating for both aerobic and anaerobic microorganisms.

  2. Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.

    SciTech Connect

    Garner, P. L.; Hanan, N. A.

    2011-06-07

    The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

  3. AC-3-irradiation test of sphere-pac and pellet (U,Pu)C fuel in the US Fast Flux Test Facility

    NASA Astrophysics Data System (ADS)

    Bart, G.; Botta, F. B.; Hoth, C. W.; Ledergerber, G.; Mason, R. E.; Stratton, R. W.

    2008-05-01

    The objective of the AC-3 bundle experiment in the Fast Flux Test Facility (FFTF) was to evaluate a fuel fabrication method by 'direct conversion' of nitrate solutions into spherical uranium-plutonium carbide particles and to compare the irradiation performance of 'sphere-pac' fuel pins prepared at Paul Scherrer Institute (PSI) with standard pellet fuel pins fabricated at Los Alamos National Laboratory (LANL). The irradiation and post test examination results show that mixed carbide pellet fuel produced by powder methods and sphere-pac particle fuel developed by internal gelation techniques are both valuable advanced fuel candidates for liquid metal reactors. The PSI fabrication process with direct conversion of actinide nitrate solutions into various sizes of fuel spheres by internal gelation and direct filling of spheres into cladding tubes is seen as more easily transferable to remote operation, showing a significant reduction of process steps. The process is also adaptable for the fabrication of carbonitrides and nitrides (still based on a uranium matrix), as well as for actinides diluted in a (uranium-free) yttrium stabilized zirconium oxide matrix. The AC-3 fuel bundle was irradiated in the Fast Flux Test Facility (FFTF) during the years 1986-1988 for 630 full power days to a peak burn up of ˜8 at.% fissile material. All of the pins, irradiated at linear powers of up to 84 kW/m, with cladding outer temperatures of 465 °C appeared to be in good condition when removed from the assembly. The rebirth of interest for fast reactor systems motivated the earlier teams to report about the excellent, still perfectly relevant results reached; this paper focusing on the sphere-pac fuel behaviour.

  4. UV testing of solar cells: Effects of antireflective coating, prior irradiation, and UV source

    NASA Technical Reports Server (NTRS)

    Meulenberg, A.

    1993-01-01

    Short-circuit current degradation of electron irradiated double-layer antireflective-coated cells after 3000 hours ultraviolet (UV) exposure exceeds 3 percent; extrapolation of the data to 10(exp 5) hours (11.4 yrs.) gives a degradation that exceeds 10 percent. Significant qualitative and quantitative differences in degradation were observed in cells with double- and single-layer antireflective coatings. The effects of UV-source age were observed and corrections were made to the data. An additional degradation mechanism was identified that occurs only in previously electron-irradiated solar cells since identical unirradiated cells degrade to only 6 +/- 3 percent when extrapolated 10(exp 5) hours of UV illumination.

  5. Design, fabrication, and operation of capsules for the irradiation testing of candidate advanced space reactor fuel pins

    NASA Technical Reports Server (NTRS)

    Thoms, K. R.

    1975-01-01

    Fuel irradiation experiments were designed, built, and operated to test uranium mononitride (UN) fuel clad in tungsten-lined T-111 and uranium dioxide fuel clad in both tungsten-lined T-111 and tungsten-lined Nb-1% Zr. A total of nine fuel pins was irradiated at average cladding temperatures ranging from 931 to 1015 C. The UN experiments, capsules UN-4 and -5, operated for 10,480 and 10,037 hr, respectively, at an average linear heat generation rate of 10 kW/ft. The UO2 experiment, capsule UN-6, operated for 8333 hr at an average linear heat generation rate of approximately 5 kW/ft. Following irradiation, the nine fuel pins were removed from their capsules, externally examined, and sent to the NASA Plum Brook Facility for more detailed postirradiation examination. During visual examination, it was discovered that the cladding of the fuel pin containing dense UN in each of capsules UN-4 and -5 had failed, exposing the UN fuel to the NaK in which the pins were submerged and permitting the release of fission gas from the failed pins. A rough analysis of the fission gas seen in samples of the gas in the fuel pin region indicated fission gas release-to-birth rates from these fuel pins in the range of .00001.

  6. Testing of Performance of Optical Fibers Under Irradiation in Intense Radiation Fields, When Subjected to Very High Temperatures

    SciTech Connect

    Blue, Thomas; Windl, Wolfgang; Dickerson, Bryan

    2013-01-03

    The primary objective of this project is to measure and model the performance of optical fibers in intense radiation fields when subjected to very high temperatures. This research will pave the way for fiber optic and optically based sensors under conditions expected in future high-temperature gas-cooled reactors. Sensor life and signal-to-noise ratios are susceptible to attenuation of the light signal due to scattering and absorbance in the fibers. This project will provide an experimental and theoretical study of the darkening of optical fibers in high-radiation and high-temperature environments. Although optical fibers have been studied for moderate radiation fluence and flux levels, the results of irradiation at very high temperatures have not been published for extended in-core exposures. Several previous multi-scale modeling efforts have studied irradiation effects on the mechanical properties of materials. However, model-based prediction of irradiation-induced changes in silica's optical transport properties has only recently started to receive attention due to possible applications as optical transmission components in fusion reactors. Nearly all damage-modeling studies have been performed in the molecular-dynamics domain, limited to very short times and small systems. Extended-time modeling, however, is crucial to predicting the long-term effects of irradiation at high temperatures, since the experimental testing may not encompass the displacement rate that the fibers will encounter if they are deployed in the VHTR. The project team will pursue such extended-time modeling, including the effects of the ambient and recrystallization. The process will be based on kinetic MC modeling using the concept of amorphous material consisting of building blocks of defect-pairs or clusters, which has been successfully applied to kinetic modeling in amorphized and recrystallized silicon. Using this procedure, the team will model compensation for rate effects, and the

  7. Experimental tests of irradiation-anneal-reirradiation effects on mechanical properties of RPV plate and weld materials

    SciTech Connect

    Hawthorne, J.R.

    1996-01-01

    The Charpy-V (C{sub V}) notch ductility and tension test properties of three reactor pressure vessel (RPV) steel materials were determined for the 288{degree}C (550{degree}F) irradiated (I), 288{degree}C (550{degree}F) irradiated + 454{degree}C (850{degree}F)-168 h postirradiation annealed (IA), and 288{degree}C (550{degree}F) reirradiated (IAR) conditions. Total fluences of the I condition and the IAR condition were, respectively, 3.33 {times} 10{sup 19} n/cm{sup 2} and 4.18 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV. The irradiation portion of the IAR condition represents an incremental fluence increase of 1. 05 {times} 10{sup 19} n/cm{sup 2}, E > 1 MeV, over the I-condition fluence. The materials (specimens) were supplied by the Yankee Atomic Electric Company and represented high and low nickel content plates and a high nickel, high copper content weld deposit prototypical of the Yankee-Rowe reactor vessel. The promise of the IAR method for extending the fluence tolerance of radiation-sensitive steels and welds is clearly shown by the results. The annealing treatment produced full C{sub V} upper shelf recovery and full or nearly full recovery in the C{sub V} 41 J (30 ft-lb) transition temperature. The C{sub V} transition temperature increases produced by the reirradiation exposure were 22% to 43% of the increase produced by the first cycle irradiation exposure. A somewhat greater radiation embrittlement sensitivity and a somewhat greater reirradiation embrittlement sensitivity was exhibited by the low nickel content plate than the high nickel content plate. Its high phosphorus content is believed to be responsible. The IAR-condition properties of the surface vs. interior regions of the low nickel content plate are also compared.

  8. Charpy impact test results of four low activation ferritic alloys irradiated at 370{degrees}C to 15 DPA

    SciTech Connect

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

    1996-10-01

    Miniature CVN specimens of four low activation ferritic alloys have been impact tested following irradiation at 370{degrees}C to 15 dpa. Comparison of the results with those of control specimens indicates that degradation in the impact behavior occurs in each of these four alloys. The 9Cr-2W alloy referred to as GA3X and the similar alloy F82H with 7.8Cr-2W appear most promising for further consideration as candidate structural materials in fusion energy system applications. These two alloys exhibit a small DBTT shift to higher temperatures but show increased absorbed energy on the upper shelf.

  9. Cellular regulation of basal and FSH-stimulated cyclic AMP production in irradiated rat testes

    SciTech Connect

    Kangasniemi, M.; Kaipia, A.; Toppari, J.; Mali, P.; Huhtaniemi, I.; Parvinen, M. )

    1990-05-01

    Basal and follicle-stimulating hormone (FSH)-stimulated cyclic AMP (cAMP) productions by seminiferous tubular segments from irradiated adult rats were investigated at defined stages of the epithelial cycle when specific spermatogenic cells were low in number. Seven days post-irradiation, depletion of spermatogonia did not influence the basal cAMP production, but FSH response increased in stages II-VIII. Seventeen days post-irradiation when spermatocytes were low in number, there was a small increase in basal cAMP level in stages VII-VIII and FSH-stimulated cAMP production increased in stages VII-XII and XIII-I. At 38 days when pachytene spermatocytes and round spermatids (steps 1-6) were low in number, a decreased basal cAMP production was measured in stages II-VI and IX-XII. FSH-stimulated cAMP output increased in stages VII-XII but decreased in stages II-VI. At 52 days when all spermatids were low in number, basal cAMP levels decreased in all stages of the cycle, whereas FSH response was elevated only in stages VII-XII. All spermatogenic cell types seem to have an effect on cAMP production by the seminiferous tubule in a stage-specific fashion. Germ cells appear to regulate Sertoli cell FSH response in a paracrine way, and a part of cAMP may originate from spermatids stimulated by an unknown FSH-dependent Sertoli cell factor. The FSH-dependent functions may control such phenomena as spermatogonial proliferation, final maturation of spermatids, and onset of meiosis.

  10. Re-weldability tests of irradiated austenitic stainless steel by a TIG welding method

    NASA Astrophysics Data System (ADS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Kalinin, George

    2000-12-01

    Austenitic stainless steel (SS) is widely used for the in-vessel and ex-vessel components of fusion reactors. In particular, SS316L(N)-IG (IG-ITER Grade) is used for the vacuum vessel (VV), pipe lines, blanket modules, branch pipe lines connecting the module coolant system with the manifold and for the other structures of ITER. One of the most important requirements for the VV and the water cooling branch pipelines is the possibility to repair different defects by welding. Those components which may require re-welding should be studied carefully. The SS re-weldability issue has a large impact on the design of in-vessel components, in particular, the design and efficiency of radiation shielding by the modules. Moreover, re-welded components should operate for the lifetime of the reactor. This paper deals with the study of re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. Tungsten inert gas (TIG) welding was used for re-welding of the SS.

  11. Safety testing of AGR-2 UO2 compacts 3-3-2 and 3-4-2

    SciTech Connect

    Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.; Montgomery, Fred C.

    2015-09-01

    Post-irradiation examination (PIE) is in progress on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2) [Collin 2014]. The AGR-2 PIE will build upon new information and understanding acquired throughout the recently-concluded six-year AGR-1 PIE campaign [Demkowicz et al. 2015] and establish a database for the different AGR-2 fuel designs.

  12. /sup 14/C-lactose breath tests during pelvic radiotherapy: the effect of the amount of small bowel irradiated

    SciTech Connect

    Weiss, R.G.; Stryker, J.A.

    1982-02-01

    Thirty patients who were undergoing pelvic radiotherapy had /sup 14/C-lactose breath tests performed in the first and fifth weeks of treatment. In Group I (21 patients), a significant portion of the small intestine was irradiated, and in Group II (9 patients), only a small portion of the small intestine was irradiated. In Group I, the average reductions in the excretion of ingested /sup 14/C between the first- and fifth-week tests were 41.5% at 1/2 hour postingestion (p less than 0.05), and 21.8% at 1 hour postingestion (p less than 0.05). In Group II, the percentage reductions were 11.8% and 3.7% at 1/2 and 1 hour, respectively (p greater than 0.05). The data suggest that lactose malabsorption is a factor in the etiology of the nausea, vomiting, and diarrhea experienced by patients who are undergoing pelvic radiotherapy, and that the amount of bowel included in the treatment volume significantly influences the degree of malabsorption.

  13. /sup 14/C-lactose breath tests during pelvic radiotherapy: the effect of the amount of small bowel irradiated

    SciTech Connect

    Weiss, R.G.; Stryker, J.A.

    1982-02-01

    Thirty patients who were undergoing pelvic radiotherapy had /sup 14/C-lactose breath tests performed in the first and fifth weeks of treatment. In Group I (21 patients), a significant portion of the small intestine was irradiated, and in Group II (9 patients), only a small portion of the small intestine was irradiated. In Group I, the average reductions in the excretion of ingested /sup 14/C between the first- and fifth-week tests were 41.5% at 1/2 hour postingestion (p<0.05), and 21.8% at 1 hour postingestion (p<0.05). In Group II, the prercentage reduction were 11.8% and 3.7% at 1/2 and 1 hour, respectively (p>0.05). The data suggest that lactose malabsorption is a factor in the etiology of the nausea, vomiting, and diarrhea experienced by patients who are undergoing pelvic radiotherapy, and that the amount of bowel included in the treatment volume significantly influences the degree of malabsorption.

  14. Design and Nuclear-Safety Related Simulations of Bare-Pellet Test Irradiations for the Production of Pu-238 in the High Flux Isotope Reactor using COMSOL

    SciTech Connect

    Freels, James D; Jain, Prashant K; Hobbs, Randy W

    2012-01-01

    The Oak Ridge National Laboratory (ORNL)is developing technology to produce plutonium-238 for the National Aeronautics and Space Administration (NASA) as a power source material for powering vehicles while in deep-space[1]. The High Flux Isotope Reactor (HFIR) of ORNL has been utilized to perform test irradiations of incapsulated neptunium oxide (NpO2) and aluminum powder bare pellets for purposes of understanding the performance of the pellets during irradiation[2]. Post irradiation examinations (PIE) are currently underway to assess the effect of temperature, thermal expansion, swelling due to gas production, fission products, and other phenomena

  15. Altered energy metabolism in an irradiated population of lizards at the Nevada Test Site

    SciTech Connect

    Nagy, K.A.; Medica, P.A.

    1985-07-01

    Field metabolic rates (via doubly labeled water), body compartmentalization of energy stores, and energy assimilation efficiencies were measured to assess all avenues of energy utilization in Uta stansburiana living in a low-level ..gamma..-irradiated plot in Rock Valley, Nevada. Comparison of energy budgets for radiation-sterilized females with those of nonirradiated control lizards revealed several substantial differences. Sterile females were heavier, mainly because they had extraordinarily large energy (fat) storage depots. Sterile females had much lower rates of energy expenditure via respiration and lower rates of energy intake by feeding. These differences are interpreted as indirect responses to radiation-induced sterility. There is little evidence of direct radiation effects on physiological functions other than reproduction.

  16. Estrogen-regulated genes in rat testes and their relationship to recovery of spermatogenesis after irradiation.

    PubMed

    Zhou, Wei; Bolden-Tiller, Olga U; Shao, Shan H; Weng, Connie C; Shetty, Gunapala; AbuElhija, Mahmoud; Pakarinen, Pirjo; Huhtaniemi, Ilpo; Momin, Amin A; Wang, Jing; Stivers, David N; Liu, Zhilin; Meistrich, Marvin L

    2011-10-01

    Despite numerous observations of the effects of estrogens on spermatogenesis, identification of estrogen-regulated genes in the testis is limited. Using rats in which irradiation had completely blocked spermatogonial differentiation, we previously showed that testosterone suppression with gonadotropin-releasing hormone-antagonist acyline and the antiandrogen flutamide stimulated spermatogenic recovery and that addition of estradiol (E2) to this regimen accelerated this recovery. We report here the global changes in testicular cell gene expression induced by the E2 treatment. By minimizing the changes in other hormones and using concurrent data on regulation of the genes by these hormones, we were able to dissect the effects of estrogen on gene expression, independent of gonadotropin or testosterone changes. Expression of 20 genes, largely in somatic cells, was up- or downregulated between 2- and 5-fold by E2. The unexpected and striking enrichment of transcripts not corresponding to known genes among the E2-downregulated probes suggested that these might represent noncoding mRNAs; indeed, we have identified several as miRNAs and their potential target genes in this system. We propose that genes for which expression levels are altered in one direction by irradiation and in the opposite direction by both testosterone suppression and E2 treatment are candidates for controlling the block in differentiation. Several genes, including insulin-like 3 (Insl3), satisfied those criteria. If they are indeed involved in the inhibition of spermatogonial differentiation, they may be candidate targets for treatments to enhance recovery of spermatogenesis following gonadotoxic exposures, such as those resulting from cancer therapy.

  17. Heterogeneous uptake of NO2 on Arizona Test Dust under UV-A irradiation: An aerosol flow tube study

    NASA Astrophysics Data System (ADS)

    Dupart, Yoan; Fine, Ludovic; D'Anna, Barbara; George, Christian

    2014-12-01

    The uptake rate of NO2 on Arizona Test Dust aerosols was measured using an aerosol flow tube (AFT). While the uptake rate in the dark could not be measured, the uptake under UV-A irradiation was enhanced, with values in the range from (0.6 ± 0.3) × 10-8, (2.4 ± 0.4) × 10-8. The observed gas phase products were HONO and NO, with yields of at 30% and 9.6%, respectively. The difference between these measurements and those previously reported on macroscopic films are discussed and differences highlighted. Interestingly, a reasonable agreement is observed between the uptake kinetics of NO2 on Arizona Test Dust macroscopic films and aerosols, despite the different experimental approaches. The simplest approach i.e. thin films having a significant porosity, provides similar uptake kinetics to the more complex and realistic AFT approach.

  18. Comparison of fracture behavior for low-swelling ferritic and austenitic alloys irradiated in the Fast Flux Test Facility (FFTF) to 180 DPA. Revision 1

    SciTech Connect

    Huang, F.H.

    1992-02-01

    Fracture toughness testing was conducted to investigate the radiation embrittlement of high-nickel superalloys, modified austenitic steels and ferritic steels. These materials have been experimentally proven to possess excellent resistance to void swelling after high neutron exposures. In addition to swelling resistance, post-irradiation fracture resistance is another important criterion for reactor material selection. By means of fracture mechanics techniques the fracture behavior of those highly irradiated alloys was characterized in terms of irradiation and test conditions. Precipitation-strengthened alloys failed by channel fracture with very low postirradiation ductility. The fracture toughness of titanium-modified austenitic stainless steel D9 deteriorates with increasing fluence to about 100 displacement per atom (dpa), the fluence level at which brittle fracture appears to occur. Ferritic steels such as HT9 are the most promising candidate materials for fast and fusion reactor applications. The upper-shelf fracture toughness of alloy HT9 remained adequate after irradiation to 180 dpa although its ductile- brittle transition temperature (DBTT) shift by low temperature irradiation rendered the material susceptible to brittle fracture at room temperature. Understanding the fracture characteristics under various irradiation and test conditions helps reduce the potential for brittle fracture by permitting appropriate measure to be taken.

  19. Comparison of fracture behavior for low-swelling ferritic and austenitic alloys irradiated in the Fast Flux Test Facility (FFTF) to 180 DPA

    SciTech Connect

    Huang, F.H.

    1992-02-01

    Fracture toughness testing was conducted to investigate the radiation embrittlement of high-nickel superalloys, modified austenitic steels and ferritic steels. These materials have been experimentally proven to possess excellent resistance to void swelling after high neutron exposures. In addition to swelling resistance, post-irradiation fracture resistance is another important criterion for reactor material selection. By means of fracture mechanics techniques the fracture behavior of those highly irradiated alloys was characterized in terms of irradiation and test conditions. Precipitation-strengthened alloys failed by channel fracture with very low postirradiation ductility. The fracture toughness of titanium-modified austenitic stainless steel D9 deteriorates with increasing fluence to about 100 displacement per atom (dpa), the fluence level at which brittle fracture appears to occur. Ferritic steels such as HT9 are the most promising candidate materials for fast and fusion reactor applications. The upper-shelf fracture toughness of alloy HT9 remained adequate after irradiation to 180 dpa although its ductile- brittle transition temperature (DBTT) shift by low temperature irradiation rendered the material susceptible to brittle fracture at room temperature. Understanding the fracture characteristics under various irradiation and test conditions helps reduce the potential for brittle fracture by permitting appropriate measure to be taken.

  20. Input Correlations for Irradiation Creep of FeCrAl and SiC Based on In-Pile Halden Test Results

    SciTech Connect

    Terrani, K. A.; Karlsen, T. M.; Yamamoto, Yukinori

    2016-05-01

    Swelling and creep behavior of wrought FeCrAl alloys and CVD-SiC, two candidate accident tolerant fuel cladding materials, are being examined using in-pile tests at the Halden reactor. The outcome of these tests are material property correlations that are inputs into fuel performance analysis tools. The results are discussed and compared with what is available in literature from irradiation experiments in other reactors or out-of-pile tests. Specific recommendation on what correlations should be used for swelling, thermal, and irradiation creep for each material are provided in this document.

  1. Interaction between age of irradiation and age of testing in the disruption of operant performance using a ground-based model for exposure to cosmic rays.

    PubMed

    Rabin, Bernard M; Joseph, James A; Shukitt-Hale, Barbara; Carrihill-Knoll, Kirsty L

    2012-02-01

    Previous research has shown a progressive deterioration in cognitive performance in rats exposed to (56)Fe particles as a function of age. The present experiment was designed to evaluate the effects of age of irradiation independently of the age of testing. Male Fischer-344 rats, 2, 7, 12, and 16 months of age, were exposed to 25-200 cGy of (56)Fe particles (1,000 MeV/n). Following irradiation, the rats were trained to make an operant response on an ascending fixed-ratio reinforcement schedule. When performance was evaluated as a function of both age of irradiation and testing, the results showed a significant effect of age on the dose needed to produce a performance decrement, such that older rats exposed to lower doses of (56)Fe particles showed a performance decrement compared to younger rats. When performance was evaluated as a function of age of irradiation with the age of testing held constant, the results indicated that age of irradiation was a significant factor influencing operant responding, such that older rats tested at similar ages and exposed to similar doses of (56)Fe particles showed similar performance decrements. The results are interpreted as indicating that the performance decrement is not a function of age per se, but instead is dependent upon an interaction between the age of irradiation, the age of testing, and exposure to HZE particles. The nature of these effects and how age of irradiation affects cognitive performance after an interval of 15 to 16 months remains to be established.

  2. Developing and testing solar irradiance forecasting techniques in the Hawaiian Islands region

    NASA Astrophysics Data System (ADS)

    Matthews, D. K.

    2015-12-01

    The Hawaíi Natural Energy Institute (HNEI) is developing an operational solar forecasting for the Hawaiian Islands. The system comprises the following three components, covering forecasting horizons from seconds to days ahead. (i) A ground-observation driven advection model, using sky imagery and cloud height data. (ii) A satellite-image based advection model, primarily driven by Geostationary Operational Environmental Satellite (GOES) imagery. (iii) A coupled ocean-atmosphere model, using the Regional Ocean Modeling System (ROMS) model and the Weather Research and Forecasting (WRF) model, including newly available microphysics, shallow convection parameterization, and radiative transfer physics options. The satellite and NWP components provide coverage for the entire island chain, however, lack the resolution in time and space, to accurately forecast ramp events (large changes in irradiance that occur over a short period of time). Knowledge of the magnitude, duration and timing of ramp events are particularly important in Hawaíi due to the small size of the electric grids. Currently, HNEI employs a sky imager and ceilometer installed on the University of Hawaíi campus for high resolution forecasting, however, instrument design and cost limit widespread deployment. We discuss the development and preliminary validation of a new forecasting system based on inexpensive, panoramic (large FOV), off-the-shelf cameras with a cloud base height retrieval algorithm that does not require additional instrumentation.

  3. Physics Characterization of TLD-600 and TLD-700 and Acceptance Testing of New XRAD 160 Biological X-Ray Irradiator

    NASA Astrophysics Data System (ADS)

    Cao, Yanan

    2: Acceptance testing of new X-RAD 160 Biological X-Ray Irradiator. Purpose: An X-RAD 160 Biological X-Ray Irradiator was recently installed at Duke University to serve as a key device for cellular radiobiology research. The purpose of this study is to perform acceptance testing on the new irradiator for operator radiation safety and irradiation specifications. Methods: The acceptance testing included tests of the following components: (1) Leakage radiation survey, (2) Half-value layer (beam quality), (3) Uniformity, (4) KVp accuracy, (5) Exposure at varying mA (linearity of mA), (6) Exposure at varying kVp, (7) Inverse square measurements, (8) Field size measurement, (9) Exposure constancy. The irradiation parameters for each components of first round of acceptance testing performed on September 21, 2012 were: Leakage radiation survey (none, 160 kVp, 18 mA, 200s), Beam quality (40cm, 50-140 kVp in 10 kVp incensement, 1 mA, 10s, none), Uniformity (40cm, 160 kVp, 18 mA, 15s, F1), KVp accuracy (40cm, 50-150 kVp in 10 kVp incensement, 10 mA, 15s, none), Linearity of mA (40cm, 160 kVp, 2-18 mA, 15s, none), Inverse square measurements (20-63cm, 160 kVp, 1mA, 30s, none), Field size measurement (40cm, 160 kVp, 10 mA, 15s, none), Exposure constancy (40cm, 160 kVp, 18 mA, 20s, none). The irradiation parameters for each components for each components of second round of acceptance testing performed on November 18, 2012 were: Beam quality (40cm, 35-150 kVp, 1 mA, 10s, F1), KVp accuracy (40cm, 35-150 kVp, 1 mA, 10s, F1), Variation of kVp (40cm, 160 kVp, 18 mA, 30s, F1), Linearity of mA (40cm, 160 kVp, 1-18 mA, 30s, F1), Uniformity (40cm, 160 kVp, 18 mA, 30s, F1), Inverse square measurements (20-63cm, 160 kVp, 18 mA, 30s, F1). Results: The first round of acceptance testing performed on September 21, 2012 failed due to the fact that the measured exposure along the X-axis was significantly non-uniform; the exposure greatly decreases going in the left direction, which is a clear

  4. Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

    NASA Astrophysics Data System (ADS)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

    2016-06-01

    Low-enriched (U-235 <20 pct) U-Mo dispersion fuel is being developed for use in research and test reactors. In most cases, fuel plates with Al or Al-Si alloy matrices have been tested in the Advanced Test Reactor to support this development. In addition, fuel plates with Mg as the matrix have also been tested. The benefit of using Mg as the matrix is that it potentially will not chemically interact with the U-Mo fuel particles during fabrication or irradiation, whereas with Al and Al-Si alloys such interactions will occur. Fuel plate R9R010 is a Mg matrix fuel plate that was aggressively irradiated in ATR. This fuel plate was irradiated as part of the RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

  5. Investigation of the Feasibility of Utilizing Gamma Emission Computed Tomography in Evaluating Fission Product Migration in Irradiated TRISO Fuel Experiments

    SciTech Connect

    Jason M. Harp; Paul A. Demkowicz

    2014-10-01

    In the High Temperature Gas-Cooled Reactor (HTGR) the TRISO particle fuel serves as the primary fission product containment. However the large number of TRISO particles present in proposed HTGRs dictates that there will be a small fraction (~10-4 to 10-5) of as manufactured and in-pile particle failures that will lead to some fission product release. The matrix material surrounding the TRISO particles in fuel compacts and the structural graphite holding the TRISO particles in place can also serve as sinks for containing any released fission products. However data on the migration of solid fission products through these materials is lacking. One of the primary goals of the AGR-3/4 experiment is to study fission product migration from failed TRISO particles in prototypic HTGR components such as structural graphite and compact matrix material. In this work, the potential for a Gamma Emission Computed Tomography (GECT) technique to non-destructively examine the fission product distribution in AGR-3/4 components and other irradiation experiments is explored. Specifically, the feasibility of using the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) Precision Gamma Scanner (PGS) system for this GECT application is considered. To test the feasibility, the response of the PGS system to idealized fission product distributions has been simulated using Monte Carlo radiation transport simulations. Previous work that applied similar techniques during the AGR-1 experiment will also be discussed as well as planned uses for the GECT technique during the post irradiation examination of the AGR-2 experiment. The GECT technique has also been applied to other irradiated nuclear fuel systems that were currently available in the HFEF hot cell including oxide fuel pins, metallic fuel pins, and monolithic plate fuel.

  6. HIGH TEMPERATURE IRRADIATION RESISTANT THERMOCOUPLES – A LOW COST SENSOR FOR IN-PILE TESTING AT HIGH TEMPERATURES

    SciTech Connect

    Joy L. Rempe; Darrell L. Knudson; Keith G. Condie; S. Curtis Wilkins; Joshua E. Daw

    2008-06-01

    Several options have been identified to improve recently-developed Idaho National Laboratory (INL) High Temperature Irradiation Resistant ThermoCouples (HTIR-TCs) for in-pile testing. These options have the potential to reduce fabrication costs and allow HTIR-TC use in higher temperature applications (up to at least 1800 °C). The INL and the University of Idaho (UI) investigated these options with the ultimate objective of providing recommendations for alternate thermocouple designs that are optimized for various applications. This paper summarizes results from these INL/UI investigations. Specifically, results are reported about several options found to enhance HTIR-TC performance, such as improved heat treatments, alternate geometries, alternate fabrication techniques, and the use of copper/nickel alloys as soft extension cable.

  7. Radiation effects on rat testes. VIII. Kinetic properties of hydrolases following partial body gamma irradiation of rats.

    PubMed

    Gupta, G S; Bawa, S R

    1975-05-01

    Kinetic properties such as Michaelis constant (Km), maximum velocity (Vmax), temperature coefficient (Q10) and energy of activation (Ea) for hydrolysis of adenosine-5'-phosphate at pH 9.5 and sodium pyrophosphate at pH 8.35 by normal and radiated testes supernatants have been described. Kinetic parameters are related to respective phosphohydrolases (phosphatases). (1) Km values for 5'nucleotidase and inorganic pyrophosphatase of normal testis were determined as 1.25 X 10(-3)M and 0.81 X 10(-3)M respectively; (II) Vmax correspond to 318 mug P/15 min and 430 mug P/15 min for 100 mg tissue respectively; (III) Q10 for 5 nucleotidase is 1.7 and for inorganic pyrophosphatase is 4.2 at a temperature 10-30degreesC; (IV) Ea for hydrolysis of AMP and sodium pyrophosphate were calculated by Arrhenius plots as 17000 and 9000 cal/mol. (V) Km values for irradiated enzymes are similar to the control values suggesting that the binding capacities of these enzymes with their substrates remain unaffected after radiation; (VI) Vmax for radiated enzymes correspond to a value of 500 mug P/100 mg tissue/15 min for 5'nucleotidase and 118 mug P/100 mug tissue/15 min for inorganic pyrophosphatase; (VII) 110 for 5'nucleotidase is 2.2 and inorganic pyrophosphatase 1.16 at 10-30degreesC; (VIII) Ea for irradiated 5'nucleotidase is comparable to those of normal rats whereas for inorganic pyrophosphatase Ea is moderately declined. The observed changes have been related to the different types of metabolic activity in germinal and nongerminal cells of testes.

  8. On tests of equal effect per fraction in microcolony assays of survival after fractionated irradiations.

    PubMed

    Taylor, J M

    1985-01-01

    H. D. Thames, Jr. and H. R. Withers [Br. J. Radiol. 53, 1071-1077 (1980)] propose a test of an equal effect per fraction in microcolony assays after fractionated radiation, in which the total effect is measured by counting microcolonies derived from surviving cells in a tissue. The factors considered to influence the cytocidal effect per fraction are incomplete repair, repopulation, and synchrony. The statistics used in the method are criticized and conditions are given under which the test should not be used. An alternative method of testing for an equal effect per fraction is proposed. The pros and cons of each test are discussed and compared using some mouse jejunal crypt cell survival data.

  9. Imminent: Irradiation Testing of (Th,Pu)O{sub 2} Fuel - 13560

    SciTech Connect

    Kelly, Julian F.; Franceschini, Fausto

    2013-07-01

    Commercial-prototype thorium-plutonium oxide (Th-MOX) fuel pellets have been loaded into the material test reactor in Halden, Norway. The fuel is being operated at full power - with instrumentation - in simulated LWR / PHWR conditions and its behaviour is measured 'on-line' as it operates to high burn-up. This is a vital test on the commercialization pathway for this robust new thoria-based fuel. The performance data that is collected will support a fuel modeling effort to support its safety qualification. Several different samples of Th-MOX fuel will be tested, thereby collecting information on ceramic behaviours and their microstructure dependency. The fuel-cycle reasoning underpinning the test campaign is that commercial Th- MOX fuels are an achievable intermediate / near-term SNF management strategy that integrates well with a fast reactor future. (authors)

  10. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    SciTech Connect

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

  11. Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

    SciTech Connect

    Palmer, A. J.; Haggard, DC; Herter, J. W.; Swank, W. D.; Knudson, D. L.; Cherry, R. S.; Scervini, M.

    2015-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly

  12. Fabrication of ORNL Fuel Irradiated in the Peach Bottom Reactor and Postirradiation Examination of Recycle Test Elements 7 and 4

    SciTech Connect

    Long, Jr. E.L.

    2001-10-25

    Seven full-sized Peach Bottom Reactor. fuel elements were fabricated in a cooperative effort by Oak Ridge National Laboratory (ORNL) and Gulf General Atomic (GGA) as part of the National HTGR Fuel Recycle Development Program. These elements contain bonded fuel rods and loose beds of particles made from several combinations of fertile and fissile particles of interest for present and future use in the High-Temperature Gas-Cooled Reactor (HTGR). The portion of the fuel prepared for these elements by ORNL is described in detail in this report, and it is in conjunction with the GGA report (GA-10109) a complete fabrication description of the test. In addition, this report describes the results obtained to date from postirradiation examination of the first two elements removed from the Peach Bottom Reactor, RTE-7 and -4. The fuel examined had relatively low exposure, up to about 1.5 x 10{sup 21} neutrons/cm* fast (>0.18 MeV) fluence, compared with the peak anticipated HTGR fluence of 8.0 x 10{sup 21}, but it has performed well at this exposure. Dimensional data indicate greater irradiation shrinkage than expected from accelerated test data to higher exposures. This suggests that either the method of extrapolation of the higher exposure data back to low exposure is faulty, or the behavior of the coated particles in the neutron spectrum characteristic of the accelerated tests does not adequately represent the behavior in an HTGR spectrum.

  13. Are fern stomatal responses to different stimuli coordinated? Testing responses to light, vapor pressure deficit, and CO2 for diverse species grown under contrasting irradiances.

    PubMed

    Creese, Chris; Oberbauer, Steve; Rundel, Phil; Sack, Lawren

    2014-10-01

    The stomatal behavior of ferns provides an excellent system for disentangling responses to different environmental signals, which balance carbon gain against water loss. Here, we measured responses of stomatal conductance (gs ) to irradiance, CO2 , and vapor pressure deficit (VPD) for 13 phylogenetically diverse species native to open and shaded habitats, grown under high- and low-irradiance treatments. We tested two main hypotheses: that plants adapted and grown in high-irradiance environments would have greater responsiveness to all stimuli given higher flux rates; and that species' responsiveness to different factors would be correlated because of the relative simplicity of fern stomatal control. We found that species with higher light-saturated gs had larger responses, and that plants grown under high irradiance were more responsive to all stimuli. Open habitat species showed greater responsiveness to irradiance and CO2 , but lower responsiveness to VPD; a case of plasticity and adaptation tending in different directions. Responses of gs to irradiance and VPD were positively correlated across species, but CO2 responses were independent and highly variable. The novel finding of correlations among stomatal responses to different stimuli suggests coordination of hydraulic and photosynthetic signaling networks modulating fern stomatal responses, which show distinct optimization at growth and evolutionary time-scales.

  14. In-pile irradiation research at NRI Rez for corrosion and material testing

    SciTech Connect

    Kysela, J.

    1995-09-01

    Experimental test facility enabling exposition of materials to radiation environment is described. The facility enables research on behavior of materials and water chemistry of PWRs and BWRs. High pressure water loop RVS-3 is operated in LVR-15 test research reactor. The loop was designed as a universal facility providing wide experimental possibilities. Loop is designed for 16.7 MPa pressure, 334 C temperature, 10 t/hour water flow rate and 1.10{sup 14}cm{sup {minus}2}{center_dot}s{sup {minus}1} neutron flux. Water chemistry used corresponds the PWRs with boric acid and pH control by potassium/lithium, hydrogen control by gas or ammonia. BWR loop is designed for investigation of structural materials and Water chemistry in conditions of boiling water power reactors.

  15. Effect of electron beam irradiation and microencapsulation on the flame retardancy of ethylene-vinyl acetate copolymer materials during hot water ageing test

    NASA Astrophysics Data System (ADS)

    Sheng, Haibo; Zhang, Yan; Wang, Bibo; Yu, Bin; Shi, Yongqian; Song, Lei; Kundu, Chanchal Kumar; Tao, Youji; Jie, Ganxin; Feng, Hao; Hu, Yuan

    2017-04-01

    Microencapsulated ammonium polyphosphate (MCAPP) in combination with polyester polyurethane (TPU) was used to flame retardant ethylene-vinyl acetate copolymer (EVA). The EVA composites with different irradiation doses were immersed in hot water (80 °C) to accelerate ageing process. The microencapsulation and irradiation dose ensured positive impacts on the properties of the EVA composites in terms of better dimensional stability and flame retardant performance. The microencapsulation of APP could lower its solubility in water and the higher irradiation dose led to the more MCAPP immobilized in three dimensional crosslinked structure of the EVA matrix which could jointly enhance the flame retardant and electrical insulation properties of the EVA composites. So, the EVA composites with 180 kGy irradiation dose exhibited better dimensional stability than the EVA composites with 120 kGy due to the higher crosslinking degree. Moreover, the higher irradiation dose lead to the more MCAPP immobilizated in crosslinked three-dimensional structure of EVA, enhancing the flame retardancy and electrical insulation properties of the EVA composites. After ageing test in hot water at 80 °C for 2 weeks, the EVA/TPU/MCAPP composite with 180 kGy could still maintain the UL-94 V-0 rating and the limiting oxygen index (LOI) value was as high as 30%. This investigation indicated the flame retardant EVA cable containing MCAPP could achieve stable properties and lower electrical fire hazard risk during long-term hot water ageing test.

  16. Pretransplant pulmonary function tests predict risk of mortality following fractionated total body irradiation and allogeneic peripheral blood stem cell transplant

    SciTech Connect

    Singh, Anurag K. . E-mail: singan@mail.nih.gov; Karimpour, Shervin E.; Savani, Bipin N.; Guion, Peter M.S.; Hope, Andrew J.; Mansueti, John R.; Ning, Holly; Altemus, Rosemary M. Ph.D.; Wu, Colin O.; Barrett, A. John

    2006-10-01

    Purpose: To determine the value of pulmonary function tests (PFTs) done before peripheral blood stem cell transplant (PBSCT) in predicting mortality after total body irradiation (TBI) performed with or without dose reduction to the lung. Methods and Materials: From 1997 to 2004, 146 consecutive patients with hematologic malignancies received fractionated TBI before PBSCT. With regimen A (n = 85), patients were treated without lung dose reduction to 13.6 gray (Gy). In regimen B (n = 35), total body dose was decreased to 12 Gy (1.5 Gy twice per day for 4 days) and lung dose was limited to 9 Gy by use of lung shielding. In regimen C (n = 26), lung dose was reduced to 6 Gy. All patients received PFTs before treatment, 90 days after treatment, and annually. Results: Median follow-up was 44 months (range, 12-90 months). Sixty-one patients had combined ventilation/diffusion capacity deficits defined as both a forced expiratory volume in the first second (FEV{sub 1}) and a diffusion capacity of carbon dioxide (DLCO) <100% predicted. In this group, there was a 20% improvement in one-year overall survival with lung dose reduction (70 vs. 50%, log-rank test p = 0.042). Conclusion: Among those with combined ventilation/diffusion capacity deficits, lung dose reduction during TBI significantly improved survival.

  17. Thermal Cycling and High Temperature Reverse Bias Testing of Control and Irradiated Gallium Nitride Power Transistors

    NASA Technical Reports Server (NTRS)

    Patterson, Richard L.; Boomer, Kristen T.; Scheick, Leif; Lauenstein, Jean-Marie; Casey, Megan; Hammoud, Ahmad

    2014-01-01

    The power systems for use in NASA space missions must work reliably under harsh conditions including radiation, thermal cycling, and exposure to extreme temperatures. Gallium nitride semiconductors show great promise, but information pertaining to their performance is scarce. Gallium nitride N-channel enhancement-mode field effect transistors made by EPC Corporation in a 2nd generation of manufacturing were exposed to radiation followed by long-term thermal cycling and testing under high temperature reverse bias conditions in order to address their reliability for use in space missions. Result of the experimental work are presented and discussed.

  18. JESD57 Test Standard, Procedures for the Measurement of Single-Event Effects in Semiconductor Devices from Heavy-Ion Irradiation Revision Update

    NASA Technical Reports Server (NTRS)

    Lauenstein, Jean-Marie

    2016-01-01

    The JEDEC JESD57 test standard, Procedures for the Measurement of Single-Event Effects in Semiconductor Devices from Heavy-Ion Irradiation, is undergoing its first revision since 1996. This presentation will provide an overview of some of the key proposed updates to the document.

  19. Displacement damage induced in iron by gammas and neutrons under irradiation in the IFMIF test cell

    NASA Astrophysics Data System (ADS)

    Simakov, S. P.; Fischer, U.

    2011-10-01

    This work presents a complete comparative analysis of the radiation damage induced in iron-based materials in IFMIF by photons and neutrons. The gamma induced damage takes into account, for the first time, both photonuclear and photoatomic reaction mechanisms. The relevant cross sections were taken from available data evaluations. The gamma and neutron radiation fields were calculated by the McDeLicious Monte Carlo code using a 3-D geometry model. Finally the gamma and neutron induced damages in the iron have been assessed inside the IFMIF test cell and the surrounding concrete walls. It was found that the photoatomic mechanism dominates the photonuclear with at least one hundred times higher damage rates. The ratio of the gamma and the neutron induced displacement damage was found to be 10 -3 inside the concrete wall and 10 -5 in the components close to d-Li source. This fraction may increase a few times due to the uncertainty of the evaluated γ-dpa cross sections and the different surviving probabilities for defects produced by gammas and neutrons, nevertheless unlikely exceed 1%.

  20. Genetic screening test for psoriatic arthritis and UVB irradiation potential responders: A new tool to identify psoriasis subpopulation patients?

    PubMed Central

    Lotti, Torello; Tognetti, Linda; Galeone, Massimiliano; Bruscino, Nicola; Moretti, Silvia; Giorgini, Simonetta

    2011-01-01

    Psoriatic arthritis (PsA) is a psoriasis-associated inflammatory disease of the joints and enthuses. The occurrence of PsA is linked to the complex interplay of gene environment, and immune system. Genetic factors have long been recognized to play an important role in PsA. Genes within the major histocompatibility complex (MHC) region have been shown to be associated with PsA. These include genes coded in the HLA region, (especially Class I antigens) and non-HLA genes (i.e., MHC class I chain-related antigen A, MICA, and TNF-α genes). Association studies in PsA have also identified a number of genes outside MHC region, including interleukin-1 (IL-1) gene cluster, killer-cell immunoglobulin-like receptors (KIRs), and IL-23R genes. Established systemic treatments for moderate-severe psoriasis and PsA may be potentially dangerous and usually time consuming for the patient and often expensive for the National Health Systems. Tests which could predict which subset of psoriatic patients could develop the most severe forms of the disease (i.e., PsA) or will respond to well-established (UVB irradiation) or other systemic treatments are now required. The goal of genetic test screening is to rapidly and safely identify subjects for preventive or early treatment or extended surveillance prior to the onset of signs and symptoms. Genetic tests today represent a reliable investigation procedure which could rapidly and consistently improve the diagnostic ability of the dermatologist and contribute to the early and correct treatment of the different subsets of PsA. PMID:23130225

  1. Quantity of 135I released from the AGR-1, AGR-2, and AGR-3/4 experiments and discovery of 131I at the FPMS traps during the AGR-3/4 experiment

    SciTech Connect

    Scates, Dawn M.

    2014-09-01

    A series of three Advanced Gas Reactor (AGR) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2014, these experiments supported the development and qualification of the new U.S. tristructural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each AGR experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The HPGe detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About 2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that its production rate very nearly equals the decay rate of the

  2. Test Standard Revision Update: JESD57, "Procedures for the Measurement of Single-Event Effects in Semiconductor Devices from Heavy-Ion Irradiation"

    NASA Technical Reports Server (NTRS)

    Lauenstein, Jean-Marie

    2015-01-01

    The JEDEC JESD57 test standard, Procedures for the Measurement of Single-Event Effects in Semiconductor Devices from Heavy-Ion Irradiation, is undergoing its first revision since 1996. In this talk, we place this test standard into context with other relevant radiation test standards to show its importance for single-event effect radiation testing for space applications. We show the range of industry, government, and end-user party involvement in the revision. Finally, we highlight some of the key changes being made and discuss the trade-space in which setting standards must be made to be both useful and broadly adopted.

  3. U.S. Contribution 1994 Summary Report Task T12: Compatibility and irradiation testing of vanadium alloys

    SciTech Connect

    Smith, D.L.

    1995-03-01

    Vanadium alloys exhibit important advantages as a candidate structural material for fusion first wall/blanket applications. These advantages include fabricability, favorable safety and environmental features, high temperature and high wall load capability, and long lifetime under irradiation. Vanadium alloys with (3-5)% chromium and (3-5)% titanium appear to offer the best combination of properties for first wall/blanket applications. A V-4Cr-4Ti alloy is recommended as the reference composition for the ITER application. This report provides a summary of the R&D conducted during 1994 in support of the ITER Engineering Design Activity. Progress is reported for Vanadium Alloy Production, Welding, Physical Properties, Baseline Mechanical Properties, Corrosion/Compatibility, Neutron Irradiation Effects, Helium Transmutation Effects on Irradiated Alloys, and the Status of Irradiation Experiments. Separate abstracts have been prepared for individual reports from this publication.

  4. Miniaturized disk-bend testing, nano-indentation and the microstructure of ion-irradiated titanium aluminides

    SciTech Connect

    Petouhoff, N.L.; Ardell, A.J.; Oliver, W.C.; Lucas, B.N.

    1993-09-01

    Effect of ion irradiation on microstructures and mechanical properties of Ti-52Al (TiAl) and Ti-26Al (Ti{sub 3}Al) were investigated. The alloys were irradiated with 2 MeV protons and Ar{sup +} ions at low temperatures ({minus}175 to {minus}135C) to max fluences of 4.5 {times} 10{sup 15} Ar{sup +}/cm{sup 2} and 2 {times} 10{sup 17} H{sup +}/cm{sup 2}. Yield strengths of unirradiated TiAl and the fracture strength of unirradiated Ti{sub 3}Al were 367 {plus_minus} 33 MPa and 536 {plus_minus} 30 MPa, respectively, in excellent agreement with published data. Yield strength of TiAl and fracture strength of Ti{sub 3}Al increased as a result of irradiation. Strengths of both alloys were lowest for the samples irradiated to the highest Ar{sup +}-ion dose, but otherwise there was no correlation of strength with dose. The nanohardness of irradiated specimens generally increased with dose, but influence of dose on Young`s modulus was erratic. Plate-shaped defects vacancy in character, and helical dislocations were observed in irradiated TiAl by TEM. The Ar{sup +}-ion irradiation-induced microstructure of Ti{sub 3}Al contained defects producing mottled contrast at 1 dpa and black-spot contrast at 5 dpa. Irradiation-induced loss of long-range order was also observed in both alloys. Influence of these microstructural variables on mechanical behavior is discussed.

  5. An in vitro cell irradiation protocol for testing photopharmaceuticals and the effect of blue, green, and red light on human cancer cell lines.

    PubMed

    Hopkins, S L; Siewert, B; Askes, S H C; Veldhuizen, P; Zwier, R; Heger, Michal; Bonnet, Sylvestre

    2016-05-11

    Traditionally, ultraviolet light (100-400 nm) is considered an exogenous carcinogen while visible light (400-780 nm) is deemed harmless. In this work, a LED irradiation system for in vitro photocytotoxicity testing is described. The LED irradiation system was developed for testing photopharmaceutical drugs, but was used here to determine the basal level response of human cancer cell lines to visible light of different wavelengths, without any photo(chemo)therapeutic. The effects of blue (455 nm, 10.5 mW cm(-2)), green (520 nm, 20.9 mW cm(-2)), and red light (630 nm, 34.4 mW cm(-2)) irradiation was measured for A375 (human malignant melanoma), A431 (human epidermoid carcinoma), A549 (human lung carcinoma), MCF7 (human mammary gland adenocarcinoma), MDA-MB-231 (human mammary gland adenocarcinoma), and U-87 MG (human glioblastoma-grade IV) cell lines. In response to a blue light dose of 19 J cm(-2), three cell lines exhibited a minimal (20%, MDA-MB-231) to moderate (30%, A549 and 60%, A375) reduction in cell viability, compared to dark controls. The other cell lines were not affected. Effective blue light doses that produce a therapeutic response in 50% of the cell population (ED50) compared to dark conditions were found to be 10.9 and 30.5 J cm(-2) for A375 and A549 cells, respectively. No adverse effects were observed in any of the six cell lines irradiated with a 19 J cm(-2) dose of 520 nm (green) or 630 nm (red) light. The results demonstrate that blue light irradiation can have an effect on the viability of certain human cancer cell types and controls should be used in photopharmaceutical testing, which uses high-energy (blue or violet) visible light activation.

  6. Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor

    SciTech Connect

    J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

    1998-10-01

    The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

  7. Irradiation of Metallic Fuels with Rare Earth Additions for Actinide Transmutation in the Advanced Test Reactor. Experiment Description for AFC-2A and AFC-2B

    SciTech Connect

    Hayes, Steven L.

    2006-12-01

    The U.S. Advanced Fuel Cycle Initiative (AFCI), now within the broader context of the Global Nuclear Energy Partnership (GNEP), seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter-lived fission products, thereby dramatically decreasing the volume of material requiring disposal and the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository. One important component of the technology development is actinide-bearing metallic transmutation fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. The proposed AFC-2A and AFC-2B irradiation experiments are a continuation of the metallic fuel test series in progress in the ATR. This report documents the experiment description and test matrix of the proposed experiments and the Post Irradiation Examination (PIE) and fabrication schedule.

  8. Tests of the radiation hardness of VLSI Integrated Circuits and Silicon Strip Detectors for the SSC (Superconducting Super Collider) under neutron, proton, and gamma irradiation

    SciTech Connect

    Ziock, H.J.; Milner, C.; Sommer, W.F. ); Carteglia, N.; DeWitt, J.; Dorfan, D.; Hubbard, B.; Leslie, J.; O'Shaughnessy, K.F.; Pitzl, D.; Rowe, W.A.; Sadrozinski, H.F.W.; Seiden, A.; Spencer, E. . Inst. for Particle Physics); Ellison, J.A. ); Ferguson, P. ); Giubellino

    1990-01-01

    As part of a program to develop a silicon strip central tracking detector system for the Superconducting Super Collider (SSC) we are studying the effects of radiation damage in silicon detectors and their associated front-end readout electronics. We report on the results of neutron and proton irradiations at the Los Alamos National Laboratory (LANL) and {gamma}-ray irradiations at UC Santa Cruz (UCSC). Individual components on single-sided AC-coupled silicon strip detectors and on test structures were tested. Circuits fabricated in a radiation hard CMOS process and individual transistors fabricated using dielectric isolation bipolar technology were also studied. Results indicate that a silicon strip tracking detector system should have a lifetime of at least one decade at the SSC. 17 refs., 17 figs.

  9. Activity testing of alveolar macrophages and changes in surfactant phospholipids after irradiation in bronchoalveolar lavage: Experimental and clinical data

    SciTech Connect

    Steinberg, F.; Rehn, B.; Kraus, R.; Quabeck, K.; Bruch, J.; Beelen, D.W.; Schaefer, U.W.; Streffer, C. )

    1992-07-01

    This study presents results of bronchoalveolar lavage (BAL) after irradiation to the lungs in mice as well as clinical data. The number of BAL cells, mainly macrophages, lymphocytes, and granulocytes, changed in a time-dependent manner. The phagocytic activity of the macrophages measured as the phagocytosis of microbeads and measured as the esterase activity also showed a strong time-dependent increase during the acute phase up to 21 days after irradiation. The contents of surfactant phospholipids (SF) and sphingomyelin (SPH; as a parameter for cell death) were quantified by HPLC. Both were significantly changed between day 2 and 21 after irradiation. Three BALs of a patient with idiopathic interstitial pneumonitis, who had received an allogenic bone marrow graft after total body irradiation with 10 Gy, showed similar effects in the cellular and surfactant parameters. These data indicate that there are positive interactions between the number of different BAL cells, macrophage activity, and SF and SPH content in the preclinical model of the mouse as well as in the clinical situation after lung irradiation. 30 refs., 7 figs., 3 tabs.

  10. Shear punch testing of {sup 59}Ni isotopically-doped model austenitic alloys after irradiation in FFTF at different He/dpa ratios

    SciTech Connect

    Hankin, G.L.; Faulkner, R.G.; Hamilton, M.L.; Garner, F.A.

    1998-03-01

    A series of three model alloys, Fe-15Cr-25Ni, Fe-15Cr-25Ni-0.04P and Fe-15Cr45Ni were irradiated side-by-side in FFTF-MOTA in both the annealed and the cold worked condition in each of two variants, one using naturally occurring isotopic mixtures, and another doped with {sup 59}Ni to generate relatively high helium-to-dpa ratios. Previous papers in this series have addressed the influence of helium on radiation-induced evolution of microstructure, dimensional stability and mechanical properties, the latter using miniature-tensile specimens. In the final paper of this experimental series, three sets of irradiations conducted at different temperatures and displacement rates were examined by shear punch testing of standard microscopy disks. The results were used to determine the influence of helium generation rate, alloy starting condition, irradiation temperature and total neutron exposure. The results were also compared with the miniature tensile data obtained earlier. In general, all alloys approached saturation levels of strength and ductility that were relatively independent of He/dpa ratio and starting condition, but were sensitive to the irradiation temperature and total exposure. Some small influence of helium/dpa ratio on the shear strength is visible in the two series that ran at {approximately}490 C, but is not evident at 365 C.

  11. Effects of low energy E-beam irradiation on graphene and graphene field effect transistors and raman metrology of graphene on split gate test structures

    NASA Astrophysics Data System (ADS)

    Rao, Gayathri S.

    2011-12-01

    to the graphene Dirac bands thereby reducing the inelastic scattering and inhibiting the phonon decay medicated by SiO2 surface polar phonons (SPP). This model also explains the enhancement of n-type doping in GFETS observed for multi-step irradiation. These results highlight the impact of substrate defects and interaction of induced defectivity with the e-beam along with the role of interfacial water in impacting graphene device performance. The thesis also presents data on Raman-based characterization of graphene including layer number determination and carrier concentration measurement. Determination of layer number for graphene exfoliates focused on the splitting of the 2D Raman band. In addition, an alternate Raman-based thickness metrology was evaluated for CVD-based, polycrystalline graphene. Both were carried out on split gate test structures as a method for monolayer or bilayer confirmation in device geometries. In addition, carrier concentration measurements of exfoliates on 300nm SiO2 and split-gate test structure substrate have also been characterized with back gate biasing. These measurements made use of the stiffening of the Raman G-band with doping and the narrowing of the G-band FWHM. These results were important for validating conclusions from the e-beam irradiation experiments mentioned above regarding carrier doping.

  12. Results from irradiation tests on D0 Run 2a silicon detectors at the Radiation Damage Facility at Fermilab

    SciTech Connect

    Gardner, J.; Cerber, C.; Ke, Z.; Korjanevsky, S.; Leflat, A.; Lehner, F.; Lipton, R.; Lackey, J.; Merkin, M.; Rapidis, P.; Rykalin, V.; Shabalina, E.; Spiegel, L.; Stutte, L.; Webber, B.; /Kansas U. /Kansas State U. /Illinois U., Chicago /Fermilab /Moscow State U. /Zurich U. /NICADD, DeKalb

    2006-03-01

    Several different spare modules of the D0 experiment Silicon Microstrip Tracker (SMT) have been irradiated at the Fermilab Booster Radiation Damage Facility (RDF). The total dose received was 2.1 MRads with a proton flux of {approx} 3 {center_dot} 10{sup 11} p/cm{sup 2} sec. The irradiation was carried out in steps of 0.3 or 0.6 MRad, with several days between the steps to allow for annealing and measurements. The leakage currents and depletion voltages of the devices increased with dose, as expected from bulk radiation damage. The double sided, double metal devices showed worse degradation than the less complex detectors.

  13. Genetic studies: dominant lethal study, sex linked recessive lethal, ames mutagenicity, and heritable translocation test of thermal processed, frozen, electron irradiated, and gamma irradiated chicken. Final report Jun 76-Aug 83

    SciTech Connect

    Sullivan, D.; Lusskin, R.M.; Thomson, G.M.; Kuzdas, C.D.; Ronning, D.C.

    1983-01-01

    Four samples of chicken meat identified as the frozen control, thermally processed, gamma sterilized (5.9 Mrad), and electron sterilized (5.9 MeV), along with negative and positive controls, were evaluated for genetic activity. The samples were evaluated for ability to induce dominant lethal mutations in spermatid and spermatozoan stages of spermatogenesis in mice fed 35 percent chicken meat. The test meat samples were not observed to have an effect on the incidence of the dominant lethal mutations. However, the positive control failed to give a positive response. The meat samples were investigated for mutagenic activity employing Drosophila melanogaster in the sex linked recessive lethal test. The samples were determined to be nonmutagenic in this test and the positive control gave a significant response. Reduced production of offspring in cultures of Drosophila reared on gamma irradiated chicken which could not be overcome by the addition of vitamins was observed. Pre-incubation tests with and without added mutagens revealed that in no case was a positive result observed in the Ames test from chicken meat without an added mutagen. The manner in which chicken meat was processed had no effect upon the response to the Ames test. A heritable translocation test in mice failed to reveal any cytological evidence of translocation heterozygosity in any of the chicken-containing diets.

  14. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    NASA Astrophysics Data System (ADS)

    Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

    2016-05-01

    The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The MFF fuel operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in EBR-II experiments. Data from the MFF-3 and MFF-5 assemblies are most comparable to the data obtained from the EBR-II X447 experiment. The two X447 pin breaches were strongly influenced by fuel/cladding chemical interaction (FCCI) at the top of the fuel column. Post irradiation examination data from MFF-3 and MFF-5 are presented and compared to historical EBR-II data.

  15. Improved solar irradiance forecast with Weather Research and Forecasting model: A Sensitivity test of shallow cumulus clouds to the turbulence process

    NASA Astrophysics Data System (ADS)

    Kim, C. K.; Betterton, E. A.; Leuthold, M.; Holmgren, W.; Cronin, A.

    2014-12-01

    Accurate forecasts of solar irradiance are required for electric utilities to economically integrate substantial amounts of solar power into their power generation portfolios. A common failing of numerical weather models is the prediction of shallow cumulus clouds which are generally difficult to be resolved due to complicated processes in the planetary boundary layer. The present study carried out the sensitivity test of turbulence parameterization for better predicting solar irradiance during the shallow cumulus events near the state of Arizona by using the Weather Research and Forecasting model. The results from the simulations show that increasing the exchange coefficient leads to enhanced vertical mixing and a deeper mixed layer. At the top of mixed layer, an adiabatically ascending air parcel achieved the water vapour saturation and finally shallow cumulus is generated. A detailed analysis will be discussed in the upcoming conference.

  16. The protozoan, Paramecium primaurelia, as a non-sentient model to test laser light irradiation: The effects of an 808nm infrared laser diode on cellular respiration.

    PubMed

    Amaroli, Andrea; Ravera, Silvia; Parker, Steven; Panfoli, Isabella; Benedicenti, Alberico; Benedicenti, Stefano

    2015-07-01

    Photobiomodulation (PBM) has been used in clinical practice for more than 40 years. Unfortunately, conflicting literature has led to the labelling of PBM as a complementary or alternative medicine approach. However, past and ongoing clinical and research studies by reputable investigators have re-established the merits of PBM as a genuine medical therapy, and the technique has, in the last decade, seen an exponential increase in the numbers of clinical instruments available, and their applications. This resurgence has led to a clear need for appropriate experimental models to test the burgeoning laser technology being developed for medical applications. In this context, an ethical model that employs the protozoan, Paramecium primaurelia, is proposed. We studied the possibility of using the measure of oxygen consumption to test PBM by irradiation with an infrared or near-infrared laser. The results show that an 808nm infrared laser diode (1W; 64J/cm²) affects cellular respiration in P. primaurelia, inducing, in the irradiated cells, a significantly (p < 0.05) increased oxygen consumption of about 40%. Our findings indicate that Paramecium can be an excellent tool in biological assays involving infrared and near-infrared PBM, as it combines the advantages of in vivo results with the practicality of in vitro testing. This test represents a fast, inexpensive and straightforward assay, which offers an alternative to both traditional in vivo testing and more expensive mammalian cellular cultures.

  17. Beam Test Studies of 3D Pixel Sensors Irradiated Non-Uniformly for the ATLAS Forward Physics Detector

    DTIC Science & Technology

    2013-02-21

    d’Altes Energies (IFAE), Barcelona, Spain bCentro Nacional de Microelectronica, CNM-IMB (CSIC), Barcelona , Spain cFondazione Bruno Kessler, FBK-CMM, Trento...Devices that show good electri - cal behavior before irradiation are able to sustain the voltage needed to achieve excellent efficiency (> 98% at...Research (ONR). The work at SCIPP was supported by Department of Energy , grant DE-FG02-04ER41286. References [1] The ATLAS Collaboration, “The ATLAS

  18. Remote fabrication and irradiation test of recycled nuclear fuel prepared by the oxidation and reduction of spent oxide fuel

    NASA Astrophysics Data System (ADS)

    Jin Ryu, Ho; Chan Song, Kee; Il Park, Geun; Won Lee, Jung; Seung Yang, Myung

    2005-02-01

    A direct dry recycling process was developed in order to reuse spent pressurized light water reactor (LWR) nuclear fuel in CANDU reactors without the separation of sensitive nuclear materials such as plutonium. The benefits of the dry recycling process are the saving of uranium resources and the reduction of spent fuel accumulation as well as a higher proliferation resistance. In the process of direct dry recycling, fuel pellets separated from spent LWR fuel rods are oxidized from UO2 to U3O8 at 500 °C in an air atmosphere and reduced into UO2 at 700 °C in a hydrogen atmosphere, which is called OREOX (oxidation and reduction of oxide fuel). The pellets are pulverized during the oxidation and reduction processes due to the phase transformation between cubic UO2 and orthorhombic U3O8. Using the oxide powder prepared from the OREOX process, the compaction and sintering processes are performed in a remote manner in a shielded hot cell due to the high radioactivity of the spent fuel. Most of the fission gas and volatile fission products are removed during the OREOX and sintering processes. The mini-elements fabricated by the direct dry recycling process are irradiated in the HANARO research reactor for the performance evaluation of the recycled fuel pellets. Post-irradiation examination of the irradiated fuel showed that microstructural evolution and fission gas release behavior of the dry-recycled fuel were similar to high burnup UO2 fuel.

  19. Irradiation Creep in Graphite

    SciTech Connect

    Ubic, Rick; Butt, Darryl; Windes, William

    2014-03-13

    An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.

  20. Phytosanitary Irradiation

    PubMed Central

    Hallman, Guy J.; Blackburn, Carl M.

    2016-01-01

    Phytosanitary treatments disinfest traded commodities of potential quarantine pests. Phytosanitary irradiation (PI) treatments use ionizing radiation to accomplish this, and, since their international commercial debut in 2004, the use of this technology has increased by ~10% annually. Generic PI treatments (one dose is used for a group of pests and/or commodities, although not all have been tested for efficacy) are used in virtually all commercial PI treatments, and new generic PI doses are proposed, such as 300 Gy, for all insects except pupae and adult Lepidoptera (moths). Fresh fruits and vegetables tolerate PI better than any other broadly used treatment. Advances that would help facilitate the use of PI include streamlining the approval process, making the technology more accessible to potential users, lowering doses and broadening their coverage, and solving potential issues related to factors that might affect efficacy. PMID:28231103

  1. Evaluation of critical resolved shear strength and deformation mode in proton-irradiated austenitic stainless steel using micro-compression tests

    NASA Astrophysics Data System (ADS)

    Jin, Hyung-Ha; Ko, Eunsol; Kwon, Junhyun; Hwang, Seong Sik; Shin, Chansun

    2016-03-01

    Micro-compression tests were applied to evaluate the changes in the strength and deformation mode of proton-irradiated commercial austenitic stainless steel. Proton irradiation generated small dots at low dose levels and Frank loops at high dose levels. The increase in critical resolved shear stresses (CRSS) was measured from micro-compression of pillars and the Schmid factor calculated from the measured loading direction. The magnitudes of the CRSS increase were in good agreement with the values calculated from the barrier hardening model using the measured size and density of radiation defects. The deformation mode changed upon increasing the irradiation dose level. At a low radiation dose level, work hardening and smooth flow behavior were observed. Increasing the dose level resulted in the flow behavior changing to a distinct heterogeneous flow, yielding a few large strain bursts in the stress-strain curves. The change in the deformation mode was related to the formation and propagation of defect-free slip bands. The effect of the orientation of the pillar or loading direction on the strengths is discussed.

  2. Study of the effects of low-fluence laser irradiation on wall paintings: Test measurements on fresco model samples

    NASA Astrophysics Data System (ADS)

    Raimondi, Valentina; Cucci, Costanza; Cuzman, Oana; Fornacelli, Cristina; Galeotti, Monica; Gomoiu, Ioana; Lognoli, David; Mohanu, Dan; Palombi, Lorenzo; Picollo, Marcello; Tiano, Piero

    2013-11-01

    Laser-induced fluorescence is widely applied in several fields as a diagnostic tool to characterise organic and inorganic materials and could be also exploited for non-invasive remote investigation of wall paintings using the fluorescence lidar technique. The latter relies on the use of a low-fluence pulsed UV laser and a telescope to carry out remote spectroscopy on a given target. A first step to investigate the applicability of this technique is to assess the effects of low-fluence laser radiation on wall paintings. This paper presents a study devoted to investigate the effects of pulsed UV laser radiation on a set of fresco model samples prepared using different pigments. To irradiate the samples we used a tripled-frequency Q-switched Nd:YAG laser (emission wavelength: 355 nm; pulse width: 5 ns). We varied the laser fluence from 0.1 mJ/cm2 to 1 mJ/cm2 and the number of laser pulses from 1 to 500 shots. We characterised the investigated materials using several diagnostic and analytical techniques (colorimetry, optical microscopy, fibre optical reflectance spectroscopy and ATR-FT-IR microscopy) to compare the surface texture and their composition before and after laser irradiation. Results open good prospects for a non-invasive investigation of wall paintings using the fluorescence lidar technique.

  3. Simulated Irradiation of Samples in HFIR for use as Possible Test Materials in the MPEX (Material Plasma Exposure Experiment) Facility

    SciTech Connect

    Ellis, Ronald James; Rapp, Juergen

    2014-01-01

    The importance of Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) facility will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. The project presented in this paper involved performing assessments of the induced radioactivity and resulting radiation fields of a variety of potential fusion reactor materials. The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR; generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. These state-of-the-art simulation methods were used in addressing the challenge of the MPEX project to minimize the radioactive inventory in the preparation of the samples for inclusion in the MPEX facility.

  4. Food irradiation: Activities and potentialities

    NASA Astrophysics Data System (ADS)

    Doellstaedt, R.; Huebner, G.

    After the acceptance of food irradiation up to an overall average dose of 10 kGy recommended by the Joint FAO/IAEA/WHO Expert Committee on the Wholesomeness of Irradiated Food in October 1980, the G.D.R. started a programme for the development of techniques for food irradiation. A special onion irradiator was designed and built as a pilot plant for studying technological and economic parameters of the irradiation of onions. The new principle of bulk-cargo irradiation allows the integration of this technology into the usual harvest technology for onions on the way from field to storage. Scientific and applied research work has been carried out in the past 3 yr on the irradiation of spices, potatoes, eviscerated chicken, animal feeds, fodder yeast, drugs and vaccines. In connection with the irradiation of eviscerated chicken, fodder yeast and animal feeds the basis of an antisalmonella programme has been discussed. Germ-count-reduced spices were employed for the production of test charges of preserves and tinned products. The results have led to the decision to design and build a new multipurpose irradiator for food irradiation. In order to cover the legal aspects of food irradiation the Ministry of Health issued regulations concerning the recommendation of irradiated food in the G.D.R.

  5. Development and high temperature testing by 14 MeV neutron irradiation of single crystal diamond detectors

    NASA Astrophysics Data System (ADS)

    Pilotti, R.; Angelone, M.; Pagano, G.; Loreti, S.; Pillon, M.; Sarto, F.; Marinelli, M.; Milani, E.; Prestopino, G.; Verona, C.; Verona-Rinati, G.

    2016-06-01

    In the present paper, the performances of single crystal diamond detectors "ad hoc" designed to operate at high temperature are reported. The detectors were realized using commercial CVD single crystal diamond films, 500 micron thick with metal contacts deposited by sputtering method on each side. The new detector layout is based upon mechanical contacts between the diamond film and the electric ground. The detector was first characterized by measuring the leakage current as function of temperature and applied biasing voltage (I-V characteristics). The results obtained using two different metal contacts, Pt and Ag respectively, while irradiated with 14 MeV neutrons at the Frascati neutron generator (FNG) are reported and compared. It is shown that diamond detectors with Ag metal contacts can be properly operated in spectrometric mode up to 240oC with energy resolution (FWHM) of about 3.5%.

  6. Food irradiation in perspective

    NASA Astrophysics Data System (ADS)

    Henon, Y. M.

    1995-02-01

    Food irradiation already has a long history of hopes and disappointments. Nowhere in the world it plays the role that it should have, including in the much needed prevention of foodborne diseases. Irradiated food sold well wherever consumers were given a chance to buy them. Differences between national regulations do not allow the international trade of irradiated foods. While in many countries food irradiation is still illegal, in most others it is regulated as a food additive and based on the knowledge of the sixties. Until 1980, wholesomeness was the big issue. Then the "prerequisite" became detection methods. Large amounts of money have been spent to design and validate tests which, in fact, aim at enforcing unjustified restrictions on the use of the process. In spite of all the difficulties, it is believed that the efforts of various UN organizations and a growing legitimate demand for food safety should in the end lead to recognition and acceptance.

  7. Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

    SciTech Connect

    Carmack, W. J.; Chichester, H. M.; Porter, D. L.; Wootan, D. W.

    2016-05-01

    Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.

  8. Slag recycling of irradiated vanadium

    SciTech Connect

    Gorman, Patrick K.

    1995-04-05

    An experimental inductoslag apparatus to recycle irradiated vanadium was fabricated and tested. An experimental electroslag apparatus was also used to test possible slags. The testing was carried out with slag materials that were fabricated along with impurity bearing vanadium samples. Results obtained include computer simulated thermochemical calculations and experimentally determined removal efficiencies of the transmutation impurities. Analyses of the samples before and after testing were carried out to determine if the slag did indeed remove the transmutation impurities from the irradiated vanadium.

  9. Distribution of Pd, Ag & U in the SiC Layer of an Irradiated TRISO Fuel Particle

    SciTech Connect

    Thomas M. Lillo; Isabella J. van Rooyen

    2014-08-01

    The distribution of silver, uranium and palladium in the silicon carbide (SiC) layer of an irradiated TRISO fuel particle was studied using samples extracted from the SiC layer using focused ion beam (FIB) techniques. Transmission electron microscopy in conjunction with energy dispersive x-ray spectroscopy was used to identify the presence of the specific elements of interest at grain boundaries, triple junctions and precipitates in the interior of SiC grains. Details on sample fabrication, errors associated with measurements of elemental migration distances and the distances migrated by silver, palladium and uranium in the SiC layer of an irradiated TRISO particle from the AGR-1 program are reported.

  10. ORNL irradiation creep facility

    SciTech Connect

    Reiley, T.C.; Auble, R.L.; Beckers, R.M.; Bloom, E.E.; Duncan, M.G.; Saltmarsh, M.J.; Shannon, R.H.

    1980-09-01

    A machine was developed at ORNL to measure the rates of elongation observed under irradiation in stressed materials. The source of radiation is a beam of 60 MeV alpha particles from the Oak Ridge Isochronous Cyclotron (ORIC). This choice allows experiments to be performed which simulate the effects of fast neutrons. A brief review of irradiation creep and experimental constraints associated with each measurement technique is given. Factors are presented which lead to the experimental choices made for the Irradiation Creep Facility (ICF). The ICF consists of a helium-filled chamber which houses a high-precision mechanical testing device. The specimen to be tested must be thermally stabilized with respect to the temperature fluctuations imposed by the particle beam which passes through the specimen. Electrical resistance of the specimen is the temperature control parameter chosen. Very high precision in length measurement and temperature control are required to detect the small elongation rates relevant to irradiation creep in the test periods available (approx. 1 day). The apparatus components and features required for the above are presented in some detail, along with the experimental procedures. The damage processes associated with light ions are discussed and displacement rates are calculated. Recent irradiation creep results are given, demonstrating the suitability of the apparatus for high resolution experiments. Also discussed is the suitability of the ICF for making high precision thermal creep measurements.

  11. [Food irradiation].

    PubMed

    Migdał, W

    1995-01-01

    A worldwide standard on food irradiation was adopted in 1983 by Codex Alimentarius Commission of the Joint Food Standard Programme of the Food and Agriculture Organization (FAO) of the United Nations and the World Health Organization (WHO). As a result, 41 countries have approved the use of irradiation for treating one or more food items and the number is increasing. Generally, irradiation is used to: food loses, food spoilage, disinfestation, safety and hygiene. The number of countries which use irradiation for processing food for commercial purposes has been increasing steadily from 19 in 1987 to 33 today. In the frames of the national programme on the application of irradiation for food preservation and hygienization an experimental plant for electron beam processing has been established in Institute of Nuclear Chemistry and Technology. The plant is equipped with a small research accelerator Pilot (19MeV, 1 kW) and an industrial unit Elektronika (10MeV, 10 kW). On the basis of the research there were performed at different scientific institutions in Poland, health authorities have issued permission for irradiation for: spices, garlic, onions, mushrooms, potatoes, dry mushrooms and vegetables.

  12. Tissue irradiator

    DOEpatents

    Hungate, F.P.; Riemath, W.F.; Bunnell, L.R.

    1975-12-16

    A tissue irradiator is provided for the in-vivo irradiation of body tissue. The irradiator comprises a radiation source material contained and completely encapsulated within vitreous carbon. An embodiment for use as an in- vivo blood irradiator comprises a cylindrical body having an axial bore therethrough. A radioisotope is contained within a first portion of vitreous carbon cylindrically surrounding the axial bore, and a containment portion of vitreous carbon surrounds the radioisotope containing portion, the two portions of vitreous carbon being integrally formed as a single unit. Connecting means are provided at each end of the cylindrical body to permit connections to blood- carrying vessels and to provide for passage of blood through the bore. In a preferred embodiment, the radioisotope is thulium-170 which is present in the irradiator in the form of thulium oxide. A method of producing the preferred blood irradiator is also provided, whereby nonradioactive thulium-169 is dispersed within a polyfurfuryl alcohol resin which is carbonized and fired to form the integral vitreous carbon body and the device is activated by neutron bombardment of the thulium-169 to produce the beta-emitting thulium-170.

  13. Proton irradiation SEU test results for the SEDS MIL-STD-1773 fiber optic data bus: integrated optoelectronics

    NASA Astrophysics Data System (ADS)

    LaBel, Kenneth A.; Stassinopoulos, E. G.; Marshall, Paul W.; Petersen, Ed L.; Dale, Cheryl J.; Crabtree, Christina M.; Stauffer, Craig A.

    1993-09-01

    The Small Explorer Data System (SEDS) is a spaceflight command and data handling system for the small explorer (SMEX) program at Goddard Space Flight Center (GSFC). A key component in this system is the SEDS MIL-STD-1773 Fiber Optic Multiplexed Data Bus. The 1773 bus provides a means of passing telemetry and commands between spacecraft subsystems. This bus is currently being considered for additional spaceflight programs inside and outside of the NASA realm. The SEDS 1773 bus uses integrated optoelectronics as part of its electrical subsystem (or user) to optical interface. Generic proton and heavy ion test results have been previously reported. Herein is presented proton test results for continuing this investigation under actual subsystem interface conditions (MIL-STD-1773) as well as for generic devices using the proton test facilities at University of California, Davis (UCD). This testing was undertaken as a joint effort between NASA/GSFC and the Naval Research Laboratories (NRL).

  14. FY 2013 Summary Report: Post-Irradiation Examination of Zircaloy-4 Samples in Target Capsules and Initiation of Bending Fatigue Testing for Used Nuclear Fuel Vibration Integrity Investigations

    SciTech Connect

    Howard, Richard H.; Yan, Yong; Wang, Jy-An John; Ott, Larry J.; Howard, Rob L.

    2013-10-01

    This report documents ongoing work performed at Oak Ridge National Laboratory (ORNL) for the Department of Energy, Office of Fuel Cycle Technology Used Fuel Disposition Campaign (UFDC), and satisfies the deliverable for milestone M2FT-13OR0805041, “Data Report on Hydrogen Doping and Irradiation in HFIR.” This work is conducted under WBS 1.02.08.05, Work Package FT-13OR080504 ST “Storage and Transportation-Experiments – ORNL.” The objectives of work packages that make up the S&T Experiments Control Account are to conduct the separate effects tests (SET) and small-scale tests that have been identified in the Used Nuclear Fuel Storage and Transportation Data Gap Prioritization (FCRD-USED-2012-000109). In FY 2013, the R&D focused on cladding and container issues and small-scale tests as identified in Sections A-2.9 and A-2.12 of the prioritization report.

  15. Irradiation subassembly

    DOEpatents

    Seim, O.S.; Filewicz, E.C.; Hutter, E.

    1973-10-23

    An irradiation subassembly for use in a nuclear reactor is described which includes a bundle of slender elongated irradiation -capsules or fuel elements enclosed by a coolant tube and having yieldable retaining liner between the irradiation capsules and the coolant tube. For a hexagonal bundle surrounded by a hexagonal tube the yieldable retaining liner may consist either of six segments corresponding to the six sides of the tube or three angular segments each corresponding in two adjacent sides of the tube. The sides of adjacent segments abut and are so cut that metal-tometal contact is retained when the volume enclosed by the retaining liner is varied and Springs are provided for urging the segments toward the center of the tube to hold the capsules in a closely packed configuration. (Official Gazette)

  16. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    SciTech Connect

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  17. Schedule and status of irradiation experiments

    SciTech Connect

    Rowcliffe, A.F.; Grossbeck, M.L.

    1997-04-01

    To provide an updated summary of the status of irradiation experiments for the neutron-interactive materials program. The current status of reactor irradiation experiments is presented in tables summarizing the experimental objectives, conditions, and schedule. Currently, the program has two irradiation experiments in reactor; and 8 experiments in the planning or design stages. Postirradiation examination and testing is in progress on 18 experiments.

  18. Irradiated foods

    MedlinePlus

    ... it reduces the risk of food poisoning . Food irradiation is used in many countries. It was first approved in the U.S. to prevent sprouts on white potatoes, and to control insects on wheat and in certain spices and seasonings.

  19. The RBE-LET relationship for rodent intestinal crypt cell survival, testes weight loss, and multicellular spheroid cell survival after heavy-ion irradiation

    NASA Technical Reports Server (NTRS)

    Rodriguez, A.; Alpen, E. L.; Powers-Risius, P.

    1992-01-01

    This report presents data for survival of mouse intestinal crypt cells, mouse testes weight loss as an indicator of survival of spermatogonial stem cells, and survival of rat 9L spheroid cells after irradiation in the plateau region of unmodified particle beams ranging in mass from 4He to 139La. The LET values range from 1.6 to 953 keV/microns. These studies examine the RBE-LET relationship for two normal tissues and for an in vitro tissue model, multicellular spheroids. When the RBE values are plotted as a function of LET, the resulting curve is characterized by a region in which RBE increases with LET, a peak RBE at an LET value of 100 keV/microns, and a region of decreasing RBE at LETs greater than 100 keV/microns. Inactivation cross sections (sigma) for these three biological systems have been calculated from the exponential terminal slope of the dose-response relationship for each ion. For this determination the dose is expressed as particle fluence and the parameter sigma indicates effect per particle. A plot of sigma versus LET shows that the curve for testes weight loss is shifted to the left, indicating greater radiosensitivity at lower LETs than for crypt cell and spheroid cell survival. The curves for cross section versus LET for all three model systems show similar characteristics with a relatively linear portion below 100 keV/microns and a region of lessened slope in the LET range above 100 keV/microns for testes and spheroids. The data indicate that the effectiveness per particle increases as a function of LET and, to a limited extent, Z, at LET values greater than 100 keV/microns. Previously published results for spread Bragg peaks are also summarized, and they suggest that RBE is dependent on both the LET and the Z of the particle.

  20. The irradiation effects on zirconium alloys

    NASA Astrophysics Data System (ADS)

    Negut, Gh.; Ancuta, M.; Radu, V.; Ionescu, S.; Stefan, V.; Uta, O.; Prisecaru, I.; Danila, N.

    2007-05-01

    Pressure tube samples were irradiated under helium atmosphere in the TRIGA Steady State Research and Material Test Reactor of the Romanian Institute for Nuclear Research (INR). These samples are made of the Zr-2.5%Nb alloy used as structural material for the CANDU Romanian power reactors. After irradiation, mechanical tests were performed in the Post Irradiation Examination Laboratory (PIEL) to study the influence of irradiation on zirconium alloys mechanical behaviour. The tensile test results were used for structural integrity assessment. Results of the tests are presented. The paper presents, also, pressure tube structural integrity assessment.

  1. Status of Post Irradiation Examination of FCAB and FCAT Irradiation Capsules

    SciTech Connect

    Field, Kevin G.; Yamamoto, Yukinori; Howard, Richard H.

    2016-09-29

    A series of irradiation programs are ongoing to address the need for determining the radiation tolerance of FeCrAl alloys. These irradiation programs, deemed the FCAT and FCAB irradiation programs, use the High Flux Isotope Reactor (HFIR) to irradiate second generation wrought FeCrAl alloys and early-generation powder-metallurgy (PM) oxide dispersion-strengthened (ODS) FeCrAl alloys. Irradiations have been or are being performed at temperatures of 200°C, 330°C, and 550°C from doses of 1.8 dpa up to 16 dpa. Preliminary post-irradiation examination (PIE) on low dose (<2 dpa) irradiation capsules of tensile specimens has been performed. Analysis of co-irradiated SiC thermometry have shown reasonable matching between the nominal irradiation temperatures and the target irradiation temperatures. Room temperature tensile tests have shown typical radiation-induced hardening and embrittlement at irradiations of 200°C and 330°C, but a propensity for softening when irradiated to 550°C for the wrought alloys. The PM-ODS FeCrAl specimens showed less hardening compared to the wrought alloys. Future PIE includes high temperature tensile tests on the low dose irradiation capsules as well as the determination of reference fracture toughness transition temperature, To, in alloys irradiated to 7 dpa and higher.

  2. Key differences in the fabrication, irradiation and high temperature accident testing of US and German TRISO-coated particle fuel, and their implications on fuel performance

    SciTech Connect

    Petti, David Andrew; Buongiorno, Jacopo; Maki, John Thomas; Hobbins, Richard Redfield

    2003-06-01

    Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the US. German fuel generally has displayed gas release values during irradiation three orders of magnitude lower than US fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the US and Germany and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the US fuel has not faired as well, and what process/production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer US irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, degree of acceleration) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance.

  3. Enhancement of Irradiation Capability of the Experimental Fast Reactor Joyo

    NASA Astrophysics Data System (ADS)

    Maeda, Shigetaka; Serine, Takashi; Aoyama, Takafumi; Suzuki, Soju

    2009-08-01

    The experimental fast reactor Joyo is the first sodium-cooled fast reactor in Japan. One of its primary missions is to perform irradiation tests of fuel and structural materials to support the development of fast reactors. The MK-III high performance core upgrade to enhance the irradiation testing capabilities was completed in 2003. In order to expand Joyo's capabilities for innovative irradiation testing applications, neutron spectrum tailoring, lower irradiation temperature, movable sample devices and fast neutron beam holes are being considered. This program responds to existing irradiation needs and aims to further expand capabilities for a variety of irradiation tests.

  4. NSUF Irradiated Materials Library

    SciTech Connect

    Cole, James Irvin

    2015-09-01

    The Nuclear Science User Facilities has been in the process of establishing an innovative Irradiated Materials Library concept for maximizing the value of previous and on-going materials and nuclear fuels irradiation test campaigns, including utilization of real-world components retrieved from current and decommissioned reactors. When the ATR national scientific user facility was established in 2007 one of the goals of the program was to establish a library of irradiated samples for users to access and conduct research through competitively reviewed proposal process. As part of the initial effort, staff at the user facility identified legacy materials from previous programs that are still being stored in laboratories and hot-cell facilities at the INL. In addition other materials of interest were identified that are being stored outside the INL that the current owners have volunteered to enter into the library. Finally, over the course of the last several years, the ATR NSUF has irradiated more than 3500 specimens as part of NSUF competitively awarded research projects. The Logistics of managing this large inventory of highly radioactive poses unique challenges. This document will describe materials in the library, outline the policy for accessing these materials and put forth a strategy for making new additions to the library as well as establishing guidelines for minimum pedigree needed to be included in the library to limit the amount of material stored indefinitely without identified value.

  5. Key Differences in the Fabrication, Irradiation, and Safety Testing of U.S. and German TRISO-coated Particle Fuel and Their Implications on Fuel Performance

    SciTech Connect

    Petti, David Andrew; Maki, John Thomas; Buongiorno, Jacopo; Hobbins, Richard Redfield

    2002-06-01

    High temperature gas reactor technology is achieving a renaissance around the world. This technology relies on high quality production and performance of coated particle fuel. Historically, the irradiation performance of TRISO-coated gas reactor particle fuel in Germany has been superior to that in the United States. German fuel generally displayed in-pile gas release values that were three orders of magnitude lower than U.S. fuel. Thus, we have critically examined the TRISO-coated fuel fabrication processes in the U.S. and Germany and the associated irradiation database with a goal of understanding why the German fuel behaves acceptably, why the U.S. fuel has not faired as well, and what process/ production parameters impart the reliable performance to this fuel form. The postirradiation examination results are also reviewed to identify failure mechanisms that may be the cause of the poorer U.S. irradiation performance. This comparison will help determine the roles that particle fuel process/product attributes and irradiation conditions (burnup, fast neutron fluence, temperature, and degree of acceleration) have on the behavior of the fuel during irradiation and provide a more quantitative linkage between acceptable processing parameters, as-fabricated fuel properties and subsequent in-reactor performance.

  6. Passive SiC irradiation temperature monitor

    SciTech Connect

    Youngblood, G.E.

    1996-04-01

    A new, improved passive irradiation temperature monitoring method was examined after an irradiation test at 627{degrees}C. The method is based on the analysis of thermal diffusivity changes during postirradiation annealing of polycrystalline SiC. Based on results from this test, several advantages for using this new method rather than a method based on length or lattice parameter changes are given.

  7. New facility for post irradiation examination of neutron irradiated beryllium

    SciTech Connect

    Ishitsuka, Etsuo; Kawamura, Hiroshi

    1995-09-01

    Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800{degrees}C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and {sup 60}Co;7.4 MBq/day.

  8. Batch-Equilibrium Hot-Cell Tests of Caustic-Side Solvent Extraction (CSSX) with SRS Simulant Waste and Internal 137Cs Irradiation

    SciTech Connect

    Spence, R.D.

    2001-09-17

    The solvent was loaded with {sup 137}Cs and subsamples were stored on a shaker table while in contact with the extract, scrub, or strip aqueous phases. Evidence of solvent degradation was evaluated at exposure times of 0, 20, 54, and 83 days. This resulted in estimated solvent doses ranging up to 1.24 Mrad, equivalent to the dose expected to be received during 16.5 years of operation at the plant proposed for the Savannah River Site. The break times and distribution of cesium of the batch samples remained constant within experimental error; in addition, no third-phase formation was observed. The solvent concentrations of calix[4]arene-bis-(tert-octylbenzo-crown-6) and 1-(2,2,3,3-tetra-fluoroproproxy)-3-(4-sec-butylphenoxy)-2-propanol remained constant within experimental error. Solvent degradation with irradiation was evidenced by a decrease in the trioctylamine (TOA) concentration in the solvent and an increase in the solvent concentration of the degradation product 4-sec-butylphenol. No decline in extraction or scrubbing performance of the irradiated solvents was observed. The stripping performance of the solvent was seriously impaired with irradiation; however, a mild caustic wash and replenishment of the TOA concentration restored the ability to strip the irradiated solvent.

  9. Interaction between age of irradiation and age of testing in the disruption of operant performance using a ground-based model for exposure to cosmic rays

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Exposure to HZE particles produces deficits in cognitive performance. While previous research has shown a progressive deterioration in cognitive performance in radiated rats as a function of age, the present experiment was designed to evaluate the effects of age of irradiation independently of the ...

  10. Irradiation test of tungsten clad uranium carbide-zirconium carbide ((U,Zr)C) specimens for thermionic reactor application at conditions conductive to long-term performance

    NASA Technical Reports Server (NTRS)

    Creagh, J. W. R.; Smith, J. R.

    1973-01-01

    Uranium carbide fueled, thermionic emitter configurations were encapsulated and irradiated. One capsule contained a specimen clad with fluoride derived chemically vapor deposited (CVD) tungsten. The other capsule used a duplex clad specimen consisting of chloride derived on floride derived CVD tungsten. Both fuel pins were 16 millimeters in diameter and contained a 45.7-millimeter length of fuel.

  11. Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor

    SciTech Connect

    Khericha, Soli T

    2002-06-01

    This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to ~42 GWd/MT burnup (+ 2.5% as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: ~50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies (@ ~40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches ~40 GWd/MT burnup per MCNP-predicted values.

  12. Experiment Safety Assurance Package for the 40- to 50-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-Hole Positions in the Advanced Test Reactor

    SciTech Connect

    Khericha, S.T.

    2002-06-30

    This experiment safety assurance package (ESAP) is a revision of the last MOX ESAP issued in February 2001(Khericha 2001). The purpose of this revision is to identify the changes in the loading pattern and to provide a basis to continue irradiation up to {approx}42 GWd/MT burnup (+ 2.5%) as predicted by MCNP (Monte Carlo N-Particle) transport code before the preliminary postirradiation examination (PIE) results for 40 GWd/MT burnup are available. Note that the safety analysis performed for the last ESAP is still applicable and no additional analysis is required (Khericha 2001). In July 2001, it was decided to reconfigure the test assembly using the loading pattern for Phase IV, Part 3, at the end of Phase IV, Part 1, as the loading pattern for Phase IV, Parts 2 and 3. Three capsule assemblies will be irradiated until the highest burnup capsule assembly accumulates: {approx}50 GWd/MT burnup, based on the MCNP code predictions. The last ESAP suggests that at the end of Phase IV, Part 1, we remove the two highest burnup capsule assemblies ({at} {approx}40 GWd/MT burnup) and send them to ORNL for PIE. Then, irradiate the test assembly using the loading pattern for Phase IV, Part 2, until the highest burnup capsule reaches {approx}40 GWd/MT burnup per MCNP-predicted values.

  13. Electron spin resonance identification of irradiated fruits

    NASA Astrophysics Data System (ADS)

    Raffi, Jacques J.; Agnel, Jean-Pierre L.

    The electron spin resonance spectrum of achenes, pips, stalks and stones from irradiated fruits (strawberry, raspberry, red currant, bilberry, apple, pear, fig, french prune, kiwi, water-melon and cherry) always displays, just after γ-treatment, a weak triplet ( aH≈30 G) due to a cellulose radical; its left line (lower field) can be used as an identification test of irradiation, at least for strawberries, rapsberries, red currants or bilberries irradiated in order to improve their storage time.

  14. Use of Irradiated Foods

    NASA Technical Reports Server (NTRS)

    Brynjolfsson, A.

    1985-01-01

    The safety of irradiated foods is reviewed. Guidelines and regulations for processing irradiated foods are considered. The radiolytic products formed in food when it is irradiated and its wholesomeness is discussed. It is concluded that food irradiation processing is not a panacea for all problems in food processing but when properly used will serve the space station well.

  15. RERTR-8 Irradiation Summary Report

    SciTech Connect

    D. M. Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-12-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-8, was designed to test monolithic mini-fuel plates fabricated via hot isostatic pressing (HIP), the effect of molybdenum (Mo) content on the monolithic fuel behavior, and the efficiency of ternary additions to dispersion fuel particles on the interaction layer behavior at higher burnup. The following report summarizes the life of the RERTR-8 experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.

  16. RERTR-6 Irradiation Summary Report

    SciTech Connect

    D. M. Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-12-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-6 was designed to evaluate several modified fuel designs that were proposed to address the possibility of breakaway swelling due to porosity within the (U. Mo) Al interaction product observed in the full-size plate tests performed in Russia and France1. The following report summarizes the life of the RERTR-6 experiment through end of irradiation, including as-run neutronic analyses, thermal analyses and hydraulic testing results.

  17. Tensile properties of irradiated surveillance coupons

    SciTech Connect

    Huang, F.H.; Blackburn, L.D.

    1994-06-01

    Tensile testing of austenitic steel and superalloy samples irradiated in the HMO 13 assembly was performed in support of the Fast Flux Test Facility (FFTF) Surveillance Program. Postirradiation yield stress, ultimate tensile stress, uniform elongation, total elongation, and reduction in area of 304 stainless steel (SS), 308 SS weld, 316 SS, A286, In718, and In718 weld were determined. Results showed the strength of austenitic steels increased while the ductility decreased as a result of irradiation. Low irradiation exposure produced little property change in In718. Overall, the tensile properties of HMO 13 surveillance coupons showed a lower magnitude of irradiation-induced property change than was expected based on earlier studies. Results from these tests gave no indications of unexpectedly severe irradiation damage to FFTF components.

  18. Study of irradiation creep of vanadium alloys

    SciTech Connect

    Tsai, H.; Strain, R.V.; Smith, D.L.

    1997-08-01

    Thin-wall tubing was produced from the 832665 (500 kg) heat of V-4 wt.% Cr-4 wt.% Ti to study its irradiation creep behavior. The specimens, in the form of pressurized capsules, were irradiated in Advanced Test Reactor and High Flux Isotope Reactor experiments (ATR-A1 and HFIR RB-12J, respectively). The ATR-A1 irradiation has been completed and specimens from it will soon be available for postirradiation examination. The RB-12J irradiation is not yet complete.

  19. Healing burns using atmospheric pressure plasma irradiation

    NASA Astrophysics Data System (ADS)

    Hirata, Takamichi; Kishimoto, Takumi; Tsutsui, Chihiro; Kanai, Takao; Mori, Akira

    2014-01-01

    An experiment testing the effects of plasma irradiation with an atmospheric-pressure plasma (APP) reactor on rats given burns showed no evidence of electric shock injuries upon pathology inspection of the irradiated skin surface. In fact, the observed evidence of healing and improvement of the burns suggested healing effects from plasma irradiation. The quantities of neovascular vessels in the living tissues at 7 days were 9.2 ± 0.77 mm-2 without treatment and 18.4 ± 2.9 mm-2 after plasma irradiation.

  20. Cherry Irradiation Studies. 1984 annual report

    SciTech Connect

    Eakin, D.E.; Hungate, F.P.; Tingey, G.L.; Olsen, K.L.; Fountain, J.B.; Burditt, A.K. Jr.; Moffit, H.R.; Johnson, D.A.; Lunden, J.D.

    1985-04-01

    Fresh cherries, cherry fruit fly larvae, and codling moth larvae were irradiated using the PNL cobalt-60 facility to determine the efficacy of irradiation treatment for insect disinfestation and potential shelf life extension. Irradiation is an effective disinfestation treatment with no significant degradation of fruit at doses well above those required for quarantine treatment. Sufficient codling moth control was achieved at projected doses of less than 25 krad; cherry fruit fly control, at projected doses of less than 15 krad. Dose levels up to 60 krad did not adversely affect cherry quality factors tested. Irradiation above 60 krad reduced the firmness of cherries but had no significant impact on other quality factors tested. Irradiation of cherries below 80 krad did not result in any significant differences in sensory evaluations (appearance, flavor, and firmness) in tests conducted at OSU. Irradiation up to 200 krad at a temperature of about 25/sup 0/C (77/sup 0/F) did not measurably extend shelf life. Irradiation at 500 krad at 25/sup 0/C (77/sup 0/F) increased mold and rotting of cherries tested. There is no apparent advantage of irradiation over low-temperature fumigation.

  1. Reprocessing technology development for irradiated beryllium

    SciTech Connect

    Kawamura, H.; Sakamoto, N.; Tatenuma, K.

    1995-09-01

    At present, beryllium is under consideration as a main candidate material for neutron multiplier and plasma facing material in a fusion reactor. Therefore, it is necessary to develop the beryllium reprocessing technology for effective resource use. And, we have proposed reprocessing technology development on irradiated beryllium used in a fusion reactor. The preliminary reprocessing tests were performed using un-irradiated and irradiated beryllium. At first, we performed beryllium separation tests using un-irradiated beryllium specimens. Un-irradiated beryllium with beryllium oxide which is a main impurity and some other impurities were heat-treated under chlorine gas flow diluted with Ar gas. As the results high purity beryllium chloride was obtained in high yield. And it appeared that beryllium oxide and some other impurities were removed as the unreactive matter, and the other chloride impurities were separated by the difference of sublimation temperature on beryllium chloride. Next, we performed some kinds of beryllium purification tests from beryllium chloride. And, metallic beryllium could be recovered from beryllium chloride by the reduction with dry process. In addition, as the results of separation and purification tests using irradiated beryllium specimens, it appeared that separation efficiency of Co-60 from beryllium was above 96%. It is considered that about 4% Co-60 was carried from irradiated beryllium specimen in the form of cobalt chloride. And removal efficiency of tritium from irradiated beryllium was above 95%.

  2. Schedule and status of irradiation experiments

    SciTech Connect

    Rowcliffe, A.F.; Grossbeck, M.L.; Robertson, J.P.

    1998-09-01

    The current status of reactor irradiation experiments is presented in tables summarizing the experimental objectives, conditions, and schedule. Currently, the program has one irradiation experiment in reactor and five experiments in the design or construction stages. Postirradiation examination and testing is in progress on ten experiments.

  3. Schedule and status of irradiation experiments

    SciTech Connect

    Rowcliffe, A.F.; Grossbeck, M.L.; Robertson, J.P.

    1998-03-01

    The current status of reactor irradiation experiments is presented in tables summarizing the experimental objectives, conditions, and schedule. Currently, the program has four irradiation experiments in reactor, and five experiments in the design or construction stages. Postirradiation examination and testing is in progress on ten experiments.

  4. Methods for detection of irradiation of spices.

    PubMed

    Sjöberg, A M; Manninen, M; Härmälä, P; Pinnioja, S

    1990-02-01

    Three types of methods for the identification of irradiation of spices were tested as potential control methods. The methods were microbiological, combining a direct epifluorescent filter technique (DEFT) with a total aerobic plate count (APC), a chemiluminescence method and chemical gas-chromatographic (GC) and GC mass-spectrometric (MS) methods for analysis of volatile oils of spices isolated by steam distillation. Twelve samples of spices, mainly peppers, were analysed before and after gamma-irradiation with doses of 10 and 50 kGy. The chemiluminescence measurements were performed before the irradiation and 10 and 100 days after the irradiation. The best methods for control purposes were the microbiological (DEFT + APC) methods combined with chemiluminescence measurements. No differences were detected between the irradiated and non-irradiated samples with the chemical methods.

  5. Observations on cattle schistosomiasis in the Sudan, a study in comparative medicine. III. Field testing of an irradiated Schistosoma bovis vaccine

    SciTech Connect

    Majid, A.A.; Bushera, H.O.; Saad, A.M.; Hussein, M.F.; Taylor, M.G.; Dargie, J.D.; Marshall, T.F.; Nelson, G.S.

    1980-05-29

    Previous work has shown that cattle can acquire a strong resistance to Schistosoma bovis infection following repeated natural exposure. Partial resistance to a laboratory challenge with S. bovis has also been demonstrated in calves after immunization with an irradiated schistosomular or cercarial vaccine. The aim of the present study was to see whether this type of caccine could protect calves under the very different conditions of natural exposure to S. bovis in the field. Thirty 6- to 9-month-old calves were each immunized with 10,000 irradiated S. bovis schistosomula by intramuscular injection and 8 weeks later were released into an enzootic area along with 30 unvaccinated animals. The calves were followed up for 10 months, during which period protection was evidenced by a lower mortality rate, a slower rate of acquisition of infection, and lower fecal egg counts in the vaccinated calves. Necropsy of the survivors showed 60 to 70% reductions in worm and tissue egg counts of the vaccinated calves as compared to those not vaccinated.

  6. Production of modified starches by gamma irradiation

    NASA Astrophysics Data System (ADS)

    Kang, Il-Jun; Byun, Myung-Woo; Yook, Hong-Sun; Bae, Chun-Ho; Lee, Hyun-Soo; Kwon, Joong-Ho; Chung, Cha-Kwon

    1999-04-01

    As a new processing method for the production of modified starch, gamma irradiation and four kinds of inorganic peroxides were applied to commercial corn starch. The addition of inorganic peroxides without gamma irradiation or gamma irradiation without the addition of inorganic peroxides effectively decreased initial viscosity, but did not sufficiently keep viscosity stable. The combination of adding ammonium persulfate (APS) and gamma irradiation showed the lowest initial viscosity and the best stability out of the tested four kinds of inorganic peroxides. Among the tested mixing methods of APS, soaking was found to be more effective than dry blending or spraying. Therefore, the production of modified starch with low viscosity as well as with sufficient viscosity stability became feasible by the control of gamma irradiation dose levels and the amount of added APS to starch.

  7. ATF Neutron Irradiation Program Technical Plan

    SciTech Connect

    Geringer, J. W.; Katoh, Yutai

    2016-03-01

    The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post-irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.

  8. Results of Charpy V-Notch Impact Testing of Structural Steel Specimens Irradiated at ~30°C to 1 x 1016 neutrons/cm2 in a Commercial Reactor Cavity

    SciTech Connect

    Iskander, S. K.; Stoller, R. E.

    1997-04-01

    A capsule containing Charpy V-notch (CVN) and mini-tensile specimens was irradiated at ~30°C (~ 85°F) in the cavity of a commercial nuclear power plant to a fluence of 1 x 1016 neutrons/cm2 (>1 MeV). The capsule included six CVN impact specimens of archival High Flux Isotope Reactor A212 grade B ferritic steel and five CVN impact specimens of a well-studied A36 structural steel. This irradiation was part of the ongoing study of neutron-induced damage effects at the low temperature and flux experienced by reactor supports. The plant operators shut down the plant before the planned exposure was reached. The exposure of these specimens produced no significant irradiation-induced embrittlement. Of interest were the data on unirradiated specimens in the L-T orientation machined from a single plate of A36 structural steel, which is the same specification for the structural steel used in some reactor supports. The average CVN energy of five unirradiated specimens obtained from one region of the plate and tested at room temperature was ~ 99 J, while the energy of 11 unirradiated specimens from other locations of the same plate was 45 J, a difference of ~ 220%. The CVN impact energies for all 18 specimens ranged from a low of 32 J to a high of 111 J. Moreover, it appears that the University of Kansas CVN impact energy data of the unirradiated specimens at the 100-J level are shifted toward higher temperatures by about 20 K. The results were an example of the extent of scatter possible in CVN impact testing. Generic values for the CVN impact energy of A36 should be used with caution in critical applications.

  9. Food irradiation regulatory development in the U.S.

    NASA Astrophysics Data System (ADS)

    Miller, Sandford A.; Coleman, Eugene C.

    The Food and Drug Administration's involvement in food irradiation dates back more than thirty years. The agency has been involved with the wholesomeness testing of the irradiated foods from both nutritional and toxicological standpoints. Knowledge about the nutritional and toxicological aspects of irradiated foods is fundamental in the development of a regulatory strategy for assuring the safe use of such foods.

  10. Fracture surfaces of irradiated composites

    NASA Technical Reports Server (NTRS)

    Milkovich, Scott M.; Sykes, George F., Jr.; Herakovich, Carl T.

    1987-01-01

    Electron microscopy was used to analyze the fracture surfaces of T300/934 graphite/epoxy unidirectional off-axis tensile coupons which were subjected to 1.0-MeV electron radiation at a rate of 50 Mrad/h for a total dose of 10 Grad. Fracture surfaces from irradiated and nonirradiated specimens tested at 116 K, room temperature, and 394 K were analyzed to assess the influence of radiation and temperature on the mode of failure and variations in constituent material as a function of environmental exposure. Micrographs of fracture surfaces indicate that irradiated specimens are more brittle than nonirradiated specimens at low temperatures. However, at elevated temperatures the irradiated specimens exhibit significantly more plasticity than nonirradiated specimens.

  11. Commercial food irradiation

    SciTech Connect

    Black, E.F.; Libby, L.M.

    1983-06-01

    Food irradiation is discussed. Irradiation exposes food to gamma rays from a cobalt-60 or a cesium-137 source, or to high-energy electrons emitted by an electron accelerator. A major advantage is that food can be packaged either before or after treatment. FDA regulations with regard to irradiation are discussed. Comments on an 'Advance Notice' on irradiation, published by the FDA in 1981 are summarized.

  12. Testing zinc chloride as a new catalyst for direct synthesis of cellulose di- and tri-acetate in a solvent free system under microwave irradiation.

    PubMed

    El Nemr, Ahmed; Ragab, Safaa; El Sikaily, Amany

    2016-10-20

    This research demonstrates the effect of ZnCl2 as a catalyst on the esterification of commercial cotton cellulose using acetic anhydride in order to obtain di- and tri-cellulose acetates under microwave irradiation. It was discovered that microwave irradiation significantly increased the yield and reduced the reaction time. It was found that the maximum yield for cellulose triacetates was 95.83% under the reaction conditions that were as follows: 3min reaction time, 200mg of ZnCl2 catalyst and 20ml of Ac2O for 5g cellulose. However, the cellulose acetate obtained in this manner had the highest DS (2.87). The cellulose di-acetate was produced with the maximum yield of 89.97% and with the highest DS (2.69) using 25ml Ac2O, 200mg of ZnCl2 for 5g cellulose and in 3min reaction time. The effect of some factors such as the amount of used catalyst, the quantity of acetic acid anhydride and the reaction time of the esterification process have been investigated. The production of di- and tri-cellulose acetate and the degree of substitution were confirmed using Fourier-transform infrared (FTIR) and nuclear magnetic resonance (NMR). The thermal stability was investigated using thermo gravimetric analysis (TGA), differential thermal analysis (DTA) and differential scanning calorimetry (DSC). The molecular weight and the degree of polymerization were obtained using Gel Permeation Chromatography (GPC). The analysis confirmed the successful synthesis of di- and tri-cellulose acetate without degradation during the reaction; these results were found to be in contrast to some recent studies. The present study reveals that ZnCl2 is a new catalyst; it is effective as well as inexpensive and is a low toxicity catalyst for usage in cellulose esterification.

  13. Food irradiation and sterilization

    NASA Astrophysics Data System (ADS)

    Josephson, Edward S.

    Radiation sterilization of food (radappertization) requires exposing food in sealed containers to ionizing radiation at absorbed doses high enough (25-70 kGy) to kill all organisms of food spoilage and public health significance. Radappertization is analogous to thermal canning is achieving shelf stability (long term storage without refrigeration). Except for dry products in which autolysis is negligible, the radappertization process also requires that the food be heated to an internal temperature of 70-80°C (bacon to 53°C) to inactivate autolytic enzymes which catalyze spoilage during storage without refrigeration. To minimize the occurence of irradiation induced off-flavors and odors, undesirable color changes, and textural and nutritional losses from exposure to the high doses required for radappertization, the foods are vacuum sealed and irradiated frozen (-40°C to -20°C). Radappertozed foods have the characteristic of fresh foods prepared for eating. Radappertization can substitute in whole or in part for some chemical food additives such as ethylene oxide and nitrites which are either toxic, carcinogenic, mutagenic, or teratogenic. After 27 years of testing for "wholesomeness" (safety for consumption) of radappertized foods, no confirmed evidence has been obtained of any adverse effecys of radappertization on the "wholesomeness" characteristics of these foods.

  14. Decline in Tested and Self-Reported Cognitive Functioning After Prophylactic Cranial Irradiation for Lung Cancer: Pooled Secondary Analysis of Radiation Therapy Oncology Group Randomized Trials 0212 and 0214

    SciTech Connect

    Gondi, Vinai; Paulus, Rebecca; Bruner, Deborah W.; Meyers, Christina A.; Gore, Elizabeth M.; Wolfson, Aaron; Werner-Wasik, Maria; Sun, Alexander Y.; Choy, Hak; Movsas, Benjamin

    2013-07-15

    Purpose: To assess the impact of prophylactic cranial irradiation (PCI) on self-reported cognitive functioning (SRCF), a functional scale on the European Organization for Research and Treatment of Cancer Core Quality of Life Questionnaire (EORTC QLQ-C30). Methods and Materials: Radiation Therapy Oncology Group (RTOG) protocol 0214 randomized patients with locally advanced non-small cell lung cancer to PCI or observation; RTOG 0212 randomized patients with limited-disease small cell lung cancer to high- or standard-dose PCI. In both trials, Hopkins Verbal Learning Test (HVLT)-Recall and -Delayed Recall and SRCF were assessed at baseline (after locoregional therapy but before PCI or observation) and at 6 and 12 months. Patients developing brain relapse before follow-up evaluation were excluded. Decline was defined using the reliable change index method and correlated with receipt of PCI versus observation using logistic regression modeling. Fisher's exact test correlated decline in SRCF with HVLT decline. Results: Of the eligible patients pooled from RTOG 0212 and RTOG 0214, 410 (93%) receiving PCI and 173 (96%) undergoing observation completed baseline HVLT or EORTC QLQ-C30 testing and were included in this analysis. Prophylactic cranial irradiation was associated with a higher risk of decline in SRCF at 6 months (odds ratio 3.60, 95% confidence interval 2.34-6.37, P<.0001) and 12 months (odds ratio 3.44, 95% confidence interval 1.84-6.44, P<.0001). Decline on HVLT-Recall at 6 and 12 months was also associated with PCI (P=.002 and P=.002, respectively) but was not closely correlated with decline in SRCF at the same time points (P=.05 and P=.86, respectively). Conclusions: In lung cancer patients who do not develop brain relapse, PCI is associated with decline in HVLT-tested and self-reported cognitive functioning. Decline in HVLT and decline in SRCF are not closely correlated, suggesting that they may represent distinct elements of the cognitive spectrum.

  15. The effects of irradiation and EtO-treatment on ultrahigh molecular weight polyethylene acetabular cups following accelerated aging: Degradation of mechanical properties and morphology changes during hip simulator tests

    NASA Astrophysics Data System (ADS)

    Taddei, Paola; Affatato, Saverio; Rocchi, Mirko; Fagnano, Concezio; Viceconti, Marco

    2008-03-01

    The present study was aimed at investigating the effects of the sterilization method and accelerated aging on the wear and morphology of ultrahigh molecular weight polyethylene. In a first test, gamma- and EtO-sterilized acetabular cups were tested in a hip joint simulator for two million cycles. After the test, the cups underwent an accelerated aging treatment (80 °C, 4 weeks in air furnace) and were newly tested for another three million cycles. Wear was evaluated by gravimetric measurements, morphology by micro-Raman spectroscopy. During the first test, the EtO-sterilized cups underwent a significantly higher wear than the gamma-sterilized ones (62 and 30 mg/million cycles, respectively). No significant crystallinity changes were observed. Upon accelerated aging, the crystallinity increase of the gamma-sterilized cups was more pronounced than for the EtO-sterilized cups, due to chain scission and oxygen incorporation. In the second test, the wear rate of EtO-sterilized cups decreased to 38 mg/million cycles, while for gamma-irradiated cups it increased to 84 mg/million cycles. At the same time, the latter cups underwent significant increases in temperature and crystallinity, due to the higher friction. For the EtO-sterilized cups a significant decrease in crystallinity was observed, due to the occurrence of an orthorhombic → monoclinic phase transformation.

  16. Associations of Pd, U and Ag in the SiC layer of neutron-irradiated TRISO fuel

    NASA Astrophysics Data System (ADS)

    Lillo, T. M.; van Rooyen, I. J.

    2015-05-01

    Knowledge of the associations and composition of fission products in the neutron irradiated SiC layer of high-temperature gas reactor TRISO fuel is important to the understanding of various aspects of fuel performance that presently are not well understood. Recently, advanced characterization techniques have been used to examine fuel particles from the Idaho National Laboratory's AGR-1 experiment. Nano-sized Ag and Pd precipitates were previously identified in grain boundaries and triple points in the SiC layer of irradiated TRISO nuclear fuel. Continuation of this initial research is reported in this paper and consists of the characterization of a relatively large number of nano-sized precipitates in three areas of the SiC layer of a single irradiated TRISO nuclear fuel particle using standardless EDS analysis on focused ion beam-prepared transmission electron microscopy samples. Composition and distribution analyses of these precipitates, which were located on grain boundaries, triple junctions and intragranular precipitates, revealed low levels, generally <10 atomic %, of palladium, silver and/or uranium with palladium being the most common element found. Palladium by itself, or associated with either silver or uranium, was found throughout the SiC layer. A small number of precipitates on grain boundaries and triple junctions were found to contain only silver or silver in association with palladium while uranium was always associated with palladium but never found by itself or in association with silver. Intergranular precipitates containing uranium were found to have migrated ∼23 μm along a radial direction through the 35 μm thick SiC coating during the AGR-1 experiment while silver-containing intergranular precipitates were found at depths up to ∼24 μm in the SiC layer. Also, Pd-rich, nano-precipitates (∼10 nm in diameter), without evidence for the presence of either Ag or U, were revealed in intragranular regions throughout the SiC layer. Because not

  17. Associations of Pd, U and Ag in the SiC layer of neutron-irradiated TRISO fuel

    SciTech Connect

    Lillo, Thomas; Rooyen, Isabella Van

    2015-05-01

    Knowledge of the associations and composition of fission products in the neutron irradiated SiC layer of high-temperature gas reactor TRISO fuel is important to the understanding of various aspects of fuel performance that presently are not well understood. Recently, advanced characterization techniques have been used to examine fuel particles from the Idaho National Laboratory’s AGR-1 experiment. Nano-sized Ag and Pd precipitates were previously identified in grain boundaries and triple points in the SiC layer of irradiated TRISO nuclear fuel. Continuation of this initial research is reported in this paper and consists of the characterization of a relatively large number of nano-sized precipitates in three areas of the SiC layer of a single irradiated TRISO nuclear fuel particle using standardless EDS analysis on focused ion beam-prepared transmission electron microscopy samples. Composition and distribution analyses of these precipitates, which were located on grain boundaries, triple junctions and intragranular precipitates, revealed low levels, generally <10 atomic %, of palladium, silver and/or uranium with palladium being the most common element found. Palladium by itself, or associated with either silver or uranium, was found throughout the SiC layer. A small number of precipitates on grain boundaries and triple junctions were found to contain only silver or silver in association with palladium while uranium was always associated with palladium but never found by itself or in association with silver. Intergranular precipitates containing uranium were found to have migrated ~23 μm along a radial direction through the 35 μm thick SiC coating during the AGR-1 experiment while silver-containing intergranular precipitates were found at depths up to ~24 μm in the SiC layer. Also, Pd-rich, nano-precipitates (~10 nm in diameter), without evidence for the presence of either Ag or U, were revealed in intragranular regions throughout the SiC layer. Because not all

  18. Welding irradiated stainless steel

    SciTech Connect

    Kanne, W.R. Jr.; Chandler, G.T.; Nelson, D.Z.; Franco-Ferreira, E.A.

    1993-12-31

    Conventional welding processes produced severe underbead cracking in irradiated stainless steel containing 1 to 33 appm helium from n,a reactions. A shallow penetration overlay technique was successfully demonstrated for welding irradiated stainless steel. The technique was applied to irradiated 304 stainless steel that contained 10 appm helium. Surface cracking, present in conventional welds made on the same steel at the same and lower helium concentrations, was eliminated. Underbead cracking was minimal compared to conventional welding methods. However, cracking in the irradiated material was greater than in tritium charged and aged material at the same helium concentrations. The overlay technique provides a potential method for repair or modification of irradiated reactor materials.

  19. Irradiation response on mechanical properties of neutron irradiated F82H

    NASA Astrophysics Data System (ADS)

    Shiba, K.; Suzuki, M.; Hishinuma, A.

    1996-10-01

    Tensile and Charpy impact properties of neutron irradiated F82H (Fe8Cr2WVTa) with and without boron have been investigated to obtain the basic irradiation response on mechanical properties in low damage regime less than 1 dpa at the temperature ranging from 300° to 590°C. Boron-doped steel was used for the helium effect due to (n, α) reaction. Typical irradiation hardening was observed at 300°C. The irradiation above 520°C did not reveal increase in yield stress, but the specimen irradiated at 590°C showed some reduction in elongation in room temperature tensile testing. Slight difference in the tensile properties between boron-doped and boron-free were observed at 590°C. No changes in ductile brittle transition temperature (DBTT) occurred at a temperature between 335° and 460°C by Charpy impact testing.

  20. AGC-1 Post Irradiation Examination Status

    SciTech Connect

    David Swank

    2011-09-01

    The Next Generation Nuclear Plant (NGNP) Graphite R&D program is currently measuring irradiated material property changes in several grades of nuclear graphite for predicting their behavior and operating performance within the core of new Very High Temperature Reactor (VHTR) designs. The Advanced Graphite Creep (AGC) experiment consisting of six irradiation capsules will generate this irradiated graphite performance data for NGNP reactor operating conditions. All six AGC capsules in the experiment will be irradiated in the Advanced Test Reactor (ATR), disassembled in the Hot Fuel Examination Facility (HFEF), and examined at the INL Research Center (IRC) or Oak Ridge National Laboratory (ORNL). This is the first in a series of status reports on the progress of the AGC experiment. As the first capsule, AGC1 was irradiated from September 2009 to January 2011 to a maximum dose level of 6-7 dpa. The capsule was removed from ATR and transferred to the HFEF in April 2011 where the capsule was disassembled and test specimens extracted from the capsules. The first irradiated samples from AGC1 were shipped to the IRC in July 2011and initial post irradiation examination (PIE) activities were begun on the first 37 samples received. PIE activities continue for the remainder of the AGC1 specimen as they are received at the IRC.

  1. Rapid detection of irradiated frozen hamburgers

    NASA Astrophysics Data System (ADS)

    Delincée, Henry

    2002-03-01

    DNA comet assay can be employed as a rapid and inexpensive screening test to check whether frozen ground beef patties (hamburgers) have been irradiated as a means to increase their safety by eliminating pathogenic bacteria, e.g. E. coli O157:H7. Such a detection procedure will provide an additional check on compliance with existing regulations, e.g. enforcement of labelling and rules in international trade. Frozen ready prepared hamburgers from the market place were `electron irradiated' with doses of 0, 1.3, 2.7, 4.5 and 7.2kGy covering the range of potential commercial irradiation. DNA fragmentation in the hamburgers was made visible within a few hours using the comet assay, and non-irradiated hamburgers could be easily discerned from the irradiated ones. Even after 9 months of frozen storage, irradiated hamburgers could be identified. Since DNA fragmentation may also occur with other food processes (e.g. temperature abuse), positive screening tests shall be confirmed using a validated method to specifically prove an irradiation treatment, e.g. EN 1784 or EN 1785.

  2. Hair dosimetry following neutron irradiation.

    PubMed

    Lebaron-Jacobs, L; Gaillard-Lecanu, E; Briot, F; Distinguin, S; Boisson, P; Exmelin, L; Racine, Y; Berard, P; Flüry-Herard, A; Miele, A; Fottorino, R

    2007-05-01

    Use of hair as a biological dosimeter of neutron exposure was proposed a few years ago. To date, the (32)S(n,p)(32)P reaction in hair with a threshold of 2.5 MeV is the best choice to determine the fast neutron dose using body activation. This information is essential with regards to the heterogeneity of the neutron transfer to the organism. This is a very important parameter for individual dose reconstruction from the surface to the deeper tissues. This evaluation is essential to the adapted management of irradiated victims by specialized medical staff. Comparison exercises between clinical biochemistry laboratories from French sites (the CEA and COGEMA) and from the IRSN were carried out to validate the measurement of (32)P activity in hair and to improve the techniques used to perform this examination. Hair was placed on a phantom and was irradiated at different doses in the SILENE reactor (Valduc, France). Different parameters were tested: variation of hair type, minimum weight of hair sample, hair wash before measurement, delivery period of results, and different irradiation configurations. The results obtained in these comparison exercises by the different laboratories showed an excellent correlation. This allowed the assessment of a dose-activity relationship and confirmed the feasibility and the interest of (32)P measurement in hair following fast neutron irradiation.

  3. Mechanical properties of UV irradiated rat tail tendon (RTT) collagen.

    PubMed

    Sionkowska, Alina; Wess, Tim

    2004-04-01

    The mechanical properties of RTT collagen tendon before and after UV irradiation have been investigated by mechanical testing (Instron). Air-dried tendon were submitted to treatment with UV irradiation (wavelength 254 nm) for different time intervals. The changes in such mechanical properties as breaking strength and percentage elongation have been investigated. The results have shown, that the mechanical properties of the tendon were greatly affected by time of UV irradiation. Ultimate tensile strength and ultimate percentage elongation decreased after UV irradiation of the tendon. Increasing UV irradiation leads to a decrease in Young's modulus of the tendon.

  4. Irradiation performance of full-length metallic IFR fuels

    SciTech Connect

    Tsai, H.; Neimark, L.A.

    1992-07-01

    An assembly irradiation of 169 full-length U-Pu-Zr metallic fuel pins was successfully completed in FFTF to a goal burnup of 10 at.%. All test fuel pins maintained their cladding integrity during the irradiation. Postirradiation examination showed minimal fuel/cladding mechanical interaction and excellent stability of the fuel column. Fission-gas release was normal and consistent with the existing data base from irradiation testing of shorter metallic fuel pins in EBR-II.

  5. Van de Graaff Irradiation of Materials

    SciTech Connect

    Quigley, Kevin; Chemerisov, Sergey; Tkac, Peter; Vandegrift, George F.

    2016-10-01

    Through irradiations using our 3 MeV Van de Graaf accelerator, Argonne is testing the radiation stability of components of equipment that are being used to dispense molybdenum solutions for use as feeds to 99mTc generators and in the 99mTc generators themselves. Components have been irradiated by both a direct electron beam and photons generated from a tungsten convertor.

  6. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K. Hartwell; John b. Walter

    2010-10-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  7. Fission Product Monitoring of TRISO Coated Fuel For The Advanced Gas Reactor -1 Experiment

    SciTech Connect

    Dawn M. Scates; John K Hartwell; John B. Walter

    2008-09-01

    The US Department of Energy has embarked on a series of tests of TRISO-coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burn up of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B’s) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  8. Influences of Microwave Irradiation on Environment

    NASA Astrophysics Data System (ADS)

    Murakami, H.; Abe, Y.; Iwata, T.; Kudo, I.; Saito, K.; Okuda, T.

    2004-12-01

    An experimental facility to evaluate the long-duration influence of microwave to environment, a so-called long duration microwave exposure facility (LDMEF), was constructed in Tsukuba in 1994, and so far irradiation tests on plants accumulated over 40,000 hours have been conducted with the aid of 2.45 GHz magnetron. The LDMEF consists of a pair of outdoor electromagnetically isolated areas, one under the influence of microwave irradiation with a 500 W magnetron and one without microwave irradiation. The growth rates of plants in both areas were compared and evaluated with the experimental data for the temperature distribution in the soil and power distribution of microwave. Although any appreciable influence of microwave was not noticed in the power density less than 10 mW/cm2 , the experimental results showed a significant growth rate enhancement when the power density became over 10 mW/cm2 . However, the growth was rather depressed when the power density increased over 15 mW/cm2 . These effects are well explained by the temperature and moisture in the soil which are also under an appreciable influence of microwave irradiation [1,2]. In this context, we newly constructed an indoor irradiation facility, in which the growth conditions of plants under a constant soil temperature can be maintained. In addition, irradiation with a 5.8 GHz magnetron will be conducted in the new facility. In parallel to a series of indoor and outdoor irradiation tests on plants, the influence of microwave irradiation on the growth pattern of albino mouse will be conducted. This experiment will be the first experimental evaluation for the influence of microwave irradiation on animals.

  9. Effect of low-level laser irradiating point on immunity

    NASA Astrophysics Data System (ADS)

    Cai, ChangSong; Qi, Qiong-fang; Xin, Jiang

    1993-03-01

    This paper reports that cellular immune function was observed when He-Ne laser was used to irradiate `zusanli' point in rats using various power, time, and periods. The indicator was a lymphocyte transformation test (LTT) by MTT colorimetric analysis. The best irradiating condition was determined, the effect and both virtues and defects of the laser were compared with those of electropuncture. The results show (1) LTT was enhanced in the group of laser irradiating point, but LTT was not enhanced in non-point (t' test, P < 0.01). (2) Lower power -- 2 mW or 5 mW of irradiating for 15 - 20 min, was better; 10 mW or 20 mW of irradiating for 10 - 15 min was suitable. Prolonged irradiating time did not enhance the immune function of the rats. On the contrary, immune function was inhibited. (3) A 7-day period of irradiating was best (once a day, 10 mW for 10 min). Enhanced LTT was not seen when irradiation days were added (SNK, P > 0.05). (4) Laser irradiation point and electropuncture were compared with vehicle control, LTT in the former two groups was enhanced significantly (ANOVA, P < 0.01), and laser irradiating point and electropuncture had the same effect (SNK, P > 0.05). The data suggest that laser irradiating point was able to enhance cell immunity and the enhancement of LTT had a point specific characteristic. The best condition of laser irradiating point was 2 mW for 15 - 20 min, and 10 mW or 20 mW for 10 - 15 min. The best period was 7-day irradiation. The results show laser irradiating the point may activate the main and collateral channels system, then modify the immune function of the body. Our observations provide experimental evidence for proper clinical application of laser irradiating points. The paper theoretically discusses and analyzes the experiment results in detail.

  10. Irradiation performance of FFTF drivers using the D9 alloy

    SciTech Connect

    Pitner, A.L.; Gneiting, B.C.; Bard, F.E.

    1994-12-31

    Five test assemblies similar in design to the Fast Flux Test Facility driver fuel assembly , but employing the alloy D9 in place of stainless steel 316 for duct, cladding, and wire wrap compnents were irradiated to demonstrate the improved performance of the new design. Results of post-irradiation examinations are discussed.

  11. Comminuting irradiated ferritic steel

    DOEpatents

    Bauer, Roger E.; Straalsund, Jerry L.; Chin, Bryan A.

    1985-01-01

    Disclosed is a method of comminuting irradiated ferritic steel by placing the steel in a solution of a compound selected from the group consisting of sulfamic acid, bisulfate, and mixtures thereof. The ferritic steel is used as cladding on nuclear fuel rods or other irradiated components.

  12. Perspective on food irradiation

    SciTech Connect

    Not Available

    1987-02-01

    Recent US Food and Drug Administration approval of irradiation treatment for fruit, vegetables and pork has stimulated considerable discussion in the popular press on the safety and efficacy of irradiation processing of food. This perspective is designed to summarize the current scientific information available on this issue.

  13. MASSIVE LEAKAGE IRRADIATOR

    DOEpatents

    Wigner, E.P.; Szilard, L.; Christy, R.F.; Friedman, F.L.

    1961-05-30

    An irradiator designed to utilize the neutrons that leak out of a reactor around its periphery is described. It avoids wasting neutron energy and reduces interference with the core flux to a minimum. This is done by surrounding all or most of the core with removable segments of the material to be irradiated within a matrix of reflecting material.

  14. Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor

    SciTech Connect

    Field, Kevin G.; Howard, Richard H.; Yamamoto, Yukinori

    2015-06-26

    The objective of the FeCrAl embrittlement screening tests being conducted through the use of Oak Ridge National Laboratories (ORNL) High Flux Isotope Reactor is to provide data on the radiation-induced changes in the mechanical properties including radiation-induced hardening and embrittlement through systematic testing and analysis. Data developed on the mechanical properties will be supported by extensive microstructural evaluations to assist in the development of structure-property relationships and provide a sound, fundamental understanding of the performance of FeCrAl alloys in intense neutron radiation fields. Data and analysis developed as part of this effort will be used to assist in the determination of FeCrAl alloys as a viable material for commercial light water reactor (LWR) applications with a primary focus as an accident tolerant cladding.

  15. Food irradiation development in Pakistan

    NASA Astrophysics Data System (ADS)

    Khan, I.

    The large scale trials were held to extend the storage life of potatoes, onions and dry fruits by gamma radiation. It was concluded that radiation preservation of potatoes and onions was much cheaper as compared to conventional methods. A dose of 1 kGy can control the insects in dry fruits and nuts. The consumers' acceptability and market testing performed during the last four years are also conducive to the commercialization of the technology in this country. The Government of Pakistan has accorded clearance for the irradiation of some food items like potatoes, onions, garlic and spices for human consumption. The Pakistan Radiation Services (PARAS), the commercial irradiator (200 Kci) at Lahore, has already started functioning in April, 1987. It is planned to start large scale sterilization of spices by gamma radiation in PARAS shortly.

  16. Identification of gamma-irradiated papaya, melon and watermelon

    NASA Astrophysics Data System (ADS)

    Marín-Huachaca, Nélida S.; Mancini-Filho, Jorge; Delincée, Henry; Villavicencio, Anna Lúcia C. H.

    2004-09-01

    Ionizing radiation can be used to control spoilage microorganisms and to increase the shelf life of fresh fruits and vegetables in replacement for the treatment with chemical fumigants. In order to enforce labelling regulations, methods for detecting the irradiation treatment directly in the produce are required. Recently, a number of detection methods for irradiated food have been adopted by the Codex Comission. A rapid screening method for qualitative detection of irradiation is the DNA Comet Assay. The applicability of the DNA Comet Assay for distinguishing irradiated papaya, melon, and watermelon was evaluated. The samples were treated in a 60Co facility at dose levels of 0.0, 0.5, 0.75, and 1.0kGy. The irradiated samples showed typical DNA fragmentation whereas cells from non-irradiated ones appeared intact. In addition to the DNA Comet Assay also the half-embryo test was applied in melon and watermelon to detect the irradiation treatment.

  17. AGC-2 Specimen Post Irradiation Data Package Report

    SciTech Connect

    Windes, William Enoch; Swank, W. David; Rohrbaugh, David T.; Cottle, David L.

    2015-08-01

    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

  18. Irradiated Shellfish: Identification by Photostimulated Luminescence

    PubMed Central

    Di Schiavi, Maria Teresa; Falconi, Grazia; Verità, Francesca Della; Cavallina, Roberta

    2014-01-01

    The irradiation of food is a technology used in the industry to prevent the deterioration of foodstuff in some countries. The European Community legislation states that each Member State must carry out annual checks on the products during commercialisation. The Istituto Zooprofilattico Sperimentale delle regioni Lazio e Toscana (Rome, Italy) has developed and validated the screening method of photostimulated luminescence UNI EN 13751:2009 to identify irradiated shellfish. A total of 30 tests of shellfish samples, consisting of 22 certified as irradiated and 8 not-irradiated samples, were performed. The validation procedure was based on sensitivity and specificity; the compatibility between the screening method and the reference standard EN 13751:2009 was evaluated. Data were processed: 100% sensitivity and 100% specificity were obtained. Results obtained in our laboratory were perfectly compatible with the reference standard. For this reason, the method has been validated and proved to be suitable for its intended use. PMID:27800335

  19. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    SciTech Connect

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000ºC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  20. Neutron Irradiation Resistance of RAFM Steels

    SciTech Connect

    Gaganidze, Ermile; Dafferner, Bernhard; Aktaa, Jarir

    2008-07-01

    The neutron irradiation resistance of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 and international reference steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) have been investigated after irradiation in the Petten High Flux Reactor up to 16.3 dpa at different irradiation temperatures (250-450 deg. C). The embrittlement behavior and hardening are investigated by instrumented Charpy-V tests with sub-size specimens. Neutron irradiation-induced embrittlement and hardening of EUROFER97 was studied under different heat treatment conditions. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement vs. hardening behavior of RAFM steels within a proper model in terms of the parameter C={delta}DBTT/{delta}{sigma} indicates hardening-dominated embrittlement at irradiation temperatures below 350 deg. C with 0.17 {<=} C {<=} 0.53 deg. C/MPa. Scattering of C at irradiation temperatures above 400 deg. C indicates non hardening embrittlement. A role of He in a process of embrittlement is investigated in EUROFER97 based steels, that are doped with different contents of natural B and the separated {sup 10}B-isotope (0.008-0.112 wt.%). Testing on small scale fracture mechanical specimens for determination of quasi-static fracture toughness will be also presented in a view of future irradiation campaigns. (authors)

  1. Irradiation creep of vanadium-base alloys

    SciTech Connect

    Tsai, H.; Billone, M.C.; Strain, R.V.; Smith, D.L.; Matsui, H.

    1998-03-01

    A study of irradiation creep in vanadium-base alloys is underway with experiments in the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) in the United States. Test specimens are thin-wall sealed tubes with internal pressure loading. The results from the initial ATR irradiation at low temperature (200--300 C) to a neutron damage level of 4.7 dpa show creep rates ranging from {approx}0 to 1.2 {times} 10{sup {minus}5}/dpa/MPa for a 500-kg heat of V-4Cr-4Ti alloy. These rates were generally lower than reported from a previous experiment in BR-10. Because both the attained neutron damage levels and the creep strains were low in the present study, however, these creep rates should be regarded as only preliminary. Substantially more testing is required before a data base on irradiation creep of vanadium alloys can be developed and used with confidence.

  2. Microarchitecture of irradiated bone: comparison with healthy bone

    NASA Astrophysics Data System (ADS)

    Bléry, Pauline; Amouriq, Yves; Guédon, Jeanpierre; Pilet, Paul; Normand, Nicolas; Durand, Nicolas; Espitalier, Florent; Arlicot, Aurore; Malard, Olivier; Weiss, Pierre

    2012-03-01

    The squamous cell carcinomas of the upper aero-digestive tract represent about ten percent of cancers. External radiation therapy leads to esthetic and functional consequences, and to a decrease of the bone mechanical abilities. For these patients, the oral prosthetic rehabilitation, including possibilities of dental implant placement, is difficult. The effects of radiotherapy on bone microarchitecture parameters are not well known. Thus, the purpose of this study is to assess the effects of external radiation on bone micro architecture in an experimental model of 25 rats using micro CT. 15 rats were irradiated on the hind limbs by a single dose of 20 Grays, and 10 rats were non irradiated. Images of irradiated and healthy bone were compared. Bone microarchitecture parameters (including trabecular thickness, trabecular number, trabecular separation, connectivity density and tissue and bone volume) between irradiated and non-irradiated bones were calculated and compared using a Mann and Whitney test. After 7 and 12 weeks, images of irradiated and healthy bone are different. Differences on the irradiated and the healthy bone populations exhibit a statistical significance. Trabecular number, connectivity density and closed porosity are less important on irradiated bone. Trabecular thickness and separation increase for irradiated bone. These parameters indicate a decrease of irradiated bone properties. Finally, the external irradiation induces changes on the bone micro architecture. This knowledge is of prime importance for better oral prosthetic rehabilitation, including implant placement.

  3. Alaskan Commodities Irradiation Project

    SciTech Connect

    Zarling, J.P.; Swanson, R.B.; Logan, R.R.; Das, D.K.; Lewis, C.E.; Workman, W.G.; Tumeo, M.A.; Hok, C.I.; Birklid, C.A.; Bennett, F.L.

    1988-12-01

    The ninety-ninth US Congress commissioned a six-state food irradiation research and development program to evaluate the commercial potential of this technology. Hawaii, Washington, Iowa, Oklahoma and Florida as well as Alaska have participated in the national program; various food products including fishery products, red meats, tropical and citrus fruits and vegetables have been studied. The purpose of the Alaskan study was to review and evaluate those factors related to the technical and economic feasibility of an irradiator in Alaska. This options analysis study will serve as a basis for determining the state's further involvement in the development of food irradiation technology. 40 refs., 50 figs., 53 tabs.

  4. RERTR-7 Irradiation Summary Report

    SciTech Connect

    D. M. Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-12-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-7A, was designed to test several modified fuel designs to target fission densities representative of a peak low enriched uranium (LEU) burnup in excess of 90% U-235 at peak experiment power sufficient to generate a peak surface heat flux of approximately 300 W/cm2. The RERTR-7B experiment was designed as a high power test of 'second generation' dispersion fuels at peak experiment power sufficient to generate a surface heat flux on the order of 230 W/cm2.1 The following report summarizes the life of the RERTR-7A and RERTR-7B experiments through end of irradiation, including as-run neutronic analyses, thermal analyses and hydraulic testing results.

  5. RERTR-10 Irradiation Summary Report

    SciTech Connect

    D. M. Perez

    2011-05-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-10 was designed to further test the effectiveness of modified fuel/clad interfaces in monolithic fuel plates. The experiment was conducted in two campaigns: RERTR-10A and RERTR-10B. The fuel plates tested in RERTR-10A were all fabricated by Hot Isostatic Pressing (HIP) and were designed to evaluate the effect of various Si levels in the interlayer and the thickness of the Zr interlayer (0.001”) using 0.010” and 0.020” nominal foil thicknesses. The fuel plates in RERTR-10B were fabricated by Friction Bonding (FB) with two different thickness Si layers and Nb and Zr diffusion barriers.1 The following report summarizes the life of the RERTR-10A/B experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  6. Thermally activated deformation of irradiated reactor pressure vessel steel

    NASA Astrophysics Data System (ADS)

    Böhmert, J.; Müller, G.

    2002-03-01

    Temperature and strain rate change tensile tests were performed on two VVER 1000-type reactor pressure vessel welds with different contents of nickel in unirradiated and irradiated conditions in order to determine the activation parameters of the contribution of the thermally activated deformation. There are no differences of the activation parameters in the unirradiated and the irradiated conditions as well as for the two different materials. This shows that irradiation hardening preferentially results from a friction hardening mechanism by long-range obstacles.

  7. Licensing a new industrial irradiator.

    PubMed

    Bates, Nicolas K; Entwistle, Frederick B

    2010-02-01

    After nearly three decades of medical product sterilization, 3M launched a major new project to build and license an irradiator facility. 3M Corporate Health Physics was responsible for the licensing aspect of this project. The licensing process consisted of six amendments, over 30 submissions to the U.S. Nuclear Regulatory Commission (U.S. NRC) and four U.S. NRC site visits. It took approximately 22 months to complete. The six license amendments are reviewed and several of the submissions are discussed. These include 3M's response to the U.S. NRC's interest in the shielding calculations used for the bioshield, the development of a protocol of radiation safety system test methods, and an analysis to show that a dropped cask during loading operations would not fall on sealed sources. A number of lessons were learned during the course of licensing the new irradiator. Among these were the importance of understanding the U.S. NRC license reviewer's perspective, the need to thoroughly review the irradiator manufacturer's licensing package during project negotiations, the benefits of leaving the Health Physics Office and meeting with the non-health physicists involved in the project, and the necessity of maintaining the solid relationships that already existed with the site Radiation Safety Officer and Sterilization Engineer.

  8. [The irradiation process].

    PubMed

    Barillot, I; Chauvet, B; Hannoun Lévi, J M; Lisbona, A; Leroy, T; Mahé, M A

    2016-09-01

    The purpose of this article is to describe the regulatory framework of the radiotherapy practice in France, the external irradiation and brachytherapy process and the guidelines for patient follow-up.

  9. Total lymphoid irradiation and discordant cardiac xenografts

    SciTech Connect

    Kaplan, E.; Dresdale, A.R.; Diehl, J.T.; Katzen, N.A.; Aronovitz, M.J.; Konstam, M.A.; Payne, D.D.; Cleveland, R.J. )

    1990-01-01

    Total lymphoid irradiation can prolong concordant cardiac xenografts. The effects of total lymphoid irradiation in a discordant xenograft model (guinea pig to rat) were studied with and without adjuvant pharmacologic immunosuppression. Inbred Lewis rats were randomly allocated to one of four groups. Group 1 (n = 6) served as a control group and rats received no immunosuppression. Group 2 (n = 5) received triple-drug therapy that consisted of intraperitoneal azathioprine (2 mg/kg), cyclosporine (20 mg/kg), and methylprednisolone (1 mg/kg) for 1 week before transplantation. Group 3 animals (n = 5) received 15 Gy of total lymphoid irradiation in 12 divided doses over a 3-week period. Group 4 (n = 6) received both triple-drug therapy and total lymphoid irradiation as described for groups 2 and 3. Complement-dependent cytotoxicity assay was performed to determine if a correlation between complement-dependent cytotoxicity and rejection-free interval existed. Rejection was defined as cessation of graft pulsation and was confirmed by histologic test results. Only groups 1 and 2 showed a difference in survival (group 1, 6.9 +/- 1.0 minutes; group 2, 14.2 +/- 2.7 minutes, p = 0.02). Although total lymphoid irradiation did decrease complement-dependent cytotoxicity, linear regression revealed no correlation between complement-dependent cytotoxicity and graft survival (coefficient of correlation, 0.30). Unlike concordant cardiac xenografts, total lymphoid irradiation with or without triple-drug therapy does not prolong graft survival.

  10. Total lymphoid irradiation

    SciTech Connect

    Sutherland, D.E.; Ferguson, R.M.; Simmons, R.L.; Kim, T.H.; Slavin, S.; Najarian, J.S.

    1983-05-01

    Total lymphoid irradiation by itself can produce sufficient immunosuppression to prolong the survival of a variety of organ allografts in experimental animals. The degree of prolongation is dose-dependent and is limited by the toxicity that occurs with higher doses. Total lymphoid irradiation is more effective before transplantation than after, but when used after transplantation can be combined with pharmacologic immunosuppression to achieve a positive effect. In some animal models, total lymphoid irradiation induces an environment in which fully allogeneic bone marrow will engraft and induce permanent chimerism in the recipients who are then tolerant to organ allografts from the donor strain. If total lymphoid irradiation is ever to have clinical applicability on a large scale, it would seem that it would have to be under circumstances in which tolerance can be induced. However, in some animal models graft-versus-host disease occurs following bone marrow transplantation, and methods to obviate its occurrence probably will be needed if this approach is to be applied clinically. In recent years, patient and graft survival rates in renal allograft recipients treated with conventional immunosuppression have improved considerably, and thus the impetus to utilize total lymphoid irradiation for its immunosuppressive effect alone is less compelling. The future of total lymphoid irradiation probably lies in devising protocols in which maintenance immunosuppression can be eliminated, or nearly eliminated, altogether. Such protocols are effective in rodents. Whether they can be applied to clinical transplantation remains to be seen.

  11. Brain fibronectin expression in prenatally irradiated mice

    SciTech Connect

    Meznarich, H.K.; McCoy, L.S.; Bale, T.L.; Stiegler, G.L.; Sikov, M.R. )

    1993-01-01

    Activation of gene transcription by radiation has been recently demonstrated in vivo. However, little is known on the specificity of these alterations on gene transcription. Prenatal irradiation is a known teratogen that affects the developing mammalian central nervous system (CNS). Altered neuronal migration has been suggested as a mechanism for abnormal development of prenatally irradiated brains. Fibronectin (FN), an extracellular glycoprotein, is essential for neural crest cell migration and neural cell growth. In addition, elevated levels of FN have been found in the extracellular matrix of irradiated lung. To test whether brain FN is affected by radiation, either FN level in insoluble matrix fraction or expression of FN mRNA was examined pre- and postnatally after irradiation. Mice (CD1), at 13 d of gestation (DG), served either as controls or were irradiated with 14 DG, 17 DG, or 5,6, or 14 d postnatal. Brain and liver were collected from offspring and analyzed for either total FN protein levels or relative mRNAs for FN and tubulin. Results of prenatal irradiation on reduction of postnatal brain weight relative to whole are comparable to that reported by others. Insoluble matrix fraction (IMF) per gram of brain, liver, lung, and heart weight was not significantly different either between control and irradiated groups or between postnatal stages, suggesting that radiation did not affect the IMF. However, total amounts of FN in brain IMF at 17 DG were significantly different (p < .02) between normal (1.66 [+-] 0.80 [mu]g) and irradiated brains (0.58 [+-] 0.22 [mu]g). FN mRNA was detectable at 13, 14, and 17 DG, but was not detectable at 6 and 14 d postnatal, indicating that FN mRNA is developmentally regulated. 41 refs., 4 figs., 3 tabs.

  12. Blood irradiation: Rationale and technique

    SciTech Connect

    Lewis, M.C. )

    1990-01-01

    Upon request by the local American Red Cross, the Savannah Regional Center for Cancer Care irradiates whole blood or blood components to prevent post-transfusion graft-versus-host reaction in patients who have severely depressed immune systems. The rationale for blood irradiation, the total absorbed dose, the type of patients who require irradiated blood, and the regulations that apply to irradiated blood are presented. A method of irradiating blood using a linear accelerator is described.

  13. Evaluation of Neutron Irradiated Silicon Carbide and Silicon Carbide Composites

    SciTech Connect

    Newsome G, Snead L, Hinoki T, Katoh Y, Peters D

    2007-03-26

    The effects of fast neutron irradiation on SiC and SiC composites have been studied. The materials used were chemical vapor deposition (CVD) SiC and SiC/SiC composites reinforced with either Hi-Nicalon{trademark} Type-S, Hi-Nicalon{trademark} or Sylramic{trademark} fibers fabricated by chemical vapor infiltration. Statistically significant numbers of flexural samples were irradiated up to 4.6 x 10{sup 25} n/m{sup 2} (E>0.1 MeV) at 300, 500 and 800 C in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Dimensions and weights of the flexural bars were measured before and after the neutron irradiation. Mechanical properties were evaluated by four point flexural testing. Volume increase was seen for all bend bars following neutron irradiation. Magnitude of swelling depended on irradiation temperature and material, while it was nearly independent of irradiation fluence over the fluence range studied. Flexural strength of CVD SiC increased following irradiation depending on irradiation temperature. Over the temperature range studied, no significant degradation in mechanical properties was seen for composites fabricated with Hi-Nicalon{trademark} Type-S, while composites reinforced with Hi-Nicalon{trademark} or Sylramic fibers showed significant degradation. The effects of irradiation on the Weibull failure statistics are also presented suggesting a reduction in the Weibull modulus upon irradiation. The cause of this potential reduction is not known.

  14. Effect of Electron Beam Irradiation on Tensile Strength of Polypropylene

    NASA Astrophysics Data System (ADS)

    Yamada, Hiroshi; Ikeda, Masayuki; Shimbo, Minoru; Miyano, Yasushi

    In this paper, the effects of the intensity of electron beam and the variation with time after irradiation of electron beam on the tensile strength of the polypropylene (PP), which is widely used as medicine containers, were investigated. PP with and without colorants were used first and samples irradiated under various intensity of EB. A tensile test on the irradiated samples with elapsed time after the irradiation of the electron beam was carried out. The effects of those factors on the tensile strength were discussed. The following results were obtained (1) The tensile strength of PP decreased due to the influence of the electron beam irradiation, however the rate of the decrease in strength was small compared with the original one. Furthermore, the rate of the decrease in strength was very small owing to the variation with time after the EB irradiation. (2) The tensile rupture strength of PP increased and the rupture strain owing to the influence of the electron beam irradiation compared with the original one. In addition, these rupture strength increased and the rupture strain decreased along with time after the irradiation of the electron beam. (3) The tensile rupture strain energy of PP decreased owing to the influence of the electron beam irradiation compared with the original one. In addition, the strain energy decreases with time after the irradiation of the electron beam. Moreover, the strength characteristics of PP with colorants received greater influence of electron beam compared with the one without colorants.

  15. An in vitro cell irradiation protocol for testing photopharmaceuticals and the effect of blue, green, and red light on human cancer cell lines† †Electronic supplementary information (ESI) available. See DOI: 10.1039/c5pp00424a Click here for additional data file.

    PubMed Central

    Hopkins, S. L.; Siewert, B.; Askes, S. H. C.; Veldhuizen, P.; Zwier, R.; Heger, Michal

    2016-01-01

    Traditionally, ultraviolet light (100–400 nm) is considered an exogenous carcinogen while visible light (400–780 nm) is deemed harmless. In this work, a LED irradiation system for in vitro photocytotoxicity testing is described. The LED irradiation system was developed for testing photopharmaceutical drugs, but was used here to determine the basal level response of human cancer cell lines to visible light of different wavelengths, without any photo(chemo)therapeutic. The effects of blue (455 nm, 10.5 mW cm–2), green (520 nm, 20.9 mW cm–2), and red light (630 nm, 34.4 mW cm–2) irradiation was measured for A375 (human malignant melanoma), A431 (human epidermoid carcinoma), A549 (human lung carcinoma), MCF7 (human mammary gland adenocarcinoma), MDA-MB-231 (human mammary gland adenocarcinoma), and U-87 MG (human glioblastoma-grade IV) cell lines. In response to a blue light dose of 19 J cm–2, three cell lines exhibited a minimal (20%, MDA-MB-231) to moderate (30%, A549 and 60%, A375) reduction in cell viability, compared to dark controls. The other cell lines were not affected. Effective blue light doses that produce a therapeutic response in 50% of the cell population (ED50) compared to dark conditions were found to be 10.9 and 30.5 J cm–2 for A375 and A549 cells, respectively. No adverse effects were observed in any of the six cell lines irradiated with a 19 J cm–2 dose of 520 nm (green) or 630 nm (red) light. The results demonstrate that blue light irradiation can have an effect on the viability of certain human cancer cell types and controls should be used in photopharmaceutical testing, which uses high-energy (blue or violet) visible light activation. PMID:27098927

  16. Development of microstructure and irradiation hardening of Zircaloy during low dose neutron irradiation at nominally 358 C

    SciTech Connect

    Cockeram, Brian V; Smith, Richard W; Leonard, Keith J; Byun, Thak Sang; Snead, Lance Lewis

    2011-01-01

    Wrought Zircaloy-2 and Zircaloy-4 were neutron irradiated at nominally 358 C in the high flux isotope reactor (HFIR) at relatively low neutron fluences between 5.8 1022 and 2.9 1025 n/m2 (E > 1 MeV). The irradiation hardening and change in microstructure were characterized following irradiation using tensile testing and examinations of microstructure using Analytical Electron Microscopy (AEM). Small increments of dose (0.0058, 0.11, 0.55, 1.08, and 2.93 1025 n/m2) were used in the range where the saturation of irradiation hardening is typically observed so that the role of microstructure evolution and hai loop formation on irradiation hardening could be correlated. An incubation dose between 5.8 1023 and 1.1 1024 n/m2 was needed for loop nucleation to occur that resulted in irradiation hardening. Increases in yield strength were consistent with previous results in this temperature regime, and as expected less irradiation hardening and lower hai loop number density values than those generally reported in literature for irradiations at 260 326 C were observed. Unlike previous lower temperature data, there is evidence in this study that the irradiation hardening can decrease with dose over certain ranges of fluence. Irradiation induced voids were observed in very low numbers in the Zircaloy-2 materials at the highest fluence.

  17. Formation of hazardous by-products resulting from the irradiation of natural organic matter: comparison between UV and VUV irradiation.

    PubMed

    Buchanan, W; Roddick, F; Porter, N

    2006-05-01

    The use of ultraviolet (UV) or vacuum ultraviolet (VUV) photo-oxidation followed by biological treatment for the removal of natural organic matter (NOM) in drinking water is a potential water treatment technique under investigation. This paper reports on the trihalomethane formation potential (THMFP), the haloacetic acid formation potential (HAAFP), and formation of nitrite and peroxide following both UV and VUV irradiation of NOM prior to biological treatment. The total THMFP was found to decrease with increasing UV and VUV irradiation dose, although there was a linear increase in bromoform formation. Determination of the THMFP of NOM fractions obtained after irradiation, showed that the hydrophobic fraction was dominated by chlorinated species which accounted for the majority of the total THMFP, while bromoform was observed only in the hydrophilic fraction of NOM. VUV irradiation reduced the HAAFP with increasing dose, in contrast, UV irradiation had a limited effect on the overall HAAFP. Following UV or VUV irradiation, the chlorinated species accounted for the majority of HAAFP; however, significant formation of brominated haloacetic acid (HAA) was observed. The nitrate concentration of the untreated water directly influenced the concentration of nitrite produced as a consequence of UV and VUV irradiation. Hydrogen peroxide formation was greater during VUV irradiation than during UV irradiation. Samples exposed to various doses of UV or VUV irradiation (up to 138 J cm(-2)) were deemed non-cytotoxic (African green monkey kidney cells) and non-mutagenic (Ames test).

  18. Correlation of irradiation data using activation fluences and irradiation temperature.

    NASA Technical Reports Server (NTRS)

    Lynch, J. H.

    1972-01-01

    A new method of correlating radiation damage data is tested using actual measured data taken from the open literature. This method, the activation fluence method, was found to be as accurate as other contemporary models with which it was compared. The new method also has several advantages over the other methods. The method employs a new entity, the activation fluence (time-integrated specific activation rate), as the independent variables in a regression model. Temperature at which the irradiation takes place is also a variable. Although the method was tested for a specific type of damage (change in nil-ductility transition temperature for A302-B steel) it has no inherent restrictions and is limited only by the imagination of the user.

  19. The present situation of the irradiation application industry and irradiation facilities in Japan

    NASA Astrophysics Data System (ADS)

    Mizusawa, K.; Baba, T.

    2003-08-01

    The irradiation application industry and irradiation facilities in Japan have been making slow but steady progress for the past 2-3 years. Beside conventional applications, new ones such as carbon fibers and membrane filters have come into the market. There are a lot of new applications about to emerge. PE tubing, already is in the European market, is being evaluated by end users in Japan. Cleaning of dioxin in exhaust gas was successfully tested at a pilot plant. Cross-linked PTFE and polyamide are waiting customers' evaluations as an engineering plastic. Surface cross-linking of artificial polycarbonate teeth has yielded remarkable experimental results. Cross-linking of polycaprolactone will be useful for biodegradable products. Being aware of the future growth of irradiation industry, contract service providers opened new facilities or increased their capability. Beside in-house facilities, there are now three Co-60 facilities and nine EB facilities available for contract irradiation in Japan.

  20. Evaluation of irradiation hardening of ion-irradiated V-4Cr-4Ti and V-4Cr-4Ti-0.15Y alloys by nanoindentation techniques

    NASA Astrophysics Data System (ADS)

    Miyazawa, Takeshi; Nagasaka, Takuya; Kasada, Ryuta; Hishinuma, Yoshimitsu; Muroga, Takeo; Watanabe, Hideo; Yamamoto, Takuya; Nogami, Shuhei; Hatakeyama, Masahiko

    2014-12-01

    Irradiation hardening behavior of V-4Cr-4Ti and V-4Cr-4Ti-0.15Y alloys after Cu-ion beam irradiation were investigated with a combination between nanoindentation techniques and finite element method (FEM) analysis. The ion-irradiation experiments were conducted at 473 K with 2.4 MeV Cu2+ ions up to 7.6 dpa. For the unirradiated materials, the increase in nanoindentation hardness with decreasing indentation depth, so-called indentation size effect (ISE), was clearly observed. After irradiation, irradiation hardening in the measured depth was identified. Hardening behavior of bulk-equivalent hardness for V-4Cr-4Ti-0.15Y alloy was similar to that for V-4Cr-4Ti alloy. Y addition has little effect on irradiation hardening at 473 K. Adding the concept of geometrically necessary dislocations (GNDs) to constitutive equation of V-4Cr-4Ti alloy, the ISE was simulated. A constant value of α = 0.5 was derived as an optimal value to simulate nanoindentation test for ion-irradiated V-4Cr-4Ti alloy. Adding the term of irradiation hardening Δσirrad. to constitutive equation with α = 0.5, FEM analyses for irradiated surface of V-4Cr-4Ti alloy were carried out. The analytic data of FEM analyses based on neutron-irradiation hardening equivalent to 3.0 dpa agreed with the experimental data to 0.76 dpa. The comparison indicates that irradiation hardening by heavy ion-irradiation is larger than that by neutron-irradiation at the same displacement damage level. Possible mechanisms for extra hardening by heavy ion-irradiation are the processes that the injected Cu ions could effectively produce irradiation defects such as interstitials compared with neutrons, and that higher damage rate of ion-irradiation enhanced nucleation of irradiation defects and hence increased the number density of the defects compared with neutron-irradiation.

  1. Vanadium irradiation at ATR - neutronics aspects

    SciTech Connect

    Gomes, I.C.; Smith, D.L.

    1995-04-01

    Calculations were performed to estimate damage and transmutation rates in vanadium irradiated in the ATR (Advanced Test Reactor) located in Idaho. The main focuses of the study are to evaluate the transmutation of vanadium to chromium and to explore ways to design the irradiation experiment to avoid excessive transmutation. It was found that the A-hole of ATR produces damage rate of {approximately} 0.2%/dpa of vanadium to chromium. A thermal neutron filter can be incorporated into the design to reduce the vanadium-to-chromium transmutation rate to low levels. A filter 1-2 mm thick of gadolinium or hafnium can be used.

  2. Irradiation embrittlement characterization of the EUROFER 97 material

    NASA Astrophysics Data System (ADS)

    Kytka, M.; Brumovsky, M.; Falcnik, M.

    2011-02-01

    The paper summarizes original results of irradiation embrittlement study of EUROFER 97 material that has been proposed as one candidate of structural materials for future fusion energy systems and GEN IV. Test specimens were manufactured from base metal as well as from weld metal and tested in initial unirradiated condition and also after neutron irradiation. Irradiation embrittlement was characterized by testing of toughness properties at transition temperature region - static fracture toughness and dynamic fracture toughness properties, all in sub-size three-point bend specimens (27 × 4 × 3 mm 3). Testing and evaluation was performed in accordance with ASTM and ESIS standards, fracture toughness KJC and KJd data were also evaluated with the "Master curve" approach. Moreover, J- R dependencies were determined and analyzed. The paper compares unirradiated and irradiated properties as well as changes in transition temperature shifts of these material parameters. Discussion about the correlation between static and dynamic properties is also given. Results from irradiation of EUROFER 97 show that this steel - base metal as well as weld metal - is suitable as a structural material for reactor pressure vessels of innovative nuclear systems - fusion energy systems and GEN IV. Transition temperature shifts after neutron irradiation by 2.5 dpa dose show a good agreement in the case of EUROFER 97 base material for both static and dynamic fracture toughness tests. From the results it can be concluded that there is a low sensitivity of weld metal to neutron irradiation embrittlement in comparison with EUROFER 97 base metal.

  3. FOOD IRRADIATION REACTOR

    DOEpatents

    Leyse, C.F.; Putnam, G.E.

    1961-05-01

    An irradiation apparatus is described. It comprises a pressure vessel, a neutronic reactor active portion having a substantially greater height than diameter in the pressure vessel, an annular tank surrounding and spaced from the pressure vessel containing an aqueous indium/sup 1//sup 1//sup 5/ sulfate solution of approximately 600 grams per liter concentration, means for circulating separate coolants through the active portion and the space between the annular tank and the pressure vessel, radiator means adapted to receive the materials to be irradiated, and means for flowing the indium/sup 1//sup 1//sup 5/ sulfate solution through the radiator means.

  4. Fuel or irradiation subassembly

    DOEpatents

    Seim, O.S.; Hutter, E.

    1975-12-23

    A subassembly for use in a nuclear reactor is described which incorporates a loose bundle of fuel or irradiation pins enclosed within an inner tube which in turn is enclosed within an outer coolant tube and includes a locking comb consisting of a head extending through one side of the inner sleeve and a plurality of teeth which extend through the other side of the inner sleeve while engaging annular undercut portions in the bottom portion of the fuel or irradiation pins to prevent movement of the pins.

  5. Instability of nanoscale metallic particles under electron irradiation in TEM

    NASA Astrophysics Data System (ADS)

    Chen, X. Y.; Zhang, S. G.; Xia, M. X.; Li, J. G.

    2016-03-01

    The stability of nano metallic glass under electron beam in transmission electron microscope (TEM) was investigated. The most common voltage of TEM used in metallic materials characterization was either 200 kV or 300 kV. Both situations were investigated in this work. An amorphous metallic particle with a dimension of a few hundred nanometers was tested under 300 keV electron irradiation. New phase decomposed from the parent phase was observed. Moreover, a crystal particle with the same composition and dimension was tested under 200 keV irradiation. Decomposition process also occurred in this situation. Besides, crystal orientation modification was observed during irradiation. These results proved that the electron beam in TEM have an effect on the stability of nanoscale samples during long time irradiation. Atomic displacement was induced and diffusion was enhanced by electron irradiation. Thus, artifacts would be induced when a nanoscale metallic sample was characterized in TEM.

  6. AFIP-3 Irradiation Summary Report

    SciTech Connect

    Danielle M Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-05-01

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-3 was designed to evaluate the performance of monolithic fuels at a prototypic scale of 2.25 inches x 21.5 inches x 0.050 inches (5.75 cm x 54.6 cm x 0.13cm). The AFIP-3 experiment was fabricated by hot isostatic pressing (HIP) and consists of two plates, one with a zirconium (Zr) diffusion barrier and one with a silicon (Si) enhanced fuel/clad interface1,2. The following report summarizes the life of the AFIP-3 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

  7. AFIP-3 Irradiation Summary Report

    SciTech Connect

    Danielle M Perez

    2011-04-01

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-3 was designed to evaluate the performance of monolithic fuels at a prototypic scale of 2.25 inches x 21.5 inches x 0.050 inches (5.75 cm x 54.6 cm x 0.13cm). The AFIP-3 experiment was fabricated by hot isostatic pressing (HIP) and consists of two plates, one with a zirconium (Zr) diffusion barrier and one with a silicon (Si) enhanced fuel/clad interface1,2. The following report summarizes the life of the AFIP-3 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

  8. AFIP-3 Irradiation Summary Report

    SciTech Connect

    Danielle M Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2012-03-01

    The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-3 was designed to evaluate the performance of monolithic fuels at a prototypic scale of 2.25 inches x 21.5 inches x 0.050 inches (5.75 cm x 54.6 cm x 0.13cm). The AFIP-3 experiment was fabricated by hot isostatic pressing (HIP) and consists of two plates, one with a zirconium (Zr) diffusion barrier and one with a silicon (Si) enhanced fuel/clad interface1,2. The following report summarizes the life of the AFIP-3 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

  9. ACL reconstruction with BPTB autograft and irradiated fresh frozen allograft*

    PubMed Central

    Sun, Kang; Tian, Shao-qi; Zhang, Ji-hua; Xia, Chang-suo; Zhang, Cai-long; Yu, Teng-bo

    2009-01-01

    Objective: To analyze the clinical outcomes of arthroscopic anterior cruciate ligament (ACL) reconstruction with irradiated bone-patellar tendon-bone (BPTB) allograft compared with non-irradiated allograft and autograft. Methods: All BPTB allografts were obtained from a single tissue bank and the irradiated allografts were sterilized with 2.5 mrad of irradiation prior to distribution. A total of 68 patients undergoing arthroscopic ACL reconstruction were prospectively randomized consecutively into one of the two groups (autograft and irradiated allograft groups). The same surgical technique was used in all operations done by the same senior surgeon. Before surgery and at the average of 31 months of follow-up (ranging from 24 to 47 months), patients were evaluated by the same observer according to objective and subjective clinical evaluations. Results: Of these patients, 65 (autograft 33, irradiated allograft 32) were available for full evaluation. When the irradiated allograft group was compared to the autograft group at the 31-month follow-up by the Lachman test, the anterior drawer test (ADT), the pivot shift test, and KT-2000 arthrometer test, statistically significant differences were found. Most importantly, 87.8% of patients in the autograft group and just only 31.3% in the irradiated allograft group had a side-to-side difference of less than 3 mm according to KT-2000. The failure rate of the ACL reconstruction with irradiated allograft (34.4%) was higher than that with autograft (6.1%). The anterior and rotational stabilities decreased significantly in the irradiated allograft group. According to the overall International Knee Documentation Committee (IKDC), functional and subjective evaluations, and activity level testing, no statistically significant differences were found between the two groups. Besides, patients in the irradiated allograft group had a shorter operation time and a longer duration of postoperative fever. When the patients had a fever, the

  10. Progress on performance assessment of ITER enhanced heat flux first wall technology after neutron irradiation

    NASA Astrophysics Data System (ADS)

    Hirai, T.; Bao, L.; Barabash, V.; Chappuis, Ph; Eaton, R.; Escourbiac, F.; Giqcuel, S.; Merola, M.; Mitteau, R.; Raffray, R.; Linke, J.; Loewenhoff, Th; Pintsuk, G.; Wirtz, M.; Boomstra, D.; Magielsen, A.; Chen, J.; Wang, P.; Gervash, A.; Safronov, V.

    2016-02-01

    ITER first wall (FW) panels are irradiated by energetic neutrons during the nuclear phase. Thus, an irradiation and high heat flux testing programme is undertaken by the ITER organization in order to evaluate the effects of neutron irradiation on the performance of enhanced heat flux (EHF) FW components. The test campaign includes neutron irradiation (up to 0.6-0.8 dpa at 200 °C-250 °C) of mock-ups that are representative of the final EHF FW panel design, followed by thermal fatigue tests (up to 4.7 MW m-2). Mock-ups were manufactured by the same manufacturing process as proposed for the series production. After a pre-irradiation thermal screening, eight mock-ups will be selected for the irradiation campaigns. This paper reports the preparatory work of HHF tests and neutron irradiation, assessment results as well as a brief description of mock-up manufacturing and inspection routes.

  11. Tensile properties of CLAM steel irradiated up to 20.1 dpa in STIP-V

    NASA Astrophysics Data System (ADS)

    Ge, Hongen; Peng, Lei; Dai, Yong; Huang, Qunying; Ye, Minyou

    2016-01-01

    Specimens of China low activation martensitic steel (CLAM) were irradiated in the fifth experiment of SINQ Target Irradiation Program (STIP-V) up to 20.1 dpa/1499 appm He/440 °C. Tensile tests were performed at room temperature (R.T) and irradiation temperatures (Tirr) in the range of 25-450 °C. The tensile results demonstrated strong effect of irradiation dose and irradiation temperature on hardening and embrittlement. With Tirr below ˜314 °C, CLAM steel specimens tested at R.T and Tirr showed similar evolution trend with irradiation dose, compared to other reduced activation ferritic/martensitic (RAFM) steels in similar irradiation conditions. At higher Tirr above ˜314 °C, it is interesting that the hardening effect decreases and the ductility seems to recover, probably due to a strong effect of high irradiation temperature.

  12. Emulation of reactor irradiation damage using ion beams

    SciTech Connect

    Was, G. S.; Jiao, Z.; Getto, E.; Sun, K.; Monterrosa, A. M.; Maloy, S. A.; Anderoglu, O.; Sencer, B. H.; Hackett, M.

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide, irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.

  13. Emulation of reactor irradiation damage using ion beams

    DOE PAGES

    Was, G. S.; Jiao, Z.; Getto, E.; ...

    2014-06-14

    The continued operation of existing light water nuclear reactors and the development of advanced nuclear reactor depend heavily on understanding how damage by radiation to levels degrades materials that serve as the structural components in reactor cores. The first high dose ion irradiation experiments on a ferritic-martensitic steel showing that ion irradiation closely emulates the full radiation damage microstructure created in-reactor are described. Ferritic-martensitic alloy HT9 (heat 84425) in the form of a hexagonal fuel bundle duct (ACO-3) accumulated 155 dpa at an average temperature of 443°C in the Fast Flux Test Facility (FFTF). Using invariance theory as a guide,more » irradiation of the same heat was conducted using self-ions (Fe++) at 5 MeV at a temperature of 460°C and to a dose of 188 displacements per atom. The void swelling was nearly identical between the two irradiation and the size and density of precipitates and loops following ion irradiation are within a factor of two of those for neutron irradiation. The level of agreement across all of the principal microstructure changes between ion and reactor irradiation establishes the capability of tailoring ion irradiation to emulate the reactor-irradiated microstructure.« less

  14. The results and analysis of irradiation experiments conducted on reactor vessel plate and weld materials

    SciTech Connect

    Biemiller, E.C.; Carter, R.G.; Rosinski, S.T.

    1996-09-01

    This paper documents the extensive amount of experimental work on radiation damage to reactor vessel materials carried out by Yankee Atomic Electric Company (YAEC) and others in support of a licensing effort to restart the Yankee Rowe nuclear power plant. The effect of plate nickel content and microstructure on irradiation damage sensitivity was assessed. Typical reactor pressure vessel plate materials each containing 0.24% (by weight) copper, but different nickel contents at 0.19% and 0.63% were heat treated to produce different microstructures. A Linde 80 weld containing 0.30% copper and 1.00% nickel was produced and heat treated to test microstructure effects on the irradiation response of weld metal. Materials taken from plate surface locations (vs 1/4%) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from a rapid quench, is maintained after irradiation. Irradiations were conducted at two irradiation temperatures, 500 F (260 C) and 550 F (288 C), to determine the effect of irradiation temperature on embrittlement. The results of this irradiation testing and additional data from a DOE/Sandia National Laboratories irradiation study show an irradiation temperature effect that is not consistent, but varies with the materials tested. The test results demonstrate that for nickel bearing steels, the superior toughness of plate surface material is maintained even after irradiation to high fluences, and for the copper content tested, nickel has little effect on irradiation response. A mixed effect of microstructure/heat treatment on the materials` irradiation response was noted. Phosphorus potentially played a role in the irradiation response of the low nickel material irradiated at 500 F (288 C) but did not show prominence in the irradiations for the same material conducted at 500 F (260 C).

  15. Irradiating insect pests

    Technology Transfer Automated Retrieval System (TEKTRAN)

    This is a non-technical article focusing on phytosanitary uses of irradiation. In a series of interview questions, I present information on the scope of the invasive species problem and the contribution of international trade in agricultural products to the movement of invasive insects. This is foll...

  16. Update on meat irradiation

    SciTech Connect

    Olson, D.G.

    1997-12-01

    The irradiation of meat and poultry in the United States is intended to eliminate pathogenic bacteria from raw product, preferably after packaging to prevent recontamination. Irradiation will also increase the shelf life of raw meat and poultry products approximately two to three times the normal shelf life. Current clearances in the United States are for poultry (fresh or frozen) at doses from 1.5 to 3.0 kGy and for fresh pork at doses from 0.3 to 1.0 kGy. A petition for the clearance of all red meat was submitted to the Food and Drug Administration (FDA) in July 1994. The petition is for clearances of fresh meat at doses from 1.5 to 4.5 kGy and for frozen meat at {approximately}2.5 to 7.5 kGy. Clearance for red meat is expected before the end of 1997. There are 28 countries that have food irradiation clearances, of which 18 countries have clearances for meat or poultry. However, there are no uniform categories or approved doses for meat and poultry among the countries that could hamper international trade of irradiated meat and poultry.

  17. Phytosanitary applications of irradiation

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Phytosanitary treatments are used to disinfest agricultural commodities of quarantine pests so the commodities can be shipped across quarantine barriers to trade. Ionizing irradiation is a promising treatment that is increasing in use. Almost 19,000 tons of sweet potatoes and several fruits, plus ...

  18. Generic phytosanitary irradiation treatments

    Technology Transfer Automated Retrieval System (TEKTRAN)

    The history of the development of generic phytosanitary irradiation (PI) treatments is discussed beginning with its initial proposal in 1986. Generic PI treatments in use today are 150 Gy for all hosts of Tephritidae, 250 Gy for all arthropods on mango and papaya shipped from Australia to New Zeala...

  19. Experiment Safety Assurance Package for the 40- to 52-GWd/MT Burnup Phase of Mixed Oxide Fuel Irradiation in Small I-hole Positions in the Advanced Test Reactor

    SciTech Connect

    S. T. Khericha; R. C. Pedersen

    2003-09-01

    This experiment safety assurance package (ESAP) is a revision of the last mixed uranium and plutonium oxide (MOX) ESAP issued in June 2002). The purpose of this revision is to provide a basis to continue irradiation up to 52 GWd/MT burnup [as predicted by MCNP (Monte Carlo N-Particle) transport code The last ESAP provided basis for irradiation, at a linear heat generation rate (LHGR) no greater than 9 kW/ft, of the highest burnup capsule assembly to 50 GWd/MT. This ESAP extends the basis for irradiation, at a LHGR no greater than 5 kW/ft, of the highest burnup capsule assembly from 50 to 52 GWd/MT.

  20. Design of unique pins for irradiation of higher actinides in a fast reactor

    SciTech Connect

    Basmajian, J.A.; Birney, K.R.; Weber, E.T.; Adair, H.L.; Quinby, T.C.; Raman, S.; Butler, J.K.; Bateman, B.C.; Swanson, K.M.

    1982-03-01

    The actinides produced by transmutation reactions in nuclear reactor fuels are a significant factor in nuclear fuel burnup, transportation and reprocessing. Irradiation testing is a primary source of data of this type. A segmented pin design was developed which provides for incorporation of multiple specimens of actinide oxides for irradiation in the UK's Prototype Fast Reactor (PFR) at Dounreay Scotland. Results from irradiation of these pins will extend the basic neutronic and material irradiation behavior data for key actinide isotopes.

  1. Irradiation creep of various ferritic alloys irradiated {approximately}400 C in the PFR and FFTF reactors

    SciTech Connect

    Toloczko, M.B.; Garner, F.A.; Eiholzer, C.R.

    1998-03-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400 C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400 C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 {times} 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

  2. The volume effect in irradiated mouse colorectum

    NASA Astrophysics Data System (ADS)

    Skwarchuk, Mark William

    1997-11-01

    Damage of the colorectum is the dose-limiting normal tissue complication following radiotherapy of prostate and cervical cancers. One approach for decreasing complications is to physically reduce the treatment volume. Mathematical models have been previously developed to describe the change in associated toxicity with a change in irradiated volume, i.e. the 'volume effect', for serial-type normal tissues including the colorectum. The first goal of this thesis was to test the hypothesis that there would not be a threshold length in the development of obstruction after irradiation of mouse colorectum, as predicted by the Probability model of the volume effect. The second goal was to examine if there were differences in the threshold and in the incidence of colorectal obstruction after irradiation of two mouse strains, C57B1/6 (C57) and C3Hf/Kam (C3H), previously found to be fibrosis-prone and-resistant, respectively, after lung irradiation due, in part, to genetic differences. The hypothesis examined was that differences in incidence between strains were due to the differential expression of the fibrogenic cytokines TGF/beta and TNF/alpha. Various lengths of C57 and C3H mouse colorectum were irradiated and the incidence of colorectal obstruction was followed up to 15 months. A threshold length was observed for both mouse strains, in contradiction of model predictions. The mechanism of the threshold was epithelial regeneration after irradiation. C57 mice had significantly higher incidence of colorectal obstruction compared to C3H mice, especially at smaller irradiated lengths. Colorectal tissue was obtained at various times after irradiation and prepared for histology, immunohistochemistry and RNase protection assay for measurement of TGF/beta 1, 2, 3 and TNF/alpha mRNA. Distinct strain differences in the histological time of appearance and spatial locations of fibrosis were observed. However, there were no consistent strain difference in mRNA levels or

  3. Effects of gamma irradiation on deteriorated paper

    NASA Astrophysics Data System (ADS)

    Bicchieri, Marina; Monti, Michela; Piantanida, Giovanna; Sodo, Armida

    2016-08-01

    Even though gamma radiation application, also at the minimum dosage required for disinfection, causes depolymerization and degradation of the paper substrate, recently published papers seemed, instead, to suggest that γ-rays application could be envisaged in some conditions for Cultural Heritage original documents and books. In some of the published papers, the possible application of γ-rays was evaluated mainly by using mechanical tests that scarcely reflect the chemical modifications induced in the cellulosic support. In the present article the effect of low dosage γ-irradiation on cellulosic substrates was studied and monitored applying different techniques: colorimetry, spectroscopic measurements, carbonyl content and average viscometric degree of polymerization. Two different papers were investigated, a non-sized, non-filled cotton paper, and a commercial permanent paper. To simulate a real deteriorated document, which could need γ-rays irradiation, some samples were submitted to a hydrolysis treatment. We developed a treatment based on the exposition of paper to hydrochloric acid vapors, avoiding any contact of the samples with water. This method induces a degradation similar to that observed on original documents. The samples were then irradiated with 3 kGy γ-rays at a 5258 Gy/h rate. The aforementioned analyses were performed on the samples just irradiated and after artificial ageing. All tests showed negative effects of gamma irradiation on paper. Non-irradiated paper preserves better its appearance and chemical properties both in the short term and after ageing, while the irradiated samples show appreciable color change and higher oxidation extent. Since the Istituto centrale restauro e conservazione patrimonio archivistico e librario is responsible for the choice of all restoration treatments that could be applied on library and archival materials under the protection of the Italian State (http://www.icpal.beniculturali.it/allegati/DM-7

  4. Intellectual, educational, and behavioural sequelae after cranial irradiation and chemotherapy.

    PubMed Central

    Anderson, V; Smibert, E; Ekert, H; Godber, T

    1994-01-01

    Cognitive and educational sequelae are inconsistently reported in children treated with cranial irradiation for acute lymphoblastic leukaemia. This study investigated differences in these skills after cranial irradiation, controlling the effects of chemotherapy and psychosocial factors. Three groups were evaluated: 100 children diagnosed with acute lymphoblastic leukaemia and treated with cranial irradiation and chemotherapy; 50 children diagnosed with acute lymphoblastic leukaemia or other cancers and treated with chemotherapy alone; and a healthy control group of 100 children. Children in the clinical groups stopped treatment at least two years before evaluation and had no history of relapse. Children were aged between 7 and 16 at the time of assessment. Evaluation included cognitive, educational, and behavioural measures. Analyses found that children receiving cranial irradiation and chemotherapy performed more poorly than non-irradiated groups on intellectual and educational tests, with verbal and attentional deficits most pronounced. Children receiving chemotherapy alone performed similarly to controls, suggesting such treatment is not associated with adverse neurobehavioural sequelae. PMID:8048815

  5. ELECTRON IRRADIATION OF SOLIDS

    DOEpatents

    Damask, A.C.

    1959-11-01

    A method is presented for altering physical properties of certain solids, such as enhancing the usefulness of solids, in which atomic interchange occurs through a vacancy mechanism, electron irradiation, and temperature control. In a centain class of metals, alloys, and semiconductors, diffusion or displacement of atoms occurs through a vacancy mechanism, i.e., an atom can only move when there exists a vacant atomic or lattice site in an adjacent position. In the process of the invention highenergy electron irradiation produces additional vacancies in a solid over those normally occurring at a given temperature and allows diffusion of the component atoms of the solid to proceed at temperatures at which it would not occur under thermal means alone in any reasonable length of time. The invention offers a precise way to increase the number of vacancies and thereby, to a controlled degree, change the physical properties of some materials, such as resistivity or hardness.

  6. BIOLOGICAL IRRADIATION FACILITY

    DOEpatents

    McCorkle, W.H.; Cern, H.S.

    1962-04-24

    A facility for irradiating biological specimens with neutrons is described. It includes a reactor wherein the core is off center in a reflector. A high-exposure room is located outside the reactor on the side nearest the core while a low-exposure room is located on the opposite side. Means for converting thermal neutrons to fast neutrons are movably disposed between the reactor core and the high and low-exposure rooms. (AEC)

  7. Position of the American Dietetic Association: food irradiation.

    PubMed

    Wood, O B; Bruhn, C M

    2000-02-01

    Food irradiation has been identified a sa safe technology to reduce the risk of foodborne illness as part of high-quality food production, processing, handling, and preparation. Food irradiation's history of scientific research , evaluation, and testing spans more than 40 countries around the world and it has been endorsed or support by numerous national and international food and organizations and professional groups. Food irradiation utilizes a source of ionizing energy that passes through food to destroy harmful bacteria and other organism. Often referred to as "cold pasteurization," food irradiation offers negligible loss of nutrients or sensory qualities in food as it does not substantially raise the temperature of the food during processing. Food irradiation does not replace proper food production, processing, handling, or preparation, nor can it enhance the quality of or prevent contact with foodborne bacteria after irradiation. In the United States, manufacturers are required to identify irradiated food sold to consumers with an international symbol (Radura) and and terminology describing the process on product labels. In addiction, food irradiation facilities are thoroughly regulated and monitored for worker and environmental safety. Members of The American Dietetic Association (ADA) and other food, nutrition, and health professionals have a responsibility to educate consumers, food processors, manufacturers and retailers about the safety and application of the technology. When consumers are educated about food irradiation, many prefer irradiated products because of their increased safety. It is the position of ADA that food irradiation enhances the safety and quality of the food supply and helps protect consumers from foodborne illness. The ADA encourages the government, food manufactures, food commodity groups, and qualified food and nutrition professionals to work together to educate consumers about this additional food safety tool and make this choice

  8. Commissioning and quality assurance of Dynamic WaveArc irradiation.

    PubMed

    Sato, Sayaka; Miyabe, Yuki; Takahashi, Kunio; Yamada, Masahiro; Nakamura, Mitsuhiro; Ishihara, Yoshitomo; Yokota, Kenji; Kaneko, Shuji; Mizowaki, Takashi; Monzen, Hajime; Hiraoka, Masahiro

    2015-03-08

    A novel three-dimensional unicursal irradiation technique "Dynamic WaveArc" (DWA), which employs simultaneous and continuous gantry and O-ring rotation during dose delivery, has been implemented in Vero4DRT. The purposes of this study were to develop a commissioning and quality assurance procedure for DWA irradiation, and to assess the accuracy of the mechanical motion and dosimetric control of Vero4DRT. To determine the mechanical accuracy and the dose accuracy with DWA irradiation, 21 verification test patterns with various gantry and ring rotational directions and speeds were generated. These patterns were irradiated while recording the irradiation log data. The differences in gantry position, ring position, and accumulated MU (EG, ER, and EMU, respectively) between the planned and actual values in the log at each time point were evaluated. Furthermore, the doses delivered were measured using an ionization chamber and spherical phantom. The constancy of radiation output during DWA irradiation was examined by comparison with static beam irradiation. The mean absolute error (MAE) of EG and ER were within 0.1° and the maximum error was within 0.2°. The MAE of EMU was within 0.7 MU, and maximum error was 2.7 MU. Errors of accumulated MU were observed only around control points, changing gantry, and ring velocity. The gantry rotational range, in which EMU was greater than or equal to 2.0 MU, was not greater than 3.2%. It was confirmed that the extent of the large differences in accumulated MU was negligibly small during the entire irradiation range. The variation of relative output value for DWA irradiation was within 0.2%, and this was equivalent to conventional arc irradiation with a rotating gantry. In conclusion, a verification procedure for DWA irradiation was designed and implemented. The results demonstrated that Vero4DRT has adequate mechanical accuracy and beam output constancy during gantry and ring rotation.

  9. Irradiation strongly reduces tumorigenesis of human induced pluripotent stem cells.

    PubMed

    Inui, Shoki; Minami, Kazumasa; Ito, Emiko; Imaizumi, Hiromasa; Mori, Seiji; Koizumi, Masahiko; Fukushima, Satsuki; Miyagawa, Shigeru; Sawa, Yoshiki; Matsuura, Nariaki

    2017-03-03

    Induced pluripotent stem (iPS) cells have demonstrated they can undergo self-renewal, attain pluripotency, and differentiate into various types of functional cells. In clinical transplantation of iPS cells, however, a major problem is the prevention of tumorigenesis. We speculated that tumor formation could be inhibited by means of irradiation. Since the main purpose of this study was to explore the prevention of tumor formation in human iPS (hiPS) cells, we tested the effects of irradiation on tumor-associated factors such as radiosensitivity, pluripotency and cell death in hiPS cells. The irradiated hiPS cells showed much higher radiosensitivity, because the survival fraction of hiPS cells irradiated with 2 Gy was < 10%, and there was no change of pluripotency. Irradiation with 2 and 4 Gy caused substantial cell death, which was mostly the result of apoptosis. Irradiation with 2 Gy was detrimental enough to cause loss of proliferation capability and trigger substantial cell death in vitro. The hiPS cells irradiated with 2 Gy were injected into NOG mice (NOD/Shi-scid, IL-2 Rγnull) for the analysis of tumor formation. The group of mice into which hiPS cells irradiated with 2 Gy was transplanted showed significant suppression of tumor formation in comparison with that of the group into which non-irradiated hiPS cells were transplanted. It can be presumed that this diminished rate of tumor formation was due to loss of proliferation and cell death caused by irradiation. Our findings suggest that tumor formation following cell therapy or organ transplantation induced by hiPS cells may be prevented by irradiation.

  10. Surface segregation during irradiation

    SciTech Connect

    Rehn, L.E.; Lam, N.Q.

    1985-10-01

    Gibbsian adsorption is known to alter the surface composition of many alloys. During irradiation, four additional processes that affect the near-surface alloy composition become operative: preferential sputtering, displacement mixing, radiation-enhanced diffusion and radiation-induced segregation. Because of the mutual competition of these five processes, near-surface compositional changes in an irradiation environment can be extremely complex. Although ion-beam induced surface compositional changes were noted as long as fifty years ago, it is only during the past several years that individual mechanisms have been clearly identified. In this paper, a simple physical description of each of the processes is given, and selected examples of recent important progress are discussed. With the notable exception of preferential sputtering, it is shown that a reasonable qualitative understanding of the relative contributions from the individual processes under various irradiation conditions has been attained. However, considerably more effort will be required before a quantitative, predictive capability can be achieved. 29 refs., 8 figs.

  11. Tensile behavior of irradiated manganese-stabilized stainless steel

    SciTech Connect

    Klueh, R.L.

    1996-10-01

    Tensile tests were conducted on seven experimental, high-manganese austenitic stainless steels after irradiation up to 44 dpa in the FFTF. An Fe-20Mn-12Cr-0.25C base composition was used, to which various combinations of Ti, W, V, B, and P were added to improve strength. Nominal amounts added were 0.1% Ti, 1% W, 0.1% V, 0.005% B, and 0.03% P. Irradiation was carried out at 420, 520, and 600{degrees}C on the steels in the solution-annealed and 20% cold-worked conditions. Tensile tests were conducted at the irradiation temperature. Results were compared with type 316 SS. Neutron irradiation hardened all of the solution-annealed steels at 420, 520, and 600{degrees}C, as measured by the increase in yield stress and ultimate tensile strength. The steel to which all five elements were added to the base composition showed the least amount of hardening. It also showed a smaller loss of ductility (uniform and total elongation) than the other steels. The total and uniform elongations of this steel after irradiation at 420{degrees}C was over four times that of the other manganese-stabilized steels and 316 SS. There was much less difference in strength and ductility at the two higher irradiation temperatures, where there was considerably less hardening, and thus, less loss of ductility. In the cold-worked condition, hardening occured only after irradiation at 420{degrees}C, and there was much less difference in the properties of the steels after irradiation. At the 420{degrees}C irradiation temperature, most of the manganese-stabilized steels maintained more ductility than the 316 SS. After irradiation at 420{degrees}C, the temperature of maximum hardening, the steel to which all five of the elements were added had the best uniform elongation.

  12. Temperature dependence of fracture toughness in HT9 steel neutron-irradiated up to 145 dpa

    SciTech Connect

    Baek, Jong-Hyuk; Byun, Thak Sang; Maloy, S; Toloczko, M

    2014-01-01

    The temperature dependence of fracture toughness in HT9 steel irradiated to high doses was investigated using miniature three-point bend (TPB) fracture specimens. These specimens were from the ACO-3 fuel duct wall of the Fast Flux Test Facility (FFTF), in which irradiation doses were in the range of 3.2 144.8 dpa and irradiation temperatures in the range of 380.4 502.6 oC. A miniature specimen reuse technique has been established for this investigation: the specimens used were the tested halves of miniature Charpy impact specimens (~13 3 4 mm) with diamond-saw cut in the middle. The fatigue precracking for specimens and fracture resistance (J-R) tests were carried out in a MTS servo-hydraulic testing machine with a vacuum furnace following the standard procedure described in the ASTM Standard E 1820-09. For each of five irradiated and one archive conditions, 7 to 9 J-R tests were performed at selected temperatures ranging from 22 C to 600 C. The fracture toughness of the irradiated HT9 steel was strongly dependent on irradiation temperatures rather than irradiation dose. When the irradiation temperature was below about 430 C, the fracture toughness of irradiated HT9 increased with test temperature, reached an upper shelf of 180 200 MPa m at 350 450 C and then decreased with test temperature. When the irradiation temperature 430 C, the fracture toughness was nearly unchanged until about 450 C and decreased with test temperature in higher temperature range. Similar test temperature dependence was observed for the archive material although the highest toughness values are lower after irradiation. Ductile stable crack growth occurred except for a few cases where both the irradiation temperature and test temperature are relatively low.

  13. Susceptibility of irradiated bovine aortic endothelial cells to injury

    SciTech Connect

    Zhou, M.H.; Dong, Q.; Ts'ao, C.

    1988-11-01

    Using cultured bovine aortic endothelial cells (BAEC), the authors attempted to determine whether prior irradiation would alter the susceptibility of these cells to three known injurious stimuli and, if so, whether the alteration would be related to radiation dose. BAEC were irradiated with 0, 5, or 10 Gy of gamma rays and, on the third postirradiation day, exposed to fibrin, nicotine, or bacterial endotoxin (lipopolysaccharide, LPS). Release of prelabeled 51Cr, representing cell lysis, cell detachment, or a combination of the two, was determined. Significant differences between irradiated and control cells were determined by using paired Student's t-tests. Irradiation did not appear to have altered the sensitivity of BAEC to fibrin-induced injury. Cells irradiated with 10 Gy of gamma rays, but generally not those irradiated with half this dose, showed a heightened susceptibility to nicotine. Contrary to the nicotine results, irradiated cells showed less cell detachment and lysis after exposure to LPS. These results suggest that the susceptibility of irradiated BAEC to harmful stimuli depends largely on the nature of the stimulus as well as the radiation dose.

  14. Surface, structural and tensile properties of proton beam irradiated zirconium

    NASA Astrophysics Data System (ADS)

    Rafique, Mohsin; Chae, San; Kim, Yong-Soo

    2016-02-01

    This paper reports the surface, structural and tensile properties of proton beam irradiated pure zirconium (99.8%). The Zr samples were irradiated by 3.5 MeV protons using MC-50 cyclotron accelerator at different doses ranging from 1 × 1013 to 1 × 1016 protons/cm2. Both un-irradiated and irradiated samples were characterized using Field Emission Scanning Electron Microscope (FESEM), X-ray Diffraction (XRD) and Universal Testing Machine (UTM). The average surface roughness of the specimens was determined by using Nanotech WSxM 5.0 develop 7.0 software. The FESEM results revealed the formation of bubbles, cracks and black spots on the samples' surface at different doses whereas the XRD results indicated the presence of residual stresses in the irradiated specimens. Williamson-Hall analysis of the diffraction peaks was carried out to investigate changes in crystallite size and lattice strain in the irradiated specimens. The tensile properties such as the yield stress, ultimate tensile stress and percentage elongation exhibited a decreasing trend after irradiation in general, however, an inconsistent behavior was observed in their dependence on proton dose. The changes in tensile properties of Zr were associated with the production of radiation-induced defects including bubbles, cracks, precipitates and simultaneous recovery by the thermal energy generated with the increase of irradiation dose.

  15. Experimental plan for irradiation experiment HRB-21

    SciTech Connect

    Goodin, D. T.; Kania, M. J.; Patton, B. W.

    1989-04-01

    Irradiation experiment HRB-21 is the first in a series of test capsules that are designed to provide a fuel-performance data base to be used for the validation of modular high-temperature gas-cooled reactor (MHTGR) coated-particle fuel performance models under MHTGR normal operating conditions and specific licensing basis events. Capsule HRB-21 will contain an advanced TRISO-P UCO/ThO{sub 2} - coated-particle fuel system with demonstrated low defective-particle fraction ({le}5 {times} 10{sup {minus}5}) and a heavy metal-contamination fraction ({le}1 {times} 10{sup {minus}5}) that meets MHTGR quality specifications. The coated particles and fuel compacts were fabricated in laboratory-scale facilities using MHTGR reference procedures at General Atomics (GA). Nearly 150,000 fissile and fertile particles will be irradiated in capsule HRB-21 at a mean volumetric fuel temperature of 975{degree}C and will achieve a peak fissile burnup of 26% fissions per initial metal atom (FIMA) while accumulating a fast neutron fluence of about 4.5 {times} 10{sup 25} neutrons/m{sup 2}. This experiment is a cooperative effort between the US Department of Energy (DOE) and the Japan Atomic Energy Research Institute (JAERI). The participants are the Oak Ridge National Laboratory (ORNL), GA, and the Tokai Research Establishment. Capsule HRB-21 will contain the US MHTGR fuel specimens, and a companion capsule, HRB-22, will contain the JAERI fuel. The irradiation will take place in the removable beryllium reflector facility of the High Flux Isotope Reactor (HFIR) at ORNL. The performance of the fuel during irradiation will be closely monitored through on-line fission gas release measurements. Detailed postirradiation examination and conduction cooldown simulation testing will be performed on the irradiated fuel compacts from both the HRB-21 and HRB-22 capsules. 5 refs., 9 figs., 6 tabs.

  16. FDA perspective on food irradiation

    SciTech Connect

    Pauli, G.H.

    1994-12-31

    The Center for Food Safety and Applied Nutrition (CFSAN) monitors the safety of food irradiation. A few limited uses are regulated, and occasionally CFSAN receives a petition for a new use. Despite extensive studies (more than 400) showing the safety of food irradiation, a cloud of suspicion continues to hang over this issue in the mind of the public. People perceive food irradiation and direct body irradiation as having similar implications. Food irradiation is banned in two states in the United States. Food is irradiated for the following purposes: delay of ripening, prevention of sprouting, eradication of pests and sterilization, and allowing commodities to be stored unrefrigerated for long periods of time. The dosage depends on the purpose of the irradiation. Radiolytic products are formed during irradiation and during storage afterward. Most of these products are also formed during conventional preservation. In 1980, CFSAN, then the Bureau of Foods, introduced the term unique radiolytic products for compounds not identified in foods after conventional processing. Although the existence of URPs was never proven chemically, the term has caused anxiety. Irradiation of foods in the commercially useful range does not generate radioactivity above natural background. Because radiolytic products formed from beef, chicken, and pork are primarily the same, irradiated foods of similar food groups may be evaluated generically.

  17. Post-irradiation-examination of irradiated fuel outside the hot cell

    SciTech Connect

    Dawn E. Janney; Adam B. Robinson; Thomas P. O'Holleran; R. Paul Lind; Marc Babcock; Laurence C. Brower; Julie Jacobs; Pamela K. Hoggan

    2007-09-01

    Because of their high radioactivity, irradiated fuels are commonly examined in a hot cell. However, the Idaho National Laboratory (INL) has recently investigated irradiated U-Mo-Al metallic fuel from the Reduced Enrichment for Research and Test Reactors (RERTR) project using a conventional unshielded scanning electron microscope outside a hot cell. This examination was possible because of a two-step sample-preparation approach in which a small volume of fuel was isolated in a hot cell and shielding was introduced during later stages of sample preparation. The resulting sample contained numerous sample-preparation artifacts but allowed analysis of microstructures from selected areas.

  18. Progress report on the design of a varying temperature irradiation experiment for operation in HFIR

    SciTech Connect

    Qualls, A.L.; Muroga, T.

    1997-04-01

    The purpose of this experiment is to determine effects of temperature variation during irradiation on microstructure and mechanical properties of potential fusion reactor structural materials. A varying temperature irradiation experiment is being performed under the framework of the Japan-USA Program of Irradiation Tests for fusion Research (JUPITER) to study the effects of temperature variation on the microstructure and mechanical properties of candidate fusion reactor structural materials. An irradiation capsule has been designed for operation in the High Flux Isotope Reactor at Oak Ridge National Laboratory that will allow four sets of metallurgical test specimens to be irradiated to exposure levels ranging from 5 to 10 dpa. Two sets of specimens will be irradiated at constant temperature of 500{degrees}C and 350{degrees}C. Matching specimen sets will be irradiated to similar exposure levels, with 10% of the exposure to occur at reduced temperatures of 300{degrees}C and 200{degrees}C.

  19. Replacement of 137Cs irradiators with x-ray irradiators.

    PubMed

    Dodd, Brian; Vetter, Richard J

    2009-02-01

    Self-shielded 137Cs irradiators have been used for many years to irradiate blood products to prevent graft vs. host disease and to irradiate cells and small animals in research. A report by the National Academy of Sciences recommends that careful consideration be given to replacement of 137Cs irradiators with x-ray irradiators. Several manufacturers and users of x-ray irradiators were contacted to determine costs of replacing and maintaining 137Cs irradiators with x-ray units and to assess users' experience with x-ray irradiators. Purchase costs of x-ray units are similar to 137Cs irradiators, but maintenance costs are significantly higher if annual service contracts are used. Performance of the two irradiator types appears to be equivalent, but in some cases x-ray irradiations may need to be performed in multiple configurations to achieve adequate uniformity in dose. No literature reports were found that evaluated the biological effectiveness of x rays vs. 137Cs gamma rays; therefore, a careful study should be conducted to determine the biological effectiveness of x rays vs. 137Cs gamma rays for biological responses relevant to transfusion medicine and immunological research. Throughput may be problematic for large transfusion medicine programs, and back-up plans may be necessary in case the x-ray unit needs to be taken out of service for extended maintenance. Disposition of a 137Cs irradiator will add to the cost of replacement with an x-ray unit, but disposal may be possible through the U.S. Department of Energy's Off-Site Source Recovery Program.

  20. In vitro antimicrobial activity of LED irradiation on Pseudomonas aeruginosa.

    PubMed

    Petrini, Morena; Trentini, Paolo; Tripodi, Domenico; Spoto, Giuseppe; D'Ercole, Simonetta

    2017-03-01

    Pseudomonas aeruginosa is an opportunistic pathogen responsible of many deaths due to nosocomial pneumonia each year. It is particularly resistant to many different classes of antibiotics and disinfectants. For all these reasons, there is the necessity to find novel approaches of treatment. The aim of this study was to evaluate the effect of 880nm light emitting diodes (LED) irradiation on P. aeruginosa, in vitro. Different LED irradiation parameters (time, energy output and the addition of methylene blue and chlorhexidine) have been tested in order to evaluate the effects on this bacterium. After treatment, the colony forming units per milliliter (CFU mL-1) were recorded and the data were submitted to ANOVA and Bonferroni post hoc tests at a level of significance of 5%. A statistical significant reduction of bacterial count has been registered after 5min of LED irradiation. The antibacterial effect was directly proportional to irradiation time and the output energy. The pre-treatment with methylene blue, seems to be not effective against P. aeruginosa, independently from irradiation parameters. On the contrary, the contemporary action of LED and chlorhexidine has shown a great reduction of bacterial count that was statistical significant respect chlorhexidine and LED alone. The effect of LED irradiation was visible also after 24h, when a lower bacterial count characterized all irradiated samples respect controls.

  1. Effect of thermal and irradiation aging simulation procedures on polymer properties

    SciTech Connect

    Bustard, L.D.; Minor, E.; Chenion, J.; Carlin, F.; Alba, C.; Gaussens, G.; LeMeur, M.

    1984-04-01

    Prior to initiating a qualification test on safety-related equipment, the testing sequence for thermal and irradiation aging exposures must be chosen. Likewise, the temperature during irradiation must be selected. Typically, U.S. qualification efforts employ ambient temperature irradiation, while French qualification efforts employ 70/sup 0/C irradiations. For several polymer materials, the influence of the thermal and irradiation aging sequence, as well as the irradiation temperature (ambient versus 70/sup 0/C), has been investigated in preparation for Loss-of-Coolant Accident simulated tests. Ultimate tensile properties at completion of aging are presented for three XLPO and XLPE, five EPR and EPDM, two CSPE (HYPALON), one CPE, one VAMAC, one polydiallylphtalate, and one PPS material. Bend test results at completion of aging are presented for two TEFZEL materials. Permanent set after compression results are presented for three EPR, one VAMAC, one BUNA N, one Silicone, and one Viton material.

  2. Irradiation creep and swelling of various austenitic alloys irradiated in PFR and FFTF

    SciTech Connect

    Garner, F.A.; Toloczko, M.B.

    1996-10-01

    In order to use data from surrogate neutron spectra for fusion applications, it is necessary to analyze the impact of environmental differences on property development. This is of particular importance in the study of irradiation creep and its interactions with void swelling, especially with respect to the difficulty of separation of creep strains from various non-creep strains. As part of an on-going creep data rescue and analysis effort, the current study focuses on comparative irradiations conducted on identical gas-pressurized tubes produced and constructed in the United States from austenitic steels (20% CW 316 and 20% CW D9), but irradiated in either the Prototype Fast Reactor (PFR) in the United Kingdom or the Fast Flux Test Facility in the United States. In PFR, Demountable Subassemblies (DMSA) serving as heat pipes were used without active temperature control. In FFTF the specimens were irradiated with active ({+-}{degrees}5C) temperature control. Whereas the FFTF irradiations involved a series of successive side-by-side irradiation, measurement and reinsertion of the same series of tubes, the PFR experiment utilized simultaneous irradiation at two axial positions in the heat pipe to achieve different fluences at different flux levels. The smaller size of the DMSA also necessitated a separation of the tubes at a given flux level into two groups (low-stress and high-stress) at slightly different axial positions, where the flux between the two groups varied {le}10%. Of particular interest in this study was the potential impact of the two types of separation on the derivation of creep coefficients.

  3. Continuous wave laser irradiation of explosives

    SciTech Connect

    McGrane, Shawn D.; Moore, David S.

    2010-12-01

    Quantitative measurements of the levels of continuous wave (CW) laser light that can be safely applied to bare explosives during contact operations were obtained at 532 nm, 785 nm, and 1550 nm wavelengths. A thermal camera was used to record the temperature of explosive pressed pellets and single crystals while they were irradiated using a measured laser power and laser spot size. A visible light image of the sample surface was obtained before and after the laser irradiation. Laser irradiation thresholds were obtained for the onset of any visible change to the explosive sample and for the onset of any visible chemical reaction. Deflagration to detonation transitions were not observed using any of these CW laser wavelengths on single crystals or pressed pellets in the unconfined geometry tested. Except for the photochemistry of DAAF, TATB and PBX 9502, all reactions appeared to be thermal using a 532 nm wavelength laser. For a 1550 nm wavelength laser, no photochemistry was evident, but the laser power thresholds for thermal damage in some of the materials were significantly lower than for the 532 nm laser wavelength. No reactions were observed in any of the studied explosives using the available 300 mW laser at 785 nm wavelength. Tables of laser irradiance damage and reaction thresholds are presented for pressed pellets of PBX9501, PBX9502, Composition B, HMX, TATB, RDX, DAAF, PETN, and TNT and single crystals of RDX, HMX, and PETN for each of the laser wavelengths.

  4. Results and analyses for irradiation/anneal experiments conducted on Yankee Rowe reactor pressure vessel surrogate materials. Final report

    SciTech Connect

    Biemiller, E.C.

    1995-12-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate and weld materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. The effect of plate nickel content and microstructure on irradiation damage sensitivity was assessed. Typical reactor vessel plate materials each containing 0.24% (by weight) copper, but different nickel contents at 0.63% and 0.19%, were heat treated to produce different microstructures in the test materials. A Linde 80 weld containing 0.30% copper and 1.00% nickel was produced and heat treated to test microstructure effects on the irradiation response of weld metal. Materials taken from plate surface locations (vs 1/4 thickness) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from a rapid quench, is maintained after irradiation. Irradiations were conducted at two irradiation temperatures (500{degrees}F and 550{degrees}F) to determine the effect of irradiation temperature on embrittlement. An annealing test matrix was also initiated to study the potential for a 650{degrees}F anneal. A major irradiation/annealing/reirradiation study was conducted by the DOE`s LWR Technology Center at Sandia National Laboratories (SNL), using an irradiation temperature of 550{degrees}F and a 850{degrees}F anneal. The results of the irradiation testing and the DOE/SNL annealing study show an irradiation temperature effect that is not consistent but, varies with the materials tested. The test results demonstrate that for nickel bearing steels, the superior toughness of plate surface material is maintained even after irradiation to high fluences and for the copper content tested, nickel has little effect on irradiation response.

  5. Preliminary analysis of irradiation effects on CLAM after low dose neutron irradiation

    NASA Astrophysics Data System (ADS)

    Peng, Lei; Huang, Qunying; Li, Chunjing; Liu, Shaojun

    2009-04-01

    To investigate the irradiation effects on a new version of reduced activation ferritic/martensitic steels (RAFMs) i.e. China Low Activation Martensitic steel (CLAM), neutron irradiation experiments has been being carried out under wide collaboration in China and overseas. In this paper, the mechanical properties of CLAM heats 0603A, 0408B, and 0408D were investigated before and after neutron irradiation to ˜0.02 dpa at 250 °C. The test results showed that ultimate strength and yield stress of CLAM HEAT 0603A increased about 10-30 MPa and ductile to brittle transition temperature (DBTT) shift was about 5 °C. For CLAM heats 0408B and 0408D, ultimate strength and yield stress increased about 80-150 MPa.

  6. Microstructure and mechanical behavior of neutron irradiated ultrafine grained ferritic steel

    SciTech Connect

    Ahmad Alsabbagh; Apu Sarkar; Brandon Miller; Jatuporn Burns; Leah Squires; Douglas Porter; James I. Cole; K. L. Murty

    2014-10-01

    Neutron irradiation effects on ultra-fine grain (UFG) low carbon steel prepared by equal channel angular pressing (ECAP) has been examined. Counterpart samples with conventional grain (CG) sizes have been irradiated alongside with the UFG ones for comparison. Samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to 1.24 dpa. Atom probe tomography revealed manganese, silicon-enriched clusters in both ECAP and CG steel after neutron irradiation. X-ray quantitative analysis showed that dislocation density in CG increased after irradiation. However, no significant change was observed in UFG steel revealing better radiation tolerance.

  7. Craniospinal irradiation techniques

    SciTech Connect

    Scarlatescu, Ioana Avram, Calin N.; Virag, Vasile

    2015-12-07

    In this paper we present one treatment plan for irradiation cases which involve a complex technique with multiple beams, using the 3D conformational technique. As the main purpose of radiotherapy is to administrate a precise dose into the tumor volume and protect as much as possible all the healthy tissues around it, for a case diagnosed with a primitive neuro ectoderm tumor, we have developed a new treatment plan, by controlling one of the two adjacent fields used at spinal field, in a way that avoids the fields superposition. Therefore, the risk of overdose is reduced by eliminating the field divergence.

  8. Effects of the SARM ACP-105 on rotorod performance and cued fear conditioning in sham-irradiated and irradiated female mice.

    PubMed

    Dayger, Catherine; Villasana, Laura; Pfankuch, Timothy; Davis, Matthew; Raber, Jacob

    2011-03-24

    Female mice are more susceptible to radiation-induced cognitive changes than male mice. Previously, we showed that, in female mice, androgens antagonize age-related cognitive decline in aged wild-type mice and androgens and selective androgen receptor modulators (SARMs) antagonize cognitive changes induced by human apolipoprotein E4, a risk factor for developing age-related cognitive decline. In this study, the potential effects of the SARM ACP-105 were assessed in female mice that were either sham-irradiated or irradiated with ¹³⁷Cesium at a dose of 10Gy. Behavioral testing started 2 weeks following irradiation. Irradiation impaired sensorimotor function in vehicle-treated mice but not in ACP-105-treated mice. Irradiation impaired cued fear conditioning and ACP-105 enhanced fear conditioning in sham-irradiated and irradiated mice. When immunoreactivity for microtubule-associated protein 2 was assessed in the cortex of sham-irradiated mice, there was a brain area × ACP-105 interaction. While ACP-105 reduced MAP-2 immunoreactivity in the sensorimotor cortex, there was a trend towards increased MAP-2 immunoreactivity in the enthorhinal cortex. No effect on MAP-2 immunoreactivity was seen in the irradiated cortex or sham-irradiated or irradiated hippocampus. Thus, there are relatively early radiation-induced behavioral changes in female mice and reduced MAP-2 levels in the sensorimotor cortex following ACP-105 treatment might contribute to enhanced rotorod performance.

  9. Understanding the Irradiation Behavior of Zirconium Carbide

    SciTech Connect

    Motta, Arthur; Sridharan, Kumar; Morgan, Dane; Szlufarska, Izabela

    2013-10-11

    -induced microstructures mapped spatially and temporally, microstructural evolution during post-irradiation annealing, and atomistic modeling of defect formation and transport energetics will provide new, critical understanding about property changes in ZrC. The behavior of materials under irradiation is determined by the balance between damage production, defect clustering, and lattice response. In order to predict those effects at high temperatures so targeted testing can be expanded and extrapolated beyond the known database, it is necessary to determine the defect energetics and mobilities as these control damage accumulation and annealing. In particular, low-temperature irradiations are invaluable for determining the regions of defect mobility. Computer simulation techniques are particularly useful for identifying basic defect properties, especially if closely coupled with a well-constructed and complete experimental database. The close coupling of calculation and experiment in this project will provide mutual benchmarking and allow us to glean a deeper understanding of the irradiation response of ZrC, which can then be applied to the prediction of its behavior in reactor conditions.

  10. Mechanical properties and plasticity size effect of Fe-6%Cr irradiated by Fe ions and by neutrons

    NASA Astrophysics Data System (ADS)

    Hardie, C. D.; Odette, G. R.; Wu, Y.; Akhmadaliev, S.; Roberts, S. G.

    2016-12-01

    The mechanical behaviour of Fe6%Cr in the un-irradiated, self-ion irradiated and neutron irradiated conditions was measured and compared. Irradiations were performed to the same dose and at the same temperature but to very different damage rates for both methods. The materials were tested using nanoindentation and micromechanical testing, and compared with microstructural observations from Transmission Electron Microscopy (TEM) and Atom Probe Tomography (APT) reported elsewhere. Irradiated and un-irradiated micro-cantilevers with a wide range of dimensions were used to study the interrelationships between irradiation hardening and size effects in small-scale plasticity. TEM and APT results identified that the dislocation loop densities were ∼2.9 × 1022m-3 for the neutron irradiated material and only 1.4 × 1022m-3 for the ion irradiated material. Cr segregation to loops was only found for the neutron-irradiated material. The nanoindentation hardness increase due to neutron irradiation was 3 GPa and that due to ion irradiation 1 GPa. The differences between the effects of the two irradiation types are discussed, taking into account inconsistencies in damage calculations, and the differences in PKA spectra, dose rate and transmutation products for the two irradiation types.

  11. Bulk charging and breakdown in electron-irradiated polymers

    NASA Technical Reports Server (NTRS)

    Frederickson, A. R.

    1981-01-01

    High energy electron irradiations were performed in an experimental and theoretical study of ten common polymers. Breakdowns were monitored by measuring currents between the electrodes on each side of the planar samples. Sample currents as a function of time during irradiation are compared with theory. Breakdowns are correlated with space charge electric field strength and polarity. Major findings include evidence that all polymers tested broke down, breakdowns remove negligible bulk charge and no breakdowns are seen below 20 million V/m.

  12. Irradiation of northwest agricultural products

    NASA Astrophysics Data System (ADS)

    Eakin, D. E.; Tingey, G. I.

    1985-02-01

    Irradiation of food for disinfestation and preservation is increasing in importance because of increasing restrictions on various chemical treatments. Irradiation treatment is of particular interest in the Northwest because of a growing supply of agricultural products and the need to develop new export markets. Several products have, or could potentially have, significant export markets if stringent insect ocntrol procedures are developed and followed. Due to the recognized potential benefits of irradiation, this program was conducted to evaluate the benefits of using irradiation on Northwest agricultural products. Commodities currently included in the program are cherries, apples, asparagus, spices, hay, and hides.

  13. Food irradiation and the consumer

    NASA Astrophysics Data System (ADS)

    A Thomas, P.

    The poster presents a review of research work undertaken on the perception and understanding that consumers have of food irradiation. Food irradiation is not a revolutionary new food processing technique, in fact it is probably one of the most investigated methods presently available. Many countries such as Belgium, France, Denmark, Italy, Spain, the Netherlands and the United States of America permit food irradiation. In Britain it is presently banned although this is currently under review. Awareness of food irradiation by the general public in Britain, although not extensively researched would appear to be increasing, especially in the light of recent media coverage. New quantitative and qualitative work indicates that the general public are concerned about the safety and effectiveness of food irradiation. Research has shown that a large proportion of consumers in Britain, if given the opportunity to purchase irradiated food, would not do so. Further exploration into this response revealed the fact that consumers are confused over what food irradiation is. In addition, there is concern over the detection of irradiated food. The views presented in this paper, of the consumer reaction to irradiated food are of great importance to those involved in the food industry and industries allied to it, which are ultimately dependent on the consumer for their commercial survival.

  14. Wholesomeness of irradiated food

    NASA Astrophysics Data System (ADS)

    Ehlermann, Dieter A. E.

    2016-12-01

    Just with the emergence of the idea to treat food by ionizing radiation, the concerns were voiced whether it would be safe to consume such food. Now, we look back on more than hundred years of research into the 'wholesomeness', a terminology developed during those efforts. This review will cover the many questions which had been raised, explaining the most relevant ones in some detail; it will also give place to the concerns and elucidate their scientific relevance and background. There has never been any other method of food processing studied in such depth and in such detail as food irradiation. The conclusion based on science is: Consumption of any food treated at any high dose is safe, as long as the food remains palatable. This conclusion has been adopted by WHO, also by international and national bodies. Finally, this finding has also been adopted by Codex Alimentarius in 2003, the international standard for food. However, this conclusion has not been adopted and included at its full extent in most national regulations. As the literature about wholesomeness of irradiated food is abundant, this review will use only a few, most relevant references, which will guide the reader to further reading.

  15. Fatigue behavior of type 316 stainless steel following neutron irradiation inducing helium

    SciTech Connect

    Grossbeck, M.L.; Liu, K.C.

    1980-01-01

    Since a tokamak fusion reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially the first wall and blanket. Type 316 stainless steel in the 20% cold-worked condition has been irradiated in the HFIR in order to introduce helium as well as displacement damage. A miniature hourglass specimen was developed for the reactor irradiations and subsequent fully reversed low cycle fatigue testing. For material irradiated and tested at 430/sup 0/C in vacuum to a damage level of 7 to 15 dpa and containing 200 to 1000 appm He, a reduction in life by a factor of 3 to 10 was observed. An attempt was made to predict irradiated fatigue life by fitting data from irradiated material to a power law equation similar to the universal slopes equation and using ductility ratios from tensile tests to modify the equation for irradiated material.

  16. Protecting fibrinogen with rutin during UVC irradiation for viral inactivation.

    PubMed

    Marx, G; Mou, X; Freed, R; Ben-Hur, E; Yang, C; Horowitz, B

    1996-04-01

    Fibrinogen solutions were irradiated with UVC (254 nm) to inactivate contaminating viruses. In order to protect fibrinogen during UVC irradiation, 0.5 mM rutin was added prior to UVC exposure and subsequently removed during processing. Viral kill by 0.1 J/cm2 UVC resulted in the following inactivation values (log 10): non-lipid-enveloped viruses: Parvo > or = 5.5; encephalomyocarditis virus > or = 6.5; hepatitis A virus > or = 6.5: lipid-enveloped viruses: human immunodeficiency virus > or = 5.7; vesicular stomatitis virus > or = 5.7. Fibrinogen irradiated with 0.5 mM rutin did not significantly differ from unirradiated material in terms of clot time and breaking strength. In the absence of rutin, UVC irradiation of fibrinogen at similar fluence led to loss of solubility, increased clot time and the cleavage of fibrino-peptides that reacted with dinitrophenyl hydrazine as a test for ketonic carbonyl groups. High-performance liquid chromatography and mass spectrometry data showed that rutin exposed to UVC formed numerous breakdown, oxidation and combinational products. Experiments with 3H-rutin showed that after UVC irradiation, subsequent processing by a C18 resin and alcohol precipitation removed > 99% rutin, representing < 10 ppm rutin in the final fibrinogen preparations. Residual 3H-rutin was not covalently bonded to the fibrinogen. Immunochemical studies with rabbit antisera to UVC irradiated (with rutin) fibrinogen showed the absence of neoimmungens. By all measures, rutin prevents fibrinogen degradation during virucidal UVC irradiation.

  17. AGC-1 Pre-Irradiation Data Report Status

    SciTech Connect

    William Windes

    2011-08-01

    The Next Generation Nuclear Plant (NGNP) Graphite R&D program is currently measuring irradiated material property changes in several grades of nuclear graphite for predicting their behavior and operating performance within the core of new Very High Temperature Reactor (VHTR) designs. The Advanced Graphite Creep (AGC) experiment consisting of six irradiation capsules will generate this irradiated graphite performance data for NGNP reactor operating conditions. All samples in the experiment will be fully characterized before irradiation, irradiated in the Advanced Test Reactor (ATR), and then re-examined to determine the irradiation induced changes to key materials properties in the different graphite grades. The information generated during the AGC experiment will be utilized for NRC licensing of NGNP reactor designs, shared with international collaborators in the Generation IV Information Forum (GIF), and eventually utilized in ASME design code for graphite nuclear applications. This status report will describe the process the NGNP Graphite R&D program has developed to record the AGC1 pre-irradiation examination data.

  18. Degradation of poly(carbonate urethane) by gamma irradiation

    NASA Astrophysics Data System (ADS)

    Özdemir, T.; Usanmaz, A.

    2007-06-01

    Poly(carbonate urethane) (PCU), is a valuable commercial engineering polymer. In order to understand the possible use of PCU in radioactive waste management as a solidifying agent or as a disposal container, radiation stability of the PCU is studied by Co-60 gamma irradiations at two different dose rates of 1540 and 82.8 Gy/h. The total dose of irradiation was up to 6.24 MGy. Degradation nature was tested by studying the changes in mechanical and thermal properties with rate and total dose of irradiation. Ultimate tensile strength and toughness first increased and then decreased with the irradiation dose. Half value dose (HVD) for elongation was 4010 kGy and for tensile strength 6010 kGy at the dose rate of 1540 Gy/h. The non-irradiated PCU transparent color changed to yellow and then brown with increased irradiation dose. The FTIR spectral analysis showed a random scission of polymer with irradiation. From the experimental observation, it was shown that PCU can be used for embedding radioactive waste for about 300 years.

  19. Identification of irradiated refrigerated pork with the DNA comet assay

    NASA Astrophysics Data System (ADS)

    Araújo, M. M.; Marin-Huachaca, N. S.; Mancini-Filho, J.; Delincée, H.; Villavicencio, A. L. C. H.

    2004-09-01

    Food irradiation can contribute to a safer and more plentiful food supply by inactivating pathogens, eradicating pests and by extending shelf-life. Particularly in the case of pork meat, this process could be a useful way to inactivate harmful parasites such as Trichinella and Taenia solium. Ionizing radiation causes damage to the DNA of the cells (e.g. strand breaks), which can be used to detect irradiated food. Microelectrophoresis of single cells (``Comet Assay'') is a simple and rapid test for DNA damage and can be used over a wide dose range and for a variety of products. Refrigerated pork meat was irradiated with a 60Co source, Gammacell 220 (A.E.C.L.) installed in IPEN (Sa~o Paulo, Brazil). The doses given were 0, 1.5, 3.0 and 4.5kGy for refrigerated samples. Immediately after irradiation the samples were returned to the refrigerator (6°C). Samples were kept in the refrigerator after irradiation. Pork meat was analyzed 1, 8 and 10 days after irradiation using the DNA ``Comet Assay''. This method showed to be an inexpensive and rapid technique for qualitative detection of irradiation treatment.

  20. UVA system for human cornea irradiation

    NASA Astrophysics Data System (ADS)

    Pereira, Fernando R. A.; Stefani, Mario; Otoboni, José A.; Richter, Eduardo H.; Rossi, Giuliano; Mota, Alessandro D.; Ventura, Liliane

    2009-02-01

    According to recent studies, an increase in corneal stiffness is a promising alternative for avoiding ectasias and for stagnating keratoconus of grades 1 and 2. The clinical treatment consists essentially of instilling Riboflavin (vitamin B2), in the cornea and then irradiating the corneal tissue, with UVA (365nm) radiation at 3mW/cm2 for 30min. This procedure provides collagen cross-linking in the corneal surface, increasing its stiffness. This work presents a system for UVA irradiation of the corneas at a peak wavelength of 365nm with adjustable power up to 5mW. The system has closed loop electronics to control the emitted power with 20% precision from the sated power output. The system is a prototype for performing corneal cross-linking and has been clinically tested. The closed loop electronics is a differential from the equipments available on the market.

  1. Commercial implementation of food irradiation

    NASA Astrophysics Data System (ADS)

    Welt, M. A.

    In July 1981, the first specifically designed multi-purpose irradiation facility for food irradiation was put into service by the Radiation Technology, Inc. subsidiary Process Technology, Inc. in West Memphis, Arkansas. The operational experience gained, resulted in an enhanced design which was put into commercial service in Haw River, North Carolina, by another subsidiary, Process Technology (N.C.), Inc. in October 1983. These facilities have enabled the food industry to assess the commercial viability of food irradiation. Further impetus towards commercialization of food irradiation was gained in March 1981 with the filing in the Federal Register, by the FDA, of an Advanced Proposed Notice of Rulemaking for Food Irradiation. Two years later in July 1983, the FDA approved the first food additive regulation involving food irradiation in nineteen years, when they approved the Radiation Technology, Inc. petition calling for the sanitization of spices, onion powder and garlic powder at a maximum dosage of 10 kGy. Since obtaining the spice irradiation approval, the FDA has accepted four additional petitions for filing in the Federal Register. One of the petitions which extended spice irradiation to include insect disinfestation has issued into a regulation while the remaining petitions covering the sanitization of herbs, spice blends, vegetable seasonings and dry powdery enzymes as well as the petition to irradiate hog carcasses and pork products for trichinae control at 1 kGy, are expected to issue either before the end of 1984 or early in 1985. More recently, food irradiation advocates in the United States received another vote of confidence by the announcement that a joint venture food irradiation facility to be constructed in Hawaii by Radiation Technology, is backed by a contractual committment for the processing of 40 million pounds of produce per year. Another step was taken when the Port of Salem, New Jersey announced that the Radiation Technology Model RT-4104

  2. Triacylglycerols profiling as a chemical tool to identify mushrooms submitted to gamma or electron beam irradiation.

    PubMed

    Fernandes, Ângela; Barreira, João C M; Antonio, Amilcar L; Martins, Anabela; Ferreira, Isabel C F R; Oliveira, M Beatriz P P

    2014-09-15

    In order to define irradiation treatment as a routine conservation methodology, it is imperative to develop chemometric indicators with the ability to distinguish irradiated from unirradiated foodstuffs. Electron spin resonance, photostimulated luminescence and thermoluminescence methods were employed to monitor radiation-induced markers, as well as different chemical compounds produced from the lipidic fraction of different foodstuffs. Apart from these methods, the specificity of triacylglycerol profiles has previously been detected in mushroom species, as has the effect of irradiation treatment in the triacylglycerol profiles of chestnut. Accordingly, the feasibility of using this as a chemometric indicator of irradiated mushrooms was evaluated. In line with the obtained results in literature, the effects of each type of irradiation were significantly different, as can be concluded from the correlations among discriminant functions and variables within each statistical test. Triacylglycerol profiling proved to be a useful tool to detect irradiated mushrooms, independently of the species or irradiation source, especially for doses above 1 kGy.

  3. Pallet irradiators for food processing

    NASA Astrophysics Data System (ADS)

    McKinnon, R. G.; Chu, R. D. H.

    This paper looks at the various design concepts for the irradiation processing of food products, with particular emphasis on handling the products on pallets. Pallets appear to offer the most attractive method for handling foods from many considerations. Products are transported on pallets. Warehouse space is commonly designed for pallet storage and, if products are already palletized before and after irradiation, then labour could be saved by irradiating on pallets. This is also an advantage for equipment operation since a larger carrier volume means lower operation speeds. Different pallet irradiator design concepts are examined and their suitability for several applications are discussed. For example, low product holdup for fast turn around will be a consideration for those operating an irradiation "service" business; others may require a very large source where efficiency is the primary requirement and this will not be consistent with low holdup. The radiation performance characteristics and processing costs of these machines are discussed.

  4. Post irradiation examination of thermal reactor fuels

    NASA Astrophysics Data System (ADS)

    Sah, D. N.; Viswanathan, U. K.; Ramadasan, E.; Unnikrishnan, K.; Anantharaman, S.

    2008-12-01

    The post irradiation examination (PIE) facility at the Bhabha Atomic Research Centre (BARC) has been in operation for more than three decades. Over these years this facility has been utilized for examination of experimental fuel pins and fuels from commercial power reactors operating in India. In a program to assess the performance of (U,Pu)O 2 MOX fuel prior to its introduction in commercial reactors, three experimental MOX fuel clusters irradiated in the pressurized water loop (PWL) of CIRUS up to burnup of 16 000 MWd/tU were examined. Fission gas release from these pins was measured by puncture test. Some of these fuel pins in the cluster contained controlled porosity pellets, low temperature sintered (LTS) pellets, large grain size pellets and annular pellets. PIE has also been carried out on natural UO 2 fuel bundles from Indian PHWRs, which included two high burnup (˜15 000 MWd/tU) bundles. Salient investigations carried out consisted of visual examination, leak testing, axial gamma scanning, fission gas analysis, microstructural examination of fuel and cladding, β, γ autoradiography of the fuel cross-section and fuel central temperature estimation from restructuring. A ThO 2 fuel bundle irradiated in Kakrapar Atomic Power Station (KAPS) up to a nominal fuel burnup of ˜11 000 MWd/tTh was also examined to evaluate its in-pile performance. The performance of the BWR fuel pins of Tarapur Atomic Power Stations (TAPS) was earlier assessed by carrying out PIE on 18 fuel elements selected from eight fuel assemblies irradiated in the two reactors. The burnup of these fuel elements varied from 5000 to 29 000 MWd/tU. This paper provides a brief review of some of the fuels examined and the results obtained on the performance of natural UO 2, enriched UO 2, MOX, and ThO 2 fuels.

  5. Principles and practice of a bellows-loaded compact irradiation vehicle

    NASA Astrophysics Data System (ADS)

    Byun, Thak Sang; Li, Meimei; Snead, Lance L.; Katoh, Yutai; Burchell, Timothy D.; McDuffee, Joel L.

    2013-08-01

    This article describes the key design principles and application of a mini-bellows loaded irradiation creep frame technology developed for use in the high flux isotope reactor (HFIR). For this irradiation vehicle, the bellows, frame, sample, and temperature monitor are contained within a hydraulic or fixed "rabbit" capsule of a few inches in length. Of critical importance and key to this technology is the viability and stability of the metallic bellows under the elevated temperature irradiation environment. Conceptual design and supporting analysis have been performed for tension and compression specimens. Benchtop verification has substantiated the modeling regarding the ability of the bellows to produce sufficient stress to induce irradiation creep in subsize specimens. Discussion focuses on the possible stress ranges in specimens induced by the miniature gas-pressurized bellows and the limitations imposed by the size and structure of thin-walled bellows. A brief discussion of pre- and post-irradiation measurement of the integrity of load frames irradiated to 4.4 × 1025 n/m2 (E > 0.1 MeV) is presented. Following this protocol, the pre-irradiation loading to a sample is determined and post-irradiation loading inferred. An in-reactor creep testing technology using pressurized mini-bellows has been established for irradiation creep tests in tensile or compressive mode using the HFIR rabbit capsule design. Results from theoretical calculation and in-furnace tests confirmed that the pressurized bellows-loaded miniature creep frame can produce enough thrust force to induce irradiation creep in subsize specimens. Bellows materials, types and dimensions were selected considering in-reactor integrity, load transferring function, weldability, and in-reactor stability. Both stainless steel and IN 718 mini-bellows were proven to be capable of irradiation creep testing. A practical process for the testing and evaluation of applied stress has been developed and applied to in

  6. Postnatal neurophysiologic effects of prenatal X-irradiation.

    PubMed

    Jensh, R P; Eisenman, L M; Brent, R L

    1995-02-01

    Histological and neurophysiological effects of in utero irradiation were examined following exposure of pregnant Wistar rat to 2.0 Gy X-irradiation or sham-irradiated on the 17th day of gestation. The 234 newborns were monitored for the age of appearance of four selected physiologic markers and the age of acquisition of five selected reflexes. Offspring were evaluated as young adults using selected behavioural tests. Postnatal growth was monitored weekly. Selected offspring were autopsied to determine the presence of morphologic central nervous system alterations. The results indicated that 2.0 Gy X-irradiation during the foetal period in rat gestation caused permanent alterations in the mature adult organism, which include non-recuperable growth retardation, morphologic changes in the brain such as microcephaly, abnormal cerebellar cortical cellular patterns, and alterations in the cell architecture of the hippocampus; diminished attainment of selected reflexes; alterations in the appearance of selected physiologic markers; and changes in adult test performance indicating significant hyperactivity among the irradiated offspring. Such exposure to X-irradiation during this period results in behavioural and morphologic alterations, which persist throughout life.

  7. Impact toughness of irradiated reduced-activation ferritic steels*1

    NASA Astrophysics Data System (ADS)

    Klueh, R. L.; Alexander, D. J.

    1994-09-01

    Eight chromium-tungsten steels ranging from 2.25 to 12 wt% Cr were irradiated at 365°C to 13-14 dpa in the Fast Flux Test Facility. Post irradiation Charpy impact tests showed a loss of toughness for all steels, as measured by an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy. The most irradiation-resistant steels were two 9% Cr steels: the DBTT of a 9Cr-2W-0.25V-0.1C steel increased 29°C, and for the same composition with an addition of 0.07% Ta the DBTT increased only 15°C. This is the smallest shift ever observed for such a steel irradiated to these levels. The other steels developed shifts in DBTT of 100 to 300°C. A 2.25% Cr steel with 2% W, 0.25% V, and 0.1% C was less severely affected by irradiation than 2.25% Cr steels with 0.25% V and no tungsten, 2% W and no vanadium, and with 1% W and 0.25% V. Irradiation resistance appears to be associated with microstructure, and microstructural manipulation may lead to improved properties.

  8. Prediction of yield stress in highly irradiated ferritic steels

    NASA Astrophysics Data System (ADS)

    Windsor, Colin G.; Cottrell, Geoff; Kemp, Richard

    2008-03-01

    The design of any fusion power plant requires information on the irradiation hardening of low-activation ferritic/martensitic steels beyond the range of most present measurements. Neural networks have been used by Kemp et al (J. Nucl. Mater. 348 311-28) to model the yield stress of some 1811 irradiated alloys. The same dataset has been used in this study, but has been divided into a training set containing the majority of the dataset with low irradiation levels, and a test set which contains just those alloys which have been irradiated above a given level. For example some 4.5% of the alloys were irradiated above 30 displacements per atom. For this 'prediction' problem it is found that simpler networks with fewer inputs are advantageous. By using target-driven dimensionality reduction, linear combinations of the atomic inputs reduce the test residual below that achievable by adding inputs from single atoms. It is postulated that these combinations represent 'mechanisms' for the prediction of irradiated yield stress.

  9. Combined Effects of Temperature and Irradiation on Concrete Damage

    DOE PAGES

    Le Pape, Yann; Giorla, Alain; Sanahuja, Julien

    2016-01-01

    Aggregate radiation-induced volumetric expansion (RIVE) is a predominant mechanism in the formation of mechanical damage in the hardened cement paste (hcp) of irradiated concrete under fast-neutron flux (Giorla et al. 2015). Among the operating conditions difference between test reactors and light water reactors (LWRs), the difference of irradiation flux and temperature is significant. While a temperature increase is quite generally associated with a direct, or indirect (e.g., by dehydration) loss of mechanical properties (Maruyama et al. 2014), we found that it causes a partial annealing of irradiation amorphization of α-quartz, hence, reducing RIVE rate. Based on data collected by Bykovmore » et al. (1981), an incremental RIVE model coupling neutron fluence and temperature is developed. The elastic properties and coefficient of thermal expansion (CTE) of irradiated polycrystalline quartz are interpreted through analytical homogenization of experimental data on irradiated α-quartz published by Mayer and Lecomte (1960). Moreover, the proposed model, implemented in the meso-scale simulation code AMIE, is compared to experimental data obtained on ordinary concrete made of quartz/quartzite aggregate (Dubrovskii et al. 1967). Substantial discrepancy, in terms of damage and volumetric expansion developments, is found when comparing irradiation scenarios assuming constant flux and temperature, as opposed to more realistic test reactor operation conditions.« less

  10. Combined Effects of Temperature and Irradiation on Concrete Damage

    SciTech Connect

    Le Pape, Yann; Giorla, Alain; Sanahuja, Julien

    2016-01-01

    Aggregate radiation-induced volumetric expansion (RIVE) is a predominant mechanism in the formation of mechanical damage in the hardened cement paste (hcp) of irradiated concrete under fast-neutron flux (Giorla et al. 2015). Among the operating conditions difference between test reactors and light water reactors (LWRs), the difference of irradiation flux and temperature is significant. While a temperature increase is quite generally associated with a direct, or indirect (e.g., by dehydration) loss of mechanical properties (Maruyama et al. 2014), we found that it causes a partial annealing of irradiation amorphization of α-quartz, hence, reducing RIVE rate. Based on data collected by Bykov et al. (1981), an incremental RIVE model coupling neutron fluence and temperature is developed. The elastic properties and coefficient of thermal expansion (CTE) of irradiated polycrystalline quartz are interpreted through analytical homogenization of experimental data on irradiated α-quartz published by Mayer and Lecomte (1960). Moreover, the proposed model, implemented in the meso-scale simulation code AMIE, is compared to experimental data obtained on ordinary concrete made of quartz/quartzite aggregate (Dubrovskii et al. 1967). Substantial discrepancy, in terms of damage and volumetric expansion developments, is found when comparing irradiation scenarios assuming constant flux and temperature, as opposed to more realistic test reactor operation conditions.

  11. Proton irradiation effects on beryllium – A macroscopic assessment

    SciTech Connect

    Simos, Nikolaos; Elbakhshwan, Mohamed; Zhong, Zhong; Camino, Fernando

    2016-07-01

    Beryllium, due to its excellent ne